ML043130398

Holtec Licensing Report HI-2043162, Revision 1, Spent Fuel Storage Expansion at Diablo Canyon Power Plant for Pacific Gas and Electric Co

ML043130398
Person / Time
Site:	Diablo Canyon
Issue date:	10/15/2004
From:	Holtec
To:	Office of Nuclear Reactor Regulation
References
DCL-04-149, Holtec Project 1368 HI-2043162
Download: ML043130398 (233)

Text

Enclosure 6 PG&E Letter DCL-04-149 Holtec Licensing Report HI-2043162 Rev. I (Non-proprietary)

SPENT FUEL STORAGE EXPANSION at DIABLO CANYON POWER PLANT for PACIFIC GAS AND ELECTRIC CO.

HOLTEC PROJECT NO. 1368 HOLTEC REPORT HI-2043162 REPORT CATEGORY: A REPORT CLASS: SAFETY RELATED COMPANY PRIVATE This document version has all proprietary information removed and has replaced those sections, figures, and tables with highlighting and/or notes to designate the removal of such information. This document is to be used only in connection with the performance of work by Holtec International or its designated subcontractors. Reproduction, publication or presentation, in whole or in part, for any other purpose by any party other than the Client is expressly forbidden.

HOLTEC INTERNATIONAL DOCUMENT NUMBER: HI-2043162 PROJECT NUMBER: 1368 DOCUMENT ISSUANCE AND REVISION STATUS DOCUMENT NAME: -Spent Fuel Storage Expansion __________

_at Diablo Canyon Power Plant DOCUMENT CATEGORY: El GENERIC 3 PROJECT SPECIFIC REVISION No. _0 REVISION No. _I REVISION No.

Author's Date Author's Date VIR i Author's Date VIR No. Document Initials Approved VIR l Initials Approved Initials Approved

1. CHAPOI SP 10115/04 806077 CWB 10/27/04 542637

2. CHAP02 SP 10/15/04 314657 CWB 10/27104 343808

3. CHAP03 SP 10/15/04 139839 CWB 10/27/04 680474

4. CHAP04 SPA 10/15/04 320072 ..

5. CHAP05 ER 10/15/04 393867 --

6. CHAP06 SP 10/15/04 814221

7. CHAP07 SP 10/15/04 730005 CWB 10/27/04 721102

8. CHAP08 SP 10/15/04 534945 ...

9. CHAP09 VJB 10115/04 313258

10. CHAPIO SP 10/15/04 157764 II CHAPI I SP 10/15/04 758318 __ .

12.

tt Chapter or section number.

Page 1 of 2

HOLTEC INTERNATIONAL DOCUMENT NUMBER: _HI-2043 162 PROJECT NUMBER: 1368 _

DOCUMENT CATEGORIZATION In accordance with the Holtec Quality Assurance Manual and associated Holtec Quality Procedures (HQPs), this document is categorized as a:

E Calculation Package 3 (Per HQP 3.2) Z Technical Report (Per HQP 3.2)(Such as a Licensing Report)

El Design Criterion Document (Per HQP 3.4) 0I Design Specification (Per HQP 3.4)

E] 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 E Nonproprietary 3 Holtec Proprietary E] Privileged Intellectual Property (PIP)

Documents labeled Privileged Intellectual Property 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 responsibility to safeguard it against loss or duplication.

Notes:

1. This document has been subjected to review, verification and approval process set forth in the Holtec Quality Assurance Procedures Manual. Password controlled signatures of Holtec personnel who participated in the preparation, review, and QA validation of this document are saved in the N-drive of the company's network. The Validation Identifier Record (VIR) number is a random number that is generated by the computer after the specific revision of this document has undergone the required review and approval process, and the appropriate Holtec personnel have recorded their password-controlled electronic concurrence to the document.

2. A revision to this document will be ordered by the Project Manager and carried out if any of its contents is materially affected during evolution of this project. The determination as to the need for revision will be made by the Project Manager with input from others, as deemed necessary by him.

3. Revisions to this document may be made by adding supplements to the document and replacing the "Table of Contents", this page and the "Revision Log".

Page 2 of 2

SUMMARY

OF REVISIONS Revision I contains the following pages:

COVER PAGE I page DOCUMENT ISSUANCE AND REVISION STATUS 2 page

SUMMARY

OF REVISIONS 1 page TABLE OF CONTENTS 8 pages

1.0 INTRODUCTION

8 pages 2.0 CASK PIT STORAGE RACKS AND SUPPORT PLATFORMS 27 pages 3.0 MATERIAL AND HEAVY LOADS CONSIDERATIONS 14 pages 4.0 CRITICALITY SAFETY ANALYSES 26 pages

- APPENDIX 4A 26 pages 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 26 pages 6.0 STRUCTURAL/SEISMIC CONSIDERATIONS 43 pages 7.0 FUEL HANDLING AND CONSTRUCTION ACCIDENTS 19 pages 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 10 pages 9.0 RADIOLOGICAL EVALUATION 8 pages 10.0 INSTALLATION 8 pages 11.0 ENVIRONMENTAL COST/BENEFIT ASSESSMENT 5 pages TOTAL 232 pages Revision 0 Initial issue.

Revision I Minor editorial changes made on pages RI, 1-3, 2-1, 2-2, 2-14, 3-5, 3-7, and 7-8.

Holtec Report HI-2043 162 RI Project 1368

TABLE OF CONTENTS

1.0 INTRODUCTION

.............................................. 1-1 1.1 References ................................................. 1-6 2.0 CASK PIT STORAGE RACKS AND SUPPORT PLATFORMS ..... ..... 2-1 2.1 Introduction ................................................. 2-1 2.2 Summary of Principal Rack Design Criteria ........ ................... 2-3 2.3 Applicable Codes and Standards ............... ..................... 2-4 2.4 Quality Assurance Program . ....................................... 2-10 2.5 Mechanical Design .............................................. 2-11 2.6 Rack Fabrication .............................................. 2-12 2.6.1 Region 11 Rack Module .......................................... 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 Characteristics of MetamicT. ...................................... 3-4 3.4 Compatibility with Environment .............. ...................... 3-5 3.5 Heavy Load Considerations for the Proposed Rack Installations ..... ...... 3-5 3.6 References ................................................. 3-12 4.0 CRITICALITY SAFETY ANALYSES .......... ..................... 4-1 4.1 Introduction and Summary ......................................... 4-1 4.2 Acceptance Criteria .............................................. 4-3 4.3 Assumptions .............................................. 4-4 4.4 Design and Input Data ............................................ 4-6 4.4.1 Fuel Assembly and Fuel Insert Specification ........ ................... 4-6 4.4.2 Cask Area Storage Rack Specification .......... ..................... 4-7 4.5 Methodology ........ 4-7 4.6 Analysis ........ 4-10 4.6.1 Bounding Fuel Assemblies ...................................... 4-10 4.6.2 Moderator Temperature Effect ...................................... 4-10 4.6.3 Pool Water Temperature Effects ................................... 4-11 4.6.4 Uncertainties due to Manufacturing Tolerances ......... .............. 4-11 4.6.5 Uncertainty in Depletion Calculations ................. .............. 4-12 4.6.6 Isotopic Compositions . ...................................... 4-12 4.6.7 Eccentric Fuel Assembly Positioning ................. .............. 4-13 4.6.8 Reactivity Effect of Axial Bumup Distribution ......... ............... 4-13 4.6.9 Bumup Versus Enrichment Curve .................................. 4-13 4.6.10 Interfaces ............. ............................ 4-14 4.6.11 Soluble Boron Concentration for Maximum kcff of 0.95 ...... ........... 4-15 4.6.12 Abnormal and Accident Conditions ................................. 4-16 4.6.12.1 Temperature and Water Density Effects .............. ............... 4-16 4.6.12.2 Dropped Assembly - Horizontal ................................... 4-16 Project 1368 Holtec Report Holtec HI-2043 162 Report HI-2043162 i Project 1368

TABLE OF CONTENTS 4.6.12.3 Dropped Assembly - Vertical ...................................... 4-17 4.6.12.4 Abnormal Location of a Fuel Assembly ........... .................. 4-17 4.6.12.4.1 Misloaded Fresh Fuel Assembly . .................................. 4-17 4.6.12.4.2 Mislocated Fresh Fuel Assembly . .................................. 4-18 4.6.13 Code Comparison............................................... 4-18 4.7 References ..... 4-19 Appendix 4A "Benchmark Calculations" ....... Total of 26 Pages including cover and 6 figures 4A.1 Introduction and Summary ................. 4A-1 4A.2 Effect of Enrichment ................. 4A-3 4A.3 Effect of `1B 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 THERMAL-HYDRAULIC CONSIDERATIONS ....... ............... 5-1 5.1 Introduction ................................................... 5-1 5.2 Cooling Systems Description ....................................... 5-2 5.3 Offload Scenarios . ......................................... 5-2 5.4 Maximum Pool Bulk Temperatures . .................................. 5-3 5.5 Minimum Time-to-Boil and Maximum Boil-off Rate ....... ............. 5-6 5.6 Maximum SFP Local Water Temperature ........... .................. 5-7 5.7 Fuel Rod Cladding Temperature . .................................. 5-10 5.8 Results ................................................... 5-12 5.8.1 Maximum Pool Bulk Temperatures ................ ................. 5-12 5.8.2 Minimum Time-to-Boil and Maximum Boil-off Rate ....... ............ 5-12 5.8.3 Local Water and Fuel Cladding Temperatures ........ ................ 5-13 5.9 References .............. ........................... 5-14 6.0 RACK STRUCTURAL INTEGRITY CONSIDERATIONS ..... ......... 6-1 6.1 Introduction .............. ........................... 6-1 6.2 Structural Characteristics of New Cask Pit Rack ........ ................ 6-1 6.3 Applicable Load Combinations ..................................... 6-3 6.4 Synthetic Time-Histories ......................................... 6-4 6.5 Analysis Methodology ......................................... 6-5 6.5.1 Overview . ................................................ 6-5 6.5.2 Rack/Platform Coupled Dynamic Model ............ .................. 6-9 6.5.3 Governing Equations of Motion . .................................. 6-12 6.6 Acceptance Criteria ......................................... 6-13 6.6.1 Kinematic and Stress Acceptance Criteria .......... .................. 6-13 6.6.2 Dimensionless Stress Factors ...................................... 6-16 HI-2043 162 ii Project 1368 Holtec Report HI-2043162 Holtec Report ii Project 1368

TABLE OF CONTENTS 6.6.3 Allowable g-Load on the Stored Fuel .......... ..................... 6-17 6.7 Key Input Data and Dynamic Simulations ........ .................... 6-17 6.7.1 Key Input Data ................................................ 6-17 6.7.2 Dynamic Simulations ............................................ 6-19 6.8 Margins Against Kinematic and Stress Limits ....... ................. 6-20 6.8.1 Satisfaction of Kinematic Acceptance Criteria ....... ................. 6-20 6.8.2 Satisfaction of Stress Acceptance Criteria ........ .................... 6-21 6.9 Assessment of Margin Against Fatigue Failure ....... ................. 6-24 6.10 References ................................................ 6-27 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 7.5.1 Shallow Drop Event .............................................. 7-5 7.5.2 Deep Drop Events ............................................... 7-5 7.5.3 Rack Drop Event ................................................ 7-5 7.5.4 Rack Platform Drop Event . ......................................... 7-6 7.6 Conclusion ................................................. 7-6 7.7 References ................................................. 7-7 8.0 POOL STRUCTURE INTEGRITY CONSIDERATIONS ..... ........... 8-1 8.1 Introduction ................................................ 8-1 8.2 Description of the Cask Pit Cavity Structure ........ ................... 8-1 8.3 Structural Consequences of the Cask Pit Rack to the Cask Pit Cavity ....... 8-2 8.4 Load Cases and Combinations ............... ....................... 8-4 8.4.1 Normal Load Conditions . ......................................... 8-4 8.4.2 Abnormal Load Conditions . ....................................... 8-5 8.4.3 Affected Load Cases ............................................. 8-6 8.5 Strength Evaluation .............................................. 8-7 8.6 Conclusion ................................................. 8-8 8.7 References ................................................. 8-9 9.0 RADIOLOGICAL AND SHIELDING CONSIDERATIONS ..... ......... 9-1 9.1 Fuel Handling Accident ............................................ 9-1 9.2 Solid Radwaste ................................................ 9-1 9.3 Gaseous Releases ................................................ 9-2 9.4 Liquid Radiological Releases ................. ...................... 9-3 9.5 Personnel Exposures ............................................. 9-3 9.6 Anticipated Exposure During Rack Installation ....... ................. 9-4 9.7 Rack Removal and Storage Prior to Cask Handling Operations ..... ....... 9-6 10.0 INSTALLATION ............................................... 10-1 Holtec Report HI-2043162 iii Project 1368 I

TABLE OF CONTENTS 10.1 Introduction . ............................................ 10-1 10.2 Rack Arrangement ............................................ 10-4 10.3 Rack Interferences ............................................ 10-4 10.4 SFP Cooling ............................................ 10-5 10.5 Installation of Cask Pit Rack . ..................................... 10-5 10.6 Safety, Health Physics, and ALARA Methods ........ ................ 10-6 10.6.1 Safety . ............................................ 10-6 10.6.2 Health Physics ............................................. 10-7 10.6.3 ALARA .... 10-7 10.7 Radwaste Material Control ........................................ 10-8 11.0 ENVIRONMENTAL AND ECONOMIC CONSIDERATIONS ..... ..... 11-1 11.1 Need for Temporary Spent Fuel Storage Capacity ....... .............. 11-1 11.2 Environmental Considerations ...................................... 11-1 11.3 Project Alternatives ............................................ 11-4 11.4 Design Basis Events ............................................ 11-5 11.5 Summary . ............................................ 11-5 11.6 References . ............................................ 11-5 iv Project 1368 Holtec Report HI-2043 162 Report HI-2043162 iv Project 1368

TABLE OF CONTENTS Tables 2.1.1 Geometric and Physical Data for Cask Pit Storage Racks ....... ............... 2-16 2.5.1 Module Data for Region II Cask Pit Racks ............... .................. 2-17 3.5.1 Heavy Load Handling Compliance Matrix (NUREG-0612) ....... ............. 3-14 4.1.1 Summary of the Criticality Safety Analyses for Holtec Cask Area Rack without Soluble Boron . ................................................ 4-21 4.1.2 Summary of the Criticality Safety Analyses for Holtec Cask Area Rack with Soluble Boron ................................................... 4-22 4.4.1 PWR Fuel Assembly Specifications ........... ........................... 4-23 4.4.2 Core Operating Parameter for Depletion Analyses ........................... 4-24 4.4.3 BPRA Design Specifications (based on [4.111 except for bumup) ..... .......... 4-24 4.4.4 Fuel Rack Specifications . .............................................. 4-25 4A.1 Summary of Criticality Benchmark Calculations ..... ............. 4A-9 thru 4A-13 4A.2 Comparison of MCNP4a and Keno5a 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 KENO5a 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-1 8 4A.7 Calculations for Critical Experiments with MOX Fuel ...................... 4A-19 5.4.1 Historic and Projected Fuel Offload Schedule ............................... 5-15 5.4.2 Key Input Data for Bulk Temperature Evaluation ............................ 5-16 5.5.1 Key Input Data for Time-to-Boil Evaluation ................................ 5-17 5.6.1 Key Design Data for Local Temperature Evaluation .......................... 5-18 5.8.1 Result of Transient Bulk Temperature Evaluations ........................... 5-19 5.8.2 Results of Loss of Forced Cooling Evaluations .............................. 5-20 5.8.3 Results of Maximum Local Water and Fuel Cladding Temperature Evaluations .... 5-21 6.4.1 ZPAValues (in g's) for the Applicable Earthquakes ........................... 6-5 6.5.1 Partial Listing of Fuel Rack Applications Using DYNARACK ..... ... 6-6 through 6-8 6.5.2 Degrees-of-Freedom ................................................. 6-11 6.7.1 Principal Input Data ................................................. 6-17 6.7.2 Rack Material Property Data (ASME - Section II, Part D) ...... ................ 6-18 6.7.3 Dynamic Simulation Scenarios ......... ...................... 6-19 through 6-20 6.8.1 Maximum Stress Factors at Critical Locations in the Cask Pit Rack .. 6-22 through 6-23 6.8.2 Maximum Stress Factors in the Platform Under Limiting Interface Loads Holtec Report HI-2043162 v Project 1368

TABLE OF CONTENTS from the Table 6.7.3 Dynamic Simulations ............. .................... 6-24 7.4.1 Impact Event Data . ............................................ 7-8 7.4.2 Material Definition ............................................. 7-9 8.2.1 Reinforced Concrete Wall and Basemat Data for the Cask Pit Cavity Region (Unit One) .............. .................... 8-2 8.5.1 Safety Factors for the Liner and Liner Anchorage .......... ................... 8-7 8.5.2 Safety Factors for the Cask Pit Concrete Wall and Floor ....................... 8-7 Holtec Report HI-2043162 vi Project 1368

TABLE OF CONTENTS Figures 1.0.1 Cask Pit Layout for Diablo Canyon Power Plant Unit I 1.0.2 Cask Pit Layout for Diablo Canyon Power Plant Unit 2 2.1.1 Pictorial View of Typical Region II Rack Module 2.1.2 Rack Platform in Isometric View 2.1.3 Recess Configuration in the Cask Pit Platform for Confining the Rack Support Pedestal 2.1.4 Rack to Platform Connection in Isometric 2.1.5 Rack to Platform Connector Link 2.1.6 Base Plate with Pedestal 2.6.1 Composite Box Assembly 2.6.2 Typical Array of Storage Cells (Non-Flux Trap Construction) 2.6.3 Elevation View of Storage Rack Module 2.6.4 Support Pedestals for DCPP Cask Pit Racks 4.1.1 Cask Pit Rack Loading Curve 4A.1 MCNP Calculated k-eff Values for Various Values of the Spectral Index 4A.2 KENO5a 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 KENO5a Calculated k-eff Values at Various U-235 Enrichments 4A.5 Comparison of MCNP and KENO5a Calculations for Various Fuel Enrichments 4A.6 Comparison of MCNP and KENO5a Calculations for Various Boron-1O Areal Densities 5.6.1 CFD Model Isometric View 5.8.1 Scenario I - Partial Core Discharge 5.8.2 Scenario 2 - Full Core Discharge 5.8.3 Scenario 3 - Emergency Full Core Discharge 5.8.4 CFD Model with Converged Temperature Contours 6.4.1 Acceleration Time History DE-X 6.4.2 Acceleration Time History DE-Y 6.4.3 Acceleration Time History DE-VT 6.4.4 Acceleration Time History DDE-E/W 6.4.5 Acceleration Time History DDE-N/S 6.4.6 Acceleration Time History DDE-VT 6.4.7 Acceleration Time History HOSGRI-X 6.4.8 Acceleration Time History HOSGRI-Y 6.4.9 Acceleration Time History HOSGRI-VT 6.4.10 Acceleration Time History LTSP-EW 6.4.11 Acceleration Time History LTSP-NS Holtec Report HI-2043162 Vii Project 1368 Holtec Report HI-2043 162 vii Project 1368

TABLE OF CONTENTS 6.4.12 Acceleration Time History LTSP-VT 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 Schematic of the Dynamic Model of Cask Pit Rack Platform Used in DYNARACK 7.2.1 Finite Element Model of the "Shallow" Drop Event 7.2.2 Schematics of the"Deep" Drop Scenario 7.2.3 Schematics of the Rack Drop Event 7.2.4 Finite Element Model of the Rack Platform Drop Event 7.5.1 "Shallow" Drop: Maximum Plastic Strain 7.5.2 "Deep" Drop: Maximum Baseplate Vertical Displacement 7.5.3 Rack Drop: Maximum Plastic Strain - Liner 7.5.4 Rack Drop: Cracks in the CLA Slab Concrete 7.5.5 Rack Platform Drop: Cracks in the CLA Slab Concrete 7.5.6 Rack Platform Drop: Maximum Plastic Strain - Liner 8.3.1 Cask Pit Cavity Isometric Project 1368 Holtec Report HI-2043 162 viii Viii Project 1368

1.0 INTRODUCTION

The Diablo Canyon Nuclear Plant (DCPP) is operated by the Pacific Gas and Electric Company (PG&E). The plant is located on the central California coast in San Luis Obispo County, approximately 12 miles west-southwest of the city of San Luis Obispo. The plant consists of two Westinghouse pressurized water reactors (PWR) nuclear units. Unit I has been in commercial operation since 1985 and Unit 2 since 1986.

Each unit utilizes a spent fuel pool (SFP) for the storage of irradiated and un-irradiated nuclear fuel in order to maintain a sub-critical array, remove decay heat and provide radiation shielding.

The DCPP SFPs are designed as mirror images of one another and have identical storage capabilities. Each SFP is currently licensed for 1,324 fuel assembly storage locations, arranged in 16 distinct rack modules.

Until 2002, the DCPP SFP racks were licensed for storage utilizing a two-region configuration with one region utilizing Boraflex as a neutron absorber. In 2002, a License Amendment was approved which allowed for credit of soluble boron in the SFP and resulted in a single region rack configuration without relying on the use of Boraflex.

PG&E is currently undertaking a project to license and construct an Independent Spent Fuel Storage Installation (ISFSI) at DCPP. It is planned for the ISFSI facility to be in operation by early 2007 to maintain a Full Core Offload Capability (FCOC) for both Units at DCPP.

However, to avoid the potential for the loss of FCOC due to unforeseen design, permitting or construction delays with the ISFSI Project, PG&E intends to temporarily expand the spent fuel storage capacity by adding a new fuel storage rack within the Cask Pit of each Unit. Based on the current inventory of fuel assemblies stored in the spent fuel pools and the anticipated future discharges of spent fuel, Unit I will lose full core reserve capacity in 2007 and Unit 2 will lose full core reserve capacity in 2008. The addition of the new racks will provide a two cycle extension to the loss of full core offload capability.

Holtec Report H1-2043162 1-1 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

PG&E intends to expand spent fuel storage capacity by adding one new 13 by 12 rack (154 total cells due to the elimination of two comer cells in each rack) Region II style rack within the Cask Pit of the SFP in Unit I by May 2007 and in Unit 2 by February 2008. Details of the Region 1I rack design are discussed in Section 2.6.1. This modification would temporarily increase the licensed storage capacity in each Unit from the current 1,324 storage cells to 1,478 storage cells.

However, other that an unexpected event requiring a full core offload, the only times when the actual inventory of fuel in the pools would exceed the current licensing basis of 1324 would be during the short duration when the core is offloaded during the 14th and 15th refueling outages on each Unit.

In order to simplify the design and licensing of the cask pit racks, the following general design considerations are employed:

From a criticality standpoint, the cask pit racks are designed for use as a "single zone" rack and will not rely on defined loading patterns within the rack.

Storage of fuel assemblies with only relatively low initial enrichments and/or high bumups will be licensed for storage within the racks. New, "unirradiated" fuel will not be stored within the racks.

The cask pit racks will not be used for general fuel storage operations. The racks are to be fully loaded once, freeing up storage locations in the existing SFP racks, and unloaded and removed from the cask pit when the dry cask storage facility (ISFSI) is implemented.

This minimizes fuel movements associated with the addition of the racks.

The cask pit racks are only to be used for storage of fuel with relatively long decay times

(>10 years) and low thermal output.

This report provides the design basis, analysis methodology, and results for the proposed temporary cask pit storage racks at Diablo Canyon to support the licensing process.

Holtec Report HI-2043 162 - 1-2 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

The physical description of the cask pit racks is provided in Section 2 of this report. The functional fuel storage capabilities and safety margins are discussed in Section 4 of this report.

The proposed fuel storage rack arrays for Units I and 2 are shown in the plan views provided by Figures 1.0.1 and 1.0.2, respectively. Note that two corner cells in each rack are eliminated to avoid interference with the cask restraint system.

The new storage racks are similar in construction to other racks previously licensed by the NRC and installed at numerous plants. The racks are not seismically braced at the top, but do have seismic restraints at the base pedestal connections to the supporting platform, as discussed in Section 6 of this report. The principal construction materials for the cask pit racks are ASME SA-240 Type 304L stainless steel sheet and plate stock, and ASME SA564-630, Condition H 1I00 (precipitation hardened stainless steel) for the adjustable support spindles. The only non-stainless material utilized in the rack is the neutron absorber material, which is a boron carbide and aluminum-composite available under the patented product name MetamicTm.

The racks are designed to the stress limits of, and analyzed in accordance with, Section 1II, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel (B&PV) Code [1]. The material procurement, analysis, fabrication, and installation of the rack modules conform to IOCFR50 Appendix B requirements.

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 [2].

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

Holtec Report HI-2043 162 1-3 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Section 4 provides a summary of the methods and results of the criticality evaluations performed for the cask pit 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 borated water at a temperature corresponding to the highest reactivity. In addition, it is demonstrated that kfr is less than 1.0 under the assumed accident of the loss of soluble boron in the pool water, i.e.

assuming unborated water in the spent fuel pool. 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 length and the amount of B' 0 per unit area (i.e., loading density) of the neutron absorber panel for the cask pit racks.

Thermal-hydraulic considerations require that local boiling will not occur in the cask pit racks, and that the pool bulk temperature will remain within the licensing basis described in the DCPP FSAR Update to satisfy the pool structural strength, operational, and regulatory requirements.

The rack module structural analysis requires that the primary stresses in the rack module structure will remain below the ASME B&PV Code (Subsection NF) [1] allowables.

Demonstrations of seismic and structural adequacy are presented in Section 6. 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 cask pit area of the SFP structure to withstand the new rack loads. The radiological considerations are documented in Section 9. Section 10 discusses installation of the cask pit racks. Section 11 discusses a costlbenefit 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.

Holtec Report H--2043162 1-4 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

t The analyses presented herein clearly demonstrate that the cask pit racks possess wide margins of safety with respect to all design considerations specified in the OT Position Paper [2], namely, nuclear subcriticality, thermal-hydraulic safety, seismic and structural adequacy, radiological compliance, and mechanical integrity.

Holtec Report HI-2043 162 1-5 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

1.1 References

[1] American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel Code, Section 111, 1989 Edition, Subsection NF, and Appendices.

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

Holtcc Report HI-2043 1f62 1-6 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

-CASK

- - .AREA zDCD z '<_

C nU r-)

%D -

I -_ __ _ - ___ -__ __ I1 EXISTING RACKS FIGURE 1.0.1 CASK PIT LAYOUT FOR DIABLO CANYON POWER PLANT UNIT 1.

H1368l HI-2043162 G'\DRAWINGS\O-FIGURES\1368\FIGI 119.625 NOM.

108.396 RACK DIM. -5.86 2.39 .NrM3 9NOM.

CASK I N - = ii AREA-1111fl 21111

+/- C CD'_

EXISTING RACKS FIGURE 1.0.2 CASK PIT LAYOUT FOR DIABLO CANYON POWER PLANT UNIT 2.

H1368]l HI-2043162 rG\DRAVINGS\O-FIGLRES\1368\FIGLO0-2

2.0 CASK PIT STORAGE RACKS AND SUPPORT PLATFORMS 2.1 Introduction The Diablo Canyon Power Plant (DCPP) cask pit fuel storage racks will each be 13 x 12 cell Region II style racks with a storage capacity of 154 assemblies, after the elimination of two comer cells as noted in Table 2.1.1. Each rack will be 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 II storage rack module. The cask pit rack modules for Units I and 2 are designed and constructed to have identical configurations. Data on the cross sectional dimensions, weight and cell count for the two rack modules are presented in Table 2.1.1.

The baseplates on both cask pit rack modules extend approximately 1/4" beyond the rack module periphery wall such that the contiguous edges of the plates establish a geometric separation between the rack and the surrounding racks or wall. The rack module support pedestals are engineered to accommodate minor level adjustments. Each rack is supported by four threaded pedestals, which are remotely height-adjustable. Thus, the racks can be made plumb and the top of the racks can easily be made co-planar with the racks in the adjacent Spent Fuel Pool (SFP).

Each of the two pools at the DCPP site will be equipped with a cask pit platform (CPP) in the cask pit cavity. The CPP, custom engineered to ASME Section III Subsection NF, provides the staging and mounting surface for the cask pit rack. The CPP is an all-stainless steel weldment of rectangular cross-section manufactured from SA240-304 plate and bar stock. A gridwork of plates in both orthogonal directions provides the required level of in-plane load resistance capacity to the platform. An isometric of the CPP is provided in Figure 2.1.2 and detailed cross-section and cut-away isometric views of the rack pedestal mating and attachment to the CPP are provided in Figures 2.1.3 and 2.1.4. Figure 2.1.5 provides an isometric of the connector link that provides the seismic restraint that holds the rack down onto the platform. Figure 2.1.6 provides Holtec Report HI-2043 162 2-1 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION .

a view of the baseplate and upper (female) portion of the rack pedestal that will receive the connector link.

The platform serves five major operational functions pertaining to the use of the cask pit as a fuel storage and fuel loading space:

i. The baseplate structure of the fuel rack is fastened to the CPP through eight "connector links" that are configured to be installed and removed remotely. The connector links provide a robust connectivity between the CPP and the fuel rack structure, allowing for a minimal side sway of the rack under seismic conditions.

ii. With the aid of lateral wedges (to be installed between the recess on the periphery of the CPP shown in Figure 2.1.2 and the liner), the platform is rendered essentially immovable in the horizontal direction, which eliminates the potential of platform-to-cask pit wall impacts during a seismic event. The relatively modest state of stress condition that may develop in the platform from an elevated pool bulk temperature (discussed in Chapter 5) is accounted for in the stress analysis of the CPP structure. (The structural analysis of the CPP entails the consideration of the combined effect of thermal and seismic (inertial) loads.)

iii. The design of the CPP/fuel rack interface ensures that the rack structure will not be subject to "restraint-of-free end expansion" under the most severe thermal conditions that may obtain in the cask pit region.

iv. While the CPP is structurally unconstrained to lift off in the vertical direction, the considerable nominal dead weight of the platform (22,825 Ibs) and the location of the above-mentioned "wedges" (a total of 16 in number) and kinematic coupling of the CPP with the fuel rack through the connector links, collectively help ensure that the platform will not lift off during any of the applicable earthquakes for the DCPP pool. Thus, the incidence of uplift of the fuel rack and the resulting vertical impact load are completely eliminated.

v. The CPP is designed to stage a transfer cask, if necessary, to faciliate the transfer of the plant's used fuel in dry storage, or off-site transport, if required. The CPP is designed to be physically compatible with the HI-TRAC transfer cask (reference Docket No. 72-1014). Thus, while the CPP is intended to be installed as a permanent appurtenance in the cask pit, it is nevertheless equipped with lifting and handling facilities to enable convenient and ALARA installation into and removal from the cask pit.

Structural attributes of the cask pit platform are further discussed in Section 6.2.

Holtec Rcport HI-2043 162 2-2 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

The overall design of the rack modules is similar to those presently in service in the spent fuel pools at many other nuclear plants, such as Davis-Besse, Callaway, Byron and Braidwood.

Altogether, Holtec International has provided over fifty thousand storage cells of this design to various nuclear plants around the world.

2.2 Summary of Principal Rack Design Criteria The key design criteria for the cask pit 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: Both cask pit rack modules are designed without bracing to the wall at the top of the rack. The base pedestals are restrained from lateral and vertical movement by connections to the supporting platform, as discussed later in this section and in greater detail in section 6.

b. Kinematic Stability: Each 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. Thermal-Hydraulic Compliance: With forced cooling available (i.e., Spent Fuel Pool Cooling System operating) the maximum SFP bulk temperature for a partial core offload must not exceed 140'F, the maximum SFP bulk temperature for a full core offload must not exceed 1740 F (the current licensing basis value) and the maximum temperature for an emergency core offload must be less than 2121F.

These temperature limits are consistent with the DCPP FSAR.

Holtec Report HI-2043162 2-3 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

e. Criticality Compliance: The cask pit racks must be able to store spent fuel of 4.1 weight percent (w/o) maximum enrichment while maintaining the reactivity (Kdff) less than 0.95 with the storage racks fully loaded with fuel of the highest l permissible reactivity and the pool flooded with borated water at a temperature corresponding to the highest reactivity. In addition, it is demonstrated that krff is less than 1.0 under the assumed accident of the loss of soluble boron in the pool water, i.e. assuming unborated water in the spent fuel pool. The maximum calcu-lated reactivities include a margin for uncertainty in reactivity calculations, including manufacturing tolerances, and are calculated with a 95% probability at a 95% confidence level.

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. Credit for borated water is allowed under accident scenarios.

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. Desiizn Codes (1) American Institute of Steel Construction (AISC) Manual of Steel Construction, 1969 Edition.

(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).

Holtec Report HI-2043162 24 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

(3) American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel (BP&V) Code,Section III, 1988 Edition through 2000 Addenda.

(4) American Society for Nondestructive Testing SNT-TC-IA, 2001, Personnel Qualifications and Certification in Non-destructive Testing.

(5) American Concrete Institute Building Code Requirements for Reinforced Concrete (ACI 318-63).

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

(7) ASME Y14.5M, Dimensioning and Tolerancing for Engineering Drawings and Related Documentation Practices (8) ASME B&PV Code,Section II, Parts A and C, 1989 Edition.

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

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.

Holtec Report HI-2043162 2-5 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

(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 El 90 - Standard Test Method for Guided Bend Test for Ductility of Welds.

c. Welding Code:

(1) ASME B&PV Code,Section IX - Welding and Brazing Qualifications, 1989.

d. Quality 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).

(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. 1.116).

(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, 2001 Edition.

Holtec Report HI-2043162 2-6 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

(10) ANSI N16.9 Validation of Calculation Methods for Nuclear Criticality Safety.

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 by USNRC letter dated 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) ANSLIANS 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 Facilities - 1971.

(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 N 14.6 - 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 Personnel Engaged in ASME Boiler and Pressure Vessel Code Section 111, Div. 1,-Certifying Activities.

(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.

Holtec Report HI-2043162 2-7 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

(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) 10CFR50.68 - Criticality Accident Requirements

h. Regulatorv Guides (RG)

(1) RG 1.13 - Spent Fuel Storage Facility Design Basis (Proposed Revision 2, dated 12/81).

(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).

(4) RG 1.33 - Quality Assurance Program Requirements.

(5) RG 1.29 - Seismic Design Classification, Rev. 3 - September 1978.

(6) RG 1.31 - Control of Ferrite Content in Stainless Steel Weld Metal, Rev. 3

-April 1978.

(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, Rev. I - 1973.

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

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

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

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

Holtec Report HI-2043162 2-8 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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

(14) RG 1.84 - Design and Fabrication Code Case Acceptability (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).

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

(21) RG 1.144 - Auditing of Quality Assurance Programs for Nuclear Power Plants, January 1979 revision (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.

Holtec Report HI-2043 162 2-9 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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

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

(2) AWS D1.3 - 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 and Guides for Qualification and 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.

2.4 Oualitv Assurance Program The governing quality assurance requirements for design and fabrication of the spent fuel racks are stated in IOCFR50 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.

Holtec Report HI-2043162 2-10 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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 Diablo Canyon site-specific requirements.

2.5 Mechanical Design The DCPP cask pit 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 cask pit 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 cask pit rack modules in the context of the fabrication methodology.

Holtec Report HI-2043162 2-11 Project 1368 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 cask pit 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 2.6.1 Region 11 Rack Module As stated in Section 2.1 the DCPP cask pit racks are Region 11 style racks. As such the cask pit racks are comprised of five significant components: (a) the composite box subassembly, (b) the baseplate, (c) the neutron absorber material, (d) the sheathing, and (e) the support pedestals.

Fabrication details for each of these components is discussed below.

Holtec Report HI-2043162 2-12 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

a. Composite box subassembly: 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. The minimum weld penetration is 80% of the box metal gage. The nominal inside dimension of the fabricated box (or cell) is 8.75 inches. This process results in a square box as shown in Figure 2.6.1. Metal sheathing is then attached to select sides of each box, and the poison material is installed in the sheathing cavities.

The square cross section box with neutron absorber panels affixed to its external surfaces is referred to as the "composite box assembly." Each composite box has 1-inch diameter holes punched near its bottom edge to provide auxiliary flow.

Region II storage cells have two holes at the base aligned in one direction.

The composite boxes are arranged in a checkerboard array and welded edge-to-edge to form an assemblage of storage cell locations as shown in Figure 2.6.2.

Filler panels and comer angles are welded to the edges of boxes at the outside boundary of the rack to complete the formation of the peripheral cells. Some of the cell inside dimension at these locations are larger by the thickness of the adjacent cell walls (approximately 0.09") to become 8.85", because the filler panels and comer angles are welded to the outside of the adjacent fabricated cell walls, as seen in Figure 2.6.2. The inter-box welding and pitch adjustment are accomplished by small longitudinal connectors. The connectors are sized and placed to ensure that the inside cell dimension for developed boxes is maintained after inclusion of any reductions from the sheathing. This assemblage of composite boxes results in a honeycomb structure with axial, flexural and torsional rigidity, which depend on the extent of inter-cell welding. Figures 2.6.2 shows that all four comers of each interior box are connected to the adjacent boxes, which creates a well-defined path for "shear flow."

Holtec Report HI-2043 162 2-13 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

As stated in Section 1.0, two comer cells are eliminated from the rectangular cell array. Figure 2.6.2 indicates a typical array, but has been adjusted to indicate the manner in which these two comer cells are to be eliminated. In the instance shown in the upper left of the figure, a normally placed comer angle will be eliminated. In the lower right, the missing cell will be eliminated through the use of a larger comer angle fabricated to enclose two cells instead of the usual one cell. The baseplate will be cut at a 45-degree angle at the comer beneath the area of the eliminated cells.

A typical elevation view of Region II storage cells is shown in Figure 2.6.3.

b. Baseplate: A 1-inch thick baseplate provides a continuous horizontal surface for supporting the fuel assemblies. The baseplate has 5-1/4 inch diameter holes at each cell location, except at lift locations. At the four lift locations, a 5-1/2 inch by 2-5/8 inch rectangular cut-out is centered over a 3-13/16 inch circular flow hole to allow insertion and engagement of the lifting rig. The location of all baseplate holes coincides with the cell centerlines. The baseplate is attached to the base of the cell assemblage by fillet welds and extends horizontally beyond the periphery of the rack cells (refer to Figure 2.6.3).

c. Neutron absorber material: For the DCPP cask pit racks, every interior cell wall and boundary wall are equipped with one integral neutron absorber panel. The neutron absorber material is a boron carbide and aluminum-composite available under the patented product name MetamicT`I. A further discussion of the neutron absorber material is provided in Section 3.

Holtec Report HI-2043162 2-14 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

fI

d. Sheathing: As described earlier, metal sheathing is fastened to all four sides of each box wall. The design objective calls for attaching the neutron absorber material tightly on the box surface. This is accomplished by die forming the internal and external sheathing to provide end flares with smooth edges, as shown in Figure 2.6.1. The flanges of the sheathing are welded to the box using skip welds and spot welds. The sheathing serves to locate and position the poison sheet accurately and to preclude its movement under seismic conditions. The sheathing also provides a vented enclosure for the neutron absorber material.

e. Support pedestals: All support pedestals are the adjustable type as shown in Figure 2.6.4. The top (female threaded) portion is made of austenitic stainless steel. The bottom (male threaded) portion is made of SA564-630 stainless steel.

Each support pedestal is equipped with a readily accessible socket to enable remote leveling of the rack after its placement in the cask pit. The support pedestals are located at the centerlines of cells to ensure accessibility of the leveling tool through the 5-1/4 inch diameter flow hole in the baseplate. A vertical hole is drilled in the bottom portion of the support pedestals to allow for cooling flow.

The assembly of the rack modules is carried out by welding the composite boxes in a vertical fixture with the precision fabricated baseplate serving as the bottom positioner.

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 pedestal welds, in accordance with the design drawings.

Holtec Report HI-2043 162 2-15 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 2.1.1 Geometric and Physical Data for Cask Pit Storage Racks RACK RACK CELL-TO-CELL NO. OF CELLS MODULE DRY NO. OF UNIT I.D. TYPE PITCH ENVELOPE WEIGHT CELLS SIZE S(lb) PER RACKt N-S E-W N-S E-W N-S E-W (in) (in) Direction Direction (in) (in) 1 CPI Region II 8.946 8.946 12 13 108.4 117.3 26,000 154 2 CP2 Region II 8.946 8.946 12 13 108.4 117.3 26,000 154 t Two corner cells are eliminated from each rack, as shown in Figure 1.0.1, 1.0.2 and 2.6.2, to avoid interference with the cask restraint system.

Holtec Report HI-2043 162 2-16 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION.

Table 2.5.1 MODULE DATA FOR REGION II CASK PIT RACKS t Storage cell inside nominal dimension 8.75 in ft Cell pitch (North-South) 8.964 in Cell pitch (East-West) 8.964 in Storage cell height (above the baseplate) 169 in Baseplate hole size ttt (except for lift locations and two tie-down restraint locations on each pedestal) 5.25 in Baseplate thickness 1.0 in Female support pedestal height 4.0 in +/- 1/16 in.

Male support pedestal height (protrusion below female pedestal) 3.0 in +/- 1/2 in.

Support pedestal type Remotely adjustable pedestals w/

vertical and horizontal seismic restraints Number of support pedestals per rack 4 Number of cell walls containing 1-inch diameter flow 2 (aligned in one direction) holes at base of cell wall Remote lifting and handling provisions Yes Poison length 150 in Poison width 7.5 in (7.0 inches wide at all boundary wall locations) t All dimensions indicate nominal values tt As discussed in section 2.6.1(a), cells formed by comer angles and filler panels will have dimensions enlarged by the thickness of the adjacent cell walls to 8.85".

ttt There are two examples of exceptions to the standard baseplate hole diameter. The rack is lifted at four points using irregular shaped holes placed at the centerline of specific cells in lieu of the standard baseplate hole. The lifting holes are sized with approximately the same area as the standard holes. The second exception is the two baseplate holes that align with the pedestal vertical seismic restraint (hold-down) connector links, shown in Figure 6.2.3. These holes are 7.25 inches in diameter and the lower portion is a rectangular opening sized 6.5 by 3.25 inches.

Holtec Report HI-2043 162 2-17 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

SUPPORT-PEDESTAL (REFER TO FIGURE 2.1.3 FOR ADDITIONAL DETAILS)

FIGURE 2.1.1; PICTORIAL VIEW OF ATYPICAL REGION IIRACK MODULE NOTE: THE NUMBER OF CELLS SHOWN IS NOT INTENDED TO DEPICT ACTUAL RACK SIZES HI-2043162 l c IH136

FIGURE 2.1.2 - Rack Platform in Isometric View H1322l lHOLTEC REPORT HI-2043 162

RACK CELL FEMALE RACK CELL WALLS PEDESTi *AL BASE PLATE TOP OF PLATFORM J /l Ill i , , i.. -. . I ; :: : . - 1! -2" ; :1 - : 1: ~I - - I Il I. .-.

Ii CONNECTOR LINK I MALE PEDESTAL FIGURE 2.1.3 - Recess configuration in the "Cask Pit Platform" for confining the Rack Support Pedestal H 1322 IHOLTEC REPORT HI-2043162 H1322 HOLTEC REPORT HI-2043 162

U 4

... I I

FIGURE 2.1.4 - Rack to Platform Connection in Isometric H1 3 12 . I HOLTEC REPORT HI-2043162

THREADS NOT SHOWN FOR to P "

CLARITY FIGURE 2.1.5 - Rack to Platformn "Connector Link" H1322 l I HOLTEC REPORT HI-2043 162

litJ FIGURE 2.1.6 - Base Plate with Pedestal 11322 HOLTEC REPORT HI-2043162

-. 4..

CELL WALL NEUTRON ABSORBER PANEL

-TAPERED WELDED AND SMOOTHED ENDS FI GURE COMPOSITE BOX ASSEMBLY HI-2 431 r7 2 D\DRAVINGS\O-FIGQRES\1368\REPDRT-HI2043162\CHfT..2\FIG2-61 IH 136.8

-J ii -J Oj LU Cw co Cn- LLj v ln II1 FIGURE 2.6.2; TYPICAL ARRAY OF STORAGE CELLS (NON-FLUX TRAP CONSTRUCTION)

,I NOTE THIS FIGURE ISINTENDED TO INDICATE THE TYPICAL COMPONENTS OF THE RACK AND ISNOT INTENDED TO ACCURATELY H13681 REFLECT THE NUMBER OF CELLS OR SPECIFY LOCATIONS OF FABRICATED VS FORMED CELLS. HI-20243162 G:\DRhWMS\O-FIURES\1368\REPORT-HI-2043162\MURES2.6.2

I CELL I FPITCH __

DEVELOPED CELL I I " .. ,

I I

lI

- r II 11 -- POISON PANEL I I CELL Ifl Ie I:

ACTIVE POISON LENGTH LE ENGTH I

I iII I I  ! SHEATHING r- 1_ ,

t-I FLOW HOLE I V ~(TYP.)

.I I BASEPLATE P r - -., ------.--

I I - 4BASE PLATE HOLE FIGURE 2.6.3; ELEVATION VIEW OF STORAGE RACK MODULE NOTE: DEPICTION OF STORED FUEL ASSEMBLY IS NOT INTENDED TO BE ACCURATE.

H13681 IHI-2043162 I

G:\DRAWINGS\0-RGURES\I 368\REPORT-HI-20431 62\FGURE2.6.3

LEVELING TOOL SOCKET CELL 0 \ / -BASE PLATE 0

' \I I'ix I V I

I I I

-z tX I\ \ \ \ .

I I I I I , I I

- .§ I I I I I I I I I I I I I I I I I

~'.4DRARIN HO,LE TYPICAL ELEVATION VIEW TYPICAL B011TM MEWS OF RACK BASEPLATE CORNER

-FIGURE 2.6.4; SUPPORT PEDESTALS FOR DCPP CASK PIT RACKS Hl13681 lHI-20431 62 G:\DRANGS\0-F1WRES\1 368\REPORT-H-2043162\FIGURES2.6.4

3.0 MATERIAL AND HEAVY LOAD CONSIDERATIONS 3.1 Introduction Safe storage of nuclear fuel in the SFP requires that the materials utilized in the rack fabrication be of proven durability and compatible with the pool water environment. This section provides a synopsis of the considerations with regard to long-term design service life of 40 years.

3.2 Structural Materials The following structural materials will be utilized in the fabrication of the DCPP cask pit racks and associated support platforms:

a. ASME SA240 Type 304L for all sheet metal stock and baseplate

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

c. Extemally threaded support spindle: ASME SA564-630, Condition HI 100 (precipitation hardened stainless steel)

d. Weld material - austenitic stainless steel 3.3 Neutron Absorbing Material Holtec Report HI-2043162 3-1 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Holtec Report HI-2043 162 3-2 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Holtec Report HI-2043 162 3-3 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

3.3.1 Characteristics of MetamicTm Holtec Report HI-2043162 3-4 Project 1368 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 DCPP SFP, including the cask pit area. Austenitic stainless steel (Type 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. The materials used in Metamic, 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 Handling Building (FHB) Crane will be used to install the cask pit racks and associated support platforms in both units. The Fuel Building Crane is designed as Seismic Category I equipment. The capacity of the main hook is 125 tons. The FHB Crane is designed for spent fuel cask handling operations. More specifically, it is used to place casks within the cask pit area for removal of spent fuel from the plant.

Holtec Renort HI-2043162 3-5 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION I

Safe handling of heavy loads by the FHB Crane will be ensured by following the defense in depth approach guidelines of NUREG 0612:

• Defined safe load paths in accordance with the Plant Staff Review Committee approved procedures

• Supervision of heavy load lifts by designated qualified 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, as supplemented by CALOSHA requirements.

• Ensuring the design of the FHB Crane meets the intent of requirements of CMAA-70 [3.5.4] and ANSI/ASME B30.2-1976

• Reliability of special lifting devices by application of design safety margins, and periodic inspection and examinations using approved procedures, in accordance with ANSI N14.6 The salient features of the lifting rigs and associated procedures are described as follows:

a. Safe Load Paths and Procedures Safe load paths will be defined for moving the new racks and platforms into and within the Fuel Handling Building. All movements of the racks and platforms within the FHB will be performed using the FHB Crane. During the receipt, inspection and placement of the platforms and racks into the cask pit area, heavy loads will not be carried over any portions of the existing storage racks containing active fuel assemblies. All phases of rack installation activities will be conducted in accordance with written procedures, which will be reviewed and approved by the owner.

Holtec Report HI-2043162 3-6 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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

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. Crane operators will also be trained and qualified in accordance with PG&E's training program.

d. Lifting Devices Design and Reliability The FHB Crane can access the truck bay, the adjacent laydown area, and the cask pit area of both SFPs. The FHB Crane has sufficient capacity to handle all lifts during the reracking process. The FHB crane is anticipated to be upgraded to single failure proof capability before the rack and support platform installation commences.

Holtec Report HI-2043 162 3-7 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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

Item Weight (lb)

Rack 26,000 (max.)

Lift Rig 1,800 Rigging 500 Total Lift 28,300 The following table determines the maximum lift weight during the installation of the rack platforms.

Item Weight (lb)

Platform 22,825 (max.)

Lift Rig 2,300 Rigging 500 Total Lift 25,625 It is clear, based on the heaviest rack or platform weight to be lifted, that the heaviest load will be well below the 125-ton rating of the FHB Crane Main Hook.

Remotely engaging lift rigs, meeting all requirements of NUREG-0612, will be used to lift the racks and support platforms. The rack and platform 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 Holtec Report HI-2043162 3-8 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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.

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.

• A stress analysis of each of the rigs is carried out and the primary stress limits postulated in ANSI N14.6 [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 FHB Crane is maintained functional per the DCPP 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 Holtec Report HI-2043 162 3-9 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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.

Rigging failure: The lifting rigs designed for handling and installing the new racks and support platforms have 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 2, Ulchin 2, 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 produce high quality installations 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 platforms 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.

Holtec Report HI-2043162 3-10 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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-2043 162 3-11 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

3.6 References

[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 61 BT (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.

[3.3.11] "Use of METAMICt in Fuel Pool Applications," Holtec Information Report No.

HI-2022871, Revision 1 (2002).

[3.3.12] "Sourcebook for MetamicTm Performance Assessment" by Dr. Stanley Turner, Holtec Report No. HI-2043215 (2004).

Holtec Report HI-2043162 3-12 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

[3.5.1] DCPP Final Safety Analysis Report Update (FSAR).

[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-2043162 3-13 Project 1368 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 crane 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-2043162 3-14 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.0 CRITICALITY SAFETY ANALYSES 4.1 Introduction and Summary Overview This section documents the criticality safety evaluation for the storage of PWR spent nuclear fuel in Holtec Region 2 style spent fuel storage racks at the Diablo Canyon Power Plant (DCPP),

Unit I and 2. These racks are designed to be placed into the cask pits of the spent fuel pools of the units. The racks for the two units are identical in all characteristics relevant to criticality control. Therefore, no distinction needs to be made between the racks for the two units in this section, and all results apply to both units. The scope of the analysis is limited to fuel assemblies 235 with an initial enrichment of no more than 4.1 wt% U from cycles I through 7 of each unit.

The objective of this analysis is to ensure that the effective neutron multiplication factor (kfn) 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 borated water at a temperature corresponding to the highest reactivity. In addition, it is demonstrated that kff is less than 1.0 under the assumed accident of the loss of soluble boron in the pool water, i.e. assuming unborated water in the spent fuel pool.

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% con-fidence level [4.1].

A curve for the minimum required fuel burnup as a function of the initial fuel enrichment is developed, that ensures that the requirements stated above are met.

Reactivity effects of abnormal and accident conditions are also been evaluated to assure that under all credible conditions, the reactivity will not exceed the regulatory limit of 0.95, considering the presence of soluble boron.

Holtec Report H1-2043162 4-1 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Summary of Results The analysis demonstrates that for the defined acceptance criteria that are summarized below, the effective neutron multiplication factor (kff) is less than 1.00 with unborated water and less than 0.95 with 500 ppm of soluble boron, at a 95% probability with a 95% confidence level. Further, the reactivity effects of abnormal and accident conditions have been evaluated. To assure that under credible abnormal and accident conditions the reactivity will not exceed 0.95 at a 95%

probability with a 95% confidence level, a soluble boron level of 800 ppm must be maintained.

Calculations have been performed to qualify the racks for storage of burned fuel assemblies with a maximum nominal enrichment of 4.1 wt% 23 5U and a minimum burnup of 28.53 GWd/mtU, from cycles I through 7 from DCPP Unit 1 and Unit 2. Results for this condition are summarized in Table 4.1.1 and Table 4.1.2 for unborated and borated water, respectively. For other enrichments, the minimum burnup can be determined with the function BU=A+B*E+C* E2 with BU Minimum Burnup in GWd/mtU E Initial Enrichment in wt% 235U (2.0 through 4.1 wt%)

A Coefficient, = -40.627 B Coefficient, = 24.903 C Coefficient,= -1.960 This function is also shown as a graph in Figure 4.1.1.

Holtec Report HI-2043162 4-2 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.2 Acceptance Criteria The objective of this analysis is to ensure that the effective neutron multiplication factor (krff) 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 borated water at a temperature corresponding to the highest reactivity. In addition, it is demonstrated that kdf is less than 1.0 under the assumed accident of the loss of soluble boron in the pool water, i.e. assuming unborated water in the spent fuel pool.

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% con-fidence level [4.1].

Applicable codes, standard, 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, Rev. 3

-July 1981.

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

• L. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T.

Collins, August 19, 1998.

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

Holtec Report HI-2043162 4-3 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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

• Code of Federal Regulation IOCFR50.68, Criticality Accident Requirements (for soluble boron) 4.3 Assumptions To assure the true reactivity will always be less than the calculated reactivity, the following conservative design criteria and assumptions were employed:

1) Moderator is borated or unborated water at a temperature that results in the highest reactivity, as determined by the analyses.

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

3) The effective multiplication factor of an infinite radial array of fuel assemblies is calculated in the analyses.

4) As cooling time increases, decay ofPu-241 and build-up ofAm-241 and Gd-155 result in a continuous decrease in reactivity, which provides an increasing subcriticality margin with time.

No credit is taken for this decrease in reactivity, i.e. calculations are performed for 0 cooling time. This is a significant conservatism, since the assemblies qualified for the cask pit rack (discharged in cycles I through 7) have already cooling times between 8 and 18 years by mid 2004.

Holtec Report H1-2043162 4-4 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

5) The presence of burnable poison rod assemblies (BPRAs) during fuel depletion potentially increases the krffof an assembly [4.11]. Therefore, depletion calculations with and without BPRAs were performed to determine the most reactive condition. Two major assumptions were made when performing calculations with BPRAs: a) All assemblies are assumed tof contain a burnable poison rod assembly (BPRA) during depletion; and b) all BPRAs are assumed to contain the maximum number of poison rods. Both assumptions represent a major conservatism, since in reality a) only a small fraction of fuel assemblies were exposed to a BPRA; and b) most BPRAs contain less than the maximum number of poison rods.

6) No credit is taken for the potential presence of BPRAs in fuel assemblies when the assemblies are in the spent fuel racks.

7) Higher soluble boron concentrations during fuel depletion result in a higher reactivity of the assemblies due to spectrum hardening. Therefore, the highest cycle average soluble boron concentration of all applicable cycles is used in the depletion analyses.

8) To bound the reactivity effect associated with the axial burnup distribution in a fuel assembly, calculations with an axial burnup profile and calculations with an axially constant bumup were performed in the design basis calculations, and the higher of the two resulting reactivities is used for each case.

9) The moderator temperature during depletion has a significant impact on reactivity, with higher temperatures resulting in higher kcff values. Conservatively, the core average exit temperature is used as the moderator temperature in the depletion calculations.

10) For power density and fuel temperature, average values from cycles I through 7 are used in the depletion calculations. This is acceptable, since these values have a comparably small effect on reactivity. The potential effect due to variations in these parameters is more than compensated Holtec Report HI-2043162 4-5 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

by the effect of the other conservative assumptions, including neglecting cooling time and assuming there are BPRAs with the maximum number of rods in all assemblies during depletion.

Additionally, there are a small number of modeling assumptions, as listed below. These assumptions have a negligible effect on the calculated reactivities.

1) The single instrument tube in each fuel assembly is modeled with the same dimensions as the 24 guide tubes in the assembly.

2) Some minor variations of guide tube dimensions in some assemblies from the values used in the analyses are neglected.

3) All zirconium alloys are modeled as pure zirconium.

4) The majority of the rack cells have a cell ID of 8.75 Inches, while a few cells on the rack periphery have an ID of 8.85 Inches. In the calculations, all rack cells are modeled with a cell ID of 8.75 Inches 4.4 Design and Input Data 4.4.1 Fuel Assembly and Fuel Insert Specification The design specifications for the fuel assemblies, which were used for this analysis, are given in Table 4.4.1. The assembly types listed in this table, i.e. STD (LOPAR) and OFA (Vantage 5),

are the only types analyzed. Therefore, these are the only assembly types allowed to be stored in the cask pit racks.

The operating parameters used in the depletion analysis are given in Table 4.4.2. The data is applicable to cycles I through 7 for both units. For data that differ slightly between the two units, the higher, i.e. more conservative data is listed in Table 4.4.2 and used in the analysis. For the soluble boron level, the highest cycle average value for any of the cycles I through 7 in both units is used.

Holtec Report HI-2043162 4-6 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Specification of Burnable Poison Rod Assemblies (BPRAs) assumed in the assemblies during depletion are listed in Table 4.4.3. For the BPRAs, only nominal specifications are used, since these devices are only used in the depletion calculations to establish a bounding approach to represent the conservative effect of poison inserts. Any effect of the tolerances or small variations of these specifications would be small, and would be bounded by the effect of assuming a poison insert with the maximum number of poison rods in every assembly during depletion.

4.4.2 Cask Area Storage Rack Specification The storage cell characteristics which are used in the criticality evaluations are summarized in Table 4.4.4.

4.5 Methodology The principal method for the criticality analysis of the high-density storage racks is the three-dimensional Monte Carlo code MCNP4a [4.2]. MCNP4a is a continuous energy three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory. MCNP4a was selected because it has been previously used and benchmarked for criticality analyses and has all of the necessary features for this analysis. The critical benchmark experiments considered the effects of varying fuel enrichment, boron-10 loading, lattice spacing, fuel pellet diameter, and soluble boron concentration. The experimental data are sufficiently diverse to establish that the method bias and uncertainty will apply to the DCPP temporary rack conditions. MCNP4a calculations used continuous energy cross-section data based on ENDF/B-V and ENDF/B-VI, as distributed with the code [4.2]. Exceptions are two lumped fission products and one individual fission product calculated by the CASMO4 depletion code (see below), that do not have corresponding cross sections in MCNP. For these isotopes, the CASMO-4 cross sections are used in MCNP.

This approach has been validated by showing that the cross sections result in the same reactivity Holtec Report HI-2043162 4-7 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

effect in both CASMO-4 and MCNP. Independent verification calculations were performed with KENOV.a [4.8], which is a three-dimensional multigroup Monte Carlo code developed at the Oak Ridge National Laboratory. The KENOV.a calculations used the 238-group cross-section library, which is based on ENDF/B-V data and is distributed as part of the SCALE-4.3 package [4.9], in association with the NITAWL-II program [4.10], which adjusts the uranium-238 cross sections to compensate for resonance self-shielding effects.

Benchmark calculations, presented in Appendix A, indicate a bias of 0.0030 with an uncertainty of+

0.0012 for KENOV.a and 0.0009 +/- 0.0011 for MCNP4a, both evaluated with a 95% probability at the 95% confidence level [4.1]. The calculations for this analysis utilize the same computer platform and cross-section libraries used for the benchmark calculations discussed in Appendix A.

The convergence of a Monte Carlo criticality problem is sensitive to the following parameters:

(1) number of histories per cycle, (2) the number of cycles skipped before averaging, (3) the total number of cycles and (4) the initial source distribution. The MCNP4a criticality output contains a great deal of useful information that may be used to determine the acceptability of the problem convergence. This information has been used in parametric studies to develop appropriate values for the aforementioned criticality parameters to be used in storage rack criticality calculations. Based on these studies, a minimum of 10,000 histories were simulated per cycle, a minimum of 20 cycles were skipped before averaging, a minimum of 100 cycles were accumulated, and the initial source was specified as uniform over the fueled regions (assemblies). Further, the output was reviewed to ensure that each calculation achieved acceptable convergence. These parameters represent an acceptable compromise between calculational precision and computational time.

Fuel depletion analyses during core operation were performed with CASMO-4 (using the 70-group cross-section library), a two-dimensional multigroup transport theory code based on capture probabilities [4.3-4.5]. CASMO-4 is used to determine the isotopic composition of the spent fuel.

In addition, the CASMO4 calculations are restarted in the storage rack geometry yielding the tvo-Holtec Report HI-2043162 4-8 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

dimensional infinite multiplication factor (kinf) for the storage rack to determine the reactivity effect of fuel and rack tolerances, and to perform various studies. For all calculations in the spent fuel pool racks, the Xe-135 concentration in the fuel is conservatively set to zero.

The following equation was used to perform the kjff calculations:

kcff = kcaic + Akbis. + Aktl, + Ak.c.cct where kcaic = kcff under nominal conditions Akbi. = bias determined from benchmark calculations Akt,,v = temperature bias Akwm = statistical summation of tolerance and uncertainty components

= [toll 2 + tol22 + ... + uncertl2 + uncert2 2 + . ..*I2 As allowed in [4.6], the methodology employed to calculate the tolerance effects combine both the worst-case bounding value and sensitivity study approaches. For each individual parameter associated with a tolerance, the full tolerance value is utilized to determine the maximum reactivity effect. All of the Ak values from the various tolerances are then statistically combined to determine the final reactivity allowance for manufacturing tolerances.

In the geometric models used for the calculations, each fuel rod and its cladding were described explicitly and reflecting boundary conditions were used in the radial direction which has the effect of creating an infinite radial array of storage cells.

Holtec Report HI-2043162 4-9 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

I 4.6 Analysis This section describes the calculations that were used to determine the acceptable storage criteria for the racks and summarizes the results. In addition, this section discusses the possible abnormal and accident conditions.

Unless otherwise stated, all calculations assumed nominal characteristics for the fuel and the fuel storage cells. The effect of the manufacturing tolerances is accounted for with a reactivity adjustment as discussed below.

4.6.1 Bounding Fuel Assemblies To determine the bounding assembly, calculations are performed for both assembly types listed in Table 4.4.1 and for assemblies with and without BPRAs during depletion. Calculations are performed for three enrichments, 2.0 wt%, 3.0 wt% and 4.1 wt%, and a representative burnup for each enrichment. The results show that the LOPAR assemblies with a BPRA has the highest reactivity. This assembly type is therefore used in all subsequent calculations, and all depletion calculations are performed assuming the presence of a BPRA during the initial fuel burnup.

4.6.2 Moderator Temperature Effect For the depletion calculations, the core exit temperature is used as the moderator temperature (see Table 4.4.2). However, the residual fuel assembly reactivity that would be stored in the rack is dominated by the area slightly below the top of the active fuel region, where the moderator temperature in the core is lower. Since the calculated residual reactivity increases significantly with increasing moderator temperature, the assumption used in the depletion calculations is conservative.

Holtec Report HI-2043162 4-10 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.6.3 Pool Water Temperature Effects Pool water temperature effects on reactivity have been calculated with CASMO-4. The results demonstrate that the spent fuel pool temperature coefficient of reactivity is negative, i.e. a lower temperature results in a higher reactivity. Consequently, all CASMO-4 calculations are evaluated at 38 0F, which is well below the actual minimum SFP water temperature.

In MCNP4a, the Doppler treatment and cross-sections are valid only at 300K (80.6 IF).

Therefore, a Ak is determined in CASMO4 from 80.6 0F to 38 0F, and is included in the final kcfr calculation as a bias.

4.6.4 Uncertainties Due to Manufacturing Tolerances In the calculation of the final kin, the effect of manufacturing tolerances on reactivity must be included. CASMO-4 was used to perform these calculations. As allowed in [4.6], the methodology employed to calculate the tolerance effects combine both the worst-case bounding value and sensitivity study approaches. The evaluations include tolerances of the rack dimensions (see Table 4.4.4) and tolerances of the fuel dimensions (see Table 4.4.1). In addition to the tolerances specified in these tables, an enrichment tolerance of 0.05 wt% is analyzed. As for the bounding assembly, calculations are performed for different enrichments and burnups.

The reference condition is the condition with nominal dimensions and properties. To determine the Ak associated with a specific manufacturing tolerance, the kinf calculated for the reference condition is compared to the king from a calculation with the tolerance included. Note that for the individual parameters associated with a tolerance, no statistical approach is utilized. Instead, the full tolerance value is utilized to determine the maximum reactivity effect. All of the Ak values from the various tolerances are statistically combined (square root of the sum of the squares) to determine the final reactivity allowance for manufacturing tolerances.

Holtec Report HI-2043162 4-11 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Calculations are performed for three enrichments, 2.0 wt%, 3.0 wt% and 4.1 wt%, and a representative burnup for each enrichment. The maximum of the statistically combined values are used in the final rffcalculations (see Tables 4.1. 1. and 4.1.2).

4.6.5 Uncertainty in Depletion Calculations CASMO-4 was used to perform the depletion calculations. Since critical experiment data with spent fuel is not available for determining the uncertainty in burnup-dependent reactivity calculations, an allowance for uncertainty in reactivity was assigned based upon other considerations. Assuming the uncertainty in depletion calculations is less than 5% of the total reactivity decrement, a burnup dependent uncertainty in reactivity for burnup calculations may be assigned. This uncertainty is statistically combined with the other reactivity allowances in the determination of the maximum kff.

4.6.6 Isotopic Compositions To perform the criticality evaluation for spent fuel in MCNP, the isotopic composition of the fuel is calculated with the depletion code CASMO-4 and then specified as input data in the MCNP4A run.

The CASMO-4 calculations to obtain the isotopic compositions for MCNP4Awere performed generically, with one calculation for each enrichment, and burnups in increments of 2.5 GWD/MTU or less. The isotopic composition for any given burnup is then determined by linear interpolation.

Holtec Report HI-2043 162 4-12 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.6.7 Eccentric Fuel Assembly Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell.

Nevertheless, MCNP4a calculations were made with the fuel assemblies assumed to be in the comer of the storage rack cell (four-assembly cluster at closest approach). These calculations indicated that eccentric fuel positioning results in a decrease in reactivity.

4.6.8 Reactivity Effect of Axial Bumup Distribution Generic analytic results of axial bumup effects have been provided by Turner [4.7] based upon calculated and measured axial bumup distributions. These analyses confirm a minor and generally negative reactivity effect of the axially distributed bumup, becoming positive at bumups greater than about 30 GWD/MTU. The required bumup for all enrichments analyzed here is less than 30 GWD/MTU. Therefore, a positive reactivity effect of the axially distributed bumup would not be expected. However, for additional assurance, all design basis calculations are performed with an axial bumup distribution and with an axially constant bumup, and the higher of the two resulting reactivities is used.

4.6.9 Bumup Versus Enrichment Curve To establish a bumup versus enrichment curve (loading curve), calculations were performed at selected enrichments between 2.0 wt% and 4.1 wt%, and for bumup values slightly above and below the expected loading curve. Points on the proposed loading curve are then calculated by linear interpolation for each enrichment, based on an appropriate target value for the reactivity.

The loading curve is then determined by a curve fit to these calculated bumup values. The curve fit results in the following equation:

BU=A+B*E+C*E 2 Holtec Report HI-2043162 , 4-13 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

with BU Minimum Burnup in GWd/mtU E Initial Enrichment in wt% 235U A Coefficient, = -40.627 B Coefficient, = 24.903 C Coefficient, =-1.960 The loading curve is also shown graphically in Figure 4.1.1.

To confirm the validity of this curve for all enrichments, calculations are performed for a total of seven enrichment values between 2.0 wt% and 4.1 wt%, and the corresponding burnups determined from the loading curve. Results of these calculations demonstrate that the maximum kcff including all applicable biases and uncertainties is well below the regulatory limit of 1.0 for the entire enrichment range. The results for 4.1 wt% initial enrichment (corresponding burnup is 28.53 GWd/mtU) are listed in Table 4.1.1.

4.6.10 Interfaces The cask area racks face the pool wall on two sides, and face existing Region 2 racks on the other two sides. To evaluate whether there is the possibility of any neutronic coupling between the new cask area racks and the existing Region 2 racks, the following needs to be noted:

• The distance between the racks is more than 12 inches, and there is an existing structure in the pool which precludes the racks from being closer than 12 inches. The mean free path of neutrons in steel or water, as determined in the MCNP4A calculations, is about I inch or less. The separation of more than 12 inches, i.e. more than 12 times the mean free path, is therefore already sufficient to preclude any neutronic interaction.

Holtec Report HI-2043162 4-14 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

• In addition, the existing Region 2 racks are designed as non-poisoned racks, with a distance between individual cells of about 2 inches. These racks are analyzed as infinite arrays of cells. A fuel assembly would have to be placed closer than two inches on the outside of these racks in order to have an impact on the reactivity. However, the minimum possible distance is about 12 inches, as discussed above. Therefore, the placement of the cask area racks has no effect on the reactivity of the existing racks in the spent fuel pool.

• The new cask area racks contain poison panels on all four sides of the racks, and are conservatively analyzed as infinite cell arrays. Even the inadvertent mis-placement of an individual fuel assemblies placed directly on the outside of a rack would not affect the calculated reactivity. The existing racks are at least 12 inches away from the new cask area racks and do not affect the reactivity of the cask area racks. Therefore, there is no need to consider the existing racks in the criticality calculations presented here.

In summary, the new and existing racks are sufficiently separated to preclude any neutronic interaction, and no specific criticality calculations of any interface configuration are required.

4.6.11 Soluble Boron Concentration for Maximum kff of 0.95 The calculations crediting soluble boron in the spent fuel pool to ensure that the reactivity does not exceed 0.95 are performed for several enrichments at a soluble boron level of 500 ppm. In all cases, the maximum kff including all applicable biases and uncertainties is well below the regulatory limit of 0.95. The results for 4.1 wt% initial enrichment (corresponding burnup is 28.53 GWd/mtU) are also listed in Table 4.1.2.

Holtec Report HI-2043162 4-15 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.6.12 Abnormal and Accident Conditions The effects on reactivity of credible abnormal and accident conditions are examined in this section. None of the abnormal or accident conditions that have been identified as credible cause the reactivity of the storage racks to exceed the limiting reactivity value of kcr = 0.95, considering the presence of soluble boron. The double contingency principle of ANSI N16.1-1975 (and the USNRC letter of April 1978) specifies that it shall require at least two unlikely independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

4.6.12.1 Temperature and Water Density Effects Temperature effects on reactivity have been calculated with CASMO-4. The results show that the spent fuel pool temperature coefficient of reactivity is negative, and that introducing voids in the water internal to the storage cell (to simulate boiling) further decreased reactivity. Using the lowest temperature (38 0 F) in the analysis, therefore, assures that the true reactivity will always be lower than the calculated value regardless of temperature.

4.6.12.2 Dropped Assembly - Horizontal For the case in which a fuel assembly is assumed to be dropped on top of a 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 as discussed in section 4.6.10). Maximum expected deformation under seismic or accident conditions will not reduce the minimum spacing to less than 12 inches.

Consequently, the horizontal fuel assembly drop accident will not result in a significant increase in reactivity.

Holtec Report HI-2043162 4-16 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.6.12.3 Dropped Assembly - Vertical It is also possible to vertically drop an assembly into a location that might be occupied by another assembly or that might be empty. Such a vertical impact would at most cause a small compression of the stored assembly, if present, or result in a small deformation of the baseplate for an empty cell. These deformations could potentially increase reactivity. However, the reactivity increase would be small compared to the reactivity increase created by the misloading of a fresh assembly discussed in the following section. The vertical drop is therefore bounded by this misloading accident and no separate calculation is performed for the drop accident.

4.6.12.4 Abnormal Location of a Fuel Assembly 4.6.12.4.1 Misloaded Fresh Fuel Assembly The misplacement of a fresh unburned fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (kff of 0.95). This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt%) were to be inadvertently misloaded into a storage cell intended for spent fuel. The reactivity consequence of this situation was investigated. The corresponding calculational model consists of a 5x5 array of cells with a fresh assembly in the center cell. The model is surrounded by periodic boundary conditions, which generates an infinite arrangement of 5x5 arrays with a misloaded assembly. Calculations are performed with 500 ppm and 1000 ppm soluble boron, and the minimum allowable soluble boron concentration is determined by linear interpolation. As the target values for kfr, the maximum value from the design basis calculations is used To assure that the regulatory limit is not exceeded under this condition, a soluble boron concentration of at least 800 ppm in the spent fuel pool is required.

Holtec Report HI-2043162 4-17 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.6.12.4.2 Mislocated Fresh Fuel Assembly The mislocation of a fresh unburned fuel assembly, i.e. the accidental placement of an assembly outside of a storage rack adjacent to other fuel assemblies, has also been considered. There are two locations at opposite comers of the racks where a single assembly could be placed on the outside of the racks. The racks have been designed to prevent this event. Nevertheless, this event was conservatively evaluated. The assemblies would still be separated from the assemblies inside the rack by a neutron absorber panel, since all peripheral walls of the racks are equipped with such panels, although with a slightly smaller panel width than on the walls inside the racks. Further, the assemblies in these locations would still be more than 12 inches away from the existing Region 2 racks, which precludes any significant neutron interaction (see also Section 4.6.10, first item in bulleted list). The effect of the slightly smaller panel width (7 inches instead 7.5 inches) would be more than compensated by the increased neutron leakage on the periphery of the racks, which is not credited in the design basis calculations. In summary, the mislocation of an assembly would not result in any significant increase in reactivity, compared to placement of the same assembly inside the racks. The mislocation accident of a fresh assembly is therefore bounded by the misloading accident discussed in the previous section, and no specific calculations were performed for this mislocation accident.

4.6.13 Code Comparison A comparison between MCNP4A and KENOV.A is performed for fresh fuel at 2.0 wt%

enrichment. The comparison shows a good agreement between the two codes, with a difference in reactivity (including Bias) of only 0.00 13 Ak. Of the two codes, MCNP shows the higher value.

Holtec Report HI-2043162 4-18 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.7 References 4.1 M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

4.2 J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-12625, Los Alamos National Laboratory (1993).

4.3 M. Edenius, K. Ekberg, B.H. Forssdn, and D. Knott, "CASMO-4 A Fuel Assembly Burnup Program User's Manual," Studsvik/SOA-95/1, Studsvik of America, Inc.

and Studsvik Core Analysis AB (proprietary).

4.4 D. Knott, "CASMO-4 Benchmark Against Critical Experiments", SOA-94/13, Studsvik of America, Inc., (proprietary).

4.5 D. Knott, "CASMO-4 Benchmark Against MCNP," SOA-94/12, Studsvik of America, Inc., (proprietary).

4.6 L.I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L.

Kopp to T. Collins, August 19, 1998.

4.7 S.E. Turner, "Uncertainty Analysis - Burnup Distributions", presented at the DOE/SANDIA Technical Meeting on Fuel Burnup Credit, Special Session, ANS/ENS Conference, Washington, D.C., November 2, 1988.

Holtec Report HI-2043162 4-19 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.8 L.M. Petrie and N.F. Landers, "KENO Va - An Improved Monte Carlo Criticality Program with Supergrouping," Volume 2, Section Fl I from "SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation" NUREG/CR-0200, Rev. 4, January 1990.

4.9 "SCALE 4.3: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluations," NUREG-CR-0200, Rev. 5, Oak Ridge National Laboratory (1995).

4.10 N.M. Greene, L.M. Petrie and R.M. Westfall, "NITAWL-II: Scale System Module for Performing Shielding and Working Library Production," Volume 1, Section F1 from "SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation" NUREG/CR-0200, Rev. 4, January 1990.

4.11 "Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit",

NUREG/CR-6761, ORNL/TM-2000/373, March 2002 Holtec Report HI-2043162 4-20 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.1.1 Summary of the Criticality Safety Analyses for Holtec Cask Area Rack without Soluble Boron Design Basis Burnup at 4.1 wt% 235U 28.53 GWd/mtU Soluble Boron Concentration 0 ppm Uncertainties and Tolerances Bias Uncertainty (95%/95%) +/- 0.0011 Calculational Statistics (95%/95%, 2.Oxa) +/- 0.0014 Depletion Uncertainty +/- 0.0105 Fuel Eccentricity negative Fuel Tolerances Fuel Density +0.0021 Enrichment +0.0068 Rod Pitch +/-0.0014 Clad-OD +/-0.0008 Minimum Clad Thickness +0.0006 Fuel Rod OD +/-0.0004 Guide Tube OD +/-0.0000 Minimum G.T. Thickness +0.0005 Fuel Tolerances, statistically combined +/- 0.0073 Rack Tolerances Cell ID/Pitch +/-0.0016 Wall Thickness +/-0.0003 Sheathing Thkns +/-0.0001 Poison Thickness +/-0.0001 Poison Loading +/-0.0025 Rack Tolerances, statistically combined +/- 0.0029 Statistical Combination of Uncertainties +/- 0.0132 and Tolerances' Reference kcff (MCNP4a) 0.9721 Total Uncertainty (above) 0.0132 Calculational Bias (see Appendix A) 0.0009 Temperature Bias 0.0046 Maximum keff 0.9908 Regulatory Limiting keff 1.0000 lSquare root of the sum of the squares.

Holtec Report HI-2043162 4-21 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

i Table 4.1.2 Summary of the Criticality Safety Analyses for Holtec Cask Area Rack with Soluble Boron Design Basis Burnup at 4.1 wt% 235U 28.53 GWd/mtU Soluble Boron Concentration 500 ppm Uncertainties Bias Uncertainty (95%/95%) +/- 0.0011 Calculational Statistics (95%/o/95%, 2.Oxa) +/- 0.0014 Depletion Uncertainty +/- 0.0099 Fuel Eccentricity negative Fuel Tolerances Fuel Density +0.0021 Enrichment +0.0068 Rod Pitch +/-0.0014 Clad-OD +/-0.0008 Minimum Clad Thickness +0.0006 Fuel Rod OD +/-0.0004 Guide Tube OD +/-0.0000 Minimum G.T. Thickness +0.0005 Fuel Tolerances, statistically combined +/- 0.0073 Rack Tolerances Cell ID/Pitch +/-0.0016 Wall Thickness +/-0.0003 Sheathing Thkns +/-0.0001 Poison Thickness +/-0.0001 Poison Loading +/-0.0025 Rack Tolerances, statistically combined +/- 0.0029 Statistical Combination of Uncertainties' +/- 0.0127 Reference kff (MCNP4a) 0.9162 Total Uncertainty (above) 0.0127 Calculational Bias (see Appendix A) 0.0009 Temperature Bias 0.0046 Maximum keff 0.9344 Regulatory Limiting keff 0.9500 2 Square root of the sum of the squares.

Holtec Report HI-2043162 4-22 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.4.1 PWR Fuel Assembly Specifications Parameter Value Assembly type STD (LOPAR) OFA (Vantage 5)

Rod Array Size 17x17 17x17 Rod Pitch, Inches 0.496 L. 0.496 ';>

Active Fuel Length, Inches 144 144 Stack Density 95% [a*; of theoretical 95% 1 of (10.96 g/cm 3 ) theoretical (10.96 g/cm 3 )

Total Number of Fuel Rods 264 264 Fuel Rod Outer Diameter, Inches 0.374 1 - 0.360 ;. l Fuel Rod Inner Diameter, Inches 0.329 j, 0.315 _

Minimum Cladding Thickness, Inches Cladding Material Zr-4 Zr-4 or Zirlo Pellet Diameter, Inches 0.3225 1 0.3088 1 Number of Guide/Instrument Tubes 24/1 24/1 Guide Tube Outer Diameter, Inches 0.482 ! _ 0.474[

Guide Tube Inner Diameter, Inches 0.450 i 0.442 -

Guide Tube Minimum Wall Thickness, Inches Guide Tube Material Zr-4 Zr-4 or Zirlo Holtec Report HI-2043162 4-23 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.4.2 Core Operating Parameter for Depletion Analyses Parameter Value Soluble Boron Concentration (cycle average), ppm 820 Average Linear Power, W/ft 5450 Average Fuel Temperature, 0F 1310 Average Core Exit Temperature, 0F 614.8 In-Core Assembly Pitch, Inches 8.466 Table 4.4.3 BPRA Design Specifications (based on [4.11] except for burnup)

Parameter Value Boron Loading 0.0062 g B-0/cm Outer Clad OD (cm) 0.96774 Outer Clad ID (cm) 0.87376 Poison OD (cm) 0.85344 Poison ID (cm) 0.48260 Inner Clad OD (cm) 0.46101 Inner Clad ID (cm) 0.42799 Assembly Burnup when Absorber is removed, GWD/MTU 22 Holtec Report HI-2043162 4-24 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.4.4 Fuel Rack Specifications Parameter Value Cell ID, Inches 8.75L[7*

(8.85 on some peripheral cells, see text)

Box Wall Thickness, Inches 0.09 Lv.

Sheathing Thickness, Inches Poison Pocket Thickness, Inches L Poison Thickness, Inches Poison Width, Inches 7.5 min.

(7.0 min on rack exterior, see text)

Cell Pitch, Inches 8.946['.

Poison Loading, gm B-10/cm 2l Holtec Report HI-2043162 . 4-25 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

30.00 25.00 20.00 S.

M CD 15.00 10.00 5.00 0.00 2 2.5 3 3.5 4 Initial Enrichment (wt% U-235)

Figure 4.1.1: Cask Pit Rack Loading Curve Holtec Report H1-2043 162 4-26 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Appendix 4A Benchmark Calculations (total number of pages: 26 including this page)

Holtec Report HI-2043162 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

APPENDIX 4A: BENCHMARK CALCULATIONS 4A.1 INTRODUCTION AND

SUMMARY

Benchmark calculations have been made on selected critical experiments, chosen, in so far as possible, to bound the range of variables in the rack designs. Two independent methods of analysis were used, differing in cross section libraries and in the treatment of the cross sections. MCNP4a [4A.1] is a continuous energy Monte Carlo code and KEN05a [4A.2]

uses group-dependent cross sections. For the KEN05a analyses reported here, the 238-group library was chosen, processed through the NITAWL-fl [4A.2] program to create a working library and to account for resonance self-shielding in uranium-238 (Nordheim integral treatment). The 238 group library was chosen to avoid or minimize the errors' (trends) that have been reported (e.g., [4A.3 through 4A.5]) for calculations with collapsed cross section sets.

In rack designs, the three most significant parameters affecting criticality are (1) the fuel enrichment, (2) the `0B loading in the neutron absorber, and (3) the lattice spacing (or water-gap thickness if a flux-trap design is used). Other parameters, within the normal range of rack and fuel designs, have a smaller effect, but are also included in the analyses.

Table 4A. 1 summarizes results of the benchmark calculations for all cases selected and analyzed, as referenced in the table. The effect of the major variables are discussed in subsequent sections below. It is important to note that there is obviously considerable overlap in parameters since it is not possible to vary a single parameter and maintain criticality; some other parameter or parameters must be concurrently varied to maintain criticality.

One possible way of representing the data is through a spectrum index that incorporates all of the variations in parameters. KENO5a computes and prints the "energy of the average lethargy causing fission' (EALF). In MCNP4a, by utilizing the tally option with the identical 238-group energy structure as in KENO5a, the number of fissions in each group may be collected and the EALF determined (post-processing).

t Small but observable trends (errors) have been reported for calculations with the 27-group and 44-group collapsed libraries. These errors are probably due to the use of a single collapsing spectrum when the spectrum should be different for the various cases analyzed, as evidenced by the spectrum indices.

Appendix 4A, Page 1

Figures 4A. 1 and 4A.2 show the calculated k~f for the benchmark critical experiments as a function of the EALF for MCNP4a and KEN05a, respectively (U0 2 fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental errorl in performing the critical experiments within each laboratory, as well as between the various testing laboratories. The B&W critical experiments show a larger experimental error than the PNL criticals. This would be expected since the B&W criticals encompass a greater range of critical parameters than the PNL criticals.

Linear regression analysis of the data in Figures 4A. 1 and 4A.2 show that there are no trends, as evidenced by very low values of the correlation coefficient (0.13 for MCNP4a and 0.21 for KEN05a). The total bias (systematic error, or mean of the deviation from a kfn of exactly 1.000) for the two methods of analysis are shown in the table below.

Calculational Bias of MCNP4a and KEN05a MCNP4a 0 .0009i0.0011 KEN05a l 0.0030+/-t0.0012 The bias and standard error of the bias were derived directly from the calculated kff values in Table 4A.1 using the following equadonstt, with the standard error multiplied by the one-sided K-factor for 95% probability at the 95% confidence level from NBS Handbook 91 [4A. 18] (for the number of cases analyzed, the K-factor is -2.05 or slightly more than 2).

= L k, (4A.1) n t A classical example of experimental error is the corrected enrichment in the PNL experiments, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods.

tt These equations may be found in any standard text on statistics, for example, reference

[4A.6] (or the MCNP4a manual) and is the same methodology used in MCNP4a and in KENOSa.

Appendix 4A, Page 2

-1.=

.L1 2 -(I 1

k1 9 ', (4A.2) k n (n-i)

Bias = (1- k) d K Ai (4A.3) where k, are the calculated reactivities of n critical experiments; a, is the unbiased estimator of the standard deviation of the mean (also called the standard error of the bias (mean)); K is the one-sided multiplier for 95% probability at the 95% confidence level (NBS Handbook 91 [4A. 18]).

Formula 4.A.3 is based on the methodology of the National Bureau of Standards (now NIST) and is used to calculate the values presented on page 4.A-2. The first portion of the equation, ( 1- k ), is the actual bias which is added to the MCNP4a and KENO5a results.

The second term, Kai, is the uncertainty or standard error associated with the bias. The K values used were obtained from the National Bureau of Standards Handbook 91 and are for one-sided statistical tolerance limits for 95 % probability at the 95 % confidence level. The actual K values for the 56 critical experiments evaluated with MCNP4a and the 53 critical experiments evaluated with KENO5a are 2.04 and 2.05, respectively.

The bias values are used to evaluate bhe maximum ken values for the rack designs.

KENO5a has a slightly larger systematic error than MCNP4a, but both result in greater precision than published data [4A.3 through 4A.5] would indicate for collapsed cross section sets in KENOSa (SCALE) calculations.

4A.2 Effect of Enrichment The benchmark critical experiments include those with enrichments ranging from 2.46 w/o to 5.74 w/o and therefore span the enrichment range for rack designs. Figures 4A.3 and 4A.4 show the calculated K.e values (Table 4A. 1) as a function of the fuel enrichment reported for the critical experiments. Linear regression analyses for these data confirms that there are no trends, as indicated by low values of the correlation coefficients (0.03 for MCNP4a and 0.38 for KENOSa). Thus, there are no corrections to the bias for the various enrichments.

• .Appendix 4A, Page 3

As further confirmation of the absence of any trends with enrichment, a typical configuration was calculated with both MCNP4a and KENOSa for various enrichments.

The cross-comparison of calculations with codes of comparable sophistication is suggested in Reg. Guide 3.41. Results of this comparison, shown in Table 4A.2 and Figure 4A.5, confirm no significant difference in the calculated values of K. for the two independent codes as evidenced by the 450 slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence of an enrichment effect (trend) in the bias.

4A.3 Effect of `13 Loading Several laboratories have performed critical experiments with a variety of thin absorber panels similar to the Boral panels in the rack designs. Of these critical experiments, those performed by B&W are the most representative of the rack designs. PNL has also made some measurements with absorber plates, but, with one exception (a flux-trap experiment),

the reactivity worth of the absorbers in the PNL tests is very loW and any significant errors that might exist in the treatment of strong thin absorbers could not be revealed.

Table 4A.3 lists the subset of experiments using thin neutron absorbers (from Table 4A. 1) and shows the reactivity worth (Ak) of the absorber.t No trends with reactivity worth of the absorber are evident, although based on the calculations shown in Table 4A.3, some of the B&W critical experiments seem to have unusually large experimental errors. B&W made an effort to report some of their experimental errors. Other laboratories did not evaluate their experimental errors.

To further confirm the absence of a significant trend with °B concentration in the absorber, a cross-comparison was made with MCNP4a and KEN05a (as suggested in Reg.

Guide 3.41). Results are shown in Figure 4A.6 and Table 4A.4 for a typical geometry.

These data substantiate the absence of any error (trend) in either of the two codes for the conditions analyzed (data points fall on a 45° line, within an expected 95% probability limit).

t The reactivity worth of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental (Ak) change in reactivity due to the absorber.

Appendix 4A, Page 4

4A.4 Miscellaneous and Minor Parameters 4A.4.1 Reflector Material and Spacings PNL has performed a number of critical experiments with thick steel and lead reflectors.t Analysis of these critical experiments are listed in Table 4A.5 (subset of data in Table 4A.1). There appears to be a small tendency toward overprediction of kf at the lower spacing, although there are an insufficient number of data points in each series to allow a quantitative determination of any trends. The tendency toward overprediction at close spacing means that the rack calculations may be slightly more conservative than otherwise.

4A.4.2 Fuel Pellet Diameter and Lattice Pitch The critical experiments selected for analysis cover a range of fuel pellet diameters from 0.311 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches. In the rack designs, the fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.496 to 0.580 inch lattice spacing) for PWR fuel and from 0.3224 to 0.494 inches O.D. (0.488 to 0.740 inch lattice spacing) for BWR fuel. Thus, the critical experiments analyzed provide a reasonable representation of power reactor fuel. Based on the data in Table 4A.1, there does not appear to be any observable trend with either fuel pellet diameter or lattice pitch, at least over the range of the critical experiments applicable to rack designs.

4A.4.3 Soluble Boron Concentration Effects Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KEN05a) calculations are shown in Table 4A.6. Analyses of the very high boron concentration experiments (> 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.

t 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 Buel 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 KENO5a 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 KENO~a calculated reactivities are below 1.00, suggesting that a small trend may exist with KEN05a. 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 Kln over a wide range of the spectral index (energy of the average lethargy causing fission).

Appendix 4A, Page 6

4A.6 Reference

[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-2/U2/R5, 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-1810, Babcock and Wilcox Company, April 1984.

Appendix 4A, Page 7

[4A.1O] 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. I1] S.R. Bierman and E.D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o 235 U 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 23U Enriched U0 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 5U 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 199O.

[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-f, E~AU t (eV)

Reference Identification Enrich. MCNP4a KENO5a MCNP4a KCENO5a 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-14.4 (4A.7) Core 11 2.46 1.0008 +/- 0.0011 1.0015 +/- 0.0005 0.2553 0.2446 3 B&W-1484 (4A.7) Core III 2.46 1.0010 +/- 0.0012 1.0005 +/- 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 NI 2.46 0.9978 +/- 0.0012 1.0005 +/- 0.0005 0.2031 0.1947 7 B&W-1484 (4A.7) Core XII 2.46 0.9988 +/- 0.0011 0.9978 +/- 0.0006 0.1718 0.1662 8 B&W-1484 (4A.7) Core Xm 2.46 1.0020 +/- 0.0010 0.9952 +/- 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 f 0.0006 0.2092 0.2014 11 B&W-1484 (4A.7) Core XVI tt 2.46 0.9935 +/- 0.0010 0.9889 +/- 0.0006 0.1757 0.1713 12 B&W-1484 (4A.7) Core XVII 2.46 0.9962 +/- 0.0012 0.9942 +/- 0.0005 0.2083 0.2021 13 B&W-1484 (4A.7) Core XVm 2.46 1.0036 +/- 0.0012 0.9931 +/- 0.0006 0.1705 0.1708 Appendix 4A, Page 9

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

Reterence Identification Enrich. MCNP4a KENO5a MCNP4a KENOSa 14 B&W-1484 (4A.7) Core XIX 2.46 0.9961 i 0.0012 0.9971 +/- 0.0005 0.2103 0.2011 IS B&W-1484 (4A.7) Core XX 2.46 1.0008 +/- 0.0011 0.9932 +/- 0.0006 0.1724 0.1701 16 B&W-1484 (4A.7) Core XXI 2.46 0.9994 i 0.0010 0.9918

• 0.0006 0.1544 0.1536 17 B&W-1645 (4A.8) S-type Fuel, w/886 ppm B 2.46 0.9970 +/- 0.0010 0.9924

• 0.0006 1.4475 1.4680 18 B&W-164S M4A.8) S-type Fuel, w/746 ppm B 2.46 0.9990 +/- 0.0010 0.9913 +/- 0.0006 1.5463 1.5660 19 B&W-1645 (4A.8) SO-type Fuel, w/IlS6 ppm B 2.46 0.9972 +/- 0.0009 0.9949 0.0005 0.4241 0.4331 20 B&W-1810 (4A.9) Case 1 133 7 ppm B 2.46 1.0023 +/- 0.0010 NC 0.1531 NC 21 B&W-1810 (4A.9) Case 12 1899 ppm B 2.4614.02 1.0060 +/- 0.0009 NC 0.4493 NC 22 French (4A.10) Water Moderator 0 gap 4.75 0.9966 +/- 0.0013 NC 0.2172 NC 23 French (4A.10) Water Moderator 2.5 cm gap 4.75 0.9952 +/- 0.0012 NC 0.1778 NC 24 French (4A.10) Water Moderator 5 cm gap 4.75 0.9943 +/- 0.0010 NC 0.1677 NC 25 French (4A.10) Water Moderator 10 cm gap 4.75 0.9979 +/- 0.0010 NC 0.1736 NC 26 PNL-3602 (4A.11) Steel Reflector, 0 separatIon 2.35 NC 1.0004 +/- 0.0006 NC 0.1018

• -Appendix 4A, Page 10

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

Reference Identification Enrich. MCNP4a KEN05a MCNP4a IKENO5a 27 PNL-3602 (4A.11) Steel Reflector, 1.321 cm sepn. . 2.35 0.9980 +/- 0.0009 0.9992 +/- 0.0006 0.1000 0.0909 28 PNL-3602 (4A.11) Steel Reflector, 2.616 cm sepn 2.35 0.9968 +/- 0.0009 0.9964 +/- 0.0006 0.0981 0.0975 29 PNL-3602 (4A.11) Steel Reflector, 3.912 cm sepn. 2.35 0.9974 +/- 0.0010 0.9980 +/- 0.0006 0.0976 0.0970 30 PNI,3602 (4A.11) Steel Reflector, Infinite sepn. 2.35 0.9962 +/- 0.0008 0.9939 +/- 0.0006 0.0973 0.0968 31 PNL3602 (4A.11) Steel Reflector, 0 cm sepn. 4.306 NC 1.0003 +/- 0.0007 NC 0.3282 32 PNIL3602 (4A.11) Steel Reflector, 1.321 cm sepn. 4.306 0.9997 +/- 0.0010 1.0012 +/- 0.0007 0.3016 0.3039 33 PNL-3602 (4A.1l) Steel Reflector, 2.616 cm sepn. 4.306 0.9994 +/- 0.0012 0.9974 +/- 0.0007 0.2911 0.2927 34 PNL-3602 (4A.11) Steel Reflector, 5.405 cm sepn. 4.306 0.9969 +/- 0.0011 0.9951 +/- 0.0007 0.2828 0.2860 35 PNL-3602 (4A.11) Steel Reflector, Infinite sepo. 1' 4.306 0.9910 +/- 0.0020 0.9947 +/- 0.0007 0.2851 0.2864 36 PNI,3602 (4A.11) Steel Reflector, with Boral Sheets 4.306 0.9941 +/- 0.0011 0.9970 +/- 0.0007 0.3135 0.3150 37 PNL-3926 (4A.12) Lead Reflector, 0 cm sepn. 4.306 NC 1.0003 +/- 0.0007 NC 0.3159 38 PNI-3926 (4A.12) Lead Reflector, 0.55 cm sepn. 4.306 1.0025 +/- 0.0011 0.9997 +/- 0.0007 0.3030 0.3044 39 PNL-3926 (4A.12) Lead Reflector, 1.956 cm sepn. 4.306 1.0000 +/- 0.0012 o.ms +/- 0.0007 0.2883 -02930 Appendix 4A, Page 11

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k, EALF t W)V Identification Enrich. MCNP4a KENOSa MCNP4a KENOSa Reference 40 PNL-3926 (4A.12) Lead Reflector, 5.405 cm sepn. 4.306 0.9971 +/- 0.0012 0.9946 i 0.0007 0.2831 0.2854 Experiment 004/032 - no absorber 4.306 0.9925 +/- 0.0012 0.9950 +/- 0.0007 0.1155 0.1159 41 PNL-2615 (4A.13)

Experiment 030 Zr plates 4.306 NC 0.9971 0.0007 NC 0.1154 42 PNL-2615 (4A.13) -

Experiment 013 - Steel plates 4.306 NC 0.9965 +/- 0.0007 NC 0.1164 43 PNL-2615 (4A.13)

Experiment 014 - Steel plates 4.306 NC 0.9972 +/- 0.0007 NC 0.1164 44 PNI-2615 (4A.13)

Exp. 009 1.05% Boron-Steel plates 4.306 0.9982 +/- 0.0010 0.9981 +/- 0.0007 0.1172 0.1162 45 PNL-2615 (4A.13)

Exp. 012 1.62% Boron-Steel plates 4.306 0.9996 +/- 0.0012 0.9982 +/- 0.0007 0.1161 0.1173 46 PNL-2615 (4A.13)

Exp. 031 - Boral plates 4.306 0.9994 +/- 0.0012 0.9969 i 0.0007 0.1165 0.1171 47 PNI,2615 (4A.13)

Experiment 214R - with flux trap 4.306 0.9991 +/- 0.0011 0.9956 0.0007 0.3722 0.3812 48 PNL-7167 (4A.14)

Experiment 214V3- with flux trap 4.306 0.9969 +/- 0.0011 0.9963 i 0.0007 0.3742 0.3826 49 PNL-7167 (4A.14)

Case 173 - 0 ppm B 4.306 0.9974 +/- 0.0012 - NC 0.2893 NC 50 PNL-4267 (4A.15)

Case 177 - 2550 ppm B 4.306 1.007 0.0010 NC 0.5509 NC 51 PNL-4267 (4A.15)

MOX Fuel - Type 3.2 Exp. 21 20% Pu 1.0041 +/- 0.0011 1.0046 +/- 0.0006 0.9171 0.8868 52 PNL-5803 (4A.16)

Appendix 4A, Page 12

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

Enrich. MCNP4a KINO5a MCNP4a ltENOSa Rererence Identification 20% Pu 1.0058 +/- 0.0012 1.0036 +/- 0.0006 0.2968 0.2944 53 PNI-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 43 20% Pu 1.0083 +/- 0.0011 0.9989 +/- 0.0006 0.1665 0.1706 54 PNl-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 13 20% Pu 1.0079 +/- 0.0011 0.9966 +/- 0.0006 0.1139 0.1165 55 PNL-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-338S (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 PuO2 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.

the possibility of unusually large experimental tt These experimental results appear to be statistical outliers (i3a) suggesting retained in determining the calculational error. Although they could justifiably be excluded, for conservatism, they were basis.

Appendix 4A, Page 13

Table 4A.2 COMPARISON OF MCNP4a AN)D KENO5a CALCULATED REACTIVITIESt FOR VARIOUS ENRICHMENTS Calculated kff +/- la Enrichment MCNP4a KENO5a 3.0 0.8465 +/- 0.0011 0.8478 i 0.0004 3.5 0.8820 +/- 0.0011 0.8841 +/- 0.0004 3.75 0.9019 +/- 0.0011 0.8987 i 0.0004 4.0 0.9132 +/- 0.0010 0.9140 i 0.0004 4.2 0.9276 +/- 0.0011 0.9237 i 0.0004 4.5 0.9400 +/-i0.001 1 0.9388 +/- 0.0004 t Based on the GE Wx8R fuel assembly.

Appendix 4A, Page 14

Table 4A.3 MCNP4a CALCULATED REACTIVITIES FOR CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS Ak MCNP4a Worth of Calculated EALFt Ref. Experiment Absorber Ir (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 XrV 0.1575 0.9953+/-0.0011 0.2022 4A.7 B&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 15

Table 4A.4 COMPARISON OF MCNP4a AND KENO5a CALCULATED REACTIVrIDESt FOR VARIOUS '0B LOADINGS Calculated kff + la

'0 B, glcm 2 MCNP4a KENOSa 0.005 1.0381 +/- 0.0012 1.0340 +/- 0.0004 0.010 0.9960 +/- 0.0010 0.9941 +/- 0.0004 0.015 0.9727 i 0.0009 0.9713 +/- 0.0004 0.020 0.9541 +/- 0.0012 0.9560 +/- 0.0004 0.025 0.9433 +/- 0.0011 0.9428 +/- 0.0004 0.03 0.9325 i 0.0011 0.9338 +/- 0.0004 0.035 0.9234 i 0.0011 0.9251 +/- 0.0004 0.04 0.9173 i 0.0011 0.9179 +/- 0.0004 Based on a 4.5% enriched GE WxSR fuel assembly.

Appendix 4A, Page 16

Table 4A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH THICK LEAD AND STEEL REFLECTORSt Separation, Ref. Case E, wt% cm MCNP4a kff KEN05a ky, 4A.11 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. I1 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 t 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 kff 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 CRMTICAL EXPERIMENTS WlTH MOX FUEL MCNP4a 05NOSa Reference Case' kff EALj t kdt EALF' PNL-5803 MOX Fuel - Exp. No. 21 1.0041+/-0.0011 0.9171 1.0046+/-0.0006 0.8868

[4A.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 CD 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+/-i.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 a,

-I0-CD

'4-

'4-1.000

.4-la 0)

C-4 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 a,

X 1.000

.4--

U,

-Y

'a la 0.995 0

C-,

0.990 0.985 0.1 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 U

~0 aI

.4-v 1.000

-a a 0.99 U

0.9 a

0.995 0.990 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

• 0

.4-CD 1.000 ___ _ -1

.-Y c

Ua I- ______ - -- - - - -

Bias T

.4-

.2 0.995 0ca I 0 I

0.990 0.985 111111111 111111111 111111111 111111111 ilillilli illigijil gggjgggjj Z.U Z.t) 15.u 3.5 4.0 4.5 5.0 5.5 6.0 Enrichment, w/o U-235 FIGURE 4A.4. KENO CALCULATED k-eff VALUES AT VARIOUS U-235 ENRICHMENTS

0.94 E 0.92 0

0 0.90 0

0f 0

4-

'4 a,

C 0.88 0)

I LU1 0.86 0.84 MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KEN05A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS

i-ns /c vr

1. 02 za. 1.01 3: _ it.010 g/ams

0. 99

.4-la 0.913_

4-,

0.96-a U,

0.96_-_ 0.07.0 g/o sq

_ X 0.0I /Gmxq 0.945 0.94-0.035 g/omsq B.c2- _

0.04 g/c 4rq A.91 0.0I 0.900 I I I 0.920 l I I 0.940

-I I iA 0.9 60 Il l I 0.980 i II I I 1.000 I II4 I I ' I 1.020 1.040 Reactivity Calculated with KEN05a

.. I.0 FIGURE 4A.6. COMPARISON OF MCNP AND KEN05a CALCULATIONS FOR VARIOUS BORON-10 AREAL DENSITIES

5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5.1 Introduction This section provides a summary of the analyses performed to demonstrate the compliance of the Diablo Canyon Units 1 and 2 spent fuel pools (SFPs) and their attendant cooling systems with the appropriate provisions of USNRC Standard Review Plan (SRP) 9.1.3 (Spent Fuel Pool Cooling and Cleanup System, Rev. 1, July 1981),Section III of the USNRC "OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications," (April 14, 1978),

and the existing Diablo Canyon Power Plant (DCPP) licensing bases. Similar methods of thermal-hydraulic analysis have been used in the licensing evaluations for other SFP capacity expansion projects.

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

i. Evaluation of the maximum SFP bulk temperatures for the design-basis offload scenarios, to establish that maximum bulk temperature limits are not exceeded.

ii. Evaluation of loss-of-forced cooling scenarios, to establish minimum times to boil, maximum boil-off rates and the minimum time for the SFP level to reach 10 feet above the stored fuel. This establishes the minimum times to perform corrective actions, and the associated makeup water requirements.

iii. Determination of the maximum local water temperature, at the instant when the bulk temperature reaches its maximum value, to establish that localized boiling in the fuel storage racks is not possible while forced cooling is operating.

iv. Evaluation of the maximum fuel rod cladding temperature, at the instant when the bulk temperature reaches its maximum value, to establish that nucleate boiling is not possible while forced cooling is operating.

The following sections present plant system descriptions, analysis methodologies and assumptions, a synopsis of the input data employed and summaries of the calculated results.

Holtec Rcport HI-2043162 5-1 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

5.2 Cooling Systems Description The existing spent fuel pool cooling system (SFPCS) licensing basis for both units requires that the system must maintain the SFP bulk temperature below the following limits:

• Following a partial core offload of 76 fuel assemblies the SFP bulk temperature must not exceed 140.0 0 F and following a partial core offload of 96 fuel assemblies the SFP bulk temperature must not exceed 147.3°F.

• Following a full core offload (either planned or unplanned), the SFP bulk temperature must not exceed 174.3 0 F.

The SFPCS at each unit consists of two parallel cooling pumps discharging to a single shell and tube heat exchanger. SFP water is circulated through the heat exchanger tubes and heat is transferred to component cooling water circulating through the shell side. Only one pump is operated at a time, providing for a single active failure without any reduction in system capacity.

5.3 Offload Scenarios Three offload scenarios are postulated in this evaluation. These scenarios are as follows:

Scenario 1 A partial core offload is comprised of 96 assemblies offloaded into the SFP, filling all but 142 storage locations when the cask pit rack is in place. All 96 offloaded assemblies are assumed to have 52000 MWD/MTU of burnup. The minimum decay time of the previously offloaded fuel assemblies for this offload scenario is 18 months.

Holtec Report HI-2043 162 5-2 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Scenario 2 A full core offload is comprised of 193 assemblies offloaded into the SFP, filling all but 45 storage locations when the cask pit rack is in place. The 193 offloaded assemblies are separated into two distinct groups: 101 assemblies with 52000 MWDIMTU of burnup and 92 assemblies with 25000 MWD/MTU of burnup. The minimum decay time of the previously offloaded fuel assemblies for this scenario is 18 months.

Scenario 3 A planned refueling is completed in 30 days, leaving 104 recently discharged assemblies in the SFP at restart. After 36 days of operation at 100% power an emergency full core offload is performed, resulting in a total SFP inventory of 1537 assemblies, which conservatively exceeds the SFP capacity. The 193 offloaded assemblies are separated into two distinct groups: 113 assemblies with 40000 MWD/MTU of burnup and 80 assemblies with 3000 MWD/MTU of burnup.

Each of these offload/cooling scenarios is evaluated to determine the peak SFP bulk temperature.

5.4 Maximum Pool Bulk Temperatures In this section, we present the methodology for calculating the maximum SFP bulk temperatures for the scenarios presented in the preceding section.

The transient thermal response of the SFP and the attendant cooling systems is governed by a first-order, ordinary differential equation. The governing differential equation can be written by utilizing conservation of energy as:

dT C d=Q(r)- Q1xfI () - QEVV (T (5-1) dr where:

C = SFP thermal capacity, Btu/0 F T = SFP bulk temperature, 'F T= Time after reactor shutdown, hr Holtec Report HI-2043162 5-3 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Q(r) = Time varying decay heat generation rate, Btu/hr Qiix(T) = Temperature dependent SFPCS heat rejection rate, Btu/hr QENv (T) = Temperature dependent passive heat loss to the environment, Btu/hr Q,1x(T) in Equation 5-1 is a function of the SFP bulk temperature and the coolant water flow rate and temperature, and can be written in terms of the temperature effectiveness (p) as follows:

QL C(T) = IV, C, p (T - I,) (5-2) where:

NMt = Coolant water flow rate, lb/hr Ct = Coolant water specific heat capacity, Btu/(lb-0 F) p = SFPCS heat exchanger(s) temperature effectiveness T = SFP bulk water temperature, IF ti = Coolant water inlet temperature, 0 F The temperature effectiveness, a measure of the heat transfer efficiency of the heat exchanger(s),

is defined as:

(5-3) where to is the coolant outlet temperature (0 F) and all other terms are as defined above. The SFPCS heat exchanger(s) coolant outlet temperature (to) for various SFP bulk temperatures (T) are determined using the Holtec QA validated computer program STER [5.4.7].

Holtec Report HI-2043162 5-4 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

The following conservatisms are applied in the maximum SFP bulk temperature calculations:

• The reactor thermal power level is increased by 5% to account for burnup uncertainty.

• Appropriate parameters (i.e., burnup, batch size, assembly uranium weight and initial enrichment) are used for all projected offloads.

• The thermal performance of the SFPCS heat exchanger(s) is determined with all heat transfer surfaces fouled to their design-basis maximum levels.

• The thermal performance of the SFPCS heat exchanger(s) is determined assuming 5%

tube plugging.

• The thermal inertia (thermal capacity) of the SFP is based on the net water volume only.

This conservatively neglects the considerable thermal inertia of the fuel assemblies and the SFP structure.

• Every fuel assembly in the SFP is assumed to be the heavier LOPAR design with a control rod inserted. This yields a conservative value for the calculated net water volume within the SFP.

The differential equation that defines the transient thermal response of the SFP (Equation 5-1) is solved numerically. The decay heat load from previously offloaded fuel assemblies is assumed to be constant and is calculated using Holtec's QA validated LONGOR computer program [5.4.3].

This program incorporates the ORIGEN2 computer code [5.4.4] to perform the decay heat calculations. The transient decay heat loads and SFP bulk temperatures are calculated using Holtec's QA validated BULKTEM computer program [5.4.5], which also incorporates the ORIGEN2 computer code. The maximum SFP bulk temperatures are extracted from the results Holtcc Report HI-2043162 5-5 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

of the transient evaluations. Table 5.4.1 presents the historic and projected offload schedule used for these analyses. The other major input values for these analyses are summarized in Table 5.4.2.

5.5 Minimum Time-to-Boil and Maximum Boiloff Rate In this section, we present the methodology for calculating the minimum time-to-boil and corresponding maximum boiloff rate for the scenarios presented in Section 5.3.

The governing enthalpy balance equation for this condition, subject to these conservative assumptions, can be written as:

C(r) d = Q(r + rO) - QENV (T) (5-5) where:

C(r) = Time-varying SFP thermal capacity T= Time after cooling is lost (hr)

To = Loss of cooling time after shutdown (hr)

All other terms in this equation are the same as defined for Equation 5-1 in Section 5.4.

The following conservatisms and assumptions are applied in the time-to-boil and boiloff rate calculations:

• During the loss of forced cooling evaluations, no credit is taken for any makeup water.

• The loss of forced cooling is assumed to occur coincident with the peak SFP bulk temperature. Maximizing the initial temperature will conservatively minimize the calculated time-to-boil.

• The decay heat load in the SFP is assumed constant that the value that corresponds to the peak SFP bulk temperature, neglecting the reduction in decay heat with time.

Holtec Report HI-2043162 5-6 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

• The passive heat losses from the SFP surface to the building air are evaluated assuming the relative humidity of the building air is 100%. This minimizes the evaporation driving force and reduces the resulting heat losses.

• The thermal inertia (thermal capacity) of the SFP is based on the net water volume only.

This conservatively neglects the considerable thermal inertia of the fuel assemblies and the SFP structure.

Equation 5-5 is solved using a numerical solution technique to obtain the bulk pool temperature as a function of time. The time-to-boil, boil-off rate and water depth versus time are calculated using Holtec's QA validated TBOIL program [5.4.6]. Calculations also determined the time required for the water level to reduce to within 10' of the top of the spent fuel storage racks. The major input values for these analyses are summarized in Table 5.5.1.

5.6 Maximum SFP Local Water Temperature In this section, a summary of the methodology for evaluating the maximum local water temperatures within the fuel racks in the SFP is presented. The results of these evaluations are maximum local water temperatures.

To demonstrate adequate cooling of hot fuel in the SFP, it is necessary to rigorously quantify the coupled velocity and temperature fields created by the interaction of buoyancy driven and forced water flows. A Computational Fluid Dynamics (CFD) analysis for this demonstration is required.

The objective of this study is to demonstrate that the thermal-hydraulic criterion of ensuring local subcooled conditions in the SFP is met for all postulated fuel offload/cooling alignment scenarios. The local thermal-hydraulic analysis is performed such that partial cell blockage and slight fuel assembly variations are bounded. An outline of the CFD approach is described in the following.

There are several significant geometric and thermal-hydraulic features of the Diablo Canyon SFPs that need to be considered for a rigorous CFD analysis. From a fluid flow modeling standpoint, there are two regions to be considered. One region is the SFP bulk region where the Holtec Report HI-2043162 5-7 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

classical Navier-Stokes equations [5.6.1] are solved, with turbulence effects included. The other region is the fuel storage racks containing heat generating fuel assemblies, located near the bottom of the SFP. In this region, water flow is directed vertically upwards due to buoyancy forces through relatively small flow channels formed by rods of the fuel assemblies in each rack cell. This situation is modeled as a porous solid region with pressure drop in the flowing fluid governed by Darcy's Law as:

a-=-2-v.-Cp jvi (5-6) aX, K(i) 2 where aP/8X; is the pressure gradient, K(i), V3 and C are the corresponding permeability, velocity and inertial resistance parameters and g is the fluid viscosity. These terms are added to the classic Navier-Stokes equations. The permeability and inertial resistance parameters for the rack cells loaded with fuel assemblies are determined based on friction factor correlations for the laminar flow conditions that would exist due to the low buoyancy induced velocities and the small size of the flow channels.

The Diablo Canyon SFP geometries require an adequate portrayal of both large scale and small scale features, spatially distributed heat sources in the racks and water inlet/outlet piping.

Relatively cooler bulk water normally flows down between the fuel racks outline and wall liner, a clearance known as the downcomer. Near the bottom of the racks the flow turns from a vertical to horizontal direction into the bottom plenum, supplying cooling water to the rack cells. Heated water issuing out of the top of the racks mixes with the bulk water. An adequate modeling of these features on the CFD program involves meshing the large scale bulk SFP region and small scale downcomer and bottom plenum regions with a sufficient number of computational cells to capture both the global and local features of the flow field.

The distributed heat sources in the racks are modeled by identifying distinct heat generation zones considering recently offloaded fuel, bounding peaking effects, and the presence of background decay heat from previous offloads. Three heat generating zones are identified. The first consists of background fuel from previous offloads. The second and third zones consist of Holtec Report HI-2043162 5-8 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

fuel from recently offloaded fuel assemblies. The t vo recent offload zones are differentiated by one zone with higher than average decay heat generation (hottest assemblies in full core) and the other with less than average decay heat generation (remainder of full core). This is a conservative model, since all of the fuel with higher than average decay heat is placed in a contiguous area. A uniformly distributed heat generation rate was applied throughout each distinct zone (i.e., there were no variations in heat generation rate within a single zone).

In order to determine an upper bound on the maximum local water temperature, a series of conservative assumptions are made. The most important of these assumptions are:

• The walls and floor of the SFP are all modeled as adiabatic surfaces, thereby neglecting conduction heat loss through these items.

• Heat losses by thermal radiation and natural convection from the hot SFP surface to the environment are neglected.

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

- The hydraulic resistance of every fuel storage rack cell is determined based on the most hydraulically limiting (i.e., highest hydraulic resistance) fuel assembly type.

• The hydraulic resistance parameters for the rack cells, permeability and inertial resistance, are conservatively adjusted by 5%.

• The bottom plenum heights used in the model are less than the actual heights to conservatively reduce the area for cooling water flow to the inlets of the rack cells.

• The hydraulic resistance of every fuel storage rack cell is determined based on the most restrictive water inlet geometry of the cells over rack support pedestals (i.e., all baseplate holes are completely blocked). These cells have a reduced water entrance area, caused by the pedestal blocking the baseplate hole, and a correspondingly increased hydraulic resistance.

• The hydraulic resistance of every fuel storage rack cell includes the effects of blockage due to an assumed dropped fuel assembly lying horizontally on top of the racks.

Holtec Report HI-2043162 5-9 v 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

• The fuel assemblies with the highest decay heat generation rates are grouped together in the center of the model. This conservatively maximizes the distance between these highest heat fuel assemblies and the rack-to-wall downcomers, so the cooled water from the SFPCS must travel farther along the SFP floor to cool them. Discharge of these assemblies into any rack locations that are closer to a downcomer is bounded by the analyzed configuration.

The CFD analysis was performed on the commercially available FLUENT [5.6.2] computational fluid dynamics program, which has been benchmarked under Holtec's QA program. The FLUENT code enables buoyancy flow and turbulence effects to be included in the CFD analysis.

Buoyancy forces are included by specifying a temperature-dependent density for water and applying an appropriate gravity vector. Turbulence effects are modeled by relating time-varying Reynolds' Stresses to the mean bulk flow quantities with the standard k-C turbulence model.

Some of the major input values for this analysis are summarized in Table 5.6.1. An isometric view of the assembled CFD model for the DCPP units is presented in Figure 5.6. 1.

5.7 Fuel Rod Cladding Temperature In this section, the method to calculate the temperature of the fuel rod cladding is presented. The maximum local water temperature and peak heat flux are considered to occur coincidentally. The superposition of these two maximum values ensures that the calculated peak fuel cladding temperature bounds the fuel cladding temperature anywhere along the length of the fuel assembly. As previously stated in Section 5.1, the maximum fuel rod cladding temperature is determined to establish that nucleate boiling is not possible while forced cooling is operating.

This requires demonstrating that the highest fuel rod cladding temperatures are less than the local saturation temperature of the adjacent SFP water. The saturation temperature of water increases with increasing pressure and, subsequently, increasing depth. The critical location for localized boiling in the fuel racks is at the top of the active fuel length. The minimum depth of water at the top of the active fuel length is greater than 24 feet. At a water depth of 24 feet, the saturation temperature of water is approximately 24 0 'F.

Holtec Report HI-2043162 5-10 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

A fuel rod can produce F2 times the average heat emission rate over a small length, where F2 is the axial peaking factor. The axial heat distribution in a rod is generally a maximum in the central region, and tapers off at its two extremities. Thus, peak cladding heat flux over an infinitesimal rod section is given by the equation:

= Q x F2 (5-7)

Ac where Q is the rod average heat emission and A, is the total cladding external heat transfer area in the active fuel length region. The axial peaking factor is obtained by dividing the total assembly peaking factor by the radial assembly peaking factor, both given in Table 5.6.1.

As described previously, the maximum local water temperature was computed. Within each fuel assembly sub-channel, water is continuously heated by the cladding as it moves axially upwards under laminar flow conditions. Rohsenow and Hartnett [5.7.1] report a Nusselt-number for' laminar flow heat transfer in a heated channel. The film temperature driving force (ATr) at the peak cladding flux location is calculated as follows:

,&Tf= q, h= (5-8) h = Nu-Ki where hf is the waterside film heat transfer coefficient, Dh is the sub-channel hydraulic diameter, Kw is the water thermal conductivity and Nu is the Nusselt number for laminar flow heat transfer.

In order to introduce some additional conservatism in the analysis, we assume that the fuel cladding has a crud deposit resistance R, (equal to 0.0005 ft2-hr-°F/Btu) which covers the entire surface. Thus, including the temperature drop across the crud resistance, the cladding to water local temperature difference (AT,) is given by the equation AT, = ATf+ R, x q, Holtec Report HI-2043 162 5-11 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

5.8 Results This section contains results from the analyses performed for the postulated offload scenarios.

5.8.1 Maximum Pool Bulk Temperatures For the offload/cooling scenarios described in Section 5.3, the maximum calculated bulk temperatures are summarized in Table 5.8.1. Given the conservatisms incorporated into the calculations, actual bulk temperatures will be lower than these calculated values. Figures 5.8.1 through 5.8.3 each present profiles of net. decay heat load, passive heat losses and bulk temperature versus time for the evaluated transient scenarios.

The results presented in Table 5.8.1 demonstrate that calculated bulk temperatures for all scenarios remain below the allowable bulk temperature limit. The maximum bulk temperature, the time after reactor shutdown that the maximum temperature is reached and the coincident net heat load are given for three core offload scenarios.

5.8.2 Minimum Time-to-Boil and Maximum Boiloff Rate For the offload/cooling described in Section 5.3, the calculated times-to-boil and maximum boil-off rates are summarized in Table 5.8.2. Given the conservatisms incorporated into the calculations, actual times-to-boil will be higher than these calculated values and actual boil-off rates will be lower than calculated. The results show that the minimum time-to-boil is greater than the current licensing basis value of 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and the maximum boil-off rate is less than the current licensing basis value of 44,905 lb/hr (93.6 gpm).

Holtec Report HI-2043162 5-12 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

5.8.3 Local Water and Fuel Cladding Temperatures Consistent with the approach to make conservative assessments of temperature, the local water temperature calculations are performed for a bounding condition with the maximum SFP bulk temperature (from Scenario 3), the maximum coincident total SFP decay heat load (also from Scenario 3) and the maximum coincident individual fuel assembly decay heat load (from Scenario 1). The peak fuel clad superheat (i.e., the maximum clad-to-local water temperature difference) is determined for a bounding scenario. The resultant bounding superheat value was then used to calculate a bounding maximum fuel clad temperature.

The numeric results of the maximum local water temperature and the bounding fuel cladding temperature evaluations are presented in Table 5.8.3. Figure 5.8.4 presents converged temperature contours in a vertical slice through the hot fuel region of the SFP.

The maximum local water temperature and the bounding maximum fuel clad temperature are both lower than the 2401F local boiling temperature at the top of the racks and are also both below the current licensing basis values of 219.50 F and 254.30 F, respectively. These results demonstrate that boiling, including nucleate boiling on clad surfaces cannot occur.

Holtec Report HI-2043162 5-13 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

5.9 References

[5.4.1] "Heat Loss to the Ambient from Spent Fuel Pools: Correlation of Theory with Experiment", Holtec Report HI-90477, Revision 0, April 3, 1990.

[5.4.2] "An Improved Correlation for Evaporation from Spent Fuel Pools", Holtec Report HI-971664, Revision 0.

[5.4.3] "QA Documentation for LONGOR," Holtec Report HI-951390, Revision 0.

[5.4.4] A.G. Croff, "ORIGEN2 - A Revised and Updated Version of the Oak Ridge Isotope Generation and Depletion Code," ORNL-5621, Oak Ridge National Laboratory, 1980.

[5.4.5] "QA Documentation for BULKTEM," Holtec Report HI-951391, Revision 1.

[5.4.6] "QA Documentation for TBOIL," Holtec Report HI-92832, Revision 5.

[5.4.7] "QA Documentation for STER", Holtec Report HI-92776, Revision 11.

[5.6.1] Batchelor, G.K., "An Introduction to Fluid Dynamics", Cambridge University Press, 1967.

[5.6.2] "Validation of FLUENT Version 6.1.18", Holtec Report HI-2032998, Revision 0.

[5.7.1] Rohsenow, N.M., and Hartnett, J.P., "Handbook of Heat Transfer", McGraw Hill Book Company, New York, 1973.

Holtec Report HI-2043162 5-14 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 5.4.1 Historic and Projected Fuel Offload Schedule 5

Initial 23U Average Number of Assembly "U Offload Date Enrichment Burnup Assemblies Weight (kgU) (dd/mmlyyyy)

(wt.%) (NIWD/MTU)

Previously Discharged Fuel Assemblies 2.123 16165 37 460.1 3/4/1987 2.795 29192 74 459.8 17/9/1988 3.199 33512 84 462.5 4/3/1990 3.419 37609 113 463.3 31/8/1991 3.552 37328 88 436.4 6/3/1993 4.162 44578 88 424.8 24/9/1994 4.289 45117 80 424.5 6/4/1996 4.320 48085 69 425.3 14/2/1998 4.685 50328 101 425.1 26/9/1999 4.444 49065 93 418.7 28/4/2001 4.512 49048 101 417.7 3/2/2003 4.396 52000 104 423.0 3/8/2004 4.396 52000 104 423.0 3/2/2006 4.396 52000 104 423.0 3/8/2007 Partial Core Offload Scenario I 4.7 52000 1 96 1 472.6 [ 3/2/2009 Full Core Offload Scenario 2 4.7 52000 101 472.6 3/2/2009 4.7 25000 92 472.6 3/2/2009 Emergency Core Offload Scenario 3 4.7 52000 104 472.6 3/2/2009 4.7 40000 113 472.6 10/4/2009 4.7 3000 80 472.6 10/4/2009 Holtec Report HI-2043162 5-15 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 5.4.2 Key Input Data for Bulk Temperature Evaluatioh Parameter Value Number of Storage Cells in SFP 1478 Reactor Thermal Power 3411 MWt Reactor Core Size 193 assemblies Burnup Uncertainty 5%

SFPCS HX Coolant (CCW) Inlet Temperature 750F SFPCS HX Coolant (CCW) Flow Rate 3400 gpm SFPCS HX SFP Water Flow Rate 1,140,000 lb/hr In-Core Hold Time Before Transfer 100 hrs Fuel Assembly Transfer Rate 4 per hr Holtec Report HI-2043162 5-16 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 5.5.1 Key Input Data for Time-To-Boil Evaluation Parameter Value SFP Surface Area 1394 ft2 Minimum Pool Water Depth 38.3 ft Fuel Racks Displaced Volume 771 ft3 Fuel Assemblies Displaced Volume 5,840 fl3 SFP Net Water Volume 47,215 ft3 Holtec Report HI-2043 162 5-17 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 5.6.1 Key Design Data for Local Temperature Evaluation Parameter Value Radial Assembly Peaking (Nuclear Enthalpy 1.7 Rise Hot Channel) Factor Total Assembly Peaking (Heat Flux Hot 2.7 Channel) Factor Cooled SFP Water Flow Rate through SFPCS 1,140,000 lb/hr Heat Exchange Hydraulically Limiting Fuel Assembly Westinghouse I7x17 Standard (LOPAR)

Fuel Rod Outer Diameter 0.374 in Active Fuel Length 144 in Number of Fuel Rods per Assembly 264 Rack Cell Inner Dimension (Min.) 8.75 in Rack Cell Length 169 in Modeled Bottom Plenum Height 2 in Holtec Report HI-2043162 5-18 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 5.81 Result of Transient Bulk Temperature Evaluations Scenario Maximum Bulk Time After Reactor Coincident Net Heat Temperature (IF) Shutdown (hrs) Load (Btu/hr)

I - Partial Core 127.21 139 22.92x106 2-FullCore 156.99 159 36.67x106 3 -Emergency Core 161.50 159 (X) 38.71x106 (1) For the Emergency Core Offload (Scenario 3) only, the time after reactor shutdown is measured from the second reactor shutdown.

Holtec Report HI-2043162 5-19 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 5.8.2 Results of Loss-of-Forced Cooling Evaluations Scenario Minimum Time-to- Maximum Boiloff Minimum Time to Boil (hrs) Rate (gpm) Reach 10' Above Racks (hrs)

I - Partial Core 11.27 50.34 58 2 - Full Core 4.35 82.45 33 3 - Emergency Core 3.76 87.37 30 Holtec Report HI-2043 162 5-20 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 5.8.3 Results of Maximum Local Water and Fuel Cladding Temperature Evaluations Maximum Local Water Temperature 188°F Bounding Fuel Clad Superheat 25°F Bounding Fuel Clad Temperature 213°F Holtec Report HI-2043162 5-21 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Grid Jun 01, 2004 FLUENT 6.1 (3d, dp, segregated, ske)

Figure 5.6.1 - CFD Model Isometric View Holtec Report HI-2043162 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

2.50E+07 130 125

-..- 2.OOE+07 - 120 115 0

to 0 1.50E+07 110 C, 0~

-4

.105 (U

0 C)

- 100 C) (U CI *In

- 95 M .OOE+07 90

. 85 I , I .- 80 0 40 80 120 160 200 240 280 320 360 400 440 480 Time After Reactor Shutdown (hr)

Figure 5.8.1 -Scenario I - Partial Core Discharge Holtec Report HI-2043162 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.OOE+07 160 tSFP BulHa Teperature 3.50E+07 150 m 3.OOE+07 -140 0

-j (l ~Net Heat Lad cn

, 2.50E+07 _,

j

_ 130 0

0 G 2.00E+07 120 0

0.

ox CD 0

0 1.50E+07 110 7C 0

0

-j 0

0) 1.00E+07 - 100 z

x 5.OOE+06 /11

. 1.

Passive Heat Loss

90 0.OOE+00 80 0 40 80 120 160 200 240 280 320 360 400 440 4130 Time After Reactor Shutdown (hr)

Figure 5.8.2 -Scenario 2 - Full Core Discharge Holtec Report HI-2043162 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.OOE+07 170 3.50E+07 - 160

• 1E

- 150 3 3.OOE+07 I

0 ta (A

-JI

% 2.50E+07 IS 4 AX I'+qu CO)

C co 002.OOE+07 KI

. _ \ \ e t Heat Load

/i 130 5 120 0 C

3 C

Ca VcU 1.50E407 0

-j 110 '

0 1.00E407 C) 100 z SFP Bul Terrperature /

5.OOE+06 90 Passive Heat Loss 0.00E+00 80 a 200 400 600 800 1000 1200 1400 1600 1800 Time After Initial Reactor Shutdown (hr)

Figure 5.8.3 - Scenario 3 - Emergency Full Core Discharge Holtec Report HI-2043162 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

3 0oe-J2 3 53e-32 11 - 4 3':7e-32 3 55e-)2 V. 3 " e-32 3 5e-32 3 ERe-D2 3 48e-32 3 46e-32 3 4t.-32 3 43e-32 34'e-32 3 39e-J2 3 37e-32 3 36e-32 3nLe-32 3 32e-32

-3re-D2 3 29e-32 3 we-32 3 25o-32 Contours of Static Temperature k Sep 02, 2004 FLUENT 6.1 (3d, dp, segregated, ske)

Figure 5.8.4 -CFD Model with Converged Temperature Contours Holtec Rcport HI-2043162 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

6.0 RACK STRUCTURAL INTEGRITY CONSIDERATIONS 6.1 Introduction The structural integrity evaluations performed on the new cask pit rack in each of the two DCPP pools is presented in this section. The analyzed storage rack configuration is depicted in Figure 1.1.1. Inasmuch as the two DCPP pools are mirror images of each other, Figure 1.1.1 represents both pools with appropriately assigned local coordinates.

The analyses undertaken to confirm the structural integrity of the new racks are performed in compliance with the USNRC Standard Review Plan [6.1.1] and the OT Position Paper [6.1.2]. A summary of the key input data, an abstract of the methodology, modeling assumptions, key results, and quantification of the minimum kinematic and stress margins for each of the applicable load combinations is provided in this section.

6.2 Structural Characteristics of the New Cask Pit Rack The Cask Pit region of the DCPP fuel pools is integral to the fuel pool cavity space wherein 16 freestanding modules containing a total of 1,324 storage locations are presently employed. The cask pit region, situated in a corner of the rectangular pool planform space, is surrounded by reinforced concrete walls on two sides and by a welded steel framework (intended to serve as a barrier against excessive sliding of the fuel pool racks adjacent to the pit) on the two other sides.

The cask pit floor liner is approximately 51 inches lower in elevation than the floor liner in the main pool. Below the elevation of the main pool floor, the cask pit is surrounded on all four sides by reinforced concrete lined with 1/4" thick stainless steel liner plate.

To ensure that the top of the new rack will be coplanar with the existing racks (to facilitate fuel handling), a stainless steel platform is provided that is engineered to be emplaced in the pit with a snug fit with the four stainless liner-equipped reinforced concrete walls in the cask pit cavity.

This so-called "cask pit platform", which is discussed in Section 2.1 and illustrated in Figure 2.1.2, is thus a laterally immovable and extremely stiff weldment that is engineered to serve as a staging surface for the cask pit rack.

Holtec Report HI-2043 162 6-1 1368

The four comers of the platform are equipped with precisely machined recesses (see Figure 2.1.3) that are coaxial with and approximately 3/4" larger in diameter than the 5" diameter rack support pedestals, shown in Figure 2.6.4. Thus, when staged on the platform, the sliding of the rack (under inertia loads) is severely restricted while the clearance between the recess and the rack pedestal precludes any differential thermal expansion-related concerns. Finally, to prevent a vertical lift-off of the rack from the platform, two specially machined "connector links" made of stainless steel at each comer of the rack are installed through storage cells adjacent to the pedestal location. The design of the "connector link" is intended to prevent a separation between the rack and the platform and thus prevent excessive sidesway and uplift of the rack during a seismic event. Thus, the combined effect of the lateral confinement of the support pedestals and vertical fixity (Figure 2.1.4) is to limit the kinematic movement of the cask pit rack under seismic events. Because the fluid coupling forces increase with a rack's motion, a chief structural consequence of the above rack staging configuration is to virtually eliminate any significant hydrodynamic coupling between the cask pit racks and the existing array of racks in the main pool. The virtual elimination of a cross coupling between the new rack and the neighboring existing racks in the main pool is further helped by the large gap between them (minimum gap in excess of 20 inches).

The high bending rigidity of the honeycomb cellular design of the racks (described in Section 2) prevents any significant flexural bending of module during the seismic event, thus ensuring that the cask pit rack will not impact any adjacent structures (reinforced concrete wall or the seismic restraint structure). Thus, while the existing racks in the pool are permitted to impact each other, the new rack is engineered to preclude impact with any external surface. This helps mitigate stress levels and maximize the computed safety margins. Nevertheless, in recognition of the rather severe earthquakes postulated for DCPP, the metal thicknesses of the storage cell walls in the cask pit modules is at least 20% greater than a similar rack of recent vintage at any other nuclear plant. The benefits of the design approach summarized in the foregoing become apparent in the stress levels computed for the cask pit rack under the applicable seismic events, as presented in a later subsection herein.

Holtec Report HI-2043162 6-2 1368

6.3 Applicable Load Combinations The applicable loads and their combinations that must be considered in structural qualification of the cask pit rack module are excerpted from References [6.1 .1] and [6.1.2] in the following:

Load Combination Service Level Per the ASME Code [6.3. 11 D+L Level A D + L + T0, (Normal Operating Condition)

D + L + T,0 + E _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

D + L + T. + E Level B D + L + T, + E (Upset Condition)

D + L + Ta + E' Level D D + L + To + Fd (Faulted Condition) where:

D = Dead weight-induced loads (including stored spent nuclear fuel and control elements)

L = Live Load Pf = Force on the racks caused by the removal of a 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)

T. = Differential thermal expansion induced loads Ta = Differential thermal expansion induced loads under postulated abnormal conditions As for the rack itself, T. and Ta produce thermal stresses that are of the self-limiting (secondary) genre since the rack is free to expand or contract due to the existence of gaps between the rack support pedestals and the recessed holes in the platform. Therefore, T, and Ta are not germane to the stress analysis of the cask pit rack. As for the cask pit platform, the thermal loads (T. and Ta) produce primary stresses in the platform since the platform is restrained from thermal growth by 6-3 1368 Holtcc HT-2043 162 Report HI-2043162 1-Toltec Report 6-3 1368

the lateral support shims. Therefore, thermal loads are a contributing factor in the platform stress analysis.

The inertia loading corresponding to the Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) in [6.1.1] are termed Design Earthquake (DE) and Double Design Earthquake (DDE), respectively, in the DCPP FSAR.

In addition to DE and DDE, the plant's owner (PG&E) has also defined two other earthquakes, named Hosgri earthquake (HE) and LTSP (acronym for the Long Term Seismic Program), for which the cask pit rack assembly is analyzed and the resulting stresses are compared with Level D condition limits.

In accordance with the recommendation of Section IV.2 of the OT Position Paper [6.1.2] the cask pit rack will be designed in accordance with ASME Code Section III, Subsection NF limits.

Because of the honeycomb construction of the rack, the effect of mechanical loads such as Pf and Fd result in local stresses and possibly plastic strains that are not within the purview of the stress limits for Class 3 "NF" linear structures. Therefore, mechanical loads are treated separately, mainly with the goal to quantify the extent of local damage sustained by the rack under localized loads, as discussed in Section 7 of this report.

6.4 Synthetic Time-Histories Because the fuel rack is a highly non-linear structure, the applicable earthquakes must be cast in terms of acceleration time-histories for analysis purposes.

The synthetic time-histories in three orthogonal directions (N-S, E-W, and vertical) were accordingly generated in accordance with the provisions of SRP 3.7.1 [6.1.1] for each of the four earthquakes (DE, DDE, HE, and LTSP). The structural damping value used in the generation of the synthetic time-histories utilizes the values prescribed in the DCPP FSAR. No credit for material damping (hysterisis) or fluid damping is incorporated in the time-history generation algorithm for conservatism.

6-4 1368 HoItcc 1-11-2043162 Report HI-2043162 Holtec Report 6-4 1368

As suggested in SRP 3.7.1 [6.1.1], both the response spectrum and the power spectral density corresponding to the generated acceleration time-history should envelope their target (design basis) counterparts with only finite enveloping infractions. The time-histories for the pool slab for all four earthquakes have been generated to satisfy the above enveloping criteria. Finally, the three orthogonal time-histories for each earthquake are required to meet the criterion of statistical independence (i.e., cross correlation factor between any two components, c). The conventional limit for c is 0.15.

Figures 6.4.1 through 6.4.12 provide plots of the time-history accelerograms corresponding to the four discrete earthquakes (DE, DDE, HE and LTSP). These artificial time-histories are used in the non-linear dynamic simulations of the racks to quantify the structural margins of safety.

For reference purposes, the peak accelerations for the twelve time-histories are summarized in Table 6.4.1 below.

Table 6.4.1 ZPA VALUES (in g's) FOR THE APPLICABLE EARTHQUAKES Earthquake Horizontal (N-S) Horizontal (E-W) Vertical DE 0.334 0.260 0.130 DDE 0.520 0.520 0.270 HE 0.710 0.750 0.565 LTSP 0.846 0.987 0.670 6.5 Analysis Methodology 6.5.1 Overview The response of a free-standing rack module (free to move to the extent of available gaps between the rack support pedestal and the recessed holes in the platform) to seismic inputs is highly nonlinear and involves a complex combination of motions (sliding, rocking, twisting, and turning), resulting from an interplay of inertia, friction, impact, and damping effects. Some of the unique attributes of the rack dynamic behavior include a large fraction of the total structural Holtec Report HI-2043 162 6-5 1368

mass in a confined rattling motion, frictional effect at the bottom surface of the pedestal, and fluid coupling effects due to deep submergence. Evidently, 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 acceleration time-histories applied as the forcing functions acting simultaneously.

The DYNARACK solver [6.5.1], the vehicle previously utilized in virtually all rerack O.L.

amendment requests over the past two decades (see Table 6.5.1 below) to simulate the dynamic behavior of the complex storage rack structures, is also used to analyze the DCPP cask pit rack and platform. The analytical premise of DYNARACK is provided in a published technical paper

[6.5.2].

Table 6.5.1 PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK PLANT DOCKET NUMBER(s) YEAR Enrico Fermi Unit 2 USNRC 50-341 1980 Quad Cities I & 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 1 & 2 USNRC 50-275, 50-323 1986 Byron Units I & 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 I USNRC 50-335 1987 I-20 3 16 Re ort HI-2043162 6-6136 HoltecHoltc Report 6-6 1368

Table 6.5.1 PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK PLANT DOCKET NUMBER(s) YEAR Millstone Point Unit 1 USNRC 50-245 1989 Chinshan Taiwan Power Company 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 50401 1991 Hope Creek USNRC 50-354 1990 Kuosheng Units I & 2 Taiwan Power Company 1990 Ulchin Unit 2 Korea Electric Power Co. 1990 Laguna Verde Units I & 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 I & 2 KINS 1996 6-7 1368 Holtec Report H Holtec Report 1-2043 162 Hl-2043162 6-7 1368

Table 6.5.1 PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK PLANT DOCKET NUMBER(s) YEAR Kori4 KINS 1996 Connecticut Yankee USNRC 50-213 1996 Angra Unit I Brazil 1996 Sizewell B United Kingdom 1996' Waterford 3 USNRC 50-382 1996 Vogtle USNRC 50-424 1997 J. A. Fitzpatrick USNRC 50-333 1997 Vermont Yankee USNRC 50-271 1998 Callaway USNRC 50-483 1998 Nine Mile USNRC 50-220 1998 Chin Shan Taiwan Power Company 1998 Millstone 3 USNRC 50-423 1998 Byron/Braidwood USNRC 50-454,50-455, 1999 50-567, 50457 Wolf Creek USNRC 50482 1999 Plant Hatch Units I & 2 USNRC 50-321, 50-366 1999 Harris Pools C and D USNRC 50401 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 162 6-8 1368 Report HI-2043 Holtec Report HI-2043162 6-8 1368

Recognizing that the analysis 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 rack/platform coupled models suitable for a time-history analysis.

Include all fluid coupling interactions and mechanical coupling appropriate to performing an accurate non-linear simulation.

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.

6.5.2 Rack/Platform Coupled Dynamic Model The dynamic modeling of the rack structure is prepared with special consideration of all nonlinearities and fluid effects inherent to the physical problem. Particulars of modeling details and assumptions for the analysis of racks are summarized below.

a. The cask pit 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 with appropriate bending and shear springs [6.5.3] providing the structural connectivity between the lumped masses. The cask pit platform is modeled, beneath the rack, as a six degree-of-freedom rigid body with appropriate overall dimensions and mass properties.

b. Fuel assemblies that are free to rattle within the rack are modeled by five lumped masses located at H, 0.75H, 0.5H, 0.25H, and at the rack base (H is the rack height measured above the baseplate). Each lumped fuel mass has two horizontal displacement degrees-of-freedom. Vertical motion of the fuel assembly mass is assumed equal to rack vertical motion at the baseplate level. The centroid of each fuel assembly mass can be located off-center, relative to the rack structure centroid at that level, to simulate a partially loaded rack (see Figure 6.5.1).

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.

6-9 I 368 Report H Holtec Rcport 1-2043 162 HI-2043162 6-9 1368

d. Fluid coupling between rack and fuel assemblies, and between rack and wall, is simulated by appropriate inertial coupling in the system kinetic energy. Inclusion of these effects uses the methods of references [6.5.4-6.5.8].

e. Fluid damping and form drag are conservatively neglected.

f. Because the top of the rack is over 25' below the pool's free water surface, the sloshing effect is negligible in the water mass surrounding the rack and, therefore, neglected in the rack dynamic model.

g. Potential impacts between the cell walls of the new rack and the contained fuel assemblies are accounted for by appropriate compression-only gap elements between masses involved (see Figure 6.5.2).

h. The cask pit rack is connected to the cask pit platform structure at each comer by a pair of tension-only springs that represent the axial stiffnesses of the connector links.

i. Each rack support pedestal is linked to its support structure by two piece-wise linear orthogonally-disposed friction springs. In addition, the lateral motion of each of the four rack support pedestals in the horizontal direction is constrained by the clearance gap with the recessed holes in the platform.

j. 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.

Table 6.5.2 provides a complete listing of each of the 28 degrees-of-freedom for the rack/platform 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 I, 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. The five masses are connected to each other by an axially rigid member, which enables the fuel masses to vibrate in unison with the rack in the vertical direction. Therefore, the masses represented by the fuel assemblies vibrate independently in the horizontal direction and are driven by the inertia loads and local impact loads. The five fuel masses are connected to the rack model via impact gap elements. Impact loads between the fuel masses and the rack cell wall are obtained upon closure of this gap element. The gap dimensions are determined at each time step by establishing the independent' displacements of the fuel masses and the rack geometric centerline displacement corresponding Holtec Report Hr-2043 162 6-10 1368

to the same elevation. Therefore, the outer boxes surrounding the fuel masses shown in Figure 6.5.2 depict the inside of the fuel cell. The final six degrees-of-freedom are assigned to the platform. The interface between the cask pit rack and the platform is modeled by a series of gap elements and friction springs at the correct locations, which simulate the rack support pedestals, connector links, and recessed holes in the platform.

Table 6.5.2 Degrees-of-Freedom DISPLACEMENT l ROTATION LOCATION (Node) U. U, U, Ox 0f 0 1 Pi P2 P3 q4 q5 q6 2 P7 P8 . Psq q10 gil q12 3 P23 P24 P25 q2 6 q27 q2_

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.

Node 3 is assumed to be attached to the platform at its centroid.

Refer to Figures 6.5.1 and 6.5.3 for node identification.

2 P13 P14 3 P15 P16 4 P17 P18 5 P1i P20 1 P21 P22 where the relative displacement variables qi are defined as:

pi qi(t) + U1(t) i = 1,7, 13, 15,17, 19,21,23

= q(t) + U,(t) i = 2,8,14, 16, 18, 20, 22,24

= q(t) + U(t) i = 3, 9,25 q(t) i=4, 5,6,10,11,12,26,27,28 pi denotes absolute displacement (or rotation) with respect to inertial space qi denotes relative displacement (or rotation) with respect to the floor slab

• denotes fuel mass nodes U(t) are the three known earthquake displacements 162 6-11 1368 Holtec Report HI-2043 Hl-2043162 6-11 1368

Finally, to eliminate the uncertainty in rack/platform dynamic analyses because of the uncertainty in the extent of friction at the pedestal/platform and platform/liner interfaces, multiple simulations are performed by adjusting the friction coefficient. These friction coefficients are chosen consistent with the two bounding extremes from Rabinowicz's data

[6.5.9]t.

6.5.3 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 [6.5.2, 6.5.3]. 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:

[AM] [ +/-.~J = [Q2 + [G]

where:

[M] - total mass matrix (including structural and fluid mass contributions). The size of this matrix will be 28 x 28.

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 t The coefficient of friction (p) between the pedestal supports and the bearing pad is indeterminate.

According to Rabinowicz [6.5.9], results of 199 tests performed on austenitic stainless steel plates submerged in water show a mean value of itto be 0.503 with standard deviation of 0.125. Upper and lower bounds (based on twice standard deviation) are 0.753 and 0.253, respectively. Analyses are therefore performed for coefficient of friction values of 0.2 (lower limit) and for 0.8 (upper limit). The bounding values of 1 = 0.2 and 0.8 have been found to envelope the upper limit of module response in previous rerack projects.

6-12 1368 Holtcc Report HI-2043 Holtec Report 162 HI-2043162 6-12 1368

The above column vectors have a length of 28. The equations can be rewritten as follows:

[d 12 ]=[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 [6.5.1].

6.6 Acceptance Criteria 6.6.1 Kinematic and Stress Acceptance Criteria There are two sets of criteria to be satisfied by the rack modules:

a. Kinematic Criteria Although the OT Position Paper [6.1.2] does not prohibit impact between the rack and neighboring structures, such impact is not permitted under any of the four earthquake scenarios under the kinematic acceptance criteria conservatively set down for the new rack.

b. Stress Criteria The stress limits applicable to the rack structure are derived from the ASME Code, Section 111, Subsection NF [6.3.1]. The terminology used hereunder is in accordance with the ASME Code.

(i) Normal and Unset Conditions (Level A or Level B)

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.

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

F,= .4 Sy 6-13 1368 Holtcc H1-2043 162 Report HI-2043162 Holtec Report 6-13 1368

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

F = Sy .47 -

-444 r) 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:

= 1 (simple support both ends)

= 2 (cantilever beam)

= 0.5 (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:

k + C-x fb., + Cmy fby<

F. DFFb. DyFby 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 Cmy = 0.85 DX = I - (f/F'cx)

Dy = I - (fa/F'cy)

F',x.Y = (n: E)/(2.15 (k r)Xb Holtec Report HI-2043 162 6-14 1368

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

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

a", + bhe+ < 1.0 0.6 Sy F&x 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 S, 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 the service temperature.

'(ii) Level D Service Limits Based on Section F-1334 (ASME Section III, Appendix F) [6.6.1], the limits for the Level D condition are the smaller of 1.2 (SY/Ft) or (0.7SU/Ft) times the corresponding limits for the Level A condition. Su is ultimate tensile stress at the specified rack design temperature.

Exceptions to the above general multiplier are the following:

a) Stresses in shear shall not exceed the lesser of 0.72SY or 0.42Su. In the case of the material used here, 0.72SY governs.

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

Holtec Report H1-2043 162 6-15 1368

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',, and F'y 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:

F, = 0.3 S, f where:

f= (Level D shear stress limit)/(Level A shear stress limit) 6.6.2 Dimensionless Stress Factors As is customary in fuel rack stress analysis, the stress results are presented in dimensionless form. Dimensionless stress factors are defined as the ratio of the actual developed stress to the specified limiting value. The limiting value of each stress factor is 1.0.

The applicable stress factors are:

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)

R6 = 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-2043162 6-16 1368

6.6.3 Allowable g-Load on the Stored Fuel The permissible lateral load on an irradiated fuel assembly has been studied by the Lawrence Livermore National Laboratory. The LLNL report [6.6.2] states that "...for the most vulnerable fuel assembly, the axial buckling load 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.

6.7 Key Input Data and Dynamic Simulations 6.7.1 Key Input Data The principal input data used in the seismic analysis of the cask pit rack and platform are summarized in Table 6.7.1 below.

Table 6.7.1 PRINCIPAL INPUT DATA Item Value Weight of empty rack, (lb) 27,200 Height of rack above baseplate, H (inch) 169 Number of storage cells in the N-S direction t 12 Number of storage cells in the E-W direction t 13 Nominal height of support pedestal Per Table 2.5.1 Storage rack prismatic envelope tt, (inch) 107.9 x 116.8 Storage cell prismatic (square) opening, (inch) 8.75 Weight of fuel assembly, (lb) 1,616 Material properties of structural components Per Table 6.7.2 Structural damping (percent of critical) DE 2%

Structural damping (percent of critical) DDE and 4%

HOSGRI Structural damping (percent of critical) LTSP 5%

Weight of platform, (lb) 28,775 Height of platform, (inch) 54.1 Planar dimensions of platform, (inch) 117 x 117 t Two comer cells are eliminated from each rack, as shown in Figures 1.0.1 and 1.0.2, to avoid interference with the cask restraint system.

tt The prismatic envelope is defined as the rectangular envelope, including verticality tolerances, but not including poison, sheathing or baseplate extensions.

Holtec Report HI-2043 162 6-17 1368

The input quantities presented in Table 6.7.1, wherever applicable, are bounding values rather than nominal data to maximize response. Thus, for example, assuming an upper bound weight of the rack, platform, and fuel are cited because increased inertia translates into increased inertial response to seismic excitations.

The racks will be constructed of stainless steel material. The cell wall, baseplate, sheathing and female pedestal materials will all be made form SA 240, Type 304L. The male pedestals are subjected to higher stresses than the remainder of the rack components, since these members support the weight of the loaded rack and are often subjected to simultaneous axial and bending moments. The male support pedestals and the connector links will be fabricated from SA 564-630, which provides greater strength. The mechanical properties of the rack material are defined in Table 6.7.2 below.

Table 6.7.2 RACK MATERIAL PROPERTY DATA*

(ASME - Section II. Part D)

Young's Modulus Yield Strength Ultimate Strength Material E SY Su (psi) (psi) (psi)

SA240, Type 304L S.S. 27.6 x 106 21,300 66,200 SUPPORT MATERIAL DATA (200 0F)

SA240, Type 304L (upper 27.6 x 106 21,300 66,200 part of support feet) 6 SA-564-630 (lower part of 28.5 x 106 106,300 140,000 support feet; age hardened at 1100IF)

CONNECTOR LINK MATERIAL DATA (200 0F)

SA-564-630 (connector link; 28.5 x 106 106,300 140,000 age hardened at 1100 0F)

SA240, Type 304L 27.6 x 106 21,300 66,200 (connector link nut) I _____ I

• Properties correspond to the reference metal temperature = 200TF.

HoItcc Report HI-2043162 6-18 1368

6.7.2 Dynamic Simulations The plurality of earthquakes, the wide range of coefficients of friction, and the potential of a variable extent of storage cells loaded suggest the need for multiple dynamic simulations. Table 6.7.3 lists the individual scenarios that are analyzed to establish the margins of safety.

Table 6.7.3 DYNAMIC SIMULATION SCENARIOS Scenario Earthquake Fuel Storage Configuration Coefficient of ASME Code Service Number I.D. Friction Condition

1. DE Full 0.2 Level B t

2. DE Full 0.8 Level B

3. DE Half Loaded (Diagonally) 0.2 Level B 4 DE Half Loaded (Diagonally) 0.8 Level B

5. DE Half Loaded (E-W) 0.2 Level B

6. DE Half Loaded (E-W) 0.8 Level B

7. DE Half Loaded (N-S) 0.2 Level B

8. DE Half Loaded (N-S) 0.8 Level B

9. DE Nearly Empty 0.2 Level B

10. DE Nearly Empty 0.8 Level B I1. DDE Full 0.2 Level D

12. DDE Full 0.8 Level D

13. DDE Half Loaded (Diagonally) 0.2 Level D

14. DDE Half Loaded (Diagonally) 0.8 Level D

15. DDE Half Loaded (E-W) 0.2 Level D

16. DDE Half Loaded (E-W) 0.8 Level D

17. DDE Half Loaded (N-S) 0.2 Level D

18. DDE Half Loaded (N-S) 0.8 Level D

19. DDE Nearly Empty 0.2 Level D

20. DDE Nearly Empty 0.8 Level D

21. HOSGRI Full 0.2 Level D

22. HOSGRI Full 0.8 Level D

23. HOSGRI Half Loaded (Diagonally) 0.2 Level D

24. HOSGRI Half Loaded (Diagonally) 0.8 Level D

25. HOSGRI Half Loaded (E-W) 0.2 Level D

26. HOSGRI Half Loaded (E-W) 0.8 Level D

27. HOSGRI Half Loaded (N-S) 0.2 Level D t The evaluation conservatively used normal (Level A) condition allowable strcsses for all DE seismic cases.

Holtec Report HI-2043162 6-19 1368

Table 6.7.3 DYNAMIC SIMULATION SCENARIOS Scenario Earthquake Fuel Storage Configuration Coefficient of ASME Code Service Number I.D. Friction Condition

28. HOSGRI Half Loaded (N-S) 0.8 Level D

29. HOSGRI Nearly Empty 0.2 Level D

30. HOSGRI Nearly Empty 0.8 Level D

31. LTSP Full 0.2 Level D

32. LTSP Full 0.8 Level D

33. LTSP Half Loaded (Diagonally) 0.2 Level D

34. LTSP Half Loaded (Diagonally) 0.8 Level D

35. LTSP Half Loaded (E-W) 0.2 Level D

36. LTSP Half Loaded (E-W) 0.8 Level D

37. LTSP Half Loaded (N-S) 0.2 Level D

38. LTSP Half Loaded (N-S) 0.8 Level D

39. LTSP Nearly Empty 0.2 Level D

40. LTSP Nearly Empty 0.8 Level D

41. HOSGRI Full 0.2 Level D

42. *HOSGRI Full 0.8 Level D Scenario Nos. 41 and 42 are identical to Scenario Nos. 21 and 22 except that the compression-only gap elements that represent the sixteen (16) platform lateral shims (refer to Figure 6.5.3) are preloaded to simulate the restraint of thermal expansion due to a bounding pool water temperature rise of 181'F.

6.8 Margins Against Kinematic and Stress Limits 6.8.1 Satisfaction of Kinematic Acceptance Criteria All of the dynamic simulations listed in Table 6.7.3 show that the cask pit rack does not impact the neighboring cask pit walls or the restraint structure at any point in time.'The ratio of the available gap to the maximum kinematic excursion towards the proximate structure is in excess of 3 in every simulation. Therefore, the prescribed criterion of no impact with adjacent structure is satisfied with large margins.

6-20 1368 Holtec Report HI-2043 162 Report HI-2043162 6-20 1368

Finally, the cask pit platform structure, which serves to stage the rack, is not found to lift off the liner surface due to existence and arrangement of the 16 support shims located between the platform and cask pit wall liner.

6.8.2 Satisfaction of Stress Acceptance Criteria As is typical of all "honeycomb connected construction" modules [6.5.2] the locations of maximum stress in the cask pit rack are (i) the base of the cell wall, just above the baseplate, (ii) the support pedestal-to-baseplate junction (welds) and (iii) the threaded joint in the adjustable pedestals. In addition, in the case of the cask pit rack, the axial stress in the connector links is a candidate for potential stress exceedance. Finally, for the stored fuel, the maximum deceleration "g" load sustained by the fuel due to its rattling in the storage cell during the seismic event is a matter of safety concern.

As described in Subsection 6.6, the margin of safety in fuel rack dynamic simulations is expressed in terms of dimensionless ratios of actual stress to allowable stress, referred to as stress factors. Table 6.8.1 provides the maximum value of stress factors for each of the Table 6.7.3 simulation scenarios at each of the above-mentioned vulnerable locations in the fuel rack.

Holtec Report HI-2043 162 6-21 1368

Table 6.8.1 MAXIMUM STRESS FACTORS AT CRITICAL LOCATIONS IN THE CASK PIT RACK Maximum Computed Stress Factor Maximum g-load experienced Scenario Seismic (Maximum Allowable Value = 1.0) by the fuel Event Base of Rack Female Support Female Rack-to-Platform (Allowable is 63g's, per Cell Walls Pedestal Pedestal Connector Links Subsection 6.6.4)

Threads

1. DE 0.425 0.069 0.679 0.145 1.042

2. DE 0.434 0.067 0.704 0.173 0.978

3. DE 0.220 0.040 0.369 0.120 0.874

4. DE 0.244 0.043 0.402 0.150 0.869

5. DE 0.218 0.040 0.365 0.113 0.965

6. DE 0.242

• 0.043 0.405 0.157 0.983

7. DE 0.216 0.040 0.366 0.116 0.871

8. DE 0.239 0.042 0.401 0.154 0.965

9. DE 0.049 0.008 0.080 0.031 0.938

10. DE 0.050 0.010 0.085 0.040 0.896

11. DDE 0.259 0.042 0.436 0.183 1.785

12. DDE 0.275 0.043 0.455 0.215 1.817

13. DDE 0.125 0.022 0.242 0.107 1.614

14. DDE 0.157 0.031 0.273 0.149 1.624

15. DDE 0.131 0.023 0.237 0.105 1.542

16. DDE 0.166 0.034 0.282 0.157 1.713

17. DDE 0.130 0.022 0.223 0.102 1.540

18. DDE 0.157 0.030 0.273 0.145 1.582

19. DDE 0.029 0.005 0.047 0.025 1.572

20. DDE 0.039 0.008 0.081 0.063 1.597

21. HOSGRI 0.362 0.054 0.555 0.282 2.587 16 6-2 _1368. ___

Rcport .I24 HoUcc Ror Hte HI-2043162 6-22 1368

Table 6.8.1 MAXIMUM STRESS FACTORS AT CRITICAL LOCATIONS IN THE CASK PIT RACK Maximum Computed Stress Factor Maximum g-load experienced Scenario Seismic (Maximum Allowable Value = 1.0) by the fuel Event Base of Rack Female Support Female Rack-to-Platform (Allowable is 63g's, per Cell Walls Pedestal Pedestal Connector Links Subsection 6.6.4)

Threads

22. HOSGRI 0.386 0.077 0.612 0.413 2.926

23. HOSGRI 0.198 0.028 0.287 0.152 2.500

24. HOSGRI 0.243 0.058 0.391 0.293 2.606

25. HOSGRI 0.168 0.025 0.273 0.128 2.490

26. HOSGRI 0.236 0.058 0.375 0.285 2.834

27. HOSGRI 0.196 0.027 0.273 0.125 2.465

28. HOSGRI 0.241 0.059 0.389 0.299 2.639

29. HOSGRI 0.031 0.005 0.048 0.028 2.649

30. HOSGRI 0.048 0.011 0.085 0.065 2.649

31. LTSP 0.341 0.063 0.636 0.308 2.010

32. LTSP 0.410 0.085 0.761 0.469 1.993

33. LTSP 0.176 0.034 0.349 0.165 2.079

34. LTSP 0.247 0.048 0.470 0.277 2.163

35. LTSP 0.177 0.034 0.349 0.165 2.092

36. LTSP 0.247 0.046 0.470 0.277 2.327

37. LTSP 0.177 0.034 0.346 0.165 2.092

38. LTSP 0.247 0.047 0.470 0.277 2.193

39. LTSP 0.035 0.006 0.063 0.043 1.980

40. LTSP 0.064 0.015 0.130 0.109 2.017

41. HOSGRI 0.332 0.058 0.586 0.323 2.490

42. HOSGRI 0.387 0.089 0.640 0.481 2.941

.oteReor

_4 16 6.2 136 Holtec Report HI-2043162 6-23 1368

It is evident from the summary results compiled in Table 6.8.1 that there is considerable margin in the stress factors against the allowable limit, and that the maximum "g" load experienced by the spent fuel is a fraction of the allowable.

The interface loads between the racks and the platform, obtained from the dynamic analyses, were applied to the platform to evaluate its stress compliance. Results of the stress analysis using the maximum interface loads from the totality of the dynamic simulations are presented, again, in terms of dimensionless stress factors in Table 6.8.2, which indicate large margins of safety.

Table 6.8.2 MAXIMUM STRESS FACTORS IN THE PLATFORM UNDER LIMITING INTERFACE LOADS FROM THE TABLE 6.7.3 DYNAMIC SIMULATIONS Region Maximum Stress Factor Corner Support Plates 0.556 Corner Lock Plates 0.833 Welds 0.833 Mainframe 0.769 6.9 Assessment of Margin Against Fatigue Failure The dynamic simulations described in the foregoing showed that the cask pit rack behaves 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/platform 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 pV, where p is the limiting interface coefficient of friction and V is the concomitant vertical thrust on the support surface (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/platform interface, F, is given by F S p N (r)

Holtec Report HT-2043 162 6-24 1368

where N (vertical thrust) is the time-varying function ofT. F does not always equal gN; rather, gN 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. If the above inequality is not satisfied then the pedestal slides, resulting in a confinement reaction from the platform recess surface. In summary, the reaction force at the pedestal/support 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 rack support pedestals and connector links) 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 rack safety evaluation.

The time-history method of analysis, deployed in the evaluation of the rack structure, 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=y ni Ni 6-25 1368 Holtec H1-2043 162 Report HI-2043162 Holtec Report 6-25 1368

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

Using the archived results of the spent fuel rack dynamic analyses 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 = 0.513 due to the combined effect of one SSE (see note below) and twenty OBE events. This is well below the ASME Code limit of 1.0.

Note: LTSP results, where governing, have conservatively been used in lieu of Hosgri and DDE results in determining the value for U.

Holtec Report HT-2043 162 6-26 1368

6.10 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.

[6.3.1] ASME Boiler & Pressure Vessel Code, Section 111, Subsection NF, 1989 Edition.

[6.5.1] A.I. Soler, "DYNARACK Validation Manual", Holtec Report No. HI-91700 (1991).

[6.5.2] Soler, A.I. and Singh, K.P., "Seismic Responses of Free Standing Fuel Rack Constructions to 3-D Motions", Nuclear Engineering and Design, Vol. 80, pp. 315-329 (1984).

[6.5.3] Levy, S. and Wilkinson, J.P.D., "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering,"

McGraw Hill, 1976.

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

[6.5.5] 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.

[6.5.6] Paul, B., "Fluid Coupling in Fuel Racks: Correlation of Theory and Experiment", (Proprietary), Holtec Report HI-88243.

[6.5.7] 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.

[6.5.8] 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).

[6.5.9] Rabinowicz, E., "Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," MIT, a report for Boston Edison Company, 1976.

Holtec Report HI-2043 162 6-27 1368

[6.6.1] ASME Boiler & Pressure Vessel Code,Section III, Appendices, 1989 Edition.

[6.6.1] UCID-21246, "Dynamic Impact Effects on Spent Fuel Assemblies, Lawrence Livermore National Laboratory, dated October 20, 1987.

162 6-28 1368 Rcport H[-2043 Holt= Report Holtec HI-2043162 6-28 1368

0.4-0.3 - _. _ -_ -_

0.2-z 0.1-0.0-

-J o-0.1-

-0.2

-0.2 - - _ _ _ _ _ _ _ _ _ _; 0 000 - f0 0? :

-0.3-

- ::- --l-Peak Acceleration =0.334

-0.4-0 8 16 24 TIME (sec)

FIGURE 6.4.1 F'roject 1368 ACCELERATION TIME HISTORY DE-X Report No 204316: 2 i2

0.3 0.2-0.1-0.

-0.0-

- Peak Acceleration 0.260

-0.3-0 8 .16 24 TIME (see)

FIGURE 6.4.2 IProject 1368 ACCELERATION TIME HISTORY DE-Y Report No 204316 2

0.2 0.

0.0- - - . --

O 1

-Peak Acceleration =0.130

-0.2 0 8 16 24 TIME (sec)

FIGURE 6.4.3 project 1368 Fp ACCELERATION TIME HISTORY DE-VT Report No 204316 ,2

0.6 0.4 z

0.

-0.

-04

-Peak Acceleration = .520

-0.6-0 8 TIME (sec) 1.6 .24 FIGURE 6A44 Project 1368 . ACCELERATION TIME HISTORY DDE-ER Reportt 2043162

0.6 -

0.4 w

F-C. -0.2

-0.47

-Peak Acceleration =0.520

-0.6 0 8 16 24 TIME (sec)

FIGURE 6.4.5 F'roject 1368 ACCELERATION TIME HISTORY DDE-N/S Report No 204319 38

0.3 - -Peak Acceleration 0.270,.

0.2 1 .0  ;. X ,- -r.- l ... .- ;_ _ _, '. _ _ _^

z 0.

-0.2- . - --

-0.3 0 8 16 24 TIME (sec)

FIGURE 6.4.6 F'roiect 1368 ACCELERATION TIME HISTORY DDE-VT Report No 204316 2

0.8-

-Peak Acceleration = 0.710 0.6-0.4 w

-0.4 -- - - - - -- _ __

-0.6 - _ 7 __ __777_ __ __ _

-0.8 0 8 16 24 TIME (sec)

FIGURE 6.4.7 Fpproject 1368 ACCELERATION TIME HISTORY HOSGRI-X Report No 204316:;2

0.8-0.6-

-~0.4 -

0.2-0.0---

-0.8

-0.4-

-0.6- -eak Acceleration

=0.750

-0.8 0 8 16 24 TIME (sec)

FIGURE 6.4.8 Fproject 1368 ACCELERATION TIME HISTORY HOSGRI-Y Report No 204316 )2

0.6 0.4 -

0.

0.0 - -

-JI w

0.2-Pe kA- - ceeaio

-; .55l- -; -  :- - -^_;_ _- _ ~~~~

-0 .4 -_ _ _ _ _ _ _ _

--PeakAcceleration =0.565

-0.6 0 8 16 24 TIME (sec)

FIGURE 6.4.9

'roject 1368 UP ACCELERATION TIME HISTORY HOSGRI-VT Report No 204316 2

0.8-0.6-0.4-0 -

I-0.0-

-0.

-0.6-  : - :-  :::-  ;-

-0.8 - ... ..

-- Peak Acceleration = 0.987

-1.0 0 8 16 24 TIME (sec)

FIGURE 6.4.10 Project 1368 ACCELERATION TIME HISTORY LTSP-EW Report No 2043162

1.0 l~~-Peak Acceleratio .4  :  :- -- . --- - - -

-0.4 0.6 10.2 W .

U o-0.2

-0.4 -

-0.6

-0.8 0 8 16 24 F O2:-.,... . .;

TIME (sec)

FIGURE 6.4.11 F'roject 1368 ACCELERATION TIME HISTORY LTSP-NS Reirort No 204316 2

,2

-~0.4-z 02

-0.2 0.2 - - --

- 0.0

-0A

-- .;Fv

• -_ -- a-#.a4 w0 2

-0.6 0 8 16 24 TIME (sec)

FIGURE 6.4.12 Iroject 1368 UP ACCELERATION TIME HISTORY LTSP-VT. Report No 204316.2

FIGURE 6.5.1; SCHEMATIC OF THE DYNAMIC MODEL OF A SINGLE RACK MODULE USED IN DYNARACK H13681 IREPDRT HI-2043162 CIAVINGSIO-RA RES113681RET-HI-2043162ICHP 60WE6.51

y

-FUEL ASSEMBLY/CELL GAP/ELEMENT IMPACT I- XE .

I FIGURE 6.5.2; FUEL-TO-RACK GAP/IMPACT ELEMENTS AT LEVEL OF RATTLING MASS MR68 IHI-2043162 G,\DRAVINGS\O-FIGURES\1368\REPORT-HI-2043162\CHP..6\FIGURE6.5.2

PLATFORM/CASK PiT SHIM LOCATIONS FRICTION ELEMENTir GAP ELEMENT (TYP.)

(SEE DETAIL) (TYP.)

VERTICAL-FRICTION ELEMENT RACK PEDESTAL TO PLATFORM -CASK PIT RACK SHIM LOCATION DETAIL SUPPORT PEDESTAL GAP/IMPACT (TYPICAL)

ELEMENT ALIGNMENT HOLE DETAIL (TYPICAL)

FIGURE 6.5.3; SCHEMATIC OF THE DYNAMIC MODEL OF CASK PIT RACK PLATFORM USED IN DYNARACK IqhJl W I~RPnPT Hf-PQ4fl4 1,P

.IDRAVIN5SIO-FIGiJESI1368lREPLJRT-llI-2O4316I6?CHP61FIG65.3

7.0 MECHANICAL ACCIDENTS 7.1 Introduction The USNRC OT position paper [7.1.1] specifies that the rack design must ensure the functional integrity of the spent fuel racks and the pool under all credible mechanical 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 Spent Fuel Pool (SFP) Cask Loading Area (CLA) in Diablo Canyon Power Plant (DCPP), Units I and 2.

The proposed change does not alter assumptions or results under the current DCPP 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 cask pit rack, the pool liner, and the concrete slab in the cask pit area subsequent to the impact of a fuel assembly, a rack or a rack platform under various drop scenarios. Four categories of fuel assembly, rack and rack 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. 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 Metamic neutron absorber.

Stated in quantitative terms, this criterion implies that the plastic deformation of the cell walls should not extend more than 1515/16 inches (downwards) from the top. In order to utilize an upper bound of kinetic energy at impact, the impact is assumed to be a fuel assembly and its handling tool (3,000 Ibs) and the free-fall height is conservatively assumed to be 3 feet [7.2.1].

Holtec Report HI-2043162 7-1 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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 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. Figure 7.2.1 depicts the finite element model used to evaluate this scenario.

The second class of drop event postulates that a fuel assembly falls through an empty storage cell impacting the fuel assembly support surface (i.e., rack baseplate) as schematically shown in Fig. 7.2.2.

This so-called "deep" drop event threatens the structural integrity of the baseplate. If the baseplate is pierced, and the fuel assembly impacts the rack platform or 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 and that the maximum lowering of the baseplate is shown to be acceptable by the criticality evaluations (see Section 4 for further discussion). Note that a dropped fuel assembly may hit the rack baseplate over a supporting pedestal resulting in significant impact on the floor. However, this deep drop scenario is bounded by the rack drop accident as described below in terms of damage to the SFP floor, and therefore is not directly evaluated.

As discussed in Section 3.5, DCPP plan to upgrade the FHB crane to meet single failure proof requirements prior to installation of the new cask pit rack and support platform. Although not required to be postulated as drop accidents, the third and fourth drop events were analyzed for defense-in-depth.

The third and fourth classes of drop events assume that a lifted empty rack and a rack platform fall from the top of the SFP water level and impact the floor of the cask pit, respectively. The structural integrity of the SFP liner and the concrete slab in the cask pit area must be demonstrated to be maintained in the rack and rack platform drop events. The acceptance criterion for both the third and fourth drop events is that the pool structure must not be damaged (i.e., develop through-thickness cracks in the concrete slab or punch through the liner and the underlying steel plate) leading to uncontrolled water leakage.

Figures 7.2.3 and 7.2.4 depict the rack and platform drop events, respectively.

Holtec Repon H1-2043162 7-2 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

7.3 Incident Impact Velocity Holtec Rerort HI-2043 162 7-3 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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.

The physical properties of material types undergoing deformation in the postulated impact events are summarized in Table 7.4.2.

Holtec Report HI-2043162 74 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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 15.5 inches, which is slightly smaller than the design limit of 1515/16 inches. It should be pointed out that the weld connection between the inner surface of the impacted cell wall and the adjacent cell is not considered in the model; the actual plastically crushed depth would be significantly smaller than the predicted value if this additional weld connection were considered in the rack model. Therefore, it can be concluded that the damage of the fuel cell due to a shallow drop event 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 instantaneously lowered by a maximum of 3.55 inches during the impact process. The deformation of the rack 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.

7.5.3 Rack Drop Event The liner plate experiences a maximum plastic strain of 0.019 in the rack drop accident, as shown in Figure 7.5.3, which is much smaller than the failure strain of the liner material (0.38). The damage to the slab is local and limited to the small area on the concrete surface as demonstrated in Figure 7.5.4, which imposes no structural threat to the cast pit slab. 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-2043162 7-5 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

7.5.4 Rack Platform Drop Event The platform drop does not result in through-thickness cracks in the slab as shown in Figure 7.5.5, indicating that the pool floor can maintain its overall integrity. It should be pointed out that the predicted concrete response to the platform impact is very conservative, since the slab reinforcement, which is not considered in the analysis, would significantly suppress the cracks especially at the slab periphery. The liner plate experiences a maximum plastic strain of 0.014 in the rack platform drop accident, as shown in Figure 7.5.6, which is much smaller than the failure strain of the liner material (0.38). Since the liner is not breached in the drop event, there will be no abrupt or uncontrollable loss of water from the pool.

7.6 Conclusion The drop events postulated for the DCPP SFP 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 maximum downward displacement of the baseplate is limited to 3.55 inches, which ensures that fuel will remain in a subcritical condition. As previously stated, the consequences of a deep-drop scenario where a fuel assembly enters a cell directly above a rack support pedestal are bounded by the rack drop evaluation.

The analyzed rack drop event 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. Similar to the rack drop, the postulated rack platform drop event is also found to be acceptable, since the structural integrity of both the liner plate and concrete slab is not compromised by the accident. In conclusion, the new Holtec spent fuel rack for the DCPP cask pit possess acceptable margins of safety under the postulated mechanical accidents.

Holtec Report HI-2043162 7-6 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

7.7 References

[7.1.1] "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978 as amended by the USNRC letter dated 1/18/1979.

[7.2.1] Diablo Canyon Nuclear Plant Specification No. 10029-N-NPG, Rev. 3.

Holtec Report HI-2043162 7-7 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 7.4.1 IMPACT EVENT DATA Impactor Impactor Drop Impact Case Weight (lb) Type Height Velocity WegtI) ye(in) (in/sec)

Fuel assembly

1. Shallow drop event 3,000 & 36 149.3 handling tool Fuel assembly

2. Deep drop event 3,000 & 205 282.5 handling tool

3. Rack drop event 27,2O Empty rack 531 217.6

4. Rack platform drop event 22,825 Rack platform 531 373.0 t Bounds maximum empty rack weight of 26,000 lb.

Holtec Report HI-2043 162 7-8 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 7.4.2 MATERIAL DEFINITION Elastic Stress Strain Material Name Material Density Modulus Type (pc) (psi) First Yield (psi) Failure (psi) Elastic Failure Stainless Steel SA240-304L 501 2.787e+07 2.315e+04 6.81 Oe+04 8.306e-04 3.800e-01 Stainless Steel SA240-304 501 2.76e+07 2.500e+04 7.100e+04 9.058e-04 3.800e-01 (Liner) I _ _ __ _ _

Stainless Steel SA564-630 501 2.86e+07 1.092e+05 1.400e+05 3.818e-03 1.400e-01 Carbon Steel SA-36 490 2.88e+07 3.300e+04 5.800e+04 1.146e-03 2.000e-01 Concrete fc=5,000 psi 140 4.03 e+06 Holtcc Report H1-2043162 7-9 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

DIABLO CANYON "SHALLOW DROP" OF FUEL Fig. 7.2.1 Finite Element Model of the "shallow" drop event Holtec Report H1-2043162 7-10 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

-FLsL ASSEMBLY

~> p-PAEI, e

't!¶'AET PEGIdO1 Fig. 7.2.2 Schematics of the "deep" drop scenario 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-2043162 7-11 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Fig. 7.2.3 Schematics of the rack drop event Note: This figure is primarily provided to indicate the impact scenario. The configuration of the rack and the SFP is not intended to be accurate.

Holtec Report HI-2043162 7-12 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

DIABLO CANYON "PLATFORM DROP' Fig. 7.2.4 Finite element model of the rack platform drop event Holtec Report HI-2043162 7-13 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Fig. 7.5.1 "Shallow" Drop: Maximum Plastic Strain Holtec Report HI-2043162 7-14 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

DIABLO CANYON "DEEP DROP" Time = 0.024 Fringe Levels Contours of Z-displacement min=-3.5457, at node# 48214 9.384e max=0.938385, at node# 52527 4.900e-01 4.1 57e412 I1I

-4.068e-01_

-8.552e-01 _

-1.304e#00 -

-1.752e#0O

-2.200e*00-

-2.649e#00 -

-3.097c#00

-3.546e#00 -

i Fig. 7.5.2 "Deep" Drop: Maximum Baseplate Vertical Displacement Holtec Report HI-2043162 7-15 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

DlABLO CANYON 'RACK DROP' Timed 0.005 Fringe Levels Contours of Effective Plastic Strain 1.934e-002 max Ipt value min=0. at eleml 80004 1.740e-002I max=0.0193364, at eleml 80467 1.547L-002 _

I 1.354e-002 _

1.1 60e-002 _

9.668e"03 _

7.735e-003 _

5.801 e-003 -

3.867e-003 I

1.934e-003 O.O00e+0OOO 1

i z

Fig. 7.5.3 Rack Drop: Maximum Plastic Strain - Liner Holtec Report HI-2043162 7-16 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Fig. 7.5.4 Rack Drop: Cracks in the CLA Slab Concrete Holtec Report HI-2043162 7-17 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Fig. 7.5.5 Rack Platform Drop: Cracks in the CLA Slab Concrete Holtec Renort HI-2043162 7-18 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

DIABLO CANYON "PLATFORM DROP Tlime 0.0050002 Fringe Levels Contours of Effective Plastic Strain mInO. at elernm 1 1.380c-002-max=0.01 38039. at elernt 931 1.242e..002 _

9.6 63c-003 _.

8.282e-003-6.902e-003 -

S.S22e-0303 _.

2.761 e-4003 1.3BOe-003 0.00 0e 000 yZ Fig. 7.5.6 Rack Platform Drop: Maximum Plastic Strain - Liner Holtec Report HI-2043162 7-19 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

8.0 POOL STRUCTURE INTEGRITY CONSIDERATIONS 8.1 Introduction The reinforced concrete spent fuel pool structure serves to withstand all service condition loadings and is designated as a seismic Category I structure in the DCPP FSAR. The consequences of the addition of a fuel storage rack in the cask pit cavity must be evaluated to quantify any increase (or decrease) in the margins of safety available in the presently licensed configuration. This section provides a summary of the evaluations required and performed to support the required structural safety evaluation. The structural loads imposed by the new rack and platform are confined to the floor and walls of the cask pit cavity area. It will be demonstrated that the rack will not be in contact with any portion of the walls of the spent fuel pool at any elevation above the main floor of the pool. Thus, the discussion and evaluation are limited to that portion of the overall pool structure.

8.2 Description of Cask Pit Cavity Structure The DCPP fuel pool structure is founded on bedrock. In both DCPP units, the cask pit is integral to the fuel pool, occupying a comer of the pool (the southwest comer in Unit One, and northwest comer in Unit 2). The two pools are mirror image of each other. As is typical of many Westinghouse PWRs, the cask pit cavity is deeper than the fuel pool proper: at DCPP, the cask pit is deeper than the main fuel pool by 4'-3". Thus, the bottom 51 inches of the reinforced concrete walls is fully supported from the outside by the hard subgrade.

The entire surface of the cask pit cavity area (floor and walls), is equipped with a 1/4" inch thick austenitic stainless steel liner that provides confinement of the pool water and protects the reinforced concrete pool walls and base mat from direct contact with the pool's borated water.

The cask pit floor liner plate also has an additional 3/" thick carbon steel backing plate, underlain by a 4'-9" thick reinforced concrete slab founded on bedrock (topped by 5" of concrete cover).

Two of the four reinforced concrete walls of the 51 inch deep cavity in the comer of the pool are 8-I 1368 Holtec Report 162 HI-2043 162 Report HI-2043 8-1 1368

merely extensions of the much larger spent fuel pool walls, which are 6'-O" thick. The remaining two 51 inch high walls defining the cask pit depression are 5'-0" thick.

The key data on the reinforced concrete base mat and walls in the cask pit cavity regiont germane to the structural strength of the pool enclosure is provided in Table 8.2.1 herein.

Table 8.2.1 REINFORCED CONCRETE WALL AND BASEMAT DATA FOR THE CASK PIT CAVITY REGION (Unit Oneo 't Base Mat South Wall West Wall Total thickness, inch 5'-2" 6'-0" 69-0" Concrete Compressive DE/DDE: 5000 psi DE/DDE: 5000 psi DE/DDE: 5000 psi Strength, psi Hosgri: 5650 psi Hosgri:5650 psi Hosgri: 5650 psi LTSP: 0.95(5650)= LTSP: 0.95(5650)= LTSP: 0.95(5650)=

5367 psi 5367 psi 5367 psi Rebar size and spacing in N- #11 @ 6" #11 @6" # @6" S direction (horizontal for the wall, near water surface)

Rebar size and spacing in E- #11 @ 6" #11 @ 7" #11 @ 7" W direction (vertical for the wall) 8.3 Structural Consequences of the Cask Pit Rack to the Cask Pit Cavity As discussed in Subsection 2.1, the cask pit rack is designed to be situated on a cask pit platform (CPP), which fits snugly in the 51 -inch deep cask pit cavity of the spent fuel pool and is totally confined in the subterranean portion of the cask pit cavity (see Figure 8.3.1). The platform is actually wedged against the sidewalls of the cask pit cavity to eliminate any differential movement between the pool walls and the platform. This ensures that the difference in seismic motion between the cask pit cavity structure and the top of the platform is negligible.

t Throughout this document, the term cask pit cavity region refers to the 4'-3" deep subterranean portion of the fuel pool where the cask pit platform is situated.

t Unit 2 is identical in all structural attributes, except it is a mirror image of Unit 1.

Koltec 1-11-2043162 Report HI-2043162 8-2 1368 Mice Report 8-2 1368

The rack is mounted on the platform and connected to it through "connector links" (see Subsection 2.1 and Figures 2.1.4 and 2.1.5) that act to restrain the rack in the vertical upward direction. The rack pedestals are inset into recessed holes in the top of the platform (see Figure 2.1.3), which provide lateral rack restraint against sliding and limit lateral motion. Thus, under seismic events, the cask pit rack response produces negligible hydrodynamic forces on the cask pit walls. A free body diagram of the cask pit platform makes it evident that the reaction to the rack and platform inertia forces during an earthquake event will be provided principally by:

• the friction forces between the platform and the cask pit floor liner and if the friction forces are inadequate to resist the lateral forces on the platform, then by:

• the compressive contact forces between the cask pit walls, and the platform, and

• the friction forces between the platform shims and the cask pit wall liner.

These contact forces act on the 51" tall region of the cask pit walls, which are below grade and thus result in the structural response of the wall that simulates a plate-on-an-elastic foundation.

Inasmuch as the structural response of a locally loaded plate-on-elastic foundation produces localized flexure and shear [8.2.1], it follows that the platform-to-cask pit wall reaction forces will produce local moments and shear in the loaded region of the cask pit cavity.

In contrast to the earthquake-induced loading of the reinforced concrete structure by the cask pit racks, the high-density racks, licensed and installed in the 1980s, exert impact loads as well as substantial hydrodynamic loads on the pool walls. These loads, however, act at locations on the reinforced concrete walls that are far removed and structurally apart from the region loaded by the cask pit racks. Thus, the response of the reinforced concrete structure of the plant to the loadings from the existing racks do not interact, i.e., they are completely uncoupled. The moments and shears on the reinforced concrete structure from the loadings due to the cask pit rack can, therefore, be analyzed independently from the loadings produced by the existing array of racks (which do not exert any loads on the cask pit cavity region). Therefore, the addition of 8-3 1368 Report HI-2043 Holtec Rcport 162 HI-2043162 8-3 1368

the temporary rack and its support platform in the cask pit area will not have an adverse effect on the overall structural integrity of the spent fuel pool 8.4 Load Cases and Combinations The DCPP reinforced concrete pool structure was designed using ACI 318-63 Code. The applicable load combinations are taken from reference [8.3.1].

8.4.1 Normal Load Conditions For the normal load conditions the pool structure was evaluated using the Working Stress Design (WSD) method with the following numbered load case combinations.

(1) C=D+L+DE+T.+Ro (2) C=D+L+W +T.+R&

where:

C= Required Load Capacity of the Section D= Dead Weight L= Live Load DE= Design Earthquake W= Normal Wind To= Normal Thermal Loads R.= Normal Pipe Reactions 8-4 1368 Holtec Report HI-2043 162 Report HI-2043162 84 1368

8.4.2 Abnormal Load Conditions For the abnormal load conditions the pool structure was evaluated using the Ultimate Stress Design (USD) method with the following numbered load case combinations.

(3) U=D+L+TA+RA+ 1.5PA (4) U=D+L+TA+RA+1.25PA+ I.O(YJ+YNI+YR)+ 1.25DE (5) U=D+L+TA+RA+ 1.OPA + 1.O(Yj+Ym+YR)+DDE (6) U=D+L+TA+RA+ 1.0PA +1.0(Yj+Ym+YR)+HE (7) U=D+L+TA+RA+ 1.0PA + 1.0(YJ+YN+YR)+LTSP where:

U = Required Load Capacity of the Section TA= Abnormal Thermal Loads RA= Abnormal Pipe Reactions PA= Abnormal Pressure Loads Yj= Jet Impingement Loads YN¶= Missile Impact Loads YR= rupture restraint Reactions DDE= Double Design Earthquake HE= Hosgri Earthquake LTSP= Long Term Seismic Program Earthquake Obviously, many of the load cases listed above, such as wind and piping loads, do not apply to the cask pit cavity design.

Holtec Report HI-2043162 8-5 1368

8.4.3 Affected Load Cases The following load conditions are commonly evaluated for pool structures where racks are being installed or removed. Each condition is assessed below to address any potential influence from the addition of the new cask pit rack and associated support platform.

a. Dead Load: The cask pit cavity's base mat is subjected to the dead load of the loaded cask pit rack and its platform. This load is applied as a bearing pressure on the platform/cavity liner contact surface.

b. Hydrostatic Load: This load, exerted by the column of water (43 feet nominal height) acts on the cask pit cavity liner as well as the walls. This considered as a dead load and is not a consequence of the introduction of the cask pit racks.

c. Thermal Load: Thermal-hydraulic calculations presented in Section 5 of this report show that the bulk pool temperature of the water mass in the pool will remain below the existing licensed limit. However, there will be normal forces at shim locations on the cask pit walls due to thermal expansion of the platform.

d. Seismic Loads: Any of the four earthquake events described in Section 6 will, if sufficiently strong, produce lateral reaction between the platform and the cask pit cavity walls. The base mat will also experience increased transient loads.

It is concluded from the above that only an earthquake in combination with thermal and the rack dead weight would produce additional loadings on the reinforced concrete structure and that the remaining loading (hydrostatic) remains unchanged.

Further, the effect of the seismic loading on the cask pit cavity is to produce local bending and shear in contrast to the global bending and shear produced by other loads in the aboveground portion of the pool structure. Therefore, the structural assessment of the consequences of adding the cask pit rack resolves to evaluating the local moment and shear in the cask pit cavity base mat and walls and comparing them against the available corresponding capacities. Therefore, the addition of the temporary rack and its support platform in the cask pit area will not have an adverse effect on the overall structural integrity of the spent fuel pool.

8-6 1368 Holtec Rport 1-11-2043162 Report HI-2043162 8-6 1368

8.5 Strength Evaluation The cask pit cavity walls, basemat, and liner plate and anchorages are evaluated using the maximum reaction loads obtained from the cask pit rack/platform dynamic analyses, which are discussed in Section 6. Local bearing and shear stresses in the reinforced concrete walls and basemat are compared with the appropriate ACI Code stress limits. Finite element calculations are performed to determine the strain levels in the cask pit liner due to friction loads.

The Cask Pit floor and wall liners, and liner anchorage computed safety factors are provided in Table 8.5.1.

Table 8.5.1 Safety Factors for the Liner and Liner Anchorage Cask Pit Floor Cask Pit Wall Considered Load Combinations Liner Anchor Liner Anchor I and4 1.8 1.7 1.4 1.1 5 through 7 2.8 2.0 3.3 2.3 Safety factors against punching shear, diagonal edge shear, and bearing on the Cask Pit concrete wall and floor are summarized in Table 8.5.2.

t Note that the reported safety factors are the bounding values from the listed load case combinations.

Holtec Report HI-2043162 8-7 1368

Table 8.5.2 Safety Factors for the Cask Pit Concrete Wall and Floor Cask Pit Wall Cask Pit Floor Diagonal Edge Considered Load Combinations Punching Beang Punching Bearing Shear

________ Shear Bang Shear Bern I and 4 9.5 1.2 23.8 6.8 1.3 5 through 7 10.5 2.0 13.4 5.5 1.5 8.6 Conclusion It is concluded from the results presented in this report that the available structural capacity in the loaded region of the cask pit is incomparably greater than the stress resultants arising from the applied loads. Therefore, the spent fuel pool structure is adequate to withstand the additional loads imposed by the addition of the cask pit rack and associated support platform.

The auxiliary building, which includes the Spent Fuel Pool structures, has been seismically qualified using the criteria outlined in Chapter 3 of the Diablo Canyon FSAR update. As described in FSAR Section 3.7.2.1.7.1, the seismic inertia loads were obtained using time-history analyses of spring and lumped mass models (two horizontal and one vertical) of the auxiliary building. A detailed analytical static model of the auxiliary building was then used to distribute seismic inertia forces and moments to various walls, diaphragms, and columns, as described in FSAR Section 3.8.2.4. The effect of the change in weight on the seismic models due to the cask pit rack is considered to be insignificant, since the increase in global mass is determined to be less than 1.5%. Therefore, there is no change in the seismic responses and forces reported in the FSAR.

8-8 1368 Holiec Report Holtec 162 HI-20433162 Report Hl-204 8-8 1368

8.7 References

[8.2.1] Timoshenko and Goodier, Theory of Elasticity, Section §139, 3rd Edition, McGraw Hill.

[8.3.1] Diablo Canyon Power Plant Units I & 2, Design Criteria Memorandum No. T-2, Auxiliary Building, dated January 6, 2000.

[8.3.2] DCPP Adminstrative Work Procedure E-015, Rev. 0 "Application of the Long-Term Seismic Program to Diablo Canyon Power Plant Additions and Modifications".

Holtec Report HI-2043 162 8-9 1368

4- /7/f" STAINEESS S/EEL LINER PLA7F

-3,4" CARBON SEEZ BACK/NC REINFORMCED COKNCREE-F/GYIRz-i 6J./ 6AS43/f 1CAW ('4 /SOA/t7/C NTo? FULZ AlE/C/GT NEAR POOL H'ALLS NOT S110M FOR CLAR/IT

///,7Ky I 110Z 12-T IfM-Olf7 111-,?aM'162' C I 0-F/CURES l A36E IREPORT 1/1-204362 I C S I/7 a3/

9.0 RADIOLOGICAL AND SHIELDING CONSIDERATIONS This Section will address the dose exposure consequences associated with the installation of temporary cask pit racks and supporting platforms at the Diablo Canyon Power Plant (DCPP).

9.1 Fuel Handling Accident The installation and removal of the temporary cask pit racks and support platforms will not result in a change in the consequences of a fuel handling accident (FHA) previously analyzed in DCPP FSAR Update Section 15.5. The temporary cask pit rack and support platform will be installed at the same water depth as the existing SFP racks, providing the same iodine decontamination factors assumed in the FHA analysis. Furthermore, the fuel stored in the temporary cask pit rack will have at least a 10-year decay time. Thus, the radioactivity source term for the fuel that will be stored in the temporary cask pit rack is much less than the conservative assumption made in the existing DCPP FSAR Update Section 15.5 analysis (freshly discharged fuel). Hence, the existing DCPP FSAR Update Section 15.5 FHA analysis is bounding.

9.2 Solid Radwaste The only time that the number of fuel assemblies stored in either SFP will exceed the current licensing basis will be during the 14th and 15th refueling outages when the core is fully offloaded or in the unlikely event a full core offload is required outside of a refueling outage.

For each Unit, this amounts to a total of approximately 2 months. The necessity for resin replacement is determined primarily by the requirement for water clarity, and the resin is normally changed about once a year. Because of the small time frame over which the number of fuel assemblies stored in the SFP will exceed the existing licensing basis, no significant increase in the volume of resin generated solid radioactive waste from either unit is expected to result from the addition of the temporary racks.

Holtec Report HI-2043162 9-1 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Once removed from the spent fuel pools, the temporary cask pit racks may be shipped to a vendor for volume disposal, disposed of as radioactive waste, stored for temporary use, or stored until decommissioning. The disposal of the used temporary rack will result in a one-time incremental increase in solid waste. Since the temporary cask pit rack will be packaged, transported, and disposed of in a manner consistent with NRC regulations, there will be no significant radiological impact on the environment. Because ongoing volume reduction efforts have effectively minimized the amount of waste generated, the incremental increase from disposal of the temporary cask pit racks is bounded by the plant's original licensing basis described in the Final Environmental Statement and therefore, is acceptable.

9.3 Gaseous Releases The storage of additional spent fuel assemblies in the spent fuel pool (SFP) during the 14th and 15th refueling outages is not expected to significantly affect the releases of radioactive gases from the SFP. Gaseous fission products such as krypton-85 and iodine-13 1 are produced by the fuel in the reactor core during reactor operation. Tritium is routinely produced in an operating reactor core through neutron capture by deuterium present in the reactor coolant. Small amounts of krypton-85 and iodine-13 1 fission gases are released to the reactor coolant from the small number of fuel assemblies that develop leaks during reactor operation. During refueling operations, some of these gaseous fission products enter the SFP and are subsequently released into the air. There will not be an increase in the amounts of gaseous fission products released to the atmosphere as a result of the temporary increase in SFP storage capacity because the frequency of refueling and the number of freshly off-loaded spent fuel assemblies stored in the SFP, at any one time, will not increase. The dominant source of the tritium inventory in the SFP at DCPP originates from the Reactor Cooling System (RCS). Normally, DCPP recycles RCS waste and recovers the boron and the high purity water. This water contains moderate concentrations of tritium from the RCS. This water is used to replenish SFP water inventory losses due to evaporation, and is the major source of tritium in the SFP. The contribution of tritium produced from neutron capture in the boron-10 of the storage racks is insignificant to the total inventory. Therefore, the additional racks will have a negligible effect on the overall tritium level in the SFP.

Holtec Renort HI-2043162 9-2 Project

-1 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION -

Normally, the contributions from the fuel storage areas are negligible comipared to the other sources of gaseous releases and no significant increases are expected in either unit as a result of the expanded storage capacity.

9.4 Liquid Radiological Releases The release of radioactive liquids will not be affected directly as a result of the temporary rack installation. The SFP ion exchanger resins remove soluble radioactive materials from the SFP water.

When the resins are changed out, the small amount of resin sluice water is processed by the radioactive waste system before release to the environment. The frequency of resin change out may.

increase slightly during the installation and removal of the temporary rack and installation of the support platform. However, the amount of liquid effluents released to the environment as a result of the temporary rack and support platform is expected to be negligible.

9.5 Personnel Exposures Personnel exposures in the vicinity of the SFP at each unit result principally from radionuclides in the pool water and from fuel in transit underwater. The radionuclides in the water derive generally from: 1) the mixing of primary system water with the pool water and 2) the spalling of crud deposits from the spent fuel assemblies as they are moved in the storage pool during refueling operations.

Although the overall storage capacity of each pool is temporarily being increased, the movement of fuel during refueling is independent of storage capacity. Similarly, the dose rate from fuel in transit does not increase with increased storage capacity. The use of the temporary cask pit rack will require fuel transits paths in the vicinity of the southwest corner of the Unit I SFP and the northwest corner of the Unit 2 SFP, similar to transit paths that would be used during cask loading. These transit paths will not increase the dose rate beyond that already experienced at these locations during placement of fuel into racks adjacent to the Cask Pit. Similarly, the dose rate from fuel stored in the temporary cask pit racks is expected to be comparable to the dose rate from fuel stored in the existing spent fuel pool racks adjacent to the Cask Pit, because the exterior wall thickness surrounding the Cask Pit is the same as the wall thickness surrounding the pool.

Holtec Report HI-2043 162 9.3 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

An evaluation has been performed to demonstrate that although the water gap between the temporary cask pit racks and the cask pit pool wall is slightly smaller than the gap between the existing racks and the spent fuel pool wall, there will be no change in the previously analyzed dose rates at the exterior wall surface. This is attributed to a source term strength reduction due to a minimum cooling time restriction of 10 years for all assemblies stored in the temporary cask pit racks.

Operating experience has shown that there have been negligible concentrations of airborne radioactivity during normal operations, and no increases are expected as a result of the expanded storage capacities or during the temporary cask pit .rack installation or removal. However, continuous air monitors are used in the immediate vicinity of the spent fuel pools during periods of underwater activities.

No significant increase in radiation exposure to operating personnel of either unit is expected as a result of adding a temporary cask pit rack; therefore, neither the current health physics programs nor the area monitoring systems need to be modified.

9.6 Anticipated Exposure During Rack Installation All operations involved in installing a temporary cask pit rack in each unit will utilize detailed procedures prepared with full consideration of As Low As Reasonably Achievable (ALARA) principles. Similar operations have been performed in a number of facilities in the past, and experience indicates that the relatively simple task of installing these cask pit racks and support platforms in locations not previously occupied by other racks can be accomplished with minimum radiation exposure to personnel. Diving operations are not required based on a physical survey of the existing spent fuel pool configuration and the design of the temporary cask pit racks and support platforms, which allow remote installation.

Holtec Report HI-2043162 9-4 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

The combined occupational exposure for installing a rack in each unit is estimated to be a total of approximately 0.3 person-rem. This estimated dose is based on the following.

ESTIMATED MANHOURS PERSON-REM OPERATION PER UNIT EXPOSURE PER UNIT Clean and Vacuum 20 0.050 Cask Pit Area Install Support Platform 30 0.075 Install Platform Shims (16 per Unit) 16 0.040 Install Cask Pit Rack 30 0.075 Install Connector Links (8 per Unit) 20 0.050 Typical dose rates above the SFPs during normal operations are approximately 2.5 mrem or less.

The existing radiation protection programs at DCPP are adequate for the rack installation and removal operations. Radiation Work Permits (RWPs) will govern activities, and personnel monitoring equipment will be issued to each individual. As a minimum, this will include thermoluminescent dosimeters (TLDs) and self-reading dosimeters (SRDs), which are normally personnel electronic dosimeters. All work in the SFP will be directly monitored and controlled by a radiation protection technician and exposures maintained ALARA.

Holtec Report HI-2043 162 9-5 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

9.7 Rack Removal and Storage Prior to Cask Handling Operations Based on the projected storage capacity of DCPP, PG&E does not expect to remove the temporary racks until 2010 for Unit I and 2011 for Unit 2. In preparation for dry cask loading operations, the spent fuel assemblies must be removed from the temporary cask pit racks followed by removal of the cask pit racks from the SFP. The support platforms will remain in place for cask loading activities.

Thereafter, the temporary racks may be shipped to a vendor for volume reduction, disposed of as radioactive waste, stored for temporary use, or permanently stored until decommissioning.

Based on Holtec experience with rack module removal and decontamination projects, the removal and storage process will not create significant radiological waste or personnel exposure. The removal and decontamination process should not result in more than 0.2 person-rem exposure based on a pool surface dose rate of 2.5 mrem/hr, an estimated rack contact surface dose rate of 20 mrem/hr, and ajob estimate of 80 man-hours. Typically, the surfaces of a rack module as well as the cell interiors can be decontaminated to lower the dose and contamination levels of the rack. During all handling operations, appropriate radiological controls will be in place for all rack handling and storage activities.

During the installation and removal of the racks, exposures will be maintained ALARA consistent with the requirements of IOCFR20 and DCPP's existing ALARA program. The radiation protection section will prepare one or more radiation work permits (RWPs) for various in-pool and out-of-pool activities. The RWPs and supporting documentation will establish requirements in the following areas:

• Frequency of radiation, contamination, and airborne surveys.

• Individual monitoring devices, typically a TLD and SRD.

• Protective clothing

• Access and work controls

• Contamination controls including controls for radioactive materials that could involve significant shallow dose equivalent and effective dose equivalent exposures.

Holtec Report HI-2043162 9-6 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

• Contamination and radiation limits at various steps requiring the evaluation of further controls or activities. (e.g., results of monitoring the rack as it breaks the water may "trigger" additional underwater efforts including pressure washing.

Radiation protection technicians will be involved in the movement, decontamination, packaging, and storage of the temporary racks. Constant coverage by a radiation protection technician will be required for all activities involving work in the SFP including the removal of items from the spent fuel pools. In addition to work activities in the SFP, additional periods of continuous coverage by the technicians will be as specified in the RNWPs covering the work activities. During SFP activities and periods of work with high levels of contamination, a portable continuous air monitor with alarming capability will be used to provide warning in the event of increasing airborne radioactivity in the immediate work areas. Respiratory protection may be used if shown to be TEDE ALARA consistent with the requirements of IOCFR20 and plant procedures.

The DCPP radiation protection section will perform the following actions prior to and during rack removal:

• Vacuuming of the Cask Pit area prior to rack installation.

• Underwater survey of accessible areas of the rack.

• Underwater pressure wash or rinse as appropriate.

• Monitoring and rinsing of the rack as it breaks the water surface as appropriate.

• Survey of the rack prior to wrapping in plastic or bagging.

Once wrapped or bagged, the rack will either be stored on site protected from the elements or placed into an approved shipping container for shipment to a vendor for volume reduction and/or disposal.

All items or tools used in areas posted as contaminated will either be surveyed and released if free of contamination or treated as radioactive material and handled in accordance with DCPP procedures.

Holtec Report HI-2043162 9-7 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Each member of the project team will receive radiation protection training consistent with the requirements of IOCFRI9. Project specific information such as the potential for extremity doses when removing and decontaminating items from the spent fuel pool and operating experience with spent fuel pool activities will be discussed in pre-job briefings. Radiation protection technicians, involved in the project, will participate in pre-job briefings involving their activities associated with the work.

Holtec Report H1-2043162 9-8 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

10.0 INSTALLATION 10.1 Introduction The installation phase of the Diablo Canyon cask pit rack project will be executed by Holtec International's Field Services Division. Holtec, serving as the installer, is also responsible for performance of specialized services. All installation work at Diablo Canyon is performed in compliance with NUREG-0612 (refer to Section 3.0), Holtec Quality Assurance Procedure 19.2, Diablo Canyon project specific procedures, and applicable Diablo Canyon 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 Diablo Canyon personnel, who are trained in accordance with Diablo Canyon procedures.

Rack lifting rigs are required for installation of the cask pit platform and the rack. These two lifting rigs are almost identical and are designed to remotely engage and disengage on lift points at the bottom of each component. The lifting rigs 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 implemented as part of the installation of the cask pit rack and platform. A set of inspection points, which have been proven to be effective in producing high quality work in previous rack projects, is implemented by the installer.

Due to the design of the cask pit rack and associated platform, underwater diving operations are not required for this project.

Holtec Report HI-2043162 10-1 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Holtec International developed procedures, to be used in conjunction with the Diablo Canyon procedures, which cover the scope of activities for the installation effort. Similar procedures have been utilized and successfully implemented by Holtec on previous rack installation 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 Diablo Canyon administrative procedures prior to use on site. The following is a list of the Holtec procedures, used in addition to the Diablo Canyon procedures to implement the installation phase of the project.

A. Installation/Handling Procedure:

This procedure provides direction for the handling/installation of the rack platforms and new storage rack modules in the Cask Pit. This procedure delineates the steps necessary to receive the rack on site, the proper method for unloading and uprighting the rack, staging the rack prior to installation, and installation of the platforms and racks. The cask pit platform has a cross section that closely matches the rack but a shorter height and lower weight. The process for placement of the cask pit platforms is similar to that for the racks. The procedure provides for the installation of cask pit platform, positioning and leveling of the cask pit platform, if necessary to compensate for any Cask Pit floor uneveness, and installation of the wedges. This procedure also provides for the installation of the rack onto the platform, including the leveling of the rack, installation of the connector links between the rack and the platform, 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 the rack modules and cask pit platforms after their arrival on site. The receipt inspection includes dimensional measurements, cleanliness inspection, visual weld examination, and verticality measurements.

Holtec Report HI-2043162 10-2 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

C. Cleaning Procedure:

This procedure provides for the cleaning of rack module and cask pit platform, if required. The components are to meet the requirements of ANSI N45.2.1, Level B, prior to placement in the Cask Pit. Methods and limitations on cleaning materials to be utilized are provided.

D. Pre- and Post-Installation Drag Test Procedure:

These tvo procedures stipulate the requirements for performing a functional test on the rack module prior to and following installation. The procedures provide direction for inserting and withdrawing an insertion gage or dummy fuel assembly into designated cell locations, and establishes an acceptance criteria 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 Diablo Canyon 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.

Holtec Report HI-2043162 10-3 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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 limiting conditions. The weld procedure is qualified in accordance with ASME Section XI , or may be qualified to an alternate code accepted by Pacific Gas &

Electric and Holtec International.

10.2 Rack Arrangement The rack installation process will not directly require any fuel shuffling. During refueling outages prior to the rack installation, final placements of fuel assemblies within the spent fuel pool will maximize the open storage locations adjacent to the Cask Pit. The final rack arrangement allows for a 12 by 13 cell Region II style rack to be installed in both the Unit I and Unit 2 Cask Pit Areas. Schematic plan views depicting Cask Pit Area storage rack configurations are shown in Figures 1.0.1 and 1.0.2.

10.3 Rack Interferences A survey was conducted to identify any objects which would interfere with cask pit rack and platform installation or prevent usage of any storage locations. Prior to installation of the platform, PG&E will remove two lighting brackets from each of the spent fuel pools in the vicinity of the Cask Pits and will perform a minor modification to the comer bracket of the Cask Restraint to provide clearance for the rack/platform installation. Any existing miscellaneous equipment that is temporarily stored within the Cask Pit will be removed followed by vacuuming prior to installation of the cask pit rack and platform.

Holtec Report HI-2043162 10-4 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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 activities, such as platform placement, may require the temporary shutdown of the Spent Fuel Pool cooling system.

Prior to any shutdown of the Spent Fuel Pool cooling system, the estimated time after shutdown to increase the pool bulk coolant temperature to a selected value of <120 IF will be determined.

A temperature of < 120 IF is chosen with enough margin such that cooling may be restored to ensure the pool bulk temperature will not exceed 150 IF.

10.5 Installation of Cask Pit Rack Installation of the cask pit rack, supplied by Holtec International, involves the following activities. The rack is delivered in the horizontal position. The 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 rack is up-righted into a vertical position.

The cask pit rack lifting rig 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.lB & D). The cask pit rack is lifted with the 125-ton Fuel Handling Building (FHB) Crane and transported along the pre-established safe load paths. The cask pit platform is handled in a similar fashion and staged for installation into the Cask Pit.

The Cask Pit Area floor is inspected and any debris, which may inhibit the installation of the cask pit platform, is removed. The floor elevations in the Cask Pit have been carefully measured and leveling shims are installed, if necessary, at the cask pit platform load bearing points.

Holtec Report HI-2043162 10-5 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

The cask pit platform is lifted via the lifting rig and lowered by the FHB Crane into position slightly above the Cask Pit floor. Following a final check of positioning, the cask pit platform is lowered into position. The platform is equipped with adjustable removable wedge shims which secure it in the Cask Pit cavity. The wedge shims are installed using long handled tools from above the SFP surface. Elevation readings are taken to confirm that the platform is level and to pre-set the extensions on the rack pedestals prior to rack installation. In addition, platform-to-wall offset distances are also measured. Adjustments are made as necessary to ensure compliance with design documents. The lifting rig is disconnected from the cask pit platform and removed.

The rack module is carefully lowered into the Cask Pit Area and positioned slightly above the cask pit platform. After final positioning, the rack is lowered onto the cask pit platform. The rack pedestals are then adjusted using the remote adjustment tool, if necessary, to ensure that the final position is level and upright.

A set of connector links are installed in the rack using long-handled tools to secure the rack to the platform. The connector links are an "I" shaped key (see Figure 2.1.5), which when rotated 90 degrees, engages the bottom against small tabs attached to the platform. Continued turning allows the threaded end on the connector link to tighten against the stops in the platform.

The lifting rig is then disengaged and removed from the Cask Pit Area under Health Physics direction. As directed by procedure, post-installation free path verification is performed on all rack cells using a dummy fuel assembly.

10.6 Safety, Health Physics, and ALARA Methods 10.6.1 Safety During the installation phase of the cask pit rack project, personnel safety is of paramount importance. All work shall be carried out in compliance with applicable approved procedures.

Holtec Report HI-2043162 10-6 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

10.6.2 Health Physics Health Physics is carried out per the requirements of the Diablo Canyon Radiation Protection Program.

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, support leveling device, connector link installation tool, and a lift rod disengagement device have been developed to execute numerous activities from the SFP surface, where dose rates are relatively low.

Holtcc Report HI-2043162 10-7 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Shielding During the course of the cask pit rack project, primary shielding is provided by the water in the Spent Fuel 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.

10.7 Radwaste Material Control Radioactive waste generated from the rack installation will be controlled in accordance with established Diablo Canyon procedures.

Holtec Report HI-2043 162 10-8 Project 1368 SHADED AREAS DENOTE PROPRIETARY INFORMATION

11.0 ENVIRONMENTAL AND ECONOMIC CONSIDERATIONS Article V of the USNRC OT Position Paper [11.1] specifies environmental and economic considerations as essential predicates for a fuel storage capacity enhancement. This section provides justification for selecting installation of an additional rack in each of the DCPP cask pits as fulfilling the directive of the OT Position Paper.

11.1 Need for Temporary Spent Fuel Storage Capacity PG&E is currently undertaking a project to construct an Independent Spent Fuel Storage Installation (ISFSI) at DCPP. It is planned for the ISFSI facility to be in operation in early 2007 to maintain a Full Core Offload Capability (FCOC) for both Units at DCPP. However, to avoid the potential for the loss of FCOC due to any ISFSI Project delays, PG&E intends to temporarily expand the spent fuel storage capacity by adding an additional fuel storage rack within the Cask Pit of each Unit. Based on the current inventory of fuel assemblies stored in the spent fuel pools (SFPs) and the anticipated future discharges of spent fuel, Unit I will lose full core reserve capacity in 2007 and Unit 2 will lose full core reserve capacity in 2008.

The proposed temporary cask pit rack and support platform would be installed in Units I and 2 for Cycles 14 through 16. The temporary cask pit rack would allow the temporary storage of an additional 154 spent fuel assemblies per unit. The total spent fuel pool storage capacity per unit would be increased to 1478 spent fuel assemblies for Cycles 14 through 16.

11.2 Environmental Considerations The proposed temporary increase in SFP storage capacity described and analyzed in this report, will not result in a significant impact to the environment as described below.

Holtc Rpor HI-043162Il-IProect136 Holtec Report HI-2043162 11-1 Project 1368

i. Thermal Impact Thermal effects on the environment due to adding a temporary cask pit rack to each unit will be negligible. Any additional heat load attributed to the installation of a cask pit rack only occurs during relatively short periods of time during an outage when the reactor core is offloaded. During non-outage periods, the maximum number of spent fuel assemblies stored in the SFP will not exceed the existing licensing basis, resulting in no additional heat load imposed on the environment.

The most significant contributor to the SFP decay heat load is from the number of assemblies discharged during a reactor offload to the SFP. Since the size of planned refueling discharges are unchanged, there will be no significant increase in SFP heat load.

A small temporary heat load will be imposed on the SFP cooling system from the oldest spent fuel that is allowed to remain in the SFP longer due to the additional storage capacity provided by the temporary cask pit rack. However, this additional decay heat load will be a small fraction of the existing SFP heat load and insignificant when compared to the total heat rejected to the environment by the plant.

ii. Consumption of Raw Materials It is estimated conservatively that the manufacturing of the two cask pit racks and support platforms will require the following raw material resources:

Stainless Steel: 80,000 lbs Aluminum: 5,000 lbs Boron Carbide: 4,000 lbs Stainless Steel Weld Wire and Flux: 500 lbs Argon Gas (for welding): 1,000 SCF The annual worldwide consumption of each of the above materials is at least 100,000 times as much as the quantities needed for the cask pit rack project implementation.

Therefore, it is readily concluded that this project will have a negligible effect on the world's production and availability of the above-mentioned raw materials.

162 11-2 Project 1368 Report HI-2043 Holtec Report Holtec HI-2043162 11-2 Project 1368

iii. Generation of Radioactive Waste The radioactivity and impurities in the SFP water are not expected to increase as a result of the addition of the temporary cask pit storage rack and support platform. Replacement of filters or demineralizers would offset any anticipated increase of the radioactivity and impurity level of the water. The temporary increase in spent fuel storage capacity during fueling outages is not expected to result in a significant change in long-term generation of solid radioactive waste.

Prior to installing the cask pit racks and support platforms, equipment temporarily stored in the cask pit will be removed and the pit floor will be cleaned using an underwater vacuum. These activities are expected to generate a small volume of low-level solid radwaste that will be captured underwater in vacuum filter cartridges and properly disposed of using DCPP radwaste handling procedures.

Storing and removing fuel assemblies from the cask pit racks is not expected to generate any additional solid or gaseous radwaste from either unit compared to the current practice of storing fuel assemblies in the SFP. Because each cask pit rack must eventually be removed for cask handling operations, rack contamination will be minimized by a fuel loading process that will preferentially select non-failed fuel for storage in this rack.

The disposal of the used temporary rack will result in a one-time incremental increase in solid waste. Because ongoing volume reduction efforts have effectively minimized the amount of waste generated, this incremental increase is bounded by the plant's original licensing basis described in the Final Environmental Statement and therefore, is acceptable.

iv. Gaseous Emissions and Liquid Discharges As discussed in Section 9, the temporary storage of additional spent fuel assemblies in the SFP during Cycles 14-16 is not expected to significantly affect the releases of radioactive gases from the SFP.

Holtec Report HI-2043162 11-3 Project 1368

The release of radioactive liquids will not be affected as a result of the temporary rack installation. The SFP ion exchanger resins remove soluble radioactive materials from the SFP water. When the resins are changed out, the small amount of resin sluice water is processed by the radioactive waste system before release to the environment. The frequency of resin change out may increase slightly during the installation and removal of the temporary rack and installation of the support platform. However, the amount of liquid effluents released to the environment as a result of the temporary rack and platform is expected to be negligible.

v. Hazardous Human Activity Manufacturing and installation of fuel racks in fuel pools is a routine and commonplace human activity. Thousands of racks have been manufactured and installed world-wide.

As discussed in Section 9, there is no significant increase in occupational radiation exposure. There is no scientific evidence (or assertion by any group) that the manufacture or use of fuel racks entails any human risk factors.

vi. Transport of Hazardous Materials The temporary use of the cask pit racks does not involve off-site movement of spent nuclear fuel or any other radiologically significant cargo. It is therefore concluded that the proposed temporary rack installation is environmentally benign in all respects.

11.3 Project Alternatives Alternatives to the proposed action will be discussed in detail in PG&E's license amendment request. As previously discussed, the installation of the temporary rack is to provide interim storage during Cycles 14 through 16 while outstanding issues are being resolved to provide dry cask storage for long-term spent fuel storage.

162 11-4 Project 1368 Report HI-2043 Holtec Report HI-2043162 11-4 Project 1368

11.4 Design Basis Events This Licensing Report has addressed both the safety and environmental aspects of a fuel handling accident, an event which bounds the potential adverse consequences of accidents attributable to operation of a spent fuel pool with high-density racks. A fuel handling accident is a design basis event that the pool and its associated structures, systems and components are designed and constructed to prevent and mitigate. The environmental impacts of this accident were found to be bounded by the previous licensing basis in the DCPP FSAR Update Section 15.5.

11.5 Summary The installation of the temporary cask pit rack and support platform, temporary storage of spent fuel in the rack, and the removal of the rack will not significantly increase the probability or consequences of accidents. There are no significant changes in the types or significant increase in the quantities of effluents that may be released offsite and there is no significant increase in occupational or public radiation exposure.

11.6 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", by K.P. Singh, Power Generation Technology, Sterling Publishers, pp. 137-140, U.K. (November 1990).

[11.4] USNRC, "Environmental Assessment and Finding of No Significant Impact Related to the Construction and Operation of the Diablo Canyon Independent Spent Fuel Storage Installation", October 24, 2003.

162 11-5 Project 1368 Report HI-2043 Holtec Report Holtec HI-2043162 11-5 Project 1368