ML042390181

Holtec International Report No. HI-2033124 Spent Fuel Storage Expansion at Clinton Power Station.

ML042390181
Person / Time
Site:	Clinton
Issue date:	07/16/2004
From:	Holtec
To:	Office of Nuclear Reactor Regulation
References
1342, RS-04-113 HI-2033124
Download: ML042390181 (257)

Text

ATTACHMENT 6 Holtec International Report No. HI-2033124 "Spent Fuel Storage Expansion at Clinton Power Station" (Non-Proprietary Version)

Eu... Holtec Center, 555 Lincoln Drive West, Mariton, NJ 08053 Telephone (856) 797-0900 HOLTEC INTERNATIONAL Fax (856) 797-0909 SPENT FUEL STORAGE EXPANSION at CLINTON POWER STATION for AMERGEN HOLTEC PROJECT NO. 1342 HOLTEC REPORT HI-2033124 REPORT CATEGORY: A REPORT CLASS: SAFETY RELATED r ,,.,, 7= I* .-.

This document contains proprietary informiation which is the property of 1-loitec Intemational and itsClient. The '

proprietaiy information in the' text has been denoted by backshading or by the'additi6h' 'ofa "Holtec Proprietary". note in instances'where enitirc pages contain proprietary information. 'This'doc'ument is to`blejused 'onlyjin connection with the perforsntances of w6rk'by-Holteec- International orits d bignatyed pacontractohar hReaprodction, publication or*

resetation, in whole or inprfor an other purpose by n pryohr thnteleti xrsl o'iden__

HOLTEC INTERNATIONAL DOCUMENT ISSUANCE AND REVISION STATUS' DOCUMENT NAME: Spent Fuel Storage Expansion at Clinton Power Station DOCUMENT NO.: I HI-2033124 CATEGORY: L GENERIC PROJECT NO.: 1 1342 Z PROJECT SPECIFIC Rev. Date Author's Rev. Date Author's No. 2 Approved Initials VIR # No. Approved Initials VIR #

0 1/30/04 SP 550600 1 6/18/04 SP -666423 2 7/16/04 _ SP 1583678 1 DOCUMENT CATEGORIZATION In accordance with the Holtec Quality Assurance Manual and associated Holtec Quality Procedures (HQPs), this document is categorized as a:

R Calculation Package 3 (Per HQP 3.2) Z Technical Report (Per HQP 3.2)

(Such as a Licensing Report)

R Design Criterion Document (Per HQP 3.4) LI Design Specification (Per HQP 3.4)

Fli Other (Specify):

DOCUMENT FORMATTING The formatting of the contents of this document is in accordance with the instructions of HQP 3.2 or 3.4 except as noted below:

DECLARATION OF PROPRIETARY STATUS This document is labeled:

LI Nonproprietary Z Holtec Proprietary TOP SECRET Documents labeled TOP SECRET contain extremely valuable intellectual/commercial property of Holtec International. They cannot be released to external organizations or entities without explicit approval of a company corporate officer. The recipient of Holtec's proprietary or Top Secret document bears full and undivided responsib to safeguard it against loss or duplication.

Thls~ doumn ha bee reotes  ; .- -::t:-c-n1.e' sdocmenthasbeensubjected t review vericionand approva process set forthin'thi HoltecQ6uaiit urance rocedures Manual. Password controlled signatuiesof Hoitec personnel who participated in the preparation, review, and QA validation of this documfient are saved in the N-drive of the company's network The' Validation IdentifierRecord (VIR) number is a'unique six-digitraindom number that is generated by the computer after the specific irevision' bf this document has undergone the requiredrei'levw and approval process,and the. -

iappropriate Holtec personnei ha've recorded their pa'ss"word-controlled elecfrrnic concurrence to the document.-

2. . - revision to' this doi'ent wilI bee ordered by the Project Manager and carried out -if any of its contents is.

aterially affected 'during evolution of this :roject.:The determination as to the need for revision will 'be mnade by,

'Project Manag'e'r with input.frm others, as deemed necessary'by him. - ythe

.3. -. 'Revisions to Calculation Packges may be made by adding'supplements to the document and replacing the i-Tabeof Contents the 'Revew and Certification page and the" Revision Log"`.:-, -

I.-

SUMMARY

OF REVISIONS Revision 2 contains the folloiving pages:

COVER PAGE 1 page DOCUMENT ISSUANCE AND REVISION STATUS 1 page

SUMMARY

OF REVISIONS 1 page TABLE OF CONTENTS 9 pages

1.0 INTRODUCTION

7 pages 2.0 CASK PIT STORAGE RACKS 20 pages 3.0 MATERIAL AND HEAVY LOADS CONSIDERATIONS 12 pages 4.0 CRITICALITY SAFETY ANALYSES 31 pages

-- APPENDIX 4A 26 pages 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 33 pages 6.0 STRUCTURAL/SEISMIC CONSIDERATIONS 66 pages 7.0 FUEL HANDLING AND CONSTRUCTION ACCIDENTS 16 pages 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 13 pages 9.0 RADIOLOGICAL EVALUATION 5 pages 10.0 INSTALLATION 8 paes 11.0 ENVIRONMENTAL COST/BENEFIT ASSESSMENT 7 pages TOTAL 256 pages Revision I incorporates client comments. Added reference in Section 2.6 (5) and Reference Section 2.7. Revised the word "skip" to the word "intermittent" in Section 2.6.1. Corrected footers and pagination on pages 6-13 thru 6-18.

Corrected wording "Cask Loading" to "Fuel Cask Storage" throughout Section 11. Corrected typographical error on page 11-4. Revised Code versions in Section 6.13 to be consistent with plant design basis. Revised Section 5 to reflect results of updated thermal-hydraulic evaluations. Revised Section 3 and other references to clarify neutron absorber material selection.

Revision 2 incorporates client comments. The word "borated" on page 1-2 is corrected to "unborated". Dimensions on Figures 1.1.1 to 1.1.3 were adjusted to reflect nominal values. The software ORIGEN2 has been added as a reference on pages 5-5 and 5-15.

Ri 1342 Holtec Report HI-2033 HI-2033124124 RI 1342

TABLE OF CONTENTS

1.0 INTRODUCTION

............................................. 1-1 1.1 References . ............................................. 1-4 2.0 STORAGE RACKS DESCRIPTION ........... ..................... 2-1 2.1 Introduction . ............................................. 2-1 2.2 Summary of Principal Design Criteria ........... ..................... 2-2 2.3 Applicable Codes and Standards ................ .................... 2-3 2.4 Quality Assurance Program ......................................... 2-9 2.5 Mechanical Design ............................................. 2-9 2.6 Rack Fabrication ............................................. 2-10 2.6.1 Rack Modules ............................................. 2-10 2.7 References . ............................................. 2-12 3.0 MATERIAL AND HEAVY LOAD CONSIDERATIONS ...... .......... 3-1 3.1 Introduction . ............................................. 3-1 3.2 Structural Materials ............................................. 3-1 3.3 Neutron Absorbing Material . ...................................... 3-1 3.3.1 MetamicTM' Neutron Absorber Panels ............ .................... 3-1 3.3.2 Characteristics of MetamicTm . ..................................... 3-3 3.4 Compatibility with Environment ................ .................... 3-4 3.5 Heavy Load Considerations for the Proposed Rack Installations ..... ...... 3-4 3.6 References . ............................................. 3-10 4.0 CRITICALITY SAFETY EVALUATION ......... ................... 4-1 4.1 Introduction . ............................................. 4-1 4.1.1 Purpose ............... .............................. 4-1 4.1.2 Design Criteria and Assumptions ............... .................... 4-3 4.2 Summary and Conclusions ......................................... 4-4 4.3 Input Parameters ............................................. 4-6 4.3.1 Fuel Assembly Specifications . ...................................... 4-6 4.3.2 Storage Rack Cell Specifications ................ .................... 4-6 4.4 Analytical Methodology ........................................... 4-7 4.4.1 Computer Codes and Benchmarking ............ ..................... 4-7 4.4.2 CASMO4 Validation ............................................. 4-8 4.4.3 Calculation of the k. in the SCCG ............... .................... 4-8 4.4.4 Gadolina Effects and Burnup . ...................................... 4-9 4.4.5 Gadolina Rod Locations ........................................... 4-9 4.5 Criticality Analyses and Tolerance Variations ........ ................. 4-11 4.5.1 Nominal Design Case ........................................... 4-11 4.5.1.1 Enrichment Limit Criterion . ...................................... 4-11 4.5.1.2 Maximum kinr in the Standard Cold Core Geometry ...... .............. 4-11 4.5.1.3 Criteria for Minimum Gadolinia Loading ......... ................... 4-12 4.5.2 Uncertainties Due to Manufacturing Tolerances ....... ................ 4-13 4.5.2.1 Rack Manufacturing Tolerances ............... .................... 4-13 Holtec Report HI-2033124 i 1342 Holtec Report HI-2033 124 1342

TABLE OF CONTENTS 4.5.2.2 Boron-1O Loading Variation .............. ......................... 4-13 4.5.2.3 Neutron Absorber Width Tolerance Variation. ......................... 4-14 4.5.2.4 Lattice Spacing Variation .............................. ............ 4-14 4.5.2.5 Stainless Steel Thickness Tolerances ..................... ............ 4-14 4.5.2.6 Zircaloy Flow Channel ................................ ...... ..... 4-14 4.5.3 Fuel Tolerances (Enrichment and Density Uncertainties). ...... ..... 4-14 4.5.4 Uncertainty in Depletion Calculations .................... ...... ..... 4-15 4.5.5 Existing Fuel Assemblies at the Clinton Station ............ ...... ..... 4-15 4.6 Abnormal and Accident Conditions ...................... ...... ..... 4-16 4.6.1 Temperature and Water Density Effects .................. ...... ..... 4-16 4.6.2 Abnormal Location of a Fuel Assembly .................. ...... ..... 4-16 4.6.3 Eccentric Fuel Assembly Positioning .................... ...... ..... 4-17 4.6.4 Dropped Fuel Assembly ............................... ...... ..... 4-17 4.6.5 Fuel Rack Lateral Movement ........................... ...... ..... 4-18 4.7 References ......................................... ...... ..... 4-19 Appendix 4A "Benchmark Calculations" ........ Total of 26 Pages including cover anc1 6 figures 4A.1 Introduction and Summary .................................... ... 4A-I 4A.2 Effect of Enrichment .............. ....... ... 4A-3 4A.3 Effect of ' 0B Loading .............. ....... ... 4A-4 4A.4 Miscellaneous and Minor Parameters ....... ... 4A-5 4A.4.1 Reflector Material and Spacings ..... ....... ... 4A-5 4A.4.2 Fuel Pellet Diameter and Lattice Pitch ....... ... 4A-5 4A.4.3 Soluble Boron Concentration Effects ....... ... 4A-5 4A.5 MOX Fuel ...................... ....... ... 4A-6 4A.6 References ...................... ....... ... 4A-7 5.0 TUTDXM A T VT1YnD A TT TO OnXMcn~T71 ArTTIXTQ

.1 11 anN'V1AL ......

5.1 Introduction ...... ..... .

5.2 Acceptance C^riteria ................................... ...... ..... .

5.3 Assumptions and Input Data ............................ ...... ..... .

5.3.1 Assumptions ;......................................... ...... ..... .

5.3.2 Design Data ......................................... ...... ..... .

5.4 Bulk Pool At ialysis Methodology ........................ ...... ..... .

5.4.1 Time-to-Boil Calculation ............................... ...... 5-7 .

5.4.2 Bulk Pool ter nperature Results ........................... ...... ..... .

5.4.3 Time-to-Boil Results .................................. ...... 5-9 .

5.5 Local Analys,is Methodology .......... ...... ..... .

5.5.1 Peak Clad Te :mperature Calculation ...................... ...... .... 5-11 5.5.2 Local Analys'is Results ................................. ...... .... 5-13 5.6 References ,......................................... ...... .... 5-15 6.0 STRUCTURAL/SEISMIC CONSIDERATIONS ..... ..... .

6.1 Introduction .................................. ..... .

6.2 Overview of Rack Structural Analysis Methodology . . ..... .

ii 1342 Holtec HI-2033 124 Holtec Report Hl-2033124 ii 1342

TABLE OF CONTENTS 6.2.1 Background of Analysis Methodology ............... ................ 6-2 6.3 Description of Racks . ....................................... 6-6 6.4 Synthetic Time-Histories ....................................... 6-7 6.5 WPMR Methodology ....................................... 6-8 6.5.1 Model Details for Spent Fuel Racks ................................. 6-8 6.5.1.1 Assumptions ....................................... 6-8 6.5.1.2 Element Details ................. ...................... 6-10 6.5.2 Fluid Coupling Effect ....................................... 6-11 6.5.2.1 Multi-Body Fluid Coupling Phenomena ............. ................ 6-12 6.5.3 Stiffness Element Details ....................................... 6-12 6.5.4 Coefficients of Friction ....................................... 6-13 6.5.5 Governing Equations of Motion .................................... 6-14 6.6 Structural Evaluation of Spent Fuel Rack Design ........ .............. 6-15 6.6.1 Kinematic and Stress Acceptance Criteria ............ ................ 6-15 6.6.2 Stress Limit Evaluations ....................................... 6-16 6.6.3 Dimensionless Stress Factors ...................................... 6-20 6.6.4 Loads and Loading Combinations for Spent Fuel Racks ...... ........... 6-21 6.7 . Parametric Simulations ....................................... 6-22 6.8 Time History Simulation Results ................................... 6-30 6.8.1 Rack Displacements ........................................ 6-30 6.8.2 Pedestal Vertical Forces ....................................... 6-31 6.8.3 Pedestal Friction Forces ....................................... 6-31 6.8.4 Rack Impact Loads . ....................................... 6-31 6.8.4.1 Impacts External to the Rack ...................................... 6-31 6.8.4.2 Impacts Internal to the Rack ....................................... 6-32 6.9 Rack Structural Evaluation ....................................... 6-33 6.9.1 Rack Stress Factors . ....................................... 6-33 6.9.2 Pedestal Thread Shear Stress ..................................... 6-34 6.9.3 Local Stresses Due to Impacts ..................................... 6-34 6.9.4 Assessment of Rack Fatigue Margin ................................. 6-35 6.9.5 Weld Stresses ............. .......................... 6-37 6.9.6 Bearing Pad Analysis ........................................ 6-39 6.10 Level A Evaluation . ....................................... 6-41 6.11 Hydrodynamic Loads on Pool Walls ................. ............... 6-41 6.12 Local Stress Considerations ....................................... 6-42 6.12.1 Cell Wall Buckling . ....................................... 6-42 6.12.2 Analysis of Welded Joints in the Racks .............. ................ 6-42 6.14 References ....................................... 6-44 7.0 MECHANICAL ACCIDENTS ..................................... 7-1 7.1 Introduction ....................................... 7-1 7.2 Description of Mechanical Accidents ................. ............... 7-1 7.3 Incident Impact Velocity ....................................... 7-3 7.4 Mathematical Model . ....................................... 7-4 7.5 Results .................. ............................ 7-5 Holtec Report HI-2033124 iii 1342

TABLE OF CONTENTS 7.5.1 Shallow Drop Event .......................................... 7-5 7.5.2 Deep Drop Events . ......................................... 7-5 7.6 Conclusion ............. ............................ 7-6 7.7 References ............. ............................ 7-7 8.0 AUXILIARY BUILDING STRUCTURAL INTEGRITY EVALUATIONS .. 8-1 8.1 Introduction ....... 8-1 8.2 Description of the SFP West Wall and Fuel Cask Storage Pool South Wall .. 8-3 8.3 Analysis Procedures .......................................... 8-3 8.4 Definition of Loads Included in Structural Evaluation ....... ............ 84 8.4.1 Static Loading ................. ........................ 8-4 8.4.2 Seismic ......................................... 8-4 8.4.3 Thermal Loading . ......................................... 8-5 8.4.4 Load Combinations and Acceptance Criteria ......... ................. 8-5 8.5 Results of Reinforced Concrete Analyses ........... .................. 8-6 8.6 Pool Liner Evaluation ......................................... 8-7 8.7 Bearing Evaluation .......................................... 8-7 8.8 Conclusions .............. ........................... 8-8 8.9 References . ............................................ 8-9 9.0 RADIOLOGICAL EVALUATION .................................. 9-1 9.1 Fuel Handling Accident ......................................... 9-1 9.2 Solid Radwaste .................. ....................... 9-1 9.3 Gaseous Releases . ......................................... 9-1 9.4 Personnel Exposures ......................................... 9-1 9.5 Anticipated Exposure During Storage Expansion ........ ............... 9-3 9.6 References ... ......................................... 9-4 10.0 INSTALLATION ............................................ 10-1 10.1 Introduction ............... .......................... 10-1 10.2 Rack Arrangement ......................................... 10-4 10.3 Rack Interferences .......................................... 10-4 10.4 SFP Cooling ......................................... 10-4 10.5 Installation of New Racks and Removal of Existing Racks ..... ......... 10-5 10.6 Safety, Health Physics, and ALARA Methods ........ ................ 10-6 10.6.1 Safety ......................................... 10-6 10.6.2 Health Physics . ......................................... 10-6 10.6.3 ALARA ............ ............................. 10-7 10.7 Radwaste Material Control ....................................... 10-8 11.0 ENVIRONMENTAL COST/BENEFIT ASSESSMENT ...... .......... 11-1 11.1 Introduction ................... ...................... 11-1 11.2 Imperative for Additional Spent Fuel Storage Capacity ...... ........... 11-1 11.3 Appraisal of Alternative Options ................................... 11-1 iv 1342 111-2033124 Holtec Report HI-2033124 iv 1342

TABLE OF CONTENTS 11.3.1 Alternative Option Cost Summary ........ ............... 11-4 11.4 Cost Estimate ....................... 11-4 11.5 Resource Commitment ....................... 11-5 11.6 Environmental Considerations ....................... 11-5 11.7 References ....................... 11-7 Holtc ReortHI-233 V134 24 Holtec Report HI-2033124 v 1342

TABLE OF CONTENTS Tables 2.1.1 Geometric and Physical Data for Storage Racks ............................. 2-13 2.5.1 Module Data for New BWR Racks ........... .................. 2-14 3.5.1 Heavy Load Handling Compliance Matrix (NUREG-0612) .3-13 4-1 Fuel Assembly Design Specifications Used in the Analyses .4-20 4-2 Summary of Criticality Safety Analyses for Fuel of 3.3% Enrichment w/out Gd2O3 or Burnup .4-21 4-3 Summary of Criticality Safety Analyses at Design Basis Enrichment and Burnup ... 4-22 4-4 Reactivity Effects of Abnormal and Accident Conditions .4-23 4-5 Reactivity Uncertainties Due to Manufacturing Tolerances .4-24 4-6 Comparison of CASMO4 and MCNP4a Calculations .4-25 4-7 Reactivity Effects of Gd 2 O3 Rod Locations .4-26 4-8 Effect of temperature and Void on Calculated Reactivity of the Storage Rack . 4-27 4A.1 Summary of Criticality Benchmark Calculations ..... ............. 4A-9 thru 4A-13 4A.2 Comparison of MCNP4a and KenoSa Calculated Reactivities for Various Enrichments ................ .................. 4A-14 4A.3 MCNP4a Calculated Reactivities for Critical Experiments with Neutron Absorbers ................ .................. 4A-15 4A.4 Comparison of MCNP4a and KENOSa Calculated Reactivities for Various

' 0B Loadings ................. ................. 4A-16 4A.5 Calculations for Critical Experiments with Thick Lead and Steel Reflectors ..... 4A-17 4A.6 Calculations for Critical Experiments with Various Soluble Boron ,

Concentrations ..................................................... 4A-18 4A.7 Calculations for Critical Experiments with MOX Fuel ......... ............. 4A-19 5.1.1 Partial Listing of Rerack Applications Using Similar Methods of Thermal-Hydraulic Analysis ........... 5-16 through 5-17 5.3.1 Inputs for Bulk Pool Analysis ........... 5-18 5.3.2 Inputs for Local Analysis ........... 5-19 5.3.3 Fuel Bundle Data ........... 5-20 5.4.1 Summary of Bulk Pool Temperature Results ........... 5-21 5.4.2 Summary of Time-to-Boil Analysis Results ........... 5-22 5.5.1 Results of Local temperature Analysis Results ........... 5-23 6.2.1 Partial Listing of Fuel Rack Applications Using DYNARACK ...... 6-46 through 6-48 6.3.1 Rack Material Data (2007F) (ASME - Section II, Part D) ..... ...... 6-49 6.4.1 Time-History Statistical Correlation Results ........... 6-50 6.5.1 Degrees-of-Freedom ........... 6-51 6.9.1 Comparison of Bounding Calculated Loads/Stresses vs. Code Allowables at Impact Locations and at Welds ........... 6-52 Holtec Report HI-2033124 Vi 1342 Holtec Report HI-2033 124 vi 1342

TABLE OF CONTENTS 7.4.1 Impact Event Data .......................... 7-8 7.4.2 Material Definition .......................... 7-9 8.1.1 Reinforced Concrete Properties .......................... 8-10 9.1 Preliminary Estimate of Person-REM Exposures During the Expansion of Fuel-Storage Capacity ....................... 9-5 Holtec Report HI-2033124 ..i 1342 Holtc Reortvii134 1-23312

TABLE OF CONTENTS Figures 1.1.1 Phase I - Fuel Cask Storage Pool Rack Layout 1.1.2 Phase 2 - Spent Fuel Pool Rack Layout 1.1.3 Phase 2 - Fuel Cask Storage Pool Existing Racks Layout 2.1.1 Pictorial View of Typical BWR Rack Module 2.6.1 Fabricated Square Cross-Section Cell Box 2.6.2 Composite Box Assembly 2.6.3 Typical array of Storage Cells 2.6.4 Elevation View of Cells 2.6.5 Support Pedestals for BWR Racks 4-1 Storage Cell Calculation Model 4-2 Correlation Between k-Infinite in the SCCG and k-Infinite in the Storage Rack 4-3 Variation in Rack reactivity with Burnup for 4.8% Fuel with Gd 2O3 4-4 Correlation of Reactivity in Rack and k-Infinite in the SCCG for Fuel of Various Designs 4A.l MCNP Calculated k-eff Values for Various Values of the Spectral Index 4A.2 KEN05a Calculated k-eff Values for Various Values of the Spectral Index 4A.3 MCNP Calculated k-eff Values at Various U-235 Enrichments 4A.4 KEN05a Calculated k-eff Values at Various U-235 Enrichments 4A.5 Comparison of MCNP and KEN05a Calculations for Various Fuel Enrichments 4A.6 Comparison of MCNP and KENO5a Calculations for Various Boron-I 0 Areal Densities 5.4.1 Spent fuel Pool Cooling Model 5.4.2 Normal Discharge Scenario Bulk Pool Temperature Plot 5.4.3 Full Core Discharge Scenario Bulk Pool Temperature Plot 5.4.4 Normal Discharge Scenario Fuel Pool Decay Heat Plot 5.4.5 Full Core Discharge Scenario Pool Decay Heat Plot 5.4.6 Water Depth Plot in a Loss of Cooling Event (Case I) 5.4.7 Water Depth Plot in a Loss of Cooling Event (Case II) 5.5.1 Plan View of CFD Model (Full Core Discharge) 5.5.2 Hot Region Temperature Contours (Case I Discharge Scenario) 5.5.3 Hot Region Temperature Contours (Case II Discharge Scenario) 6.4.1 Time History Accelerogram CPS Auxiliary-Fuel Bldg. At Elev. 712', SSE 4% Damping (X - East-West Direction) 6.4.2 Time History Accelerogram CPS Auxiliary-Fuel Bldg. At Elev. 712', SSE 4% Damping (Y - North-South Direction) 6.4.3 Time History Accelerogram CPS Auxiliary-Fuel Bldg. At Elev. 712', SSE 4% Damping (Z - Vertical Direction)

Holtec Report HI-2033124 . .i. 1342 Holtc iii134 RportHI-033124

TABLE OF CONTENTS 6.4.4 Time History Accelerogram CPS Auxiliary-Fuel Bldg. At Elev. 712', OBE 2% Damping (X - East-West Direction) 6.4.5 Time History Accelerogram CPS Auxiliary-Fuel Bldg. At Elev. 712', OBE 2% Damping (Y - North-South Direction) 6.4.6 Time History Accelerogram CPS Auxiliary-Fuel Bldg. At Elev. 712', OBE 2% Damping (Z - Vertical Direction) 6.5.1 Schematic of the Dynamic Model of a Single Rack Module Used in DYNARACK 6.5.2 Fuel-to-Rack Gap/Impact Elements at Level of Rattling Mass 6.5.3 Two Dimensional View of the Spring-Mass Simulation 6.5.4 Rack Degrees-of-Freedom with Shear and Bending Springs 6.5.5 Rack Periphery Gap/Impact Elements 6.9.1 Isometric of ANSYS Model 6.9.2 Stress Distribution in Bearing Pad 6.12.1 Loading on Rack Wall 6.12.2 Welded Joint in Rack 7.2.1 Finite Element Model of the "Shallow" Drop Event 7.2.2 Schematic of the"Deep" Drop Scenario 1 7.2.3 Schematic of the"Deep" Drop Scenario 2 7.5.1 "Shallow" Drop: Maximum Plastic Strain 7.5.2 "Deep" Drop Scenario 1: Maximum Vertical Displacement 7.5.3 "Deep" Drop Scenario 2: Maximum Von Mises Stress - Liner 7.5.4 "Deep" Drop Scenario 2: Maximum Compressive Strain - Concrete 8.1 Plan of Clinton Spent Fuel Pool Area 8.2 SFP West Wall Finite Element Model 8.3 Fuel Cask Storage Pool South Wall Finite Element Model Holtec Report HI-2033 124 1X 1342 Holtec Report HI-2033 124 ix 1342

1.0 INTRODUCTION

The Clinton Power Station is operated by Amergen Energy Company, LLC (Amergen). The single unit nuclear power facility is located in Harp Township, DeWitt County, approximately 6 miles east of the city of Clinton, Illinois. Unit I of the Clinton Power Station has a boiling water reactor nuclear steam supply system with 624 fuel assemblies in the core. The plant was designed by General Electric and is designated as a BWR/6 unit and has a licensed rated power level of 3473 MWt. The unit currently uses a Spent Fuel Pool (SFP) for storage of irradiated nuclear fuel between refueling outages in order to maintain a subcritical array, remove decay heat and provide radiation shielding.

The SFP is currently licensed for 2,512 fuel assembly storage locations and 10 failed fuel container storage locations, arranged in twenty-two distinct rack modules. An additional storage capacity of 160 assemblies exists in the Upper Containment Storage Pool. However, the upper pool capacity is commonly relied upon only during refueling and is not considered for long-term fuel storage. Therefore, full core offload will be lost once the SFP contains 2,512 less 624 fuel assemblies, or 1,888 assemblies.

Based on the current inventory of 1,312 fuel assemblies stored in the spent fuel pool and the anticipated future discharges of spent fuel, loss of full core reserve capacity will occur during the scheduled February 2006 refueling outage when an anticipated 312 fuel assemblies are permanently discharged and new fuel is loaded into the SFP during Operating Cycle 11.

To maintain prudent storage reserve, Amergen intends to expand spent fuel storage capacity in two phases. Phase I consists of adding two new 15 by 12 cell racks within the Fuel Cask Storage Pool by January 2005. The completed Fuel Cask Storage Pool configuration is shown in Figure 1.1.1. This modification would increase the licensed storage capacity from the current 2,512 storage cells to 2,872 storage cells. During Phase 2 the two racks in the Fuel Cask Storage Pool will be relocated into the SFP along with 14 more new racks. During this second phase, 12 of the existing racks will be removed from the SFP and three of these existing racks to be removed will be placed within the Fuel Cask Storage Pool. The final, Phase 2, configurations of the SFP and the Fuel Cask Storage Pool are shown in Figures 1.1.2 and 1.1.3, respectively.

Holtec Report HI-2033124 1-1 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

This report provides the design basis, analysis methodology, and evaluation results for the proposed storage racks at Clinton Power Station to support the licensing process. The rack design and analysis methodologies employed are a direct evolution of previous license applications. This report documents the design and analyses performed to demonstrate that the racks meet 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 [1].

The new Cask Area storage racks are freestanding and self-supporting. The principal construction materials for the SFP 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 discussed in Section 3. The racks are designed to the stress limits of, and analyzed in accordance with,Section III, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel (B&PV) Code [2]. The material procurement, analysis, fabrication, and installation of the rack modules conform to I OCFR50 Appendix B requirements.

Sections 2 and 3 of this report provide an abstract of the design and material information on the racks.

Section 4 provides a summary of the methods and results of the criticality evaluations performed for the Cask Area storage racks. The criticality safety analysis requires that the effective neutron multiplication factor (kff) is less than or equal to 0.95 with the storage racks fully loaded with fuel of the highest permissible reactivity and the pool flooded with unborated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivities include a margin for uncertainty in reactivity calculations, including manufacturing tolerances, and are calculated with a 95% probability at a 95%

confidence level. The criticality safety analysis sets the requirements on the neutron absorber panel length and the amount of B' 0 per unit area (i.e., loading density) for the new racks.

Thermal-hydraulic consideration requires that fuel cladding will not fail due to excessive thermal stress, and that the steady state pool bulk temperature will remain within the limits prescribed for the Fuel Cask Storage Pool and Spent Fuel Pool to satisfy the pool structural strength, operational, and regulatory Holtec Report HI-2033124 1-2 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

requirements. The thermal-hydraulic analyses carried out in support of this storage expansion effort are described in Section 5.

Rack module structural analysis requires that the primary stresses in the rack module structure will remain below the ASME B&PV Code (Subsection NF) [2] allowables. Demonstrations of seismic and structural adequacy are presented in Section 6.0. The structural qualification also requires that the subcriticality of the stored fuel will be maintained under all postulated accident scenarios. The structural consequences of these postulated accidents are evaluated and presented in Section 7 of this report.

Section 8 discusses the evaluation of the Fuel Cask Storage Pool and Spent Fuel Pool structures to withstand the new rack loads. The radiological considerations are documented in Section 9.0. Section 10 discusses the salient considerations in the installation of the new racks.Section I I discusses a cost/benefit and environmental assessment to establish the acceptability of the wet storage expansion option.

All computer programs utilized to perform the analyses documented in this report are benchmarked and verified. Holtec International has utilized these programs in numerous license applications over the past decade.

The analyses presented herein clearly demonstrate that the new racks possess wide margins of safety with respect to all considerations of safety specified in the OT Position Paper [I], namely, nuclear subcriticality, thermal-hydraulic safety, seismic and structural adequacy, radiological compliance, and mechanical integrity.

Holtec Report HI-2033124 1-3 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

1.1 References

[1] USNRC, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14, 1978, and Addendum dated January 18, 1979.

[2] American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel Code,Section III, 1977 Edition, Subsection NF, and Appendices.

Holtec Report HI-2033124 1-4 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

LF 294.U REr.

EA REf.- 94.227 PtU. VAIL

,/ -INSIDE DIMENSIEW (REr3

,I TH I I I I I I I I 1, 1 93.8

_ _o _ _ _0 (REF.)

I__L_No I I I I II fr1Dnrtx A 1 T __

0104D REfl 753 se _ ___1 8) ICE L9I 163D REF.

t7D REf. 71 REF.d LO

_ Q - - -o ne~~~~~~ I 0C l l¢ 1~

1 0111 I 1~ 11dIL

-101 - 1 E I1_ 1110 111 75r I IrlI

---

1-P- lr I

-YIU, 4 I1 111

-T Id~1°l1 1 IC

- I 11 1°0b (RErJ

- .1. -

I \-rUEL. TRAN~fERCASK FIGURE 1.1.1; PHASE 1 - FUEL CASK STORAGE POOL RACK LAYOUT DIMENSIONS ARE NOMINAL VALUES FOR REFERENCE ONLY IHI-2033124 H1342 HI-2033124 MVDUn2MVWlAM-N40V(Wj=.U

FIGURE 1.1.2; PHASE 2 - SPENT FUEL POOL RACK LAYOUT H1342l DIMENSIONS ARE NOMINAL VALUES SHOWN FOR REFERENCE ONLY lHI-2033124 r.DRAVD\3f134tREtT-MW7124\1PT-lG- J2

F04 nV 2' RCF.

1.0 R.-

7t9gRS -i 70iqz9 1 I.L II

'- .. . . . . . . I' . . . .. . . . . . 1- . . . . .-

P' R~F Z .L%~ .Lk. L .- L -J.

_1_ 1 1 Il I 'T 51 u

_ ux S __ng. .1,

_~~-

_ __ B@1tI __i __ __rii __ __ __ g, __

5.9 REr.

0104' REF.

165.0' I 10  := = = = = = = = = = =

I I, I I I 1 I1 I M--TRAC1253 rUEL TRANSFERCASK I IC!TI:I 9318 REU.) - 16 REF.

77.422 111X ICE-115 TbY 9I2I 1 L

53 REF.

I I FIGURE 1.1.3; PHASE 2 -

FUEL CASK STORAGE POOL RACK LAYOUT DDlMW NE G1 WE tm FM FE tE H1342 lHI-2033124

&\DRAVIM O-FIWRS\1342T fRT-HI2033124\OPj\FIG113

2.0 STORAGE RACKS DESCRIPTION 2.1 Introduction The new Clinton Power Station (CPS) fuel storage racks will be freestanding modules, made primarily from Type 304L austenitic stainless steel containing honeycomb storage cells interconnected through longitudinal welds. Neutron absorber panels containing a high areal loading of the boron- 10(B-10) isotope provide appropriate neutron attenuation between adjacent storage cells.

Figure 2.1.1 provides an isometric schematic of a typical Region 2 storage rack module. Data on the cross sectional dimensions, weight and cell count for the rack modules are presented in Table 2.1.1. All of the new rack modules are designed and constructed to have identical configurations, except for the number of storage cells.

The baseplates on all spent fuel rack modules extend approximately 1/2" beyond the rack module periphery wall such that the plate protrusions act to maintain rack-to-rack distances. Each rack is supported by four pedestals, which are remotely height-adjustable. The rack module support pedestals are engineered to accommodate minor level adjustments. Thus, the racks can be made plumb during installation. The cell height and overall height of the racks is chosen to be similar to the existing racks.

The similar cell height ensures that the existing fuel grapple can properly engage bundles and provides sufficient access for storage and removal of fuel channels. The overall rack height similarity coupled with the adjustable pedestal heights also ensures that the top of the racks will be approximately co-planar with all other new racks and the existing racks in the pool.

The overall design of the rack modules is similar to those presently in service in the spent fuel pools at many other nuclear plants, among them Hatch and J.A. FitzPatrick. Altogether, over 50 thousand storage cells of this design have been provided by Holtec International to various nuclear plants around the world.

Holtec Report HI-2033124 2-1 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

2.2 Summary of Principal Design Criteria The key design criteria for the new racks are set forth in the USNRC memorandum entitled "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14, 1978 as modified by amendment dated January 18, 1979. The individual sections of this report address the specific design bases derived from the above-mentioned "OT Position Paper". The design bases for the new racks are summarized in the following:

a. Disposition: New rack modules are required to be free-standing.

b. Kinematic Stability: Each freestanding module must be kinematically stable (against tipping or overturning) if a seismic event is imposed.

c. Structural Compliance: All primary stresses in the rack modules must satisfy the limits postulated in Section III subsection NF of the ASME B & PV Code.

d. Thernal-Hydraulic Compliance: The spatial average bulk pool temperature is required to remain below 150F in the wake of a normal partial core offload or a full core offload.

e. Criticality Compliance: The Holtec high-density spent fuel storage racks are designed to assure that the neutron multiplication factor (kff) is equal or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with unborated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations and in manufacturing tolerances, statistically combined, giving assurance that the true kff will be equal to or less than 0.95 with a 95% probability at a 95% confidence level. Reactivity effects of the abnormal and accident conditions have also been evaluated to assure that under credible abnormal and accident conditions, the reactivity will be maintained less than 0.95.

Holtec Report HI-2033 124 2-2 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

f. Accident Events: In the event of postulated drop events (uncontrolled lowering of a fuel assembly, for instance), it is necessary to demonstrate that the subcritical geometry of the rack structure is not compromised.

The foregoing design bases are further articulated in Sections 4 through 7 of this licensing report.

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

a. Design Codes (1) American Institute of Steel Construction (AISC) Manual of Steel Construction, 9 th Edition, 1989.

(2) American National Standards Institute/ American Nuclear Society ANSI/ANS-57.2-1983, "Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants" (contains guidelines for fuel rack design).

(3) ASME B&PV Code Section 111, 1977 Edition; ASME Section IX, 1977 Edition.

(4) American Society for Nondestructive Testing SNT-TC-IA, June 1980, Recommended Practice for Personnel Qualifications and Certification in Non-destructive Testing.

(5) American Concrete Institute Building Code Requirements for Reinforced Concrete (ACI 318-71).

(6) Code Requirements for Nuclear Safety Related Concrete Structures, ACI 349-76/ACI 349R-76, and ACI 349.1R-80.

(7) ASME Y14.5M, Dimensioning and Tolerancing (8) ASME B&PV Code, Section Il-Parts A and C, 1977 Edition.

(9) ASME B&PV Code NCA3800 - Metallic Material Organization's Quality System Program.

Holtec Report HI-2033124 2-3 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

b. Standards of American Society for Testing and Materials (ASTM)

(1) ASTM E165 - Standard Test Method for Liquid Penetrant Examination.

(2) ASTM A240 - Standard Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet and Strip for Pressure Vessels.

(3) ASTM A262 - Standard Practices for Detecting Susceptibility to Intergranular Attack in Austenitic Stainless Steel.

(4) ASTM A276 - Standard Specification for Stainless Steel Bars and Shapes.

(5) ASTM A479 - Standard Specification for Stainless Steel Bars and Shapes for use in Boilers and other Pressure Vessels.

(6) ASTM A564 - Standard Specification for Hot-Rolled and Cold-Finished Age-Hardening Stainless Steel Bars and Shapes.

(7) ASTM C750 - Standard Specification for Nuclear-Grade Boron Carbide Powder.

(8) ASTM A380 - Standard Practice for Cleaning, Descaling, and Passivation of Stainless Steel Parts, Equipment and Systems.

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

(10) ASTM E3 - Standard Practice for Preparation of Metallographic Specimens.

(11) ASTM El90 - Standard Test Method for Guided Bend Test for Ductility of Welds.

c. Welding Code:

ASME B&PV Code,Section IX - Welding and Brazing Qualifications, 1977.

d. Qualitv Assurance, Cleanliness. Packaging. Shipping. Receiving. Storage. and Handling (1) ANSI N45.2.1 - Cleaning of Fluid Systems and Associated Components during Construction Phase of Nuclear Power Plants - 1973 (R.G. 1.37).

(2) ANSI N45.2.2 - Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants - 1972 (R.G. 1.38).

(3) ANSI N45.2.6 - Qualifications of Inspection, Examination, and Testing Personnel for the Construction Phase of Nuclear Power Plants - 1978. (R.G. 1.58).

Holtec Report HI-2033124 2-4 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

(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 - 1975 (R.G. 1.11 6).

(5) ANSI N45.2.11 - Quality Assurance Requirements for the Design of Nuclear Power Plants - 1974 (R.G. 1.64).

(6) ANSI N45.2.12 - Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants - 1977 (R.G. 1.144).

(7) ANSI N45.2.13 - Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants - 1976 (R. G. 1.123).

(8) ANSI N45.2.23 - Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants - 1978 (R.G. 1.146).

(9) ASME B&PV Code,Section V, Nondestructive Examination, 1977 Edition.

(10) ANSI N 16.9 Validation of Calculation Methods for Nuclear Criticality Safety.

(11) ASME NQA- I - Quality Assurance Program Requirements for Nuclear Facilities.

(12) ASME NQA Quality Assurance Requirements for Nuclear Power Plants.

e. USNRC Documents (1) "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.

(2) 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) ANSI/ANS 8.1 - Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.

(2) ANSI/ANS 8.17 - Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors.

(3) ANSI N45.2 - Quality Assurance Program Requirements for Nuclear Power Plants - 1977.

Holtec Report HI-2033124 2-5 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

(4) ANSI N45.2.9 - Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants - 1974.

(5) ANSI N45.2.10 - Quality Assurance Terms and Definitions - 1973.

(6) ANSI N14.6 - American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500 kg) or more for Nuclear Materials - 1993.

(7) ANSI/ASME N626 Qualification and Duties of Specialized Professional Engineers.

(8) ANSI/ANS- 57.3 - Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants.

g. Code-of-Federal Regulations (CFR)

(1) IOCFR20 - Standards for Protection Against Radiation.

(2) 10CFR21 - Reporting of Defects and Non-compliance.

(3) 10CFR50 Appendix A - General Design Criteria for Nuclear Power Plants.

(4) 10CFR50 Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.

(5) 10CFR61 - Licensing Requirements for Land Disposal of Radioactive Waste.

(6) 10CFR71 - Packaging and Transportation of Radioactive Material.

(7) 10CFR100 - Reactor Site Criteria (8) 10CFR50.68 ""Criticality Accident Requirements"

h. Regulatory Guides (RG)

(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 for Boiling and Pressurized Water Reactors, Rev. 0 - March, 1972.

(3) RG 1.28 - Quality Assurance Program Requirements - Design and Construction, Rev. 2 - February, 1979 (endorses ANSI N45.2).

Holtec Report HI-2033124 2-6 1342

'SHADED AREAS DENOTE PROPRIETARY INFORMATION

(4) RG 1.33 - Quality Assurance Program Requirements.

(5) RG 1.29 - Seismic Design Classification, Rev. 2 - February, 1976.

(6) RG 1.31 - Control of Ferrite Content in Stainless Steel Weld Metal.

(7) RG 1.38 - Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants, Rev. 2 -

May, 1977 (endorses ANSI N45.2.2).

(8) RG 1.44 - Control of the Use of Sensitized Stainless Steel.

(9) RG 1.58 - Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel, Rev. 1 - September 1980 (endorses ANSI N45.2.6).

(10) RG 1.60 - Design Response Spectra for Seismic Design of Nuclear Power Plants.

(11) RG 1.61 - Damping Values for Seismic Design of Nuclear Power Plants, Rev. 0, 1973.

(12) RG 1.64 - Quality Assurance Requirements for the Design of Nuclear Power Plants, Rev. 2 - June, 1976 (endorses ANSI N45.2.1 1).

(13) RG 1.71 - Welder Qualifications for Areas of Limited Accessibility.

(14) RG 1.74 - Quality Assurance Terms and Definitions, Rev. 2 - February, 1974 (endorses ANSI N45.2.10).

(15) RG 1.85 - Materials Code Case Acceptability - ASME Section III, Division 1.

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

(17) RG 1.92 - Combining Modal Responses and Spatial Components in Seismic Response Analysis, Rev. I - February, 1976.

(18) RG 1.116 - Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems, Rev. 0-R - May,1977 (endorses ANSI N45.2.8-1975)

(19) RG 1.123 - Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants, Rev. I - July, 1977 (endorses ANSI N45.2.13).

Holtec Report HI-2033124 2-7 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

(20) RG 1.124 - Service Limits and Loading Combinations for Class I Linear-Type Component Supports, Revision 1, January, 1978.

(21) RG 1.144 - Auditing of Quality Assurance Programs for Nuclear Power Plants, Rev.I - September, 1980 (endorses ANSI N45.2.12-1977)

(22) RG 3.4 - Nuclear Criticality Safety in Operations with Fissionable Materials at Fuels and Materials Facilities.

(23) RG 8.8 - Information Relative to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be as Low as Reasonably Achievable (ALARA).

(24) IE Information Notice 83 Fuel Binding Caused by Fuel Rack Deformation.

(25) RG 8.38 - Control of Access to High and Very High Radiation Areas in Nuclear Power Plants, June, 1993.

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

j. American Welding Society (AWS) Standards (1) AWS Dl.l - Structural Welding Code - Steel.

(2) AWS Dl .3 - Structure Welding Code - Sheet Steel.

(3) AWS D9.1 - Sheet Metal Welding Code.

(4) AWS A2.4 - Standard Symbols for Welding, Brazing and Nondestructive Examination.

(5) AWS A3.0 - Standard Welding Terms and Definitions.

(6) AWS A5.12 - Specification for Tungsten and Tungsten Alloy Electrodes for Arc-Welding and Cutting (7) AWS QCI - Standard for AWS Certification of Welding Inspectors.

(8) AWS 5.4 - Specification for Stainless Steel Electrodes for Shielded Metal Arc Welding.

(9) AWS 5.9 - Specification for Bare Stainless Steel Welding Electrodes and Rods.

Holtec Report HI-2033124 2-8 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

2.4 Quality Assurance Program The governing quality assurance requirements for design and fabrication of the spent fuel racks are stated in I OCFR50 Appendix B. Holtec's Nuclear Quality Assurance program complies with this regulation and is designed to provide a system for the design, analysis and licensing of customized components in accordance with various codes, specifications, and regulatory requirements.

The manufacturing of the racks will be carried out by Holtec's designated manufacturer, U.S. Tool &

Die, Inc. (UST&D). The Quality Assurance system enforced on the manufacturer's shop floor shall provide for all controls necessary to fulfill all quality assurance requirements. UST&D has manufactured high-density racks for over 60 nuclear plants around the world. UST&D has been audited by the nuclear industry group Nuclear Procurement Issues Committee (NUPIC), and the Quality Assurance branch of the USNRC Office of Nuclear Material Safety and Safeguards (NMSS) with satisfactory results.

The Quality Assurance System that will be used by Holtec to install the racks is also controlled by the Holtec Nuclear Quality Assurance Manual and by the CPS site-specific requirements.

2.5 Mechanical Design The CPS rack modules are designed as cellular structures such that each fuel assembly has a square opening with conforming lateral support and a flat horizontal-bearing surface. All of the storage locations are constructed with multiple cooling flow holes to ensure that redundant flow paths for the coolant are available. The basic characteristics of the racks are summarized in Table 2.5.1.

A central objective in the design of the new rack modules is to maximize structural strength while minimizing inertial mass and dynamic response. Accordingly, the rack modules have been designed to simulate multi-flange beam structures resulting in excellent de-tuning characteristics with respect to the applicable seismic events. The next subsection presents an item-by-item description of the rack modules in the context of the fabrication methodology.

Holtec Report HI-2033124 2-9 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

2.6 Rack Fabrication The object of this section is to provide a brief description of the rack module construction activities, which enable an independent appraisal of the adequacy of design. The pertinent methods used in manufacturing the racks may be stated as follows:

1. The rack modules are fabricated in such a manner that the storage cell surfaces, which would come in contact with the fuel assembly, will be free of harmful chemicals and projections (e.g., weld splatter).

2. The component connection sequence and welding processes are selected to reduce fabrication distortions.

3. The fabrication process involves operational sequences that permit immediate accessibility for verification by the inspection staff.

4. The racks are fabricated per the UST&D Appendix B Quality Assurance program, which ensures, and documents, that the fabricated rack modules meet all of the requirements of the design and fabrication documents.

5. The corners of these storage cells are connected to each other using austenitic stainless steel connector elements, which lead to a honeycomb lattice construction. The extent of welding is selected to "detune" the racks from the seismic input motion [I].

2.6.1 Rack Modules This section describes the constituent elements of the new CPS rack modules in the fabrication sequence. Figure 2.1.1 provides a schematic view of a typical BWR rack.

The rack module manufacturing begins with fabrication of the "box". The boxes are fabricated from two precision formed channels by seam welding in a machine equipped with copper chill bars and pneumatic clamps to minimize distortion due to welding heat input. Figure 2.6.1 shows the box. The minimum Holtec Report HI-2033124 2-10 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

weld seam penetration is 80% of the box metal gage, which is 0.075 inch (14 gage). 3/4 inch diameter holes are punched on at least two sides near the end of the box to provide the redundant flow holes.

Each box constitutes a storage location. Each external box side is equipped with a stainless steel sheathing, which holds one integral neutron absorber sheet (poison material) on each side, except the boxes on the south and west peripheries of each rack, which only have neutron absorber panels on the interior sides. This is because it is only necessary to have neutron absorber panels on one of the two sides of facing racks. The existing racks have neutron absorber panels on all four of the outer sides. The design objective calls for attaching neutron absorber panels tightly on the box surface. This is accomplished by die forming the box sheathings, as shown in Figure 2.6.2. The flanges of the sheathing are attached to the box using intermittent welds and spot welds. The sheathings serve to locate and position the poison sheet accurately, and to preclude its movement under seismic conditions.

Having fabricated the required number of composite box assemblies, they are joined together in a fixture using connector elements in the manner shown in Figure 2.6.3. Figure 2.6.4 shows an elevation view of two storage cells of a BWR rack module. Joining the cells by the connector elements results in a well-defined shear flow path, and essentially makes the box assemblage into a multi-flanged beam-type structure. The "baseplate" is attached to the bottom edge of the boxes. The baseplate is a 5/8 inch thick austenitic stainless steel plate stock which has 3-5/8 inch diameter holes (except at four lift locations, which are modified to accept the lifting rig lug) cut out in a pitch identical to the box pitch. The 3-5/8 inch diameter flow holes are specifically designed to accept the bottom nozzle of the BWR style fuel assembly. The baseplate is attached to the cell assemblage by fillet welding the box edge to the plate.

In the final step, adjustable leg support pedestals (shown in Figure 2.6.5) are welded to the underside of the baseplate. The top (female threaded) portion is made of austenitic steel material. The bottom male threaded part is made of 17:4 Ph series stainless steel to avoid galling problems. All support legs are the adjustable type (Figure 2.6.5), which provide a + 1/2-inch vertical height adjustment at each leg location for leveling the rack. Each support leg is equipped with a readily accessible socket to enable remote leveling of the rack after its placement in the pool.

Appropriate NDE (nondestructive examination) occurs on all welds including visual examination of sheathing welds, box longitudinal seam welds, box-to-baseplate welds, and box-to-box connection welds; and liquid penetrant examination of support leg welds, in accordance with the design drawings.

Holtec Report HI-2033124 2-11 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

2.7 References

[1] Holtec Position Paper WS-1 10, The Detuned Honeycomb Rack Module, Revision 0 dated November 7, 1996.

Holtec Report HI-2033124 2-12 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 2.1.1 Geometric and Physical Data for Storage Racks RACK NO. OF CELLS MODULE ENVELOPE SIZE WEIGHT NO. OF I.D. N-S Direction E-W Direction N-S E-W (Ibs) CELLSPER RACK (in.) (in.)

FI thru F3 10 15 62.9 94.1 14,142 150 GI thruGIO 12 15 75.4 94.1 16,737 180 H 10 11 62.9 69.2 10,637 110 Ji 12 11 75.4 69.2 12,558 132 J2 12 15 75.4 94.1 13,769 144 t t A 4 by 9 array of cells has been removed to provide clearance for existing equipment within the pool, as shown in Figure 1.1.2.

Holtec Report HI-2033 124 2-13 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION.

Table 2.5.1 MODULE DATA FOR NEW BWR RACKS t Storage cell inside nominal dimension 6.05 in.

Cell pitch 6.243 in.

Storage cell height (above the baseplate) 168 in.

Baseplate hole size (except for lift and pedestal locations) 3.625 in.

Baseplate thickness 0.625 in.

Support pedestal height 5.5 in. +/- 0.5 in.

Support pedestal type Remotely adjustable pedestals Number of support pedestals per rack 4 Number of cell walls containing 0.75" diameter flow At Least Two Cell Walls holes at base of cell wall Remote lifting and handling provisions Yes Poison material Metamic Poison length 152 in.

Poison width 4.75 in.

t All dimensions indicate nominal values unless noted.

Holtec Report HI-2033124 2-14 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

STOlRAGE C[LLS

-EXTERIOR NEUTIRON ABSORBER SHEATHING SUPPORT-PEDESTAL (REFER TO FIGURE 2.6.5 FOR ADDITONAL DETAILS)

FIGURE 2.1.1; PICTORiAL viEw OF ATYPICAL BER RACK MODULE NOTE, THE NUMBER OF CELLS SHOWN IS NOT INTENDED TO DEPICT ACTUAL RACK SIZES I - n,,-

PRECISION FORMED CHANNELS FIGURE 2.6.1; FABRICATED SQUARE CROSS-SECTION CELL BOX HI-20331241l G:\RAVINGs\o-rIGKS\1342\REPT-H[2033124\CHT-2\rl62-6-1 I H HI-2033124 &\BRAwINGs\-rIEs\l342\R[PmT-H[?o33l24\ofL2\rIt61 H134 13 4 22

CELL WALL

-TAPERED WELDED AND SMOOTHED ENDS FIGURE 2.6.2; COMPOSITE BOX ASSE4MB LY HI-20331241 H~i13 4 G:\DRAWINGS\Bl-FIGURES\1342\REPOIRT-H[2033124\CIPT _2\FlG_6_2

FABRICATED COMPOSITE BOX CELL FIGURE 2.6.3; TYPICAL ARRAY OF STORAGE CELLS NOTE: THE NUMBER OF CELLS SHOWN IS NOT INTENDED TO DEPICT ACTUAL RACK CELLS HI-20331 24 H1342 G:\DRAY~tNGS\O-FlCURES\1 342\REPORT-H12033124\CHIPT_2\FIG2-6_3

FABRICATED

'COMPOSITE BOX' DEVELOPED CELL CELL 6.243' / SPACER (TYP, FUEL - CELL FOR INNER CELL ASSEMBLY ace I I.

\r- PITCH~1 I

I -

WALLS)

I I I I I I

I POISON PANEL- a..

I II I i I I SHEATHING II 152' ACTIVE POISON LENGTH FLOW HOLE (TYP)

BASE Pi-mATF-I , -

BASEPL ATE HOLE FIGURE 2.6.4: ELEVATION VIEW OF A TYPICAL BWR STORAGE RACK MODULE NOTE: FUEL ASSEMBLY IS NUT INTENDED TO DEPICT ACTUAL CONFIGURATION

\PROJECTS\1342\H12033124\F2_6 4

SUPPORT AND BASEPLATE HOLE MUST BE ALIGNED WITHIN 1/8" BASEPLATE MATL. SA-240-304L 3 1/2

+/-1/16 5 1/2"

+/-1/2"l MATL.: SA-564-630 (AGE-HARDENED 1100*F)

FIGURE 2.6.5; SUPPORT PEDESTALS FOR BWR RACKS H I-20 33 124 I G:\DRAWINGS\O-FIGURES\1342\REPORT-H12033124\CHPT 2\FIG2-6-5l H1342 HI-2033124 G:\]JRAWIN5S\O-FIURES\1342\R[PHRI-HI2O33124\CHPI2\FIG265 13 42

3.0 MATERIAL AND HEAVY LOAD CONSIDERATIONS 3.1 Introduction A primary consideration in the design of the racks proposed in this amendment request is that the materials introduced in the pool water be of proven durability and compatible with the pool water environment. This section provides a synopsis of the considerations that provide the assurance that the rack structural materials and the neutron absorber panels will perform their intended function for the design life of the fuel racks.

3.2 Structural Materials The following structural materials will be utilized in the fabrication of the Clinton fuel racks:

a. ASTM A240-304L for all sheet metal stock and baseplate

b. Internally threaded support legs: ASTM A240-304L

c. Externally threaded support spindle: ASTM A564-630 precipitation hardened stainless steel (heat treated to I1 00 0 F)

d. Weld material - austenitic stainless steel 3.3 Neutron Absorbing Material 3.3.1 Metamicm Neutron Absorber Panels Holtec Report HI-2033124 3-1 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

-' '4 4 -

- 4 1 444 4 4 4

- 4 4.4 4 4 44 4 4444 44 4 4 I 4 >4 -

444 4 4- -'

4 4 44 C

4 4 44 4 4 44 4 444 .4 4 4 4 4 44 4 4 4 4 4 4 4 4 4 4 4 4 4 44 .4 '4 I 4 4 4 4 4 4 4 4444 44 -

4 4 4 4 4 4 4 4 4 4 i, r i: . .

I - L

, T',

4 44 .4 4 1- 44 4 4 4 .4

.4 4 44 4 444 0;: 4 4 4 4 4 4 44 4 4 4 4 .4 4 4 4 4 4 4 4 4 44

.4 4 4 4 .4 4 4 4 4 4 4 44 4 4 , 1. -  : - Z 4 4 4 4 .4 .4 4 .4 4 .4 - 4 4 4 4 4 4 4 44 44.4 4 4 4 4 4 4 4 4 4 4 4 4 .4.4 4 44 4 4 44 4 4 4 4 4 4 .4 44 4 4 4

. 7 1 - " - I I ; 1: ,

44 4 4

-4 Holtec Report HI-2033 124 3-2 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Holtec Report HI-2033D124 3-3 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

3.4 Compatibility with Environment All materials used in the construction of the Holtec racks have been determined to be compatible with the Clinton Power Station Spent Fuel Pool and Fuel Cask Storage Pool, and have an established history of in-pool usage. Austenitic stainless steel (304L) is a widely used stainless alloy in nuclear power plants and has a proven in-service performance as the liner material for many fuel pools in the U.S. and abroad.

Both of the constituent materials used in the neutron absorber, namely boron carbide and aluminum, are chemically compatible and ideally suited for long-term use in the radiation, thermal and chemical environment of a spent fuel pool.

3.5 Heavy Load Considerations for the Proposed Rack Installations The Fuel Building Crane will be used for installing the new racks and removing the existing racks. The Fuel Building Crane is designed as Seismic Category I equipment. The capacity of the main hoist is 125 tons. However, the hoist has been derated to a single failure capacity equivalent to 62 tons to comply with NUREG 0612. The Fuel Building Crane is designed for spent fuel cask handling operations. More specifically, it is used to place casks within the Fuel Cask Storage Pool for removal of spent fuel from the plant. Cask drop accidents have been evaluated previously, as discussed within the plant USAR

[3.5.1]. The cask drop evaluations will remain valid and will not be compromised by placement of racks Holtec Renort HI-2033124 3-4 1342

- SHADED AREAS DENOTE PROPRIETARY INFORMATION

within the Fuel Cask Storage Pool. All fuel will be removed from the Fuel Cask Storage Pool racks before placement of any cask in this vicinity.

The Fuel Building Crane will also be used to lift new and existing racks between the truck bay and the operating deck to enable rack access and egress from the building. However, physical travel limits of the Fuel Building Crane preclude use of the main hook over the east end of the spent fuel pool. Therefore, the Fuel Building Crane cannot be used to install and remove all of the racks within the Spent Fuel Pool during Phase 2 of the project. To overcome this constraint, a low profile temporary crane will be required to install and remove the racks along the east wall. The Fuel Building Crane will be used to lower racks into the pool and place racks within the range of accessibility and to remove racks from the SFP. The temporary crane will be used to lift racks from the pool floor and move the racks horizontally with a limited lift height above the pool floor. The Fuel Building Crane will be used to assemble the temporary crane on the operating deck.

The temporary low profile crane will have a sufficient rated lifting capacity to lift each of the new and old racks, including any additional lifting hardware (i.e., rack lift rig, hoist block, and rigging).

Safe handling of heavy loads by the Fuel Building Crane and temporary crane will be ensured by following the defense in depth approach guidelines of NUREG 0612:

• Defined safe load paths in accordance with approved procedures

• Supervision of heavy load lifts by designated individuals

• Crane operator training and qualification that satisfies the requirements of ANSI/ASME B30.2-1976 [3.5.2]

• Use of lifting devices (slings) that are selected, inspected and maintained in accordance with ANSI B30.9-1971 [3.5.3]

• Inspection, testing and maintenance of cranes in accordance with ANSI/ASME B30.2-1976

• Ensuring the designs of the Fuel Building Crane and the temporary crane meet the requirements of CMAA-70 [3.5.4] and ANSI/ASME B30.2-1976 Holtec Report HI-2033124 3-5 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

• Reliability of special lifting devices by application of design safety margins, and periodic inspection and examinations using approved procedures The salient features of the lifting devices and associated procedures are described as follows:

a. Safe Load Paths and Procedures Safe load paths will be defined for moving the new rack into the Fuel Building. The racks will be lifted by the main hook of the Fuel Building Crane and enter the Fuel Building operating deck through the opening designed for ingress and egress of spent fuel casks. The rack will enter the building at a location adjacent to the area of placement and will not be carried over any portions of the existing storage racks containing active fuel assemblies. A staging area will be setup on the operating deck as a laydown area for racks. The staging area location also will not require any heavy load to be lifted over the pools or any safety-related equipment.

All phases of rack installation activities will be conducted in accordance with written procedures, which will be reviewed and approved by the owner.

b. Supervision of Lifts Procedures used during the installation of the racks require supervision of heavy load lifts by a designated individual who is responsible for ensuring procedural compliance and safe lifting practices. Holtec personnel experienced in similar rack installations will supervise the initial installation of the racks.

c. Crane Operator Training All crew members involved in the use of the lifting and upending equipment will be given training by Holtec International using a videotape-aided instruction course which has been utilized in previous rack installation operations.

Holtec Report HI-2033124 3-6 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

d. Lifting Devices Design and Reliability The Fuel Building Crane is located at the west end of the Fuel Building, where it can access the opening to the truck bay, adjacent laydown area, Fuel Cask Storage Pool, and the west end of the SFP. Physical limitations prevent movement of the hook to the eastern portion of the SFP. The Crane is single failure proof with sufficient capacity to handle all lifts during the reracking process.

The following table determines the maximum lift weight during the installation of the new racks.

Item Weight (Ibs)

Rack 16,737 (max.)

Lift Rig 1,100 Rigging 500 Total Lift 18,337 It is clear, based on the heaviest rack weight to be lifted, that the heaviest load will be well below the 62 ton rating of the Fuel Building Crane main hoist The temporary hoist to be used to maintain the main hoist hook in a dry condition and lift racks into the pool will be selected to provide an adequate load capacity and comply with NUREG-0612.

Remotely engaging lift rigs, meeting all requirements of NUREG-0612, will be used to lift the rack modules. The rack lift rigs consist of four independently loaded traction rods in a lift configuration. The individual lift rods have a safety factor of greater than

10. If one of the rods break, the load will still be supported by at least two rods, which will have a safety factor of more than 5 against ultimate strength. Therefore, the lift rigs comply with the duality feature called for in Section 5.1.6 (3) of NUREG 0612.

Holtec Report HI-2033124 3-7 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

The lift rigs have the following attributes:

• The traction rod is designed to prevent loss of its engagement with the rig in the locked position. Moreover, the locked configuration can be directly verified from above the pool water without the aid of an underwater camera.

• The stress analysis of the rig is carried out and the primary stress limits postulated in ANSI N 14.6 [3.5.5] are met.

• The rigs are load tested with 300% of the maximum weight to be lifted. The test weight is maintained in the air for 10 minutes. All critical weld joints are liquid penetrant examined to establish the soundness of all critical joints.

e. Crane Maintenance The Fuel Building Crane is maintained functional per the Clinton Power Station preventive maintenance procedures.

The proposed heavy load lifts will comply with the guidelines of NUREG-0612, which calls 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 rack installation 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, comprehensive training will be provided to the installation crew.

All training shall be in compliance with ANSI B30.2.

Holtec Report HI-2033124 3-8 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Rigging failure: The lifting device designed for handling and installing the new racks has redundancies in the lift legs and lift eyes such that there are four independent load members in the new rack lift rig.

Failure of any one load bearing member would not lead to uncontrolled lowering of the load. The rig complies with all provisions of ANSI N14.6-1993, including compliance with the primary stress criteria, load testing at 300% of maximum lift load, and dye examination of critical welds.

The rig designs are 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 adequate inspection: The designer of the racks has developed a set of inspection points that have been proven to eliminate any incidence of rework or erroneous installation in numerous prior rerack projects. Surveys and measurements are performed on the storage racks prior to and subsequent to placement into the pool to ensure that the as-built dimensions and installed locations are acceptable.

Measurements of the pool and floor elevations are also performed to determine actual pool configuration and to allow height adjustments of the pedestals prior to rack installation. These inspections minimize rack manipulation during placement into the pool. Preoperational crane testing will verify proper function of crane interlocks prior to rack movement.

Inadequate procedures: Procedures will be developed to address operations pertaining to the rack installation effort, including, but not limited to, mobilization, rack handling, upending, lifting, installation, verticality, alignment, dummy gage testing, site safety, and ALARA compliance. The procedures will be the successors of the procedures successfully implemented in previous projects.

Table 3.5.1 provides a synopsis of the requirements delineated in NUREG-06 12, and its intended compliance.

Holtec Report HI-2033 124 3-9 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

3.6 References

I I . 11:

, , ': :, , .I .i

, . t .

I I. Z, , ,

0

. A_.. ..... _.fi_.__. ........................................ av_*_Urow - :

[3.3.7] "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Holtec International Report HI-2022871 Regarding Use of Metamic in Fuel Pool Applications,"

Facility Operating License Nos. DPR-51 and NPF-6, Entergy Operations, Inc., Docket No. 50-313 and 50-368, USNRC, June 2003.

[3.3.8] USNRC Docket No. 72-1004, NRC's Safety Evaluation Report on NUHOMS 61BT (2002).

[3.3.9] "METAMIC 6061 + 40% boron Carbide Metal Matrix Composite Test Program for NAC International, Inc.", California Consolidated Technology, Inc. (2001).

[3.3.10] "METAMIC" Qualification Program for Nuclear Fuel Storage Applications, Final Test Results", Report NET 152-03, Prepared for Reynolds Metal Company, Inc. by Northeast Technology Corporation.

[3.5.1] Clinton Power Station, Updated Safety Analysis Report (USAR), Revision 7.

Holtec Report HI-2033124 3-10 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

[3.5.2] ANSI/ASME B30.2, "Overhead and Gantry Cranes, (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist)," American Society of Mechanical Engineers, 1976.

[3.5.3] ANSI B30.9, "Safety Standards for Slings," 1971.

[3.5.4] CMMA Specification 70, "Electrical Overhead Traveling Cranes," Crane Manufacturers Association of America, Inc., 2000.

[3.5.5] ANSI N14.6-1993, Standard for Special Lifting Devices for Shipping Containers Weighing 10000 Pounds or more for Nuclear Materials," American National Standard Institute, Inc.,

1978.

[3.5.6] ANSI/ASME B30.20, "Below-the-Hook Lifting Devices," American Society of Mechanical Engineers, 1993.

Holtec Report HI-2033124 3-11 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 3.5.1 HEAVY LOAD HANDLING COMPLIANCE MATRIX (NUREG-0612)

Criterion Compliance

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

2. Will procedures be developed to cover: Yes 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 and Yes qualified?

4. Will special lifting devices meet the Yes guidelines of ANSI 14.6-1993?

5. Will non-custom lifting devices be Yes installed and used in accordance with ANSI B30.20 [3.5.6], latest edition?

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

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

Holtec Report HI-2033 124 3-12 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.0 CRITICALITY SAFETY EVALUATION

4.1 INTRODUCTION

4.1.1 Purpose The purpose of the present study is to document the analyses supporting the criticality safety of the Holtec spent fuel storage racks in the Clinton Station. The Holtec high-density spent fuel storage racks for the Clinton Nuclear Power Station are designed to assure that the neutron multiplication factor (kfir) is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with (clean) unborated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations and in mechanical tolerances, statistically combined, giving assurance that the true kfr will be less than 0.95 with a 95% probability at a 95% confidence level. Reactivity effects of abnormal and accident conditions are also evaluated to assure that under credible abnormal or accident conditions, the reactivity will be maintained less than 0.95.

The rack design philosophy has been described in a paper by Cummings and Turner [4.1.1]. In this philosophy, the racks are designed to assure a kdf less than 0.95 for fuel of a specified k., in the standard cold core geometry* (SCCG) for a given fuel assembly average enrichment and configuration, without consideration of the gadolinia burnable poison normally used in BWR reactors. The specified assembly configuration, average enrichment and maximum kc. (SCCG) define a discharge fuel bumup above which any fuel assembly can be safely stored in the racks.

Once the rack design is completed for the specified design basis, credit may be taken for the gadolinia in the assembly and a minimum Gd 2O3 loading and number of fuel rods with Gd 2 O3 defined that will protect against the storage of any fuel assembly with burnup less than the design basis limit. There is also a limiting enrichment below which any fuel assembly may be safely stored without consideration of the bumup or Gd 2 O3 loading.

• The kin in the standard cold core geometry is defined as the infinite multiplication factor (kin) in a 6-inch core lattice at 20'C without void or control rods.

4-I 1-loltec Project No. 1342 Holtec Report 1-11-2033124 Report Hl-2033124 4-1 Holtec Project No. 1342

The rack design and criticality analyses utilize Metamic as the neutron absorber material.

The fuel used as the design basis for the racks is a GE-14 (IOx10) assembly with an (uniform) initial enrichment of 4.8 % U-235. All fuel assembly types to be stored in the high-density racks were explicitly analyzed to confirm their acceptability for storage.

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

• Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criterion 62, "Prevention of Criticality in Fuel Storage and Handling".

• USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage.

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

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

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

• L. Kopp, "Guidance On The Regulatory Requirements For Criticality Analysis Of Fuel Storage At Light-Water Reactor Power Plants", USNRC Internal Memorandum, L. Kopp to Timothy Collins, August 19, 1998.

Holtec Report HI-2033124 4-2 Holtec Project No. 1342

4.1.2 Design Criteria and Assumptions To assure the true reactivity will always be less than the calculated reactivity, the following conservative design criteria and assumptions were made.

• The racks are assumed to contain the most reactive fuel authorized to be stored in the facility without any control rods or uncontained burnable poison.

• Moderator is pure, unborated water at a temperature (40 C) within the design basis range corresponding to the highest reactivity.

• Criticality safety analyses are based upon an infinite radial array of storage cells. No credit is taken for radial neutron leakage, except in the assessment of certain abnormal/accident conditions where neutron leakage is inherent. A finite axial length of the design basis fuel assembly is used, with an effectively infinite (30 cm) water reflector top and bottom.*

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

• An average void content (40 %) in the reactor core was assumed in the depletion calculation.

• No credit is taken for axial blankets of natural or depleted U0 2 .

• The neutron absorber material, Metamic, is an alloy of aluminum encapsulating boron-carbide.

• The 30 cm axial reflectors effectively bounds any water and stainless steel (and fittings) that would exist.

Holtcc Report Hl-2033124 4-3 Holtcc Project No. 1342

4.2

SUMMARY

AND CONCLUSIONS The fuel assembly used as the design basis for the racks is a standard GE IOx1O array (GE-14) of BWR fuel rods containing U0 2 clad in Zircaloy, and assumes uniform initial enrichments of 4.8 wt% U-235. Explicit analyses of all other fuel assembly types (see Table 4-1) were performed to confirm their acceptability for storage in the high-density racks. The effects of calculational and manufacturing tolerances were evaluated and added in determining the maximum keff in the storage rack.

In BWR fuel, there are a wide variety of designs, including enrichment distribution and gadolinia loading, which often vary in the axial direction. Three different criteria, which bound fuel acceptable for safe storage, are defined as follows.

1. Any fuel with an average enrichment less than or equal to 3.3% U-235, independent of burnup or the gadolinia normally used in BWR fuel, may be safely stored with assurance that the maximum reactivity in the storage rack will be less than the regulatory limit.

2. For fuel assemblies with an average initial enrichment above 3.3% and up to 4.8%, a minimum discharge burnup of 12 MWD/KgU is acceptable. Any fuel equal to or greater than 12 MWD/KgU may be safely stored with assurance that the maximum reactivity in the storage rack will be less than the regulatory limit.

3. A maximum nominal enrichment of 4.8 wt% U-235, with a maximum planar kcnr in the standard cold core geometry (SCCG) of 1.33, where the SCCG is defined as the multiplication factor (Lk) for an infinite array of fuel assemblies on a 6-inch lattice spacing, at 20 'C without voids or control rods. A minimum of 3.0 wt% Gd2 O3 in at least 6 fuel rods is required to assure a k0. (SCCG) less than 1.33 or to assure the criticality safety in storage of fuel with less than 12 MWD/KgU burnup.

Any one of the three criteria is sufficient to determine the acceptability of fuel for safe storage in the spent fuel racks. Based on these criteria, all of the fuel assemblies currently at the Clinton Station are acceptable for storage in the spent fuel racks with assurance that the maximum krff is below the regulatory limit.

Holtc RportHI-03324 -4 olte PrjectNo.134 Holtec Report HI-2033124 44 Mice Project No. 1342

For conservatism, these criteria should be applied to the axial (planar) region of highest reactivity. Each planar region should be separately evaluated to assure that the planar region of highest reactivity is assessed. If fuel vendor calculations are used to determine the ki (SCCG) of the spent fuel, it is suggested that the vendor value be increased by 0.01 Ak to compensate for any difference in vendor calculations and those reported herein.

The basic calculations supporting the criticality safety of the Clinton Station fuel storage racks are summarized in Table 4-2 and Table 4-3.

Abnormal and accident conditions were also evaluated. None of the abnormal or accident conditions that have been identified as credible will result in exceeding the limiting reactivity (kf' of 0.95). The effects on reactivity of credible abnormal and accident conditions are summarized in Table 4-4. The double contingency principle of ANSI N16.1-1975 (and in the principal USNRC guidelines) specifies that it shall require at least two unlikely independent and concurrent events to produce a criticality accident. This principle precludes consideration of the simultaneous occurrence of multiple accident conditions. Other hypothetical events were considered and no credible occurrences or configurations have been identified that might have any adverse effect on the storage rack criticality safety.

Holtec Report HI-2033124 4-5 Holtec Project No. 1342

4.3 INPUT PARAMETERS 4.3.1 Fuel Assembly Specifications The design basis fuel assembly is a standard GE-14 assembly with a l~xl0 array of BWR fuel rods containing U0 2 clad in Zircaloy (see Table 4-1). The GE-14 fuel exhibits the highest reactivity compared to all other types of fuel design used at the Clinton Power Station.

The other designs evaluated (see Table 4-1), are listed below:

A GE-6 design 8x8 assembly with 62 fuel rods and 2 water rods, A GE-7 design 8x8 assembly with 62 fuel rods and 2 water rods, A GE-8 design 8x8 assembly with 62 fuel rods and 2 water rods, A GE-10 design 8x8 assembly with 60 fuel rods and 1 large water rod.

4.3.2 Storage Rack Cell Specifications The high-density storage rack cells consist of an egg-crate structure, as is illustrated in Figure 4-1, with fixed neutron absorber material positioned between the fuel assembly storage cells in a Dinch channel. The neutron absorber material is Metamic, with an areal density of

= areal density. This arrangement provides a nominal center-to-center lattice spacing of 6.243 = inches. Manufacturing tolerances used in evaluating uncertainties in reactivity are indicated in Table 4-5. The 0.075-in. stainless-steel box that defines the fuel assembly storage cell has a nominal inside dimension of 6.05 in. This allows adequate clearance for inserting/removing the fuel assemblies, with or without the Zircaloy flow channel. The neutron absorber panels are 152 inches long, 4.75 inches wide, and = inches thick. Neutron absorber panels are not needed or used on the exterior walls of modules facing non-fueled regions, i.e., the pool walls where there is insufficient room to mis-load a fuel assembly. Similarly, neutron absorber panels are used on only one exterior surface of the modules that face each other across the small water gap between the modules.

Holtec Report HI-2033124 4-6 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.4 ANALYTICAL METHODOLOGY 4.4.1 Computer Codes and Benchmarking In the fuel rack evaluation, criticality analyses of the high-density spent fuel storage racks were performed with the MCNP4a code [4.4.1] (a continuous energy Monte Carlo code developed by the Los Alamos National Laboratory). Independent calculations were made with the CASMO4 code [4.4.2], a two-dimensional multi-group transport theory code.

Benchmark calculations are presented in Appendix 4A and indicate a bias of 0.0009 +/- 0.0011 for MCNP4a evaluated at the 95% probability, 95% confidence level [4.4.3][4.4.4]. These benchmark calculations have previously been reviewed and accepted by the USNRC in numerous licensing applications. In the geometric model used in both CASMO4 and MCNP4a calculations, each fuel rod and its cladding were explicitly described and reflecting boundary conditions (zero neutron current) were used at the effective centerline of the neutron absorber and steel plate between storage cells. These boundary conditions have the effect of conservatively creating an infinite array of storage cells in the radial direction. MCNP4a is a 3-dimensional code and the calculations assumed a 30 cm water reflector (effectively infinite) on the top and bottom of the assemblies.

The CASMO4 computer code was used as a means of evaluating small reactivity increments associated with manufacturing tolerances. Depletion calculations were also performed with CASMO4, using the restart option to describe spent fuel in the storage cell.

MCNP4a was used as the primary method of criticality analysis and to assess the reactivity consequences of eccentric fuel positioning and abnormal locations of fuel assemblies.

In the CASMO4a depletion calculations, the assumed fuel and moderator temperatures were 832 K (1038 0 F) and 560 K (5490 F) respectively. A typical average of 40% moderator void is usually assumed in BWR core during depletion calculations. To investigate the sensitivity of the Holtec Report HI-2033124 4-7 Holtec: Project No. 1342

calculations to void content, an average core void of 45% was also calculated. Results showed an increase in reactivity of 0.0006 Ak, which illustrates the insensitivity to core void.

CASMO4 depletion calculations include three nuclides (Pm-148m and two lumped fission products, 401 and 402) that are not included in the MCNP library. These nuclides amount to 0.0050 Ak at 12 MWD/KgU and are compensated by an equivalent B-10 concentration. A check calculation confirmed the accuracy of this calculational approach.

4.4.2 CASMO4 Validation The CASMO4 calculations were validated against MCNP4a for the specific fuel assemblies and geometries involved. Comparison calculations are listed in Table 4-6 and Table 4-7. (MCNP4a results are bias corrected and listed at the 95%/95% level)*. These data confirm the CASMO4 calculations, and show that MCNP4a calculations tend to be more conservative than the corresponding CASMO4 calculations.

4.4.3 Calculation of the k- in the SCCG The k. in the SCCG was calculated (CASMO4) as a function of fuel burnup, evaluated in the storage rack reference design at 4.8 % initial enrichment. Results are shown in Figure 4-2. At the design basis limit of a k0. in the SCCG of 1.33, the limiting reactivity in the storage rack at 40 C is 0.9267 (CASMO4) or 0.9285+0.0012 (MCNP4a, uncertainties not included). This is the fundamental design basis of the storage rack. The uncertainties are evaluated separately and the maximum rack keff is shown in Table 4-3.

Figure 4-4 shows a similar correlation for all of the fuel types considered (calculated without gadolinia) and confirms that the GE-14 assembly type at 4.8% enrichment bounds all other fuel types considered at a kinr (SCCG) of 1.33. The values shown in Figure 4-4 were also evaluated at K-factor for one-sided tolerance at 95%/ 95% from NBS Handbook 91 [4] is 1.70.

4-8 Holtec Project No. 1342 1-loltec Report Holtec 1-11-2033124 Report HI-2033124 4-8 Holtec Project No. 1342

4.8% enrichment although, in practice, the other assemblies would usually have a significantly lower enrichment.

4.4.4 Gadolinia Effects and Burnup Once the storage rack design is completed (based on the k0. (SCCG)) and the minimum burnup is determined, credit for gadolinia is taken to assure that the design basis reactivity is not exceeded at bumups below the design limit (12 MWD/KgU). Gadolinia (Gd 2O 3 ) is used in almost all BWR fuel designs as a means of augmenting reactivity control in core operations. Gadolinium has a higher cross-section than U-235 and the reactivity of an assembly generally increases with burnup, reaching a maximum at some point in burnup where the gadolinium is virtually depleted.

For fuel of 4.8% average enrichment, Figure 4-3 illustrates the reactivity variation with burnup for several illustrative gadolinia loadings with GE-14 fuel of 4.8% enrichment, evaluated in the spent fuel storage rack. Also shown in Figure 4-3 is the reference calculation without any Gd2 O3 ,

which define the design basis burnup.

From the data shown in Figure 4-3, it is concluded that all of the cases considered are bounded by the reference design basis bumup curve calculated without any gadolinia. The limiting (minimum) design burnup is 12 MWD/KgU and credit for the gadolinia actually present is taken with fuel of less than 12 MWD/KgU burnup to assure that the regulatory limit on the maximum kdf (0.95) is not exceeded.

4.4.5 Gadolinia Rod Locations A number of alternative locations of the gadolinia rods were calculated to evaluate the potential effect of rod locations. As part of this investigation, corresponding calculations were made with MCNP4a at zero burmup to compensate for any discrepancy that might exist in the CASM04 estimates of self-shielding in the strongly absorbing gadolinia-bearing fuel rods. Results of these analyses are shown in Table 4-7 and indicate that the MCNP4a calculations are slightly more conservative (higher reactivity) than the corresponding CASM04 calculations although the difference is not large. The calculations that are most important are the calculations with 6 Holtec Report HI-2033124 4-9 Holtec Project No. 1342

gadolinia rods per assembly since 8 Gd 2 O3 rods in an assembly are well below the reference reactivity. A larger number of Gd 2O3 rods or higher loading of Gd 2 03 would be even more conservative.

In general, the MCNP4a calculations tend to be slightly more conservative than the corresponding CASM04 calculations at zero burnup. The design basis MCNP4a calculation gave a calculated keff of 0.9285 +/- 0.0012 (see Table 4-6). For 2 of the cases in Table 4-7, the MCNP4a calculated kff is higher than that of the reference design. These two cases are the 3.0%

and 3.5% Gd 2 O3 loadings (cases 9 and 10 in Table 4-7) where the Gd 2O3 rods are in close proximity to each other. These two cases suggest the possibility of a small bias in the calculations where Gd 2O3 rods are close in proximity and mutually shield each other. Cases 9 andlO represent the largest difference between the MCNP4a calculations at zero bumup and the design basis reactivity, amounting to a bias of +0.003OAk at 3% Gd 2 O3 and 0.0007Ak at 3.5%

Gd 2O3 . Although the bias values would apply only to 3.0% and 3.5% Gd 2O3 in a unique arrangement, for conservatism, the largest bias (+0.0030) is included in the evaluation of the maximum rack reactivity in Table 4-3.

With this correction bias, the data in Table 4-3 and Table 4-7 indicates that for the GE14 fuel at 4.8 % enrichment, a minimum Gd 2 03 loading of 3 % in 6 rods is required to protect against the possible loading of fuel with less than 12 MWD/KgU burnup.

4-10 1-loltec Project No. 1342 Holtec Report HI-2033 124 Report HI-2033124 4-10 HoItcc Project No. 1342

4.5 CRITICALITY ANALYSES AND TOLERANCE VARIATIONS 4.5.1 Nominal Design Case 4.5.1.1 Enrichment Limit Criterion One criterion for acceptable storage is an enrichment of 3.3 % (or less), which would not require any credit for gadolinia or burnup. Results of calculations with 3.3% enrichment are summarized in Table 4-2 and confirm that the maximum reactivity is below the regulatory limit. The criterion of an enrichment of 3.3% is therefore acceptable, without any consideration of the Gadolinia normally present in the fuel, or the discharge fuel burnup.

4.5.1.2 Maximum kinrin the Standard Cold Core Geometry It is conventional practice for the fuel vendor, in developing a specific assembly design, to provide values for kinr in the SCCG for each planar (axial) region of significantly different composition or arrangements. These kinf values are usually provided at 0% void (core inlet),

40% void (core average), and 70% void (exit condition). The 40% core average is the most meaningful since the 0% and 70% void cases would apply only to the ends of the assemblies (small volume and high neutron leakage).

The initial design Gd 2 O3 loading enters into the fuel vendor's calculations of the bumup at which the peak reactivity occurs. At this burnup, the gadolinium is essentially depleted. Consequently, calculations of the reactivity in the storage rack do not need to include gadolinium, but only the average enrichment, burnup and fuel design. Calculations are provided illustrating this fact and correlating the reactivity in the storage rack to the kinf in the SCCG. Figure 4-2 illustrates the variation in reactivity of the storage rack with values of the kinf in the SCCG. The acceptance criterion for safe storage of spent fuel is that the kinf in the SCCG must be equal to or less than 1.33 for the planar region of highest reactivity. Thus, the kefT values in the storage rack are determined by the limiting kinf in the SCCG (1.33).

Holtec Report HI-2033124 4-1 1 Holtec Project No. 1342

The design basis fuel assembly is the GE-14. For each of the fuel assembly types considered (listed in Table 4-1), the kif in the rack for a kinr in the SCCG were calculated and the results are summarized in Figure 4-4. These data confirm that all other fuel assembly types are less reactive and therefore bounded by the GE-14 fuel type, and at an enrichment of 4.8 %. In fact, the other fuel types have average enrichment significantly less than 4.8 % and would therefore exhibit even lower reactivity.

The maximum kerf value, including calculational and manufacturing uncertainties, is listed in Table 4-3. These calculations confirm that, for fuel of 4.8% enrichment or less, the criteria of a maximum kinf of 1.33 in the SCCG will assure compliance with the regulatory guidelines (kefr limit of 0.95).

4.5.1.3 Criteria for Minimum Gadolinia Loading Gadolinia (Gd 2 O3 ) is normally used in BWR fuel to augment reactivity control during in-core operation. A very wide variety of Gd 2 O3 loadings are commonly used - often differing in axial (planar) regions. Furthermore, the Gd2 O3 loadings for fuel of 4.8% average enrichment have not yet been developed. However, it is possible to define criteria for the minimum Gd 2O3 loadings required to assure that the highest reactivity over burnup is always less than the regulatory limit.

Results of the analysis of a series of cases with various Gd 2O3 loadings were described earlier (Section 4.4.4 and 4.4.5) and summarized in Figure 4-3, for fuel of 4.8 % initial average enrichment. For fuel of 4.8 % enrichment, a burnup of at least 12 MWD/KgU is required and a bumup lower than 12 MWD/KgU will require credit for Gd 2O3 .

Figure 4-3 shows the effect of Gd 2 O3 concentration for 6 and 8 Gd 2O3 rods per assembly and indicates that a minimum of 6 Gd 2 O3 rods of at least 3% loading is required. The 6-rod case bounds cases of 8 or more Gd 2O3 rods per assembly. At enrichments less than 4.8%, the peak reactivity will be even lower, and the criteria of 3 % Gd 2 03 in at least 6 rods will be even more conservative. Thus, the criterion of a minimum Gadolinia loading (including tolerance on Gd 2 O3 content), evaluated at the axial planar region of the highest reactivity, will assure safe storage of Holtc RportHI-03324 -12 oltc Prjec No.134 Holtec Report HI-2033124 4-12 HoItcc Project No. 1342

all fuel assemblies and conformance with the NRC guidelines. For fuel of 3.3 % average enrichment (Table 4-2), no Gd2 O3 or burnup is necessary and any assembly of 3.3 % enrichment or less may be safely stored in the rack. For enrichments less than or equal to 3.3%, the maximum kinf in the rack is less than 0.95 (including uncertainties and allowances) regardless of gadolinia content or the kinr in the SCCG.

4.5.2 Uncertainties Due to Manufacturing Tolerances 4.5.2.1 Rack Manufacturing Tolerances The GE-14 assembly was used for the evaluation of incremental reactivity changes due to manufacturing tolerances. The reactivity effects associated with manufacturing tolerances are listed in Table 4-5 and discussed below. Values at 12 MWD/KgU bumup (the approximate burnup for a kinf of 1.33 in the SCCG) were used, although the reactivity uncertainties are nearly insensitive to burnup.

4.5.2.2 Boron- IOLoading Variation The Metamic absorber panels are nominally ='J-inch thick, with a design B-10 areal density of I=. The manufacturing tolerance limit is l in B-10 content, including both thickness ;T) and composition ( I tolerances. This assures that the minimum boron-l 0 areal density will not be less than Go in any location or any panel.

Differential CASM04 calculations provide the incremental reactivity uncertainty shown in Table 4-5.

Holtec Renort HI-2033124 4-13 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.5.2.3 Neutron Absorber Width Tolerance Variation The reference storage cell design uses a neutron absorber panel width of 4.75 inches. The tolerance on the panel width isl' .- , and the reactivity effect is shown in Table 4-5.

4.5.2.4 Lattice Spacing Variation The design storage cell lattice spacing is 6.243 inches. Decreasing the lattice spacing (by decreasing the box I.D.) increases reactivity. The manufacturing tolerance is [ and the corresponding uncertainties in reactivity are listed in Table 4-5.

4.5.2.5 Stainless Steel Thickness Tolerances The nominal thickness of the stainless steel box is 0.075 inches with i for the steel sheath. The maximum positive reactivity effects of the expected stainless steel thickness tolerance were calculated with CASM04 and are listed in Table 4-5.

4.5.2.6 Zircaloy Flow Channel Elimination of the Zircaloy flow channel results in a small decrease in reactivity (see listing below).

4.5.3 Fuel Tolerances (Enrichment and Density Uncertainties)

CASM04 calculations of the sensitivity to small changes in fuel enrichment and U0 2 density gave uncertainties in reactivity listed below:

Holtec Report H1-2033124 4-14 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Reactivity Effects of Enrichment and Density Tolerances Case Description Tolerance Reactivity effect, Ak 3.3 % Enrichment +0.05 % in Enrichment 0.0035 3.3 % Enrichment 0.20 increase in Fuel Density 0.0024 4.8 % Enrichment @ 12 MWD/KgU +0.05 % in Enrichment 0.0028 4.8 % Enrichment @ 12 MWD/KgU 0.20 increase in Fuel Density 0.0021 4.5.4 Uncertainty in Depletion Calculations Since critical experiment data with spent fuel is not available for determining the uncertainty in depletion calculations; a conservative allowance for uncertainty in reactivity was assigned based upon the Kopp memorandum (reference 5). In the Clinton racks, the reactivity decrement (from fresh to the design basis bumup) in the absence of gadolinium is 0.0854 Ak. It is assumed that the uncertainty in depletion calculations is less than 5% of the total reactivity decrement or an uncertainty +/-0.0043Ak.

4.5.5 Existini Fuel Assemblies at the Clinton Station Fuel assemblies from cycles I through 7 are presently in storage at the Clinton Station. All of these assemblies may be safely stored in the fuel racks in accordance with the acceptance criteria established in this report. Examination of the G.E. fuel design reports reveal that fuels from Cycles 1 through 4 has average enrichments less than 3.3%. In Cycles 5 through 7, the number and loading of the gadolinia rods is well above the minimum required and the fuel assemblies are therefore acceptable for storage.

Holtec Report HI-2033124 4-15 Holtec Project No. 1342

4.6 ABNORMAL AND ACCIDENT CONDITIONS 4.6.1 Temperature and Water Density Effects The moderator temperature coefficient of reactivity is negative. Using the minimum temperature of 4 'C (maximum possible water density) therefore assures that the true reactivity will always be lower than the calculated value regardless of the temperature. Temperature effects on reactivity have been calculated and the results are shown in Table 4-8. Introducing voids in the water in the storage cells (to simulate boiling) decreased reactivity, as shown in the table.

Boiling at the submerged depth of the racks would occur at approximately 120'C. The only significance of these calculations is to confirm that the temperature and void coefficients of reactivity are negative and the reference temperature of 4 'C is conservative.

4.6.2 Abnormal Location of a Fuel Assembly It is hypothetically possible to suspend a fuel assembly of the highest allowable reactivity outside and adjacent to the fuel rack, although such an accident condition is highly unlikely. The exterior walls of the rack modules facing the outside (where such an accident condition might be conceivable) are regions of high neutron leakage, which more than compensates for the extra fuel assembly. For comparison to the reference kef, calculations were performed (MCNP4a) for this condition with an extraneous fuel assembly present. The calculations were performed with the GE-14 fuel assembly. Conservatively, no neutron absorber panels were assumed to be present on the exterior rack wall. With the inherent neutron leakage included, the kdf with an extraneous fuel assembly of the maximum reactivity located outside and adjacent to the fuel rack, is less than the reference kff. Thus it is concluded that the abnormal location of a fuel assembly will have a negligible reactivity effect (see listing below).

Holtec Report H1-2033124 4-16 Holtcc Project No. 1342

Description Calculated kff (Bias and Tolerance not included)

Normal Condition - explicit model 0.9258 without extraneous fuel assembly Accident Condition - with extraneous 0.9229 assembly 4.6.3 Eccentric Fuel Assemblv Positioning The fuel assembly is normally located in the center of the storage rack cell with bottom fittings and spacers that mechanically restrict lateral movement of the fuel assemblies. Nevertheless, calculations were performed with the fuel assembly moved into the corner of the storage rack cell (four-assembly cluster at closest approach) to investigate the effect on reactivity. Two cases were considered: (1) fuel of 3.3% enrichment and (2) fuel of 4.8% burned to 12 MWD/KgU was used. These (MCNP4a) calculations (see below) resulted in a small negative reactivity effect.

Thus, the nominal case, with the fuel assembly positioned in the center of the storage cell, yields the higher reactivity.

Fuel Assembly Reference kir Eccentric kinf GE-14 @ 3.3 % Enrichment 0.9266 0.9191 GE-14 @ 4.8 % burned to 12 MWD/KgU 0.9258 0.9192 4.6.4 Dronned 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 active fuel region of more than 12 inches, which is sufficient to preclude neutron coupling (i.e., an effectively infinite separation). Maximum expected deformation under seismic or accident conditions will not reduce this spacing to less than 12 inches.

Holtec Report HI-2033124 4-17 Holtec Projcct No. 1342

It is conceivable that a dropped assembly might penetrate a storage cell in a vertical position, impacting and causing local deformation of the base plate. The maximum expected local deformation is about 2 inches, with smaller deformations in the eight immediately adjacent cells.

Conservative calculations, using a bounding deviation of 2.5 inches of exposed fuel in all cells everywhere, showed that the krf of the rack was increased only slightly and the maximum keff would remain less than 0.95.

Description Calculated keff (bias and uncertainties not added)

Normal Condition 0.9258 +/- 0.0002 Accident Condition - with 2.5 inches 0.9259 +/- 0.0002 of exposed fuel These values are within the expected statistical variation of MCNP4a calculations and confirm that a dropped fuel assembly will not have an adverse effect on reactivity and the maximum kff will remain below the regulatory limit.

4.6.5 Fuel Rack Lateral Movement Neutron absorber panels are installed in the rack wall along one side of the water gap between adjacent racks. With this configuration, the maximum reactivity of the storage rack is not depen-dent upon the water-gap spacing between modules. Thus, misalignment of the racks or seismically induced movement will not affect the maximum reactivity of the rack.

4-18 Holtec Project No. 1342 Holtcc 1-H-2033 124 Report Hl-2033124 1-foltec Report 4-18 Holtec Project No. 1342

4.7 REFERENCES

[4.1.1] K. Cummings and S. E. Turner, "Design of Wet Storage racks for Spent BWR Fuel", Proceedings of the Topical Meeting on Practical Implementation of Nuclear Criticality Safety, November 2001, ANS Meeting - Reno, Nevada.

[4.4.1] J.F. Briesmeister, Ed., "MCNP- A General Purpose Monte Carlo N-Particle Transport Code, Version 4A", Los Alamos National Laboratory, LA-12625-M (1993).

[4.4.2] A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis," ANS Transactions, Vol. 26, p. 604, 1977.

M. Edenius, et al., "CASMO4, A Fuel Assembly Burnup Program, Users Manual", Studsvik Report SOA/95/1, Studsvik Report (proprietary).

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

[4.4.4] L.I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Plants", USNRC memorandum, Kopp to Collins, August 1998.

Holtec Report H--2033124 4-19 Ho1tcc Project No. 1342

Table 4-1 FUEL ASSEMBLY DESIGN SPECIFICATIONS USED IN THE ANALYSES FUEL ROD DATA GE-6 GE-7 GE-S GE-lO GE-14 Cladding outside diameter, in. 4 tr F7 iL Cladding inside diameter, in.

1 JJ Li Cladding material 4 at, Pellet diameter, in.

.. ... r Enrichment (design basis) Aa4 U0 2 density (stack), g/cc U0 2 $ift ~~iS;afX I...

aa L

WATER ROD 1{ a a aIJ DATA 4 1 l, --; - - -I' I!

'a Number of water [ '..  ;

rods II Inside diameter, inch Outside diameter Material F.,.. '

FUEL ASSEMBLY a A a a ;I.;j.

DATA a..'

I Fuel rod array .1 Number of fuel rods C:.1 Fuel rod pitch, inch [17 Fuel channel, material Inside dimension, inch A

• .1 Channel Thickness, inch rL____

Holtec Report H1-2033124 4-20 1342 The Shaded Areas are Proprietary to GE Company

Table 4-2

SUMMARY

OF CRITICALITY SAFETY ANALYSES FOR FUEL OF 3.3 % ENRICHMENT W/OUT Gd 2O3 OR BURNUP Reference Fuel Type GE-14 Temperature assumed for analysis 20 0 C Fuel Enrichment (average) 3.3%

Gd 2O3 loading % N/A MCNP4a Calculated kff 0.9266 Calculational bias, Ak 0.0009 Temperature Correction to 4 'C, Ak 0.0025 Allowance for Gd 2O3 rod location N/A Reference kff 0.9300 Uncertainties Removal of flow channel negative Eccentric assembly location negative Uncertainty in bias +/-0.0011 Tolerances (Table 4-5) +/- 0.0073 Uncertainty in Depletion calculations N/A MCNP Statistics +/- 0.0007 Total Uncertainties +/- 0.0074 Maximum Reactivity 0.9374 Regulatory Limit 0.95 4-1 Holtc Rpor 1-1-20312 HltecProectNo.134 Holtec Report HI-2033124 4-21 Holtec Project No. 1342

Table 4-3

SUMMARY

OF CRITICALITY SAFETY ANALYSES AT DESIGN BASIS ENRICHMENT AND BURNUP Reference Fuel Type GE-14 Temperature assumed for analysis 20 0 C Initial Fuel Enrichment (average) 4.8 Maximum kinf in SCCG 1.33 Design basis Bumup, MWD/KgU 12 Calculated kf 0.9258 Calculated bias 0.0009 Temperature correction to 4 'C 0.0018 Allowance for Gd 2O3 rod location 0.0030 Reference kdf 0.9315 UNCERTAINTIES Removal of flow channel negative Eccentric assembly location negligible Uncertainty in bias +/-0.0011 Tolerances (Table 4-5) +/- 0.0069 Uncertainty in Depletion calculations +/- 0.0043 MCNP Statistics +/- 0.0003 Total Uncertainty +/-0.0082 MAXIMUM REACTIVITY 0.9397 **

Regulatory Limit 0.95

• • For an average core void of 45%, the maximum kyr would increase to 0.9403.

Holtec Report HI-2033124 4-22 Holtec Project No. 1342

Table 4-4 REACTIVITY EFFECTS OF ABNORMAL AND ACCIDENT CONDITIONS Accident/Abnormal Condition Reactivity Effect Temperature increase Negative (Table 4-8)

Void (boiling) Negative (Table 4-8)

Assembly dropped on top of rack Negligible Misplacement of a fuel assembly Negligible Seismic Movement Negligible Holtec Report 11-2033124 4-23 Holtec Project No. 1342

Table 4-5 REACTIVITY UNCERTAINTIES DUE TO MANUFACTURING TOLERANCES Reactivity Effects of Manufacturing Tolerances

(,3.3%E (~i0 4.8 % E 0, Ouantitv Nominal Value Tolerance MWD/KgU 12 MWD/KgU CASMO Case Number B-10 Loading - Metamic +/-0.0046 +/-0.0045 EX Metamic Width 4.75 inches +/- 0.0025 +/-0.0024 Lattice spacing (Box 6.243 inches +/-0.0024 +/-0.0028 I.D.)

SS thickness 0.075 /77 +/-10% +/-0.0008 +/-0.0008 Fuel enrichment +/-0.0035 +/-0.0028 Fuel density +/-0.0024 +/-0.0021 Statistical combination of tolerance

+/-0.0073 +/-0.0069 uncertainties Holtec Report HI-2033124 4-24 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4-6 COMPARISON OF CASMO4 AND MCNP4a CALCULATIONS MCNP4a l CASMO4 I

C..ase Diescription 1.) Uniform 3.3 % 0.9292 0.9326+/-0.0012 (1)

Enrichment, no Gadolinia 2.) 4.8 % E Fuel at 12 I

MWD/KgU, no Gadolinia 0.9260 0.9287+/-0.0012 (2),(3)

I (1) With bias (0.0009) and temperature correction (0.0025Ak) to 40 C.

(2) With bias (0.0009) and temperature correction (0.0018Ak) to 40 C.

(3) 0.9285 +/- 0.0012 with axial leakage (Design Basis)

Holtec Report HI-2033124 4-25 Holtec Project No. 1342

Table 4-7 Reactivity Effects of Gd 2O3 Rod Locations MCNP4a (1) CASMO4 Case Description k.i k~.

Nominal GE Locations

1) 6 GD)Rods of 3% Gd2 O3 0.9235+/-0.0013 0.9230

2) 6 GD Rods of 3.5% Gd2 O3 0.9211+/-0.0013 0.9211

3) 6 GD Rods of 4% Gd 2O 3 0.9199+0.0013 0.9194

4) 8 GD Rods of 3% Gd2 O3 0.8917+/-0.0013 0.8958

5) 8 GD Rods of 3.5% Gd2 O3 0.8887+/-0.0013 0.8933

6) 8 GD Rods of 4% Gd 2O 3 0.8865+/-0.0013 0.8911 Alternate Configurations

7) 6 GD Rods of 3.5% Gd 2 O3 0.9247+/-0.0013 0.9234

8) 6 GD Rods of 3.5% Gd 2O3 0.9263+/-0.0013 0.9249

9) 6 GD Rods of 3.5% Gd 2 O3 0.9292+/-0.0013 0.9283

10) 8 GD Rods of 3.5% Gd 2 O3 0.9315_0.0013 0.9300 (1) With bias (0.0009) and temperature correction (O.0018Ak) to 40 C Holtec Rcport HI-2033124 4-26 Holtec Projcct No. 1342

Table 4-8 EFFECT OF TEMPERATURE AND VOID ON CALCULATED REACTIVITY OF THE STORAGE RACK Case Incremental Reactivity Change, Ak GE-14 GE-14

() 3.3 % En 0 Bumup 4.8 %E ().12 MWD/KgU 40 C (39 0 F) Reference Reference 200 (68 0 F) -0.0025 -0.0018 400 C (122 0 F) -0.0062 -0.0047 80 0 C (1760 F) -0.0146 -0.0116 1000 C (2121F) -0.0196 -0.0157 1200 C (2480 F) -0.0249 -0.0200 120'C + 20% void -0.0450 -0.0405 Holtec Report HI-2033124 4-27 Holtec Project No. 1342

Neutron Absorber 4.75 Width (GE-14 Assembly Illustrated)

Figure 4-1 Storage Cell Calculational Model 4-28 1 81A' 1 51I Holtec Report HI-2033124 SHADED AREAS DENOTE PROPRIETARY INFORMATION

0.940 0.935 0.930 0 Design Bas s 0 0.925

.4-C 0.920 T

0.915 0.910 0.905 rT-r-r-r-=r nr ......m .... rr 1.30 1.31 1.32 1.33 1.34 1.35 k-Infinite in the SCCG Fig. 4-2 Correlation Between k-Infinite in the SCCG and k-Infinite In the Storage Rack (Note: Bias and Uncertainties not Included) 4-29 Holtec Project No. 1342 111.2033124 Holtec Report HI-2033124 Holtcc 4-29 HoRce Project No. 1342

I 0.94 - _ ___ ___ _£ _ _

0.93CASMO Limit 0.92 a) o - \sk6Gd3.!5 0.91 6CI 0.89 l 1300 ppm) show a tendency to slightly overpredict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the racks with higher soluble boron concentrations could be slightly conservative.

Parallel experiments with a depleted uranium reflector were also performed but not included in the present analysis since they are not pertinent to the Holtec rack design.

Appendix 4A, Page 5

4A;5 MOX Fuel The number of critical experiments with PuO2 bearing fuel (MOX) is more limited than for U0 2 fuel. However, a number of MOX critical experiments have been analyzed and the results are shown in Table 4A.7. Results of these analyses are generally above a kff of 1.00, indicating that when Pu is present, both MCNP4a and KEN05a overpredict the reactivity. This may indicate that calculation for MOX fuel will be expected to be conservative, especially with MCNP4a. It may be noted that for the larger lattice spacings, the KEN05a calculated reactivities are below 1.00, suggesting that a small trend may exist with KENO5a. It is also possible that the overprediction in kff for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and Am-241 growth. This possibility is supported by the consistency in calculated k. over a wide range of the spectral index (energy of the average lethargy causing fission).

Appendix 4A, Page 6

4A.6 Refererinm

[4A. 1] J.F. Briesmeister, Ed., "MCNP4a - A General Monte Carlo N-Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1993).

[4A.2] SCALE 4.3, "A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", NUREG-0200 (ORNL-NUREG-CSD-21U21R5, Revision 5, Oak Ridge National Laboratory, September 1995.

[4A.3] M.D. DeHart and S.M. Bowman, "Validation of the SCALE Broad Structure 44-G Group ENDF/B-Y Cross-Section Library for Use in Criticality Safety Analyses", NUREG/CR-6102 (ORNL/TM-12460)

Oak Ridge National Laboratory, September 1994.

[4A.4] W.C. Jordan et al., "Validation of KENOV.a", CSD/TM-238, Martin Marietta. Energy Systems, Inc., Oak Ridge National Laboratory, December 1986.

[4A.5] O.W. Hermann et al., "Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analysis", ORNL-TM-12667, Oak Ridge National Laboratory, undated.

[4A.6] R.J. Larsen and M.L. Marx, An Introduction to Mathematical Statistics and its Applications, Prentice-Hall, 1986.

[4A.7] M.N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, Babcock and Wilcox Company, July 1979.

[4A.8] G.S. Hoovier et al., Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins, BAW-1645-4, Babcock & Wilcox Company, November 1991.

[4A.9] L.W. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-18 10, Babcock and Wilcox Company, April 1984.

Appendix 4A, Page 7

[4A. 10] J.C. Manaranche et al., "Dissolution and Storage Experimental Program with 4.75 w/o Enriched Uranium-Oxide Rods, " Trans.

Am. Nucl. Soc. 33: 362-364 (1979).

[4A. 1] S.R. Bierman and E.D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o "5U Enriched U0 2 Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981.

[4A. 12] S.R. Bierman et al., Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o ` 5U Enriched Ut 2 Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December, 1981.

[4A. 13] S.R. Bierman et al., Critical Separation Between Subcritical Clusters of 4.31 w/o 2`U Enriched U0 2 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977.

[4A. 14] S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990.

[4A.15] B.M. Durst et al., Critical Experiments with 4.31 wt % 235U Enriched U0 2 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982.

[4A. 16] S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1981.

[4A. 17] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium Core, WCAP-3385-54, Westinghouse Electric Corp., Atomic Power Division, December 1965.

[4A. 18] M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

Appendix 4A, Page 8

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k ff. EALF' (eV)

Reference Identification Enrich. MCNP4a KENOMa MCNP4a KEN05a 1 B&W-1484 (4A.7) Core I 2.46 0.9964 +/- 0.0010 0.9898+/- 0.0006 0.1759 0.1753 2 B&W-1484 (4A.7) Core II 2.46 1.0008 +/- 0.0011 1.0015 +/- 0.0005 0.2553 0.2446 3 B&W-1484 (4A.7) Core Hi 2.46 1.0010 +/- 0.0012 1.0005 i 0.0005 0.1999 0.1939 4 B&W-1484 (4A.7) Core IX 2.46 0.9956 +/- 0.0012 0.9901 +/- 0.0006 0.1422 0.1426 5 B&W-1484 (4A.7) Core X 2.46 0.9980 +/- 0.0014 0.9922 +/- 0.0006 0.1513 0.1499 6 B&W-1484 (4A.7) Core XI 2.46 0.9978 +/- 0.0012 1.0005 i 0.0005 0.2031 0.1947 7 B&W-1484 (4A.7) Core XI 2.46 0.998B +/- 0.0011 0.9978 +/- 0.0006 0.1718 0.1662 8 B&W-1484 (4A. Core XII 2.46 1.0020 +/- 0.0010 0.9952 :1 0.0006 0.1988 0.1965 9 B&W-1484 (4A.7) Core XIV 2.46 0.9953 +/- 0.0011 0.9928 +/- 0.0006 0.2022 0.1986 10 B&W-1484(4A.7) Core XV 2.46 0.9910-+/- 0.0011 0.9909 0.0006 0.209 0.2014 11 D&W-1484 (4A.7) Core XV ' 2.46 0.9935 +/- 0.0010 0.9889 + 0.0006 0.1757 0.1713 12 B&W-1484 (4A.7) Core XVft 2.46 0.9962 +/- 0.0012 0.9942 i 0.0005 0.2083 0.2021 13 B&W-1484 (4A.7) Core XVm 2.46 1.0036 +/- 0.0012 0.9931 i 0.0006 0.1705 0.1708 Appendix 4A, Page 9

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated kl,, EALFt (eV)

Reference Identification Enrich. MCNP4a KENO~a MCNP4a KENOSa B&W-1484 (4A.7) Core XIX 2.46 0.9961 +/-0.0012 0.9971 0.0005 0.2103 0.2011 14 B&W-1484 (4A.7) Core XX 2.46 1.0008 i 0.0011 0.9932 i 0.0006 0.1724 0.1701 15 B&W-1484 (4A.7) Core X 2.46 0.9994 i 0.0010 0.9918 +/- 0.0006 0.1544 0.1536 16 B&W-1645 (4A.8) S-type Fuel, wl886 ppm B 2.46 0.9970 +/-0.0010 0.9924 i 0.0006 1.4475 1.4680 17 18 B&W-1645 (4A.8) S-type Fuel, w174M ppm B 2.46 0.9990 0.0010 0.9913 i0.0006 1.5463 1.5660 19 B&W-1645 (4A.8) SO-type Fuel, w/1156 ppm B 2.46 0.9972

• 0.0009 0.9949 i 0.0005 0.4241 0.4331 B&W-1810 (4A.9) Case1 1337 ppm B 2.46 1.0023 i 0.0010 NC 0.1531 NC 20 B&W-1810 (4A.9) Case 12 1899 ppm B 2.4614.02 1.0060 +/- 0.0009 NC 0.4493 NC 21 French (4A.10) Water Moderator 0 gap 4.75 0.9966 +/- 0.0013 NC 0.2172 NC 22 23 French (4A.10) Water Moderator 2.5 cm gap 4.75 0.9952 +/- 0.0012 NC 0.1778 NC French (4A.10) Water Moderator 5 cm gap 4.75 0.9943 +/- 0.0010 NC 0.1677 NC 24 25 French (4A.10) Water Moderator 10 cm gap 4.75 0.9979 +/- 0.0010 NC 0.736 NC PNI-3602 (4A.11) Steel Reflector, 0 separatIon 2.35 NC 1.0004 i 0.0006 NC 0.1018 26

-~ - - * * -

Appendix 4A, Page 10

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k, .EALF t (eV)

Enrich. MCNP4a KEN05a MCNP4a KEN05a Reference Identification 2.35 0.9980 +/- 0.0009 0.9992 +/- 0.0006 0.1000 0.0909 27 PNIL3602 (4A.11) Steel Reflector, 1.321 cm sepn. .

2.35 0.9968 +/- 0.0009 0.9964 +/- 0.0006 0.0981 0.0975 28 PNL-3602 (4A.11) Steel Reflector, 2.616 cm sepn 2.35 0.9974 +/- 0.0010 0.9980 : 0.0006 0.0976 0.0970 29 PNL-3602 (4A.11) Steel Reflector, 3.912 cm sepn.

2.35 0.9962 +/- 0.0008 0.9939 i 0.0006 0.0973 0.0968 30 PNI-3602 (4A.11) Steel Reflector, Inflalte sepn.

4.306 NC 1.0003 i 0.0007 NC 0.3282 31 PNI,3602 (4A.11) Steel Reflector, 0 cm sepn.

4.306 0.9997 i 0.0010 1.0012 i 0.0007 0.3016 0.3039 32 PNL-3602 (4A.11) Steel Reflector, 1.321 cm sepn.

4.306 0.9994

• 0.0012 0.9974 +/- 0.0007 0.2911 0.2927 33 PNI,3602 (4A.11) Steel Reflector, 2.616 cm sepn.

4.306 0.9969 +/-0.0011 0.9951 t 0.0007 0.2828 0.2860 34 PNL3602 (4A.11) Steel Reflector, 5.405 cm sepn.

4.306 0.9910 +/- 0.0020 0.9947

• 0.0007 0.2851 0.2864 35 PNL-3602 (4A.11) Steel Reflector, Innite sepn.

4.306 0.9941 +/- 0.0011 0.997h i 0.0007 0.3135 0.3150 36 PNLr3602 (4A.11) Steel Reflector, with Boral Sheets 4.306 NC 1.0003 i 0.0007 NC 0.3159 37 PNL-3926 (4A.12) Lead Reflector, 0 cm sepn.

4.306 1.0025 +/- 0.0011 0.9997 i 0.0007 0.3030 0.3044 38 PNL3926 (4A.12) Lead Reflector, 0.55 cm sepn.

4.306 1.0000 +/- 0.0012 0.9985 i 0.0007 0.2883 0.2930 39 PNI,3926 (4A.12) Lead Reflector, 1.956 cm sepn.

.. Appendix 4A, Page 11

Table 4A.1 Summary of Criticality Benchmark Calculations llated k EALF t (eV)

Reference Identification Enrich. MCNP4a KENOSa MCNP4a KENO5a 40 PNL-3926 (4A.12) Lead Reflector, 5.405 cm sepn. 4.306 0.9971 +/-0.0012 0.9946 +/- 0.0007 0.2831 0.2854 41 PNL-2615 (4A.13) Experiment 004/032 - no absorber 4.306 0.9925 + 0.0012 0.9950 +/- 0.0007 0.1155 0.1159 42 PNI,2615 (4A.13) Experiment 030 - Zr plates 4.306 NC 0.9971 +/- 0.0007 NC 0.1154 43 PNL-2615 (4A.13) Experiment 013 - Steel plates 4.306 NC 0.9965 +/- 0.0007 NC 0.1164 44 PNT,2615 (4A.13) Experiment 014 - Steel plates 4.306 NC 0.9972 +/- 0.0007 NC 0.1164 45 PNL-2615 (4A.13) Exp. 009 1.05% Boron-Steel plates 4.306 0.9982 +/- 0.0010 0.9981 +/- 0.0007 0.1172 0.1162 46 PNL-2615 (4A.13) 9Boron-Steel plates Exp. 012 1.62% 4.306 0.9996 +/- 0.0012 0.9982 +/- 0.0007 0.1161 0.1173 47 PNL-2615 (4A.13) Exp. 031 - Boral plates 4.306 0.9994 +/- 0.0012 0.9969 +/- 0.0007 0.1165 0.1171 48 PNM.7167 (4A.14) Experiment 214R - with flux trap 4.306 0.9991 +/- 0.0011 0.9956 +/- 0.0007 0.3722 0.3812 49 PNT,7167 (4A.14) Experiment 214V3 - with flux trap 4.306 0.9969 +/- 0.0011 0.9963 +/- 0.0007 0.3742 0.3826 50 PNL-4267 (4A.15) Case 173 - 0 ppm B 4.306 0.9974 +/- 0.0012 NC 0.2893 NC 51 PNt-4267 (4A.15) Case 177 - 2550 ppm B 4.306 1.0057 +/- 0.0010 NC 0.5509 NC 52 PNL-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 21 20% Pu 1.0041 +/- 0.0011 1.0046 +/- 0.0006 0.9171 0.8868 Appendix 4A, Page 12

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k.f EALF t (eVl Enrich. MCNP4a KENOSa MCNP4a KENOSa Reference Identification 20% Pu 1.0058 +/- 0.0012 1.0036 +/- 0.0006 0.2968 0.2944 53 PNL-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 43 20% Pu 1.0083 +/- 0.0011 0.999 +/- 0.0006 0.1665 0.1706 54 PNL-5.03 (4A.16) MOX Fuel - Type 3.2 Exp. 13 20% Pu 1.0079 +/- 0.0011 0.9966 +/- 0.0006 0.1139 0.1165 55 PNI-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 32 6.6% Pu 0.9996 +/- 0.0011 1.0005 +/- 0.0006 0.8665 0.8417 56 WCAP-3385 (4A.17) Saxton Case 52 PuO2 0.52" pitch 5.74 1.0000 +/- 0.0010 0.9956 +/- 0.0007 0.4476 0.4580 57 WCAP-3385 (4A.17) Saxton Case 52 U 0.52" pitch 6.6% Pu 1.0036 +/- 0.0011 1.0047 +/- 0.0006 0.5289 0.5197 58 WCAP-3385 (4A.17) Saxton Case 56 PuO2 0.56" pitch 6.6% Pu 1.0008 +/- 0.0010 NC 0.6389 NC 59 WCAP-3385 (4A.17) Saxton Case 56 borated PuO2 5.74 0.9994 +/- 0.0011 0.9967 +/- 0.0007 0.2923 0.2954 60 WCAP-3385 (4A.17) Saxton Case 56 U 0.56" pitch 6.6% Pu 1.0063 +/- 0.0011 1.0133 +/- 0.0006 0.1520 0.1555 61 WCAP-3385 (4A.17) Saxton Case 79 Pu02 0.79" pitch 5.74 1.0039 +/- 0.0011 1.0008 +/- 0.0006 0.1036 0.1047 62 WCAP-3385 (4A.17 Saxton Case 79 U 0.79" pitch Notes: NC stands for not calculated.

t EALF is the energy of the average lethargy causing fission.

tt These experimental results appear to be statistical outliers (>3a) suggesting the possibility of unusually large experimental determining the calculational error. Although they could justifiably be excluded, for conservatism, they were retained in basis.

Appendix 4A, Page 13

Table 4A.2 COMPARISON OF MCNP4a AND KENO5a CALCULATED REACTIVITIES FOR VARIOUS ENRICHMENTS Calculated kg +/- la Enrichment MCNP4a KENO5a 3.0 0.8465 i 0.0011 0.8478 +/- 0.0004 3.5 0.8820 i 0.0011 0.8841 i 0.0004 3.75 .9019 i 0.0011 0.8987 i 0.0004 4.0 0.9132 i 0.0010 0.9140 i 0.0004 4.2 0.9276 +/- 0.0011 0.9237 +/- 0.0004 4.5 0.9400 +/- 0.0011 0.9388 +/- 0.0004 Based on the GE 8x8R fuel assembly.

Appendix 4A, Page 14

Table 4A.3 MCNP4a CALCULATED REACTIVITIES FOR CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS Ak MCNP4a Worth of Calcuilated EALFt Ref. Experiment Absorber kw (eV) 4A.13 PNL-2615 Boral Sheet 0.0139 0.9994+/-0.0012 0.1165 4A.7 B&W-1484 Core XX 0.0165 1.0008+/-0.0011 0.1724 4A.13 PNL-2615 1.62% Boron-steel 0.0165 0.9996+/-0.0012 0.1161 4A.7 B&W-1484 Core XIX 0.0202 0.9961+/-0.0012 0.2103 4A.7 B&W-1484 Core XXI 0.0243 0.9994+/-0.0010 0.1544 4A.7 B&W-1484 Core XVII 0.0519 0.9962+/-0.0012 0.2083 4A.11 PNL-3602 Boral Sheet 0.0708 0.9941+/-0.0011 0.3135 4A.7 B&W-1484 Core XV 0.0786 0.9910+/-0.0011 0.2092 4A.7 B&W-1484 Core XVI 0.0845 0.9935+/-0.0010 0.1757 4A.7 B&W-1484 Core XIV 0.1575 0.9953+/-0.0011 0.2022 4A.7 B1&W-1484 Core XIII 0.1738 1.0020+/-0.0011 0.1988 4A.14 PNL-7167 Expt 214R flux trap 0.1931 0.9991+/-0.0011 0.3722 tEALF is the energy of the average lethargy causing fission.

Appendix 4A, Page 1S

Table 4A.4 COMPARISON OF MCNP4a AND KENOSa CALCULATED REACTiVITIESt FOR VARIOUS '0B LOADINGS Calculated kf i la 10B, glcm2 MCNP4a KENO5a 0.005 1.0381 +/-0.0012 1.0340 0.0004 0.010 0.9960 +/- 0.0010 0.9941 i 0.0004 0.015 0.9727 +/- 0.0009 0.9713 +/- 0.0004 0.020 0.9541 i 0.0012 0.9560 i 0.0004 0.025 0.9433 +/- 0.0011 0.9428 +/- 0.0004 0.03 0.9325 i 0.0011 0.9338 i 0.0004 0.035 0.9234 i 0.0011 0.9251 i 0.0004 0.04 0.9173 i 0.0011 0.9179 +/- 0.0004 t Based on a 4.5% enriched GE 8x8R fuel assembly.

Appendix 4A, Page 16

Table 4A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH THICK LEAT) AND STEEL REFLECTORSt Separation, Ref. Case E, wt% cm MCNP4a kff KEN05a k,,

4A. I I Steel 2.35 1.321 0.9980+/-0.0009 0.9992+/-0.0006 Reflector 2.35 2.616 0.9968+/-0.0009 0.9964+/-0.0006 2.35 3.912 0.9974+/-0.0010 0.9980+/-0.0006 2.35 - 0.9962+/-0.0008 0.9939+/-0.0006 4A. I I Steel 4.306 1.321 0.9997+/-0.0010 1.0012+/-0.0007 Reflector 4.306 2.616 0.9994+/-0.0012 0.9974+/-0.0007 4.306 3.405 0.9969+/-0.0011 0.9951+/-0.0007 4.306 co 0.9910+/-0.0020 0.9947+/-0.0007 4A.12 Lead 4.306 0.55 1.0025+/-0.0011 0.9997+/-0.0007 Reflector 4.306 1.956 1.0000+/-0.0012 0.9985+/-0.0007 4.306 5.405 0.9971+/-0.0012 0.9946+/-0.0007 t Arranged in order of increasing reflector-fuel spacing.

Appendix 4A, Page 17

Table 4A.6 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS Calculated kf Boron Concentration, Reference Experiment ppm MCNP4a KENO5a 4A.15 PNL-4267 0 0.9974 +/- 0.0012 4A.8 B&W-1645 886 0.9970 +/- 0.0010 0.9924 +/- 0.0006 4A.9 B&W-1810 1337 1.0023 +/- 0.0010 4A.9 B&W-1810 1899 1.0060 +/- 0.0009 4A.15 PNL-4267 2550 1.0057 +/- 0.0010 _

Appendix 4A, Page 18

Table 4A.7 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNP4a KCENOSa Reference k EAL.Fjtt EALFI PNL-5803 MOX Fuel - Exp. No. 21 1.0041+/-:O.0011 0.9171 1.0046+/-0.0006 0.8868 14A. 16]

MOX Fuel - Exp. No. 43 1.0058+/-0.0012 0.2968 1.0036+/-0.0006 0.2944 MOX Fuel - Exp. No. 13 1.0083+/-0.0011 0.1665 0.9989+/-0.0006 0.1706 MOX Fuel - Exp. No. 32 1.0079+/-0.0011 0.1139 0.9966+/-0.0006 0.1165 WCAP- Saxton @ 0.52" pitch 0.9996+/-0.0011 0.8665 1.0005+/-0.0006 0.8417 3385-54

[4A.17] Saxton @ 0.56' pitch 1.0036+/-0.0011 0.5289 1.0047+/-0.0006 0.5197 Saxton @ 0.56" pitch borated 1.0008+/-0.0010 0.6389 NC NC Saxton @ 0.79" pitch 1.0063+/-0.0011 0.1520 1.0133+/-0.0006_ 0.1555 Note: NC stands for not calculated t Arranged in order of increasing lattice spacing.

tt EALF is the energy of the average lethargy causing fission.

Appendix 4A, Page 19

- --- Linear Regression with Correlation Coefficient of 0.13 1.010 1.005 0

ID 0

'4-4-

CD vx 1.000

'a 0

0 0

0.995 0.990 0.1 1 Energy of Average Lethargy Causing Fission (Log Scale)

FIGURE 4A.1 MCNP CALCULATED k-eff VALUES for

VARIOUS VALUES OF THE SPECTRAL INDEX

- - - - Linear Regression with Correlation Coefficient of 0.21 1.010 1.005 0

GO1.000

'4-4-

Qa S 0.995 0

0.990 0.985 0.1 Energy of Average Lethargy Causing Fission (Log Scale)

FIGURE 4A.2. KEN05a CALCULATED k-eff VALUES FOR VARIOUS VALUES OF THE SPECTRAL INDEX

-- - Linear Regression with Correlation Coefficient of 0.03 1.010 -

1.005 -

0

.4-0 4-U 0 1.000 - - :3 C.

U 0

0.995 -

0.990 -

2.0' Enrichment, w/o U-235 FIGURE 4A.3. MCNP CALCULATED k-eff VALUES AT VARIOUS U-235 ENRICHMENTS

Linear Regression with Correlation Coefficient of 0.38 1.010 1.005 a) l 1.000 I-o l

a)c 20 0.995 U

0.990 0.985 Enrichment, w/o U-235 FIGURE 4A.4. KENO CALCULATED k-eff VALUES AT VARIOUS U-235 ENRICHMENTS

0.94 E 0.92 C

0 0.90

-0 C.,

0 0

'-Y 0.88 0

z Lii 0.86 0.84 MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KEN05A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS

CI ., '..

I .*r j X Inn= n /-.

1 .03 t -- ---- - - - - - -

Ai 1.01 _ _ .

0-z 1 .0- d .

.. 010 g/ama

• 0 0.99- _

0 C'-, 0.96-Cm

._- 0.97 . . 7 3 X0.15 g/amuq 0

C.)

0.96- . --

0 e 0.020 g/on sq 0.96-0.025 /emaq 0.949-0.93- ___7___.030 g/cma l _ _ _ __ _ _ _

0.92-_

0.035 g/omwq 3j 0.04 g/o4q a-0, Q1 4 Ai I 1.

I i I 1 1 I

L I

. 9I a I I I II ' I I I I I Iji I I I l l Ii 0.900 0.920 0.940 0.960 0.9E0 1.000 1.020 1.040 Reactivity Calculated with KEN05a FIGURE 4A.6' COMPARISON OF MCNP AND KENO5a CALCULATIONS FOR VARIOUS BORON-10 AREAL DENSITIES

5.0 THERMAL-HYDRAULIC EVALUATION 5.1 Introduction The storage capacity at Clinton Power Station (CPS) is proposed to be expanded by adding racks to the open spaces in the Cask Storage Pool (CSP) and re-racking of the Spent Fuel Pool (SFP) with high density storage racks. To ensure the thermal-hydraulic adequacy of the expanded storage configuration it must be demonstrated that the Spent Nuclear Fuel (SNF) in the SFP and CSP racks is adequately cooled.

This section provides a summary of the methods, models, analyses and numerical results to demonstrate the CPS spent fuel pool meets the thernal-hydraulic requirements for safe storage of SNF set forth in Sub-section 5.2 herein. Similar thermal-hydraulic analyses have been used in spent fuel pool licensing at many nuclear plants worldwide (see Table 5. 1.1 for a partial list). Specifically, the following analyses are required:

1. Calculation of the SNF decay heat. The decay heat contributions from both previously stored fuels and freshly discharged fuels must be considered.

2. Determination of the SFP bulk thermal response versus time in accordance with each discharge scenario.

3. Calculation of time-to-boil during a postulated loss of forced cooling event for each discharge scenario.

4. Rigorous Computational Fluid Dynamics (CFD) based study to conservatively quantify the peak local water temperature in the fuel storage cells.

5. Determination of the maximum fuel clad temperature.

The CPS thermal-hydraulic evaluation is presented in the following.

Holtec Report HI-2033124 5-1 Report 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

5.2 Acceptance Criteria Applicable codes, standards and regulations include the following:

a. NUREG-0800, Standard Review Plan, Section 9.1.3.

b. USNRC OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Application, 4/78 [5.1. 11.

The design of the rack modules must ensure that fuel assemblies are adequately cooled by natural circulation of water. For this purpose bounding discharge scenarios are defined and evaluated. The storage of SNF in the SFP and CSP fuel racks is evaluated for two cases:

Case I for Normal (end of cycle) fuel batch discharge and Case II for Abnormal (full core) discharge. The case description and acceptance criteria are:

1. Under a Normal discharge scenario (Case I), a batch of 312 bundles is transferred to the SFP after 24 months of reactor operation. The SFP is cooled with one Spent Fuel Pool Cooling System (SFPCS) train operating with component cooling water (CCW) at 105TF. The bulk pool water temperature shall be limited to 1400 F and water in the rack cells in a sub-cooled condition.

2. Under an Abnormal discharge scenario (Case 11) a full core (624 bundles) is transferred to the SFP after 24 months of reactor operation. The SFP is cooled with one SFPCS train operating with component cooling water (CCW) at 1050F. The bulk pool water temperature shall be limited to 150 0F and water in the rack cells in a sub-cooled condition.

Holtec Report HI-2033124 5-2 Report 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

5.3 Assumptions and Inputs 5.3.1 Assumptions The Clinton thermal-hydraulic analysis embeds an array of assumptions to render a conservative portrayal of spent fuel pool temperatures. A numbered list of assumptions are provided in the following:

Bulk Pool Temperature Analyses

1. Heat loss by the conduction through pool walls is neglected.

2. The thermal capacity of the SFP is computed on the water volume above fuel. This assumption neglects thermal capacity of fuel, cladding, racks metal and water in the racks resulting in faster computed heat-up rate, higher pool temperatures and shorter times-to-boil.

3. As an added measure of conservatism, thermal inertia of water in the upper containment pool and cask storage pool are neglected.

4. The decay heat load contribution of previously discharged fuel assemblies is assumed constant during all discharge scenarios.

5. All refueling batches discharged previously and freshly discharged fuels are assumed to be at a bounding 43000 MWD/MTU burnup. This maximizes the decay heat load associated with all fuels stored in the SFP.

6. For minimizing the cooling time of old fuel, a short (18 month) cycle is assumed for previously discharged fuel batches.

7. The maximum SFP building temperature is assumed for evaporative heat loss calculations. This conservatively minimizes credit for evaporative cooling.

8. All of the radioactive decay energy is assumed to heat the fuel pool water, including that from gamma radiation. This conservatively maximizes heat input to the SFPCS, yielding conservative bulk temperatures.

Holtec Report HI-2033 124 5-3 Report 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

9. For time-to-boil calculations a loss of forced cooling is assumed to occur at the instant of maximum bulk SFP temperature for each discharge scenario. This conservatively minimizes the time-to-boil.

Local Temperature Analyses

1. Passive heat losses (i.e., conduction through walls and slab) are neglected.

2. Calculations employ conservative bounding hydraulic resistances of fuel. This understates natural circulation cooling and maximizes water temperatures.

3. Downcomer flow between rack modules is neglected. This conservatively minimizes natural circulation flow.

4. The peripheral downcomer area is understated in the CFD modeling. This conservatively maximized the downcomer flow resistance.

5. The hottest fuel assemblies are assumed to be located together at the center of the spent fuel pool, conservatively maximizing the local decay heat generation rates.

6. An additional heat transfer resistance (0.005 (hrxfi 2 xoF)/Btu) is conservatively assumed to the outside of the fuel rods to maximize clad temperatures.

7. For peak-clad temperature calculations the maximum local water temperature (at the fuel rack cell exit) and peak heat flux (typically near the mid-height of the active fuel region) are assumed to occur co-incidentally. The superposition of these two maximum values conservatively maximizes the computed fuel clad temperatures.

5.3.2 Design Data Bulk Pool Analysis The principal input data employed to determine the maximum bulk temperature of CPS fuel pools is summarized in Table 5.3.1. To maximize decay heat, a bounding burnup, maximum uranium weight and minimum fuel enrichment reported in Table 5.3.1 are assumed for all fuel discharged in the CPS pool. The cumulative decay heat load from old fuel batches is Holtec Report HI-2033 124 5-4 Report 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

determined using ORIGEN 2 [5.3.2], as incorporated in Holtec's QA validated computer program DECOR [5.3.1], for a sufficient number of discharged batches to fill the racks in the SFP and CSP up to a reserve of one normal batch.

Local Temperature Analysis The principal inputs for local analysis, namely the racks construction data and fuel bundle data are presented in Tables 5.3.2 and 5.3.3.

5.4 Bulk Pool Analysis Methodologw This analysis is performed to determine bulk temperature of water and decay heat profiles under postulated discharge scenarios. The mathematical formulation for this analysis can be explained with reference to the simplified heat exchanger alignment shown in Figure 5.4.1.

Referring to the Spent Fuel Pool Cooling System (SFPCS), the governing differential equation for bulk pool temperature can be written by utilizing conservation of energy as:

dT C X d- = Pcot. + Q(v) - QHX (T) - QEVAP (T, TA ) + QP - Q, where:

C thermal capacity of water in the pool, BtulPF PcoIvs heat generation rate from previous discharges, Btu / hr Q(r) heat generation rate from recently discharged fuel as a function of time, Btu/ hIr QHx (T) heat removed by the SFP heat exchangers, Btu / hr QEVAP (T, TA) evaporative heat loss, Btu / hr QP pump heat, Btu / hr QW. heat loss from pool walls, Btu / hr (conservatively neglected)

Holtec Report HI-2033124 5-5 Report 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

T bulk pool temperature, 'F TA fuel building ambient temperature, 'F In bulk pool thermal-hydraulics, it is recognized that pump heat (Qp) is very small in comparison with pool decay heat (about 2 orders of magnitude smaller). As a result it's effect on pool temperatures is negligible. Accordingly, the contribution of pump heat to fuel pool energy input is ignored in the evaluation.

The heat removed by the SFP heat exchangers, QHX (T) is defined by the following equation:

QHX (T) = WC -CC *p *(T - Tc) where:

JVc coolant flow rate, l-Cc coolant specific heat, Blu/ThF p temperature effectiveness of heat exchanger T bulk pool temperature, 'F TC coolant inlet water temperature, 'F The equation used to determine the temperature effectiveness, p of the SFPCS heat exchanger is as follows:

p= Tco - TC.,

Tp -TcJ where:

Tc,i coolant inlet water temperature, F TC.o coolant outlet water temperature, F T7 SFP water temperature at heat exchanger inlet, 'F Holtec Report HI-2033124 5-6 Report 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

The temperature effectiveness, p is determined from the SFP heat exchanger specifications.

The SFP decay heat contribution from all previously stored fuels is held constant during the entire analysis because its decrease with decay time after shutdown is conservatively neglected. The decay heat generation, Q(r) of the freshly discharged fuel decays exponentially with elapsed time after reactor shutdown. The evaporative heat loss, QEVAP (T, TA) is a nonlinear function of pool temperature, T and ambient temperature, TAX and includes heat loss from the pool surface by natural convection and radiation. The evaporation heat loss is a function of rate of moisture generation from the pool surface which is obtained using a proprietary Holtec correlation [5.4.1]. For a conservative assessment of evaporation cooling of the fuel pool, the ambient temperature is assumed to be at it's design (i.e. maximum) temperature and relative humidity is 100%.

The Holtec's QA validated BULKTEM computer program [5.4.2] is used to numerically solve the bulk pool thermal hydraulic equations described in the foregoing. Inputs to the BULKTEM computer program include the coolant flow rate and temperature, the temperature effectiveness of SFP cooling heat exchanger, the discharge batch transfer time from reactor to SFP, in-core exposure times, the average assembly operating power, "old" fuels decay heat contribution, SFP thermal capacity, initial SFP water temperature, SFP surface area, and ambient temperature in fuel building. Outputs from BULKTEM include the bulk water temperature, evaporative cooling loss and net decay heat - all as functions of time.

5.4.1 Time-To-Boil Calculation This analysis is used to determine the time it takes for pool water to reach boiling temperature in case all forced cooling trains become unavailable. Clearly, the most critical instant of loss-of-cooling is when the SFP pool water temperature has reached its maximum value. The following governing differential equation can be derived from the differential Holtec Report HI-2033124 5-7 Report 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

thermal response equation defined in Section 5.4 without including the SFPCS heat rejection term.

C X dT=Pcon + Q(T) QE AP (T, TA )

dr where:

C is thermal capacity of the pool, Bf10F Pcom.v is heat generation rate from "old" fuel, Btu / hr Q(r) is heat generation rate from recently discharged fuel as a function of time, Buu / hr QEVAP (TTA) is evaporative heat loss, Ball / hr T is bulk pool temperature, F TA is fuel building ambient temperature, 'F The Holtec QA validated TBOIL computer program [5.4.3] is used to numerically integrate the foregoing equation and time-to-boil, boil-off rate, and pool water depth versus time profile obtained.

5.4.2 Bulk Pool Temperature Results As required by Sub-section 5.2, the maximum bulk pool temperature analyses are performed for the two cases, namely Case I (Normal) and Case 11 (Abnormal) discharge scenarios. As stated previously in Sub-section 5.3.2, for decay heat inputs the bulk pool evaluations use bounding assumptions for fuel burnup, uranium weight and enrichment.

The time dependent bulk pool temperature profiles are presented in Figures 5.4.2 and 5.4.3.

The time dependent pool decay heat profiles are presented in Figures 5.4.4 and 5.4.5.

Holtec Report HI-2033124 5-8 Report 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

The results of the maximum bulk pool temperature analyses are summarized in Table 5.4.1.

The results of the analysis confirm that the bulk pool water temperatures are below the prescribed limits set forth in Section 5.2 (1400F (Normal), 150'F (Abnormal)).

5.4.3 Time-to-Boil Results Under a loss of all forced pool cooling, the pool water temperature will rise to reach the boiling temperature of water (-212'F). The time to reach boiling temperature will be the shortest when the loss of forced cooling occurs coincident with the maximum bulk pool temperature. Although the probability of a loss-of-cooling event coinciding with the instant when the pool water has reached its peak value is extremely remote, the calculations are performed under this scenario to obtain a bounding result. The results with the additional proviso that no makeup water is added to the pool are summarized in Table 5.4.2. The water depth profiles in the SFP as a function of time after loss of cooling are presented in Figure 5.4.6 and 5.4.7.

5.5 Local Analysis Methodologv 7~  ;..I I;'.i ff,.S. . I .

.. ' '-:; .  :.'i .*- '*.11;.-,.w1, j~....-.

a~~~~~~ n .I ! ,E,2i Holtec Report HI-2033124 5-9 Report 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

1'

.t;>.>- . ' '. ,'_ '. ' ;,.; ',;i.; -

-Jt is~~~~~-'7 ZW<',, I i: ~24-hr In-core 105- Hold mie 0 25 50 75 100 125 150 175 200 225 250 275 300 325 Time After Reactor Shutdown (hr)

FIGURE 5.4.2: NORMAL DISCHARGE SCENARIO BULK POOL TEMPERATURE PLOT X .

CD

-0 0

134.95 0 F Max Temperature Mi CI(

aJ 0

4-1 0.6 E2 cii C-0 25 50 75 100 125 150 175 200 225 250 275 300 325 Time After Reactor Shutdown (hr)

-a.

.2

!I FIGURE 5.4.3: FULL CORE DISCHARGE SCENARIO BULK POOL TEMPERATURE PLOT Ci)

N)

P:.s 30- _ _ _ ._

0 CAj

-:5 m c co m

z C o2 O- ,_ _,,..,.

_ _ _odTme.,..,,,

Ca 0=

m0 0_ 25 5_510_2_5 7 20 25 20 7 0 2 0 a

> ~ 0 o 0 25 50 75 .100 125 150 175 200 '225 250 275 300 325 Time After Reactor Shutdown (hr)

FIGURE 5.4.4: NORMAL DISCHARGE SCENARIO FUEL POOL DECAY HEAT PLOT 0

0,

CD 0

I 0 45-04 35-

• 40

~30-__

25_ \_ _ _ _ _ _ _ lli _lll_ _

5.25 z

-20 35- ____Q _ __ in-cor __ _

o CD Ho,,.,,

odm,

-n 0

24 hr in-core

-I Hold Time 0 25 50 75 100 125' 150 175 200 225 250 275 300 325 Time After Reactor Shutdown (hr)

FIGURE 5.4.5: FULL CORE DISCHARGE SCENARIO POOL DECAY HEAT PLOT C,

CD 0

00 em Cn Top of 0 -

Racks cuj o ~ ~~0 1 03 05 07 Fnj 10-

~I1 5-0 10 20 30 40 50 60 70 Time After Loss of Cooling (hr)

CD FIGURE 5.4.6: WATER DEPTH PLOT IN A LOSS OF COOLING EVENT (CASE I) vi

3a 03 N,5 00 (A)-

. 5- .<~ \ 1Top of 10-5-

0 5 10 15 20 25 30 35 40 45 Time After Loss of Cooling (hr)

H.T CnD FIGURE 5.4.7: WATER DEPTH PLOT IN A LOSS OF COOLING EVENT (CASE II)

N,

o.

0 0

W.8

tJ 0

w t~'3

-P

,z Figure 5.5.1 z

Proprietary Fgr 0

?Tl Po 0

0 z

10 A.

O 0

0 w

w z

Z Figure 5.5.2

Proprietary Figure 0

0

~T1 z

0

0 w

z z

!O 10 0

Figure 5.5.3

'a Proprietary Figure o

In X

0 3 C)

6.0 STRUCTURAL/SEISMIC CONSIDERATIONS 6.1 Introduction This section considers the structural adequacy of the new Spent Fuel Pool (SFP) maximum density spent fuel racks under all loadings postulated for normal, seismic, and accident conditions at the Clinton Power Station (CPS). New racks will be placed in both the SFP and the Fuel Cask Storage Pool in two phases, as discussed in Section 1.0. The module layouts for both phases are illustrated in Figures 1.1.1 thru 1.1.3.

In order to evaluate interim configurations encountered during periods wherein the full complement of racks is not yet installed, additional single rack evaluations are performed.

During the structural evaluation of the racks, the input parameters affecting the dynamic rack response are varied to determine the set of characteristics resulting in greatest displacement at the top of the racks for both the Operating Basis Earthquake (OBE) and the Safe Shutdown Earthquake (SSE). Following this determination, single rack analyses with increased earthquake excitation are performed for those cases to conservatively establish the overturning safety factor of the racks.

The analyses undertaken to confirm the structural integrity of the racks, are performed in compliance with the USNRC Standard Review Plan (SRP) [6.1.1] and the OT Position Paper

[6.1.2]. An abstract of the methodology, modeling assumptions, key results, and summary of the parametric evaluation is presented. Delineation of the relevant criteria is discussed in the text associated with each analysis.

Holtec Report H--2033124 6-1 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.2 Overview of Rack Structural Analysis Methodolory The response of a free-standing rack module to seismic inputs is highly nonlinear and involves a complex combination of motions (sliding, rocking, twisting, and turning), resulting in potential impacts and friction effects. Some of the unique attributes of the rack dynamic behavior include a large fraction of the total structural mass in a confined rattling motion, friction support of rack pedestals against lateral motion, and large fluid coupling effects due to deep submergence and independent motion of closely spaced adjacent structures.

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. An accurate simulation is obtained only by direct integration of the nonlinear equations of motion with the three pool slab acceleration time-histories applied as the forcing functions acting simultaneously.

Whole Pool Multi-Rack (WPMR) analysis is the vehicle utilized in this project to simulate the dynamic behavior of the complex storage rack structures. The following sections provide the basis for this selection and discussion on the development of the methodology.

6.2.1 Background of Analysis Methodology Reliable assessment of the stress field and kinematic behavior of the rack modules calls for a conservative dynamic model incorporating all key attributesof the actual structure. This means that the model must feature the ability to execute the concurrent motion forms compatible with the free-standing installation of the modules.

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 Holtec Report HI-2033124 6-2 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

modeled 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 6.2.1) indicates that an upper bound value of the coefficient of friction often maximizes the computed rack displacements as well as the corresponding elastostatic stresses.

In short, there are a large number of parameters with potential influence on the rack kinematics.

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 some 50 rerack projects since that time (Table 6.2.1) addresses most of the abovementioned array of parameters. The details of this methodology are also published in the permanent literature [6.2.1]. Despite the versatility of the 3-D seismic model, the accuracy of the single rack simulations has been suspect due to one key element; namely, hydrodynamic participation of water around the racks. During dynamic rack motion, hydraulic energy is either drawn from or added to the moving rack, modifying its submerged motion in a significant manner. Therefore, the dynamics of one rack affects the motion of all others in the pool.

A dynamic simulation, which treats only one rack, or a small grouping of racks, is intrinsically inadequate to predict the motion of rack modules with any quantifiable level of accuracy. Three-dimensional Whole Pool Multi-Rack analyses carried out on several previous plants demonstrate that single rack simulations under predict rack displacement during seismic responses [6.2.2].

Briefly, the 3-D rack model dynamic simulation, involving one or more spent fuel racks, handles the array of variables as follows:

Holtec Report HI-2033124 6-3 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

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. Simulations are also performed with the array of pedestals having randomly chosen coefficients of friction in a Gaussian distribution with a mean of 0.5 and lower and upper limits of 0.2 and 0.8, respectively. In the fuel rack simulations, the Coulomb friction interface between rack support pedestal and liner is simulated by piecewise linear (friction) elements. These elements function only when the pedestal is physically in contact with the pool liner or bearing pad.

Rack Beam Behavior Rack elasticity, relative to the rack base, is included in the model by introducing linear springs to represent the elastic bending action, twisting, and extensions.

Impact Phenomena Compression-only gap elements are used to provide for opening and closing of interfaces such as the pedestal-to-bearing pad interface, and the fuel assembly-to-cell wall interface. These interface gaps are modeled using nonlinear spring elements. The term "nonlinear spring" is a generic term used to denote the mathematical representation of the condition where a restoring force is not linearly proportional to displacement.

Fuel Loading Scenarios The fuel assemblies are conservatively assumed to rattle in unison which obviously exaggerates the contribution of impact against the cell wall.

Fluid Coupling Holtec International 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 [6.2.4] which handles simultaneous simulation of all racks in the pool as a Whole Pool Multi-Rack 3-D analysis. This development was first utilized in Chinshan, Oyster Creek, and Harris plants [6.2.1, 6.2.3] and, subsequently, in numerous other rerack projects. The Holtec Report HI-2033124 6-4 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

WPMR analyses have corroborated the accuracy of the single rack 3-D solutions in predicting the maximum structural stresses, and also serve to improve predictions of rack kinematics.

For closely spaced racks, demonstration of kinematic compliance is verified by including all modules in one comprehensive simulation using a WPMR model. In WPMR analysis, all rack modules in the pool are modeled and evaluated simultaneously and the coupling effect due to this multi-body motion is included in the analysis. Due to the superiority of this technique in predicting the dynamic behavior of closely spaced submerged storage racks, the Whole Pool Multi-Rack analysis methodology is used to evaluate the configurations of the storage racks subsequent to the completion of the reracking process, as shown in Figures 1.1.1 through 1. 1.3.

Additional, more conservative, single rack analyses are performed to confirm kinematic stability under the most adverse conditions such as fuel loading eccentricities and interim reracking configurations.

Holtec Report H1-2033124 6-5 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.3 Description of Racks The racks in the SFP and the Fuel Cask Storage Pool are analyzed as follows:

RACK WEIGHT DATA Empty Rack Rack #/Module I.D. Cells/Module Array Size Dry Weight (Ibs) 1-3/F1-F3 150 15x1O 14,142 4/H 110 llxlO 10,637 5-7,9-11,13-16/GI-GIO 180 15x12 16,737 8/JI 132 1lx12 12,558 12/J2 144 15x12 13,769 17-19,22-24/B I -B3,A2-A3,C1 120 12xlO 14,467 20,21,25,26/D1-D4 110 llxlO 13,381 CPl,CP3/A4,B5 132 11x12 15,761 CP2/B4 99 11x9 12,190 For the purpose of analytical modeling, the racks in all cases addressed are numbered. Rack #1 is in the northwest corner of either the SFP or Fuel Cask Storage Pool. The numbering progresses west to east, continuing with the west most rack in the next row to the south, etc..

Thus for the SFP case module H, in the northeast corner, is Rack #4 and module D4 in the southeast corner is Rack #26. In Fuel Cask Storage Pool cases, modules G7 and G8 are Rack #1 and Rack #2, respectively, in Phase 1. In Phase 2, referring to Figure 1.1.3, the existing rack identities of CPI thru CP3 are given above.

Rack material is defined in Table 6.3.1.

The cartesian coordinate system utilized within the rack dynamic model has the following nomenclature:

x = Horizontal axis along plant East y = Horizontal axis along plant North z = Vertical axis upward from the rack base Holtec Report HI-2033124 6-6 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.4 Svnthetic Time-Histories The synthetic time-histories in three orthogonal directions (N-S, E-W, and vertical) are generated in accordance with the provisions of SRP Section 3.7.1 [6.1. 1]. In order to prepare an acceptable set of acceleration time-histories, Holtec International's proprietary code GENEQ [6.4.1] is utilized.

A preferred criterion for the synthetic time-histories in SRP 3.7.1 calls for both the response spectrum and the power spectral density corresponding to the generated acceleration time-history to envelope their target (design basis) counterparts with only finite enveloping infractions. The time-histories for the pools have been generated to satisfy this preferred criterion. The seismic files also satisfy the requirements of statistical independence mandated by SRP 3.7.1.

Figures 6.4.1 through 6.4.3 provide plots of the time-history accelerograms which were generated over a 20 second duration for the SSE event. Figures 6.4.4 through 6.4.6 provide plots of the time-history accelerograms which were generated over a 20 second duration for the OBE event.

These artificial time-histories are used in all non-linear dynamic simulations of the racks.

Results of the correlation function of the three time-histories are given in Table 6.4.1. Absolute values of the correlation coefficients are shown to be less than 0.15, indicating that the desired statistical independence of the three data sets has been met.

Holtec Report HI-2033124 6-7 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.5 WPMR Methodologv Recognizing that the analytical work effort must deal with both stress and displacement criteria, the sequence of model development and analysis steps that are undertaken are summarized in the following:

a. Prepare 3-D dynamic models suitable for a time-history analysis of the new maximum density racks. These models include the assemblage of all rack modules in each pool. Include all fluid coupling interactions and mechanical coupling appropriate to performing an accurate non-linear simulation. This 3-D simulation is referred to as a Whole Pool Multi-Rack model.

b. Perform 3-D dynamic analyses on various physical conditions (such as coefficient of friction and extent of cells containing fuel assemblies). Archive appropriate displacement and load outputs from the dynamic model for post-processing.

c. Perform stress analysis of high stress areas for the limiting case of all the rack dynamic analyses. Demonstrate compliance with ASME Code Section III, Subsection NF limits on stress and displacement.

6.5.1 Model Details for Spent Fuel Racks The dynamic modeling of the rack structure is prepared with special consideration of all nonlinearities and parametric variations. Particulars of modeling details and assumptions for the Whole Pool Multi-Rack analysis of racks are given in the following:

6.5.1.1 Assumptions

a. The fuel rack structure motion is captured by modeling 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 manner, the response of the module, relative to the base-plate, is captured in the dynamic analyses once suitable springs are introduced to couple the rack degrees-of-freedom and simulate rack stiffness.

b. Rattling fuel assemblies within the rack are modeled by five lumped masses located at H, .75H, .5H, .25H, and at the rack base (H is the rack height measured Holtec Report HI-2033124 6-8 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

above the base-plate). 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 base-plate 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.

c. 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.

d. Fluid coupling between the rack and fuel assemblies, and between the rack and wall, is simulated by appropriate inertial coupling in the system kinetic energy.

Inclusion of these effects uses the methods of [6.5.2, 6.5.3] for rack/assembly coupling and for rack-to-rack coupling.

e. Fluid damping and form drag are conservatively neglected.

f. Sloshing is found to be negligible at the top of the rack and is, therefore, neglected in the analysis of the rack.

g. 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 the masses involved. The possible incidence of rack-to-wall or rack-to-rack impact is simulated by gap elements at the top and bottom of the rack in two horizontal directions. Bottom gap elements are located at the base-plate elevation.

The initial gaps reflect the presence of baseplate extensions, and the rack stiffnesses are chosen to simulate local structural detail.

h. Pedestals are modeled by gap elements in the vertical direction and as "rigid links" for transferring horizontal stress. The base of each pedestal support is linked to the pool liner (or bearing pad) by two friction springs. The spring rate for the friction springs includes any lateral elasticity of the pedestals. Local pedestal vertical spring stiffness accounts for floor elasticity and for local rack elasticity just above the pedestal.

i. 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 in order to provide a conservative measure of fluid resistance to gap closure.

j. The model for the rack is considered supported, at the base level, on four pedestals modeled as non-linear compression only gap spring elements and eight piecewise linear friction spring elements. These elements are properly located Holtec Report HI-2033124 6-9 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

with respect to the centerline of the rack beam, and allow for arbitrary rocking and sliding motions.

6.5.1.2 Element Details Figure 6.5.1 shows a schematic of the dynamic model of a single rack. The schematic depicts many of the characteristics of the model including all of the degrees-of-freedom and some of the spring restraint elements.

Table 6.5.1 provides a complete listing of each of the 22 degrees-of-freedom for a rack model.

Six translational and six rotational degrees-of-freedom (three of each type on each end) describe the motion of the rack structure. Rattling fuel mass motions (shown at nodes 1, 2, 3', 4%, and 5' in Figure 6.5.1) are described by ten horizontal translational degrees-of-freedom (two at each of the five fuel masses). The vertical fuel mass motion is assumed (and modeled) to be the same as that of the rack baseplate.

Figure 6.5.2 depicts the fuel to rack impact springs (used to develop potential impact loads between the fuel assembly mass and rack cell inner walls) in a schematic isometric. Only one of the five fuel masses is shown in this figure. Four compression only springs, acting in the horizontal direction, are provided at each fuel mass.

Figure 6.5.3 provides a 2-D schematic elevation of the storage rack model, discussed in more detail in Section 6.5.3. This view shows the vertical location of the five storage masses and some of the support pedestal spring members.

Figure 6.5.4 shows the modeling technique and degrees-of-freedom associated with rack elasticity. In each bending plane a shear and bending spring simulate elastic effects [6.5.4].

Linear elastic springs coupling rack vertical and torsional degrees-of-freedom are also included in the model.

Figure 6.5.5 depicts the inter-rack impact springs (used to develop potential impact loads between racks or between rack and wall).

Holtec Report HI-2033124 6-10 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.5.2 Fluid Coupling Effect In its simplest form, the so-called "fluid coupling effect" [6.5.2, 6.5.3] can be explained by considering the proximate motion of two bodies under water. 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:

(ml + M,,)A, + M12 A2 = applied forces on mass ml + 0 (Xi2 )

M2, A, + (m2 + M22) A2 = applied forces on mass M2 + 0 (X2 )

A, and A2 denote absolute accelerations of masses ml and M2 , respectively, and the notation O(X2 ) denotes nonlinear terms.

Ml , M,2 , M21, and M22 are fluid coupling coefficients which depend on body shape, relative disposition, etc. Fritz [6.5.3] gives data for Mij for various body shapes and arrangements. The fluid adds mass to the body (MI, to mass ml), 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 storage location encounters this effect. For example, fluid coupling behavior will be experienced between nodes 2 and 2* in Figure 6.5.1. The rack analysis also contains inertial fluid coupling terms, which model the effect of fluid in the gaps between adjacent racks.

Terms modeling the effects of fluid flowing between adjacent racks in a single rack analysis suffer from the inaccuracies described earlier. These terms are usually computed assuming that all racks adjacent to the rack being analyzed are vibrating in-phase or 1800 out of phase. The WPMR analyses do not require any assumptions with regard to phase.

Rack-to-rack gap elements have initial gaps set to 100% of the physical gap between the racks or between outermost racks and the adjacent pool walls.

Holtec Report HI-2033124 6-11 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.5.2.1 Multi-Body Fluid Coupling Phenomena During the seismic event, all racks in the pool are subject to the input excitation simultaneously.

The motion of each free-standing module would be autonomous and independent of others as long as they did not impact each other and no water were present in the pool. While the scenario of inter-rack impact is not a common occurrence and depends on rack spacing, the effect of water (the so-called fluid coupling effect) is a universal factor. As noted in Ref. [6.5.2, 6.5.4], the fluid forces can reach rather large values in closely spaced rack geometries. It is, therefore, essential that the contribution of the fluid forces be included in a comprehensive manner. This is possible only if all racks in the pool are allowed to execute 3-D motion in the mathematical model. For this reason, single rack or even multi-rack models involving only a portion of the racks in the pool, are inherently inaccurate. The Whole Pool Multi-Rack model removes this intrinsic limitation of the rack dynamic models by simulating the 3-D motion of all modules simultaneously. The fluid coupling effect, therefore, encompasses interaction between every set of racks in the pool, i.e., the motion of one rack produces fluid forces on all other racks and on the pool walls. Stated more formally, both near-field and far-field fluid coupling effects are included in the analysis.

The derivation of the fluid coupling matrix [6.5.5] relies on the classical inviscid fluid mechanics principles, namely the principle of continuity and Kelvin's recirculation theorem. The derivation of the fluid coupling matrix has been verified by an extensive set of shake table experiments

[6.5.5].

6.5.3 Stiffness Element Details Three element types are used in the rack models. Type 1 are linear elastic elements used to represent the beam-like behavior of the integrated rack cell matrix. Type 2 elements are the piece-wise linear friction springs used to develop the appropriate forces between the rack pedestals and the supporting bearing pads. Type 3 elements are non-linear gap elements, which model gap closures and subsequent impact loadings i.e., between fuel assemblies and the storage cell inner walls, and rack outer periphery spaces.

Holtec Report HI-2033124 6-12 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

If the simulation model is restricted to two dimensions (one horizontal motion plus one vertical motion, for example), for the purposes of model clarification only, then Figure 6.5.3 describes the configuration. This simpler model is used to elaborate on the various stiffness modeling elements.

Type 3 gap elements modeling impacts between fuel assemblies and racks have local stiffness K; in Figure 6.5.3. Support pedestal spring rates Ks are modeled by type 3 gap elements. Local compliance of the concrete floor is included in Ks. The type 2 friction elements are shown in Figure 6.5.3 as Kf. The spring elements depicted in Figure 6.5.4 represent linear type I elements.

Friction at support/liner interface is modeled by the piecewise linear friction springs with suitably large stiffness Kf up to the limiting lateral load pIN, 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/bearing pad, or a compressive finite value) is computed.

The gap element Ks, modeling the effective compression stiffness of the structure in the vicinity of the support, includes stiffness of the pedestal, local stiffness of the underlying pool slab, and local stiffness of the rack cellular structure above the pedestal.

The previous discussion is limited to a 2-D model solely for simplicity. Actual analyses incorporate 3-D motions.

6.5.4 Coefficients of Friction To eliminate the last significant element of uncertainty in rack dynamic analyses, multiple simulations are performed to adjust the friction coefficient ascribed to the support pedestal/pool bearing pad interface. These friction coefficients are chosen consistent with the two bounding extremes from Rabinowicz's data [6.5.1]. Simulations are also performed by imposing intermediate value friction coefficients, both 0.5 and those developed by a random number generator with Gaussian normal distribution characteristics. The assigned values are then held Holtec Report H1-2033124 6-13 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

constant during the entire simulation in order to obtain reproducible results.t Thus, in this manner, the WPMR analysis results are brought closer to the realistic structural conditions.

The coefficient of friction (p) between the pedestal supports and the pool floor is indeterminate.

According to Rabinowicz [6.5.1], results of 199 tests performed on austenitic stainless steel plates submerged in water show a mean value of p to be 0.503 with standard deviation of 0.125.

Upper and lower bounds (based on twice standard deviation) are 0.753 and 0.253, respectively.

Analyses are therefore performed for coefficient of friction values of 0.2 (lower limit), 0.5 and 0.8 (upper limit), as well as for random friction values clustered about a mean of 0.5. The bounding values of p = 0.2 and 0.8 have been found to envelope the upper limit of module response in previous rerack projects.

6.5.5 Governing Equations of Motion Using the structural model discussed in the foregoing, equations of motion corresponding to each degree-of-freedom are obtained using Lagrange's Formulation [6.5.4]. 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:

[MI [ diq] d = [I/t2 +[G

+ [G]

where:

[M] - total mass matrix (including structural and fluid mass contributions). The size of this matrix will be 22n x22n for a WPMR analysis (n = number of racks in the model).

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 computer 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.

Holtec Report HI-2033124 6-14 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

q the nodal displacement vector relative to the pool slab displacement (the term with q indicates the second derivative with respect to time, i.e., acceleration)

[G] a vector dependent on the given ground acceleration

[Q] a vector dependent on the spring forces (linear and nonlinear) and the coupling between degrees-of-freedom The above column vectors have length 22n. The equations can be rewritten as follows:

E dt2] [M] [Q + [M [G]

This equation set is mass uncoupled, displacement coupled at each instant in time. The numerical solution uses a central difference scheme built into the proprietary computer program DYNARACK [6.2.4].

6.6 Structural Evaluation of Spent Fuel Rack Design 6.6.1 Kinematic and Stress Acceptance Criteria There are two sets of criteria to be satisfied by the rack modules:

a. Kinematic Criteria An isolated fuel rack situated in the middle of the storage cavity is most vulnerable to overturning because such a rack would be hydrodynamically uncoupled from any adjacent structures. Therefore, to assess the margin against overturning, a single rack module is evaluated.Section IV(6) of Reference [6.1.2]

Holtec Report HI-2033124 6-15 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

refers to the SRP for safety factors against rack overturning. According to SRP Section 3.8.5.11-5 [6.1. 1], the minimum required safety margins under the OBE and SSE events are 1.5 and 1.1, respectively. In order to ensure that these safety factors are met, the simulations resulting in the highest top of rack displacements were re-performed with an earthquake excitation multiplier of, conservatively, 1.5 for both OBE and SSE. The maximum rotations of the rack (about the two principal axes) are obtained from a post processing of the rack time history response output. The ratio of the rotation required to produce incipient tipping in either principal plane to the actual maximum rotation in that plane from the time history solution is the margin of safety. Since the factors of safety are conservatively embedded in the earthquake multipliers, meeting the acceptance criteria is established by the ratio described above being greater than 1.0.

b. Stress Limit Criteria Stress limits must not be exceeded under the postulated load combinations provided herein.

6.6.2 Stress Limit Evaluations The stress limits presented below apply to the rack structure and are derived from the ASME Code,Section III, Subsection NF, 2001 [6.6.1] for the new racks, and 1977 [6.6.3] for existing racks. Parameters and terminology are in accordance with the ASME Code. Material properties are obtained from the ASME Code Appendices [6.6.2], and are listed in Table 6.3.1.

(i) Normal Conditions (Level A)

a. Allowable stress in tension on a net section is:

Ft = 0.6 Sy Where, Sy = yield stress at temperature, and Ft is equivalent to primary membrane stress.

Holtec Report H1-2033124 6-16 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

b. Allowable stress in shear on a net section is:

F = 0.4 Sy

c. Allowable stress in compression on a net section is:

FaSy(.47 rl) where kl/r for the main rack body is based on the full height and cross section of the honeycomb region and does not exceed 120 for all sections.

I = unsupported length of component k length coefficient which gives influence of boundary conditions. The following values are appropriate for the described end conditions:

I (simple support both ends) 2 (cantilever beam)

%2(clamped at both ends) r = radius of gyration of component

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 bending and compression on a net section satisfies:

f[ + C.x f' b+ C,,,y f by<

Fa Dx Fbx Dy Fky where:

fa = Direct compressive stress in the section fbx = Maximum bending stress along x-axis fby = Maximum bending stress along y-axis Cmx = 0.85 C..y = 0.85 DX = 1 - (fa/F'ex)

Dy = I - (fa/F'ey)

F'exey = (n2 E)/(2.15 (kl/r)',,y)

Holtec Report HI-2033124 6-17 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

E = Young's Modulus and subscripts x,y reflect the particular bending plane.

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

f[ + fbr + fby <1 0 0.6Sy Fbx Fby The above requirements are to be met for both direct tension or compression.

g. Welds Allowable maximum shear stress on the net section of a weld is given by:

F = 0.3 Su where Su is the weld material ultimate strength at temperature. For fillet weld legs in contact with base metal, the shear stress on the gross section is limited to O.4SY, where Sy is the base material yield strength at temperature.

h. Bearing Allowable maximum stress for bearing on a contact area is given by:

Fp = 0.9 S, (ii) Level B Service Limits (Upset Conditions. including OBE)

Section NF-3321 (ASME Section III, Subsection NF [6.6.1]) states that, for the Level B condition, the allowable stresses for those given above in (i) may be increased by a factor of 1.33.

(iii) Level D Service Limits (including SSE)

Section F-1334 (ASME Section III, Appendix F [6.6.2]), states that limits for the Level D condition are the smaller of 2 or 1.167SU/Sy times the corresponding limits for the Level A condition if S, > I.2Sy, or 1.4 if Su, less than or equal l.2Sy except for requirements specifically listed below. Su ,Sy are the ultimate strength and yield strength at the Holtec Report HJ-2033124 6-18 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

specified rack design temperature. Examination of material properties for 304L stainless demonstrates that 1.2 times the yield strength is less than the ultimate strength.

Therefore, the Level D stress limits are double the correponding Level A limits.

Exceptions to the above general multiplier are the following:

a) Stresses in shear shall not exceed the lesser of 0. 7 2 SY or 0.42S,. In the case of the Austenitic Stainless material used here, 0.72SY governs.

b) Axial Compression Loads shall be limited to 2/3 of the calculated buckling load.

c) Combined Axial Compression and Bending - The equations for Level A conditions shall apply except that:

Fa = 0.667 x Buckling Load/ Gross Section Area, and the terms F',x and F'ey may be increased by the factor 1.65.

d) For welds, the Level D allowable maximum weld stress is not specified in Appendix F of the ASME Code. An appropriate limit for weld throat stress is conservatively set here as:

FR = (0.3 SJ)x factor where:

factor = (Level D shear stress limit)/(Level A shear stress limit)

Holtec Report HI-2033124 6-19 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.6.3 Dimensionless Stress Factors For convenience, the stress results are in dimensionless form. Dimensionless stress factors are defined as the ratio of the actual developed stress to the specified limiting stress value. The limiting value of each stress factor is 1.0. Stress factors are determined as follows:

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

R2 = Ratio of gross shear on a net section in the x-direction to its allowable value R3 = Ratio of maximum x-axis bending stress to its allowable value for the section R4 = Ratio of maximum y-axis bending stress to its allowable value for the section R5 = Combined flexure and compressive factor (as defined in the foregoing)

= Combined flexure and tension (or compression) factor (as defined in the foregoing)

R7 = Ratio of gross shear on a net section in the y-direction to its allowable value Holtec Report HI-2033124 6-20 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.6.4 Loads and Loading Combinations for Spent Fuel Racks The applicable loads and their combinations, which must be considered in the seismic analysis of rack modules, are excerpted from the OT Position [6.1.3] and SRP, Section 3.8.4 [6.1.2]. The load combinations considered are identified below:

Loading Combination Service Level D+L Level A D+L+TO D+L+To+E D+L+Ta+E Level B D+L + T,+Pf D+L+Ta+E' Level D D+L+To+Fd The functional capability of the fuel racks must be demonstrated. This load case is discussed in Section 7.0.

Where:

D = Dead weight-induced loads (including fuel assembly weight)

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

Pf = Upward force on the racks caused by postulated stuck fuel assembly Fd = Impact force from accidental drop of the heaviest load from the maximum possible height.

E = Operating Basis Earthquake (OBE)

E' = Safe Shutdown Earthquake (SSE)

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

Ta = Differential temperature induced loads (the highest temperature associated with the postulated abnormal design conditions)

Holtec Report HI-2033124 6-21 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Ta and T. 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 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.

6.7 Parametric Simulations The multiple rack models employ the fluid coupling effects for all racks in the pool, as discussed above, and these simulations are referred to as WPMR evaluations. In addition, single rack models are also developed for additional study of the effect of various parameters on rack displacement. The models are described as follows:

( I) Whole Pool Multi Rack Model Three models are develop for WPMR analysis. For the phase 1 case, two racks and the fuel transfer cask are present in the Fuel Cask Storage Pool. For phase 2, an array of twenty-six racks in the SFP and three racks in the Fuel Cask Storage Pool are modeled. In these cases, proper interface fluid gaps and a coefficient of friction at the support interface locations with the bearing pad generated by a Gaussian distribution random number generator with 0.5 as the mean and 0.15 standard deviation are implemented. The response to both SSE and OBE seismic excitation is determined.

(II ) Single Rack Models: Two models are employed for studying the structural behavior of a single rack. A model is developed for the largest rack and another for the rack with the maximum aspect ratio (defined as the rack exhibiting the maximum ratio of the height to the smaller of the length or width). In both these models, the rack is modeled as fully loaded (to act as a baseline),

half loaded (east-west, north-south and diagonally) and nearly empty. The coefficient of friction between male pedestal and bearing pad is taken as one of four possibilities: 0.2, 0.5, 0.8 or as selected by a Gaussian random number generator as introduced in the prior section. For these models, either in phase or opposed phase motion is assumed. The in phase case is implemented by assuming that the maximum actual water gaps that exist between the racks and the four walls of the SFP surround the single rack, in the same north-east-south-west orientation. This reflects the behavior that would occur if all the racks moved in unison. For the opposed phase case, one-half Holtec Report HI-2033124 6-22 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

the actual gap is attributed to each rack side. All single rack cases in the study are done for both SSE and OBE excitation.

( III ) Single Rack Overturning Check Model This model is developed to study the potential for rack overturning. The SSE case which had the maximum displacement in the study is run, subjected to 1.5 times the SSE excitation and the OBE case which had the maximum displacement is run, subjected to 1.5 times the OBE excitation.

The Whole Pool and Single Rack simulations listed on the following tables have been performed to investigate the structural integrity of the rack arrays.

LIST OF WPMR SIMULATIONS Case Load Case COF Event I Fuel Cask Storage Pool Phase 1 All Racks Random SSE 2 Fuel Cask Storage Pool Phase 1 All Racks Random OBE 3 SFP - All Racks Fully Loaded Random SSE 4 SFP - All Racks Fully Loaded Random OBE 5 Fuel Cask Storage Pool Phase 2 All Racks Random SSE 6 Fuel Cask Storage Pool Phase 2 All Racks Random OBE LIST OF SINGLE RACK SIMULATIONS Case Motion Load Case COF Event 1 IN PHASE Largest Rack Fully Loaded Random SSE 2 IN PHASE Largest Rack Fully Loaded 0.2 SSE 3 IN PHASE Largest Rack Fully Loaded 0.5 SSE 4 IN PHASE Largest Rack Fully Loaded 0.8 SSE 5 IN PHASE Largest Rack Half Loaded (E-W) Random SSE 6 IN PHASE Largest Rack Half Loaded (E-W) 0.2 SSE 7 IN PHASE Largest Rack Half Loaded (E-W) 0.5 SSE 8 IN PHASE Largest Rack Half Loaded (E-W) 0.8 SSE 9 IN PHASE Largest Rack Half Loaded (N-S) Random SSE 10 IN PHASE Largest Rack Half Loaded (N-S) 0.2 SSE Holtec Report H1-2033124 6-23 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

LIST OF SINGLE RACK SIMULATIONS Case Motion Load Case COF Event 11 IN PHASE Largest Rack Half Loaded (N-S) 0.5 SSE 12 IN PHASE Largest Rack Half Loaded (N-S) 0.8 SSE 13 IN PHASE Largest Rack Half Loaded (Diag) Random SSE 14 IN PHASE Largest Rack Half Loaded (Diag) 0.2 SSE 15 IN PHASE Largest Rack Half Loaded (Diag) 0.5 SSE 16 IN PHASE Largest Rack Half Loaded (Diag) 0.8 SSE 17 IN PHASE Largest Rack Nearly Empty Random SSE 18 IN PHASE Largest Rack Nearly Empty 0.2 SSE 19 IN PHASE Largest Rack Nearly Empty 0.5 SSE 20 IN PHASE Largest Rack Nearly Empty 0.8 SSE 21 OPPOSED Largest Rack Fully Loaded Random SSE 22 OPPOSED Largest Rack Fully Loaded 0.2 SSE 23 OPPOSED Largest Rack Fully Loaded 0.5 SSE 24 OPPOSED Largest Rack Fully Loaded 0.8 SSE 25 OPPOSED Largest Rack Half Loaded (E-W) Random SSE 26 OPPOSED Largest Rack Half Loaded (E-W) 0.2 SSE 27 OPPOSED Largest Rack Half Loaded (E-W) 0.5 SSE 28 OPPOSED Largest Rack Half Loaded (E-W) 0.8 SSE 29 OPPOSED Largest Rack Half Loaded (N-S) Random SSE 30 OPPOSED Largest Rack Half Loaded (N-S) 0.2 SSE 31 OPPOSED Largest Rack Half Loaded (N-S) 0.5 SSE 32 OPPOSED Largest Rack Half Loaded (N-S) 0.8 SSE 33 OPPOSED Largest Rack Half Loaded (Diag) Random SSE 34 OPPOSED Largest Rack Half Loaded (Diag) 0.2 SSE 35 OPPOSED Largest Rack Half Loaded (Diag) 0.5 SSE 36 OPPOSED Largest Rack Half Loaded (Diag) 0.8 SSE 37 OPPOSED Largest Rack Nearly Empty Random SSE 38 OPPOSED Largest Rack Nearly Empty 0.2 SSE 39 OPPOSED Largest Rack Nearly Empty 0.5 SSE Holtec Report HI-2033124 6-24 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

LIST OF SINGLE RACK SIMULATIONS Case Motion Load Case COF Event 40 OPPOSED Largest Rack Nearly Empty 0.8 SSE 41 IN PHASE Largest Rack Fully Loaded Random OBE 42 IN PHASE Largest Rack Fully Loaded 0.2 OBE 43 IN PHASE Largest Rack Fully Loaded 0.5 OBE 44 IN PHASE Largest Rack Fully Loaded 0.8 OBE 45 IN PHASE Largest Rack Half Loaded (E-W) Random OBE 46 IN PHASE Largest Rack Half Loaded (E-W) 0.2 OBE 47 IN PHASE Largest Rack Half Loaded (E-W) 0.5 OBE 48 IN PHASE Largest Rack Half Loaded (E-W) 0.8 OBE 49 IN PHASE Largest Rack Half Loaded (N-S) Random OBE 50 IN PHASE Largest Rack Half Loaded (N-S) 0.2 OBE 51 IN PHASE Largest Rack Half Loaded (N-S) 0.5 OBE 52 IN PHASE Largest Rack Half Loaded (N-S) 0.8 OBE 53 IN PHASE Largest Rack Half Loaded (Diag) Random OBE 54 IN PHASE Largest Rack Half Loaded (Diag) 0.2 OBE 55 IN PHASE Largest Rack Half Loaded (Diag) 0.5 OBE 56 IN PHASE Largest Rack Half Loaded (Diag) 0.8 OBE 57 IN PHASE Largest Rack Nearly Empty Random OBE 58 IN PHASE Largest Rack Nearly Empty 0.2 OBE 59 IN PHASE Largest Rack Nearly Empty 0.5 OBE 60 IN PHASE Largest Rack Nearly Empty 0.8 OBE 61 OPPOSED Largest Rack Fully Loaded Random OBE 62 OPPOSED Largest Rack Fully Loaded 0.2 OBE 63 OPPOSED Largest Rack Fully Loaded 0.5 OBE 64 OPPOSED Largest Rack Fully Loaded 0.8 OBE 65 OPPOSED Largest Rack Half Loaded (E-W) Random OBE 66 OPPOSED Largest Rack Half Loaded (E-W) 0.2 OBE 67 OPPOSED Largest Rack Half Loaded (E-W) 0.5 OBE 68 OPPOSED Largest Rack Half Loaded (E-W) 0.8 OBE Holtec Report HI-2033124 6-25 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

LIST OF SINGLE RACK SIMULATIONS Case Motion Load Case COF Event 69 OPPOSED Largest Rack Half Loaded (N-S) Random OBE 70 OPPOSED Largest Rack Half Loaded (N-S) 0.2 OBE 71 OPPOSED Largest Rack Half Loaded (N-S) 0.5 OBE 72 OPPOSED Largest Rack Half Loaded (N-S) 0.8 OBE 73 OPPOSED Largest Rack Half Loaded (Diag) Random OBE 74 OPPOSED Largest Rack Half Loaded (Diag) 0.2 OBE 75 OPPOSED Largest Rack Half Loaded (Diag) 0.5 OBE 76 OPPOSED Largest Rack Half Loaded (Diag) 0.8 OBE 77 OPPOSED Largest Rack Nearly Empty Random OBE 78 OPPOSED Largest Rack Nearly Empty 0.2 OBE 79 OPPOSED Largest Rack Nearly Empty 0.5 OBE 80 OPPOSED Largest Rack Nearly Empty 0.8 OBE 81 IN PHASE Aspect Rack Fully Loaded Random SSE 82 IN PHASE Aspect Rack Fully Loaded 0.2 SSE 83 IN PHASE Aspect Rack Fully Loaded 0.5 SSE 84 IN PHASE Aspect Rack Fully Loaded 0.8 SSE 85 IN PHASE Aspect Rack Half Loaded (E-W) Random SSE 86 IN PHASE Aspect Rack Half Loaded (E-W) 0.2 SSE 87 IN PHASE Aspect Rack Half Loaded (E-W) 0.5 SSE 88 IN PHASE Aspect Rack Half Loaded (E-W) 0.8 SSE 89 IN PHASE Aspect Rack Half Loaded (N-S) Random SSE 90 IN PHASE Aspect Rack Half Loaded (N-S) 0.2 SSE 91 IN PHASE Aspect Rack Half Loaded (N-S) 0.5 SSE 92 IN PHASE Aspect Rack Half Loaded (N-S) 0.8 SSE 93 IN PHASE Aspect Rack Half Loaded (Diag) Random SSE 94 IN PHASE Aspect Rack Half Loaded (Diag) 0.2 SSE 95 IN PHASE Aspect Rack Half Loaded (Diag) 0.5 SSE 96 IN PHASE Aspect Rack Half Loaded (Diag) 0.8 SSE 97 IN PHASE Aspect Rack Nearly Empty Random SSE Holtec Report HI-2033124 6-26 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

LIST OF SINGLE RACK SIMULATIONS Case Motion Load Case COF Event 98 IN PHASE Aspect Rack Nearly Empty 0.2 SSE 99 IN PHASE Aspect Rack Nearly Empty 0.5 SSE 100 IN PHASE Aspect Rack Nearly Empty 0.8 SSE 101 OPPOSED Aspect Rack Fully Loaded Random SSE 102 OPPOSED Aspect Rack Fully Loaded 0.2 SSE 103 OPPOSED Aspect Rack Fully Loaded 0.5 SSE 104 OPPOSED Aspect Rack Fully Loaded 0.8 SSE 105 OPPOSED Aspect Rack Half Loaded (E-W) Random SSE 106 OPPOSED Aspect Rack Half Loaded (E-W) 0.2 SSE 107 OPPOSED Aspect Rack Half Loaded (E-W) 0.5 SSE 108 OPPOSED Aspect Rack Half Loaded (E-W) 0.8 SSE 109 OPPOSED Aspect Rack Half Loaded (N-S) Random SSE 110 OPPOSED Aspect Rack Half Loaded (N-S) 0.2 SSE 111 OPPOSED Aspect Rack Half Loaded (N-S) 0.5 SSE 112 OPPOSED Aspect Rack Half Loaded (N-S) 0.8 SSE 113 OPPOSED Aspect Rack Half Loaded (Diag) Random SSE 114 OPPOSED Aspect Rack Half Loaded (Diag) 0.2 SSE 115 OPPOSED Aspect Rack Half Loaded (Diag) 0.5 SSE 116 OPPOSED Aspect Rack Half Loaded (Diag) 0.8 SSE 117 OPPOSED Aspect Rack Nearly Empty Random SSE 118 OPPOSED Aspect Rack Nearly Empty 0.2 SSE 119 OPPOSED Aspect Rack Nearly Empty 0.5 SSE 120 OPPOSED Aspect Rack Nearly Empty 0.8 SSE 121 IN PHASE Aspect Rack Fully Loaded Random OBE 122 IN PHASE Aspect Rack Fully Loaded 0.2 OBE 123 IN PHASE Aspect Rack Fully Loaded 0.5 OBE 124 IN PHASE Aspect Rack Fully Loaded 0.8 OBE 125 IN PHASE Aspect Rack Half Loaded (E-W) Random OBE 126 IN PHASE Aspect Rack Half Loaded (E-W) 0.2 OBE Holtec Report HI-2033124 6-27 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

LIST OF SINGLE RACK SIMULATIONS Case Motion Load Case COF Event 127 IN PHASE Aspect Rack Half Loaded (E-W) 0.5 OBE 128 IN PHASE Aspect Rack Half Loaded (E-W) 0.8 OBE 129 IN PHASE Aspect Rack Half Loaded (N-S) Random OBE 130 IN PHASE Aspect Rack Half Loaded (N-S) 0.2 OBE 131 IN PHASE Aspect Rack Half Loaded (N-S) 0.5 OBE 132 IN PHASE Aspect Rack Half Loaded (N-S) 0.8 OBE 133 IN PHASE Aspect Rack Half Loaded (Diag) Random OBE 134 IN PHASE Aspect Rack Half Loaded (Diag) 0.2 OBE 135 IN PHASE Aspect Rack Half Loaded (Diag) 0.5 OBE 136 IN PHASE Aspect Rack Half Loaded (Diag) 0.8 OBE 137 IN PHASE Aspect Rack Nearly Empty Random OBE 138 IN PHASE Aspect Rack Nearly Empty 0.2 OBE 139 IN PHASE Aspect Rack Nearly Empty 0.5 OBE 140 IN PHASE Aspect Rack Nearly Empty 0.8 OBE 141 OPPOSED Aspect Rack Fully Loaded Random OBE 142 OPPOSED Aspect Rack Fully Loaded 0.2 OBE 143 OPPOSED Aspect Rack Fully Loaded 0.5 OBE 144 OPPOSED Aspect Rack Fully Loaded 0.8 OBE 145 OPPOSED Aspect Rack Half Loaded (E-W) Random OBE 146 OPPOSED Aspect Rack Half Loaded (E-W) 0.2 OBE 147 OPPOSED Aspect Rack Half Loaded (E-W) 0.5 OBE 148 OPPOSED Aspect Rack Half Loaded (E-W) 0.8 OBE 149 OPPOSED Aspect Rack Half Loaded (N-S) Random OBE 150 OPPOSED Aspect Rack Half Loaded (N-S) 0.2 OBE 151 OPPOSED Aspect Rack Half Loaded (N-S) 0.5 OBE 152 OPPOSED Aspect Rack Half Loaded (N-S) 0.8 OBE 153 OPPOSED Aspect Rack Half Loaded (Diag) Random OBE 154 OPPOSED Aspect Rack Half Loaded (Diag) 0.2 OBE 155 OPPOSED Aspect Rack Half Loaded (Diag) 0.5 OBE Holtec Report HI-2033124 6-28 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

LIST OF SINGLE RACK SIMULATIONS Case Motion Load Case COF Event 156 OPPOSED Aspect Rack Half Loaded (Diag) 0.8 OBE 157 OPPOSED Aspect Rack Nearly Empty Random OBE 158 OPPOSED Aspect Rack Nearly Empty 0.2 OBE 159 OPPOSED Aspect Rack Nearly Empty 0.5 OBE 160 OPPOSED Aspect Rack Nearly Empty 0.8 OBE 161 IN PHASE Largest Rack Half Loaded (Diag) 0.5 l.lxSSE 162 IN PHASE Largest Rack Fully Loaded 0.2 1.5xOBE where Random = Gaussian distribution with a mean coeff. of friction of 0.5.

(upper and lower limits of 0.8 and 0.2, respectively) and COF = Coefficient of Friction Holtec Report HI-2033124 6-29 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.8 Time History Simulation Results The results from the DYNARACK runs may be seen in the raw data output files. However, due to the huge quantity of output data, a post-processor is used to scan for worst case conditions and develop the stress factors discussed in subsection 6.6.3. Further reduction in this bulk of information is provided in this section by extracting the worst case values from the parameters of interest; namely displacements, support pedestal forces, impact loads, and stress factors. This section also summarizes additional analyses performed to develop and evaluate structural member stresses which are not determined by the post processor.

6.8.1 Rack Displacements The maximum rack displacements are obtained from the time histories of the motion of the upper and lower four corners of each rack in each of the simulations. The maximum absolute value of displacement in the two horizontal directions, relative to the pool slab, is determined by the post-processor for each rack, at the top and bottom corners. The maximum displacement in either direction reported from the WPMR analyses occurred at the top of module B4 in the Phase 2 Fuel Cask Storage Pool configuration. The maximum displacement is 2.276 inches for the SSE scenario and 0.336 inches for the OBE scenario. The maximum displacement in either direction reported from the single rack analyses is 0.8686" from simulation 104, which was performed for module Fl.

To assess the kinematic stability safety margin, the maximum displacement single rack cases were run again, using 1.5 times the SSE excitation and 1.5 times the OBE excitation, respectively. These are single rack cases 161 and 162. The maximum displacements from these runs was still less than the displacement results from the WPMR runs. Therefore, the bounding displacement result of 2.276 inches is obtained from the WPMR scenario. The result for module B4 from WPMR case 5 is used to compute the safety factor against overturning. It was shown to be more than 22, which far exceeds the acceptance criteria of 1.0.

Holtec Report HI-2033124 6-30 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.8.2 Pedestal Vertical Forces The maximum vertical pedestal force obtained in the WPMR simulations was 138,000 Ibf for module G9, one of the 15 x 12 racks in the Phase I Fuel Cask Storage Pool SSE simulation. The maximum vertical pedestal force obtained in the OBE simulation was 81,200 Ibf for module GI, another 15 x 12 rack, in the SFP simulation.

6.8.3 Pedestal Friction Forces The maximum interface shear force value in any direction bounding all pedestals in all simulations is 51,500 lbf for module G8 in the Phase I Fuel Cask Storage Pool SSE case.

6.8.4 Rack Impact Loads A freestanding rack, by definition, is a structure subject to potential impacts during a seismic event. Impacts arise, in some instances, from localized impacts between the racks, or between a peripheral rack and the pool wall and from rattling of the fuel assemblies in the storage rack locations. The following sections discuss the bounding values of these impact loads.

6.8.4.1 Impacts External to the Rack Gap elements track the potential for rack-to-rack or rack-to-pool wall impacts. The result of each gap spring element in terms of impact force is printed in a list format in the output file produced by the DYNARACK program. The lists from each simulation are scanned for non-zero values.

A non-zero value indicates an impact and vice versa. Naturally impacts are expected between racks which are initially in contact at the baseplates. These include all new racks which are designed for installation in contact with a neighboring new rack and for impact loads resulting from seismic events. Impacts also occur at the top elevation as well. The maximum rack impact is between modules A4 and B4 (both existing modules) in the Fuel Cask Storage Pool Phase 2 case, with impact of 35,160 1bf at the top of the north end of their interface and 27,250 1bf at the south. The impact site is at the top of the rack which would be above the tops of any stored fuel assemblies. It has been determined that there is no permanent deformation of the rack material Holtec Report HI-2033124 6-31 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

from this impact, with the rack returning to its normal configuration when the impact load is removed. Fuel configuration and poison areas remain unaffected. Therefore these impacts are acceptable.

6.8.4.2 Impacts Internal to the Rack A review of all simulations performed allows determination of the maximum instantaneous impact load between fuel assembly and fuel cell wall at any modeled impact site. The maximum fuel/cell wall impact loads occur in the SFP WPMR analyses in module D3 and are 815 lbf in the SSE case and 380 lbf for the OBE case. The cell wall integrity under the 815 lbf impact load has been evaluated and shown to remain intact with no permanent damage.

The permissible lateral load on an irradiated spent fuel assembly has been studied by the Lawrence Livermore National Laboratory. The LLNL report [6.8.1] states that "...for the most vulnerable fuel assembly, axial buckling varies from 82g's at initial storage to 95g's after 20 years' storage. In a side drop, no yielding is expected below 63g's at initial storage to 74g's after 20 years' [dry] storage". The most significant load on the fuel assembly arises from rattling during the seismic event.

X-, ',

a permissible lateral acceleration in g's (a = 63)

Therefore, the limiting lateral load, Fe= 12,600 lbs The maximum fuel-to-storage cell rattling force from the WPMR runs is 815 lbs. Therefore, the nominal factor of safety against fuel failure is computed to be more than 15.

Holtec Report HI-2033124 6-32 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.9 Rack Structural Evaluation 6.9.1 Rack Stress Factors The time history results from the DYNARACK solver provide the pedestal normal and lateral interface forces, which may be converted to the limiting bending moment and shear force at the bottom baseplate-pedestal interface. In particular, maximum values for the previously defined stress factors are determined for every pedestal in the array of racks. With this information available, the structural integrity of the pedestal can be assessed and reported. The net section maximum (in time) bending moments and shear forces can also be determined at the bottom baseplate-rack cellular structure interface for each spent fuel rack in the pool. Using these forces and moments, the maximum stress in the limiting rack cell (box) can be evaluated.

The stress factor results for male and female pedestals, and for the entire spent fuel rack cellular cross-section just above the bottom casting has been determined. These factors are reported for every rack in each simulation, and for each pedestal in every rack. These locations are the most heavily loaded net sections in the structure so that satisfaction of the stress factor criteria at these locations ensures that the overall structural criteria set forth in Section 6.6 are met.

An evaluation of the stress factors for all of the WPMR simulations performed leads to the conclusion that all stress factors, as defined in Section 6.6.3, are less than the mandated limit of 1.0 for the load cases examined. All of the maximum stress factors occurred at the base of the rack cell walls under the WPMR scenarios. For the new racks, the bounding stress factor was determined to be 0.243 (R6) for the Phase 2 SFP OBE simulation, occurring in module H. The maximum calculated SSE stress factor for the new racks occurred during the Phase 2 SFP simulation and was 0.167 (R6) for module Jl. For the existing racks, the bounding stress factor was determined to be 0.303 (R6) for the Phase 2 Fuel Cask Storage Pool OBE simulation, occurring in the cell region of module A4. The maximum calculated SSE stress factor for the existing racks was 0.210 (R6) for module B4. Relevant stress factors are for cell wall stresses above the baseplate, since these control over the pedestal stress factors. The values for all other Holtec Report HI-2033124 6-33 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

defined stress factors are archived and show that the requirements of Section 6.6 are indeed satisfied for the load levels considered for every limiting location in every rack in the array.

6.9.2 Pedestal Thread Shear Stress The maximum thread engagement stresses under faulted conditions for every pedestal for every rack in the WPMR simulations run was 8,268 psi for the Fuel Cask Storage Pool Phase 1 SSE run and 4,865 psi for the SFP OBE run. By ASME code section NF-3321, the Level A allowable stress is 0.4*Fy = 0.4(25,000) = 10,000 psi. Referring to section 6.6.4, for Level B (OBE), the allowable is increased by the factor 1.33 from table NF-3523(b), resulting in an allowable stress of 13,300 psi, which exceeds both calculated stresses.

6.9.3 Local Stresses Due to Impacts Impact loads at the pedestal base produce stresses in the pedestal for which explicit stress limits are prescribed in the Code. However, impact loads on the cellular region of the racks, as discussed in subsection 6.8.4.2 above, produce stresses which attenuate rapidly away from the loaded region. This behavior is characteristic of secondary stresses.

Even though limits on secondary stresses are not prescribed in the Code for class 3 NF structures, evaluations are made to ensure that the localized impacts do not lead to plastic deformations in the storage cells which affect the sub-criticality of the stored fuel array.

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 which would lead to gross permanent deformation. As shown in Table 6.9.1, the limiting impact load (of 2,826 lbf, including a safety factor of 2.0) is much greater than the highest calculated impact load value (of 815 Ibf, see subsection 6.8.4.2) obtained from any of the rack analyses. Therefore, fuel impacts do not represent a significant concern with respect to fuel rack cell deformation.

Holtec Report HI-2033124 6-34 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

b. Impacts Between Adiacent Racks As may be seen from subsection 6.8.4.1, the storage racks will impact each other at a few locations during seismic events. Since the loading produces distributed stresses shown to be less than the yield stress, local deformation will be negligible. The impact loading will be distributed over a large area (a significant portion of the entire face width of the rack of about 58 inches). The resulting compressive stress from the highest combined impact load of 62,410 lbs distributed over 10 cell walls is 17,336 psi, which is less than the cell wall material yield strength of 21,300 psi. Therefore, any deformation will not affect the configuration of the stored fuel.

6.9.4 Assessment of Rack Fatigue Margin Deeply submerged high density spent fuel storage racks arrayed in close proximity to each other in a free-standing configuration behave primarily as a nonlinear cantilevered structure when subjected to 3-D seismic excitations. In addition to the pulsations in the vertical load at each pedestal, lateral friction forces at the pedestal/ bearing pad interface, which help prevent or mitigate lateral sliding of the rack, also exert a time-varying moment in the baseplate region of the rack. The friction-induced lateral forces act simultaneously in x and y directions with the requirement that their vectorial sum does not exceed [tV, where p. is the limiting interface coefficient of friction and V is the concomitant vertical thrust on the bearing pad (at the given time instant). As the vertical thrust at a pedestal location changes, so does the maximum friction force, F, that the interface can exert. In other words, the lateral force at the pedestal/bearing pad interface, F, is given by F < p N (r) where N (vertical thrust) is the time-varying function of '. F does not always equal AIN; rather, MN is the maximum value it can attain at any time; the actual value, of course, is determined by the dynamic equilibrium of the rack structure.

Holtec Report HI-2033124 6-35 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

In summary, the horizontal friction force at the pedestal/bearing pad interface is a function of time; its magnitude and direction of action varies during the earthquake event.

The time-varying lateral (horizontal) and vertical forces on the extremities of the support pedestals produce stresses at the root of the pedestals in the manner of an end-loaded cantilever.

The stress field in the cellular region of the rack is quite complex, with its maximum values located in the region closest to the pedestal. The maximum magnitude of the stresses depends on the severity of the pedestal end loads and on the geometry of the pedestal/rack baseplate region.

Alternating stresses in metals produce metal fatigue if the amplitude of the stress cycles is sufficiently large. In high density racks designed for sites with moderate to high postulated seismic action, the stress intensity amplitudes frequently reach values above the material endurance limit, leading to expenditure of the fatigue "usage" reserve in the material.

Because the locations of maximum stress (viz., the pedestal/rack baseplate junction) and the close placement of racks, a post-earthquake inspection of the high stressed regions in the racks is not feasible. Therefore, the racks must be engineered to withstand multiple earthquakes without reliance of nondestructive inspections for post-earthquake integrity assessment. The fatigue life evaluation of racks is an integral aspect of a sound design.

The time-history method of analysis, deployed in this report, provides the means to obtain a complete cycle history of the stress intensities in the highly stressed regions of the rack. Having determined the amplitude of the stress intensity cycles and their number, the cumulative damage factor, U, can be determined using the classical Miner's rule:

U=_ ni Ni Holtec Report H1-2033124 6-36 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

where ni is the number of stress intensity cycles of amplitude a2 , and N2 is the permissible number of cycles corresponding to a1 from the ASME fatigue curve for the material of construction. U must be less than or equal to 1.0.

To evaluate the cumulative damage factor, a finite element model of a portion of the spent fuel rack in the vicinity of a support pedestal is constructed in sufficient detail to provide an accurate assessment of stress intensities. The finite element solutions for unit pedestal loads in three orthogonal directions are combined to establish the maximum value of stress intensity as a function of the three unit pedestal loads. Using the archived results of the spent fuel rack dynamic analyses (pedestal load histories versus time) enables a time-history of stress intensity to be established at the most limiting location. This permits establishing a set of alternating stress intensity ranges versus cycles. Following ASME Code guidelines for computing U, it is found that U =0.104 due to the combined effects of one SSE and twenty OBE events. This is well below the ASME Code limit of 1.0.

6.9.5 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. Bounding values of resultant loads are used to qualify the connections.

Holtec Report HI-2033124 6-37 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

a. Baseplate-to-Rack Cell Welds For Level A or B conditions, Ref. [6.6.1] permits an allowable weld stress oft = .3 S" =

19,860 psi (multiplied by 1.33 for Level B). As stated in subsection 6.6.2, the allowable may be increased for Level D by an amplification factor which is equal to 1.8 (=

.72SY/.4Sy). The allowable stress increase factor of 1.8 greatly exceeds the ratio of maximum SSE to OBE stresses. Therefore, Level B becomes the governing condition.

Weld dimensionless stress factors are produced through the use of a simple conversion (ratio) factor applied to the corresponding stress factor in the adjacent rack material.

Addressing the new racks, the RATIO, 2.079, is developed from the differences in material thickness and length versus weld throat dimension and length:

RATIO = I =

The highest predicted weld stress for OBE is calculated from the highest cell wall (above the baseplate) R6 value, 0.243, (corresponding to the same simulation as reported in subsection 6.9.1) as follows:

R6 * [(0.6) Fy]

• RATIO = 0.243 * [0.6

• 21300]

• 2.079 = 6,456 psi This value is less than the Level B allowable weld stress value, which is 1.33 x 19,860 26,414 psi. Therefore, all weld stresses between the baseplate and cell wall base are acceptable.

b. Baseplate-to-Pedestal Welds The weld between baseplate and support pedestal is checked using finite element analysis to determine that the maximum stress is 8,767 psi under a Level D event. This calculated stress value can be shown to be acceptable by comparing to the normal stress allowable of 19,860 psi.

Holtec Report HI-2033124 6-38 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

c. Cell-to-Cell Welds Cell-to-cell connections are by a series of connecting welds along the cell height.

Stresses in storage cell to cell welds develop due to fuel assembly impacts with the cell wall. These weld stresses are conservatively calculated by assuming that fuel assemblies in adjacent cells are moving out of phase with one another so that impact loads in two adjacent cells are in opposite directions; this tends to separate the two cells from each other at the weld.

Table 6.9.1 gives the computed results for the maximum allowable load that can be transferred by these welds based on the available weld area. The upper bound on the applied load transferred is also given in Table 6.9.1. This upper bound value is conservatively obtained by applying the bounding rack-to-fuel impact load from any simulation in two orthogonal directions simultaneously, and multiplying the result by 2 to account for the simultaneous impact of two assemblies in adjacent cells moving in opposing directions. An equilibrium analysis at the connection then yields the upper bound load to be transferred. As shown in Table 6.9.1, the calculated stress of 4,031 psi is below the allowable stress of 8,520 psi.

6.9.6 Bearing Pad Analysis To protect the pool slab from highly localized dynamic loadings, bearing pads are placed between the pedestal base and the slab. Fuel rack pedestals impact on these bearing pads during a seismic event and pedestal loading is transferred to the liner. Bearing pad dimensions are set to ensure that the average pressure on the slab surface due to a static load plus a dynamic impact load does not exceed the American Concrete Institute, ACI-349 [6.9.1] limit on bearing pressures. Section 10.17 of [6.9.2] gives the design bearing strength as fb = ¢ (.85 fe') E Holtec Report HI-2033124 6-39 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

where 4 = .7 and f,' is the specified concrete strength for the spent fuel pool. E = I except when the supporting surface is wider on all sides than the loaded area. In that case, E = (A2/A,) 5, but not more than 2. Al is the actual loaded area, and A2 is an area greater than Al and is defined in

[6.9.2]. Using a value of E > 2 includes credit for the confining effect of the surrounding concrete. It is noted that this criterion is in conformance with the ultimate strength primary design methodology of the American Concrete Institute in use since 1971. For CPS, f,' = 3,500 psi and the allowable static bearing pressure is fib = 4,165 psi, assuming full concrete confinement. This allowable bearing pressure is utilized because concrete confinement is not compromised in the leak chase region due to the large slab dimensions in both lateral and thickness directions (therefore the supporting area A2 below the slab surface is much larger than the loaded area AI at the bearing pad/slab interface).. The primary objective of the bearing pad analysis is to show that this primarily compressive component remains in the elastic range.

The analyses are performed with ANSYS using finite element models, which place a bearing pad over two perpendicular leak chases. For conservatism the pedestal is centered at the intersection of these two leak chases that produces maximum stress. This configuration is selected with the intent of bounding all other possible bearing pad/pool floor interfaces. The analysis applies the maximum total vertical pedestal load from results for all bearing pads, scanned from the time-history solution from the SSE simulation. The maximum vertical pedestal load over a leak chase (which is modeled to remove a 2" wide strip of concrete from under the bearing pad) is taken to be 140 kips on a 12" x 12" bearing pad.

The bearing pads in the SFP will be 1.5" thick. All bearing pads will be made from austenitic stainless steel plate stock. Bearing pad models were prepared to evaluate all possible configurations. Figure 6.9.1 provides an isometric of the controlling ANSYS finite element model (leak chase condition). The model permits the bearing pad to deform and lose contact with the liner, if the conditions of elastostatics so dictate. Figure 6.9.1 shows the bearing pad and underlying leak chases located within the supporting concrete. The slab is modeled as an elastic foundation. The average pressure at the pad to liner interface is computed and compared against the above-mentioned limit. Calculations show that the average pressure at the slab / liner Holtec Report HI-2033124 6-40 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

interface is 3,500 psi, which is well below the allowable value of 4,165 psi, providing a factor of safety of 1.19. The stress distribution in the bearing pad is also evaluated, with the results shown in Figure 6.9.2 (top and bottom views). The peak stress in the bearing pad during a Level D event is 27,964 psi. ASME Section 1II, Appendix F, Section 1334.10 states that bearing stresses need not be evaluated for Level D conditions. Nevertheless, an appropriate stress limit may be considered by maintaining the increase factor of two above normal condition allowables, as discussed earlier. The material yield strength of 25,000 psi at 200'F then provides an allowable stress of 2*0.9Sy (i.e., 45,000 psi) producing a factor of safety against yield of about 1.61. A similar evaluation performed for Level B conditions shows that the Level D controls. Therefore, the bearing pad design devised for the CPS SFP is deemed appropriate for the prescribed loadings.

6.10 Level A Evaluation The Level A condition is not a governing condition for spent fuel racks since the general level of imposed loading is far less than Level B or D loading. The stress allowable for Level B loading is only approximately 1/3 greater than the corresponding Level A stress allowable. The ratio of the loading increase from Level A to B loading far exceeds this 1/3 value. Therefore Level A is acceptable by comparison.

6.11 Hvdrodvnamic Loads on Pool Walls The hydrodynamic pressures that develop between adjacent racks and the pool walls can be developed from the archived results produced by the WPMR analysis. Of the racks next to the SFP walls, the one that resulted in the maximum displacement generates the maximum hydrodynamic load on its adjacent wall. Time dependent hydrodynamic pressures are determined for subsequent analysis as discussed in Section 8.0.

Holtec Report HI-2033124 6-41 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.12 Local Stress Considerations This section presents the results of evaluations for the possibility of cell wall buckling and the secondary stresses produced by temperature effects.

6.12.1 Cell Wall Buckling The allowable local buckling stresses in the fuel cell walls are obtained by using classical plate buckling analysis using a model as shown in Figure 6.12.1. The evaluation for cell wall buckling is based on the applied stress being uniform along the entire length of the cell wall. In the actual fuel rack, the compressive stress comes from consideration of overall bending of the rack structures during a seismic event, and as such is negligible at the rack top, and maximum at the rack bottom.

The critical buckling stress is determined to be 22,580 psi. The computed compressive stress in the cell wall, based on the R5 stress factor, is 4,090 psi. Therefore, there is a large margin of safety against local cell wall buckling.

6.12.2 Analysis of Welded Joints in the Racks Cell-to-cell welded joints are examined under the loading conditions arising from thermal effects due to an isolated hot cell in this subsection. This secondary stress condition is evaluated alone and not combined with primary stresses from other load conditions.

A thermal gradient between cells will develop when an isolated storage location contains a fuel assembly emitting maximum postulated heat, while surrounding locations are empty. We obtain a conservative estimate of weld stresses along the length of an isolated hot cell by considering a beam strip uniformly heated by 750 F, and restrained from growth along one long edge. This temperature rise is based on thermal-hydraulic evaluations discussed in Section 5.0, which show that a conservative upper bound for the difference between local cell maximum temperatures and Holtec Report HI-2033124 6-42 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

the bulk temperature in the pool is less than this value. The analyzed configuration is shown in Figure 6.12.2.

Using shear beam theory, as discussed in Holtec generic calculation HI-89330 [6.9.3], and subjecting the strip to a uniform temperature rise AT = 750 F, we can calculate an estimate of the maximum value of the average shear stress in the strip. The strip is subjected to the following boundary conditions.

a. Displacement U,, (x,y) = 0 at x = 0, at y = H, all x.

b. Average force M., acting on the cross section Ht = 0 at x = 1, all y.

The final result for wall shear stress, maximum at x = 1,is found to be given as EaAT Tmax =0.931 where E = 27.6 x 106 psi, ax = 9.5 x 10 6 in/in 'F and AT = 75°F.

Therefore, we obtain an estimate of maximum weld shear stress in an isolated hot cell, due to thermal gradient, as

¶ max = 21,122 psi Since this is a secondary thermal stress, we use the allowable shear stress criteria for faulted conditions (0.42*Su=27,804 psi) as a guide to indicate that this maximum shear is acceptable.

Therefore, there is a margin of safety of 24% against cell wall shear failure due to secondary thermal stresses from cell wall growth under the worst case hot cell conditions.

Holtec Report HI-2033124 643 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

6.13 References

[6.1.1] USNRC NUREG-0800, Standard Review Plan, June 1987.

[6.1.2] (USNRC Office of Technology) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14, 1978, and January 18, 1979 amendment thereto.

[6.2.1] 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).

[6.2.2] 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.

[6.2.3] 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] Holtec Proprietary Report HI-961465 - WPMR Analysis User Manual for Pre&Post Processors & Solver, August, 1997.

[6.4.1] Holtec Proprietary Report HI-89364 - Verification and User's Manual for Computer Code GENEQ, January, 1990.

[6.5.1] Rabinowicz, E., "Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," MIT, a report for Boston Edison Company, 1976.

[6.5.2] 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.

Holtec Report HI-2033124 6-44 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

[6.5.3] 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.

[6.5.4] Levy, S. and Wilkinson, J.P.D., "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering,"

McGraw Hill, 1976.

[6.5.5] Paul, B., "Fluid Coupling in Fuel Racks: Correlation of Theory and Experiment", (Proprietary), NUSCO/Holtec Report HI-88243.

[6.6.1] ASME Boiler & Pressure Vessel Code,Section III, Subsection NF, 2001 Edition up to and including 2003 Addenda.

[6.6.2] ASME Boiler & Pressure Vessel Code,Section III, Appendices, 2001 Edition up to and including 2003 Addenda.

[6.6.3] ASME Boiler & Pressure Vessel Code,Section III, Subsection NF, 1977 Edition.

[6.8.1] Chun, R., Witte, M. and Schwartz, M., "Dynamic Impact Effects on Spent Fuel Assemblies," UCID-21246, Lawrence Livermore National Laboratory, October 1987.

[6.9.1] ACI 349-76, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan, 1976.

[6.9.2] ACI 318-71, Building Code requirements for Structural Concrete,"

American Concrete Institute, Detroit, Michigan, 1971.

[6.9.3] Holtec report HI-89330, Rev. 1, "A Seismic Analysis of High Density Fuel Racks; Part III: Structural Design Calculations - Theory."

Holtec Report HI-2033124 6-45 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Table 6.2.1:

PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK PLANT DOCKET NUMBER(s) YEAR Enrico Fermi Unit 2 USNRC 50-341 1980 Quad Cities 1 & 2 USNRC 50-254, 50-265 1981 Rancho Seco USNRC 50-312 1982 Grand Gulf Unit 1 USNRC 50-416 1984 Oyster Creek USNRC 50-219 1984 Pilgrim USNRC 50-293 1985 V.C. Summer USNRC 50-395 1984 Diablo Canyon Units I & 2 USNRC 50-275, 50-323 1986 Byron Units 1 & 2 USNRC 50-454,50-455 1987 Braidwood Units I & 2 USNRC 50-456,50-457 1987 Vogtle Unit 2 USNRC 50-425 1988 St. Lucie Unit 1 USNRC 50-335 1987 Millstone Point Unit 1 USNRC 50-245 1989 Chinshan Taiwan Power 1988 D.C. Cook Units I & 2 USNRC 50-315, 50-316 1992 Indian Point Unit 2 USNRC 50-247 1990 Three Mile Island Unit I USNRC 50-289 1991 James A. FitzPatrick USNRC 50-333 1990 Shearon Harris Unit 2 USNRC 50-401 1991 Hope Creek USNRC 50-354 1990 Kuosheng Units I & 2 Taiwan Power Company 1990 Holtec Report HI-2033124 6-46 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

- ~~Table 6.2.1 --:

PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK PLANT DOCKET NUMBER(s) YEAR Ulchin Unit 2 Korea Electric Power Co. 1990 Laguna Verde Units 1 & 2 Comision Federal de 1991 Electricidad Zion Station Units 1 & 2 USNRC 50-295, 50-304 1992 Sequoyah USNRC 50-327,50-328 1992 LaSalle Unit I USNRC 50-373 1992 Duane Arnold Energy Center USNRC 50-331 1992 Fort Calhoun USNRC 50-285 1992 Nine Mile Point Unit I USNRC 50-220 1993 Beaver Valley Unit I USNRC 50-334 1992 Salem Units I & 2 USNRC 50-272,50-311 1993 Limerick USNRC 50-352,50-353 1994 Ulchin Unit I KINS 1995 Yonggwang Units 1 & 2 KINS 1996 Kori-4 KINS 1996 Connecticut Yankee USNRC 50-213 1996 Angra Unit I Brazil 1996 Sizewell B United Kingdom 1996 Waterford 3 USNRC 50-382 1997 J.A. Fitzpatrick USNRC 50-333 1998 Callaway USNRC 50-483 1998 Holtec Report HI-2033124 6-47 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Table 6.2.1

- : : PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK PLANT DOCKET NUMBER(s) YEAR Nine Mile Unit I USNRC 50-220 1998 Chin Shan Taiwan Power Company 1998 Vermont Yankee USNRC 50-271 1998 Millstone 3 USNRC 50-423 1998 Byron/Braidwood USNRC 50-454, 50-455, 1999 50-567, 50-457 Wolf Creek USNRC 50-482 1999 Plant Hatch Units I & 2 USNRC 50-321, 50-366 1999 Harris Pools C and D USNRC 50-401 1999 Davis-Besse USNRC 50-346 1999 Enrico Fermi Unit 2 USNRC 50-341 2000 Kewaunee USNRC 50-305 2001 V.C. Summer USNRC 50-395 2001 St. Lucie USNRC 50-335, 50-389 2002 Turkey Point USNRC 50-250, 251 2002 Holtec Report HI-2033124 6-48 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Table 6.3.11-RACK MATERIAL DATA (200 0 F):. ..- : -

(ASME - Section II, Part D)-

Stainless Steel Young's Modulus Yield Strength Ultimate Strength Material E SY S, (psi) (psi) (psi)

SA240, Type 304L (cell 27.6 x 106 21,300 66,200 boxes)

SUPPORT MATERIAL DATA (200 0 F)

SA240, Type 304L 27.6 x 106 21,300 66,200 (upper part of support feet &

Bearing Pads)

SA-564-630 (lower part of 28.5 x 106106,300 140,000 support feet; age hardened at 11001F)

Holtec Report HI-2033124 649 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Table 6.4.1

'TIME-HISTORY STATISTICAL CORRELATION RESULTS.:..

OBE Data 1 to Data2 0.078 Datal to Data3 0.003 Data2 to Data3 0.008 SSE Datal to Data2 0.083 Data I to Data3 0.007 Data2 to Data3 0.005 Data 1 corresponds to the time-history acceleration values along the X axis (South)

Data2 corresponds to the time-history acceleration values along the Y axis (East)

Data3 corresponds to the time-history acceleration values along the Z axis (Vertical)

Holtec Report HI-2033124 6-50 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

-Table 6.5.1

_ . ' De rees-of-freedom  :-,.,______._:::.':-'-_:

LOCATION (Node) DISPLACEMENT ROTATION Ux Uy Uz Ox oy I 1 PI P2 P3 q4 qs q6 2 P7 P8 P9 qto q1l q12 Node 1 is assumed to be attached to the rack at the bottom most point.

Node 2 is assumed to be attached to the rack at the top most point.

Refer to Figure 6.5.1 for node identification.

2 P13 P14 3 P15 P16 4 P17 P18 5 P19 P20 1 P21 P22 where the relative displacement variables qi are defined as:

pi = qi(t) + U.(t) i = 1,7,13,15,17,19,21

= q1(t) + Uy(t) i = 2,8,14,16,18,20,22

= qi(t) + U,(t) i = 3,9

= qi(t) i = 4,5,6,10,11,12 pi denotes absolute displacement (or rotation) with respect to inertial space qj denotes relative displacement (or rotation) with respect to the floor slab

• denotes fuel mass nodes U(t) are the three known earthquake displacements Holtec Report HI-2033124 6-51 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Table 6.9.1 COMPARISON OF BOUNDING CALCULATED LOADS/STRESSES VS. CODE-ALLOWABLES

AT IMPACT LOCATIONS AND AT WELDS SSE* or OBEI ItemlLocation Calculated Allowable Fuel assembly/cell wall impact, lbf. 815 2,826" Rack/baseplate weld, psi 6,456 26,414 Female pedestal/baseplate weld, psi

• 8,767* 35,748*

Cell/cell welds, psi, based on impact loads 4,03 ltt" 8,520 t Loads and allowables given are for the more limiting of OBE or SSE (When applicable to SSE case, it is denoted by an asterisk, *).

ft Based on the limit load for a cell wall. The allowable load on the fuel assembly itself may be less than this value (see discussion in Section 6.8.4.3), but is greater than 815 lbs.

ttt Based on the base metal stresses adjacent to weld placements resulting from the maximum shear flow developed between two adjacent cells.

Holtec Report H--2033124 6-52 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Time History Accelerogram ClintonPower Station Auxiliary-Fuel Building at Elev. 712' SSE 4% Damping (X - East-West Direction) 040 0.20 - I I II I .I I I I 11 . I I

-6 z

0 0.00 liI fiA

-J lyl II 41 UJ C.

C.

-0.20 II A Af% I

-U.qU 0 500 1000 1500 TIME (SEC x 100)

Figure 6.4.1 Holtec Report H--2033124 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Time History Accelerogram ClintonPower Station Auxiliary-Fuel Building at Elev. 712' SSE 4% Damping (Y - North-South Direction) 0.40 0.20 z

0 w 0.00 w

-J Ul C.)

-0.20

-0.40 0 500 1000 1500 TIME (SEC x 100)

Figure 6.4.2 Holtec Report HI-2033124 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Time History Accelerogram ClintonPower Station Auxiliary-Fuel Building at Elev. 712' SSE 4% Damping (Z - Vertical Direction) 0.40 0.20 z

0 0.00 - I!

-J

'L C.)

0.

-0.20

-0.40 0 500 1000 1500 TIME (SEC x 100)

Figure 6.4.3 Holtec Report HI-2033124 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Time History Accelerogram ClintonPower Station Auxiliary-Fuel Building at Elev. 712' OBE 2% Damping (X - East-West Direction) 0.30 0.10 I . 1 .1 I.,,-1.I I, I I I 11 I I- ..

z 0

'MI I I I I AidlUlIkki.,

-j w I I I I 11" I III 'if 171 VW -

C.

C.

-0.10 I TI , 11 I I I I, I I

-0.30 0 500 1000 1500 TIME (SEC x 100)

Figure 6.4.4 Holtec Report HI-2033124 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Time History Accelerogram ClintonPower Station Auxiliary-Fuel Building at Elev. 712' OBE 2% Damping (Y - North-South Direction) 0.30 0.10 z h I 0

-J w

C.

-0.10

-0.30 0 500 1000 1500 TIME (SEC x 100)

Figure 6.4.5 Holtec Report HI-2033124 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

Time History Accelerogram ClintonPower Station Auxiliary-Fuel Building at Elev. 712' OBE 2% Damping (Z - Vertical Direction) 0.25 0.15 z 0.05 0

I-

. -0.05

-0.15

-0.25 0 500 1000 1500 TIME (SEC x 100)

Figure 6.4.6 Holtec Report HI-2033124 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

FIGURE 6.5 . SCHEMATIC OF THE DYNAMIC MODEL OF A SINGE RACK MODULE USED IN DYNARACK H134,21 &lDRAVIN5S10-FI6URES1134?2REPDRT-HI-20331241CHPT61FIGURE6.5 11REP[IRT HI-2033124

/-FUEL ASSEMBLY/CELL Y / GAP/ELEMENT IMPACT

//,

//

B/

x/

L-FIGURE 6.5.2; FUEL-TO-RACK GAP/IMPACT ELEMENTS AT LEVEL OF RATTLING MASS H13421 G\DRAvINS\O-FIGRES\1342\REPoRT-HI-2033124\CHfT-6\FGURE6.52 HI-2033124

FUEL ASSY./CELL \

GAP/IMPACT ELEMENT, Ki II H/4 H/2 TYPICAL FUEL RACK C.G.-\ RATTLING MASS I H/4 RACK PERIPHERY-GAP/IMPACT ELEMENTS, Kg H/4 H/2 I

H/4 HWv FRICTION INTERFACE GAP/IMPACT ELEMENT, Kf ELEMENT, Ks

\ 777 FIGURE 6.5.3; TWO DIMENSIONAL VIEW OF THE SPRING-MASS SIMULATION H13421 W]WES3 IRLPURi V til-2033124 H13421 \IRAv:NEis\fl-rIwRs\I342\P[PnRT-HI-?D321?4\cfPT \VICJR[&E3 iRLPUR I III -2033W4

, H/2 Au ql 0 14 H/2 J FOR Y-Z PLANE BENDING 7

SL ~H/2 (q1 H/2 FOR X-Z PLANE BENDING FIGURE 6.5.4; RACK DEGREES-OF-FREEDOM WITH SHEAR AND BENDING SPRINGS H13421 C.:\DRAY4NGS\0-FMES\1342\REPORT-HI-203124\CHPT_6\FIGURE6.5.4 IHI-2033 124 H13421 C:\DRAWINGS\O-RGURES\1342\REPORT-HI-2033124\CHFL6\F1GtJRE6.5.4 1HI -20331 24

TYPICAL TOP -

IMPACT ELEMENT RACK STRUCTURE TYPICAL BOTTOM IMPACT ELEMENT RACK B PLATE R 5 REL 6 G FIGURE 6.5.5; RACK PERIPHERY GAP/IMPACT ELEMENTS H1342l H G'.\DRAWINGS\(-FWB\1342\UM-H-2033124\CC-6\WR655 C\RWN\-E\32RPfT1-O32\lL\UE6.5IRP *32 IREPORTT HI-2033124 HI-032

Figure 6.9.1 Isometric of ANSYS Model Holtec Report HI-2033124 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

I.

NODAL SOLUTION ANSYS JAN '6 2004 STE P-l 08:41:34 SUB =1 SEQV (AVG)

D!IV -. 010998 SIDC' =86.996 SIX =27964

- -- 7 1 I- - 7,_ -. I---

- ... 7, 17 7 86.996 6282- 1247.7, 24866

3184 9379 *15574w .21769 27964 Bear ing Pad. Andlysiis Figure 6.9.2 Stress Distribution in Bearing Pad Holtec Report HI-2033124 Project 1342 SHADED AREAS DESIGNATE PROPRIETARY INFORMATION

a

--aB-. II

-- - .0-q ---I-b -a-q

--I- I T a >4 b

FIGURE 6.12.1; LOADING ON RACK WALL

- r-t Heated Celt Watt N- x H LI

\\\ \ N\\\\\\\ \ \ \ \\\\\\\

I  %

L

'We Id Line y

FIGURE 6.12.2; WELDED JOINT IN RACK H1348 H13421 ~G:\DRAWINGS\O-VIGUR[S\1342\REPORT-HI-2033124\CHPI 6\FIGURE~6.1~221 REIPORTI HI1-2033124

7.0 MECHANICAL ACCIDENTS 7.1 Introduction The USNRC OT position paper [7.1.1] specifies that the design of the rack must ensure the functional integrity of the spent fuel racks and the pool under all credible fuel assembly drop events. This chapter contains synopses of the analyses carried out to demonstrate the regulatory compliance of the proposed racks under postulated accidental drop events germane to the Clinton Power Station Unit I spent fuel pool and cask storage pool.

The proposed change does not alter assumptions or results under the current licensing basis on the potential fuel damage due to mechanical accidents.

7.2 Description of Mechanical Accidents Analyses are performed to evaluate the damage to the new racks and the pool liner subsequent to a fuel assembly impact under various drop scenarios. Two categories of accidental drop events are considered.

In the so-called "shallow" drop event, a fuel assembly, along with the portion of handling tool, which is severable in the case of a single element failure, is assumed to drop vertically and hit the top of a rack cell and subsequently the fuel assembly stored in the cell. Inasmuch as the new racks are of honeycomb construction, the deformation produced by the impact is expected to be confined to the region of collision. However, the "depth" of damage to the affected cell walls must be demonstrated to remain limited to the portion of the cell above the top of the "active fuel region", which is essentially the elevation of the top of the neutron absorber. Stated in quantitative terms, this criterion implies that the plastic deformation of the cell walls should not extend more than 12 inches (downwards) from the top.

In order to utilize an upper bound of kinetic energy at impact, the free-fall height is conservatively assumed to be 6 feet [7.2.1].

It is readily apparent from the description of the rack modules in Section 3 that the impact resistance of a rack at its periphery is considerably less than its interior. Accordingly, the limiting shallow drop Holtec Report HI-2033124 7-1 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

scenario, which would produce maximum cell wall deformation, consists of the case where the fuel assembly impacts the peripheral wall of a cell on the periphery of the rack. Furthernore, the dropped fuel assembly can only achieve first contact with the top of the rack at this location because of the orientation of the fuel assembly handle that sticks out of the top of rack. Other impact sites would require the falling fuel assembly to impact the stored fuel assembly handle first. These other impact scenarios would tend to reduce the damage to the rack and are, thus, not considered here, since the goal is to maximize penetration depth to compare against the acceptance criteria. Figure 7.2.1 depicts the finite element model used to evaluate this scenario.

The second class of fuel drop event postulates that the impactor falls through an empty storage cell impacting the fuel assembly support surface (i.e., rack baseplate). This so-called "deep" drop event threatens the structural integrity of the baseplate. If the baseplate is pierced or sufficiently deformed to allow the fuel assembly to impact the liner, then the liner integrity is at risk and water could leak from the pool. The deformed baseplate may also lead to an abnormal condition of the enriched zone of fuel assembly outside the "poisoned" space of the fuel rack. To preclude damage to the pool liner and to avoid the potential of an abnormal fuel storage configuration in the aftermath of a deep drop event, it is required that the baseplate remain unpierced, the baseplate not impact the liner, and that the maximum lowering of the baseplate is shown to be acceptable by the criticality evaluations (see Section 4 for further discussion).

The deep drop event can be classified into two scenarios, namely, a drop in an interior cell away from the support pedestal, as shown in Figure 7.2.2, and a drop through a cell located above a support leg, as shown in Figure 7.2.3. In deep drop scenario 1, the fuel assembly impacts the baseplate away from the support pedestal, where it is more flexible. A gross severing or large deflection of the baseplate leading to a secondary impact with the pool liner is unacceptable. In deep drop scenario 2, the baseplate is buttressed by the support pedestal and presents a hardened impact surface, resulting in a high load. The principal design objective is to ensure that the support pedestal does not tear the liner that overlays the reinforced concrete pool slab.

A review of the proposed cask storage pool layouts (shown in Figures 1.1.1 and 1.1.3) indicates that a transfer cask may be placed adjacent to the storage racks. However, administrative controls will ensure Holtec Report Hl-2033124 7-2 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

that all fuel will be removed from the Cask Loading Pool prior to any casks being moved over this area.

Controls will be implemented via the plant change package prepared to support the physical modifications. The plant change package will ensure that cask movement and/or fuel movement procedures are modified as necessary to preclude fuel storage in racks located within Cask Loading Pool during fuel cask movement in the vicinity of the Cask Loading Pool. Therefore, there is no need to evaluate any new cask drop scenarios for this fuel storage expansion project.

7.3 Incident Impact Velocity 7 7 444 474 I-_tec R H 7-3V2d3_1.4 142 4 49 4 , 4;,.,.';:;,-; '---.

I  ;.40 a"':L0  :; 0 X 44d adH -S ~0n,*;

, 4.. ' '  ; . ' ;' ;".5 FSAE ARA DEOT PRPITR INFORMEATION  :' :Ed t 44... 4~. .4 4 . ¢i.

hW4ue_;_i._

7- 134 Hote Hl2332 Reor SHDE ARA DEOT PRPITR INFRMAIO

4444 444 444 4 44

- .4

.444444

- - 44 1

444 4 4.

4 4 44 4 4 a 4 4 4 4 44 4 44 4 444 4 4 4 4 4 4 44 44 44 44 44 4 4 44 444 44 .4 4 4 4 4 a

-' a 4 44

.. a 44 4 4 4 4 4 4 4.4 4 444 4 4 44 4 4 a 4 44 4 4 4 4,

.4

.4 4 4 4 4444 4 44 4 4 4 4 4 4 4 4 4 7.4 Mathematical Model In the first step of the solution process, the velocity of the dropped object (impactor) is computed for the condition of underwater free fall in the manner of the formulation presented in the above section. Table 7.4.1 contains the computed velocities for the various drop events.

4 4 4 4' 44 44 4 4 4 .4 44 .4 4 4 4' '4 4 4 4 44 4 4 4 .4 4 4 4 4 4 4 4 .4 4 4 4 4 44 '4 4 4 4 4 .4 4 4. 44 4 4 4 4' 4 44 Holtec Report HI-2033124 7-4 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

E. E The physical properties of material types undergoing deformation in the postulated impact events are summarized in Table 7.4.2.

7.5 Results 7.5.1 Shallow Drop Event For the shallow drop event, the dynamic analysis shows that the top of the impacted region undergoes localized plastic deformation. Figure 7.5.1 shows an isometric view of the post-impact geometry of the rack. The maximum depth of plastic deformation is limited to 4.75 inches, which is less than the design limit of 12 inches. Therefore, the damage does not extend into the active fuel region of any stored fuel.

7.5.2 Deep Drop Events The deep drop through an interior cell does produce some deformation of the baseplate with local severing of the baseplate/cell wall welds. Figure 7.5.2 shows the deformed baseplate configuration.

The fuel assembly support surface is lowered by a maximum of 2.6 inches, which is much less than the gap (5.4 inches) between the fuel bottom and the pool liner. The deformation of the baseplate has been determined to be acceptable with respect to lowering the fuel seating position and the resulting criticality consequences, as discussed in Chapter 4.0.

The deep drop event, wherein the impact region is located directly above the support pedestal, is found to produce a maximum plastic strain of 0.0863 in the liner, which is much smaller than the failure strain of the liner material, as shown in Figure 7.5.3. Finally, the concrete slab is found to experience very limited local damage as shown by the predicted cracks and the effective strain distribution in Figure 7.5.4. Since the pool floor can maintain its overall integrity and the liner is not breached in the drop event, there will be no abrupt or uncontrollable loss of water from the pool.

Holtec Report HI-2033124 7-5 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

7.6 Conclusion The drop events postulated for the Clinton Power Station spent fuel/cask storage pools were analyzed and found to produce localized damage within the design limits for the racks. The shallow drop event is found to produce some localized plastic deformation in the top of the storage cell, but the region of permanent strain is limited to the portion of the rack structure situated above the top of the active fuel region. The analysis of the deep drop event at cell locations selected to maximize baseplate deformation indicates that the downward displacement of the baseplate is limited to 2.6 inches, which ensures that fuel will remain in a subcritical condition. The deep drop case analyzed for the scenario to produce maximum pedestal force indicates that the pedestal axial load precludes liner damage and prevents a breach in the integrity of the concrete floor slab. Therefore, there will be no uncontrollable loss of pool water inventory. In conclusion, the new Holtec high-density spent fuel racks for the Clinton Power Station spent fuel/cask storage pools possess acceptable margins of safety under the postulated mechanical accidents.

Holtec Report HI-2033124 7-6 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

7.7 References for Chapter 7.0

[7.1.1] "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications,"

dated April 14, 1978, and addendum dated 1979.

[7.2.1] Clinton Power Station USAR, Revision 10.

Holtec Report HI-2033124 7-7 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 7.4.1 IMPACT EVENT DATA Drop Impact Case Impactor Impactor Height Velocity Weight (ib) Type (in) (in/sec) 614 (fuel, wet weight) Fuel assembly I. Shallow drop event 450 (tool, dry weight) h t2 205_7

2. Deep drop event 614 (fuel, wet weight) Fuel assembly scenario 1 (away 40(oldr egt)& 240 343.6 from pedestal) handling tool

3. Deep drop event 614 (fuel, wet weight) Fuel assembly scenario 2 (above 450 (tool, dry weigh& t 240 299.9 pedestal) 45 (todywih) handling tool _____________

Holtec Report HI-2033124 7-8 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 7.4.2 MATERIAL DEFINITION Elastic Stress Strain Material Name Material Density Modulus (psi) First Yield (psi) Failure (psi) Elastic Failure Stainless Steel SA240-304L 501 2.787e+07 2.3 15e+04 6.81 Oe+04 8.306e-04 3.800e-01 Carbon Steel SA564-630 490 2.86e+07 1.092e+05 1.400e+05 3.818e-03 1.400e-0l Concrete fc=3,500 psi 140 3.372e+06 Holtec Report H1-2033124 7-9 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Fig. 7.2.1 Finite Element Model of the "shallow" drop event Holtec Report H1-2033124 7-10 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

FIEL ASSEMLY IMPAfCT EGIOi Fig. 7.2.2 Schematics of the "deep" drop scenario 1 Note: This figure is primarily provided to indicate the impact zone for this scenario. The configuration of the rack is not intended to be accurate.

Holtec Report HI-2033124 7-11 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

L; FUEL ASSEMLY IMPACT FEG5ON Fig. 7.2.3 Schematics of the "deep" drop scenario 2 Note: This figure is primarily provided to indicate the impact zone for this scenario. The configuration of the rack is not intended to be accurate.

Holtec Report HI-2033124 7-12 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Fig. 7.5.1 "Shallow" Drop: Maximum Plastic Strain Holtec Report HI-2033124 7-13 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Fig. 7.5.2 "Deep" Drop Scenario 1: Maximum Vertical Displacement Holtec Report HI-20331224 7-14 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Fig. 7.5.3 "Deep" Drop Scenario 2: Maximum Von Mises Stress - Liner Holtec Report H1-2033124 7-15 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Fig. 7.5.4 "Deep" Drop Scenario 2: Maximum Effective Strain - Concrete Holtec Report HI-2033124 7-16 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

8.0 FUEL BUILDING STRUCTURAL INTEGRITY EVALUATIONS Structural integrity evaluations of the regions of the reinforced concrete structure affected by the proposed fuel storage capacity expansion are summarized in this section. The introduction of racks into the Fuel Cask Storage Pool during Phases I and 2 and installation of new racks in the Spent Fuel Pool (SFP) during Phase 2, as shown in Figures 1.1.1 through 1.1.3 will affect the structure by imposing a hydrodynamic pressure on the walls adjacent to the new racks and additional loads on the pool floor through the pedestal bearing pads.

8.1 Introduction Figure 8.1 shows a plan view of the SFP and adjacent Fuel Cask Storage Pool. The SFP and Fuel Cask Storage Pool each consist of four walls with floor slabs on grade. The floors of both pools are constructed above 21'-10" of controlled compacted fill material and consist of a 6 inch mud mat overlain with 9'-8" of reinforced concrete, topped with 6 inches of grout and covered with a 1/4" thick liner. This brings the level of the floor up to the 712'-0" plant elevation. The concrete floor provides substantial strength, due to its thickness, placement on grade and the fact that it is heavily reinforced with #11 and # 18 reinforcement bars.

A review of Figure 8.1 indicates that the western wall of the SFP will control the evaluation for the four walls surrounding this pool. The nominal rack-to-wall dimension of 2.5 inches (See Figure 1.1.2) is less for this wall than any of the other three walls. This wall is longer than the north and south walls and does not have any intermediate walls or floors providing support. Therefore, an evaluation of the western wall of the SFP for the increased rack-to-wall hydrodynamic coupling pressures is performed. The adjacent transfer pool is considered to be empty for conservatism.

Similarly, by observation of Figure 8.1, the 36 inch thick south and west walls of the Fuel Cask Storage Pool are much thinner than the other two 72 inch thick walls. The southern wall controls, because this wall is longer than the west wall and contains less reinforcement in the horizontal direction.

Holtec Report HI-2033124 8-1 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

The structural evaluations of the affected portions of the SFP and Fuel Cask Storage Pool are conducted using finite element models of the controlling walls extending the full width of the pools.

Figures 8.2 and 8.3 show the SFP west wall and Fuel Cask Storage Pool south wall finite element models, respectively.

Results for individual load components are combined using the factored load combinations mandated by the Clinton Power Station USAR [8.1], and are based on the "ultimate strength" design method.

Bending moment capabilities are checked for appropriate sections on each wall in each direction (vertical and horizontal) for concrete structural integrity. Due to the compressive nature of the axial loads, the relationships between bending moment capacity and compression loads are conservatively neglected. Shear capability is evaluated along all sections of the affected walls. Load combinations and structural capacity assessments follow requirements of the plant USAR [8.1] and the American Concrete Institute Code (ACI 318) [8.2].

The SFP and Fuel Cask Storage Pool floor slabs are not included in the finite element model. The global adequacy of the slab and underlying subsoil remains satisfactory to withstand the additional loading imposed by the rack. This assertion is derived from the fact that the slab is a continuous structure with the same capacities in the Fuel Cask Storage Pool as in the SFP in areas beneath existing racks with similar fuel storage density and associated loading. Local stresses in the liner and underlying concrete imposed by the rack pedestals are addressed in Sections 6.9.6, 8.6, and 8.7.

The thermal loading in the reinforced concrete structure is considered in the manner specified in the applicable codes. The temperature gradients considered (see Section 8.4.3) are those defined in Section 5.0 and the plant USAR [8.1]. Consistent with standard design practices, the temperature gradient established for the pool walls is intended to subsume local thermal effects such as direct heat deposition into the concrete from the absorption of gamma radiation from the stored spent fuel.

Holtec Report HI-2033124 8-2 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

8.2 Description of the SFP West Wall and Fuel Cask Storage Pool South Wall The west wall of the SFP is 6' thick, 36' long and 43' high spanning in elevation from the top of the slab at 712' to the mezzanine floor at 755'. This wall is shared by the Transfer Pool and has no lateral support against out of plane displacements. The transfer gate from the SFP to the Transfer Pool creates a 4' wide by 27' deep discontinuity in the SFP west wall and is considered in the model, as shown in Figure 8.2. The horizontal reinforcement in this wall increases approaching the floor slab. The vertical reinforcement remains consistent through the height except for a local area near the floor slab in the middle of the base of the wall.

The south wall of the Fuel Cask Storage Pool is 3' thick, 17' long and 43' high spanning in elevation from the top of the floor slab at 712' to the mezzanine floor at 755'. This wall has lateral support against out of plane displacements from a 3' thick floor at Elevation 739'. The horizontal reinforcement is consistent throughout the height of the wall. The vertical reinforcement varies along the length and height of the wall.

In both analyses, the walls are fixed against out-of-plane and horizontal in-plane translations and three rotational degrees of freedom, but are free to displace in the vertical direction for all dead and seismic loadings. The boundary conditions are modified to allow for in-plane displacements due to thermal expansion.

8.3 Analysis Procedures The reinforced concrete walls are subjected to individual "unit" load cases covering the service conditions (the structural weight of the concrete structure, the hydro-static water pressure and the temperature gradient) and seismic induced loads (structural inertial loads, hydro-dynamic water loads, and rack-structure interaction dynamic loads) for operating basis earthquake (OBE) and safe shutdown earthquake (SSE) conditions. The service condition loads are considered as static acting loads; the seismic induced loads for both OBE and SSE seismic events are obtained from the application of acceleration spectra provided in the plant USAR [8.1] with input seismic acceleration Holtec Report HI-2033124 8-3 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

amplifiers defined on the basis of a frequency analysis of the structure. Results from the seismic load cases are combined using the square root of the sum of the squares (SRSS) and then combined with the static load.

The reinforced concrete is considered elastic and isotropic. The elastic characteristics of the concrete are independent of the reinforcement contained in each structural element for the mechanical load cases when un-cracked cross-sections are assumed. This assumption is valid for all load cases with the exception of the thermal loads, where, for a more realistic description of the reinforced concrete cross-section behavior, the assumption of cracked concrete is used. To simulate the cracking patterns, the original elastic modulus of the concrete is reduced in accordance with the methodology suggested by ACI 349 [8.3]. Table 8.1 summarizes the concrete properties employed in the structural evaluation of the SFP west wall and Fuel Cask Storage Pool south wall.

8.4 Definition of Loads Included in Structural Evaluation These definitions apply to both the SFP west wall and Fuel Cask Storage Pool south wall.

8.4.1 Static Loadina (D = Dead Loads)

I) Dead weight includes the weight of the wall.

2) The hydrostatic water pressure acting on the wall.

8.4.2 Seismic (E= OBE: E' = SSE)

I) Horizontal hydrodynamic inertia loads due to the contained water mass and sloshing loads in the entire SFP (considered in accordance with [8.4]) that arise during a seismic event.

2) Horizontal hydrodynamic pressures between spent fuel rack and pool wall caused by rack motions during a seismic event.

3) Vertical hydrodynamic pressure due to acceleration of the contained water mass.

4) Seismic inertia force of the walls from the wall mass.

Holtec Report HI-2033124 8-4 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

8.4.3 Thermal Loading (T)

A steady-state thermal gradient is defined by the bulk pool temperature of 1340 F (Section 5) and the Fuel Building ambient of 70'F given in [8.1]. The bulk pool temperature is applied to both the SFP and Fuel Cask Storage Pool.

8.4.4 Load Combinations and Acceptance Criteria No live loads are defined for the areas under consideration. Results from a suite of unit load analyses are used to form appropriate load cases and then combined in accordance with the load combinations specified in Table 3.8-1.2 of the plant USAR [8.1].

The final load combinations evaluated for structural integrity are:

Normal Loading Combination 1.4 D + 1.7 T0 Severe Environmental Loading Combination l.4D+ 1.7 T.+ l.9E Abnormal Loading Combination 1.0 D + 1.0 T. +1.0 E' Extreme Environmental Loading Combination l.OD+ 1.25E Abnormal / Severe and Extreme Loading Combination 1.0 D + 1.0 E' Note that seismic loads, after the SRSS combination, are directed so that they add to the hydrostatic pressure and wall self-weight.

Moments and shears computed for each load combination are compared with their respective capacities. Consistent with the intent of the guidance provided in the ACI literature, and recognizing that there is always load re-distribution occurring in a concrete structure designed in accordance with ultimate strength methods, characteristic section widths (horizontal and vertical) are established over which moments are averaged and then compared with the averaged section capacity. Similarly, the transverse shear is averaged over the same section width to define the "section shear".

Holtec Report H1-2033124 8-5 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

The ratios of the moment and shear capacities to their respective "section" values are referred to as the safety factor (SF). In computing the SF for section moments and shears, the presence of in-plane compressive loads, which act to increase capacities, are conservatively neglected.

8.5 Results of Reinforced Concrete Analyses The structural integrity of the SFP west wall and Fuel Cask Storage Pool south wall are evaluated and the axial forces, the bending moments and the shear forces were computed for all load combinations. The reinforced concrete cross-sectional capacities were determined and used to obtain the safety factors of the structural elements for each load combination considered. Safety factors are acceptable if the safety factor exceeds 1.0. The calculated minimum safety factors for the sections of the walls for each load combination are:

SFP West Wall Item / Direction Safety Factor Load Combination Moment / Horizontal 1.23 Severe Environmental Moment / Vertical 1.24 Normal Shear / Horizontal 1.42 Severe Environmental Shear / Vertical 1.26 Normal Fuel Cask Storage Pool South Wall Item / Direction Safety Factor Load Combination Moment / Horizontal 1.07

• Severe Environmental Moment / Vertical 2.23 Normal Shear / Horizontal 1.62 Severe Environmental Shear / Vertical 1.52 Normal

• The bending safety factors conservatively neglect the additional load carrying capacity induced by the presence of axial compression.

Holtec Report H1-2033124 8-6 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

8.6 Pool Liner Evaluation The freestanding racks are supported on a 1/4" thick stainless steel liner plate, which separates the bearing pad from the concrete pool floor. During a seismic event the racks may undergo a series of motions developing friction forces between the bearing pad and the pool liner. The friction loads shall not buckle or tear the liner plate or cause the liner seam welds to rupture. A strength assessment is performed to show that the stresses in the liner plate and the seam welds comply with ASME Code Section III stress limits [8.5]. Since the pedestal loads occurring during the seismic event are repetitive, a fatigue assessment of the liner using Miner's Rule is also performed. The fatigue assessments conservatively consider 20 OBE and one SSE events.

The calculated minimum safety factors for the SFP liner are:

Principal Stress in Liner 5.9 Shear Stress in Liner Weld 3.8 Cumulative Usage Factor 1.05 x 10-3 The stresses in the liner comply with ASME Subsection NF stress limits. In accordance with ASME Subsection NB [8.6], the cumulative usage factor is below the limit of 1. Therefore, fatigue failure does not occur after twenty OBE and one SSE events.

8.7 Bearing Evaluation Bearing pads are placed between rack pedestals and the spent fuel pool (SFP) floor to reduce the otherwise high local stresses in the SFP concrete slab by spreading the concentrated load of each pedestal over a larger concrete contact area. The vertical pedestal loads generated by the peak load are obtained from the dynamic analyses of the spent fuel racks subject to seismic loads.

The calculated minimum safety factors for local bearing stress on the concrete is:

Bearing Stress in Concrete Slab 1.19 Holtec Report HI-2033124 8-7 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

8.8 Conclusions Regions affected by the installation of racks in the Fuel Cask Storage Pool and the SFP are examined for structural integrity under bending and shearing action and bearing stress. It is determined that adequate safety margins exist when the factored load combinations are checked against the appropriate structural design strengths. For the most limiting load combination, the minimum safety factor remained above 1.0. Finally, it is also shown that local loading on the liner does not compromise liner integrity.

Holtec Report HI-2033124 B-8 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

8.9 References

[8.1] Clinton Power Station USAR, Revision 10, October 2001.

[8.2] ACI 318-77, Building Code Requirements for Reinforced Concrete, American Concrete Institute, 1977.

[8.3] ACI 349-01, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, 2001.

[8.4] Nuclear Reactors and Earthquakes, TID-7024, United States Atomic Energy Commission, Division of Reactor Development, August 1963.

[8.5] ASME Code Section III, 1977 Edition.

[8.6] ASME Code Section III, Subsection NB, 1977 Edition.

Holtec Report HI-2033124 8-9 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 8.1 Reinforced Concrete Properties Concrete Strength 4,000 psi Reinforced Concrete Density (typical) 150 lbf/ IV Un-cracked Concrete Modulus of Elasticity 3.6 x 106 psi Concrete Poisson's Ratio 0.17 Concrete Coefficient of Thermal Expansion 5.5 x 10-6 in/in/PF Reinforcement Strength 60,000 psi Reinforcement Modulus of Elasticity 29 x 10 psi Reference Temperature 70 0F Holtec Report HI-2033124 8-10 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

I I

_ _ __ _ _ - I I - I SPENT FUEL POOL 6' MUD SLAB tw BENEATH 9'-8' THICK .

c Li FLOOR ON so COMPACTED FILL ELEV. 711'-6' I/4 LINER SURGE 4 i TANK FIGURE 8.1 PLAN OF CLINTON SPENT FUEL POOL AREA HI233124 IPROJECT 1342

1:.

EZEMENT S AN. .;

JAN 30 2004 09:40:34 iA I

.1 A

'77 -L-Ill T.,V I-F I I 1, k Y

Figure 8. 2Spent Fuel Pool West` Wall Finite Element'Model, Proj 1342, HI-2033124

1.

ELEMEN"T'S JAN4 30 ,od4l 09:34

- .~ . I I YIi F-igure -,8.3 -Fuel.Cask, storage. Pool 'South 'Wall FE- Moidel, Proj~ 1342,- HI-`2033124

9.0 RADIOLOGICAL EVALUATION 9.1 Fuel Handling Accident Increases in the fuel-storage capacity at Clinton do not require updating the analysis of the fuel-handling accident, for these doses were calculated in the recent past [9.1. 1]. The factors affecting the doses, such as the depth of the pool, the exposure of the fuel elements considered when the accident occurs, etc. have not changed 9.2 Solid Radwaste The necessity for resin replacement is determined primarily by the requirement for water clarity, and the resin is normally changed about once a year. No significant increase in the volume of solid radioactive wastes is expected with the expanded storage capacity. During fuel-storage expansion operations, small amounts of additional resins may be generated by the pool cleanup system on a one-time basis.

9.3 Gaseous Releases Gaseous releases from the fuel storage building are combined with other plant exhausts. Normally, the contribution from the fuel storage building is negligible compared to the other releases, and no significant increases are expected as a result of the expanded storage capacity.

9.4 Personnel Exposures During normal operations, personnel on the working level of the fuel storage area are exposed to radiation from the spent fuel pool. The dose rates experienced by personnel are not expected to increase with the increased storage capacity of the pools because the dose rate from the fuel in storage is negligible. The water above the stored fuel is sufficiently deep that the dose rate from that fuel is orders of magnitude lower than the low dose rate from the radionuclides in the pool water itself and the dose rate from a fuel assembly in transit. Consequently, though the dose rate from stored fuel will increase because more Holtec Report HI-2033124 9-1 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

spent fuel assemblies are stored, it will not increase to levels comparable to those attributable to other sources.

The radionuclide concentrations in the pool water are not expected to increase significantly, for the levels are determined principally from the mixing of primary system water with the pool water and the spalling of crud deposits from the spent fuel assemblies as they are moved in the storage pool during refueling operations. Although the overall capacities of the pools are being increased, the movement of fuel during given refuelings is independent of storage capacity.

There will be no change in the dose rate from a fuel assembly in transit, for the fuel parameters have not changed and the depth of water above the active portion of the assembly has not changed.

The concrete side walls of the fuel pool are six feet thick, providing sufficient shielding that the maximum dose rate at the outside surface of the concrete, from stored spent fuel, is two mr/hr if the pool is completely filled with fuel that has cooled only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The dose rate at the outer surface of the three-foot shield for the Fuel Cask Storage Pool would be high - 26 rem/hr -- in the hypothetical situation in which the racks in the Fuel Cask Storage Pool were filled with fuel cooled only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the three rows of storage cells closest to the wall were filled with old fuel that contributes negligibly to the dose rate at the outer surface, and the rest of the cells were filled with 24-hour fuel, the dose rate through the three-foot shield is reduced to 4.4 mr/hr.

This fuel storage management scheme will be controlled by administrative controls implemented with procedural changes.

Operating experience has shown that there have been negligible concentrations of airborne radioactivity, and no increases are expected as a result of the expanded storage capacities. Area monitors for airborne activities are available in the immediate vicinities of the pools.

In summary, no increases in radiation exposure to operating personnel are expected. Consequently, neither the current health-physics program nor the area monitoring system needs to be modified.

Holtec Report HI-2033124 9-2 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

9.5 Anticipated Exposures During Storage Expansion All of the operations involved in increasing the storage capacity will utilize detailed procedures prepared with full consideration of ALARA principles. Similar operations have been performed in numerous facilities in the past, and there is every reason to believe that the expansion of storage capacity can be safely and efficiently accomplished at Clinton, with minimum radiation exposure to personnel.

Total occupational exposure for the operations required to increase storage capacity is estimated to be between 7 and 14 person-rem, as shown in Table 9.1. While individual task efforts and exposures may differ from those in Table 9. 1, the total is believed to be a reasonable estimate for planning purposes.

The existing radiation protection programs at the plant are adequate for the storage-expansion operations. Where there is a potential for significant airborne activity, continuous air samplers will be in operation. Personnel will wear protective clothing and, if necessary, respiratory protective equipment. Activities will be governed by Radiation Work Permits, and personnel monitoring equipment will be issued to each individual. As a minimum, this will include pocket dosimeters.

Additional personnel monitoring equipment (i.e., extremity badges or alarming dosimeters) may be utilized as required. Work, personnel traffic, and the movement of equipment will be monitored and controlled to minimize contamination and to assure that exposures are maintained ALARA.

At Clinton, some of the existing storage racks will be removed from service and washed down in preparation for packaging and offsite shipment. Estimates of the person-rem exposures associated with washdown and readying the old racks for shipment are included in Table 9.1. 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.

Holtec Report HI-2033124 9-3 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

9.6 References

[9.1.1] Site Boundary and Control Room Dose following a FHA in Containment using Altemative Source Terms, Stone & Webster Engineering Corporation calculation No. C-022, March 26, 2002.

Holtec Report HI-2033124 9-4 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 9.1 t PRELIMINARY ESTIMATE OF PERSON-REM EXPOSURES DURING THE EXPANSION OF FUEL-STORAGE CAPACITY Estimated Number of Person-Rem Operation Personnel Hours Exposure ft Shuffle fuel 4 160 1.6 to 3.2 Remove/relocate existing racks 6 100 1.5 to 3.0 Clean and vacuum pool(s) 4 20 0.3 to 0.6 Remove underwater appurtenances and 4 24 1.9to3.8 assist in rigging racks for moving (Divers)

Install/relocate new racks 6 60 1.0 to 2.0 Wash and decon old racks 3 20 0.2 to 0.3 Prepare old racks for shipment ttt 4 60 0.6 to 1.2 TOTAL PERSON-REM EXPOSURE 7 to 14 I This tabulation presents the estimated exposures associated with the operations necessary to reposition the existing racks and install the new racks. The exposures are based on the experience obtained in increasing the capacities of many storage pools.

It Assumes a dose rate of 2.5 mr/hr (minimum) to 5 mr/hr (maximum), except for pool cleaning and vacuuming operations, which assume 4 to 8 mr/hr, and diving operations, which assume 20 to 40. mr/hr.

Itt Maximum expected exposure, although details of preparation and packaging of old racks for shipment have not yet been determined.

Holtec Report HI-2033124 9-5 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

10.0 INSTALLATION 10.1 Introduction The installation phase of the Clinton fuel storage rack project will be executed by Holtec International's Field Services Division. Holtec, serving as the installer, is responsible for performance of specialized services, such as underwater diving and welding operations, as necessary. All installation work at Clinton is performed in compliance with NUREG-0612 (refer to Section 3.0), Holtec Quality Assurance Procedure 19.2, Clinton rack installation project specific procedures, and applicable Clinton procedures.

Crane and fuel bridge operators are trained in the operation of overhead cranes per the requirements of ANSI/ASME B30.2, and the plant's specific training program. Consistent with the installer's past practices, a videotape aided training session is presented to the installation team, all of whom are required to successfully complete a written examination prior to the commencement of work. Fuel handling bridge operations are performed by Clinton personnel, who are trained in accordance with Clinton procedures.

Rack lifting devices are required for the handling of new racks and existing racks. The lifting devices are designed to engage and disengage on lift points at the bottom of the racks. The lifting devices comply with the provisions of ANSI N14.6-1978 and NUREG-0612, including compliance with the design stress criteria, load testing at a multiplier of maximum working load, and nondestructive examination of critical welds.

A surveillance and inspection program shall be maintained as part of the installation of the racks. A set of inspection points, which have been proven to eliminate any incidence of rework or erroneous installation in previous rack projects, is implemented by the installer.

Underwater diving operations will be required to support some aspects of this project including, but not limited to, wall hanger/bracket interference removal and rack handling support.

Holtec Report HI-2033 124 10-1 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Holtec International developed procedures will be used in conjunction with the Clinton procedures to cover the scope of activities for the rack installation and removal effort. Similar procedures have been utilized and successfully implemented by Holtec on previous rerack projects. These procedures are written to include ALARA practices and provide requirements to assure equipment, personnel, and plant safety. These procedures are reviewed and approved in accordance with Clinton administrative procedures prior to use on site. The following is a list of the Holtec procedures, used in addition to the Clinton procedures to implement the installation phase of the project.

A. Installation/Removal and Handling Procedure:

This procedure provides direction for the installation, removal, and handling of the new and existing storage rack modules in the Spent Fuel Pool and Fuel Cask Storage Pool, as applicable. This procedure delineates the steps necessary to receive the new racks on site, the proper method for unloading and uprighting the racks, staging the racks prior to installation, installation of the racks, and removal and packaging of existing racks. The procedure provides for the installation of the new racks, their height and level adjustments of the rack pedestals and verification of the as-built field configuration to ensure compliance with design documents.

B. Receipt Inspection Procedure:

This procedure delineates the steps necessary to perform a thorough receipt inspection of a new rack module after its arrival on site. The receipt inspection includes dimensional measurements, cleanliness inspection, visual weld examination, and verticality measurements.

C. Cleaninz Procedure:

This procedure provides for the cleaning of a new rack module, if required. The modules are to meet the requirements of ANSI N45.2. 1, Level B, prior to placement in the Fuel Cask Storage Pool or Spent Fuel Pool. Methods and limitations on cleaning materials to be utilized are provided.

Holtec Report HI-2033 124 10-2 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

D. Pre- and Post-Installation Drag Test Procedure:

These two procedures stipulate the requirements for performing a functional test on a new rack module prior to and following installation. The procedures provide direction for inserting and withdrawing an insertion gage into designated cell locations, and establishes an acceptance criterion in terms of maximum drag force.

E. ALARA Procedure:

Consistent with Holtec International's ALARA Program, this procedure provides guidance to minimize the total man-rem received during the rack installation project, by accounting for time, distance, and shielding. This procedure will be used in conjunction with the Clinton ALARA program.

F. Liner Inspection Procedure:

In the event that a visual inspection of any submerged portion of the pool liner is deemed necessary, this procedure describes the method to perform such an inspection using an underwater camera and describes the requirements for documenting any observations.

G. Leak Detection Procedure:

This procedure describes the method to test the pool liner for potential leakage using a vacuum box.

This procedure may be applied to any suspect area of the liner.

H. Liner Repair and Underwater Welding Procedure:

In the event of a positive leak test result, underwater welding procedures may be implemented which provide for a weld repair, or placement of a stainless steel repair patch, over the area in question. The procedures contain appropriate qualification records documenting relevant variables, parameters, and Holtec Report HI-2033 124 10-3 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

limiting conditions. The weld procedure is qualified in accordance with ASME Section XI , or may be qualified to an alternate code accepted by Amergen and Holtec International.

10.2 Rack Arrangement The rerack project at Clinton will occur in two phases, as initially discussed in Section 1.0. In the initial phase, two new racks, to eventually be relocated to the Spent Fuel Pool in the second phase of the project, will be temporarily installed in the Fuel Cask Storage Pool. In the second phase of the project, the entire Spent Fuel Pool rack array will be changed, with the final array in the Spent Fuel Pool being comprised of both new and existing racks. This configuration requires new and existing racks to be placed closer to the pool walls. Therefore, the existing sparger pipes will need to be modified by removal of some portions. This modification is address by the pool thermal-hydraulic assessment, as discussed in Section 5.5 Additionally, three existing racks will be permanently installed in the Fuel Cask Storage Pool.

10.3 Rack Interferences A survey was conducted to identify any objects which would interfere with rack installation or prevent usage of any storage locations. There are several permanently installed components interfering with the installation of the racks in the Fuel Cask Storage Pool and the Spent Fuel Pool. Those protrusions that require modification or removal to support the new rack array will be address during the rerack project.

Removal or modification of wall protusions shall be done with the aid of diver, as necessary.

10.4 SFP Cooling The pool cooling system shall be operated in order to maintain the pool water temperature at an acceptable level. It is anticipated that none of the installation activities will require the temporary shutdown of the Spent Fuel Pool cooling system.

Holtec Report HI-2033124 104 1342

'SHADED AREAS DENOTE PROPRIETARY INFORMATION

If a temporary shutdown of the Spent Fuel Pool cooling system were required, the estimated time after shutdown to increase the pool bulk coolant temperature to a selected value of <120 'F will be determined. A temperature of < 120 'F is chosen with enough margin such that cooling may be restored to ensure the pool bulk temperature will not exceed 150 'F.

10.5 Installation of New Racks and Removal of Existing Racks Installation of the new racks, supplied by Holtec International, involves the following activities. The racks are delivered in the horizontal position. A new rack module is removed from the shipping trailer using a suitably rated crane, while maintaining the horizontal configuration. The rack is placed on the up-ender and secured. Using two independent overhead hooks, or a single overhead hook and a spreader beam, the module is up-righted into a vertical position.

The new rack lifting device is engaged in the lift points at the bottom of the rack. The rack is then transported to a pre-leveled surface where, after leveling the rack, the appropriate quality control receipt inspection is performed. (See l0.1B & D.)

The Fuel Cask Storage Pool or Spent Fuel Pool floor, as applicable, is inspected and any debris, which may inhibit the installation of the racks, is removed. The new rack module is lifted with the Fuel Building Crane and transported along the pre-established safe load path. The rack module is carefully lowered into the Fuel Cask Storage Pool or the Spent Fuel Pool. For the installation of racks along the eastern edge of the Spent Fuel Pool, Fuel Building Crane travel limits will preclude this crane from installing the racks past this travel limit. For these racks, a low profile crane is anticipated to be used to locate the rack to its final design location in the Spent Fuel Pool after its initial installation into the Spent Fuel Pool by the Fuel Building Crane. The use of a temporary crane in support of rerack operations is consistent with the process utilized at previous rerack projects including Beaver Valley, Callaway, Wolf Creek, V.C. Summer, McGuire, Davis Besse, and Three Mile Island.

Holtec Report HI-2033 124 10-5 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Elevation readings are taken to confirm that the module is level and the pedestal heights are adjusted as necessary to achieve level. In addition, rack-to-wall and rack-to-rack off-set distances (gaps) are also measured. Adjustments are made as necessary to ensure compliance with design documents. The lifting device is then disengaged and removed from the Fuel Cask Storage Pool or Spent Fuel Pool under Health Physics direction. As directed by procedure, post-installation free path verification of individual cells is performed using an inspection gage.

For existing rack removal from the Spent Fuel Pool, the racks will be cleaned via pressure washing and surveyed by Health Physics prior to removal from the Spent Fuel Pool. As is the case for new rack placement, rack handling shall be completed by the Fuel Building Crane except for those where crane travel limits preclude this ability. In these instances, the low profile temporary crane will be used to lift and move the existing rack to a location in the Spent Fuel Pool that will allow access by the Fuel Building Crane for the ultimate removal of the existing rack from the Spent Fuel Pool. Safe movement of heavy loads is ensured by the processes discussed in Section 3.5.

10.6 Safetl. Health Physics. and ALARA Methods 10.6.1 Safety During the installation phase of the fuel storage rack project, personnel safety is of paramount importance. All work shall be carried out in compliance with applicable approved procedures.

10.6.2 Health Physics Health Physics is carried out per the requirements of the Clinton Radiation Protection Program.

Holtec Report HI-2033124 10-6 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

10.6.3 ALARA The key factors in maintaining project dose As Low As Reasonably Achievable (ALARA) are time, distance, and shielding. These factors are addressed by utilizing many mechanisms with respect to project planning and execution.

Time Each member of the project team is trained and provided appropriate education and understanding of critical evolutions. Additionally, daily pre-job briefings are employed to acquaint each team member with the scope of work to be performed and the proper means of executing such tasks. Such pre-planning devices reduce worker time within the radiological controlled area and, therefore, project dose.

Distance Remote tooling such as lift fixtures, pneumatic grippers, a support leveling device and a lift rod disengagement device have been developed to execute numerous activities from the SFP surface, where dose rates are relatively low.

Shielding During the course of the fuel storage rack project, primary shielding is provided by the water in the Spent Fuel Pool and Fuel Cask Storage Pool. The amount of water between an individual at the surface and an irradiated fuel assembly is an essential shield that reduces dose. Additionally, other shielding may be employed to mitigate dose when work is performed around high dose rate sources. If necessary, additional shielding may be utilized to meet ALARA principles.

Holtec Report HI-2033124 10-7 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

10.7 Radwaste Material Control Radioactive waste generated from the rack installation will be controlled in accordance with established Clinton procedures.

Holtec Report HI-2033 124 10-8 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

11.0 ENVIRONMENTAL COST / BENEFIT ASSESSMENT 11.1 Introduction Article V of the USNRC OT Position Paper [11.1] requires the submittal of a cost/benefit analysis for a fuel storage capacity enhancement. This section provides justification for selecting installation of additional racks in the Fuel Cask Storage Pool and Spent Fuel Pool (SFP) as the most cost effective alternative.

11.2 Imperative for Additional Spent Fuel Storage Capacity The specific need to increase the limited existing storage capacity of the Clinton Power Station (CPS)

Spent Fuel Pool is based on the continually increasing inventory in the pool, the prudent requirement to maintain full-core offload capability, and a lack of viable economic alternatives.

Based on the current inventory of 1,312 fuel assemblies stored in the SFP and the anticipated future discharges of spent fuel, loss of full core reserve capacity will occur during the scheduled February 2006 refueling outage when an anticipated 312 fuel assemblies are permanently discharged and new fuel is loaded into the SFP during Operating Cycle 11. The projected loss of storage capacity in the pool would affect the owner's ability to operate the reactor.

11.3 Appraisal of Alternative Options Adding fuel storage space to the Clinton Power Station SFP is the most viable option for increasing spent fuel storage capacity.

The key considerations in evaluating the alternative options included:

• Safety: Minimize the risk to the public.

Economy: Minimize capital and O&M expenditures.

• Security: Protection from potential saboteurs, natural phenomena.

Holtec Report HI-2033124 11-1 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

• Non-intrusiveness: Minimize required modifications to existing plant systems.

• Maturity: Extent of industry experience with the technology.

• ALARA: Minimize cumulative dose.

• Schedule: Minimize time to implement a plan which will maintain full-core offload capability for the distant future.

• Risk Management: Maximize probability of completing the expansion to support fuel storage needs.

Rod Consolidation Option Rod consolidation has been shown to be a potentially feasible technology. Rod consolidation involves disassembly of a fuel assembly and the disposal of the fuel assembly skeleton outside of the pool (this is considered a 2:1 compaction ratio). The rods are stored in a stainless steel can that has the outer dimensions of a fuel assembly. The can is stored in the spent fuel racks. The top of the can has an end fixture that matches up with the spent fuel handling tool. This permits moving the cans in an easy fashion.

Rod consolidation pilot project campaigns in the past have consisted of underwater tooling that is manipulated by an overhead crane and operated by a maintenance worker. This is a very slow and repetitive process.

The industry experience with rod consolidation has been mixed thus far. The principal advantages of this technology are: the ability to modularize, moderate cost, no need of additional land and no additional required surveillance. The disadvantages are: potential gap activity release due to rod breakage, potential for increased fuel cladding corrosion due to some of the protective oxide layer being scraped off, potential interference of the (prolonged) consolidation activity which might interfere with ongoing plant operation, and lack of sufficient industry experience. The drawbacks associated with consolidation are expected to diminish in time. However, it is Amergen's view that rod consolidation technology has not matured sufficiently to make this a viable option for the present CPS SFP limitations.

Holtec Report HI-2033 124 11-2 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

On-Site Dry Cask Storage Option Dry cask storage is a method of storing spent nuclear fuel in a high capacity container. The cask provides radiation shielding and passive heat dissipation. Typical capacities for BWR fuel are in the range of 68 assemblies that have been removed from the reactor for at least five years. The casks, once loaded, dried, and sealed are then stored outdoors on a seismically qualified concrete pad.

The U.S. DOE has embraced the concept of multi-purpose canisters (MPCs) obsolescing all existing licensed cask designs. Work is also continuing by several companies, including Holtec International, to improve licensed MPC systems that are capable of long storage, transport, and final disposal in a repository. However, it is noted that a cask system makes substantial demands on the resources of a plant. For example, the plant must provide for a decontamination facility where the outgoing cask can be decontaminated for release.

Several plant modifications may be required to support cask use, including: (1) tap-ins to the gaseous waste system, (2) chilled water to support vacuum drying of the spent fuel, and (3) piping to return cask water back to the Spent Fuel Pool/Fuel Cask Storage Pool. A seismic concrete pad would be needed to store the loaded casks. This pad may require a security fence, surveillance protection, a diesel generator for emergency power and video surveillance for the duration of fuel storage, which may extend beyond the life of the adjacent plant.

Other Storage Options Other options such as Modular Vault Dry Storage and a new Fuel Storage Pool are overly expensive as compared to placing new racks in the Fuel Cask Storage Pool and Spent Fuel Pool. Due to the complexity of implementation, these options could not meet the required schedule for extending full-core offload capability.

Holtec Report HI-2033 124 11-3 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

11.3.1 Alternative Option Cost Summarv An estimate of relative costs in 2004 dollars for the aforementioned options is provided in the following:

Rack Installation: $15 million Rod consolidation: $25 million Dry Storage Horizontal Silo: $35-45 million Dry Storage Modular vault: $56 million Dry Storage Metal cask (MPC): $68-100 million New fuel pool: $150 million The above estimates are consistent with estimates by EPRI and others [11.2, 11.3].

To summarize, based on the required short time schedule, the status of the dry spent fuel storage industry, and the storage expansion costs, the most acceptable alternative for increasing the on-site spent fuel storage capacity at CPS is expansion of the wet storage capacity. First, there are no commercial independent spent fuel storage facilities operating in the United States. Second, the adoption of the Nuclear Waste Policy Act (NWPA) created a de facto nuclear fuel cycle requiring disposal. Since the cost of spent fuel reprocessing is not offset by the salvage value of the residual uranium, reprocessing represents an added cost for the nuclear fuel cycle which already includes the NWPA Nuclear Waste Fund fees. In any event, there are no domestic reprocessing facilities. Third, at over $Y2 million per day replacement power cost, shutting down Clinton Power Station is many times more expensive than addition of high density racks to the Fuel Cask Storage Pool and SFP.

11.4 Cost Estimate The plant modification proposed for the CPS fuel storage expansion utilizes freestanding, poisoned spent fuel racks installed the Fuel Cask Storage Pool during Phase I of the project and the same style racks installed in the SFP during Phase 2.

Holtec Report HI-2033124 11-4 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

The total capital cost is estimated to be approximately $15 million as detailed below.

Phase I Phase 2 Engineering, design, project management: $1 million $1 million Rack fabrication: $2 million $6 1/2 million Rack installation: $X2 million $4 million As described in the preceding section, other fuel storage expansion technologies were evaluated prior to deciding on the use of additional racks. Storage rack capacity expansion provides a cost advantage over other technologies.

11.5 Resource Commitment The expansion of the CPS spent fuel storage capacity via augmentation of racks in the SFP is expected to require the following primary resources per Unit:

Stainless steel: 1 12 tons Neutron absorber: 12 tons, of which 5 ton is Boron Carbide powder and 7 tons are aluminum.

The requirements for stainless steel and aluminum represent a small fraction of total world output of these metals (less than 0.001%). Although the fraction of world production of Boron Carbide required for the fabrication is somewhat higher than that of stainless steel or aluminum, it is unlikely that the commitment of Boron Carbide to this project will affect other alternatives. Experience has shown that the production of Boron Carbide is highly variable, depends upon need, and can easily be expanded to accommodate worldwide needs.

11.6 Environmental Considerations The proposed rack installation results in an additional heat load burden to the Spent Fuel Pool Cooling and Cleanup System due to increased spent fuel pool inventory, as discussed in Section 5.0. The Holtec Report HI-2033 124 11-5 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

maximum bulk pool temperature will be limited to less than 150'F under normal refueling scenarios.

The peak heat load from the spent fuel pool is less than 45 million Btulhr, which is a minuscule fraction of the total operating plant heat loss to the environment and is well within the capability of the SFP cooling system. Consequently, the short duration of increased heat loading during an outage is not expected to have any significant impact on the environment.

The increased peak bulk pool temperature during a refueling results in a slightly higher increased pool water evaporation rate for a short period of time. This increase is within the Fuel Building HVAC system capacity and does not necessitate any hardware modifications for the HVAC system. Therefore, the environmental impact resulting from the increased heat loss and water vapor generation at the pool surface is negligible.

Holtec Report HI-2033 124 11-6 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION

11.7 References

[11.1] OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications, USNRC (April 1978).

[11.2] Electric Power Research Institute, Report No. NF-3580, May 1984.

[11.3] "Spent Fuel Storage Options: A Critical Appraisal", Power Generation Technology, Sterling Publishers, pp. 137-140, U.K. (November 1990).

Holtec Report HI-2033 124 11-7 1342 SHADED AREAS DENOTE PROPRIETARY INFORMATION