ML18100A340

Licensing Rept for Spent Fuel Storage Capacity Expansion Pse&G,Salem Generation Station,Units 1 & 2.

ML18100A340
Person / Time
Site:	Salem
Issue date:	04/28/1993
From:	HOLTEC INTERNATIONAL
To:
Shared Package
ML18100A338	List: ML18100A337 ML18100A339 ML18100A340
References
HI-92950, NUDOCS 9305100027
Download: ML18100A340 (394)

Text

HOLTEC INTERNATIONAL LICENSING REPORT for SPENT FUEL STORAGE CAPACI'IY EXPANSION .

PUBLIC SERVICE ELECTRIC & GAS C01\1PANY

• SALEM GENERATING STATION UNITS 1 & 2 USNRC DOCKET NOS. 50-272 & 50-311

• _

Prepared by HOLTEC INTERNATIONAL 2060 Fairfax Avenue Cherry Hill, NJ 08003-1666 REPORT ID-92950

( 9305100027 930428 PDR *ADOCK 05000272 P PDR

--1 TABLE OF CONTENTS SECTION*

1.0 INTRODUCTION

1-1 2.0 MODULE LAYOUT FOR INCREASED STORAGE 2.1 Background 2-1 2.2 Multi-Region Storage 2-1 2.3 Material Considerations 2-3 2.3.1 Introduction 2-3 2.3.2 Structural Materials 2-3 2.3.3 Poison Material 2-4 2.3.4 Compatibility with Coolant 2-6 2.4 Retained Exxon Rack Modules and Proposed 2-6 Reracking Operation 2.5 Heavy Load Considerations for the 2-7 Proposed Reracking Operation 3.0 RACK FABRICATION AND APPLICABLE CODES 3.1 Fabrication Objective 3-1 3.2 Rack Module for Region I 3-1

• 4.0 3.3 3.4 (Retained Exxon Racks)

Rack Module for Region II Codes, Standards, and Practices for the*

Salem Nuclear Generating Station Spent Fuel Pool Racks CRITICALITY SAFETY ANALYSIS 3-2 3-3 4.1 Design Bases 4-1 4.2 Summary of Criticality Analyses 4-4 4~2.1 Region 1 storage Cells 4-4 4.2.2 Region 2 Storage Cells 4-5 4.2.3 Abnormal and Accident 4-8 Conditions 4.3 Reference Fuel Storage Cells 4-9 4.3.1 Reference Fuel Assembly 4-9 4.3.2 Region 1 Fuel Storage Cells *4-9 4.3.3 Region 2 Fuel Storage Cells 4-9 4.4 Analytical Methodology 4-10 4.4.1 Reference Design Calculations 4-10 4.4.2 Fuel Burnup Calculations 4-11 and Uncertainties 4.4.3 Effect of Axial Burnup 4-12 Distribution i

4.4.4 Long-Term Changes in Reactivity 4-13 4.5 Region 1 Criticality Analyses and 4-13 Tolerances 4.5.1 Nominal Design Case 4-13 4.5.2 Peripheral Cells 4-14 4.5.3 With IFBA Assemblies 4-14 4.5.4 With Credit for Fuel Burnup 4-16 4.5.5 Uncertainties Due to 4-16 Tolerance in Region 1 4.5.5.1 Boron Loading Tolerances 4-16 4.5.5.2 Tolerance in Cell Lattice Spacing 4-17 4.5.5.3 Stainless Steel Thickness 4-17 Tolerance 4.5.5.4 Fuel Enrichlitent and Density 4-17 Tolerances 4.5.6 Eccentric Fuel Positioning 4-17 4.6 Region 2 Criticality Analyses and 4-18 Tolerances 4.6.1 Nominal Design Case 4-18 4.6.2 Uncertainties due to Tolerances 4-19 4.6.2.1 Boron Loading Tolerances 4-19 4.6.2.2 Boral Width Tolerance 4-19 4.6.2.3 Tolerance in Cell Lattice 4-19 Spacing 4.6.2.4 Stainless Steel Thickness 4-19 Tolerance 4.6.2.5 Fuel Enrichment and Density 4-20

• 4.6.3 4.6.4 4.6.4.1 4.6.4.2 4.6.4.3 Tolerances Eccentric Fuel Positioning Alternate Storage Configurations Checkerboard Loading Peripheral Cell Locations 3x3 Array with Fresh Fuel 4-20 4-20 4-21 4-21 4-22 4.6.5 Summary of Region 2 Criteria 4-23 4.7 Abnormal and Accident Conditions 4-24 4.7.1 Temperature and Water Density 4-24 Effects 4.7.2 Dropped Fuel Assembly 4-25 4.7.3 Lateral Rack Movement 4-25 4.7.4 Abnormal Location of a Fuel 4-25 Assembly 4.8 References 4-26 Appendix A to Section 4 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5.1 Introduction 5-1 5.2 Spent Fuel Cooling and Cleanup System 5-2 Description 5.2.1 Design Bases 5-2 ii

5.2.2 System Description 5-2 5.2.3 Components 5-4 5.2.4 Design Evaluation 5-6 5.3 Decay Heat Load Calculations 5-7 5.4 Discharge Scenarios 5-9 5.5 Bulk Pool Temperatures 5-10 5.6 Local Pool Water Temperature 5-14 5.6.1 Basis 5-14 5.6.2 Model Description 5-15 5.7 Cladding Temperature 5-16 5.8 Results 5-18 5.8.1 Bulk Pool Temperature 5-18 5.8.2 Time-to-Boil 5-19 5.8.3 Local Water and Cladding 5-20 Temperature 5.9

• Ref er enc es 5-20 6.0 STRUCTURAL/SEISMIC CONSIDERATIONS 6.1 Introduction 6-1 6.2 Analysis Outline 6-2 6.3 Synthetic Seismic Time-Histories 6-7 6.3.1 - Acceptance Criteria for 6-7 Synthetic Time-Histories 6.3.2 Procedure for Time-History 6-8 Generation 6.3.3 Broadened Response Spectra 6-9 of Salem.Spent Fuel Pool Slab 6.3.4 Synthetic Time-Histories 6-9 6.4 Modelling for Dynamic Simulations 6-11 6.4.1 General Remarks 6-11 6.4.2 The 3-D 22 DOF Model for 6-13 Single Rack Module Analysis of New Maximum Density Racks 6.4.2.1 Assumptions 6-13 6.4.2.2 Model Details for New.Fuel Racks. 6-14 6.4.2.3 Fluid Coupling Details 6-15 6.4.2.4 Stiffness Element Details 6-16 6.4.3 Whole Pool Multi-Rack (WPMR) Model 6-18 6.4.3.1 General Remarks 6-18 6.4.3.2 Whole Pool Fluid Coupling 6-18 6.4.3.3 Coefficients of Friction 6-19 6.4.3.4 Modelling Details 6-19 6.5 Acceptance criteria, Stress Limits 6-20 and Material Properties 6.5.1 Acceptance Criteria 6-20 6.5.2 Stress Limits for Various 6-22 Conditions 6.5.2.1 Normal and Upset Conditions 6-22 (Level A*or Level B) iii

6.5.2.2 Level D Service Limits 6-24 6.5.2.3 Dimensionless Stress Factors 6-24 6.5.3 Material Properties 6-25 6.6 Governing Equations of Motion 6-25 6.7 Results of 3-D Nonlinear Analyses 6-27 of Single Racks 6.7.1 Summary Results for New Racks 6-28 in the Fuel Pool 6.7.1.1 Impact Analyses 6-29 6.7.1.2 Weld stresses 6-30 6.8 Results from Whole Pool Multi-Rack 6-31 (WPMR) Analyses 6.9 Bearing Pad Analysis 6-33 6.10 Fatigue Considerations 6-35 6.10.1 Introductory Remarks 6-35 6.10.2 Methodology and Assumptions 6-35 6.10.2.1 Governing Codes and Analysis 6-35 Procedure 6.10.2.2 Model 6-36 6.11 Conclusion 6-37 6.12 Re:t;erences 6...;39 7.0 ACCIDENT ANALYSIS AND MECHANICAL INTEGRITY CONSIDERATIONS 7.1 Introductiori 7-1 7.2 Refueling Accidents 7-1 7.2.1 Dropped Fuel Assembly 7-1 Accident (Deep Drop Scenario) 7.2.2 Dropped Fuel Assembly 7-2 Accident (Shallow Drop Scenario) 7.3 Conclusion 7-2 7.4 References 7-3 8.0 FUEL POOL STRUCTURAL INTEGRITY CONSIDERATIONS 8.1 Introduction 8-1 8.2 Description of Spent Fuel 8-3 Pool Structure 8.3 Definition of Loads 8-3 8.3.1 Static Loading 8-3 8.3.2 Dynamic Loading 8-4 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 8-5 8.4.2.1 Application of Loads 8-5 8.4.2.2 Concrete Cracking 8-6 iv

8.4.3 Load Combinations 8-6 8.5 Results of Analyses 8-7 8.6 Pool Liner 8-8 8.7 Bearing Pads 8-9 8.8 Conclusions 8-9 8.9 References 8-10 9.0 RADIOLOGICAL EVALUATION 9.1 Fuel Handling Accident 9-1 9.1.1 Assumptions and Source 9-1 Term Calculations 9.1.2 Results 9-3 9.2 Solid Radwaste 9-4 9.3 Gaseous Releases 9-4 9.4 Personnel Exposures 9-4 9.5 Anticipated Exposure During Reracking 9-6 9.6 References 9-7 10.0 BORAL SURVEILLANCE PROGRAM 10.1 Purpose 10-1 10.2 Coupon Surveillance Program 10-2 10.2.1 Coupon Description 10-2 10.2.2 Surveillance Coupon Testing 10-3 Schedule 10.2.3 Measurement Program 10-4 10.2.4 Surveillance Coupon Acceptance 10-5 Criteria 10.3 References 10-6 11.0 ENVIRONMENTAL COST-BENEFIT ASSESSMENT 11.1 Introduction 11-1 11.2 Need for Increased On-Site Storage 11-1 Capacity 11.3 Appraisal of Alternate Options 11-2 11.4 Economic Analysis of Viable Options 11-10 11.5 Decision to.Rerack Salem Pools 11-12 11.6 Resource Commitment 11-12 11.7 Environmental Considerations 11-13 11.8 References 11-13 v

LIST OF TABLES

• 1.1 1.2 Pertinent Plant Data Previous and Projected Fuel Discharge Schedule Salem Unit 1 1.3 Previous and Projected Fuel Discharge Schedule Salem Unit 2 2.1.1 Module Data for Salem Maximum Density Racks 2.2.1 Total Cell Data 2.2.2 Common Module Data for Holtec Racks 2.3.1 Baral Experience List (Domestic and International) 2.3.2 1100 Alloy Aluminum Physical Properties 2.3.3 Chemical Composition - Aluminum (1100 Alloy) 2.3.4 Boron carbide Chemical Composition, Weight %

Boron Carbide Physical Properties 2.5.1 Heavy Load Handling Compliance Matrix (NUREG-0612) 4.2.1 Summary of Criticality Safety Analyses - Region 1 4.2.2 Summary of Criticality Safety Analyses - Region 2 4.2.3 Reactivity Effects of Abnormal and Accident Conditions 4.3.1 Design Basis Fuel Assembly Specifications 4.4.1 Limiting Burnup Values for Various Initial Enrichments -

Region 1 4.4.2 Limiting Burnup Values for Various Initial Enrichments -

Region 2 4.5.1 Effect of Temperature and Void on Calculated Reactivity of Storage Rack 5.1.1 Partial Listing of Rerack Applications Using Similar Methods of Thermal-Hydraulic Analysis 5.3.l Fuel Discharge Scenario: Salem 1 vi

Fuel Discharge Scenario: Salem 2 5.3.2 5.4.1 Data for Discharge Scenarios 5.4.2 Fuel Specific Power and Pool Capacity Data 5.6.1 Data for Local Temperature Evaluation 5.7.1 Peaking Factors 5.8.1 SFP Bulk Pool Temperature 5.8.2 Results of Loss-of-Cooling 5.8.3 Maximum Local Pool Water and Fuel Cladding Temperature for the Limiting Case (Case 2) 6.1.1 Partial Listing of Fuel Rack Applications Using DYNARACK 6.3.1 Cross-Correlation Coefficients in Each Set and Between the Sister Components in the Four Sets of Time-Histories Generated From the Broadened Design Basis Earthquake (DBE) Response Spectra 6.3.2 Cross-Correlation Coefficients Between Components of Time-Histories Generated from the Average Response Spectra of Four Sets of Time-Histories for Design Basis Earthquake (DBE) 6.3.3 Cross-Correlation Coefficients in Each Set and Between the Sister Components in the Four Sets of Time-Histories Generated From the Broadened Operating Basis Earthquake (OBE) Response Spectra 6.3.4 Cross-Correlation Coefficients Between Components of Time-Histories Generated from the Average Response Spectra of Four Sets of Time-Histories for Operating Basis Earthquake (OBE) 6.4.1 Degrees-of-Freedom 6.4.2 Numbering System for Gap Elements and Friction Elements 6.4.3 Rack Details 6.5.1 Rack Material Data (200°F); Support Material Data (200°F) 6.7.1 Results of Single Rack Analyses - List of All Runs vii

6.7.2 Summary of Worst Results from 16 Runs of Single Rack Analysis For Holtec Racks in Salem Pool (Loaded with 1700# Regular Fuel Assemblies; Controlling Seismic: DBE x 1.1) 6.7.3 Summary of Worst Results from 12 Runs of Single Rack Analysis For Holtec Racks in Salem Pool (Loaded with 1700# Regular Fuel Assemblies; Controlling Seismic: OBE x 1.1) 6.7.4 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-Al; Holtec Run I.D. draldbeo.rfS 6.7.5 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-Al; Holtec Run I.D. draldbeo.rf2 6.7.6 summary Results of 3-D Single Rack Analysis for Rack Module: Rack-Al; Holtec Run I.D. draldbeo.rhS 6.7.7 Summary Results of 3-D Single, Rack Analysis for Rack Module: Rack-Al; Holtec Run I.D. draldbeo.rh2 6.7.8 summary Results of 3-D Single Rack Analysis for Rack Module: Rack-Al; Holtec Run I.D. draldbeo.res 6.7.9 summary Results of 3-D single Rack Analysis for Rack

• 6.7.10 6.7.11 Module: Rack-Al; Holtec Run I.D. draldbeo.re2 Summary Results of 3-D single Rack Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4dbeo.rf8 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4dbeo.rf2 6.7.12 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4dbeo.rh8 6.7.13 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4dbeo.rh2 6.7.14 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4dbeo.re8 6.7.15 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4dbeo.re2 6.7.16 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-A3; Holtec Run I.D. dra3dbeo.rf8 6.7.17 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-A3; Holtec Run I.D. dra3dbeo.rf2

• viii

6.7.18 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-A3; Holtec Run I.D. dra3dbei.rf8 6.7.19 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-A3; Holtec Run I.D. dra3dbei.rf2 6.7.20 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-A1; Holtec Run I.D. dra1obeo.rf8 6.7.21 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-A1; Holtec Run I.D. dra1obeo.rf2 6.7.22 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-A1; Holtec Run I.D. dralobeo.rh8 6.7.23 Summary Results of 3-D single Rack Analysis for Rack Module: Rack-A1; Holtec Run I.D. dra1obeo.rh2 6.7.24 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-Al; Holtec Run I.D. dralobeo.res 6.7.25 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-A1; Holtec Run I.D. dralobeo.re2 6.7.26 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4obeo.rf8 6.7.27 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4obeo.rf2 6.7.28 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4obeo.rh8 6.7.29 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4obeo.rh2 6.7.30 Summary Results of 3-D Single Rack Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4obeo.re8 6.7.31 Summary Results of 3-D Single Rack. Analysis for Rack Module: Rack-B4; Holtec Run I.D. drb4obeo.re2 6.7.32 Comparison of Calculated and Allowable Loads/Stresses at Impact Locations and at Welds for Regular Fuel Loadings 6.8.1 Maximum Displacements from Whole Pool Multi-Rack Runs 6.8.2 Maximum Impact Force of Each Gap Element 6.8.3 Maximum Pedestal stress Factors of All Racks in Pool ix

• 6.8.4 6.8.5 6.8.6 6.8.7 Results of Pool Wall Dynamic Pressures Static Load and Dynamic Load Adder for Each Pedestal Total Static Load and Dynamic Adder on the Whole Slab Maximum Displacements from Whole Pool Multi-Rack Runs 6.8.8 Maximum Impact Force of Each Gap Element 6.8.9 Maximum Pedestal Stress Factors of All Racks in Pool 6.8.10 Results of Pool Wall Dynamic Pressures 6.8.11 Static Load and Dynamic Load Adder for Each Pedestal 6.8.12 Total Static Load and Dynamic Adder on the Whole Slab 6.8.13 Maximum Displacements from Whole Pool Multi-Rack Run 6.8.14 Maximum Impact Force of Each Gap Element 6.8.15 Maximum Pedestal Stress Factors of All Racks in Pool 6.8.16 Results of Pool Wall Dynamic Pressures 6.8.17 Static Load and Dynamic Load Adder for Each Pedestal 6.8.18 Total Static Load and Dynamic Adder on the Whole Slab 6.8.19 Comparison of Limiting Results from Single Rack Analyses and WPMR Analyses 8.5.1 Calculation of Equivalent Slab Vertical Pressures 8.5.2 Bending Strength Evaluation 8.5.3 Shear Strength Evaluation 9.1 Results of Origen-2 Calculations for Radionuclides of Iodine, Krypton, and Xenon at 168-Hours Cooling Time; Activity Released in the Fuel-Handling Accident

9. 2 Properties of Radionuclides Included in Fuel Handling Accident Analyses Performed by Holtec

9. 3 Data and Assumptions for the. Evaluation of the Fuel Handling Accident

• x

of Radionuclides 9.4 Concentrations in Salem Spent Fuel Pool Water 9.5 Preliminary Estimate of Person-Rem Exposures during Rerackinq xi

• 1.1 2.1.1 Unit 1 LIST OF FIGURES Plan View of Pool Area Fuel Handling Bldg., Salem Pool Layout Existing Racks (1170 Cells)

Salem Units 1 and 2 2.1.2 Increased Capacity Pool Layout (1632 Cells)

Salem Units 1 and 2 2.5.1 Fuel Handling Building Crane 3.3.1 Seam Welding Precision Formed Channels 3.3.2 Composite Box Assembly 3.3.3 Elevation View of Region II Rack Module 3.3.4 Typical Array of Region II Cells (Non-Flux Trap Construction) 3.3.5 Adjustable Support.

4.2.1 Definition of Acceptable Burnup Domain in Region 1

• 4.2.2 4.3.1 4.3.2 Definition of Acceptance Burnup Domains in Region 2 for Three Alternative Storage Configurations Region 1 Fuel Storage Cell Cross-Section of Storage Cell for Salem Spent Fuel Storage Rack 5.2.1 Spent Fuel Cooling System Simplified Flow Diagram 5.4.1 Salem Units 1 and 2 Spent Fuel Pool Discharge Scenario Case 1 5.4.2 Salem Units 1 and 2 Spent Fuel Pool Discharge Scenario Case 2 5.4.3 Salem Units 1 and 2 Spent Fuel Pool Discharge Scenario Case 3 5.5.1 Spent Fuel Pool Cooling Model 5.6.1 Salem Spent Fuel Pool Units 1 and 2 - Idealization of Rack Assembly xii

5.6.2 Thermal Chimney Flow Model 5.6.3 Salem Spent Fuel Pool - Convection currents in the Pool 5.8.1(a} Heat Load Curves for Case 1 5.8.1(b} Bulk .Pool Temperature Profile for Case 1 5.8.2(a} Heat Load Curves for Case 2 5.8.2(b} Bulk Pool Temperature Profile for Case 2 5.8.3(a} Heat Load Curves for Case 3 5.8.3(b} Bulk Pool Temperature Profile for Case 3 6.1.1 Multi-Rack Analysis - Rack and Pedestal Numbers Shown 6.3.1 Horizontal Response Spectra at Floor Elevation 89. 5' Design Basis Earthquake (20% G}

6.3.2 Vertical Response Spectra at Floor Elevation 89. 5' Design Basis Earthquake (13.3% G}

• 6.3.3 Horizontal Response Spectra at Floor Elevation 89. 5' Operating Basis Earthquake (10% G}

6.3.4 Vertical Response Spectra at Floor Elevation 89.5' Operating Basis Earthquake (6.67% G}

6.3.5 Original Response Spectrum DBE; Horizontal; Floor Ele.

89.5'; 3% Damping and the Corresponding Broadened Response Spectrum 6.3.6 Original Response Spectrum DBE; Vertical; Floor Ele.

89.5'; 3% Damping and the Corresponding Broadened Response Spectrum 6.3.7 Original Response Spectrum OBE; Horizontal; Floor Ele.

89.5'; 1% Damping and the Corresponding Broadened Response Spectrum 6.3.8 Original Response Spectrum OBE; Vertical; Floor Ele.

89.5'; 1% Damping and the Corresponding Broadened Response Spectrum 6.3.9 Salem Generating Station Units 1 and 2, PSEG, Pool Slab, El. 89.5' (1) The Broadened Floor Response Spectrum (DBE, Horizontal) (2} The Average Response Spectrum of 4 Synthetical DBE Accel. Time-Histories (3} The Re-Generated Response Spectrum from time-history: A-T-DBE.H1 xiii

6.3.10 Salem Generating Station Units 1 and 2, PSEG, Pool Slab, El. 89.5 1 (1) The Broadened Floor Response Spectrum (DBE, Horizontal) (2) The Average Response Spectrum of 4 Synthetical DBE Accel. Time-Histories (3) The Re-Generated Response Spectrum from time-history: A-T-DBE.H2 6.3.11 Salem Generating Station Units 1 and 2, PSEG, Pool Slab, El. 89. 5' (1) The Broadened Floor Response Spectrum (DBE, Vertical (2) The Average Response Spectrum of 4 Synthetical DBE Accel. Time-Histories (3) The Re-Generated Response Spectrum from time-history: A-T-DBE. VT 6.3.12 Synthetical DBE Acceleration Time-History for Pool Slab (Final), a-t-dbe.hl; Duration: 20 sec.

6.3.13 Synthetical DBE Acceleration Time-History for Pool Slab (Final), a-t-dbe.h2; Duration: 20 sec.

6.3.14 Synthetical DBE Acceleration Time-History for Pool Slab (Final), a-t-dbe.vt; Duration: 20 sec.

6.3.15 Salem Generating Station Units 1 and 2, PSEG, Pool Slab, El. 89.5 1 (1) The Broadened Floor Response Spectrum (OBE, Horizontal Damping 1%) (2) The Average Response Spectrum of 4 Synthetical OBE Accel. Time-Histories (3) The Re-Generated Response Spectrum from time-history: A-T-OBE.Hl 6.3.16 Salem Generating Station Units 1 and 2, PSEG, Pool Slab, El. 89.5 1 (1) The Broadened Floor Response Spectrum (OBE, Horizontal Damping 1%) (2) The Average Response Spectrum of 4 Synthetical OBE Accel. Time-Histories (3)- The Re-Generated Response Spectrum from time-history: A-T-OBE.H2 6.3.17 Salem Generating station Units 1 and 2, PSEG, Pool Slab, El. 89.5 1 (1) The Broadened Floor Response Spectrum (OBE, Vertical Damping 1%) (2) The Average Response Spectrum of 4 Synthetical OBE Acee!. Time-Histories ( 3) The Re-Generated Response Spectrum from time-history: A-T-OBE. VT.

6.3.18 Synthetical OBE Acceleration Time-History for Pool Slab (Final), a-t-obe.hl; Duration: 20 sec.

6.3.19 Synthetical OBE Acceleration Time-History for Pool Slab (Final), a-t-obe.h2; Duration: 20 sec.

6.3.20 Synthetical OBE Acceleration Time-History for Pool Slab (Final), a-t-obe.vt; Duration: 20 sec.

6.4.1 Pictorial View of Rack Structure 6.4.2 Schematic Model for DYNARACK 6.4.3 Rack-to-Rack Impact Springs xiv

6.4.4 Fuel-to-Rack Impact Springs

• 6.4.5 Degrees-of-Freedom Modeling Rack Motion 6.4.6 Rack Degrees-of-Freedom for Y-Z Plane Bending 6.4.7 Rack Degrees-of-Freedom for x-z Plane Bending 6.4.8 2-D View of Rack Module 6.8.1 Gap Time History, Gap Between Rack-2 and North Wall; East Corner, Top 6.8.2 Gap Time History, Gap Between Rack-2 and Rack-5; East Corner, 'rop 6.8.3 Gap Time History, Gap Between Rack-5 and Rack-6; South Corner, Top 6.8.4 Gap Time History, Gap Between Rack-5 and Rack-8, East Corner, Top 6.8.5 Gap Time History, Gap Between Rack-8 and Rack-11; East Corner, Top 6.8.6 Gap Locations 6.10.1 Finite Element Model 16 Locations 8.1.1 Salem Unit 1 SFP 8.2.1 SGS SFP Finite Element Model, View from Southeast 8.2.2 SGS SFP Finite Element Model, View from Northwest 8.2.3 SGS SFP Finite Element Model, View from Southeast, Showing External Roof and Wall Structure xv

1.0 INTRODUCTION

• The Salem Nuclear Generating Station (SNGS or Salem) is a twin unit pressurized water nuclear power reactor installation operated by the Public Service Electric and Gas Company (PSE&G). The facility is jointly owned by the Atlantic City Electric Company, the DelMarva Power and Light Company, the Philadelphia Electric company, and the Public Service Gas and Electric Company. The plant is situated eight miles southwest of the city of Salem, New Jersey.

Salem Unit 1 received its construction permit from the NRC (formerly AEC) in September, 1968 and its full power operating license in April, 1977. Salem Unit 2 received its construction permit in September, 1968 and full power operating license in October, 1981. The two reactors went into commercial operation in 1977 (Unit 1) and 1980 (Unit 2), respectively. The Salem spent fuel storage system consists of a spent fuel pool approximately 444 inches long and 342 inches wide with a separate cask laydown area for each unit (Figure 1.1). Each of the two pools is presently

• licensed to contain 1170 spent fuel storage locations.

Other pertinent data on the two Salem units is abstracted in Table 1.1. Information on the previous and projected spent fuel discharges into the two Salem pools is summarized in Tables 1.2 and 1.3. Referring to Table 1.2, the Unit 1 pool will contain 716 fuel assemblies in the summer of 1994 when construction for reracking the Unit 1 pool is scheduled to begin. In other words, the pool will be close to 2/3rds full at the time of rack change-out, which is typical of PWR pools. The Unit 2 pool will have a smaller inventory of spent fuel (564 fuel assemblies) at the time of its reracking (ca. 1995).

since all systems and planned modifications for the two pools are identical, this license document provides the supporting analyses and safety considerations for reracking both the pools.

1-1

Accordingly, the purpose of this submittal is to request the authorization to rerack the two Salem pools and equip them with poisoned maximum density racks (MDRs or Holtec racks) providing a total capacity of 1632 cells in each pool.

Referring to Tables 1.2 and 1.3, it is evident that the Unit 1 pool will lose its operational full core reserve (OFCR)* at the end of Cycle 14 in March, 1998 and Unit 2 by March, 2002. The loss of OFCR deadlines will be extended to September, 2008 and September, 2012 for Unit 1 and Unit 2, respectively, when the pools are equipped with the Holtec racks. This data explains the imperative to commence the reracking of the Unit 1 pool by 1994, and the Unit 2 pool shortly thereafter.

The principal owner of Salem, the Public Service Electric and Gas Company, entered into a contract with Hol tee International of Cherry Hill, New Jersey, in July, 1992 to design, procure material, fabricate, deliver and install Holtec racks for both Salem units.

This licensing document has been prepared by PSE&G and its contractor, Hol tee International,.

After the proposed modification, twelve free-standing poisoned rack modules positioned in the spent fuel pool with a prescribed and geometrically controlled gap between them will contain a total of 1632 storage cells. Of these, three existing {lOxlO) modules of

• Operational Full Core Reserve (OFCR): As established by PSE&G Nuclear Department, Operational Full Core Reserve {OFCR) is defined to be approximately 300 storage locations to maintain operational flexibility. An OFCR is larger than a simple full core reserve (i.e., 193 storage locations) as it allows for storage locations required for a full core offload and additional locations that are typically required for storage of non-fuel bearing components, unusable cells and maneuverability during refueling.

1-2

flux trap type* containing a total of 300 cells will serve as Region 1 storage. All new modules scheduled to be added to the pool are of Region 2 type. The design and construction of the storage cells is described in Section 3 of this document.

The new Holtec racks are free-standing and self-supporting. The principal construction materials for the new racks are SA240-Type 304L stainless steel sheet and plate stock, and SA564-630 (precipitation hardened stainless steel) for the adjustable support spindles. The only non-stainless material utilized in the rack is the neutron absorber material which is a boron carbide and aluminum-composite sandwich available under the patented product name "Baral".

The new Holtec racks are designed and analyzed in accordance with Section III, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel (B&PV) Code. The material procurement, analysis, and fabrication of the rack modules conform to 10CFR 50 Appendix B requirements.

This Licensing Report documents the design and analyses performed to demonstrate that the new Hol tee 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", USNRC (1978) and 1979 Addendum thereto.

The safety assessment of the proposed rack modules involved demonstration of their criticality, thermal-hydraulic, seismic, structural, and radiological adequacy as well as all USNRC specified criteria.

• A flux trap construction implies that there is a water gap between adjacent storage cells such that the neutrons emanating from a fuel assembly are thermalized before reaching an adjacent fuel assembly.

1-3

The criticality safety adequacy requires that the neutron multiplication factor for the stored fuel array be bounded by the USNRC limit of 0.95 (OT Position Paper) under assumptions of 95%

probability and 95% confidence. The criticality safety analysis synopsized in Section 4 sets the requirements on the Boral panel length and the areal B-10 density for the new Holtec racks. The criticality evaluation of the existing racks is also performed to establish the permissible maximum initial enrichment with and without Integral Fuel Burnable Absorber (IFBA) credit.

Hydrothermal adequacy requires that fuel cladding will not fail due to excessive thermal stress, and that the steady state pool bulk temperature wil~ remain within the limits prescribed for the spent fuel pool to satisfy the pool structural strength constraints. The thermal-hydraulic analyses is described in Section 5.0.

Demonstration of structural adequacy primarily.involves analyses showing that the free-standing rack modules will not impact with each other in the cellular region or the pool walls under the postulated Design Basis Earthquake (DBE) and Operating Basis Earthquake (OBE) events, and that the primary stresses in the rack module structure will remain below the ASME B&PV Code (subsection NF) allowables. The structural qualification also includes analytical demonstration that the subcriticality of the stored fuel will be maintained under all postulated accident scenarios in the Salem Updated Final Safety Analysis Report (UFSAR)

• The structural consequences of these postulated accidents are evaluated and presented in Sections 6, 7 and 8 of this report.

The radiological analyses are documented* in Section 9 and the Boral surveillance Program is described in Section 10. A summary of the cost/benefit analyses (Section 11. 0) demonstrating reracking as the most cost effective approach to increase the on-site storage capacity of the Salem Nuclear Generating station is also included in this report.

1-4

A11 computer programs utilized to perform the analyses documented in this licensing report are identified in the appropriate sections. All computer codes are benchmarked and verified in accordance with Hol tee International' s Nuclear Quality Program. All have been utilized in numerous rerack applications in recent years.

The analyses presented herein clearly demonstrate that the rack module arrays possess wide margins of safety from all five vantage points: criticality, thermal-hydraulic, seismic, structural and radiological. The No-Significant-Hazard Considerations submitted to the Commission along with this Licensing Report is based on the descriptions and analyses synopsized in the subsequent sections of this report .

• 1-5

/

Table 1.1 PERTINENT PLANT DATA ITEM UNIT 1 UNIT 2 Docket Number 50-372 50-311 Rated reactor 3411 3411 operating power '

MW(t)

Core size (number 193 193 of assemblies)

Pool size 37 1 x 28.5 1 37 1 x 28.5' Present licensed 1170 1170 storage capacity '

(number of fuel assemblies)

Present fuel cycle 11 7 (ca. February, 1993)

• Table 1.2 PREVIOUS AND PROJECTED FUEL DISCHARGE SCHEDULE SALEM UNIT 1 NUMBER OF OPEN STORAGE LOCATIONS With Present After Number of Total Capacity Reracking cycle End Month Assemblies Inventory (1170 (1632 No. & Year Discharged in the Pool Cells) Cells) 1 4/79 40 40 1130 1592 2 9/80 64 104 1066 1528 3 1/82 56 160 1010 1472 4 10/82 52 212 958 1420 5 2/84 84 296 874 1336 6 3/86 84 380 790 1252 10/87 84 464 706 1168 3/89 76 540 630 1092 2/91 48 588 582 1044 10 4/92 68 656 514 976 11 10/93 60 716 454 916 12 3/95 68 784 386 848 16 3/01 68 1048 122 584 17 9/02 68 1116 54 516 18 3/04 68 1184 448 19 9/05 68 1252 380 20 3/07 68 1320 312 23 9/11 68 1524 108 24 3/13 68 1592 40

• Loss of Operational Full Core Reserve without Reracking will occur in March, 1998.

• • Loss of Operational Full Core Reserve with Reracking will occur in September, 2008.

f i

Table 1.3 PREVIOUS AND PROJECTED FUEL DISCHARGE SCHEDULE SALEM UNIT 2 NUMBER OF OPEN STORAGE LOCATIONS With Present After Number of Total Capacity Reracking Cycle End Month Assemblies Inventory (1170 (1632 No. & Year Discharged in the Pool Cells) Cells) 1 1/83 72 72 1098 1560 2 10/84 68 140 1030 1442 3 10/86 84 224 946 1408 4 8/88 84 308 862 1324 5 3/90 76 384 786 1248 6 11/91 52 436 734 1196 7 3/93 64 500 670 1132 8 9/94 64 564 606 1068 9 3/96 68 632 538 1000 10 9/97 64 696 474 936 1 3/99 68 764 406 868 9/00 64 828 342... 804 15 9 / os 68 1028 142 604 16 9/06 64 1092 78 540 17 9/08 68 1160 10 472 18 9 /09 64 1124 408 19 9/11 68 1292 340 22 9/1s 64 1488 144 23 9/17 68 1556 76 24 9/18 64 1620 12

• Loss of Operational Full Core Reserve without Reracking will occur in March, 2002.

Loss of Operational Full Core Reserve with Reracking will occur in September, 2012

• 2'-2" 3' -10" 3' -10" APPROX 8 APPROX 11

.:- ... .'.' .. ':.- ".* ...*::._.,_..._._-. ,.:-" .-...* -" .:- ~ ...

4'-0" "

<AT TOP) 7' -8" 2' -2" 17' -4" 6' - 7" 39'-0" NEW FUEL 2' -2" 16' -0" l l' -8" STORAGE VAULT SPENT FUEL 23'-6" TRANSFER <COVERED)

STORAGE POOL POOL 8" APPROX CURB I' -0" 9' - 7" 8" APPROX 7' -l" CONCRETE FLOOR ( ALL AROUND ) APPROX .....

ELEV. 130' -0" -- .

~~

~

2' -6"

... 30' -0" -. " 2' -6" I. 47' -0" APPROX.

3' -0" 3' -0" FIGURE 101 PLAN VIEW OF POOL AREA FUEL HANDLING BLDGo. SALEM UNIT 1

2.0 MODULE LAYOUT FOR INCREASED STORAGE 2.1 Background The fuel pools in the Salem Nuclear Generating Station (SNGS or Salem) are of rectangular planform (approximately 444 inches x 342 inches) section. The nominal elevation of the pool liner is 89 1 -6 11

• The existing arrangement of the rack modules in the Salem Unit 1 pool is illustrated in Figure 2.1.1. The layout in Unit 2 pool is a mirror image of Unit 1. Figure 2.1.2 shows the proposed layout with increased storage capacity. This layout features a total of 1632 storage cells in twelve free-standing rack modules. The new densified storage in the Salem pools calls for no major physical alterations within the pools. The module layouts* presented in Figures 2.1.1 and 2.1.2 use as-built pool dimensions and account for the deviations in the pool opening dimensions as established.by the 3-D pool cavity surveys of both Units 1 and 2 pools.

Table 2.1.1 provides key module data for both Region I and Region

• II racks. Table 2.2.2 contains key design data for the new rack modules. It is noted from Table 2.1.1 that the cross-section of the new rack modules for the Salem pools have been proportioned to have aspect ratios (length to width ratio) in the range of 1 to 1.5 range, resulting in kinematically stable free-standing structures.

2.2 Multi-Region Storage The maximum density racks (MDRs or Holtec racks) in conjunction with the existing (Exxon) modules in each of the two Salem pools will provide up to 1632 storage locations and maintain the stored fuel, having an initial enrichment of up* to 5.0 wt% U-235 (by taking credit for integral burnable poison rods), in a safe, 2-1

coolable, and subcritical configuration during normal, abnormal and postulated conditions under the classical multi-region storage arrangement.

All rack modules for the Salem pools will be of the "free-standing" type such that .the modules do not attach to the pool floor nor do they require any lateral braces or restraints. These rack modules will be placed in the pool in their designated locations, and the support legs will be remotely leveled using a telescopic removable handling tool. The design of the modules enables remote levelling and alignment without the use of divers.

The existing racks in the Salem pools, three of which will be retained in each pool, were equipped with lateral restraints with prescribed gaps. The seismic analyses of the racks performed in support of the previous rerack (amendment number 33 for Unit 1, February 1981) of the Salem pool indicated that the module displacements under the postulated seismic events were less than the clearances provided, demonstrating that the lateral restraints were unnecessary. These redundant lateral restraints will be removed from the Salem pools during the reracking effort.

The racks will be arranged in two regions in the pool. The retained Region I Exxon racks, having 300 storage iocations in each pool, will be located adjacent to the fuel transfer canal and be capable of storing unirradiated fuel of up to 4. 25 wt% U-235 initial enrichment without credit for IBFAs. With credit for only 64 IBFAs the allowable initial enrichment increases to 5%.

Additional details are provided in Section 4. Table 2.2.1 provides cell data for Region I and Region II racks. The Region II racks will have 1332 storage locations

. and are designed to store spent fuel of various initial enrichments and corresponding minimum burnups. The enrichment and burnup criteria for Region II fuel 2-2

storage has been developed as part of the rack design. Certain locations in Region II can also be used to store fresh fuel as long as a sufficient number of fuel assemblies with a prescribed minimum burnup surround it.

Each new rack module is supported by a minimum of four legs which are remotely adjustable. Thus, the racks can be made vertical and the top of the racks can easily be made co-planar with each other.

The rack module support legs are engineered to accommodate variations in the pool floor flatness. The placement of the racks in the Salem pools has been designed to preclude any support legs from being located over the liner weld seams and existing obstructions on the pool floor. However, the proposed rack layout results in one of the support legs of the retained Exxon racks to be positioned on top of the sump cover in each pool. This condition has been analyzed and requires the replacement of the existing sump covers with new covers in both the pools.

2.3 Material Considerations 2.3.1 Introduction Safe storage of nuclear fuel in the Salem pools requires that t~e materials utilized in the rack fabrication be of proven durability and be compatible with the pool water environment. This section provides the necessary information on this subject.

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

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

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

2-3

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.

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

Boral is a thermal neutron poison material composed of boron carbide and 1100 all,oy 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 lightweight 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 compatlble and ideally suited for long-term use in the radiation, thermal and chemical environment of a nuclear reactor or a spent fuel pool.

Boral has been exclusively used in fuel rack applications in recent years. Its use in the spent fuel pools as the neutron absorbing material can be attributed to its proven performance (over 150 pool years of experience) and the following unique characteristics:

(i) The content and placement of boron carbide provides a very high removal cross-section for thermal neutrons.

(ii) Boron carbide, in the form of fine particles, is homogeneously dispersed throughout the central layer of the Boral panels .

2-4

(iii) The boron carbide and aluminum materials in Baral do not degrade as a result of long-term exposure to radiation.

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

(v) Baral is stable, strong, durable, and corrosion resistant.

Holtec International's Q.A. program ensures that Baral is manufactured by AAR Brooks & Perkins under the control and surveillance of 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 2.3.1, Baral has been licensed by the USNRC for use in numerous BWR and PWR spent fuel storage racks and has been extensively used in international nuclear installations.

Baral 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 has atomic number of 13, atomic weight of 26.98, specific gravity of 2.69 and valence of 3. The physical, mechanical and chemical properties of the 1100 alloy aluminum are listed in Tables 2.3.2 and 2.3.3.

The excellent corrosion resistance of the 1100 alloy aluminum is provided by the protective oxide film that develops on its surface from exposure to the atmosphere or water. This film prevents the loss of metal from general corrosion or pitting corrosion.

2-5

Boron Carbide: The boron carbide contained in Baral 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 conf arms to the chemical composition and properties listed in Table 2.3.4.

2.3.4 Compatibility with Coolant All materials used in the construction of the Holtec 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 2.3.1, Baral has been successfully used in both vented and unvented configurations in fuel pools. Austenitic stainless steel (304L) is perhaps the most widely used stainless alloy in nuclear power plants.

2.4 Retained Exxon Rack Modules and Proposed Reracking Operation The reracking of the Salem Unit 1 pool will be carried out in the latter half of 1994. At the time of the reracking operation, there will be 716 spent fuel assemblies in the Unit 1 pool, leaving only 454 open locations. For Unit 2, the reracking will be carried out in the July, 1995 - December,1995 time frame. At the time of the reracking operation, there will be 564 spent fuel assemblies in the Unit 2 pool, leaving only 606 open locations. A rack change-out and fuel shuffle scheme is being developed which eliminates the potential of any damage to the stored fuel during the handling operations associated with the rerack work effort. All activities in the rerack construction effort are intended to comport with the provisions of NUREG-0612. The details of the "defense-in-depth" approach to eliminate the potential for construction accidents are provided in the next subsection .

• 2-6 L

It is PSE&G's intention to perform all field operations remotely

• insofar as it is consistent with the ALARA objectives. In other words, divers will be utilized to carry out an in-pool operation only if such an approach demonstrably minimizes the total radiation exposure to the personnel. All diving operations will comply with Draft Regulatory Guide DG-8006, "Control of Access to High and Very High Radiation Areas in Nuclear Power Plants".

2.5 Heavv Load Considerations for the Proposed Reracking Operation There are two cranes for handling spent fuel racks. The fuel handling crane is currently rated for five tons, and its principal direction of movement is north-south in the spent fuel pool area.

The cask crane in the fuel pool area is currently rated for 110 tons and its principal direction of movement is east-west. This cran~ can be used to handle racks from the ground level to the fuel pool area, but it cannot access the spent fuel pool pit due to its movement constraints. The maximum weight for new racks is 21,100 lbs and that for the old racks is 33,800 lbs. As a result, the fuel handling crane which covers the spent fuel pool area cannot be used for handling racks unless it is upgraded and qualified for a higher capacity.

Considerations of available opening, structural design, heavy load handling, ALARA and convenience in the reracking operation suggest upgrading and qualifying the existing fuel handling crane to handle the new and existing modules in and out of the spent fuel pool.

This crane was previously upgraded during the earlier rerack and equipped with a 20-ton hoist to handle the existing racks. Since NUREG-0612 was not in effect during the period of previous reracking, supporting calculations to demonstrate compliance with the provisions of NUREG-0612 were not made at that time. This condition is being corrected in the proposed rerack effort. A 2-7

summary of the steps utilized to upgrade the fuel handling crane for a design basis load of 20 tons (which exceeds the weight of the heaviest rack and the lift gear) is presented in the following section.

Figure 2. 5. 1 illustrates the schematic appearance of the fuel handling crane. The L-shape construction* of the crane maximizes its span by utilizing the protrusion (corbel) on one pool wall as the load bearing surface. Upgrading this crane calls for installing a lifting system consisting of four hoists in parallel, which together have a rated capacity of over 20 tons. All four hoists will be operated from a single electronic signal device.

The implementation of NUREG-0554 in accordance with Appendix c to NUREG-0612 is realized as follows:

(i) NUREG-0554 recommends that the crane be designed to the Maximum Critical Load. stress analyses using the finite element Code ANSYS are carried out to demonstrate that an amplified load of 100 tons (five times the design* load) produces primary stresses which are less than 90% of the ultimate strength of the structural components of the crane.

Additionally, a postulated load of 30 tons (1.5 times the design load) is shown to produce primary stresses which are less than the material yield strength of the crane structural components.

(ii) The minimum ambient temperature in the fuel pool region will be kept above 70°F during the load handling operations, thus obviating the need for a cold proof test.

(iii) Since it is not possible to establish whether preheat and post-weld heat treatment temperatures were specified in the original welding operations during manufacturing of the crane, the alternative approach of non-destructive examination (visual examination) o*f crane structural welds will be followed.

2-8

(iv) Operability of the crane travel limit controllers and interlocks will be checked and verified before beginning the reracking operation.

(v) The crane is analyzed to ensure that it is capable of stopping and holding the load during a seismic event equal to a Safe Shutdown Earthquake applicable to the Salem site. This analysis purports to show that the primary stresses in crane structural members will remain below the Code allowable (Subsection NF) for level B condition.

Furthermore, the brake drums in the hoists are rated to stop and hold a vertical load of 20 (1 +

a) tons where a is the vertical SSE acceleration multiplier.

(vi) While not required by NUREG-0612, analyses were also performed to demonstrate that the design strength of the corbel is not exceeded if the factored dead and live load combination per ACI318-71 is applied on the crane. The procedures of ACI318 supplemented by a finite element analysis were used to establish the adequacy of the corbel.

(vii) The compliance of the crane to the provisions of CMAA-70 for a design load of 20 tons is also analytically confirmed.

(viii) The assembled crane/hoist system will be load tested with 25 tons dead load prior to use.

A remotely engagable lift rig, meeting NUREG-0612 stress criteria, will be used to lift the empty modules (old and new). The rig designed for handling the Salem racks is identical in its physical attributes to the rigs utilized to rerack Millstone Unit One (1988), Vogtle Unit Two (1989), Indian Point Unit Two (1990),

Ulchin Unit Two (1990), Hope Creek (1990), Laguna Verde Unit One (1990), Kuosheng (1991), Three Mile Island Unit 1 (1992), Zion (1993), and J.A. FitzPatrick (1992). The rig consists of independently loaded lift rods with a "cam type" lift configuration which ensures that failure of one traction rod will not result in 2-9

uncontrolled lowering of the load being carried by the rig (which complies with the duality feature called for in Section 5.1.6(1) of NUREG 0612). The rig has the following additional attributes:

a. The stresses in the lift rods are self limiting inasmuch as an increase in the magnitude of the load reduces the eccentricity between the upward force and downward reaction (moment arm).

b. The traction rod is designed to prevent loss of 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 14.6 {1978) are shown to be met.

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

Pursuant to the defense-in-depth approach of NUREG-0612, the following additional measures of safety will be undertaken for the reracking operation.

(i) The cranes used in the project will be given a preventive maintenance checkup and inspection per the Salem maintenance procedures before the beginning of the reracking operation.

(ii} The old fuel racks will be lifted no more than 6 inches above the pool floor and held at that elevation for approximately 10 minutes before maklng any further rack movements.

(iii) Safe load paths will be developed for moving the old and new racks in the Fuel Handling Building. The "old" or "new" racks will not be carried over any region of the pool containing fuel.

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

2-10

(v) All crew members involved in the use of the lifting and upending equipment will be given training.

In addition to the above design, testing, and operation measures, the consequences of a postulated rack drop were also considered on the integrity of the pool structure. The following analyses were performed.

a. The heaviest rack module (out of all existing and new racks) was postulated to free fall from the top of the water surface level to the pool floor.

b. The fall of a rack is assumed to occur in its normal vertical configuration which minimizes the retarding effect of water drag.

c. The falling rack is assumed to impact the pool slab undergoing an elastic/plastic impact.

d. The maximum impact load is compared with the gross seismic slab impact load during the DBE event (presented in Section 6 of this report) .

The results of the analyses show that the maximum additional load on the pool structure is less than the DBE seismic adder. The pool structure has been qualified for the latter, as reported in Section 8 of this report. Therefore, the integrity of the pool structure under the postulated rack drop event is ensured.

The fuel shuffle scheme for the spent fuel pool has been developed based on the following criteria:

(1) No heavy load (rack or rig) with a potential to drop on a rack has less than 3 feet lateral free zone clearance from active fuel.

(2) All heavy loads are lifted in such a manner that the C.G.

of the lift point is aligned with the c.G. of the load being lifted.

4lt 2-11

(3) Turnbuckles rather than slings are utilized to "fine tune" the verticality of the rack being lifted.

All phases of the reracking activity will be conducted in accordance with written procedures which will be reviewed and approved by Holtec International and PSE&G.

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

The NUREG-0612 guidelines cite four major causes of load handling accidents, namely (i) operator errors (ii) rigging failure (iii) lack of adequate inspection (iv) inadequate procedures The Salem rerack program ensures maximum emphasis on mitigating the potential load drop accidents by implementing measures to eliminate shortcomings in all aspects of the operation including the four aforementioned areas. A summary of the measures specifically planned to deal with the major causes is provided below.

Operator errors: As mentioned above, PSE&G plans to provide comprehensive training to the installation crew.

Rigging failure: The lifting device designed for handling and installation of the racks in the Salem pools has redundancies in the lift legs and lift eyes such that there are four independent load members. Failure of any one load bearing member would not lead to uncontrolled lowering of the load. The rig complies with all provisions of ANSI 14. 6 - 1978, including compliance with the primary stress criteria, load testing at 150% of maximum lift load, and dye examination of critical welds

• 2-12

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

Lack of adeauate inspection: The designer of the racks will develop a set of inspection points which have been proven to I eliminate any incidence of rework or erroneous installation in numerous prior rerack projects. I Inadequate procedures: PSE&G's contractor, Holtec International, plans to develop over twenty operating procedures to cover the entire gamut of operations pertaining to the rerack effort, including but not limited to, mobilization, rack handling, upending, lifting, installation, verticality, alignment, dummy gage testing, site safety, and ALARA compliance. Procedures for handling both the old racks and new racks will be developed.

The series of operating procedures planned for Salem rerack are the successors of the procedures successfully implemented in previous projects.

Table 2. 5. 1 provides a synopsis_, of the requirements delineated in NUREG-0612, and our intended compliance.

In summary, the measures implemented in Salem reracking are identical to the those utilized in the most recent successful projects such as Indian Point Unit 2, completed in October, 1990, Hope Creek, completed in March, 1992, and Three Mile Island, Unit 1 completed in September, 1992 *

• 2-13

Table 2.1.1 MODULE DATA FOR SALEM MAXIMUM DENSITY RACKS MOOULE ENVELOPE NUMBER OF SIZE Cinches) SHIPPING MODULE N-S E-W CELLS WEIGHT l.D. DIRECTION DIRECTION PER RACIC N-S E-W (lbs.) MODULE TYPE A1 12 13 156 108.84 117.91 21, 100 Region II A2 12 13 156 108.84 117 .91 21,100 Region II A3 12 13 156 108.84 117 .91 21, 100 Region II B1 12 12 144 108.84 108.84 19,500 Region II B2 12 12 144 108.84 108.84 19,500 Region II B3 12 12 144 108.84 108.84 19,500 Region II B4 12 12 144 108.84 108.84 19,500 Region II BS 12 12 144 108.84 108.84 19,500 Region II B6 12 12 144 108.84 108.84 19,500 Region II E1 10 10 100 .109.5 109.S 33,800 Region I E2 10 10 100 109.5 109.S 33,800 Region I E3 10 10 100 109.5 109.5 33,800 Region I

Table 2.2.1 TOTAL CELL DATA TOTAL NUMBER REGION NUMBER OF RACKS OF CELLS I (Flux-Trap) 3 300 II (Non-Flux Trap) 9 1332

• GRAND TOTAL: 12 1632

Table 2.2.2 COMMON MODULE DATA FOR HOLTEC RACKS Storage cell inside dimension (nominal): 8.86 inch Storage cell height 168-1/2 inch (above the baseplate):

Baseplate thickness: 0.75 inch Support leg height: 14-3/4 inch (nominal)

Support leg type: Remotely adjustable legs with lateral gussets.

Number of support legs: 4 (minimum)

Remote lifting and handling provision: Yes Poison material: Boral Poison length: 144 inch Poison width: 7.5 inch Cell Pitch: 9.052 +/- .04 inch

Table 2.3.1 BORAL EXPERIENCE LIST (Domestic and International)

Pressurized Water Reactors Vented Construc- Mfg.

Plant Utility t ion Year Bellefonte 1,2 Tennessee Valley Authority No 1981 Donald c. Cook Indiana & Michigan 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 Sequoyah (rerack) Tennessee Valley Authority Yes 1992 D.C. Cook (rerack) American Electric Power Yes 1992 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 Duane Arnold Iowa Elec. Light & Power Yes 1993 Pilgrim Boston Edison Company Yes 1993 LaSalle Unit 1 Commonwealth Edison Company Yes 1992

Table 2.3.1 (continued)

INTERNATIONAL INSTALLATIONS USING BORAL FRANCE 12 PWR Plants Electricite de France SOUTH AFRICA Koeberg 1,2. ES COM SWITZERLAND Beznau 1, 2 Nordostschweizerische Kraftwerke AG Gosgen Kernkraftwerk Gosgen-Daniken AG TAIWAN Chin-Shan 1,2 Taiwan Power Company Kuosheng 1,2 Taiwan Power Company MEXICO Laguna Verde Comision Federal de Electricidad Units 1 & 2

Table 2.3.2 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 deg. F) 0.53 cal/sec/sq cm/deg. C/cm Coef. of Thermal 13.1 x 10- 6 in/in., °F Expansion 23. 6 x 10- 6 cm/cm, 0 c (68-212 deg. F)

Specific heat 0.22 BTU/lb/deg. F (221 deg. F) 0.23 cal/gm/deg. C Modulus of 10xl06 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 2.3.3 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 2.3.4 BORON CARBIDE CHEMICAL COMPOSITION. WEIGHT %

Total boron 70.0 min.

B10 isotopic content in 18.0 natural boron Boric oxide 3.0 max.

Iron 2.0 max.

Total boron plus 94.0 min.

total carbon BORON CARBIDE PHYSICAL PROPERTIES Chemical formula Boron content (weight) 78.28%

Carbon content (weight) 21.72%

Crystal Structure rombohedral Density 2.51 gm./cc-0.0907 lb/cu. in.

Melting Point 2450°C - 4442°F Boiling Point 3500°C-6332°F Microscopic Capture 600 barn cross-section

• Table 2.5.1 HEAVY LOAD HANDLING COMPLIANCE MATRIX (NUREG-0612)

Criterion Compliance

1. Are safe load paths defined for Yes the movement of heavy loads to minimize the potential of impact, if dropped on irradiated fuel?

2. Will procedures be developed to Yes 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 guidelines of ANSI 14.6-1978?

s. Will non-custom lifting devices Yes be installed and used in accordance with ANSI B30.9-1971?

6. Will the cranes be inspected and Yes tested prior to use in rerack?

7. Does the crane meet the intent of Yes ANSI BJ0.2-1976 and CMMA-70?

N

<J======i UNIT 1 (UNIT 2 MIRROR IMAGE) 444"

( 37' -0" )

10 x 10 10 x 10 10 x 10 9 x 10 A-10 A-7 A A-1 10 x 10 10 x 10 10 x 10 9 x 10 342"

( 28' -6" )

A-11 A-8 A-5 A-2 10 x 10 10 x 10 10 x 10 9 x 10 A-12 A-9 A-6 A-3 FIGURE 20101 POOL LAYOUT EXISTING RACKS Q 1170 CELLS)

SALEM UNKTS 1 AND 2 I*

I

N

<J=

UNIT 1 (UNIT 2 IS MIRROR IMAGE)

,___ ________ ~

444.00 - - - - - - - - - - -

NDM 0 0 0 0 0 342' 00 NDM 0

0 0 0 0 0 0 FIGURE 2.1.2 INCREASED CAPACITY POOL LAYOUT ( 1632 CELLS)

SALEM UNITS 1 AND 2

2 ' -IO l/B'_'- - - - - - - - - .- - - ~LOW PDI~T-DF ROOF - - -----i R TD R TOP OF RAIL TOTAL 7'-0" 3, -O" Li I jcrnc1 I I LOW POINT OF WALK GIRDER TRUCK 3' -6" APPROX VERTICAL CORBEL LEG 27' -1 l/B" RAJL TD RAIL 911 APPROX I WALL I

I 3'-5" ~ TRUCK l - 1 1 1 -4" ws--l

- - - - - - - - 33' -3" SPAN--------j 4'-10 112" GIRD.

5'-0" END CONN.

<A) VIEW OF UNIT I CRANE LOOKING SOUTH (8) ENO VIEW FIGURE 2.5.n FUEL HANDLING BUILDING CRANE

3.0 RACK FABRICATION ANO APPLICABLE CODES

• The object of this section is to provide a description of rack module construction for the Salem pool to enable an independent appraisal of the adequacy of design. A list of applicable codes and standards is also presented.

3.1 Fabrication Objective The requirements in manufacturing the maximum density racks (MDRs or Holtec racks) for the Salem pool may be stated in four interrelated points:

( 1) The rack module is fabricated in such a manner that there is Il.Q weld splatter on the storage cell surfaces which would come in contact with the fuel assembly.

( 2) The storage locations are constructed so that redundant flow paths for the coolant are available.

(3) The fabrication process involves operational sequences which permit immediate verification by the inspection staff.

(4) The storage cells are connected to each other by austenitic stainless steel corner welds which leads to a honeycomb lattice construction. The extent of welding is selected to "detune" the racks from the postulated seismic input motion.

3.2 Rack Module for Region I (Retained Exxon Racks)

PSE&G plans to retain three existing rack modules which are of poisoned flux trap construction to serve as Region I of the densified storage. Each rack module retained in the pool for this purpose consists of one hundred storage cells and has cross-sectional dimensions of 109. 5 11 x 109. 5 11

• The locations of the existing rack modules are indicated in Figure 2.1.2.

3-1

3.3 Rack Module for Region II A total of nine Region II storage rack modules of maximum density genre will be installed in each of the two Salem pools. Region II storage cell locations have a single poison panel between adjacent austenitic stainless steel surfaces. The significant components (discussed below) of the Region II racks are: (1) the storage box subassembly, (2) picture frame sheathing, (3) the neutron absorber material, (4) the baseplate, and (5) support legs.

(1) Storage cell box subassembly: The rack module manufacturing begins with fabrication of the box. The "boxes" are fabricated from two precision formed channels by continuous seam welding in a machine equipped with copper chill bars and pneumatic clamps to minimize distortion due to welding heat input. Figure 3. 3 .1 shows the "box".

The minimum weld penetration will be 80% of the box metal gage which is 0.075 11 (14 gage). The boxes are manufactured to 8. 86-inch inside dimension (nominal).

Each box constitutes a storage location.

(2) Picture Frame Sheathing The design objective calls for installing Boral with minimal surf ace loading. This is accomplished by die forming a "picture frame sheathing" as shown in Figure

3. 3. 2. This sheathing is made to precise dimensions such that the offset is .010 to .005 inches greater than the poison material thickness.

As shown in Figure 3.3.3, the(pois~n material is placed in the customized flat depression region of the sheathing, which is next laid on a side of the "box".

The* precision of the shape of the sheathing obtained by die forming guarantees that the poison sheet installed in it will not be subject to surface compression. The flanges of the sheathing {on all four sides) are attached to the box using skip welds. The sheathing serves to locate and position the poison sheet accurately, and to preclude its movement under seismic conditions.

The edges of the sheathing and. the box are welded together to form a smooth edge. The box, with integrally connected sheathing, is referred to as the "composite box". Each box has two lateral holes punched near its bottom edge to provide auxiliary flow holes.

3-2

(

The "composite boxes" are arranged in a checkerboard array to form an assemblage of storage cell locations (Figure 3.3.4). Flat plates are welded to the edges of the boxes at the outside boundary of the rack to make the periphery cells. The inter-box welding and pitch adjustment are accomplished by small longitudinal connectors.

This assemblage of box assemblies is welded edge-to-edge as shown in Figure 3. 3. 4, resulting in a honeycomb structure with axial, flexural and torsional rigidity depending on the extent of intercell welding provided.

It can be seen from Figure 3.3.4 that two edges of each interior box are connected to the contiguous boxes resulting in a well defined. path for .. "shear flow".

( 3) The neutron absorber material: Boral is used as the neutron absorber material.

(4) Baseplate: The baseplate provides a continuous horizontal surface for supporting the fuel assemblies.

The baseplate is attached to the cell assemblage by fillet welds. The baseplate in each storage cell has a 5 11 diameter flow hole.

(5) Support Legs: All support legs are the adjustable type (Figure 3.3.5). The top portion is made of austenitic stainless steel material. The bottom part is made of 17: 4 Ph series stainless steel to avoid galling problems.

Each support leg is equ:j.pped with a readily accessible socket to enable remote leveling of the rack after its placement in the pool. Lateral holes in the support leg provide.the requisite coolant flow path.

The height of the support legs in the new Hol tee racks is engineered to ensure that the top of the racks, as well as the top of fuel assemblies, will be coplanar in the Exxon and Holtec racks.

3.4 Codes, Standards. and Practices for the Salem Nuclear Generating Station Spent Fuel Pool Racks The fabrication of the rack modules for Salem will be performed under a strict quality assurance system suitable for manufacturing and complying with the provisions of 10CFR50 Appendix B *

• 3-3

The fallowing codes, standards and practices are used as applicable for the design, construction, and assembly of the spent fuel storage racks. Additional specific references related to detailed analyses are given in each section.

a. Design Codes (1) AISC Manual of Steel Construction, 1964 Edition and later.

(2) ANSI N210-1976, "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, 1986 Edition . (governing material procurement, fabrication and NDE); ASME Section V, 1989 edition.

(4) ASNT-TC-1A June, 1980 American Society for Nondestructive Testing (Recommended Practice for Personnel Qualifications}.

(5) American Concrete Institute Building Code Requirements for Reinforced Concrete (ACI318-63) and (ACI318-71}.

(6} Code Requirements for Nuclea~ Safety Related Concrete Structures, ACI349-85/ACI349R-85, and ACI349.1R-80.

(7) ASME NQA-2-1989, Quality Assurance Requirements for Nuclear Facility Applications.

(8) ACI Detailing Manual - 1980.

b. Material Codes - Standards of ASTM (1) E165 Standard Methods for Liquid Penetrant Inspection.

(2) A240 - Standard Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless steel Plate, Sheet and Strip for Fusion-Welded Unfired Pressure Vessels.

( 3} A2 62 - Detecting Susceptibility to Intergranular Attack in Austenitic Stainless Steel.

3-4

( 4) A276 - Standard Specification for Stainless and

• (5)

(6)

Heat-Resisting Steel Bars and Shapes.

A479 - Steel Bars for Boilers & Pressure Vessels.

C750 Standard Specification for Boron Carbide Powder.

Nuclear-Grade (7) A380 - Recommended Practice for Descaling,. Cleaning and Marking Stainless Steel Parts and Equipment.

(8) C992 Standard Specification for Boron-Based Neutron Absorbing Material Systems for Use in Nuclear Spent Fuel Storage Racks.

(9) American Society of Mechanical Engineers (ASME),

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

(10) NCA3800 Metallic Material Manufacturer's and Material Supplier's Quality System Program.

c. Welding Codes ASME Boiler and Pressure Vessel Code,Section IX -

Welding and Brazing Qualifications, 1986 Edition.

d. Quality Assurance, Cleanliness, Packaging, Shipping, Receiving, Storage, and Handling Requirements (1) ANSI 45.2.1 Cleaning of Fluid Systems and Associated Components during Construction Phase of Nuclear Power Plants.

(2) ANSI N45.2.2 Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants (During the Construction Phase).

(3) ANSI - N45. 2. 6 - Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants (Regulatory Guide 1.58).

(4) ANSI-N45.2.8, Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and systems for the Construction Phase of Nuclear Plants.

(5) ANSI - N45.2.ll, Quality Assurance Requirements for the Design of Nuclear Power Plants.

(6) ANSI-N45.2.12, Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants.

3-5

(7) ANSI N45.2.13 - Quality Assurance Requirements for Control of Procurement of Equipment Materials and Services for Nuclear Power Plants (Regulatory Guide 1.123) .

(8) ANSI N45.2.15-18 Hoisting, Rigging, and Transporting of Items For Nuclear Power Plants.

(9) ANSI N45.2.23 - Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants (Regulatory Guide 1.146).

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

(11) ANSI - N16.9-75 Validation of Calculation Methods for Nuclear Criticality Safety.

(12) ASME Boiler and Pressure Vessel Code NCA3550 Requirements for Design Documents, 1986 Edition.

(13) ASME Boiler and Pressure Vessel Code NCA4000 -

Quality Assurance, 1986 Edition.

e. Governing NRC Design Documents (1) NUREG-0800, Radiological Consequences of Fuel Handling Accidents.

(2) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978, and the modifications to this document of January 18, 1979.

(3) NUREG 0612, "Control of Heavy Loads at Nuclear Power Plants", USNRC, Washington, D.C., July, 1980.

f. Other ANSI Standards (not listed in the preceding)

(1) N16 .1 - Nuclear criticality Safety in Operations with Fissionable Materials Outside Reactors (2) N45.2 - Quality Assurance Program Requirements for Nuclear Facilities - 1971 (3) N45.2.9 - Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants - 1974 (4) N45.2.10 - Quality Assurance Terms and Definitions

-1973

• 3-6

(5) N210 - Design Objective for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants

( 6) N14. 6 - American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500 kg) or more for Nuclear Materials

g. Code-of-Federal Regulations (1) 10CFR20 Standards for Protection Against Radiation (2) 10CFR21 - Reporting.of.Defects and Non-compliance (3) 10CFRSO - Appendix A - General Design Criteria for Nuclear Power Plants (4) 10CFRSO - Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants (5) 10CFR61 - Licensing Requirements for Land Disposal of Radioactive Material (6) 10CFR71 Packaging and Transportation of Radioactive Material

h. Regulatory Guides (1) RG 1.13 - Spent Fuel Storage Facility Design Basis (Revision 2 Proposed)

(2) RG 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 - (ANSI N45.2) - Quality Assurance Program Requirements (4) RG 1.29 - Seismic Design Classification (Rev. 3)

(5) RG 1.31 - Control of Ferrite Content in Stainless Steel Weld Material (5) RG 1.38 (ANSI N45.2.2) Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants 3-7

_I

(6) -RG 1.44 Control of the Use of Sensitized Stainless Steel (7) RG 1.58 - (ANSI N45.2.6) Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel (8) RG 1. 61 - Damping Values for Seismic Design of Nuclear Power Plants, Rev. o, 1973 (9) RG 1.64 (ANSI N45.2.11) Quality Assurance Requirements for the Design of Nuclear Power Plants (10) RG 1.71 Welder Qualifications for Areas of Limited Accessibility .

(11) RG 1.74 - (ANSI N45.2.10) Quality Assurance Terms and Definitions (12) RG 1.85 - Materials Code Case Acceptability - ASME Section 3, Div. 1 (13) RG 1.88 - (ANSI N45.2.9) Collection, Storage and Maintenance of Nuclear Power Plant Quality Assurance Records

( 14) RG 1. 92 * - Combining Modal Responses and Spatial Components in Seismic Response Analysis.

(15) RG 1.122 - Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or C~mponents (16) RG 1.123 (ANSI N45.2.13) Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants (17) RG 1.124 - Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports, Revision 1, 1978 (18) RG 3.41 - Validation of Calculational Methods for Nuclear Criticality Safety,.Revision 1, 1977 (19) DG-8006, "Control of Access to High and Very High Radiation Areas in Nuclear Power Plants"

i. Branch Technical Position (1) CPB 9.1 Criticality in Fuel Storage Facilities

( 2) ASB 9-2 - Residual Decay Energy for Light-Water Reactors for Long-Term Cooling 3-8

j . Standard Review Plan (1) SRP 3.2.1 - Seismic Classification (2) SRP 3.2.2 - System Quality Group Classification (3) SRP 3.7.1 - Seismic Design Parameters (4) SRP 3.7.2 - Seismic System Analysis (5) SRP 3.7.3 - Seismic Subsystem Analysis (6) SRP 3.8.4 - Other Seismic Category I Structures (including Appendix D) , Technical Position on Spent Fuel Rack (7) SRP 3. 8. 5 - Foundations for Seismic Category I Structures, Revision 1, 1981 (8) SRP 9.1.2 - Spent Fuel Storage, Revision 3, 1981 (9) SRP 9.1.3 - Spent Fuel Pool Cooling and Cleanup System (10) SRP 9.1.4 - Light.Load Handling System (11) SRP 9.1.5 - Heavy Load Handling System

( 12) SRP 15. 7. 4 - Radiological Consequences of Fu_el Handling Accidents 3-9 l_

AUXILIARY FLOW HOLE (TYPICAL)

WELD SEAM FI GLJRE 3 3 1 I I SEAM WELDED PRECISION FORMED CHANNELS

I* *

• SHEATHING FIGURE 3.3.2 COMPOSITE BOX ASSEMBLY

~ BOX CELLS DEVELOPED CELL

~ CELL I I p ITCH ----i BORAL PANEL SHEATHING POISON LENGTH CELL LENGTH ONE INCH LATERAL BASEPALTE FLOW HOLE C TY.P )

FIGURE 30303 ELEVATION VIEW OF REGION II RACK MODULE

FORMED CELL BOX CELL i i I i i i I  ;(

i I I i i i

- -------f------- ------i------ ------i------ -------+------- ------1------- ----+-------

i i i i I i i i i i i i

- f-------f------- ------i----- f-------t ------ -------+ ----- ' ------+------ --------i--------

i i

i i

i i

- f-------t------- -------j ------- ----~ ------ --------t------- --------r------- -------r-------

i i i i i i i i I I i i i i i I i i i i i i i i

- f-------r------- -----1------ ----~------ ------i------- -------t------- -------r-------

i i i i i i i i i i i i i i i i i i I I I I I I i i I i i i I i I ______i i______ _ i i

- f-------t------ ---------]*----- -------i------ -------r------ -----1-------

i i i i i i i i i i i i i i i i i i I I I I I I I I i i I i I

- f-------r------

I c--------i------ i

---~,-------

_____Ti ______ _______ TI ______ ------1*------

i i i i i i i i i i i i i i i i i i i

!  !  !  !  ! I

!  !  ! I I i I I I i i i

- _______ii_____ -----1--------

i i

---

,--------- i

---

1------- i ____ j_ _______

---

,------- I i i i i [ I i i i i i I i i i i i i FIGURE 3.3.4 TYPICAL ARRAY OF REGION KI CELLS

<NON-FLUX TRAP CONSTRUCTION)

• CELL GUSSET

• ( A) ELEVATION (B) PLAN VIEW IN MINIATURE FIGURE 3.3.5 ADJUSTABLE SUPPORT

• 4.0 4.1 CRITICALITY SAFETY ANALYSES Design Bases The maximum density spent fuel storage racks for Salem Units 1 and 2 are designed to assure that the effective neutron multiplication factor (keff) is equal to or less than O. 9 5 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with unborated water at the temperature within the operating range corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including mechanical tolerances. All uncertainties are statistically combined, such that the final kett will be equal to or less than 0.95 with a 95% probability at a 95% confidence level [4.1.1].

The storage rack analyses encompass three of the existing Exxon storage racks which will be retained as Region 1, and a new Region 2 in which credit is taken for fuel burnup. Applicable codes, standards, and regulations or pertinent sections thereof, include the following:

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

USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev. 3 - July 1981 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.

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

4-1

---1 ANSI ANS-8.17-1984, criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.

USNRC guidelines and the applicable ANSI standards specify that the maximum effective multiplication factor, "kett", including uncertainties, shall be less than or equal to 0.95. The infinite multiplication factor, "k..,", is calculated for an infinite array, neglecting neutron losses due to leakage from the actual storage rack, and therefore results in a higher and more conservative value. In the present evaluation of criticality safety in the Salem storage racks, the system limiting design basis criterion is based upon an infinite radial array of assemblies with a finite axial dimension, except for certain cases (alternate configurations and accident conditions) that inherently involve radial neutron leakage.

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

Moderator is unborated water at a temperature of 4°C, corresponding to the maximum possible moderator density and results in the highest reactivity.

In all cases (except for the assessment of certain conditions where neutron leakage is inherent), the infinite multiplication factor, k.., was used rather than the effective multiplication factor, keff (i.e. , neutron loss from radial leakage neglected).

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

4-2

The design basis fuel assembly is a standard Westinghouse 17 x 17 fuel assembly containing U0 2 at an initial enrichment of 4.5 wt%

U-235 by weight, corresponding to 57. 54 grams U-235 per axial centimeter of fuel assembly. A Westinghouse Vantage-SH fuel assembly, identical in dimensions to the standard assembly but with a burnable poison (IFBA) coating on some fuel rods was also considered with enrichments up to 5.0%. Evaluations of alternate storage configurations are also presented.

The water in the spent fuel storage pool normally contains soluble boron which would result in large subcriticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 be evaluated in the absence of soluble boron. The double contingency principle of ANSI N-16.1-1975 and of the April 1978 NRC letter allows credit for soluble boron under other abnormal or accident conditions since o~ly a single accident need be considered at one time. Consequences of abnormal and accident conditions have also been evaluated, where "abnormal" refers to conditions (such as higher water temperatures resulting from full-core discharge) which may reasonably be expected to occur during the lifetime of the plant and "accident" refers to conditions which are not expected.

to occur but nevertheless must be protected against.

4-3

4.2

SUMMARY

OF CRITICALITY ANALYSES 4.2.1 Region 1 Storage Cells In Region 1 (existing Exxon storage racks), analysis showed that the racks can safely accommodate fuel of 4.25% enrichment, with a maximum k..* of O. 949, including calculational and manufacturing uncertainties (95%/95%) as summarized in Table 4. 2. 1. In addition, the following storage configurations were also evaluated.

Peripheral Cells - Because of neutron leakage from the rack, the peripheral cells of Region 1 can accommodate fresh fuel of

5. 0% enrichment (rack otherwise filled with fuel of 4. 25%

enrichment or equivalent), with a maximum keff (3-dimensional KEN0-5a calculation) of 0.945, including uncertainties.

IFBA Assemblies - Fuel assemblies of 5. 0% enrichment con-taining boron-coated fuel pellets (Integral Fuel Burnable Absorber, or IFBA, with 2.355 mg. B-10/inch) may be stored in Region 1 without restriction, provided the assembly contains a minimum of 32 IFBA rods. A larger number of IFBA rods (64 or 104 usually used) woulcrprovide a greater margin below the 0.95 kett limit. For enrichments between 4.25% and 5%, the minimum required number of IFBA rods may be determined by interpolation (e.g, 4.6% fuel would require a minimum of 15 IFBA rods).

Spent Fuel Assemblies with initial enrichments up to 5.0% are acceptable for storage in Region 1 provided the enrichment burnup combination is within the acceptable domain of Figure 4.2.1. The limiting burnup (Bu) defining the acceptable domain has been fitted to the initial enrichment ( E%) as follows:

• For an infinite radial array of finite length *

• 4-4

4.2.2 Region 2 Storage Cells Region 2 storage cells are designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups within the acceptable domain depicted in Figure 4.2.2. Alternate storage configurations considered {with burnup limits also shown in Figure 4.2.2) include

{1) The use of the peripheral cells as a "Region 3 11 by taking credit for neutron leakage from these peripheral cells, and

{2) A loading pattern consisting of a fresh fuel assembly

( 4. 5% enrichment) in the center of a 3 x 3 array, surrounded by fuel burned to 50 MWD/KgU.

In addition, a checkerboard loading pattern of fresh fuel (5.0%

enrichment) and empty cells was also found acceptable. The presence of IFBA rods does not affect the limiting storage requirements for Region 2, since at the higher burnups, the B-10 in the IFBA coating is burned out.

The criticality analyses for each of the separate arrangements in the spent fuel storage pool are summarized in Table 4.2.2 assuming the design basis single accident condition of the loss of all soluble boron. The calculated maximum reactivities in Region 2 include a burnup-dependent allowance for uncertainty in depletion calculations and incorporate the effect of the estimated axial burnup distribution. As cooling time increases in long-term storage, decay of Pu-241 will result in a significant decrease in reactivity, which will provide an increasing subcriticality margin with storage time.

After the analyses were completed, the B-10 loading was increased from 0.0216 gm/cm2 to 0.0227 gm/cm2

• This provides an additional conservatism by 0.0021 ok below the results reported here.

4-5

The checkerboard loading pattern with fresh 5. 0% fuel has no penalty for burnup and the maximu,m keff was calculated to be O. 8462, including uncertainties.

Region 2 can safely accommodate fuel of any initial enrichment and discharge fuel burnup, provided the combination falls within the acceptable domain illustrated by the solid line in Figure 4.2.2.

For convenience, the minimum (limiting) burnup data in Figure 4.2.2 for unrestricted storage in Region 2 can be described as a function of the initial enrichment, E, in weight percent U-235 by a fitted polynomial expression that conservatively bounds the calculated values.

The storage cells along the periphery of the racks facing the concrete pool wall are in an area of high n~utron leakage and can accommodate fuel of lower burnup than that for unrestricted storage. Calculations for the peripheral~a~ea (called Region 3) confirmed that these peripheral cells could accommodate fuel with burnups corresponding the lower curve in Figure 4. 2. 2. These burnup limits have been fitted to a polynomial expression that conservatively bound the calculated limits.

4-6

Similarly, calculations were made for fresh fuel of 5.0% enrichment in the central cell of a 3 x 3 array (9-cells) with the eight surrounding cells filled with fuel of 4.5% enrichment burned to 50 MWD/KgU (or equivalent). For other enrichments, the same reactivity would exist at burnups indicated by the upper curve in Figure 4.2.2. These data have also bee~ fitted to a polynomial expression to conservatively bound the calculated values .

The burnup criteria identified above for acceptable storage in the three arrangements can be implemented in appropriate administrative procedures to assure verified burnup as specified in the proposed Regulatory Guide 1.13, Revision 2. Administrative procedures will also be employed to confirm and assure the presence of soluble poison in the pool water during fuel handling operations, as a further margin of safety and as a precaution in the event of fuel misplacement during fuel hanqling operations *

• 4-7

• 4.2.3 Abnormal and Accident Conditions Al though credit for the soluble poison normally present in the spent fuel pool water is permitted under abnormal or .accident conditions, most abnormal or accident conditions will not result in exceeding the limiting reactivity (keff of O. 9 5) even in the absence of soluble poison. The effects on reactivity of credible abnormal and accident conditions are presented in detail in Section 4.7 and briefly summarized in Table 4. 2. 3. Of these abnormal/accident conditions, only one has the potential for a more than negligible positive reactivity effect.

The inadvertent misplacement of a fresh fuel assembly has the potential for exceeding the limiting reactivity, should there be a concurrent and independent accident condition resulting in the loss of all soluble boron. Administrative procedures to assure the presence of soluble poison during fuel handling operations will preclude the possibility of the simultaneous occurrence of the two independent accident conditions. The largest reactivity increase would occur if a new fuel assembly of the highest permissible reactivity were to be positioned in an otherwise fully loaded Region 2 storage rack module. Under this accident condition, credit for the presence of soluble poison is permitted by NRC guidelines*, and it is estimated, for the worst case scenario, that a minimum boron concentration of about 600 ppm boron is adequate to assure that the limiting keff of O. 95 is not exceeded for any of the alternative storage arrangements in either Region 1 or Region 2.

• oouble contingency principle of ANSI NlG.1-1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Reg. Guide 1.13 (Section 1.4, Appendix A).

4-8

• 4.3 4.3.1 Reference Fuel Storage Cells Reference Fuel Assembly The design basis fuel assembly, illustrated in Figure 4.3.1, is a 17 x 17 array of fuel rods with 25 rods replaced by control rod guide tubes and an instrument thimble. The standard Westinghouse (and Vantage-SH) fuel assembly design specifications are summarized in Table 4.3.1 with the expected range of the significant manufac-turing tolerances.

4.3.2 Region 1 Fuel Storage Cells The Region 1 fuel storage cell, illustrated in Figure 4.3.1, uses a flux-trap water-gap to enhance reactivity control. A formed Boral panel surrounds the inner 0.033 inch stainless steel inner box and is backed up by a 0.083 inch stainless steel outer box.

Measurements of samples of a large number of the Boral panels show a mean B-10 loading of 0.0252 gm/cm2 with a minimum of 0.022 gm/cm2

• The cell lattice spacing is 10.50 inches (nominal) with a 1.118

+/- O.165 inch water-gap (flux-trap). These cells are assembled into lOxlO array modules, with three modules being used in the storage pool.

4.3.3 Region 2 Fuel Storage Cells The design basis for Region 2 storage cells is fuel of 4.5 wt% u-235 initial enrichment burned to 38 MWD/KgU A single Boral absorber panel is positioned between the o. 075 inch stainless steel box and the 0.035 inch sheath that forms the wall of adjacent storage cells. These cells, shown in Figure 4.3.2, are located on a lattice spacing of 9.052 +/- 0.04 inches. The Boral absorber has a thickness of 0.075 +/- 0.004 inches and a nominal B-10 areal

• 4-9

• density of 0.0216 g/cm2

• After the analyses were completed, the B-10 areal density was increased to 0.0227 gm/cm2 which provides an additional conservatism of 0.0021 ok to the results reported here.

4.4 Analytical Methodology 4.4.1 Reference Design Calculations In the fuel rack analyses, the primary criticality analyses of the high density spent fuel storage racks were performed with a two-dimensional multi-group transport theory technique, using the CASM0-3 computer code [4.4.1]. Independent verification calculations were made with a Monte Carlo technique utilizing the KENO-Sa computer code [ 4. 4. 2], with the 27-group SCALE* cross-section library [ 4. 4. 3] and the NITAWL subroutine for U-238 resonance shielding effects (Nordheim integral treatment).

Benchmark calculations, presented in Appendix A, indicate a bias of 0.0000 with an uncertainty of +/- 0.0024 for CASM0-3 and 0.0101 +/-

0.0017 (95%/95% [4.4.4]) for NITAWL-KENO-Sa. CASMO was also used both for burnup calculations and for evaluating the small reactivity increments associated with manufacturing tolerances.

In tracking long-term (30-year) reactivity effects of. spent fuel stored in Region 2 of the fuel storage rack, previous [ 4. 4. 5,

4. 4. 6 J calculations have confirmed a continuous reduction in reactivity with time (after Xe decay) due primarily to Pu-241 decay and Am-241 growth.

In the geometric model used in the CASM0-3 calculations, each fuel rod and its cladding were described explicitly and reflecting boundary conditions (zero neutron* current) were used in the axial

• 11 sCALE" is an acronym for B,tandardized Computer Analysis for Licensing Evaluation, a standard cross-section set developed by ORNL for the USNRC .

• 4-10

• direction and at the centerline of the Boral and steel plates between storage cells. These boundary conditions nave the effect of creating an infinite array of storage cells in all directions.

A similar geometric model was. also used for the KENO-Sa calculations except that the finite fuel length was used in the axial direction.

NITAWL-KENO-Sa Monte Ca~lo calculations inherently include a statistical uncertainty due to the random nature of neutron tracking. To minimize the statistical uncertainty of the KENO-calculated reactivity, a minimum of soo,ooo neutron histories in 1000 generations of 500 neutrons each are accumulated in each calculation. KENO-Sa calculations were made at 20°c with a temperature correction to 4°C determine.d by CASM0-3.

4.4.2 Fuel Burnup Calculations and Uncertainties CASM0-3 was used for burnup calculations . in the hot operating condition. CASMO has been extensively benchmarked (Appendix A and References [4.4.1 and 4.4.5]) against cold, clean, critical experiments (including *plutonium-bearing fuel) , Monte Carlo calculations, reactor operations, and heavy-element concentrations in irradiated fuel.

Since critical experiment data with spent fuel is not available for determining the uncertainty in depletion calculations, the uncer-tainty in reactivity was assumed to be adequately represented by the assumption of 5% of the reactivity decrement from zero to the burnup of concern. For the Salem Station storage racks at the design basis burnup of 3 8 MWD /KgU, the uncertainty of 5% in reactivity decrement corresponds to 0.0142 ok uncertainty in reactivity. Table 4. 4. 1 summ.arizes results of the burnup analyses and allowances for uncertainties at other burnups. The reactivity

• 4-11

uncertainty due to burnup may be either positive or negative, it is combined statistically with other uncertainties*. Allowance for uncertainty in burnup calculations is believed to be a conservative

• estimate, particularly in view of the substantial reactivity decrease with aged fuel.

4.4.3 Effect of Axial Burnup Distribution Initially, fuel loaded into the reactor will burn with a slightly skewed cosine power distribution. As burnup progresses, the burnup distribution will tend to flatten, becoming more highly burned in the central regions than in the upper and lower ends. This effect may be clearly seen in the curves compiled in Reference [4.4.6]. At high burnup, the more reactive fuel near the

• ends of the fuel assembly (less than average burnup) occurs in regions of lower reactivity worth due to neutron leakage. Consequently, it would be expected.that over most of the burnup history, distributed burnup fuel assemblies would exhibit a slightly lower reactivity than that calculated for the average burnup. As burnup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the* axial power distribution, precluding the existence of large regions of significantly reduced burnup.

Among others, Turner [4.4.4] has provided generic analyses of the axial burnup effect based upon calculated and measured axial burnup distributions. These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnups at values less than . about 28-30 MWD/KgU. The trends observed, however, suggest the possibility of a small positive reactivity effect at higher burnup values. KENO-Sa calculations were made, based on a representative axial burnup distribution in 10 axial zones using burnup-equivalent enrichments in each of the axial zones. Results of these calculations indicate the effect of the 4-12

axial distribution in burnup would not be greater than 0.0070 ok at the design basis burnup of ~8 MWD/KgU, increasing to 0.0165 ok at 45 MWD/kgU and 0.0268 ok at 50 MWD/kgU burnup. In configurations such as the 1-in-9 array with fresh fuel, the fresh fuel dominates the axial variation in reactivity and no positive effect is seen from the axial variation in burnup in fuel of the surrounding eight cells.

4.4.4 Long-Term Changes in Reactivity since the fuel racks in Region 2 are intended to contain spent fuel for long periods of time, consideration was given to the long-term changes in reactivity of spent fuel. Published data [ 4

• 4. 5, 4. 4

• 6]

confirm that reactivity continuously decreases as the spent fuel in storage ages. Early in the decay period, Xenon grows from Iodine decay (reducing reactivity) and subsequently decays, with the reactivity reaching a maximum at 100-200 hours. The decay of Pu-241 (13-year half-life) and growth of Am-241 substantially reduce reactivity during long term storage. The design of the Salem racks do not take *credit for this long-term reduction in reactivity, other than to indicate an increasing subcriticality margin in Region 2 of the spent fuel storage pool.

4.5 Region 1 Criticality Analyses and Tolerances 4.5.1 Nominal Design Case Region 1 consists of three of the existing Exxon racks as described in Figure 4.3.1. The principal method of analysis was the CASM0-3 code, supplemented by confirming calculations with NITAWL-KENO-Sa.

Calculations were made for several initial enrichments and,. by trial-and-error, it was determined that the Exxon racks could safely accommodate fuel of 4.25% initial enrichment with a maximum

• 4-13

k..* of o. 9 4 9, including uncertainties, as shown in Table 4

• 2. 1.

The CASM0-3 calculation for fuel of 4.25% enrichment *gave a k,,. of o

• 9 3 2 5 compared to a bias-corrected KENO-Sa result of O

• 9 3 2 5 +/-

0.0007 Jc.. Storage of fuel with enrichments greater than 4.25%

requires additional factors as described in the following sections.

4.5.2 Peripheral Cells The peripheral cells of Region 1 facing the concrete walls of the spent fuel pool constitute an area of high neutron leakage and therefore, can accommodate fuel with enrichments greater than 4.25%. KENO-Sa calculations with 5.0% enriched fuel in the peripheral cells and 4.25% fuel in the internal cells gave a bias-corrected keff of O. 9 2 51 +/- O. OO07. With the temperature correction (0.0016) and uncertainties (+0.0188 *ok) added, the maximum keff is 0.946. Therefore, fuel of 5.0% enrichment may be safely stored in the peripheral cells of Region 1 facing the concrete walls of the pool, with assurance that the maximum keff will be less than the O. 95 keff regulatory *1imit.

4.5.3 With IFBA Assemblies Some fuel assemblies may contain integral fuel burnable absorber (IFBA) rods consisting of a thin coating of boron on the surface of the fuel pellets. Since the boron-10 has a higher cross-section than the fuel, it burns out more rapidly. Consequently, the reactivity effect must be evaluated at the point of high~st reactivity over burnup. CASM0-3 calculations show that the h~ghest reactivity occurs at zero burnup for assemblies with less than 64 IFBA rods and at higher burnups for assemblies with more than 64 IFBA rods.

• Infinite radial array of finite length .

• 4-14

For the evaluation of cases with IFBA rods, the boron coating was assumed to cover only the central 10 feet of the axial length (i.e., 12-inch cutback top and bottom). Three-dimensional KENO-Sa calculations gave the following bias-corrected results for fuel of S.0% enrichment.

0 IFBA k = 0.9S72 +/- 0.0007 32 IFBA k.. = 0.914S +/- 0.0007 64 IFBA k.. = 0.889S +/- 0.0007 With 32 IFBA rods and with the temperature correction and uncertainties added, the maximum keff is O. 93S, which confirms that S% fuel assemblies with 32 IFBA rods are acceptable for unrestricted storage in Region 1. The margin in reactivity below the 0.9S limit constitutes an allowance for possible uncertainties

. that might be associated with the boron loading in the coating with arrangements of IFBA rods within the assembly. For example, a 10%

uncertainty in boron loading in the coating could result in a 0.004 cS'k uncertainty in reactivity which would increase the maximum reactivity to 0.939.

For enrichments between 4.2S% (no IFBA) and S% (32 IFBA rods) the required number of IFBA rods may be determined by interpolation.

For example, 4.6% enriched fuel would require approximately 16 IFBA rods. A KENO-Sa calculation with a conservative arrangement of IFBA rods gave a maximum reactivity of o. 94S for fuel of 4. 6%

enrichment.

To summarize, the fuel to be stored in Region 1 shall be restricted to Standard or Vantage SH fuel which conform to one of the following limiting conditions.

a) fuel assemblies with a nominal average enrichment of 4.2S U-23S or less.

4-1S

I b) fuel assemblies with a nominal average enrichment (E) between 4.25 w/o and 5.0 w/o U-235, provided the assembly includes integral fuel burnable absorber ( IFBA) rods with at least 2.355 mg B-10 on the surface of the U02 pellets and the number of such IFBA rods equals or exceeds the number, N, determined by the following equation:

N = 42.67 (E - 4.25) rounded to the nearest whole number of rods.

c) fuel assemblies with a nominal average enrichment (E) between 4. 25 w/o and 5. o w/o U-235, provided the assembly has attained the minimum burnup (Bu, in MWD/KgU) determined by the following equation:

Bu = -26.212 + 6.1677E 4.5.4 With Credit For Fuel Burnup

• Fuel with enrichments greater than 4.25% (and no IFBA rods) may be safely stored in Region 1 provided the enrichment-burnup combination is within the acceptable domain of Figure 4. 2 .1. Table 4.4.1 lists the calculated limits on burnup for initial enrichments above 4.25%. For these cases, the maximum keff' including uncertainties is 0.949 which is less than the regulatory limit.

4.5.5 Uncertainties Due to Tolerance in Region 1 4.5.5.1 Boron Loading Tolerances The Boral absorber used *in the Region 1 storage cells are 0.091 inch thick. Measurements on a large number of samples from the Baral *sheets confirmed that the as-built B-10 loading is O. 0252 gm/cm2 with a minimum of 0.022 gm/cm2

• Differential CASMO calculations indicate that this tolerance limit results in an incremental reactivity uncertainty of +/- .0048 ok.

4-16

• 4.5.5.2 Tolerance in Cell Lattice Spacing The manufacturing tolerance on inner box dimension (+/- o. 0625 inches) affects the storage cell lattice spacing between fuel assemblies. This corresponds to an uncertainty in reactivity of +/-

0.0012 cS'k.

4.5.5.3 Stainless Steel Thickness Tolerance The nominal thickness of the stainless steel box walls are 0.033 and 0.083 inch with tolerances of +/- 0.003 and +/- 0.005 inches, respectively. These tolerances result in combined uncertainty in reactivity of +/- 0.0004 cS'k.

4.5.5.4 Fuel Enrichment and Density Tolerances Uncertainties in reactivity due to tolerances on fuel enrichment and U02 density were calculated using CASM0-3 in the Region 1 storage rack configuration. These calculations resulted in the following uncertainties:

+/- o.os in % Enrichment +/- 0.0017 cS'k Uncertainty

+/- 0.20 in U0 2 Density +/- 0.0020 ok Uncertainty 4.5.6 Eccentric Fuel Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell. A calculation was also made with the fuel assemblies assumed to be in the corner of the storage rack cell (four-assembly cluster at closest approach). This calculation indicated that eccentric fuel positioning results in a decrease in reactivity as determined by KENO-Sa calculations. The highest reactivity, therefore, corresponds to the reference design with the fuel assemblies positioned in the center of the storage cells.

4-17

4.6 Region 2 Criticality Analyses and Tolerances 4.6.l Nominal Design Case The principal method of analysis was the CASMO code, using the restart option to analytically transfer fuel of a specified burnup into the storage rack configuration at a reference temperature of 4°C. Calculations were made for fuel of several different initial enrichments and, at each enrichment, a limiting k.. value was established which included an additional factor for uncertainty in the burnup analyses and for the axial burnup distribution (if positive). The restart CASMO calculations. (cold, no-Xenon, rack geometry) were then interpolated to define the burnup value yielding the limiting k.. value for each enrichment, as indicated in Table 4.4.2. These converged burnup values define the boundary of the acceptable domain shown*in Figure 4.2.2.

At a burnup of 38 MWD/KgU, the sensitivity to burnup is calculated to be 0. 0067 ok per MWD/KgU. During long-term storage, the k...

values of the Region 2 fuel rack will decrease continuously as indicated in Section 4.4.4.

An independent AMPX-KENO calculation was used to provide additional confidence in the reference Region 2 criticality analyses. Fuel of 1.593 wt% initial enrichment (equivalent to the reference rack design for burned fuel without burnup uncertainty) was analyzed by NITAWL-KEN0-5a and by the CASMO model used.for the Region 2 rack analysis. For this case, the CASMO k... (0.9060) at 1.593 wt%

enrichment was slightly higher than the bias-corrected KENO value (0.9019 +/- 0.0019, 95%/95%) obtained in the AMPX-KENO calculations, confirming the CASMO calculation.

4-18

4.6.2 Uncertainties Due to Tolerances 4.6.2.1 Boron Loading Tolerances The Boral absorber panels used in the Region 2 storage cells are 0.075 inch thick with a nominal B-10 areal density of 0.0227 g/cm2

• The manufacturing limit of +/- 0.0027 g/cm2 in B-10 loading assures that at any point the minimum B-10 areal density will not be less than 0.020 g/cm2

• Differential CASMO calculations indicate that this tolerance limit results in an incremental reactivity uncertainty of +/- .0030 ok.

4.6.2.2 Boral Width Tolerance The reference storage cell design for Region 2 (Figure 4.3.2) uses a Boral absorber width of 7. 50 +/- o. 06 inches. This tolerance results in a reactivity uncertainty of +/- 0.0003 ok.

4.6.2.3 Tolerance in Cell Lattice Spacing The manufacturing tolerance on inner box dimension (+/- 0.04 inches) affects the storage-cell lattice spacing between fuel assemblies.

This corresponds to an uncertainty in reactivity of +/- 0.0009 ok.

4.6.2.4 Stainless Steel Thickness Tolerance The nominal thickness of the stainless steel box wall is 0.075 inch with a tolerance of +/- 0.005 inches, resulting in an uncertainty in reactivity of +/- o.oooi ok.

4-19

---1 4.6.2.S Fuel Enrichment and Density Tolerances Uncertainties in reactivity due to tolerances on fuel enrichment and U0 2 density were calculated using CASM0-3 to deplete the fuel and to restart in the Region 2 geometry. These calculations resulted in the following uncertainties:

+/- o.os in % Enrichment +/- 0.0037 ok Uncertainty

+/- 0.20 in U0 2 Density +/- 0.0022 ok Uncertainty 4.6.3 Eccentric Fuel Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell. Calculations were also made with the fuel assemblies assumed to be in the corner of the storage rack cell (four-assembly cluster at closest approach). These calculations indicated that eccentric fuel positioning results in a decrease in reactivity as determined by KENO-Sa calculations. The highest reactivity, therefore, corresponds to the reference design with the fuel assemblies positioned in the center of the storage cells.

4.6.4 Alternate Storage Configurations Three alternate configurations were also evaluated, including Checkerboard loading of fresh S.0% enriched fuel with empty cells, Use of the Peripheral cells as a Region 3, with fuel of 30 MWD/KgU burnup or equivalent, Location of a fresh fuel assembly (S.0% enrichment) in the center of a 3 x 3 array of 4

• S% enriched fuel assemblies burned to SO MWD/KgU or equivalent.

4-20

4.6.4.1 Checkerboard Loading Calculations were made for a checkerboard loading pattern of fresh

5. 0% enriched fuel with the alternate cells remaining empty of fuel. This results in a low reactivity*, calculated (KENO-Sa) to be 0.8462 (including uncertainties), which is well below the regulatory limit (0.-95 keff).

4.6.4.2 Peripheral Cell Locations .

The outer peripheral cells facing non-fueled areas (i.e., walls of the storage pool) are in a region of high neutron leakage and can therefore accommodate fuel of lower burnup or higher enrichment without significant increase in system reactivity. For a design basis burnup of 30 MWD/KgU for 4.5% enriched fuel, the calculated reactivity was a maximum of 0.929, including uncertainties, which is within the regulatory limit (0.95 kett>* Therefore the use of the peripheral cells for the higher reactivity fuel is an acceptable configuration. Additional calculations for other initial fuel enrichments were made to define the corresponding burnup limits. These burnup limits are shown as the lower curve in Figure 4.2.1 and tabulated in Table 4.4.2.

4-21

4.6.4.3 3x3 Array with Fresh Fuel Calculations were made (KENO-Sa) for a fresh fuel assembly centered in a configuration of 9-cells in a 3 x 3 array, as illustrated below:

Spent Spent Spent Fuel Fuel Fuel Spent Fresh Spent Fuel Fuel Fuel Spent Spent Spent Fuel Fuel Fuel With fuel of 4.5% initial enrichment burned to 50 MWD/KgU in the surrounding 8 cells, the k.. was calculated to be a maximum of 0.944, including uncertainties (infinite radial array, finite axial length)

• Similar calculations for fuel of other enrichments (CASM0-3 with restart option in the storage cells) were made to establish th~ limiting burnups shown.in Figure 4.2.1 and tabulated in Table 4.4.1. With these restrictions, fresh fuel may be stored in the center of a 3 x 3 array with a maximum *keff within the regulatory limit.

A subset of the 1-in-9 array occurs along the edge of the rack where the configuration is that described below.

Spent Spent Fuel Fuel Spent Fresh Fuel Fuel Reflector Spent Spent Fuel Fuel 4-22

• Due to leakage from the periphery, this arrangement will have a lower k*ff than the 1-in-9 array occurring within the bulk of the storage rack. Thus, this arrangement may safely be used to store fresh fuel intermixed with fuel in the acceptable domain for the 1-in-9 array of Figure 4.2.2.

4.6.5 Summary of Region 2 Storage Criteria Region 2

• is designed to accommodate spent fuel and shall be maintained with a nominal 4. 05 inch center-to-center spacing between fuel assemblies in storage. The fuel to be stored in Region 2 shall be restricted to Standard or Vantage SH fuel which conform to one of the following limiting conditions:

a) fuel assemblies of an average enrichment, E, which have attained a minimum average burnup (Bu) determined by the following equation:

Bu= -32.06 + 25.21E - 3.723 E2 + 0.3535E3 may be stored in any Region 2 cell except those in (c) and (d).

b) fuel assemblies of an average enrichment, E, which have attained a minimum average burnup (Bu) determined by the following equation:

Bu = -25.56 + 15.14E - 0.602E2 may be stored in any peripheral cell facing an unfueled region.

c) unirradiated assemblies with a uniform average enrichment of 5. o w/ o U-2 3 5 or less provided the assemblies are stored in a checkerboard pattern with the intermediate cells containing only water or non-fissile bearing material.

4-23

• d) unirradiated fuel assemblies with a uniform average enrichment (E) of 5.0 w/o U-235 or less may be stored in the central cell of any 3x3 array of cells provided the surrounding eight cells are empty or contain assemblies with minimum burnups (Bu) determined by the following equation:

Bu= -15.48 + 17.SOE - 0.7038E2 In this configuration, none of the nine cells in any 3x3 array shall be common to cells in any other similar 3x3 array. Along the rack periphery, the non-fueled region is equivalent to three outer cells in a 3x3 array.

4.7 Abnormal and Accident Conditions 4.7.1 Temperature and Water Density Effects The moderator temperature coefficient of reactivity in both Region 1 and Region 2 is negative; a moderator temperature of 4°C was assumed, which assures that the true reactivity will always be lower over the expected range of water temperatures. Temperature effects on reactivity have been calculated and the results are shown in Table 4. 5. 1.. Introducing voids in the water internal to the storage cell (to simulate boiling) decreased reactivity, as shown in the table. These calculations were made with CASM0-3 to avoid the deficiency in the NITAWL cross-section library.* The incremental temperature effect from 20°c to 4°C is +o. 0016 in Region 1 and +0.0018 in Region 2.

With soluble poison present, the temperature coefficients of reactivity would differ from those inferred from the data in Table 4

• 5. 1. However, the reactivities would also be substantially

• See USNRC Information Notice 91-66.

4-24

lower at all temperatures with soluble boron_present, and the data in Table 4. 5. 1 is pertinent to the higher-reactivity unborated case.

4.7.2 Dropped Fuel 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 fuel in the rack of more than 12 inches, including any deformation under seismic or accident conditions. At this separation distance, the effect on reactivity would be insignificant (<0.0001 &k). Furthermore, soluble boron in the pool water would substantially reduce the reactivity and assure that the true reactivity is always __ less than the limiting value for any conceivable dropped fuel accident.

4.7.3 Lateral Rack Movement Lateral motion of the rack modules under seismic conditions could potentially alter the spacing between rack modules. However, Region 2 storage cells are insensitive to the spacing between modules. Furthermore, soluble poison would assure that a reactivity less than the design limitation is maintained under all accident or abnormal conditions.

4.7.4 Abnormal Location of a Fuel Assembly The abnormal location of a fresh unirradiated fuel assembly of 5.0 wt% enrichment could, in the absence of soluble poison, result in exceeding the design reactivity limitation (k.., of o. 95)

• This could occur if a fresh fuel assembly of 5.0% enrichment were to be inadvertently loaded into a Region 1 or Region 2 storage cell 4-25

otherwise filled to capacity, for any of the configurations.

Soluble boron in the spent fuel pool water, for which credit is permitted under these accident conditions, would assure that the reactivity is maintained substantially less than the design limitation. Calculations (KENO-Sa) for the worst case configuration confirmed that 600 ppm soluble boron is adequate to compensate for a mis-located fuel accident. The Salem plant procedures require that the boron concentration in the pool be maintained between 2300 and 2500 ppm during operating modes, confirmed by weekly surveillance measurements. This is more than enough to compensate for any postulated accident.  :-.*

4.8 References

[4.1.1] M.G. Natrella, Experimental Statistics National Bureau of Standards, Handbook 91, August 1963.

[4.4.1] 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. 2 6, p.

604, 1977.

M. Edenius et al., "CASMO Benchmark Report," Studsvik/ RF-78-6293, Aktiebolaget Atomenergi, March 1978.

[4.4.2] Green, Lucious, Petrie, Ford, White, Wright, "PSR-63/AMPX-1 (code package),

AMPX Modular Code System for Generating Coupled Mul tigroup Neutron-Gamma Libraries from ENDF/B," ORNL-TM-3706, Oak Ridge National Laboratory, March 1976.

[4.4.3] R.M. Westfall et al., "SCALE: A Modular Code System for performing Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-0200, 1979.

4-26

[4.4.4] s. E. Turner, "Uncertainty Analysis Burnup D:i:stributions", in Proceedings of a Workshop on the use of Burnup Credit in Spent Fuel Transport Casks, Sandia Report SAND-89-0018, October 1989.

[4.4.5] c. V. Parks, "Parametric Neutronic Analyses Related to Burnup Credit Cask Design" in Proceedings of a Workshop on the use of Burnup Credit in Spent Fuel Transport Casks, Sandia Report SAND-89-0018, October 1989.

[4.4.6] H. Richings, Some Notes on PWR (W) Power Distribution Probabilities for LOCA Probabilistic Analyses, Internal NRC Memorandum to P. s. Check, dated July 5, 1977.

4-27

Table 4.2.1

SUMMARY

OF CRITICALITY SAFETY ANALYSES REGION 1 Reference Peripheral 32 IFBA @ 4.63 Region 1 Cells Rods MWD/KgU Initial 4.25 5.0 5.0 5.0 Enrichment, %

Temperature 20°C 4°C ok Correction to 4° 0.0016 0.0016 Reference k.., 0.9325 0.9150 0.9044 0.9286 (CASMO) (KENO-Sa) (KENO-Sa) (CASMO)

Bias 0.0000 0.0101 0.0101 0.0000 Correction for axial -0.0023 leakage certainties and Allowances (Assumed same for all Configurations)

Bias +/- 0.0024 B-10 loading +/- 0.0048 Water gap +/- 0.0178 Inner box dimension +/- 0.0012 SS thickness +/- 0.0004 Fuel enrichment +/- 0.0017 Fuel density +/- 0.0020 Eccentric position Negative Statistical combination +/- 0.0188 of uncertainties*

Allowance for Burnup +0.0016 Uncertainty (5% of reactivity decrement)

Total 0.9302 0.9267 0.9161 0.9302

+/-0.0188 +/-0.0188 +/-0.0188 +/-0.0188 Maximum 0.9490 0.9455 0.9349 0.9490 Reactivity (k..,)

• Square root of sum of squares.

Table 4.2.2

SUMMARY

OF CRITICALITY SAFETY ANALYSES REGION 2 Reference Peripheral 3X3 <1 >

Region 2 Cells Array Design Basis Burnup 38 MWD/KgU 30 MWD/KgU 50 MWD/KgU 4.5% initial enrichment Temperature for 4°C 20°C 20°C analysis ok Correction to 4°C +0.001.8 +0.0018 Reference k.., 0.9060 0.8904 0.9169 Bias 0.0000 0.0101 0.0101 Uncertainties and Allowances (Assumed same for all Configurations)

Bias +/- 0.0024 B-10 loading +/- 0.0030 Boral width +/- 0.0003 Inner box dimension +/- 0.0009 SS thickness +/- 0.0001 Fuel enrichment +/- 0.0037 Fuel density +/- 0.0022 Eccentric position Negative Statistical combination +/- 0.0058 of uncertainties <2 >

Axial Burnup Distribution +0.0070 + 0.0070 o<3>

Allowance for Burnup +0.0142 + 0.0142 + 0. 0078( 3 )

Uncertainty Total 0.9272 0.9235 0.9366

+/-0.0058 +/-0.0058 +/-0.0058 Maximum Reactivity (k..,) 0.9330 0.9293 0.9424

<1 > For fresh fuel assembly of 5% enrichment in 1-in-9 array.

(2)

Square root of sum of squares.

3

<> In the 3x3 array, the axial burnup distribution and the burnup certainty are dominated by the fresh fuel assembly resulting in the ues shown.

Table 4.2.3 REACTIVITY EFFECTS OF ABNORMAL AND ACCIDENT CONDITIONS Accident/Abnormal Conditions Reactivity Effect Temperature increase (above 0°C) Negative (Table 4.5)

Void (boiling) Negative (Table 4.5)

Assembly dropped on top of rack Negligible (<0.0001 ok)

Lateral rack module movement Negligible (<0.0001 ok)

Misplacement of a fuel assembly Positive (controlled by soluble boron)

Table 4.3.1 DESIGN BASIS FUEL ASSEMBLY SPECIFICATIONS FUEL ROD DATA Outside diameter, in. 0.374 Cladding thickness, in. 0.0225 Cladding inside diameter, in~ 0.329 Cladding material Zr-4 Pellet density, % T.D. 95.0 Stack density, g U02 /cc 10.41 +/- 0.20 Pellet diameter, in. 0.3225 Maximum enrichment, wt % U-235 5.0 +/- 0.05

• FUEL ASSEMBLY DATA Fuel rod array Number of fuel rods 17x17 264 Fuel rod pitch, in. 0.496 Number of control rod guide and 25 instrument thimbles Thimble O.D., in. (nominal) 0.482 (0.474)*

Thimble I.D., in. (nominal) 0.450 (0.442)*

• Dimensions of the guide tubes differ slightly for the Standard Vantage-SH fuel assemblies, with insignificant impact on the

• culated reactivities.

I Table 4.4.1 LIMITING BURNUP VALUES FOR VARIOUS INITIAL ENRICHMENTS - REGION 1 MWD/KgU Limit For Initial Enrichment Unrestricted Region 1 4.25% 0 4.6% 2.03 5.0% 4.36

I Table 4.4.2 LIMITING BURNUP VALUES FOR VARIOUS INITIAL ENRICHMENTS - REGION 2 MWD/KgU Limit For Initial Unrestricted Peripheral 1-in-9 Cells Enrichment Region 2 Cells

• 3 x 3 Array 2.0 6.05 1.83 16.87 2.5 13.08 8.53 24.54 3.0 19.51 14.29 31.58 3.5 25.72 19.71 37.88 4.0 31.50 24.96 43.99 4.5 38.12 30.00 50.00 5.0 44.96 34.96 55.82 l

Table 4.5.1 EFFECT OF TEMPERATURE AND VOID ON CALCULATED REACTIVITY OF STORAGE RACK Incremental Reactivity Change, cS'k Case Region 1 Region 2

• 4°C (39 °F) 20°c (68°F) 65.6°C (150°F}

Reference

-0.0016

-0.0108 Reference

-0.0018

-0.0078 100°c(212°F) -0.0201 -0.0139 122°C {248°F) -0.0262 -0.0183 122°C (248°Fr + 10% void -0.0571 -0.0390

4.5 4.0

- - v I

- I

J

~

0' 3'

C) 3.5 3.0 ACCE bTABLE BURN UP DOMAJ N I

v I:

2.5 I

I 2.0 I I

1. 5 I

I

-- UNA1 :CEPT1 ~BLE EURNUP DOt AIN

1. 0

- /

0.5

-- I I

0.0 I I I

I I I I I I I I l I I I I I I I I I I I I I I I I l l l I I 5.0 4.2 4.4 4.6 4.8 INITIAL ENRICHMENT, ~U-235

• FIGURE 4.2.1 DEFINITION OF ACCEPT ABLE BURNUP DOMAIN IN REGION 1

0.207" 0.033 +/- 0.003 SS (INNER WALL)

---+------ I 0.091 +/- 0.007 BORAL

• ____ ----- ______ - - - - - - - - - ______ l _ ..

0.083 +/- 0.005 SS (OUTER WALL)

WESTINGHOUSE VS FUEL FUEL OD 0.3225 IN.

DENSITY 10.41 G/CC NUMBER 264 (17x17)

CLAD OD 0.374 IN.

CLAD ID 0.329 IN.

PITCH 0.496 THIMBLES OD 0.474 IN.

ID 0.442 IN.

No, 25 WATER GAP 1.118 +/- 0.165" 10.50 +/- 0.18"

---~-------------------------~--------------------------,-----

• . NOT TO SCALE I FIGURE 4.3.1 REGION 1 FUEL STORAGE CELL

BORAL 7.50 +/- 0.06" 2

0.075" THK 0.020 Min gmB-10/ cm h FUEL OD 0.3225 in.

0~ DENSITY 10.41 G/CC 00<:.... NUMBER 264 (17x17) 000~\. CLAD OD 0.374 In. 0.035" SHEATH 0000'- CLAD ID 0.329 in.

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0.075" SS BOX 9.052 +/- 0.040" LATIICE SPACING NOT TO SCALE FIGURE 4.3.2 *CROSS-SECTION OF STORAGE CELL FOR SALEM SPENT FUEL STORAGE RACK

I APPENDIX A TO SECTION 4 CRITICALITY BENCHMARK CALCULATIONS by Stanley E. Turner, PhD, PE HOLTEC INTERNATIONAL November, 1992

1.0 INTRODUCTION

AND

SUMMARY

The objective of this benchmarking study is to verify both the NITAWL-KENOsaC1.2> methodology with the 27-qroup SCALE cross-section library and the CASM0-3 code~* 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.

Results 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) NITAWL-KENOSa calculations consistently under-predict the critical eigenvalue by 0.0101 +/- 0.0018 ok (with a 95%

probability at a 95% confidence level) for critical experiments~

that are as representative as possible of realistic spent fuel storage rack configurations and poison worths.

Extensive benchmarking calculations of critical experi-ments with CASM03 have also been reported<S),

• giving a mean keff of 1.0004 +/- 0.0011 for 37 cases. With a K-factor of 2.14<6) for 95%

probability at a 95% confidence level, and conservatively neglect-ing the small overprediction, the CASM03 bias then becomes 0.0000

+/- O. 0024. CASM03 and NITAWL-KEN05a 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 +/- 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 core isotopics, agreeing with the measured values within experimental error.

A - 1

The benchmark calculations reported here confirm that either the 27-group (SCALE) NITAWL-KENO or CASM03 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. CASM03, however, is not reliable when large water gaps ( > 2 or 3 inches) are present.

2.0 NITAWL-KENO 5a BENCHMARK CALCULATIONS Analysis of a series of Babcock & Wilcox critical experiments<4>, including some with absorber panels typical of a poisoned spent fuel rack, is summarized in Table 1, as calculated with NITAWL-KEN05a using the 27-group SCALE cross-section library and the Nordheim resonance integral treatment in NITAWL. Dancoff factors for input to NITAWL were calculated with the Oak Ridge SUPERDAN routine (from the SCALE~ system.of codes). The mean for these calculations is 0.9899 +/- 0.0028 (1 a standard deviation of the population). With a one-sided tolerance factor corresponding to 95% probability at a 95% confidence leve1<6>, the calculational bias is + 0.0101 with an uncertainty of the mean of +/- 0.0018 for the sixteen critical experiments analyzed.

Similar calculational deviations have been reported by ORNLm for some 54 critical experiments (mostly clean critical without strong absorbers), obtaining a mean bias of 0.0100 +/- 0.0013 (95%/95%). These published results are in good agreement with the results obtained in the present analysis and lend further credence to the validity of the 27-group NITAWL-KEN05a calculational model for use in criticality analysis of high density spent fuel storage racks. No trends in kcff 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 experiments~ at 4.75% enrichment and for several of the BNWL er i ti ca ls with 4 . 2 6 % enriched fuel.

A - 2

• Analysis of the French criticals (Table 2) showed a tendency to overpredict the reactivity, a result also obtained by ORNLn~.

calculated kdf values showed a trend toward higher values with decreasing core size. In the absence of a significant enrichment The effect (see Section 3 below), this trend and the overprediction is attributed to a small inadequacy in NITAWL-KENOSa in calculating neutron leakage from very small assemblies.

Similar overprediction was also observed for the BNWL series of critical experimentsC11>, 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 kdf of 0.9959 +/- 0.0013 (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 failure of NITAWL-KENOSa 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-KENOSa is not significant. Furthermore, omitting results of the French and BNWL 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 is 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 CASM03, based on the asswnption 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

agreement establishes the validity of the interpolation routine, in conjunction with NITAWL-KEN05a,

• in calculating reactivities at temperatures other than 20°C (the reference library temperature).

The deficiency in the hydrogen scattering matrix does not appear except in the presence of a large water gap where the scattering matrix is important. However, the absolute value of the k~ from CASM03 is not reliable in the presence of a large water gap, although the relative values should be accurate. 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 ARRAYS The BAW close-packed series of critical experiments(l2> intended to simulate consolidated fuel, were analyzed with NITAWL-KEN05a.

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 close-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 0RNL<13>.

5.0 CASM03 BENCHMARK CALCULATIONS 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 practice is to treat the racks as an infinite medium of storage cells, neglecting leakage effects.

A - 4

CASM03 is a modification of the CASM0-2E code and has been extensively benchmarked against both mixed oxide and hot and cold critical experiments by studsvik Energiteknik~. Reported ana-lyses of 37 critical experiments indicate a mean kdf of 1.0004 +/-

0.0011 (1a). To independently confirm the validity of CASM03 (and to investigate any effect of enrichment), a series of calculations were made with CASM03 and with NITAWL-KENOSa on identical poisoned storage cells representative of high-density spent fuel storage racks. Results of these intercomparison calculations* (shown in Table 5 and in *Figure 2r are within the normal statistical variation of KENO calculations and confirm the bias of 0.0000 +/-

0.0024 (95%/95%} for CASM03.

Since two independent methods of analysis would not be expected to have the same error function with enrichment, results of the intercomparison analyses (Table 5) indicate that there is no significant effect of fuel enrichment over the range of enrich-ments involved in power reactor fuel. Furthermore, neglecting the French and BNWL cri~ical benchmarking in the determination of bias is a conservative approach.

A second series of CASM03-KEN05a intercomparison calculations consisting of five cases from the BAW critical experiments analyzed for the central cell only. The calculated results, also shown in Table 5, indicate a mean difference within the 95% confidence limit of the KEN05a calculations. This lends further credence to the recommended bias for CASM03 .

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

A - 5

• 6.0 1.

REFERENCES TO APPENDIX A Green, Lucious, Petrie, Ford, White, and Wright, "PSR /NITAWL-1 (code package) NITAWL Modular Code System For Generating coupled Mul tigroup Neutron-GAmma Libraries from ENDF /B", ORNL-TM-3706, Oak Ridge National Laboratory, November 1975.

2. R.M. Westfall et. al., "SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation",

NUREG/CR-0200, 1979.

3. A. Ahlin, M. Edenius, , .. and .. H.. Haggblom, "CASMO - A Fuel Assembly Burnup Program", AE-RF-76-4158, Studsvik report.

A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis", ANS Transactions, Vol. 26,

p. 604, 1977.

"CASM03 A Fuel Assembly Burnup Program, Users Manual",

Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986

4. M.N. Baldwin et al., "Critical Experiments Supporting Close Proximity Water storage of Power Reactor Fuel", BAW-1484-7, The Babcock & Wilcox Co., July 1979.

5. M. Edenius and A. Ahlin, "CASM03: New Features, Benchmarking, and Advanced Applications", Nuclear Science and Engineering, 100, 342-351, (1988)

6. M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

7. R.W. Westfall and J. H. Knight, "SCALE System Cross-section Validation with Shipping-cask Critical Experiments", ANS Transactions, Vol. 33, p. 368, November 1979

8. S. E. Turner and M.K. Gurley, "Evaluation of NITAWL-KENO Benchmark Calculations for High Density Spent Fuel Storage Racks", Nuclear Science and Engineering, 80(2) :230-237, February 1982.

9. J. c. Manaranche, et. al., "Dissolution and Storage Experiment with 4. 75% U-235 Enriched uo 2 Rods", Nuclear Technology, Vol.

so, pp 148, September 1980

10. A.M. Hathout, et. al., "Validation of Three Cross-section Libraries Used with the SCALE System for Criticality Analy-sis", Oak Ridge National Laboratory, NUREG/CR-1917, 1981 *

• A - 6

11. S.R. Bierman, et. al., acritical Separation between Sub-critical Clusters of 4.29 Wt. % 23Su Enriched U02 Rods in Water with Fixed Neutron Poisons", Battelle Pacific Northwest Laboratories, NUREG/CR/0073, May 1.978 (with August 1979 errata).

12. G.S. Hoovler, et al., "critical Experiments supporting Underwater storage of Lightly Packed Configurations of Spent Fuel Pins", BAW-1645-4, Babcock & Wilcox Company (1981).

13. R.M. Westfall, et al., "Assessment of criticality Computation-al Software for the U.S. Department of Energy Office of civilian Radioactive Waste Management Applications", Section 6, Fuel Consolidation Applications, ORNL/CSD/TM-24 7 (undated) .

• A - 7

Table 1 RESULTS OF 27-GROUP (SCALE) NITAWL-KEN05a CALCULATIONS OF B&W CRITICAL EXPERIMENTS Experiment Calculated (J Number kdf I 0.9922 +/- 0.0006 II 0.9917 +/- 0.0005 III 0.9931 +/- 0.0005 IX 0.9915 +/- 0.0006 x 0.9903 +/- 0.0006 XI 0.9919 +/- 0.0005 XII 0.9915 +/- 0.0006 XIII 0.9945 +/- 0.0006 XIV 0.9902 +/- *o.0006 xv 0.9836 +/- 0.0006 XVI 0.9863 +/- 0.0006 XVII 0.9875 +/- 0.0006

• • 'XVIII 0.9880 +/- 0.0006 XIX 0.9882 +/- 0.0005 xx 0.9885 +/- 0.0006 XXI 0.9890 +/- 0.0006 Mean 0.9899 +/- 0. 0007<1>

Bias (95%/95%) 0.0101 +/- 0.0018 (1)

Standard Deviation of the Mean, calculated from the kcff values *

• A - 8

Table 2 RESULTS OF 27-GROUP (SCALE) NITAWL-KEN05a CALCULATIONS OF FRENCH and BNWL CRITICAL EXPERIMENTS French Experiments Separation critical Calculated Distance, cm Height, cm kctf 0 23.8 1.0302 +/- 0.0008 2.5 24.48 1.0278 +/- 0.0007 5.0 31.47 1.0168 +/- 0.0007 10.0 64.34 0.9998 +/- 0.0007 BNWL_Experiments Calculated Case Expt. No. kcff No Absorber 004/032 0.9942 +/- 0.0007 SS Plates (1..05 B) 009 0.9946 +/- 0.0007 SS Plates ( l.. 62 B) 01.l. 0.9979 +/- 0.0007 SS Plates ( l.. 62 B} 012 0.9968 +/- 0.0007 SS Plates 013 0.9956 +/- 0.0007 SS Plates 014 0.9967 +/- 0.0007 Zr Plates 030 0.9955 +/- 0.0007 Mean 0.9959 +/- 0.001.3

• A 9

Table 3 Intercomparison of NITAWL-KENOSa (Interpolated) and CASM03 Calculations at Various Temperatures Temperature CASMO 3 W-N-KEN05a<->

4°C 1. 2276 1.2345 +/- 0.0014 17.5°C 1.2322 1.2328 +/- 0.0015 25°C 1. 234 7 1.2360 +/- 0.0013 50°C 1. 2432 1.2475 +/- 0.0014 75°C 1. 2519 1.2569 +/- 0.0015 1.2 0 °C 1. 2701 1.2746 +/- 0.0014

• corrected for bias A - 10

Table 4 Reactivity Calculations for Close-Packed Critical Experiments Cale. BAW Pin Module Boron Calculated No. Expt. Pitch Spacing Cone. kcff No. cm cm ppm KS01 2SOO Square 1.792 . 11S6 0.9891 +/- a.coos

1. 4097 KS02 2SOS Square 1. 792 1068 0.9910 +/- a.coos

1. 4097 KS1 248S Square 1.778 886 0.984S +/- o.ooos Touching KS2 2491 Square 1. 778 746 0.9849 +/- o.ooos Touching KT1 24S2 Triang. 1.86 43S 0.984S +/- 0.0006 Touching KT1A 24S7 Triang. 1.86 33S 0.986S +/- 0.0006 Touching KT2 2464 Triang. 2.62 361 0.9827 +/- 0.0006 Touching KT3 2472 Triang. 3.39 121 1. 0034 +/- 0.0006 Touching A 11

Table 5 RESULTS OF CASM03 AND NITAWL-KEN05a BENCHMARK (INTERCOMPARISON) CALCULATIONS EnrichmentCl> koo Wt. % U-235 NITAWL-KEN05aa> CASM03 lokl f\

2.5 0.8376 +/- 0.0010 0.8386 0.0010 3.0 0.8773 +/- 0.0010 0.8783 0.0010 3.5 0.9106 +/- 0.0010 0.9097 0.0009 4.0 0.9367 +/- 0.0011 0.9352 0.0015 4.5 0.9563" +/- 0.0011 0.9565 0.0002 5.0 0~*9744 +/- 0.0011 0.9746 0.0002 Mean 0.0008 Expt. No. C3>

XIII 1.1021 +/- 0.0009 1.1008 0.0013 XIV 1.0997 "+/- 0.0008 1.1011 0~0014 xv 1.1086 +/- 0.0008 1.1087 0.0001 XVII 1.1158 +/- 0.0007 1.1168 0.001.0 XIX 1.1215 +/- 0.0007 1.1237 0.0022 Mean 0.001.2 (1)

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

(2) kao from NITAWL-KEN05a corrected for bias.

(3) central Cell from BAW Critical Experiments A 12

I

1. 26

..,,CD

~

c:

~

'I- 1. 25 c:

.,J

..w:

1.22~~.--.-+-..-..,....................+--.--.-.-......+-...............,...........-+-.--.--.--i---.-.-........+--.-.-.............~

0 20 40 60 80 100 120 140 Temperature, Degrees C Ftg. 1 COMPARISON OF CASM0-3 end KEN05c TEMPERATURE DEPENDENCE A - 13

w 1-H zH

~ 0.90-+-~~-+-~~-+--~+-~~--1~~-+~~-+-~~-1 H

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CASM KENO Sci 0.80--+-.-.-..--.-+-.--.-.-.-1-T--..-.---+--.-.--.-.-+-._-........-~-.-...........-i--.--.......~

2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 FUEL ENRICHMENT, WT~ U-235 Ftg. 2 COMPARISON OF CASMO AND KENO-Sa CALCULATIONS AT VARIOUS ENRICHMENTS IN REPRESENTATIVE FUEL STORAGE RACK A - 14

5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5.1 Introduction This section provides a summary of the methods, models, analyses and numerical results to demonstrate the compliance of the reracked Salem spent fuel pool cooling systems with the provisions of Section I I I of the USNRC "OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications", (April 14, 1978).

Similar methods of thermal-hydraulic analysis have been used in other rerack licensing projects (see Table 5.1.1).

The thermal-hydraulic qualification analyses for the rack array may be broken down into the following categories:

( i) Evaluation of the pool decay heat load and pool bulk temperature as a function of time.

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

(iii) Evaluation of the maximum fuel cladding temperature to establish that bulk nucleate boiling at any location around the fuel is not possible with cooling available.

(iv) Evaluation of the time-to-boil if all forced heat rejection paths from the pool are lost.

(v) Compute the effect of a blocked fuel cell opening on the local water and maximum cladding temperature.

The following sections present the plant system description, analysis assumptions, a synopsis of the analysis methods employed and final results.

5-1

5.2 Spent Fuel Cooling and Cleanup System Description

• 5.2.1 Design Bases The Spent Fuel Pool Cooling System is designed to remove the heat generated by stored spent fuel elements. The system serves the spent fuel pool located in the Fuel Handling Building adjacent to the Containment Building. A secondary function is to clarify and purify spent fuel pool, transfer pool, and refueling water. The system design considers the possibility that during the life of the plant it will become necessary to totally unload a reactor at the time when spent fuel is in the fuel pool.

The system design incorporates redundant active components. System piping is arranged so that failure of any pipe line does not drain the spent fuel pool below the top of the stored fuel elements.

The spent fuel pool water is limited to 180°F with one pump in operation in the unloading of a full core. Boron concentration in the pool fluid is maintained at a minimum of 2000 ppm.

5.2.2 System Description The Spent Fuel Pool Cooling System consists of three subsystems:

the Cooling System, the Purification System, and the Skimmer System. A simplified system flow diagram is shown on Figure s.2.1.

Austenitic stainless steel piping is used in the Spent Fuel Pool Cooling System. The cooling loop consists of the spent fuel pool pumps and the SFP heat exchanger. The purification loop consists of the SFP pump, the SFP filter, the SFP demineralizer, the refueling water purification pump, and the refueling water purification filter. The skimmer loop consists of the skimmer pump, strainer, and filter.

5-2.

I i

During the heat removal operation, fuel pool water is pumped through the tubeside of the heat exchanger and is returned to the pool. The suction line *is located at an elevation 4 feet below the pool normal water level. The return line terminates in the pool at an elevation approximately 6 feet above the top of the fuel assemblies. If the SFP pump fails, the second pump supplies 100%

backup.

The SFP Cooling system has its maximum duty during the refueling operation when the decay heat from the spent fuel is the highest.

The system is normally placed in operation prior to the transfer of any fuel and is continued in operation as long as required to maintain temperature at the required level and water purity.

Installed piping and valves allow the Units 1 and 2 heat exchangers to be cross connected. During normal plant operation, the heat exchangers operate independently to meet the cooling requirements of the individual units. The cross connect also allows one heat

• exchanger to be used to alternatively cool the spent fuel pools in both units during times when one heat exchanger is out for maintenance.

While the heat removal operation is in progress, a portion *of the SFP water, 100 gpm, may be diverted through the spent fuel pool demineralizer and SFP filter to maintain pool water clarity and purity. This purification loop is sufficient for removing fission products and other contaminants which may be introduced if a leaking fuel assembly is transferred to the SFP.

The demineralizer may be isolated, by manual valves, from the heat removal portion of the SFP Cooling System. This allows the refueling water purification filter to clean and purify the refueling water while spent fuel pool heat removal operations 5-3

proceed. Connections are provided to the isolated loop such that

• the refueling water may be pumped from either the refueling water storage tank (RWST) or the refueling cavity, through the demineralizer and filter, and discharged to either the refueling cavity or the RWST.

To further assist in maintaining spent fuel pool water clarity, the water surface is cleaned by a skimmer loop. This system consists of two skimmers, a skimmer pump, a strainer and a filter. Water is removed from the surf ace by the skimmer, pumped through the strainer and filter, and returned to the pool surface at three locations remote to the skimmers.

The SFP is initially filled with water at the same boron concentrations as the RWST. Borated water may be supplied from the RWST via the refueling water purification pump connection (rated at 100 gpm at 125 ft TDH), or from the boric acid blender in the eves.

Demineralized water is also added to the pool for makeup by a connection in the recirculation return line.

5.2.3 Components Spent Fuel Pool Heat Exchanger The SFP heat exchanger is a shell and u-tube type with the tubes welded to the tubesheet. Component Cooling water circulates through the shell, and SFP water circulates through the tubes. The tubes are austenitic stainless steel and the shell is carbon steel.

Spent Fuel Pool Pumps The SFP pumps circulate water in the SFP Cooling system. All wetted surfaces of the pumps are austenitic stainless steel, or equivalent corrosion resistant material. The pumps are operated manually from a local station.

5-4 j

Spent Fuel Pool Filter The SFP Filter removes particulate matter larger than 5 microns from the SFP water. The filter cartridge is of synthetic fiber and the vessel shell is austenitic stainless steel.

Spent Fuel Pool Strainer A stainless steel strainer is located at the inlet of the spent fuel pool cooling suction line for removal of re lati vely large particles which might otherwise clog the spent fuel pool demineralizer.

Spent Fuel Pool Demineralizer The demineralizer is sized to pass 100 gpm of the loop circulation flow to provide adequate purification of the fuel pool water for unrestricted access to the working area and to maintain optical clarity.

Refueling Water Purification Pump The refueling water purification pump circulates water in a loop between the RWST, the spent fuel pool demineralizer, and the refueling water purification filter. All wetted surfaces of the pump are austeni tic stainless steel. The pump is operated manually from a local station.

Refueling Water Purification Filter The refueling water purification filter removes particulate matter larger than 5 microns from the refueling water purification flow.

Spent Fuel Pool Cooling System Valves Manual stop valves are used to isolate equipment and lines,

• and manual throttle valves provide flow control. Valves in contact with spent fuel pool water are austeni tic stainless steel or equivalent corrosion resistant material.

Spent Fuel Pool Cooling System Piping All piping in contact with spent fuel pool water is austenitic stainless steel. The piping is welded except where flanged connections are used to facilitate maintenance.

5-5

Spent Fuel Pool Skimmers Two spent fuel pool skimmers are provided to remove water from the surface of the spent fuel pool. The skimmer heads are manually positioned to take water from any elevation from the water surface to 4 inches below the surface. The elevation of the skimmers head can be manually adjusted over a total range of 2 feet.

Spent Fuel Pool Skimmer Pump The spent fuel pool skimmer pump circulates surface water through a strainer, a filter, and returns it to the pool.

Spent Fuel Pool Skimmer Strainer The spent fuel pool skimmer strainer i$ designed to remove debris from the skimmer process flow.

Spent Fuel Pool Skimmer Filter The spent fuel pool skimmer filter is designed to remove insoluble particles which are not removed by the strainer.

5.2.4 Design Evaluation The most serious failure of this system would be complete loss of water in the spent fuel pool. To protect against this possibility, the spent fuel pool cooling suction connection enters near the normal water level so that the pool cannot be gravity-drained. The cooling water return lines contain anti-siphon holes to prevent the possibility of gravity draining the pool. There are no drains or permanently connected systems to the SFP (seismic class I) which, in the event of failure, could cause loss of coolant from the pool that would uncover the fuel. Also, provisions have been made to supply makeup to the spent fuel pool as noted below.

Water loss from the spent fuel pool due to the accidental opening of a sluice gate when the transfer pool is empty will not occur due to the redundancy in the sluice gates. Two sluice gates separate the spent fuel pool from the transfer pool.

5-6

The normal source of makeup water to the spent fuel pool is the Demineralized water System which distributes water from two 500,000 gallon demineralized water storage tanks. The tanks and the distribution system do_not have seismic classification. Makeup is also available from the Primary Water Storage Tank via the makeup pumps (seismic class II) and from the eves holdup tanks via the holdup tank recirculation pump (seismic class II) rated at 500 gpm.

Valves have been installed on the existing 6 inch spare nozzles on both RWSTs (contained volume is 364, 500-400, 000 gallons each).

These tanks are Seismic Class I. A portable pump, with appropriate suction and discharge connections and hose are provided with the capability to deliver approximately 100 gpm makeup water flow from one of the RWSTs directly to the spent fuel pool. The valves installed on the RWSTs are locked, closed and capped, and* under administrative control. The portable pump and hose are also under administrative control to ensure constant and timely availability.

If a leaking fuel assembly is stored in the spent fuel pool, a small quantity of fission products may enter the cooling water.

Fission products and other contaminants are removed by the* spent fuel pool purification loop.

5.3 Decay Heat Load Calculations The decay heat load calculation is conservatively perf armed in accordance with the provisions of USNRC Branch Technical Position ASB9-2, "Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July, 1981.

5-7

All calculations are performed at the point in time when the spent fuel pool has accumulated maximum possible inventory. Tables 5. 3

• 1 and 5.3.2 show the past discharge chronology and future projected discharges in Unit 1 and Unit 2 pools, respectively. For conservatism, all calculations are performed assuming that the pool has accumulated maximum possible fuel inventory. For the Unit 1 pool, Table 5. 3 .1 indicates that the pool will have 1592 fuel assemblies in Cycle 24, which will leave the pool with insufficient capacity to accept a normal discharge batch (68 bundles). The background decay heat load is calculated with reference to the inventory of 1592 bundles. Since the decay heat load from the old assemblies varies very slowly as a function of time, it is assumed to remain constant for the duration when the discharge scenarios described in Section 5.4 are considered. Although the projected discharge patterns into the two pools are slightly different, calculations showed slightly high "background" heat load (decay heat at the time of final discharges) from Unit 1. Therefore, the case of Unit 1 pool, presented in depth in this section, will bound the results for Unit 2 pool.

It is noted that assuming 1592 previously stored fuel assemblies in the pool followed by a normal batch (68 assemblies) and a full core discharge (193 assemblies) implies a cumulative pool capacity of 1853 assemblies. Since the installed rack capacity in each pool is 1632 locations, this implies that the assumed heat loads are predicated on a larger fuel inventory than can be accommodated in the pool. Alternatively, the assumption of the gross count of 1853 assemblies implies that 221 locations (1853 minus 1632) hold consolidated (with 2:1 consolidation ratio) canisters. The assumption of a higher cell count than requested in this licensing application is also followed in the pool structure evaluation wherein (Section 8) a slightly different level of conservatism in the cell count has been utilized *

• 5-8

5.4 Discharge Scenarios Three discharge scenarios are considered in cycle 25. The first scenario deals with the limiting case of normal discharge into the pool. The remaining two pertain to full core off load into the pool.

Case 1: Normal Discharge The reactor is shutdown and is cooled for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (in-core).

Then, a batch of 88 assemblies with 1642. 5 days full power operation is discharged to the pool at a rate of 7 assemblies per hour (see Figure 5.4.1).

Case 2: EOL Full Core Offload At the scheduled refueling discharge of Case 1, the full core of 193 assemblies begins to transfer to the pool 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> after-reactor-shutdown. All assemblies in the core are assumed to have 1642.5 days full power operation in the core. The fuel transfer rate is 7 assemblies per hour (see Figure 5.4.2).

Case 3: Emergency Full Core Offload The rea,ctor is back to operation in Cycle 25 after the normal refueling shutdown for 45 days (Cycle 24) . In the Cycle 24 refueling outage, 68 assemblies are assumed to be discharged into the pool. Thirty days later, the reactor experiences an unplanned shutdown. The full core of 193 assemblies is transferred to the pool 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> after-reactor-shutdown. sixty eight assemblies in the core are assumed to have 30 days full power operation and 125 assemblies are assumed to have 1642.5 days full power operation (see Figure 5.4.3).

The background heat load from the previously stored fuel is computed using the data summarized in Table 5. 3 .1. It is recognized that the future discharge dates may vary from those assumed in Table 5.3.1. However, the effect of such variations on the heat load and pool water temperature profile is small and has infinitesimal effect on the results of the analysis *

• 5-9

Key common data for all cases may be found in Tables 5.4.1 and 5.4.2.

5.5 Bulk Pool Temperatures In this section, we present the methodology for calculating the bulk pool temperature as a function of the time coordinate. The method used to calculate the rate of pool water temperature rise and the time-to-boil, when all forced cooling paths are unavailable, is also presented.

In order to perform the conservative analysis, the heat exchangers are assumed to be fouled to their design maximum (0.0005 shellside and 0.000575 tubeside). Thus, the temperature effectiveness, p, for the heat exchanger utilized in the analysis is the lowest postulated value calculated from heat exchanger thermal-hydraulic codes. The temperature effectiveness p is assumed to remain constant for the duration of the bulk temperature evaluation.

The mathematical formulation can be explained with reference to the simplified heat exchanger alignment of Figure 5.5.1.

Ref erring to the spent fuel pool cooling system, the governing differential equation can be written by utilizing conservation of energy:

dT C- = QL-QHX (5.5.1) ck 5-10

where:

c: Thermal capacitance of the pool (net water volume times water density and times heat capacity), Btu/°F.

Heat load to the heat exchanger, Btu/hr.

Q (T) : Heat generation rate from recently discharged fuel, which is a specified function of time, T, Btu/hr.

Peons = B Po: Heat generation rate from "old" fuel, Btu/hr. It is also termed as "background" heat load. (P 0 =* average assembly operating power, Btu/hr.)

Heat removal rate by the heat exchanger, Btu/hr.

Heat loss to the surroundings, which is a function of pool temperature T and ambient temperature ta, Btu/hr.

QHx is a non-linear function of time if we assume the temperature effectiveness p is constant during the calculation. QHX can, however, be written in terms of effectiveness p as follows:

aHX = wt ct p(T-t) (5.5.2) where:

wt: coolant flow rate, lb./hr.

~: Coolant specific heat, Btu/lb. 0 P.

p: Temperature effectiveness of heat exchanger.

T: Pool water temperature, 0 P

ti: Coolant inlet temperature, 0 p

t 0

Coolant outlet temperature, 0 P

s-11

The temperature effectiveness p is obtained by rating the heat exchanger on a Holtec proprietary thermal-hydraulic computer code.

Q(T) 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(T) is a function of decay time, number of assemblies, and in-core exposure time. During the fuel transfer, the heat load in the pool will increase With respect to the rate Of fuel transfer and equals Q(T) after the fuel transfer.

QEV is a non-linear function of pool temperature and ambient temperature. QEv contains the heat evaporation loss through the pool surface, natural convection from the pool surface and heat conduction through the pool walls and slab. Experiments show that the heat conduction takes only about 4% of the total heat loss (5.5.1], and can therefore be conservatively neglected. The evaporation heat loss and natural convection heat loss can be expressed as:

{5.5.3) where:

m: Mass evaporation rate, lb./hr. ft. 2 r: Latent heat of pool water, Btu/lb.

Pool surface area, ft. 2 Convection heat transfer coefficient at pool surface, Btu/ft. 2 hr. 0 p e - T-ta: The temperature difference between pool water and ambient air, op 5-12

The mass evaporation rate m can be obtained as a non-linear function of 9. We, therefore, have m = h0 (6) (Wps-Was) (5.5.4) where:

Humidity ratio of saturated moist air at pool water surface temperature T.

Humidity ratio of saturated moist air at ambient temperature ta ho (9) : Diffusion coefficient at pool water surface. ho is a non-linear function of e, lb./hr. ft. 2 °F The non-linear single order differential equation (5.5.1) is solved using Holtec 1 s Q.A. validated numerical integration code 11 0NEPOOL 11

• The next step in the analysis is to determine the time-to-boil if all forced cooling paths become unavailable.

Clearly, the most critical instant of loss-of-cooling is when pool water temperature has reached its maximum value. It is assumed that makeup water is added at the rate of G lb./hr. The makeup water is at temperature,

• tcooi

• The governing enthalpy balance equation for this condition can be written as dT *

[C + G (CJ ('t--rJ] - . =Peons+ Q ('t+'tm) + G (CJ (tcoo1-n (5.5.5) ch .. .

where water is assumed to have specific heat of unity, and the time coordinate T" is measured from the instant maximum pool water temperature is reached. T 0 is the time coordinate when the makeup water application is begun. T" ins is the time coordinate measured from the instant of reactor shutdown to when maximum pool water temperature is reached. T is the dependent variable (pool water temperature). For conservatism, Q~ is assumed to remain constant after pool water temperature reaches and rises above 170°F.

5-13

A Q.A. validated numerical quadrature code is used to integrate the

• foregoing equation. The pool water heat up rate, time-to-boil, and subsequent water evaporation-time profile are generated and compiled for safety evaluation.

5.6 Local Pool Water Temperature In this section, a summary of the methodology for evaluating the local pool water temperature is presented.

5.6.1 Basis The local water temperature analysis uses the bulk pool temperature as the datum and establishes the maximum incremental water temperature which may exist adjacent to the most heat emissive fuel assembly in the pool.

In order to determine an upper bound on the maximum local water 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 (T 8 ) . Since the heat emission falls off rapidly with increasing T 8 , it is conservative to assume that all fuel assemblies are from the latest batch discharged simultaneously in the shortest possible time and they all have had the maximum postulated years of operating time in the reactor. The heat emission rate of each fuel assembly is assumed to be equal and maximum.

As shown in the pool layout drawings, the modules occupy an irregular floor space in the pool. For the hydrothermal analysis, a circle circumscribing the actual rack floor space is drawn (Figure 5.6.1). It is further assumed that the cylinder with this circle as its base is packed with fuel assemblies at the nominal layout pitch.

5-14

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 (Figures 5.6.2 and 5.6.3) (i.e., minimum gap between the pool wall and rack module) .

No downcomer flow is assumed to exist between the rack modules.

5.6.2 Model Description Using the assumptions of the foregoing section'- 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 5.6.2 '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

• : (Ref. 5. 6. 1)

• Wherever discrepancies in reported values exist, the conservative values are consistently used. Reference 5.6.2 gives the details of mathematical analysis used in this solution process.

After 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.

5-15

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 temper~ture is obtained. In view of these aforementioned assumptions, the temperatures calculated in this manner overestimate the temperature rise that will actually occur in the pool. Holtec' s proprietary computer code THERPOOL, based on the theory of Ref. 5. 6. 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. The *major input data are given in Table 5.6.1.

5.7 Cladding Temperature In this section, the method to calculate the temperature of the fuel cladding is presented.

The maximum specific power or a fuel array ~ can be given by:

(5.7.1)

• THERPOOL has been used in qualifying the spent fuel pools for Enrico Fermi Unit 2 (1980), Quad Cities 1 and 2 (1981), Oyster Creek (1984), Virgil c. Summer (1984), Rancho Seco (1983), Grand Gulf Unit 1 (1985), Diablo Canyon 1 and 2 (1986), st. Lucie Unit One (1988), J.A. FitzPatrick (1991), Three Mile Island Unit One (1992), among others.

5-16

where:

Fxy = radial peaking factor q = average fuel assembly specific power The peaking factors are given in Table 5.7.1. The maximum temperature rise of pool water in the most disadvantageously placed fuel assembly, defined as one which is subject to the highest local pool water temperature, 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 Fz times the average heat emission rate over a small length, where Fz 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

" (a + x~

q(x) = ~ sin (5.7.2) t + 2a where:

t: active fuel length a: chopped length at both extremities in the power curve x: axial coordinate with origin at the bottom of the active fuel region The value of a is given by t z a =

1 - 2Z 5~11

where:

1 1 1 + __g_ ]112 Z=-F.--( 2~ 1t2 1t z 1t rz where Fz is the axial peaking factor.

The cladding temperature Tc is governed by a third order differential equation which has the form of d3T +<<1 d2T - <<2 dT =f (x) dx3 dx2 dx where a1, 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 with .005 °F-sq.ft.-hr/Btu thermal resistance, which covers the entire surface.

5.8 Results This section contains results culled from the analyses performed for each of the three postulated discharge scenarios.

5.8.1 Bulk Pool Temperature The net heat load* in the spent fuel pool and the bulk pool temperature profile as a function of time (after-reactor~shutdown) are shown in Figures 5.8.l{a-b) to 5.8.3{a-b) for the three discharge scenarios postulated in Section 5. 4. As would be

• The net heat load is defined as the heat generation rate less the heat lost to the ambient through evaporation.

5-18 I

L

expected from physical considerations, the thermal inertia of the

• pool water causes the bulk pool temperature to reach its maximum value within a short time after the occurrence of the peak decay heat load, the lag is a direct result of the system thermal capacitance. The maximum values of the pool bulk temperature and the coincident heat load to the spent fuel pool cooler are summarized in Table 5.8.1. The fact that the coincident heat load is not the maximum decay heat load (due to the aforementioned system inertia) warrants emphasis; since this distinction is often neglected in evaluation of the system performance data. The design basis temperature of the purification loop is 140°F which is governed by the integrity of the resins. The Salem plant procedures limit the purification loop temperature to 130°F with alarm set point at 125°F. The purification loop will be manually isolated whenever the spent fuel pool temperature reaches 125°F.

The maximum bulk temperatures tabulated in Table 5.8.1 demonstrate the compliance of the Salem cooling system with the latest USNRC acceptance criteria. Therefore, no physical modification of the spent fuel pool cooling and purification system is necessary.

An assessment of the baseline design basis HVAC system of the Fuel Handling Building indicates that the incremental psychromatric load on the HVAC system

• is very small and no upgrading of the HVAC system is necessary*due to the proposed capacity expansion.

5.8.2 Time-to-Boil If all heat exchanger assisted forced pool cooling becomes unavailable, then the pool water will begin to rise in temperature and eventually will reach the bulk boiling temperature, nominally 212°Fo The time to reach the boiling point will be the shortest when the loss of forced cooling occurs at the point in time when the pool bulk temperature is at its maximum calculated value.

Although the probability of the loss-of-cooling event coinciding

• 5-19

with the instant when the pool water has reached its peak value is

• extremely remote, the calculations are performed under this extremely unlikely scenario. Table 5. 8. 2 contains the results with the additional proviso that no makeup water was added to the pool.

The time-to-boil results for Salem are comparable to other PWR pools with densified fuel storage.

5.8.3 Local Water and Cladding Temperature Consistent with our approach to make the most pessimistic assessments of temperature, the local water temperature calculations are performed when the pool is at its peak bulk temperature. Thus, the local water temperature evaluation is a calculation of the temperature increment over the theoretical.

spatially uniform value due to local hot spots (due to the presence of a highly heat emissive fuel bundle).

The maximum local water temperature for the limiting case (Case 2) is calculated to be 236.5°F and the maximum local fuel cladding temperature is 269.8°F. If the limiting cells are 50% blocked on the top, the maximum local water temperature becomes 237.6°F and the maximum fuel cladding temperature is 270.7°F (see Table 5.8.3).

While no nucleate boiling is indicated for the standard storage condition, assuming 50% .cell blockage indicates that a highly localized two phase condition near the top of the fuel may exist.

The corresponding thermal stress in the fuel cladding is, however, less than 7000 psi, which is considerably lower than the endurance limit of the cladding material. It is therefore concluded that the reracked Salem fuel pools comply with all thermal-hydraulic regulatory criteria *

• 5-20

5.9 References

• (5.5.1] Wang, Yu, "Heat Loss to the Ambient From Spent Fuel Pools: Correlation of Theory* with Experiment", Holtec Report HI-90477, Rev. o, April 3, 1990.

(S.6.1] General Electric Corporation, R&D Data Books, "Heat Transfer and Fluid Flow", 1974 and updates.

(5.6.2] Singh, K.P. et al., "Method for Computing the M~ximum Water Temperature in a Fuel Pool Containing Spent Nuclear Fuel", Heat Transfer Engineering, Vol. 7, No. 1-2, pp. 72-82 (1986)

• 5-21

Table 5.1.1 PARTIAL LISTING OF RERACK APPLICATIONS USING SIMILAR METHODS OF THERMAL-HYDRAULIC ANALYSIS PLANT DOCKET NO.

Enrico Fermi Unit 2 USNRC 50-341 Quad Cities 1 and 2 USNRC 50-254, 50-265 Rancho Seco USNRC 50-312 Grand Gulf Unit 1 USNRC 50-416 Oyster Creek USNRC 50-219 Pilgrim USNRC 50-293 v.c. Summer USNRC 50-395 Diablo Canyon Units 1 and 2 USNRC 50-275, 50-323 Byron Units 1 & 2 USNRC 50-454, 50-455 Braidwood Units 1 & 2 USNRC 50-456, 50-457 Vogtle *unit 2 USNRC 50-425 St. 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. FitzPatrick USNRC 50-333 Shearon Harris Unit 2 USNRC 50-401 Hope Creek USNRC 50-354 Kuosheng Units 1 & 2 Taiwan Power Company Chin Shan Units 1 & 2 Taiwan Power Company

Table 5.1.1 (continued)

PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK PLANT DOCKET NO.

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 La Salle Unit One USNRC 50-373 Duane Arnold USNRC 50-331 Fort Calhoun USNRC 50-285 Nine Mile Point Unit One USNRC 50-220 Beaver Valley Unit One USNRC 50-334

Table 5.3.1 FUEL DISCHARGE SCENARIO: SALEM 1 TOTAL MONTHS ASSEMB- NUMBER OF CYCLE START END AFTER PRE- LIES IN ASSEMBLIES NUMBER DATE DATE DISCHARGE THE POOL DISCHARGED 1 12/18/76 04/03/79 0 40 40 2 12/27/79 09/19/80 17.5 104 64 3 12/25/80 01/01/82 15.5 160 56 4 04/19/82 10/15/82 9.5 212 52 5 05/23/83 02/24/84 16.5 296 84 6 10/20/84 03/21/86 13.0 380 84 7 05/06/86 10/02/87 18.0 464 84 8 02/28/88 03/23/89 18.0 540 76 9 06/14/89 02/08/91 22.0 588 48 10 04/11/91 04/03/92 14.0 656 68 11 08/14/92 10/01/93 18.0 716 60 12 12/13/93 03/24/95 18.0 784 68 13 05/09/95 09/20/96 18.0 848 64 4 11/25/96 03/27/98 18.0 916 68 05/12/98 09/24/99 18.0 980 64 6 11/14/99 03/23/01 18.0 1048 68 17 05/02/01 09/06/02 18.0 1116 68 18 11/09/02 03/17/04 18.0 1184 68 19 05/02/04 09/06/05 18.0 1252 68 20 11/09/05 03/17/07 18.0 1320 68 21 05/02/07 09/06/08 18.0 1388 68 22 11/09/08 03/17/10 18.0 1456 68 23 05/02/10 09/06/11 18.0 1524 68 24 11/09/11 03/17/13 18.0 1592 68 25 05/02/13 09/06/14 18.0 68,88 or 193

Table 5.3.2 FUEL DISCHARGE SCENARIO: SALEM 2 TOTAL MONTHS ASSEMB- NUMBER OF CYCLE START END AFTER PRE- LIES IN ASSEMBLIES NUMBER DATE DATE DISCHARGE THE POOL DISCHARGED 1 09/01/81 01/21/83 0 72 72 2 07/30/83 10/04/84 20.0 140 68 3 04/10/85 10/02/86 24.0 224 84 4 12/20/86 08/31/88 23.0 308 84 5 11/28/88 03/30/90 19.0 384 76 6 06/15/90 11/09/91 19.0 436 52 7 05/09/92 03/26/93 17.0 500 64 8 05/21/93 09/16/94 18.5 564 64 9 11/06/94 03/15/96 18.0 632 68 10 04/30/96 09/12/97 18.0 696 64 11 11/02/97 03/12/99 18.0 764 68 12 04/27/99 09/15/00 18.0 828 64 13 11/20/00 03/15/02 18.0 896 68 14 04/30/02 09/12/03 18.0 960 64 15 11/01/03 03/15/05 18.0 1028 68 04/30/05 09/12/06 18.0 1092 64 11/01/06 03/15/08 18.0 1160 68 04/30/08 09/12/09 18.0 1224 64 19 11/01/09 03/15/11 18.0 1292 68 20 04/30/11 09/03/12 18.0 1356 64 21 11/01/12 03/15/14 18.0 1424 68 22 04/30/14 09/03/15 18.0 1488 64 23 11/01/15 03/15/17 18.0 1556 68 24 04/30/17 09/03/18 18.0 1620 64,88,193

Table 5.4.1 DATA FOR DISCHARGE SCENARIOS Number of assemblies in refueling batch: 68 (88 maximum)

Number of assemblies in full core: 193 Number of fuel pool coolers in the SFPCS: 1 Fuel normal exposure time, hrs.: 39420 Fuel transfer rate, assemblies per hr.: 7 Time after reactor shutdown before the 168 fuel transfer, hrs

• Table 5.4.2 FUEL SPECIFIC POWER AND POOL CAPACITY DATA Net water volume of pool, gal. 239,000 Fuel pool thermal capacity, 10 6 Btu/°F 1.96 Average operating power of a fuel assembly, 10 6 Btu/hr: 63.69 SFP cooler coolant inlet temperature, °F: 99 SFP Cooler coolant flow rate, 10 6 lb/hr: 1.49

Table 5.6.1 DATA FOR LOCAL TEMPERATURE EVALUATION Type of fuel assembly PWR 17x17 Fuel cladding outer diameter, inches 0.36 Fuel cladding inside diameter, inches 0.31 storage cell inside dimension, inches 8.86 Active fuel length, inches 144 Number of fuel rods/assembly 264 Operating power per fuel assembly, P 0 , 63.69 10 6 Btu/hr Cell pitch, inches 9.052 Cell height, inches 168.5 Bottom plenum height, inches 16 Plenum radius, feet (see Fig. 5.6.1) 23.352 Peripheral average rack-to-wall 2.396 gap, inches

• Table 5. 7. 1 PEAKING FACTORS FACTOR VALUE Radial 1.70 Total 2.45

Table 5.8.1 SFP BULK POOL TEMPERATURE Coincident. Coincident Coincident Maximum Time . After Heat Load Evaporation Pool Reactor Shut- to6 SFP Cooler, ~t Iosses, Temp., OF doWn, hrs. 10 Btu/hr 10 Btu/hr Discharge 148.94 195 23.80 0.86 in Case 1 Discharge 179.93 205 38.57 3.87 in Case 2 Discharge 178.79 204 38.07 3.64 in Case 3

Table 5.8.2 RESULTS OF LOSS-OF-COOLING Time-to-Boil (Hours) Maximum Case (Without Evaporation Number Makeup Water) Rate (GPM)

1. 4.61. 52.64

• 2 3

1..28 1.38 92.27 89.38 Note: The maximum available makeup from the RWST is 1.00 gpm (see Section 5.2.4).

i

Table 5.8.3 MAXIMUM LOCAL POOL WATER AND FUEL CLADDING TEMPERATURE FOR THE LIMITING CASE (Case 2)

Maximum Local Maximum Local Pool Water Fuel Cladding Temp. °F Temp., °F No Blockage 236.5 269.8 50% Blockage 237.6 270.7

SPEHT n.EJ..

PIT SKllAliR SPENT P\a PrT IE/-T ElCOW<<Et ITlllZ lil'll !" liTRAltel Sl'Et i;FUfT l'\IEl. PlT

&Ila.El ,.._

5RIQ C'tlCS HIT AECIRC PU.P SKllASIS lil'U .. 5rEB 6P£NT nE.. PIT REJ'lELJNJ WA~

l'Ulll'ICATION Pl.IP L-~S'ZZ-F11-~~,~--1..~.~11~.~~--

lin:I axr~ caJ~

_ I ..

DUTSllJE INS!

llD aNrAINENT i;a.fl"AJ~

FIGURE 5.2.1 SPENT FUEL COOLING SYSTEM SIMPLIFIED FLOW DIAGRAM

HOLTEC INTERNATIONAL NORMAL DISCHARGE

....J 0

~ POOL FILLED WITH 1592 FUEL BUNDLES w

j!:

~ 168 HOURS fl:

~

z

~ 88 FAS DISCHARGED AT 7 FAS/HR

....J w

u.

REACTOR SHUTDOWN TIME FIGURE 5.4.1 SALEM UNITS 1 AND 2 SPENT FUEL, POOL DISCHARGE SCENARIO CASE 1

HOLTEC INTERNATIONAL END OF CYCLE (EOC) FULL CORE OFFLOAD

_J 0

2 w

POOL FILLED WITH 1592 FUEL BUNDLES

~

z 168 HOURS fl:

~

z

~

_J 193 FAS DISCHARGED AT 7 FAS/HR w

J LL REACTOR SHUTDOWN TIME FIGURE 5.4.2 SALEM UNITS 1 AND 2 SPENT FUEL POOL DISCHARGE SCENARIO CASE 2

HOLTEC INTERNATIONAL BEGINNING OF CYCLE (BOC) FULL CORE OFFLOAD a

0 NORMAL DISCHARGE ONE FULL CORE OFFLOAD n.

w i!:

~

fi: REACTOR

~ 68 FAS DISCHARGED AT 7 FAS/HR OPERATING .

~

FULL CORE OFFLOAD 193 FAS AT 7 FAS/HR 45DAYS 30DAYS REACTOR SHUTDOWN END OF OUTAGE REACTOR SHUTDOWN FIGURE 5.4.3 SALEM UNITS 1 AND 2 SPENT FUEL POOL DISCHARGE SCENARIO CASE 3

• EVAPORATIVE LOSS SPENT FUEL POOL HEAT EXCHANGER p

COOLANT ti Wt FIGURE 50501 SPENT FUEL POOL COOLING MOIDEL

IDEALIZED OUTLINE OF RACK ASSEMBLY

• • -- -*- _,_ 1 - - -- - - -f- -- - - - f - -

-- --- IA I/ IA!>'

-- 1 ~

j

, ,")I" ,

,'/. liT-

• • • • -* ... **101 ** kl fl!_ I 1

(

'" ON Ill ~o E s I ~G N I

- -f-II ,"/ I 11.

II Jr I

~ II

.,'n<- - - -- - - -*- 1 -l !;J -- - - * *- -

o-=-1~ 1=-=-' -=-oJ qECIO I I

__! ~/_()_ _l .!.. - - -- -- -- -

,; !1 l!fd II - -*

10 GISIN,.

I J,X;lllN ~I I I I I mm CLLLLLLLI I I fil,f

( fx: ON I I

__ -- II G ON II I Jl

-- rn CJ ~NII -- II~

I I I I I I ACTUAL OUTLINE OF POOL

~ASSUMED FUEL ASSEMBLIES ADDED IDEALIZED 0UTLINE OF POOL BOU ND ARY FilGURE 5.6.n SALEM SPENT FUEL POOL UNITS 1 AND 2-IDEALilZATION OF RACK ASSEMBLY

r------a----

I v

UY I T OUT Po T /: I IN -r\ I Pi I ~I I I~

l:Z H I~

r I

~Q HEAT ADDITION I

I I I I

1

v I

I I

~-

FIGURE 5.6.2 THERMAL CHIMJVEY FLOW MODEL

?

p

• p

• p '

'  ?

?

• p

?

RACK  ?

• ?

1~111  ?

? *

• '  ?

I I I I ' p DDW~l

• ?

p p '

• p COMER p  ?

p

. p p' '

? ' p

• ? *  ?

? ,

p p

• p

?

P'

'?

, P' , P'

, P" , P' ', P' ,' , P' P' , P' , P' , , , , ;

__ ,,,,, ,,'P' r

~'~ p'  ?

\_BOTTOM PLENUM FIGURE 5. 6.3 SALEJJ SPENT FUEL POOL-CONVECTION CURRENTS IN THE POOL

HCl..TEC INTERNATIONAL SALEM UNITS 1 & 2 SFP, NORMAL DISCHARGE CEl3 FAS> - CASE 1 REACTOR st-UllX>WN I-

~ 1 .00E+7 _.,_~~~~~~~~~-+-~~~~~~~~~~~~~

EVAPORATION HEAT LOSSES 0.00E+0-1-....._____..__....,_____..__ ~__,,__.._... __.-.-_......___.._..__.-.-_......___...........~

0 100 200 300 "400 TIME AFTER REACTOR SHUTDCJ,.N, I-RS FIGURE 5.8.1 (a) HEAT LOAD CURVES FOR CASE 1

• HOLTEC INTERNATIONAL SALEM UNITS 1 & 2 SFP, NORMAL DISCHARGE CEE FASl - CASE 1 REACTOR SHUTDOWN 160---------------------------------------------------------------------------

l.L

~ 1'10 ffi

~130-+-------------------------------+---------------------------------------1

~

~ 120-4---------1------------t 110-+-............--..-.-...-.---'l"""'T"...........,,.............._..--.-'l"""'T"...........,,............"T"""'I"_ _.......................,...........,._,._ _.......................,...........,._,.~

0 100 ~ 300 "400 TIME AFTER REACTOR SHUTDOWN. HRS FIGURE 5.8.1 (b) BULK POOL TEMPERATURE PROFILE FOR CASE 1

HOLTEC INTERNATIONAL SALEM ~ITS 1 & 2 SFP, EOC FULL CORE OFFLOAD C193 FAS> - CASE 2 REACTOR SHUTDOWN 4.ee£+7-r------------------------------------.-------------------------------------~

~2.00E+7-+--------------------------------+----------------------------------------~

~

EVAPORATION HEAT LOSSES 0.00E+0-l-_______.._________..._.,.i:;.............,,__.._..________........,.._________..........~

• 0 100 200 300 '400 TIME AFTER REACTOR SHUTDa..N, ~

FIGURE 5.8.2(a) HEAT LOAD CURVES FOR CASE 2

HCLTEC INTERNATIONAL.

SALEM UNITS 1 & 2 SFP, EOC FULL CORE OFFLOAD C 193 FASl - CASE 2 REACTOR SHUTDOWN 100---~~___,...~~~~~~~~---..-~~~~~~~~~~~---.

LL r68 t~e~-----__,,

~~

5m120-+-~~~~~~~~~--1~~~~~~~~~~~~~-1 100--'-...-.-............--..--.........------..........--............__..,__.._...____..........._____...,.........__...........__..____.

0 100 200 300 "400 TIME AFTER REACTOR SHUTDOWN. HRS FIGURE 5.8.2(b) BULK POOL TEMPERATURE PROFILE FOR CASE 2

HOLTEC INTERNATIONAL SALEM l.1'1ITS 1 & 2 SFP, BOC FULL CORE OFFLOAD C 193 FAS) - CASE 3 REACTOR SHJTDOWN REACTOR SH.JTIXl..N 4.0e£+7--~~~~~~~~~~~~~~~~~--~~~~~~~

~2.0eE+7--~---~~~~~~~~~~~~~~~--~--~~~~~

~

EVAPORATION HEAT LOSSES 500 1000 1600 2000 Til"'E AFTER REACTCR FIRST SH.JTIXl..N, ~

FIGURE 5.8.3(a) HEAT LOAD CURVES FOR CASE 3

HOLTEC INTERNATIONAL SALEM ~ITS 1 & 2 SFP, BOC FU..L CORE OFFLOAD C193 FAS> - CASE 3 REACTOR SHJTDOWN REACTOR SH.JTIXl.N 100--~~~~~~~~~~~~~~~~---....-~~~~~---.

lJ..

~ 160 t~---------

~

M

~ 120-t---t-----------+------t 100...,._,.--.,,_...,,_................,.-.-,....,..........,.........._,...................-r-_.,.....-r-.......-r-r-~~.--.-t-.--..............................................._,...~

0 600 1000 1500 201210 2600 TIME AFTER REACTOR FIRST SHUTDO~Ng HRS FIGURE 5.8.3(b) BULK POOL TEMPERATURE PROFILE FOR CASE 3

6.0 STRUCTURAL/SEISMIC CONSIDERATIONS 6.i Introduction The structural adequacy of the maximum density spent fuel rack under seismic loadings postulated for the Salem Units 1 and 2 spent fuel pools is considered in this section. All analyses and subsequent evaluations are in compliance with the requirements of the OT Position Paper,Section IV * [ 6. 1. 1] , and fallow the USNRC Standard Review Plan (SRP) [6.1.2]. The dynamic analyses employ a time-history simulation code used in numerous previous licensing efforts (see Table 6. 1. 1)

• This section provides an abstract of the method of analysis, modelling assumptions, key results, and parametric evaluations performed to establish the required margins of safety.

The module layout for the reracked Salem pool Unit 1 .is illustrated in Figure 6. 1. 1. The module layout for Salem pool Unit 2 is a mirror image of the Unit 1 layout. A total of 12 modules, nine new Holtec maximum density racks and three existing flux trap racks, constitute the reracked module array. As can be*

inferred from Figure 6.1.1 and the existing module layout presented in Section 2, the existing racks are required to be repositioned in the pool to realize the new layout arrangement.

In the process, the interties which were installed between the modules will be removed. These interties were originally modelled in the design basis seismic qualification [ 6 .1. 4]. That design basis qualification showed that the interties were redundant and never loaded because of initial thermal gaps. Therefore, the existing analysis continues to be the governing document for the continuing utilization of these racks in the Salem pool. However, it is necessary to ensure that there is sufficient spacing between the new and old racks to enable all cells in the latter to be used for storing fresh fuel. The composite dynamic simulation wherein all racks in the pool - old and new - are simultaneously analyzed,

.6-1

referred to as the Whole Pool Multi-Rack (WPMR) analysis, is utilized to study the relative motion and inter-rack impact phenomena in the entire assemblage of racks. Finally, while this licensing application is confined to demonstrating the structural adequacy of the racks with intact fuel, supplementary seismic analyses of the Salem rack modules with consolidated fuel (2:1 consolidation) have also been carried out to establish their adequacy to store consolidation canisters.

6.2 Analysis Outline The existing Salem spent fuel racks, referred to as the ENC racks, were designed and licensed as seismic Category I equipment. The new racks are also designed as seismic Category I as required by [6.2.1].

The response of a free-standing 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 more accurate simulation is obtained only by direct integration of the nonlinear equations of motion using actual pool slab acceleration time-histories as the forcing function.

Therefore, the initial step in spent fuel rack qualification is to develop synthetic time-histories for three orthogonal directions which comply with the guidelines of the 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 interations between racks occur when rack-to-rack impacts take place during the 6-2

event. To demonstrate structural qualification of new racks, and the kinematic compliance of the remaining existing racks, it is required to show that stress resultants are within allowable limits, that displacements remain within the constraints of the contemplated design layout for the pool, and that when any rack-to-rack impact in the cellular region occurs, it is restricted to locations away from the active fuel region.

Reliable assessment of the stress field and kinematic behavior of the rack modules calls for a conservative dynamic model incorporating 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 compatible with the free-standing installation of the modules. Furthermore, the model must possess the capability to effect momentum transfers which occur due to rattling of fuel assemblies inside storage cells and the capability to simulate lift-off and subsequent impact of support pedestals with the pool liner (or bearing pad). The contribution of the water mass in the interstitial spaces around the rack modules and within the storage cells must be *modelled in an accurate manner since erring in quantification of fluid coupling on either side of the actual value is no guarantee of conservatism. 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,

µ, often maximizes the computed rack displacements as well as the equivalent elastostatic stresses. Finally, the analysis must consider that a rack module may be fully or partially loaded with fuel assemblies or may be 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

• 6-3

influence on the rack motion. The comprehensive structural evaluation must deal with all of these without sacrificing conservatism.

The three-dimensional single rack dynamic model introduced by Holtec International in the Enrico Fermi Unit 2 rack project (ca.

1980) and used in over 25 rerack projects since that time (Table

6. 1. 1) , addresses the above mentioned array of parameters. 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, the single rack 3-D model handles the array of variables as follows:

Interface Coefficient of Friction Parametric runs are made with upper bound and lower bound values of the coefficient of friction. The limiting values are based on experimental data which have been found to be bounded by the values 0.2 and 0.8 *

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

Fuel Loading Scenarios The fuel assemblies are conservatively assumed to rattle in unison which obviously exaggerates the contribution of impact against the cell wall. The different patterns of possible fuel assembly loadings in the rack are simulated by orienting the center of gravity column of the assemblage of fuel assemblies with respect to the module geometric center of gravity in an appropriate manner.

Fluid Coupling 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 when dealing with a single rack is that the adjacent racks vibrate out-of-phase with respect to the rack being analyzed.

6-4

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 safety analyses reported in previous dockets (Table 6. 1. 1) over the past decade have relied on the 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 relative motion of 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 successfully extended Fritz's classical two-body fluid coupling model to multiple bodies and utilized it to perform the first two-dimensional multi-rack analysis (Diablo Canyon, ca. 1987). Subsequently, laboratory experiments were conducted to validate the multi-rack fluid coupling theory. This technology was incorporated in the computer code DYNARACK which now could handle simultaneous simulation of all racks in the pool. This development marked a pivotal expansion in rack structural modelling capability and was first utilized in Chin Shan, Oyster Creek and Shearon Harris plants

[ 6. 2. 3, 6. 2. 4]. The Whole Pool Multi-Rack (WPMR) 3-D analyses have corroborated the accuracy of the single rack 3-D solutions in predicting the maximum structural stresses,

• and also serve to improve predictio~s of rack kinematics.*

.The Whole Pool Multi-Rack analysis methodology is ideally tailored to establish the presence or absence of specific rack-to-rack impacts during the seismic event. This method is utilized to ensure the kinematic stability of the existing Salem racks.

Fortunately, sufficient information regarding the stiffness characteristics of the existing racks is available from reference

[6.1.4] to enable the WPMR model to be constructed.

For the new Holtec racks, where a background analysis in the manner of the existing racks does not exist, the analysis work effort must deal with both stress and displacement criteria.

Towards this end, the sequence of model development and analysis steps that are undertaken are summarized in the following.

Subsequent subsections provide model detail, limiting criteria for stress and displacement, and results of the analyses.

a. Prepare 3-D dynamic models suitable for a time-history analysis of the new maximum density racks.

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 and extent of cells containing fuel assemblies)

• 6-5

c. Perform stress analysis of high stress areas for the limiting case of all the single rack dynamic analysis runs made in the fore going 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.

e. Prepare a whole pool multi-rack dynamic model which includes the RDOFM's of all existing and new rack modules in the pool, and includes all fluid coupling interactions among them, as well as fluid coupling interactions between racks and pool walls. This 3-D simulation is referred to as a Whole Pool Multi-Rack (WPMR) model.

f. Perform 3-D Whole *Pool* Multi-Rack (WPMR) analyses to demonstrate that all kinematic criteria for the spent fuel rack modules are satisfied, and that resultant structure loads confirm the validity of the single rack structural qualification. The principal kinematic criteria are (1) no rack-to-pool wall impact, and (2) no rack-to-rack impact in the cellular region of the racks containing active fuel.

The rack-to-rack gaps for fluid coupling considerations are set a

0. 5 inch as shown in Ref. [ 6. 2. 5]. The baseplates, however, project by 3/16" (each), leading to a maximum gap of 1/8" at the baseplate location. Similarly, 3/16" thick bumper bars at the top of the rack reduce the maximum gaps to 1/8". The baseplate extensions and the top bumper bars ensure that any rack-to-rack impa~ts do not occur in the active fuel region.

The gap between the new Holtec and the existing ENC racks is 1".

The baseplate of the new racks are extended to make contiguous contact with the ENC racks. Bumper bars at the top of Holtec rack sides facing the ENC racks are also added to define definitive

.6-6

• locations for impact during seismic events. The structure of the existing ENC racks is such as to ensure no impacts in the cellular regions.

The rack-to-wall gaps corresponding to Unit 2 Pool "as-surveyed" dimension 'which are assessed to be limiting are used for gauging the potential for rack-to-wall impacts. For fluid coupling purposes, the larger gaps computed with reference to the Salem construction drawings are used so as to minimize the ameliorative effect of fluid coupling.

6.3 Synthetic Seismic Time-Histories 6.3.1 Acceptance Criteria for Synthetic Time-Histories Section 3. 7 .1 of the USNRC Standard Review Plan ( SRP) [ 6 .1. 2]

provides guidelines for establishing seismic time-histories. In particular, subsection 3.7.1.II.1.b gives applicable criteria for generation of time-histories from design response spectra.

There are two options for generati~g 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 this licensing application, Option 2 is used for generations of seismic time-histories from Operating Basis Earthquake ( OBE) response spectra and from Design Basis .Earthquake (DBE) response spectra.

A *total time duration between 10 seconds and 25 seconds is required to adequately match the design response spectra at 0. 4 Hz. The corresponding stationary phase strong-motion duration should be between 6 seconds and 15 seconds.

6-7

.-~

With Option 2, as a minimum, four time-histories are to be used for analyses. The response spectra re-calculated from each individual time-history need not envelop the original response spectra, but the multiple time-histories are acceptable if the average calculated response spectra re-generated from these four time-histories envelop the original design response spectra.

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. For three-dimensional dynamic analysis, a set of artificial time-histories consists of three orthogonal acceleration time-histories, one in the vertical direction and two in the horizontal plane. In a set of time-histories, the three generated artificial time-histories must also be statistically independent. Any two time-histories

• are considered to be statistically independent if their normalized correlation coefficient is less than 0.15.

6.3.2 Procedure for Time-History Generation The procedure for generation of synthetic seismic time-histories for Salem follows a more conservative approach than that, used in previous dockets. It is summarized as follows:

a. The response spectra provided in the PSE&G Rerack Specification (Ref. [6.3.1]) is broadened(+/- 15%) per the provision of Reg. Guide 1.122 (Ref. [6.3.2]) before any synthetic time-histories are generated.

b. Four sets of time-histories are generated corresponding to each broadened spectra.

c. The three components of accelerations in each set must be statistically independent with respect to each other (correlation factors 0.15). Further, each component in a set must be statistically independent with respect to its sister component in the other sets.

6-8

d. The average of the four spectra re-generated using the generated time-histories in each direction is developed.

The average spectra must meet the SRP (Ref. [ 6

• 1. 2] )

enveloping requirements.

e. A new synthetic, statistically independent, time-history set corresponding to the averaged spectra is generated.

These new time-histories are used as inputs to the rack seismic analysis. A multiplier of 1.1 is applied to each new time-history to ensure conservatism.

6.3.3 Broadened Response Spectra of Salem Spent Fuel Pool Slab Figures 6.3.1-6.3.4 show the governing response spectra applicable to the spent fuel pool for Salem Generating Station Units 1 and 2.

The response spectra are taken from pages 61 to 64 of Ref. [6.3.l]

as specified by Item 7. 1. 1. 3 of Ref. [ 6. 3. 1]

• Figures 6. 3. 1 and 6.3.2 are for the Design Basis Earthquake (DBE) and Figures 6.3.3 and 6.3.4 are for the Operating Basis Earthquake (OBE). As specified in Appendix A of Ref. [6.3.l], the damping factors used for obtaining the data of response spectra are three percent of critical damping for DBE and one percent for OBE.

The response spectra taken from Figures 6.3.1-6.3.4 are then broadened by +/- 15% of the peak frequencies. The original response spectra for DBE and the broadened spectra are shown in Figures

6. 3. 5 and 6. 3. 6. The original response spectra for OBE and the broadened spectra are in Figures 6.3.7 and 6.3.8. The broadened spectra are used as the basic input data for generating synthetical seismic time-histories.

6.3.4 Synthetic Time-Histories (A) The Holtec Proprietary program GENEQ [ 6. 3. 3] is used to generate synthetic time-histories for both the DBE and OBE from the broadened response spectra shown in Figures 6.3.5-6.3.8. Each set consists of three statistically independent time-histories for two horizontal, and the vertical directions, respectively *

. 6-9

• (B) The average calculated response spectra from these time-histories are obtained by Hol tee Proprietary program AVESPC [ 6. 3. 4] , and are shown in Figures 6.3.9-6.3.11 and 6.3.15-6.3.17 as curve (2). The corresponding broadened spectra are also shown in these figures as curve (1) to demonstrate that the broadened spectra are enveloped by the corresponding average response spectra.

In regenerating the response spectra from the time-histories, the total numbers of periods at equal intervals on a logarithmic scale are 830, 920, 870 for the two horizontal and the vertical directions, respectively. The intervals are much smaller than the acceptable set provided by Table 3. 7 .1-1 in Ref * [ 6

• L 2 ]

• It is clear from Figures 6. 3. 9-6. 3 .11 and 6. 3 .15-

6. 3 .17 that the average response spectra meet the requirement of enveloping the corresponding broadened spectra.

The normalized correlation coefficients Pij between any two time-histories i, j in each set ana between any two sister components of the four sets are provided in Table 6.3.1 for the DBE event and in Table 6. 3. 3 for the OBE. Tables 6. 3 .1 and 6. 3. 3 show that the generated time-histories meet the requirements of the statistical independence.

(C) A new set of synthetical time-histories corresponding to each averaged spectra is then generated using Hol tee Proprietary program GENEQ

[6.3.3] for DBE and OBE, respectively. Figures 6.3.12 to 6.3.14 show the plots of the three components of the final time-history set for DBE and Figures 6.3.18-6.3.20 show the components for QBE.

The response spectra corresponding to the new sets of time-histories are shown in Figures 6.3.9-6.3.11 and 6.3.15-6.3.17 as curve (3). It is demonstrated that the re-generated spectra envelop the corresponding averaged spectra. The normalized correlation coefficients Pij between any two time-histories i and j in the se~ are provided in Tables

6. 3. 2 and 6. 3. 4. It shows that these time-histories meet the requirement of the statistical independence. The corresponding new sets of time-histories, multiplied by a factor of 1.1, are conservatively used as the seismic simulation input of the pool slab for the reracking project.

6-10

• 6.4 6.4.1 Modelling for Dynamic Simulations General Remarks Figure 6.4.1 shows a pictorial view of a typical new rack module for the Salem reracking project. The baseplate extends beyond the cellular region and bumper bars are welded to rack top corners to ensure that inter-rack impacts, can only occur at these two impact hardened regions.

A rack may be completely loaded with fuel assemblies (which corresponds to greatest total mass),. or it may be completely empty. The coefficient of friction, µ, between pedestal supports and pool floor is indeterminate. According to Rabinowicz

[6.4.l], results of 199 tests performed on austenitic stainless steel plates submerged in water show a mean value of µ 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 µ = 0. 2 and 0. 8 have been found to bracket the upper limit of module response in previous rerack projects.

Lift-off of support pedestals and subsequent liner impacts are modelled 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 beam-like action, twisting, and extensions. These special attributes of rack dynamics require strong emphasis on modelling of linear and nonlinear springs, dampers, and compression-only gap elements. The term "nonlinear spring" is a generic term to denote the mathematical element representing the case where restoring force

.6-11

• is not linearly proportional* to displacement.

simulations, In. the fuel the Coulomb friction interface between rack support pedestal and liner is typical of a nonlinear spring.

rack Three-dimensional dynamic analyses of single rack modules require a key modelling assumption. This relates to location and relative motion of neighboring racks. The gap between a peripheral rack and adjacent pool wall is known, with motion of the pool wall prescribed. However, another rack, adjacent to the rack being analyzed, is also free-standing and subject to motion during a seismic event. To conduct the seismic analysis of a given rack, its physical interface with neighboring modules must be specified.

The standard procedure in analysis of a single rack module is to specify that neighboring racks move 180° out-of-phase in relation to the subject rack. Thus, . the available gap before inter-rack impact occurs is 50% of the physical gap. This "opposed-phase motion" assumption increases the likelihood of intra-rack impacts and is thus conservative. However, it also increases the relative contribution of fluid coupling, which depends on fluid gaps and relative movements of bodies, making overall conservatism a less certain assertion. Three-dimensional Whole Pool Multi-Rack analyses carried out on several previous plants demonstrate that single rack simulations predict smaller rack displacement during seismic responses [ 6. 2. 4]. Nevertheless, 3-D analyses of single rack modules permit detailed evaluation of stress fields, and serve as a kinematic benchmark check for the much more involved WPMR analysis.

Particulars of modelling details and assumptions for the 3-D Single Rack analysis for the new fuel racks and for the Whole Pool Multi-Rack analysis for the entire array of racks are given in the following subsections *

• 6-12

6.4.2 The 3-D 22-DOF Model for Single Rack Module Analysis of New Maximum Density Racks 6.4.2.1 Assumptions

a. The fuel rack structure motion is captured by modelling the rack as a 12 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. In this way, the elastic beam-like character of the assemblage, relative to the baseplate, is captured in the dynamic analyses once suitable springs are introduced to couple the rack degrees-of-freedom. Rattling fuel assemblies within the rack are modelled by five lumped masses located ***.1 at H, .75H, .SH, .25H, and at the rack base (H is the rack height measured above the baseplate). Each lumped fuel mass has two horizontal displacement degrees-of-freedom. Vertical motion of the fuel assembly mass is assumed equal to rack vertical motion at the baseplate level. The centroid of each fuel assembly mass can be located off-center, relative to the rack structure centroid at that level, to simulate a partially loaded rack.

b. Seismic motion of a fuel rack is characterized by random rattling of fuel assemblies in their individual storage locations. All fuel assemblies are assumed to move in-phase within a rack. This exaggerates computed dynamic loading on the rack structure and, therefore, yields conservative results.

c. Fluid coupling between rack and fuel assemblies, and between rack and wall, is simulated by appropriate inertial coupling in the system kinetic energy. Inclusion of these effects 'uses the methods of [6.4.2, 6.4.3] for rack/assembly coupling and for rack~to-rack coupling, respectively. Fluid coupling terms for rack-to-rack coupling are based on opposed-phase motion of adjacent modules.

d. Fluid damping and form drag is conservatively neglected.

e. Sloshing is shown to be negligible at the top of the rack and is therefore neglected in the analysis of the rack.

6-13

/

f. Potential impacts between the cell walls of the new racks and the contained fuel assemblies are accounted for by appropriate compression-only gap elements between masses involved. The possible incidence of rack-to-wall or rack-to-rack impact is simulated by gap elements at top and bottom of the rack in two horizontal directions. Bottom elements are located at the baseplate elevation. The initial gaps reflect the presence of baseplate extensions and bumper bars.

g. Pedestals are modelled 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. The spring rate for the friction springs includes any lateral elasticity of the stub pedestals. Local pedestal vertical spring stiffness accounts for floor elasticity and for local rack e,lasticity 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 for New Fuel Racks Figure 6. 4. 2 shows a schematic of the dynamic model. Si ( i. =

1, *** ,4) represent support locations, Pi 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.

Table 6.4.1 lists the degrees-of-freedom for the single rack model. Translational and rotational degrees-of-freedom 1-6 and 17-22 describe the rack motion; rattling fuel masses (nodes 1*, 2*,

3*, 4*, 5* in Figure 6.4.2) are described by translational degrees-of-freedom 7-16. Ui ( t) represents pool floor slab displacement seismic time-history.

6-14

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 ~ssembly/storage cell impact springs at one location of rattling fuel assembly mass.

Figures 6.4.5-6.4.7 show the modelling 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 elastic springs coupling rack vertical and torsional degrees-of-freedom are also included in the model.

Additional details concerning fluid coupling and determination of stiffness elements are provided below.

6.4.2.3 Fluid Coupling 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:

II n

• n n X1 1 X2 denote absolute accelerations of masses mi and m2, respectively, and the notation ocx2) denotes nonlinear terms.

Mi11 Mi2, M211 and M22 are fluid coupling coefficients which depend on body shape, relative disposition, etc. Fritz [ 6. 4. 3]

gives data for Mij for various body shapes and arrangements. The fluid adds mass to the body (M11 to mass mi), and an inertial 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 inter-body gap, reaching large values for small gaps. Lateral motion of a fuel assembly inside a 6-15

the rack top. Support pedestal spring rates Ks are modelled by elements 1 through 4 in Table 6

• 4

• 2

• Local compliance of the concrete floor is included in Ks. 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 modelled by the piecewlse linear friction springs with suitably large stiffness Kf up to the limiting lateral load, µN, where N is the current compression load at the interface between support and liner. At every time-step during transient analysis, the current value of N (either zero if the pedestal has lifted off the liner, or a compressive finite value) is computed. Finally, support rotational friction springs*

KR reflect any rotational limitations (such as edging as the rack rotates) that may be offered by the foundation. The rotational friction spring rate is calculated using a modified Bousinesq equation [6.4.4] and is included to simulate resistive moment by the slab to counteract rotation of the rack pedestal in a vertical plane. The nonlinearity of these springs (friction elements 9, 11,

• 13, and 15 in Table 6.4.2) reflects the edging limitation imposed on the base of the rack support pedestals and the shift in location of slab resistive load as the rack pedestal rotates.

The gap element K5, modelling 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.

• The term "rotational friction spring" is used to connote the parallel between the force which opposes lateral movement of a compressively loaded pedestal due to Coulomb friction and the moment offered by the foundation against rotation of a compressively loaded pedestal. The resistive moment reaches its limiting value when the pedestal contact reaches the so-called "edging" condition, just as the friction reaches a limiting value and remains constant thereafter *

.6-17

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 modelling effects of fluid flowing between adjacent racks are computed assuming that all racks adjacent to the rack being analyzed are vibrating 180° out-of-phase from the rack being analyzed. Thus, the modelled 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 planf orm y = Horizontal coordinate along the long direction of the rack rectangular planf orm z = Vertical coordinate upward from the rack base Table 6. 4. 2 lists all spring elements used in the 3-D 22-DOF single rack model.

If the simulation model is restricted to two dimensions (one horizontal motion plus vertical motion, for example), for the purposes of model clarification only, then Figure 6.4.8 describes the configuration. This simpler model is used to elaborate on the various stiffness modelling elements.

Gap elements modelling impacts between fuel assemblies and rack have local stiffness Kr in Figure 6. 4. 8. In Table 6. 4. 2, for example, gap elements 5 through 8 act on the rattling fuel mass at 6-16

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

6.4.3 Whole Pool Multi-Rack (WPMRl Model 6.4.3.1 General Remarks The single rack 3-D (22-DOF) models for the new racks outlined in the preceding subsection are used to evaluate structural integrity and physical stability 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 including all modules in one comprehensive simulation using a Whole Pool Multi-Rack (WPMR) model. In WPMR analysis, all racks (new and existing) are modelled and their correct fluid interaction is included in the model.

6.4.3.2 Whole Pool Fluid Coupling The presence of fluid moving in the narrow gaps between racks and between racks and pool walls causes both near and far 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 racks. 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 mechanics. 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 was reviewed and approved by the Nuclear Regulatory Commission during post-licensing multi-rack analyses 6-18

• for the Diablo Canyon Units 1 and 2 reracking project (ca. 1987).

The fluid flow model used to obtain the whole pool hydrodynamic effect reflects actual gaps and *rack locations in the spent fuel pool.

6.4.3.3 Coefficients of Friction To eliminate the last significant element of uncertainty in rack dynamic analyses, the friction coefficient as.cribed to the support pedestal/pool bearing pad interface consistent are made consistent with Rabinowicz's data [6.4.l]. 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 WP.MR analysis can simulate the effect of different coefficients of friction at adjacent rack pedestals

• 6.4.3.4 .Modelling Details Figure 6.1.1 shows a view of the spent fuel pool which shows rack and pedestal numbering scheme used for the WP.MR analysis. The details on number of cells per rack and on rack and fuel weights are shown in Table 6.4.3 in this report.

In Whole Pool .Multi-Rack analysis, a .reduced degree-of-freedom (ROOF) set is used to model each rack plus contained fuel. The rack structure is modelled by six 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

• It is noted 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.

.6-19

/ ---

contained fuel is modelled by_ two horizontal degrees-of-freedom.

Thus, the WPMR model involves all racks in the spent fuel pool with each individual rack and its fuel modelled as an 8-DOF structure.

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 Acceptance Criteria, Stress Limits, and Material Properties 6.5.1 Acceptance Criteria There are two sets of criteria to . be satisfied by the rack modules:

a. Kinematic Criteria In order to be qualified as a physically stable structure it is necessary to demonstrate that an isolated rack in water exhibits no overturning tendency when a seismic event of magnitude 1.1 times DBE is applied [6.1.2].

b. Stress Limit Criteria Stress limits must not be exceeded under the postulated load combinations. The following loading combinations are applicable [6.1.3].

6-20

Loading Combination Service Level D + L Level A D + L + To D + L + To + E D + L + Ta + E Level B D + L + To + Pf D + L + Ta + E' Level D D + L + Fd 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 OT Position Paper on "Review and Acceptance of Spent Fuel Storage and Handling Applications" section:

D = Dead weight-induced internal moments

{including fuel assembly weight)

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

= Force caused by the accidental drop of the heaviest load from the maximum possible height specified in the PSE&G Specification [6.3.l].

Pf = Upward force on the racks caused by postulated stuck fuel assembly (see Section 7)

E = Operating Basis Earthquake (OBE)

E' = Design Basis Earthquake (DBE)

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

= Differential temperature induced loads {the highest temperature associated with the postulated abnormal design conditions)

Ta and T 0 produce local thermal stresses. The worst thermal stress field in a fuel rack is obtained when an isolated storage location has a fuel assembly generating heat at maximum postulated rate and

• 6-21

surrounding storage locations contain no fuel. Heated water makes unobstructed contact with the inside of the storage walls, thereby producing maximum possible temperature difference between adjacent cells. Secondary stresses produced are limited to the body of the rack; that is, support pedestals do not experience secondary (thermal) stresses.

6.5.2. Stress Limits for Various Conditions Stress limits are derived from the ASME Code, Section III, Subsection NF [ 6 .1. 3, 6. 5 .1]

• Parameters and terminology are in ,.

accordance with the ASME Code.

6.5.2.1 Normal and Upset Conditions (Level A or Level Bl

a. Allowable stress in tension on a net section is:

Ft = 0.6 Sy (Sy =yield stress at temperature)

(Ft is equivalent to primary membrane stress)

b. Allowable stress in shear on a net section is:

Fv = .4 Sy

c. Allowable stress in compression on a net section

[ 1 -

(kl )2 r2

/2Cc 2 J Fa =

5 kl kl 3 3

{ ( - ) + [3 ( - - ) /8Cc] - [ ( - - ) /SC c ]}

3 r r where:

(2rr2 E) 1;2 Cc =[ ]

s 1 = unsupported length of component k = length coefficient which gives influence of boundary conditions; e.g *

.6-22

• k E

=

=

=

=

1 (simple support both ends) 2 (cantilever beam) 1/2 (clamped at both ends)

Young's Modulus r = radius of gyration of component kl/r for the main rack body is based on the full height and cross-section of the honeycomb region.

d. Maximum allowable bending stress at the outermost fiber of a net section, due to flexure about one plane of symmetry is:

Fb = 0.60 Sy (equivalent to primary bending)

e. Combined flexure and compression on a net section satisfies:

fa Cmx f bx Cmyf by

+ + < 1 Fa DxFbx DyFby where:

fa = Direct compressive stress in the section f bx = Maximum flexural stress along x-axis f by = Maximum flexural stress along y-axis Cmx = Cmy = 0.85 fa Dx = 1 -

F'ex fa Dy = 1 -

F'ey 12 rr2 E F'ex1ey = 2 kl 23 ( )

r x,y

• .6-23

and subscripts x,y reflect the particular bending plane.

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

+ + < 1.0 0.6Sy The above requirements are to be met for both direct tension or compression.

6.5.2.2 Level D Service Limits Section F-1370 (ASME Section III, Appendix F), states that limits for the Level D condition are the minimum of 1. 2 (Sy/Ft) or (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. 2 Sy is less than 0. 7 Su, then the multi plier on the Level A limits, to obtain Level D limits, is 2.0.

6.5.2.3 Dimensionless Stress Factors Stress results are presented in dimensionless form. Dimensionless stress factors are defined as the ratio of the actual developed stress to the specified limiting value. The limiting value of each stress factor is 1. 0 for OBE and 2. 0 (or less) for the DBE condition. Stress factors reported are:

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

= Ratio of gross shear on a net section in the x-direction to its allowable value

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

.6-24

• R4 Rs

=

=

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

R1 = Ratio of gross shear on a net section in the y-direction to its allowable value 6.5.3 Material Properties Physical properties of the rack and support materials, obtained from the AS.ME Boiler & Pressure Vessel Code [6.5.1] 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 Levels A and, D, corresponding to conditions in Section 6.5.2 above, are evaluated using given yield strength data *

.6.6 Governing Equations of Motion 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 degree-of-freedom are obtained using Lagrange's Formulation

[ 6. 6. 1]

• The system kinetic energy includes contributions from solid structures and from trapped and surrounding fluid. The final system of equations obtained have the matrix form:

[M] {q"} = {Q} + {G}

where:

[M] total mass matrix (including structural and fluid mass contributions)

.6-25

• {q}

{G}

the nodal displacement vector relative to the pool slab displacement (double prime stands for second derivatives with respect to time) 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.S]. As indicated earlier, this program has been used in the licensing effort for a large number of reracking projects.

DYNARACK has been validated against exact solutions, experimental data, and solutions obtained using alternate numerical schemes

[6.6.5]. These solutions are chosen to exercise all features of .

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

The application of DYNARACK to the spent fuel rack analysis requires the establishment of a time-step to ensure convergence and stability of the results. DYNARACK utilizes the classical central difference algorithm [6.4.4]. Stability of the results is assured as long as the time-step is significantly below the smallest period of the equivalent linear problem. Convergence is obtained by performing a series of rack analyses with different time-steps to ascertain the upper limit on time-step that will provide converged results. This is done by taking a typical rack

• .6-26

• 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, DYNARACK automatically tracks maximum values of dimensionless factors R1 to R1 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.

6.7 Results of 3-D Nonlinear Analyses of Single Racks This section focuses on results from all 3-D single rack analyses.

In the next 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 Rack Al (the largest rack and with maximum aspect ratio), Rack B4, and Rack A3 (facing existing Rack El).

Altogether, 28 runs are carried out using Holtec proprietary computer program DYNARACK [ 6. 6. 3, 6. 6. 4, 6. 6. 5]

• Results are abstracted from output files and presented here. Analyses have been carried out conservatively using a fuel assembly dry weight of 1700 lbs. Results for either Unit 1 or Unit 2 can be used for

• 6-27

evaluations for the postulated in~rack fuel loading. No rack-to-wall impacts occur in either unit. Results for Unit 1 are detailed here and are used to demonstrate structural integrity and kinematic compliance.

6.7.1 Summa:r:y Results for New 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 6.7.1-6.7.31. Table 6.7.1 lists all runs carried out. Tables 6.7.2 and 6. 7. 3, respectively, present the bounding results from all runs for DBE seismic event, and QBE seismic events. Single rack runs are carried out assuming both in-phase and out-of-phase motion of adjacent racks. For out-of-phase motion of adjacent racks, the maximum top corner displacement is 0.3058" (Table

6. 7. 6)

• The maximum displacement for the as_sumed case of adjacent racks moving in-phase is 1.22" {Table 6.7.2). This case is only used to establish a bounding comparison value for the results of whole pool analyses. Tables 6. 7. 4-6. 7. 31 give details for each individual run. The tabular results for each run give maximax (maximum in time and in space) values of stress factors at important locations in the rack *. Results are given for maximum rack displacements (see Section 6.4.2.2 for x,y orientation),

maximum impact forces at pedestal-liner interface, and rack cell-to-fuel, rack-to-rack, arid rack-to-wall impact forces.

By virtue of the symmetry assumption for the assumed opposed-phase rack motion case, impact 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. It is seen from Tables 6. 7. 2 and 6. 7. 3 that impacts between the rack top corners may occur during a DBE seismic event as well as during an QBE seismic event. Such impact potential is further confirmed in Whole Pool Multi-Rack ( WPMR)

.6-28

• analysis. As stated earlier, the top corners of the new racks are structurally modified to serve as the designated loctation for potential impacts (see Figure 6.4.1).

Structural integrity at various rack sections is considered by computing the appropriate stress factors Ri* Results corresponding to the DBE event yield the highest stress factors. Limiting stress factors for pedestals are at the upper section of the support and are to be compared with the bounding value of 1. 0 ( OBE) or 2. 0 (DBE). 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 are well below the allowable limits.

Additional investigation of important structural elements is carried out and results are summarized in Table 6.7.32. A discussion of these items* follows:

6.7.1.1 Impact Analyses

a. Impact Loading 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 6.7.32 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 6.7.4-6.7.31.

b. Impacts Between Adjacent Racks From Tables 6.7.4-6.7.31, it is found that in many cases impacts at rack top between adjacent racks occur during a DBE seismic event as well as during an OBE seismic event. The impacts between racks are also confirmed in WPMR analysis reported in Section 6.8, where the maximum impact load at rack top between new racks is found to be 5669 lbf (Table 6.8.2, Spring no. 128, between Racks 5 and

6) and the maximum impact load at rack top between new racks and existing rack is found to be 2293 lbf. (Table 6. 8. 8, Spring no. 150, between Rack 8

.6-29 L

and 11). Rack top corners are hardened by welding 3/16" thick, 8" wide bumper bars and serve as the designated.impact location (Figure 6.4.1).

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 rack and multi-rack dynamic analyses (Section 6.8) are surveyed and the maximum loadings are used to qualify the welds.

a. Baseplate-to-Rack Cell Welds and Baseplate-to-Pedestal Welds Reference [6.1.3] (ASME Code Section III, Subsection NF) permits, for the DBE event, an allowable weld stress ~ =

.42 Su* A comparison of this allowable value with the highest weld stress predicted is given in.Table 6.7.32.

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

The weld between baseplate and support pedestal is checked using limit analysis techniques [ 6. 7 .1]. The structural weld at that location is considered safe if the interaction curve between net force and moment is such that:

G = Function(F/Fy,M/My) < 1.0 Fy, 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, M are absolute values of actual force and moments applied to the weld section. The calculated

• value of G for the pedestal/baseplate weld is presented in Table 6.7.32 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 skip 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 6-30

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 6.7.32 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. 32 and is much lower than the allowable load. This upper bound value is obtained by using the highest rack-to-fuel impact load from Table 6.7.3 (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 CWPMR) Analyses Figure 6.1.1 shows the Salem spent fuel pool Unit 1 with nine new spent fuel racks and three existing racks (Racks 10, 11 and 12).

Three runs for WPMR analysis have been carried out:

• Run No. 1: Nine new* racks and three existing racks are all fully loaded with 17 OOi fuel assemblies; coefficients of friction of iack pedestals are random coefficients between 0. 2-0. 8 with the mean value being 0.5; controlling DBE seismic excitations as defined in Section 6.3.

Run No. 2: New Racks A3, B3 and B6 are half loaded, others are. *fully loaded with 1700# fuel assemblies; random friction coefficients; DBE seismic excitations. (This run is configured to maximize the possibility of impact between new racks and existing racks.)

Run No. 3: Nine new racks and three existing racks are all fully loaded with 1700# fuel assemblies; random coefficients; OBE seismic excitations.

Results of Run No. 1 are sununarized in Tables 6.8.1-6.8.6. 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 .1.1) of Run *No. 1. Table 6. 8. 2 summarizes the maximum impact force in each gap spring of Run No. 1. For each rack in 6-31

• Table 6. 8. 2, the first 4 springs (for Racks 10, 11 and 12, the first seven springs) are the pedestal vertical springs; the last 4 springs are the fuel-to-cell impact springs. The springs from No.

106 to No. 167 are rack-to-rack/wall impact springs at rack top, and the springs from No. 168 to No. 229 are rack-.to-rack/wall impact springs at baseplate level. Non-zero values are found for some impact springs between racks indicate that there are some impacts between racks during a DBE seismic event. However, no non-zero values are found for impact springs between racks and walls indicating that there is no impact between any rack and the pool walls during a DBE seismic event. Table 6.8.3 shows the maximum pedestal stress factors for each of the racks in the pool from the WPMR analysis. . Table 6. 8. 4 shows the hydrodynamic pressures on the pool walls. Table 6.8.5 shows the static load and the effective dynamic load adder for each pedestal. Table 6.8.6 gives the total static load and the effective dynamic load adder on the whole pool slab *

• Results of WPMR Run No. 2 and Run No. 3 are summarized in Tables 6.8.7-6.8.12 and 6.8.13-6.8.18, respectively.

results the following conclusions are drawn:

From the WPMR (1) No rack-to-pool wall impacts are predicted during DBE or OBE seismic events.

( 2) In Run No. 2, impacts at rack top corners are found between new racks and existing racks during a DBE event.

(3) Impacts are found to occur at rack top corners among the new racks during a DBE event as well as during an OBE event.

Figures 6. 8. 1-6. 8. 5 show the time-histories of rack-to-wall or rack-to-rack gaps on rack top at typical locations (see Figure

6. 8. 6 for gap locations) for WP.MR Run No. 1. Figure 6. 8 .1 shows that there is no

• impact between Rack 2 . and the north wall.

Figures 6.8.2-6.8.4 show that there are a series of impacts between new racks during a DBE seismic event. Figure 6.8.5 shows 6-32

• no impact between new Rack 8 and the existing Rack 11 in Run No. 1 but demonstrates that there are instants when the gap is nearly closed. *In Run No. 2, however, there are impacts between Rack 8 and Rack 11.

Impacts between the rack modules produce local membrane and bending stresses which ai*e not governed by any stress limits in the ASME Code for Class 3 of NF Structures. Nevertheless, the design criteria imposed on the new and existing Salem racks is that the local stresses remain below the material yield point.

All designated and potential impact regions are found to satisfy this criterion in both new and existing Salem racks.

In Table 6.8.19, the maximum displacement, pedestal vertical loads, and pedestal stress factors obtained from the WPMR analyses are compared with results obtained from single rack analyses assuming that the single ~ack undergoes in-phase motion and is not

• affected by adjacent racks. Note that the single rack displacement of 1. 22" obtained from the assumption of in-phase motion bounds the realistic displacement value of 0.87" obtained from the WPMR analysis.

6.9 Bearing Pad Analysis To protect the slab from high localized dynamic loadings, bearing pads are placed between the pedestal base and the slab. Fuel rack pedestals impact on these bearing pads during a seismic event and pedestal loading is transferred to the liner. Bearing pad dimensions are set to ensure that the average pressure on the slab surf ace due to a static load plus a dynamic impact load does not exceed the American Concrete Institute [ 6. 9 .1] limit on bearing pressures. Pedestal locations are set to avoid overloading of leak chase and weld seam regions under the slab. Time-history results

• 6-33

• from dynamic simulations for each pedestal are used to generate appropriate static and dynamic pedestal loads, which are then used to develop the bearing pad size.

Section io of [6.9.i] gives the design bearing strength as fb = - (.85 fc') e where ~ = .7 and fc' is the specified concrete strength for the spent fuel pool. E = i except when the su,Pporting surface is wider on all sides than the loaded area. In that case, E =

(A2/Ai)*s, but not more than 2. Ai is the actual loaded area, and A2 is an area greater than Ai and is defined in [6.9.i]. Using a value of E > i includes credit for the confining effect of the surrounding concrete.

Bearing pads are sized so as to provide sufficient margin on the average bearing pressure in comparison to the applicable

• allowables. For fc' = 3500 psi, e = 2 I . the allowable bearing pressure is fb = 4i65 psi assuming full concrete confinement and fb = 2082.5 psi assuming no concrete confinement.

From the results of WP.MR analysis summarized in Tables 6.8.5 and 6.8.ii, it is found that the maximum effective static pedestal load (static plus dynamic adder) is 62600 + 39277.6 = ioi877.6 lbf at Rack 9, pedestal 2. While the load used is less than the peak instantaneous loads from either single or Whole Pool Multi-Rack analyses, the use of an effective static load here, in lieu of a peak dynamic load, reflects the fact that the ACI criteia is based on static concepts.

The maximum value of the average bearing pressure on the bearing pad is computed to be less than 710 psi. This is less than the allowable value (2082.5 psi) based on the no-concrete confinement

• .6-34

• situation, indicating that no bearing strength problem exists in the slab due to iompacts of the rack pedestals during a seismic event.

6.10 Fatigue Considerations 6.10.1 Introductory Remarks The NF classification used for free-standing fuel racks poses stress limits on primary stresses only; no limits on secondary or peak stresses are set. The structural integrity assessment is limited to ensuring that gross structural collapse will not occur.

As an additional check on the conservative nature of the design, a fatigue analysis of the highly loaded area of a new fuel rack is carried out. The number of imposed seismic events is one DBE and twenty QBE earthquakes. It is important to emphasize that this specification is beyond the requirements set forth by the OT Position Paper. The purpose of this analysis is solely to provide an assessment of the fat~gue life of the area of the spent fuel rack deemed to be subjected to significant local stress range due to seismic events.

6.10.2 Methodology and Assumptions 6.10.2.1 Governing Codes and Analysis Procedure It is assumed that the AS.ME Section III, Subsection NB-3222. 4

[ 6 .10 .1] for Class I components is the reference code section which outlines the procedure for fatigue analysis. The section is written for Class 1 components but is applied here for the Class 3 Fuel racks. The *procedure outlined makes use of fatigue data in Appendix I of the AS.ME Section III Code, and also refers to sections NB-3222.2, NB-3228.5, NB-3215, and NB-3216. The analysis method outlined in the above sections of the Code is briefly summarized here *

• .6-35

Develop a *model of* the area to*be investigated for fatigue damage. The finite element code ANSYS

[6.10.2] is used in this report.

For the given loading history, perform bounding stress analyses to. establish peak stress amplitudes at the critical locations.

• Calculate critical stress intensities taking into account different principal stress directions, and find the peak stress intensities that characterize the critical cycles. The alternating stress amplitudes used for fatigue damage analysis are 50%

of the stress intensity ranges that are found from the analyses.

• The magnitudes of the alternating stress intensities are adjusted upwards to account for plasticity effects, and the appropriate Code fatigue curve used to obtain a damage factor accounting for the actual.number of cycles ~xpected during the postulated events.

6.10.2.2 Model For free-standing fuel racks, the geometry .is such that the area of concern is the region in the rack cellular structure above the baseplate, in the vicinity of a support pedestal. During a seismic

. event, the free-standing rack *is expected to slide, rock, etc.,

leading to the highest local loadings near the pedestal. The maximum values for these local loads only occurs during a limited period of time during a seismic event. During most of the seismic event duration, *the loadings, while present, are significantly reduced. The spent fuel rack analyses provide results for the time-history behavior of the vertical and horizontal loads at the pedestal liner interface. In the fatigue analysis, these time-histories are used to develop a bounding load scenario for stress analysis and also to establish the number of stress cycles which are to be considered in the fatigue analysis *

• .6-36

The region of interest that needs to be modelled for the detailed stress analysis need only be a small region in the immediate vicinity of the loaded pedestal. A section of 16 cells, extending 32" above the baseplate, is modelled in detail to obtain the limiting stress distributions. Eight actual cell boxes plus eight developed boxes are modelled. The applied.loads on the pedestal are assumed to be balanced by inertia forces in the portions of the rack not modelled. Figure 6 .10 .1 gives an ANSYS plot of the extent of the model.

Fatigue analysis is carried out on the representative new spent *-*

fuel rack which experiences the highest vertical support pedestal loading as characterized by the

• Bol tee whole pool multi-rack analysis. Stress cycles are characterized by the -cyclic life of the direct compressive load and the two friction loads acting on the pedestal. Bounding loads and number of cycles are obtained by examination of the relevant load time-histories for the pedestal chosen for detailed examination.

The cumulative damage factor calculated in conformance with the AS.ME Code indicates that it is less than 50% of the Code allowable value of 1. 0 a 6.11 Conclusion The results of the exhaustive set of seismic analyses with both DBE and OBE excitations indicate that the reconfigured Salem fuel pools with nine Boltec maximum density racks and three existing flux-trap racks constitute a kinematically stable and dynamically safe system. 3-D analyses have been performed on racks modelled individually ( 3-D single rack analysis) .and in a much more comprehensive model, on the entire module array (all twelve racks in the pool) modelled simultaneously. The main conclusions are summarized as follows:

.6-37

• a.

b.

The maximum primary stress levels in the cellular region of the rack have a margin (ratio of allowable to actual maximum) of over 1.9 in both DBE and OBE events.

The maximum primary stresses in the pedestal region has a margin of over 1.25 in both DBE and OBE events.

c. The more accurate Whole Pool Multi-Rack analysis (WPMR) predicts higher stresses and displacements than does the classical opposed-phase single rack 3-D analysis. It is recalled that, in previous analyses (such as those reported on the dockets of Zion, Sequoyah, and LaSalle Unit 1), the WPMR analysis predicted larger rack movements than the single rack opposed-phase analysis but not higher stresses. Recognizing the inherent limitations of the single rack 3-D model, the Salem results continue to cast its usefulness .as an accurate predictive tool even for effects such as time-history of pedestal vertical and shear loads into doubt. In other words, a true bounding assessment of the kinematics and load levels in

• the racks can only be assured if all racks in the pool are modelled with three dimensional degrees-of-freedom models and complete fluid coupling interactions in one integrated time-history simulation.

d* The WPMR analyses indicate .rack-to-rack impacts, although the impact forces are only a fraction of those computed in other PWR plants such as Diablo Canyon I and II, Byron and Braidwood.

e. No rack-to-pool wall impacts are indicated. Therefore, the only loads transmitted by the racks to the fuel pool walls are the hydrodynamic loads (which are fully accounted for in the pool structural considerations summarized in Section 8).

f. The overall stress and displacement factor of safety in the Salem racks are comparable to other PWR plants most recently licensed, such as TMI Unit 1, Indian Point Unit Two, *o.c. Cook, Vogtle Unit Two, and Zion.

g. The maximum value of the cumulative damage factor due to the oscillatory .stressing of the rack during the seismic event is less than 1. 0, indicating that the critical sections and welds in . the racks will not fail due to fatigue during and in the wake of the postulated seismic events.

h. The whole pool analysis,- which reflects reality, predicts displacements which are less than the single rack in-phase solution which is used as a bounding calculation.

6-38

• 6.12

i. The above conclusions apply to racks in both the Unit 1 and Unit 2 spent fuel pools.

References

[6.1.11 "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14, 1978, and January 18, 1979 amendment thereto.

[6.1.2] USNRC Standard Review Plan, NUREG-0800 (SRP 3.7.1, Rev. 2, 1989).

[ 6 .1. 3] ASME Boiler* & Pressure Vessel Code Section III, Subsection NF, (1986).

[6.1.4] Design Analysis Report - Replacement of Spent Fuel Storage Modules for Public Service Electric and Gas Co. , Salem Generating Station, Uni ts 1 and 2, Exxon Nuclear Co. ,

Report XN-NS-018, April, 1978.

[6.2.1] USNRC Regulatory Guide 1.29, "Seismic Design Classification," Rev. 3, 1978.

[6.2.2] Soler, A.I. and Singh, K.P., "Seismic Responses of Free Standing Fuel Rack Constructions to 3-D Motions", Nuclear Engineering and Design, Vol. 80, pp. 315-329 (1984).

[6.2.3] Singh, K.P. and Soler, A.I., "Seismic Qualification of Free Standing Nuclear Fuel Storage Racks - the Chin Shan Experience, Nuclear Engineering International, UK (March 1991).

[6.2.4] Soler, A.I. and Singh, K.P., "Some Results from Simultaneous Seismic Simulations of All Racks in a .Fuel Pool", INNM Spent Fuel Management Seminar X, January, 1993.

[6.2.5] Boltec International Pool Layout Drawing Nos.

1086, Rev. 1 and 1094, Rev. O.

[6.3.l] Technical Specification, Maxlmum Density Spent Fuel Storage Racks for Salem Generating Station Uni ts 1 and 2, Spec. No. S-C-FBB-SGS-0154.

6-39

[6.3.2] Regulatory Guide 1.122, "Development of Floor Design Response Spectra for Seismic Design of Floor Supported Equipment or Components",

U.S. Nuclear Regulatory *commission, Rev. 1, February, 1978.

[6.3.3] Boltec Proprietary Report - Verification and User's Manual for Computer Code GENEQ, Report BI-89364, January, 1990.

[6.3.4] Boltec Proprietary Report - User's Manual and Validation of Program AVESPC.FOR, Report BI-92871, 1992.

[6.4.l] Rabinowicz, E., "Friction Coefficients of Water Lubrica~ed Stainless Steels for a Spent Fuel Rack 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.

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

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

Correlation of Theory and Experiment",

(Proprietary)*, NUSCO/Boltec Report BI-88243.

[6.5.1] ASME Boiler & . Pressure Vessel Code,Section III, appendices (1986).

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

Penzien, McGraw Hill (1975).

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

.6-40

• [6.6.3] Soler, A.I.,

Proprietary February, 1990.

"Theoretical Background for Single and Multiple Rack Analysis*, Boltee Report BI-90439, Rev. 0,

[6.6.4] Soler, A. I. , DYNARACK Theoretical Manual",

Bol tee Proprietary Report BI-8 716 2 , Rev. 1, January, 1988.

[6.6.S] Soler, A.I., "DYNARACK Validation Manual, Boltec Proprietary Report BI-91700, Rev. O, October, 1991.

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

[6.9.l] ACI349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan, 1985.

[6.10.l] ASME Code,Section III, Subsection NB, 1989.

[6.10.2] ANSYS Finite Element Code, Version 4.4A, Swanson Analysis Systems, 1990.

6-41

Table 6 .1.1 PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK PLANT DOCKET NO.

Enrico Fermi Unit 2 USNRC 50-341 Quad Cities 1 and 2 USNRC 50-254, 50-265 Rancho Seco USNRC 50-312 Grand Gulf Unit 1 USNRC 50-416 Oyster Creek USNRC 50-219 Pilgrim USNRC 50-293 v.c. Summer USNRC 50-395 Diablo Canyon Units 1 and 2 USNRC 50-275, 50-323 Byron Units 1 & 2 USNRC 50-454, 50-455 Braidwood Units 1 & 2 USNRC 50-456, 50-457 Vogtle Unit 2 USNRC 50-425 St. 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. FitzPatrick USNRC 50-333 Shearon Harris Unit 2 USNRC 50-401 Hope Creek USNRC 50-354 Kuosheng Units 1 & 2 Taiwan Power Company

Table 6.1.1 (continued)

PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK PLANT DOCKET NO.

Ulchin Unit 2 Korea Electric Power Laguna Verde Units 1 & 2 Comision Federai de Electricidad Zion Station Units 1 & 2 USNRC 50-295, 50-304 Sequoyah USNRC 50-327, 50-328 La Salle Unit One USNRC 50-373 Duane Arnold USNRC 50-331 Fort Calhoun USNRC 50-285 Nine Mile Point Unit One USNRC 50-220 Beaver Valley Unit One USNRC 50-334

Table 6.3.1 CROSS-CORRELATION COEFFICIENTS IN EACH SET AND BETWEEN THE SISTER COMPONENTS IN THE FOUR SETS OF TIME-HISTORIES GENERATED FROM THE BROADENED DESIGN BASIS EARTHQUAKE (DBE) RESPONSE SPECTRA SET-1 SET-2 a-tdbe.h11 a-tdbe.h12 a-tdbe.vt1 a*tdbe.h21 a-tdbe.h22 a-tdbe.vt2 a-tdbe.h11 a-tdbe.h12 - .014725 a-tdbe.vt1 -.061709 -.094488 a-tdbe.h21 .016391 a-tdbe.h22

• 080574 - .009205 a-tdbe.vt2 -.089348 .061563 -.083913 a-tdbe.h31 .026088 -.059937 a-tdbe.h32 .091398 .106358 a-tdbe.vt3 -.087192 -.064129 a-tdbe.h41 .048469 .088356 a-tdbe.h42 .109665 -.020324 a-tdbe.vt4 -.085047 -.019103 SET-3 SET-4 a-tdbe.h11 a-tdbe.h12 a-tdbe.vt1

• a-tdbe.h21 a-tdbe.h22 a-tdbe.vt2 a-tdbe.h11 a-tdbe.h12 a-tdbe.vt1 a-tdbe.h21 a-tdbe.h22 a-tdbe.vt2 a-tdbe.h31 a-tdbe.h32 -.061458 a-tdbe.vt3 -.013844 -.113745 a-tdbe.h41 .097970 a-tdbe.h42 -.117685 -.053673 a-tdbe.vt4 .061951 -.040076 .036925

Table 6.3.2 CROSS-CORRELATION COEFFICIENTS BETWEEN COMPONENTS OF TIME-HISTORIES GENERATED FROM THE AVERAGE RESPONSE SPECTRA OF FOUR SETS OF TIME-HISTORIES FOR DESIGN B~SIS EARTHQUAKE (DBE)

A-T-DBE.Hl to A-T-DBE.H2 = -.03662610 A-T-DBE.Hl to A-T-DBE.VT = -.06734440 A-T-DBE.H2 to A-T-DBE.VT = .03564046

• Table 6.3.3 CROSS-CORRELATION COEFFICIENTS JN EACH SET AND BETWEEN THE SISTER COMPONENTS JN THE FOUR SETS OF TIME-HISTORIES GENERATED FROM THE BROADENED OPERATING BASIS EARTHQUAKE COBE) RESPONSE SPECTRA SET-1 SET-2 a-tobe.h11 a-tobe.h12 a-tobe.vt1 a-tobe.h21 a-tobe.h22 a-tobe.vt2 a-tobe.h11 a-tobe.h12 -.052851

• a-tobe.vt1 -.074957 -.111497 a-tobe.h21 .005254 a-tobe.h22 -.070156 *.051633 a-tobe.vt2 -.038223 *.089976 -.009521 a-tobe.h31 -.006887 .036917 a-tobe.h32 .076004 .061111 a-tobe.vt3 .120404 -.021014 a-tobe.h41 .043770 -.083823 a-tobe.h42 *.002985 - .000324 a-tobe.vt4 -.063687 .089888 SET-3 SET-4 a-tobe.h31 a-tobe.h32 a-tobe.vt3 a-tobe.h41 a-tobe.h42 a-tobe.vt4 a-tobe.h11 a-tobe.h12 a-tobe.vt1 a-tobe.h21 a-tobe.h22 a-tobe.vt2 a-tobe.h31 a-tobe.h32 -.021500 a*tobe.vt3 *.055189 -.037978 a-tobe.h41 -.084680 a-tobe.h42 0.135475 -.022059 a-tobe.vt4 -.018658 -.115870 -.018393

Table 6.3.4 CROSS-CORRELATION COEFFICIENTS BETWEEN COMPONENETS OF TIME-HISTORIES GENERATED FROM THE AVERAGE RESPONSE SPECTRA OF FOUR SETS OF TIME-HISTORIES FOR OPERATING BASIS EARTHQUAKE (OBE)

A-T-OBE.Hl TO A-T-OBE.H2 = -.06540158 A-T-OBE.Hl TO A-T-OBE.VT = .07796821 A-T-OBE.H2 TO A-T-OBE.VT = .1186225

• Table 6 .4 .1 DEGREES-OF-FREEDOM Displacement Rotation Location (Node) 1 Pl P2 P3 2 P17 PlB Pl9 Point 2 is assumed attached to rigid rack at the top most point.

P7 PB P9 PlO Pll Pl2 s* P13 P14 PlS P16 where the relative displacement variables qi are defined as:

Pi = qi(t) + U1 ( t) i = 1,7,9,11,13,15,17

= qi(t) + U2(t) i = 2,8,10,12,14,16,18

= qi(t) + U3(t) i = 3,19 Ui(t) are the 3 known earthquake displacements.

l

• Table 6.4.2 (continued)

NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS II. Friction Elements (16 total)

Number Node Location Description 1 Support Sl X direction friction 2 Support Sl 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 Sl X Slab moment 10 Sl 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 I

l

Table 6.4.2 NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS I. Nonlinear Springs (Gap Elements) (64 Total)

Number Node Location Description 1 Support Sl Z compression only element 2 Support S2 Z compression only element 3 Support S3 Z compression only element 4 Support S4 Z 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

• 3 * , 4

• and 5

• masses for nodes 1.,

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 Continued on Next Page

Table 6.4.3 RACK DETAILS Rack Number Rack Fuel Assembly I.D. of Cells Weight (lbf) Weight (lbf)

UNIT 1 Al 1S6 21100 1700 A2 1S6 21100 1700 A3 1S6 21100 1700 Bl 144 19SOO 1700 B2 144 19SOO 1700 B3 144 19SOO 1700 B4 144 19SOO 1700 BS 144 19SOO 1700 B6 144 19SOO 1700 El (Reg. 1) 100 33800 1700 E2 100 33800 1700 E3 100 33800 1700 UNIT 2 A4 1S6 21100 1700 AS 1S6 21100 1700 A6 1S6 21100 1700 B7 144 19SOO 1700 BB 144 19SOO 1700 B9 144 19SOO 1700 BlO 144 19SOO 1700 Bll 144 19SOO 1700 Bl2 144 19SOO 1700 E4 100 33800 1700 ES 100 33800 1700 E6 100 33800 1700

Table 6.5.1 RACK MATERIAL DATA (200°F)

(ASME - Section II, Part D)

Young's Yield Ultimate Modulus Strength Strength Material E (psi) Sy (psi) Su (psi)

ASTM 240 304L S.S. 27.6x106 21,300 66,200 Continued on Next Page

Table 6.5.1 (continued)

SUPPORT MATERIAL DATA (200°F)

Young's Yield Ultimc:ite Modulus Strength Strength Material E (psi) Sy (psi) Su {psi) 1 SA-240, 27.6x106 21,300 66,200 Type 304L (upper part of support feet) 2 SA-564-630 27.6x106 106,300 140,000 (lower part of support feet; age hardened at 1100°F)

• Holtec Rack Fuel Table 6.7.1

• RESULTS OF SINGLE RACK ANALYSES List of All Runs Fuel Loading Seismic Coefficient Motion Run I.D. I.D. I.D. .condition Loading of Friction Mode dra1dbeo.rf8 Al 1700# Fully Loaded DBExl.1 0.8 opposed regular 156 cells phase dra1dbeo.rf2 Al 1700# Fully* Loaded DBEx1.1 0.2 opposed regular 156 cells phase dra1dbeo.rh8 A1 1700# Half Loaded DBExl.1 0.8 opposed regular 78 cells phase dra1dbeo.rh2 Al 1700# Half Loaded DBExl.1 0.2 opposed regular 78 cells phase dra1dbeo.re8 Al 170# II Empty II DBEx1.1 0.8 opposed assumed 12 cells loaded phase draldbeo.re2 Al 170# II Empty II DBExl.1 0.2 opposed assumed 12 cells loaded phase drb4dbeo.rf8 B4 1700# Fully Loaded DBExl.1 0.8 opposed regular 144 cells phase drb4dbeo.rf2 B4 1700# Fully Loaded DBExl.1 0.2 opposed regular 144 cells phase drb4dbeo.rh8 B4 1700# Half Loaded DBExl.1 0.8 opposed regular 72 cells phase drb4dbeo.rh2 B4 1700# Half Loaded DBExl.1 0.2 opposed regular 72 cells phase drb4dbeo.re8 B4 170# II Empty II DBExl.1 0.8 opposed assumed 1.2 cells loaded phase drb4dbeo.re2 B4 1.70# II Empty II DBExl.1 0.2 opposed assumed 12 cells loaded phase 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 dra3dbeo.rf8 A3 1700# Fully Loaded DBEx1.1 o.a opposed regular 156 cells phase dra3dbeo.rf2 A3 1700# Fully Loaded DBEx1.1 0.2 opposed regular 156 cells phase dra3dbei.rf8 A3 1700# Fully Loaded DBEx1.1 0.8 in-regular 156 cells phase dra3dbei.rf2 A3 1700# Fully Loaded DBEx1.l. 0.2 in-regular 1.56. cells phase dral.obeo.rf8 Al. 1.700# Fully Loaded OBEx1.l. 0.8 opposed regular 1.56 cells phase draiobeo.rf2 Al. 1700# Fully Loaded OBEx1 . .l 0.2 opposed regular 1.56 cells phase Partially dral.obeo.rh8 Al. 1.700# Load ea OBEx1.l. 0.8 opposed regular 78 cells phase Partially dral.obeo.rh2 Al. 1.700# Loaded OBEx1.l. 0.2 opposed regular 78 cells phase dralobeo.re8 Al. 1.70# II Empty II OBEx1.1 0.8 opposed assumed 1.2 cells loaded phase dral.obeo.re2 Al. 1.70# II Empty 11 OBEx1.l. 0.2 opposed assumed 1.2 cells loaded phase drb4obeo.rf8 B4 1700# Fully Loaded OBEx1.1 0.8 opposed regular 1.44 cells phase drb4obeo.rf2 B4 1.700# Fully Loaded OBEx1.1 0.2 opposed regular 1.44 cells phase

( to be continued )

L

( 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 drb4obeo.rh8 B4 1700# Half Loaded OBEx1.1 0.8 opposed regular 72 cells phase drb4obeo.rh2 B4 17.00.# .. Half Loaded . OBEx1.1 0.2 opposed regular 72 cells phase drb4obeo.re8 B4 170# II Empty II OBEx1.1 0.8 opposed assumed 12 cells loaded phase drb4obeo.re2 B4 170# II Empty II OBEx1.1 0.2 opposed assumed 12 cells loaded phase 1

Table 6.7.2

SUMMARY

OF WORST RESULTS FROM 16 RUNS OF SINGLE RACK ANALYSIS FOR HOLTEC RACKS IN SALEM POOL

( LOADED WITH 1700# REGULAR FUEL ASSEMBLIES; CONTROLLING SEISMIC: DBE x 1.1 )

Item Value Run I.D.

1. Maximum total vertical pedestal load: 381,938 lbs. dra3dbei.rf8

2. Maximum vertical load in any single pedestal: 313,465 lbs. dra3dbei.rf8

3. Maximum shear load in any single pedestal: 102,544 lbs. dra3dbei.rf8

4. Maximum fuel assembly-to-cell wall impact load at one local position: 611 lbs. drb4dbeo.rh8

5. Maximum rack-to-wall impact load at baseplat level: 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: 13.2 lbs. drb4dbeo.rh2

9. Maximum corner displacements Top corner in x direction: 1.2233 in. dra3dbei.rf8 in y direction: 0.7265 in. dra3dbei.rf2 Baseplate corner in x direction: 0.3581 in. dra3dbei.rf2 in y direction: 0.4408 in. dra3dbei.rf2

10. Maximum stress factors Above baseplate: 0.521 (R6) dra3dbei.rf8 Support pedestals: 0.789 (R6) dra3dbei.rf8

• Table 6.7.3

SUMMARY

OF WORST RESULTS FROM 12 RUNS OF SINGLE RACK ANALYSIS FOR HOLTEC RACKS IN SALEM POOL

( LOADED WITH 1700# REGULAR FUEL ASSEMBLIES; CONTROLLING SEISMIC: OBE x 1.1 )

Item Value Run I.D.

1. Maximum total vertical pedestal load: 314,612 lbs. dra1obeo.rf8

2. Maximum vertical load in any single pedestal: 112,480 lbs. dra1obeo.rf8

3. Maximum shear load in any single pedestal: 19,438 lbs. dra1obeo.rh8

4. Maximum fuel assembly-to-cell wall impact load at one local position: 814 lbs. drb4obeo.re2

5. Maximum rack-to-wall impact load at baseplat level: 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: 6.1 lbs. drb4obeo.rh2

9. Maximum corner displacements Top corner in x direction: 0.1237 in. drb4obeo.rh8 in y direction: 0.1358 in. dra1obeo.rh8 Baseplate corner in x direction: 0.0063 in. drb4obeo.rf8 in y direction: 0.0073 in. dra1obeo.rf8

10. Maximum stress factors Above baseplate: 0.135 (R6) dra1obeo.rf2 Support pedestals: 0.176 (R6) dra1obeo.rf2

Table 6.7.4

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-Al Holtec Run I.D.: draldbeo.rf8 Seismic Loading: DBExl.1 Fuel Assembly I.D. and Weight: 1700#l7x17 .

I 1 7 oo

• o ( lbs . )

Fuel Loading: 156 cells loaded; Fuel centroid X,Y: *0 I . o (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logf ile: C:/racks/dynamO/dynamO.fov $

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$Revision: 3.36 $

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DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 381279.6 (2) Maximum vertical load in any single pedestal: 155439.5 (3) Maximum shear load in any single pedestal: 40713.7 (4) Maximum fuel-cell impact at one local position: 597.9 (5) Maximum rack-to-wall impact at baseplate: *0 (6) Maximum rack-to-wall impact at rack top: *0 (7) Maximum rack,...to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at rack top: 9.8 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .1763 .2557 Baseplate corner: .0134 .0192 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 R5 R6 R7 Above baseplate: .042 .033 .198 .124 .216 .249 .041 Support pedestal: .190 .050 .181 .120 .296

• 318 .075

• See Section 6.5.2.3 of the Licensing Report for definitions.

• Table 6.7.5

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-A1 Holtec Run I.D.: dra1dbeo.rf2 Seismic Loading: DBEx1.1 Fuel Assembly I.D. and Weight: 1700#17x17 .

I 1 7 oo

• o ( lbs * )

Fuel Loading: 156 cells loaded; Fuel centroid X,Y: *0 I . o (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logf ile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynas1.fov $

$Revision: 3.36 $

$Logf ile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 381279.6 (2) Maximum vertical load in any single pedestal: 154451.5 (3) Maximum shear load in any single pedestal: 29922.0 (4) Maximum fuel-cell impact at one local position: 598.0 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at :tack top: .o (7) Maximum rack-to-rack impact at baseplat_e: .o (8) Maximum rack-to-rack impact at rack top: 9.7 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .1793 .2521 Baseplate corner: .0142 .0184 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: . 042 .031 .197 .126 .216 .248 .039 Support pedestal: .189 .043 .129 .104 .297 .317 .053

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.6

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-A1 Holtec Run I.D.: dra1dbeo.rh8 Seismic Loading: DBEx1.1 Fuel Assembly I.D. and Weight: 1700#17x17 ,. 1 7 oo

• O ( lbs * )

Fuel Loading: 78 cells loaded; Fuel centroid X,Y:

• O, 2 9. 3 (in. )

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynas1.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 198421. 0 (2) Maximum vertical load in any single pedestal: 104583.3 (3) Maximum shear load in any single pedestal: 49935.4

( 4 ). Maximum fuel-cell impact at one local position: 586.6 (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:. .o (8) Maximum rack-to-rack impact at rack top: 7.2 MAXIMUM CORNER DISPLACEMENTS {in.)

Location: X-direction Y-direction Top.corner: .1571 .3058 Baseplate corner: .0099 .0182 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 RS R6 R7 Above baseplate: .023 .021 .092 .084 .130 .150 .024 support pedestal: .128 .047 .221 .113 .309 .342 .091

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.7

SUMMARY

RESULTS.OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-A1 Holtec Run I.D.: dra1dbeo.rh2 Seismic Loading: DBEx1.1 Fuel Assembly I.D. and Weight: 1700#17x17 ,. 1 7 oo

• o ( lbs * )

Fuel Loading: 78 cells loaded; Fuel centroid X,Y: . O, 2 9

• 3 ( in . )

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logf ile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 199644.4 (2) Maximum vertical load in any single pedestal: 104104.7 (3) Maximum shear load in any single pedestal: 19956.2 (4) Maximum fuel-cell impact at one local position: 586.8 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to...:rack impact at baseplate: 1.5 (8) Maximum rack-to-rack impact at rack top: 7.8 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .1404 .2476 Baseplate corner: .0186 .0626 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: .023 .016 .093 .077 .121 .141

• 020 Support pedestal: .127 .032 .088 .078 .198 .211 .036

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.8

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-A1 Holtec Run I.D.: dra1dbeo.re8 Seismic Loading: DBEx1.1 Fuel Assembly I.D. and Weight: 12x170# ,. 1 7 o* o ( lbs * )

Fuel Loading: 12 cells loaded; Fuel centroid X,Y:

• QI 54 o 4 (in* )

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 25764.5 (2) Maximum vertical load in any single pedestal: 17226.3 (3) Maximum shear load in any single pedestal: 6079.1 (4) Maximum fuel-cell impact at one local position: 113.7 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at rack top: .o MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0355 .0512 Baseplate corner: .0027 .0039 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 RS R6 R7 Above baseplate: .007 .003 .022 .019 .030 .034 .004 Support pedestal: .021 .008 .025 .020 .038 .042 .010

• 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-A1 Holtec Run I.D.: dra1dbeo.re2 Seismic Loading: DBEx1.1 Fuel Assembly I.D. and Weight: 12x170# . 1 7 o

• O ( lbs * )

Fuel Loading: 12 cells loaded; Fuel centroid X,Y: . o, 54. 4 (in. )

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynam0/dynas1.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 25077.7 (2) Maximum vertical load in any single pedestal: 15393.9 (3) Maximum shear load in any single pedestal: 3074.5 (4) Maximum fuel-cell impact at one local position: 108.0 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum raqk-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at .baseplate: .o (8) Maximum rack-to-rack impact at rack top: .o MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0403 .0418 Baseplate corner: .0230 .0219 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 RS R6 R7 Above baseplate: .007 .003 .019 .019 .029 .033 .003 Support pedestal: .019 .005 .012 .012 .030 .033 .005

• See Section 6.5.2.3 of the Licensing Report for definitions.

• Table 6.7.10

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-B4 Holtec Run I.D.: drb4dbeo.rf8 Seismic Loading: DBExl.1 Fuel Assembly I.D. and Weight: 1700#17x17 .

I 1100.0 (lbs.)

Fuel Loading: 144 cells loaded; Fuel centroid X,Y: .o, . o (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 351881.1 (2) Maximum vertical load in any single pedestal: 147848.9 (3) Maximum shear load in any single pedestal: 29829.9 (4) Maximum fuel-cell impact at one local position: 583.1 (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: .o (8) Maximum rack-to-rack impact at rack top: 10.0 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .2431 .1920 Baseplate corner: .0184 .0143 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: .041 .040 .142 .186 .223 .258 .032 Support pedestal: .181 .055 .099 .133 .263 .281 .041

• See Section 6.5.2.3 of the Licensing Report for definitions.

• Table 6.7.11

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-B4 Holtec Run I.D.: drb4dbeo.rf2 Seismic Loading: DBEx1.1 Fuel Assembly I.D. and Weight: 1700#17x17 . 1 7 OO* O ( lbs * )

Fuel Loading: 144 cells loaded; Fuel centroid X,Y: . 0' .o (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: . C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynamO/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 351881.1 (2) Maximum vertical load in any single pedestal: 147155.9

• (3) Maximum shear load in any single pedestal:

(4) Maximum fuel-cell impact at one local position:

(5) Maximum rack-to-wall impact at baseplate:

28796.3 583.1

• 0

( 6 ), Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at rack top: 10.0 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .2421 .1926 Baseplate corner: .0185 .0148 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 RS R6 R7 Above baseplate: .041 .039 .141 .184 .225 .260 .030 Support pedestal: .180 .048 .116 .116 .285 .304 .048

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.12

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-B4

'Holtec Run I.D.: drb4dbeo.rh8 Seismic Loading: DBExl.1 Fuel Assembly I.D. and Weight: 1700#17x17 ,. 1700.0 (lbs.)

Fuel Loading: 72 cells loaded; Fuel centroid X,Y: 27.2, . o (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logf ile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logf ile: C:/racks/dynamO/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 176688.0 (2) Maximum vertical load in any single pedestal: 97474.8 (3) Maximum shear load in any single pedestal: 20971.6 (4) Maximum fuel-cell impact at one local position: 610.7 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at rack top: 13.1 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .1937 .1307 Baseplate corner: .0079 .0090 MAXIMUM STRESS FA,CTORS

• Stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: .020 .021 .083 .076 .124 .143 .018 Support pedestal: .119 .037 .075 .090 .185 .198 '* 031

• 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 MODULE: RACK-B4 Holtec Run I.D.: drb4dbeo.rh2 Seismic Loading: DBEx1.1 Fuel Assembly I.D. and Weight: 1700#17x17 .

I 1700.0 (lbs.)

Fuel Loading: 72 cells loaded; Fuel centroid X,Y: 27.2, . o (in.)

coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logf ile: C:/racks/dynamO/dynas1.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 177509.7 (2) Ma.ximum vertical load in any single pedestal: 97303.7 (3) Maximum shear load in any single pedestal: 17390.4 (4) Maximum fuel-cell impact at one local position: 610.1 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at ~ack top: 13.2 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .1944 .1267 Baseplate corner: .0158 .0138 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 RS R6 R7 Above baseplate: .020 .020 .082 .074 .117 .135 .016 support pedestal: .119 .032 .058 .077 ~184 .195 .024

• See Section 6.5.2.3 of the Licensing Report for definitions.

• Table 6.7.14

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-B4 Holtec Run I.D.: drb4dbeo.re8 Seismic Loading: DBEx1.1 Fuel Assembly I.D. and Weight: 12x170# . 1 7 o. o ( lbs . )

Fuel Loading: 12 cells loaded; Fuel centroid X,Y: 49.9, *o (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logf ile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logf ile: C:/racks/dynam0/dynas1.fov $

$Revision: 3.36 $

$Logf ile: C:/racks/dynamO/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

( 1) Maximum total vertical pedestal load: .23747.6 (2) Maximum vertical load in any single pedestal: 15597.4 (3) Maximum shear load in any single pedestal: 5403.5 (4) Maximum fuel-cell impact at one local position: 111.0 (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: .o MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0336 .0429 Baseplate corner: .0027 .0032 MAXIMUM STRESS FACTORS

• stress factor: R1 R2 R3 R4 RS R6 R7 Above baseplate: .007 .003 .019 .020 .029 .033 .003 Support pedestal: .019 .007 .021 .016 .037 .041 .009

• See Section 6.5.2.3 of the Licensing Report for definitions.

• Table 6.7.15

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-B4 Holtec Run I.D.: drb4dbeo.re2 Seismic Loading: DBExl.1 Fuel Assembly I.D. and Weight: 12x170# . 1 7 o. o {lbs . )

Fuel Loading: 12 cells loaded; Fuel centroid X,Y: 49.9, . o {in. )

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logf ile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logf ile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS {lbs.)

{1) Maximum total vertical pedestal load: 23573.1

{2} Maximum vertical load in any single pedestal: 12647.4

{3) Maximum shear load in any single pedestal: 2503.4

{4) Maximum fuel-cell impact at one local position: 111.0 (5) Maximum rack-to-wall impact at baseplate: .o (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: .o MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0391 .0248 Baseplate corner: .0161 .0068 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 R5 R6 R7 Above baseplate: .007 .003 .015 .019 .025 .029 .002 Support pedestal: .015 .004 .010 .010 .025 .026 .004

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.16

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-A3 Holtec Run I.D.: dra3dbeo.rf8 Seismic Loading: DBExl.1 Fuel Assembly I.D. and Weight: 1700#17x17 .

1 7 oo

• O ( lbs * )

Fuel Loading: 156 cells loaded; Fuel centroid X,Y: .o, .o (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logf ile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynamO/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

( 1) Maximum total vertical pedestal load: 381279.6 (2) Maximum vertical load in any single pedestal: 159628.2 (3) Maximum shear load in any. single pedestal: 31154.2 (4) Maximum fuel-cell impact at one local position: 597.5 (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: 9.8 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .2843 .1722 Baseplate corner: .0213 .0130 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 R5 R6 R7 Above baseplate: .042 .040 .137 .191 .221 .256 .046 Support pedestal: .195 .054 .124 .130 .291 .310 .051

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.17

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-A3 Holtec Run I.D.: dra3dbeo.rf2 Seismic Loading: DBExl.1 Fuel Assembly I.D. and Weight: 1700#17x17 . 1 7 oo

• o ( lbs * )

Fuel Loading: 156 cells loaded; Fuel centroid X,Y: .o, . o (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynamO/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

( 1) Maximum total vertical pedestal load: 381279.6 (2) Maximum vertical load in any single pedestal: 160076.8 (3) Maximum shear load in any single pedestal: 30574.4 (4) Maximum fuel-cell impact at one local position: 597.8 (5) Maximum rack-to-wall impact at baseplate: *0 (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at baseplate: *0 (8) Maximum rack-to-rack impact at rack top: 9.8 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .2875 .1742 Baseplate corner: .0251 .0167 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: .042 .038 .136 .194 .220 .252 .035 support pedestal: .195 .054 .116 .131 .305 .325 .048

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.18

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-A3 Holtec Run I.D.: dra3dbei.rf8 Seismic Loading: DBExl.l Fuel Assembly I.D. and Weight: 1700#17xl7 .

I 1 7 oo

• O ( lbs . )

Fuel Loading: 156 cells loaded; Fuel centroid X,Y: *0 I . o (in.)

coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logf ile: C:/racks/dynamO/dynamO.fov $

$Revision: 2 .'5 $ .

$Logfile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS ** (lbs.)

(1) Maximum total vertical pedestal load: 381938.4 (2) Maximum vertical load in any single pedestal: 313464.7 (3) Maximum shear load in any single pedestal: 102543.8 (4) Maximum fuel-cell impact at one local position: 579.1 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at rack top: .o MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: 1.2233 .6114 Baseplate corner: .0993 .0487 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 R5 R6 R7 Above baseplate: .042 .052 .247 .282 .446 .521 .053 Support pedestal: .384 .149 .355 .360 .729 .789 .147

• See Section 6.5.2.~ of the Licensing Report for definitions.

• Table 6.7.i9

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-A3 Holtec Run I.D.: dra3dbei.rf2 Seismic Loading: DBEx1.1 Fuel Assembly I.D. and Weight: 1700#17x17 . 1 7 oo

• o ( lbs * )

Fuel Loading: 156 cells loaded; Fuel centroid X,Y: .o, . o (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

. $Revision: 3.46 $

$Logflle: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logf ile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logf ile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 378236.5 (2) Maximum vertical load in any single pedestal: 273148.2 (3) Maximum shear load in any single pedestal: 54537.7 (4) Maximum fuel-cell impact at one local position: 566.5 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at rack top: .o MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .9297 .7265 Baseplate corner: .3581 .4408 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: .041 .042 .240 .268 .394 .460 .039 Support pedestal: .334 .089 .197 .215 .540 .577 .081

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.20

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-A1 Holtec Run I.D.: dra1obeo.rf8 Seismic Loading: OBEx1~1 Fuel Assembly I.D. and Weight: 1700#17x17 .

I 1700.0 (lbs.)

Fuel Loading: 156 cells loaded; Fuel centroid X,Y: .o,

• O (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynas1.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

. ( 1) Maximum total vertical pedestal load: 314612.1 (2) Maximum vertical load in any single pedestal: 112480.2 (3) Maximum shear load in any single pedestal: 12946.1 (4) Maximum fuel-cell impact at one local position: 448.9 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at baseplate: *0 (8) Maximum rack-to-rack impact at rack top: 2.7 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0792 .0971 Baseplate corner.: .0060 .0073 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 RS R6 R7 Above baseplate: .017 .022 .079 .062 .117 .135 .019 Support pedestal: .138 .018 .038 . 044 .164 .173 .016

• See Section 6.5.2.3 of the Licensing Report for definitions.

-* Table 6.7.21

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-A1 Holtec Run I.D.: dra1obeo.rf2 Seismic Loading: OBEx1.1 Fuel Assembly I.D. and Weight: 1700#17x17 .

I 1 7 oo . o ( lbs . )

Fuel Loading: 156 cells loaded; Fuel centroid X,Y: *0 I . o (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logf ile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logf ile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logf ile: C:/racks/dynamO/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 314612.1 (2) Maximum vertical load in any single pedestal: 112480.2 (3) Maximum shear load in any single pedestal: 13631.0 (4) Maximum fuel-cell impact at one local position: 448.9 (5) Maximum rack-to-wall impact- at baseplate: .o (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: 2.7 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0792 .0971 Baseplate corner: .0060 .0073 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 R5 R6 R7 Above baseplate: .017 .022 .079 .062 .117 .135 .018 Support pedestal: .138 .021 .041 .050 .166 .176 .017

• See Section 6.5.2.3 of the Licensing Report for definitions.

• Table 6.7.22

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-A1 Holtec Run I.D.: dra1obeo.rh8 Seismic Loading: OBEx1.1 Fuel Assembly I.D. and Weight: 1700#17x17 .

I 1 7 oo

• o ( lbs * )

Fuel Loading: 78 cells loaded; Fuel centroid X,Y:

• o , 2 9

• 3 ( in . )

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logf ile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynas1.fov $

$Revision: 3.36 $

$Logf ile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 165815.4 (2) Maximum vertical load in any single pedestal: 82500.2 (3) Maximum shear load in any single pedestal: 19437.6 (4) Maximum fuel-cell impact at one local position: 444.7 (5) Maximum rack-to-wall impact at baseplate: .o (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: .0559 .1.358 Baseplate corner: .0039 .0040 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 RS R6 R7 Above baseplate: .01.1 .009 .042 .038 .075 .087 .011 Support pedestal: .101 .022 .086 .053 .162 .1.75 .036

• 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-Al Holtec Run I.D.: dralobeo.rh2 Seismic Loading: OBExl.1 Fuel Assembly I.D. and Weight: 1700#17x17 .

I 1 7 oo

• o ( lbs . )

Fuel Loading: 78 cells loaded; Fuel centroid X,Y: *0I 2 9

• 3 (in* )

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 165816.2 (2) Maximum vertical load in any single pedestal: 82494.1 (3) Maximum shear load in any single pedestal: 15882.3 (4) Maximum fuel-cell impact at one local position: 444.7 (5) Maximum rack-to-wall impact at baseplate: .o (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: .0565 .1358 Baseplate corner: .0051 .0048 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 R5 R6 R7 Above baseplate: .011 .010 .041 .037 .075 .087 .009 Support pedestal: .101 .018 .069 .044 .161 .172 .028

• See Section 6.5.2.3 of the Licensing Report for definitions.

• Table 6.7.24

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-Al Holtec Run I.D.: dralobeo.re8 Seismic Loading: OBExl.l Fuel Assembly I.D. and Weight: 12x170# .

I 1 7 O* O ( lbs . )

Fuel Loading: 12 cells loaded; Fuel centroid *x,Y: *0I 54

• 4 (in* )

Coefficient of friction at the bottom of suppor~ pedestal: 0.8

$Revision: 3.46 $

$Logf ile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynamO/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 22807.6 (2) Maximum vertical load in any single pedestal: 10035.3 (3) Maximum shear load in any single pedestal: 1313.5 (4) Maximum fuel-cell impact at one local position: *6 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at rack top: .o MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0109 .0149 Baseplate corner: .0007 .0009 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: .006 .001 .010 .007 .017 .019 .002 Support pedestal: .012 .002 .006 .004 .015 .016 .002

• 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 MODULE: RACK-A1 Holtec Run I.D.: dralobeo.re2 Seismic Loading: OBExl.1 Fuel Assembly I.D. and Weight: 12Xl70# . 1 7 o*o ( lbs * )

Fuel Loading: 12 cells loaded; Fuel centroid X,Y: .o, 54.4 (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: c: /racks/dynamO/dynasl. fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynamO/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

( 1) Maximum total vertical pedestal load: 22807.6 (2) Maximum vertical load in any single pedestal: 10035.2 (3) Maximum shear load in any single pedestal: 1368.5 (4) Maximum fuel-cell impact at one local position: .6 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0109 .0149 Baseplate corner: .0007 .0009 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: .006 .001 .010 .007 .017 .019 .002 Support pedestal: .012 .002 .006 .004 .017 .018 .002

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.26

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS.FOR RACK MODULE: RACK-B4 Holtec Run I.D.: drb4obeo.rf8 Seismic Loading: OBExl.1 Fuel Assembly I.D. and Weight: 1700#17x17 .

I 1700. o (lbs.)

Fuel Loading: 144 cells loaded; Fuel centroid X,Y: *0 I . O (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logf ile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynamO/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 290024.9 (2) Maximum vertical load in any single pedestal: 99722.0 (3) Maximum shear load in any si~gle pedestal: 9070.4 (4) Maximum fuel-cell impact at one local position: 463.0 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at rack top: 2.1 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0836 .0707 Baseplate corner: .0063 .0054 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: .017 .019 .056 .064 .101 .117 . 013 Support pedestal: .122 .015 .025 .035 .135 .140 .010

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.27

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-B4 Holtec Run I.D.: drb4obeo.rf2 Seismic Loading: OBEx1.1 Fuel Assembly I.D. and Weight: 1700#17x17 .

I 1 7 oo

• o ( lbs . )

Fuel Loading: 144 cells loaded; Fuel centroid X,Y: *0 I . O (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynam0/dynas1.fov $

$Revision: 3.36 $ .

$Logfile: C:/racks/dynamO/dynas2.fov * $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 290024.9 (2) Maximum vertical load in any single pedestal: 99722.0 (3) Maximum shear load in any single pedestal: 9070.4 (4) Maximum fuel~cell impact at one local position: 463.0 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at rack top: 2.1 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0836 .0707 Baseplate corner: .0063 .0054 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 RS R6 R7 Above baseplate: .017 .019 .056 .064 .101 .117 .013 Support pedestal: .122 .015 .025 .035 .135 .140 .010

• See Section 6.5.2.3 of the Licensing Report for definitions.

---

- - ---~--

Table 6.7.28

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-B4 Holtec Run I.D.: drb4obeo.rh8 Seismic Loading: OBEx1.1 Fuel Assembly I.D. and Weight: 1700#17x17 .

I 1 7 o o

• o (lbs * )

Fuel Loading: 72 cells loaded; Fuel centroid X,Y: 27.2, .o (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logf ile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 152918.1 (2) Maximum vertical load in any single pedestal: 73539.9 (3) Maximum shear load in any single pedestal: 5047.2 (4) Maximum fuel-cell impact at one local position: 450.6 (5) Maximum rack-to-wall impact at baseplate: .o (6) Maximum rack-to-wall impact at rack top: .o (7) Maximum rack-to-rack impact at baseplate: .o (8) Maximum rack-to-rack impact at rack top: 6.1 MAXIMUM CORNER DISPLACEMENTS {in.)

Location: X-direction Y-direction Top corner: .1237 .0400 Baseplate corner: .0033 .0028 MAXIMUM STRESS FACTORS

• stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: .011 . oc:i8 .029 .033 .057 .065 .007 Support pedestal: .090 .008 .016 .020 .098 .100 .007

• See Section 6.5.2.3 of the Licensing Report for definitions.

• Table 6.7.29

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-B4 Holtec Run I.D.: drb4obeo.rh2 Seismic Loading: OBEx1.1 Fuel Assembly I.D. and Weight: 1700#17x17 .

I 1 7 OO* O ( lbs * )

Fuel Loading: 72 cells loaded; Fuel centroid X,Y: 27.2, .o (in.)

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov * $

$Revision: 2.5 $

$Logf ile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logf ile: C:/racks/dynamO/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 152918.1 (2) Maximum vertical load in any single pedestal: 73540.0 (3) Maximum shear load in any single pedestal: 4979.1 (4) Maximum fuel-cell impact at one local position: 450.6 (5) Maximum rack-to-wall impact at baseplate: .o (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: 6.1 MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .1237 .0398 Baseplate corner: .0033 .0028 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: .011 .008 .029 .034 .057 .066 .007 Support pedestal: .090 .008 .018 .018 .099 .101 .007

• See Section 6.5.2.3 of the Licensing Report for definitions.

/

Table 6.7.30

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-B4 Holtec Run I.D .. : drb4obeo.re8 Seismic Loading: OBEx1.1 Fuel Assembly I.D. and Weight: 12x170# .

I 1 7 O. O ( lbs . )

Fuel Loading: 12 cells loaded; Fuel centroid X,Y: 49.9, . o (in.)

Coefficient of friction at the bottom of support pedestal: 0.8

$Revision: 3.46 $

$Logf ile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynam0/dynas1.fov $

$Revision: 3.36 $

$Logf ile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

(1) Maximum total vertical pedestal load: 21250.3 (2) Maximum vertical load in any single pedestal: 9111.2 (3) Maximum shear load in any single pedestal: 658.9 (4) Maximum fuel-cell impact at one local position: .6 (5) Maximum rack-to-wall impact at baseplate: .o (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: .0129 .0094 Baseplate corner: .0007 .0006 MAXIMUM STRESS FACTORS

• Stress factor: R1 R2 R3 R4 RS R6 R7 Above baseplate: .006 .001 .006 .007 .016 .018 .001 Support pedestal: .011 .001 .003 .003 .013 .013 .001

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.31

SUMMARY

RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: RACK-B4 Holtec Run I.D.: drb4obeo.re2 Seismic Loading: OBExl.1 Fuel Assembly I.D. and Weight: 12xl70# .

I 17 o. o ( lbs * )

Fuel Loading: 12 cells loaded; Fuel centroid X,Y: 49.9, . O (in. )

Coefficient of friction at the bottom of support pedestal: 0.2

$Revision: 3.46 $

$Logfile: C:/racks/dynamO/dynamO.fov $

$Revision: 2.5 $

$Logfile: C:/racks/dynamO/dynasl.fov $

$Revision: 3.36 $

$Logfile: C:/racks/dynam0/dynas2.fov $

DYNAMIC IMPACT LOADS (lbs.)

{l} Maximum total vertical pedestal load: 21250.3

{2} Maximum vertical load in any single pedestal: 9111.2

{3} Maximum shear load in any single pedestal: 814.1 (4) Maximum fuel-cell impact at one local position: .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: .o MAXIMUM CORNER DISPLACEMENTS (in.)

Location: X-direction Y-direction Top corner: .0129 .0094 Baseplate corner: .0007 .0006 MAXIMUM STRESS FACTORS

• Stress factor: Rl R2 R3 R4 RS R6 R7 Above baseplate: .006 .001 .006 .007 .016 .018 .001 Support pedestal: .011 .001 .003 .003 .013 .013 .001

• See Section 6.5.2.3 of the Licensing Report for definitions.

Table 6.7.32 COMPARISON OF CALCULATED AND ALLOWABLE LOADS/STRESSES AT IMPACT LOCATIONS AND AT WELDS FOR REGULAR FUEL LOADING VALUES ITEM/LOCATION CALCULATED ALLOWABLE Fuel assembly/cell 814 2761 wall impact, lbs.

Rack/baseplate weld 14300 27800 psi Pedestal/baseplate 0.864 1. 0 weld (dimensionless limit load ratio)

Cell/cell welds, lbs. 1243 9829

Table 6.8.1 MAXIMUM DISPLACEMENTS FROM WHOLE POOL MULTI RACK RUNS SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling DBE( =1.lxDBE );

Run I.D.: dwpmr405.rfr Rack uxt uyt uxb uyb No (in.) (in.) (in.) (in.)

1 .3660E+OO .6139E+OO .1069E+OO .1188E+OO 2 .4822E+OO .5608E+OO .2662E+OO .3397E+OO 3 .3301E+OO .4422E+OO .1072E+OO .1097E+OO 4 .4109E+OO .3831E+OO .6578E-Ol .7133E-Ol 5 .3113E+OO .2231E+OO .9315E-Ol .9255E-Ol 6 .5953E+OO .6087E+OO .3323E+OO .3587E+OO 7 .3699E+OO .3490E+OO .3387E-Ol .2900E-Ol 8 .6080E+OO .6381E+OO .3364E+OO .3371E+OO 9 .6560E+OO .6259E+OO .4767E+OO .2948E+OO 10 .4168E+OO .6340E+OO .2653E+OO .2429E+OO 11 .2710E+OO .9601E+OO .2728E-Ol .8746E-Ol 12 .4479E+OO .5769E+OO .3012E+OO .3455E+OO

$Revision: 1.8 $

$Logf ile: C:/racks/multirac/maxdisp.fov $

( Table 6

• 8 . 2 , file: impacts . rfr ) -~:.* _ 1 Table 6.8.2 MAXIMUM IMPACT FORCE OF EACH GAP ELEMENT SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling DBE( =1.lxDBE ) ;

Run I.D.: dwpmr405.rfr GAP ELEMENT MAX.FORCE TIME (lb.) (sec.)

RACK-1:

1 1.506E+05 5.989E+OO 2 1.703E+05 1. 251E+Ol 3 1. 443E+05 8.428E+OO 4 1.803E+05 1. 853E+Ol 5 9.117E+04 1. 722E+Ol 6 8.811E+04 5.219E+OO 7 7.038E+04 9.829E+OO 8 8.514E+04 9.461E+OO RACK-2:

9 1.523E+05 5.362E+OO 10 1. 625E+05 1.448E+Ol 11 1. 602E+05 1.273E+Ol 12 1. 734E+05 1.605E+Ol 13 8.186E+04 5.559E+OO 14 8.472E+04 5.222E+OO 15 9.626E+04 1.174E+Ol 16 9.259E+04 1. 206E+Ol RACK-3:

17 1. 839E+05 1.227E+Ol 18 1.479E+05 1. 297E+Ol 19 1. 843E+05 5.123E+OO 20 l.444E+05 1. 736E+Ol 21 8.221E+04 1.257E+Ol 22 7.884E+04 5.252E+OO 23 7.816E+04 1.392E+Ol 24 7.439E+04 9.434E+OO

( Table 6.8.2, file: impacts.rfr, continued ) 2 RACK-4:

25 1.964E+05 1.136E+Ol 26 1.519E+05 1.566E+Ol 27 1. 899E+05 1.075E+Ol 28 1.707E+05 1.498E+Ol 29 8.419E+04 5.555E+OO 30 8.610E+04 5.223E+OO 31 7.009E+04 9.886E+OO 32 7.871E+04 1.308E+Ol RACK-5:

33 1.424E+05 9.084E+OO 34 1.481E+05 1.387E+Ol 35 1. 388E+05 8.511E+OO 36 1. 375E+05 1.736E+Ol 37 8.573E+04 5.556E+OO 38 1. 031E+05 9.637E+OO 39 9.380E+04 1.401E+Ol 40 8.386E+04 1.434E+Ol RACK-6:

41 1. 611E+05 1.107E+Ol 42 1. 818E+05 1.351E+Ol 43 1.622E+05 1.274E+Ol 44 1. 882E+05 1.298E+Ol 45 8.843E+04 1.147E+Ol 46 1.087E+05 9.662E+OO 47 7.796E+04 8.882E+OO 48 8.782E+04 1.304E+Ol RACK-7:

49 1. 905E+05 9.182E+OO 50 1. 350E+05 1.306E+Ol 51 1.799E+05 1.065E+Ol 52 1.486E+05 1.351E+Ol 53 8.031E+04 5.585E+OO 54 1. 019E+05 9.639E+OO 55 6.482E+04 1.271E+Ol 56 8.146E+04 1.306E+Ol RACK-8:

57 1. 992E+05 1.127E+Ol 58 1. 649E+05 1.661E+Ol 59 1.836E+05 1.074E+01 60 1.634E+05 1.294E+Ol 61 8.690E+04 1.250E+01 62 7.885E+04 1.216E+Ol 7.924E+04 9.879E+OO 63 64 8.702E+04 9.538E+OO

• • ( Table 6.8.2, file:

RACK-9:

65 66 impacts.rfr, continued )

1.557E+05 2.205E+05 1.786E+01

1. 352E+01 3

67 1.883E+05 1.147E+01 68 1. 995E+05 1.295E+Ol 69 9.006E+04 5.570E+OO 70 8.893E+04 5.233E+OO 71 1. 061E+05 1.403E+Ol 72 9.707E+04 1. 436E+Ol RACK-10 { EXISTING RACK ) :

73 1. 340E+05 1.008E+Ol 74 1. 246E+05 1.134E+Ol 75 1.218E+05 1.155E+Ol 76 1.152E+05 1.180E+Ol 77 7.026E+04 1.118E+Ol 78 3.376E+04 1.422E+Ol 79 6.511E+04 1.180E+Ol 80 7.014E+04 9.215E+OO 81 7.468E+04 9.491E+OO 82 7.610E+04 l.389E+Ol 83 6.627E+04 1.297E+Ol RACK-11 { EXISTING ~CK ) :

84 1. 615E+05 1.218E+Ol 85 1. 488E+05 9.899E+OO 86 1.412E+05 7.734E+OO 87 1. 396E+05 l.032E+Ol 88 6.096E+04 9.934E+OO 89 3.267E+04 7.370E+OO 90 6.338E+04 5.971E+OO 91 5.495E+04 1.709E+Ol 92 6.175E+04 l.381E+Ol 93 6.998E+04 1.166E+Ol 94 7.193E+04 1.138E+Ol RACK-12 { .EXISTING RACK ) :

95 1.233E+05 1. 052E+Ol 96 l.489E+05 1. 670E+Ol 97 1.163E+05 1.146E+Ol 98 l.169E+05 1.633E+Ol 99 6.083E+04 1. 694E+Ol 100 3.380E+04 9.481E+OO 101 6.826E+04 1.134E+Ol 102 7.202E+04 1. 712E+Ol 103 6.468E+04 9.718E+OO 104 5.485E+04 5.0SOE+OO 105 6.240E+04 4.782E+OO

( Table 6.8.2, file: impacts.rfr, continued ) 4 IMPACT SPRINGS BETWEEN RACKS AND RACK/WALL AT RACK TOP:

106 O.OOOE+OO O.OOOE+OO RACK/WALL 107 O.OOOE+OO O.OOOE+OO R/W 108 O.OOOE+OO O.OOOE+OO R/W 109 O.OOOE+OO O.OOOE+OO R/W 110 O.OOOE+OO O.OOOE+OO R/W 111 O.OOOE+OO O.OOOE+OO R/W 112 O.OOOE+OO O.OOOE+oo* R/W 113 4.710E+03 1. 808E+Ol 114 1~362E+03 5.261E+OO 115 O.OOOE+OO O.OOOE+OO R/W 116 O.OOOE+OO O.OOOE+OO R/W 117 3.048E+03 1.268E+Ol 118 1. 357E+03 5.261E+OO 119 O.OOOE+OO O.OOOE+OO R/W 120 4.238E+03 l.201E+Ol 121 4.643E+03 i.201E+Ol 122 3.471E+03 l.622E+Ol 123 3.527E+03 1.612E+Ol 124 2.989E+03 l.351E+Ol 125 5.603E+03 l.351E+Ol 126 O.OOOE+OO O.OOOE+OO R/W 127 2.990E+03 1.140E+Ol 128 5.669E+03 1. 392E+Ol 129 O.OOOE+OO O.OOOE+OO R/W 130 O.OOOE+OO O.OOOE+OO R/W 131 2.662E+03 1.140E+Ol 132 3.363E+03 1.285E+Ol 133 O.OOOE+OO O.OOOE+OO R/W 134 2.384E+03 8.818E+OO 135 2.384E+03 8.818E+OO 136 O.OOOE+OO O.OOOE+OO 137 2.019E+03 1. 058E+Ol 138 1. 752E+03 5.641E+OO 139 1. 789E+03 5.642E+OO 140 O.OOOE+OO O.OOOE+OO R/W 141 5.550E+03 1.301E+Ol 142 2.196E+03 1.142E+Ol 143 O.OOOE+OO O.OOOE+OO R/W 144 O.OOOE+OO O.OOOE+OO R/W 145 3.875E+03 1. 304E+Ol 146 3.763E+03 1.142E+Ol 147 O.OOOE+OO O.OOOE+OO R/W

• ( Table 6.8.2, file: impacts.rfr, continued )

148 149 150 O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO HOLTEC/EXIST.

O.OOOE+OO O.OOOE+OO H/E 5

H/E 151 O.OOOE+OO O.OOOE+OO H/E 152 ,Q. OOOE+OO O.OOOE+OO H/E 153 O.OOOE+OO O.OOOE+OO H/E 154 O.OOOE+OO O.OOOE+OO R/W 155 O.OOOE+OO O.OOOE+OO 156 O.OOOE+OO O.OOOE+OO 157 O.OOOE+OO O.OOOE+OO R/W 158 O.OOOE+OO O.OOOE+OO R/W 159 O.OOOE+OO O.OOOE+OO 160 O.OOOE+OO O.OOOE+OO 161 O.OOOE+OO O.OOOE+OO R/W 162 O.OOOE+OO O.OOOE+OO R/W 163 O.OOOE+OO O.OOOE+OO R/W 164 O.OOOE+OO O.OOOE+OO R/W 165 O.OOOE+OO O.OOOE+OO R/W 166 O.OOOE+OO O.OOOE+OO R/W 167 O.OOOE+OO O.OOOE+OO R/W IMPACT SPRINGS BETWEEN RACKS AND RACK/WALL AT RACK BOTTOM:

168 O.OOOE+OO O.OOOE+OO R/W 169 O.OOOE+OO O.OOOE+OO R/W 170 O.OOOE+OO O.OOOE+OO R/W 171 O.OOOE+OO O.OOOE+OO R/W 172 O.OOOE+OO O.OOOE+OO R/W 173 O.OOOE+OO O.OOOE+OO R/W 174 O.OOOE+OO O.OOOE+OO R/W 175 1.139E+04 1.685E+Ol 176 O.OOOE+OO O.OOOE+OO 177 O.OOOE+OO O.OOOE+OO R/W 178 O.OOOE+OO O.OOOE+OO R/W 179 O.OOOE+OO O.OOOE+OO 180 O.OOOE+OO O.OOOE+OO 181 O.OOOE+OO O.OOOE+OO R/W 182 O.OOOE+OO O.OOOE+OO 183 O.OOOE+OO O.OOOE+OO 184 9.409E+03 l.634E+Ol 185 3.893E+03 1.951E+Ol 186 O.OOOE+OO O.OOOE+OO 187 8.126E+03 l.351E+Ol 188 O.OOOE+OO O.OOOE+OO R/W 189 O.OOOE+OO O.OOOE+OO 190 1. 093E+04 l.391E+Ol

• ( Table 6.8.2, file: impacts.rfr, continued )

191 O.OOOE+OO O.OOOE+OO R/W 6

192 O.OOOE+OO O.OOOE+OO R/W 193 O.OOOE+OO O.OOOE+OO 194 O.OOOE+OO O.OOOE+OO 195 O.OOOE+OO O.OOOE+OO R/W 196 O.OOOE+OO O.OOOE+OO 197 O.OOOE+OO O.OOOE+OO 198 O.OOOE+OO O.OOOE+OO 199 O.OOOE+OO O.OOOE+OO 200 O.OOOE+OO O.OOOE+OO 201 O.OOOE+OO O.OOOE+OO 202 O.OOOE+OO O.OOOE+OO R/W 203 9.959E+03 1.089E+Ol 204 O.OOOE+OO O.OOOE+OO 205 O.OOOE+OO O.OOOE+OO R/W 206 O.OOOE+OO O.OOOE+OO R/W 207 6.378E+03 1.597E+Ol 208 6.076E+03 1.838E+Ol 209 O.OOOE+OO O.OOOE+OO R/W 210 O.OOOE+OO O.OOOE+OO HOLTEC/EXIST.

211 O.OOOE+OO O.OOOE+OO H/E 212 9.211E+03 1.745E+Ol H/E 213 6.324E+03 1.578E+Ol H/E 214 2.698E+03 1.873E+Ol H/E 215 8.826E+03 1.686E+Ol H/E 216 O.OOOE+OO O.OOOE+OO R/W 217 O.OOOE+OO O.OOOE+OO 218 O.OOOE+OO O.OOOE+OO 219 O.OOOE+OO O.OOOE+OO R/W 220 -0.000E+OO O.OOOE+OO R/W 221 O.OOOE+OO O.OOOE+OO 222 O.OOOE+OO O.OOOE+OO 223 O.OOOE+OO O.OOOE+OO R/W 224 O.OOOE+OO O.OOOE+OO R/W 225 O.OOOE+OO O.OOOE+OO R/W 226 O.OOOE+OO O.OOOE+OO R/W 227 O.OOOE+OO O.OOOE+OO R/W 228 O.OOOE+OO O.OOOE+OO R/W 229 O.OOOE+OO O.OOOE+OO R/W FILE INFORMATION FOR THIS RUN Input File dwpmr405.rfr Plot File fwpmr.rfr X-Seismic a-t-dbe.hl Y-Seismic a-t-dbe.h2 Z-Seismic a-t-dbe.vt Output File owpmr405.rfr SALEM,UNIT-1,WPMR,12 Racks,df=dwpmr405.rfr,1.1xDBE,dt=.00005

• 1 Table 6.8.3 MAXIMUM PEDESTAL STRESS FACTORS OF ALL RACKS IN POOL SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: controlling DBE( =l.lxDBE );

Run I.D.: dwpmr405.rfr SALEM,UNIT-1,WPMR,3-exist.+ 9-Holtec racks; wpmr405.rfr.

• • • • • • • • • • • • • • • • • • INPUT DATA ******************

File name of FX Time history frictiox File name of FY Time history frictioy File name of FV Time history pltfwlO

$Revision: 1.0 $

$Logfile: C:/racks/multirac/sfmr2.fov $

$Date: 28 May 1992 18:08:26 $

File name of result output sfwpmr.rfr Number of racks in the pool 12 Height of the pedestal, in. 15.25 Offset of FV from center, in. 1.30 Area of female pedestal, in**2. 63.86 Inertia of female pedestal, in**4. :1208.00 Distance of extrame fiber in X, in. 4.50 Distance of extrame fiber in Y, in. 4.50 Yield stress of female pedestal, psi. 21300.

Number of pedestals of each rack 4 4 4 4 4 4 4 4 4 ~ 6 6 MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR ALL RACK PEDESTALS Maximum Values of Rl -- R7 Rl,R3,R4 for Ma.x.R6 Rl R2 R3 R4 RS R6 R7 R1R6M R3R6M R4R6M

.270 .158 .327 .316 .660 .730 .163 .265 .280 .185 Rack No.:

9 6 8 6 9 9 8 Pedestal No. :

2 2 1 2 2 2 1 Time (sec.):

13.520 13.390 10.200 13.380 13.510 13.510 10.200 L ___ _

( Table 6.8.3, continued ) 2 l) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 1 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.185 .068 .214 .122 .346 .378 .109 .164 .214 .000 Time (sec.):

5.990 8.150 8.860 8.190 8.860 8.860 8.860 Pedestal-2:

.208 .077 .145 .134 .362 .391 .086 .200 .076 .116 Time (sec. ) :

12.510 13.690 15.610 13.690 18.070 18.070 12.510 Pedestal-3:

.176 .099 .196 .196 .332 .363 .100 .156 .029 .179 Time (sec. ) :

8.430 7.660 15.550 7.660 7.640 7.640 15.550 Pedestal-4:

.221 .116 .212 .224 .501 .554 .105 .194 .142 .219 Time (sec.):

18.530 17.350 5.900 17.340 17.480 17.480 5.900

2) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 2 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 R1R6M R3R6M R4R6M Pedestal-1:

.186 .102 .211 .191 .443 .490 .109 .173 .199 .119 Time (sec. ) :

5.360 5.380 14.130 5.380 14.140 14.140 14.130 Pedestal-2:

.198 .087 .244 .171 .431 .474 .120 .185 .200 .089 Time (sec. ) :

14.480 13.300 9.530 19.790 16.620 16.620 9.510 Pedestal-3:

.196 .124 .194 .239 .530 .589 .099 .196 .154 .239 Time (sec. ) :

12.730 12.730 15.680 12.730 12.730 12.730 16.620 Pedestal-4:

.212 .066 .154 .107 .343 .366 .091 .212 .154 .000 Time (sec.):

16.050 13.860 16.050 13.850 16.050 16.050 16.050

• ( Table 6.8.3, continued )

3) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, Maximum Values of Rl -- R7

~OR EACH PEDESTAL OF RACK- 3 Rl,R3,R4 for Max.R6 3

Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.225 .108 .196 .203 .469 .Sl4 .107 .212 .162 .140 Time (sec. ) :

12.270 12.190 12.360 12.160 12.330 12.330 12.360 Pedestal-2:

.181 .102 .216 .196 .365 .397 .112 .181 .216 .000 Time (sec.):

12.970 12.160 12.970 12.160 12.970 12.970 12.970 Pedestal-3:

.22S .145 .224 .288 .535 .S93 .118 .200 .223 .170 Time (sec. ) :

5.120 11. 730 4.040 11. 730 4.080 4.050 4.050 Pedestal-4:

.176 .117 .23S .243 .493 .552 .117 .158 .208. .186 Time (sec. ) :

17.360 11. 640 12.450 11.640 12.590 12.590 12.450 4} MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 4 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.240 .099 .217 .174 .433 .467 .110 .240 .063 .164 Time (sec. ) :

11.360 11.380 9.010 11.460 11.360 11.360 9.010 Pedestal-2:

.186 .071 .205 .133 .351 .381 .103 .172 .181 .029 Time (sec. ) :

15.660 14.310 9.490 14.310 15.520 15.510 9.490 Pedestal-3:

.233 .099 .247 .184 .483 .529 .130 .223 .231 .076 Time (sec.) :

10.7SO 11.840 10.640 11.840 10.740 10.740 10.640 Pedestal-4:

.208 .099 .209 .189 .436 .477 .106 .201 .103 .173 Time (sec.):

14.980 lS.080 8.860 5.850 lS.100 15.100 8.860

• ( Table 6.8.3, continued )

7) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 7 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 5

Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.233 .086 .170 .138 .476 ~519 .096 .229 .153 .138 Time (sec.):

9.180 9.170 S.460 9.170 9.170 9.170 9.190 Pedestal-2:

.165 .083 .180 .155 .318 .351 .092 .134 .180 .037 Time (sec.):

13.060 13.100 16.680 13.100 13.100 16.680 16.680 Pedestal-3:

.219 .149 .168 .305 .sos .556 .090 .216 .111 .230 Time (sec.): -

10.650 10.620 4.100 10.620 10.740 *10.740 16.600 Pedestal-4:

.181 .099 .210 .198 .388 .426 .107 .169 .186 .071 Time (sec.):

10.520 12. 620 17.120 12.620 17.250 17.250 17.120

8) MAXIMUM VALUE.S OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 8 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 - R7 RlR6M R3R6M R4R6M Pedestal-1:

.244 .089 .327 .149 .623 .694 .163 .218 .327 .149 Time (sec.):

11.270 10.200 10.200 10.200 10.200 10.200 10.200 Pedestal-2:

.202 .119 .256 .243 ~S14 .570 .131 .195 .255 .120 Time (sec.):

16.610 14.360 16.730 14.360 16.690 16.690 16.710 Pedestal-3:

.224 .099 .281 .178 .S34 .588 .144 .224 .244 .120 Time (sec.) :

10.740 11.820 10.630 11.820 10.740 10.740 10.630 Pedestal-4:

.200 .06S .091 .095 .281 .295 .062 .200 .ooo .095 Time (sec.>):

12.940 12.940 17.360 12.940 12.940 12.940 17.360

( Table 6.8.3, continued ) 4 S) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- S Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.174 *.089 .101 .16S .304 .329 .OS6 .164 .000 .16S Time (sec. ) :

9.080 16.940 14.250 16.940 16.940 16.940 14.2SO Pedestal-2:

.181 .104 .121 .209 .316 .347 .063 .138 .000 .209 Time (sec.):

13.870 12.060 17.020 12.060 12.060 12.060 16.940 Pedeatal-3:

.170 .069 .069 .12S .274 .294 .044 .14S .027 .123 Time (sec.):

8.SlO 8.600 4.080 17.360 8.S20 8.600 4.080 Pedestal-4:

.168 .113 .lSl .232 .416 .460 .078 .132 .096 .232 Time (sec.):

17.360 12. 590 17.160 12.560 17.360 12.560 17.160

6) MAXIMUM VALUES OF.STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 6 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.197 .096 .164 .176 .339 .365 .090 .186 .014 .165 Time (sec.):

11.070 12.310 10.070 12.310 12.300 12.300 10.070 Pedestal-2:

.223 .158 .262 .316 .637 .710 .128 .221 .194 .296 Time (sec.):

13.510 13.390 9.500 13.380 13.520 13.520 9.SOO Pedestal-3:

.198 .ass .138 .14S .322 .* 344 .081 .198 .ooo .14S Time (sec.):

12.740 12.740 9.500 12.740 12.740 12.740 9.SOO Pedestal-4:

.229 .142 .212 .280 .533 .586 .112 .229 .083 .274 Time (sec.) :

12.970 12.980 14.030 12.950 12.980 12.980 14.010

( Table 6.8.3, continued ) 6

9) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 9 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.191 .110 .205 .212 .409 .448 .106 .183 .061 .205 Time (sec.):

17.860 12.000 5.190 12.070 11.980 11.980 5.190 Pedestal-2:

.270 .137 .280 .276 .660 .730 .150 .265 .280 .185 Time (sec.):

13.520 12.080 13.510 12.080 13.510 13.510 13.520 Pedestal-3:

.230 .135 .205 .255 .484 .529 .107 .230 .043 .255 Time (sec.):

11.470 11.470 9.430 11.470 11.470 11.470 9.430 Pedestal-4:

.244 .088 .231 .162 .489 .532 .127 .241 .177 .114 Time (sec.):

12.940 17.230 12.960 17.230 12.930 12.930 12.960 (10) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R.7, FOR EACH PEDESTAL OF RACK-10 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 R1R6M R3R6M R4R6M Pedestal-1:

.164 .081 .170 .146 .371 .408 .091 .161 .142 .105 Time (sec.):

10.080 10.190 10.080 10.190 10.100 10.100 10.080 Pedestal-2:

.152 .084 .161 .161 .380 .420 .084 .151 .132 .138 Time (sec.):

11.330 11.320 14.400 11.310 11.350 11.350 14.400 Pedestal-3:

.149 .074 .165 .137 .369 .408 .086 .149 .163 .096 Time (sec.):

11.550 11. 620 9.500 11. 620 11. 550 . 11. 550 11. 550 Pedestal-4:

.141 .075 .108 .139 .315 .346 .058 .141 .072 .132 Time (sec.):

11.800 11. 790 9.970 11. 790 11.800 11.800 14.960 Pedestal-5:

.086 .052 .064 .099 .184 .201 .034 .083 .030 .088 Time (sec.):

11.180 11.190 11.330 11.190 11.180 11.170 10.210 Pedestal-6:

.041 .024 .* 044 .044 .092 .101 .023 .040 .025 .036 Time (sec.):

14.220 14.220 13.770 14.220 14.220 10.290 13. 770

• ( Table 6.8.3, Rl R2 continued )

(11) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK-11 Maximum Values of Rl -- R7 R3 R4 RS R6 R7 Rl,R3,R4 for Max.R6 RlR6M R3R6M R4R6M 7

Pedestal-1:

.197 .071 .239 .12S .460 .S06 .124 .197 .239 .071 Time (sec. ) :

12.180 9 .010 12.190 9.990 12.180 12.180 12.190 Pedestal-2:

.183 .090 .206 .164 .488 .S41 .108 .183 .204 .lSS Time (sec. ) :

9.900 9.910 9.880 9.920 9.900 9.900 9.900 Pedestal-3:

.173 .099 .214 .194 .* 409 .452 .111 .161 .15S .137 Time (sec.):

. 7. 730 12.730 7.730 12.730 12.630 12.630 7.730 Pedestal-4:

.170 .07S .276 .133 .491 .S47 .136 .170 .276 .101 Time (sec. ) :

10.320 10.370 10.320 10.370 10.320 10.320 10.320 Pedestal-S:

.074 .OSl .OS7 .100 .160 .175 .030 .074 .000 .100 Time (sec. ) :

9.930 9.930 10. 020 9.930 9.930 9.930 10.020 Pedestal-6:

.040 .026 .047 .oso .089 .098 .024 .037 .041 .021 Time (sec.):

7.370 13.120 16.020 13.120 S.lSO 5.150 16.020 (12) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK-12 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.151 .067 .193 .117 .411 .458 .099 .148 .192 .117 Time (sec.):

10.520 10.530 10.520 10.530 10.530 10.530 10.520 Pedestal-2:

.183 .104 .193 .19S .426 .469 .100 .183 .091 .195 Time (sec.):

16.700 16.700 10.980 16.700 16.700 16.700 10.980 Pedestal-3:

.142 .078 .138 .149 .354 .391 .074 .141 .131 .119 Time (sec.):

11.460 11.440 7. 720 11.440 11.S40 11.540 7. 720 Pedestal-4:

.143 .071 .160 .133 .324 .356 .084 .143 .140 .073 Time (sec.):

10.300 10.190 10.310 10.180 16.330 16.330 10.310 Pedestal-S:

.074 .060 .062 .122 .187 .208 .033 .068 .049 .091 Time (sec. ) :

16.940 16.940 16.590 16.940 16.790 16.790 16.S90 Pedestal-6:

.041 .026 .OSl .OS2 .100 .112 .025 .037 .038 .037 Time (sec. ) :

9.480 12.lSO 16.030 12.lSO S.090 S.090 16.030

/*

Table 6.8.4 RESULTS OF POOL WALL DYNAMIC PRESSURES SALEM NUCLEAR GENERATING STATION, ONIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling DBE( =1.lxDBE );

Run I.D.: dwpmr405.rfr

$Revision: 1.0 $

$Logfile: C:/racks/multirac/wallpres.fov $

(1) AVERAGE DYNAMIC PRESSURES ON POOL WALLS (psi.) :

Average dynamic pressure on (-x) wall: -1.105161E-02 Average dynamic pressure on (+x) wall: -1.105161E-02 Average dynamic pressure on (-y) wall: 6.286292E-03 Average dynamic pressure on (+y) wall: 6.286292E-03 (2) PEAK DYNAMIC PRESSURES ON POOL WALLS (psi.):

Positive peak pressure on (-x) wall: 5.520000 Negative peak pressure on (-x) wall: -6.470000 Positive peak pressure on (+x) wall: 5.520000 Negative peak pressure on (+x) wall: -6.470000 Positive peak pressure on (-y) wall: 5.660000 Negative peak pressure on (-y) wall: -5.450000 Positive peak pressure on (+y) wall: 5.660000 Negative peak pressure on (+y) wall: -5.450000 (3) DYNAMIC PRESSURE ADDERS ON POOL WALLS (psi.>":

Positive pressure adder on (-x) wall: 1. 530816 Negative pressure adder on (-x) wall: -1. 635854 -

Positive pressure adder on (+x) wall: 1.530816 Negative pressure adder on (+x) wall: -1. 635854 Positive pressure adder on (-y) wall: 1. 539086 Negative pressure adder on (-y) wall: -1. 633054 Positive pressure adder on (+y) wall: 1.539086 Negative pressure adder on (+y) wall: -1.633054 (4) NUMBER OF TIME POINTS:

• Total number of time points in file: 2001

• Number of time points where dynamic pressure on (-x) wall was positive: 1027

• Number of time points where dynamic pressure on (-x) wall was negative: 974

• Number of time points where dynamic pressure on (+x) wall was positive: 1027

• Number of time points where dynamic pressure on (+x) wall was negative: 974

• Number of time points where dynamic pressure on (-y) wall was positive: 1034'

• Number of time points where dynamic pressure on (-y) wall was negative: 967

• Number of time points where dynamic pressure on (+y) wall was positive: 1034

• Number of time points where dynamic pressure on (+y) wall was negative: 967

( Table 6.8.5; file: ftload.rfr) 1 Table 6.8.5 STATIC. LOAD AND DYNAMIC LOAD ADDER FOR EACH PEDESTAL SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling DBE( =l.lxDBE );

Run I.D.: dwpmr405.rfr RACK FOOT NCOUNT DYNAMIC ADDER STATIC LOAD No. No. (lbs.) (lbs.)

1 1 918 24689.980000 57800.000000 1 2 973 31551.390000 57800.000000 1 3 977 26044.220000 57800.000000 1 4 982 32942.460000 57800.000000 2 1 998 28771.840000 57800.000000 2 2 973 30634.330000 57800.000000 2 3 938 29957.250000 57800.000000 2 4 933 26631.300000 57800.000000 3 1 979 33996.830000 62600.000000 3 2 937 22261.470000 62600.000000 3 3 983 35843.840000 62600.000000 3 4 904 21053.100000 62600.000000 4 1 1020 27619.020000 57800.000000 4 2 939 26726.200000 57800.000000 4 3 931 29940.600000 57800.000000 4 4 952 2115a.820000 57800.000000 5 1 962 24503.330000 57800.000000 5 2 926 19584.230000 57800.000000 5 3 1010 22663.960000 57800.000000 5 4 1045 20148.520000 57800.000000 6 1 934 31651.390000 62600.000000 6 2 966 32242.550000 62600.000000 6 3 909 29225.960000 62600.000000 6 4 972 31062.140000 62600.000000 7 1 919 33628.950000 57800.000000 7 2 873 21945.360000 57800.000000 7 3 1045 32772.340000 57800.000000 7 4 999 23191.390000 57800.000000

( Table 6.8.5; file: ftload.rfr, continued ) 2 RACK FOOT NCOUNT DYNAMIC ADDER STATIC LOAD No. No. (lbs.) (lbs.)

8 1 984 26959.550000 57800.000000 8 2 954 28019.710000 57800.000000 8 3 905 26673.260000 57800.000000 8 4 889 26021. 040000 57800.000000 9 1 932 36662.020000 62600.000000 9* 2 933 39277.600000 62600.000000 9 3 927 34834.630000 62600.000000 9 4 984. 35500.410000 62600.000000 10 1 976 26574.490000 23300.000000 10 2 1078 22923.190000 23300.000000 10 3 1106 29275.590000 23300.000000 10 4 952 20176.890000 23300.000000 10 5 963 9589.512000 23300.000000 10 6 1039 2832.916000 23300.000000 11 1 997 44468.910000 23300.000000 1.1 2 1078 28803.340000 23300.000000 11 3 1063 46112.420000 23300.000000 11 4 928 30424.250000 23300.000000 11 5 746 9653.753000 23300.000000 11 6 664 2921.084000 23300.000000 12 1 974 23005.850000 23300.000000 12 2 1060 28467.260000 23300.000000 12 3 1070 27786.920000 23300.000000 12 4 987 25925.130000 23300.000000 12 5 954 11476.520000 23300.000000 12 6 921 2862.649000 23300.000000

• Table 6.8.6 TOTAL STATIC LOAD AND DYNAMIC ADDER ON THE WHOLE SLAB SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling DBE( =1.lxDBE ) ;

Run I.D.: dwpmr405.rfr (1) THE TOTAL STATIC LOAD OF RACKS AND FUEL ON SLAB IS:

2627700.00 lbs.

(2) THE TOTAL DYNAMIC LOAD ADDER ON THE SLAB IS:

301805.36 lbs.

Notes:

(1) Total static load is modified to include the static loads of the 7-th pedestals of the three existing racks:

2557800.0 + 3 x 23300.0 = 2627700.0 lbs.

(2) Total dynamic load adder is conservatively modified to include the dynamic adders of the 7-th pedestals of the three existing racks:

267375.8 + 3 x 11476.52 = 301805.36 lbs.

where 23300.0 is the static vertical load of each pedestal of the existing racks; and 11476.52 is the maximum vertical load adder of each pedestal of the existing racks (data from file: ftload.rfr ).

Table 6.8.7 MAXIMUM DISPLACEMENTS FROM WHOLE POOL MULTI-RACK RUN SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Holtec Racks A3, B3 and B6 Half loaded; Others Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling DBE( =l.lxDBE );

Run I.D.: dwpmr405.rhr Rack uxt uyt uxb uyb No. (in.) (in.) (in.) (in.)

1 *. 3105E+OO .4430E+OO .2747E-Ol .3701E-Ol 2 .3931E+OO .4383E+OO .2279E+OO .2392E+OO 3 .2586E+OO .3860E+OO .3459E-Ol .3901E-Ol 4 .2695E+OO .3999E+OO .4754E-Ol .5704E-Ol 5 .2051E+OO .1912E+OO .1832E-Ol .1642E-Ol 6 .3351E+OO .3463E+OO .7908E-Ol .lOSOE+OO 7 .7146E+OO .3760E+OO .1296E+OO .1028E+OO 8 .7618E+OO .6039E+OO .3998E+OO .2599E+OO 9 .8691E+OO .5483E+OO .6401E+OO .2586E+OO 10 .2820E+OO .3940E+OO .2400E-Ol .3301E-Ol 11 .2762E+OO .1213E+Ol .5619E-Ol .1093E+OO 12 .2064E+OO .5187E+OO .2888E-Ol .4868E-Ol

$Revision: 1.8 $

$Logfile: C:/racks/multirac/maxdisp.fov $

1 Table 6.8.8 MAXIMUM IMPACT FORCE OF EACH GAP ELEMENT SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Holtec Racks A3, BJ and B6 Half loaded; Others Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling DBE( =1.1xDBE );

Run I.D.: dwp:mr405.rhr GAP ELEMENT MAX.FORCE TIME (lb.) (sec.)

RACK-1:

1 1. 295E+05 5.979E+OO 2 1. 767E+05 1.808E+Ol 3 1. 297E+05 7.610E+OO 4 1. 749E+05 1. 746E+Ol 5 8.799E+04 5.547E+OO 6 9.513E+04 9.675E+OO 7 7.917E+04 1.404E+Ol 8 7.635E+04 1. 439E+Ol RACK-2:

9 1.428E+05 5.358E+OO 10 1.432E+05 9.583E+OO 11 1. 331E+05 4.995E+OO 12 1. 386E+05 1.724E+Ol 13 9.696E+04 1.254E+Ol 14 9.340E+04 9.692E+OO 15 7.788E+04 1. 411E+Ol 16 8.313E+04 1. 303E+Ol RACK-3:

17 1.788E+05 4.567E+OO 18 1. 343E+OS 1. 509E+Ol 19 1. 748E+05 5.123E+OO 20 l.418E+05 1. 052E+Ol 21 8.986E+04 1.727E+Ol 22 7.760E+04 5.257E+OO 23 8.099E+04 8.921E+OO 24 7.794E+04 1. 313E+Ol

( Table 6.8.8, contonued ) 2 RACK-4:

25 1. 492E+05 1.128E+Ol 26 1. 366E+05 1.678E+Ol 27 1. 771E+05 1.325E+01 28 1.507E+05 1.736E+01 29 8.679E+04 1.250E+Ol 30 8.391E+04 5.230E+OO 31 7.744E+04 1.388E+Ol 32 6.395E+04 8.401E+OO RACK-5:

33 1. 469E+05 1.207E+Ol 34 1.272E+05 6.223E+OO 35 1. 232E+05 l.259E+Ol 36 1.173E+05 8.633E+OO 37 8.647E+04 5.549E+OO 38 8.763E+04 5.219E+OO 39 9.242E+04 l.403E+Ol 40 8.763E+04 1.436E+Ol RACK-6:

41 l.419E+05 l.007E+Ol

. 42 1.689E+05 1.249E+Ol 43 1.494E+05 1. 064E+Ol 44 1. 696E+05 1.198E+Ol 45 8.734E+04 1. 266E+Ol 46 9.598E+04 9.731E+OO 47 8.214E+04 1.177E+Ol 48 8.640E+04 9.575E+OO RACK-7:

49 1.272E+05 1. 021E+Ol 50 1. 095E+05 1.241E+Ol 51 1. 228E+05 8.641E+OO 52 1. 263E+05 1. 369E+Ol 53 4.841E+04 5.427E+OO 54 5.930E+04 5.130E+OO 55 4.852E+04 1.177E+Ol 56 4.697E+04 1.208E+Ol

• ( Table 6.8.8, contonued )

RACK-8:

3 57 1. 350E+05 1.020E+01 58 1. OOOE+OS 1.660E+Ol 59 1.314E+05 1.075E+01 60 1. 016E+05 1.604E+Ol 61 4.949E+04 1.717E+Ol 62 4.281E+04 1.379E+Ol 63 4.436E+04 5.092E+OO 64 4.970E+04 4.791E+OO RACK-9:

65 1. 330E+05 l.119E+Ol 66 1.104E+05 1.249E+Ol 67 1. 304E+05 7.622E+OO 68 1.217E+05 1.297E+Ol 69 5.482E+04 9.182E+OO 70 5.528E+04 8.878E+OO 71 4.764E+04 5.lOOE+OO 72 5.290E+04 4.795E+OO RACK-1.0:

73 1. 041.E+OS 1.134E+Ol.

74 9.134E+04 1.447E+Ol.

75 1.424E+05 l..085E+Ol 76 8.446E+04 8.997E+OO 77 5.703E+04 5.255E+OO 78 3.220E+04 9.273E+OO 79 6.895E+04 1.097E+Ol 80 6.814E+04 1.242E+Ol 81 7.230E+04 1.214E+Ol 82 6.935E+04 l.401E+Ol 83 6.170E+04 4.780E+OO RACK-11:

84 1. 530E+05 9.084E+OO 85 1.426E+05 9.926E+OO 86 1. 353E+05 8.641E+OO 87 1.463E+05 1. 044E+Ol 88 6.833E+04 l..OOlE+Ol.

89 3.651.E+04 8.409E+OO 90 6.665E+04 1.063E+Ol 91 5.211.E+04 l..710E+Ol.

92 5.921E+04 1.380E+Ol 93 6.073E+04 1.164E+Ol 94 6.857E+04 l.133E+Ol

l

• ( Table 6.8.8, contonued ) 4 RACK-12:

95 1.135E+05 1.510E+Ol 96 1. 355E+05 1.107E+Ol 97 l.158E+05 1.180E+Ol 98 1. 257E+05 1.052E+Ol 99 5.240E+04 4.454E+OO 100 3.369E+04 1. 658E+Ol 101 5.512E+04 1. 604E+Ol 102 5.625E+04 1.708E+Ol 103 5.710E+04 1.210E+Ol 104 6.401E+04 1. 399E+Ol 105 6.260E+04 4.779E+OO RACK-TO-RACK/WALL IMPACT SPRINGS AT RACK TOP:

106 O.OOOE+OO O.OOOE+OO RACK/WALL 107 O.OOOE+OO O.OOOE+OO R/W 108 O.OOOE+OO O.OOOE+OO R/W 109 O.OOOE+OO O.OOOE+OO R/W 110 O.OOOE+OO O.OOOE+OO R/W 111 O.OOOE+OO O.OOOE+OO R/W 112 O.OOOE+OO O.OOOE+OO R/W 113 4.276E+03 1. 817E+Ol 114 8.129E+02 5.231E+OO 115 O.OOOE+OO O.OOOE+OO R/W 116 O.OOOE+OO O.OOOE+OO R/W 117 1. 854E+03 1. 817E+Ol 118 1. 845E+03 1.982E+Ol 119 O.OOOE+OO O.OOOE+OO R/W 120 3.073E+03 1.426E+Ol 121 2.775E+03 '1. 207E+Ol 122 3.359E+03 1.0lSE+Ol 123 1.772E+03 1.0lOE+Ol 124 2.639E+03 5.758E+OO 125 2.816E+03 5.772E+OO 126 O.OOOE+OO O.OOOE+OO R/W 127 1.482E+03 1. 267E+Ol 128 1.607E+03 1.205E+Ol 129 O.OOOE+OO O.OOOE+OO R/W 130 O.OOOE+OO O.OOOE+OO R/W 131 1. 567E+03 1.267E+Ol 132 2.441E+03 1. 205E+Ol 133 O.OOOE+OO O.OOOE+OO R/W

• ( Table 6.8.8, contonued )

134 1. 979E+03 l.126E+Ol 135 3.015E+03 1.126E+Ol 136 1. 545E+03 1. 090E+01 137 3.617E+03 1. 087E+Ol 138 1.127E+03 5.623E+OO 139 1. 526E+03 1. OOOE+Ol 140 O.OOOE+OO O.OOOE+OO R/W 141 3.858E+03 1. 321E+Ol 142 4.444E+03 1.151E+Ol 143 O.OOOE+OO O.OOOE+OO R/W 144 O.OOOE+OO O.OOOE+OO R/W 145 4.163E+03 1. 321E+Ol 146 4.687E+03 1.149E+Ol 147 O.OOOE+OO O.OOOE+OO R/W 148 O.OOOE+OO O.OOOE+OO HOLTEC/EXIST.

149 O.OOOE+OO O.OOOE+OO H/E 150 2.293E+03 9.782E+OO H/E 151 1. 361E+03 9.782E+OO H/E 152 O.OOOE+OO O.OOOE+OO H/E 153 O.OOOE+OO O.OOOE+OO H/E 154 O.OOOE+OO O.OOOE+OO R/W 155 O.OOOE+OO O.OOOE+OO 156 O.OOOE+OO O.OOOE+OO 157 O.OOOE+OO O.OOOE+OO R/W 158 O.OOOE+OO O.OOOE+OO R/W 159 O.OOOE+OO O.OOOE+OO 160 O.OOOE+OO O.OOOE+OO 161 O.OOOE+OO O.OOOE+OO R/W 162 O.OOOE+OO O.OOOE+OO R/W 163 O.OOOE+OO O.OOOE+OO R/W 164 O.OOOE+OO O.OOOE+OO R/W 165 O.OOOE+OO O.OOOE+OO R/W 166 O.OOOE+OO O.OOOE+OO R/W 167 O.OOOE+OO O.OOOE+OO R/W RACK-TO-RACK/WALL IMPACT SPRINGS AT RACK BOTTOM:

168 O.OOOE+OO O.OOOE+OO R/W 169 O.OOOE+OO O.OOOE+OO R/W 170 O.OOOE+OO O.OOOE+OO R/W 171 O.OOOE+OO O.OOOE+OO R/W 172 O.OOOE+OO O.OOOE+OO R/W 173 O.OOOE+OO O.OOOE+oo R/W 174 O.OOOE+OO O.OOOE+OO R/W

• ( Table 6.8.8, contonued ) 6 175 5.825E+03 1.817E+Ol 176 O.OOOE+OO O.OOOE+OO 177 O.OOOE+OO O.OOOE+OO R/W 178 O.OOOE+OO O.OOOE+OO R/W 179 O.OOOE+OO O.OOOE+OO 180 O.OOOE+OO O.OOOE+OO 181 O.OOOE+OO O.OOOE+OO R/W 182 O.OOOE+OO O.OOOE+OO 183 O.OOOE+OO O.OOOE+OO 184 5.619E+03 1.940E+Ol 185 O.OOOE+OO O.OOOE+OO 186 O.OOOE+OO O.OOOE+OO 187 O.OOOE+OO O.OOOE+OO 188 O.OOOE+OO O.OOOE+OO R/W 189 O.OOOE+OO O.OOOE+OO 190 O.OOOE+OO O.OOOE+OO 191 O. OOOE+OO. O.OOOE+OO R/W 192 O.OOOE+OO O.OOOE+OO R/W 193 O.OOOE+OO O.OOOE+OO 194 O.OOOE+OO O.OOOE+OO 195 O.OOOE+OO O.OOOE+OO R/W 196 O.OOOE+OO O.OOOE+OO 197 O.OOOE+OO O.OOOE+OO 198 O.OOOE+OO O.OOOE+OO 199 O.OOOE+OO O.OOOE+OO 200 O.OOOE+OO O.OOOE+OO 201 6.569E+03 9.935E+OO 202 O.OOOE+OO O.OOOE+OO R/W 203 1.159E+04 1. 682E+Ol 204 5.681E+03 1.248E+Ol 205 O.OOOE+OO O.OOOE+OO R/W 206 O.OOOE+OO O.OOOE+OO R/W 207 1. 343E+04 1.195E+Ol 208 1. 047E+04 1.178E+Ol 209 O.OOOE+OO O.OOOE+OO R/W 210 O.OOOE+OO O.OOOE+OO HOLTEC/EXIST.

211 O.OOOE+OO O.OOOE+OO H/E 212 2.178E+03 1.720E+Ol H/E 213 4.767E+03 1.254E+Ol H/E 214 1. 794E+03 1.963E+Ol H/E 215 4.144E+03 1.544E+Ol H/E 216 O.OOOE+OO O.OOOE+OO R/W 217 O.OOOE+OO O.OOOE+OO 218 O.OOOE+OO O.OOOE+OO 219 O.OOOE+OO O.OOOE+OO R/W 220 O.OOOE+OO O.OOOE+OO R/W

• ( Table 6.8.8, contonued ) 7 221 O.OOOE+OO O.OOOE+OO 222 O.OOOE+OO O.OOOE+OO 223 O.OOOE+OO O.OOOE+OO R/W 224 O.OOOE+OO O.OOOE+OO R/W 225 O.OOOE+OO O.OOOE+OO R/W 226 O.OOOE+OO O.OOOE+OO R/W 227 O.OOOE+OO O.OOOE+OO R/W 228 O.OOOE+OO O.OOOE+OO R/W 229 O.OOOE+OO O.OOOE+OO R/W FILE INFORMATION FOR THIS RUN Input File dwpmr405.rhr Plot File fwpmr.rhr x-seismic a-t-dbe.hl Y-Seismic a-t-dbe.h2 z-seismic a-t-dbe.vt Output File owpmr405.rhr SALEM, UNIT-1, WPMR, 12 Racks,df=dwpmr405.rhr,1. lxDBE,A3 ,BJ ,B6 half loaded.

• Table 6.8.9 1

MAXIMUM PEDESTAL STRESS FACTORS OF ALL RACKS IN POOL SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Holtec Racks A3, B3 and B6 Half loaded; others Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling DBE( =l.lxDBE );

Run I.D.: dwpmr405.rhr SALEM,UNIT-1,WPMR,3-exist.+ 9-Holtec racks; wpmr405.rhr.

• • • • • • • • • • • • • • • • • • INPUT DATA ******************

File name of FX Time history frictiox File name of FY Time history frictioy File name of FV Time history pltfwlO

$Revision: 1.0 $

$Logfile: C:/racks/multirac/sfmr2.fov $

$Date: 28 May 1992 18:08:26 $

File name of result output sfwpmr.rhr Number of racks in the pool 12 Height of the pedestal, in. 15.25 Offset of FV from center, in. 1.30 Area of female pedestal, in**2. 63.86 Inertia of female pedestal, in**4. :1208.00 Distance of extrame fiber in X, in. 4.50 Distance of extrame fiber in Y, in. 4.50 Yield stress of female pedestal, psi. 21300.

Number of pedestals of each rack 4 4 4 4 4 4 4 4 4 6 6 6 MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR ALL RACK PEDESTALS Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 R1R6M R3R6M R4R6M

.218 .116 .298 .226 .536 .592 .146 .214 .163 .215 Rack No.:

3 1 6 3 1 1 6 Pedestal No. :

1 4 2 3 4 4 2 Time (sec. ) :

4.570 17.450 9.500 10.600 17.460 17.460 9.500

( Table 6.8.9, continued ) 2

1) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- l Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.1S8 .081 .148 .1S8 .30S .332 .078 .137 .138 .OS6 Time (sec. ) :

S.980 13.620 8.810 13.620 8.860 8.910 8.860 Pedestal-2:

.217 .083 .21s .147 .414 .449 .llS .214 .162 .073 Time (sec. ) :

18.080 16.830 14.740 16.830 18.090 18.090 14.740 Pedestal-3:

.1S9 .095 .171 .188 .360 .398 .090 .145 .095 .1S9 Time (sec.):

7.610 7.670 12.480 7.670 13.370 13.370 12.480 r

Pedestal-4:

.214 .116 .188 .216 .S36 .S92 .101 .214 .163 .21s Time (sec.):

17.460 17.450 12.220 17.410 17.460 17.460 12.220

2) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 2 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 R1R6M R3R6M R4R6M Pedestal-1:

.. .l7S .080 .215 .142 .430 .476 .109 .172 .210 .095 Time (sec. ) :

5.360 5.340 10.060 5.310 S.420 S.420 5.420 Pedestal-2:

.17S .075 .267 .13S .422 .466 .133 .172 .267 .027 Time (sec. ) :

9.580 17.730 9.570 14.360 9.570 9.570 9.570 Pedestal-3:

.163 .072 .182 .133 .354 .392 .093 .137 .166 .089 Time (sec.) :

4.990 5.930 16.910 5.930 16.970 16.970 16.890 Pedestal-.4:

.170 .059 .118 .091 .27S .295 .069 .162 .048 .085 Time (sec.):

17.240 17.220 9.090 17.210 17.270 17.270 9.090

( Table 6.8.9, continued ) 3

3) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 3 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.218 .080 .197 .138 .416 .451 .105 .217 .110 .124 Time (sec.):

4.570 4.590 5.640* 4.s20 4.560 4.560 5.640 Pedestal-2:

.164 .068 .120 .121 .268 .293 .064 .121 .063 .109 Time (sec.):

15.090 12 .260 9.520 12.260 10.960 10.940. 9.510 Pedestal-3:

.214 .113 .216 .226 .507 .559 .112 .208 .204 .147 Time (sec.):

5.120 11.820 4.040 10.600 4.070 4.070 4.040 Pedestal-4:

.174 .094 .209 .186 .356 .392 .108 .136 .202 .054 Time (sec.):

10.520 11. 790 s. 770 11. 790 5.790 5.820 s. 770

4) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 4 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 R1R6M R3R6M R4R6M Pedestal-1:

.183 .052 .215 .097 .362 .395 .109 .175 .161 .059 Time (sec.):

11.280 6.520 8.990 13.500 13.790 13.790 8.990 Pedestal-2:

.167 .066 .210 .124 .355 .393 .105 .135 .143 .116 Time (sec.):

16.700 13.380 9.510 6.520 13.390 13.390 9.510 Pedestal-3:

.217 .081 .236 .143 .473 .519 .125 .211 .188 .120 Time (sec.):

13.250 13.220 13.260 13.170 13.230 13.230 13.250 Pedestal-4:

.184 .088 .236 .163 .391 .428 .113 .184 .099 .145 Time (sec.):

17.360 17.470 10.240 17.470 17.360 17.360 10.240

( Table 6.8.9, continued ) 4 S) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- S Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.179 .a67 .101 .1aa .3Sl .382 .a64 .173 .1a1 .1aa Time (sec.):

12.a7a 12.a6a 12.a6a 12.a6a 12.a6a 12.a6a 12.a6a Pedestal-2:

.156 .ass .091 .092 .233 .252 .as1 .129 .a31 .a92 Time (sec. ) :

6.220 7.asa 4.04a 7.a7a 7.02a 7.07a 4.a40 Pedestal-3:

.lSl .a72 .1a6 .133 .2ss .276 .ass .136 .a24 .116 Time (sec. ) :

12.S9a 7.7aa 4.040 7.7aa 7.6sa 7.6sa 4.a40 Pedestal-4:

.143 .a72 .1S6 .140 .294 .328 .a76 .099 .a94 .135 Time (sec.):

8.630 7.64a 7.7aO 7.640 7.660 7.66a 7.7aa

6) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 6 Maxunum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.174 .a84 .148 .1S3 .292 .314 .aa1 .162 .aaa .153 Time (sec. ) :

10.070 12.3Sa 9.010 12.3SO 12.3SO 12.3Sa 9.a10 Pedestal-2:

.207 .a9S .298 .17S .466 .S17 .146 .179 .163 .174 Time (sec.):

12.49a 6. 72a 9.sao 6.74a 6.71a 6. na 9.5aa Pedestal-3:

.183 .a66 .130 .113 .302 .325 .a77 .17S .ass .a62 Time (sec. ) :

la.640 l2.S3a 9.saa l2.S3a la.620 la.62a 9.saa Pedestal-4:

.207 .1a4 .2a2 .2ao .431 .477 .1a2 .169 .139 .17a Time (sec. ) :

ll.97a l2.93a 8.980 l2.93a 11.810 11.a1a 8.9Sa

• ( Table 6.8.9, continued )

7) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 7 5

Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 R1R6M R3R6M R4R6M Pedestal-1:

.156 .083 .132 .154 .382 .422 .071 .156 .121 .146 Time (sec.):

10.210 10.190 5.420 10.190 10.210 10.210 10.220 Pedestal-2:

.134 .071 .161 .129 .318 .353 .080 .119 .139 .095 Time (sec.):

12.400 12.420 5.090 12.420 13.370 13.370 5.090 Pedestal-3:

.151 .101 .161 .198 .424 .472 .084 .147 .150 .175 Time (sec. ) :

8.640 11.170 11.190 11.170 11.180 11.180 11.190 Pedestal-4:

.154 .103 .206 .208 .501 .562 .105 .154 .206 .201 Time . (sec. ) :

11.890 13. 770 13. 770 17.340 13.770 13. 770 13.770

8) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 8 Max.unum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.165 .080 .177 .146 .422 .469 .094 .156 .168 .145 Time (sec.):

10.200 10.170 10.200 11.470 10.170 10.170 10.200 Pedestal-2:

.123 .069 .160 .136 .308 .341 .082 .123 .160 .058 Time (sec. ) :

16.600 11.130 16.600 13.510 16.600 16.600 16.600 Pedestal-3:

.161 .073 .168 .129 .390 .430 .087 .161 .154 .115 Time (sec. ) :

10.750 10.880 10.640 10.880 10.750 10.750 10.640 Pedestal-4:

.124 .028 .059 .041 .174 .183 .040 .124 .059 .000 Time (sec.):

16.040 14.830 16.040 14.830 16.040 16.040 16.040

• ( Table 6.8.9, continued )

9) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7 I FOR EACH PEDESTAL OF RACK- 9 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 6

Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.163 .088 .144 .165 .323 .352 .076 .156 .037 .159 Time (sec. ) :

11.190 11.200 9.160 11.200 11.170 11.170 9.160 Pedestal-2:

.135 .086 .147 .171 .341 .378 .076 .127 .130 .120 Time (sec.) :

12.490 16.610 13.510 16.610 13.470 13.470 13.510 Pedestal-3:

.159 .078 .158 .i43 .361 .396 .085 .159 .145 .092 Time (sec. ) :

7.620 11. 630 7.730 11.630 7.620 7.620 7.730 Pedestal-4:

.148 .064 .134 .117 .285 .309 .074 .146 .128 .035 Time (sec. ) :

12.970 15.930 12.970 15.930 12.970 12.960 12.970 (10) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK-10 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.127 .044 .136 .083 .246 .268 .072 .122 .133 .013 Time (sec. ) :

11.340 11.970 10.440 11.970 10.510 10.510 10.440 Pedestal-2:

.112 .039 .131 .077 .229 .252 .067 .101 .131 .020 Time (sec.) :

14.470 11.970 9.520 11.970 9.550 9.550 9.550 Pedestal-3:

.174 .096 .140 .178 .436 .482 .080 .174 .140 .168 Time (sec. ) :

10.840 10.860 10.850 10.880 10.850 10.850 10.850 Pedestal-4:

.103 .oso .120 .095 .204 .224 .062 .088

• 051. .086 Time (sec. ) :

9.000 5.610 10.360 5.610 s.sso 5.550 10.360 Pedestal-5:

.069 .042 .042 .081 .147 .160 .024 .069 .027 .064 Time (sec. ) :

5.250 9.270 5.410 9.270 5.250 5.250 11.340 Pedestal-6:

.039 .022 .040 .042 .086 .095 .021 .037 .029 .029 Time (sec. ) :

9.270 9.270 17.670 9.270 5.150 5.150 17.670

( Table 6.8.9, continued ) 7 (11) MAXIMUM VALUES OF *STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK-11 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.187 .084 .216 .1S4 .460 .sos .112 .187 .177 .143 Time (sec.):

9.080 9.060 9.190 10.070 9.080 9.080 9.180 Pedestal-2:

.174 .09S .211 .183 .463 .514 .109 .172 .208 .135 Time (sec.):

9.920 10.000 9.920 10.000 9.970 9.970 9.920 Pedestal-3:

.16S .07S .194 .141 .380 .418 .100 .162 .150 .106 Time (sec.):

8.640 5.890 9.SlO S.890 8.660 8.660 9.SlO Pedestal-4:

.179 .081 .212 .1s2 .447 .49S .109 .17S .194 .125 Time (sec.):

10.440 10.610 10.420 10.610 10.S30 10.S30 10.420 Pedestal-S:

.083 .054 .07S .105 .192 .212 .041 .078 .062 .072 Time (sec.):

10.010 10.020 10.080 10.020 10.070 10.070 10.080 Pedestal-6:

.045 .023 .047 .044 .086 .095 .024 .036 .034 .* 025 Time (sec.):

8.410 6.340 8.840- 6.340 4.300 4.300 5.140 (12) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK-12 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 R1R6M R3R6M R4R6M Pedestal-1:

.138 .049 .163 .098 .277 .302 .085 .138 .163 .ooo Time (sec.):

15.100 10.490 15.100 10.490 15.100 .15 .100 15.100 Pedestal-2:

.165 .055 .193 .092 .362 .396 .101 .163 .193 .041 Time (sec.):

11.070 14.460 11.060 14.440 11.060 11.060 11.060 Pedestal-3:

.142 .092 .153 .179 .409 .456 .079 .142 .140 .174 Time (sec.):

11.800 11. 780 11.840 11. 780 11.800 11.800 11.820 Pedestal-4:

.154 .osa .171 .100 .300 .326 .091 .154 .171 .ooo Time (sec.) :

10.520 16.040 10.520 16.040 10.S20 10.520 10.520 Pedestal-5:

.064 .041 .066 .082 .156 .172 .034 .062 .042 .067 Time (sec.):

4.450 16.700 13.690 16.700 14.470 14.470 13.690 Pedestal-6:

.041 .026 .052 .053 .100 .111 .026 .038 .046 .026 Time (sec. ) :

16.580 10.750 15.820 10.750 5.090 5.090 5.100

Table 6.8.10 RESULTS OF POOL WALL DYNAMIC PRESSURES SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Holtec Racks A3, 83 and 86 Half loaded; others Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling DBE( =1.lxDBE );

Run I.D.: dwpmr405.rhr

$Revision: 1.0 s

$Logfile: C:/racks/multirac/wallpres.fov $

(1) AVERAGE DYNAMIC PRESSURES ON POOL WALLS (psi.):

Average dynamic pressure on c-x) wall: -3.803498E-03 Average dynamic pressure on (+x) wall: -3.803498E-03 Average dynamic pressure on (-y) wall: 6.343786E-03 Average dynamic pressure on (+y) wall: 6.343786E-03 (2) PEAK DYNAMIC PRESSURES ON POOL WALLS (psi.):

Positive peak pressure on (-x) wall: 5.180000 Negative peak pressure on* (-x) wall: -5.640000 Positive peak pressu*re on (+x) wall: 5.180000 Negative peak pressure on (+x) wall: -5.640000 Positive peak pressure on (-y) wall: 4.780000 Negative peak pressure on (-y) wall: -4.850000 Positive peak pressure on (+y) wall: 4.780000 Negative peak pressure on (+y) wall: -4.850000 (3) DYNAMIC PRESSURE ADDERS ON POOL WALLS (psi.) :

Positive pressure adder on (-x) wall: 1. 305517 Negative pressure adder on (-x) wall: -1.360913 Positive pressure adder on (+x) wall: 1.305517 Negative pressure adder on (+x) wall: -1.360913 Positive pressure adder on (-y) wall: 1.384353 Negative pressure adder on (-y) wall: -1.428963 Positive pressure adder on (+y) wall: 1.384353 Negative pressure adder on (+y) wall: -1.428963

( 4) NUMBER OF TIME POINTS:

• Total number of time points in file: 2001

• Number of time points where dynamic pressure on (-x) wall was positive: 1018

• Number of time points where*dynamic pressure on (-x) wall was negative: 983

• Number of time points where dynamic pressure on (+x) wall was positive: 1018

• Number of time points where dynamic pressure on (+x) wall was negative: 983

• Number of time points where dynamic pressure on (-y) wall was positive: 1021

• Number of time points where dynamic pressure on (-y) wall was negative: 980

• Number of time points where dynamic pressure on (+y) wall was positive: 1021

• Number of time points where dynamic pressure on (+y) wall was negative: 980

• Table 6.8.11 STATIC LOAD AND DYNAMIC LOAD ADDER FOR EACH PEDESTAL SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Holtec Racks AJ, BJ and B6 Half loaded; Others Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling DBE( =1.lxDBE ) ;

Run I.D.: dwpmr405.rhr RACK FOOT NCOUNT DYNAMIC ADDER STATIC LOAD No. No. (lbs.) (lbs.)

1 1 828 21068.720000 57800.000000 1 2 948 35436.600000 57800.000000 1 3 940 21032.770000 57800.000000 1 4 1032 34993.700000 57800.000090 2 1 984 24041.260000 57800.000000 2 2 1007 28394.740000 57800.000000 2 3 985 22396.550000 57800.000000 2 4 935 26405.560000 57800.000000 3 1 991 31990.010000 62600.000000 3 2 969 19674.920000 62600.000000 3 3 992 33010.080000 62600.000000 3 4 935 18656.150000 62600.000000 4 1 997 29366.30'0000 57800.000000 4 2 995 22186.330000 57800.000000 4 3 960 30684.060000 57800.000000 4 4 940 24692.980000 57800.000000 5 1 1020 18032.160000 57800.000000 5 2 943 18897.460000 57800.000000 5 3 947 17329.360000 57800.000000 5 4 1040 18525.480000 57800.000000 6 1 993 23856.090000 62600.000000 6 2 1013 27077.790000 62600.000000 6 3 933 23926.470000 62600.000000 6 4 969 28663.470000 62600.000000 7 1 341 18345.460000 55900.000000 7 2 1164 31371.050000 15100.000000 7 3 1352 31878.030000 15100.000000 7 4 431 20555.220000 55900.000000 8 1 368 18054.620000 55900.000000 8 2 1355 27532.990000 15100.000000 8 3 1335 29705.990000 15100.000000 8 4 257 8297.665000 55900.000000

• ( to be continued )

• ( Table 6.8.11, continued )

RACK No.

FOOT No.

NCOUNT DYNAMIC ADDER (lbs.)

STATIC LOAD (lbs.)

9 1 400 18373.500000 60500.000000 9 2 1326 33074.730000 16300.000000 9 3 1339 29959.670000 16300.000000 9 4 352 16189.200000 60500.000000 10 1 919 22852.340000 23300.000000 10 2 1076 18681.410000 23300.000000 10 3 1184 23952.530000 23300.000000 10 4 1020 15726.080000 23300.000000 10 5 1040 9437.692000 23300.000000 10 6 1167 2826.393000 23300.000000 11 1 974 36352.570000 23300.000000 11 2 1079 27351.710000 23300.000000 11 3 1085 37561.570000 23300.000000 11 4 924 27424.780000 23300.000000 11 5 853 9094.373000 23300.000000 11 6 817 2777.234000 23300.000000 12 1 1024 22200.880000 23300.000000

. 12 2 1103 29093.290000 23300.000000 12 3 1047 28545.650000 23300.000000 12 4 954 29133.440000 23300.000000 12 5 917 8850.709000 23300.000000 12 6 909 2805.611000 23300.000000

Table 6.8.12 TOTAL STATIC LOAD AND DYNAMIC ADDER ON THE WHOLE SLAB SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Holtec Racks A3, B3 and B6 Half loaded; Others Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling DBE( =1.lxDBE );

Run I.D.: dwpmr405.rhr (1) THE TOTAL STATIC LOAD OF RACKS AND FUEL ON SLAB IS:

2352500.00 lbs.

(2) THE TOTAL DYNAMIC LOAD ADDER ON THE SLAB IS:

231788.98 lbs.

Not~s:

(1) Total static load is modified. to include the static loads of the 7-th pedestals of the three existing racks:

2282600.0 + 3 x 23300.0 = 2352500.0 lbs.

(2) Total dynamic load adder is conservatively modified to include the dynamic adders of the 7-th pedestals of the three existing racks:

203475.9 + 3 x 9437.692 = 231788.976 lbs.

where 23300.0 is the static vertical load of each pedestal of the existing racks; and 9437.692 is the maximum vertical load adder of the 5-th and 6-th pedestals of the existing racks ( data from file: ftload.rhr ) "

Table 6.8.13 MAXIMUM DISPLACEMENTS FROM WHOLE POOL MULTI-RACK RUN SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling OBE( =1.lxOBE );

Run I.D.: dwpmrobe.rfr Rack uxt uyt . uxb uyb No. (in.) (in.) (in.) (in.)

1 .1826E+OO .2162E+OO .1520E-Ol .1830E-Ol 2 .1407E+OO .1760E+OO .1170E-Ol .1460E-Ol 3 .1251E+OO .2220E+OO .1040E-Ol .1850E-Ol 4 .1383E+OO .1132E+OO .1150E-Ol .9400E-02 5 .1435E+OO .9620E-Ol .1190E-Ol .8000E-02 6 .1839E+OO .1115E+OO .1530E-Ol .9300E-02 7 .1479E+OO .1535E+OO .1230E-Ol .1270E-Ol 8 .1444E+OO .1167E+OO .1200E-Ol .9700E-02 9 .1764E+OO .2120E+OO .1460E-Ol .1800E-Ol 10 .6740E-Ol .1368E+OO .. 5601E-02 .1140E-Ol 11 .1101E+OO .1959E+OO .9100E-02

• l660E-0*1 12 .9360E-Ol .3090E+OO .7800E-02 .2600E-Ol

$Revision: 1.8 $

$Logfile: C:/racks/multirac/maxdisp.fov $

1 Table 6.8.14 MAXIMUM IMPACT FORCE OF EACH GAP ELEMENT SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling OBE( =1.1xOBE );

Run I.D.: dwpmrobe.rfr GAP ELEMENT MAX.FORCE TIME (lb.) (sec.)

RACK-1:

1 1. 096E+05 1.771E+Ol 2 1.110E+05 1. 812E+Ol 3 1.131E+05 1.736E+Ol 4 1.117E+05 1.769E+01 5 6.589E+04 6.624E+OO 6 6.331E+04 6.232E+OO 7 4.853E+04 1.205E+Ol 8 5.299E+04 1.162E+Ol RACK-2:

9 1.133E+05 1.432E+Ol 10 1. 065E+05 9.267E+OO 11 1.114E+05 1. 477E+Ol 12 1.037E+05 9.698E+OO 13 5.981E+04 6.625E+OO 14 5.799E+04 6.206E+OO 15 6.975E+04 1.624E+Ol 16 6.983E+04 1.591E+Ol RACK-3:

17 1.267E+05 1.643E+Ol 18 1.102E+05 1.592E+Ol 19 1.264E+05 1.390E+Ol 20 1. 018E+05 1.551E+Ol 21 6.397E+04 6.645E+OO 22 5.876E+04 6.221E+OO 23 6.113E+04 1.127E+Ol 24 7.318E+04 1.598E+Ol

{ Table 6.8.14, continued } 2 RACK-4:

25 l.062E+05 l.442E+Ol 26 1. 052E+05 9.263E+OO 27 1. 080E+05 1.484E+01 28 1. 012E+05 9. 693-E+OO 29 4.134E+04 1.478E+01 30 4.433E+04 1.436E+01 31 5.505E+04 1.121E+Ol 32 5.798E+04 l.160E+01 RACK-5:

33 9.332E+04 1.697E+Ol 34 9.567E+04 1.503E+01 35 9.182E+04 1. 54.4E+01 36 l.008E+05 1.460E+01 37 5.809E+04 1.456E+01 38 5.803E+04 1.417E+Ol 39 6.726E+04 1.625E+01 40 6.649E+04 -1.591E+01 RACK-6:

41 1. 276E+05 1.433E+01 42 l.080E+05 l.332E+01 43 1. 252E+05 1.390E+01 44 1. 046E+05 1. 376E+01 45 7.757E+04 6.651E+OO 46 7.123E+04 6.279E+OO 47 5.579E+04 1. 206E+Ol 48 5.953E+04 1.164E+01 RACK-7:

49 9.796E+04 l.027E+Ol 50 1. 099E+05 1.493E+Ol 51 1.058E+05 1.082E+01 52 1.161E+05 1.452E+01 53 6.150E+04 1.453E+Ol 54 6.331E+04 1. 414E+01 55 7.425E+04 1. 622E+01 56 7.068E+04 1.589E+Ol

• ( Table 6. 8. 14, continued )

• 3 RACK-8:

57 1.088E+05 1.632E+Ol 58 l.031E+05 7.271E+OO 59 9.934E+04 1.288E+Ol 60 9.870E+04 8.700E+OO 61 6.885E+04 6.636E+OO 62 6.436E+04 6.252E+OO 63 6.204E+04 1.628E+Ol 64 6.639E+04 1.594E+Ol RACK-9:

65 1. 312E+05 1.688E+Ol 66 1.126E+05 1.536E+Ol 67 1.289E+05 1.469E+Ol 68 1.103E+05 1.493E+Ol 69 6.588E+04 *6.623E+OO 70 6.447E+04 6.208E+OO 71 5.463E+04 1.208E+Ol 72 5.640E+04 1.166E+Ol RACK-10:

73 6.748E+04 1.849E+Ol 74 6.389E+04 1.148E+Ol 75 7.297E+04 1.884E+Ol 76 5.439E+04 1.195E+Ol 77 4.066E+04 1. 625E+Ol 78 2.859E+04 1.424E+Ol 79 4.387E+04 1.194E+Ol 80 3.908E+04 4.457E+OO 81 3.987E+04 4.885E+OO 82 4.486E+04 1. 624E+Ol 83 4.266E+04 1.168E+Ol RACK-11:

84 8.617E+04 1.431E+Ol 85 6.779E+04 1.155E+Ol 86 8.804E+04 1. 470E+Ol 87 6.429E+04 1. 201E+Ol 88 4.806E+04 7.270E+OO 89 2.835E+04 7.265E+OO 90 4.496E+04 1. 285E+Ol 91 4.234E+04 1. 374E+Ol 92 4.365E+04 1.412E+Ol 93 4.426E+04 1. 437E+Ol 94 3.763E+04 7.009E+OO

( Table 6. 8 . 14, cont.inued ) 4 RACK-12:

95 6.121E+04 l.632E+Ol 96 9.538E+04 1.599E+Ol 97 7.637E+04 1.593E+Ol 98 9.433E+04 1. 641E+Ol 99 4.802E+04 7.279E+OO 100 2.904E+04 7.267E+OO 101 4.110E+04 1.474E+Ol 102 4.204E+04 1. 461E+Ol 103 4.032E+04 1.497E+Ol 104 5.289E+04 l.619E+Ol 105 5.259E+04 l.588E+Ol RACK-TO-RACK/WALL IMPACT SPRINGS AT RACK TOP:

106 O.OOOE+OO O.OOOE+OO RACK/WALL 107 O.OOOE+OO O.OOOE+OO R/W 108 O.OOOE+OO O.OOOE+OO R/W 109 O.OOOE+OO O.OOOE+OO R/W 110 O.OOOE+OO O.OOOE+OO R/W 111 O.OOOE+OO O.OOOE+OO R/W 112 O.OOOE+OO O.OOOE+OO R/W 113 O.OOOE+OO O.OOOE+OO 114 1.813E+Ol 1. 619E+Ol 115 O.OOOE+OO O.OOOE+OO R/W 116 O.OOOE+OO O.OOOE+OO R/W 117 O.OOOE+OO O.OOOE+OO 118 1. 813E+Ol 1.619E+Ol 119 O.OOOE+OO O.OOOE+OO R/W 120 3.947E+02 1. 770E+Ol 121 3.947E+02 1. 770E+Ol 122 8.302E+Ol 1. 949E+Ol 123 8.302E+Ol 1. 949E+Ol 124 1. 074E+03 1. 547E+Ol 125 1. 074E+03 l.547E+Ol 126 b.OOOE+OO O.OOOE+OO R/W 127 6.030E+02 1. 727E+Ol 128 7.949E+02 1.496E+Ol 129 O.OOOE+OO O.OOOE+OO R/W 130 O.OOOE+OO O.OOOE+OO R/W 131 6.030E+02 1. 727E+Ol 132 7.949E+02 1.496E+Ol 133 O.OOOE+OO O.OOOE+OO R/W

• ( Table 6.8.14, continued ) 5 134 1. 945E+02 1. 770E+01 135 1.945E+02 1. 770E+Ol 136 O.OOOE+OO O.OOOE+OO 137 O.OOOE+OO O.OOOE+OO 138 5.749E+02 1.544E+Ol 139 5.749E+02 1.544E+Ol 140 O.OOOE+OO O.OOOE+OO R/W 141 6.076E+02 1.492E+Ol 142 1.214E+02 1.725E+01 143 O.OOOE+OO O.OOOE+OO R/W 144 O.OOOE+OO O.OOOE+OO R/W 145 6.076E+02 1. 492E+01 146 1.214E+02 1.725E+Ol 147 O.OOOE+OO O.OOOE+OO R/W 148 O.OOOE+OO O.OOOE+OO HOLTEC/EXIST.

149 O.OOOE+OO O.OOOE+OO H/E 150 O.OOOE+OO O.OOOE+OO H/E 151 O.OOOE+OO O.OOOE+OO H/E 152 O.OOOE+OO O.OOOE+OO H/E 153 O.OOOE+OO O.OOOE+OO H/E 154 O.OOOE+OO O.OOOE+OO R/W 155 O.OOOE+OO O.OOOE+OO 156 O.OOOE+OO O.OOOE+OO 157 O.OOOE+OO O.OOOE+OO R/W 158 O.OOOE+OO O.OOOE+OO R/W 159 O.OOOE+OO O.OOOE+OO 160 O.OOOE+OO O.OOOE+OO 161 O.OOOE+OO O.OOOE+OO R/W 162 O.OOOE+OO O.OOOE+OO R/W 163 O.OOOE+OO O.OOOE+OO R/W 164 O.OOOE+OO O.OOOE+OO R/W 165 O.OOOE+OO O.OOOE+OO R/W 166 O.OOOE+OO O.OOOE+OO R/W 167 O.OOOE+OO O.OOOE+OO R/W RACK-TO-RACK/WALL IMPACT SPRINGS AT RACK BOTTOM:

168 O.OOOE+OO O.OOOE+OO R/W 169 O.OOOE+OO O.OQOE+OO R/W 170 O.OOOE+OO O.OOOE+OO R/W 171 O.OOOE+OO O.OOOE+OO R/W 172 O.OOOE+OO O.OOOE+OO R/W 173 O.OOOE+OO 0.000E+OO R/W 174 O.OOOE+OO O.OOOE+OO R/W

/*

( Table 6.8.14, continued ) 6 175 O.OOOE+OO O.OOOE+OO 176 O.OOOE+OO O.OOOE+OO 177 O.OOOE+OO O.OOOE+OO R/W 178 O.OOOE+OO O.OOOE+OO R/W 179 O.OOOE+OO O.OOOE+OO 180 O.OOOE+OO O.OOOE+OO 181 O.OOOE+OO O.OOOE+OO R/W 182 O.OOOE+OO O.OOOE+OO 183 O.OOOE+OO O.OOOE+OO 184 O.OOOE+OO O.OOOE+OO 185 O.OOOE+OO O.OOOE+OO 186 O.OOOE+OO O.OOOE+OO 187 O.OOOE+OO O.OOOE+OO 188 O.OOOE+OO O.OOOE+OO R/W 189 O.OOOE+OO O.OOOE+OO 190 O.OOOE+OO O.OOOE+OO 191 O.OOOE+OO O.OOOE+OO R/W 192 O.OOOE+OO O.OOOE+OO R/W 193 O.OOOE+OO O.OOOE+OO 194 O.OOOE+OO O.OOOE+OO 195 O.OOOE+OO O.OOOE+OO R/W 196 O.OOOE+OO O.OOOE+OO 197 O.OOOE+OO O.OOOE+OO 198 O.OOOE+OO O.OOOE+OO 199 O.OOOE+OO O.OOOE+OO 200 O.OOOE+OO O.OOOE+OO 201 O.OOOE+OO O.OOOE+OO 202 O.OOOE+OO O.OOOE+OO R/W 203 O.OOOE+OO O.OOOE+OO 204 O.OOOE+OO O.OOOE+OO 205 O.OOOE+OO O.OOOE+OO R/W 206 O.OOOE+OO O.OOOE+OO R/W 207 O.OOOE+OO O.OOOE+OO 208 O.OOOE+OO O.OOOE+OO 209 O.OOOE+OO O.OOOE+OO R/W 210 O.OOOE+OO O.OOOE+OO HOLTEC/EXIST.

211 O.OOOE+OO O.OOOE+OO H/E 212 O.OOOE+OO O.OOOE+OO H/E 213 O.OOOE+OO O.OOOE+OO H/E 214 O.OOOE+OO O.OOOE+OO H/E 215 O.OOOE+OO O.OOOE+OO H/E 216 O.OOOE+OO O.OOOE+OO R/W 217 O.OOOE+OO O.OOOE+OO 218 O.OOOE+OO O.OOOE+OO

( Table 6.8.14, continued ) 7 219 O.OOOE+OO O.OOOE+OO R/W 220 O.OOOE+OO O.OOOE+OO R/W 221 O.OOOE+OO O.OOOE+OO 222 O.OOOE+OO O.OOOE+OO 223 o.oooE+oo O.OOOE+OO R/W 224 O.OOOE+OO O.OOOE+OO R/W 225 O.OOOE+OO O.OOOE+OO R/W 226 O.OOOE+OO O.OOOE+OO R/W 227 O.OOOE+OO O.OOOE+OO R/W 228 O.OOOE+OO O.OOOE+OO R/W 229 O.OOOE+OO O.OOOE+OO R/W FILE INFORMATION FOR THIS RUN Input File dwpmrobe.rfr Plot File fwpmr.rfr x-seismic a-t-obe.hl Y-Seismic a-t-obe.h2 Z-Seismic a-t-obe.vt output File owpmrobe.rfr SALEM,UNIT-1,WPMR,12 Racks,df=dwpmrobe.rfr,1.lxDBE,dt=.00005 *

/

1 Table 6.8.15 MAXIMUM PEDESTAL STRESS FACTORS OF ALL RACKS IN POOL SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling OBE( =1.lxOBE );

Run I.D.: dwpmrobe.rfr SALEM,UNIT-1,WPMR,3-exist.+ 9-Holtec racks; wpmrobe.rfr.

• • • • • • • • • • • • • • • • • • INPUT DATA ******************

File name of FX Time history frictiox File name of FY Time history frictioy File name of FV Time history pltfwlO SRevision: 1.0 S SLogfile: C:/racks/multirac/sfmr2.fov $

SDate: 28 May 1992 18:08:26 $

File name of result output sfwpmrob.rfr Number of racks in the pool 12 Height of the pedestal, in. 15.25 Offset of FV from center, in. 1.30 Area of female pedestal, in**2. 63.86 Inertia of female pedestal, in**4. :1208.00 Distance of extrame fiber in X, in. 4.50 Distance of extrame fiber in Y, in. 4.50 Yield stress of female pedestal, psi. 21300.

Number of pedestals of each rack 4 4 4 4 4 4 4 4 4 6 6 6

• • • • ~*******************************************

MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR ALL RACK PEDESTALS

~aximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS ~6 R7 R1R6M R3R6M R4R6M

.161 . 073 .121 .130 .277 .300 .069 .151 .121 .028 Rack No.:

9 9 3 9 3 3 3 Pedestal No. :

1 1 1 1 1 1 1 Time (sec.):

16.870 7.260 15.440 7.260 15.440 15.440 15.440

( Table 6.8.15, continued ) 2

1) MAXIMUM VALUES OF STRESS FACTORS, R1 -- R7, FOR EACH PEDESTAL OF RACK- 1 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.13S .044 .110 .072 .223 .239 .062 .132 .107 .000 Time (sec.):

17.710 17.080 17.670 17.090 17.690 17.690 17.670 Pedestal-2:

.136 .OS4 .101 .092 .219 .234 .OS9 .134 .101 .000 Time (sec. ) :

18.120 17.040 18.090 17.010 18.090 18.090 18.090 Pedestal-3:

.138 .040 .111 .063 .21S .232 .061 .121 .111 .000 Time (sec.):

17.3SO 17.420 18.090 19.630 18.090 18.090 18.090 Pedestal-4:

.137 .038 .092 .060 .216 .230 .OS6 .136 .092 .002 Time (sec.):

17.690 7.760 17.690 17.400 17.680 17.680 17.690

2) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 2 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.138 .043 .072 .082 .211 .22s .044 .130 .OSl .044 Time (sec.):

14.320 17.100 19.S30 17.100 14.360 14.360 19.SlO Pedestal-2:

.130 .027 .044 .042 .1S6 .162 .032 .123 .039 .000 Time (sec.):

9.260 13.400 19.910 13.410 19.070 19.060 19.000 Pedestal-3:

.136 .030 .080 .040 .193 .203 .047 .136 .04S .022 Time (sec. ) :

13.890 13.840 19.060 16.780 14.780 14.780 19.040 Pedestal-4:

.127 .027 .064 .039 .179 .188 .042 .12s .063 .000 Time (sec. ) :

9.700 7.690 19.490 13.840 19.SlO 19.SlO 19.490

( Table 6.8.15, continued ) 3

3) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 3 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 R1R6M R3R6M R4R6M Pedestal-1:

.156 .046 .121 .066 .277 .300 .069. .151 .121 .028 Time (sec. ) :

16.430 13.380 15.440 13.380 15.440 15.440 15.440 Pedestal-2:

.135 .030 .083 .047 .202 .214 .051 .134 .080 .000 Time (sec.):

15.910 16.580 15.860 16.580 15.900 15.900 15.860 Pedestal-3:

.154 .040 .120 .060 .245 .262 .067 .153 .098 .010 Time (sec. ) :

13.900 12.820 15.860 13.730 14.960 14.960 1s;060 Pedestal-4:

.125 .025 .102 .039 .205 .219 .056 .125 .094 .000 Time (sec.) :

15.500 12.820 15.430 11.530 15.500 15.500 15.430

4) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 4 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.130 .033 .036 .052 .153 .160 .026 .116 .008 .036 Time (sec.):

14.410 17.210 15.430 16.130 15.300 15.300 15.420 Pedestal-2:

.129 .029 .036 .043

• 135 .138 .022 .117 .000 .021 .

Time (sec.):

9.250 10.320 14.470 10.320 7.260 7.260 14.470 Pedestal-3:

.132 .032 .052 .047 .182 .192 .034 .129 .038 .025 Time (sec.):

14.840 15.700 16. 210 15.660 14.870 14.870 14.910 Pedestal-4:

.124 .024 .043 .044 .133 .136 .028 .11.6 .000 .019 Time (sec. ) :

9.690 7.760 14.470 7.230 9.710 9.730 14.470

( Table 6.8.15, continued ) 4

5) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 5 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-!:

.114 .027 .021 .035 .140 .145 .015 .114 .coo .031 Time (sec.):

16.970 16.970 18.010 6.800 16.970 16.970 14.360 Pedestal-2:

.117 .024 .022 .039 .136 .140 .016 .116 .coo .023 Time (sec.):

15.030 15~040 14.360 6.790 15.040 15.040 18.490 Pedestal-3:

.112 .026 .021 .042 .130 .134 .016 .112 .coo .021

'Time (sec.):

15.440 17.330 6.510 7.260 15.430 15.430 18.010 Pedestal-4:

.124 .024 .018 .043 .134 .136 .015 .122 .000 .014 Time (sec.):

14.600 17.330 7.050 7.230. 14.590 14.590 7.050

6) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 6 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.157 .068 .047 .121 .251 .267 .039 .157 .047 .064 Time (sec.):

• 14.330 13.370 14.330 13.380 14.330 14.330 14.330 Pedestal-2:

.132 .051 .035 .oas .204 .217 .023 .132 .004 .080 Time (sec.):

13.310 13.380 13.930 13.380 13.330 13.330 13.930 Pedestal-3:

.153 .053 .031 .oas .214 .226 .027 .143 .000 .083 Time (sec.):

13.890 13.800 18.090 13.730 13.800 13.800 14.750 Pedestal-4:

.129 .051 .051 .ass .196 .209 .029 .125 .000 .084 Time (sec. ) :

13.760 13.800 14.330 13.800 13.790 13.790 14.330

( Table 6~8.15, continued ) 5

7) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 7 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.120 .02S .OS7 .042 .1S4 .162 .036 .110 .OS2 .ooo Time (sec.):

10.270 14.0SO 14.460 14.0SO 11.230 11.230 11.170 Pedestal-2:

.13S .038 .068 .OS8 .221 .237 .041 .132 .OS4 .OSl Time (sec.):

14.920 14.900* lS.740 17.120 14.900 14.900 lS.740 Pedestal-3:

.130 .032 .048 .ass .160 .16S .031 .130 .03S .ooo Time (sec.):

10.810 14.520 5.140 14.520 10.810 10.810 10.770 Pedestal-4:

.142 .03S .OS6 .041 .212 .226 .039 .134 .OS2 .040 Time (sec. ) :

14.S20 14.520 lS.420 14.S40 14.S20 14.470 14.470

8) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK- 8 Maximum Values Of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 R1R6M R3R6M R4R6M Pedestal-1:

.134 .042 .033 .06S .174 .184 .022 .118 .000 .06S Time (sec.) :

16.320 13.380 8.320 13.380 13.380 13.380 18.400 Pedestal-2:

.126 .028 .030 .035 .150 .1S4 .022 .125 .000 .029 Time (sec.) :

7.270 7.260 19.790 12.920 7.260 7.260 8.320 Pedestal-3:

.122 .030 .043 .OSl .163 .171 .032 .116 .041 .OlS Time (sec. ) :

12.880 12.920 18.020 7.260 .15.870 lS.870 *.15.870 Pedestal-4:

.121 .034 .038 .oso .161 .168 .028 .121 .029 .018 Time (sec.):

8.700 13.800 10.160 13.840 8.700 8.700 8.700 L

( Table 6.8.15, continued ) 6

9) MAXIMUM VALUES OF STRESS FACTORS, Rl" -- R7, FOR EACH PEDESTAL OF RACK- 9 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.161 .073 .08S .130 .276 .298 .ass .147 .038 .114 Time (sec.):

16.870 7.260 16.920 7.260 14.270 14.270 16.910 Pedestal-2:

.138 .052 .073 .092 .195 .208 .046 .122 .coo .087 Time (sec.):

15.360 7.310 15.460 7.310 7.260 7.260 15.450 Pedestal-3:

.158 .061 .082

• 101 .253 .271 . .050 .156 .ass .057 Time (sec.):

8.590 7.760 18.110 7.760 8.SBO 8.580 8.480 Pedestal-4:

.13S .047 .091 .079 .181 .195 .OSl .100 .091 .004 Time (sec.):

14.920 7.760 16.910 7.760 7.760 16.910 16.910

• (10) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK-10 Rl Pedestal-1:

.083 Time (sec.):

R2

.017 Maximum Values of Rl -- R7 R3

.031 R4

.031 RS

.108 R6

.113 R7

.023 Rl,R3,R4 for Max.R6 RlR6M

.083 R3R6M

.030 R4R6M

.000 18.480 6.250 14.270 6.250 18.490 18.490 18.490 Pedestal-2:

.078 .016 .037 .026 .103 .107 .023 .076 .029 .003 Time (sec.):

11.480 16.270 18.150 16.2SO 11.480 11.530 18.190 Pedestal-3:

.089 .018 .031 .032 .115 .120 .024 .089 .030 .000 Time (sec. ) :

18.840 11.950 18.150 11.960 18.840 18.840 18.150 Pedestal-4:

.067 .020 .038 .037 .097 .102 .023 .067 .014 .021 Time (sec.):

11.950 12.910 19.180 12.910 11.950 11.950 19.180 Pedestal-S:

.oso .016 .037 .025 .092 .100 .020 .047 .033 .021 Time (sec. ) :

16.240 17.200 19.180 17.210 18.490 18.490 19.180 Pedestal-6:

.035 .012 .023 .021 .060 .065 .013 .031 .021 .014 Time (sec.):

14.240 12.910 D.780 12.910 18.490 18.490 13.780 e

• ( Table 6.8.15, continued )

(11) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK-11 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 R1R6M R3R6M R4R6M Pedestal-1:

.106 .023 .ass .033 .164 .17S .036 .106 .054 .015 Time (sec.):

14.310 7.290 14.270 7.290 14.310 14.310 14.270 Pedestal-2:

.083 .022 .064 .043 .13S .144 .036 .081 .063 .000 Time (sec.):

11.SSO 7.280 11.600 7.280 11. S70 ll.S70 ll.S70 Pedestal-3:

.108 .020 .OS2 .033 .147 .lSS .033 .102 .036 .017 Time (sec .* ) :

14.700 13.880 11. 600. 10.970 13.880 13.880 11.600 Pedestal-4:

.079 .022 .048 .043 .112 .119 .028 .077 .041 .000 Time (sec. ) :

12.010 14.260 11.130 14.270 11. 9SO 11.9SO 11.130 Pedestal-5:

.OS9 .02S .OS6 .043 .141 .156 .030 .058 .055 .043 Time (sec.):

7.270 14.240 14.2SO 14.250 14.240 14.240 14.240 Pedestal-6:

.03S .018 .036 .034 .07S .083 .018 .031 .028 .02S Time (sec.):

11.130 14.560 11.940 14.560 14.180 14.180 17 .110 (12) MAXIMUM VALUES OF STRESS FACTORS, Rl -- R7, FOR EACH PEDESTAL OF RACK-12 Maximum Values of Rl -- R7 Rl,R3,R4 for Max.R6 Rl R2 R3 R4 RS R6 R7 RlR6M R3R6M R4R6M Pedestal-1:

.075 .030 .064 .053 .139 .152

• 035 .067 .064 .021 Time (sec.):

16.320 7.330 16.410 7.330 16.410 16.410 16.340 Pedestal-2:

.117 .028 .* 117 .052 .216 .233 .063 .116 .117 .000 Time (sec. ) :

15.990 7.290 16.050 7.310 16.020 16.020 16.020 Pedestal-3:

.093 .020 .103 .031 .180 .196 .054 .093 .103 .000 Time (sec. ) :

15.930 7.650 15.940 7.670 '15.940 15.940 15.940 Pedestal-4:

.116 .022 .108 .039 .220 .239 .059 .115 .108 .016 Time (sec.):

16.410 16.420 16.420 7.650 16.420 16.420 16.410 Pedestal-5:

.059 .024 .052 .039 .104 .115

• 026 .042 .050 .023 Time (sec. ) :

7.280 7.310 9.110 7.310 9.180 9.180 9.110 Pedestal-6:

.036 .024 .047 .047 .076 .084 .023 .033 .006 .045 Time (sec.):

7.270 7.310 16.240 7.310 7.360 7.360 16.240

Table 6.8.16 RESULTS OF POOL WALL DYNAMIC PRESSURES SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling OBE( =l.lxOBE );

Run I.D.: dwpmrobe.rfr

$Revision: 1.0 $

$Logfile: C:/racks/multirac/wallpres.fov $

(1) AVERAGE DYNAMIC PRESSURES ON POOL WALLS (psi.):

Average dynamic pressure on (-x) wall: 1. 031610E-02 Average dynamic pressure on (+x) wall: l.031610E-02 Average dynamic pressure on (-y) wall: 2.639219E-04 Average dynamic pressure on (+y) wall: 2.639219E-04 (2) PEAK DYNAMIC PRESSURES ON POOL WALLS (psi.):

Positive peak pressure on (-x) wall: 3.380000 Negative peak pressure* on (-x) wall: -3.660000 Positive peak pressure on (+x) wall: 3.380000 Negative peak pressure on (+x) wall: -3.660000 Positive peak pressure on (-y) wall: 2.550000 Neg~tive peak pressure on (-y) wall: -3.150000 Positive peak pressure on (+y) wall: 2.550000 Negative peak pressure on (+y) wall: -3.150000 (3) DYNAMIC PRESSURE ADDERS ON POOL WALLS (psi. ) :

Positive pressure adder on (-x) wall: 1.009628 Negative pressure adder on (-x) wall: -1.037137 Positive pressure adder on (+x) wall: 1.009628 Negative pressure adder on (+x) wall: -1.037137 Positive pressure adder on (-y) wall: 8.091071E-Ol Negative pressure adder on (-y) wall: -8.196703E-01 Positive pressure adder on (+y) wall: 8.091071E-Ol Negative pressure adder on (+y) wall: -8.196703E-Ol (4) NUMBER OF TIME POINTS:

• Total number of time points in file: 2001

• Number of time points where dynamic pressure on (-x) wall was positive: 1024

• Number of time points where dynamic pressure on (-x) wall was negative: 977

• Number of time points where dynamic pressure on (+x) wall was positive: 1024

• Number of time points where dynamic pressure on (+x) wall was negative: 977 Number of time points where dynamic

• pressure on (-y) wall was positive: 1008 Number of time points where dynamic

• pressure on (-y) wall was negative: 993

• Number of time points where dynamic pressure on (+y) wall was positive: 1008

• Number of time points where dynamic pressure on (+y) wall was negative: 993

• Table 6.8.17 STATIC LOAD AND DYNAMIC LOAD ADDER FOR EACH PEDESTAL SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling OBE( =1.lxOBE ) ;

Run I.D.: dwpmrobe.rfr RACK FOOT NCOUNT DYNAMIC ADDER STATIC LOAD No. No. (lbs.) (lbs.)

1 1 965 12341.350000 57800.000000 1 2 965 19642.070000 57800.000000 1 3 936 12594.760000 57800.000000 1 4 946 20841.230000 57800.000000 2 1 919 13999.020000 57800.000000 2 2 958 15242.380000 57800.000000 2 3 984 13584.350000 57800.000000 2 4 940 15601.060000 57800.000000 3 1 935 20221.070000 62600.000000 3 2 992 12132.060000 62600.000000 3 3 982 19360.690000 62600.000000 3 4 928 13196.980000 62600.000000 4 1 941 15361.640000 57800.000000 4 2 973 13979.140000 57800.000000 4 3 948 15846.310000 57800.000000 4 4 943 14784.520000 57800.000000 5 1 907 11048.950000 57800.000000 5 2 988 10678.640000 57800.000000 5 3 961 10716.540000 57800.000000 5 4 959 11029.300000 57800.000000 6 1 921 16136.260000 62600.000000 6 2 948 13399.050000 62600.000000 6 3 948 15986.390000 62600.000000 6 4 988 13066.500000 62600. 000000 (

7 1 928 13730.930000 57800.000000 7 2 963 18461.580000 57800.000000 7 3 973 13519.840000 57800.000000 7 4 969 17654.390000 57800.000000

( to be continued )

( Table 6.8.17, continued )

RACK FOOT NCOUNT DYNAMIC ADDER STATIC LOAD No. No~ (lbs.) (lbs.)

8 1 924 13765.150000 57800.000000 8 2 969 11402.170000 57800.000000 8 3 974 12772.380000 57800.000000 8 4 950 12340.000000 57800.000000 9 1 930 21179.250000 62600.000000 9 2 926 14368.360000 62600.000000 9 3 994 20463.780000 62600.000000 9 4 993 13755.990000 62600.000000 10 1 985 10766.600000 23300.000000 10 2 1242 11396.300000 23300.000000 10 3 1253 13931.050000 23300.000000 10 4 1050 9141.333000 23300.000000 10 5 1235 4972.712000 23300.000000 10 6 1776 2090.372000 23300.000000 11 1 981 16215.700000 23300.000000 11 2 1196 13659.110000 23300.000000 11 3 1176 19850.850000 23300.000000 11 4 1011 11245.200000 23300.000000 11 5 1162 6506.196000 23300.000000 11 6 1564 2092.775000 23300.000000 12 1 965 12005.490000 23300.000000 12 2 1214 15876.610000 23300.000000 12 3 1234 15093.840000 23300.000000 12 4 1058 14249.050000 23300.000000 12 5 1158 .5273.662000 23300.000000 12 6 1640 2011.098000 23300.000000

• Table 6.8.18 TOTAL STATIC LOAD AND DYNAMIC ADDER ON THE WHOLE SLAB SALEM NUCLEAR GENERATING STATION, UNIT-1 PUBLIC SERVICE ELECTRIC & GAS COMPANY WPMR Analysis, 3 Existing + 9 Holtec Racks in Pool, Fully Loaded with 1700# Reg.Fuel; Random Friction; Seismic: Controlling OBE( =1.lxOBE ) ;

Run I.D.: dwpmrobe.rfr (1) THE TOTAL STATIC LOAD OF RACKS AND FUEL ON SLAB IS:

2627700.00 lbs.

(2) THE TOTAL DYNAMIC LOAD ADDER ON THE SLAB IS:

121538.79 lbs .

• Notes:

(1) Total static load is modified to include the static loads of the 7-th pedestals of the three existing racks:

2557800.0 + 3 x 23300.0 = 2627700.0 lbs.

(2) Total dynamic load adder is conservatively modified to include the dynamic adders of the 7-th pedestals of the three existing racks:

102020.2 + 3 x 6506.196 = 121538.79 lbs.

where 23300.0 is the static vertical load of each pedestal of the existing racks; and 6506.196 is the maximum vertical load adder of the 5-th and 6-th pedestals of the existing racks (data from file: ftload.rob )

• Table 6.8.19 COMPARISON OF LIMITING RESULTS FROM SINGLE RACK ANALYSES AND WPMR ANALYSES Maximum Pedestal Maximum Maximum Vertical Pedestal Displacement Load Stress (inches) (lbf) Factor Single Rack 1.2233 313465 0.789 Analysis (Rack 9 (Rack 9) (Rack 9) in-phase motion assumed)

WPMR 0.8691 220500 0.730 Analysis (Rack 9, (Rack 9, (Rack 9, Run No. 2) Run No. 1) Run No

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<( o-ldba.h11, o-ldba.h21, o-ldbe.hJ1, o-ldba.h41; Domplng: ]%.

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<( o-ldbe.h12, o-ldbe.h22, o-ldbe.h32, o-ldbe.h42; Dornplng: 3%.

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v - 0. 00 --l-+-+-4-lf-~i-.---1++1 a-.__...,._.,,,,_.. -l-+-tll~-*-ll*-++-+-+-i.-f--+--~-11!--l--jl.-4-jlMJ*-..__,.,__._,._._

(_

Q)

Q) 0 0

<{

-0.20

-0.40-+-r-~.-.-.~--.-r--.----r-r-T-r~-.--.--.----r-r-T-r~-.--.--.----r-r-T-r~-.--.--,-,--,

0.00 5.00 *10 00 a 20.00 Tl me sec p

• FIGURE 6.3.13

0.40 SALEM GENERATING STATION UNITS 1 AND 2, PSEG. POOL SLAB, EL.89.5' Sy n l he l t c o I DBE Ac c e I e r o l lo n T l me - H ls lo r y f or Po o I S I o b , ( FINAL )

o - l - db e

• v t. ; Ou r o t. lo n : 20 s e c .

. 0. 20 c

0

._.J

..__)

'° - 0. 00 -

L Q)

Q) 0 0

<{

-0.20

-0.40~-r--.--.----.----.----r--r-,--r--.--.-.----.----.----r--r-,--r--.--.-~--.--,.----.---.--.----.--.----.----.--.-r-,--,-~

0.00 5.00 10.00 15.00 20.00 Tl me , sec

• FIGURE 6.3.14

(2) ( 3)

../

c 0

-+--'

0 L

Q)

Cl) ..-

u CS) SALEM GENERATING STATION UNITS 1 AND 2, PSEG. POOL SLAB, EL.89.5' u0- ( 1) THE BROADENED FLOOR RESPONSE SPECTRUM ( OBE, Hor lzon le I, Dcmp tng 1sti )

(2) THE AVERAGE RESPONSE SPECTRUM OF 4 SYNTHETICAL OBE ACCEL.TIME-HISTORIES:

<C c-lobe.h11, c-lobe.h21, c-lobe.h31, c-lobe.h41 (J) THE RE-GENERATED RESPONSE SPECTRUM from ltme-htsLory: A-T-OBE.H1.

0. 1 1 10 Frequency, Hz.

FIGURE 6.3.15

(3)

(J) c 0

.SALEM GENERATING STATION UNITS 1 AND 2, PSEG. POOL SLAB, EL.89.5' C1) THE BROADENED FLOOR RESPONSE SPECTRUM C OBE, Hor tzonLol, Oomplng 1%)

C2) T~E AVERAGE RESPONSE SPECTRUM OF 4 SYNTHETICAL OBE ACCEL.TIME-HISTORIES:

o-Lobe.h12, o-Lobe.h22, o-lobe.h32, o-Lobe.h42 CJ) THE RE-GENERATED RESPONSE SPECTRUM from ltme-htslory: A-T-OBE.H2.

0. 1 1 10 Frequency, Hz.

FIGURE 6.3.16

(/)

O\_

... (S) c 0

-+--'

0 L

Q)

Q) 0 u

<( SALEM GENERATING STATION UNITS 1 AND 2, PSEG. POOL SLAB, EL.89.5'

( 1) THE BROADENED FLOOR RESPONSE SPECTRUM ( OBE, Var l t c o I , Do mp t ng 1~ )

(2) THE AVERAGE RESPONSE SPECTRUM OF 4 SYNTHETICAL OBE ACCEL.TIME-HISTORIES:

o-lobo.vL1, o-Lobo.vl2, o-Lobo.vt.3, o-Lobo.vL4 (3) THE RE-GENERATEd RESPONSE SPECTRUM from Ltrne-htsLory: A-T-OBE.VT.

1 10 Frequency, Hz.

FIGURE 6.3.17

• 0.20 -

- SALEM GENERATING STATION UNITS 1 AND 2p PSEG. POOL SLAB, EL.89.5'

- Syn t. he t. l co I OBE Ac c e I er o t. lo n T l me - Hls lo r y for Po o I SI ob , ( FINAL )

- o - t. - o b e

• h 1 ; Du r o l lo n : 20 s e c .

. 0. 10 -

(./)

O>

c -

0

~

~ .J

~

([) - 0. 00 =

....._)

. AL ~.1A 1

(_

-"y 11' r r r* \.J Q)

Q) -

0 0

<( -

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- 0

• 20 I I I I I I I I I I I I I I I I I I I I I I I I I I I. I I I I I I I I I I I I I 0.00 5.00 10.00 15.00 20.00 Tl me p sec

• FIGURE 6.3.18 l

0.20 SALEM GENERATING STATION UNITS 1 AND 2, PSEG. POOL SLAB, EL.89.5' Syn t. he t. t co I OBE Ac c e I er o t. ton T t me - Hts t. or y for Po o I SI ob , C FINAL )

o-t.-obe.h2; Durol ton: 20 sec.

. 0. 10 c

0

• .J co - 0 . 00 -t;--t-t+-\-rt-t--ittN--t~-t-U-<rt-ti tt-P'l-tt-+--l L

Q)

Q) 0 0

<(

-0. 10

-0.20-+-.--.-r-~c-T-T-~~~-.-.-~~~~~~~-.-.-~~

0.00 5.00 10.00 15.00 20.00 Tl me p sec

• FIGURE 6.3.19

• 0.20 SALEM GENERATING STATION UNITS 1 AND 2, PSEG. POOL SLAB, EL.89.5' Syn l he l t co I OBE Ac c e I er o l ton T t me - Hts lo r y for Po o I SI ob , C FINAL )

o- l - obe *v l ; Du r o l lo n :

• 20 s e c *

. 0. 10 c

0

._J

_,_)

V - 0

• 00 ~f-Hllll\+lllj-~_...ll'H-IH~ .n&-Ull-4-IUl.111111-1i-3-+-+-11.

(_

G)

Q) 0 0

<(

-0. 10

~0.20-t--r-~..-.-.~-.--.--~..-.-.~-.--.--~..-.-.~~~r-T-T~~~

0.00 5.00 10.00 15.00 20.00 Tl me p sec

• FIGURE 6.3.20

FIGURE 6.4.1 PICTORIAL VIEW OF RACK STRUCTURE

z

~ q22 PI9 2I PlB q21 o/-- -p1P8 Pl7 2* ,'

H/4 3*~

pyi I

w.o:

rmE!Rl[ H/4 CBflB!.IIE H

4*w:g Pll/i

' H/4

~ !1S*k Q

IHRT Dlli!ITllJI ~

P3 7t31 H/4 P2 I S4 q5 LI

. I

';I

• - /p15 (

J [?

,foe j, I b X LINi mmrrn SimRT S2 TYPIDL FRICTfill 8.E!ENT FIG. 6.4.2 SCHEMATIC MODEL FDR DYNARACK

IYPff AL Ill' JIU 1ACT ELEHENI RACK SJRl[JIRE FIGURE 6.4,3 RACK-TO-RACK IMPACT SPRINGS

CELL WALL

~ASS I FUEL ASSCMBLY/CELL IMPACT SPRING L--x0*_ _ J x FIGURE 8.4.4 FUEL-TO-RACK IMPACT SPRINGS

• ~--------- L __________,

z FIGURE 6.4.5 DEGREES-OF-FR~EDOM MODELING RACK MOTION

L/2 FIGURE 6.4.6 RACK DEGREE-OF-FREEDOM FOR Y-Z PLANE BENDING

• • ~

L/2 L/2 FIGURE 6.4. 7 RACK DEGREE-OF-FREEDOM FOR X-Z PLANE BENDING

FlEL ASSY..1ELL ltrA[J SPRING, KI 0,l-jf 1112 0.2'."ll 0.2~1 IV2 FRICTHJN 0.2~1 INIERrA[E SPRING, Kf

-SIPP~f l.EG Sl~ING, Ks

._ ~- FllIDATllW ROT ATIMAL

((fflJMl:E SPRING, KR FIGURE 6.4.8 2-D VIEW OF RACK MODULE

3.00 GAP TIME HISTORY, SALEM GENERATING STATION, PSE&G Go p bet hie en Rock - 2 on d Nor th Wo I I , Eo s t corn er , Top ,

From WPMR o n o I y s ls : d w p mr 405. r fr ; 1 . 1 x DBE ; 1 700*r e g f u e I ; f u I I ;

Fr l c t lo n c o e ff l c le n t = r o n d o m ( me o n = 0

• 5 ) . F l I e : g hi - 2r

• d o t .

2.00 1 . 00 0.00--+-r----r-r--~~-.--,-~~~~~-.--,-~~~~~~~-

0.00 5.00 10.00 15.00 20.00 Tl me 9 sec

• FIGURE 6.8.l

1 00 D

GAP TIME HISTORY, SALEM GENERATING STATION , PSE&G Gop between Rock-2 ond Rock-5, Eosl Corner, Top, From WPMR onolysls: dwpmr405.rfr; 1.1xDBE; 1700#reg fuel ;ful I; Fr l c l lo n co e ff l c le n l = r on do m ( me on = f2J. 5 )

• F l I e : g 2- 5r l

• do l .

0.80

• 0.60 c

,._J 0.

ID (9

0.40 0.20 0.00~-.-r---.-.--.-.-r-.-.--l,J-,-~-.-.----.-.-~~--.-U.-~--.-L-.-J,-+-L,-l.-,l..Ll-,-----.-L.-L.--,

0.00 5.00 10.00 15.00 20.00 Tl me p sec

• FIGURE 6.8.2

1 00 D

GAP TIME HISTORY, SALEM GENERATING STATION , PSE&G Go p be l ween Ro ck - 5 on d Ro ck - 6, Sou l h Corn ~ r , Top ,

From WPMR on o I ys ls: dwpmr405. rfr; 1 .1 xDBE; 1700#reg fue I ; fu I I ;

Fr l c t lo n co e ff l c le n t = r on do m ( me on = 0. 5 )

• F l I a : g 5- 61 t . do t .

0.80

• 0.60 c

~

Q_

ID

~

0.40 0.20 0.00-i-r--.---r--.-r-'~,L--,--Jr~-'-r-+-~-.-1,-L..--.-,r--r-r~~--,-L-.J--r-1--.J--y-L-,-L,~-+-L,--1,--L, 0.00 5.00 10.00 15.00 20.00 Tl me p sec

• FIGURE 6.8.3

• 1

• 00 GAP TIME HISTORY, SALEM GENERATING. STATION , PSE&G I

Go p b s t ws s n Ro c k - 5 on d Ro c k - 8 , Eo s t Corn s r , Top ,

From WPMR onolysls: dwpmr405.rf'r; 1.1xDBE; 17012l#rsg fuel ;ful I; Fr l c t lo n c o s f f l c ls n t = r o n d o m ( ms o n = 0

• 5 )

• F l I s : g 5- Br t . d o t .

0.80

• 0.60 c

o-J 0..

v l'.)

0.40 0.20 0.00---t-r--,.-,-~-.-.-.-.----.---,-.---.-.-----.--.--,-,-J~~Y-T--,.-,-~-,-,--,---,--,~--.--.---.-.

0.00 5.00 10.00 15.00 20.00 Tl me p sec

• FIGURE 6.8.4 J

2.00 GAP TIME HISTORY, SALEM GENERATING STATION , PSE&G

-Go p be t.1-1 e en Rod - 8 on d Rod - 1 1 , Eo s t. Co r n e r

• Top ,

From WPMR onolysls: dwpmr405.rfr; 1.1xDBE; 1700#reg fuel ;ful I; Fr l c t. lo n c o e ff l c la n t. = r o n d o m ( me o n = 0

• 5 )

• F l I e : g 8- 11 r

• d o t.

• 1 . 50

.. 1 . 00 0.50 0.00--t--r--,--,--.---.---r-r-.----.----r-r-T-r-r----..-.-.----.-----.--.-r-r-r-Y--.---,--,--.---.---r-r-,---,-,---,---,-----.---r--r-.

0.00 5.00 10.00 15.00 20.00 T lme p sec .

FIGURE 6.8.5

• For Gap Time-Histories, Plea ee:

Figure 6.8.1, Figure 6.8.2, Figure 6.8.3,r Figure 6.8.4, Figure 6.8.5

-NORTH (UNIT 1) >-

FIGURE 6.8.6 GAP LOCATIONS

i I

I I

I

.FIGURE 6.10.l FINITE ELEMENT MODEL 16 LOCATIONS

7.0 ACCIDENT ANALYSIS AND MECHANICAL INTEGRITY CONSIDERATIONS 7.1 Introduction This section provides results of fuel drop accident analyses performed to demonstrate that the maximum density racks comply with regulatory requirements. The refueling accidents considered include the drop of one fuel assembly onto the top of a rack or through a cell to the baseplate.

7.2 Refueling Accidents Although this licensing application is not for consolidated fuel assemblies, all mechanical drop accident evaluations were performed with an assumed consolidated fuel plus handling tool mass of 3810 lbs. The mechanical accident analysis results bound those for a standard PWR assembly. The assumed high mass fuel assembly including the handling tool are referred to as a "heavy fuel assembly"_. It is assumed that the bottom of the fuel assembly as it is being moved over stored fuel is 36" above the top of the new spent fuel racks [7.2.1).

7.2.1 Dropped Fuel Assembly Accident (Deep Drop Scenario)

A 3810 lb. heavy fuel assembly is dropped from 36 11 above the top of a storage location and impacts the base of the module. While local failure of the baseplate is acceptable, a gross structural failure is not permissible. Further, the stored fuel array must remain subcritical. If a fuel assembly drops through a cell not located over a pedestal, calculated results show that the spacing between cells is unaffected and the stored fuel array remains subcritical.

Local baseplate deformation in the vicinity of the impact occurs, but the dropped assembly is contained within the rack and does not impact the liner*. The impact results in a maximum baseplate movement toward the liner which is less than 3.40". Since the 7-1 J

• • baseplate height above the liner is over four times greater than this calculated movement (16.75 inches}, a wide margin of safety exists against any potential impact with the pool liner.

If a fuel assembly drops through a cell located over a pedestal, the impact load transmitted through the support to the liner is well below the loads caused by seismic events (presented in Section 6). Therefore, the concrete bearing pressures calculated for the seismic events and reported in Section 6 bound those due to the drop accident.

7.2.2 Dropped Fuel Assembly Accident (Shallow Drop Scenario)

One heavy fuel assembly 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 must be limited to the top region. The rack cross-sectional geometry at the elevation of the top of the active fuel (and below} must not be altered. Postulating a physically plausible region of impact, permanent deformation is restricted to a depth less than or equal to 6.52 11 from the top of the rack. This is considerably less than the available cell length above the active fuel region (approximately 20 11 ) .

7.3 Conclusion The Salem UFSAR postulated refueling accidents have been considered to determine whether the proposed racks meet the essential criteria of subcriticality and structural ruggedness. The subcriticality criterion requires that the center-to-center spacing and other design basis parameters in the active fuel region of the racks are not altered due to a postulated fuel assembly drop accident.

Analyses conclude that, under both "shallowH and "deep drop" scenarios, the stored spent fuel array remains subcritical. These

• 7-2

conclusions were obtained considering the "heavy fuel" in the drop simulation which is considerably heavier than a typical PWR fuel assembly.

The structured ruggedness criterion seeks to ensure that the postulated drop

• accident does not result in secondary damage.

Examples of secondary damage are baseplate piercing leading to an impact between the fuel assembly and the pool liner during a "deep drop" scenario, or extensive plastic deformation of the rack top, cushioning the active fuel region but leading to Baral neutron absorber damage. Analyses conclude that large margins of safety exist against all such gross structural damage scenarios.

In conclusion, the new maximum density racks meet all required mechanical and functional integrity criteria under postulated fuel handling accidents *

7.4 REFERENCES

[7.2.1) Technical Specification, Maximum Density Spent Fuel Storage Racks for Salem Generating Station, Units 1 and 2, Spec. No. s-c-FHB-SGS-0154.

7-3

• 8.0 8.1 FUEL POOL STRUCTURAL INTEGRITY CONSIDERATIONS Introduction The Salem Units 1 and 2 spent fuel pools are safety related, seismic category I, reinforced concrete structures. This section describes the analysis used to demonstrate the structural adequacy of the pool structures, as required by Section IV of the USNRC OT Position Paper [8.1.1]. Since the Salem Units 1 and 2 pools are geometric mirror images of each other, only one need be analyzed.

The remainder of this section describes analysis of the Unit 1 spent fuel pool; all results apply to the Unit 2 pool as well.

(Rack dynamic loads, which differ slightly between Units 1 and 2, were both considered in analysis of the Unit 1 geometry). Figure 8.1.1 shows a schematic view of the Unit 1 spent fuel pool.

The Salem spent fuel pool region was analyzed using the finite element method. Individual load component results are combined using the factored load combinations described in SRP 3.8.4 [8.1.2]

based on the "ultimate strength" design method. These load combinations are generally more conservative than those in the American Concrete Institute (ACI) Code [ 8. 1. 3] , which is the current design basis. Structural integrity is maintained with the fuel pool assumed to be fully loaded with new, free-standing, maximum density fuel racks and all storage locations occupied by standard fuel assemblies.

The fuel pool regions examined are the pool slab and the highly loaded wall sections surrounding the pool. Both moment and shear capacities were checked for concrete structural integrity. Slab local punching and bearing integrity, in the vicinity of a rack module support pedestal pad, was evaluated. All structural capacity calculations were made using design formulas meeting the requirements of ACI 318-71 [8.1.3] and ACI 349-85 [8.1.4] *

• 8-1.

The structural analysis incorporates several improvements. In particular, the conservatism and accuracy of the solution methodology have been refined as follows:

(i) The pool structure is modelled in three dimensions via a 3-D finite element model, rather than in the customary two dimensions in which the solution relies on approximations to account for the interaction between adjacent walls.

(ii) The plant design spectra for Salem UFSAR DBE and OBE events have been broadened per the provisions of Reg. Guide 1.122 [8.1.5].

(iii) The response spectrum method is used to determine the self-excitation loadings on the pool structure.

This eliminates the use of simplified static methods that have uncertain accuracies.

(iv) The thermal gradient across the pool slab and the pool

• walls is also computed using the finite element method. The effect of interaction between the ambient, pool water, and grade temperatures is fully incorporated in the analysis.

(v) The pressure on the lower portion of the wall during a seismic event undergoes a cyclic pulsation due to the hydrodynamic coupling between the racks and the pool walls. This loading, often neglected in pool structural evaluation, is fully quantified using the Whole Pool Multi-Rack analysis described in Section 6 of this report. This loading, ref erred to as hydrodynamic pressure, is fully accounted for in this analysis.

(vi) The "ultimate strength" design method permitted by SRP 3.8.4 [8.1.2] is applied.

While not required by NUREG-0612, analyses are also performed to demonstrate that the design strength of the corbel (supporting the semi-gantry fuel handling crane) is not exceeded if the factored dead and live load combinations of ACI 318-71 are applied to the crane.

8-2

• In summary, the pool structure analysis carried out in support of the rerack application utilizes a rigorous mathematical model and a comprehensive simulation of all applicable loads.

8.2 Description of Spent Fuel Pool Structure The Salem spent fuel pool is a rectilinear reinforced concrete (RC) structure with a stainless steel liner. The pool structure is supported by a fill concrete substructure. The spent fuel pool (SFP) slab is 11 ft. thick. The top of the SFP floor slab is at EL. 89 1 -0 11

• The fill concrete under the Containment, Fuel Handling, and Auxiliary Buildings is continuous. The RC walls are 41'~0" high above the top of the pool slab. The south wall of the pool separates the SFP from the fuel transfer pool. This wall is 4'-0" thick at the upper 22'-6" of its height and 6'-0" thick at the lower 16'-0" of its height. The upper and lower sections are separated by a 2' -O" high transition section. The south wall is notched near the southeast corner of the pool to allow for the transfer of fuel into and out of the pool.

The fuel transfer pool is located adjacent to the south end of the SFP. This pool is assumed to be dry for this evaluation. This assumption will provide the most limiting load condition for the south wall of the SFP. Figures 8.2.1, 8.2.2, and 8.2.3 show general views of the finite element mesh used to model the SFP structure.

8.3 Definition of Loads Pool structural loading involves the fallowing discrete components:

8.3.1 Static Loading

1) Dead weight of pool structure plus 41' -0 11 (high water limit) of water (including hydrostatic pressure on the pool walls). Combining the hydrostatic pressure and structure dead weight is in conformance with ACI 349-85 [8.1.4).

8-3

2) Dead weight of maximum density rack modules and standard fuel assemblies stored in the modules.

8.3.2 Dynamic Loading

1) Vertical loads transmitted by the rack support pedestals to the slab during a DBE 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 motion in the pool during a seismic event.

8.3.3 Thermal Loading Mean temperature rise and temperature gradient across the pool slab and the pool walls due to temperature differential between the pool water and the atmosphere external to the slab and walls.

8.4 Analysis Procedures 8.4.1 Finite Element Analysis Model The finite element model of the SFP structure is illustrated in Figures 8.2.1 and 8.2.2. The wall and roof structure external to the SFP, illustrated in Figure 8. 2. 3, is modelled only to the extent that the appropriate interactions are accounted for.

The finite element code ANSYS 4.4A [8.4.1) is used for all structural analyses. AN SYS 4. 3 [ 8. 4. 2] is used for all heat convection/conduction analyses. The three-dimensional finite element model is constructed using STIF45 solid elements for the main pool wall and slab, and STIF63 shell elements to model the external structure. The element thickness in the various regions of the structure is the actual thickness of the structure at that location. The finite element model analyzes the mechanical and thermal loads. The effect of reinforcement and concrete cracking is reflected in element properties assigned to various locations during the simulations.

8-4

---

~* -- --

• The intent of the model is to provide a conservative evaluation of the moment distribution in the slab and in the surrounding pool walls. The rotational resistance from any lateral external supporting member not explicitly modelled by solid elements is neglected.

8.4.2 Analysis Methodology In Section 6 of this report, the results of Whole Pool Multi-Rack analyses are documented.

• The results (for DBE and OBE seismic events) establish pedestal load time histories on the pool slab and hydrodynamic pressure-tim.e histories for the wall structure adjacent to the racks.

For spent fuel pool re-qualification, the pool is assumed to contain 12 free-standing, fully loaded, spent fuel racks. A total of 1776 fuel assemblies having a dry weight of 1700 lbs per assembly are conservatively assumed to be stored in the pool. This loading is greater than the actual storage capacity (1632 cells) and the actual dry weight of the PWR fuel (1500 lbs per assembly).

8.4.2.1 Application of Loads The major structural loadings discussed in Section 8.3 are imposed on the finite element model in the following manner:

a. Dead weight of the structure is simulated by a 1-g vertical gravitational load imposed on the model.

b. Loads (static and dynamic) from the racks and fuel are imposed on the slab as effective uniform pressure loads.

Effective dynamic adders are obtained from the results of the Whole Pool analyses. The dynamic adders are defined as equivalent static loads giving the same total impulse as the actual time varying dynamic loads during the time period when the total load is greater than the static load.

c. Water loads are imposed on the walls and slab as a hydrostatic pressure and as effective uniform horizontal pressures to simulate the action of various hydrodynamic effects.

8-5

d. The pool water temperatures (section 8.3.3) are assumed to be uniform in space. Heat conduction/convection analyses on representative wall and slab sections are performed to obtain actual concrete surface temperatures which are then imposed on the elements to simulate a mean temperature plus a temperature gradient. The maximum pool water temperature is 180°F. Ambient air temperature is assumed to be 60°F and the steady state grade temperature is assumed to be 50°F at an elevation 28 feet below the bottom of the slab, for the purpose of establishing concrete temperature distribution through the concrete.

The temperatures present a limiting thermal load under both normal and abnormal operating conditions.

e. Self-excitation loads are imposed on the structure by adding the results of a response spectrum analysis and a zero-period acceleration (ZPA) analysis. The response spectrum analysis includes all modes with frequencies below 40 Hz.

8.4.2.2 Concrete Cracking

.The effect of concrete cracking (permitted in analyses per the ACI

. Code [ 8. 1. 4] ) is simulated by reduction of the element Young's Moduli in appropriate regions. Load re-distribution that occurs in the walls and slab is accounted for in the model. Concrete cracking is only considered under the thermal load condition.

8.4.3 Load Combinations The factored load combinations applicable to this analysis are in accordance with [8.1.2). The following factored loading combinations are potentially limiting (those loads which are not applicable to the Salem spent fuel pool have been deleted):

1) 1.40 +/- 1.9E

2) .75 (1.40 + 1.7T0 +/- 1.9E)

3) D + Ta +/- E'

4) D + Ta +/- 1.25E In the above defined combinations, the notation of [8.1.2) is used:

D = dead loads and hydrostatic loads E = OBE resultant loads E' = DBE resultant loads To = thermal loads due to postulated normal thermal condition; to be conservative, we assume T0 = Ta*

Ta = thermal loads due to postulated abnormal thermal condition 8-6

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 arbitrary directional nature of the seismic loading. Both directions are evaluated to establish the limits at various structural locations.

The dead load and the thermal load tend to reduce each other in some of the areas. Per SRP 3.8.4, adjustments to load coefficients were made and additional combinations were computed when a load acted in a direction that reduced the effect of other loads. For considerations involving dead and thermal loadings in conjunction with the seismic load, additional combinations are computed with reduced coefficients on the dead load and/or thermal load.

8.5 Results of Analyses The AN SYS postprocessing capability is used to form the appropriate load combinations and to establish the limiting bending moments and shear forces in various sections of the pool structure. Section limit strength formulas for bending and shear stresses are computed using appropriate concrete and reinforcement strengths.

To assess the potential importance of certain load components on the Salem pool structure, calculation of an equivalent static uniform pressure on the pool slab is made by combining each of the major distributed mechanical loads. Table 8.5.1 shows that the major contributions are from.the water and the fully-loaded racks, and that the dead load contribution is significantly higher than the seismic contribution.

Tables 8.5.2 and 8.5.3 indicate results from potentially limiting load combinations for the bending and shear strengths of the slab and walls. For each section, we define the limiting safety margins as the limit strength bending moment or shear force defined by ACI for that structural section divided by the calculated bending 8-7

moment or shear force {from the finite element analyses) . The major regions of the pool structure consist of eight different regions corresponding to different reinforcements and section thicknesses.

Each region is searched independently for the maximum bending moments and shear forces in the various coordinate directions.

Safety margins were determined from the calculated maximum bending moments and shear forces based on the respective local strengths.

The procedures were repeated for all the potential limiting load combinations.

Table 8.5.2 demonstrates that the limiting safety margins in bending for all sections are above 1.0. Table 8.5.3 demonstrates that the limiting safety margins in shear for all sections are above 1.0.

Subsequent to completion of the pool structure analysis with intact fuel, the consequences of loading every storage cell with 2: 1 consolidation canisters on the integrity of the pool structure were also analyzed. Since the pool slab is premised on base mat, the effect of the increased fuel mass in the racks on the structural safety margin occurs only through a slight increase in the hydrodynamic pressure loads on the walls during the postulated seismic events. The incremental loading, however, is quite small and the margins of safety are found to undergo negligible change.

8.6 Pool Liner The integrity of the Salem pool liner under the seismic event is analyzed and confirmed. The pool liner is subject to in-plane strains due to movement of the rack support feet during the seismic event. Analyses were performed to establish that the liner will not tear or rupture under limiting fuel pool loading conditions *

• 8-8

l

• The liner was subjected to limiting vertical and horizontal pedestal loads and shown not to have a cumulative damage factor less than 1.0 under 1 DBE and 20 OBE events [8.6.1, Section 7.3].

The cumulative damage factor was calculated in accordance with ASME Code,Section III, Fatigue Analysis Method. The number of strain cycles is based on results from Whole Pool Multi-Rack analysis pedestal load time-histories. Finite element analysis was used to model the liner plate in the vicinity of a pedestal and determine the calculated stress under the input pedestal load (horizontal and vertical components)

• These analyses are based on loadings imparted from the most highly loaded pedestal in the pool assumed to be positioned in the most unfavorable position. The analyses demonstrate that the pool liner will not rupture due to cyclic loadings induced by the postulated ground motions for the Salem fuel pools.

8.7 Bearing Pads Each support pedestal in the pool is emplaced on a bearing pad that is sized to ensure that the average bearing pressure under steady state and seismic events complies with the provisions of [8.1.4].

No rack pedestals or bearing pads encroach on any leak chase lanes in the pool, eliminating the need to evaluate surface discontinuity stresses. The bearing pads are constructed from ASTM 240-304 austenitic stainless plate stock.

8.8 Conclusions The Salem spent fuel pool structure was analyzed using a 3-D finite element model. All applicable steady state and dynamic loadings were considered. The time variant loads due to sloshing of pool water, and the hydrodynamic interaction between the assemblage of racks and the pool walls were incorporated into the factored load

• 8-9

combinations. The effect of the thermal gradient across the pool walls and the pool slab was also evaluated using the finite element method.

The factored load combinations from SRP 3.8.4 [8.1.2] were utilized to compare the margins against the section design strengths throughout the pool structure. The minimum values of available margins in bending and shear {i.e., the margin of safety at the most stressed locations in the structure) indicate large structural margins.

8.9 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] NUREG-0800, SRP-3.8.4, Rev .. 1, July 1981.

• [8.1.3]

[8.1.4]

ACI 318-71, Building Reinforced Concrete Code Requirements ACI 349-85, Code Requirements for Concrete Nuclear Safety Related Concrete Structures, Concrete Institute, Detroit, Michigan.

for American

[8.1.5] USNRC Regulatory Guide 1.122, Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components.

[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 Systems, Rev.

4.4A.

[8.4.2] ANSYS User's Manual, Swanson Analysis Systems, Rev.

4o3

[8.6.1] PSE&G Specification S-C-FHB-SGS-0154.

8-10

Table 8 *. s .1 CALCULATION OF EQUIVALENT SLAB VERTICAL PRESSURES Slab Weight (average thickness 11 1 ) 11.5 psi Hydrostatic Load (41' Water) 17.77 psi Rack plus Fuel Dead Weight 19.84 psi Dynamic Adder (Water, Racks and Fuels) (DBE) 2.00 psi Dynamic Adder (Water, Racks and Fuels) (OBE) 0.81 psi

Table 8.5.2 BENDING STRENGTH EVALUATION Critical Limiting Load Combination Location Safety Margin (see Section 8.4.3)

East wall 1.22 2 West wall 1.25 2 North wall 1.44 2 South wall, 1.60 1 upper section, South wall, 1.89 1 middle section South wall, 1.37 2 lower section Slab 3.11 1

_j . l I

Table 8.5.3 SHEAR STRENGTH EVALUATION Critical Limiting Load Combination Location Safety Margin Csee Section 8.4.3)

East wall 2.22 2 West wall 2.91 2 North wall 2.33 2 South wall, 3.81 2 upper section South wall, 2.11 2 middle section South wall, 2.42 2 lower section Slab 3.02 2 I

I I_ J

EL 130' -

EL 89' -

EL 78' -

• FIGURE 8. 1. 1 SALEM UNIT 1 SFP

J ANSYS-PC 4.4A1 MAR 2 1993 9:52:2a

• PLOT NO. 2 PHEP7 ELEMENTS MAT NUM XU =-1 YU =1 zu =1 DIST=778.162 XF =264.5 YF =406 .5 .

ZF =522 UUP =Z

.PRECISE HIDDEN SGS SFP FE MODEL - UIEU FROM SOUTHEAST FIGURE 8

• 2

• 1 SGS SFP FINITE E.LEMENT MODEL, VIEW FROM SOUTHEAST

ANSYS-PC -4.-4A1 MAR 2 1993 9:5e:2s PLOT HO. 1 PREP7 ELEMENTS MAT HUM XU =1 YU =-1 zu =1 DIST=770.162 XF =264.5 VF =406 .5 ZF =522 UUP =2 PRECISE HIDDEN SGS SFP FE MODEL - VIEW FROM NORTHWEST FIGURE 8.2.2 SGS SFP FINITE ELEMENT MODEL, VIEW FROM NORTHWEST

J*

RNSYS-PC -4.-4R1 MAR 2 1993 9:52:04 PLOT NO. 1 PREP7 ELEMENTS MAT HUM XU =-1 YU =1 zu =1 DIST=770.162 XF =264.5 VF =406.5 ZF =522 UUP =Z PRECISE HIDDEN SGS SFP FE MODEL - UIEW FROM SOUTHEAST FIGURE 8.2.3 SGS SFP FINITE ELEMENT MODEL, VIEW FROM SOUTHEAST, SHOWING EXTERNAL ROOF AND WALL STRUCTURE

9.0 RADIOLOGICAL EVALUATION 9.1 Fuel Handling Accident The potential radiological consequences at the Salem exclusion area boundary (EAB) resulting from a fuel handling accident have been determined.

9.1.1 Assumptions and Source Term Calculations Evaluations of the accident. were* -based* on fuel of 4. 5 wt% initial enrichment* burned to 65,000 Mwd/mtu. The reactor was assumed to have been operating at 3600 Mw thermal power (105.5% of rated power) prior to shutdown, with an average specific power of 40.45 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 all the fuel .rods (264) in a peak-power fuel assembly. Gap inventories of fission products available for release were estimated using the release fractions identified in Regulatory Guide 1.25 [9.1.1) except for Iodine-131. The release fraction for I-131 increased 20% in accordance with NUREG/CR-5009

[9.1.2). The failed fuel cooling time prior to the accident was 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 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 reach saturation inventories during in-core operation. These inventories depend primarily on the fuel specific power over the few months immediately preceding reactor* shutdown. The specific power and iodine and xenon inventory will be proportional to the radial power peaking factor (1.70).

• An enrichment of 5.0 wt% was also investigated. The 4.5 wt% value shown here was determined to be more conservative .

• 9-1

After 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> of cooling time, most of the thyroid dose comes from Iodine-131, while most of the whole-body dose comes from Xenon-133. 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 atmospheric diffusion factor Cx/Q) and filter efficiency values previously specified by PSE&G. Core specific fission product inventories (Curies per metric ton of uranium) were estimated using the ORIGEN-2 code [9.1.3].

Assumptions were as follows: specific power of 40.45 kw/kgU, initial enrichment of 4.5 wt% U, burnup of 65,000 Mwd/mtu, and a cooling time of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />. The results of the ORIGEN calculations for isotopes considered in the determination of the thyroid and whole-body doses are provided in Table 9. 1. Table 9. 2 lists pertinent data for the isotopes of interest. Data and assumptions used in the dose calculations are given in Table 9.3.

The following equation1 from Reg Guide 1. 25, was used to calculate the thyroid dose (D, in rads) from the inhalation of radioiodine. Values for many of the terms in the equation are given in Table 9.2 and Table 9.3.

Dose =L Fq Ii F p B ~ x/Q ,where i DFP DFf Fq = fraction of fuel rod iodine Ri= dose conversion inventory in gap space factor (rads per curie)

Ii = core iodine radionuclide x/Q = atmospheric inventory at time of the diffusion factor accident (curies) (sec per cubic meter) 9-2

F = fraction of core damaged so DF P = effective iodine as to release iodine in the decon. factor for rod gap pool water p = core peaking factor B = breathing rate (cubic meters DFf = effective iodine per second) decon. factor for 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.

Doser = L o. 25 (X/Q) F P Gi Ey 1 i

G1 is the gap inventory of the gaseous radionuclides of xenon and krypton, while the E11 term is the average energy (Mev) per disintegration of each radionuclide as stated 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.

9.1.2 Results The doses at the Salem EAB from the specified fuel handling accident are listed below. The doses are based on the release of all gaseous fission product activity in the gaps of all 264 fuel rods in a highest-power assembly.

Thyroid dose, rad = 0.307 Whole-body dose, rem = 0.483

• 9-3

These potential doses are "well within" the exposure guideline values of 10 CFR Part 100, paragraph 11. As defined in Standard Review Plan 15.7.4, Radiological Consequences of Fuel Handling 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 whole-body doses.

9.2 Solid Radwaste Resin replacement is primari:l.y determined*bythe requirement for water clarity. Normally the resin is changed about.once a year.

A significant increase in the volume of solid radioactive wastes is not expected as a result of expanded storage capacity. Fuel pool storage expansion activities may result in the generation of a small amount of additional resins due to pool cleanup requirements.

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 releases; therefore, significant increases are not expected as a result of the expanded storage capacity.

9.4 Personnel Exposures 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 at Salem, which originate primarily from radionuclides in the pool water, are generally 1. o mrem/hr or less around the pool, and 2 . o mrem/hr around the Salem pool bridge. The maximum dose rate at the Salem pool surface from a fuel assembly in transit was calculated to be 1.2 mrem/hr

• 9 - 4

Radiation levels in zones surrounding the pool are not expected to be affected significantly except, possibly, on the outer surface of the pool east wall.

• Existing shielding around the north and west sides of the pool -- water and concrete walls --

provides more than adequate protection {dose rates less than 0.002 mrem/hr). The east wall is less thick than the other walls.

It is calculated that dose rates up to about 9 mrem/hr could be experienced at its surface. The very conservative calculations yielding this result were based on a set of conditions that are highly improbable. The dose point was* directly opposite the mid-height of fuel assemblies from a radial "hot" zone in the core.

All fuel assemblies were c.ooled for the minimum cooling time {168 hrs). If the same fuel were qooled an average time of one year, the dose rate on the surf ace of the east wall would be less than 1.0 mrem/hr.

Typical pool water radionuclide concentrations are

• listed in Table 9. 4, which shows activities for outage and non-outage periods. During fuel reload operations, the concentrations are expected to increase due to crud deposits spalling from spent fuel assemblies, and radionuclides carried into the pool from the primary system. The data in Table 9. 4 reflects historical values.

Calculations based on average concentrations during an outage period result in a pool surface dose rate of less than 3. o mrem/hr.

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. Airborne activity monitors 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 radiation protection program nor the area monitoring system requires modification.

9 - 5

I 9.5 Anticipated Exposure During Reracking All of the operations involved in reracking will utilize detailed procedures prepared with full consideration of AI.ARA principles.

Similar reracking operations have been performed in a number of facilities with minimal radiation exposure to personnel.

Total occupational exposure for the reracking operation is estimated to be between 6 and 12 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 possible exposure.,

It is PSE&G's *intention to perform all field operat~ons remotely insofar as it is consistent with the AI.ARA objectives. In other words, divers will be utilized to carry out an in-pool operation only if such an approach demonstrably minimizes the total radiation exposure to the personnel. All diving operations will comply with Draft Regulatory Guide DG_-8006, "Control of Access to High and Very High Radiation Areas in Nuclear Power Plants".

The existing radiation protection program at Salem is adequate for the reracking operations. Where there is a potential for significant airborne activity, continuous air samplers will be in operation. Personnel will wear protective clothing, personnel monitoring equipment (TLD and pocket dosimeters) and, if necessary, respiratory protective equipment. Activities will be governed by a Radiation Work Permit. Additional personnel monitoring equipment (ioe., extremity badges) may be utilized.

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

• The existing storage racks will be removed, then washed down in preparation for packaging and shipment. Estimates of the person-rem exposures associated with washdown and readying the old racks for shipment are indicated in Table 9. 5. Shipping containers and procedures will conform to Federal DOT regulations and to the requirements of any state through which the shipment may pass, as set forth by the State DOT office.

9.6 References

[9.1.1] Regulatory Guide 1. 2 5 ( AEC Safety Guide 2 5) ,

"Assumptions tJsed for Evaluating The Potential Radiological Consequences Of A Fuel Handling Accident in The Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors", March 23, 1972.

[9.1.2] C. E. Beyer, et al. , 11 Assessment of the Use of Extended Burnup Fuel in Light Water Reactors",

NUREG/CR-5009, Pacific Northwest Laboratory, February, .1988.

[9.1.3] "ORNL Isotope GENeration and Depletion", ORNL/TM-7175, Oak Ridge National Laboratory, July, 1980 *

• 9 - 7

• Table 9.1 RESULTS OF ORIGEN-2 CALCULATIONS FOR RADIONUCLIDES OF IODINE, KRYPTON, AND XENON AT 168-HOURS COOLING TIME; ACTIVITY RELEASED IN THE FUEL-HANDLING ACCIDENT Radionuclide curies per mtu Activity Released. Ci I-131 6.231 x 10 5 5.860 x 10 4 I-132 3.586 x 10 5 2.810 x 10 4 I-133 8.077 x 10 3 6.330 x 10 2 I-134 ----- -----

I-135 4.515 x 10- 2 3.538 x 10- 3 Kr-85 1.646 x 10 4 2.276 x 10 3 Kr-85m 1. 083 x 10- 6 8.487 x 10- 8 Kr-87 ----- -----

Kr-88 8.124 x 10-13 6.367 x 10-14 Xe-131m 1.135 x 10 4 8.895 x 10 2 Xe-133 1.033 x 10 6 8.096 x 10 4 Xe-133m 1.180 x 10 4 9.248 x 10 2 Xe-135 1.575 x 101 1.234 x 10° Xe-135m 7.232 x 10- 3 5.668 x 10- 4

• Table 9.2 PROPERTIES OF RADIONUCLIDES INCLUDED IN FUEL HANDLING ACCIDENT ANALYSES PERFORMED BY HOLTEC

[Rev. 0, 1 -93]

Thyroid Dose Conversion(

Radionuclide Rads/CUrie 1 > EP (Mev/dis) E, (Mev/dis)

Iodine-131 1.48 x 1011 0.191< 2 > 0.381< 2 >

Iodine-132 5.35 x 10" 0. 392 <3 > 2.295< 3 >

Iodine-133 4.0 x 105 0. 409< 2 > 0. 612< 2 >

Iodine-134 2.5 x 10 4 0.455< 3 > 1.840<3 >

Iodine-135 1.24 x 10 5 0.309< 3 > 1. 775<3 >

Krypton-S5 ----- 0.251< 2 > 0. 002< 2 >

Krypton-S5m ----- 0.213< 3 > 0.151< 3 >

Krypton-S7 ----- 1. 050< 3 > 1.374<3 >

Krypton-SS ----- 0.341< 3 > 1.745<3 >

Xenon-13lm ----- 0.142< 2 > 0.020< 2 >

Xenon-133 ----- 0 .136< 2 > o. 047< 2 >

Xenon-133m ----- 0.195< 2 > 0.040< 2 >

Xenon-135 ----- 0.303< 3 > 0. 246< 3 >

Xenon-135m ----- 0.095< 2 > 0.42s< 2 >

<1 > Regulatory Guide 1.25 .(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 Pressurized Water Reactors, USNRC, 3/23/72.

<2 > Personal communication, c. w. Reich, INEL, to s. E. Turner, Holtec International, August 16, 1991.

<3 > c. M. Lederer, et al., Table of Isotopes, sixth Edition, John Wiley & Sons, New York, 1967.

Table 9.3 DATA AND ASSUMPTIONS FOR*THE EVALUATION OF THE FUEL HANDLING ACCIDENT Core power level, Mw(t) 3600 Fuel enrichment, wt% u 4 .s*

Fuel burnup, Mwd/mtu 65,000 Specific power, kw/kgU 40.45 Fuel cooling time, hrs 168 Power peaking factor 1.70 No. of failed fuel rods 264 Core inventory released to gap, %

Iodine-131 12 Other iodines 10 Krypton-BS 30 Xenons 10 10 Pool decontamination factors Iodine Noble gases 500 1

Filter decontamination factors Iodine Noble gases 100 1

Atmospheric diffusion factor Cx/Q), sec/~

5.00 x 10" 4 Breathing rate, m /sec 3

3.47 x 10" 4

• An enrichment of s.o wt% was also investigated. The 4.5 wt% value shown here was determined to be more conservative.

Table 9.4 CONCENTRATIONS OF RADIONUCLIDES IN SALEM SPENT FUEL POOL WATER*

Concentration During Concentration During Nuclide Outage Period 1 ~Cilml Non-outage Period 1 MCilml Co-58 2.0 x 10" 3 8.9 x 10* 5 Co-60 3.6 x 10* 4 2.0 x 10" 4 Nb-95 1.2 x 10" 4 5.9 x 10" 6 Sb-125 3.5 x 10* 5 3.3 x 10* 5 Cs-137 1.2 x 10" 5 4.9 x 10" 6

• Averages.of values for samples ta$en at inlets to both pools.

• Table 9.5 PRELIMINARY ESTIMATE OF PERSON-REM EXPOSURES DURING RERACKING Estimated Number of Person-Rem Step Personnel Hours Exposure

• Remove empty racks 5 40 0.5 to 1.0 Wash racks 3 10 0.08 to 0.2 Clean and Vacuum Pool 3 25 0.3 to 0.6 Remove underwater 4 5 0.4 to 0.8 appurtenances Partial installation 5 20 0.25 to 0.5 of new rack modules Move fuel to new racks 2 150 0.8 to 1.5 Remove remaining racks 5 120 1.5 to 3.0 wash racks 3 30 0.2 to 0.4 Install remaining new 5 35 0.4 to 0.8 rack modules Decon and prepare old racks for shipment 4 80 1.0 to 2. o**

Total Exposure, person-rem 6 to 12

• 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 operations, which assume 20 to 40 mrem/hr.

• • Maximum expected exposure, al though details of preparation and packaging of old racks for shipment have not yet been determined.

BORAL SURVEILLANCE PROGRAM 10.1 Purpose tm Boral , the neutron absorbing material incorporated in the spent fuel storage rack design to assist in controlling system reactivity, consists of finely divided particles of boron carbide (B4C) 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 environments of the spent fuel pool have demonstrated the stability and chemical inertness of Baral (References [10.1.1]-[10.1.3]). The accumulated dose to the Baral over the expected rack lifetime is

. 10 ,,

• estimated 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 large amounts of in-pool data, Boral is considered a satisfactory material for reac-tivity control in spent fuel storage racks and is expected to fulfill its design function over the lifetime of the racks.

Nevertheless, the USNRC requires the licensee to establish a surveillance program to continually monitor the Boral's integrity and performance* of Boral 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."

10-1

The purpose of the surveillance program presented herein is to characterize certain properties of the Baral to assess the capability of the Baral panels in the racks to continually perform their intended function. This program is in consonance with the ongoing Baral surveillance in the Salem pools with respect to the existing modules.

The proposed surveillance program is capable of detecting the onset of any significant degradation with ample time to take any necessary corrective action.

In response to the need for a comprehensive Baral 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 primarily depends on monitoring representative coupon samples to evaluate the perfor-mance of the absorber material without disrupting the integrity of the storage system. The principal parameter to be measured is the thickness (to monitor for swelling) followed by boron content if thickness measurement indicates any anamoly.

10.2 COUPON SURVEILLANCE PROGRAM 10.2.1 coupon Description The coupon measurement program consists of jacketed coupons suspended on a mounting (called a "tree"), placed in a designated cell, and surrounded by spent fuel. Coupons will be removed from the array on a prescribed schedule to measure certain physical and chemical properties from which the stability and integrity of the Baral in the storage cells may be inferred. The existing coupon holders in the pool may be used for this purpose or new coupon holders may be utilized.

10-2

• Each new surveillance coupon will mimic the dimensions of the existing coupons. The coupon surveillance program will use a total of 10 test coupons for each of the two pools.

Each coupon will be carefully pre-characterized prior to pool insertion to provide reference, initial values for comparative measurements after irradiation. The surveillance coupons will be pre-characterized for weight and physical dimensions. In addition, two coupons will be preserved as archive samples for subsequent comparative test coupon measurements. Wet chemical analyses of samples from the same Boral lot will be provided for B10 content reference data.

10.2.2 Surveillance Coupon Testing Schedule The coupon tree will be surrounded by freshly discharged fuel assemblies 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 may remain in place for the remaining lifetime of the racks.

At the time of the first fuel off-load following installation of the coupon tree, the (8) storage cells surrounding the tree shall be loaded with freshly-discharged fuel assemblies from among those which are not scheduled to be returned to the core. Effort will be made to surround the coupon assemblage with freshly discharged fuel each time a new batch of fuel is unloaded from the reactor.

However, from the fifth cycle on, the fuel assemblies in the (8) surrounding cells may remain in place. Coupons will then be pulled on a set, incremental schedule such that the last coupon will be pulled during the last several refuelings *

• 10-3

• Evaluation of the removed coupons will provide information on the effects of the radiation, thermal and chemical environments of the pool and by inference, comparable information on the Boral panels in the racks. Over the duration of the coupon testing program, the coupons will have accumulated more radiation dose than the expected lifetime dose for normal storage cells.

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, should any defects be found.

10.2.3 Measurement Program 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

. * (optional)

Neutron Attenuation Dimensional Measurements (particularly thickness of the coupon; other dimensional measurements may be made at the discretion of.

Plant Management)

Weight and

• Wet-chemical analysis* (optional)

Photography (optional)

• Neutron attenuation and wet chemistry test will be performed only if the coupon shows visible signs of degradation.

10-4

• The most significant measurement is thickness (to monitor for swelling). The neutron attenuation test is a precise standby tool to measure the neutron absorption effectiveness of the coupon.**

If loss of boron is observed or suspected, the data may be further augmented by wet-chemical analysis (a destructive gravimetric technique for total boron only).

10.2.4 surveillance Coupon Acceptance Criteria Of the measurements to be performed on the Boral surveillance coupons, the most important are (1) the thickness measurement (to monitor potential swelling) and (2) the neutron attenuation measurements (to verify the continued presence of the boron).

Acceptance criteria for these measurements are as follows:

An increase in thickness at any point should not exceed 10% of the initial thickness at that point *

• A decrease of no more than 5% in Boron-10 content, as determined by neutron attenuation, is acceptable. (This is tantamount to a requirement for no loss in boron within the accuracy of the measurement.)

The thickness measurement is the principal vehicle for monitoring.

Our past experience with Baral indicates that the thickness criteria, in most probability, will not be violated. However, in the remote event that the thickness criterion is exceeded, then the confirmatory evaluation of neutron attenuation will be utilized.

• • Neutron attenuation measurement is a precise instrumental method of chemical analysis 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 Baral panel.

10-5

• Additional coupons may be retrieved and tested if deemed necessary.

This will provide corroborative evidence whether or not the indicated change(s) is real. If the deviation is determined to be real, an engineering evaluation shall be performed to identify necessary further testing or any corrective action.

The remaining measurement parameters serve a supporting role and should be examined for early indications of potential Boral degradation that would suggest a need for further attention 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 measurement accuracy. The ongoing coupon surveillance program for the existing Exxon racks will also be updated to accord with the acceptance criteria presented herein.

Procedures for coupon surveillance measurement have been developed

• by the rack supplier and provided to the Licensee for future use in in-service surveillance.

10.3 References

[10.1.1] "Spent Fuel Storage Module Corrosion Report",

Brooks & Perkins Report 554, June 1, 1977.

[10.1.2] "Suitability of Brooks & Perkins Spent Fuel Storage Module for Use in PWR Storage Pools",

Brooks & Perkins Report 578, July 7, 1978.

[10.1.3] "Boral Neutron Absorbing/Shielding Material -

Product Performance Report", Brooks & Perkins Report 624, July 20, 1982.

[10 .1. 4] USNRC Letter to All Power Reactor Licensees, transmitting the "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applicat.ionsn, April 14, 1978

• 10-6

• 11.0 11.1 ENVIRONMENTAL COST-BENEFIT ASSESSMENT Introduction Article V of the USNRC OT Position Paper [11.1] requires the submittal of a cost-benefit analysis for the selected option of expanding on-site spent nuclear fuel storage capacity. This section describes the analyses and evaluations made by PSE&G which justified reracking Salem 1 and Salem 2 spent fuel pools as the most viable alternative to meet the near-term spent fuel storage requirements for Salem 1 and 2.

11.2 Need for Increased On-Site Storage Capacity The immediate need to increase the limited capacity of the Salem 1 and 2 spent fuel pools is based on the continuously increasing spent nuclear fuel supply and high level waste inventories, the prudent requirement to maintain Operational Full Core Reserve

{OFCR)*, delays announced by the U.S. Department of Energy {DOE) in their Monitored Retrievable Storage {MRS) and geological repository program for permanent disposal of spent nuclear fuel and the lack of viable alternatives.

Salem 1 and 2 projected spent fuel discharges are tabulated in Section 1, Tables 1. 2 and 1. 3-. As indicated, without on-site expansion, the Salem 1 and 2 pools will lose OFCR in March, 1998 and September, 2002, respectively. This projected loss of storage capacity would affect PSE&G's ability to operate each reactor.

• Loss of Operational Full Core Reserve Without Reracking will occur in March, 1998.

Loss of Operational Full Core Reserve with Reracking will occur in September, 2008.

11-1

Neither Salem 1 nor Salem 2 has an existing or planned contractual arrangement for third party spent fuel storage or spent fuel reprocessing. Since the cost of spent fuel reprocessing is not offset by the salvage value of the residual uranium, reprocessing represents an added cost to the nuclear fuel cycle which already includes the NWPA's Nuclear Waste Fund fees. Shutting down Salem Generating Station would result in replacement power costs that far exceed the cost of increasing on-site spent fuel storage capacity.

Thus, there are no acceptable alternatives to increasin~ on-site spent fuel storage capacity at Salem. Moreover, to maintain regulatory prudence over-e)Cpansion of on-site storage capacity must be avoided.

11.3 Appraisal of Alternate Options The key guidelines that PSE&G considered which lead to the decision to rerack the Salem pools are listed below:

1. To protect the public health and safety and the quality of the environment by implementing a technically well proven and an existing NRC licensed technology.

2. To increase on-site storage capacity in a timely manner to maintain plant operability.

3. To minimize licensing risk and increase public acceptance as Salem 1 was previously reracked during 1978-1981.

4. To maintain maximum flexibility, avoid over-expansion and over-commitment of funds to increase on-site storage capacity. This protects the interests of the ratepayers by ensuring prudent expenditures.

5. To meet the near and long-term storage needs and maintain an incremental expansion capability. This would allow PSE&G to maintain sufficient pressure on the Federal Program in meeting its contractual obligation. It would also protect the interests of the ratepayers by avoiding excess expenditures for on-site spent fuel storage.

11-2

The following is an overview of alternate technologies that PSE&G considered with respect to the above criteria. The feasibility of each option was determined by evaluating their technical characteristics, licensing status, regulatory requirements, industry utility experience, applicability to PSE&G site specific situations, impact on plant operations, implementation schedule, and estimated total life cycle cost.

Intrasite Shipment of Spent Fuel Between Salem Units Intrasite shipment involves transferring spent fuel from the Salem 1 to the Salem 2 pool. All such shipments would occur within the existing secured area. Intrasite shipment only provides temporary relief from the overall spent fuel* storage problem, as it would improve the storage situation at one site at the expense of the other. At most, the estimated time extension until loss of OFCR occurs would be one cycle.

Several utilities have executed this option and shipped spent fuel between pools located on the same site and also between pools at two different sites. These shipments allowed utilities to fully utilize their existing resources, gain time to develop a spent fuel storage strategy, follow emerging storage technologies, and delay (not avoid) new on-site storage expenditures. Since the value of this option is extremely limited for the Salem spent fuel storage situation, it has been disqualified from further consideration at the present time.

In-Pool Rod Consolidation The primary purpose of in-pool rod consolidation is to reduce the volume of spent fuel assemblies thereby increasing the spent fuel pool capacity . Rod consolidation involves removing the spent fuel

• 11-3

• rods from their assemblies and loading them into metal canisters in a close-packed array. This process can be performed robotically or mechanically. The remaining non-fuel bearing components (NFBC) of the fuel assembly hardware (i.e., grid spacers, guide tubes, end fittings, etc.) are sheared, compacted and. separately stored in another container. Both the fuel rod and NFBC canisters would be suitable for storage within the existing Salem pool racks.

The technical issues associated with rod consolidation are primarily related to operational considerations which in turn affect the economics of this technology. These include extracting the rods from the assemblies, loading them into canisters in a tight array, compacting the NFBC, and prudently performing all of these operations without impacting plant operations.

Under existing design conditions, the rod consolidation and NFBC compaction ratios would increase the number of available storage spaces by 8 for every 20 consolidated PWR fuel assemblies and 7 for every 20 consolidated BWR fuel bundles (BWR fuel channels must also be compacted and stored which requires one additional storage cell). To date, rod consolidation has been demonstrated at six utility sites. However, al though the design rod consolidation ratios have been achieved, design NFBC compaction ratios have had limited success in a production environment. Moreover, there are additional operational issues, such as spent fuel pool contamination during the rod consolidation operation, and the speed of the rod consolidation operation, that have not yet been satisfactorily resolved to make this an attractive technology for current large scale utility use.

Although this option was determined to be comparable in cost to reracking, the plant operational impact was the primary disqualifying factor.

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• *conversion of Hope Creek Unit 2 Reactor Building into Spent Fuel Storage Pool A feasibility study was performed to address the conversion of the abandoned Hope Creek Unit. 2 reactor building into an Independent Spent Fuel Storage Installation (ISFSI). The wet storage concept (i.e., a spent fuel pool) was investigated in detail because such a conversion has the potential to meet life of plant storage, including life extension, if necessary, for Salem 1, Salem 2, and Hope Creek. However, utilizing dry storage technology in this ISFSI (cask storage) has not been eliminated*from*consideration.

Although the wet spent fuel storage at the ISFSI and the Salem/Hope Creek pools would be similar, the design requirements for the ISFSI are less stringent. The requirements associated with reactor spent fuel pool designs are not applied to an ISFSI because only spent fuel which has aged at least ten years would be planned for storage in the ISFSI.

The existing HC-U2 reactor building structure was determined to be adequate for the loads imposed by the new use of the structure. The ISFSI could be constructed in two phases with little demolition or rework of the existing structure. Phase 1 would provide approximately ten years worth of storage capacity, while Phase 2 would complete the facility and provide life of plant storage capacity for the Salem and Hope Creek uni ts. Support systems necessary for the safe underwater spent fuel storage were also determined to be similar to the traditional, proven systems utilized by Salem's and Hope Creek's existing facilities.

The ISFSI would be licensed in accordance with 10CFRPart 72 as a non-safety-related nuclear facility; however, the existing structure is seismically designed and constructed to an augmented

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• QA program. PSE&G would be required to submit an application for a specific license to receive, transfer, and possess spent fuel at the ISFSI.

The feasibility study results indicated that the existing Hope creek Unit 2 reactor building could be converted into an ISFSI to provide.safe, underwater storage of spent fuel from Salem 1, Salem 2, and Hope Creek. Moreover, underwater spent fuel storage technology is well understood, proven within the nuclear industry and it offers low technical risks. As such, this option remains open for consideration as a supplement to the reracking option should the future spent fuel storage situation dictate the need for additional storage capacity.

Cask Storage Spent fuel storage in metal casks is one of the most mature on-site dry storage methods available at the present time. It has been tested, demonstrated, licensed, and used in the United States since 1986 and it continues to gain industry acceptance. The dry storage technique involves loading intact or consolidated spent fuel into casks which would be stored on a concrete platf orl'il in a secured area. This installation would be classified as an ISFSI and therefore, would be required to meet licensing under 10CFRPart 72.

A dry cask ISFSI is a passive storage system requiring no auxiliary equipment such as pumps, fans, motors, etc. Aside from the casks and a cask transporter, the ISFSI would require lighting, monitored security fencing, a backup diesel generator and an alarm panel for cask monitoring, but it would not have to be staffed on a continuous basis.

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• Present generation casks have been designed for storage only. Dual purpose casks are currently being designed to serve both storage and transport functions. Metal cask designs, which have been used since 1986 can be modified to obtain approval under 10CFRPart 71 for transporting spent fuel. such a dual purpose cask would eliminate the need to prepare another shipping cask.

Spent fuel cask storage provides many benefits. The fuel loading and cask placement on an on-site storage pad is not expected to be technically complex. It would not require large amounts of station labor and the accumulated radiation dose from these activities is expected to be a very small fraction of the total station radiation dose. Cask storage has little effect on* plant operations and would require minimal plant support. This technology also allows modular expansion of the on-site storage facility which will spread out the high expenditures. This degree of freedom would also allow PSE&G to take advantage of technological progress in cask design.

Furthermore, . cask storage is* expected to reduce compatibility concerns with the eventual DOE system.

The cost uncertainties associated with metal casks are primarily due to market conditions rather than technical factors. As the concrete and dual purpose cask technologies continue to evolve, their estimated costs are subject to some uncertainty. In general, cask costs are expected to come down with competition and demand, but presently, that remains to be seen.

Horizontal Concrete Modules or Vaults In the horizontal concrete module dry storage option, the spent fuel is kept in the fuel basket of a stainless steel canister which is shielded, vacuum dried, sealed and *filled with helium or nitrogen to prevent fuel oxidation. These canisters are then

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• stored in concrete modules or vaults which provide adequate shielding during storage. The heat generated by the spent fuel is removed via radiation, conduction and natural convection through air channels in the concrete module. The NUTECH Horizontal Modular Storage (NUHOMS) system is presently the only system utilized to facilitate this type of storage option. Its major components include the concrete horizontal storage module, a spent fuel transfer cask, and a special purpose cask transfer trailer.

This storage option woulg require an ISFSI and would have to meet the licensing criteria of 10CFRPart 72. This NUHOMS system ISFSI could be located on-site and meet the integrated spent fuel storage needs of the Salem and Hope Creek units. However, the seismic qualifications required* for the concrete pad and the storage modules place this option at a disadvantage compared to other options. The PSE&G site specific geology would make construction of such a facility costly because of the need for deep geologic drilling, excavation, dewatering, pouring concrete foundation, etc.

Transferring spent fuel from the storage pool into the horizontal modules is a technically complex operation compared to cask storage. A cask would also be needed to facilitate transportation of the spent fuel canister to a DOE offsite facility and therefore, would have to closely interface with the eventual DOE system.to eliminate the need for secondary fuel handling. Additionally, even though this option allows for modular expansion, like cask storage, it is not expected to be easily expandable because of the need to reactivate on-site construction.

Since this option is expected to be as expensive as, and does not appear to have the same degree of flexibility, operational simplicity, and modular expansibility as the cask option, it does not currently appear to be desirable for the PSE&G site *

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• Vertical Concrete Modules or Vaults The vertical concrete module or vault option stores spent fuel in sealed metal tubes housed in a concrete structure. Each tube is vertically arranged and stores one fuel assembly under a cover of nitrogen. The tubes are shielded and protected on all sides by the concrete structure. A group of such tubes makes up one* module. The current design allows storage of 83 PWR fuel assemblies or 150 BWR fuel bundles. Each fuel tube penetrates the upper concrete shield that opens into the floor of a fuel handling bay and is sealed by a removable shield plug. A shielded fuel handling machine is used to transfer the fuel assemblies from cask to fuel tube. The bottom of each fuel tube is connected*to a common manifold of a cover gas filling system. The spent fuel is cooled via convection facilitated by the cooling channels built into the concrete structure.

Nominal costs are expected to be comparable to those for the metal cask and horizontal concrete vault options. Seismic qualification requirements for the concrete housing structure raise similar concerns to those for horizontal concrete modules because of the PSE&G site geology. This also increases cost estimate uncertainties. Moreover, although the design allows for modular expansion, the intense construction involved would make this a difficult task.

No clear benefits appear from the postulated use of this technology. There are no striking features which make it more desirable than the other dry storage options like ease of construction, simplicity of operation, low cost, etc. Furthermore, the uncertainties associated with its high cost estimates would not

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• make this a prudent choice. Therefore, this option was disqualified; but, it can be reevaluated as the technology progresses and cost uncertainties decrease.

11.4 Economic Analysis of Viable Options The economic analysis included developing the costs of each storage option separately and then combining the costs of various options, as necessary, to determine the total life cycle costs of the different strategic plans. All known costs (in 1991 dollars) associated with each option were incorporated into the analysis.

Cash flows associated with the capital and O&M costs were analyzed to determine the overall present value cost of a given option. It should be noted that these costs are preliminary estimates that do not include 100% of expected project costs; however, they were considered sufficient to perform a comparative analysis of all the options. At a minimum, the six long-term options provide enough additional storage capacity to support Salem 1, Salem 2, and Hope Creek operation until normal shutdown (2016, 2020, and 2026, respectively). The eight options analyzed and their respective, estimated present value costs were as follows:

PRESENT VALUE COSTS ITEM NEAR-TERM OPTIONS (1991 DOLLARS)

(1) Rerack Sl/S2 Spent Fuel 8.6 Million Pools (2) Convert HC-U2 - Phase 1 21.6 Million (provides storage for S1/S2 fuel) 11-10

• ITEM (3)

LONG-TERM OPTIONS Rerack S1 and S2 followed by conversion of HC-U2 facility in two phases:

PRESENT VALUE COSTS (1991 DOLLARS).

27.7 Million Begin Phase 1 in 2001 (provides storage for S1/S2 fuel)

Begin Phase 2 in 2012 (provides storage of S1/S2/HC Fuel)

(4) Rerack Sl and S2 followed 40.8 Million by concrete casks Begin cask storage option in 2002 (provides storage of Sl/S2/HC fuel)

(5) Convert HC-U2 - Phase 1 28.3 Million in 1992 (provides storage for Sl/S2 fuel)

Begin Phase 2 in 2006 (provides storage for Sl/S2/HC fuel)

(6) Convert HC-U2 - 32.2 Million Phases 1 & 2 together (provides storage for Sl/S2/HC fuel)

(7) Concrete Casks beginning 52.7 Million in 1992 (provides storage for Sl/S2/HC fuel)

(8) NUHOMS beginning in 1992 71.2 Million (provides storage for Sl/S2/HC fuel) 11-11

• 11.5 Decision to Rerack Salem Pools As shown in Section 11.4, reracking the Salem 1 and 2 spent. fuel pools is clearly the least expensive near-term option. It protects our ratepayers interests by minimizing over-expansion and over-commitment of funds.

When the Salem 1 rack change-out is projected *to occur, the Salem 1 pool will be filled to approximately 61% of its capacity. To allow for safe and efficient fuel handling operations during reracking, it is.necessary that the rack change-out occur prior to the pool inventory reaching approx~mately 67% of its capacity.

Thus, expeditious implementation of this reracking is essential to ensure safety and minimize the risks to public health.

The reracking option will provide a timely addition of on-site spent fuel storage capacity and flexibility to adjust our strategy depending on DOE progress or availability of newer and better options with advancing technology. As underwater spent fuel storage is a proven and accepted technology, the public health and safety as well as the quality of the environment will be protected.

11.6 Resource Commitment The proposed rerack of the two Salem pools will consume approximately 300 tons of stainless steel and 60 tons of Boron carbide. These quantities represent less than 0.01 percent of the world output of stainless steel and Boron carbide. Therefore, - it is concluded that the key raw material resources committed in the reracking of the Salem pools represent a minuscule demand on the supply of these materials .

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1I

• 11.7 Environmental Considerations Salem reracking results in additional heat load due to increased spent fuel pool inventory. The anticipated maximum bulk pool temperature, assuming only one fully-degraded spent fuel pool heat exchanger (fully-fouled, maximum allowable number of tubes plugged), is approximately 180°F. Assuming only one maximum fouled heat exchanger in service is . extremely conservative given the actual condition of the Salem heat exchangers. The total heat load (worst case) is less than 40 million Btu/hr, which is less than 0.05% of the total plant heat loss to the environment, and well within the capability of the plant cooling system.

The increased bulk pool temperature results in an increased pool water evaporation rate. This has been calculated to increase the Fuel Handling Building relative humidity by less than 10%. This increase is within the capacity of both (normal and emergency) existing Salem HVAC systems and does not necessitate any hardware modifications for the HVAC system. The environmental impact resulting from the increased heat loss and water vapor emission is negligible.

11.8 References

[11.1] OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications, USNRC (April, 1978).

[11.2] Spent Fuel Storage Strategy to Meet Near and Long-Term Requirements for Salem and Hope Creek Units, NFU-0129, October 28, 1991.

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