ML20128P373

Spent Fuel Storage Capacity Expansion

ML20128P373
Person / Time
Site:	Pilgrim
Issue date:	01/05/1993
From:	HOLTEC INTERNATIONAL
To:
Shared Package
ML20128P367	List: BECO-93-016, Proposed TS 3.5.C,D & E Re k-infinity Factor,Spent Fuel Pool Storage Capacity & Max Loads Allowed to Travel Over Fuel Assemblies,Respectively ML20128P365 ML20128P373
References
HI-92925, NUDOCS 9302240259
Download: ML20128P373 (343)

Text

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HOLTEC i N T ' E P. N A T l O N A L i

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l PILGRIM NUCLEAR POWER STATION SPENT FUEL STORAGE CAPACITY EXPANSION BOSTON EDISON COMPANY USNRC DOCIGT NO 50-293 Prepared by HOLTEC INTERNATIONAL 2060 Fairfax Avenue l Cherry Hill, NJ 08003-1666 l

REPORT HI-92925 l

January 5,- 1993 l

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T OLE OF' CONTENTS' SECTION PAGE

1.0 INTRODUCTION

1.1 References 1-4 2.0 MODULE LAYOUT FOR INCREASED STORAGE 2.1 Heavy Load Considerations for the 2-1 Proposed Reracking Operation 3.0 RACK FABRICATION AND APPLICABLE CODES 3.1 Design Objective 1 3.2 Anatomy of the Rack Module 3-2 3.3 Material Considerations 3-4 3.3.1 Introduction 3-4 i 3.3.2 Structural Materials 3-4 3.3.3 Poison Material 3-4

! 3.3.4 Compatibility with Coolant ;3-6 3.4 Codes, etandards, and Practices for the 3-7=

Spent tuel Pool Modification 4.O CRITICALITY SAFETY ANALYSIS I:

( 4.1 Introduction 4 l 4.2 Summary and Conclusions 4-2 l 4.3 Abnormal and Accident Conditions 4-4 4.4 Input Parameters 4-5 4.4.1 Fuel Assembly Design Specification;- 4-5

[ 4.4.2 Storage. Rack Cell Specifications 4-5 l 4.5 Analysis Methodology. .

4-5' i 4.6 Criticality Analyses and. Tolerance Variations 4-6 4.6.1 Nominal Design Case. 4-6 I 4.6.2 Uncertainties Due to-Manufacturing

4-7 4.6.2.1 Boron Loading Variation 4-7 4.6.2.2 Boral Width Tolerance' Variation 4-8' 4.6.2.3 Storage Cell Lattice Pitch 4-8 Variation 4.6.2.4 Stainless Steel Thickness Tolerances 4-8 4.6.2.5 Fuel Enrichment and Density. Variation 4-8

L 4.6.2.6 Zirconium Flow Channel 4-9

! 4.6.3 Uncertainty in Depletion Calculations- '4-9

-4.7 Higher Enrichments and GE-11 Fuel 4-10 i

4.8 Abnormal and Accident Conditions 4-10 4.8.1 Temperature and Water Density Effects 4 4.8.2 Abnormal Location of a Fuel Assembly 4-11 4.8.3 Eccentric Puel Assembly Positioning 4-11 4.8.4 Zirconium Puel Channel Distortion 4-11 4.8.5 Dropped Puel Assembly 4-12 4.8.6 Puel Rack Lateral Movement 4-12 4.9 Existing Spent Fuel Storage Racks 4-12 4.10 Comparison with Other Recently Licensed 4-14 US Plants 4.11 References 4-15 Appendix A to Section 4 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5.1 Introduction 5-1 5.2 Spent Fuel Pool Cooling and Cleanup 5-2

System Description

5.3 Decay Heat Load Calculations 5-5 5.4 Discharge Scenarios 5-5 5.5 Bulk Pool Temperatures 5-6 5.6 Time-to-Boil 5-8 5.7 Local Pcol Water Temperature 5-8 5.7.1 Basis 5-8 5.7.2 Model Description 5-9 5.8 Cladding Temperature 5-11 5.9 Results 5-13 5.10 References 5-14 6.0 STRUCIURAL/ SEISMIC CONSIDERATIONS 6.1 Introduction 6-1 6.2 Analysis Outline 6-1 6.3 Artificial Time-Histories 6-6 6.4 Eack Modeling for Dynamic Simulations 6-9 6.4.1 General-Remarks 6-9 6.4.2 The 3-D 22 DOF Model for Single Rack Module 6-11 6.4.2.1 Assumptions 6-11 6.4.2.2 Model Details 6-12 6.4.2.3 Fluid Coupling Details 6-13 6.4.2.4 Stiffness Element Details 6-14 6.4.3 Whole Pool Multi-Rack (WPMR) Medel 6-16 6.4.3.1 General Remarks 6-16 6.4.3.2 Whole Pool Fluid Coupling 6-16 6.4.3.3 Coefficients of Friction 6-17 6.4.3.4 Modeling Details 6-17 11

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1 6.5 Acceptance _ Criteria, Stress Limits, and- R Material Properties 6-18 o6.5.1 Acceptance Criteria 6-18 6.5.2 Stress Limits for Various Conditions 6-20 6.5.2.1 Normal-and Upset Conditions 6-20 (Level A or Level B) 6.5.2.2 Level L Service Limits 6-22 6.5.2.3 Dimensionless-Stress Factors 23.

6.5.3 -Material Properties 6-23 6.6 Governing Equations of Motion 6-24 6.7 Results of 3-D Nonlinear Analyses of 6-26 Single Racks 6.7.1 Racks in the Fuel Pool 6-26 6.7.1.1 Impact Analyses 6-28 6.7.1.2 Wald Stresses 6-28 f, . 8 Results from Wholo Pool Multi-Rack (WPMR) 6-29 ,

Analyses 6.9 References 6-32 7.0 ACCIDENT ANALYSIS AND MISCELLANEOUS STRUCTURAL EVALUATIONS 7.1 Introduction 7 7.2 Refueling Accidents 7-1 7.2.1 Heavy Fuel Assembly 7-1 Dropped on a'New Rack 7.2.2 Heavy Fuel Assembly 7-2.

Dropped on an Existing Rack 7.3 Local Buckling of Fuel Cell Walls 7-3 7.4 Analysis of Welded Joints.in Rack due to Isolated Hot Cell 7-4 7.5 References 7-5 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 8.1 Introduction 8-1 8.2 Description of Spent Fuel Pool Structure 8-2 8.3 Definition of Loads 8-3 8.3.1 Static Loading 8-3 8.3.2 Dynamic Loading 8-3 8.3.3 Thermal Loading 8-4 8.4 Analysis Procedures 8-4 8.4.1 Finite Element Analysis Model 8-4 8.4.2 Analysis Methodology 8-4 8.4.3 Load Combinations 8-5 8.5 Results of Analyses 8 8.6 Pool Liner 8-7 8.7 Conclusions 8-8 8.8 References 8-8 111

9.0 RADIOLOGICAL EVALUATION 9.1 Fuel Handling Accident 9-1 9.1.1 Assumptions and Source-Term Calculations 9-li

'9.1.2 Results 9-4 9.2 Solid Radwaste' 9-41

9. 3 - Gaseous Releases. 9 9.4 Personnel Exposures 9-5 9.5 Anticipated Exposure Luring Expansion 9-6 4 10.O 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 2,0-3 Schedule 10.2.3 Measurement Program 10-4~

10.2.4 Surveillance Coupon Acceptance 10 Criteria 10.3 References 10 11.O ENVIRONMENTAL COST-BENEFIT ASSESSMENT 11.1 Introduction- 11-1 11.2 Imperative for Adding Racks 11-1 11.3 Appraisal of Alternatives 2

-11.4 Resource Commitment 11-6 11.5 Environmental Considerations 11-6 11.6 References 11-8

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LIST.0F TABLES 1.1. Projected and Existing Discharge Schedule 2.1 Existing Racks in the Pilgrim Pool 0

2.2 Data on New Rack Modules Present & Future Campaigns 2.3 Existing and How Rack Modules Dimensional & Weight Data-2.4 Design Data for New and Existing Racks 2.5 Heavy Load Handling Compliance Matrix:(NUREG-0612) i 3.1 Boral Experience List (Domestic and Foreign) 3.2 1100 Alloy Aluminum Physical Properties 3.3 Chemical Composition - Aluminum (1100 Alloy) 3.4 Boron Carbide Chemical Composition, Weight % ,

Boron Carbide Physical Properties ,

4.2.1 Summary of Criticality Safety Analyses 4.3.1 Reactivity Etfects of Abnormal and Accident Conditions 4.4.1 Fuel Assembly Design Specifications 4.6.1 Reactivity Uncertainties due to Manuf acturing Tolerances .

4.8.1 Effect of Temperature and Void on Calculated Reactivity of Storage Rack 4.9.1 Summary of Criticality Safety Analyses - Existing Racks 5.2.1 Spent Fuel Pool Cooling System-5.2.2 RHR Heat Exchanger Data 5.3.1 Projected and Existing Discharge Schedule 5.4.1 Data for Normal Discharge (Case 1) 5.4.2 Data-for Full Core Offload Condition (Case 2) 5.8.1 Radial and Total Peaking Factors 5.8.2 Data-for Local Temperature Analysis v

T 5.9.1 Background Decay Heat and Pool Capacity Data 5.9.2 SFP Bulk Pool Temperature 5.9.3 Results of Time-to-Boil 5.9.4 Maximum Local Pool Water and Fuel Cladding Temperature for the Limiting Case (Case 1) 6.1.1 Listing of Plants Where Dynarack Was Applied ,

6.3.1 Four Sets of Time-Histories Generated From Safe Shutdown Earthquake (SSE) Responso Spectra and their Cross-Correlation Coefficients 6.3.2 Four Sets of Time-Histories Generated From Operating Basis Earthquake (OBE) Response Spectra and their C.oss-Correlation Coefficients 6.3.3 Determination of the Controlling Set of SSE Time-l Histories 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 6.7.1 Results of Single Rack Analyses - List of All Runs 6.7.2 Suumary of ' Worst Results from 36 Runs of Single Rack Analysis (Loaded with Regular Fuel Assemblies; Analysis Basis SSE Seismic) 6.7.3 Summary of Worst Results from 36 Runs of. Single Rack Analysis (Loaded with Regular Fuel Assemblies; Analysis Basis SSE Seismic) 6.7.4 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnissei.rf8 6.7.5 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnissei.rf5 6.7.6 Summary Results of 3-D Single Rack Analysis for Rack-Module: Rack-N1; Holtec Run I.D. drnissei.rf2 vi

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-6.7.7 Summary Results of 3-D Single Rack Analysis for - Rack - l Module: Rack-N1; Holtec Run I.D. drnissei.rh8 6.7.8 Summary Results of 3-D Single ' Rack Analysis for' Rack Module: Rack-N1; Holtec Run I.D.'drnissei.rh5 6.7.9 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnissei.rh2 6.7.10 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnissei.re8 6.7.11 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnissei.re5 6.7.12 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnissei.re2 6.7.13 Summary Results of 3-D Single Rack- Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnisseo.rf5 6.7.14 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnisseo.rf5 6.7.15 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnisseo.rf2 6.7.16 Summary Results of 3-D Single Rack Analysis for . Reck '

Module: Rack-N1; Holtec Run I.D. drnisseo.rh8 6.7.17 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnisseo.rh5 6.7.18 Summary Rt.sults of 3-D Single Rack Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnissea.rh2 6.7.19 Summary Results of 3-D Single Rack - Analysis for Rack Module: Rack-N1; Holtec Run I.D. drnisseo. reb 6.7.20 Summary Results of 3-D Single Rack Analysis for Rack.

Module: Rack-N1; Holtec Run I.D. drnisseo.re5 6.7.21 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N1; Holtec Run I.D. drn5ssco.re2 6.7.22 Summary Results of 3-D Single - Rack Analysis for Rack

. Module: Rack-N5; Holtec Run I.D. drn5ssei.rf8 l

l 6.7.23 Summary Results of'3-D Single Rack Analysis for Rack-

! Module: Rack-N5; Holtec Run I.D. drn5ssei.rf5.

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l 6.7.24 Summary Results of L-D Single - Rack A5alysis for Rack Module: Rack-N5; Holtec Run I.D. drn5ssei.rf2 6.7.25 Sumary Results of 3-D Single Rack Analysis for - Rack Module: Rack-N5; Holtec Run I.D. drn5ssei.rh8 6.7.26 Sumary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5ssei.rh5 6.7.27 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5ssei.rh2 6.7.28 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5ssei.re8 6.7.29 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5ssei.re5 6.7.30 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5sseo.re2 -

6.7.31 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5sseo.rf8 6.7.32 Summary Results of 3-D Single Rack Analysis for ' Rack Module: Rack-N5; Holtec Run I.D. drn5sseo.rf5 6.7.33 Summary Results of 3-D Single Rack Analysis for' Rack Module: Rack-N5; Holtec Run I.D. drn5sseo.rf2 6.7.34 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5sseo.rh8 6.7.35 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5sseo.rh5 '

6.7.36 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5sseo.rh2 6.7.37 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5sseo.re8 6.7.38 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5sseo.re5 6.7.39 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D. drn5sseo.re2 6.7.40 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-E9; Holtec Run I.D. dre9ssei.rf8 viii

6.7.41 Summary Results of 3-D Single _ Rack Analysis for Rack Module: Rack-E9;-Holtec Run,I.D. dre9ssei.rf2 6.7.42 Summary _ Results of 3-D Single Rack - Analysis : for Rack Module: Rack-E9; Holtec Run I.D. dre9ssei.rh8 6.7.43 Summary Results of 3-D Single - Rack -- Analysis for_ Rack 4 Module: Rack-E9; Holtec Run I.D. dre9ssai.rh2 6.7.44 Summary Results of 3-D Single Rack Analysis - for Rack Module: Rack-E9; Holtec Run I.D. dre9ssei.re8 6.7.45 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-E9; Holtec Run-I.D. dre9ssoi.re2 6.7.46 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-E9; Holtec Run I.D. drc9nseo.rf8 l 6.7.47 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-E9; Holtec Run I.D. drc9ssco.rf2 6.7.46 Summary Recults of 3-D Single Rack Analysis _ for Rack Module: Rack-E9; Holtec Run I.D. dre9sseo.rh8-6.7.49 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-E9; Holtec Run I.D. dre9ssco.rh2 6.7.50 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-E9; Holtec Run I.D. drc9sseo.re8 6.7.51 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-E9; Holtec Run I.D. drc9sseo.re2 6.7.52 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-N5; Holtec Run I.D._drn5ssei.12h 6.7.53 Comparison of Calculdted and Allowable Loads / Stresses at Impact Locations and at Welds.

6.7.40 Comparison of Calculated and Allowable Loads / Stresses at Impact Locations and at Welds 6.8.1 Maximum Displacements from Whole-Pool Multi-Rack Run-6.8.2 Maximum Impact Force of Each Gap Element-6.8.3 Maximum Pedestal Stress Factors!of Each Pedestal of-Each Holtec Rack in Pool ,

6.8.4 Maximum Rack Displacements, Pedestal Vertical Loads, and I. Pedestal Stress Factor ix l

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'l 6.8.5 Results of-Pool Wall Dynamic Pressures 6.8.6 Total- Sta. tic Load and Dynamic Adder on the Whole Slab 8.5.1 Breakdown of Various Contributions to Slab Load (Wetted Area) 8.5.2 Limiting Margins for Flexure-of Slab and Walls 8.5.3 Shear Capacity Check of Slab For l1.4Dl+l1.7El -

8.5.4 Underslab Beam Safety Martin Check Axial Load + Bending 8.5.2 Bending Strength Evaluation ,

8.5.3 Shear Strength Evaluation 8.5.4 Strength Evaluation for Steel Beans 9.1 Results of ORIGEN-2 Calculations for Radionuclides of Iodine, Trypton, and Xenon at 100-Hours cooling Time 9.2 Radionuclide Properties Used in the Fuel Handling Accident Analysis 9.3 Data and Assumptions for the Evaluation of the Fuel Handling Accident s 9.4 Typical Concentration of Radjonuclides in Spent Fuel Pool Water 9.5 Preliminary Estimate of Person-rem Exposures During Re-Racking 10.1 Coupon Measurement Schedule X

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t LIST OF FIGURES

.1.1 Pilgrim Spent Fuel-Pool - Existing Rack Configuration 1.2 Pilgrim Spent Fuel Pool - Rack Configuration Campaign I 1.3 Pilgrim Spent Fuel Pool - Final Reracke'l Configuration  ;

2.1 Pilgrim Spent Fuel Pool - Rack Configuration Campaign I  ;

2.2 Pilgrim Spent Fuel Pool - Final Roracked Configuration 3.1 Seam Welding Precision Formed Channels ,

3 Sheathing Shown Installed on the Box 3.3 A Cross-Sectional View of an Array of Storage Locations 3.4 Three Cwils in Elevation View 3.5 Adjustable Support 4.2.1 Infinite of Spent Fuel in the Storage Rack 4.2.2 Relationship between K-Infinite in the Rack and in the Standard Core Geometry 4.2.3 K-Infinite Limits in the Standard Core Geometry 4.4.1 Storage Cell _ Cross Section and Calculational Model 4.9.1 Reference Design Storage Cell Geometry i

5.4.1 Spent Fuel Cooling _Model 5.7.1 Idealization of Rack Assembly l- 5.7.2 Thermal Chimney Flow Model ,

5.7.3 Convection Currents in the Pool 5.9.1 Time Af ter Reactor Shutdown, Hrs. , Normal Discharge, Two SFPC Trains - Case 1 5.9.2 Time Af ter Reactor Shutdown, Hrs. , Full Core Offload, RHR Mode 2 - Case 2 i .. _ Xi= '

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5.9.3 Time Af ter Reactor Shutdown, Hrs. , Normal ~ Discharge,- Casu -

1 5.0.4 Time After Reactor Shutdown, Hrs., Full Core offload -

Case-2 6.3.1 Synthetical SSE Acceleration Time-History _ for Fuel Pool Slab, a-tsse.h11; Duration: 20 sec.

6.3.2 Synthetical SSE Acceleration Time-History for Fuel Pool Slab, a-tsse.h12; Duration: _20 sec.

6.3.3 Synthetical SSE Acceleration Time-History for Fuel Pool Slab, a-tsse.vti; Duration: 10 sec.

6.3.4 Synthetical SSE Acceleration Time-History for Fuel Pool Slab, a-tsse.h21; Duration: 20 sec.

l 6.3.5 Synthetical SSE Acceleration Time-History for Fuel Pool Slab, a-tsse.h22; Duration: 20 sec.

p 6.3.6 Synthetical SSE Acceleration Time-History for Fuel Pool Slab, a-tsse vt2; Duration: 20 sec.

6.3.7 Synthetical SSE Acceleration Time-History for Puol Pool

i. Slab, a-tsse.h33; Duration: 20 sec.

6.3.8 Synthetical SSE Acceleration Time-History for Fuel Pool I_

Slab, a-tsse.h32;-Duration: 20 sec.

i 6.3.9 Synthetical SSE Acceleration Time-History for Fuel Pool-i Slab, a-tsse.vt3; Duration: 20 sec.

6.3.10 Synthetical SSE Acceleration Time-History for Fuel Pool Slab, a-tsse.h41; Duration: 20 sec.

6.3.11 Synthetical SSE Acceleration Time-History for Fuel Pool

! Slab, a-tsse.h42; Duration: 20 sec.

6.3.12 Synthetical SSE Acceleration Time-History for Fuel Pool Slab, a-tsse.vt41; Duration: 20 sec.

6.3.13 The Responso Spectrum of Safe Shutdown Earthquake and the Average Response Spectrum of 4 Artificial SSE. '

Acceleration Time-Histories: a-tsse.h11, a-tsse.h21, a-tsse.h31, a-=tsse.h41; Damping: 2 percent: Duration:'20-seC.

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t 6.3.14 The Response Spectrum of Safe Shutdown Earthquake and the Average Response Spectrum of 4 Artificial SSE i

~

Acceleration Time-Histories a-tsse.h12, a-tssa.h22, a-tase.h12, a-=tsse.h42; Damping: 2 percent: Duration: 20 d

sec.

6.3.15 The Response Spectrum of Safe Shutdown Earthquake and the ,

Average Response . Spectrum of 4 Artificial SSE Acceleration Time-Histories a-tsse.vti, a-tsse.vt2, a-tase.vt3, a-atssa.vt4; Damping: 2 percent: Duration: 20 sec.

4 6.3.16 Synthetical OBE Acceleration Tile-History for Pool Slab, a-tsse.hll, Duration: 20 sec. -

6.3.17 Synthetical OBE Acceleration Time-History for Pool Slab, a-tsse.h12, Duration: 20 sec.

6.3.18 Synthetical OBE Acceleration Time-History for Pool Slab, I a-tsse.vti, Duration: 20 sec.

6.3.19 Synthetical OBE Acceleration Time-History for Pool Slab, I a-tsse.h21, Duration: 20 sec.

6.3.20 Synthetical OBE Acceleration Time-History for Pool Slab, a-tsse.h22, Duration: 20 sec.

6.3.21 Synthetical OBE Acceleration Time-History for Pool Slab, a-tsse.vt2, Duration: 20 sec.

6.3.22 Synthetical OBE Acceleration Time-History for Pool Slab, a-tsse.h31, Duration: 20 sec.

6.3.23 Synthetical OBE Acceleration Time-History for Pool' Slab, .

a-tsse.h32, Duration: 20 sec.

6.3.24 Synthetical OBE Acceleration Time-History for Pool Slab, a-tsse vt3, Duration: 20-sec.

6.3.25 Synthetical OBE Acceleration Time-History for Pool Slab, a-tase.h41, Duration: 20 sec.

1 6,c.26 Synthetical OBE Acceleration Time-History for Pool Slab, ,

a-tsse.h42, Duration: 20 sec.

6.3.27 Synthetical OBE Acceleration Time-History for Pool Slab, a-tsse.vt4, Duration: 20 sec.

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m _ _ - , - . - _ , _ . . _ . , - _ . _ _ , _ . . . -

4 6.3.28 The Response Spectrum of Operating Basis Earthquake and the Average- Response Spectrum of 4 Artificial ODE Accoloration Time-Historiest a-tebo.h11, a-tobo.h21, a-tobo.h31, a-tobe.h41; Damping: 1 porcent; Duration: 20 sec.

6.3.29 The Responso Spectrum of operating Basis Earthquake and the Avorage Response Spectrum of 4 Artificial ODE Accoloration ?imo-Histories: a-tobo.h12, a-tobe.h22, a-tobo.h32, a 'obo.h42; Damping: 1 percont; Duration: 20 sec.

6.3.30 The Responso Spectrum of Operating Basis Earthquake and the Average Rosponso Spectrum of 4 Artificial OBE Accoloration Time-Histories a-tobo.vci, a-t;bo.vt2, a-tobe.vt3, a-tobo.vt4; Damping: 1 percont; Duration: 20 soc. ,

6.4.1 Pictorial View of Rack Structure 6.4.2 Schematic Model for Dynarack 6.4.3 Rack-to-Rack Impact Springs 6.4.4 Fuel-to-Rack Impact Springs 6.4.5 Dogroos-of-Froodom Modoling Rack Hotion 6.4.6 Rack Degroo-of-Froodom for Y-Z Plano Banding 6 4.7 Rack Degrou-of-Froodom for X-Z Plano Bonding 6.4.8 2-D View of Rack Hodulo 6.4.9 Modulo and Podostal Numbering System for WPMR Analysis 6.8.1 Gap Timo History; Gap betwoon Rack-1 and Rack-5, West corner, Top, from WPMR Analysis; dwposo2.rfr; 1.15xSSE-2; 680/ reg. fuel;-full; Friction coefficient = random (mean = 0.5), File:gl-5r.dat.

6.8.2 Gap Timo History; Gap betwoon Rack-5 and Rack-6, South Corner, Top, from WPMR Analysis; dwpsso2.rfr; 1.15xSSE-2; 680/ reg. fuel; full; Friction coefficient = random

-(mean = 0.5), Filo:g5-6u.dat.

6.8.3 Gap Timo History; Gap betwoon Rack-5 and Rack-6, North Cornor, Top, from WPMR Analysis; dwpsse2.rfr; 1.15xSSE-2; 680/ reg. fuel; full; Friction coefficient = random (moan = 0.5), Filo:g5-61.dat.

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. . , . . , - . . - - .,.m - . . ._...__,. - _ _ , _ ~ - - - - - - _ _ . _ , - , _ . - _ . - - . _ - - _- . . _ ~ - - - .

f 6.8.4 Gap Time History; Gap between Rack-5 and Rack-9, East l Corner, Top, from WPMR Analysis; dwpsse2.rfr; 1.15xSSE- ,

2; 680# reg. fuelt-full; Friction coefficient = random (mean = 0.5), File g5-91.dat.

6.8.5 Gap Time History; Gap between Rack-5 and Rack-9, West Corner, Top, from WPMR Analysis; dwpase2.rfr; 1.15xSSE-2; 680/ reg. fuel; full; Priction coefficient = random (mean = 0.5), File g5-9r.dat.

6.8.6 Gap Time History; Gap between Rack-10 and Rack-11, South Corner, Top, from WPMR Analysis; dwpsse2.rfr; 1.1SxSSE-  !

2; 680/ reg. fuel; full; Friction coefficient = random (mean = 0.5), Filerg10-11u.dat.

6.8.7 Total Slab Vertical Load Time-History 7.4.1 Welded Joint in Rack - -

8.2.1 Exterior of Pool Looking from Underneath and West 8.2.2 Interior of Pool Looking from East 1

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1.0 INTRODUCTION

The Pilgrim Huclear Power Station (PNPS or Pilgrim) consists of ,

one Boiling Water Reactor (670 MWe not output) supplied by the General Electric Company of San Jose, California. The PNPS reactor i features a core consisting of 580 fuel assemblies. The plant is located on the shore of the Cape Code Bay at a distance of approximately 35 miles southeast of the city of Boston, ,

Massachusetts. The site is owned and operated by the Boston Edison >

Company, henceforth also referred to as the Ovner or Licensee.  !

Pilgrim received its operating license on June 0, 1972 and began commercial operation in December,1972. The Pilgrim spent fuel pool ,

is of 366" x 484" (nominal) planform section with an integral cask laydown area. The pool is presently licensed for 2320 storage cells which were installed in the pool in the manner shown in Figure 1.1 in 1985. The existing racks are of cruciform based honeycomb construction (1.1] and employ the Boraflex neutron absorber. All rack modules are free-standing. As can be seen by referring to Figure 1.1, there is a considerabic amount of open (unracked) space in the Pilgrim pool at the present time.

The open space available in the Pilgrim pool can be utilized for spent fuel storage by adding six additional storage racks 1

containing a total of 1526 storage cells. This will increase the cumulative storage capacity of the Pilgrim pool to 3859 cells and, as shown in Table 1.1, will extend the full core reserve capacity to the end of licensed life (year 2012).

• While the present OL amendment is for adding six racks containing a total of 1526 cells, at the present time only two racks, labelled .

as N1 and N2 in Figure 1.2, will be installed in the fuel pool, limiting the immediate increase in the storage capacity to 558 cella. This interim increase will enable BECo to operate Pilgrim with full core reserve margin until 2003. <

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Additional modulos, identical in design to those being procured now, will be manufactured and installed at a suitable timo before loss-of-full core reserve. Figure 1.3 illustrates the storage configuration in the Pilgrim poo* With all racks installed.

The now racks for Pilgrim havo been designed with provisions to install an overhead storago platform which servos the dual purpose of protecting the fuel assemblies stored underneath and to store miscellaneous objects (such as LPRMs) which will no longer be stored on the pool floor. The safety evaluation for thermal-hydraulics, criticality, seismic and mechanical accident scenarios considor the rack modules with and without the overhead platforms.

The Docton Edison Company has engaged floltoc International of Cherry 11111, New Jersey, to design, engineer, f abricate and deliver the racks on a turnkey basis. lloltoc International has supplied racks to every BWR project in the U.S., Mexico, and Asia in the past eight years.

The now spent fuel storage racks are froo-standing and self supporting. The principal construction materials for the now racks are ASME SA240-Type 304L stainless steel sheet and plate stock, and SA564 (precipitation hardened stainless stcol) for the adjustable support spindles. The only non-stainless material utilized in the rack in the neutron absorber material which is a boron carbido aluminum cermet manuf actured under a U.S. patent and sold under the brand namo Beral" by AAR Advanced Structures, Livonia, Michigan.

From the geometric and anatomical standpoint, the now racks are identical to the existing (old) racks. For example, both the new and old racks feature one composite panel (consisting of the neutron absorber sandwiched (betwoon two stainless steel sheets) betwoon adjacent fuel assemblics. A continuous common basoplate with a chamfered hole (for so.iting the "noso" of the fuel assembly) defines the lower extremity of the cellular region in both sets of 1-2

4I -

r racks. Finally, threaded support pedestals, arranged to - be concentric with the storage cell above them, are engineered to level the rack modules remotely in both sets of modules. However, as mentioned earlier, the longitudinal connection between the adjacent cells is effected by t5c fillet welds between the cruciforms (1.1) in the old design. The cruciform fabrication elements, however, suffer from excessive veld induced distortion which is avoided in the detuned honeycomb fabrication method utilized in the new Pilgrim, and in all Holtec maximum density racks of recent vintage. Details of the construction of the new racks are provided in Section 3 of this licensing report.

The new racks are designed and analyzed in accordance with Section III, Division 1, Subsection NF of - the ASME Boiler and Pressure Vessel Code. The t.aterial procurement and fabrication of the rack i modules conforms to 10CFR50 Appendix B requirements.

This Licensing Report documents the design and analyses performed to demonstrate that the new spent fuel racks and the existing racks satisfy all governing requirements of the applicable codes and  ;

standards, in particular, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", US!TRC (1978) and 1979 Addendum thereto.

The safety assessment of the proposed rack modules involves demonstration of its thermal-hydraulic, criticality and structural adequacy. Thermal-hydraulic 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. Demonstration of structural adequacy primarily involves analysis showing that the free-standing modules will not impact under the postulated seismic events, and that the primary stresses in the module structure will remain below the ASME Code allowables. Similar analyses for the existing racks will also be presented. The structural qualification includes 1-3

analytical demonstration that the suberiticality of the stored fuel will be maintained under accident scenarios such as fuel assembly drop, accidental misplacement of fuel outside a rack, etc.

The criticality safety analyses presented in Section 4-of this report shows that the neutron multiplication factor for the stored  ;

fuel array is bounded by the USNRC limit of 0.95'under assumptions of 95% probability and 95% confidence for both new and existing racks. Consequences of the inadvertent placement of a fuel assembly are also evaluated asi part of the criticality analyses. The criticality analysis also sets the requirements on the length of ,

the B-10 screen and the areal B-10.. density.

This Licensing Report contains documentation of the analyses performed to demonstrate the large margins of safety with respect to all USNRC specified criteria.

The analyses presented herein clearly demonstrate that the rack module arrays possess wide margins of safety from all key -

thermal-hydraulic, criticality, structural, and radiological -

vantage points. The No Significant Hazard consideration evaluation presented along with this document is based on the descriptions and 4 analyses synopsized in the subsequent sections of this report.

This document has been prepared for submission to the U.S. Nuclear tregulatory Commission for securing regulatory approval of the modification of the Pilgrim pool as proposed herein.

1.1 References

[1.1) M. Holtz and K.P. Singh, U.S. Patent No. 4,382,060,.

Radioactive Fuel Cell Storage Rack, May 3, 1983.

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Table 1.1 PROJECTED AND EXISTING DISCHARGE SCHEDULE Total l Number of Number of 1 Assemblies Assemblies-Cycle Shutdown Discharged Stored Number Date to the Pool In the Pool 1 12-28-73 20 20 2 01-29-76 132 152 3 08-06-77 428 580 4 01-04-80 92 672 5 09-25-81 232 904 6 12-09-83 224 1128 7 07-24-86 192 1320 8 05-04-91 169 1489 9 04-03-93 140 1629 10 04-01-95 168 1797 11 04-01-97 160 1957 12 04-01-99 160 2117 13 04-01-01 164 2281 14 04-01-03 160 2441 15 04-01-05 160 2601 16 04-01-07 164 2765 17 04-01-09 160 2925 18 04-01-11 164 3089 Note:- Discharges are actual- through Cycle - 8 and projections thereafter.

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i 2.0 MODULE LAYOUT FOR INCREASED STORAGE ,

This section provides general information on the new storage modules for the Pilgrim spent fuel pool. The information presented in this and the next section provide the basis for the detailed criticality, thermal-hydraulic and seismic analyses presented in the subsequent sections of this report.

As stated in the preceding section, the existingt racks in the PHPS [

pool are of cruciform honeycomb construction. These racks employed the Boraflex neutron absorber material. Nine free-standing modules comprising a total of 2333 cells at' nominal intermodule gap of-two inches are currently installed and in use in the PHPS pool (please see Table 2.1).

Although the proposed racking application is for the addition of a total of six modules, the ultimate capacity will be realized in two or more rack addition campaigns. In the present campaign two modules, denoted as N1 and N2 in Figure 2.1, will be added.

Figure 2.2 shows the fully racked PNPS pool with reference inter-rack and rack-to-wall gap dimensions. Table 2.2 providos an abstract of all new racks to be added in the present campaign (Campaign I) and future campaigns (labelled as II in Table 2.2).

The weight and dimensional data on all modules, new and old, are provided in Table 2.3. Key data on the construction of all new fuel racks as well as the existing rac.r.s is provided in Table 2.4.

2.1 Heavy Load Considerationn for the Pronosed Rerackina Ooeration The Reactor Building bridge crane features a 100 ton main hook and a five ton auxiliary hook. The crane was originally designed per the provisions of EOCI-61, " Specifications for Electric Overhead Traveling Cranes - 1961". Subsequently in 1984, BEco submitted 2-1 i

l documentation to the USNRC demonstrating that the crane is in substantial compliance with the provisions of NUREG-0612, Section 5.1.1, Guideline 7. The crane, however, is not " single failure proof".

A remotely engageable lift rig, meeting NUREG-0612 stress criteria, will be used to lift the new empty modules. The rig designed for handling the PNPS racks is identical in its physical attributes to the rigs utilized to rarack Millstone Point Unit one (1988), Vogtle Unit Two (1989), Indian Point Unit Two (1990), Ulchin Unit Two (1990), Hope Crcsk (1990), Laguna Verde Unit One (1990), Kuonheng (1991), Three Mile Island Unit 1 -(1992), Zion (1993), and J.A.

Fit: Patrick (1992). The rig consists of independently loaded lift rods with a "can type" lift configuration which ensures that failure of one traction rod will not result in uncontrolled lowering of the load being carried by the rig (which complies with the duality f eature called f or 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 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 is carried out using a finite element code, and the primary stress limits postulated in ANSI N14.6 (1978) are shown to be met.

d. The rig is loud tested with 150% of the maximum weight to be lifted. The test weight is maintained in the air for one hour. . 11 critical wold joints are liquid penetrant examined, after the load test, to establish the soundness of all critical joints.

1 2-2

i i

i Pursuant to the defense-in-depth approach of NUREG-0612, the-  :

following additional measures of safety will be undertaken for the reracking operation.

(1) The cranes used in the project receives preventive f' maintenance checkup and inspection per the PNPS maintenance procedures.

(ii) The crancs used will lif t no more than 50% of their-rated capacity at any time during the raracking operation.

(iii) Safe load patns will be developed for moving the new racks. The racks will not be carried over any  ;

region of the pool containing active fuel.

(iv) The rack upending orclaying down will be carried out in an area which is not overlapping to any safety related component.

(v) All crew members involved in the raracking operation will be given training in the use of the lif ting and upending equipment.

The installation of the new racks in the spent fuel pool is predicated on the following key 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 lif ted in such a manner that the C.G.

of the lif t point is aligned with the C.G.. of the load being lifted.

(3) Turnbuckles are utilized to " fine tune" the verticality of the rack prior to lifting.

All phases of the raracking activity will be conducted in accordance with written procedures which comply with BEco's heavy load handling procedures. All contractor developed procedures will be reviewed and approved by the. Boston Edison Company prior to their use.

2-3

- . _ __- . . . _ _ _ __. _u__-

Our proposed compliance with the objectives of NUREG-0612 follows the guidelines contained in section 5 of that document. The guidelines of NUREC-0612 call for measures to " provide an adeguate defense-in-depth for handling of heavy loads near spent fuel...".

The IMREG-0612 guidelines cite four major causes of load handling accidents, namely (1) operator errors (ii) rigging failure (iii) lack of adequate inspection (iv) inadequate procedures The PNPS rarack 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 aforamintioned areas. A summary of the measures specifically planned to deal with the major causes is provided below, ooerator errors: As mentioned above, Boston Edison Company plans to provide comprehensive training to the installation crew.

Ricaina failurg: The lifting device designed for handling and installation of the racks in the PNPS 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 crite.ria, load testing at 150%

of maximum lift load, and dye examination of critical welds.

The PNPS rig design is similar to the rigs used in the rarack of numerous other plants, such as Hope Creek, Millstone Unit 1, Indian Point Unit 2, Ulchin II, Laguna Verde, J.A. FitzPatrick and Three Mile Island Unit 1.

Lack of adeauate ingggetion: 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 rarack projects.

Inadeaunte crocedures: PNPS plans to develop procedures to cover the entire series of operations pertaining to the racking effort, such as mobilization, rack handling, upending, lifting, installation, verticality, alignment, dummy gage testing, site safety, and ALARA compliance.

2-4

The series of operating procedures planned for PNPS rerack are the successor of the procedures implemented successfully in other projects in the past. .

In addition to the above, safe load paths shall be developed as required by NUREG-0612.

Table 2.5 provides a synopsis of the requirements delineated in ,

NUREG-0612 and our intended compliance.

In summary, the measures implemented in PNPS raracking are identical to the those utilized in the most recent successful rarack 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).

Finally, the movement of the overhead platform over the fuel racks will be carried out, when necessary, using a similarly-rigorous handling procedure. The overhead platform is approximately 7 feet by 9 feet in planform section but weighs only 1500 lbs. . As a result, its momentum at impact from a 36 inch height is less than that from a standard channelled BWR fuel assembly. Nevertheless,

.the overhead platform is equipped with redundant lift points with a margin of safety of over ten (at each lift location) to ensure full compliance with NUREG-0612 requirements.

I t

2-5

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Table 2.5 HEAVY LOAD HANDLING COMPLIANCE MATRIX (NUREG-0612) ,

criterion como11ance

1. Will safe load paths be defined for Yes  ;

the movement of. heavy loads to 1 minimize the potential of impiact, if dropped on irradiated fuel?

2. Will procedures be developed-to Yes cover: identification of required equipment, inspection and acceptance criteria required before movement 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 guidulines of ANSI N14.6-19787

5. Will non-custot lifting devices Yes be installed and used in accordance with ANSI B30.9-19717

6. Does the crane meet the intent of Yes ANSI B30.2-1976 and CMMA-70?

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TABLE 2.1 EXISTING RACKS IN TILE PILGRIll P0OL l

l h000LE I D. No. DF CELLS No. OF N000LES THIS TYPE h{jb-fffg{gj[bl{ypg l

i E4 THRU E9 19X14=266 6 1596 El 19X14-( 13X14 )=214 1 214 E2 19X14-( 9X4 )=230 1 230 E3 19X14 4( 3X9 )=293 1 293 IDIAL EXISTING CELL CDUNT 2333

t TABLE 2.2 DATA DN NElf RACK AIODULES PRESENT k FUTURE Calf?AIGNS hh 7f]g Q GL{}pp h000LE I.0. No. OF CELLS No. DF ff000LES CAfrAIGN I.D.

I 288 I N1 288 1 I N2 270 1 I 270 l

tO 266 i II 266 N4 & tS 247 2 II 494 NS 208 1 II 208 l

1 DIAL GLL COUNT 1526

TABLE 2.3 EXISTING AND NEW RACK MODULES DIMENS!0HAL & HEIGHT DATA l

DIMENSI[N (IN. ) No. or i WEIGHT CELLS f N-S E4 b ID* PER 012EW[N DIPfW[N- lbs.) (FJPTY RAEK [1 u

Ni 113 100.5 29400 .288 N2 !I3 94.25- 28600 270 N3 119.25 88 27100 266 N4 119.25 81.5 25200 247 N5 119.25 81.5 25200 24 7 N6 100.5 81.5 21300 200 El 120 88.5 23600 214 E2 120 88.5 25200 239

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E3 138.75 08.5 31700 293 E4 120 86.5 29000 266 E5 120 88.5 29000 266 E6 120 08.5 29000 266 E7 120 88.5 29000 266 E8 120 88.5 29000 266 E9 120 88.5 29000 .266

"*" INDICATES DIENSIONS ARE ROUNDED TO THE NEAREST 1/4".-

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Table 2.4 DESIGN DATA FOR NEW AND EXISTING RACKS ITEM NEW RACKS EXISTING RACKS I.D. (inside 6.05" 6.05" dimension)

Cell nominal pitch G.257" 6.243" Poison matarial Soral Boraflex Poison Loading (min.

in B-10 por sq. em.) 0.015 .02 Poison plate (nom.)

vidth: 5" 5.56" Poison material thickness, inch: .075" .06H" Poison picture frame 5.125" x 144-1/2" 5.625" x 136.5" (bounding) size Poison length: 144" 136.25" Cell height: 165.375" 165.375" Baseplate thickness: 3/4" 0.625" Bottom plenum height: 6.25" (nom.) 7.25" Number of supports per Four (minimum) Four module:

Support type: Remotely Remotely adjustable adjustable

I Table 2.5 HEAVY LOAD HANDLINr. COMPLIANCE MATRIX (NUREG-0612)

Criterion Comoliance

1. Will safe load paths be defin'ed 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 covers identification of required >

equipment, inspection and acceptance criteria required before movement

• of load, steps and propar sequence for handling the load, defining the safe load paths, and special ,

preca.*tions?

3. Fill et/ane operators be trained Yes and qualified?

4. Will special lifting devices meet -'

Yes the guidelines of ANSI N14.6-19787

5. Will non-custom lifting devices Yes be installed and used in accordance '

with ANSI B30.9-19717

6. Does the crane meet the intent of Yes ANSI B30.2-1976 and CMMA-707 4

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3.0 RACK FABRICATION AND APPLICABLE CODES The purpose of this section is to provide a comprehensive resume of the concepts and features v.nderlying the design of the rack modules for the Pilgrim Nuclear Power Station spent fuel pool.

3.1 Deslan Obiective The central objective governing the design of the new high density storage racks for the PNPS fuel pool is defined in the following six criteria:

(i) The rack module is fabricated in such a manner that there is n2 weld splatter on the storage cell surfaces which would come in contact with the fuel assembly. Wald splatter on the lateral surface of the storage cell, .

which can come in contact with fuel assemblies, can be detrimental to its structural integrity.

(ii) The storage locations are designed and construct:ed in such a way that redundant flow paths for the coolant are available in case the main designated flow path is blocked.

(iii) The fabrication process based on the rack design involves operational sequences which permit immediate and convenient verification by the inspection staff to ensure that the " poison" panels are correctly installed.

(iv) The storage cells are connectea to each other by autogenously produced corner welds which leads to ' a boneycomb lattice construction. The configuration of welding is designed to "detune" the racks from the ground motion such that the rack displacements are minimized.

(v) The baseplate provides a conformal contact surface for the " nose" of the fuel assembly.

(vi) The module design affords built-in flexibility in' the fabrication process so as to maintain the desired cell pitch even if certain " boxes" are slightly oversize.

The foregoing objectives are fully realized in the module design for the PNPS racks as described in the following.

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3.2 Anatomy of the Rack Module The new high density rack module design employs storage cell locations with a single poison panel sandwiched between adjacent austenitic steel surfaces.

A complete description of the rack geometry is best presented by first introducing its constituent parts. The principal parts are denoted ass (1) the storage box subassembly (2) the baseplate (3) the neutron absorber material, (4) picture frame sheathing, and (5) support legs.

Each part is briefly described below with the aid of sketches.

(1) Storage cell box subassembly: The so-called " boxes" are fabricated from two precision formed channels by seam welding them together in a seam welding machine equipped with copper chill bars, and pneumatic clamps to minimize distortion due to welding heat input. Figure 3.1 shows the " box".

The minimum weld penetration is 80% of the box metal gage which is 16 gage. The boxes are manufactured to 6.05" nominal I.D. I,inside dimension).

As shown in Figure 3.1, each box has two lateral holes punched near its bottom edge to provide auxiliary flow holes. In the next step, a picture frame sheathing is-press formed in a precision die. The sheathing is shown attached to the box in Figure 3.2. The sheathing is made to an offset of 0.077" to ensure an unconstrained installation of Boral* plates. The " picture frame heathing" poison is attached material (Boral g)each side ofin installed thethe boxsheathing with the cavity. The top of the sheathing is connected using a smooth continuous fillet veld near the top of the box.

The edges of the sheathing and the box are welded together to form a smooth lead-in edge. The box.with integrally connected sheathing is referred to as the

" Composita box".

3-2

The "composito boxes" are arranged in a checkerboard array to form an assemblage of storage call locations (Figure 3.3) . The inter-box welding and pitch adjustment is accomplished by small longitudinal austonitic stainless cornuctors shown as small circles in Figure 3.3.

An assemblage of box assemblies thus prepared is volded edge to edge as shown in Figure 3.3 resulting in a honeycomb structure with axial, flexural and torsional rigidity depending on the extent of intercoll wolding provided. Figuro 3.3 shows that two edges of each jnterior box are connected to the contiguous boxes resulting in a well defined path for " shear flow".

(2) Basoplato: The basoplato, 3/4 inch thick, providos a continuous horizontal surface for supporting the fuel angomblics. The basoplato has a concentric hole with a 45 taper in each cell location to provido a saating surface conforming to the fuel assembly.

The basoplate is attached to the call assemblage by fillet volds.

(3) The tho

- neutron neutron absorbor absorber material:

material. Doral foral" is actured is manuf used as by AAR Brooks and Perkins of Livonia, Michigan. Horo on this material follows in the next section.

(4) Picture Frama Shoathing: The sheathing is a part of the

" composite box assembly" described earlier. The sheathing serves as the locater and retainer of the poison material. As such, it is made in repeatable preciso dimensions. This is accomplished by press-forming stainless sheet stock in a specially high toleranco die.

The schematic of the sheathing is shown in Figure 3.2.

Figure 3.4 shows three storage cells in olevation with the fuel assembly shown in phantom in one coll. The poison screen extends over 144" vertical distance, straddl'n.g the acti"o fuel length of all fuel assemblics to be used in the PNPS reactor.

(5) Support Logs: As stated earlier, all support logs are the adjustable type (Figuro 3.5). The top (female) position is made of austenitic stool material. The bottom part is made of 17:4 Ph series stainless stool to avoid galling.

Tho support log is equipped with a socket to enable remote levoling of the rack after its placement in the pool.

3-3

The baseplate projects the cellular region of the rack modules by 0.125 inch (nominal).

3.3 Material considerations 3.3.1 Introduction >

r safe storage of nuclear fuel in the PNPS 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.

3.3.2 Structural Materials The following structural materials are utilized in the fabrication of the spent fuel racks:

a. ASME SA240-304L for all sheet metal stock,

b. Internally threaded support legs: ASME SA240-304L.

c. Externally threaded support spindle: ASME SA564-630 precipitation hardened stainless steel (heat treated to 1100*F).

d. Weld material - per the following ASME specification: SFA 5.9 R308L.

3.3.3 Poison Material In addition to the structural and non-structural stainless material, the racks employ Boral*, a patented product of AAR l Brooks and Perkins, as the neutron absorber material. A brief description of Boral, and its fuel pool experience list follows.

l l

l l

3-4 -'

L

1 Boral is a thermal neutron poison materjal 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 form. The 1100 alloy aluminum is a light-weight metal with high tensile strength which is protected - from corrosion by a highly resistant oxide film. The two materials, boron carbide and aluminum, are chemically compatible and ideally suited fort long-term use in the radiation, thermal and chemical environment of a nuclear reactor or the spent fuel pool.

Boral's use in the spent fuel pools as the neutron absorbing material can be attributed to the following reasons:

(i) The content and placeLent 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 of the Boral panels.

(iii) The boron carbide and aluminum materials in Boral are totally unaffected by long-term exposure to radiation.

(iv) The neutron absorbing central layer of Boral is clad with permanently bonded surfaces of aluminum.

(v) Boral is stable, strong, durable, and corrosion recittant.

Holtec International's Q.A. program ensures that Boral is manufactured by AAR Brooks & Perkins under the control and surveillance cf a Quality Assurance / Quality Control Program that conforms to - the requirements of 10CFRSO Appendix B, " Quality Assurance Criteria for Nuclear Power Plants".

As indicated in Table 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.

3-5

Boral Material Characteristics 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 hu atomic number of.13, atomic weight of 26.08, specific gravity of 2.69 and valence of 3. The physical, mechanical and chemical properties of the 1100 alloy aluminum are listed in Tables 3.2 and 3.3.

The excellent corrosion resistance of the 1100 alloy aluminum is provided by the protective oxide film that develops on its surface from exposure to the atmosphere or water. This film prevents the loss of metal from general corrosion or pitting corrosion and the film remains stable between a pH range of 4.5 to 8.5.

Boron Carbide: The boron carbide contained in Boral is a fine 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 material conforms to the chemical composition and properties listed in Table 3.4.

3.3.4 Comcatibility with Coolant All materials used in the construction of the PNPS racks have an established history of in-pool usage. Their physical, chemical and radiological compatibility with the pool environment is an established fact at this time. As noted in Table 3.1, Boral has been used in both vented and unvented configurations in fuel pools with equal success. Austenitic stainless steel is perhaps the most widely used stainless alloy in nuclear power plants.-

3-6

3.4; C: doc, Stcndards,--cnd Practicso for th3 Sp nt Fuol Pool findifiention The ' f abrication . of the rack medules is performed under a strict quality assurance program which meets 10CFRSO Appendix 'B requirements.

The following codes, standards and practices were used for all applicable aspects of the design, construction, and assembly of the spent fuel storage racks. Additional specific references related to detailed analyses'are given in-cach section,

a. Desian Codes (1) AISC Manual of Steel Construction, 8th Edition,1980

. (provides detailed structural criteria for linear type supports).

(2) ANSI N210-197G, " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations" (contains guidelines for fuel rack design).

(3) American Society of Mechanical Engineers. (ASME) ,-

Boiler and Pressure Vessel Code, Section- III,_

Division 1, 1986 Edition.

(4) ANSI /AISC-N690-1984 -

Nuc1 car Facilities - Steel Safety Related Structure for Design, Fabrication and Erection.

(5) ASNT-TC-1A,1984 American Society for Nondestructive Testing (Recommanded Practice for Personnel-Qualifications).

(6) ACI 349 Codst Requirements-for Nuclear Safety.

Related Concrete Structures.

(7) ACI 318-63 -

Building Code Requirements for Reinforced Concrete,

b. Material Codes - Standards of ASME'or ASTM. AS NOTED:

(1) E165 - Standard Methods for Liquid Penetrant Inspection.

(2) SA240 - Standard Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless ' Steel Plate, Sheet and Strip-for Fusion-Welded Unfired Pressure Vessels.

3-7

(3)~--A262 -

Detecting Susceptibility to .Intergranular Attack in Austenitic Stainless Steel.

(4) SA276 -

Standard Specification for Stainless and Heat-Resisting Steel Bars and Shapes.

(5) SA479 - Steel Bars-for Boilers & Pressure Vessels.

(6) C750 - Standard Specification for Nuclear-Grade. ,

Boron Carbide Powder.

(7) C992 -

Standard Specification for Baron-Based Neutron Absorbing Material Systems. for Use in Nuclear Spent Fuel Storage Racks.

(8) SA312 - Specification for Seamless and ' Welded Austenitic Stainless Steel Pipe.

(9) SAS64 -

Specification for Hot Rolled and Cold-Finished Age-Hardening Stainless and Heat Resisting Steel Bars and Shapes.-

(10) American Society of Mechanical Engineers - (ASME) ,

Boiler and Pressure Vessel Code, Section II-Parts' A and C, 1986 Edition.

(11) ASTM A262 Practices A and E - Standard Recommended Practices for Detecting -Susceptibility to Intergrannular Attack in Stainless Steels.

(12) ASTM A380 -

Recommended Practice for Descaling, Cleaning and Marking Stainless Steel Parts ar.d Equipment.

c. Weldina Codes (1) ASME Boiler and Pressure Vessel Code,Section IX -

Welding and Brazing Qualifications, 1986 Edition.

(2) AWS D1.1 - Welding Standards (1989),

d. ouality Assurance, Cleanliness, Packacine, Shineine, Receivina. Storace, and-Handline Recuirements (1) NQA-2-Part 2.2 1983 _- Packaging, Shipping, Receiving, Storage, and Handling of Items for Nuclear Power Plants (During Construction Phase) .

(2) NQA-1-1983 - Basic Requirements and Supplements.

3-8 i

(3) ASME -Boiler and Pressure -Vessel, Section V, Nondestructive Examination, 1986 Edition.

(4) ANSI -

N45.2.6 -

Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants (Regulatory Guide 1.58).

e. Governina NRC Desinn Documents (1) "OT Position for Review and Acceptance of Spent Fuel Storage and Handlitig Applications," dated April 14, '

1978, and the modifications to this document of January 18, 1979.

(2) NRC Standard Review Plan Rev. 3, 1981, NUREG 0800.

- 9.1.2, Spent Fuel Storage.

(3) NRC Standard Review Plan, Rev. 2, July 1981 NUREG 0800 - 9.1.1, New Fuel Storage.

(4) NRC Standard Review Plan, Rev. 1, July 1981, NUREG 0800 - 3.8.4, Other Seismic Cate ;ry I Structures.

(5) NRC Standard Review Plan, Rev. 1, July 1981, NUREG 0800 - 3.8.5, Foundations.

f. Other ANSI Standards (not listed in the'erecedina)

(1) ANSI /ASN 8.1 - 1983, Nuclear Criticality Safety in' Operations with Fissionable Materials outside Reactors.

(2) ANSI /ASN 8.7 - 1974, Guide for Nuclear Criticality Safety in the Storage of Fissile Materials.

(3) ANSI /ANS 8.11 -

1975, Validation of Calculation Methods for Nuclear Criticality Safety.

g. Code-of-Federal Reculations (1) 10CFR21 - Reporting of Defects and Non-compliance.

(2) 10CFR50 - Appendix A - General Design Criteria for Nuclear Power Plante.

(3) 10CFR50 - Appendix B - Quality Assurance Criteria for Nuclear Power Plants and -Fuel Reprocessing Plants.

(4) 10CFR20 - Radiation Protection Standards.

3-9

4 (5) 29CFR' Section 1910.401 -

OSHA Standards for Commercial. Diving Operations,

b. Reculatory Guides (1)- RG 1.13 - Spent-Fuel' Storage Facility Design Basis.

(2) RC 1.25 -

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 - (endorses ANSI N45.2) - Quality Assurance Program Requirements, .7une, 1972.

(4) RG 1.29 - Seismic Design Classification.

(5) RG 2. 38 - (endorses ANSI N45.2.2) Quality Assurance

Regt.rements 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 - (endorses ANSI N45.2.6) Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel. Rev. 1, September, 1980.

(8) RG 1.64 - (endorses ANSI N45.2.11) Quality Assurance Requirements for the Design of Nuclear Power Plants, October, 1973.

(9) RG 1.74 - (endorses ANSI N45.2.10) Quality Assurance-Terms and Definitions,-February, 1974.

(10) RG 1.88 - (endorses ANSI N45.2.9) Collection, Storage' and Maintenance of Nuclear Power Plant Quality Assurance Records. Rev. 2, October, 1976.

(11) RG 1.92 - Combining Modal Responses and Spatial Components in Seismic Response Analysis.

(12) RG 1.'123 - (endorses ANSI N45.2.13) Quality-Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants.

(13) NRC Regulatory Guide 3.41 R e v . ,- May 1977 -

Validation of Calculation Methods for Nuclear- ,

Criticality Safety.

3-10

(14) NRC Regulatory Guide 1.26 Rev. 3, Feb.1976, Quality .

Group Classifications and Standards for Water, Steam and Radioactive Containing Components of Nuclear Power Plants.

i. Dranch Technical Position (1) CPS 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.

4.

(4) SRP 3.8.4 -

Other Seismic Category I Structures (including Appendix D).

(5) SRP 9.1.2 - Spent Fuel Storage.

(6) SRP 9.1.3 -

Spent Fuel Pool Cooling and Cleanup System.

k. Other Pilgrim Nuclear Power Station Final Safety Analysis Report (FSAR).

3-11

Table 3.1.

BORAL EXPERIENCE LIST (Domestic and Foreign)

Dressurized Water Reactors -

Vented Construc- Mfg.

Plant Utility tien Year Bellefonte 1,2 Tennessee Valley Authority No 1981 Donald C. Cook Indiana & hichigan: Electric No - 1979 Indian Point 3 NY Power Authority Yes 1987 Maine Yankee Maine Yankee Atomic 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 1964/1983 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 i 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 Dresden 2,3 Commonwealth Edison Co. Yes 1981 Duane Arnold Iowa Elec. Light & Power- No - 1979 J.A. FitzPatrick 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

9 Table.3.1 (continued)'

Foreign Installations Using Boral ~

France 12 PWR Plants Electricite.de France.

South Africa r

Koeberg 1,2 ESCOM-t Switzerland Ber.nau 1,2 Nordostschweizerische'Kraftverke AG Gosgen Kernkraftwerk Gosgen-Daniken.AG=

Taiwan Chin-Shan:-1,2 Taiwan-LPower Company Kuosheng 1,2 Taiwan Power. Company Mexico --

Laguna verde +

Comision' Federal =de Electricidad-Units.1 &.2 7

. ., , .- . , , , , .,: .e , -.. . . -

~. . .:,-

I-Table 3.2 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 dag. F) 0.53 cal /sec/sq cm/deg. C/cm Coef. of Thermel 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 10x108 psi Elasticity Tensile Strength 13,000 psi annealed (75 deg. F) 18,000 psi as rolled Yield Strength 5,000 psi annealed (75 deg. F) 17,000 psi as rolled Elongation 35-45% annealed (75 deg. F) 9-20% as rolled Hardness (Brinell) 23 annealed 32 as rolled Annealing Temperature 650 deg. F 343 deg. C

Table 3.3 CHEMICAL COMPOSITION - ALUMINUM (1100 ALLOY) 99.00% min. Aluminum 1.00% max. Silicone and Iron 0.05-0.20% max. Copper

.05% max. Manganese

.10% max. Zinc

.15% max, others each

Table 3.4.

BORON CARBIDE CHEMICAL COMPOSITION. WEIGHT %

Total boron 70.0 min.

IO B isotopic content in 18.0 natural boron Boric. oxide 3.0 max.

Iron 2.0 may.

Total boron plus 94.0 min.

total carbon BORON CARBIDE PHYSICAL PROPERTIES Chemical formula BC 4

Boron-content (weight) 78.28%

Carbon content (weight) 21.72%

Crystal Structure rombohedral.

Density 2.51 gm./cc-0.0907 lb/cu. in.

0 0 Melting Point 2450 C - 4442 F Boiling Point 35000C-6332 0F.

Microscopic Capture 600 barn cross-section

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1 l-4.0 CRITICALITY SAFETY ANALYSJJ.

4.1 Introduction The high density spent fuel storage racks for Pilgrim Nuclear Power Station are designed to assure that the neutron multiplication factor ( k,,, ) is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with unborated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations and in mechanical tolerances, statistically combined, such that the true k,,, will be equal to or less than 0.95 with a 95% probability at a 95%

confidence level. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under credible abnormal conditions, the reactivity will be less than the limiting design basis value.

The design basis fuel is an 8x8 fuel rod assembly with a uniform initial enrichment of 4% U-235 containing 2% gadolinia (Gd0) 23 in eight fuel rods. Criteria for the acceptable storage of higher enrichment fuel and for the GE-119x9 rod array were also developed based upon the k, in the standard core geometry (defined as an infinite array of fuel assemblies on a 6-inch lattice r, pacing at 20*C, without any control absorber or voids).

Applicable codes, standards, and regulations, or pertinent sections thereof, include the following:

General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling.

USNRC Standard Review Plan, NUREG-0800, Section 9.1. 2, Spent Fuel Storage.

USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979.

4-1 .,

1

USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed) , December, 1981.

. USNRC Regulatory Guide 3.41, Validation of Calculational Methods for Nuclear Criticality Safety (and related ANSI N16.9-1975).

. ANSI-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel outside Reactors.

To assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made:

The racks were assumed to contain the most reactive fuel authorized to be stored in the facility without any control rous or any uncontained burnable poison, and with the fuel at the burnup corresponding to the highest reactivity during its burnup history.

Moderator is pure, unborated water at a temperature within the design basis range corresponding to the highest reactivity.

Criticality safety analyses are based upon the infinite multiplication factor (k,,) , i.e., lattice of storage racks is assumed infinite in all directions. No credit is taken for axial or radial neutron leakage, except in the assessment of certain abnormal / accident conditions where neutron leakage is inherent.

Neutron absorption in minor structural members is neglected, i.e., spacer grids are replaced by water.

4.2 Summarv and Conclusions The design basis fuel assembly is a standard 8x8 array of BWR fuel rods containing UO 2 clad in Zircaloy, assumed to be uniform 4.0 wtt U-235 enrichment initially, with 2% gadolinia burnable poison (Gd23 0 ) in 8 fuel rods. The criticality safety was evaluated at the maximum reactivity over burnup where the gadolinium is nearly consumed. As shown in Figure 4.2.1 the maximum reactivity of the reference fuel occurs at a burnup of approximately 8 MWD /KgU with l

4-2

l a calculated k, of 0.8985 (bias corrected). Adding the effect of calculational and manuf acturing tolerances, tha maximum k, in the storage rack is 0.922 (95% probabilit _ at the 95% confidence level) including all known uncertainties and plus a conservative allowance of 0.01 6k for possible differences between fuel vendor calculations and those reported here.

The basic calculations supporting the criticality safety of the Pilgrim fuel storage racks for the design basis fuel are summarized in Table 4.2.1. For the design basis fuel, the fuel storage rack satisfies the USNRC criterion of a maximum k,,, less than or equal to 0.95, with a substantial margin. This margin is utilized in developing the acceptance criteria for storage of ruel of highar enrichment and different design in terms of the k, in the standard core geometry. Figure 4.2.1 also shows the reactivity variation with burnup for fuel of 4.6% and 4.9% enrichment, both with a fuel rods containing 2% Gd23 0 . These curvou are intended to be illustrative since the actual gadolinium content for fuel of these enrichments have yet to be spocified. Future fuel assembly designs of these higher enrichments would be expected to have higher Gd 23 0 loadings, probably with several different loadings in the same assembly.

Calculations were made for fuel of 4. 25%, 4.6%, 4.9% and the GE-ll 9x9 design at 4.5% enrichment in the storage rack and in the standard ccre geometry (6.00-in assembly pitch, 20*C). Since the higher enrichment fuel will contain a higher but as yet unspecified gadolinia content, these calculations were made to define the maximum acceptable k, in the standard core geometry for fuel of various initial (average) enrichments and design concepts, regardless or the initial gadolinia content. Calculations for the reference and higher-enriched fuel are summarized in Figure 4.2.2.

The GE-11 fuel showed the same reactivity in the rack as the BxB fuel and the curves in Figure 4.2.2 include the 9x9 fuel array at 4.6% enrichment.

4-3 l

l Figuro 4.2.3 chows the limiting k, in the etcndard coro gOomOtry for fuel of various initial enrichments. The limiting k, in the standard core geometry decreases with increasing initial average enrichment. The lower curve is for the design basis reactivity (maximum rack k, of 0.922) and the upper curve is the limit for a maximum k,,, of 0.95 in tr e storage rack, both curves with all uncertainties and allowances included. Although the upper curve in Figure 4.2.2 could be used to define the acceptable reactivity, a simple and conservative criterion for acceptable storage of fuel in the Pilgrim storage racks is that (1) the fuel must have an average enrichment of 4.9% or less and (2) the k, in the standard core geometry, calculated at the maximum over burnup, must be less than or equal to 1.32. Parallel calculations also showed that any fuel of 3.3% average enrichment or :less is acceptable for storage regardless of the gadolinium content or the k. in the standard core geometry. These criteria are also applicable to the existing racks (Section 4.9) and expected to bound all present and future fuel designs anticipated to be used at Pilgrim.

4.3 Abnormal and Accident Conditions None of the abnormal or accident conditions that have been identified will result in exceeding the limiting reactivity (k,,, of 0.95). The effects on reactivity of credible abnormal and accident conditions are summarized in Table 4.3.1. The double contingency principle of ANSI N16.1-1975 (and the USNRC letter of April 1978) specifies that it chall require at least two unlikely independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the occurrence of more than a single unlikely and independent accident condition concurrently.

Other hypothetical abnormal occurre:nces were considered and no credible occurrences or configurations have been identified that might have any adverse effect on the storage rack criticality safety.

Lattice averaged enrichment in the axial plane of highest average enrichment.

l 4-4

4.4 Inout Parameters 4.4.1 Fuel Assembly Desian Soecifications The design basis fuel assembly is a standard 8x8 array of BWR fuel rods containing 002 clad in Zircaloy (62 fuel rods with 2 vater ,

rods). For the nominal design case, fuel of uniform 4.0 wtt U-235 enrichment was assumed, with eight fuel rods cent.aining 2%

gadolinia. The GE-11 fuel design, a 9x9 array of fuel rods with 7 rods replaced by two zircaloy water channels, was also evaluated.

Design parameters for both types of fuel are summarized in Table 4.4.1.

4.4.2 Storace Rack Cell Specifications The design basis storage rack cell consists of an egg-crate structure, illustrated in Figure 4.4.1, with fixed neutron absorber material (Boral) of 0.0162 g/cm 2boron-10 areal density (0.015 gms 2

B-10/cm minimum) positioned between the fuel assembly storage cells in a 0.082 inch channel. This arrangement provides a nominal center-to-center lattice spacing of 6.257 inches. Manufacturing tolerances, used in evaluating uncertainties in reactivity, are indicated in Figure 4.4.1. The 0.090-in. stainless-steel box which defines the fuel assembly storage. cell has a nominal inside dimension of 6.05 in. This allows adequate clearance for inserting / removing the fuel assemblies, with. or without the Zircaloy flow liner. Boral panels are not needed or used on the exterior walls of modules f acing non-fueled regions, i.e. , the pool walls.

4.5 Analysis Methodoloav In the fuel rack evaluation, criticality analyses of the high density spent fuel storage racks were performed with the CASMO-3 code (4.5.1], a two-dimensional multi-group transport theory code.

Independent verification calc ulations were made with the KENO-5a 4-5

computer package (4.5.2), using the 27-group SCALE cross-section library (4.5.3) with the NITAWL subroutine f or U-238 - resonance shielding effects (Nordheim integral treatment) .

Benchmark calculations are presented in Appendix A and indicate a bias of 0.0000 1 0.0024 for CASMO-3 and 0.0101 1 0.0018 (95%/95%)

for NITAWL-KEN 05a. In the geometric model used in the calculations, each fuel rod and its cladding were explicitly described and reflecting boundary conditions (zero neutron current) were used in the axial direction and at the centerline of the Boral and steel plate between storage cells. These boundary conditiers have the effect of creating an infinite array of storage cells in all directions.

The CASMD-3 computer code was used as the primary method of analysis as well as a means of evaluating small reactivity increments associated with manufac+uring tolerances. Burnup calculations were also performed wiF CASMO-3, using the restart option to describe spent fuel in the storage cell. KENO-5a was used to assess the reactivity consequences of eccentric fuel positioning and abnormal locations of fuel assemblies.

4.6 Criticality Analyses and Tolerance variations 4.6.1 Nominal Desien Case With 2% Gd z3 o initially present in eight of the fuel rods, the reactivity of a fuel assembly of 4% enrichment increases with burnup to a maximum at 8 MWD /KgU (or slightly more) as shown in Figure 4.2.1. At 8 MWD /KgU burnup, , the nominal infinite multi-plication factor, k., in the storage racks is 0.8968. With a Ek of 0.0247 for all known uncertainties statistically combined and all allowances (Table 4.2.1) , the maximum k, is 0.922 which is less than the design basis limit for k,,, of 0.95. Reactivity effects of SCALE is an acronym for Etandardized Computer Analysis for Licensing Ivaluation, a standard cross-section se/ developed by the oak Ridge National Laboratory f or the USimC.

4-6

the natural . uranium blanket normally located at the ends of the assemblies were (conservatively) neglected since an infinite fuel length _was used.

The equivalent enrichment (fresh unburned fuel, no gadolinia) was calculated to be 2.9%. Independent check calculutions with NITAWL-KE2io at an enrichment of 2.9% gave a k of 0.8965 t 0.0010 (95%/95%) which confirms the CASMO-3 calculation at 2.9% enrichment (k. of 0.8975 without uncertainties) .

Calculations were also made at low enrichments without any gado11nia present. These results, with all uncertainties and' allowances included, are listed below, and show that the GE-119x9 array fuel exhibits a lower reactivity in the storage rack than the 8x8 fuel design for the same average enrichment.

3.3%E Ref. 8x8 Fuel 0.940 GE-11 9x9 Fuel 0.935 These data indicate that any fuel with an enrichment of 3.3% or less is acceptable for storage in the racks, regardless of the gadolinium content, fuel burnup or fuel rod array. The X-factor for 95% probability at a 95% confidence level was determined from NBS Handbook 91 (4.6.1].

4.6.2 Uncertainties Due to Manufacturina Tolerances The reactivity effects associated with manufacturing tolerances are listed in Table 4.6.1 and discussed below.

4.6.2.1 Baron Loadino Variation The Boral absorber panels used in the storage cells are nominally 0.075-inch thick, with a B-10 areal density of 0.0162 g/cm*. The 2

manufacturing tolerance limit is t 0.0012 g/cm1 in B-10 content, including both thickness and composition tolerances. This assures 4-7

that the minimum boron-10 areal density will not be less than

0. 015 . gram / cm' . Differential CASMO-3 calculations indicate-that this tolerance limit results in an incremental reactivity-uncertainty of i 0.0049 Ek.

4.6.2.2 Boral Width Tolerance Variation The reference storage cell design uses a Boral panel width of 5.00.

The tolerance on the Boral width is i 1/16 inch. Calculations (CASMO-3) showed that this tolerance corresponds to a- 0.0016 Ek uncertainty.

4.6.2.3 Storace Cell Lattice Pitch Variation The design storage cell lattice spacing between fuel assemblies is 6.257 in. Decreasing the lattice pitch increases the reactivity.

For the manufacturing tolerance of i 0.03 inches, the corresponding uneartainty in reactivity is 0.0020 ok as determined by differential CASMO calculations.

4.6.2.4 Etainless Steel Thickness Tolerances The nominal thickness of the stainless steel box is-0.090 inches and 0.035 inch for the steel backing plate. The maximum positive reactivity effect of the expected stainless steel thickness.

tolerances was calculated (CASMO-3) to be t 0.0002 ok.

4.6.2.5 Fuel Enrichment and Density Variation CASMO-3 calculations of the sensitivity to small enrichment- ,

variations yielded an average coefficient of O_2 0060 dk per 0.1 utt U-235, in the enrichment range from 4.0 to 4.25% enrichment. For an -

estimated tolerance on percent U-235 enrichment of to.05, the maximum uncertainty is t-0.0030 ok and becomes smaller for higher enriched fuel.

4-8

Calculations - were also made to determine the sensitivity to the tolerance in U0 2fuel density- (t 0.20 g/cc) . - These calculations indicate that the storage rack k, is increased by 0.0028 Sk for the expected max hum-10.61 gn/cc stack density. A lower fuel density results in correspondingly lower values of reactivity. Thus, the maximum uncertainty due to the tolerances on UO2 density is _ t 0.0028 6k.

f 4.6.2.6 Zirconium Flow Channgl Elimination of the zirconium flow channel results in a small (less than 0.008 6k) decrease in reactivity. More significant is a positive reactivity effect resulting from potential bulging of the zirconium channel, which moves the channel Wall outward toward the Boral absorber. For the maximum expected bulging based on estimates provided by GE and conservatively assumed to be uniform throughout all assemblies, an incremental reactivity of + 0.0024 Sk could result (determined by differential CASMO calculations) . Fuel assembly bowing results in a negative reactivity effect and is treated as an abnormal condition below).

4.6.3 Uncertainty in Decletion Calculations Since critical experiment data with spent fuel is not available for determining the uncertainty in depletion calculations, an allowance for uncertainty in reactivity was assigned based upon other considerations. In the Pilgrim racks, the reactivity decrement in the absence of gadolinium is apprcrimately 0.007 6k. Assuming the uncertainty in depletion calculations is less than 5% of the total reactivity decrement,- an uncertainty in reactivity' equal to 0.0050 6k would result at 10 MWD /KgU. At the burnup of maximum reactivity, the gadolinium is essentially depleted. Although the reactivity

'only that portion of the uncertainty due to burnup. Other uncertainties are accounted for elsewhere.

4-9

1 I

uncertainty due to depiction may be either positive or negative, i

for conservatism it was added to the calculated k,,, rather than being combined statistically with other uncertainties. Over long periods of storage, the reactivity will continually decrease due to the decay of Pu-241 and growth of An-241 which provides additional conservatism.

4.7 Hicher Enrichments and GE-11 Fuel Calculations were made for fuel of 4.25%, 4.6%, 4.9% and the GE-11 (9x9 array) fuel at 4.6% enrichment. The reactivities in the rack, including calculational and manufacturing uncertainties, are shown in Figure 4.2.2 as a function of the fuel assembly k, in the standard core geomntry at 20*C. Again it is evident that a k,, of 1,32 in the standard core geometry will conservatively bound both the 8x8 and 9x9 fuel designs for enrichments up to 4.9%. These curves use the uncertainties and allowances given in Tables 4.2.1 and 4.5.1.

4.8 Abnormal and Accident conditions 4.8.1 Temoerature and Water Density Effects The moderator temperature coefficient of reactivic~ y is negative.

Using the minimum temperature of 4*C therefore assures that the true reactivity will always be lower than the calculated value regardless of the temperature. Temperature effects on reactivity have been calculated and the results are shown in Table 4.8.1.

Introducing voids in the water in the storage cells (to simulate boiling) decreased reactivity, as shown in the table. Boiling at the submerged depth of the racks would occur at approximately 122*C.

4-10 i

4.8.2 Abnormal Location of a Puel Assembly It is hypothetically possible to suspend a fuel assembly of the highest allowable reactivity outside and adjacent to the fuel rack, although such an accident condition is highly unlikely. The exterior salls of the rack modules facing the outside (where such an accident condition might be postulated to exist) is a region of high neutron leakage. With neutron leakhge included, the reactivity with an extraneous fuel assembly of the maximum reac-tivity, located outside and adjacent to the fuel rack, is less than the reference k,. Thus it is concluded that the abnormal location of a fue) assembly will have a negligible reactivity effect.

4.8.3 Eccentric Fuel Assembly Positioning The fuel assembly is normally located in the center of the storage rack cell with bottom fittings and spacers that mechanically restrict lateral movement of the fuel assemblics. Nevertheless, calculations with the fuel assembly moved into the corner of the storage rack cell (four-assembly cluster at closest approach),

resulted in a small negative reactivity effect. Thus, the nominal case, with the fuel assembly positioned in the center of the storage cell, yields the higher reactivity.

4.8.4 Zirconium Puel Channel Distortion Conuoquences of bulging of the zirconium fuel channel are treated as a mechanical deviation in Section 4.6.2.6. . Bowing of the zirconium channel (including fuel rods) results in a local negative reactivity effect analogous to that of eccentric positioning the fuel assembly toward one side of the storage cell. Thus, any bowing that might occur will result in a reduction in reactivity.

4-11

4.S.5 Droceed Puel Assembly 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 active fuel region of more than the 12 inches which is sufficient to preclude neutron coupling (i.e., an effectively infinite separation). Maximum expected deformation under seismic or accident conditions will not reduce the minimum spacing to less than 12 inches. consequently, fuel assembly drop accidents will not rescult in a significant increase in reactivity due to the separation distance.

4.8.6 Fuel Rack Lateral Movement Normally, the individual rack modules in the spent fuel pool are separated by a water gap which would normally eliminate the need for poison panels between rack modules. However, as an added precaution against possible mis-alignment or seismically-induced movement, Boral panels are installed in the rack wall along one side of the water gap. With this configuration, the maximum reactivity of the storage r a k a not dependent upon the water-gap spacing between modules.

4.9 Existina Scent Fuel Storace Racks _

The existing spent fuel storage racks have previously been analyzed. Calculations and uncertaintiec supporting the criticality safety of the storage racks are summarized in Table 4.9.1. The maximum infinite multiplication factor (k.) calculated .

for the nominal design case is 0.935 including all uncertainties 4-12

. .. . , , _. - _a. . ..:--.

(95% probability at a 95% confidence level) for fuul of 3.75%

uniform enrichment (k. of fuel assembly in standard core geometry of 1.405 t 0.006) .

The design basis storage rack cell consists of an egg-crate structure, illustrated in Figure 4.6.1, with fixed neutron absorber material (Boraflex) of 0.0214 g/cm' boron-10 areal density (0.018 2

g B-10/cm minimum) positioned between the fuel assembly storage cells. This arrangement provides a nominal center-to-center lattice spacing of 6.243 in. Manufacturing tolerances, used in evaluating uncertainties in reactivity, are indicated on Figure 4.6.1. The 0.060-in, stainless-steel box which defines the fuel assembly storage call has a nominal inside dimension of 6.05 in.,

which allows adequate clearnce for inserting / removing the fuel assemblies, with or without the Zircaloy flow liner.

Reactivity effects of manuf acturing tolerances were determined, in a sensitivity study, as the difference (Ak) between cAsMo calculations with each manuf acturing tolerance independently set at its maximum deviation. Results of these calculations are eummarized in Table 4.9.1 and the uncertainties in bias and calculational methodology were combined statistically.

The acceptance criteria set forth in Section 4.2 are also applicable to the existing racks which employ a considerably higher B-10 loading (0.0214 gn/sq.cm) in the Boraflex neutron absorber.

Evaluation of abnormal / accident conditions confirms that for all conditions the reactivity remains less than the design limit.

Abnormal or accident conditions generally resulted in negligible effect.

4-13 C

i 4.10 coroarison with 0ther.Rggdh Licensed US Plants The list below gives the computed k, for recently executed high '

density storage rack projects. This listing indicates the range of j k values which h.we been previously licensed in the U.S.

Plant Year Utility Maximum P-Vogtle-2 (PWR) 1988 Georgia Power 0.936 Diablo Canyon (PWR) 1986 Pacific Gas and Electric 0.938 i St. Lucie (PWR) 1988 Florida Power & Light 0.944-Byron 1 and 2 1988 Commonwealth Edison 0.947 Oyster Creek (BWR) 1984 GPU Nuclear 0.947 MIllstono (BWR) 1989 Northeast Utilities 0.944 Kuosheng (BWR) 1991 Taiwan Power Company 0.936 FitzPatrick (BWR) 1992 New York Power  !

Authority 0.933 Laguna Verde (BWR) Comision Federal de Electricidad 0.929 Nine Mile. Point (BWR) Niagara Mohawk Power Corporation 0.935 Duano Arnold (BWR) Iowa Electric Light & Power 0.935 Pilgrim Station Boston Edison Company 0.922 l

4-14

4.11 Enferences (4.5.1) A. Ahlin, M. Edenius, H. Haggblom, "CASMO -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 Report,"

Studsvik/RF-78-6293, Aktiebolaget Atomenergi, March 1978.

(4.5.2) Green, Lucious, Patrie, Ford, White, Wright, "PSR-63 /AMPX-1 , (code package), AMPX Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B," ORNL-TM-3706, Oak Ridge National Laboratory, March 1976.

L.M. Patrie and N.F. Cross,

• KENO-IV, An Improved ~ Monte Carlo Criticality Program,"

ORNL-4938, Oak Ridge National Laboratory, November 1975.

[4.5.3) R.M. Westfall et al., " SCALE A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-0200, 1979.

[4.6.1) M.G. Natrella, Exnerimental Statistics, '

National Bureau of Standards, Handbook 91, August 1963.

P m

4-15

i-4 Table 4.2.1

SUMMARY

OF CRITICALITY SAFETY ANALYSES Temperature assumed for analysis 4*C Fuel Enrichment (average) 4.0%

Gadolinia Content, wtt Gd th t 2% in B rods Reference k,, (CASMO) 0.8968 Calculational Bias 0.0000 Uncertainties calculational io.0024 Removal of flow channel negative Eccentric assembly location negative Tolerances (Table 4.6.1) 10.0069 Statistical combinationUI of uncertainties 10.0073 Effect of Channel Bulge 0.0024 Uncertainty in Depletion calculations 0.0050 Allowance for Vendor Calculations 0.0100 Total 0.8960 1 0.0073 Maximum reactivity 0.9215- -

"3 Square root of sum of' squares of all -independent' tolerance effects.

4-16

Tablo 4.3.1 REACTIVITY EFFECTS OF ABlioRMAL AllD ACCIDDIT CollDITIOllS Accident / Abnormal Condition Reactivity Effoct Temperature incronso llegativo (Table 4.8.1)

Void (boiling) llegativo (Tablo 4.8.1)

Assembly dropped on top of rack lingligiblo Misplacement of a fuel assembly liogligible Scismic Movement liegligible r

i Table 4.4.1 FUEL ASSEMBLY DESIGN SPECIFICATIONS FUEL ROD DATA BY 8 9x9 Cladding outside diamatar, in. 0.483 0.440 Cladding inside diameter, in. 0.419 0.384 Cladding material 2r-2 Pellet diameter, inch- 0.410 0.376 Enrichment (design basis) 4.0 fD.05 -

Maximum 4.6 4.6 002 density (stack), g/cc UO 2 10.45 i 0.14 paIEB. ROD DATA Number of Water Rods 2 2 Inside diameter, inch 0.414 10.920 Outside diameter 0.484 0.980 Material Zr-2 FUEL ASSEMBLY DATA Fuel rod array 8x8 9x9 Number of fuel rods 62 74 Fuel rod pitch, inch 0.640 0.566 Fuel channel, material -Zr-2 Inside dimension, inch 5.278 5.278 Outside dimension, inch 5.478 5.408 4-18 t

, , - - - , . , , - , ,--v.- -r- -- , . - , - +

Table 4.6.1 Reactivity Uncertainties ID due to Manufacturing Tolerances Nominal Quantity Value Tolerance 6k, 3

Boron Loading 0.0162 g/cm 10.0012 g/cm 2 10.0049 Boral width 5.00 inches 11/16 inches 10.0016 Lattice spacing 6.257 inches 10.03 inches 10.0020 SS thickness 0.09 and 10.004 inches 10.0002 0.0235 inches mean Fuci enrichment 4.0% U-235 10.05% U-235 10.0030 3

Fuel density 10.41 g/cm 10.20 g/cm3 10.0028 Statistical combination '" 10.0069 of uncertainties Square root of sum of squares of all independent tolerance effects.

4-19

Table 4.8.1 EFFECT OF TEMPERATURE AND VOID ON CALCUIATED REACTIVITY OF STORAGE RACK Case Incremental Reactivity change, ok 4*C Reference 20*C -0.002 50*C . -0.007 90*C -0.015 120*C -0.023 120*C + 10% void -0.043 4

v 4-20

- _ _ , -,_ --. . . _ _ _ _ _ . _ . . - _ _ . _ _ _ _ _ _. __.~___.__-

4 Table 4.9.1

SUMMARY

OF CRITICALITY SAFETY ANALYSES OF EXISTING RACKS lioninal_Desian Cas_e k, in standard core geometry 1.405 1 0.000 k, in spent fuel storage rack (AMPX-KENO) 0.9106 Calculational bias, Ak 0.9105 Reference k, 0.9212 Uncertainties and tolerances Calculational bias 10.0048 Ak Calculation (statistical) 10.0027 Ak Boraflex thickness 10.0090 Ak B-10 concentration 10.0038 Ak Boraflex vidth 10.0013 Ak Puol enrichment 0.0009 Ak Puel density 10.0030 Ak Lattice pitch 10.0046 Ak SS thickness 10.0004 Ak Flow channel bulge iO.0058 Ak 10.0139 Statistical combination 1 0.0139 Maximum k, (95% probability 9 95% confidence level) 0.9351 Abnormal / accident conditions Temperature increase negative Ak Boiling negative Ak Reduced moderator density negative Ak Puel assembly positioning negative Ak Assembly outside rack negligible Ak Drcpped fuel assembly negligible Ak Lost / missing absorber plate +0.0027 4-21

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t 4 L a t tic , sp4 Cine HI I IIAU 6.345s 0.0529" c j- , ;,;.., ;;-

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p.

4 APPDfDIX A BENCHMARK CALCULATIONS by '

Stanley E. Turner, PhD, PE HOLTEC INTERNATIONAL November, 1992

1.0 INTRODUCTION

AND SUMM>tRY The objective of this benchmarking study is to verify both the NITAE-KDIO5aM methodology with the 27-group SCALE cross-section library and the CAsMO-3 codem for use in criticality safety calculations of high density spent fuel storage racks. Both calculational methods are based upon transport theory and have been benchmarked against critical experiments that simulate typical spent fuel storage rack designs as realistically as possible. l nosults of these benchmark calculations with both methodologies are '

consistent with corresponding calculations reported in the >

literature.

Results of the benchmark calculations show that the <

27-group (SCALE) NITAE-KDIO5a calculations consistently under-predict the critical eigenvalue by 0.0101 0.0018 sk (with a 95%

probability at a 95% confidence level) for critical experiments

• that are an representative as possible of realistic spent fuel storage rack configurations and poison vorths.

Extensive benchmarking calculations of critical experi- '

ments with CASM03 have also been reported *, giving a mean X , of 1.0004 2 0.0011 for 37 cases. With a K-factor of 2.148 for'95%

probabilit*r at a $5% confidence level, and conservatively neglect-ing the small overprediction, the CASM03 bias then becomes 0.0000 t 0.0024. CASMO3 and NITAE-KEN 05a intercomparison calculations of infinite arrays of poisoned cell configurations (representative of typical spent fuel storage rack designs) show very good agreement, confirming that 0.0000 2 0.0024 is a reasonable bias and-uncertain-ty for CASM03 calculations.

Reference 5 also documents good agreement of heavy nuclide concentrations for the Yankee cora isotopics, agreeing with the measured values within experimental error.

A-1

l The benchmark calculations reported here confirm that either the 27-group (SCALE) HITAE-KENO or CASMO3 calculations are 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. CASMO:;, however, is not reliable when large water gaps ( > 2 or 3 inches) are present.

2.0 l{.ITAE-UNO Sa BENCHMARK CALCULATIONS Analysis of a series of Babcock & Wilcox critical experiments *, including some with . absorber panels typicci of a poisoned spent fuel rack, is summarized in Table 1, as calculated with NITAWL-KEN 05a using the 27-group SCALE cross-section library and the Nordheim resonance integral treatment in NITAWL. Dancoff factors for input to NITAWL vare calculated with the Oak Ridge SUPERDAN routine (from the SCALE S system of codes). The mean for these calculations is O.9899 i O.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 1 0.0018 for the sixteen critical experiments analyzed.

Similar calculational deviations have been reported by ORNLm for some 54 critical experiments (mustly clean critical without strong absorbers) , obtaining a mean bias of 0.0100 i O.0013 (95%/95%). These published results are in good agreement with the rasults obtained in the present analysis and land further credence to the validity of the 27-group NITAWL-KEN 05a calculational model for use in criticality analysis of high density spent fuel storage racks. No trends in k, with intra-assembly water gap, with absorber panel reactivity worth, with enrichment or with poison concentration were identified.

Additional benchmarking calculations were also made for a series of French critical experimentsm at 4.75% enrichment and for several of the BNWL criticals with 4.26% enriched fuel.

A-2

i l

Analysis of the French criticals (Table 2) showed a tendency to overpredict the reactivity, a result also obtained by ORNL*. The  ;

calculated k, values showed a trend toward higher values with I 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-KD105a in calculating neutron leakage from very small assemblies.

St.milar overprediction was also observed for the DNWL series of critical experiments", 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 0.9959 i O.0013 (1 a 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 fallure of NITAWL-KD105a to adequately treat neutron leakage in very small assemblies.

Since the analysis of high-density spent fuel storage racks generally does not entail neutron leakage, the observed inadequacy of NITAWL-KDIO5a is not significant. Furthermore, omitting results of the French and DNWL critical experiment 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.

3.0 INTERPOLATION ROUTINE An interpolation routine was obtained from ORNL and la intended to interpolate the hydrogen scattering matrices for temperature in order to correct for the deficiency noted in NRC Information Notice 91-66 (October 18, 1991). Benchmark calcula- -

tions 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 3, confirm that the trend with temperature obtained by both codes are comparable. This A-3

i r

agreement establishes the validity of the interpolation routine, in conjunction Vith NITAE-KDio5a , in calculating reactivities at temperatures other than 20*C (the reference library tamperaturo).

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 CASM03 is not reliable in th6 prosence of a large water gap, although the relative values should be acetu ate. In the ,

calculations shown in Table 3 and in Figure 1, the absolute reactivity values differ somewhat but the trends with temperature are sufficiently in agreement to lend credibility to the interpola-tion routine.

4.0 CLOSE-PACKED ARPAYS The BAW close-packed series of critical experiments" intended to simulate consolidated fuel, were analyzed with NITAWL-KDio5a.

Results of these analyses, shown in Table 4, suggest a slightly higher bias than that for fuel with normal lattice spacings.

Because there are so few cases available for analysis, it is recommended that the maximum bias for closo-packed lattices be taken as 0.0155, including uncertainty. This would conservatively '

encompass all but one of the cases measured.

Similar results were obtained by ORNL".

5.O CASMO3 BENCHMARK CALCUTATIONS The CASM03 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, CASM03 is well-suited to the criticality analysis of spent fuel storage racks, since general practica is to treat the racks as an infinite medium of storage cells, neglecting leakage effects.

A-4

4 CASM03 is a modificatio.1 of the CASMO-2E code and has been extensively benchmarked againct both mixed oxide and hot and cold critical experiments by Studsvik Enargitaknik*. Reported ana-tysesm of 37 critical experinerts indicate a mean k , of 1.o004 i  !

0.0011 (la). To independently confirm thu validity of CASM03 (and to investigate any effect of enrichment), a series of calculations were made with CASM03 and with NITAWL-KDIO5a on identical poisoned storage cells representative of high-density spent fuel storage racks. Results of these intercomparison ca.lculations' (shown in i Table 5 and in Figure 2) are within the normal statistical variation of KDIO calculations and confirm the bias of 0.0000 i O.0024 (95%/95%) for CASM03.

Since two independent methods of analysis would not be expected to have the same error function with anric.hment, results of the intercomparison analyses (Table 5) indicate that there is no ,

significant effect of fuel enrichment over the range of enrich- t ments involved in power reactor fuel. Furthermore, neglecting the French and BNWL critical benchmarking in the determination of bias is a conservative approach.

P A second series of CASM03-KDiO5a intercomparison calculations consisting of five cases from the BAW critical experiments analyzed for the central cell only. The calculated results, also shown in 1 Table 5, indicate a mean difference within the 95% confidence limit of the KEN 05a calculations. This lends further credence to the recommended bias for CASM03.

• Inte comparison between analytical methods is a-technique endorsed by Reg. Guide 5.14, " Validation of Calculational Methods for Nuclear Criticality Safety".

A-5 1

,. ,,.-.g_.m...,_,_-,, . . - . . - -

. . - . . . . , - . . _ . , _ , . , _ , , , , . _- __..,,.-_,_...em. 4-

6.0 REFERDICES TO APPDfDIX A

1. Groen, Lucious, Patrie, Ford, White, and Wright, "PSR-G3-

/NITAWL-1 (code packago) HITAWL Hodular Code System For Generating coupled Multigroup Houtron-Gamma Librarien from ENDF/B", ORNL-TM-3706, Oak Ridge National Laborritory, November 1975.

2.

R.H. Wantf all ot. al. , " SCALE: A Modular System for Performing Standardized computer Analynin for Licenning Evaluation",

NUREG/CR-02OO, 1979.

3. A. Ahlin, M. Edenius, and H. 11aggblom, "CASMO -

A Puel Annombly Burnup Program", AE-RF-76-4158, Studovik report.

A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Dopidtion

p. Codo for LWR Analysin", Alls Tr.gnaActiong, Vol. 26, 604, 1977.

"CASMO3 A Fuel Annombly Burnup Program, Unors ManunI",

Studovik/HFA-87/7, Studovik Energitechnik AB, November 1986

4. M.N. Unidwin at al., " Critical Experimento Supporting Clone Proximity Water Storage of Power Ranctor Fuel", BAW-1484-7, The Babcock & Wilcox Co., July 1979.

5.

M. Edenius and A. Ahlin, "CASMO3: New Features, Benchmarking, and 100,Advanced Applications", l{nclCar Science and Engineering, 342-351, (1988)

6. H.G. Hatrolla, Experimental Statistics, National Buronu of Standards, !!andbook 91, August 1963.

7. R.W. Woutfall and J. H. Knight, " SCALE System Cross-naction Validation with Shipping-cask Critical Experiments", ALUi Transacti2DA, Vol. 33, P. 368, November 1979

8. S.E. Turner and M.K. Curley, " Evaluation of HITAWL-KENO Benchmark Calculationn for High Donnity Spent Fuel Storage Racku",

February 1982.Nuclear Science and EncTineering, 80(2):230-237, 9.

J.C. Manarancho, ot. al. , " Dissolution and Storago Experiment with 4.75% U-235 Enriched UO Rods", Nuclear Technolony, Vol.

50, pp 148, September 1980

10. A.M. Hathout, et. al.,

" Validation of Three Cross-section Librarion Used with the SCALE System for Criticality Analy-nis", Oak Ridge National Laboratory, NUREG/CR-1917, 1981.

A-G 1

11. S.R. Biarman, et. al., " Critical Separation between Sub-critical Clustars of 4.29 Wt. % "U Enriched U0, Rods in Water with Fixed Neutron Poisons", Battelle Pacific Northwest Laboratories, NUREG/CR/OO73, May 1978 (with , August 1979 arrata).

12. C.S. Hoovlar, et al., " Critical Experimento Supporting Underwater Storage of Lightly Packed Configurations of Spent Fuel Pins", BAW-1645-4, Babcock & Wilcox Company (1981).

13. R.M. Westf all, et al. , " Assessment of Criticality Computation-al Software for the U.S. Department of Energy Office of Civilian Radioactive Waste Management Applications", Section 6, Puel Consolidation Applications, ORNL/CSD/TM-247 (undated) .

A-7

Table 1 RESULT 3 0F 27-GROUP (SCALE) HITAWL-KEN 05a CALCUIATIONS OF B&W CRITICAL EXPERIMENTS ,

Experiment Calculated a Number k, I 0.9922 1 0.0006 II 0.9917 i 0.0005 III 0.9931 1 0.0005 IX 0.9915 i 0.0006 X 0.9903 i 0.0006 XI 0.9919 i 0.0005 XII 0.9915 1 0.0006 XIII 0.9945 1 0.0006 XIV 0.9902 1 0.0006 XV 0.9836 0.0006 XVI 0.9863 1 0.0006 XVII 0.9875 i 0.0006 X7III 0.9880 1 0.0006 XIX 0.9882 1 0.0005 XX 0.9885 1 0.0006 XXI 0.9890 0.0006 Mean 0.9899 i 0.0007W Bias (95%/95%) 0.0101 1 0.0016 m Standard Deviation of tha Mean, calculated from the k, values.

A-8

,t

~

Table 2 RESULTS OF 27-GROUP (SCALE) NITAWL-KEN 05a CALCULATIONS  !

OF FRENCH and BNWL CRITICAL EXPERIMENTS Prench Experiments Separation Critical Calculated Distance, cm  !!eight, 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 BNWL Experiments Calculated Case Expt. No. k, No Absorber 004/032 0.9942 i O.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 i O.0007 Zr Plates 030 0.9955 i O.0007 Mean 0.9959 i O.0013 A-9

i I

Table 3 Intercomparison of NITAWL-KEN 05a.(Interpolated) and CASM03 Calculations at Various Tamperatures Temoerature CASMO 3 W-N-KENOSaM 4'c 1.2276 1.2345 i 0.0014 17.5'C 1.2.22 1.2328 i 0.0015 25'C 1.2347 1.2360 i 0.0013 50*C 1.2432 1.2475 i 0.0014 75'c 1.2519 1.2569 i 0.0015 120*C 1.2701 1.2746 i 0.0014

• Corrected for bias A - 10 s - - .

Table 4 Reactivity Calculations for Close-Packed critical Expariments Calc. BAW Pin Module Boron calculated '

No. Expt. Pitch Spacing Conc. k, No. cm cm ppa KS01 2500 Square 1.792 1156 0.9891 1 0.0005 1.4097 KSO2 2505 Square 1.792 1068 0.9910 1 0.0005 1.4097 KS1 2485 Square 1.778 886 0.9845 1 0.0005 Touching KS2 2491 Square 1.778 746 0.9849 i 0.0005 Touching KT1 2452 Triang. 435 1.86 0.9845 i 0.0006 '

Touching KTIA 2457 Triang. 1.86 335 0.9865 i 0.0006 Touching KT2 2464 Triang. 361 2.62 0.9827-i 0.0006 Touching KT3 2472 Triang. 3.39 121 1.0034.1 0.0006 Touching ,

A - 11

Table 5 RESULTS OF CASMO3 AND NITAWL-KENO 5a 4 BENCHMARK (INTERCOMPARISON) CALCULATIONS l

~

Enrichmentm km Wt. t U-235 NITAWL-KEN 05a# CASMO3 l4%l 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 4.5 0.9563 i O.0011 0.9565 0.0002 5.0 0.9744 i O.0011 0.9746 -0.0002 Maan 0.0008 Expt. No.m XIII 1.1021 1 0.0009 1.1008 0.0013 XIV 1.0997 i 0.0008 1.1011 0.0014 XV 1.1086 i 0.0008 1.1087- 0,0001 XVII 1.1158 i 0.0007 1.1168 0.0010 XIX 1.1215 1 0.0007 1.1237 0.0022 Hean- 0.0012 m

Infinite array of assemblies typical of high-density spent fuel storage racks.

m km from NITAWL-KENO 5a corrected for bias, m Central Cell from BAW Critical Experiments .

'P A - 12

1.28 1.27- 4

. /

. l

~ e

1. 26 t- f' g A)W j

~7 3  :

a-w $ 'g - / /

i  ?

-V /

1. e4 --

\

f/ -

\. 5 -f-5 .

)- '

,5

. r; 1 4

A 1.22 ,,....

0 20 40 60 80 100 123 140 Te mp e r a t.ur e . Do ur = v. C Ftc. 1 COPPARISON OF CAST 10-3 end KENC5e TEMPERATURE CCPENDENCE A-13

- k 4 E .'

1.00-

.- i!

0. W lf<) ,

g .

e L .

z m

.i h 0,90 u -

1,4 CAST 1T KENC- Se 0.85  :

/ .

Jt-k 0.80l....2,5

+

.2 0 3.0 3.5 4.0- 4.5 5.0 5.5-FUEL ENRICHMENT.-WTsiU-235 l:

l-F tg. 2 COPPAR~ SON OF CASMO Mo KENO-5. CALC.LATIONS AT VAN 10US ENRICMMENTS IN REPRESCNTATIVE FUEL STORAGE RACI-  ;

4 v

l A-14

5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5.1 Introduction A primary objective in the design of the high density spent fuel storage racks'for the Pilgrim Station spent fuel pool is_to ensure adequate cooling of the fuel assembly cladding. In the following

~

sections, a brief synopsis of the design _ basis, the method of analysis, and the numerical results is provided.

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 Unit 1 (Docket No. 50-416), Oyster Creek (Docket No. 50-219),

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 . 50-455) , St. Lucie Unit One (Docket No. 50-335), Millstone Point I (Docket No. 50-245), Vogtle Unit 2 (Docket No. 50-425),

l Kuosheng Units 1 & 2 (Taiwan Power Company) , and Ulchin Unit 2 (Korea Electric Power Company), J.A. Fit: Patrick (Docket No. 50-333) , D.C. Cook (Dockets Nos. 50-315 and 50-316) , Zion (Docket Nos.

50-295 and 50-304), Sequoyah (Docket Nos. 50-327 and 50-328) and Three Mile Island Unit One (Docket No. 50-289), among others.

l The analyses to be carried out for the thermal-hydraulic qualification of the rack array may be broken down into the following categories:

l l (i) Pool decay heat evaluation and pool bulk temperature l

variation with time.

(ii) Determination of the maximum pool local temperature at the instant when the bulk temperature reaches its maximum value.

l 5-1

(iii) Evaluati~on 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.

(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 methods employed to perform such analyses and final results.

5.2 Spent Fuel Pool Coolina and Cleanun System Descrintion The spent fuel _ pool cooling system (SFPCS) consists of two parallel cooling trains. Each cooling train consists of one pump _and one-heat exchanger. Table 5.2.1 presents the.SFPCS heat exchanger data.

Fuel pool cooling pumps take suction from the two spent fuel pool skimmer surge tanks through a common suction header. The water is pumped through heat exchangers and through a fuel-pool filter.

Continuing the cycle, the filtered water, depending on its quality _

and conductivity, is ' then either routed through the fuel-pool demineralizer and associated basket strainer or through a bypass before passing to the two fuel pool discharge lines. The cooled water traverses across the pool,' picking up heat before' returning to start a new cycle by flowing over the weirs into the skimmer surge tanks.

Both pump and heat exchanger trains are-operated simultaneously to-take into- account high heat . loading- during and immediately following a refueling program. Flows exceeding the filter and demineralizer capacities are bypassed around these units to the fuel pool discharge lines.

5-2 t

4 During refueling outages, the SFPCS may also be ' configured to utilize an intertie connection to the_ Residual Heat Removal (RHR) system to achieve greater cooling capacity. This configuration may utilize either RHR loop and heat exchanger with any one of the four RHR pumps. There are two fundamentally different modes of operation for.the RHR/SFPCS intertie depending on whether the RHR shutdown cooling loop is available.

The first mode (Mode 1) is available when RHR shutdown cooling is operating and the reactor basin is flooded with the fuel pool gate open as is typical during refueling activities. The RHR pump takes suction from the normal shutdown cooling loop on reactor recirculation loop-A. The normal flow path is from the pump through the heat exchanger and back to the reactor vessel through either reactor recirculation loop. A portion of the shutdown cooling return flow leaving the RHR heat exchanger may be diverted through the RER/SFPCS intertie to flow up to one of the fuel pool discharge lines. This provides additional cooling to the fuel pool as the cooled water traverses the pool, picking up heat and returning to the reactor basin and the normal shutdown -cooling loop. This mode is run simultaneously with the fuel pool cooling-and cleanup system which continues drawing suction from the skimmer surge tanks and returning flow to the other fuel pool discharge line.

When the RHR shutdown cooling loop is not available, then a fundamentally different mode (Mode 2) of augmented cooling is available. The SFPCS pumps and heat exchangers are isolated and not used in this mode. The RHR pump suction is taken only from the fuel pool skimmer surge tanks through the RHR/SFPCS suction 5-3

intertie piping. The flow from the RHR heat exchanger returns through the intertie discharge piping to one or both of the fuel pool discharge lines and the reactor basin spargers if needed.

In this mode, the total RHR flow is maintained above 1800 gpm.

Table 5.2.2 shows the RHR heat exchanger data for this cooling mode. Approximately 600 gpm of the return flow through the intertie can be diverted to pass through the SFPCS filter and domineralizer to provide water cleanup and then continue up to one of the fuel pool discharge lines. The bulk of the RHR flow passes directly _

from the intertie to the opposite fuel pool discharge line. In abnormal or emergency conditions, where only fuel pool and possibly reactor basin cooling is needed, the total RHR flow may be passed directly from the intertie to one or both of the fuel pool discharge lines and, if needed, the reactor basin spargers.

The cooling water for the fuel pool and RHR heat exchangers is provided by the Reactor Building Closed Cooling Water System (RBCCW). The ultimate heat sink for the RBCCW system is the Salt Service Water (SSW) system. The normal makeup water supply to the fuel pool is from the condensate transfer system with a backup source from the plant fire water system.

The fuel pool cooling pumps are horizontal single stage centrifugal process pumps in accordance with ANSI /ASME B73.1M. The pumps are driven by 60 HP AC induction motors. During a loss of AC power, the pumps may be supplied with a temporary interconnection of AC power normally used for the reactor water cleanup pumps.

5-4

. . . . . . . . - . . -. . . ~ - .. - . - . -

4

- 'r 5.3- Decay ~ Heat' Lead Calculations- *

.The decay heat load calculation'is performed.in accordance-with the j

provisions of "USNRC Branch Technical Position ASB9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev..  ;

2, July, 1981.

For purposes of this licensing application, the decay heat. ,

calculation is based on 3859 cells of maximum storage capacity in the pool and a projected discharge schedule for.the future cycles as shown in Table 5.3.1. Since the decay heat load-is monotonic  ;

with reactor exposure time, an upper bound of 2200 ~ full -power-operation days is assumed for all discharged fuel. k The background heat load is assumed to remain invariant for the -

duration of the final discharges discussed below.

5.4 'Discharce Scenarios -

Two conditions of discharge are considered:

(i) Normal discharge (Case 1):.

In this condition, 28% of the core (164 assemblies) with 52800 hours of exposure in the reactor -- are-.

discharged to the fuel pool cooled by two spent fuel.-

pool coolers operating in parallel (Figure 5.4.1) .

The fuel transfer begins 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after-reactor-shutdown and.is carried out at the rate of-three.

assemblies per hour. .

Table 5.4.1 presents the relevant-input data.

5-5

L .

(ii) Full Core Offload (Case 2):

Full core offload is considered at the end of cycle

20. A total of 580 assemblies are transferred to the pool at the rate of three assemblies per hour after 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> in-core decay.

The RHR cooling mode 2, as described in Section 5.2, is considered to remove the decay heat load of one full core after 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> decay and the background heat load. The pool is assumed to be originally cooled by the fuel pool cooling system. The RHR heat exchanger will be operated under the reduced flow of 1800 gpm due to the limitations of the piping alignment.

The pertinent data for this case is cummarized in Table 5.4.2.

5.5 Bulk Pool Temneratures A number of simplifying assumptions were made which render the analysis conservative. These include:

The heat exchangers were assumed to have maximum design fouling. Thus, the temperature effectiveness, p, for the heat exchanger utilized in the analysis is the lowest postulated value calculated from heat exchanger technical datasheets.

No credit was taken for heat loss by evaporation of the pool water.

+

No credit was taken for heat loss to pool walls and pool floor slab.

The mathematical formulation can be explained with reference to the simplified heat exchanger alignment of Figure 5.4.1.

5-6 I

. - .. ~ .. . . . . . - , - - - ,, . - , .

)

. The basic - L energy : conservationi relationship- lfor t the Lpool -heat exchanger system yi' elds:-

dT' '

'C' _

= Qt-Qgx where:

C, = Thermal capacity of stored water _in'the pool ~.

.T = Temperature of-pool water at time, - t, '

Qt = Heat-generation rata due to-stored fuel assemblies.

in the' pool; O -

the preceding'nsection.- _is a known;. function of time,L r from:

= Heat removed in the fuel pool cooler.:

Qux Ogx. is a nonlinear. function of time if we : assume the' temperature '

ef fectiveness _ p is constant during the -calculation. . Qgx 7can,

- however, be written-in-terms of effectiveness p as follows:

Qux = W Ci p (T - t i) ~ ( 5-1)'- .

to - ti P" -

T - t, where:

W,: Coolant flow rate,-lb./hr.

C: Coolant specific heat, Btu /lb.*F.

p :- Temperature effectiveness.of. heat exchanger. ,

~

T: Pool ~ water temperature, *F.

t:g Coolant inlet temperature,-*F. ,

t:o Coolant outlet temperature, *F .

5-7

+ >

v M

p' * -

4 y . ..,n-, , S .. -- , y , y- , .E .r

l The value of p is determined by the heat exchanger design basis performance. Q 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 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 from the background heat load with respect to the rate of fuel transfer and equals Q after the fuel transfer.

5.6 Time-to-Boil Calculations were also performed to determine the time elapsed before bulk boiling of the pool water occurs if all heat exchanger cooling modes were lost at the instant when the maximum pool bulk temperature is reached.

5.7 Lp_f.al Pool Water Temnerature In this section, a summary of the methodology, calculations and results for local pool water temperature is presented.

5.7.1 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-after-shutdown ( r,) . 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.

5-8

. 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 (Figure _5.7.1). 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.

5.7.2 and 5.7.3) (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 the surroundings (wall, etc.)

5.7.2 Model Descriotion 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) . Figure 5.7.2 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 into the formulation of the integral equation are also taken from well-recognized sources (Reference [5.7.1]) and wherever 5-9 i

- ~ ,, .,

discrepancies in reported values exist, the conservative values are consistently used. Reference (5.7.2) gives the details of mathematical analysis used in this solution process.

After the axial velocity field is evaluated, it is a straight-forward matter to compute the fuel assembly cladding temperature.

The knowledge 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 these aforementioned asnunptions, the temperatures calculated in this manner overestimate the temperature rise that will actually occur in the pool. Holtec's earlier computer code THERPOOL , based on the theory of Reference [5.7.2], automates this calculation. The analysis procedure embodied in THERPOOL has been accepted by the Nuclear Regulatory 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, flux trap temperature due to gamma heating, and the clad stress calculation when a void exists, are all incorporated in THERPOOL.

THERPOOL has been used in qualifying the spent fuel pools for Enrico Fermi Unit 2 (1980), Quad Cities I and II (1981), Oyster Creek (1984), V.C. Summer (1984), Rancho Seco (1983), Grand Gulf I (1985), Diablo Canyon I and II (1986), among others.

5-10 I

l

5.8 Claddina Tomoerature The maximum specific power of a fuel array gg can be given by:

qa = q F, (5-2) where:

F,= radial peaking factor q = average fuel assembly specific power The data on radial and axial peaking factors may be found in Table 5.8.1.

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 factor. The axial heat distribution in a rod is generally a maximum in the central region, and tapers off at its two extremities.

It can be shown that the power distribution corresponding to the chopped cosine power emission rate is given by n (a + x) q(x) = gx sin t + 2a where:

L: active fuel length, a: chopped length at both extremities in the power curve.

5-11 l

x: axial coordinate with origin at the bottom-of the active fuel region.

The value of a is given by tz a=

t - 2z where:

m 2 1 1 2 <

z= - -

+

n F, w2 ,2 ,p y2

, g where F, is the axial peaking factor.

The cladding temperature T, is goverrwd by a third order differential equation which has the form of d3 T d2 T dT

+0 3 a2 "f (X) 3 2 dx dx dx where a, i 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 temperature and local water temperature profile.

In order to introduce some additional conservatism in the analysis, we assume that the fuel cladding has a crud deposit of .005 "F-sq.f t.-hr/ Btu crud resistance, which covers the entire surface.

Table 5.8.2 gives the major input for local temperature calculation.

5-12 1

l l '

Calculations are also performed assuming that 50% of the top opening in the thermally limiting storage cell is blocked due to a horizontally placed (misplaced) fuel assembly.

Finally, an assessment of the presence of the overhead platform on a spent fuel rack was also made. Calculations indicate that the open plenum between the platform and the rack is suf ficiently largo such that little additional resistance to the thermal chimney circuit is introduced. The not offect of the overhead table is equivalent to loss than 10% flow opening blockage which implies that the results corresponding to 50% blockage simulations bound the overhead platform condition.

5.9 Resultjl1 The background heat load of 19 cycles, at shown in Table 5.3.1, is calculated to be 3.31 x 10' Btu /hr. (see Table 5.9.1). It is also shown in the calculation that two normal SFPC trains maintain the bulk water temperature below 143*F during the normal refueling discharge. The decay heat load coincident to the peak water temperature is 8.ti9 x 106 Btu /hr. The RHR cooling Mode 2 is able to remove the decay heat load of one full core after 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> of decay plus the background heat load. The maximum bulk pool water temperature is 129*F. The coincident heat load is 25.90 x 10' Btu /hr. The results are shown in Table 5.9.2. The time varying bulk water temperatures and heat load are plotted vs. time-after-reactor shutdown in Figutes 5.9.1 to 5.9.4. It is demonstrated from the extreme conditions analyzed above that the RHR system, in conjunction with the normal fuel pool cooling system, will maintain the pool water temperature below 150"F during a spent fuel discharge.

5-13

l I

The postulated loss of all forced cooling events are considered.

As we noted previously, the event is assumed to occur at the instant when water temperature reaches a peak. The minimum time-to-boil is calculated to be 6.41 hrs. This occurs for Case 2, the full core of f-load, during which the cooling is provided by the RHR system. This system will function during a loss of offsite power by utilizing emergency diesel generator AC power. During a normal discharge (Case 1), the lower heat load results in a worst case time to boil of 16.17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />. For a loss of offsite power, a temporary interconnection of emergency AC power may be made to the fuel pool cooling pumps, or the intertie to the RHR system can be utilized. The maximum boil-off rate:iu determined to be 54.69 gpm.

The makeup water system in the plant is adequate to maintain the water level during a loss-of-cooling accident. The results of the analysis are summarized in Table 5.9.3.

Table 5.9.4 gives the results of the maximum local water temperature and maximum local fuel cladding temparature 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 horizontally placed (misplaced) fuel assembly. It is shown that the calculated maximum local temperaturcs will not cause local nucleate boiling at any location in the racks.

The current FSAR states that the fuel pool cooling system will maintain a fuel pool temperatcr<t of 125F with a heat load of 6.3 MBtu/Hr immediately following refueling. In this report, the maximum temperature during the actual fuel transfer is calculated to be 142F with a coincident heat load of 8.69 MBtu/Hr. The higher heat load is derived from the shorter decay time assumed and the higher background heat load from a full spent fuel pool.

5-14

The conclusion is that th; sv ent fuel pool cooling system, together with the RHR- intartie capability, provides- adequate cooling capacity with the current and proposed background heat loads for all normal and full core off-load scenarios..

5.10 References (5.7.1) General Electric Corporation, R&D Data Books,

" Heat Transfer and Fluid Flow", 1974 and updates.

[5.7.2) Singh, K.P. et al., " Method for Computing the Maximum Water ' Temperature in' a Puel Pool Containing Spent Nuclear Fuel", Heat Transfer Engineering, Vol. 7, No.1-2, pp. 72-82 (1986) . ,

I 5-15

Table 5.2.1 SPENT FUEL POOL COOLING SYSTEM Number of coolers in parallel Two Pool water flow rate through each cooler, gpm: 660 Coolant flow rate through each cooler, gpm: 500 Coolant (RBCCW) inlet temperature, 'F B0 Cooler temperature effectiveness, p: 0.2778 Heat exchanger rated capacity, 10 8Btu /hr. 3.15/ cooler

( .Il b

Table 5.2.2 RHR HEAT EXCHANGER DATA Coolant (RBCCW) flow rate, gpm'. 2700 Coolant inlet temperature, 'F: 80 Pool water flow rate, gpm: 1800 (Design 5000 gpm)

RHR heat transfer effectiveness, p: 0.3886 I

Table 5.3.1 PROJECTED AND EXISTING DISCHARGE SCHEDULE Cycle Shutdown Assemblies Assemblies Number Date Discharced Stored 1 12-28-73 20 20 2 01-29-76 132 -152 3 08-06-77 428 580 4 01-04-80 92 672 5 09-25-81 232 904 6 12-09-83 224 1128 7 07-24-86 192 1320 8 05-04-91 168 .1488 9 06-01-93 152 1640 10 06-01-95 168 1808 11 06-01-97 160 1968 12 06-01-99 160 2128 13 06-01-01 164 2292 14 06-01-03 160 2452 ,

15 06-01-05 160 2612 16 06-01-07 164 2776 17 06-01-09 160 2936 18 06-01-11 164 3100 19 06-01-13 160 3260 20 06-01-15 164 3424 All discharged assemblies are- assumed to be irradiated for 2200 effective full. power days (EFPD) for conservative decay heat calculation. The discharge schedule for cycles 8 to 20 utilizes-projected batch sizes for conservatism in the thermal-hydraulic analysisi results. The batch sizes and shutdown- dates in this table are the' actual values.used in the decay heat load calculation and are slightly different from the data presented in Table 1.1. The effect of the discrepancy between the two tables on the heat load is immeasurably small.

a Table 5.4.1 DATA FOR NORMAL DISCHARGE (CASE 1)

Number of ass smblies 164 Cooling system 2 SFPC Trains Exposure times, hrs. 52800 (2200 EFPD)

In-core decay time before transfer, hrs. 120 Fuel transfer speed, assemblies /hr. 3 f

Table 5.4.2 1

DATA FOR FULL CORE OFFLOAD CONDITION (CASE 2)

Number of assemblies in the full core 580 Cooling system RIIR Cooling Mode 2 Decay time, hrs. 120 Fuel exposure time in the full core, hrs. 52800 (2200 EFPD)

I e

Table 5.8.1 .

RADIAL AND TOTAL PEAKING FACTORS Factor -Value Radial 1.80 s Axial 1.40 Rod-to-Bundle 1.40

-Total 3.5

Table 5.8.2 DATA FOR LOCAL TEMPERATURE ANALYSIS Type of fuel assembly BWR Fuel cladding outer diameter,_ inches 0.484 Fuel cladding inside diameter, inches 0.414 Storage cell inside dimension, inches 6.05 Active fuel length, inches 150 No. of fuel rods / assembly 62 Operatiyg power per fuel assembly 11.67 P x 10 , Btu /hr Cell pitch, inches 6.257

, Cell height, inches 165.375 Plenum radius, feet 25.28 Botton height, inches 6.25

-Minimum gap between pool wall- 3.00 and outer rack periphery, inches i

The fuel bundle data corresponds to an 8x8 GE assembly.

However, the resistance characteristics of other assembly types, '

such as GE10 used in the Pilgrim pool, are only marginally different. Therefore, the ' 8x8 assembly data is used for all prototypical runs.

F Table 5.

9.1 BACKGROUND

DECAY HEAT AND POOL CAPACITY DATA Operating power _ per assembly P ; 108 Btu /hr. 11.67 Background deccy heat of 19 cycles, 10' Btu /hr. 3.31' SFP capacity, 10 8 Btu /*F 1.99

'f f

Table 5.9.2 SFP BULK POOL TEMPERATURE

~

Coincident Ileat Maximum Coincident- Genegation-Pool. ,

Time Qx10 Temp., F (hrs.) Btu /hr.

Normal Discharge 142.03 207 8.69

-(Case 1)

Full Core Offload 129.35 135 25.90 (Case 2)

P t

Table 5.9.3 RESULTS OF TIME-TO-BOIL Case Maximum Number Time-to-Boil (Hours) Boil-Off Rate. cro m 1 16.17 18.33 2 6.41 ~ 54. 69 '

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Maximum ' Maximum Local Local-Pool. Fuel Water Cladding Temo..*F Temo..-'F No Blockage 184.4 229.2 50% Blockage 193.3 236.3

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6.0 STRUCTITRAL/ SEISMIC CONSIDERATIONS 6.1 Introduction This section contains analyses to demonstrate structural adequacy of the high density spent fuel racks in the PNPS pool design under seismic' loadings postulated for the spent fuel pool slab. Analyses ^

and subsequent evaluations are in compliance with the requirements of the 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 numerous previous rarack licensing efforts (please see 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 of the analyses reported herein show that the new and existing high density spent fuel racks are structurally and kinematically adequate to meet requirements defined in references

[6.1.1), [6.1.2), and [6.1.3) with large margins of safety.

6.2 Analysis outline The PNPS spent fuel rack is designated as a seismic category I structure [6.2.1). By the nature of its construction, the rack is also a free-standing structure consisting of discrete storage cells ,

which are loaded with free-standing fuel assemblies. As a result, 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, the initial step 6-1

in spent fuel rack qualification is to develop synthetic time-histories for three orthogonal directions which comply 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 for these fuel racks. Similarly, rack-to-pool wall impacts, if engineered into the rack design (not contemplated for these racks), must be within stipulated limits. Accurate and reliable assessment of the stress field and kinematic behavior of.

the rack modules calls for a comprehensive and conservative dynamic mode) 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, it must possess the capability to effect the momentum transfers which occur due to '

rattling of the fuel assemblics inside the storage cells and impacts of support pedestals on the 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 actual value is no guarantee of conservatism.

6-2 L _ _ _ _ _ _ _ _ - - - - _ - - - - - - - - - -- ------o

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, y, often maxinizes the computed rack displacements as well as the equivalent elastostatic strenses.

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 shou.' deal with all of these without sacrificing conservatism.

The 3-0 eingle rack dynamic model introduced by Holtec 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) tackles ,

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:

Interface Coefficient of Priction Parametric runs are made with upper bound and lower bound values of the coefficient of friction. The limiting values are based on experimental data.

Ingact Phenomena Compression-only gap elements are used to provide for opening and closing of interfaces such as the podental-to-bearing pad interface.

l 6-3 E...,___________,_,..__-._____.- -- - _

1 1

Fuel Leadina Scenarlos The fuel assemblies are conservatively assumed to rattle in unisos which obviously exaggerates the contribution of impact against tho cell wall. The different patterns of possible fuel assemblp loadings in the rack are simulated by orienting the conter of gravity column of the assemblage of fuel assemblies with respect of 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). The most commonly used assumption is that the adjacent:

racks vibrate out-of-phase with respect to the rack being analyzed. i 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 aspects 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 l 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 Ucrris 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 l predictions of rack kinematics, which is the chief weakness of the 1

single rack 3-D simulations.

l In order to ensure utmost confidence in the results of structural saf ety analyses, we present results for both single rack 3-D and Whole Pool Multi-Rack (WPMR) 3-D analyses. The intent of this-parellel approach is to foster added confidence and to uncover any-l peculiarit;.es in the dynamic response which are germane to the structural safety of the storage system.

l 6-4 l

In the following, we summarize the sequence of model development 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 plant specific free-standing rack modules.

b. Perform 3-D dynamic analyses on limiting module geometry types (from all those present in 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.

d. Perform a degree-of-freedom (DOF) reduction procedure on the single rack 3-D model such that kinematic responses calculated by the Reduced DOF model (RDOFM) are in agreement with responses obtained using the baseline single rack models of step (b). The RDOFM is also truly three-dimensional.

c. Prepare a whole pool multi-rack dynamic model which includes the RDOFM's of all 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 Hulti-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) no rack-to-rack impact in the cellular region of the rackr.

6-5

6.3 Artifleial 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 synthetic 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.

The acceptance criteria for each option are specified in Ref.

[6.1.2). For our purposes, Option 2 is used in generation of' seismic time-histories from Safe Shutdown Earthquake (SSE) response; spectra and from Operating Basis Earthquake (OBE) response spectra.

The total time duration between 10 seconds and 25 seconds is required to adequately match the design response spectra at 0.4 Bz.

The corresponding stationary phase strong motion duration should be between 6 and 15 seconds. With Option 2, as a minimum, four sets of time-histories should be generated for use in analyses. The response spectra regenerated from each individual synthetic time-history need not envelop the original design response spectra. The multiple time-histories are acceptable if the averace response spectra calculated from the regenerated response spectra of these synthetic time-histories envelop the original design response spectra with the same damping factor.

The acceptance criterion for spectrum enveloping is that no more than five points of the spectrum obtained from the time-history fall 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. The nature of the spent fuel rack structure is such 6-6

<l I

i

\

that primary response is to excitations above 5-0 Hz. Within the 5-33HZ range, discreto check points are established from the above 10%

criterion.

Generated artificial time-histories must also be statistically independent. Any two time-histories are considered to be ,

statistically independent if their normalized correlation l coefficient is less than 0.15.

A set of seismic response spectra is provided for the Pilgrim Nuclear Power Station for use in qualification analyses [6.3.1). In accordance with [6.3.1), the response spectra on Sheets A-9 and A-10 of Ref. [6.3.1) with damping factor of 2% for SSE and 1% for OBE are used for generating acceleration time-histories for the spent fuel pool slab on the Reactor Building at elevation 74.25'. The reference vertical spectrum in taken as two-thirds of the horizontal spectrum.

Holtec proprietary computer program GENEQ [6.3.2) is used to '

generate four sets of synthetic SSE time-histories and four sets of synthetic OBE time-histories. Each set consists of three time-histories in orthogonal directions, two in horizontal plane and the other in vertical direction. Figures 6.3.1 to 6.3.3, 6.3.4 to 6.3.6, 6.3.7 to 6.3.9 and 6.3.10 to 6.3.12 show the time-history plots for the four SSE time-history sets, respectively. The response spectra of these time-histories are regenerated with 2.0%

damping. The average response spectra of the regenerated respense spectra in three directions are calculated respectively, and are plotted in Figures 6.3.13 to 6.3.15. The corresponding original design spectra are also plotted in Figures 6.3.13 to 6.3.15 to show that the average spectra anvelop the corresponding design spectra.

Figures 6.3.16 to 6.3.16 6.3.19 to 6.3.21, 6.3.22 to 6.3.24 and l 6.3.25 to 6.3.27 are the time-history plots for the four OBE time-history sets, respectively. The average response spectra (with 1%

damping) of the OBE time-histories, as well as the original design 6-7 I-4

b spectra, are given in Figures 6.3.28 to 6.3.30, which demonstrate; that the average spectra of the synthetic OBE time-histories ~ ment:

the envoloping requirement.

The normalized cross-correlation coefficients pij between SSE time-:

histories i and j within any one of the four sets, are provided in Table 6.3.1. The pij between OBE time-histories i and j within any!

one of the four sets are provided in' Table 6.3.2. Tables 6.3.1 andl s t are compliance with the statist cal in n requirement.

It is obvious that each of the four sets of synthetic time-histories, when applied to the same rack seismic model, will produce a particular set of displacement and stress results.- Fori conservatism, to obtain-a set of results which will bound each set of results produced by each of the four s'ets of time-historics,.the following procedure is used to establish a " controlling" or

" governing" set of time-histories from the four candidate sets:

(1) First, each of the four sets of synthetic SSE time-histories.

is applied to a typical Pilgrim rack-(Rack N1) to obtain four sets of results.

(2) By comparing the results produced by each of the four sets of time-histories, the set of-time-histories which produces the-maximum results in displacements and stresses can ~ be identified and is referred to as " Set A".

(3) A suitable amplification' factor-(in our case,-it is 1.15 ): is' applied to " Set A" to define the set of " controlling" or ,

" governing" time-histories. Application of the amplification factor ensures that the results of displacements and stresses in a Pilgrim rack produced by the controlling set of time-histories will bound the corresponding rcsults produced by each of the four sets of time-histories for-the same-rack.

(4) The controlling set of time-histories is used in our rack and pool analyses to obtain conservative results.

6-8

_. ~.._ __ . _ . _ _ . _ _. .. - __ -_ _ _ . _ . _ _ . _ . _ _ .

Table 6.3.3 shows the determination of controlling time-histories for Pilgrim SSE seismic event. SSE-Set-2, magnified by a multiplier of 1.15 (for conservatism), is defined as the controlling SSE time-history set and is used in all the analyses as seismic excitation input.

6.4 Rack Modelina for Dynamic Simulations 6.4.1 General Remarks Spent fuel storage racks are seismic class I equipment. They are required to remain functional during and after an SSE event. The racks are frec-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 typical module. The baseplate extends beyond the cellular region envelope ensuring that inter-rack impacts, if any, occur first at the baseplate elevation; this area is structurally qualifiable to withstand any large in-plane impact loads.

A rack may be completely loaded with fuel assemblics (which corresponds to greatest total mass), or it may be completely empty.

The coefficient of friction, p, 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 w. ster 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 values of p = 0.2 and 0.8 have been found to bracket the upper limit of module response in previous rarack projects.

4 6-9

Since free-standing racks are not anchored to the pool slab, noc attached to the pool walls, and not interconnected, they can execut a wide variety of motions. Racks may slide on the pool floor, one o]"

more rack support pedestals may momentarily tip and lose contae-with the floor slab liner, or racks may exhibit a combination o~

sliding and tipping. The structural models developed permi simula- of these kinematic events with inherent built- .

consors . isms. The rack models also include components fog simulation of potential inter-rack and rack-to-wall impacG phenomena. Lift-off of support pedestals and subsequent linee 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 cerm to denote the mathematical element repreienting the case where restoring force is not linearly proportional to displacement. In the fuel rack simulations, the Coulomb friction interf ace between rack support pedestal and liner is typical of a nonlinear spring.

3-D dynanic analyses of single rack modules require a key modeling assumption. This relates to location and relative motion of neighboring racks. The gap between e 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 that neighboring racks move 180*

out-of-phase in relation to the subject rack. Thus, the available gap before inter-rack 12npact occurs is 50% of the physical gap.

This " opposed phase motion- assumption increases likelihood of 6-10

intra-rack impacts and in thus conservative. However, it also increases the relative contribution of fluid coupling, which depends on fluid gaps and relative movements of bodies, making overall consnrvatism a less certain assertion. 3-D Whole Pool Hulti-Rack analyses carried out for several plants in recent years (D.C. Cook, Sequoyah, Chin Shan, Shearon Harris, etc.), demenstrate that single rack simulations predict smaller rack displacement during i.cismic 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 comprehensive 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 Sinale Rack Modulo 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 B, .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.

6-11

c. Fluid coupling between rack and fuel assembliess and between rack and wall, is simulated by appropriateinertialcouplinginthesystemkinetig energy. Inclusion of these effects uses th methods coupling of and

[6.4.2]for and rack-to-rack

[6.4.3) for rack /assembl[

coupling respectively. Fluid coupling terms for rack-to-racb coupling are based on opposed phase motion od adjacent modules.

d. Fluid damping and form drag is conservativelp neglected.

e. Sloshing is Lagligible at the top of the rack and is neglected in the analysis of the rack.

f. Potential impacts Mtween rack and fuel assemblies are accounted for a.;f appropriate " compression only" gap elements between masses involved. The possible incidence of rack-to-wall nr rack-to-rack impact is simulated by gap alsments 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' 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, p1 represent absolute degrees-of-freedom, and qi represent degrees-of-freedom relative to the slab.

H is the height of the rack above the baseplate. Not - shown . in Figure 6.4.2 are gap elements used to model pedestal / liner impact locations and impact locations with adjacent racks.

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Table 6.4.1 lists the degrees-of-freedom for the single rack model.

Translational and rctational degrees-of-freedom 1-6 and 17-22 describe the L k motion; rattling fuel masses (nodes 1*, 2*, 3*,

4*, 5* in Figure 6.4.2) are described by translational degrees-of-freedom 7-16. Ul(t) represents pool floor elab displacement seismic time-history.

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 assembly 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 plane a shear and bending spring simulate elastic effects (6.4.4).

Linear clastic 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 Pluid Counlina 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 bodies are submerged in frictionless fluid, then Newton's equations of motion for the two bodies are (mi + M11) X1+H12 X2 - applied forces on mass mi + 0 (X1) 2 H21 X1 + (m2 + H22) X2 = applied forces on mass m2 + 0 (X2) 2 7

X,1 X2 denote absolute accelerations of masses mi and m2, respectively, and the notation O(X )2 denotes nonlinear terms.

6-13

i H11, M12< H21, and H 22arefluidcouplingcoefficientswhichdependf

~

on body shape, relative disposition, etc. Fritz [6.4.3) gives data-for Hij 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 storage location encounters this effect. For example, fluid 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 1800 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 Element Details The cartesian coordinate system associated with the rack has the following nomenclature x = Horizontal coordinate along the short direction of rack rectangular planform y a Horizontal coordinate along the long direction of the rack rectangular planform z = Vertical coordinate upward from the rack base l Table 6.4.2 lists all spring elements used in the 3-D 22 DOF single

} rack model.

\

i 6-14

If the simulation model is restricted to two dimensions (one horizontal motion plus vertical motion, for example), for the purnoses of model clarification oniv, then a descriptive model of the simulated structure which includes gap and friction elements 1 shown in Figure 6.4.8. This simpler model is used to elaborate sn the various stiffness modeling elements.

Gap elements modeling impacts between fuel assemblies and rack have local stiffness Ky in Figure 6 4.8. In Table 6.4.2, for example, gap elements 5 through 8 act on the rattling fuel mass at the rack top.

Support pedestal spring rates Ks are modeled by elaments 1 through 4 in Table 6.4.2. Local compliance of the concrete floor is included in Kg. Friction elements 2 plus 8 and 4 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 friction springs with suitably large stiffness Kg 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 H (either zero if the pedestal has lifted-off the liner, or a compressive finite value) is computed. Finally, support rotational friction springs Kn reflect any rotational restraint that may be offered by the foundation. The rotational friction spring rate is calculated using a modified Bouri nesq equation [6.4.4) and is included to simulate resistive mcment by the slab to counteract rotation of 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 oZ the rack support pedestals and the shift in location of slab resistive load as the rack pedestal rotates.

The gap element Kg, modeling the effective 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 stiffness of the rack cellular structure above the pedestal.

6-15 d

l The previous discussion is limited to a 2-D model solely for simplicity. Actual analyses incorporate 3-D motions and include all stiffness elements listed in Table 6.4.2.

6.4.3 Whole Pool Multi-Rack fWPMR) 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, and to initially assess kinematic compliance (no rack-to-rack impact 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-Rack (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 Pluid Coucling The presence of fluid moving in the narrow gaps between racks and between racks and pool walls causes both near and f ar field fluid coupling effects. A single rack simulation can effectively include only hydrodynamic effects due to contiguous racks when a certain set of assumptions is used for the motion of contiguous rseks. In a Whole Pool Multi-Rack analysis, far field fluid coupling effects of all racks are accounted for using the correct model of pool fluid mechanien. The external hydrodynamic mass due to the presence of walls or adjacent racks is computed in a manner consistent with fundamental fluid mechanics principles (6.4.5) using conservative nominal fluid gaps in the pool at the beginning of the seismic event. Verification of the computed hydrodynamic effect by comparison with experiments is also provided in [6.4.5). This formulation has been reviewed and approved by the Nuclear Regulatory 6-16

l l

Commission during post-licensing multi-rack analyses for the Diablo l

Canyon Unit I and II reracking project. The fluid flow model used to i obtain the whole pool hydrodynamic effect reflects actual gaps and rack locations.

6.4.3.3 Coefficients of Priction To eliminate the last significant element of uncertainty in rack dynami' analyses, the friction coefficient is ascribed to the support pedestal / pool bearing pad interface consistent with Rabinowicz's data [6.4.1). Friction coefficients, 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 WPHR analysis can simulate the effect of different coefficients of friction at adjacent rack pedestals. The friction coefficients at the interface between rack support pedestals and pool liner is assumed distributed randemly with a mean of 0.5 and permitted to vary between the limits of 0.2 - 0.8.

6.4.3.4 Modelina Details Figure 6.4.9 shows a planform view of the spent fuel pool after Campaign III of the reracking project, 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.

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

• Note that DYNARACK has the capability to change the coefficient of friction at any pedestal at each instant of contact based on a random- reading of the PC-clock cycle. However, exercising this option would yield results that could not be reproduced. Therefore, the random choice of coefficients is made only once per run.

6-17

degrees-of-freedom. A portion of contained fuel assemblies is assumed to rattle at the top of the rack, while the remainder of the, contained fuel is assumed as a distributed mass attached to the rack. The rattling portion of the contained fuel is modeled by two hori=ontal degrees-of-freedom.

Thus, the WPMR model involves all racks in the spent fuel pool with each individual rack modeled as an 8 degree of freedom structure.

The rattling portion of fuel mass, within each rack, is chosen to insure reasonable agreement between 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 vertical compression element. Values used for spring constants for the various stiffness elements are equal to the values used in the 22 DOF model.

6.5 Acceotance Criteria, Stress Limits, ad finterial Properties 6.5.1 Acceotance criterin There are two sets of criteria to be satisfied by the rack modules:

a. Kinematic Criteria The rack must be a physically stable structure and it must be demonstrated that there are no inter-rack knpacts in the cellular region. The criteria for physical stability is that an isolated rack in water exhibit no overturning tendency -when a seismic event of magnitude 1.1 x controlling - set j SSE is applied [6.1.2).

l l

6-18

l I

f

. I

b. Stress Limit criteria Stress limits must not be exceeded under certain load combinations. The following loading combina-tions are applicable [6.1.3).

Loadina combinati2D Service Level D+L Level A D + L + To D + L + To + E D + L + Ta + E Level B D + L + To + Pg ,

D + L + Ta + E ' Level D D+L+Pd ,. The functional capability 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 Puel Storage and Handling Applications" sections D = Dead weight-induced internal moments (including fuel assumbly weight)

L =

Live Load (not applicable for the fuel rack, since there are no moving objects in the rack load path)

= Force caused by the accidental drop of the heaviest Fd load from the maximum possible he:.ght (See section 7 of this report.)

P; =

Upward force on the racks caused -by postulated st.uck fuel assembly (see Section 7)

E = Operating Basis Earthquake (OBE)

E' =

Safe Shutdown Earthquake (SSE)

To = Differential temperature induced loads (normal operating or shutdown condition based on the most critical transient or steady state condition)

Ta =

Differential temperature induced loads (the highest temperature associated with the postulated abnormal design conditions) 6-19

Ta and To cause local thermal stresses to be produced. For fuel racC analysis, only one scenario need be examined. The worst situation iq obtained when an isolated storage location has a fuel assembig generating heat at maximum postulated rate and surrounding storag(

locations contain no fuel. Beated water makes unobstructed contacG 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 (thermal) stresses.l For rack qualification, To, Ta are the same.  !

6.5.2 Stress Limits for various conditions. I I

Stress limits are derived from the ASME Code, Section III, Subsection liF [ 6.1.3 ) . Parameters and terminology are in accordancel with the ASME Code.

6.5.2.1 florinal and Doset Conditions (Level A or Level B) l

a. Allowable stress in tension on a net section iss Ft = 0.6 Sy (Sy = yield stress at temperature) l (Ft is equivalent to prima 7 membrane stress)

b. Allowable stress in shear on a not section ist i

Fy = .4 S y

c. Allowab]e stress in compression on a not section 2

l (k1)2

[1 - /2Ce Sy r2 Fa "

5 kl kl 3 3

{( ) + [3 ( ) /8Ce) - [( ) /8C e H 3 r r 6-20

,L g m _

where:

(2rr2 E) /

Ce = [ - _ -L Sy

.1 = unsupported length of component =

k = length coefficient which gives influence of-boundary conditions; e.g.

k --1 (siaple support both ends)

= 2 (cantilever beam)

L 1/2 (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 not - section, due to . flexure l:about one ~

plane =of_ symmetry'is:

Fb = 0.60 Sy (equivalent to primary bending)_

e. Combined flexure and compression --- on l 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 strees along x-axis:

fby = Maximum flexural stress along y-axis-t 6-21'

Cmx =

Cgy = 0.85 fa Dx <; .

P'ex fa Dy=1-F'ey 12 rr 2 E F;ex,ey = -

23 (' )

r x,y and subscripts x,y reflect the particular bending plane,

f. Combined flexure and compression-(or tension) on a' net section:

fa fxb fby

+ -- + < 1.0 0.6S y Px b Pby The above requirements are to bn met for both direct tension or comprr^sion.

6.5.2.2 Level D Service Limits Section F-1?70 (ASME Section III, Appendix F), states that limits for the Level D condition are the minimum of 1.2 (Sy /Ft) or (0.7Su/Ft) times the corresponding limits for- the Level A condition. Su is ultimate tensile stress at the specified rack -

design temperature. For example, if the material is such that 1.2S y is less than 0.7Su, then the multiplier on the Level A l limits, to obtain Level D limits, is 2.0.

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6.5.2.3 Dimensionless Stress Factors To facilitate the perusal of the results, stresses 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 OBE and 2.0 (or less) for the SSE condition. Stress factors reported are:

R1

= Ratio of direct tensile or compressive stress on a net section to its allowable value (note pedestals only resist compression R2

= Ratio of gross shear on a not section in the x-direction to its allowable value R3

= Ratio of maximum bending stress due to bending about the x-axis to its allowable value for the section R4

= Ratio of maximum bending stress due to bending about the y-axis to its allowable value for the section R5

= Combined flexure and compressive factor (as defined in 6.5.2.le above)

R6 = Combined flexure and tension (or compression) factor (as defined in 6.5.2.lf)

=

R7 Ratio of gross shear on a net section in the y-direction to its allowable value.

6.5.3 Material Pronerties Physical properties of the rack and support materials, obtained from the ASME Boiler t. Pressure Vessel Code, Section III, appendices, are listed in Table 6.5.1. Maximum pool bulk temperature is less than 200 F; this is used as the reference design temperature for evaluation of material properties. Stress limits for Level A,D, corresponding to conditions in Section 6.5.2 above, are evaluated using given yield strength data.

I 6-23

l l

Governino Ecuations of Motion 6.6 Using the structural model for either 22 DOF single rack analysis, or the set of simplified 8 DOF models that comprise a Whole Pool Multi-Rack model, equations of motion corresponding to each degrese of freedom are obtained using Lagrange's Formulation [6.6.1). T'ae system kinetic energy includes contributions from solid structures and from trapped and surrounding fluid. The final system of equations obtained have the matrix forms

[M) {q"} = {Q} + {G) where:

[H] -

total mass matrix (including structural and fluid mass contributions)

{q) -

the nodal displacement vector relative to the pool slab displacement; (double prime stands for second derivatives with respect to time)

{G} - a vector dependent on the given ground acceleration;

{Q} -

a vector dependent on the spring forces (linear and nonlinear) and the coupling between degrees-of-freedom The equations can be rewritten as:

{q"} = [M)~1 {Q} + [M)~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 6-24

DnIARACK. It is demonstrated there that well known classical nonlinear phenomena (subharmonic resonance, bifurcation, stick-slip) can be reproduced using DDIARACK.

The application of DntARACK 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 time steps to ascertain the upper li.mit on time step that will provide converged results. This is done by taking a typical rack 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.

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 out. Where required, DWIARACK automati: ally 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.

6-25

l 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.

From the list of racks given in Table 6.4.3, those chosen to be analyzed are Boltec rack N5 (the rack with maximum aspect ratio),

Boltec rack N1 (the largest rack), and existing rack E9. Altogether, 48 runs are carried out for governing cases using Boltec proprietary computer program DYNARACK [6.6.3, 6.6.4, 6.6.5). Results are abstracted from output files and presented here for the governing cases. Analyses have been carried out for the BWR fuel of 680 lb.

dry weight.

6.7.1 Racks in the Fuel Pool A summary of results of all analyses performed for racks in the pool, using a single rack model, is presented in summary Tables 6.7.1-6.7.51. Table 6.7.1 lists all runs carried out. Table 6.7.2 presents the bounding results from All runs for Boltec racks, and Table 6.7.3 the bounding results from all runs for existing racks.

Tables 6.7.4-6.7.51 give details for each run. The tabular results for each run give maximax (maximum in time and in ap:nce) values of stress factors at important locations in the rack. Results are given for maximum rack displacements (see Section 6.4.2.2 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. It is shown that no rack-to-rack or rack-to-wall impacts occur in the cellular region of the racks.

In the single rack analysis, kinematic criteria c.re checked by confirming that no inter-rack gap elements at the top of the rack close (see Figure 6.4.9). By virtue of the symmetry assumption for i

6-26

the assumed opposed-phase rack motion, hnpact is assumed to occur if the local horizontal displacement exceeds 50% of the actual rack-to-rack gap. For the assumed in-phase rack motion case, impact is assumed to occur if the local horizontal displacement exceeds the actual rack-to-wall gap.

Structural integrity at various rack sections is considered by computing the appropriate stress factors Ri. Results corresponding to the SSE event yield the highest stress factors. Limiting stress factors for pedestals are at the upper section of the support and are to be campared with the bounding value of 1.0 (OBE) or 2.0 (SSE). Stress factors for the lower portion of the support are not limiting and are not reported. It is seen from Tables 6.7.2 and 6.7.3 that all stress factors for Boltee racks and for existing racks are below the allowable limits. It chould be noted that because all the stress factors for the SSE event are below 1.0, and thus meet the OBE requirement, it is not necessary to carry out single rack analysis for the OBE condition.

In order to examine the stability of racks during seismic conditions, a special single rack run has been carried out. From Table 6.7.2, it is seen that the maximum rack corner displacement at rack top occurs in Run drn5ssei.rf 5 (conditions: Rack N5; in-phase motion assumed; fully loaded with regular fuel; and friction coefficient 0.5) for the controlling SSE seismic. To ensure that the rack will not be overturned during the seismic event, the'above critical case is re-run with the two horizontal SSE excitations amplified by a factor of 1.2 and the vertical SSE excitation not amplified. The results of this run are summarized in Table 6.7.52.

The small values of the maximum displacements show that the rack remains stable even when subjected to a stronger horizontal seismic excitation.

6-27 l

Additional' investigation of important structural items is carried out and results are summarized in Table 6. 7.53. ' A discussion of these items follows:

6.7.1.1 Imnact Analyses

a. Imoact Loadina 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 6.7.53 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 6.7.4-6.7.52.

b. Impacts Between Adiacent Racks No non-zaro impact loads are found for the rack-to-rack gap elements (in the cellular region), or for the rack-to-wall elements; it is concluded that no impacts between racks or between racks and walls are likely to occur during a seismic event. This is confirmed by the Whole Pool Multi-Rack results in Section 6.8.

6.7.1.2 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 are surveyed and maximum 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 SSE event,.an allowable weld stress r =-

.42 Su. A comparison of this allovable value with the highest weld stress predicted is given in Table 6.7.53.

The highest predicted weld stress is less than the allowable weld stress value.

6-28

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 not force and moment is such that:

G = Function (F/Fy ,M/Hy ) < l.0 F,

y My are the limit load and moment under direct load only and direct moment only. These values depend on the configuration and on material yield strengths. F, H 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 6.7.53 and is less than the limit value of 1.0. This calculated value is conservatively based on instantaneous peak loading. This value also conservatively neglects the gussets that are provided in the rack modules to increase pedestal area and inertia.

b. Cell-to-Cell Welds cell-to-cell connections are by a series of spot welds along 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 assemblies in adjacent cells are moving out of phase 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 6.7.53 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 6.7.53 and is less than the allowable load. This upper bound value- is-obtained by using the highest rack-to-fuel impact load f rom ' Table 6.7.2 (for any simulation), and multiplying the result by 2 (assuming that.two impact locations are supported by every weld connection).

6.8 Results from Whole Pool Multi-Rack (WPMR) Analyses Figure 6.4.9 shows the Pilgrim spent fuel pool with six new Boltee spent fuel racks (Racks N 1, N2, N3, N4, N5 and N6) and ten existing racks. In the Whole Pool Multi-Rack analysis, a reduced degree-of-freedom ( 8-DOF) model for each rack and its contained fuel is employed. The WPMR dynamic model for Pilgrim Station 6-29 4

i contains 128 degrees-of-freedom and requires e nonlinear analysis.

All racks are assumed to be fully loaded with 680-pound fuel assemblies.

Table 6.8.1 shows maximum corner absolute displacements at both the top and bottom of each rack in global x and y directions (refer to Figure 6.4.9) from the multi-rack runs. As noted previously, a random set of friction coefficients in the range of 0.2 - 0.8 with mean value being 0.5 is used. The seismic loadings consist of the controlling SSE earthquake time-histories as defined in Section 6.3. Table 6.8.2 summarizes the maximum impact force in each gap spring. For each rack in Table 6.8.2, the first 4 springs are the pedestal vertical springs; the last 4 springs are the fuel-to-cell impact springs. The springs from No. 129 to No. 208 are rack-to-rack / wall impact springs at rack top, and the springs from No. 209 to No. 288 are rack-to-rack / wall impact springs at baseplate level. No non-zero values found for impact springs from No. 129 to No. 288 indicate that there is no impact between racks and between rack and pool wall during an SSE seismic event. Table 6.8.3 showc the maximum pedestal stress factors of each of the six Boltee racks in the pool from the WEm analysis.

In Table 6.8.4, the maximum displacement, pedestal vertical loads, and pedestal stress factor obtained from the SSE multi-rack simulation are compared with the limiting single rack analyses.

From Table 6.8.4, it is seen that the maxilnum absolute displacement values are higher than those obtained from single rack analysis. This confirms our earlier speculation regarding the need to perform Whole Pool Multi-Rack analyses to verify that racks do not impact or hit the wall. Table 6.8.5 shows the dynamic pressures on the pool walls. Table 6.8.6 gives the total static load and dynamic load adder on the whole pool slab.

6-30

4 Figures 6.8.1-6.8.6 show the time-histories of rack-to-rack gaps-at typical- locations (see Figure 6. 4. 9 - for_ gap locations) . A survey of all of the rack-to-rack and rack-to-wall impact elements confirms that there are no rack-to-rack or rack-to-wall impacts in the cellular region of any rack (new Holtec racks.as well as existing PNPS racks) in the spent fuel pool. The inter-rack gap elements in the whole pool analysis have an initial gap equal to the actual gap.

Finally, modules N1 and N2 were also simulated with a fully loaded (10,000 lbs. total emplaced load) overhead platform. Analyses using the Whole Pool Multi-Rack simulation show that the rack loaded with the overhead platform does not impact other racks.

The maximum support pedestal stress factor is 0.752 which is less than 40% of the allowable value of 2 for the SSE event. The maximum vertical displacement is the minuscule amount of 0.04741 inches.

Table 6.8.5 gives the results of dynamic pressures on the pool walls. Table 6 . 6' . 6 presents the total static vertical load and dynamic vertical load adder on the whole pool slab. Figure 6.8.7 shows the time-history of the total slab vertical load. These data are used as loadings in qualifying the spent. fuel pool structure (Section 8 of this report).

l The Whole Pool Multi-Rack analyses confirms the conclusions of overall structural integrity derived from the rack simulations.

Because the values of all the stress factors obtained from WPMR analysis for SSE are less than 1.0 and no rack-to-rack / wall impacts are found, it is not necessary to perform the WPMR analysis for OBE seismic.

I l

6-31

6.9 R.a f e rences.

[6.1.1) *0T Position for Review and Acceptance c/

SpentFuel-Storageand'BandlingAppl.tcations"2 dated April 14, 1978, and January 18,. 197(

amendment theretc. i

[6.1.2) USNRC Standard Review Plan, NUREG-0800 (1981)<

[6.1.3) ASME Boiler T. Pressure- Vessel Code, SectioQ '

III, Subsection NF, appendices (1989).

[6.2.1) USNRC Regulatory Guide 1.29, " Seismic Desigg Classification," Rev. 3, 1978.

[6.2.2) Soler, A.I. and Singh, K.P., " Seismic Responses of Free Standing Fuel Rac Construction's to 3-D Motions", Nucleaa Engineering and Design, Vol. 80, pp. 315-32[

(1984).

[6.2.3] Singh, K.P. and Soler, A.I., "SeismiE Qualification of Free Standing Nuclear Fue8 Storage Racks -

the Chin Shan Experience, Nuclear Engineering International, UK (MarcQ 1991). ,

[ 6. 3.1] - " Pilgrim Unit 1 Specification for Seismi$

Response Spectra", Specification Number C-114c)

ER-Q-EO, Boston Edison Company, 1989.

[6.3.2) Boltec Proprietary Report -

Verification and User's Manual for Computer Code GENEQ, Report HI-89364, January, 1990.

[6.4.1) Rabinowicz, E., " Friction Coefficients- of Water Lubricated Stainless Steels for a Spent Fuel Rac). Facility," MIT, a report for Boston Edison Company, 1976.

[6.4.2) Singh, K.P. and Soler, A.I., " Dynamic Coupling in a Closely Spaced Two-Body System Vibrating-in Liquid Medium: The Case of Fuel Racks," 3rd International Conference on -Nuclear Power-Safety, Keswick, England, May 1982.

[6.4.3) Fritz, R.J., "The Effects of Liquids on the Dynamic Motions of Immersed Solids,"

Journal of Engineering for Industry, Trans. of.

the ASME, February 1972, pp 167- 172.

6-32

4 l

[6.4.4) Levy, S. and Wilkinson, J.P.D., "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering," McGraw Hill, 1976.

[6.4.5) Paul, B., " Fluid Coupling in Fuel Racks:

Correlation of Theory and Experiment", Holtec Proprietary Report HI-88243.

[6.6.1) " Dynamics of Structures," R.W. Clough and J.

Penzien, McGraw Hill (1975).

[6.6.2) Soler, A.I., " User Guide for PREDYNA1 and DYNAMO", Holtec Proprietary Report HI-89343, Rev. 2, March, 1990.

[6.6.3) Soler, A.I., " Theoretical Background for Single and Multiple Rack Analysis", Holtec Proprietary Report HI-90439, Rev. O, 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., "DUIARACK Validation Manual, Boltec Proprietary Report HI-91700, Rev. O, October, 1991.

[6.7.1) Singh, K.P., Soler, A.I., and Dhattacharya, S., " Design Strength of Primary Structural Welds in Free Standing Structures", ASME, Journ, of Pressure Vessel Technoloav, August, 1991.

[6.9.1) ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan, 1985.

6-33

I i

Table 6.1.1-LISTING OF PLANTS WHERE DYNARACK WAS APPLIED PLANT DOCKET-NUMBER 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 Unita 1 & 2 USNRC 50-454, 50-455 Braidwood Units 1 & 2 USNRC 50-456, 50-457 Vogtle Unit 2 USNRC 50-425 St. Lucie Unit 1 USNRC 50-335 Millstone Point Unit 1 USNRC 50-245 D.C. Cook Units 1 & 2 USNRC 50-315, 50-316 Indian Point Unit 2 USNRC 50-247 Three Mile Island Unit 1 USNRC 50-289 J.A. Fit: Patrick 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 Sequoyah USNRC 50-327, 50-328 Fort Calhoun USNRC 50-285 l Beaver Valley Unit 1 USNRC 50-334 Nine Mile Point Unit One USNRC 50-220 l

i l

m a

i

~

Table 6.3.1 ,

FOUR SETS OF TIME-HISTORIES GENERATED FROM SAFE SHUTDOWN EARTHQUAKE (SSE) RESPONSE SPECTRA AND THEIR CROSS-CORRELATION COEFFICIENTS Time Nistory File Name Crose Correletten Coefficient..-

SET WQ.

X seismic Y-selse:c- 2 selenic E-W' k S* VT xT MZ ~ Y 1-b 1 e tese.h11 a tase.h12 e tese.vt1 .115201 - . 050290 - 050434 2 a tsse.h21 a tase.h22 e tose.vt2 .114043 - .106427 .067523-3 a tsse.h31 a-tsse.h32 e tese.vt3 .108446 .082160 .088439 4 a tase.h41 a-tsee.h42 e tsse vt4 - - . 112367-

. 053399 .013001--

The directions are defined for mole Poot htti Rack Anstysis (WpMR).

t i

I-l l

l'-

..i

':3 ;

.\

3

.f

.4 Table 6.3.2

-FOUR SETS OF TIME-HISTORIES-GENERATED- 2 FOR OPERATING ~ BASIS EARTHQUAKE (OBE) RESPONSE ~ SPECTRA

.AND THEIR-- -

. CROSS-CORRELATION COEFFICIENTS .

Time Nistory File same ' Cross Correlotten Coefficient.

SET ,

h0.

X seismic' Y sefealc 2 seienic - .

E W" . N 8*

~

VT 'M*Y . XZ- z .Y2 .;

s tobe.h11

^

1 e tabe.h12 e tabe.vt1 ~.112671 .055423 .064538 2 e tobe.h21 e tabe.h22 e tabe.vt2 .111$60  ;.095217-

.061494.

3 e tobe.h31 e tobe.h32 e tobe.vt3 .107059 .105544 .072810 -_.

4 e tobe.h41 e tobe.h42 e tobe.vt4 .073081 .011613 - 012524 4

The directions are defined for Whole Poot Multi Rock Analysis'(WPMR).

Table 6.3.3 DETERMINATION OF THE CONTROLLING SET OF SSE TIME-IIISTORIES ITEM SET 1 sti 2 EET 3 sti 4 1.15xsti 2 Max. total verticat 299965.3 295860.4 300803.8 297107.0 310467.0 pedestel load, lbs.

Max. single pedestat 102653.4 110863.4 99150.0 99141.2 119326.0 vertical load, Lbs.

Max, pedestet shear 15905.9 20463.2 16373.2 17170.7 27444.2 -

toad, tbs.

kam cell fuel lapset 344.5 344.9 ' 352.8 372.6 391.5 loed, tbs.

asck wall i mact at 0 0 0 0 0 bottom, tbs.

Rack rock Irpact at 0 0 0 0 0 bottora, tbs.

Rock-wall imect at 0 0 0 0 0 top, tbs.

Beck rock treact at 0 0 0 0 0 top, lbs.

Max. displacement in .0941 .1109 .0753 .0772 .1336 x, at top, In.

Max displacement in in y, et top, in. .0613 .0895 .0914 .0837 .1074 Max. displacement in in x, et bottom, in. .0035 .0042 .0023 .0029 .0051 Max. displacement in in y, at tnttom, in. .0023 .0034 .0034 .0032 .0041 Max. stress factor 96, above baseplete .040 .047 .043 .040 .055 Max. stress factor t6, pedestal .162 .185 .162 .158 .203 Conditions: Pilgrim asek u1; fully loaded with 6808 fuel asseattles; pedestat friction coef ficient 0.8; opposed-phase motion assumed.

e

~

' Table 6.4.1:

DEGREES-OF-FREEDOM-Displacement: Rotation-

- Location Ux _. Dy- U

-z 8x By Sz (Node)-

1 P1- P2 P3 94 .45 '96 2 P17 -P18 P19 q20 _ q21. 422!-

Point 2 is assumed attached to rigid rack lat) the top most point. ,

2* P7 P8 - '

3* P9 P10  ;

4* pil P12 5* P13 P14 1* p15 P16 where:

Pi " -

qi(t)-+ U l(t) _11=fl,7,9,11,13,15,17

= qi(t);+'U 2(t) _i =:2,8,10,12,14,16,18

-=- qi(t) + U 3(t) 1'= 3,19:-

Ul(t) are the 3 known earthquake displacements.

I'

?

?

l i

4 s -.m, n. , -_m. .,- -e > ,r- 4 m

Table 6.4.2 NUMBERING-SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS I. Nonlinear Sorinos (Gao Elements) (64 Total)_

Number Node Location -peserlotion 1 Support S1 z compression only element 2 Support S2 2 compression only element 3 Support S3 z compression'only element 4 Support S4 2 compression only element 5 2,2* X rack / fuel _ assembly impact element 6 2,2* X rack / fuel assembly impact element 7 2,2* Y rack / fuel assembly impact element 8 2,2* Y rack / fuel - assembly impact element 9-24 -Other rattling masses for nodes 1*, 3*, 4* and 5*

25 Bottom cross- Inter-rack impact elements-section of rack (around edge)

Inter-rack impact elements

. 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-section Inter-rack impact elements

. of rack Inter-rack impact elements

. (around edge) Inter-rack impact elements Inter-rack impact _ elements Inter-rack impact elements Inter-rack impact elements Inter-rack impact elements 64 Inter-rack impact elements

1

. 1 Table 6.4.2-(continued)

NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS

-II. Friction Elements (16 total)

Number Node Location Descriotion 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.

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

Table 6.4.3 SPENT FUEL POOL LOADING Fuel Cell Configuration Rack Assembly (No. of Calls in Woight Weight Rack NS x EW Direction) (lb:s) (lba)

HEW RACKRt Camnaian I:

N1 18x16 29400 680 N2 18x15 28600 680 Campaian II:

N3 19x14 27100 680 N4 19x13 25200 680 N5 19x13 25200 680 N6 16x13 21300 680 XXISTING RACKS:

El 19x14-13x4 23600 680 E2 19x14-9x4 25200 600 E3 19x14-3x9 31700 680 E4 19x14 29000 680 E5 19x14 29000 680 E6 19x14 29000 680 E7 19x14 29000 680 E8 19x14 29000 680 E9 19x14 29000 680 E10

-Tablei6.5.1 RACK MATERIAL -DATA (200'F);

-Young's  ; Yield . -Ultimate-Modulus Strength' Strength Material- E (psi)-- Sy (psi). Su-(Psi)

ASME SA240-304* 27.5 x-106 - 25000 -71000 ASME .

Section III Table Table- Table Reference I-6.0 I-2.2 I-3.2 SUPPORT MATERIAL DATA Young's Yield Ultimate Modulus - Strength- Strength-Material E (psi) Sy-(psi) -Su'(Psi) 1 -ASME SA240-304* 27.6x106 25,000 71~,000 :

(upper part psi . psi- psi .

of support l feet) 2 ASME SA564-630 27.6x106 106,300 140,000- ,

(lower part of psi psi- psi' support feet;-

age hardened at 1100*F)

ASMESection III Table Table Reference I-2.1 I-3.1

'ASME SA240-304L is ordered with mechanical properties equalE to ASTM-240, Type 304.- '

k

.- = .-- ..,: , , , , , , - - . - . - - - . - . ,

, e-

.n

l l

~

Table 6.7.1

-RESULTS OF SINGLE RACK ANALYSES List of All Runs 4 Holtec Rack Fuel Fuel Loading Seismic: Coefficient Motion-Run I.D. I.D. I.D. Condition Loading of Friction -Mode / ,

drnissei.rf8 N1 regular- Fully Loaded' SSE 10.8 in-phase 288 cells drnissai.rf5 N1 regular Fully Loaded - SSE 0.5 in-phase 288 cells drnissei.rf2 N1 regular Fully Loaded SSE 0.2 Lin-phase 288 cells drnissei.rh8 N1 regular Half Loaded SSE. 0.8 .in-phase-144 cells:

drnissei.rh5 N1 regular- Half Loaded SSE 0.5 in-phase 144 cells drnissei.rh2 N1 regular Half Loaded SSE 0.2 in-phase 144 cells drnissei.re8 N1 regular " Empty " SSE 0.8 in-phase 16 cells loaded drnissei.re5 N1 regular " Empty "

SSE 0.5 in-phase:

16 cells loaded drnissei.re2 N1 regular " Empty "' _ SSE -0.2' _in-phase:-

16. cells loaded drnisseo.rf8 N1 regular " Fully Loaded: SSE 0.8 -~

opposed 288 cells' . phase:

Edrnisseo.rf5 N1 regular Fully Loaded- SSE 0.5 opposed :

288' cells _ phase. ,

-drnisseo.rf2 N1 regular _ FullyLLoaded 'SSE 0.2_  :-opposed-288 cells ~ phase

( to be continued )

k

' Table 6.7.1 '(~ continued')-

Holtec Rack Fuel Fuel Loading Seismic- Coefficient Moti Run I.D. I.D. I . D. - Condition Loading iof Frictions Mod drnissno.rh8 N1 regular Half Loaded SSE 0.8: coppe 144. cells >pha drnissoo.rh5 N1 regular. Half Loaded SSE 0. 5' 144 cells opph Eph drnisseo.rh2 N1 regular Half Loaded -SSE 0.2 opp 3 144 cells pha drnisseo.rea N1 regular. " Empty " SSE '.8 oppo!

16 cells loaded phad drnisseo.re5- N1 . regular. " Empty " . SSE 0.5 oppoy 16 cells loaded phas drnisseo.re2 N1. regular " Empty =" SSE 0.2 oppos 16 calls loaded phac drn5ssei.rf8 NS regular Fully-Loaded .SSE' O.8' in-phat 247 cells-drn5ssei.rf5 NS regular ' Fully Loaded .SSE 0.5 .in-phat 247 calls.

drn5ssei.rf2 N5 regular Fully. Loaded SSE- 0.2- in-phas 247 cells drn5ssei.rh8 N5 regular Half Loaded SSE 0.8 Ein-phat 117' cells drn5ssei.rh5 H5. regular Half Loaded- SSE 0.5 in-phat 117' cells drn5ssei.rh2 N5 regular Half Loaded SSE- -0.2 ~1n-phas 117 cells

( to~be continued _)

Table' 6.7.1 ~ ( - continued-_ )

-Holtec Rack Fuel Fuel Loading Seismic _ Coefficient Motion Run I.D. I.D. I.D. Condition Loading of Friction. Mode drn5ssei.re8 N5 regular " Empty "- SSE 0.8i :in-phasd 13 cells-loaded drn5ssei.re5 N5 regular " Empty " SSE- 0.5 in-phase 13 cells loaded drn5ssei.re2 NS regular "Emhty" . SSE 0.2 in-phasi 13 cells loaded drn5sseo.rf8 N5 regular Fully Loaded -SSE _0.8 opposeci 247-cells phase drn5sseo.rf5 N5 regular Fully Loaded- SSE 0.5 opposed i

i 247 cells phase?

drn5sseo.rf2 NS regular Fully Loaded- SSE 0.2 =. opposed 247 cells phase:

drn5sseo.rh8 N5 regular Half Loaded SSE 0.8 opposed-117 cells phase:

drn5sseo.rh5 NS regular Half Loaded- SSE 0.5 opposed 117 cells _ phase =

drn5sseo.rh2 NS regular Half Loaded SSE' O.2 opposadi 117 cells

-phase ll l

-drn5sseo.re8 NS regular " Empty " SSE- O.8 opposedi 13 cells loaded phase drn5sseo.re5- N5 regular " Empty _" SSE 0.5 opposedl 13 cells loaded - _ phase.;

drn5sseo.re2 NS regular " Empty " _SSE 0.2 -opposed 13-cells-loaded phase:

( to be continued )

Table 6.7.1 ( continued )

Holtec Rack Fuci Fuel Loading Seismic Coefficient Moti Run I.D. I.D. I.D. Condition Loading of Triction Mot:

dre9ssei.rf8 E9 regular Fully Loaded SSE 0.8 in-ph.

Existing 266 cells dre9ssei.rf2 E9 regular Fully Loaded SSE Existing 0.2 in-pha 266 colls dre9ssei.rh8 E9 regular Half Loaded SSE 0.8 in-pht Existing 126 cells dre9ssei.rh2 E9 regular Half Loaded Existing SSE 0.2 in-pha 126 cells drc9ssei.re8 E9 regular " Empty " SSE 0.8 in-pha Existing 14 cells loaded dre9ssei.re2 E9 regular " Empty " SSE 0.2 in-phc Existing 14 cells loaded dre9sseo.rf8 E9 regular Fully Loaded SSE Existing 0.8 oppose 266 cells phase dre9sseo.rf2 E9 regular Fully Loaded SSE Existing 0.2 oppose 266 cells phase dre9sseo.rh8 E9 regular Half Loaded SSE Existing 0.8 oppose 126 cells phase dre9sseo.rh2 E9 regular Half Loaded SSE Existing 0.2 opposc 126 cells phasc dre9sseo.re8 E9 regular " Empty " SSE 0.8 Existing oppose <

14 cells loaded phase drc9sseo.re2 E9 regular " Empty " SSE Existing 0.2 oppose:

14 cells loaded phase

( to be continued )

• Table 6.7.1 ( continued )

Holtec Rack Fuel Fuel Loading "eismic Coefficient Motion Run I.D. I.D. I.D. Condition Loading of Friction Mode drn5ssei.12h N5 regular Fully Loaded 1.2xSSE 0.5 in-phase:

247 cells

( Note: In this run, the two horizontal time-histories are multiplied b; a factor of 1.2 for check of overturning ).

^

Table 6.7.2

SUMMARY

OF WORST RESULTS FROM 36 RUNS OF-SINGLE RACK ANALYSIS-FOR HOLTEC RACKS

( LOADED WITH REGULAR FUEL ASSEMBLIES; ANALYSIS BASTS SSE SEISMIC )

Item Value Run I.D.

1. Maximum total vertical pedestal load: 310,467 lbs. drnissei.rf8

2. Maximum vertical load in any single pedestal: 123,803 lbs. drn5ssei.rf5

3. Laximum shear load in any single pedestal: 34,615 lbs. drn5sseo.rf5

4. Maximum fuel assembly-to-cell wall impact load at one local position: 421 lbs. drn5ssei.rh2

5. Maximum rack-to-wall impact load at baseplat level: 0 lbs.

6. Maximum rack-to-vall impact load at the top of rack: 0 lbs.

7. Maximum rack-to-rack impact load at baseplat level: 0 lbs.

8. Maximum rack-to-rack impact load at the top of rack: 0 lbs.

9. Maximum corner displacements Top corner in x direction: 0.2054 in.

in y direction: drn5ssei.rf5 0.1821 in. drn5ssei.rf2 Baseplate corner in x direction: 0.0258 in. drnissei.rf2-in y direction: -0.1000 in. drn5ssei.rf2

10. Maximum stress factors Above baseplate:

Support pedestals: 0.073 (R6) drn5ssei.rf5 0.228 (R6) drn5ssei.rf5

Tabic, 6.7.3 9UMMARY OF WORST RESULTS FTDM 36 RUNS OF SINGLE RACK AllALYSIS FOR EXISTING RACKS

( LOADED WITH REGULA't FVEL ASSEMBLIES; ANALYSIS BASIS SSE SEISMIC )

l Item Value Run I.D.

1. Max: ,.:m total vertical pedesta'. load: 293,265 lbs. dre9ssai.rf8

2. Maximum vertical load in any single pedestal: 117,819 lbs. dre9ssei.rf8

3. Maximum shear load in any single pedestal: 24,229 lbs. dre9sseo.rf8

4. Maximum fuel assembly-to-cell vall

_ impact load at one local position: 277 lbs. dre90seo.rh2

5. Maximum rack-to-vall impact load at baseplat icvel: O lbs.

6. Maximum rack-to-wall impact load at the top of rack: O lbs.

7. Maximum rack-to-rack impact load at baseplat level: O lbs.

8. Maximum rack-to-rack impact load at the top of rack: O lbs.

9. Maximum corner displacements Top corner in x direction: 0.1371 in, dre9ssei.rf2 in y direction: 0.3881 in. dre9ssei.rh8 Baseplate cerner in x direction: 0.0213 in. dre9ssei.rf2 in y direction: 0.0256 in. dre9ssei.rf2

10. Maximum stress factors Above baseplate: 0.166 (R6) dresssui.rf2 Support pedestals: 0.245 (R6) dre9ssei.rf8

Iable 6.7.4

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR TACK MODULE: RACK-N1 Holtoc Run I.D.: drnissei.rf8 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 288 cells loaded; Fuel centroid X,Y: . 0, .0 (in.).

Coefficient of friction at the bottom of lupport pedestal 0.8

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.)YNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 310467.0 (2) Maximum vertical load in any single pedestal: 104861.2 (3) Maximum shear load in any single pedestal: 25915.9 (4) Maximum fuel-cell impact at one local position: 397.3 (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-r9Ck impact at rack top: .0 MAXIMUM CORNER DISPLA IEMENTS (in.)

Location: X-direction Y-direction Top corner: .0794 .1054 Baseplate corner: .0030 .0040 MAXIMUM STRESS FACTORS

• Strecs factor: R1 R2 R3 R4 R5 R6 R7 Abovo baseplate: .012 .008 .033 .023 .040 .046 .009 Support pedestal: .154 .046 .050 .043 .168 .174 .054

• See Section 6.5.2.3 of the Licensing Report for definitions.

~

- . - ._. -.__ _ _ _ _ _ _ . . _ _ _ _ _ . - _ . _ . _ - - _ _ . - - _ .m._._-..____m._

- - . - - - _ . ~ . - - . _ - _ - - . . - - - , - - .. - - - . - -

)

Table 6.7.5

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACKRACK-q MODU Holtec Run I.D.: drnissei.rf5 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight Reg.680#  ; 680.0 (1bs.)

Fuel Loading: 288 cells loaded; Fuel centroid X,Y: . 0, .0 (in.)

Coefficient of friction at the bottom of support pedestal 0.5

$ Revision: 3.46 S SLogfile C:/ racks /dynam0/ dynamo.fov -$

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DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 310467.0 (2) Maximum vertical load in any single pedestal: 106173.2 (3) Maximum shear load in any single pedestal: 23020.2 (4) Maximum fuel-cell impact at one local position: . 397.5 (5) Maximum rack-to-wall impact at baseplater .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: .1015 Baseplate corner: . 0985

.0038 .0037 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .012 .009 .032 Support pedestal: .027 . 039 .045 .010

.156 .040 .047 .037 . 173 .181 .051

• See Section 6.5.2.3 of the Licensing Report for definitions.

1 Table 6.7.6 SUMMAPV RESULTS OF 3-D SINGLE RACK ANALYSIG FOR RACK MODULE: RACK-N1 Ho1L.; Run I.D.: drnissei.rf2 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 288 cells loaded; Fuel centroid X,Y: .0, .0 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

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~

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 310467.0 (2) Maximum vertical load in any single pedestal: 113436.9 (3) Maximum chear load in any single pedestal: 21290.0 (4) Maximum fuel-cell impact at one local position: 396.2 (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: .1322 .1088 Baseplate corner: .0258 .0276 Y.AXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .012 .007 . 030 .036 .043 .049 .009 Support pedestal: .167 .042 . 041 .039 .185 .193 .045

• See Section 6.5.2.3 of the Licensing Report for definitions.

__._______-.-______m_-___-__m_

Table 6.7.7

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N; Holtec Run I.D.: drnissei.rh8 Seismic Loading: 1.15xsSE-2 Fuel Assembly I.D. and Weight: Reg.680/ / 680.0 (lbs.)

Fuel Loading: 144 cells leaded; Puel centroid X,Y: .0,-28.3 (in.)

Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.46 S SLogfile: C:/ racks / dynamo / dynamo.fov S SRevision: 2.5 S SLogfile: C:/ racks /dynam0/dynasi.fov $

SRevision: 3.36 S SLogfile: C:/ racks /dynam0/dynac2.fov $

DYHAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 186840.3 (2) Maximum vertical load in any single pedestal: 82067.6 (3) Maximum shear load in any single pedestal: 21862.8 (4) Maximum fuel-cell impact at one local position: 368.6 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-vall 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: .0596 Baseplate corner: .1170

.0022 .0028 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .009 .005 .020 .015 .026 Support pedestal: .029 .006

.120 .027 .044 .025 .154 .162 .047

• See Section 6.5.2.3 of the Licensing Report for definitions.

T

-)

i Table 6.7.8 t i

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACX-N1 Holtec Run I.D.: drnissai.rh5 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (1bs.)

Fuel Loading: 144 cells loaded; Fuel centroid X,Y: .0,-28.3 (in.)

Coefficient of friction at the bottom of support pedestal: 0.5

$ Revision: 3.46 S

$Logfile C:/ racks /dynam0/ dynamo.fov $

$Ruvision: 2.5 $

SLogfiles C:/ racks /dynam0/dynani.foy $

$ Revision: 3.36 $

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 186840.0 (2) Maximum vertical lond in any single pedestal: 85607.1 (3) Maximum shear load in any single pedestal: 24038.5 (4) Maximum fuel-cell impact at one local position: 356.3 (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-directica Top corner: .0610 Baseplate corner .1115

.0022 .0026 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .009 .005 .020 .031 Support pedestal:

.015 .035 .006

.125 .028 .047 .026 .153 .162 .051

• See Section 6.5.2.3 of the Licensing Report for definitions.

_ _ . - _ __ _ _ _ . _ _ _ _ . . _ _ _ . _ _ - - - . _ _ ~ . _

Table 6.7.9

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N1 Holtec Run I.D.: drnissei.rh2 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680#  ; 680.0 (1bs.)

Fuel Loading: 144 cells loaded; Puol centroid X,Y: .0,-28.3 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2 SRevision: 3.46 $

$Logfile: C:/ racks /dynam0/ dynamo.fov $

$ Revision: 2.5 $

$Logfile: C / racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

,1) Maximum total vertical pedestal load: 187218.9 (2) Maximum vertical load in any single pedestal: 81920.1 (3) Maximum shear load in any single pedestal: 16230.0 (4) Maximum fuel-cell impact at one local position: 371.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-vall 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: .0865 .1114 Baseplate corner: .0144 .0157 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .009 .004 .019 .021 .035 .040 .005 Support pedestil: .120 .030 .033 .028 .140 .146 .036

• See Section 6.5.2.3 of the Licensing Report for definitions.

l l

1

_ _ - _ - _ - _ _ - _ - - - - - - . - - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ^

Table 6.7.10 1

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODUL2: RACK-N1  ;

Holtec-Run I.D.: drnissei.re8 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 16 cells loaded; Fuel centroid X,Y: .0,-53.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.46 $

$Logfile: C / racks /dynam0/ dynamo.foy $

SRevision: 2.5 $

4 SLogfile: C / racks /dynam0/dynasi.fov $

$ Revision 3.36 $

$Logfile: C:/ racks /dynam0/dynas2.foy $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 48833.0 (2) Maximum vertical load in any single pedestal: 22118.1 (3) Maximum shear load in any single pedestal: 6988.5 (4) Maximum fuel-cell impact at one local position: 396.5 ,

(5) Maximum rack-to-vall impact at baseplate: .0 (6) Maximum rack-toowall 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 I Top corner: 0241 Baseplate corner:

.0325 i

. 0008 .0009 MAXIMUM STRESS FACTORS

• Stress. factor: R1 R2 R3 R4 R5 R6 R7 L Above baseplate: .004 .001 . 007- .006 .010 .011 .002 l Support pedestal: .032 .009 014 .008

.042 .045 .015

• See Section 6.5.2.3 of the Licensing Report for definitions.

p

.,,e . - - .,--y~, .,, - , , - - - yc;-..m w~-yy- y,.,,-.,,,,-w-, --

Table 6.7.11 SUmiARY RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACE-N1 Holtec Run-I.D.: drnissei.re5 Seismic Loading: 1.15xSSE-2 Puel Assembly I.D. and Weight: Reg.680#  ; 680.0 (lbs.)

Fuel Loading: 16 cells loaded; Puel centroid X,Y: (in.)

.0,-53.4 Coefficient of friction at the bottom of support pedestal: 0.5 SRevision: 3.46 $

$Logfile C:/ racks /dynam0/ dynamo.fov $ .

$ Revision: 2.5 $

SLogfile: C / racks /dynam0/dynasi.foy $

$ Revision: 3.36 $ e SLogfile: C:/ racks /dynam0/dynas2'.fov $

DYNAMIC IMPACT LOADS (1bs.)

(1) Maximum total vertical pedestal load: 48832.9 (2) Maximum vertical load in any single pedestal: 21988.1

^

(3) Maximum shear-load in any single pedestal: 7306.5 (4) Maximum fuel-cell impact at one local position: 385.6 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-vall 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 l

Top corner: .0237 .0325 Baseplate corner: .0008 .0009 MAXIMUM STRESS FACTORS

• Stress factor:- R1 R2 R3 R4 R5 R6 R7 Above baseplate: .004 .001 .007 -.006 .010 .011 .002' Support pedestal: .032 .011 .014 .010 .041 .044 .015 ,

• See Section 6.5.2.3 of the Licensing Report for definitions.

i l -

Table 6.7.12

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACX MODULE: RACK-N1 l

\

Holtec Run I.D.: drnissei.re2 Seismic Loading: 1.15xSSE-2 l Fuel Assembly I.D. and'Waight Reg.680/  ; 680.0 (lbs.)

Puol Loading: 16 cells loaded; Puel centroid X,Y: .0,-53.4 (in.)

Coefficient of-friction at the bottom of support podestal: 0.2 '

SRevision: 3.46 S SLogfiles C:/ racks / dynamo / dynamo.fov $

$ Revision: 2.5 S

$Logfile: C:/ racks /dynam0/dynasi.foy $

$ Revision: 3.36 $

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.) ,

(1) Maximum total vertical pedestal load: 48837.4 (2) Maximum vertical load in any single pedestal: 20857.1 (3) Maximum shear load in any single podestal: 4017.2 (4) Maximum fuel-cell inpact at one local position: 370.8 (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: .0329 .0317 Baseplate corner: .0160 .0089 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .004 .001 .006 .006 .009 .010 .001 Support pedestal: .030 .008 .008 .008 .036 .037 .008

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.13

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULEt RACX-N1 Holtec Run I.D.: drnisseo.rf8 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weights Reg.680#  ; 680.0 (1bs.)

Puel Loading: 288 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

$Rovision: 2.5 S 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: 310467.0 (2) Maximum vertical lead in any single pedestal: 119326.0 (3) Maximum shear load in any single pedestal: 27444.2 (4) Maximum fuol-cell impact at one local position: 391.5-i (5) Maximum rack-to-vall impact at baseplates .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: .1336 . 1074 Baseplate corner: .0051 . 0041 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .012 .009 .033 .035 .048 .055 .010 Support pedestal: .175 .048 .050 .203

-.044 .054

.194

• See Section 6.5.2.3 of the Licensing Report for definitions.

t y , _y. , . . _ . . . . _ , , _ - - - - ,,,,__;,,

.J.

Table 6.7.14

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N1 Holtec Run I.D. drnisseo.rf5 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 288 cells loaded; Fuel centroid X,Y: . 0, .0 (in.)

Coefficient of friction at the bottom of support pedestal: 0.5 SRevision: 3.46 $

$Logfile C / racks /dynam0/ dynamo.fov $

$ Revision: 2.5 $

$Logfile C:/ racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

$Logfile: C / racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 310467.0 (2) Maximum vertical load in any single pedestal: 112673.2 (3) Maximum shear load in any single pedestal: 30300.1 (4) Maximum fuel-cell impact at one local position: 391.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 l

(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 .0911- .0961 Baseplate corner: .0035 .0036 1

MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7

Above baseplate

.012 .009 .031 .025 .041 .046 .009

Support pedestal

.165. .043 .0C2 .040 .187 .196 .067 i

• See Section 6.5.2.3 of the Licensing Report for definitions. '

l l

_ , - a,-, r-,---v '

Table 6.7.15

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N1 Holtec Run I.D.: drnisseo.rf2 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (1bs.)

Fuel Loading: 288 cells loaded; Fuel centroid X,Y: .0, .0 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$ Revision: 3.46 S

$Logfile: C:/ racks /dynam0/ dynamo.fov $

$ Revision: 2.5 $

SLogfile: C:/ racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

$Logfile: C / racks /dynam0/dynas2.fov $

DYNAMIC IMPACI' LOADS (1bs.)

(1) Maximum total vertical pedestal load: 310467.0-(2) Maximum vertical load in any single pedestal: 114227.6 (3) Maximum shear load in any single pedestal: 22428.2-(4) Maximum fuel-cell impact at one local position: 390.2 (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.)

Location: X-direction Y-direction Top corner: .1414 .1293 Baseplate corner: .0224 .0341 MAXIMUM STRESS FACIORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .012 .008 .033 .036 .043 .049 .008 Support pedestal: .168 .047 .037 .044 .199 .208 .040

• See Section 6.5.2.3 of the Licensing Report for definitions.

l l

C_._.---.-.____..~__._..-.__-.----_--:--------- - - -

?

Table 6.7.16

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N1 Holtec Run I.D.: drnisseo.rh8 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 144 cells loaded; Fuel centroid X,Y: .0,-28.3 (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

~

SRevision: 3.46 $

$Logfile: C:/ racks /dynam0/ dynamo.fov $

$ Revision: 2.5 $

$Logfile: C:/ racks /dynam0/dynasi.fov $

$ Revision: 3.36 $ ,

SLogfile: C: / racks /dynam0/dynas2. foy $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 171519.2 (2) Maximum vertical load in any single pedestal: 78964.5 (3) Maximum shear load in any single pedestal: 20257.2 (4) Maxinum fuel-cell impact at one local position: 387.1 (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 baceplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0785 .1038 Baseplate corner: .0030 .0025 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .,

.005 .021 .019 .026' -.030 .005 Support pedestal: .126 .036 .039 .033 .135 .144 .043

• See Section 6.5.2.3 of the Licensing Report for definitions.

l l.

Table 6.7.17

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOL RACK MODULE: RACK-N1 Holtec Run I.D.: drnisseo.rh5 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680#  ; 680.0 (lbs.)

Fuel Loading: 144 cells loaded; Fuel centroid X,Y: .0,-28.3 (in.)

Coefficient of friction at the bottom of support pedestal: 0.5 SRevision: 3.46 $

SLogfile: C:/ racks /dynam0/ dynamo.fov $

$ Revision: 2.5 $

SLogfile: C:/ racks /dynam0/dynasi.foy $

$ Revision: 3.36 $

SLogfile C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vartical pedestal load: 171519.2 (2) Maxir"m vertical load in any single pedestal: 78217.3 (3) Maximum shear load in any single nedestal: 17516.9-(4) Maximum fuel-cell impact at one local position: 391.5 (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 basep1F.ce: .0 (8) Maximum rack-to-rack impact at rack top: .0

~

MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction -

Top corner: .0751 .1029 Baseplate corner: .0028 .0023 MAXIMUM STRESS FACTORS

• Stress factor: R1 R3 R3 R4 R5 R6 R7 Above baseplate: .007 .005 .019 .018 . 026 .030 .005 Support pedestal: .115 .030 .031 .028 133 .140

. .034

• See Section 6.5.2.3 of the Licensing Report for definitions.

.l Table 6.7.18

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N1  !

Holtec Run I.D.: drnissoo.rh2 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/ / 680.0 (lbs.)

Fuel Loading: 144 cells loaded; Fuel centroid X,Y: .0,-28.3 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$ Revision: 3.46- S i SLogfiles C:/ racks /dynam0/ dynamo.fov $

SRevision: 2.5 S SLogfile: C / racks /dynam0/dynasti.fov $

$ Revision 3.36 $

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 171517.7 (2) Maximum vertical load in any single pedestal: 79445.0 (3) Maximum shear load in any single pedestal: 14769.7 (4) Maximum fuel-cell impact at one local position: 395.1 (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: .0698 .0828 -

Baseplate corner: .0074 .0109 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above basoplate: .007 .004 . 017 .016 .027 .030 .004 Support pedestal: .116 .031 . 029 .029 .130 .136 .031

• See Section 6.5.2.3 of the Licensing Report for definitions.

L I

4  ; . - , r---.>r---- ,v.. , ,,-r, - , , - , -- -,- = , - - , , -

Table 6.7.19

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-lO Holtec Run I.D.: drnisseo. rob Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 600.0 (lbs.)

Fuel Loading: 16 cells loaded; Fuel centroid X,Y: .0,-53.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

$ Revision: 3.46 S SLogfile: C:/ racks / dynamo / dynamo.fov $

$ Revision: 2.5 $

$Logfile: C:/ racks /dynam0/dynasi.fov $

$ Revision:

3.36 $

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (1bs.)

(1) Maximum total vertical pedestal load: 48842.6 (2) Maximum vertical load in any single pedestal: 24172.7 (3) Maximum shear load in any single pedestal: 5547.5 (4) Maximum fuel-cell impact at one local position: 406.5 (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: .0282 Baseplate corner:

.0262

.0010 .0005 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .004 .002 .005 .006 .011 .012 .001 Support pedestal: .035 .012 .009 .011 .038 .039 .010

• See Section 6.5.2.3 of the Licensing Report for definitions.

4 Table 6.7 :D

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N1 4

Holtec Run I.D.: drnisseo.re5 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (1bs.)

Fuel Loading: 16 calls loaded; ruel centroid X,Y: .0,-53.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.5 ,

$ Revision: 3.46 $

$Logfile: C: / racks /dynam0/dynam0.f ov $

$ Revision: 2.5 $

$Logfile _C / racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

$Logfile C:/ racks /dynam0/dynas2.foy $

DYNAMIC IMPACT LOADS (1bs.)

(1) Maximum total vertical pedestal load: 48842.6 (2) Maximum vertical load in any single pedestal: 23937.4 (3) Maximum shear load in any Jingle pedestal: 5592.9 (4) Maximum fuel-cell impact at o.ie local position: 369.2 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rcck-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-d'irection Y-direction Top corner: .0286 . 0263 Baseplate corner: .0010 . 0006 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .004 .002 .005 .006 .011 .012 .001 Support pedestal .035 .012 .009 .011 .039 .040 .009

• See.Section 6.5.2.3 of the Licensing Report for definitions.-

b

- . - - . -..,- -..-- .,-. - - - - , - , - , n, ., w- -

r - , - . .

Table 6.7.21

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N3 Holtec Run I.D.: drnisseo.re2 Seismic Loading: 1.15xSSE-2 Puel Assembly I.D. and Weight: Reg.680#  ; 680.0 (lbs.)

Fuel Loading: 16 cells loaded; Fuel centroid X,Y: .0,-53.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$ Revision: 3.46 S SLogfile: C: / racks /dynam0/dynam0. f ov S SRevision: 2.5 S SLogfile: C:/ racks /dynam0/dynasi.fov S SRevision: 3.36 S

$Logfile: C:/ racks /dynam0/dynas2.fov S DYNAMIC IMPACT LOADS (1bs.)

(1) Maximum total vertical pedestal load: 48844.6 (2) Maximum vertical load in any single pedestal: 21580.6 (3) Maximum shear load in any single pedestal: 4203.8 (4) Maximum fuel-cell impact at one local position: 391.5 (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 DISPLACEMDITS (in.)

Location: X-direction Y-direction Top corner: .0250 .0255 Baseplate corner: .0067 .0100 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 l Above baseplate: .004 .001 .004 .006 .009 .010 .001 Support pedestal: .032 .009 .008 .039

.008 .037 .009

~

• See Section 6.5.2.3 of the Licenhing Report for definitions.

Table 6.7.22

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N5 Holtec Run 1.D.: drn5ssei.rf8 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; G80.0 (lbs.)

Fuel Loading: 247 cells loaded; Fuel controid X,Y: .0, .0 (in.)

Coefficient of friction at the bottom of support pedestal 0.8

$ Revision: 3.46 $

SLogfile: C:/ racks /dynam0/ dynamo.fov $

$ Revision: 2.5 $

$Logfile: C:/ racks /dynam0/dynasi.fov $

SRevision: 3.36 $

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 262418.2 (2) Maximum vertical load in any single pedestal: 114997.1 (3) Maximum shear load in any single pedestal: 23460.1 (4) Maximum fuel-cell impact at one local position: 403.4 (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: .1467 .1293 Baseplate corner: .0056 .0047 MA"IMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .011 .008 .040 .033 .057 .066 .009 Support pedestal: .169 .044 .042 .040 .195 .203 .045 See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.23

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N5!

Holtec Run I.D.: drn5ssai.rf5 Seismic Loading: 1.15xSSE-2 i Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 247 calls loaded; Puol centroid X,Y: .0, .0 (in.) ;

Coefficient of friction at the bottom of support pedestal: 0.5 SRevision: 3.46 $

SLogfile C:/ racks /dynam0/ dynamo.fov $

$ Revision: 2.5 $

$Logfile: C / racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

$Logfile C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 266122.4 (2) Maximum vertical load in any single pedestal: 123833.3 (3) Maximum shear load in any single pedestal: 28294.5 (4) Maximum fuel-cell impact at one local position: 399.0 (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 MAXIMUM CORNER DISPLACEMDiTS (in.)

Location X-direction Y-direction Top corner: .2054 . 1333 Baseplate corner: .0078 . 0049 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .012 .009 .039 .045 .064 .073 .009 Support pedestal: .181 .057 .047 .052 .218 .228 .050 l

• See Section 6.5.2.3 of the Licensing Report for definitions.

k

__ . , _ , - - --~s , e *' "

. . - .. ._ - - _. -. . -- ~_ -

Table 6.7.24

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-NS Holtec Run I.D.: drn5ssei.rf2 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/ / 680.0 (lbs.)

Fuel Loading: 247 cells loaded; Fuel centroid X,Y: .0, .0 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$ Revision: 3.46 S

$Logfile: C / racks /dynam0/dynam0.foy $

$ Revision: 2.5 $

$Logfile C / racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

$Logfile C / racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 262418.1 (2) Maximum vertical load in any single pedestal 109024.2 (3) Maximum shear load in any single pedestal: 21063.7 (4) Maximum fuel-cell impact at one local position: 390.3 (5) Maximum rack-to-wall impact at baseplates .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: .1663 .1821 Baseplate corner: .0173 .1009 MAXIMUM STRESS FACTORS *

-Stress factor: R1 R2 R3 R4 R5 R6 R7 l Above baseplate: .011 .009 .031 .038 .050 .058 .009 l Support pedestal: .160 .040 .038 .037 .187 .196 .041

~~

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.25

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK HODULE: RACK-NE Moltec Run I.D.: drn5ssel.rh8 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 117 cella loaded; Fuel centroid X,Y: .0,-31.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

$ Revision: 3.46 S

$Logfile: C:/ racks /dynam0/ dynamo.fov $

$ Revision: 2.5 S SLogfile: C:/ racks /dynam0/dynasi.fov $

SRevision: 3.36 $

SLogfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 144728.2 (2) Maximum vertical load in any single pedestal: 87001.6 (3) Maximum shear load in any single pedestal: 21787.6 (4) Maximum fuel-cell impact at one local position: 383.6 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rh-% impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMDiTS (in.)

Location: X-direction Y-direction Top corner: .1453 .1613 Baseplate corner: .0054 .0045 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .007 .004 .024 .024 .038 .044 .005 Support pedestal: .127 .033 .044 .030 .142 .149 .048

• See Section 6.5.2.3 of the Licensing Repcrt for definitions.

4

.- \

Table 6.7,26 l l

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N5 l

E~oltec Run I.D.: drn5ssei.rn5 Seismic Loading: 1.15xSSE-2 '

Fuel Aesembly I.D. and Weight: Reg.680#  ; 680.0 (1bs.)

Fuel Loading: 117 cells loaded; Fuel centroid X,Y: .0,-31.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.5

$ Revision: 3.46 S

$Logfile: C:/ racks /dynam0/ dynamo.fov $

$ Revision: 2.5 $

$Logfile: C:/ racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 144727.9 (2) Maximum vertical load in any single pedestal: 92537.6 (3) Maximum shear load in any single pedestal: 25295.4 (4) Maximum fuel-cell impact at one local position: 346.7 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-vall impact at rack top: .0 (7) Maximum racksto-rack impact at baseplate: .0 i

(8) Maximum rack-to-rack impact at rack top: .0 i

MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction

Top corner

.3585 .1319 l

Baseplate corner: .0060 .0032 1

MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .007 .005 .019 .027 .039 .045 .005 Support pedestal: .135 .040 .040 .037 .145 .153 .044

• See Section 6.5.2.3 of the Licensing Report for definitions.

i i

Table 6.7.27

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RA .

Holtec Run I.D.: drn5ssel.rh2 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 117 cells loaded; Fuel centroid X,Y: . 0,-31.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2 SRevision: 3.46 S i

SLogfile C:/ racks /dynam0/ dynamo.fov $

$ Revision: 2.5 $

$Logfile: C:/ racks /dynam0/dynasi.fov $

$ Revision: 3.36 $ ,

$Logfile: C:/ racks /dynam0/dynas2.fov $ l DYNAMIC IMPACT LOADS (1bs.)

~~

(1) Maximum total vertical pedantal load: 144905.0 (2) Maximum vertical load in any single pedestal: 74962.9 (3) Maximum shear load in any single pedestal: 24985.1 ,

(4} Maximum fuel-cell impact at one local position: 421.2 (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: 0996 Baseplate corner:

.1128

. 0140 .0170 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .007 .005 .016 .019 .027 031 Support pedestal: . .004

.110 .026 .031 .024 .124 . 129 .033

• See Section 6.5.2.3 of the Licensing Report for definitions.

.g -.e +- u +-w- ----,1 y, m -

Table 6.7.28

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N5 Holtec Run 1.D.: drn5ssei.re8 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.t,80/  ; 680.0 (lbs.)

Fuel Loading: 13 cells loaded; Fuel centroid X,Y: .0,-56.4 (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 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: 41342.3 (2) Maximum vertical load in any single pedestal: 22395.8 (3) Maximum shear load in any single pedestal: 5962.9 (4) Maximum fuel-cell impact at one local position: 367.7 (5) Maximum rack-to-wall impact at baseple.te: .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 MAXIMTTM CORNER DISPLACEMENTS (in.)

Location X-direction Y-direction Top corner: .0381 .0430 Baseplate corner: .0013 .0012 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .004 .001 .007 .007 .011 .013 .002 Support pedestal .033 .008 .011 .003 .041 .043 .012

• See Section 6.5.2.3 of the Licensing Report for definitions.

1 l

1 I

i Table 6.7.29  !

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACX-N Holtec Run I.D.: drn5ssai.re5 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 13 cells loaded; Fuel centroid X,Y: .0,-56.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.5 SRevision 3.46 S

$Logfile C: / racks /dynam0/ dynamo.f oy $

$ Revision: 2.5 S GLogfile: C:/ racks /dynam0/dynasi.fov $

SRevision: 3.36 $ . >

SLogfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 41165.9 (2) Maximum vertical load in any single pedestal: 22373.0 (3) Maximum shear load in any single pedestal: 6076.0 (4) Maximum fuel-cell impact at one local position: 367.7 l (5) Maximum rack-to-vall impact at baseplate: .0 (6) Maximum rack-to-vall 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-l l

MAXIMUM CORNER DISPLACEMENTS (in.)

l Location: X-direction Y-direction l

Top corner: .0394 . 0401 Baseplate corner: .0013 . 0011 MAXIMUM STRESS PACTORS

• Stress factor: -

R1 R2 R3 R4 R5 R6 R7 Above baseplate: .004 .001 .007 .007- .011 . 013 . 002 Support pedestal: .033 .009 .011 .008 .039 . 04.1 . 012

• See Section 6.5.2.3 of the Licensing Report for definitions. ~

l

• -e- ' -

• y-.-v ww ir. -w--- .w. a.,. , c op- p,, -- w isei, i~ v- y3y-r - - - - ,

l 4

Table 6.7.30

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE! RACK-N5 Holtec Run I.D.: drn5ssei.re2 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 13 cells loaded; Puel centroid X,Y: .0,-56.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2 SRevision: 3.46 $

$Logfile C:/ racks / dynamo / dynamo.fov $

$ Revision: 2.5 $

$Logfile C:/ racks /dynam0/dynasi.foy $

$ Revision: 3.36 $ .

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPAC.T LOADS (lbs.)

(1) Maximum total vertical pedestcl load: 41301.4 (2) Maximum vertical load in any single pedestal: 21436.2 (3) Maximum shear load in any single pedestal: 4260.0 (4) Maximum fuel-cell impact at one local position: 323.0 (5) Maximum rack-to-wall impact at baseplates .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 .0476 . 0312 Baseplate corner: .0162 . 0163 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2_ R3 R4 R5 R6 R7 Above basoplate: .004 .001 .007 .006 Support pedestal:

.011 .012 .001

.031 .008 .008 .008 .038 .040 .008

• See Section 6.5.2.3 of the Licensing Report for definitions.

l 1

l l _

l Table 6.7.31

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N-Moltec Run I.D.: drn5sseo.rf8 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 247 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 $

$ Revision: 2.5 $

$Logfile: C:/ racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

SLogfile: C:/ racks /dynam0/dynaa2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 262443.9 (2) Maximum vertical load in any single pedestal: 101134.1 (3) Maximum shear load in any single pedestal: 20274.0 (4) Maximum fuel-cell impact at one local position: 387.9 (5) Maxinum 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: .1100 .0900 Baseplate corner: .0042 .0033 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .011 .009 .030 .024 .049 .055 Support pedestal: .011

.149 .045 .038 .041 .149 .153 .042

• See Section 6.5.2.3 of the Licensing Report for definitions.

i

Table 6.7.32

SUMMARY

RESULTS OF 3-D SINGLB RACK ANALYSIS FOR RACK MODUT E: RACK-N5 Holtec Run 1.D.: drn5sseo.rf5 Seismic Loading: 1.15xSSE-2 Fual Assembly I.D. and Weight: Reg.680#  ; 680.0 (lbs.)

Ftel Leading: 247 cells loaded; Fuel centroid X,Y: .0, .0 (in.)

Coefficient of friction at the bottom of support pedestal: 0.5 SRevision: 3.46 S SLogfile: C:/ racks /dynam0/ dynamo.fov S SRevision: 2.5 S

$Logfile: C:/ racks /dynam0/dynas1.fov $

SRevision: 3.36 $

SLogfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total veu-tical pedestal load: 262443.9 (2) Maximum vertical load in any single pedestal: 100370.1 (3) Maximum shear load in any single pedestal: 34614.6 (4) Maximum fuel-cell impact at one local position: 345.0 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack irnact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0910 .0939 Baseplate corner: .0034 .0034 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .011 .007 .030 .020 .044 049 .009 Support pedestal: .147 .031 .070 .028 .180 .190 .076

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.33

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N Moltec Run I.D.: drn5sseo.rf2 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680#  ; 680.0 (1bs.)

Fuel Leading: 247 cells loaded; Fuel centroid X,Y: .0, .0 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2 SRevision: 3.46 S

$Logfile: C:/ racks /dynam0/ dynamo.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: 262443.5 (2) Maximum vertical load in any single pedestal: 95789.5 (3) Maximum shear load in any single pedestal: 19139.5 (4) Maximum fuel-cell impact at one local position: 388.6 (5) Maximum rack-to-wall irpact 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: .0835 Baseplate corner: .1107

.0085 .0223 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .011 .008 .033 .019 Support pedestal: .043 .049 .009

.141 .032 .038 .030 .169 .177 .041

• See Section 5.5.2.3 of the Licensing Report for definitions.

l l

Table 6.7.34

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N5 Holtec Run I.D.: drn5sseo.rh8 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (1bs.)

Fuel Loading: 117 cells loaded; Fuel centroid X,Y: .0,-31.4 (in.)

Coefficient of friction at the botten of support pedestal: 0.8

^

SRevision: 3.46 $

$Logfile: C:/ racks /dynam0/dynam0.fov S SRevision: 2.5 $

$Logfile: C:/ racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

SLogfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 146409.3 (2) Maximum vertical load in any single pedestal: 66561.1 (3) Maximum shear load in any single pedestal: 18688.0 (4) Maximum fuel-cell impact at one local position: 343.8 (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: .0621 .1002 Baseplate corner: .0023 .0019 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .008 .004 .016 .014 .025 .028 .006 Support pedestal: .097 .022 .038 .020 .120 .127 .041

• See Section 6.5.2.3 of the Licensing Report for definitions.

I Table 6.7.35

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-NS Holtec Run I.D.: drn5sseo.rh5 Seismic Loading: 1.15xSSE 1 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 117 cells loaded; Fuel centroid X,Y: .0,-31.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.5 SFeviAlon: 3.46 S SLejfi]i: C:/ racks /dynam0/ dynamo.fov S SRevit.icn: 2.5 S SLojfile: C:/ racks /dynam0/dynasi.fov $

SRevision: 3.36 S ,

$Logfile.: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACI LOADS (lbs )

(1) Maximum total vertical pedestal loai: 146409.2 (2) Maximum vertical load in any single pedestal: 69967.4 (3) Maximum shear load in any single pedestal: 20572.2 (4) Maximum fuel-cell imptet at one local position: 371.2 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-vall 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: .0712 .1219 Baseplate corner: .0025 .0027 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .008 .004 .022 .016 .031 .036 .005 Support pedestal: .102 .025 .039 .023 .132 .139 .042~

• See Section 6.5.2.3 of the Licensing Report for definitions.

1

-i l

Table 6.7.36

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N5 Holtec P'~7 1.D.: drn5sseo.rh2 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0-(lbs.)

Fuel Loading: 117 cells loaded; Fuel centroid X,Y: .0,-31.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2 SRevision: 3.46- S SLogfile: C:/ racks /dynam0/ dynamo.fov S SRevision: 2.5 $

$Logfile: C:/ racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 146992.7 (2) Maximum vertical load in any single pedestal: 67627.6-(3) Maximum shear load in any single pedestal: 13517.5 (4) Maximum fuel-cell impact at one local position: 367.7 (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: .0486 .1080 Baseplate corner: .0072 .0094 MAXIMUM STRESS FACTORO

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .008 .004 .017 .011 .023 .026' .005 Support pedestal: .099 .023 .026 .021 .117 .122 .028

• See Section 6.5.2.3 of the Licensing Report for definitions.

+r * --y

1 i

Table 6.7.37

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-1 Moltec Run I.D.: drn5sseo.re8 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680#  ; 680.0 (lbs.)

Fuel Loading: 13 cells loaded; Fuel centroid X,Y:

.0,-56.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.46 S SLogfile: C:/ racks /dynam0/ dynamo.fov E

$ Revision: 2.5 $

S g

$Logfile: C:/ racks /t.ynam0/dynas1.fov $

SRevision: 3.36 $

SLogfile: C:/ racks / dynamo /dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 41141.8 (2) Maximum vertical load in any single pedestal: 22075.9 (3) Maximum shear load in any single pedestal: 8089.6 (4) Maximum fuel-cell impact at one local position: 393.2 (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 irpact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0235 Baseplate corner: .0347

.0007 .0009 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .004 .001 .007 .004 Support pedestal: .032

.012 .014 .002

.010 .014 .009 .041 .044 .016

• See Section 6.5.2.3 of the Licensing Report for definitions.

l

Table 6.7.38

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N5 Holtec Run I.D.: drn5sseo.re5 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680/  ; 680.0 (lbs.)

Fuel Loading: 13 cells loaded; Fuel centroid X,Y: .0,-56.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.5 SRevision: 3.46 S SLogfile: C:/ racks /dynam0/ dynamo.fov $

SRevision: 2.5 $

SLogfile: C:/ racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

SLogfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 41979.4 (2) Maximum vertical load in any single pedestal: 21897.9 (3) Maximum shear lead in any single pedestal: 8386.5 (4) Maximum fuel-cell impact at one local position: 377.3 (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 DISPLACEMFJITS (in.)

Location: X-direction Y-direction Top corner: .0257 .0336 Baseplate corner: .0007 .0009 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplato: .004 .001 .007 .004 .012 .014 .002 Support pedestal: .032 .011 .017 .010 .041 .044 .018

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.39

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MO ,

Holtec Run I.D.: drn5sseo.re2 Seismic Loading: 1.15xSSE-2 Fuel Assembly I.D. and Weight: Reg.680#  ; 680.0 (lbs.)

Fuel Leading: 13 cells loaded; Puel centroid X,Y: .0,-56.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2 SRevision: 3.46 S

$Logfile: C:/ racks /dynam0/ dynamo.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: 41102.1 (2) Maximum vertical load in any single pedestal: 18093.8 (3) Maximum shear load in any single pedestal: 3537.3 (4) Maximum fuel-cell impact at one local position: 401.3 (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: .0237 .0328 Baseplate corner: .0094 .0088 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .004 .001 .005 .003 .010 .011 .001 Support pedestal: .027 .007 .007 .007 .030 .032 .008

• See Section 6.5.2.3 of the Licensing Report for definitions.

4 Table 6.7.40

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-E9-Holtec Run I.D.: dre9ssei.rf8 Seismic Loading: 1.15xSET-2 Fuel Assembly I.D. and Weight: REG.680#  ; 680.0 (lbs.)

Fuel Loading: 266 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 $

SLogfile: C:/ racks /dynam0/dynasl.fov $

SRevision: 3.36 $ .

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 293265.4 (2) Maximum vertical load in any single pedestal: 117819.4 (3) Maximum nhear load in any single pedestal: 24229.0 (4) Maximum fuel-cell impact at one local position: 238.2 (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.)

Location: X-direction Y-direction Top corner: .1174 .2205 Baseplate corner: .0047 .0093 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .034 .024 .066 .096 .140 .160 .023 Support pedestal: .232 .061 .038 .035 .239 .245 .066

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.41

SUMMARY

RISULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK' MODULE: RACK-E2) 4 Holtec Run I.D.: dre9ssel.rf2 Seismic Loading: 1.15xSET-2 Fuel Assembly I.D. and Weight: REG.680/  ;

680.0 (lbs.)

Fuel Loading: 266 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 /dynam0/ dynamo.fov $

SRevision: 2.5 S SLogfile: C:/ racks /dynam0/dynasi.fov $

SRevision: 3.36 S

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 292387.9 (2) Maximum vertical load in any single pedestal: 113378.9

~

(3) Maximum shear load in any single pedestal: 20072.4 (4) Maximum fuel-cell impact at one local position: 242.7 (5) Maximum rack-to-vall impact at baseplate: .0 (6) Maximum rack-to-vall impact at rack top: .0 (7) Maximum rack-to-rack i= pact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEM'CNTS (in.)

Location: X-direction Y-direction Top corner: .1371 .1640 Baseplate corner: .0213 .0256 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 RS R6 R7 Above baseplate: .034 .023 .048 .104 .145 .166 .022 Support pedestal: .224 .058 .029 .033 .229 .233 .051

• See Section 6.5.2.3 of the Licensing Report for definitions.

t

_ _ 4 _- _ _ - ._ _ . _ _ _ . _. . . _ _ . .

D Table 6.7.42 j

SUMMARY

RESULTS-OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-E9-4 Holtec Run I.D.: dre9ssai.rh8 - Seismic Loading: 1.15xSET-2 Fuel Assembly I.D. and Weight: REG.680# .;

680.0.(lbs.)

, Fuel Loading: 126 cells loaded; Fuel centroid'X,Y: . 0, -31. 5 - (in. )

Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.46 S

$Logfile: C:/ racks / dynamo / dynamo.fov S-

$ Revision: 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: 161306.2 '

(2) Maximum vertical load in any single pedestal: 90617.7-(3) Maximum shear load in any single pedestal: 20766.3-(4) Maximum fuel-cell impact at one local position: 250.5 (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: .1173 . 3881 Baseplate corner: .0048 . 0114 MJUCIMUM STRESS FACTORS

• Stress factor:- R1 R2 R3- R4 R5 R6 R7 Above baseplate: .023 .016 .038- .069 .083 .094 .014-Support pedestal: .178 - .041 .031 .023 .201 .208' .055

• See Section 6.5.2.3 of the Licensing Report for definitions.

9

- - , , , , - - - - , . - . . -, - a , . -

Table 6.7.43

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-E5 Holtec Run I.D.: drc9ssei.rh2 Seismic Loading: 1.15xSET-2 Fuel Assembly I.D. a..d Weight: REG.680/  ; 680.0 (1bs.)

Fuel Loading: 126 cells loaded; Fuel centroid X,Y: .0,-31.5 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$ Revision: 3.46 S SLogfile: C:/ racks /dynam0/ dynamo.fov S

$ Revision: 2.5 S SLogfile: C:/ racks /dynam0/dynasi.fov $

SRevision: 3.36 S -

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 162067.9 (2) Maximum vertical load in any single pedestal: 92690.1 (3) Maximum shear load in any single pedestal: 18536.8 (4) Maximum fuel-cell impact at one local position: 242.2 (5) Maxirum 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 l Top corner: .0860 .3682 Baseplate corner: .0116 .0155 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 RO R4 R5 R6 R7 i

l Above baseplate: .023 .014 .036 .063 .081 .094 .012 Support pedestal: .183 .051 .027 .029 .199 .205 .048

• See Section 6.5.2.3 of the Licensing Report for definitions.

l l

I

i i

Table 6.7.44

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-E9 Holtec Run I.D.: dre9ssei.re8 Seismic Loading: 1.15xSET-2 Fuel Assembly I.D. and Weight: REG.680/  ; 680.0 (lbs.)

Fuel Loading: 14 cells loaded; Fuel centroid X,Y: .0,-56.7 (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 $

SLogfile: C:/ racks /dynam0/dynasi.fov $

$ Revision: 3.36 S '

SLogfile: C:/ racks /dynam0/dynas2.foy $

DYNAMIC IMPACT LOADS (1bs.)

(1) Maximum total vertical pedestal load: 44441.0 (2) Maximum vertical load in any single pedestal: 23909.8 (3) Maximum shear load in any single pedestal: 7273.1 (4) Maximum fuel-cell impact at one local position: 253.6 (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: .0190 .0814 Baseplate _ corner: .0008 .0024 MAXIMUM STRESS FACTORS

• Stress factor: R2 R2 R3 R4 R5 R6 R7 Above_ baseplate: .011 .004 .014 .015 .030 .034 .004-Support pedestal: .047 .013 .012 .007 .050 .052 .021

• See Section 6.5.2.3 of the Licensing Report for definitions.

_ . . ._ __-.y _ .- -_ . . . _ _ . ,_

_-_._.-m._ _

' ~

Table 6.7.45

SUMMARY

RESULTS OF 3-D ' SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-E(

4 Holtec Run I.D.: dre9ssai.re2  : Seismic Loading:.< 1.15xSET Fuel Assembly I.D. and Weight: REG.680# =;- 680.0 (lbs.)

Fuel Loading: 14 cells loaded; Fuel centroid X,Y:

- . 0, -5 6. 7 ' (in .')

i Coefficient of friction at the bottom of-support pedestal: 0.2 h

~

SRevision: 3.46 $ .

j SLogfile: C:/ racks /dynam0/ dynamo.fov S- o

$ Revision: 2.5 S SLorffile: C:/ racks / dynamo /dynasi,fov $

SRevision: 3.36 $ j SLogfile: C:/ racks /dynam0/dynas2.fov $

l DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 46'643.0 (2) Maximum- vertical load in any sir' 3 pedestal:- 22834.9 (3) Maximum shear load in any single pedestal: 4453.1 )

4 (4) Maximum fuel-cell impact at one local position: 254.8 (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: .0215 .0696 Baseplate corner:- .0075 .0121 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 .RS -R6 R7 Above baseplate:- .011 -.004 .012 .034 Support pedestal:

.015- .030 . 004-

.045 .013 .008 .007 .048 .049 . 013

• See Section 6.5.2.3 of the -Licensing Report for definitions.

~

1. ,i=, - -~ r ., e r--- = , - . ,,-- -

Table 6.7.46

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-E9 ,

Holtec Run I.D.: dre9sseo.rf8 Seismic Loading: 1.15xSET-2 Fuel Assembly I.D. and Weight: REG.680/  ; 680.0 (lbs.)

Fuel Loading: 266 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 $

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 lead: 292387.9 (2) Maximum vertical load in any single pedestal: 102615.6 (3) Maximum shear load in any single pedestal: 16085.3 (4) Maximum fuel-cell impact at one local position: 215.8 (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: .1073 .1411 Baseplate corner: .0043 .0059 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .034 .024 .044 .094 .124 .140 .023 Support pedestal: .202 .045 .024 .026 .202 .202 .042

• See Section 6.5.2.3 of the Licensing Report for definitions.

1

• i Table 6.7.47

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: PACK-E-Holtec Run I.D.: drc9sseo.rf2 Seismic Loading: 1.15xSET-2 Fuel Assembly I.D. and Weight: REG.680/  ; 680.0 (1bs.)

Fuel Loading: 266 cells loaded; Fuel centroid X,Y: . 0, .0 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

~$ Revision: 3.46 S

$Logfile: C:/ racks /dynam0/ dynamo.fov S

$ Revision: 2.S S SLogfile: C:/ racks /dynam0/dynasl.fov $

$ Revision: 3.36 $

$Logfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 292387.9 (2) Maximum vertical load in any single padestal: 96947.1 (3) Maximum shear load in any single pedestal: 18005.8 (4) Maximum fuel-cell impact at one local position: 225.5 (5) Maximum rack-to-wall impact at baseplato: .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: .1039 .1317 Baseplate corner: .0081 .0160 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .034 .022 .042 .090 .112 .127 .022 Support pedestal: .191 .049 .026 .028 .206 .212 .046 ,

• See Section 6.5.2.3 of the Licensing Report for definitions.

iI

Table 6.7.48

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-E9 Holtec Run I.D.: dre9sseo.rh8 Seismic Loading: 1.15xSET-2 Fuel Assembly I.D. and Weight: REG.680#  ; 680.0 (lbs.)

Fuel Loading: 126 cells loaded; Fuel centroid X,Y: .0,-31.5 (in.)

Coefficient of friction at the bottom of support podestal: 0.8 SRevision: 3.46 S

$Logfile: C:/ racks /dynam0/ dynamo.fov $

$ 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: 160953.8-(2) Maximum vertical load in any single pedestal: 78921.0 (3) Maximum shear load in any single pedestal: 18229.8 (4) Maximum fuel-cell impact at one local position: 272.8 (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: .0621 .2277 Baseplate corner: .0025 .0044 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 Rd R7 Above baseplate: .023 .013 .027 .047 .060 .068 .011 Support pedestal: .155 .036 .028 .021 .164 .169 .049

• See Section 6.5.2.3 of the Licensing Rept -t for definitions.

i l

i Table 6.7.49 '

SU10ERY RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-Holtec Run I.D.: dre9sseo.rh2 Seismic Loading: 1.15xSET-2 ~

Fuel Assembly I.D. and Weight: REG.680p  ;

680.0 (lbs.)

Fuel Loading: 126 cells loaded; Fuel centroid X,Y: .0,-31.5 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$ Revision: 3.46 S

$Logfile: C:/ racks / dynamo / dynamo.fov S Snevision:

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: 161684.4 (2) Maximum vertical load in any single pedestal: 81285.8

.(3) Maximum shear load in any single pedestal: 13741.3 (4) Maximum fuel-call impact at one local position: 276.9 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impant at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0515 .2323 Baseplate corner: .0055 .0064 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .023 .011 .029 .035 .061 .068 Support pedestal: .011

.160 .038 .021 .021 .160 .165 .037

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.50

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-E9 Holtec Run I.D.: dre9sseo.re8 Seismic Loading: 1.15xSET-2 Fuel Assembly I.D. and Weight: REG.680/  ; 680.0 (lbs.)

Fuel Loading: 14 cells loaded; Puel centroid X,Y: .0,-56.7 (in.)

Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.46 $

$Logfile: C:/ racks /dynam0/ dynamo.fov S SRevision: 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: 46645.9 (2) Maximum vertical load in any single pedestal: 21376.2 (3) Maximum shear load in any single pedestal: 8167.7 (4) Maximum fuel-cell impact at one local position: 221.3 (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 tiAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0228 .0490 Baseplate corner: .0009 .0010 MJLXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .011 .004 .008 .019 .026 .029 .004 Support pedestal: .042 .015 .013 .009 .044 .045 .024

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.51

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-E!

Holtec Run I.D.: dre9sseo.re2 Selsmic Loading: 1.1SXSET-2 Fuel Assembly I.D. and Weight: RIG.680/  ;

680.0 (lbs.)

Puel Loading: 14 cells loaded; Puel centroid X,Y: .0,-56.7 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2 SRevision: 3.46 S SLogfile: C:/ racks /dynam0/dynam0.fov $

$ Revision: 2.5 S

$Logfile: C:/ racks /dynam0/dynasi.fov $

$ Revision: 3.36 $

SLogfile: C:/ racks /dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 46652.0 (2) Maximum vertical load in any single pedestal: 20161.7 (3) Maximum shear load in any single pedestal: 3906.2 (4) Maximum fuel-cell impact at one local position: 240.4 (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: .0244 Baseplate corner: .0407

.0088 .0129 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .011 .004 .006 .015 .024 .026 .004 Support pedestal: .040 .011 .006 .007 .042 .043 .011

• See Section 6.5.2.3 of the Licensing Report for definitions.

. Table 6.7.52

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-N5 Holtec Run I.D.: drn5ssel.12h Saismic Loading:1.2x1.15xSSE-2 in Hl Fuel Assembly I.C. and Weight: 1.15xSSE-2 in V.

Reg.680#  ; 680.0 (lbs.)

Fuel ' Loading: 247 cells loaded; Fuel centroid X,Y: .0, .0 (in.)

Coeffico .t of friction at the bottom of support pedestal: 0.5 SRevision: 3.46 S SLogfile: C:/ racks /dynam0/ dynamo.fov $

$ Revision: 2.5 S SLogfile: 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: 262426.4 (2) Maximum vertical load in any single pedestal: 122011.6 (3) Maximum shear load in any single pedestal: 42317.2 (4) Maximum fuel-cell impact at one local position: 433.8 l (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 DISPLACEMDITS (in.)

Location: X-direction Y-direction Top corner: .1854 .1463 Baseplate corner: .0071 .0056 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .011 .011 .049 .042 .065 .075 .013 Support pedestal: .179 .059 .082 .054 .237 .252 .089

• See-Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.53 COMPARISON OF CALCULATED AND ALLOWABLE LOADS / STRESSES AT IMPACT LOCATIONS AND AT WELDS FOR REGULAR FUEL LOADING

- VALUE Item / Location Calculated Allowable Fuel assembly / 673.6 3510 cell wall impact, lbs.

Rack / Baseplate weld 1874 29820 psi Pedestal / Baseplate 0.59 1.0 weld (dl.mensionless limit load ratio)

Cell / Cell welds 1347 15180

Table 6.8.1 MAXIMUM DISPLACEMENTS FROM WHOLE POOL MULTI-RACK RUN PILGRIM NUCLEAR POWER STATION BOSTON EDISON COMPANY WPMR Analysis for Campaign-2, 16 Racks in Pool, Fully Loaded with 680/ Reg. Fuel; Random Friction; Seismic: Controlling SSE(=1.15xSSE-SET-2) ;

Run I.D.: dwpsse2.rfr Rack uxt uyt uxb uyb Holtec No. (in.) (in.) .- (in.) (in.) Rack 1 .4183E+00 .8394E+00 .3906E+00 .8209E+00 2 .5313E+00 .2825E+00 .4803E+00 .2563E+00 3 .2296E+00 .1358E+00 .1029E+00 .1081E+00-4 .3697E+00 .1555E+00 .2002E+00 .7410E-01 5 .2470E+00 .4697E+00 .4957E-01 .9763E-01 NS 6 .2794E+00 .1007E+00 .1132E+00 .8006E-01 7 .2066E+00 .1809E+00 .5707E-01 .1646E+00 8 .2199E+00 .1275E+00 .7368E-01 .7485E-01 9 .2649E+00 .3792E+00 .8981E-01 .2690E+00 N4 10 .2857E+00 .2210E+00 .2968E-01 .2897E-01 N3 11 .2380E+00 .1192E+00 .1501E+00 .9712E-01 12 .2453E+00 .2698E+00 .1284E+00 .2506E+00 13 .1403E+00 .2178E+00 .1654E-01 .1513E-01 N1 14 .3206E+00 .3082E+00 .1147E+00 .1692E+00 N2 15 .4714E+00- .6157E+00 .1255E+00 .1597E+00 N6 16 .4610E+00 .3431E+00 .2634E+00 .1549E+00 SRevision: 1.8 $

SLogfile: C:/ racks /multirac/maxdisp.fov S r ,

r r

Table 6.8.2 MAXIMUM IMPACT FORCE OF EACH GAP ELEMENT PILGRIM NUCLEAR POWER STATION BOSTON EDISON COMPANY WPMR Analysis for Campaign-2, 16 Racks in Pool, Fully Loaded with 680# Reg. Fuel; Random Friction; Seismic: Controlling SSE(=1.15xSSE-SET-2);

Run I.D.: dwpsse2;rfr GAP ELEMENT MAX. FORCE TIME (ib.) (sec.)

RACK-1:

1 2.366E+05 1.004E+01 2 2.047E+05 1.286E+01 3 2.225E+05 7.059E+00 4 1.986E+05 7.376E+00 5 3.322E+04 1.163E+01.

6 3.996E+04 4.823E+00 7 5.752E+04 6.108E+00 8 6.248E+04 5.782E+00 RACK-2:

9 1.972E+05 1.416E+01 10 2.078E+05 1.565E+01 11 1.945E+05 5.955E+00 12 2.138E+05 1.421E+01 13 3.675E+04 9.915E+00 14 5.460E+04 1.564E+01 15 5.432E+04 1.353E+01 16 5.163E404 1.182E+01 RACK-3:

17 1.961E+05 1.611E+01 18 1.771E+05 7.062E+00 19 1.797E+05 7.832E+00 20 1.991E+05 1.346E+01 21 5.272E+04 1.305E+01

'22 4.797E+04 4.764E+00 23 5.052E+04 8.453E+00 24 5.149E+04 8.094E+00

( Table 6.8.2, continued )

RACK-4:

25 1.423E+05 1.610E+01 26 1.659E+05 7.618E+00 27 1.664E+05 1.267E+01 28 1.731E+05 1.518E+01 29 4.387E+04 4.101E+00 30 4.138E+04 .

4.468E+00 31 5.398E+04 .-

1.296E+01 32 5.177E+04 1.320E+01 RACK-5 ( Holtec Rack-N5 ):

33 1.214E+05 1.399E+01 34 1.284E+05 1.288E+01 35 1.288E+05 1.246E+01 36 1.347E+05 8.715E+00 37 8.332E+04 1.696E+01 38 6.370E+04 1.095E+01 39 1.289E+05 1.019E+01 40 1.269E+05 1.046E+01 RACK-6:

41 2.024E+05 6.194E+00 42 1.943E+05 6.165E+00' 43 1.674E+05 5.465E+00 44 1.902E+05 5.396E+00 45 E.884E+04 5.971E+00 46 5.429E+04 1.296E+01 47 6.021E+04 1.617E+01 .

48 6.382E+04 1.587E+01 RACK-7:

49 1.582E+05 7.620E+00 50 1.863E+05 8.744E+00 51- 1.669E+05 6.864E+00~

52 1.927E+05 1.419E+01 53 4.642E+04 4.432E+00 54 4.823E+04 7.411E+00 55 3.125E+04 1.561E+01 56 3.518E+04 1.805E+01

_-- ___ - - __ _ _ ____ _ _______ - -__2 -_ - __

( Table 6.8.2, continued )'

RACK-8:

57 1.740E+05 1.329E+01 58 1.798E+05 6.417E+00 59 1.847E+05 7.057E+00 60 1.821E+05 7.412E+00 61 4.722E+04 1.040E+01 62 4.217E+04 1.077E+01 63 5.183E+04 ,' 1.286E+01 64 5.875E+04 1.261E+01 RACK-9 ( Holtec Rack-N4 ):

65 1.107E+05 7.598E+00 66 1.123E+05 1.418E+01 67 1.131E+05 8.758E+00 68 1.141E+05 1.525E+01 69 1.033E+05 1.452E+01 70 7.713E+04 1.746E+01 71 1.117E+05 7.629E+00 72 1.281E+05 7.353E+00 RACK-10 ( Holtec Rack-N3 ):

73 1.344E+05 1.498E+01 74 1.345E+05 1.418E+01 75 1.119E+05 9.232E+00 76 1.178E+05 1.526E+01 77 9.778E+04 7.983E+00 78 9.265E+04 1. 743 E+01 79 9.063E+04 7.826E+00 80 7.380E+04 7.203E+00 RACK-11:

81 1.381E+05 1.600E+01 82 1.765E+05 4.677E+00 83 2.085E+05 6.868E+00 84 2.160E+05 9.820E+00 85 5.777E+04 5.066E+00 86 6.545E+04 4.754E+00 87 3.511E+04 1.372E+01 88 4.072E+04 1.810E+01 1

l

F

( Table 6.8.2, continued )

RACK-12: '

89 1.825E+05 1.516E+01 90 2.123E+05 :4.456E+00-91 1.835Et05 6.828E+00 92 2.315E+05 5.117E+00 93 7.308E+04 1.782E+01 94 6.115E+04 1.809E+01 95 7.727E+04 6.059E+00 96 7.916E+04 ,-

5.727E+00 RACK-13 ( Holtec Rack-N1-):

97 1.121E+05 1.405E+01 98 1.152E+05 1.385E+01 99 1.311E+05 5.131E+00 100 1.110E+05 1.367E+01 101 6.181E+04 1.465E+01 102 6.367E+04 9.314E+00 103 1.329E+05 1.723E+01 104 9.350E+04 1.585E+01 RACK-14 (Holtec Rack-N2 ):

105 1.422E+05 9.828E+00 106 1.126E+05 1.490E+01 107 1.263E+05 8.752E+00 108 1.284E+05 1.452E+01-109 8.774E+04 1.537E+01 110 1.002E+05 1.506E+01 ,

111 1.211E+05 9.918E+00 112 1.053E+05 9.559E+00 RACK-15 ( Holtec Rack-N6 ):

113 1.384E+05 6.618E+00 114 1.42BE+05 9.604E+00-115 1.404E+05 1.418E+01 116 1.138E+05 1.017E+01 117 9.852E+04 '9.209E+00 118 8.787E+04 9.445E+00 119 1.297E+05 9.105E+00 120 1.401E+05 8.943E+00 L

( Table 6.8.2, continued )

RACK-168 121 6.532E+04 1.398E+01 122 6.882E+04 1.416E+01 123 7.368E+04 8.749E+00 124 5.841E+04 7.385E+00 125 5.368E+04 1.547E+01 126 5.368E+04 1.510E+01 ,

127 5.243E+04 1.412E+01 128 6.275E+04 .-

1.593E+01 RACK-TO-RACK /WAI4 IMPACT SPRINGS AT RACK TOP:

129 0.000E+00 0.000E+00 R-W

• 130 0.000E+00 0.000E+00 R-W 131 0.000E+00 0.000E+00 R-W 132 0.000E+00 0.000E+00 R-W '

133 0.000E+00 0.000E+00 R-W 134 0.000E+00 0.000E+00 R-W 135 0.000E+00 0.000E+00 R-W-136 0.000E+00 0.000E+00 R-W 137 0.000E+00 0.000E+00 R-W 138 0.000E+00 0.000E+00 139 0.000E+00 0.000E+00 140 0.000E+00 0.000E+00 141 0.000E+00 0.000E+00 R-W 142 0.000E+00 0.000E+00 R-W-143 0.000E+00 0.000E+00 '

144 0.000E+00- 0.000E+00  ;

145 0.000E+00 0.000E+00 146 0.000E+00 0.000E+00 R-W 147 0.000E+00 0.000E+00 148 0.000E+00 0.000E+00 149 0.000E+00 0.000E+00 150 0.000E+00 0.000E+00 151 ,

0.000E+00 0.000E+00 151 0.000E+00 0.000E+C0 153 0.000E+00 0.000E+00 154 0.000E+00 0.000E+00-155 0.000E+00 0.000E+00 R-W 156 0.000E+00 0.000E+00 157 0.000E+00 0.000E+00 R-J" indicates a rack-to-wall impact spring; others are rack-to-rack impact springs.

t

j I

( Table 6.8.2, continued ) '

158 0.000E+00 0.000E+00 159 0.000E+00 '

O.000E+00 R-W 160 0.000E+00 0.000E+00 R-W 161 0.000E+00 0.000E+00 ,

162 0.000E+00 0.000E+00 163 0.000E+00 0.000E+00  ;

164 0.000E+00 0.000E+00 R-W 165 0.000E+00 0.000E+00. l 166 0.000E+00 0.000E+00 167 0.000E+00 . 0.000E+00 168 0.000E+00 '

.." O.000E+00: 4 169 0.000E+00 O.000E+00 170 0.000E+00 0.000E+00 .

171 0.000E+00 0.000E+00 2 172 0.000E+00 0.000E+00 '

173 0.000E+00 0.000E+00 R-W 174 0.000E+00 0.000E+00 175 0.000E+00 0.000E+00 I 176 0.000E+00 -

0.000E+00 177 0.000E+00 0.000E+00 R-W 178 0.000E+00 0.000E+00: R-W #

179 0.000E+00 0.000E+00 180 0.000E+00 0.000E+00  !

181 0.000E+00 0.000E+00  !

182 0.000E+00 0.000E+00 R-W

• 183. 0.000E+00 0.000E+00 0.000E+00 184 0.000E+00-185 0.000E+00 0.000E+00 186. 0.000E+00 0.000E+00 187 0.000E+00 0.000E+00 188 0.000E+00 0.000E+00 ,1 189 0.000E+00 ~ 0.000E+00- '

190 0.000E+00 0.000E+00 191 0.000E+00 0.000E+00 R-W 192 0.000E+00 0.000E+00 193 0.000E+00 0.000E+00-194 0.000E+00 0.000E+00-- '

195 0.000E+00 0.000E+00 R-W-196 .0.000E+00 0.000E+00 R-W 197 0.000E+00 0.000E+00. ~

198 0.000E+00 0.000E+00- -l 199 0.000E+00 0.000E+00 L 200 0.000E+00 0.000E+00 R-W i

-j

[

o

. - . - . , .- , _ a_u...--_.._...2.----, . - - - . , , . .

1

( Table 6.8.2, continued )

l 201 0.000E+00 0.000E+00. R-W 202 0.00GI+00 0.000E+00- R-W 203 0.000E+00 0.000E+00 R-W.

204 0.000E+00 0.000E+00 R-W 205 0.000E+00 0.000E+00 R-W-206 0.000E+00- 0.000E+00 R-W .

207 0.000E+00 0.000Z+00: R-W .

208 'O.000E+00 0.000E+00 R-W RACK-TO-RACK / WALL IMPACT-SPRINGS AT RACK BOITOMt  !

/

209 0.000E+00

-0.000E+00 R-W 210 0.000E+00 0.000E+00 R-W -

211 0.000E+00 0.000E+00 R-W- 3 212 0.000E+00 0.000E+30 R-W '

213 0.000E+00 0.000E+00 R-W'  ;

214 0.000E+00 0.000E+00 R-W '

215 0.000E+00 0.000E+00 R-W e 216 0.000E+00 -

0.000E+00 R-W 217 0.000E+00 0.000E+00 R-W 218 0.000E+00 0.000E+00-219 0.000Et00 - 0.000E+00 220 0.000E+00 0.000E+00 '

221 0.000E+00 0.000E+00 R-W-222- 0.000E+00- 0.000E+00---R-W -

223 0.000E+00 0.000E+00 224 0.000E+00 0.000E+00 225 0.000E+00 -0.000E+00.

226 0.000E+00- 0.000E+00- R-W j 227 -0.000E+00 0.000E+00.

228 0.000E+00' 0.000E+00-229 'O.000E+00 0.000E+00 230 0.000E+00 0.000E+00 231 -0.000E+00 0.000E+00 232 0.000E+00 0.000E+00 l 233 0.000E+00' O.000E+00c ,

L 234 '

0.000E+00 0.000E+00-L 235 O.000E+00 0.000E+00 'R-W  :

236 0.000E+00~ 0.000E+00 237 0.000E+00 0.000E+00:-

238 - 0.000E+00 0.000E+00-

-239- 0.000Et00 0.000E+00 R-W t

av j

' ^

- ,- . , - , , , - , . , -,;&i

( Table 6.8.2, continued )'

i 240 0.000E+00 0.000E+00 R-W ,

241 0.000E+00 0.000E+00 '

242 0.000E+00 0.000E+00 243 0.000E+00 0.000E+00 244 0.000E+00 0.000E+00 R-W 245 0.000E+00 0.000E+00 246 0.000E+00 0.000E+00 247 0.000E+00 0.000E+00 248 0.000E+00 ,. ' O.000E+00 249 0.000E+00 0.000E+00 250 0.000E+00 0.000E+00 251 0.000E+00 0.000E+00 252 0.000E+00 0.000E+00 253 0.000E+00 0.000E+00 R-W 254 0.000E+00 0.000E+00 ,

255 0.000E+00 0.000E+00 256 0.000E+00 0.000E+00 257 0.000E+00 0.000E+00 R-W 258 0.000E+00 0.000E+00 R-W 259 0.000E+00 0.000E+00 260 0.000E+00 0.000E+00 261 0.000E+00 0.000E+00 262 0.000E+00 0.000E+00 R-W 263 0.000E+00 0.000E+00 264 0.000E+00 0.000E+00 265 0.000E+00 0.000E+00 266 0.000E+00 0.000E+00 267 0.000E+00 0.000E+00 268 0.000E+00 0.000E+00 -,

269 0.000E+00 0.000E+00 270 0.000E+00 0.000E+00 271 C.000E+00 0.000E+00 R-W 272 0.000E+00 'O.000E+00 273 0.000E+00 0.000E+00 274 , 0.000E+00 0.000E+00 275 0.000E+00 0.000E+00 R-W 276 0.000E+00 0.000E+00 R-W 277 0.000E+00 0.000E+00 278 0.000E+00 0.000E+00 279 0.000E+00 0.000E+00 280 0.000E+00 0.000E+00 R-W

l ( Table 6.8.2, continued )

281 0.000E+00 0.000E+00 R-W 282 0.000E+00 0.000E+00 R-W 283 0.000E+00 0.000E+00 R-W 284 0.000E+00 0.000E+00 R-W 285 0.000E+00 0.000E+00 R-W 286 0.000E+00 0.000E+00 R-W 287 0.000E+00 0.r00E+03 R-W 288 0.000E+00 0.000E+00 R-W

~

FILE INFORMATION FOR TlilS RUN "

Input File dupsse2.rfr .

Plot Fila fwpmr.rfr X-Seismic a-tase.h21 Y-Seismic a-tsso.h22 Z-Scismic a-te.sa.vt2 Output File owpsse2.rfr PILGRIM , WPMR ,16ta , df =dwps s e2 . rf r ; reg . ,1.15 x ( H21, H22 , VT2 ) ; . 0001" ; Kf x1 I

( file: sf sse12. rfr )

Table 6.8.3 MAXIMUM PEDESTAL STRESS FACTORS OF EACH PEDESTAL OF EACH HOLTEC RACKS IN POOL PILCRIH NUCLEAR POWrR STATION BOSTON EDISON COMPANY WPMR Analysis for Campaign-2, 16 Racks in Pool, Fully Loaded with 680# Reg. Fuel; Random Friction; Seismic Controlling SSE(=1.15xSSE-SET-2);

Run 1.D.: dwpSSE2.rfr PILCRIM, WPMR,10old+ 6new racke (No. 5,9,10,13,14,15 ) ; reg. ;1.15xSSE-2.

• • • • • • • • • • • • • • • • • • INPUT DATA ******************

File name of TX Time history ,

a frase2x.rfr File name of FY Time history a frase2y.rfr File name of TV Time history a p1wase2.rfr SRevision: 1.0 $

SLogfile: C:/ racks /multirae/sfar2.fov 5 SDate: 28 May 1992 18:08:26 $

File name of result output i sfsse12.rfr Number of racks in the pool  : 16 Height of the pedestal, in.  : 6.75 Offset of rv from center, in.  : 1.06 Area of famale pedestal, in**2.  : 51.80 Inertia of femain pedestal, in**4.  : 662.30 Distance of extrame fiber in I, _ in.  : 3.00 Distance of extrame fiber in Y, in.  : 3.00 Yield stress of female pedwatal, psi. : 25000.

Nuober of pedestals of each rack  : 4444444444444444 (1) MAXIMUM VALUES OF STRESS FACTORS,R1 -- R7,FOR EACH PEDESTAL OF'AACK-N1 Maxtmum Values of R1 - R7 R1,R3,R4 for Max.R6 R1 R2 R3 R4 R5 R6 R7 R1R6M R3R6M R4R6M Pecestal-1:

.144 .074 .058 .049 .184 .193 .087 .135 .018 .000-Time (sec.):

14.050 13.800 4.950 13.800 4.950 4.950 4.950 Pedental-2:

.148 .109 .089 .080 .224 .238 .119 .143 .046 .050 Time (sec.):

5.610 9.240 13.850 9.240 5.630 5.630 13.850 Pedestal-3:

.169 .092 .067 .066 .229 .240 .100 .163 .052 .024-Time (sec.):

5.130 8.720 15.770 8.710 5.130 5.140 15.770 Pedestal-4:

.143 .114 .087 .084 .233 .249 .112 .140 .080 .029 Time (sec.):

13.660 13.670 12.450 13.670 4.940 4.940 12.450

(, file: sfsse12.rfr, continued )

(2) MAIIMUM VALUES OF STRESS FACTORS,R1 - R7,FOR EACH PEDESTAL OF RACK-N2 Haximum Values of R1 - R7 R1,R3,R4 for Max.}

R1 R2 R3 R4 R5 R6 R7 R1R6H R3R6H 741 Pacestal-1:

.183 .119 .089 .089 .284 .303 .119 .179 .065 .059 Time (sec.):

9.830 6.610 15.010 6.610 9.840 9.840 15.010 Pedestal-2: ..

.144 .094 .071 .064 .204 .215 .101 .144 .071 .000 Time (sec.):

4.680 14.890 4.680 14.890 4.680 4.600 4.680 Pedestal-3:

.162 .082 .052 .049 .209 .217 .085 .162 .032 .023 Time (sec.):

8.750 6.890 4.690 7.810 8.750 8.750 4.690 Pedestal-4:

.165 .116 .098 .093 .225 .244 .119 .117 .046 .080

'. *1me (sec.):

14.520 10.160 4.310 10.160 7.040 7.040 4.310 (3) HAIIMUM VALUES OF STRESS FACTORS,R1 - R7,FOR EACH PEDESTAL OF RACK-N3 Haximum Values of R1 - R7 R1,R3,R4 fc Max . Rt R1 R2 R3 R4 R5 R6 R7 R1R6M R3RG. R4R$

I Pecestal-1:

.172 .097 .058 .061 .237 .248 .089 .172 .025 .051 Time (sec.):

14.980 14.970 10.640 15.000 14.980 14.980 10.640 Pedostal-2:

.172 .101 .105 .076 .246 .262 .136 .157 .105 .000 Time (sec.):

14.180 18.090 14.210 18.090 14.200 14.210 14.210 Pedestal-3:

.144 .095 .068 .067 .223 .237 .096 .144 .045 .047 Time (sec.):

9.230 9.250 9.210 13.E00 9.230 9.230 9.210 Pedestal-4:

.151 .091 .052 .059 .201 .209 .078 .151 .000 .059 Time (sec.):

15.260 15.260 4.340 15.260 15.260 15.260 4.340

(. files afsse12.rfr, continued )

(4) MAXIMUM VALUES OF STRISS FACTORS,R1 - R7,FOR EACH PEDESTAL OF RACK-N4 Maximum Values of R1 - R7 R1,RJ,R4 for Max.R6 R1 R2 R3 R4 R5 R6 R7 RIR6M R3R6M R4R6M Pedestal-1:

.142 .073 .053 .046 .194 .203 .084 .142 .032 .029 Time (sec.):

7.600 7.590 9.830 12.980 7.600 7.600 9.830 Podestal-2:

.144 .063 .057 .040 .193 .201 .088 .144 .057 .000 Time (sec.):

14.180 9.230 14.180 17.340 14.180 14.180 14.180 Pedestal-3:

.145 .100 .092 .075 .225 .240 . 12 1 .143 .084 .013 Time (sec.):

8.760 12.450 13.460 12.450 8.750 8.750 13.460 fedestal-4:

.247 .059 .051 .037 .190 .198 .083 .147 .051 .000 Time (sec.):

15.250 7.970 15.250 7.970 15.250 15.250 15.250 (5) MAXIMUM VALUES OF STRESS FACTORS,R1 -- R7,FOR EACH PEDESTAL OF RACK-N5 Maximum Values of R1 -- R7 R1,RJ,R4 for Max.R6 R1 R2 R3 R4 R5 R6 R7 R1R6M R3R6M R4R6M Pedestal-1:

.156 .097 .070 .067 .209 .218 .102 .156 .063 .000-Time (sec.):

13.990 6.600 10.040 6.600 13.990 13.990 6.610 Pedestal-2:

.265 .091 .102- .070 .231 .246 .131 .144 .102 .000 Time (sec.):

12.880 16.390 1L.630 8.980 15.630 15.630 15.630 Pedestal-3:

.166 .088 .091 .067 .262 .281 .121 .156 .089 .036 Time (sec.):

12.460 16.770 5.070 16.950 15.940 15.940 15.940 Pedestal-4:

.171 .083 .064 .056 .240 .253 .094 .171 .053 .028 Time (sec.):

8.710 18.450 10.040 18.450 8.710 8.710 10.040

- , . - - - ~ - . . - - . - . - ,

l l

(_flic: sfase12.rfr, continued )

~

1 (6) MAXIMUM VALUES OF STPISS FACTORS,R1 - R7,roR EACH PEDESTAL OF RACT.-N6 Maximum Values of R1 - R7 R1,R3,R4 for Max.h' R1 R2 R3 R4 R5 R6 R7 R1R6H R3R6H R4RI Pedestal-1:

.178 .124 .080 .097 .249 .265 .113 .157 .019 .090 Time (sec.):

6.620 6.670 7.990 6.670 6.650 6.650 7.990 Pedestal-2:

.183 .128 .076 .094 .281 .299 .104 . 175 .066 .058 Time (sec.):

9.600 5.650 3.980 5.650 5.630 5.630 5.630 Pedestal-3:

.100 .119 .077 .089 .266 .282 .108 . 175 .041 .046 Time (sec.):

14.180 8.520 11.840 8.520 10.250 10.250 10.260 Podestal-4:

.147 .111 .083 .080 .218 .233 .113 . 138 .032 .063 Time (sec.):

10.170 10.170 4.450 10.170 10.160 10.160 4.450 I

i

(

I

F Table 6.8.4 MAXIMUM RACK DISPLACEMENTS, PEDESTAL VERTICAL LOADS AND PEDESTAL STRESS FACTOR IN SINGLE RACK ANALYSIS AND IN WPMR-ANALYSIS PILGRIM NUCLEAR POWER STATION BOSTON EDISON COMPANY i

( SEISMIC SSE(1.15xSSE-SET-2); Reg.Fuelt Fully Loaded ) l Maximum i Maximms Holtec Rack Holtec Rack l Boltec *

- Pedestal Maximum i Rack corner Vertical Pedestal  ;

Displacement Load stress Run I.D. Remarks ( in.) ( lbs.)

i den 5ssei.rf5 Single rack 0.2054 123833.0 0.228 analysis in x-dir.; ( R6 }

Rack-N5(5) 0.1333 ,

cof.= 0.5 in y-dir.

denisseo.rf2 single rack 0.1414 1.14227.0 0.199 analysis in x-dir.; ( R6 )

Rack-N1(13) 0.1293 cof.= 0.2 in y-dir.

dwpses2.rfr WPMR analysis 0.4714- - 142800.0. 0.299 cof.= random in x-dir. Rack-N6(15) (:R6 )

(mean cof.=0.5) Rack-N6(15) Foot-2 Rack-N6(15) ,

0.6157 Foot-2 in y- dir . '!

Rack-N6(15) ,

4 cof.= coefficient of friction between pedestal and pool liner.

r

>= wee _. _ -- m. --w i4- 1 evgiw+.i.,s.r31we.eeg.- ,- ,%-tm.+3y. y aem .*.sw me _=,.y-. q-,... p. p .

n.-- . . . .*--g.-*e- 9,%g - Mg--re--- 0

i Table 6.8.5 RESULTS OP POOL WALL DYNAMIC PRESSURES PILCRIM NUCLEAR POWER STATION BOSTON EDISON COMPANY WPMR Analysis for Campaign-2, 16 Racks in Pool, Pully Loaded with 680# Reg.ruelt Randon Priction; ,

Seismic Controlling SSE(=1.15x$5E-SET-2);

Run I.D.: dwpase2.rfr

$ Revisions 1.0 $

$Logfile C / racks /multirsc/wallpres.for S (1) AVERAGE DYNAMIC PRESSURES ON POOL WALLS (psi.):

Average dynamic pressure on (-x) walle ~6.123008E-01 Average dynamic pressure on (+x) walls -6.123508E-01 Average dynamic pressure on (-y) walls 2.704960E-01 Average dynamic pressure on (+y) wall 3.483282E-01 (2) FEAK DYNAMIC PRESSURES ON POOL WALLS (psi.):

Positive peak pressure on (-x) wall: 15.000000 Negative peak pressure on (-x), walls -15.800000 i

Positive peak pressure on (+x) wall 15.000000 l

Negativa peak pressure on (+x) walls -15.800000 Positive peak pressure on (-y) wall: 9.140000 l Negative peak pressure on (-y) walls -11.100000 1

Positive peak pressure on (+y) wall: 11.800000 Negative peak pressure on (+y) walls -14.300000 (3) DYNAMIC PRESSURE ADDERS ON POOL WALLS (psi.):

l Positive pressure adder on (-x) wall: 5.093862 Negative pressure adder on (-x) walls -5.524945

! Positive pressure adder on (+x) wall 5.093862 l Negative pressure adder on (+x) walls -5.524945 Positive pressure adder on (-y) wall 3.668532 Neg&tive pressure adder on (-y) wall ~3.837831 l Positive pressure adder on (+y) wall: 4.725777 i Negative pressure adder on (+y) wall ~4.944138 l

(4) NUMBER OP TIME POINTS:

i

• Total number of time points Ln flie 2001 i

• Humber of time points where dynamie '

pressure on (-x) wall was positive 925

• Number of time points where dynamic pressure on_(~x) wall was negatives 1076

• Number of time points where dynsa4c pressure on (+x) wall was positives 925 l

• Number of time points where dynamic l pressure on (+x) wall was negatives 1076-

• Number of time points where dynamic pressure on (-y) wall was positive 1096

• Number of time points where dynamic pressure on (-y) wall was negative: 905

• Number of timi points where dynamic pressure on (+y) wall was positives 1096

• Number of time points where dynamic pressure on (+y) wall was negatives 905

l l

Table 6.8.6 I TOTAL STATIC-LOAD AND DYNAMIC ADDER ON THE WHOLE SLAB PILGRIM NUCLEAR POWER STATION BOSTON EDISON COMPANY {

WPMR Analysis for Campaign-2, 16 Racks in Pool, Fully Loaded with 680# Reg. Fuel; Random Friction; 1 Seismic: Controlling SSE(=1.15xSSE-SET-2); '

Run I.D.t dwpase2.rfr (1) THE TOTAL STATIC LOAD OF RACKS AND FUEL ON SLAB IS:

2716400.00 lbs. i (2) THE TOTAL DYNAMIC LOAD ADDER ON THE SLAB IS 277347.90 lbs.

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t

.s 6: PILGRIM NUCLEAR POWER STATION, BOSTON ED! SON CO.

Synthetical SSE Acceleration Time-History for Fuel Pool Slob, o-tsse.vt1; Durction: 20 sec.

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- PILGRIM ' NUCLEAR ' POWER STATION, BOSTON EDISON COMPANY:

POOL St>B. EL74.25' THE RESPONSE- SPECTRUM OF SAFE SHUTDOWN EARTHQUAK

., AND THE AVERAGE RESPONSE _ SPECTRUM.

o-tsse.h11,o-tase.h21,o-tsse.h31,o-tsse.h41;Darnping: OF 4 ARTIFICIAL SSE A

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- PILGRIM HUCLEAR POWER STATION, BOSTON EDISON COMPANY POOL SLAB. EL74.25'

- . THE RESPONSE SPECTRUM OF SAFE SHUTDOWN EARTHOUAKE ( SSE, with asterisks )

AND THE AVERAGE RESPONSE SPECTRUM OF 4 ARTIFICIAL SSE ACCELTIME-HISTORIES:

a-tsse~ h12,o-tsse.h22,a-tsse.h32 a-tsse.h42;Domping:- 2 percent; Duration: 20 sec 10 ' . .

. -. ....-i . . . . . ... . . . .. ....,

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FIGURE S.3.14 .

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- . PILGRIM ~ NUCLEAR POWER . STATION BOSTON EC SON COMPANY: POOL SLAB, ' EL74.25'

- 'S with asterisks )

THE RESPONSE SPECTRUM OF SAFE SHUTDOWN EARTHOUAK AND THE AVERAGE RESPONSE SPECTRUM OF .4 ARTIFICIAL SSE TIME-HISTORIES:

' a-tase.vt1,o-tsse.vt2,o-tese.vt3,o-tase.vt4;Domping: 2 percent: . Duration: 20 sec L10 -3 . . . . - . ....i . . . ....... . . . . ....i L10 -' '

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d- PILGRIM NUCLEAR POWER STATION, BOSTON EDISON CO.

Synthetical OBE' Acceleration' Time-History . for Pool Slob, ..

. o-tsse.h32, Duration: 20 sec. "

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- PILGRIM i NUCLEAR POWER- STATION

. THE RESPONSE ~ SPECTRUM OF . OPERATING: BASIS EARTHQUAKEBOSTON ' EDIS

' OBE.with asterisks- J AND THE AVERAGE RESPONSE SPECTRUM OF' 4 ARTIFICIAL' OBE. ACCELTIME- i o-tobe.h11,o-tobe.h21,o-tobe.h31,o-tobe.h41;Domping: .1, percant;- Duration: L20 sec 10 _3 . . . .; . .....i . ..-:.

. .....i . .. .. . ...,..i. o

' 10 ' -.

L1 ' '

11 0 : :10 *

Frequency,L Hz.1

. A mi>ae A v a? . ~ = =- -= -- - = -

10 :

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E  :

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g 10 -': :

M 'I PILGRIM NUCLEAR POWER STATION, BOSTON EDISON COMPANY: POOL SLAB, EL74.25' THE- RESPONSE SPECTRUM OF OPERATING BASIS EARTHOUAKE OBE,with - asterisks- )

AND THE AVERAGE RESPONSE SPECTRUM OF 4 ARTIFICIAL OBE(AC

_ o-tobe.h12,o-tobe.h22 o-tobe.h32,o-tobe.h42;Domping: 1 percent: Duration: 20 sac 10 -3 . . . . .. i i . . . . .ii . . . . . ...r 10 -' 1 10 10. '

Frequency. . Hz.

eini on 2,.

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M PILGRIM NUCLEAR POWER STATION BOSTON EDISON COMPANY: POOL St.AB, EL74.25' OBE,with osterisks )

THE RESPONSE SPECTRUM OF OPERATING BASIS EARTHOUAKE AND THE AVERAGE RESPONSE SPECTRUM OF 4 ARTIFICIAL OBE(ACCEL a - to b e.vt 1,o - t o b e.vt2,o-to b e.vt3,o-t o b e.vt4;Do m pin g: 1 percent: Duration: 20 see 1 0 -' - . . . . ..i . . . . . ... . . . . .....

10 -' 1 10 10

• Frequency, Hz.

FIGURE 8.3.3n , .

0

~ N ,

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f 1

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FIGURE 6.4.1 PICTORIAL VIEW OF RACK STRUCTURE

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,,- SCHEMATIC MODEL' FOR DYNARACK

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-RACK-TO-RACK: IMPACT SPRINGS . ~

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, i

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3.00- . . .

4 GAP TIME HISTORY PILGRIM STATION, BOSTON EDISON CO.

- Gop be tween R,ct a n d Re ct -6. . So u th Co r n e r . Top. . f

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.g o 3.00-GAP TIME HISTORY, PILGRIM STATION, BOSTON-EDISON CO.

Gap b e t.w e e n Ro e t -5 a n d Ro c k - 6. . No r th Co r n e r , Top., .

't

- Fr om WPf1R ' o n e l y a t s : dwpase2.cfr; 1.15x SSE-2; 680*r e g". f u e i ; f u l. I' ;-

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- . FIGURE S.S.3 _

3.00-GAP TIME HISTORY, PILGRIM STATION. BOSTON EDISON CO.

Gap be tween Rock-5 and Ro c t - 9 Ea s t Corner. Top. -

Fr o m WPMR o n e l y s t s : dupsee2.cfr; 1.15x SSE-2; 680#r o o . fu e l ; fu l l ;

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FIGURE 6.8.4

1 3.00-GAP TIME HISTORY, PILGRIM STATION, BOSTON EDISON CO.

Gap be tween Roet -5 o n d Re c t -9 We s t Co rn e r . Top, .*

Fr o m WPMR on e l y s t s : dwpase2.rfr; 1.15x SSE-2; 680*ce o . fu e I ; fu l I ;;

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. FIGURE 8.8.5

. .. _ - x- _ _ - _- - . - - - - - _ . -. .

1. *-

4

~

3.00-E -

GAP TITE HISTORY. PILGRIt1 STATION BOSTON EDISON CO.

'Gop between Re ct - 10 an d Ro c k - 11. So u th Corner. Top..

i- t Fr o m 14PtR on o f y s ts : dupese2.cfr: 1.15x SSE-2; 680*r e g . fu e l : fu l i ;

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FIGURE s.t.s . - '

. .s . . . _ , -.

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PILGRIM STATION, BOSTON EDITON CO. -

SSE-Sotomte. Fuuly Ioeded sa t th 680* cog.ruel c o n o mb i t o o .

Fr t o t. t o n coef t to ton t. = ron dom. FILE I.D.: s i o b s 2r f . d e t. .

1. 00E+ 6 ... ... .i..... ... ..... . i..... ...i 0.00 5.00 10.00 15.00 20.00 Time, sec.

FIGURE 5.8.7

ACCIDENT ANALYSIS AND MISCELIANEOUS STRUCTURAL 7.0

_ EVALUATIONS i 7.1 Introduction This section provides results of accident analyses and miscellaneous evaluations performed to demonstrate regulatory compliance of the new fuel racks.

The following limiting accident and miscellaneous structural-evaluations arc considered:

• Refueling accidents - drop of fuel assembly to top of rack or through a cell to the baseplate

- Local cell wall buckling

- Analysis of welded joints due to isolated hot cell Although this licensing application is not for consolidated fuel ,

bundles, all mechanical drop accident evaluations were performed with an assumed fuel mass of 1360 lbs. The mechanical accident' analysis results, reported herein, therefore, bound those for-a standard BWR assambly. For purposes of clarity, the assumed high mass fuel assembly will be referred to as the " heavy fuel assembly".

7.2 Refuelino Accidents 7.2.1 Heavy Fuel Assembly Dronned on a New Rack ,

The consequences of dropping a fuel assembly as it is being moved over stored fuel is discussed below. It is assumed that the lowest part of the fuel assembly being' carried is 36" above the top of the new spent fuel racks (7.2.1). The heavy _ fuel assembly weighs 1360 lbs. and associated handling equipment is assumed to weigh 140 lbs.

7-1

a. Droceed Fuel Assembly Accident (Deep Dron Scenario)

A 1500 lb. heavy fuel assembly plus handling equipment is dropped from 36" above the top of a storage location and impacts the base of the module. Local f ailu:le of the 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 occur, but the dropped assembly will be contained and not impact the liner. We show that even if there is local call-to-baseplate weld overstress in individual cells, the maximum movement of the baseplate toward the liner after the impact is at most between 1.54G" and 2.346".

The load transmitted to the liner through the support by such an accident is well below the loads caused by scismic events (given in Section 6).

b. Dronood Fuel Assembly Accident (Shallow Droo Scenario)

One fuel assembly plus the channel is dropped from 36" above the top of the rack and impacts the top of the rack. Permanent deformation of the rack is acceptable, but is required to be limited to the top region such that the rack cross-sectional geometry at the level of the top of the active fuel (and below) is not altered. Assuming 1 a minimal area of impact, it is shown that dn. mage, if it occurs, will be restricted to a depth of less than or equal to 1.591" below the top of the rack. This is above the active fuel region.

7.2.2 }Icavy fuel essembiv Dronned on an Existinct Rack The consequences of dropping a heavy fuel assembly to the existing rack as it is being moved over the stored fuel is also estimated because the existing racks have different structural parameters such as cell thickness and baseplate thickness, etc.

a. Dropoed Fuel Assembiv Accident (Deen Dron Scenario)

A 1500 lb. heavy fuel assembly plus handling equipment is dropped from 36" above the top of a storage location and impacts the base of the module. Local f ailure of the baseplate is acceptable; however, the rack design should ensure that gross structural failure does not occur and the suberiticality of the adjacent fuel assemblies is not 7-2

violated. Calculated results v.how that there will be no change in the spacing between cells. Local deformation of the basuplate in the neighborhood of the impact will occur, but the dropped assembly will be contained and not impact the liner. We show that even if there is local call-to-baseplate weld overstress in individual cells, the maximum movement of the baseplate toward the liner after the impact is at most between 1.635" and 2.57".

The load transmitted to the liner through the support by such an accident is well below the loads caused- by seismic events (given in Section 6).

b. Droceed Puel Assembly Accident (Shallow Dron Scenario)

One fuel assembly plus the channel is dropped from 36" above the top of the rack and impacts the top of the  ;

rack. Permanent deformation of the rack is acceptabic, '

but is required to be limited to the top region such that the rack cross-sectional geometry at the level of the top of the active fuel (and below) is not altered . Assuming a minimal area cf impact, it is shown that damage, if it occurs, will be restricted to a deptn of less than or equal to 1.005" below the top of the rack. This is above the active fuel region.

7.3 Local Bucklina of Fuel Cell Walls This subsection and the next one presents details on the secondary stresses produced by buckling and by temperature effects.

The allowable local buckling stresses in the fuel cell walls are obtained by using classical plate buchling analysis. The following formula for the critical stress has been used based on a width of cell "b" (7.3.1):

2 8 fr Et2 C a- "

12 b2 gy _ p 2) o, is the limiting vertical compressive stress in the tube, E=

27.6 x 108 psi, y = 0.3, (Poisson's ratio), t = .09", b = 6.05". 3 The factor 8 is suggested in (Ref. 7.3.1) to be 4.0 for a long.

panel.

e 7-3 t

,,e-,-- ,-r-+-

w .+ ,y ~ ,,.-.,--,,--.7,,

1. ,--+..c -- -,...e,  % s - , ,-- ...y- - ,.,,,-r .n.w., ,,7u,

_ _. _.m _ _ __ _ -. __ . _ . _

.__.__.____.__________..y

~

For the given data, au = 22081 psi It should be noted that this stability calculation is based on the applied stress being uniform along the entire length of the cell vall. In the actual fuel rack, the compressive stress comes from conside" tion of overall bending of the rack structures during a seismic , /ent and as such is negligible at the rack top and maximum at the rack bottom. It is conservative to apply the above equation l

to the rack call vall if we compara a, with the maximum compressive '

strecs 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 l in obtained by multiplying the limiting value of the stress factor  !

R (for the cell cross-section just abovn the basoplate) by the 6

i allowable stress (.6 x yield stress) . Thus, from Table 6.7.2, a=

R6 x allowable stress = 0.073 x (.6 x 25000) = 1095 psi..  ;

7.4 Analysis oLWelded Joint s in Rack due to Tsolated Hot Cell l

l 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 call wall) uniformly heated and restrained from growth along one long edge.

The strip is subject to a uniform temperature rise T = 63.95'F.

The temperature rise has been calculated from the difference of the maximum local water temperature and bulk water temperature in the l

7-4

-, 7 -r .~---,,----y,, e.__,- -..y-..--e_,---.% .-.s, -

v- e - --,_

spent fuel pool. (see Tables 5.9.2 and 5.9.4). Then, using a shear beam theory, we can calculate an estimate of the maximum value of the average shear stress in the strip (see Figuru 7.4.1) .

The final result for wall maximum shear stress, under conservative restraint assumptions is given as (7.5.1):

Ea T

.931 4

where a = 9.5 x 10 in/in 'F.

Therefore, we obtain an estimate of maximum weld shear stress in an isolated hot cell as tw = 18010 psi Since this is a secondary thermal stress, it is appropriate to compare this to the allowable weld shear stress for a faulted event i < .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 (7.2.1) Section 10.3, PNPS-FSAR, p. 10.3-2, Rev. 9, July, 1968.

[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-85330, Revision 1, 1989.

7-5 i

_____.__.-m. . _ . _ _ _ __

t i

Heated Cell Wall l

x H T

Y # Nav:::"M f M M f R A"X M f ~ '

L L Wald Line T

y FIGURE 7.4.1 WELDED JOINT IN RACK

.__.____.___.-.m_

4 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 8.1 Introduction The Pilgrim spent fuel pool is a safety related, seismic category-I, m inforced concrete structure with a steel beam framework for i addad structural strength under the slab and at certain locations in the walls. In this section, the analyses performed to

^

demonstrate the structural adequacy of the pool structure, as required by Section IV of the USNRC OT Position Paper (8.1.1) is abstrac t:ed.

l The Pilgrim spent fuel pool region is analyzed using the finite element method. Results for individual load components are ,

combined using factored load combinations mandated by the ACI code t (8.1.2) using the " ultimate strength" design method. It is demonstrated that for all applicable factored load combinations listed in reference (8.1.2), the structural integrity is maintained when the fuel pool is assumed to contain the maximum possible amount of dead load, i.e., when all racks are fully loaded with fuel assemblies. Satisf action of structural integrity require ments is based on limit strengths calculated per ACI [8.1.2) for concrete and AISC (8.1.3) for the steel beams.

. The regions examined in the fuel pool are the slab, the loaded wall sections adjoining the pool slab, and the steel beams and columns.

Both moment and shear capacities are checked for- concrete structural integrity. Local punching integrity of the slab in the i

vicinity of concentrated loads pad are also evaluated. All structural capacity calculations are made-using design formulas-provided in the American Concrete Institute Standards (8.1.4).

The applied mechanical loadings on the pool structure consist of steady state loads (dead weight of the racks, concrete, water. mass, fuel, etc.), as well as dynamic loads from sloshing motion, seismic excitation of the free-standing modules, hydrodynamic pressure due S

8-1

to fluid coupling between the racks and the walls in the quiescent (non-sloshing) space of the pool, arong others. The hydrodynamic coupling term, a frequently overlooked offect in pool structural analysis, has been fully incorporated in this analysis.

In addition to the mechanical loadings, the so-called self-limiting loads due to spatial thermal gradients, created by the temperature difference between the pool water and the environs surrounding the pool structure, are also included in the analysis.

8.2 Descrintion of Scent Puel Pool Structure 4

The slab and all pool valls, except for the massive East Reactor shield wall, are modelled. The top of the spent fuel pool slab is at elevation 78'-3". The bottom of the slab is at eluvation 72'-

7". The slab is supported on the east side by the reactor shield wall, which is not modelled because of its obvious strength over-capacity. The west vall of the spent fuel pool is 5'-3" thick to elevation 91'-3" and 4'-6" up to elevation 117' (top of wall).

The north and south walls of the spent tual pool are connected to the reactor shield wall on the east, end to the outside building wall on the west. The north wall in 6'-1" thick up to elevation 91-3", and 5'-1" thick from 91'-3" to 117'. The south wall is 6'-

1" thick up to elevation 91.'-3", S'-1" thick i:p to elevation 105',

and 4'-1" thick from 105' elevation up to 11/' elevation.

The extent of the vetted area of the slab is 40'-4" (N-S) and 30'-

6" (E-W).

A steel girder substructure underlies the pool slab and the surrounding walls. There are three major girders running east-west and eight girders running north-south (between the building wall and the reactor wall). The top of the steel framework is located 8-2

,i  :

at 72'-7" elevation. The pool slab and the north and south walls are also supported by steel columns (six total) which are vertically supported by cross girders at the 51' elevation.

The outer building walls are not modelled; however, the west outor building wall is assumed to provide boundary restraint to the extended north and south pool walls.

8.3 Definition of Loads' Pool structural loading involves the following discrete components:

8.3.1 Static Loadina

1) Dead weight of pool structure plus 39 feet of water (including hydrostatic pressure on the pool walls).

. Combining the hydrostatic pressure and structure dead weight is in conformance with (8.1.2].

2) Dead weight of existing and new rack modules and fuel assemblies stored in the modules.

8.3.2 Dynamic Loacling

1) Vertical loads transmitted by the rack support-pedestals to the slab during an SSE or OBE 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.

3) Hydrodynamic loads caused by rack notion in the pool during a seismic event. As discussed in Section 6 of this report, this loading is produced by the relative movement between the rack arrays and the pool dlls during the seismic event. This loading is obtained as a byproduct of the Whole Pool Multi-Rack (WPMR) analysas described-in Section 6. the importance of this loading was not recognized until recently (ca. 1987) and as a result, it was (unconnervatively) neglected in early pool structure integrity analyses.

8-3

.s %p. ---? =- - g .7 - w e w g --e.im

._. . . _ _ . . _ . _ _ _ . _ _ . = _ _ _ _ _ . . . _ - . _ _ = _ . . _ . . . _ _ _ . _ _ _ . _ _

l l

l 8.3.3 Thermal Loadinc Hean tamperature rise and temperature gradient across the ,

pool slab and the pool walls due to temperature ,

differential between the pool water and the ambient e:Kternal to the slab and, walls.

The boiling scenario calculations are based on the in-pool and ex-pool (ambient underneath the slab and adjacert to the walls) temperatures to be 212*F and 70*F, respectively. Analyses indicate that the thermal moment '

based on the linear temperature gradient bounds the moment produced in all transient states leading up to the boiling condition. The duration of boiling is assumed to be limited (less than one week) such that the strength of the zeinforced concrete is not impaired.

8.4 Analysis Procedures 8.4.1 Finite Element Analysis Model The finite element code ANSYS [8.4.1) is used .for structural-qualification. The three-dimensional linearly elastic finite element model is constructed using shell elements for the main pool walls and slab, and boa n elements to model the steel beams and columns. Figures 8.2.1 and 8.2.2 show views of the modelled structure. For clarity of presentation, the underlying beam i elements are not shown in these figures. The effect of reinforcement and concrete cracking is reflected in element properties assigned to various locations during the simulations.

l 8.4.2 Analysis Methodoloav In Section 6 of this document, the results of Whole Pool Multi-Rack analyses have been presented. The results of these analyses (for SSE and OBE seismic events) establish pedestal load time-histories on the pool slab-and the hydrodynamic pressure time-histories for the wall structure adjacent to the racks.

8-4

For purposes of this analysis, the pool is assumed to contain sixteen free-standing, fully loaded, spent fuel racks. A total of l 3859 cells are assumed-loaded conservatively with fuel having dry weight 1360 lbs per assembly. This loading bounds the actual dry l

weight of the BWR fuel (approximately 680 lb per assembly) , and permits assessment of the pool structure for heavier fuel assemblies.

The major structure loadings discussed in Section 8.3 are imposed on the finite element model using static and response spectra analysis.

9 The effect of r,oncrete cracking (permitted in analyses per the ACI Code) is simulated by appropriate reduction of the element Young's Moduli in appropriate regions where cracking is indicated. In this manner load re-distribution that will occur in the walls and slab is fully accounted for in the model.

8.4.3 Load combinations The ACI code (8.1.2) postulates a large number of the so-called factored load combinations. Out of all the mandated factored loading combinations, the following are potentially limiting (af ter deleting those loads which are not applicable to the Pilgrim spent fuel pool). These load combinations exceed those stipulatd in the PHPS FSAR which are derived from (8.1.4).

1.4D i 1.7E D+T. E' D + T, i 1.1C' D + T i E'

1. 05D + 1. 05T, i 1. 3E In the above defined combinations, the notation of [8.1.2) is used:

D = dnad loads and hydrostatic loads E = OBE resultant loads E' = SSE resultant loads T, = thermal loads due to postulated normal thermal condition ,

8-5

T, = thermal loads due to postulated abnormal thermal condition i In addition, a load combination D + Tb is also evaluated where Tb is an assumud condition of pool water boiling. This case is not one required to be considered by the ACI code.

In the load combinations involving seismic combinations, it is noted that the seismic effects act with "plus" or "minus" signs on the structural and hydrodynamic components to reflect the non-determinacy in the direction of action of the seismic loading.

Both directions are evaluated to establish the limits at various  ;

structural locations. Where any load reduces the effect of other loads, and is always present, the corresponding coefficient is taken as 0.9, as specified in (8.1.2).

8.5 Results of Analyses 1 i

The ANSYS postprocessing capability is used to form the appropriate load combinations and to determine the limiting bending moments and shear forces in various sections of the pool structure. Section limit strength moments for bending are computed using appropriate concrete and reinforcement bars present in the various regions of the pool structure.

To assess the qualitative impact of the proposed capacity expansion on the total loading on the Pilgrim pool structure, a calculation is made to evolve an equivalent static load on the wetted pool slab from each of the major distributed mechanical loads. Table 8.5.1 shows the results of the calculation and the effect of each component of load. Table 8.5.2 shown results from limiting load combinations for the bending strength of the slab and walls. For each section, the limiting safety margin is defined as the limiting strength bending moment defined by ACI for that structural section divided by the cal alated bending moment or uhear force (from the 8-6

_---__..L.._

finite element analyses) . The various regions of the pool structure contain a variety of different reinforcement patterns and section thicknesses. Each region is searched independently for the maximum bending moments in different bending directions and for the maximum shear forces. Safety margins are determined from the calculated-  !

maximum bending moments and shear forces based on the local  !

strengths. This procedure is repeated for all the potential l limiting load combinations. In this manner, the limiting safety ,

margins are determined. Table 8.5.3 shows results of shear capacity calculations for the wetted slab. Calculatod margins in Tables 4

8.5.2 and 8.5.3 should be compared with the allowable datum of 1.0.

In Table 8.5.3, the governing factored load case for slab shear is ,

the case reported.

Table 8.5.2 demonstrates that the limiting safety margins for all sections are above 1.0 as required.

Table 8.5.4 shows results of strength calculations for the underslab steel girders. The calculated resultants for each of the {

factored load cases are compared to the allowable resultants for the section calculated in accordance with [8.1.3).

Finally, the most severe factored underslab column load (in >

calculated as) 856 KIPS (compression) . The limit load is 1120 KIPS based on yield and 997 KIPS based on theoretical stability.  !

8.6 Pool Liner 7 The pool liner is subject to in-plane strains due to movement of the rack support feet during the seismic event. Analyses are performed to establish that the liner will net tear or rupture; under limiting loading conditions in the pool, and that the ASME code's cumulative damage factor criterion under the condition of-

- 1 SSE event plus 20 OBE events is met. For conservatism, all liner 1

P 0

8-7

= --

. _.~ . _- ___.._u._.-_.___ __ __. . . - _ - _ __ . . _ . . _ __

analyses are based on loadings imparted fron the most highly loaded pedestal in the pool. Bearing strength requirenants on tne most highly loaded pedestal are also shown to be satisfied.

8.7 Conclusions Regions affected by loading the fuel pool completely with existing and new high density racks ato examined for structural integrity under bending and shearing action. It is demonstrated that adequate safety margins exist assuming that all racks are fully loaded with a bounding fuel weight for all factored load

ombinations. It is also shown that cyclic loading on the liner under the seismic event does not violate the integrity of the liner (from fatigue failure) and that concrete bearing strength limits are nc.t exceeded.

8.8 References (8.1.1) OT Position for Review and Acceptance of Spent Fuel Handling Applications, by B.K. Grimes, USNRC, Washington, D.C., April 14, 1978.

(8.1.2) ACI 349-85, Code Requirements for Concrete Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan.

(8.1.3) AISC Manual of Steel Constructior., 9th Edition.

(B.1.4) ACI 318-63, Building Coda Requirements for Reinforced Concrete.

[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.

8-8

4 Table 8.5.1 BREAKDOWN OF VARIOUS CONTRIBUTIONS To SLAB LOAD (hTf7ED AREA)

Dead Weight of Slab 1318.2 KIPS Dead Weight of Water 3146.6 KIPS Dead Weight of Racks + Fuel

• 4940.6 KIPS Dynamic Addar from Racks + Fuel (SSE) 1808.2 KIPS Dynamic Adder from Racts + Fuel (OBE) 1265.5 KIPS Based on 1360/ dry weight fuel.

l t .-

. . , _ . , . _ _ , . . . - - . _ , . . . _ _ . - 4... _ . _ _ . - . - , _ , ,

Table 8.5.2 LIMITING MARGINS FOR FLEXURE OF SLAB AND WALLS LOAD 1 M, LOCATION CASE (IGP IN/IN) (KIP IN/IN) M/M, Slab - Along Reactor Wall 1.4D+1.7E 719.97 1027.1 1.427 West Wall: Lower Section - Horizontal D+T% 310.93 814.1 2.62 West Wall: Lower Section Vertical 1.4D+1.7E 393.31 814.1 2.07 i

West Wall: Upper Section - Middle Horizontal D +T,-E' 244.3 350.7 1.44 South Wall: I.ower Section Vertical D+T% 382.6 484.1 1.27

• South Wr.11: Lower Section - Horizontal D+T% 479.6 954.5 1.99 North Walh Lower ,

Section - Horizontal D+T% 480.' 954.5 1.99 North Wall: Lower Section - Vertical D+T% 386.5 484.1 1.25 .

North or South Wall Middle Section D +T,+ E' 188.2 400 2.13 North or South Wall Upper Section D+T, E' 333.7 785.9- 2.36 r

, -w.na, e .,-,~,----,-,-n, ,w,,,w---- ,---wn-- -+-n--,-w,-- --

tr*

Table 8.5.3 SHEAR CAPACITY CRECK OF-SLAB FOR l1.4Dl+l1.7El MARGIN CALCULATED ALLOWABLE OF CONDITION (RTPSi- -(KTPS) SAFETY Punching -tahear at 856 3383 3.952 contral column Two-way slab where slab 1652.7 6024.4 3.65 io concrete panel cupported by beams (242" x 69.75" panal)

Two-way niab where slab ,_

in entirts wetted area 17,344.88 23633.1 1.363 This conservatively neglects any shear capacity from . rainforcement

.and any shear capacity from the beams.

.,,irm ,, - . . , , , - . . . . . . . . . _ i -

Table 8.5.4 UNDERSLAB BEAM SAFETY MARCIN CHECK AXIAL IDAD + BENDING IDAD M,,,, P My ,

14 CATION CONDITION P (k,k,P) (KIP-IN) (KIP) (KIP) M.S. l

1. 4 D-1. 7 E 253 2041.1 997 10008. 4.59 N/S Beams

1. 05 (D* T,) 970.9 6285 1540 16560 1.59 -

E/W Central Berts -1.3E E/W Side 6919 3180 45360 13.97 Beam 1.4D+1.7E 456 l

l H.S. - margin of safety for beams calculated as 2

P,,i, Mr II ~ I II M.S. = P, H,,,,

t 6

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k !f"hY geara00.

j 3 p8 i

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$4 kb M b h;w. -f%

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.c.;. y G TERIOR OF POOL LOOXING FROM UNDERN M TH AND WEST FIGURE 8.2.1 ,

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~~. I r gg g g n# 5 lll stR dq H [j 4,l" l HEciEAinom l

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INTERIOR OF POOL LOOXING TROM EAST FIGURE 8.2.2 1

9.0 RADIOLOGICAL EVALUATION 9.1 Fuel Handlina Accident The potential radiological consequences at the Pilgrim exclusion area boundary (EAB) of a fuel handling accident in the secondary containment have been determined.

9.1.1 Assumptions and Source Term Calculations Evaluations of the accident were based on fuel of 3.54 wt% initial enrichment burned to 44,'100 Mwd /mtU. The reactor was assumed to have been operating at 2038 Mw thermal power (102% of rated power) prior to shutdown, with an average specific power of 20.07 kw/kgU. The fuel handling accident was assumed to result in the release of the gaseous fission products contained in the fuel / cladding gaps of 111 fuel rods in peak power fuel assemblies. Gap inventories of fission products available for release were estimated using the release fractions identified in Regulatory Guide 1.251 except for Iodine-131, for which the release fraction is increased 20% in accordance with NUREG/CR-50098 Cooling time for the failed fuel prior to the accident was 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The gaseous fission products that have significant impacts on the off-site doses following short fuel cooling periods are the short-lived nuclides of iodine and xenon, which 1

-Regulatory Guide 1.25 (AEC Safety Guide 25), " Assumptions Used For Evaluating The Potential Radiological Consequences Of A Fuel Handling Accident In The Fuel Handling And Storage Facility For Boiling And Pressuris:ed w a ter Reactors", March 23, 1972.

C. E. Beyer, et al., " Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors", NUREG/CR-5009, Pacific Northwest Laboratory, February, 1988.

9-1

l reach 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 assemblies, the specific power and hence the inventory of iodine and xenon will be proportional to the radial power peaking factor (conservatively taken as 2.0).

At the cooling time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> used in the Pilgrim calculations, most of the thyroid dose comes from Iodine-131, while most of the whole-body dose comes from Xenon-133 and Xenon-135. Though these iodine and xenon isotopes are the major contributors to off-site doses, the contributions from other radionuclides are calculated and included in the overall dose values.

The present evaluation uses values for atmospheric diffusion factor (x/Q) and for filter efficiencies that have been specified previously by the Boston Edison Company. Core specific inventories (Curies per metric ton of uranium) of fission products were estimated with the ORIGEN-2 code 8, based upon parameters stated earlier (specific power of 20.07 kw/kgU, initial enrichment of 3.54 wt% U, burnup of 44,100 Mwd /mtu, and a cooling time of 24 hours). The results of the ORIGEN calculations for isotopes that contribute to the thyroid and whole-body doses are given.in Table 9.1, while Table 9.2 lists pertinent data for the isotopes of interest. Data and assumption =

used in the dose calculations are given in Table 9.3.

The following equation, from Reg Guide 1.25, was used to calculate the thyroid dose (D, in rads) from the inhalation of 3

"QBNL 1sotope EHeration and Depletion", ORNL/TM-7175, Oak Ridge National Laboratory, July, 1980.

9-2

radiciodine. Values for many of the terms in the equation are given in Table 9.2 and Table 9.3.

Dose = [ , where a DF, DF, F' = fraction of fuel rod iodine inventory in gap space R* = dose factor (rads per conversion curie)

I* = core iodine radionuclide inventory at time of the X/Q= atmospheric diffusion accident (curies) factor (sec per cubic meter)

F= fraction of core damaged so as to release iodine in the D F"=

effective iodine rod gap decon. factor for pool water P= core peaking factor DF,= effective iodine B= breathing rate (cubic meters decon. factor for per second) filters The equation given below was used to calculate the external whole-body dose from gamma radiation in the cloud of radionuclides released in the fuel-handling accident. The equation contains several of the terms defined above.

Dose, = [ 0.25 (x/0) FPG 1 E,,1 In this expression, G 3 is the gap inventory of the gaseous radionuclides of xenon and krypton, while the E,,,y3,, term is the average energy per disintegration of each radionuclide (in Mev per disintegration, as given in Table 9.2). These functions assume the noble gas decontamination factors in water and the charcoal filters are 1.0. The gap inventories of radioiodine make negligible contributions to the whole body dose, D, , because of the large decontamination factors appropriate to the iodines.

9-3

9.1.2 Results The doses at the Pilgrim EAB from the specified fuel handling accident are tabulated below. The doses are based on the release of all gaseous fission product activity in the gaps of 111 fuel rods in highest-power assemblies.

Thyroid dose, rad = 0.673 Whole-body dose, rem = 0.415 These potential doses are well within the exposure guideline values of 10 CFn Part 100, paragraph 11. As defined in Standard Review Plan 15.7.4, Radiolecical Cons ecuence s of Fuel Handlina Accidents, "well within" means 25 percent or less of the 10CFR100 guidelines, or values of 75 rad for thyroid doses and 6.25 rem for wbsle-bocy doses. The potential doses at Pilgrim from the conservative scenario presented here meet the criteria for "well within."

9.2 Solid Redwaste The necessity for resin replacement is determined primarily by the requirement for water clarity, and the resin is normally changed about once a year. No significant increase in the volume of solid radioactive wastes is expected with the expanded storage capacity. During operations to expand the fuel storage capacity at Pilgrim, a smcil amount of additional resins may be generated by the pool cleanup system on a one-time basis.

9.3 Gaseous Releases Gaseous releases from the fuel storage area are combined with other plant exhausts. Normally, the contribution from the fuel storage area is negligible compared to the other 9-4

l releases and no significant increases are expected as a result of the expanded storage capacity.

9.4 Personnel Excesures 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 1.0 to 2.0 mrem /hr, with a few areas such as the pool bridge showing dose rates of 2.0 to 4.0 mrem /hr.

Radiation levels in zones surrounding the pool are not expected to be affected significantly. Existing shielding around the pool (water and concrete walls) provides more than adequate protection.

Radionuclide concentrations typical of those found in pool water are shown in Table 9.4. During fuel reload operations, the concentrations might be expected to increase due tc crud deposits spalling from spent fuel assemblies and to activities carried into the pool from the primary system. However, experi-ence to date has not indicated a major increase as a consequence I

of refueling.

Operating experience has also 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.

No increase in radiation exposure to operating personnel is expected; thus, neither the current health physics program nor the area monitoring system needs to be modified.

9-5

I 9.5 Anticipated Exoosure Durina Expansion All of the operations involved in increasing storage capacity will utilize detailed procedures prepared with full consideration of ALARA principles. Similar operations have been performed in a number of facilities in the past, and there is every reason to believe that the expansion in capacity can be safely and efficiently accomplished at Pilgrim, with minimum radiation exposure to personnel.

Total occupational exposurr for the expansion operation is estimated to be between 2 and'4 person-rem, as indicated in Table 9.5. While individual task efforts and exposures may differ from those in Table 9.5, the total is believed to be a reasonable estimate for planning purposes. Divers will be used only if necessary, but the estimated person-rem burden includes a figure for their possib',e exposure.

The existing radiation protection program at Pilgrim is adequate for operations associated with increasing the fuel-storage capacity. .n e. r e there is a potential for significant airborne activity, continuous air samplers will be 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 issued to each individual. As a minimum, this will include thermoluminescent dosimeters and pocket dosimeters. Additional personnel monitoring equipment (i.e., extremity badges or alarming dosimeters) may be utilized a- required.

Work, personnel traffic, and the movement of equipment will be monitored and controlled to minimize contamination and to assure that exposures are maintained ALARA.

9-6

Table 9.1- RESULTS OF ORIGEN-2 CALCULATIONS FOR RADIONUCLIDES OF IODINE,-KRYPTON,~AND XENON-A? 24-HOURS COOLING' TIME H3dionuclide Curies ner mtU I-131 5.076 x 10' I-132 6.385 x 105 I-133 4.934 x los I-134 2.519 x 10-2' I-135 8.157 x 10' Kr-85 1.083 x 104 Kr-85m 2.832 x 105

}: 87 4. 4 92 x 10-1 Kr-C-il 8.526 x 10:

Xe- 131m 6.084 x 103 Xe-133 4.934 x 105 Xe-133m ,

3.070 x 104 Xe-135 -2.670 x 105 Xe-135m 1.307 x 10*

9-7

e e

Table 9.2: RADIONUCLIDE PROPERTIES USED IN THE  !

FUEL-HANDLING ACCIDENT ANALYSIS

, Dose Conversion,-

Radionuclide Rads / Curie E, (Mev)

Iodine-131 1.48 x 10' -----

Iodine-132 5.35 x 10' -----

. Iodine-133 4.0  :. 1Cs _____

Iodine-134 2.5 x 10' -----

Iodine-135 1.24 x 105 -----

Krypton-85 -----

0.002 Krypton-85m -----

0.151 Krypton-87 -----

1.374 l-Krypton-88 -----

1.745 Xenon-131m -----

0.020 Xenon-133 -----

0.047 Xenon-133m- '

0.040 Xenon-135 -----

0.246-1 Xenon-135m -----

0.428- i i

lL l-9-8

~_

l]

s i

i j

. Table 9.3 DATA AND: ASSUMPTIONS FOR THE EVALUATION- I OF THE FUEL HANDLING ACCIDENT i Core power level,- Mw(t) 2038 Fuel enrichment, wt% U 3.54

-Fuel burnup, Mwd /mtU 44,100 Specific power, kw/kgU 20.07

! . Power peaking factor 2.0 p Fuel cooling time, hrs -

-24 No. of failed fuel rods 111 Core inventory released to gap, %

Iodine-131 12 Other icdines 10 Krypton-85 30 Other kryptons 10 Xenons 10 Iodine composition, %

Elemental 99.75 Organic' O.25-Pool decentamination factors Elemental lodine 133 t-Organic iodine -

1 Noble gases 1 Filter decontamination factors Elemental iodine 100 Organic lodine 100 Noble gases 1

. Atmospheric diffusion factor (%/Q), sec/m3 2.71 x 10-4 Breathing rate, m 3 /sec 3. 47 x 10-4 9-9

Tcble 9.4 TYPICAL CONCENTRATIONS OF RADIONUCLIDES IN SPENT FUEL POOL WATER Concentration, Nuclide uCi/ml Mn-54 6 x 10-'

Co-60 3 x 10-5 Cs-134 3.3 x 10 ' ->

Cs-137 2. 4 x 10-5 4

9 - 10 i

' Table-9.5 PRELIMINARY ESTIMATE OF PERSON-REM EXPOSURES DURING RACK EXPANSION 1

Estimated Number of ' Person-Rem m Personnel Hourg Exoosure m Remove underwater appurtenances (if necessary) 4 5 .0.4- to 0.8' Clean and vacuum pool (if necessary) 3 25 0.3_ to .0.6 Move fuel (if necessary) 2 150 0.8- to 1.5 -

Install new rack modules ,5 35 0.4 to 0.8 Total Exposure, person-rem 2 to 4 t

m ~

Assumes minimum dose rate of 2-1/2. mrem /hr -(expected) to a maximum of 5 mrem /hr except for pool vacuuming. operations,.

which assume 4 to 8 mrem /hr, and possible diving cperations; which assume 20 to 40 mrem /hr.

c 9 - 11

, - _ - - . - _ _ _ - . - . _ . - . ~ _ - -

l 10.0 DDRAL SURVEILLANCE PROGRILM 10.1 P_urp210 Boral", the neutron absorbing material incorporated in the spent fuel storage rac.k design to assist in controlling system reactivity, consists of finely divided particles of boron carbide (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 fual pool have demonstrated the stability and chemical inertness of Boral (References [10.1.1)-[10.1.3)) . The accumulated dose to the Borel over the expected rack lifetime is entimated to be about 3 x 10" to 1 x 10" rads depending.upon how the racks are used and the number of full-core off-loads that may be necessary.

Based upon the accelerated test programs and a large body-of in- <

pool data, Boral is considered a satisfactory material for reac-tivity control in spent fuel storage racks and is fully expected to fulfil its design :1' unction over the lifetime of. the racks.

Nevc.rtheless, the USNRC requires the Licensee 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. The April 14, 1978 USNRC letter to 'all power reactor licensees (Reference

[10.1.4)), specifies that

" Methods for verification of long-term material stability and mechanical integrity of special poison materials utilized for neutron absorption should include actual tests."

The purpose of the surveillance program presented herein is to characterize certain properties of the Boral with the objective _ of providing data necessary to assess the capability of the Boral 10-1 e

a. -

___-.__m.____-.______.___._________._.____a -_.- - .

paneln 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.

In response to the need for a comprehensive Boral surveillance program to assure that the subcriticality requirements of the stored fuel array are safely maintained, a surveillance program 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 swelling) and boron content.

10.2 COUPON SURVEILLANCE PROGRAM 10.2.1 Coubon Description The coupon measurement program includes coupons suspended on a mounting (called a " tree"), placed in a designated cell, and surrounded by spent fun 1. Coupons will be removed from the array on a prescribed schedule and certain physical and chemical properties will be measured from which the stability and integrity of the Boral in the storage cells may be inferred.

Each surveillance coupon will be api.roximately 4 inches wide and 6 inches long. The coupon surveillance program will use a total of 8 test coupons with each coupon mounted in a stainless steel jacket, simulating as nearly at possible, the actual in-service geometry,  ;

physical mounting, materials, and flow conditions of the Boral in the storage racks. The jacket (of the same alloy used in manufac-ture of the racks) will be closed by screws or clamps to allow easy opening with minimum possibility of mechanical damage to the Boral 10-2 1

l specimen inside. In mounting the coupons on the tree, the coupons will be positioned axially within the contral 8 feet of the fuel zone where the gamma flux is expor:ted to be reasonably uniform.

Ench coupon will be carefully pre-characterized prior to insertion in the pool to provide reference initial values for comparison with measurements made af ter irradiation. The surveillance coupons will be pre-characterized for weight, length, width and thickness. In addition, two coupons (which nood not be j ac.keted) will be precorved as archive samples for comparison with subsequent test coupon measurements. Het chemical analyses of samples frem the same lot of Boral will be available.from the vendor for comparison.

10.2.2 Surveillange CouPAD Testiner Schedule The coupon trea is surrounded by freshly discharged fuel assemblics at each of the first five refuelings following installation of the racks to assure that the coupons will have experienced a slightly higher radiation dose than the Boral in the racks. Beginning with the fifth load of spent fuel, the fuel assemblies will remain in place for the remaining lifetime of the racks. A sample coupon management schedule is shown in Table 10.1.

At the time of the first fuel off-load following installation of the coupon tres, the (8) storage cells surrounding the tree shall be loaded with freshly-discharged fuel assemblins that had been among the higher specific power assemblies in the core. Shortly before the second reload, the coupon tree is removed and a coupon is removed for evaluation. The coupon tree is then re-installed and, at reload, again surrounded by freshly discharged fuel assemblien. This procedure is continued for the third, fourth, and fif th off-loading of spent fucl (except that a coupon is not pulled at the fourth refueling) . From the fifth cycle on, the fuel assemblies in the (8) surrounding cells renain in place.

10-3

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 Borel 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.

Coupons which have not been destructively analyzed by wet-chemical processes, may optionally be returned to the storage pool and re-mounted on the tree. They will then be available for subsequent investigation of defects, should any be found.

10.2.3 Measurement Proctan The coupon measurement program is intended to monitor changes in physical properties of the Boral absorber material by performing the following measurements on the pre-planned schedule:

Visual Observation and Photography, a

Neutron Attenuation, Dimensional Measurements (length, width and thickness),

Weight and Specific Gravity, and Wet-chemical' analysis-(Optional).

10-4 1

The most significant measurements are thickness (to monitor for swelling) and neutron attenuation (to confirm the concentration of Boron-10 in the absorber material). In the event loss of boron in observed or suspected, the data may be augmented by wet-chemical analysis (a destructive gravimetric technique for total boron only).

10.2.4 Eurveillance Coupon Acceptante Criteria Of the measurements to be performed on the Boral survaillance coupons, the most important are (1) the neutron attenuation measurements (to verify the continued presence of the boron) and (2) the thickness measurement (as a monitor of potential swelling) .

Acceptance criteria for these measurements are as follows:

• A decrease of no more than 5% in Boron-10 con-tent, as determined by neutron attenuation, is acceptable. (This is tr.ntamount to a requirement for no loss in boro.' within the accuracy of the measurement.)

- An increase in thickness at any point should not exceed 10% of the initial thickness at that point.

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.

a Neutron attenuation measurements are a precise instrumental method of chemical snalysis for Boron-10 content using a non-destructive technique in which the percentage of thermal neutrons transmitted through the panel is measured and compared with pre-determined calibration data. Boron-10 is the nuclide of principal interest since it is the isotope responsible for neutron absorption in the Boral panel.

10-5

c 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 noud for further attention l and possibly a change in measurement schedule. These include (1) visual or photographic evidence of unusual surface pitting, corrosion or edge deterioration, or (2) unaccountable weight loss in excess of the u<4asurnment accuracy.

Procedures for coupon surveillance measurement have been provided )

to the Licensee by the rack supplier's vendor, Nusurtec, Inc. l 10.3 Referencen l

[10.1.1) "Spant Fuel Storage Module Corrosion Report",

Brooks & Perkins Report 554, June 1, 1977. j

[10.1.2) " Suitability of Brooks & Perkins Spent Fuel j Storage Module for Use in PWR Storage Pools", j Brooks & Perkins Report 578, July 7, 1978.

[10,1.3) "Boral Neutron Absorbing / Shielding Material - j Product Performance Report", Brooks & Perkins Report 624, July 20, 1932.

[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 10-6

Table 10.1 COUPON _ HEASUREMDIT SCHEDULE

'i I

Counon- Refuelina'I After Rerack/ Years 1 1 lat

2 2 nd'**

3 3rd(2) 4 5 th'I' 5 8th 6 5 '3 '

7 1 0'33 8 16'3' Remove coupons for evaluation within the 1 or-2 mcnths preceding the next refueling.

'** _ Place freshly discharged fuel in the 8 surrounding cells at the beginning of the 1st, 2nd, 3rd, 4th, and 5th refueling cycles after completion of reracking.

'33 Years after the 8th refueling. .

11.0 ENVIRONMENTAL COST-BFNEFIT ASSESSMENT 11.1 Introduction Article V of the USNRC OT Position paper (Ref. [11.1]) requires the submittal of a cost / benefit analysis for the chosen fuel storage capacity enhancement method. This section abstracts the analyses and evaluations made by Boston Zdison Co:apany before selection of adding racks as the most viable alternativ.2. Boston Edison Company had previously received Amendment No. 33 to Pilgrim Nuclear Power Station Operating License in 1976. This amendment addresso.' all relevant environmental considerations in determining the findings and conclusions for supporting Final Environmental Assessment for Amendment No. 33. All environtental related findings and conclusions of Amendment No. 33 are applicable to this proposed Spent Fuel Pool (SFP) expans. ion.

11.2 Imnerative f or Addiner Rac)gi_,

The specific need to increase the limited existing storage capacity of the PNPS spent fuel pool is based. on the continually increasing 4

inventory in the pool, the prudent practice to maintain full-core off-load capability, and a lack of viable economic alternatives.

Reference is made to Table 1.1 of Section 1 wherein the current and projected fuel placement in the PNPS spent fuel pool are tabulated. This shows the PNPS fuel pool will lose the capacity to discharge one full core (580 fuel assemblies) in 1995.

,. The projected loss of storage capacity in the PNPS por' would affect BECo's ability to operate the reactor.

11"1

11.3 Acoraisal of Alternatives BEco has determined that adding racks is by. far the most viable option for the PNPS pool in comparison to other alternatives.

The key considerations in evaluating the alternative options were: 1 Safety: minimize the number of fuel handling steps Economy: minimize total installed and OEM cost Security: protection from potential saboteurs, natural phenomena Non-intrusiveness: minimize required modification to existing systems Maturity: exte,nt of industry experience -with- the technology ALARA: minimize cumulative dose due to. handling of fuel BEco considered three urtions to provide Cf.ntinued full-core reserve in the fuel pool: rod consolidation, -dry storage or additional racks in the existing SFP. BEco found adding racks to be the most attractive option based on application of the key considerations.

An overview of the alternative technologies considered by BECo ls provided in the following:

Rod ConsolidaticD Past rod consolidation has been shown to be a feasible technology.

Rod consolidation involvet- disassembly of spent fuel, followed by the storage of the fuel rods from two assemblies into the volume of one and the disposal of the fuel assembly skeleton outside of the pool. The rods are stored in a stainless steel can that has 11-2

the outer dimensions of a fuel assembly. The can is stored'in the spent fuel racks.- The top of - the can has an end fixture that matches up with the spent fuel handling tool. This permits moving-  ;

the cans. -l 1

l Rnd consolidation pilot projects have consisted of underwater tooling manipulated by an overhead crane and operated by a maintenanen worker. This is a vary slow and repetitive process.

Industry experience with rod consolidation has been mixed. The.

principal advantages of this technology are: the ability te modularize, compatibility with DOE waste management system, no need of additional land and no additional required surveillance.- The disadvantages are: potential- gap activity release due to rod breakage, potential for increased fuel cladding corrosion due to some of the protective oxide layer being scraped off,' potential interference of the (prolonged) consolidation activity which might interfere with ongoing -plant operation, and lacx of sufficient industry e'parience resulting in a cost roughly twice that of adding racks.

Dry Fuel Storace Dry fuel storage is a method of storing spent nuclear fuel in certified steel / concrete casks or in concrete vaults. These casks / vaults provide radiation shielding and passive heat dissipation. Typical cask / vault capacity for BWR fuel range from 40 to 60 assemblies that have been removed from the reactor for at.

least five years. The casks, once loaded, are then stored outdoors on a seismically qualified concrete pad or in concrete vaults.

Such storage locations require a high level of security including frequent tours, reliable lighting, intruder detection, and 11-3 t

I continuous visual monitoring. In addition, compliance with applicable environmental regulation exceeds the scope of the-original plant licensing basis.

The casks are prasently licensed to a 20 year storage service life and are for storage only. When the 20 year certification has expired, the cask manufacturer or the utility must recertify the cask, or the utility must renove the spent fuel from the container.

Presently no cask has dual certification; it is certified for either storage or transportation. Dual purpose casks, if and when certified, will have reduced capacity, increasing the quantity of casks required.

There are several plant modifications required to support cask use, listed below. Tap-ins must be made to the gaseous waste system.

Chilled water is required to support vacuum drying of the spent fuel. Piping must be installed to return cask water back to the spent fuel pool / cask pit. A seismic concreue pad, or alternatively; vaults, must be built to store the loaded casks.

The reactor building crane must be single-failure proof in order to transport the casks. A dry fuel storage facility must have a high level of security, similar to that at the plant itself. The space between the inner and outer cask lid must be continuously mcnitored to assure there is no seal failure.

Boston Edison has estinated that dry storage costs are about two and one-half to three times the cost of additional rack storage, and offer no significant advantage reslative to our other decision criteria.

Other concerns relating to the dry fuel storage system are:

inherent eventual " repackaging" for shipment to the DOE repository, the responsibility to eventually decommission the facility at the 11-4 1

end of the useful life, and relatively large land consumption.

Implementation of this type of storage facility would require compliance original plantwith applicable licensing basis.environmental regulations exceeding the Other Ontions Constructing and licensing a new fuel pool is not a practical alternative for PNPS since such an effort may take up to 10 years .

PNPS requires expanded storage capacity beginning in 1995 . Also, a new spent Department of fuel pool can not be justified in light of the Energy's proposed plan for spent fuel rtorage at remote national locations scheduled to begin on or before the year 2000.

The shutdown of Pilgrim for lack of spent fuel storage capacity is not an alternative for consideration. Moreover, the cost of this option is prohibitively high.

To summarize, the on-site there are no acceptable alternatives to increasing storage. spent fuel storage capacity of PNPS except offsite Offsite storage is not faasible since there are no ecmmercial the U.S. independent spent fuel storage facilities operating in Second, BECo has only one nuclear power plant, making it impossible to ship and store spent fuel in the unused capacity of another system plant. .

If DOE's proposed approach to interim storage and long termrage sto of spent fuel facilities is realized within the DOE proposed schedules, BECo's proposed planned expannion of spent fuel storage capacity allows DECO's participation.

11-5

_ . ~ . - - - _

11.4 Resour.ce Coppitment The expansion of the PNPS spent fuel pool capacity is expected to require the following primary resources:

95 tons i

Stainless Steel:

Boron neutron B tons of which 6-5 tons is Boron Carbide powderabsorber:

and 1.5 tons are aluminum.

The requirements for stainless steel and aluminum represent a small fraction of total world output of these metals (less than .001%).

The Boron Carbide requirnd is a somewhat higher frhction of world it is unlikely output than that of stainless steel or aluminum, that the commitment of Baron carbide to this project will affect other alternatives.

Experience has showa that the production can beof Boron easily carbide istohighly expanded depends upon need, and variable, accommodate worldwide needs.

11.5 Environmental Considerations 1and Use The Pilgrim SFP is located next to the reactor inside the reactor building. The proposed modification will not alter the external physical geometry of the SFP. No additional conmitment of land is required.

Water Use There is no significant change in plan

• water useThe as storage a result of-of the proposed spent fuel pool storage expansion.

additional spent fuel assemblies in the SFP will slightly increase 11-6

p'n'.

l end of the useful life, and relatively large land consumption.

Implementation of this type of storage facility would require compliance with applicable environmental regulations exceeding the original plant licensing basis.

Other Ootigng constructing and licensing a new fuel pool is not a practical alternative for PNPS since such an effort may take up to 10 years.

PNPS requires expanded storage capacity beginning in 1995. Also, a new spent fuel pool can not be justified in light of the Department of Energy's proposed plan for spent fuel storage at remote national locations scheduled to begin on or before the year 2000. The shutdown of Pilgrim for leck of spent fuel storage capacity is not an alternative for consideration. Moreover, the cost of this option is prohibitively high.

To summarize, there are no acceptable alternatives to increasing the on-site spent fuel storage capacity of PNPS except offsite storage. Offsite storage is not feasible since there are no commercial independent spent fuel storage facilities operating in the U.S. Second, BECo has only one nuclear power plant, making it impossible to ship and store spent fuel in the unused capacity of another system plant.

If DOE's proposed approach to interin storage and long- term storage of spent fuel facilit.4es is realized within the DOE proposed schedules, BECo's propoued planned expansion of spent fuel storage capacity allows BECo's participation.

11-5

11.4 Researce Committent The expansion of the PNPS spent fuel pool capaci is expected to require the following primary resources:

Stainless Steel: 95 tons Baron neutron absorber: 8 tons of which 6.5 tons is Boron Carbide powder and 1.5 tons are aluminum.

The requirements for stainless steel and aluminum represent a small fraction of total world output of these metals (less than .001%).

The Boron carbide required is a nomewhat higher fraction of world output than that of stainless steel or aluminum, it is unlikely that the commitment of Boron Carbide to this project will affect other alternatives.

Experience has shown that the production at Boron Carbide is highly variable, depends upon need, and can be easily expanded to accommodate worldwide needs.

11.5 Environmental Considerations LADd VS9 The Pilgrim SFP is located next to the reactor inside the reactor building. The proposed mociification will not alter the external physical geometry of the SFP. No additional cnsmitment of land is required.

Water Use There is no significant change in plant water use as a result of the proposed spent fuel pool storage expansion. The storage of additional spent fuel assemblies in the SFP will slightly increase 11-6 I

the heat load on the SFP cooling system. The SFP coolitig system transfers hear to the Reactor Building Closed Cooling Water System then to the plant service water system. The change in SFP heat level would have an extremely small Ampact on these systems.

Due to-the additional heat-load arising from increased spent fuel-pool inventory, the anticipated maximum bulk pool temperature increases from a previously-licensed 125IF to 142IF as detailed in the calculations described in Section 5.0 of this report. This temperature is arrived at _ by assuming fully-degraded spent fuel pool coolers (fully fouled, maximum allowable number of tubes plugged). The assumption of maximum fouling is conservative given the actual condition of the PNPS coolers. The resultant total heat load (worst case) is less than 25.9 million Btu /hr, which is less than .05% of the total plant heat loss to the environment and well within the capabil'Mr of the plant cooling system.

The increased bulk pool temperature will result in an increase in the pool water evaporation rate. Calculation shows this results in relative humidity increased in the Fuel Handling Building atmosphere of less than 10%. This increase is within the capacity of the reactor building ventilation system. The net result of the increased heat loss and water vapor emission to the environment is negligible.

Imoact on_the Communitv_

i The Nuclear Regulatory Commission increased the storage capacity to the current level of 2320 cells in Amendment No. 33 to Pilgrim Nuclear Power Station Operating License. The environmental findings and conclusions reached by the NRC in granting Amendment No. 33 are applicable to the proposed spent fuel storage expansion.

11-7

The current safety assessment in support of this spent fuel storage expansion in bounded by the Environmental Impact Appraisal to Amendment No. 33 No environmental impact on the environs outside the spent fuel storage building are expected during installation of the new racks.

Activity impact within this building is expected to be limited to that normally associated with metal working activities. No significant environmental impact on the community results from the proposed action.

11.6 .teferences i

[11.1) OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications, USimC (April, 1978).

[11.2] Electric Power Research institute, Report No.

NF-3580, May, 1984.

[11.3) " Spent Fuel Storage Options: A Critical Appraisal", Power Generation Technology, Sterling Publishers, pp. 137-140, U.K.

(November , 1990) .

11-8

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