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Holtec Report No: HI-2084175, Licensing Report for Beaver Valley, Unit 2, Rerack.
	ML102940458
Person / Time
Site:	Beaver Valley
Issue date:	10/18/2010
From:	; Holtec
To:	; Office of Nuclear Reactor Regulation
References
	Holtec Project 1702, L-10-275, TAC ME1079 HI-2084175
	Download: ML102940458 (176)
	v • d • e

Enclosure D to FENOC Letter L-10-275 Licensing Report for Beaver Valley Unit 2 Rerack (Nonproprietary) 175 pages not including this cover sheet

mmmmm Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797- 0900 HOLTEC INTERNATIONAL Fax (856) 797 - 0909 I

LICENSING REPORT FOR BEAVER VALLEY UNIT 2 RERA CK FOR FIRST ENERGY Holtec Report No: HI-2084175 Holtec Project No: 1702 Sponsoring Holtec Division: NPD Report Class : SAFETY RELATED COMPANY PRIVATE This document is the propertY of Holtec International and its Client. It is to be used only in connectionwith the performance of work by Holtec International or its designated subcontractors.

Reproduction, publication or representation, in whole or in part, for any other purpose by any party other than the Client is expressly forbidden.

HOLTEC INTERNATIONAL DOCUMENT NUMBER: 2084175 PROJECT NUMBER: 1702 DOCUMENT ISSUANCE AND REVISION STATUS DOCUMENT NAME: _Licensing Report for Beaver Valley Unit 2 Rerack DOCUMENT CATEGORY: EZ GENERIC 0 PROJECT SPECIFIC REVISION No. 4 REVISION No. 5 REVISION No. 6 Author's Date Author's Date Author's Date No. Doct Initials Approved Initials Approved Initials Approved Portiontt I Chap 1 - - MTM 10/4/10 782711

2. Chap 2 CWB 08/25/10 573191 - -

3. Chap 3 - - -

4. Chap 4 BDB 06/10/10 559560 BDB 9/3/10 137707 - -

5. Chapter5 N/A N/A

.Chap 5 -- CWB 08/25/10 648728 Deleted

6. Chap 6 ER 08/20/10 129861 -

Chapter7 N/NA

7. Chap 7 CWB 08/25/10 809289 Deleted N/A N/A

8. Chap8 -8

9. Chap 9 10.

11.

12.

"" Chapter or section number.

Page 1 of 2

HOLTEC INTERNATIONAL DOCUMENT NUMBER: 2084175 PROJECT NUMBER: 1702 DOCUMENT CATEGORIZATION In accordance with the Holtec Quality Assurance Manual and associated Holtec Quality Procedures (HQPs), this document is categorized as a:

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

[- Design Criterion Document (Per HQP 3.4) E] Design Specification (Per HQP 3.4)

L-] 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 Z Nonproprietary lZ Holtec Proprietary EL 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 0 - Original Revision.

Revision 1 - Incorporated Client comments in Chapters 4 and 6.

Revision 2 - Incorporated additional client comments throughout.

Revision 3 - Incorporated additional client comments in Chapters 1, 4, 5, 6, 8 and 9.

Revision 4 - Revised Chapter 4 including tables and figures based on revisions to analyses.

Revision 5 - Updated Chapters 2, 4, 5, 6, and 7 based on RAI responses and Client comments Revision 6 - Chapters 5 and 7 have been removed. They may be added back into this report in next revision.

Holtec Report HI-2084175 i Holtec Project 1702

TABLE OF CONTENTS Section Description Page 1.0 INTROD U CTION ..................................................................................................... 1-1 1.1 References .................................................................................................................. 1-4 2.0 FUEL STORA GE RA CK S ........................................................................................ 2-1 2.1 Introduction ................................................................................................................ 2-1 2.2 Summary of Principal Design Criteria ............................... 2-2 2.3 Applicable Codes and Standards ................................... 2-3 2.4 Quality A ssurance Program ....................................................................................... 2-9 2.5 M echanical Design ................................................................................................. 2-10 2.6 Rack Fabrication ...................................................................................................... 2-10 3.0 MATERIAL CONSIDERATIONS ................................. 3-1 3.1 Introduction ................................................................................................................ 3-1 3.2 Structural M aterials .... ............................................................................................. 3-1 3.3 N eutron Absorbing M aterial ...................................................................................... 3-1 3.4 In-Service Surveillance of the N eutron Absorber ...................................................... 3-4 3.5 References .................................................................................................................. 3-7 4.0 CRITICA LITY SA FETY AN ALY SIS ...................................................................... 4-1 4.1 Introduction and Summ ary ........................................................................................ 4-1 4.2 M ethodology .............................................................................................................. 4-3 4.3 Acceptance Criteria ..................................................................................................... 4-7 4.4 A ssum ptions ............................................................................................................... 4-9 4.5 Input Data ............................................................................................................... 4-12 4.6 Com puter Codes ....................................................................................................... 4-16 4.7 Analysis ................................................................................................................... 4-17 4.8 References ............. I.......................................................................................... 4-33 5.0 Chapter 5 - To Be Provided Later.

Holtec Renort HI-2084175

...... f Holtec Project 1702 J

TABLE OF CONTENTS (continued)

Section Description Page 6.0 THERM AL-HYDRAULIC EVALUATION ............................................................ 6-1 6.1 Introduction ................................................................ I.............................................. 6-1 6.2 Acceptance Criteria .................................................................. ................................. 6-3 6.3 A ssum ptions and Design Data ................................................................................... 6-3 6.4 Fuel Rod Cladding Tem peratures .............................................................................. 6-6 6.5 Description of Spent Fuel Pool Cooling System ....................................................... 6-8 6.6 Heat Loads and Bulk Pool Temperatures .................................................................. 6-9 6.7 Local W ater and Fuel Cladding Temperatures ........................................................ 6-12 6.8 References ................................................................................................................ 6-14 7.0 Chapter 7 - To Be Provided Later 8.0 RADIOLOGICAL EVALUATION ........................................................................... 8-1 8.1 Introduction ................................................................................................................ 8-1 8.2 Acceptance Criteria ................................................................................................... 8-1 8.3 A ssumptions ............................................................................................................... 8-1 8.4 Dose Rate at the Surface of the Pool ......................................................................... 8-2 8.5 Fuel Handling Accident .............................................................................................. 8-3 8.6 Person-Rem Exposure ................................................................................................ 8-3 8.7 Conclusion ................................................................................................................. 8-3 9.0 IN STALLATION ...................................................................................................... 9-1 9.1 Introduction ................................................................................................................ 9-1 9.2 Rack Arrangement .............................................................. 9-4 9.3 Installation Sequence ................................................................................................. 9-4 Holtec Report HI-2084175 I

iii Holtec Project 1702 J

LIST OF TABLES Table Description Page 2.1.1 GEOMETRIC AND PHYSICAL DATA FOR THE NEW FUEL RACKS ........... 2-13 2.5.1 MODULE DATA FOR NON-FLUX TRAP FUEL RACKS .................................. 2-14 3.4.1 SAMPLE COUPON MEASUREMENT SCHEDULE ............................................. 3-9 4.5.1 FUEL ASSEMBLY SPECIFICATION AND DESIGN BASIS FUEL A SSEM B LY SELECTION ..................................................................................... 4-34 4.5.2 SPECIFICATION OF THE DESIGN BASIS FUEL ASSEMBLY ........................ 4-35 4.5.3 CORE OPERATING PARAMETER FOR CASMO DEPLETION ANALYSES. 4-36 4.5.4 CALCULATION OF SOLUBLE BORON FOR CASMO-4 .................................. 4-37 4 .5.5 DE L E T E D ................................................................................................................ 4-38 4.5.6 AX IAL BU RNU P PROFILES ................................................................................ 4-39 4.5.7 IFBA RO D SPECIFICA TION ............................................................................... 4-40 4.5.8 WATER DISPLACER ROD SPECIFICATION .................................................... 4-41 4.5.9 W ABA SPECIFICA TION ....................................................................................... 4-42 4.5.10 STORAGE CELL SPECIFICATION ...................................................................... 4-43 4.5.11 SPECIFICATION OF THE FUEL ROD STORAGE BASKET ............................. 4-44 4.5.12 SPECIFICATION FOR THE BORON DILUTION ANALYSIS ........................... 4-45 4.5.13 SPECIFICATION OF THE FUEL ELEVATOR ................................................... 4-46 4.7.1

SUMMARY

OF THE REGION 2 INITIAL ENRICHMENT AND BURNUP COM BINATIONS ........................................................................ 4-47 4.7.2

SUMMARY

OF THE REGION 3 INITIAL ENRICHMENT AND BURNUP COM BINATIONS ........................................................................ 4-48 4.7.3

SUMMARY

OF THE MTZR SOLUBLE BORON REQUIREMENTS FOR N ORM A L CON DITION S ............................................................................... 4-49 4.7.4 SUM M ARY OF ACCIDENT CASES .................................................................... 4-50 4.7.5 CASMO-4 CALCULATION OF THE EFFECT OF SPACER GRIDS AND BORON CONCENTRATION ON REACTIVITY ................................................. 4-51 4.7.6a CALCULATION OF THE INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR REGION 2, ENRICHED BLANKETS PROFILE .......... 4-52 4.7.6b CALCULATION OF THE INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR REGION 2, NATURAL BLANKETS PROFILE ........... 4-53 4.7.6c CALCULATION OF THE INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR REGION 2, NO BLANKETS PROFILE ........................ 4-54 4.7.7a CALCULATION OF THE INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR REGION 3, ENRICHED BLANKETS PROFILE .......... 4-55 4.7.7b CALCULATION OF THE INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR REGION 3, NATURAL BLANKETS PROFILE ........... 4-56 4.7.7c CALCULATION OF THE INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR REGION 3, NO BLANKETS PROFILE ........................ 4-57 4.7.8 RESULTS OF THE CASMO-4 CALCULATIONS OF THE IFBA BIAS ............ 4-58 Holtec Report HI-2084175 iv Holtec Project 1702

LIST OF TABLES (continued)

Table Description Page 4.7.9 RESULTS OF THE CASMO-4 CALCULATIONS OF WDR REACTIVITY DEPENDENCE ON NUMBER OF WDR RODS AND FU EL EN R IC H MEN T ....................................................................................... ..... 4-59 4.7.10 RESULTS OF THE CASMO-4 CALCULATIONS OF A FUEL ASSEMBLY WITH BOTH IFBA AND WDR BIAS DEPENDENCE ON THE NUMBER OF WDR RODS AND FUEL ENRICHMENT ....................................................... 4-60 4.7.11 RESULTS OF THE CASMO-4 CALCULATIONS OF WABA BIAS DEPENDENCE ON NUMBER OF WABA RODS AND FU EL EN RICHM EN T ............................................................................................ 4-61 4.7.12 RESULTS FOR THE CALCULATION OF THE ECCENTRIC FUEL POSITIONING REACTIVITY EFFECT ..................................................... 4-62 4.7.13a CASMO CALCULATIONS FOR FUEL ASSEMBLY MANUFACTURING TOLERANCE UNCERTAINTIES FOR FUEL STORAGE C E L L (N O C RE E P) ................................................................................................. 4-63 4.7.13b CASMO CALCULATIONS FOR FUEL ASSEMBLY MANUFACTURING TOLERANCE UNCERTAINTIES FOR FUEL STORAGE CE L L (C RE E P ) ....................................................................................................... 4-64 4.7.14 CASMO CALCULATIONS FOR MANUFACTURING TOLERANCE UNCERTAINTIES FOR FUEL STORAGE CELL ................................................ 4-65 4.7.15a CASMO CALCULATIONS FOR POOL TEMPERATURE TOLERANCE UNCERTAINTIES ......................................................................... 4-66 4.7.15b CALCULATION OF THE TEMPERATURE BIAS CASMO-4 ............................ 4-67 4.7.16 VERIFICATION OF THE INITIAL ENRICHMENT AND BURNUP COMBINATIONS AND CALCULATIONS OF SOLUBLE BO RON REQ U IREM EN TS .................................................................................... 4-68 4.7.17 CALCULATION OF THE MISLOADED FRESH FUEL ASSEMBLY A C C ID EN T ............................................................................................................. 4-70 4.7.18 CALCULATION OF THE MISLOCATED FRESH FUEL ASSEMBLY A C C IDE N T ............................................................................................................. 4-7 1 4 .7 .19 DE L E TE D ................................................................................................................ 4-72 4.7.20 MCNP4a CALCULATIONS FOR THE FUEL ROD STORAGE BASKET ......... 4-73 4.7.21 BEAVER VALLEY POWER STATION UNIT NO. 2 SPENT FUEL POOL BORON DILUTION ACCIDENT ANALYSIS .......................................... 4-74 4.7.22 CASMO CALCULATIONS FOR METAMIC MEASUREMENT UNCERTAINTY .................................................. 4-75 4.7.23 CALCULATION OF THE RACK DISPLACEMENT ACCIDENT ...................... 4-76

.4.7.24 CALCULATIONS FOR THE FUEL ELEVATOR CASE ..................................... 4-77 Holtec Report HI-2084175 V Holtec Project 1702

LIST OF TABLES (continued)

Table Description Page 6.1.1 PARTIAL LISTING OF RERACK APPLICATIONS USING SIMILAR METHODS OF THERMAL-HYDRAULIC ANALYSIS ...................................... 6-15 6.3.1

SUMMARY

OF INPUTS FOR BULK TEMPERATURE ANALYSIS ................ 6-17 6.3.2

SUMMARY

OF INPUTS FOR LOCAL TEMPERATURE ANALYSIS .............. 6-18 6.6.1

SUMMARY

OF BULK POOL TEMPERATURE RESULTS ............................... 6-19 6.6.2

SUMMARY

OF TIME-TO-BOIL RESULTS ........................................................ 6-20 6.7.1

SUMMARY

OF LOCAL TEMPERATURE RESULTS ........................................ 6-21 8.3.1 PARAMETER VALUES UTILIZED IN THE RADIOLOGICAL E V A L U A TIO N S ....................................................................................................... 8-4 8.6.1 PERSON-REM EXPOSURE FROM RE-REACKING ONE PLANT ..................... 8-4 Holtec Report HI-2084175 vi Holtec Project 1702

LIST OF FIGURES Figure Description Page 1.1 PROPOSED NEW RACK LAYOUT FOR BEAVER VALLEY POW ER STATION (BVPS) UNIT No. 2 ............................................................ 1-5 2.1.1 ISOMETRIC VIEW OF GENERIC NON-FLUX TRAP RACK ......................... 2-15 2.6.1 ISOMETRIC VIEW OF COMPOSITE BOX ASSEMBLY ................................... 2-16 2.6.2 PLAN VIEW OF GENERIC NON-FLUX TRAP RACK ARRAY ........................ 2-17 2.6.3 ADJUSTABLE PEDESTAL DESIGN .............................. 2-18 2.6.4 NON-FLUX-TRAP RACK CELLS ELEVATION VIEW ..................................... 2-19 4.5.1 128 IFBA PIN LOADING PATTERN FOR 17x 17 FUEL ASSEMBLIES ............ 4-78 4.5.2 A TWO DIMENSIONAL REPRESENTATION OF THE MZTR MCNP4a M OD E L ................................................................................... ........... 4-79 4.5.3 MZTR 9 X 12 RACK PERMISSIBLE LOADING CONFIGURATION ............. 4-80 4.5.4a MZTR 9 X 13 RACK PERMISSIBLE LOADING CONFIGURATION ............. 4-81 4.5.4b EXAMPLE OF A MZTR 9 X 13 RACK WITH A CUT-OUT SECTION PERMISSIBLE LOADING CONFIGURATION ................................................... 4-82 4.5.5 200 IFBA PIN LOADING PATTERN FOR 17x17 FUEL ASSEMBLY ............... 4-83 4.5.6 A TWO DIMENSIONAL REPRESENTATION OF THE MCNP4a F R SB M O D E L ......................................................................................................... 4-84 4.5.7 SPENT FUEL POOL MIXED ZONE THREE REGION LAYOUT ...................... 4-85 4.5.8 A TWO DIMENSIONAL REPRESENTATION OF THE REGION 3 M CNP4a SINGLE CELL M ODEL ......................................................................... 4-86 4.7.1 a REGION 2 INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR ENRICHED BLANKETS PROFILE ............................................................ 4-87 4.7.1 b REGION 2 INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR NATURAL BLANKETS PROFILE ............................................................... 4-88 4.7. l c REGION 2 INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR N O BLAN KETS PROFILE ............................................................................ 4-89 4.7.2a REGION 3 INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR ENRICHED BLANKETS PROFILE ............................................................ 4-90 4.7.2b REGION 3 INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR NATURAL BLANKETS PROFILE ............................................................... 4-91 4.7.2c REGION 3 INITIAL ENRICHMENT AND BURNUP COMBINATIONS FOR N O BLANKETS PROFILE ............................................................................ 4-92 4.7.3 A TWO DIMENSIONAL REPRESENTATION OF THE MCNP4a MISLOCATED FUEL ASSEMBLY ACCIDENT MODEL ................................ 4-93 4.7.4 REPRESENTATION OF MISLOADING ACCIDENT MCNP MODEL WITH REFLECTIVE BOUNDARY CONDITIONS ............................................. 4-94 Holtec Report HI-2084175 vii Holtec Project 1702

LIST OF FIGURES (continued) 6.6.1 SIMPLIFIED HEAT EXCHANGER ALIGNMENT ............................................ 6-22 6.6.2 FSP HEAT LOAD AND FSP BULK TEMPERATURE PROFILES -

NORMAL FULL CORE OFFLOAD - 100 HOUR REFUELING START TIME ...................................................... ......... 6-23 6.6.3 FSP HEAT LOAD AND FSP BULK TEMPERATURE PROFILES -

NORMAL FULL CORE OFFLOAD - 125 HOUR REFUELING STA R T T IME ........................................................................................................ 6-24 6.6.4 FSP HEAT LOAD AND FSP BULK TEMPERATURE PROFILES -

NORMAL FULL CORE OFFLOAD - 150 HOUR REFUELING STA R T T IM E .......................................................................................................... 6-25 6.6.5 FSP HEAT LOAD AND FSP BULK TEMPERATURE PROFILES -

ABNORMAL FULL CORE - 100 HOUR REFUELING ST A R T T IM E ........................................................................................................ 6-26 6.6.6 MAXIMUM ALLOWABLE COMPONENT COOLING WATER INLET TEMPERATURE VS. REFUELING START TIME - NORMAL ............ 6-27 6.7.1 CONTOURS OF STATIC TEMPERATURE IN A VERTICAL PLANE THROUGH THE CENER OF THE FSP ............................................................... 6-28 Holtec Report HI-2084175 viii Holtec Project 1702

1.0 INTRODUCTION

The Beaver Valley Power Station (BVPS) Unit No. 2 is a pressurized water reactor (PWR),

designed by Westinghouse, with a reactor core sized for 157 Westinghouse 17x17 fuel assemblies. The station is owned and operated by FirstEnergy Nuclear Operating Company (FENOC). The station's reactor has a thermal output of 2900 MWt (megawatts thermal).

FENOC has determined that continued operation of the station will require additional storage space to accommodate the accumulating spent nuclear fuel. This report provides a description of a proposed modification to the Spent Fuel Pool (SFP) rack configuration and a summary of the evaluations performed to support this change.

The station's SFP currently contains 1088 storage cells in seventeen spent fuel storage racks. All existing racks are of the flux trap' style. The seventeen existing racks in the SFP will be removed and replaced by fifteen new freestanding racks of the following sizes:

Rack Array Size Number of Racks Usable Storage Locations (and Rack IDs) 9by8 1 (A3) 1 x72=72 9 by 12 1 (Al) l x96=96 9 by 12 5 (A2,B2, C2, B3 & C3) 5 x 108 = 540 9by 13 1 (B1) 1 xll =111I 9by 13 1 (C1) 1 x 117= 117 9 by 14 2 (B4 & C4) 2 x 126 = 252 10 by 12 2 (D2 & D3) 2 x 120 = 240 10 by 13 1 (D1) 1 x 122 = 122 10 by 14 1 (D4) 1 x 140 = 140 1690 All of the new racks are non-flux-trap racks and are designated in a mixed-zone three-region (MZTR) array, where loading patterns are used to control criticality. The resulting fuel storage rack array proposed for the station contains 1690 storage cells and is shown in the plan view 1 Flux trap style racks contain a water-filled gap with neutron absorber on both sides, called a flux trap, between adjacent fuel storage locations.

Holtec Report HI-2084175 1-1 Holtec Project 1702

provided in Figure 1.1. The total fuel assembly storage capacity of the racks, however, depends on the storage patterns used for low-burned fuel. Figures 4.5.3 and 4.5.4 show some allowable loading patterns.

All SFP storage racks are freestanding and self-supporting. The principal construction materials for the racks are SA240 Types 304 or 304L stainless steel sheet and plate stock, and SA564-630 (precipitation hardened) stainless steel for the adjustable support pedestals. The only non-stainless material utilized in the rack is the neutron absorber material, which is a boron carbide and aluminum metal matrix composite available under the patented product name MetamicTM.

The racks are designed to the stress limits of, and analyzed in accordance with,Section III, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel (B&PV) Code [1-1]. The material procurement, analysis, and fabrication of the rack modules conform to USNRC 10CFR50 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 the "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications,"

USNRC (1978) and 1979 Addendum thereto [1-2].

This report provides the'design basis for the replacement of fuel storage racks at BVPS Unit No.

2 and is prepared to support the license amendment process for the station. This report provides discussions of the new rack design, analysis methodology, and results. Holtec prepared the design and performed the engineering evaluations, and will also fabricate and install the racks.

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

The criticality safety analysis sets the requirements on the MetamicVM panel length and the amount of B 10 per unit area (i.e., loading density) of the MetamicTM panels for the SFP racks Holtec Report HI-2084175 1-2 Holtec Pro-ject 1702

being added. The criticality safety analysis requires that the neutron multiplication factor for the stored fuel array be bounded by the USNRC keff limit of 0.95 under assumptions of 95%

probability and 95% confidence.

Rack module structural analysis requires that primary stresses in the rack module structure remain below the ASME B&PV Code (Subsection NF) [1-1] allowable stresses. Demonstrations of seismic and structural adequacy are presented in Section 5.

Thermal-hydraulic consideration requires that fuel cladding will not fail due to excessive thermal stress, and that the steady state pool bulk temperature will remain within the limits prescribed for the spent fuel pool to satisfy the pool structural strength, operational, and regulatory requirements. The thermal-hydraulic analyses carried out in support of this storage expansion effort are described in Section 6.

Mechanical accident qualification 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 results of the radiological evaluation performed to support the rerack.

Section 9 addresses the most important considerations in the installation of the racks.

All computer programs utilized to perform the analyses documented in this report are benchmarked and verified. These programs have been utilized in numerous license applications over the past decade.

The analyses presented herein clearly demonstrate that the rack module arrays possess wide margins of safety in respect to all considerations of safety specified in the USNRC OT Position Paper [1-2], namely, nuclear subcriticality, thermal-hydraulic safety, seismic and structural adequacy, radiological compliance, and mechanical integrity.

HI-2084175 1-3 Holtec Project 1702 Holtec Report HI-2084175 Holtec Report 1-3 Holtec Project 1702

1.1 References

[1-1] American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel Code,Section III, Subsection NF, latest Edition.

[1-2] USNRC, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14, 1978, and Addendum dated January 18, 1979.

HI-2084175 1-4 Holtec Project 1702 Holtec Report HI-2084175 Holtec Report 1-4 Holtec Project 1702

L 1le 92 1 1 1 -

01 1 00 0 o~ --

0 0 01 L 00 00 0~ CI - - CIO C

D _a -~2 Uo 0__ ~ 0-56 a C)- , _ _

I I 0 0 0 0 0 -' --- ------ r-.`

A]I~ AI

%5I- I II FIGURE 1.1 - PROPOSED NEW RACK LAYOUT FOR BEAVER VALLEY POWER STATION (BVPS) UNIT No. 2 Holtec Report HI-2084175 1-5 Holtec Project 1702

2.0 FUEL STORAGE RACKS 2.1 Introduction In its fully implemented configuration, the Fuel Storage Pool will contain 15 fuel racks with a maximum storage capacity of 1690 assemblies. All fuel storage rack arrays being added will consist of freestanding modules, made primarily from austenitic stainless steel containing honeycomb storage cells interconnected through longitudinal welds. A panel of Metamic metal matrix composite containing a high areal loading of the Boron-10 (B-10) isotope provides appropriate neutron attenuation between adjacent storage cells. Figure 2.1.1 provides an isometric schematic of a typical non-flux trap design storage rack module. Data on the cross sectional dimensions, weight and cell count for each rack module is presented in Table 2.1 .1.

The baseplates on all fuel rack modules extend out beyond the rack module periphery wall such that the plate protrusions act to set a required minimum separation between the facing cells in adjacent rack modules. Each fuel rack module is supported by four or five pedestals, which are remotely adjustable. The rack module support pedestals length adjustment is primarily provided to accommodate minor level variations in the pool floor flatness. Thus, the racks can be installed in a vertical position and the top of the racks can easily be made co-planar with each other. Some pedestals will be supported by the existing sub-base beams Between the rack module pedestals and the pool floor liner is a bearing pad, which serves to diffuse the dead load of the loaded racks into the reinforced concrete structure Of the pool slab. The bearing pads are part of the new rack installation.

The overall design of the rack modules is similar to those presently in service in the spent fuel pools at many other nuclear plants. Altogether, Holtec has provided over 50 thousand storage cells of this design to various nuclear plants around the world.

Holtec Report HI-2084175 2-1 Holtec Project 1702

2.2 Summary of Principal Design Criteria The key design criteria for the new fuel racks are set forth in the USNRC memorandum entitled "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications,"

dated 14 April 1978 as modified by amendment dated 18 January 1979. The individual sections of this report expound on the specific design bases derived from the above-mentioned "OT Position Paper". A brief summary of the design bases for the racks is presented in the following:

a. Disposition: All new rack modules are required to be freestanding.

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

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: The spatial average bulk pool temperature is required to remain below allowable levels subsequent to an offload, with due consideration of a worst-case single active cooling system failure.

e. Criticality Compliance: The fuel storage racks must be able to store Zircaloy clad fuel of 5.0 weight percent (w/o) maximum enrichment while maintaining the reactivity (KIf) less than 1.0 without soluble boron and 0.95 with soluble boron.

f. Bearing Pads: The bearing pad size and thickness must ensure that the pressure on the concrete continues to satisfy the American Concrete Institute (ACI)Jlimits during and after a seismic event.

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

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

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

Holtec Report HI-2084175 2-2 Holt6c PrQject 1702

a. Design Codes (1) AISC Manual of Steel Construction, 8th edition, 1980.

(2) ANSI N210-1976, "Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations" (contains guidelines for fuel rack design).

(3) American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code Section III, Subsection NF.;Section VIII; and Section XI, 1998 Edition through 2000 Addenda.

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

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

(6) Code Requirements for Nuclear Safety Related Concrete Structures, ACI349-85/AC1349R-85, and AC1349.1R-80.

(7) ASME NQA-l-1989, Quality Assurance Program Requirements for Nuclear Facilities (8) ASME NQA-2-1989, Quality Assurance Requirements for Nuclear Facility Applications.

(9) ANSI Y14.5M-1994, Dimensioning and Tolerancing for Engineering Drawings and Related Documentation Practices (10) ACI-315, Manual of standard practice for detailing reinforced concrete structures, 1965 or 1974 edition.

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

(2) A240-97a - Standard Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet and Strip for Fusion-Welded Unfired Pressure Vessels.

(3) A262-93a - Detecting Susceptibility to Intergranular Attack in Austenitic Stainless Steel.

HI-2084175 2-3 Holtec Project 1702 Report HI-2084175 Holtec Report 2-3 Holtec Project 1702

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

(5) A479-97a - Steel Bars for Boilers & Pressure Vessels.

(6) A564-95, Standard Specification for Hot-Rolled and Cold-Finished Age-Hardening Stainless and Heat-Resisting Steel Bars and Shapes.

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

(8) A380 Recommended Practice for Descaling, Cleaning and Marking Stainless Steel Parts and Equipment.

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

(10) E3-95, Preparation of Metallographic Specimens (11) E 190-92, Guided Bend Test for Ductility of Welds (12) American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section II-Parts A and C, 1998 Edition.

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

c. Welding Codes: ASME Boiler and Pressure Vessel Code,Section IX - Welding and Brazing Qualifications, 1998 Edition.

d. Quality Assurance, Cleanliness, Packaging, Shipping, Receiving, Storage, and Handling Requirements (1) ANSI 45.2.1-1980 - Cleaning of Fluid Systems and Associated Components during Construction Phase of Nuclear Power Plants.

(2) ANSI N45.2.2-1972 - Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants (During the Construction Phase).

(3) ANSI - N45.2.6-1978 - Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants (Regulatory Guide 1.58).

(4) ANSI-N45.2.8-1975, Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Plants.

Holtec Report HI-2084175 2-4 Holtec Project 1702

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

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

(7) ANSI N45.2.13-1976 - Quality Assurance Requirements for Control of Procurement of Equipment Materials and Services for Nuclear Power Plants (Regulatory Guide 1.123).

(8) ANSI N45.2.15-1981 - Hoisting, Rigging, and Transporting of Items For Nuclear Power Plants.

(9) ANSI N45.2.23-1978 - Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants (Regulatory Guide 1.146).

(10) ASME Boiler and Pressure Vessel,Section V, Nondestructive Examination, 1998 Edition.

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

(12) ASME Boiler and Pressure Vessel Code NCA3550 Requirements for Design Documents, 1998 Edition.

(13) ASME Boiler and Pressure Vessel Code NCA4000 - Quality Assurance, 1998 Edition.

(14) ASME NQA-1, Requirements for the establishment and execution of quality assurance programs for the siting, design, construction, operation, and decommissioning of nuclear facilities, 1994 Edition.

e. Governing NRC Design Documents (1) NUREG-0800, Radiological Consequences of Fuel Handling Accidents, 07/1981.

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

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

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

Holtec Report HI-2084175 2-5 Holtec Project 1702

(1) ANSI/ANS 8.1 (N16.1) - Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, 1998.

(2) ANSI/ANS 8.17-1984 (R1997), Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors (3) ANSI N45.2 - Quality Assurance Program Requirements for Nuclear 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/ANS 57.2 - Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants - 1983.

(7) ANSI N14.6 - American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500 kg) or more for Nuclear Materials- 1992 (8) ANSI/ASME N626-3-1993, Qualification and Duties of Personnel Engaged in ASME Boiler and Pressure Vessel Code Section III, Div. 1, Certifying Activities

g. Code of Federal Regulations (1) 10CFR20 - Standards for Protection Against Radiation (2) 10CFR21 - Reporting of Defects and Noh-compliance (3) 10CFR50 - Appendix A - General Design Criteria for Nuclear Power Plants (4) 10CFR50 - Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants (5) 10CFR61 - Licensing Requirements for Land Disposal of Radioactive Material (6) 10CFR71 - Packaging and Transportation of Radioactive Material

h. Regulatory Guides (RG)

(1) RG 1.13 - Spent Fuel Storage Facility Design Basis (Revision 1 - December, 1975)

Holtec Report HI-2084175 2-6 Holtec Project 1702

(2) RG 1.25 - Assumptions Used for' Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility of Boiling and Pressurized Water Reactors (Rev. 0 - March, 1972)

(3) RG 1.28 - (ANSI N45.2) - Quality Assurance Program Requirements (Rev. 2 -

February 1979)

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

(5) RG 1.31 - Control of Ferrite Content in Stainless Steel Weld Material (Rev. 3 -

April, 1978)

(6) RG 1.38 - (ANSI N45.2.2) Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants (Rev. 2 - May, 1977)

(7) RG 1.44 - Control of the Use of Sensitized Stainless Steel (Rev. 0 - May, 1973)

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

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

1973)

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

(11) RG 1.71 - Welder Qualifications for Areas of Limited Accessibility (Rev. 0 -

December, 1973)

(12) RG 1.74 - (ANSI N45.2.10) Quality Assurance Terms and Definitions (Rev. 2 -

February, 1974)

(13) RG 1.85 - Materials Code Case Acceptability - ASME Section 3, Div. 1 (Rev. 24

-June, 1986)

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

(15) RG 1.92 - Combining Modal Responses and Spatial Components in Seismic Response Analysis (Rev. 1 - February, 1976)

(16) RG 1.122 - Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components (Rev. 1 - February 1978)

HI-2084175 2-7 Holtec Project. 1702 Holtec Report HI-2084175 Holtec Report 2-7 Holtec Project. 1702

(17) RG 1.123 - (ANSI N45.2.13) Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants (Rev. 1 - July, 1977)

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

(19) RG 3.4 - Nuclear Criticality Safety in Operations with Fissionable Materials at Fuels and Materials Facilities (20) RG 3.41 - Validation of Calculational Methods for Nuclear Criticality Safety, Revision 1, 1977 (21) RG 8.8 - Information Relative to Ensuring that Occupational Radiation Exposure at Nuclear Power Plants will be as Low as Reasonably Achievable (ALARA)

(22) DG-8006, "Control of Access to High and Very High Radiation Areas in Nuclear Power Plants" (23) IE Information Notice 83 Fuel Binding Caused by Fuel Rack Deformation (24) RG 8.38 - Control of Access to High and Very High Radiation Areas in Nuclear Power Plants (June, 1993)

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

j. Standard Review Plan (1) SRP 3.2.1 - Seismic Classification, Revision 2, March 2007.

(2) SRP 3.2.2 - System Quality Group Classification, Revision 2, March 2007.

(3) SRP 3.7.1 - Seismic Design Parameters, Revision 3, March 2007.

(4) SRP 3.7.2 - Seismic System Analysis, Revision 3, March 2007.

(5) SRP 3.7.3 - Seismic Subsystem Analysis, Revision 3, March 2007.

(6) SRP 3.8.4 - Other Seismic Category I Structures (including Appendix D),

Technical Position on Spent Fuel Rack, Revision 2, March 2007.

(7) SRP 3.8.5 - Foundations for Seismic Category I Structures, Revision 2, March 2007.

Holtec Report HI-2084175 2-8 Holtec Project 1702

(8) SRP 9.1.2 - Spent Fuel Storage, Revision 3, July 1981.

(9) SRP 9.1.3 - Spent Fuel Pool Cooling and Cleanup System, Revision 2, March 2007.

(10) SRP 9.1.4 - Light Load Handling System, Revision 3, March 2007.

(11) SRP 9.1.5 - Heavy Load Handling System, Revision 1, March 2007.

(12) SRP 15.7.4 - Radiological Consequences of Fuel Handling Accidents, Revision 1, July 1981.

k. AWS Standards (1) AWS D1. 1:2008 - Structural Welding Code, Steel (2) AWS D1.3:2008 - Structure Welding Code - Sheet Steel (3) AWS D9.1:2006 - Welding of Sheet Metal (4) AWS A2.4:2007 - Standard Symbols for Welding, Brazing and Nondestructive Examination (5) AWS A3.0:2001 - Standard Welding Terms and Definitions (6) AWS A5.12:1998(R2007) - Tungsten Arc-welding Electrodes (7) AWS QC1:2007 - Standards and Guide for Qualification and Certification of Welding Inspectors 2.4 Quality Assurance Program The governing quality assurance requirements for design and fabrication of the spent fuel racks are stated in 10CFR50 Appendix B. Holtec's Nuclear Quality Assurance program complies with this regulation and is designed to provide a flexible but highly controlled system for the design, analysis and licensing of customized components in accordance with various codes, specifications, and regulatory requirements.

2-9 Holtec Project 1702 Report HI-2084175 Holtec Report HI-2084175 2-9 Holtec Project 1702

2.5 Mechanical Design The 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 fuel racks are summarized in Table 2.5.1.

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

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 fuel storage 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.

Holtec Report HI-2084175 2-10 Holtec Project 1702

4. The racks are fabricated per the Holtec 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.

2.6.1 Non-Flux-Trap Rack Module Description This section describes the constituent elements of the non-flux-trap rack module in the fabrication sequence. Figure 2.1.1 provides a schematic view of a typical non-flux-trap rack.

The rack module manufacturing begins with fabrication of the "box". The boxes are fabricated from two precision formed channels by seam welding in a machine equipped with copper chill bars and pneumatic clamps to minimize distortion due to welding heat input. Figure 2.6.1 shows the box. The minimum weld seam penetration is 80% of the box metal gauge, which is 0.075 inch (14 gauge).

Each box constitutes a storage location. Each external box side is equipped with a stainless steel sheathing, which holds one integral Metamic sheet (neutron absorber material). The design objective calls for attaching Metamic tightly on the box surface. This is accomplished by die forming the internal and external box sheathings, as shown in Figure 2.6.1. The flanges of the sheathing are attached to the box using skip welds and spot welds. The sheathings serve to locate and position the neutron absorber sheet accurately, and to preclude its movement under seismic conditions.

Having fabricated the required number of composite box assemblies, they are arranged in a checkerboard array to form an assemblage of storage cell locations (Figure 2.6.2). Filler panels and corner angles are welded to the edges of the boxes at the outside boundary of the rack to make the peripheral formed cells. The inter-box welding and pitch adjustment are accomplished by small longitudinal connectors. This assemblage of box assemblies is welded edge-to-edge as shown in Figure 2.6.2, resulting in a honeycomb structure with axial, flexural and torsional rigidity depending on the extent of inter-cell welding provided. It can be seen from Figure 2.6.2 that two edges of each interior box are connected to the contiguous boxes resulting in a well-defined path for "shear flow", and essentially makes the box assemblage into a multi-flanged beam-type structure.

Holtec Report HI-2084175 2-11 Holtec Project 1702

The "baseplate" provides a continuous horizontal surface for supporting the fuel assemblies. The baseplate is attached to the bottom edge of the boxes. The baseplate is a 3/4-inch thick austenitic stainless steel plate stock that has 5-inch diameter holes (except at lift locations that have irregular shaped cut-outs of similar area) cut out in a pitch identical to the box pitch. The baseplate is attached to the cell assemblage by fillet welding the box edge to the plate.

In the final step, adjustable leg supports (shown in Figure 2.6.3) are welded to the underside of the baseplate. The bottom part is made of A564 Gr. 630 17-4 Ph series stainless steel to avoid galling problems. The adjustable legs provide a +1/4-inch, -1/2-inch vertical height adjustment at each leg location.

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, as well as liquid penetrant examination of support leg welds, in accordance with the design drawings.

An elevation view of three contiguous rack cells is shown in Figure 2.6.4.

Holtec Report HI-2084175 2-12 Holtec Project 1702

TABLE 2.1.1: GEOMETRIC AND PHYSICAL DATA FOR THE NEW FUEL RACKS RACK RACK NO. OF CELLS MODULE ENVELOPE SIZE WEIGHT (Lbs) NO. OF CELLS PER I.D. TYPE (Nom.) (approx) RACK N-S E-W N-S E-W Direction Direction Al Non-Flux-Trap 9 12 81.565 108.655 16,200 96 A2 Non-Flux-Trap 9 12 81.565 108.655 17,800 108 A3 Non-Flux-Trap 9 8 81.565 72.535 11,700 72 BI Non-Flux-Trap 9 13 81.565 117.685 18,500 111 B2 Non-Flux-Trap 9 12 81.565 108.655 17,800 108 B3 Non-Flux-Trap 9 12 81.565 108.655 17,800 108 B4 Non-Flux-Trap 9 14 81.565 126.715 19,500 126 Cl Non-Flux-Trap 9 13 81.565 117.685 18,800 117 C2 Non-Flux-Trap 9 12 81.565 108.655 17,800 108 C3 Non-Flux-Trap 9 12 81.565 108.655 17,800 108 C4 Non-Flux-Trap 9 14 81.565 126.715 20,100 126 DI Non-Flux-Trap 10 13 90.595 117.685 19,600 122 D2 Non-Flux-Trap 10 12 90.595 108.655 19,000 120 D3 Non-Flux-Trap 10 12 90.595 108.655 19,000 120 D4 Non-Flux-Trap 10 14 90.595 126.715 21,400 140 Holtec Report HI-2084175 2-13 Holtec Project 1702

Table 2.5.1 MODULE DATA FOR NON-FLUX TRAP FUEL RACKS' Storage Cell Inside Dimension 8.8 in.

Cell Pitch 9.03 in.

Storage Cell Height (above baseplate) 168-5/8 in.

Baseplate Hole Diameter (except for lift locations) 5 in.

Baseplate Thickness 3/4 in.

Support Pedestal Height (including bearing pad) 1.1-3/8 in.

Support Pedestal Type Remotely adjustable pedestals Number Of Support Pedestals Per Rack 4 or 5 Number Of Cell Walls Containing Auxiliary Flow 4 Holes At Base Of Cell Wall Remote Lifting And Handling Provisions Yes Neutron Absorber Material Metamic Neutron Absorber Length 146 in.

Neutron Absorber Width 7-1/2 in.

1 All dimensions indicate nominal values.

Holtec Report HI-2084175 2-14 Holtec Project 1702

RACK FIGURE 2. 1.1 - ISOMETRIC VIEW OF GENERIC NON-FLUX TRAP Holtec Report HI-2084175 2-15 Holtec Project 1702

-C o

00 CELL WALL 00

~~~~POISON .. ""

PANEL ""

AUXILIARY FLOW HOLES TAPERED WELDED Ci) AND SMOOTHED ENDS o

I. I FIGURE 2.6.2 - PLAN VIEW OF GENERIC NON-FLUX TRAP RACK ARRAY Holtec Report HI-2084175 2-17 Holtec Project 1702

LEVEUNG TOOL SOCE*w k\\N \

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? I I II SII 00-R IN II LE TYPICAL ELEVATION MEW TYPICAL PLAN MIEWS OF RACK BASEPLATE CORNER FIGURE 2.6.3 - ADJUSTABLE PEDESTAL DESIGN Holtec Report HI-2084175 2-18 Holtec Project 1702

SPACER IJEVELUPEJJ CELL CELL (TYP. FIR EPITCH' INNER CELL WALLS) m "9 m*mlmm*m*mm*

fj-1 PANEL STORED I ACTIVE P*ISON CELL FUEL i LEN*iTH I ruGT H ASSENBLY I Ii I I I I 7-

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I L*I JBASEPLATE HlJLE FIGURE 2.6.4 - NON-FLUX-TRAP RACK CELLS ELEVATION VIEW Holtec Report HI-2084175 2-19 Holtec Project 1702

3.0 MATERIAL CONSIDERATIONS 3.1 Introduction Safe storage of nuclear fuel in the pool 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 service life.

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

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

b. Internally threaded support pedestals: ASTM A240 Type 304

c. Externally threaded support pedestals: ASTM A564-630 precipitation hardened stainless steel (heat treated to 11 00°F)

d. Weld material - ASTM Types 308 or 308L 3.3 Neutron Absorbing Material The MetamicTM neutron absorber material, proposed for use in the new racks, is manufactured by the Holtec Nanotec Materials Division in Lakeland, Florida. As discussed below, MetamicTM has been subjected to rigorous tests by various organizations including Holtec International, and has been approved by the USNRC in recent dry (Dockets 71-9261 and 72-1014) as well as recent wet storage applications for Arkansas Nuclear One Units 1 and 2 (Dockets 50-313 and 50-368),

Clinton (Docket 50-461), Diablo Canyon Units 1 and 2 (Dockets 50-275 and 50-323), St. Lucie' Unit 2 (Docket 50-389), Turkey Point Unit 3 (Docket 50-250) and Cooper Nuclear Station (Docket 50-298).

Metamic TM was developed in the mid-1990s by the Reynolds Metals Company [3.2.9] with the technical support of the Electric Power Research Institute (EPRI) for spent fuel reactivity control in dry and wet storage applications with the explicit objective to eliminate the performance frailties of aluminum cermet type of absorbers reported in the industry. Metallurgically, Holtec Report HI-2084175 3-1 Holtec Project 1702

MetamicTM is a metal matrix composite (MMC) consisting of a matrix of aluminum reinforced with Type 1 ASTM C-750 boron carbide. Metamic TM is characterized by extremely fine aluminum (325 mesh or smaller) and boron carbide (B 4C) powder. Typically, the average B 4C particle size is between 10 and 40 microns. The high performance and reliability of MetamicTM derives from the fineness of the B 4C particle size and uniformity of its distribution, which is solidified into a metal matrix composite structure by the powder metallurgy process. This yields excellent homogeneity and a porosity-free material.

In MetamicTM's manufacturing process, the aluminum and boron carbide powders~are carefully-blended without binders or other additives that could potentially adversely influence MetamicTM's performance. The blend of powders is isostatically compacted into a green billet under high pressure and vacuum sintered to near theoretical density. The billet is extruded and subjected to multiple rolling operations to produce sheet stock of the required thickness and a tight thickness tolerance. An array of U.S. patents discloses the unique technologies that underlie the MetamicTM neutron absorber [3.2.1-3.2.4].

In recognition of the central role of the neutron absorber in maintaining the subcriticality, Holtec International utilizes appropriately rigorous technical and quality assurance criteria and acceptance protocols to ensure satisfactory -neutron absorber performance over the service life of the fuel racks. Holtec International's Quality Assurance Program ensures that MetamicTM will be manufactured under the control and surveillance of a Quality Assurance/Quality Control Program that conforms to the requirements of 10CFR50 Appendix B, "Quality Assurance Criteria for Nuclear Power Plants." Consistent with its role in reactivity control, all neutron absorbing material procured for use in the Holtec racks is categorized as Safety Related (SR). SR manufactured items, as required by Holtec's NRC-approved Quality Assurance program, must be produced to essentially preclude, to the extent possible, the potential of an -error in the procurement of constituent materials and the manufacturing processes.

Accordingly, material and manufacturing control processes must be established to eliminate the incidence of errors, and inspection steps are implemented to serve as an independent set of Holtec Report HI-2084175 3-2 Holtec Project 1702

barriers to ensure that all critical characteristics defined for the material by Holtec's design team are met in the manufactured product.

3.3.1 Characteristics of MetamicTM Because MetamicTM is, a porosity-free material, unlike Boral, there is no capillary path through which spent fuel pool water can penetrate MetamicTM panels and chemically react with aluminum in the interior of the material to generate hydrogen. Thus, the potential of swelling and generation of significant quantities of hydrogen is eliminated.

To determine its physical stability and performance characteristics, MetamicTM was subjected to an extensive array of tests sponsored by EPRI that evaluated the functional performance of the material at elevated temperatures (up to 900'F) and radiation levels (1E+I 1 rads gamma). The results of the tests documented in an EPRI report [3.2.5] indicate that MetamicTM maintains its physical and neutron absorption properties with little variation in its properties from the unirradiated state. The main conclusions provided in the above-referenced EPRI report, which endorsed MetamicTM for dry and wet storage applications on a generic basis, are summarized below:

" The metal matrix configuration produced by the powder metallurgy process with almost a complete absence of open porosity in MetamicTM ensures that its density is essentially equal to the theoretical density.

The physical and neutronic properties of MetamicTM are essentially unaltered under exposure to elevated temperatures (750' F - 9000 F).

No detectable change in the neutron attenuation characteristics under accelerated corrosion test conditions has been observed.

Additional technical information on MetamicTM in the literature includes independent measurements of boron carbide particle distribution in MetamicTM panels, which showed extremely small particle-to-particle distance [3.2.6]. The USNRC has previously approved MetamicTM for use in both wet storage [3.2.7] and dry storage [3.2.8] applications.

Holtec Report HI-2084175 3-3 Holtec Project 1702

MetamicTM has also been subjected to independent performance assessment tests by Holtec International in the company's Florida laboratories since 2001 [3.2.9, 3.2.10]. The three-year long experimental study simulated limiting environmental conditions in wet and dry storage. No anomalous material behavior was observed in any of the tests. These independent Holtec tests essentially confirmed earlier EPRI and other industry reports cited in the foregoing with regard to MetamicTM's suitability as a neutron absorber in fuel storage applications.

3.4 In-Service Surveillance of the Neutron Absorber 3.4.1 Purpose Metamic TM , the neutron absorbing material incorporated in the spent fuel storage rack design to assist in controlling system reactivity, consists of finely divided particles of boron carbide (B 4 C) uniformly distributed in type 6061 aluminum powder. Tests simulating the radiation, thermal and chemical (i.e., boric acid solutions) environment of the spent fuel pool have demonstrated the stability and chemical inertness of Metamic.

Based upon the accelerated test programs, Metamic is considered a satisfactory material for reactivity control in spent fuel storage racks and is fully expected to fulfill its design function over the lifetime of the racks. Nevertheless, as a defense-in-depth measure, a Metamic surveillance program has been developed and will be implemented for the SFP in order to monitor the integrity and performance of Metamic.

The purpose of the surveillance program is to characterize certain properties of the Metamic with the objective of providing data necessary to assess the capability of the Metamic panels in the racks to continue to perform their intended function. The surveillance program is also capable of detecting the onset of any significant degradation with ample time to take such corrective action as may be necessary.

The Metamic surveillance program depends primarily on representative coupon samples to monitor performance of the absorber material without disrupting the integrity of the storage Holtec Report HI-2084175 3-4 Holtec Project 1702

system. The principal parameters to be measured are the thickness (to monitor for swelling) and B- 10 loading (to monitor for the continued presence of boron in the Metamic).

3.4.2 Coupon Surveillance Program 3.4.2.1 Coupon Description The coupon measurement program includes coupons suspended on a mounting (called a "tree"),

placed in a designated cell, and surrounded by spent fuel. Coupons may be removed from the array on a prescribed schedule and certain physical and chemical properties measured from which the stability and integrity of the Metamic in the storage cells may be inferred.

The coupon surveillance program uses a tree with a total of 8 to 10 test coupons. In mounting the coupons on the tree, the coupons are positioned axially within the central eight. feet (approximate) of the active fuel zone where the gamma flux is expected to be reasonably uniform.

The coupons are taken from the same lot as that used for construction of the racks. Each coupon is carefully precharacterized prior to insertion in the pool to provide reference initial values for comparison with measurements made after irradiation. As a minimum, the surveillance coupons are precharacterized for weight, dimensions (especially thickness) and B-10 loading.

3.4.2.2 Surveillance Coupon Testing Schedule To assure that the coupons will have experienced a slightly higher radiation dose than the Metamic in the racks, one of the following strategies shall be implemented:

1. The coupon tree is surrounded by freshly-discharged fuel assemblies after each of the first four refueling outages. At the time of the first fuel offload following installation of the coupon, tree, the four storage cells surrounding the tree shall be loaded with discharged fuel assemblies which are not scheduled to be returned to the core.

Holtec Report HI-2084175 3-5 Holtec Project 1702

2. At least one (1) location adjacent to the coupon tree is loaded with freshly-discharged fuel assemblies after each refueling outage for the life of the racks. At the time of the first fuel offload following installation of the coupon tree, one of the four storage cells surrounding the tree shall be loaded with a discharged fuel assembly which is not scheduled to be returned to the core.

At the scheduled test date, the coupon tree is removed and a coupon removed for evaluation.

A sample coupon measurement schedule is shown in Table 3.4.1.

Evaluation of the coupons removed will provide information of the effects of the radiation, thermal and chemical environment of the pool and by inference, comparable information on the Metamic panels in the racks. Over the duration of the coupon testing program, the coupons will have accumulated more radiation dose than the expected lifetime dose for normal storage cells.

Coupons that have not been destructively analyzed by wet-chemical processes may optionally be returned to the storage pool and remounted on the tree. They will then be available for subsequent investigation of defects, should any be found.

3.4.2.3 Measurement Program The coupon measurement program is intended to monitor changes in physical properties of the Metamic absorber material by performing the following measurements on the preplanned sche-dule:

Visual Observation and Photography

Neutron Attenuation

Dimensional Measurements (length, width and thickness)

Weight and Specific Gravity 3.4.2.4 Surveillance Coupon Acceptance Criteria Report HI-2084 175 3-6 Holtec Project 1702 Holtec Report Holtec HI-2084175 3-6 Holtec Project 1702

Of the measurements to be performed on the Metamic surveillance coupons, the most important are (1) the neutron attenuation' measurements (to verify the continued presence of the boron) and (2) the thickness measurement (as a monitor of potential swelling). Acceptance criteria for these measurements are as follows:

A decrease of no more than 5% in Boron-10 (B-10) content, as determined by neutron attenuation, is acceptable. (This is equivalent to a requirement for no loss in boron within the accuracy of the measurement.)

" An increase in thickness at any point should not exceed 10% of the initial thickness at that point.

Changes in excess of either of these two criteria requires investigation and engineering evaluation which may include early retrieval and measurement of one or more of the remaining coupons to provide corroborative evidence that the indicated change(s) is real. If the deviation is determined to be real, an engineering evaluation shall be performed to identify further testing or any corrective action that may be necessary.

The remaining measurement parameters serve a supporting role and should be examined for early indications of the potential onset of Metamic degradation that would suggest a need for further attention and possibly a change in measurement schedule. These include (1) visual or photographic evidence of unusual surface pitting, blistering, corrosion or edge deterioration, or (2) unaccountable weight loss in excess of the measurement accuracy.

3.5 References

[3.2.1] U.S. Patent # 6,332,906 entitled "Aluminum-Silicon Alloy formed by Powder",

Thomas G. Haynes III and Dr. Kevin Anderson, issued December 25, 2001.

1 Neutron attenuation measurements are a precise instrumental method of chemical analysis for Boron-10 content using a nondestructive technique in which the percentage of thermal neutrons transmitted through the panel is measured and compared with predetermined calibration data. Boron-10 is the nuclide of principal interest since it is the isotope responsible for neutron absorption in the Metamic panel.

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[3.2.2] U.S. Patent # 5,965,829 entitled "Radiation Absorbing Refractory Composition and Method of Manufacture", Dr. Kevin Anderson, Thomas G. Haynes III, and Edward Oschmann, issued October 12, 1999.

[3.2.3] U.S. Patent # 6,042,779 entitled "Extrusion Fabrication Process for Discontinuous Carbide Particulate Metal and Super Hypereutectic Al/Si Alloys", Thomas G.

Haynes III and Edward Oschmann, issued March 28, 2000.

[3.2.4] U.S. Patent Application 09/433773 entitled "High Surface Area Metal Matrix Composite Radiation Absorbing Product", Thomas G. Haynes III and Goldie Oliver, filed May 1, 2002.

[3.2.5] "Qualification of METAMIC for Spent Fuel Storage Application," EPRI, 1003137, Final Report, October 2001.

[3.2.6] "METAMIC Neutron Shielding", by K. Anderson, T. Haynes, and R. Kazmier, EPRI Boraflex Conference, November 19-20 (1998).

[3.2.7] "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.

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

[3.2.9] "Use of METAMIC in Fuel Pool Applications," Holtec Information Report No.

HI-2022871, Revision 1 (2002).

[3.2.10] "Sourcebook for MetamicTM Performance Assessment" by Dr. Stanley Turner, Holtec Report No. HI-2043215, Revision 2 (2006).

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Table 3.4.1 SAMPLE COUPON MEASUREMENT SCHEDULE Coupon Years' 1 2 2 4 3 6 4 8 5 10 6 15 7 20 8 25 9 30 10 40 I The years pertain to those after the installation of new racks.

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SHADED AREAS DENOTE PROPRIETARY INFORMATION 4.0 CRITICALITY SAFETY ANALYSIS 4.1 Introduction and Summary This chapter documents the criticality safety evaluation for the storage of fresh and spent fuel assemblies with an nominal initial enrichment of up to 5.0 wt% 235U in a Mixed-Zone Three Region (MZTR) storage arrangement (a loading pattern that includes a region of fresh fuel and two regions of spent fuel together in each storage rack) in Holtec high-density spent fuel storage racks (SFSRs) at the Beaver Valley Power Station (BVPS) Unit No. 2 operated by FirstEnergy Nuclear Operating Company (FENOC). This is a new criticality safety evaluation to support the installation of new racks to increase the storage capacity of the spent fuel pool.

The objective of this chapter is to demonstrate that the effective neutron multiplication factor (keff) is less than 1.0 with the storage racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with unborated water at a temperature corresponding to the highest reactivity. In addition, it is demonstrated that kefr is less than or equal to 0.95 with the storage racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with borated water at a temperature corresponding to the highest reactivity. The maximum calcu-lated reactivity includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than 0.95 with a 95% probability at a 95% con-fidence level. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95.

The new storage racks to be installed at BVPS Unit No. 2 spent fuel pool have storage cells that are regionalized for loading purposes into three distinct regions, with independent criteria defining each region (see also Figure 4.5.3, Figure 4.5.4a, Figure 4.5.4b, and Figure 4.5.7):

Region 1 is designed to accommodate fresh fuel with a maximum initial enrichment up to 5.0 wt% 235U. Region 1 storage cells are located on the periphery of each, rack (outer row only) and are therefore separated from other Region 1 cells in adjacent racks by the gap between the racks. Region 1 cells are additionally separated from other Region 1 cells within the same rack by Region 2 cells (including a Region 2 cell in the diagonal Holtec Report HI-2084175 4-1 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION direction, see Figure 4.5.3). Since Region 1 cells are qualified for the storage of fresh fuel, any fuel assembly (fresh or burned) meeting the maximum enrichment requirement may be stored in a Region 1 location.

" Region 2 is designed to accommodate fuel with a maximum initial enrichment of up to 5.0 wt% 235U and a high burnup defined according to the calculated Region 2 initial enrichment and burnup combination in Table 4.7.1 (see Section 4.7). Region 2 cells are located on the rack periphery (outer row) interspaced with (separating) Region 1 cells and are also located in the second row of cells (from the outside of the rack) separating the Region 1 cells from the Region 3 cells.

" Region 3 is designed to accommodate fuel with a maximum initial enrichment of up to 5.0 wt% 235U and a moderate burnup defined according to the calculated Region 3 initial enrichment and burnup combination in Table 4.7.2 (see Section 4.7). Region 3 cells are located on the interior of the rack (at least three rows in from the rack periphery) and are prohibited from being located in the outer two rows of the rack.

Additionally, reactivity effects of abnormal and accident conditions have also been evaluated. A summary of the types of accidents analyzed and the soluble boron required ensuring that the maximum keff remains below 0.95 are shown in Table 4.7.4. The most limiting accident is a misloaded fresh fuel assembly in the outer row of the rack in a Region 2 location. A minimum soluble boron requirement must be maintained in the spent fuel pool to ensure that the maximum kerr is less than 0.95 under accident conditions.

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SHADED AREAS DENOTE PROPRIETARY INFORMATION 4.2 Methodology 4.2.1 Criticality Analysis The principal method for the criticality analysis of the high-density storage racks is the use of 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 used previously and verified for criticality analyses and has all of the necessary features for this analysis. MCNP4a calculations used continuous energy cross-section data predominantly based on ENDF/B-V and ENDF/B-VI.

As discussed in [4.12], the use of the LFP may introduce an additional uncertainty which is accounted for by a 15% reactivity decrement factor that is statistically combined with the other uncertainties as described below.

Benchmark calculations, presented in [4.13], indicate a bias of 0.0013 with an uncertainty of +

0.0086 for MCNP4a, 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 [4.13].

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 Holtec Report HI-2084175 4-3 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION minimum of 50 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 calculation precision and computational time.

CASMO-4 is used in this application to determine reactivity, differences for temperature variation, manufacturing tolerances, depletion uncertainty and to calculate the isotopic inventory of the spent fuel for use in MCNP4a. References [4.5] and [4.6] are Studsvik proprietary documents related to the appropriateness of CASMO-4 for calculating the multiplication factor, kff. These references were previously provided to the NRC in support of staff approval of EMF-2158 as documented in letter "Document Control Desk ATTN: Chief, Planning, Program and Management Support Branch,

Subject:

Transmittal of Copies of CASMO-4 Benchmark Reports Relevant to EMF-2158(P) Revision 0 "Siemans Power Corporation Methodology for Boiling Water Reactors:

Evaluation and Validation of CASMO-4/MICROBURN-B2," from J.A. Umbarger, dated April 30, 1999.

Holtec International has been using the CASMO-4 code since approximately mid-1999 for calculating the reactivity effects of manufacturing tolerances, moderator temperature and depletion effects. CASMO-4 has been previously used and approved by the USNRC over the past 10 years on multiple licensing efforts by Holtec International for spent fuel storage racks. Specifically, CASMO-4 has been reviewed and approved for use on the following PWR spent fuel pool analyses for calculating the reactivity effect of moderator manufacturing tolerances, moderator temperature and depletion effects: Crystal River 3, Arkansas Nuclear 1 & 2, Harris, St. Lucie, Diablo Canyon, Turkey Point, V.C. Summer, Three Mile Island, Comanche Peak, Davis-Besse, Robinson, and Sequoyah. From the above list of plants, the following specific subset of NRC issued SERs and amendment approval references are identified where CASMO-4 and MCNP4a have been used by Holtec International for spent fuel pool criticality analyses:

F.E. Saba (NRC) to J.S. Forbes (Entergy) dated January 26, 2007, "ARKANSAS NUCLEAR ONE, UNIT NO.1 - ISSUANCE OF AMMENDMENT FOR USE OF Holtc 75 ReortHI-284 44 HltecProect 70 Holtec Report HI-2084175 4-4 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION METAMIC POISON INSERT ASSEMBLIES IN THE SPENT FUEL POOL (TAC NO.

MD2674)"

K.N. Jabbour (NRC) to C. M. Crane (Amerigen) dated October 31, 2005, "CLINTON POWER STATION, UNIT 1 - ISSUANCE OF AN AMENDMENT -RE: ONSITE SPENT FUEL STORAGE EXPANSION (TAC NO. MC4202)"

S.N. Baily (NRC) to D. E. Young (Crystal River) dated October 25, 2007, "CRYSTAL RIVER, UNIT 3 - ISSUANCE OF AMENDMENT REGARDING FUEL STORAGE PATTERNS IN THE SPENT FUEL POOL (TAC NO. MD3308)"

The BVPS Unit No. 2 spent fuel pool racks are similar in material and geometric configuration as those spent fuel racks used at other PWRs identified above. The use of CASMO-4 by Holtec International for spent fuel pool licensing activities on these PWR plants, and NRC approval of that use, provides the justification for using CASMO-4 for relative reactivity calculations for the BVPS Unit No. 2 spent fuel pool analysis.

As previously stated fuel depletion analyses during core operation were performed with CASMO-4 (using the 70-group cross-section library), a two-dimensional multi-group transport theory code based on the Method of Characteristics [4.4]-[4.6]. Detailed neutron energy spectra for each rod type are obtained in collision probability micro-group calculations for use in the condensation of the cross sections. CASMO-4 is used to determine the isotopic composition of the spent fuel. In addition, the CASMO-4 calculations are restarted in the storage rack geometry, yielding the two-dimensional infinite multiplication factor (kinf) for the storage rack to determine the reactivity effect of fuel and rack tolerances, temperature variation, 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.

Benchmark calculations, presented in [4.14], indicate a negative bias and bias uncertainty of +

0.0025 for CASMO-4 evaluated with a 95% probability at the 95% confidence level [4.1]. Since CASMO-4 is used to determine reactivity differences, the bias does not need to be applied to the results of the calculations. However, the bias uncertainty is included with the other uncertainties when determining the maximum keff values.

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SHADED AREAS DENOTE PROPRIETARY INFORMATION The maximum keff is determined from the MCNP4a calculated keff, the calculation bias, the temperature bias, and the applicable uncertainties and tolerances (bias uncertainty, calculation uncertainty, rack tolerances, fuel tolerances, depletion uncertainty, LFP uncertainty, Metamic coupon measurement uncertainty) using the following formula:

2 2 Max keff = Calculated kca]c + biases + [Yi (Uncertainty) ]"

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

When two separate, independent MCNP calculations are compared to determine a delta kc.ac, the uncertainty associated with each individual calculation is statistically combined and added to the kealc calculation according to the following equation:

Delta kcalc = (kcalc2 - kcalcl) + 2 * "1 (022 + 012) 4.2.2 Boron Dilution Analysis The methodology related to the Boron Dilution accident follows the general equation for boron dilution which is,

(_)F C,= C(-1)e F where Ct = boron concentration at time t, Co = initial boron concentration, V = credited volume of water in the pool, and F = flow rate of unborated water into the pool Holtec Report HI-2084175 4-6 . Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION This equation conservatively assumes the unborated water flowing into the pool mixes instantaneously with the water in the pool. The conservatism is with respect to the calculation of the time of dilution since any mixing rate factor is neglected. The volume of water in the SFP is conservatively set as the volume above the storage racks only. With respect to the reactivity in the spent fuel pool, pockets of unmixed unborated water may exist briefly during a dilution event and expose the fuel to an increased reactivity situation. However, the criticality calculations in Section 4.7 demonstrate that the spent fuel racks remain subcritical even with total loss of soluble boron.

For convenience, the above equation may be re-arranged to permit calculating the time required to dilute the soluble boron from its initial concentration to a specified minimum concentration, which is given below.

V t = -- ln(Co / C,)

F If V is expressed in gallons and F in gallons per minute (gpm), the time, t, will be in minutes.

4.3 Acceptance Criteria The high-density spent fuel PWR storage racks for BVPS Unit No. 2 are designed in accordance with the applicable codes and standards listed below. The objective of this evaluation is to show that the effective neutron multiplication factor, keff, is less than 1.0 with the storage racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with unborated water at a temperature corresponding to the highest reactivity. In addition, it is demonstrated that lff is less than or equal to 0.95 with the storage racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with borated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than 0.95 with a 95%

probability at a 95% confidence level [4.1]. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95 under borated conditions.

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SHADED AREAS DENOTE PROPRIETARY INFORMATION 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 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Rev. 3 - March 2007.

USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications (GL-78-01 1),

including modification letter dated January 18, 1979 (GL-79-004).

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 (Revision 1 -

December, 1975).

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

Code of Federal Regulations, Title 10, Part 50, Section 68, "Criticality Accident Requirements."

Additionally, the following NUREG's were reviewed and the criticality calculations presented here are consistent with the recommendations provided within:

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SHADED AREAS DENOTE PROPRIETARY INFORMATION I

Since credit is taken for soluble boron in the spent fuel pool (SFP), a boron dilution analysis is performed to ensure that sufficient time is available to detect and suppress the worst dilution event that can occur to reduce the SFP boron concentration from the operating concentration to the minimum required concentration as determined by this analysis. The boron dilution analysis considers all possible dilution events, sources and flow rates, as well as instrumentation and administrative procedures. The analysis justifies the surveillance interval for verifying the Technical Specification requirement concentration.

4.4 Assumptions Various assumptions used in this criticality analysis are conservative to provide an additional margin of criticality safety and to assure that the true reactivity will always be less than the calculated reactivity. The following conservative design criteria and assumptions were employed:

1) The SFP moderator is water at a temperature in the operating range that results in the highest reactivity, as determined by the analysis (see Section 4.7.8).

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2) Neutron absorption in minor structural members is neglected; spacer grids are replaced by water (see Section 4.7.1) because they have a negligible effect on reactivity.

3) The effective multiplication factor of an infinite radial array of fuel assemblies was used in the analyses, except for the assessment of certain abnormal/accident conditions and conditions where leakage is inherent.

4) The spent fuel rack neutron absorber length is conservatively modeled to be the same length as the active region of the fuel instead of the actual height of 146 inches.

5) A conservative cooling time of 0 hours0 days 0 hours 0 weeks 0 months is used along with setting the xenon concentration to zero for all CASMO-4 calculations in the rack models (see Section 4.7.4). No credit is taken for the significant cooling time of older fuel assemblies.

7) A maximum fuel pellet density (97% theoretical) is conservatively considered in the analysis over the entire fuel rod length and therefore no tolerance is needed for fuel pellet density (see Table 4.5.2). Fuel pellet dishing, champhering or pellets with an annulus are conservatively modeled as solid fuel pellet cylinders with the maximum density.

8) The guide tube dimensions used in the design basis assembly were conservatively set to the combination of diameters that provided the minimum thickness. Additionally, a tolerance is statistically considered (see Section 4.7.7).

9) The presence of burnable absorbers in fresh fuel is neglected for the design basis assembly.

This is conservative as burnable absorbers would reduce the reactivity of the fresh fuel assembly.

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SHADED AREAS DENOTE PROPRIETARY INFORMATION

10) To account for the possible positive reactivity effect of spent fuel depleted with burnable absorbers in the storage rack, a reactivity bias is determined and applied to all final reactivity calculations (see Section 4.5.4). The application of the reactivity bias to all fuel in the Region 2 MZTR calculations is conservative because it is also being applied to the fresh fuel in the modeL

11) In the depletion calculations with burnable absorbers to determine the reactivity bias discussed in (10) above (see Section 4.5.4), the burnable absorber was modeled over the entire active fuel length [4.12].

12) A very conservative segmented axial bumup profile was used in each fuel type group (i.e. enriched blankets, natural blankets, or no blankets)

(see Section 4.5.3).

13) The BVPS Unit No. 2 Cycle 14 uprated core operating parameters were conservatively applied to all the CASMO-4 depletion calculations [4.12]. These parameters are bounding and conservative because higher temperatures (maximum fuel and moderator) and boron concentrations (bounding cycle average) hardens the neutron spectrum in the core which leads to greater Pu production and therefore greater reactivity in the spent fuel pool.

14) The targeted reactivity (not including tolerances and applicable bias) by the initial enrichment and burnup combinations is conservatively set to 0.9920 for Region 3 calculations (minus relevant uncertainties and biases) and 0.9950 for Region 2 calculations (minus relevant uncertainties and biases) for the pure water calculations (instead of the limit of 1.0) and 0.9450 (minus relevant uncertainties and biases) for the borated water calculations (instead of the limit of 0.95).

15) For the boron dilution analysis low flow rate dilution accident a flow rate of 2 gpm is assumed.

This flow rate is assumed to be typical of undetectable leakage around seals and pumps.

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16) In the MCNP4a models, the fuel rod pellet to cladding gap is conservatively modeled as being essentially closed. Additionally, the cladding thickness is reduced to account for the gap closure as well as fuel rod growth (see Section 4.5.1) [4.12].

17) The fuel assemblies and fuel storage racks are assumed to have no non-integral reactivity control devices. The inclusion of such devices in the SFP reduces reactivity. Additionally, the storage of non-fissile material is not included in the reactivity calculations because they do not have a positive impact on reactivity. Although these non-fuel devices may exist and be stored in the SFP, they do not impact reactivity and therefore it is unnecessary to include them in the criticality calculations.

18) All calculations with steel use a material density of 7.84 g/cc.

4.5 Input Data 4.5.1 Design Basis Fuel Assembly Specification The spent fuel storage racks are designed to accommodate various Westinghouse designed 17x17 fuel assemblies used at BVPS Unit No. 2. Data provided by FENOC encompasses various regions of 17x17 fuel assemblies (STD, V5H, RFA, RFA-2). The design specifications for these fuel assemblies were compared (see Table 4.5.1 and Table 4.5.2) to select a design basis assembly that is conservatively bounding for the purposes of this analysis.

Additionally, during irradiation in light water reactors the fuel assemblies undergo physical changes associated with irradiation and residence time in an operating reactor as described in

[4.12]. Some of those changes are: fuel pellet densification, changes to clad geometry due to fuel rod growth, collapse of the pellet/cladding gas gap in the fuel rod (i.e. clad creep-down), and crud build up on the outside surface of the fuel rod. These fuel geometry changes are accounted for in the following way:

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Pellet densification: The spent fuel rack criticality methodology already includes an increase in fuel pellet density that covers both fresh and spent fuel. Therefore no additional calculations are required.

0 Clad geometry: Both fuel rod growth and clad creep-down have the potential to decrease the fuel-to-moderator ratio, thus potentially increasing reactivity. To address clad creep-down, the clad ID was reduced such that the pellet-clad gap was less than 0.0001 inches, while simultaneously reducing the clad OD to preserve overall clad volume. Then, to address clad thinning due to fuel rod growth, the clad OD was further reduced to allow for the maximum possible fuel rod growth of 1.31%.

0 Crud buildup: Crud buildup on the cladding increases the fuel-to-moderator ratio and therefore reduces reactivity. Thus, no additional calculations are required. Also, there has been no measureable crud-induced power shift (CIPS) at Beaver Valley Unit 2 since original startup of the plant, which indicates that crud deposition on the fuel has not been significant.

Therefore, the spent fuel calculations with MCNP4a to determine kcff include the creep model as described in [4.12].

4.5.2 Core Operating Parameters Core operating parameters are necessary for fuel depletion calculations performed with CASMO-4. The core parameters used for the depletion calculations are presented in Table 4.5.3.

The soluble boron concentration used bounds the average for BVPS Unit No. 2 Cycle 14 (see Table 4.5.4). The moderator and fuel temperatures are the maximum (most conservative) over the core for BVPS Unit No. 2 for Cycle 1 through Cycle 14. The neutron spectrum is hardened by each of these parameters, leading to a greater production of plutonium during depletion, which results in conservative reactivity results in the spent fuel pool [4.12].

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SHADED AREAS DENOTE PROPRIETARY INFORMATION 4.5.3 Axial Burnup Distribution An axial relative burnup profile was determined using data provided by FENOC for Cycles 1-14 and is specified for 24 equally-spaced axial nodes (see Table 4.5.6).

4.5.4 Integral and Non-integral Fuel Assembly Reactivity Control Devices The BVPS Unit No. 2 fuel had multiple regions that made use of integral fuel burnable absorbers (IFBAs). The design specifications for the IFBA rods are given in Table 4.5.7, Figure 4.5.1 and Figure 4.5.5 . Generic studies [10] have investigated the effect that IFBA has on the reactivity of spent fuel assemblies. These studies have concluded that there is a small positive reactivity effect associated with the presence of IFBA rods. Two separate IFBA rod loading patterns were analyzed to cover current and future fuel designs: 128 and 200 IFBA pins.

The BVPS Unit No. 2 fuel also had two fuel regions that made use of water displacement rods (WDR) that were used with and without IFBA rods. The WDR design specification is given in Table 4.5.8.

The' BVPS Unit No. 2 fuel also had two fuel regions that made use of wet annular burnable absorber (WABA) rods. The WABA rod specification is given in Table 4.5.9. Both WABA's and WDRs are present in the fuel assembly guide tubes.

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SHADED AREAS DENOTE PROPRIETARY INFORMATION 4.5.5 Storage Rack Specification The storage cell characteristics that were used in the criticality evaluations are summarized in Table 4.5.10. The spent fuel pool racks storage cells are composed of stainless steel boxes with a single fixed Metamic neutron absorber panel (attached by stainless steel sheathing), centered on each side. The stainless steel boxes are arranged in an alternating pattern such that the connection of the box comers forms the storage cells. Neutron absorber panels are installed on all exterior walls facing other racks and along the exterior of any rack location where it is physically possible to mislocate a fuel assembly.

The Region 3 design basis model is a single cell model with reflective boundary conditions through the center of the neutron absorber. See Figure 4.5.8.

4.5.6 Rack Interfaces and Cask Loading Pit The storage cell racks are separated by a 1.5 inch gap that is determined by the rack baseplate extensions and therefore is the rack wall to rack wall gap between stainless steel boxes. This gap represents the minimum possible separation and is used in the design basis models and therefore no further rack interface calculations are necessary. Additionally, one of the storage rack modules may be placed in the cask pit loading area during fuel movement operations. The results presented in this chapter are also applicable to this interim configuration.

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SHADED AREAS DENOTE PROPRIETARY INFORMATION 4.5.7 Fuel Rod Storage Basket The BVPS Unit No. 2 SFP has a Fuel Rod Storage Basket (FRSB) with the dimensions shown in Table 4.5.11 and presented in Figure 4.5.6. The FRSB is a basket that contains fuel rods, which is currently stored in a fuel storage rack cell in the SFP. For the purposes of this analysis the 23 5U.

FRSB was modeled as shown in Figure 4.5.6, fully loaded with fresh fuel pins at 5.0 wt%

The model contains fuel rods only, conservatively neglecting the steel box walls.

4.5.8 Boron Dilution Accident Evaluation The BVPS Unit No. 2 SFP has a Technical Specification Limiting Condition for Operation that requires a soluble boron concentration of 2000 ppm during normal operating conditions. The SFP volume is shown in Table 4.5.12 and conservatively considers only the volume of water above the storage racks. Under certain postulated abnormal conditions, introduction of unborated water into the spent fuel pool could reduce the spent fuel pool boron concentration below the Technical Specification requirement of 2000 ppm. The boron dilution accident evaluation considers two scenarios; the worst case high flow rate accident and an undetected low flow rate accident (see Assumption 15).

For the purposes of the boron dilution analysis it was determined that the source of water with the highest potential sustainable flow rate (worst case dilution event) was from the Service Water System, which is capable of providing 3000 gpm to the SFP.

4.5.9 Fuel Elevator The Beaver Valley Unit 2 SFP has a fuel elevator with the dimensions shown in Table 4.5.13.

For the purposes of this analysis the fuel elevator was modeled without any structural material, i.e. the cell wall was replaced by water. For conservatism, the model includes a mislocated fuel assembly directly adjacent to the fuel assembly in the fuel elevator.

4.6 Computer Codes 75 4-16 Holtec Project 1702 Holtec Report HI-20841 Holtec Report HI-2084175 4-16 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION The following computer codes were used during this analysis.

MCNP4a [2] is a three-dimensional continuous energy Monte Carlo code developed at Los Alamos National Laboratory. This code offers the capability of performing full three-dimensional calculations for the loaded storage racks. MCNP4a was run on the PCs at Holtec.

CASMO-4, Version 2.05.14 [[4.4]-[4.6]] is a two-dimensional multigroup transport theory code developed by Studsvik of Sweden. CASMO-4 performs cell criticality calculations and burnup. CASMO-4 has the capability of analytically restarting burned fuel assemblies in the rack configuration. This code was used to determine the reactivity effects of tolerances and fuel depletion. As proof of its acceptability in this application, CASMO-4 has been benchmarked [4.5, 4.6] against Monte Carlo calculations and critical experiments similar to spent fuel storage rack geometries. References 4.5 and 4.6 were previously provided to the NRC in support of staff approval of EMF-2158 for Siemens BWR methodology (see Section 4.2.1).

4.7 Analysis This section describes the calculations that were used to determine the acceptable storage criteria for the MZTR. 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 within the uncertainty analysis as discussed below.

As discussed in Section 4.2, MCNP4a was the primary code used. CASMO-4 was used to determine the reactivity effect of tolerances, reactivity bias, temperature variation and for depletion calculations. MCNP4a was used for reference cases and to determine the maximum HI-20841 75 4-17 Holtec Project 1702 Holtec Report HI-2084175 Holtec Report 4-17 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION reactivity of the spent fuel racks, eccentric fuel positioning, axial burnup distributions, and fuel misloading).

Additionally, complex 3-D geometries were modeled using MCNP4a, e.g. Figure 4.5.8 shows the single cell model used to perform Region 3 calculations. Figure 4.5.2 shows the basic model that was used in the Region 2 calculations and includes fuel assemblies from all three Regions. These pictures were created with the two-dimensional plotter and clearly indicates the explicit modeling of fuel rods in each fuel assembly. In CASMO-4, a single cell is modeled, and since CASMO-4 is a two-dimensional code, the fuel assembly hardware above and below the active fuel length is not represented. See also the discussion in [4.12] regarding MCNP4a and CASMO-4 models. The three-dimensional MCNP4a models that included axial leakage assumed approximately 30 cm of pure water above and below the active fuel length.

Additional models were generated with MCNP4a to investigate the effect of abnormal and normal conditions. These models are discussed in the appropriate section.

4.7.1 Identification of Design Basis Fuel Assembly All of the fuel assembly data provided by FENOC was analyzed as shown in Table 4.5.1 to determine which parameters would be bounding for use as the design basis fuel assembly. Many of the fuel assembly parameters were equivalent. The guide tube thickness varied between fuel assembly types and therefore a minimum thickness was conservatively chosen (see Table 4.5.2).

For all assemblies, the presence of burnable absorbers in the fuel assembly (WABA, IFBA) was neglected for determination of the design basis fuel assembly (see Section 4.7.2 for a discussion of the effect of burnable poison).

The fuel. assembly was conservatively modeled without spacer grids. However, calculations were performed to determine the reactivity effect of spacer grids in borated water. Specifically, the calculations were performed to determine if the spacer grids displace enough boron to cause a significant positive reactivity effect.

Holtec Report HI-2084175 4-18 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION This model is very conservative because it is axially infinite whereas the spacer grids are axially localized and therefore the results overestimate the reactivity effect.

The results of the calculations are shown in Table 4.7.5. For burnups under approximately 40 GWD/MTU, the spacer grid model shows a decrease in reactivity from the reference case. For burnups over approximately 40 GWD/MTU, the spacer grid model begins to show a small positive reactivity effect (approximately 0.0050 delta-kinf) that is magnified by increasing the soluble boron concentration.

This reactivity increase discussed above is due to a slight hardening of the neutron spectrum. As the fuel to moderator ratio increases, the number of fissions in the Pu increases and the number 235U decreases due to the spectral shift. In fuel with a high burnup this of fissions in the increases the reactivity of the fuel assembly because of the greater Pu concentrations. However, since the spacer grids are in reality a localized moderator displacer, the effect shown in Table 4.7.5 is further reduced. Therefore, it is acceptable to neglect the fuel assembly spacer grids for the MZTR racks because they will be primarily dominated by the high reactivity fresh fuel, where the absence of modeling the grid spacers results in negligible differences in reactivity that is more than offset by the soluble boron in the pool that is not credited (i.e. approximately 800 ppm).

4.7.2 Reactivity Effect of Fuel Assembly Reactivity Control Devices The BVPS Unit No. 2 fuel may include the use of IFBA, IFBA with WDRs, WABA or only WDRs. The reactivity effect of these fuel assembly reactivity control devices was investigated in order to determine a bias to be applied to the final calculated reactivity. Bounding cases were selected and the maximum reactivity effect over all cases was applied to the final calculated k~ff as a reactivity control bias (see Table 4.7.6 and Table 4.7.7).

HI-2084175 4-19 Holtec Project 1702 Holtec Report HI-2084175 1-loltec Report 4-19 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Depletion calculations were performed with CASMO-4 for selected configurations of IFBA, WDR and WABA rods based on the specifications in Table 4.5.7, Table 4.5.8 and Table 4.5.9 in full length absorber configurations [4.12]. For the IFBA and WABA cases, using the maximum IFBA loading and maximum number of IFBA pins or WABA rods per assembly is conservative because the reactivity effect increases with the maximum number of IFBA pins or WABA rods and decreasing enrichment. This is because the burnable absorber hardens the spectrum and increases Pu production, making it more reactive than a fuel assembly with no burnable absorber at an equivalent exposure. Therefore, the cases presented bound all possible configurations at BVPS Unit No. 2.

The WABA and WDR rod cases were modeled in the fuel assembly for 17 and 14 GWD/MTU respectively, then removed and not included for any of the rack calculations. The WDR rods were analyzed for multiple cases with and without IFBA pins. The results of these calculations are shown in Table 4.7.8 through Table 4.7.11.

4.7.3 Reactivity Effect of Axial Bumup Distribution Initially, fuel loaded into the reactor will bum with a slightly skewed cosine power distribution.

As burnup progresses, the burnup distribution will tend to flatten, becoming more highly burned in the central regions than in the upper and lower ends. At high burnup, the more reactive fuel near the ends of the fuel assembly (less than average burnup) occurs in regions of lower reactivity worth due to neutron leakage. Consequently, it would be expected that over most of the burnup history, distributed burnup fuel assemblies would exhibit a slightly lower reactivity than that calculated for the average burnup. As burnup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced burnup.

Generic analytic results of the axial burnup effect for assemblies without axial blankets have been provided by Turner [4.9] based upon calculated and measured axial burnup distributions.

These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnup compared to a flat distribution, becoming positive at burnups greater than Holtec Report HI-2084175 4-20 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION about 30 GWD/MTU. The trends observed in 4.9 suggest the possibility of a small positive reactivity effect above 30 GWD/MTU, increasing to slightly over 1% Ak at 40 GWD/MTU. The required burnup for the maximum enrichment is higher than 30 GWD/MTU. Therefore, a positive reactivity effect of the axially distributed burnup is possible.

Calculations are performed with the three axial burnup distributions shown in Table 4.5.6 (see Section 4.5.3) and with an axially constant burnup, and the higher reactivity is used in the analyses. The Region 2 kcff model includes fuel from all three Regions and therefore the axial burnup profile is consistent for fuel in both Region 2 and Region 3.

4.7.4 Isotopic Compositions To perform the criticality evaluation for spent fuel in MCNP4a, the isotopic composition of the fuel is calculated with the depletion code CASMO-4 and then specified as input data into MCNP4a.

4.7.5 Uncertainty in Depletion Calculations CASMO-4 was used for depletion. Since critical experiment data with spent fuel other than in-.

reactor criticals is not available for determining the uncertainty in burnup-dependent reactivity calculations, an allowance for uncertainty in reactivity was assigned as discussed in [4.7]. Based on the recommendation in [4.7], a burnup dependent uncertainty in reactivity for burnup Holtec Report HI-2084175 4-21 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION calculations of 5% of the reactivity decrement is used. A separate calculation is done for both Region 2 and Region 3 for each enrichment case between 2.0 wt% U-235 and 5.0 wt% U-235 in Table 4.7.6a through Table 4.7.6c and Table 4.7.7a through Table 4.7.7c. The calculated reactivity with spent fuel is then subtracted from the reactivity with fresh fuel and a 5%

allowance is statistically combined with the other reactivity allowances in the determination of the maximum keff for normal conditions where assembly burnup is credited.

4.7.6 Eccentric Fuel Assembly Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell. To investigate the potential reactivity effect of eccentric positioning of assemblies in the cells, MCNP4a calculations were performed with two separate cases. Case 1 models all the fuel assemblies eccentrically positioned away from the rack center so that there are four fresh assemblies in the four rack corners at their closest approach. Case 2 models all the fuel assemblies in four-assembly clusters repeated throughout the rack. The results of these.

calculations are presented in Table 4.7.12 and indicate that eccentric fuel positioning results in a decrease in reactivity. No additional calculations are therefore required.

4.7.7 Uncertainties Due to Manufacturing Tolerances In the calculation of the final keff, the effect of manufacturing tolerances on reactivity is included as discussed in Section 4.2. CASMO-4 was used to perform these calculations. As allowed in

[4.7], 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 4.7.14) and tolerances of the fuel dimensions (see Table 4.7.13a and Table 4.7.13b for spent fuel with and without the creep model described in Section 4.5.1).

The calculations are performed for different enrichments (2.0 to 5.0 wt% 235U) at various burnups and with a soluble boron concentration of 0 ppm and 2000 ppm. The tolerance used at Holtec Report HI-2084175 4-22 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION each enrichment case is the maximum over all burnups and soluble boron concentrations for conservatism.

To determine the Ak associated with a specific manufacturing tolerance, the kinf calculated for the reference condition is compared to the kif 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 fuel assembly and storage rack manufacturing tolerances. Only the Ak values in the positive direction (increasing reactivity) were used in the statistical combination. The fuel and rack tolerances included in this analysis are given in Table 4.5.2 and Table 4.5.10. The results of the manufacturing tolerances are shown in Table 4.7.13a, Table 4.7.13b and Table 4.7.14. The tolerance used in the final reactivity determination is shown in Table 4.7.6a through Table 4.7.6c and Table 4.7.7a through Table 4.7.7c.

Additionally, to account for the measurement uncertainty associated with the B- 10 content in the Metamic test coupons, an uncertainty of 5% on the Metamic density is included. The 235U) at various burnups calculations are performed for different enrichments (2.0 to 5.0 wt%

and with soluble boron concentrations of 0 ppm and 2000 ppm. The results of the calculations are shown in Table 4.7.22. The maximum positive reactivity effect is then combined statistically with the other uncertainties in the final reactivity determination is shown in Table 4.7.6a through Table 4.7.6c and Table 4.7.7a through Table 4.7.7c.

4.7.8 Temperature and Water Density Effects Pool water temperature effects on reactivity in the MZTR racks have been calculated with CASMO-4 [4.12]. The calculations are performed for different enrichments (2.0 to 5.0 wt%

235U) at various burnups and with soluble boron concentrations of 0 ppm and 2000 ppm. The results are presented in Table 4.7.15a. The results show that the spent fuel pool temperature Holtec Report HI-2084175 4-23 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION coefficient of reactivity is negative, i.e., a higher temperature results in a lower reactivity over the range of credited soluble boron. Consequently, all CASMO-4 calculations are evaluated at 39.2 'F, which corresponds to a moderator density of 1.0 g/cc.

In MCNP4a, the Doppler treatment and cross-sections may only be valid at 300K (80.33 'F).

Therefore, a Ak is determined in CASMO-4 from 39.2 'F to 80.33 'F, and is included in the final keff calculation as a bias (see Table 4.7.6, Table 4.7.7, and Table 4.7.15b). The calculations are 235U) at various burnups and with soluble performed for different enrichments (2.0 to 5.0 wt%

boron concentrations of 0 ppm and 2000 ppm. The temperature bias calculations using CASMO-4 use a fixed water density (i.e 1.0 g/cc). See also the related discussion in [4.12].

4.7.9 Calculations of Maximum keff Using the calculation model shown in Figure 4.5.2 and Figure 4.5.8, and the design basis fuel assembly specified in Table 4.5.2, the kcalc in the MZTR storage racks has been calculated with MCNP4a. The determination of the maximum kff values, based on,,the formula in Section 4.4.2, 235U and 5.0 wt% 235U for each of the was calculated for initial enrichments between 2.0 wt%

three axial burnup profiles shown in Table 4.5.6. The results show that the maximum keff of the MZTR rack loaded with the maximum number of fresh fuel assemblies is less than 1.0 at a 95%

probability and at a 95% confidence level without credit for soluble boron and less than or equal to 0.95 at a 95% probability and at a 95% confidence level with credit for soluble boron (see Table 4.7.3). Figure 4.5.7 shows a permissible loading configuration for the entire SFP.

Holtec Report HI-2084175 4-24 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION and summarized in Table 4.7.2 and Figure 4.7.2a through Figure 4.7.2c.

composition to the same initial enrichment as the Region 2 fuel and with the calculated bumup (does not include the 5% burnup uncertainty) determined for each axial bumup profile type (see

a through

c).

These final values are presented in

a through

c and summarized in Table 4.7.1 and Figure 4.7.1 a through Figure 4.7.1 c.

4.7.10 Interfaces Within and Between Racks The design basis model is a = array of cells representing all three Regions of the MZTR and also includes the interface of 4 MZTR rack modules at the minimum separation between racks.

The design basis model also considered the Region 1 cells from the adjacent racks to be directly across from each other, thereby considering the worst case. Therefore, no additional calculations involving interfaces within or between racks are required.

4.7.11 Verification of the Initial Burnup and Enrichment Combinations The calculations to determine the Region 2 and Region 3 initial enrichment and burnup combinations considered that the fuel in Region 2 and Region 3 had the same initial enrichment.

Holtec Report HI-2084175 4-25 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION To verify these calculations for all possible initial enrichment combinations, additional calculations were performed. The axial burnup profiles considered were the uniform profile and the no blankets profile (most reactive Region 3 profile). The initial enrichment and burnup combinations are therefore taken from the results in Table 4.7.6c and Table 4.7.7c. The axial burnup profile (segmented or uniform) that produced the highest reactivity in the Region 2 calculations presented in Tables 4.7.1c is then compared in the verification calculations. The verification calculations do not include the 5% burnup uncertainty or the polynomial fit (see Table 4.7.1 and Table 4.7.2). The reference reactivity is the target reactivity of the Region 2 enrichment and burnup combination. The results of these verification calculations are shown in Table 4.7.16 and show that the calculations are acceptable and generally within the statistical uncertainty of the calculations to the maximum reactivity from Region 2 and Region 3 and therefore validate the calculated initial enrichment and burnup combinations.

4.7.12 Soluble Boron Concentration Calculation The SFP storage racks require some amount of soluble boron to meet the acceptance criteria of keff less than or equal to 0.95 (see Section 4.3). In order to calculate the maximum soluble boron requirement required, the MCNP4a cases used for Section 4.7.11 above, i.e. the curve

verification cases, were rerun with 800 ppm soluble boron. The reactivity of the 800 ppm case was then used to interpolate the required soluble boron concentration required to obtain a reactivity of 0.945 minus the maximum total corrections for the given burnup and enrichment case from Table 4.7.6c and Table 4.7.7c. These soluble boron concentrations are conservative since they are determined for the calculated burnups and do not include the 5% burnup uncertainty or the polynomial fit. Additionally, to confirm that the interpolation was conservative, the cases in Table 4.7.16 are re-performed using the interpolated soluble boron concentration. The result is shown in Table 4.7.16 and confirm that using the interpolated soluble boron concentration is conservative. The maximum calculated soluble boron requirement is presented in Table 4.7.3.

4.7.13 Abnormal and Accident Conditions HI-2084175 4-26 Holtec Project 1702 Report HI-2084175 Holtec Report Holtec 4-26 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION The effects on reactivity of credible abnormal and accident conditions are examined in this section. This section identifies which of the credible abnormal or accident conditions will result in exceeding the limiting reactivity (kcff < 0.95). For those accident or abnormal conditions that result in exceeding the limiting reactivity, a minimum soluble boron concentration is determined to ensure that keff < 0.95. The double contingency principle of ANS-8. l/N16.1-1975 4.8 (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. For those cases where the reactivity of the accident is greater than the limit of keff < 0.95, the calculation was re-performed with soluble boron and the concentration required to meet the limit was interpolated.

4.7.13.1 Abnormal Temperature All calculations for the MZTR are performed at a pool temperature of 39.2 'F. As shown in Section 4.7.8 above, the temperature coefficient of reactivity is negative; therefore no additional calculations are required, because a further increase in temperature reduces the reactivity.

4.7.13.2 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 22 inches, which is sufficient to preclude neutron coupling (i.e., an effectively infinite separation). Consequently, the horizontal fuel assembly drop accident will not result in a significant increase in reactivity. Furthermore, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.

4.7.13.3 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 has previously been shown to cause no damage to either fuel assembly but may result in a small deformation of the baseplate Holtec Report HI-2084175 4-27 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION 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.

4.7,13.4 Abnormal Location of a Fuel Assembly 4.7,13.4.1 Misloaded Fresh Fuel Assembly The misloading of a fresh unburned fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (kef of 0.95). This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt% 235U) were to be inadvertently misloaded into a storage cell intended to be used for spent fuel. However, administrative controls make this highly unlikely. For example, on site reactor engineering determines the fuel assembly storage location based on initial enrichment and fuel assembly burnups and establishes refueling procedures for handling of fuel assemblies per Technical Specification requirements. These procedures are controlled by the I OCFR50.59 process. The reactivity consequence of this situation was investigated for various combinations of the Region 2 and Region 3 initial burnup and enrichment combinations determined above. The misloaded fuel assembly model conservatively used reflective boundary conditions, thereby simulating an infinite array of storage racks with misloaded fuel assemblies (see Figure 4.7,4). These results are summarized in Table 4.7.17 for the two relevant Regions of the MZTR. As expected, a fuel assembly placed in a Region 2 cell on the periphery of the rack between two Region 1 (fresh fuel) storage cells is the worst case. The results of the analysis are summarized in Table 4.7,4, including the soluble boron level calculated to be sufficient to ensure that the maximum kif value for this condition remains less than or equal to 0.95.

4.7,13.4.2 Mislocated Fresh Fuel Assembly The mislocation 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 Holtec Report HI-2084175 4-28 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION highest permissible enrichment (5.0 wt% 235U) were to be accidentally mislocated outside of a storage rack adjacent to other fuel assemblies. The pool layout was examined to determine a credible worst case location for this accident, and it was determined to be a junction of two racks that form an empty comer in the SFP.

The MCNP4a model consists of a cut away comer section of a storage rack (the gap between racks is not modeled for conservatism) as shown in Figure 4.7.3. The model uses reflective boundary conditions. A 5.0 wt% 235U unburned assembly is modeled adjacent to two poisoned faces of the rack that are both Region 1 cells to maximize the reactivity effect. The poison thickness along the outer rack edge -adjacent to the mislocated fuel assembly was modeled with half the normal thickness and without steel sheathing for conservatism. This accident case was investigated for various combinations of the Region 2 and Region 3 initial bumup and enrichment combinations determined above. The results of the analysis are listed in Table 4.7.18 and summarized in Table 4.7.4, including the soluble boron level sufficient to ensure that the maximum kcff value for this condition remains less than or equal to 0.95.

4.7.14 Fuel Rod Storage Basket The FRSB is modeled as described in Section 4.5 and shown in Figure 4.5.6, as bare fuel pins with fresh fuel enriched to 5.0 wt % 235U. The FRSB was fully reflected by water (at least 12 inches surrounding on all sides) and reflective boundary conditions. The reactivity of the FRSB model was compared to the reactivity of three separate fuel assemblies with the parameters described in Table 4.5.1 at each of the three burnup cases determined by the MZTR for no blankets fuel (not including the 5% bumup uncertainty): a fresh unburned Region 1 5.0 wt%

case, a Region 2 5.0 wt% fuel assembly with a bumup of 51.67 GWD/MTU case, and a Region 3 5.0 wt% fuel assembly with a bumup of 39.58 GWD/MTU case. The three cases were modeled in the same manner as the FRSB, fully reflected by water and reflective boundary conditions.

The results of these calculations are presented in Table 4.7.20 and show that the reactivity of the FRSB is less than the Region 1, Region 2 and Region 3 cases and therefore may be stored in any location in the MZTR.

Ijoltec Report HI-2084175 4-29 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION 4.7.15 Fuel Elevator The fuel elevator is modeled as described in Section 5, as a fresh 5.0 wt% fuel assembly surrounded by water. The model includes a 5.0 wt% fresh fuel assembly mislocated adjacent to the fuel elevator. The results of the calculations are presented in Table 4.7.24.

4.7.16 Boron Dilution Analysis The SFP contains a Technical Specification required minimum soluble boron concentration of 2000 ppm. Significant loss or dilution of the soluble boron concentration in the SFP is extremely unlikely, however, the guidance presented in [4.7] requires that the boron dilution analysis should determine that sufficient time is available to detect and suppress the worst dilution event that can occur to reduce the boron concentration to the level needed to maintain keff less than or equal to 0.95. Since a boron dilution accident is already an abnormal condition, additional accident conditions such as a misloaded fuel assembly do not need to be considered

[4.8].

The required minimum soluble boron concentration is provided in Table 4.7.3 under normal conditions. The volume of water in the pool is given in Table 4.7.21 along with the volume of water needed to dilute the SFP with a soluble boron concentration of 2000 ppm to the required concentration. The spent fuel pool cannot hold such a volume of water and would overflow.

However, in the event that somehow this volume of water were to dilute the SFP to the required concentration, there would be no criticality consequence since the racks are still qualified for unborated water to a keff of less than 1.0.

4.7.16.1 High Flow Rate Dilution The worst case dilution accident would occur during operator actions to align the Service Water System piping with the SFP. The service water system is capable of providing 3000 gpm of un-borated water to the SFP. The result of the boron dilution analysis is presented in Table 4.7.21.

It can be seen that the SFP volume would increase rapidly and the High Level Alarm Set Point Holtec Report HI-2084175 4-30 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION would be reached in 2.62 minutes. At this point, operators would initiate actions to mitigate the accident. The results show that in order to dilute the SFP to the required concentration (see Table 4.7.3), the Service Water System piping and pumps would need to remain operating and aligned with the SFP for 85 minutes. Upon receiving a SFP high/low level alarm, an operator would respond according to the Alarm Response Procedure and take appropriate action.

4.7.16.2 Low Flow Rate Dilution Small dilution flow around pump seals and valve stems or mis-aligned valves could possibly occur in the normal soluble boron control system or related systems. Such failures might not be immediately detected. These flow rates would be of the order of a 2 gpm maximum and the increased frequency of makeup flow might not be observed. However, an assumed loss flow-rate of 2 gpm dilution flow rate would require approximately 2.7 days in order to reach the SFP high level alarm set point and then an additional 85.8 days to reduce the boron concentration to required concentration (see Table 4.7.3). Since the SFP boron concentration is measured administratively every 7 days, there is ample time to react to both the high level alarm set point and lower soluble boron concentration and take corrective actions.

4.7.17 Rack Movement In the event of seismic activity, there is the possibility that the SFP storage racks may move.

Since the base plate extensions preclude the racks from moving closer together (see Section 4.5.5.1), lateral rack movement could potentially impact reactivity. However, as is discussed in Section 4.5.5, the design basis model is the bounding worst case scenario, with fresh fuel assemblies at their closest proximity to each other across the gap between racks. Therefore, any lateral rack movement effect on reactivity is bounded by the design basis case and no further calculations are required. Should the racks experience lateral rack movement, it is possible that the racks baseplates could collide and the rack tops touch, reducing the average space between the adjacent racks. The reactivity consequence of this accident conditions was modeled by reducing the gap between the racks to a parallel thickness less than a single baseplate extension thickness, and keeping the active length of the fuel lined up between racks. The model uses reflective boundary conditions and therefore the accident condition is conservatively modeled on Holtec Report HI-2084175 4-31 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION all four sides of the rack module at the same time. The same fuel assembly characteristics were used that were modeled in the mislocated accident case (burnup, enrichment and profile). The results of this accident are presented in Table 4.7.23 and summarized in Table 4.7.4.

4.7.18 Interim Configurations During installation of the new racks in BVPS Unit No. 2 SFP, there will be times when both the new racks and the existing racks will be present in the SFP and loaded with fuel at the same time. During these interim configurations it is important to maintain continued criticality safety by ensuring that the fuel in the new and existing racks remains neutronically decoupled. This can be assured by maintaining at least two rows of empty storage locations (approximately 20 inches; 12 inches provides effectively an infinite neutron reflector) between the fuel in the new racks and fuel in the existing racks. These two rows may be in either type of rack or split between the two racks (1 empty row in each rack). This requirement is specified in lieu of performing detailed calculations on interfaces between the new and existing racks. This requirement does not need to be imposed on fuel in racks adjacent to the same type of rack, since this is already addressed in this chapter for the new racks, and the analysis of record for the existing racks.

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SHADED AREAS DENOTE PROPRIETARY INFORMATION 4.8 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] "Lumped Fission Product and Pm148m Cross Sections for MCNP," Holtec Report HI-2033031, Rev 1.

[4.4] 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).

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

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

[4.7] 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.

[4.8] ANS-8.1/N16.1-1975, "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," April 14, 1975.

[4.9] 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.

[4.10] Study of the Effect of Integral Burnable Absorbers for PWR .Burnup Credit,"

.NUREG/CR-6760, ORNL/TM-2000-321, March, 2002.

[4.11 ] ASME Boiler & Pressure Vessel code,Section III, Subsection NB, American Society of Mechanical Engineers, 1995 with Addenda through 1997.

[4.12] HI-2104598R0, "Sensitivity Studies to Support Criticality Analysis Methodology".

[4.13] HI-2094486R0, "MCNP Benchmark Calculations".

[4.14] HI-2094370R0, "CASMO-4 Benchmark For Spent Fuel Pool Criticality Analyses".

Holtec Report HI-2084175 4-33 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.1 Fuel Assembly Specification and Design Basis Fuel Assembly Selection Parameter Value Fuel Region 1 2 2A 3 4A 4B 5A 5B 5C 6A 6B 7A Number of fuel rods 264 264 264 264 264 264 264 264 264 264 264 264 Fuel rod pitch, in. 0.496 0.496 0.496 0.496 0.496 0.496 0.496 0.496 0.496 0.496 0.496 0.496 Active Fuel Length, in. 144 144 144 144 144 144 144 144 144 144 144 144 Pellet diameter, in 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 Cladding ID, in. 0.329 0.329 0.329 0.329 0.329 0.329 0.329 0.329 0.329 0.329 0.329 0.329 Cladding OD, in. 0.374 0.374 0.374 0.374 0.374 0.374 0.374 0.374 0.374 0.374 0.374 0.374 GT/IT ID, in. 0.448 0.448 0.448 0.448 0.448 0.448 0.448 0.448 0.448 0.442 0.442 '0.442 GT/IT OD, in. 0.484 0.484 0.484 0.484 0.484 0.484 0.484 0.484 0.484 0.474 0.474 0.474 Parameter Value Fuel Region 7B 8A 8B 9A 9B 10A 10B 11A 11B 12A 12B 13A Number of fuel rods 264 264 264 264 264 264 264 264 264 264 264 264 Fuel rod pitch, in. 0.496 0.496 0.496 0.496 0.496 0.496 0.496 0.496 0.496 0.496 0.496 0.496 Active Fuel Length, in. 144 144 144 '144 144 144 144 144 144 144 144 144 Pellet diameter, in 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 Cladding ID, in. 0.329 0.329 0.329 0.329 0.329 0.329 0.329 0.329 0.329 0.329 0.329 0.329 Cladding OD, in. 0.374 0.374 0.374 0.374 0.374 0.374 0.374 0.374 0.374 0.374 0.374 0.374 GT/IT ID, in. 0.442 0.442 0.442 0.442 0.442 0.442 0.442 0.442 0.442 0.442 0.442 0.442 GT/IT OD, in. 0.474 0.474 0.474 0.474 0.474 0.474 0.474 0.474 0.474 0.482 0.482 0.482 Note: GT is "Guide Tube" and IT is "Instrument Tube" Parameter Value Parameter Design Basis FA Fuel Region 13B 14A 14B 15A 15B Min Max Number of fuel rods 264 .264 264 264 264 264 264 264 Fuel rod pitch, in. 0.496 0.496 0.496 0.496 0.496 0.496 0.496 0.496 Active Fuel Length, in. 144 144 144 144 144 144 144 144 Pellet diameter, in 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 0.3225 Cladding ID, in. 0.329 0.329 0.329 0.329 0.329 0.329 0.329 0.329 Cladding OD, in. 0.374 0.374 0.374 0.374 0.374 0.374 0.374 0.374 GT/IT ID, in. 0.442 0.442 0.442 0.442 0.442 Note: The guide tube parameters were 0.442 0.448 0.448 GT/IT OD, in. 0.482 0.482 0.482 0.482 0.482 selected to maximize reactivity. 0.474 0.484 0.474 Holtec Report HI-2084175 4-34 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.2 Specification of the Design Basis Fuel Assembly Design Basis Fuel Parameter Assembly Tolerance Number of fuel rods 264 n/a Fuel rod pitch, in.t 0.496 +0.00047,-0.0075 Active Fuel Length, in. 144 n/a Fuel Pellet diameter, in. 0.3225 +1-0.0005 Cladding ID, in. 0.329 +/-0.0015 Cladding OD, in. 0.374 +/-0.0015 Fuel pellet density, g/cc (max) 10.6312 n/a GT/IT ID, in. 0.448 +/-0.002 GT/IT OD, in. 0.474 +/-0.002 235 Enrichment, wt% U (max) 5.00 +0.05 f"The positive tolerance was calculated by dividing the maximum pitch tolerance of 0.0075 by 16.

HI-2084175 Report HI-2084175 4-35 Holtec Project 1702 Holtec Report Holtec 4-35 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.3 Core Operating Parameter for CASMO Depletion Analyses Parameter Value Bounding Cycle Average Soluble Boron 1050 Concentration, ppm Bounding Reactor Specific Power, MW/MTU 40.4 Bounding Core Fuel Temp., 'F 1516 Bounding Core Moderator Temp., 'F 638.33 In-Core Assembly Pitch, Inches 8.466 Holtec Report HI-2084175 4-36 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.4 Calculation of Soluble Boron for CASMO-4 GWD/MTU Predicted ppm 0 1506 0.15 1498 0.5 1486 1 1485 2 1488 3 1466 4 1424 6 1294 8 1126 10 936 12 735 14 530 16 325 18 125 18.8 46 19.8 -50 20 -70 Average 1031 Bounding 1050 Note: A straight average was used, neglecting the last two steps.

HI-2084175 4-37 Holtec Project 1702 Holtec Report HI-2084175 Holtec Report 4-37 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.5 deleted Holtec Report HI-2084175 4-38 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.6 Axial Burnup Profiles Node Enriched Natural No (1 is bottom) Blankets Blankets Blankets 24 0.3495 0.1537 0.4430 23 0.7140 0.6518 0.6803 22 0.8497 0.8259 0.8203 21 0.9634 0.9722 0.9485 20 1.0175 1.0384 1.0108 19 1.0380 1.0550 1.0283 18 1.0601 1.0713 1.0458 17 1.0663 1.1075 1.0801 16 1.0829 1.1129 1.0852 15 1.0700 1.0818 1.0548 14 1.0866 1.1154 1.0928 13 1.0868 1.1159 1.0946 12 1.0750 1.0992 1.0716 11 1.0868 1.1141 1.0912 10 1.0876 1.1163 1.1036 9 1.0854 1.1152 1.0975 8 1.0763 1.1041 1.0704 7 1.0898 1.1203 1.0967 6 1.0879 1.1180 1.0910 5 1.0646 1.0858 1.0471 4 1.0542 1.0722 1.0437 3 0.9935 0.9729 0.9804 2 0.8293 0.7304 0.7983 1 0.4000 0.1746 0.5084 Note: Node size is 6 inches.

4-39 Holtec Project 1702 Holtec Report HI-2084175 Holtec Report HI-2084175 4-39 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.7 IFBA Rod Specification Parameter Value Maximum number of IFBA rods 128 or 200 Max. IFBA Loading, mgB- 2.35 10/inch IFBA thickness, cm 0.00243 4-40 Holtec Project 1702 Holtec Report HI-2084175 Holtec Report HI-2084175 4-40 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.8 Water Displacer Rod Specification Parameter Value Fuel Region - Number of WDR rodlets per assembly WDR Cladding Inner Radius (cm) 0.43688 WDR Cladding Outer Radius (cm) 0.48387 Cladding Material 304 Stainless Steel Fuel Assembly Axial Layout Number of Assemblies 4A - 100 IFBA Rods - 4 WDRs 12 4A - 128 IFBA Rods - No WDRs 4 4A - 128 IFBA Rods - 8 WDRs 4 4B - 32 IFBA Rods - No WDRs 8 4B - 64 IFBA Rods - No WDRs 8 4B -100 IFBA Rods - 4 WDRs 8 4B - 128 IFBA Rods - No WDRs 4 4B - 128 IFBA Rods - 4 WDRs 8 Report HI-2084 175 4-41 Holtec Project 1702 Holtec Report Holtec HI-2084175 4-41 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.9 WABA Specification Fuel Region 2 3 Parameter Value WABA rods in FA with 3.099 wt% U-235 n/a 12,16 WABA rods in FA with 2.602 wt% U-235 12,20 n/a Inner Water Hole Radius (cm) 0.28575 0.28575 Inside Cladding Outer Radius (cm) 0.33909 0.33909 Absorber Inner Radius (cm) 0.35306 0.35306 Absorber Outer Radius (cm) 0.40386 0.40386 Outside Cladding Inner Radius (cm) 0.41783 0.41783 Outside Cladding Outer Radius (cm) 0.48387 0.48387 Cladding Material Zircaloy-4 Zircaloy-4 34 C-B14% 4 C- 14%

Absorber Material (wt%) A120 3 - 86% A120 3 -

86%

Absorber Density (g/cm 3) 2.47436 2.47436 HI-2084175 4-42 Holtec Project 1702 Holtec Report HI-2084175 Holtec Report 4-42 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.10 Storage Cell Specification Parameter Value Tolerance Cell ID, in. 8.8 Cell Wall thickness, in. 0.075 Cell Pitch, in. t 9.03 n/a Inner Sheathing Thickness, in. 0.035 Poison Thickness, in. 0.106 Poison Width, in. 7.5 Poison Gap, (min) in. 0.112 n/a Baseplate to Absorber Distance, in. 2.875 +0.125 Gap Between Racks, in. 11 n/n Metamic B4C Loading,( )%

Metamic Density, g/cc Metamic B-10, % ~1*-

Metamic B-11, %

1--

Metamic C, %

Metamic Al, %

t In the CASMO-4 Model used the cell pitch tolerance is accounted for in the cell I.D. tolerance.

Note: For the Metamic parameters the nominal and minimum values are with respect to the B4C loading. Therefore, the minimum B4C loading results in a higher density. Also, only the minimum tolerance is used since this results in an increase in reactivity.

HI-2084175 4-43 Holtec Project 1702 Report HI-2084175 Holtec Report Holtec 4-43 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.11 Specification of the Fuel Rod Storage Basket Parameter Value Number of Cells 52 Cell Pitch 0.937 inches Array Type 8x8 Basket Wall Thickness t 0.035 inches t Conservatively neglected in the analysis.

HI-2084175 4-44 Holtec Project 1702 Holtec Report HI-2084175 4-44 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.12 Specification for the Boron Dilution Analysis Parameter Value Credited Volume of SFP 182,472 gal Service Water System 3000 gpm Volume to High Level Alarm Set gal pointt 1__7850 Calculated as 1 ft of SFP volume.

HI-2084175 4-45 Holtec Project 1702 Report HI-2084175 Holtec Report 4-45 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.5.13 Specification of the Fuel Elevator Parameter Value Number of Cells 1 Cell ID (inches) 8.776 Cell Wall Thickness (inches) 0.75 HI-2084175 Report HI-2084175 4-46 Holtec Project 1702 Holtec Report Holtec 4-46 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.1 Summary of the Region 2 Initial Enrichment and Burnup Combinations Calculated Calculated Burnup Burnup (GWD/MTU)

Enrichment (GWD/MTU) + 5% and (wt% U-235) + 5% polynomial fit Enriched Blankets 2.0 10.23 11.14 2.5 19.14 19.53 3.0 25.66 27.50 3.5 35.06 35.06 4.0 41.29 42.20 4.5 47.82 48.92 5.0 54.49 55.23 BU=-0.8324x^2+20.523x-26.578 Natural Blankets 2.0 9.69 10.99 2.5 19.28 19.27 3.0 26.89 27.13 3.5 33.55 34.56 4.0 40.71 41.56 4.5 46.74 48.14 5.0 54.11 54.28 BU=-0.8553x^2+20.418x-26.425 No Blankets 2.0 10.13 11.19 2.5 19.07 19.46 3.0 27.12 27.29 3.5 34.13 34.69 4.0 40.79 41.65 4.5 46.52 48.17 5.0 54.26 54.26 BU=-0.873 1xA2+20.467x-26.25 HI-2084175 4-47 Holtec Project 1702 Hoitec Report HI-2084175 Holtec Report 4-47 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.2 Summary of the Region 3 Initial Enrichment and Bumup Combinations Calculated Calculated Burnup Burnup (GWD/MTU)

Enrichment (GWD/MTU) + 5% and (wt% U-235) + 5% polynomial fit Enriched Blankets 2.0 0.00 0.64 2.5 8.02 8.02 3.0 14.84 14.99 3.5 20.81 21.57 4.0 27.33 27.76 4.5 33.18 33.55 5.0 38.62 38.94 BU=-0.793x^2+18.315x-32.814 Natural Blankets 2.0 0.00 0.65 2.5 8.06 8.06 3.0 14.64 15.06 3.5 21.22 21.66 4.0 27.28 27.85 4.5 33.19 33.63 5.0 38.71 39.01 BU=-0.8134x^2++/-18.481x-33.0634 No Blankets 2.0 0.00 1.03 2.5 8.17 8.17 3.0 14.73 15.19 3.5 21.36 22.08 4.0 27.85 28.84 4.5 34.81 35.47 5.0 41.56 41.97 BU=-0.2574x^2+1 5.449x-28.84 1-11-2084175 4-48 Holtec Project 1702 Holtec Report 141-2084175 Holtec Report 4-48 Holtec Project 1'702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.3 Summary of the MTZR Soluble Boron Requirements For Normal Conditions Parameter Value Soluble Boron Requirement, ppm 495 Report HI-2084 175 4-49 Holtec Project 1702 Holtec Report Holtec HI-2084175 4-49 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.4 Summary of Accident Cases Soluble Boron Case Requirement (ppm)

Mislocated Fresh Fuel Assembly 1186 Misloaded Fresh Fuel Assembly in Region 2 (outer row) 1212 Misloaded Fresh Fuel Assembly in Region 2 (inner row) 1020 Misloaded Fresh Fuel Assembly in Region 3 867 Rack Displacement Accident 940 HI-2084175 4-50 Holtec Project 1702 Holtec Report HI-2084175 Holtec Report 4-50 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.5 CASMO-4 Calcuation of the Effect of Spacer Grids and Boron Concentration on Reactivity Enrichment 2 3.5 5 wt% U-235 Burnup (GWD/ Boron SpacerGrid Spacer Grid MM (ppm) ReferenceCase SpacerGrid Case Delta-k Reference Case Case Delta-k Reference Case Case Delta-k 400 0.89851 0.89131 -0.0072 1.06102 1.05182 -0.0092 1.14916 1.13908 -0.0101 800 0.83896 0.83453 -0.0044 1.00852 1.00189 -0.0066 1.10298 1.09522 -0.0078 1000 0.81245 0.80915 -0.0033 0.9845 0.97897 -0.0055 1.08153 1.0748 -0.0067 0 1200 0.7878 0.78549 -0.0023 0.96179 0.95727 -0.0045 1.06109 1.05531 -0.0058 1400 0.76483 0.76339 -0.0014 0.94029 0.93669 -0.0036 1.04157 1.03666 -0.0049 1600 0.74337 0.74271 -0.0007 0.91991 0.91713 -0.0028 1.02291 1.01881 -0.0041 1800 0.72327 0.72329 0.0000 0.90055 0.89854 -0.0020 1.00506 1.00171 -0.0033 2000 0.7044 0.70504 0.0006 0.88215 0.88083 -0.0013 0.98796 0.98531 -0.0026 400 0.77988 0.78039 0.0005 0.91468 0.90971 -0.0050 1.00865 1.00066 -0.0080 800 0.73608 0.73895 0.0029 0.87224 0.86973 -0.0025 0.96937 0.96368 -0.0057 1000 0.71649 0.72032 0.0038 0.85285 0.8514 -0.0014 0.95117 0.9465 -0.0047 20 1200 0.69823 0.7029 0.0047 0.83456 0.83404 -0.0005 0.93385 0.93011 -0.0037 1400 0.68115 0.68656 0.0054 0.81725 0.8176 0.0003 0.91733 0.91446 -0.0029 1600 0.66514 0.6712 0.0061 0.80085 0.80198 0.0011 0.90156 0.89948 -0.0021 1800 0.65011 0.65674 0.0066 0.78529 0.78713 0.0018 0.88649 0.88515 -0.0013 2000 0.63596 0.6431 0.0071 0.77051 03773 0.0025 0.87206 0.87i41 -0.0006 400 0.80457 0.80595 0.0014 0.89881 0.89515 -0.0037 800 0.76696 0.77061 0.0037 0.86268 0.8613 -0.0014 1000 0.74986 0.75446 0.0046 0.84603 0.84564 -0.0004 40 1200 0.73375 0.73921 0.0055 0.83022 0.83074 0.0005 1400 0.71856 0.72478 0.0062 0.81519 0.81653 0.0013 1600 0.70419 0.71109 0.0069 0.80087 0.80297 0.0021 1800 0.69059 0.6981 0.0075 0.78722 0.79002 0.0028 2000 0.67768 0.68575 0.0081 0.77419 0.77762 0.0034 400 0.80404 0.80609 0.0021 800 0.77056 0.77477 0.0042 1000 0.7552 0.76034 0.0051 60 1200 0.74066 0.74663 0.0060 1400 0.72688 0.7336 0.0067 1600 0.71378 0.72118 0.0074 1800 0.70132 0.70935 0.0080 2000 0.68946 0.69804 0.0086 Holtec Report 1HI-2084175 4-51 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.6a Calculation of the Initial Enrichment and Bumup Combinations for Region 2, Enriched Blankets Profile Enrichment (wt% U-235) 2.0 2.5 1 3.0 1 3.5 4.0 1 4.5 1 5.0 Bumup (GWD/MTU) 10.0 20.0 25.0 35.0 40.0 50.0 55.0 Reactivity Uniform Profile 0.9690 0.9648 0.9669 0.9647 0.9662 0.9586 0.9619 Reactivity Segmented Profile 0.9661 0.9625 0.9652 0.9594 0.9612 0.9540 0.9568 Max Reactivity 0.9690 0.9648 0.9669 0.9647 0.9662 0.9586 0.9619 Burnup (GWD/MTU) 5.0 15.0 20.0 30.0 35.0 45.0 50.0 Reactivity Uniform Profile 0.9821 0.9768 0.9800 0.9741 0.9740 0.9675 0.9685 Reactivity Segmented Profile 0.9813 0.9754 0.9786 0.9684 0.9715 0.9630 0.9639 Max Reactivity 0.9821 0.9768 0.9800 0.9741 0.9740 0.9675 0.9685 Manufacturing Tolerances UncertaintyW 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 Fuel Tolerances Uncertaintyf 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 Metamic Measurement Uncertainty" 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 CASMO-4 Bias Uncertainty 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 Calculation Uncertainty (o) 0.0007 0.0006 0.0007 0.0007 0.0007 0.0006. 0.0007 Calculation Uncertainty (2y) 0.0014 0.0012 0.0014 0.0014 0.0014 0.0012 0.0014 Depletion Uncertainty keff 0.9969 1.0304 1.0589 1.0846 1.1032 1.1228 1.1396 Depletion y 0.0007 0.0007 0.0007 0.0007 0.0007 0.0006 0.0007 Depletion Uncertainty 0.0034 0.0051 0.0066 0.0080 0.0088 0.0099 0.0109 LFP Uncertainty keff 0.9681 0.9685 0.9731 0.9687 0.9706 0.9681 0.9697 LFP y 0.0006 0.0007 0.0006 0.0007 0.0007 0.0007 0.0005 LFP Uncertainty 0.0017 0.0024 0.0028 0.0026 0.0026 0.0033 0.0029 Eccentric Fuel Positioning neg neg neg neg neg neg neg MCNP Code Uncertainty 0.0086 0.0086 0.0086 0.0086 0.0086 0.0086 0.0086 TotalUncertainty(statistiticalcombination) 0.0134 0.0141 0.0147 0.0154 0.0158 0.0166 0.0171 Code Bias 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 Temperature Biast 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 Reactivity Control Biast- 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 Total Corrections 0.0253 0.0260 0.0266 0.0273 0.0277 0.0285 0.0290 Maximum keff 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 Target k-eff(0.995-corrections) 0.9697 0.9690 0.9684 0.9677 0.9673 0.9665 0.9660 Calculated Burnup (GWD/MTU) 9.75 18.23 24.44 33.39 39.32 45.54 51.90 Calculated Burnup (GWD/MTU) + 5% 10.23 19.14 25.66 35.06 41.29 47.82 54.49 BU=-0.8324x^2+20.523x-26.578 11.14 19.53 27.50 35.06 42.20 48.92 55.23

ýt These values are the maximum from the entire burnup and enrichment range.. . . . . _

Holtec Report HI-2084175 4-52 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.6b Calculation of the Initial Enrichment and Burnup Combinations for Region 2, Natural Blankets Profile Enrichment (wt% U-235) 2.0 2.5 13.0 13.5 14.0 4.5 5.0 Burnup (GWD/MTU) 10.0 20.0 30.0 35.0 40.0 45.0 55.0 Reactivity Uniform Profile 0.9664 0.9651 0.9602 0.9613 0.9647 0.9652 0.9592 Reactivity Segmented Profile 0.9672 0.9604 0.9558 0.9564 0.9587 0.9600 0.9533 Max Reactivity. 0.9672 0.9651 0.9602 0.9613 0.9647 0.9652 0.9592 Burnup (GWD/MTU) 5.0 15.0 25.0 30.0 35.0 40.0 50.0 Reactivity Uniform Profile 0.9830 0.9770 0.9693 0.9716 0.9745 0.9779 0.9689 Reactivity Segmented Profile 0.9810 0.9732 0.9649 0.9670 0.9702 0.9694 0.9609 Max Reactivity 0.9830 0.9770 0.9693 0.9716 0.9745 0.9779 0.9689 Manufacturing Tolerances Uncertaintyt 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 Fuel Tolerances Uncertaintyf 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 Metamic Measurement Uncertaintyt 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 CASMO-4 Bias Uncertainty 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 Calculation Uncertainty (a) 0.0006 0.0007 0.0007 0.0007 0.0007 0.0008 0.0007 Calculation Uncertainty (2a) 0.0012 0.0014 0.0014 0.0014 0.0014 0.0016 0.0014 Depletion Uncertainty keff 0.9976 1.0312 1.0598 1.0821 1.1040 1.1213 1.1379 Depletion a 0.0007 0.0007 0.0007 0.0007 0.0007 0.0007 0.0007 Depletion Uncertainty 0.0034 0.0053 0.0070 0.0080 0.0089 0.0099 0.0109 LFP Uncertainty keff 0.9685 0.9672 0.9662 0.9703 0.9729 0.9749 0.9678 LFP a 0.0007 0.0006 0.0006 0.0006 0.0006 0.0007 0.0006 LFP Uncertainty '0.0020 0.0022 0.0027 0.0032 0.0031 0.0036 0.0031 Eccentric Fuel Positioning neg neg neg neg neg neg neg MCNP Code Uncertainty 0.0086 0.0086' 0.0086 0.0086 0.0086 0.0086 0.0086 Total Uncertainty (statistitical combination) 0.0135 0.0141 0.0149 0.0155 0.0160 0.0167 0.0172 Code Bias 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 Temperature Biast 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 Reactivity Control Biast 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 Total Corrections 0.0254 0.0260 0.0268 0.0274 0.0279 0.0286 0.0291 Maximum keff 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 Target k-eff.(0.995-corrections) 0.9696 0.9690 0.9682 0.9676 0.9671 0.9664 0.9659 Calculated Burnup (GWD/MTU) 9.23 18.36 25.61 31.95 38.77 44.52 51.53 Calculated Bumup (GWD/MTU) + 5% 9.69 19.28 26.89 33.55 40.71 46.74 54.11 BU=-0.8553xA2+20.418x-26.425 10.99 19.27 27.13 34.56 41.56 48.14 54.28 These values are the maximum from the entire burnup and enrichment range.

Ho-tc-Rpor 247 ---- 4-53....... Hote Proec 1702...

Holtec Report HI-2084175 4-53 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.6c Calculation of the Initial Enrichment and Bumup Combinations for Region 2, No Blankets Profile Enrichment (wt% U-235) 2.0 2.5 13.0 3.5 14.0 4.5 5.0 Burnup (GWD/MTU) 10.0 20.0 30.0 35.0 40.0 45.0 55.0 Reactivity Uniform Profile 0.9688 0.9649 0.9585 0.9633 0.9649 0.9648 0.9577 Reactivity Segmented Profile 0.9676 0.9623 0.9577 0.9613 0.9620 0.9652 0.9616 Max Reactivity 0.9688 0.9649 0.9585 0.9633 0.9649 0.9652 0.9616 Burnup (GWD/MTU) 5.0 15.0 25.0 30.0 35.0 40.0 50.0 Reactivity Uniform Profile 0.9811 0.9758 0.9703 0.9721 0.9743 0.9756 0.9682 Reactivity Segmented Profile 0.9805 0.9738 0.9668 0.9699 0.9738 0.9756 0.9682 Max Reactivity 0.9811 0.9758 0.9703 0.9721 0.9743 0.9756 0.9682 Manufacturing Tolerances Uncertainty' 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 Fuel Tolerances Uncertainty" 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 Metamic Measurement Uncertaintyf 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 CASMO-4 Bias Uncertainty 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 Calculation Uncertainty (a) 0.0006 0.0007 0.0006 0.0007 0.0007 0.0007 0.0007 Calculation Uncertainty (2y) 0.0012 0.0014 0.0012 0.0014 0.0014 0.0014 0.0014 Depletion Uncertainty keff 0.9978 1.0305 1.0594 1.0841 1.1030 1.1214 1.1385 Depletion a 0.0008 0.0007 0.0006 0.0007 0.0008 0.0007 0.0007 Depletion Uncertainty 0.0035 0.0053 0.0067 0.0080 0.0090 0.0098 0.0108 LFP Uncertainty keff 0.9689 0.9698 0.9644 0.9679 0.9724 0.9727 0.9689 LFP y 0.0006 0.0007 0.0006 0.0007 0.0007 0.0006 0.0007 LFP Uncertainty 0.0017 0.0027 0.0026 0.0027 0.0031 0.0030 0.0031 Eccentric Fuel Positioning neg neg neg neg neg neg neg MCNP Code Uncertainty 0.0086 0.0086 0.0086 0.0086 0.0086 0.0086 0.0086 Total Uncertainty (statistitical combination) 0.0134 0.0142 0.0148 0.0154 0.0160 0.0165 0.0171 Code Bias 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 Temperature Biast 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 Reactivity Control Biast 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 Total Corrections 0.0253 0.0261 0.0267 0.0273 0.0279 0.0284 0.0290 Maximum keff 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 Target k-eff(0.995-corrections) 0.9697 0.9689 0.9683 0.9677 0.9671 0.9666 0.9660 Calculated Burnup (GWD/MTU) 9.65 18.16 25.83 32.51 38.85 44.31 51.67 Calculated Bt-nup (GWD/MTU) + 5% 10.13 19.07 27.12 34.13 40.79 46.52 54.26 BU=-0.8731 x^2+20.467x-26.25 11.19 19.46 27.29 34.69 41.65 48.17 54.26 it These values are the maximum from the entire burnup and enrichment range.

Holtc RportHI-08415 454 Hlte Proect170 Holtec Repoift HI-2084175 4-54 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.7a Calculation of the Initial Enrichment and Bumup Combinations for Region 3, Enriched Blankets Profile Enrichment (wt/o U-235) 2.0 2.5 3.0 13.5 4.0 4.5 5.0 Bumup (GWD/MTU) 5.0 10.0 15.0 20.0 30.0 35.0 40.0 Reactivity Uniform Profile 0.9290 0.9472 0.9591 0.9626 0.9381 0.9429 0.9431 Reactivity Segmented Profile 0.9280 0.9455 0.9541 0.9600 0.9340 0.9403 0.9411 Max Reactivity 0.9290 0.9472 0.9591 0.9626 0.9381 0.9429 0.9431 Burnup (GWD/MTU) 0.0 5.0 10.0 15.0 25.0 30.0 35.0 Reactivity Uniform Profile 0.9651 0.9865 0.9925 0.9967 0.9679 0.9698 0.9697 Reactivity Segmented Profile 0.9651 0.9853 0.9898 0.9917 0.9639 0.9638 0.9650 Max Reactivity 0.9651 0.9865 0.9925 0.9967 0.9679 0.9698 0.9697 Manufacturing Tolerances Uncertaintyf 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 Fuel Tolerances Uncertaintyt 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 Metamic Measurement UncertaintyW 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 CASMO-4 Bias Uncertainty 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 Calculation Uncertainty (a) 0.0006 0.0006 0.0006 0.0006 0.0007 0.0006 0.0007 Calculation Uncertainty (2() 0.0012 0.0012 0.0012 0.0012 0.0014 0.0012 0.0014 Depletion Uncertainty keff 0.9651 1.0291 1.0775 1.1172 1.1482 1.1744 1.1982 Depletion y 0.0006 0.0006 0.0006 0.0006 0.0007 0.0007 0.0006 Depletion Uncertainty 0.0035 0.0058 0.0076 0.0094 0.0125 0.0134 0.0146 LFP Uncertainty keff 0.9318 0.9523 0.9652 0.9721 0.9514 0.9552 0.9572 LFP a 0.0006 0.0005 0.0007 0.0006 0.0006 0.0006 0.0006 LFP Uncertainty 0.0021 0.0023 0.0028 0.0031 0.0038 0.0035 0.0040 Eccentric Fuel Positioning neg neg neg neg neg neg neg MCNP Code Uncertainty 0.0086 0.0086 0.0086 0.0086 0.0086 0.0086 0.0086 Total Uncertainty (statistitical combination) 0.0135 0.0143 0.0152 0.0163 0.0183 0.0189 0.0199 Code Bias 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 Temperature Biast 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 Reactivity Control Biast 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 TotalCorrections 0.0254 0.0262 0.0271 0.0282 0.0302 0.0308 0.0318 Maximum keff 0.9920 0.9920 0.9920 0.9920 0.9920 0.9920 0.9920 Target k-eff(0.995-corrections) 0.9666 0.9658 0.9649 0.9638 0.9618 0.9612 0.9602 Calculated Burnup (GWD/MTU) -0.21 7.63 14.13 19.82 26.03 31.60 36.78 Calculated Burnup (GWD/MTU) + 5% Oft 8.02 14.84 20.81 27.33 33.18 38.62 BU=-0.793x^2+18.315x-32.814 0.64 8.02 14.99 21.57 27.76 33.55 38.94 it These values are the maximum from the entire burnup and enrichment range.

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Projec 1702 Holtec Report HI-2084175 4-55 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.7b Calculation of the Initial Enrichment and Burnup Combinations for Region 3, Natural Blankets Profile Enrichment (wt% U-235) 2.0 12.5 3.0 13.5 14.0 4.5 5.0 Burnup (GWD/MTU) 5.0 10.0 15.0 25.0 30.0 35.0 40.0 Reactivity Uniform Profile 0.9285 0.9473 0.9571 0.9326 0.9393 0.9428 0.9427 Reactivity Segmented Profile 0.9256 0.9420 0.9503 0.9249 0.9304 0.9347 0.9355 Max Reactivity 0.9285 0.9473 0.9571 0.9326 0.9393 0.9428 0.9427 Burnup (GWD/MTU) 0.0 5.0 10.0 20.0 25.0 30.0 35.0 Reactivity Uniform Profile 0.9623 0.9871 0.9930 0.9640 0.9673 0.9696 0.9704 Reactivity Segmented Profile 0.9623 0.9833 0.9882 0.9553 0.9586 0.9588 0.9608 Max Reactivity 0.9623 0.9871 0.9930 0.9640 0.9673 0.9696 0.9704 Manufacturing Tolerances Uncertaintyt 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 Fuel Tolerances Uncertaintyt 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 Metamic Measurement Uncertaintyt 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 CASMO-4 Bias Uncertainty 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 Calculation Uncertainty (a) 0.0007 0.0006 0.0006 0.0007 0.0007 0.0007 0.0007 Calculation Uncertainty (2a) 0.0014 0.0012 0.0012 0.0014 0.0014 0.0014 0.0014 Depletion Uncertainty keff 0.9623 1.0277 1.0785 1.1175 1.1483 1.1759 1.1973 Depletion a 0.0006 0.0006 0.0007 0.0007 0.0007 0.0007 0.0007 Depletion Uncertainty 0.0035 0.0057 0.0079 0.0112 0.0124 0.0136 0.0147 LFP Uncertainty keff 0.9316 0.9513 0.9645 0.9425 0.9502 0.9545 0.9582 LFP ( 0.0006 0.0006 0.0007 0.0006 0.0007 0.0007 0.0007 LFP Uncertainty 0.0023 0.0023 0.0030 0.0033 0.0036 0.0037 0.0043 Eccentric Fuel Positioning neg neg neg neg neg neg neg MCNP Code Uncertainty 0.0086 0.0086 0.0086 0.0086 0.0086 0.0086 0.0086 Total Uncertainty (statistitical combination) 0.0136 0.0143 0.0154 0.0174 0.0183 0.0191 0.0200 Code Bias 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 Temperature Biast 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 Reactivity Control Biast 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 TotalCorrections 0.0255 0.0262 0.0273 0.0293 0.0302 0.0310 0.0319 Maximum keff 0.9920 0.9920 0.9920 0.9920 0.9920 0.9920 0.9920 Target k-eff(0.995-corrections) 0.9665 0.9658 0.9647 0.9627 0.9618 0.9610 0.9601 Calculated Burnup (GWD/MTU) -0.63 7.67 13.94 20.21 25.98 31.61 36.86 Calculated Burnup (GWD/MTU) + 5% Ott 8.06 14.64 2.1.22 27.28 33.19 38.71 BU=-0.8134x^2++/-18.48 lx-33.067 0.65 8.06 15.06 21.66 27.85 33.63 39.01 it These values are the maximumt from the entire burnup and enrichment range. a ttThis value was. setto zero.. ..

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Holtec Report HI-2084175 4-56 Holtee Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.7c Calculation of the Initial Enrichment and Burnup Combinations for Region 3, No Blankets Profile Enrichment (wt% U-235) 2.0 2.5 13.0 13.5 14.0 4.5 5.0 Burnup (GWD/MTU) 5.0 10.0 15.0 25.0 30.0 35.0 40.0 Input File UniformAxialProflle 3n205 3n2510 3n3015 3n3525 3n4030 3n4535 3n5040 Reactivity Uniform Profile 0.9296 0.9486 0.9581 0.9329 0.9390 0.9421 0.9439 Input File Segmented AxialProflle 3m205 '3m2510 "3m3015 3m3525 T3m4030 3m4535 "3m5040 Reactivity Segmented Profile 0.9269 0.9453 0.9564 0.9383 0.9461 0.9542 0.9592 Max Reactivity 0.9296 0.9486 0.9581 0.9383 0.9461 0.9542 0.9592 Burnup (GWD/MTU) 0.0 5.0 10.0 20.0 25.0 30.0 35.0 Input File Uniform Axial Profile 3p2O5 3n255 3n3010 3n3520 3n4025 3n4530 3n5035 Reactivity Uniform Profile 0.9636 0.9874 0.9923 0.9637 0.9681 0.9691 0.9697 Input File Segmented Axial Profile 3p2O5 3m255 3m3010 3m3520 '3m4025 3m4530 3m5035 Reactivity Segmented Profile 0.9636 0.9844 0.9903 0.9648 0.9692 0.9746 0.9794 Max Reactivity 0.9636 0.9874 0.9923 0.9648 0.9692 0.9746 0.9794 Mannfacnirine Tolerance- I Jncertaintvt 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 0.0037 Fuel Tolerance- I ncertaintvy 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 0.0079 Metamic Measurement I Ineertaintvt 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 0.0028 CASMO-4 Bias I Incertaintv 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 Calcnlation I Ineertaintv (a) 0.0006 0.0006 0.0007 0.0007 0.0007 0.0006 0.0006 Calcilation IJncertaintv (2a) 0.0012 0.0012 0.0014 0.0014 0.0014 0.0012 0.0012 Denletion 1Jncertaintv Innut File 3p205 3p 2 510 3p3015 3o3525 3o4030 3o4535 3o5040 Denletion 1lneertaintv keff 0.9636 1.0293 1.0773 1.1167 1.1487 1.1752 1.1984 Denletion a 0.0006 0.0006 0.0006 0.0007 0.0006 0.0006 0.0007

enletion I Incertaintv 0.0034 0.0057 0.0078 0.0109 0.0120 0.0127 0.0138 IFPI JncertaintvlnnitFile 3r205 3r2510 3r3015 303525 304030 3q4535 3q5040

,FP llncertaintvkeff 0.9316 0.9531 0.9643 0.9457 0.9560 0.9641 0.9701

,FPpa 0.0006 0.0007 0.0007 0.0006 0.0007 0.0005 0.0007 1,FP I Jncertaintv 0.0020 0.0025 0.0029 0.0030 0.0035 0.0030 0.0035 Eccentric Fuel Positionin2 neg neg neg neg neg neg neg MCNP Code I Jnrertaintv 0.0086 0.0086 0.0086 0.0086 0.0086 0.0086 0.0086 Totail I Jncertaintv (statistitialeombinationn 0.0135 0.0143 0.0153 0.0171 0.0179 0.0184 0.0192 Code Bias 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 0.0013 Temperature Biast 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 Reactivity Control Biast 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 0.0063 Total Corrections 0.0254 0.0262 0.0272 0.0290 0.0298 0.0303 0.0311 Maximum keff 0.9920 0.9920 0.9920 0.9920 0.9920 0.9920 0.9920 Target k-eff (0.995-corrections) 0.9666 0.9658 0.9648 0.9630 0.9622 0.9617 0.9609, Calculated Burnup (GWD/MTU) -0.45 7.78 14.03 20.35 26.52 33.15 39.58 Calculated Bumup (GWD/MTU) + 5% Ott 8.17 14.73 21.36 27.85 34.81 41.56 BU=-0.2574x-2+15.449x-28.84 1.03 8.17 15.19 22.08 28.84 35.47 41.97 t These values are the maximum from the entire burnup and enrichment range.

1 ttThis value was set to zero.

Holtec Report HI-2084175 4-57 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.8 Results of the CASMO-4 Calculations of the IFBA Bias IFBA Pins 0 0 0 100 100 128 128 200 200 max delta

t105 30 38 50 30 50 30 50 38 50 kinf gwd/mtu kinf kinf kinf kinf delta kinf kinf delta kinf kinf delta kinf kinf delta kinf kinf delta kinf kinf delta kinf max delta kinf I I kinf 0.0 1.0773 1.1353 1.1954 0.8913 -0.1860 1.0446 -0.1508 0.8429 -0.2344 1.0003 -0.1951 0.8402 -0.2951 0.9274 -0.2681 -0.1508 0.1 1.0748 1.1331 1.1936 0.8919 -0.1829 1.0444 -0.1492 0.8443 -0.2305 1.0006 -0.1930 0.8420 -0.2911 0.9284 -0.2652 -0.1492 2.0 1.0556 1.1127 1-1726 0.9253 -0.1303 1.0530 -0.1197 0.8922 -0.1634 1.0185 -0.1541 0.8909 -0.2218 0.9594 -0.2133 -0.1197 4.0 1.0388 1.0964 1.1576 0.9490 -0.0898 1.0624 -0.0951 0.9270 -0.1118 1.0358 -0.1217 0.9309 -0.1655 0.9878 -0.1698 -0.0898 6.0 1.0218 1.0798 1.1423 0.9614 -0.0604 1.0673 -0.0750 0.9474 -0.0744 1.0471 -0.0952 0.9584 -0.1214 1.0085 -0.1338 -0.0604 8.0 1.0054 1.0637 1.1274 0.9659 -0.0395 1.0688 -0.0586 0.9573 -0.0480 1.0537 -0.0737 0.9763 -0.0875 1.0229 -0.1044 -0.0395 10.0 0.9897 1.0483 1.1129 0.9647 -0.0250 1.0676 -0.0453 0.9598 -0.0299 1.0565 -0.0564 0.9866 -0.0617 1.0323 -0.0806 -0.0250 11.0 0.9821 1.0408 1.1059 0.9626 -0.0195 1.0662 -0.0397 0.9590 -0.0231 1.0567 -0.0492 0.9894 -0.0514 1.0354 -0.0705 -0.0195 12.5' 0.9711 1.0300 1.0956 0.9580 -0.0131 1.0632 -0.0324 0.9558 -0.0153 1.0558 -0.0398 0.9913 -0.0386 1.0382 -0.0574 -0.0131 15.0 0.9535 1.0125 1.0791 0.9475 -0.0059 1.0563 -0.0228 0.9469 -0.0065 1.0516 -0.0275 0.9896 -0.0229 1.0391 -0.0400 -0.0059 17.5 0.9365 0.9956 1.0631 0.9347 -0.0017 1.0475 -0.0156 0.9350 -0.0015 1.0446 -0.0185 0.9831 -0.0125 1.0359 -0.0272 -0.0015 20.0 0.9201 0.9792 1.0476 0.9208 0.0007 1.0372 -0.0104 0.9214 0.0014 1.0355 -0.0121 0.9735 -0.0057 1.0297 -0.0179 0.0014 22.5 0.9042 0.9634 1.0326 0.9063 0.0021 1.0259 -0.0066 0.9072 0.0030 1.0251 -0.0075 0.9620 -0.0013 1.0213 -0.0112 0.0030 25.0 0.8890 0.9480 1.0179 0.8918 0.0029 1.0139 -0.0040 0.8928 0.0038 1.0136 -0.0042 0.9494 0.0014 1.0114 -0.0065 0.0038 27.5 0.8743 0.9330 1.0035 0.8776 0.0033 1.0014 -0.0021 0.8786 0.0043 1.0015 -0.0020 0.9362 0.0032 1.0004 -0.0031 0.0043 30.0 0.8602 0.9183 0.9893 0.8636 0.0034 0.9886 -0.0007 0.8646 0.0044 0.9889 -0.0004 0.9226 0.0043 0.9885 -0.0008 0.0044 32.5 0.8467 0.9040 0.9753 0.8502 0.0035 0.9756 0.0002 0.8512 0.0045 0.9760 0.0007 0.9090 0.0050 0.9762 0.0009 0.0050 35.0 0.8338 0.8901 0.9616 0.8373 0.0035 0.9625 0.0009 0.8383 0.0045 0.9631 0.0014 0.8955 0.0054 0.9637 0.0021 0.0054 37.5 0.8215 0.8766 0.9481 0.8249 0.0035 0.9495 0.0013 0.8259 0.0044 0.9501 0.0020 0.8822 0.0056 0.9510 0.0029 0.0056 40.0 0.8098 0.8635 0.9348 0.8132 0.0033 0.9365 0.0017 0.8141 0.0043 0.9372 0.0023 0.8692 0.0057 0.9383 0.0035 0.0057 42.5 0.7988 0.8508 0.9218 0.8020 0.0032 0.9237 0.0019 0.8029 0.0041 0.9243 0.0026 0.8566 0.0058 0.9256 0.0039 0.0058 45.0 0.7884 0.8386 0.9089 0.7915 0.0031 0.9110 0.0021 0.7924 0.0040 0.9117 0.0028 0.8444 0.0057 0.9131 0.0041 0.0057 47.5 0.7787 0.8269 0.8963 0.7817 0.0030 0.8985 0.0022 0.7825 0.0038 0.8992 0.0029 0.8326 0.0057 0.9007 0.0043 0.0057 50.0 0.7696 0.8157 0.8840 0.7725 0.0029 0.8863 0.0023 0.7733 0.0036 0.8870 0.0030 0.8213 0.0056 0.8885 0.0045 0.0056 52.5 0.7612 0.8051 0.8720 0.7639 0.0027 0.8743 0.0023 0.7647 0.0034 0.8750 0.0030 0.8106 0.0055 0.8765 0.0045 0.0055 55.0 0.7534 0.7949 0.8602 0.7560 0.0026 0.8626 0.0023 0.7567 0.0033 0.8633 0.0030 0.8003 0.0053 0.8648 0.0046 0.0053 57.5 0.7463 0.7854 0.8488 0.7487 0.0024 0.8512 0.0024 0.7493 0.0031 0.8519 0.0030 0.7905 0.0052 0.8535 0.0046 0.0052 60.0 0.7397 0.7763 0.8378 0.7419 0.0022 0.8401 0.0023 0.7425 0.0029 0.8408 0.0030 0.7813 0.0050 0.8424 0.0046 0.0050 max 0.0058 Holtec Report HI-2084175 4-58 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.9 Results of the CASMO-4 Calculations of WDR Reactivity Dependence on Number of WDR Rods and Fuel Enrichment WDR'Rods 0 4 8 n/a Enrichment 34 wt% U-235 Burnup Delta-k (GWD/MTU) Ref kinf kinf (max - ref) 0.0 1.i087 1.1087 1.1087 0.0000 0.1 1.1064 1.1064 1.1064 0.0000 2.0 1.0864 1.0864 1.0864 0.0000 4.0 1.0698 1.0697 1.0698 0.0000 6.0 1.0529 1.0529 1.0530 0.0001 8.0 1.0366 1.0366 1.0368 0.0002 10.0 1.0210 1.0211 1.0213 0.0003 11.0 1.0134 1.0135 1.0137 0.0003 12.5 1.0024 1.0025 1.0027 0.0003 15.0 0.9848 0.9847 0.9850 0.0003 17.5 0.9677 0.9678 0.9681 0.0004 20.0 0.9513 0.9514 0.9516 0.0003 22.5 0.9353 0.9354 0.9356 0.0003 25.0 0.9198 0.9199 0.9201 0.0003 27.5 0.9048 0.9049 0.9051 0.0003 30.0 0.8903 0.8903 0.8906 0.0003 32.5 0.8762 0.8763 0.8765 0.0003 35.0 0.8626 0.8627 0.8629 0.0003 37.5 0.8495 0.8496 0.8498 0.0003 40.0 0.8370 0.8370 0.8372 0.0003 42.5 0.8250 0.8250 0.8252 0.0002 45.0 0.8135 0.8136 0.8137 0.0002 47.5 0.8026 0.8027 0.8028 0.0002 50.0 0.7923 0.7924 0.7926 0.0002 52.5 0.7826 0.7827 0.7828 0.0002 Holtec Report HI-2084175 4-59 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.10 Results of the CASMO-4 Calculations of a Fuel Assembly with both IFBA and WDR Bias Dependence on the Number of WDR Rods and Fuel Enrichment WDR Rods 0 4 8 n/a Enrichment 3.4 wt% U-235 Delta-k Burnup (max -

(GWD/MTU) Ref f f ref) 0.0 1,1087 0.9100 0.9100 -0.1988 0.1 1.1064 0.9106 0.9106 -0.1958 2.0 1.0864 0.9417 0.9417 -0.1447 4.0 1.0698 0.9656 0.9656 -0.1041 6.0 1.0529 0.9794 0.9795 -0.0735 8.0 1.0366 0.9858 0.9859 -0.0507 10.0 1.0210 0.9868 0.9869 -0.0340 11.0 1.0134 0.9856 0.9858 -0.0276 12.5 1.0024 0.9824 0.9826 -0.0198 15.0 0.9848 0.9738 0.9740 -0.0107 17.5 0.9677 0.9628 0.9630 -0.0047 20.0 0.9513 0.9500 0.9502 -0.0010 22.5 0.9353 0.9363 0.9365 0.0012 25.0 0.9198 0.9222 0.9224 0.0026 27.5 0.9048 0.9080 0.9082 0.0034 30.0 0.8903 0.8939 0.8941 0.0039 32.5 0.8762 0.8801 0.8803 0.0041 35.0 0.8626 0.8666 0.8668 0.0042 37.5 0.8495 0.8536 0.8538 0.0043 40.0 0.8370 0.8410 0.8412 0.0043 42.5 0.8250 0.8290 0.8292 0.0042 45.0 0.8135 0.8175 0.8176 0.0041 47.5 0.8026 0.8065 0.8067 0.0040 50.0 0.7923 0.7961 0.7962 0.0039 52.5 0.7826 0.7862 0.7864 0:0038 Holtec Report HI-2084175 4-60 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION

___Table 4.7.11 Results of the CASMO-4 Calculations of WABA Bias Dependence onNurrber of WABA Rods arnd Fuel Enrichnrt WABARods 0 12 20 n/a 0 12 16 n/a Enrichment 2.6 3.0 vvt% U-235 Bumup Delta-k Delta-k Max (GWD/MTU) Ref k-inf k-inf (max - Ref k-inf k-inf (max - Delta-k 0.0 1.0394 1.0394 1.0394 0.0000 1.0773 1.0773 1.0773 0.0000 0.0000 0.1 1.0367 1.0367 1.0367 0.0000 1.0748 1.0748 1.0748 0.0000 0.0000 2.0 1.0188 1.0195 1.0199 0.0011 1.0556 1.0560 1.0562 0.0006 0.0011 4.0 1.0023 1.0035 1.0044 0.0021 1.0388 1.0397 1.0400 0.0012 0.0021 6.0 0.9854 0.9873 0.9885 0.0031 1.0218 1.0232 1.0236 0.0018 0.0031 8.0 0.9692 0.9715 0.9731 0.0039 1.0054 1.0072 1.0078 0.0024 0.0039 10.0 0.9537 0.9565 0.9583 0.0046 0.9897 0.9919 0.9926 0.0029 0.0046 11.0 0.9462 0.9491 0.9510 0.0048 0.9821 0.9844 0.9851 0.0031 0.0048 12.5 0.9354 0.9385 0.9406 0.0052 0.9711 0.9736 0.9744 0.0033 0.0052 15.0 0.9180 0.9216 0.9239 0.0059 0.9535 0.9564 0.9573 0.0039 0.0059 17.5 0.9014 0.9051 0.9077 0.0063 0.9365 0.9396 0.9407 0.0042 0.0063 20.0 0.8854 0.8891 0.8915 0.0060 0.9201 0.9232 0.9242 0.0041 0.0060 22.5 0.8702 0.8735 0.8758 0.0056 0.9042 0.9072 0.9081 0.0039 0.0056 25.0 0.8556 0.8588 0.8609 0.0052 0.8890 0.8918 0.8927 0.0037 0.0052 27.5 0.8417 0.8447 0.8467 0.0050 0.8743 0.8770 0.8778 0.0035 0.0050 30.0 0.8285 0.8313 0.8333 0.0047 0.8602 0.8628 0.8636 0.0034 0.0047 32.5 0.8161 0.8187 0.8205 0.0045 0.8467 0.8492 0.8500 0.0033 0.0045 35.0 0.8043 0.8068 0.8085 0.0042 0.8338 0.8362 0.8370 0.0032 0.0042 37.5 0.7932 0.7956 0.7973 0.0040 0.8215 0.8238 0.8246 0.0031 0.0040 40.0 0.7829 0.7851 0.7867 0.0038 0.8098 0.8120 0.8128 0.0030 0.0038 Holtec Report HI-2084175 4-61 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.12 Results for the Calculation of the Eccentric Fuel Positioning Reactivity Effect Parameter Enriched Blankets Natural Blankets No Blankets Region 3 Enrichment, wt% U-235 2.0 3.5 5.0 2.0 3.5 5.0 2.0 3.5 5.0 Region 3 Bumup, GWD/MTU 1.00 15.00 35.00 1.00 15.00 35.00 1.00 15.00 35.00 Reference Case Reactivity 0.956 0.9917 0.965 0.9559 0.9887 0.9608 0.9571 0.995 0.9798 Reference Case o 0.0005 0.0007 0.0006 0.0006 0.0006 0.0007 0.0006 0.0006 0.0007 Eccentric Case Reactivity 0.9503 0.988 0.962 0.9503 0.9852 0.9575 0.9492 0.9882 0.9748 Eccentric Case c1 0.0005 0.0005 0.0006 0.0006 0.0006 0.0005 0.0006 0.0007 0.0007 Delta-keff -0.0043 -0.0020 -0.0013 -0.0039 -0.0018 -0.0016 -0.0062 -0.0050 -0.0030 Holtec Report HI-2084175 4-62 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4,7,13a 4.7.3..........

Table..... .............................

.CASMO Calculations for Fuel Assembly Manufacturin Tolerance Uncertainties for Fuel Storage Cell (no cr p)

Case Fuel Pellet Fuel Pellet RefCase RodPitch-+ RodPitch- CladOD+ CladOD - CladlD+ Clad ID - OD+ Enr+

Soluble Stat. Combo Bumup Enr Boron kinf Delta kinf Delta kinf Delta kinf Delta kinf Delta kinf Delta kinf Delta kinf Delta kinf GWD/MTU 0.0 2 0 0.9648 0.0007 -0.0080 -0.0013 0.0012 0.0002 -0.0002 0.0004 00073 00074 10.0 '2 0 0.8889 0.0006 -0.0083 -0.0008 0.0008 0.0002 -0.0002 0.0005 0.0059 0.0060 20.0 '2 0 0.8277 0.0006 -0.0080 -0.0003 0.0003 0.0002 -0.0002 0.0007 0.0050 0.0051 0.0 2 2000 0.7056 0.0001 -0.0003 0.0001 -0.0001 0.0002 -0.0002 0.0008 0.0079 0.0079 10.0 2 2000 0.6753 0.0001 -0.0013 0.0004 -0.0004 0.0002 -0.0002 0.0009 0.0059 0.0060 20.0 2 2000 0.6366 0.0001 -0.0015 0.0008 -0.0008 0.0001 -0.0001 0.0010 0.0048 0.0050 0.0 3.0 0 1.0773 0.009 -0.0112 -0.0014 0.0014 0.0002 -0.0002 0.0003 0.0043 0.0046 10.0 3.0 0 0.9897 0.0008 -0.0110 -0.0011 0.0011 0.0002 -0.0002 0.0003 0.0041 0.0044 20.0 3.0 0 0.9201 0.0007 -0.0104 -0.0008 0.0008 0.0002 -0.0002 0.0004 0.0041 0.0043 30.0 3.0 0 0.8602 0.0007 -0.0097 -0.0004 0.0004 0.0002 -0.0002 0.0006 0.0039 0.0040 0.0 3.0 2000 0.8346 0.0003 -0.0031 -0.0001 0.0001 0.0002 -0.0002 0.0007 0.0052 0.0052 10.0 3.0 2000 0.7810 0.0003 -0.0035 0.0001 -0.0001 0.0002 -0.0002 0.0008 0.0046 0.0047 20.0 3.0 2000 0.7293 0.0002 -0.0033 0.0004 -0.0004 0.0002 -0.0002 0.0009 0.0043 0.0044 30.0 3.0 2000 0.6830 0.0002 -0.0030 0.0008 -0.0008 0.0002 -0.0002 0.0010 0.0039 0.0041 0.0 4.0 0 1.1471 0.0010 -0.0134 -0.0014 0.0014 0.0002 -0.0002 0.0002 0.0028 0.0033 10.0 4.0 0 1.0607 0.0010 -0.0130 -0.0013 0.0013 0.0002 -0.0002 0.0002 0.0030 0.0034 20.0 4.0 0 0.9922 0.0009 -0.0124 -0.0011 00011 0,0002 -0.0002 0.0003 0.0031 0.0034 30.0 4.0 0 0.9315 0.0008 -0.0117 -0.0008 0.0008 0.0002 -0.0002 0.0004 0.0032 0.0034 40.0 4.0 0 0.8763 0.0008 -0.0109 -0.0004 0.0004 0.0002 -0.0002 0.0005 0.0032 0.0033 50.0 4.0 0 0.8275 0.0007 -0.0102 0.0000 000 0.0002 -0.0002 0.0007 0.0029 0.0031 0.0 4.0 2000 0.9229 0.0004 -0.0054 -0.0002 0.0002 0.0002 -0.0002 0.0007 0.0037 0.0038 10.0 4.0 2000 0.8625 0.0004 -0.0055 -0.0001 0.0001 0.0002 -0.0002 0.0007 0.0036 0.0037 20.0 4.0 2000 0.8076 0.0003 -0.0051 0.0001 -0.0001 0.0002 -0.0002 0.0007 0.0035 0.0036 30.0 4.0 2000 0.7574 0.0003 -0.0046 0.0004 -0.0004 0.0002 -0.0002 0.0009 0.0034 0.0036 40.0 4.0 2000 0.7112 0.0002 -0.0041 0.0007 -0.0007 0.0002 -0.0002 0.0010 0.0033 0.0035 50.0 4.0 2000 0.6705 0.0002 -0.0037 0.0010 -0.0010 0.0002 -0.0001 0.0011 0.0030 0.0034 0.0 5.0 0 1.1954 0.0011 -0.0150 -0.0015 0.0015 0.0002 -0.0002 0.0002 0.0020 0.0028 10.0 5.0 0 1.1129 0.0011 -0.0145 -0.0014 0.0014 0.0002 -0.0002 0.0001 0.0023 0.0029 20.0 5.0 0 1.0476 0.0010 -0.0140 -0.0012 0.0012 0.0002 -0.0002 0.0002 0.0024 0.0029 30.0 5.0 0 0.9893 0.0009 -0.0133 -0.0010 0.0010 0.0002 -0.0002 0.0002 0.0026 0.0029 400 5.0 0 0.9348 0.0009 -0.0125 -0.0008 0.0008 0.0002 -0.0002 0.0004 0.0027 0.0029 50.0 5.0 0 0.8840 0.0008 -0.0118 -0.0004 0.0004 0.0002 -0.0002 0.0005 0.0027 0.0029 60.0 15.0 0 0.8378 0.0007 -0,0110 -0,0001 0.0001 0.0002 -0.0002 0.0006 0.0025 0.0027 0.0 5.0 2000 0.9877 0.0005 -0.0072 -0.0003 0.0003 0.0002 -0.0002 0.0006 0.0028 0.0030 10.0 5.0 2000 0.9260 0.0005 -0.0072 -0.0002 0.0002 0.0002 -0.0002 0.0006 0.0028 0.0029 20.0 5.0 2000 0.8715 0.0005 -0.0068 -0.0001 0.0001 0.0002 -0.0002 0.0006 0.0029 0.0030 30.0 5.0 2000 0.8212 0.0004 -0.0063 0.0001 -0.0001 0.0002 -0.0002 0.0007 0.0029 0.0030 40.0 5.0 2000 0.7736 0.0004 -0.0057 0.0004 -0.0004 0.0002 -0.0002 0.0008 0.0029 0.0031 50.0 5.0 2000 0.7292 0.0003 -0.0051 0.0007 -0.0007 0.0002 -0.0002 0.0010 0.0028 0.0031 50.0 5.0 2000 0.7292 0.0003 -0.0046 0.0010 -0.0010 0.0002 -0.0002 0.0011 1 60.0 15.0 12000 0.6890 0.0003 -0.0046 0.0010 -0.0010 0.0002 -0.0002 10.0011 0.0027 0.0027 0.0030 0.0030 HI-2084175 4-63 Holtec Project 1702 Holtec Report Holtec Report HI-2084175 4-63 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION

.... ~~ ~ ~7.17,i7ii_

~ ~ ~

Table 4.7.13b_

~ ~ .............

~ i .i i CASMO Calculations for Fuel Assemlby Manufacturing Tolerance Uncertainties for Fuel Storage Cell (cree Case Fuel Pellet Fuel Pellet RefCase Rod Pitch + Rod Pitch - Clad OD + Clad OD - Clzid ID + Clad ID - OD + Enr +

Star.Combo Bumup Soluble BuMu Enr Boron GWD/MTU aa kinf

______ Delta kinf Delta kinf Delta kinf Delta

______kinf Delta kinf Delta kinf Delta kinf Delta

___ kinf ___

. 0.0 2 0 0.9688 0.0007 -0.0076 -0.0012 0.0012 0.0010 -0.0010 0.0001 0.0073 0.0075 10.0 2 0 0.8889 0.0006 -0.0083 -0.0008 10.0008 00002 -0.0002 0.0005 0.0059 0.0060 20.0 2 0 0.8277 0.0006 -0.0080 -0.0003 0.0003 0.0002 -0.0002 0.0007 0.0050 0.0051 0.0 2 2000 0.7056 0.0001 -0.0003 0.0001 -0.0001 0.0002 -0.0002 0.0008 0.0079 0.0079 10.0 2 2000 0.6753 0.0001 -0.0013 0.0004 -0.0004 0.0002 -0.0002 0.0009 0.0059 0.0060 20.0 2 2000 0.6366 0.0001 -0.0015 0.0008 -0.0008 0.0001 -0.0001 0.0010 0.0048 0.0050 0.0 3.0 0 1.0773 0.0009 -0.0112 -0.0014 0.0014 0.0002 -0.0002 0.0003 0.0043 0.0046 10.0 3.0 0 0.9897 0.0008 -0.0110 -0.0011 0.0011 0.0002 -0.0002 0.0003 0.0041 0.0044 20.0 3.0 0 0.9201 0.0007 -0,0104 -0.0008 0.0008 0.0002 -0.0002 0.0004 0,0041 0.0043 30.0 3.0 0 0.8602 0.0007 -0.0097 -0.0004 0.0004 0.0002 -0.0002 0.0006 0.0039 0.0040 0.0 3.0 2000 0346 0.0003 -0.0031 -00001 0.0001 0.0002 -0.0002 0.0007 0.0052 0.0052 10.0 3.0 2000 0.7810 0.0003 -0.0035 0.0001 -0.0001 0.0002 -0.0002 0.0008 0.0046 0.0047 20.0 3.0 2000 0.7293 0.0002 -0.0033 0.0004 -0.0004 0.0002 -0.0002 0.0009 0.0043 0.0044 30.0 3.0 2000 0.6830 0,0002 -0.0030 0.0008 -0.0008 0.0002 -0.0002 0.0010 0.0039 0.0041 0.0 4.0 0 1.1471 0.0010 -0.0134 -0.0014 0.0014 0,0002 -0.0002 0.0002 0.0028 0.0033 10.0 4.0 0 1.0607 0.0010 -0.0130 -0.0013 0.0013 0.0002 -0.0002 0.0002 0.0030 0.0034 20.0 4.0 0 0.9922 0.0009 -0.0124 -0.0011 0.0011 0.0002 -0.0002 0.0003 0.0031 0.0034 30.0 4.0 0 0.9315 0,0008 -0.0117 -0.0008 0.0008 0.0002 -0.0002 0.0004 0.0032 0.0034 40.0 4.0 0 0.8763 0.0008 -0.0109 -0.0004 0.0004 0.0002 -0.0002 0.0005 0.0032 0.0033 50.0 4.0 0 0.8275 0.0007 -0.0102 0.0000 0.0000 0.0002 -0.0002 0.0007 0.0029 0.0031 0.0 4.0 2000 0.9229 0.0004 -0.0054 -0.0002 0.0002 0.0002 -0.0002 0.0007 0.0037 0.0038 10.0 4.0 2000 0.8625 0.0004 -0.0055 -0.0001 0.0001 0.0002 -0.0002 0.0007 0.0036 0.0037 20.0 4.0 2000 0.8076 0.0003 -0.0051 0.0001 -0.0001 0.0002 -0.0002 0.0007 0.0035 0.0036 30.0 4.0 2000 0.7574 0.0003 -0.0046 0.0004 -0.0004 0.0002 -0.0002 0.000 0.0034 0.0036 40.0 4.0 2000 0.7112 0.0002 -0.0041 0.0007 -0.0007 0.0002 -0.0002 0.0010 0.0033 0.0035 50.0 4.0 2000 0.6705 0.0002 -0.0037 0.0010 -0.0010 0.0002 -0.0001 0.0011 0.0030 0.0034 0.0 5.0 0 1.1954 0.0011 -0.0150 -0.0015 0.0015 0.0002 -0.0002 0.0002 0.0020 0.0028 10.0 5.0 0 1.1129 0.0011 -0.0145 -0.0014 0.0014 0.0002 -0.0002 0.0001 0.0023 0.0029 20.0 5.0 0 1.0476 0.0010 -0.0140 -0.0012 0.0012 0.0002 -0.0002 0.0002 0.0024 0.0029 30.0 5.0 0 0.9893 0.0009 -0.0133 -0.0010 0.0010 0.0002 -0.0002 0.0002 0.0026 0.0029 40.0 5.0 0 0.9348 0.0009 -0.0125 -0.0008 0.0008 0.0002 -0.0002 0.0004 0.0027 0.0029 50.0 5.0 0 0.8840 0.0008 -0.0118 -0.0004 0.0004 0.0002 -0.0002 0.0005 0.0027 0.0029 60.0 5.0 0 0.8378 0.0007 -0.0110 -0.0001 0.0001 0.0002 -0.0002 0.0006 0.0025 0.0027 0.0 5.0 2000 0.9877 0.0005 -0.0072 -0.0003 0.0003 0.0002 -0.0002 0.0006 0.0028 0.0030 10.0 5.0 2000 0.9260 0.0005 -0.0072 -0.0002 0.0002 0.0002 -0.0002 0.0006 0.0028 0.0029 5.0 2000 0.8715 0.0005 -0.0068 -0.0001 0.0001 0.0002 -0.0002 0.0006 0.0029 0.0030 30.0 5.0 2000 0.8212 0.0004 -0.0063 0.0001 -0.0001 0.0002 -0.0002 0.0007 0.0029 0.0030 40.0 5.0 2000 0.7736 0.0004 -0,0057 0.0004 -0.0004 0.0002 -0.0002 0.0008 0.0029 0.0031 50.0 5.0 2000 0.7292 0.0003 -0.0051 0.0007 -0.0007 0.0002 -0.0002 0.0010 0.0028 0.0031 60.0 5.0 2000 0.6890 0.0003 -0.0046 0.0010 -0.0010 0.0002 -0.0002 0.0011 0.0027 0.0030 4-64 Holtec Project 1702 Holtec Report HI-2084175 4-64 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.14 CASMO Calculations for Manufacturing Tolerance Uncertainties for Fuel Storage Cell Cas Poison Poison B-10 Loading CaseRefCase ID + ID- Width - Sheathing+ Sheathing- Box Wall + Box Wall - Thickness - Min Soluble Stat. Combo Burnup Enr Boron kinf Delta kinf Delta kinf Delta kinf Delta kinf Delta kinf Delta kinf Delta kinf Delta kinf Delta kinf GWD/MTU __ __ __ _ _

0.0 2 0 0.9648 0.0014 0.0012 0.0011 0.0002 -0.0002 0.0001 -0.0001 0.0017 0.0014 0.0028 10.0 2 0 0.8889 0.0010 0.0013 0.0011 0.0001 -0.0002 0.0001 -0.0001 0.0015 0.0013 0.0026 20.0 2 0 0.8277 0.0008 0.0012 0.0010 0.0001 -0.0002 0.0001 -0.0001 0.0014 0.0012 0.0024 0.0 2 2000 0.7056 -0.0005 0.0019 0.0006 0.0001 -0.0001 0.0001 -0.0001 0.0011 0.0010 0.0024 10.0 2 2000 0.6753 -0.0006 0.0019 0.0006 0.0001 -0.0001 0.0001 -0.0001 0.0011 0.0010 0.0024 20.0 2 2000 0.6366 -0.0007 0.0017 0.0006 0.0001 -0.0001 0.0001 -0.0001 0.0010 0.0009 0.0023 0.0 3.0 0 1.0773 0.0009 0.0017 0.0013 0.0002 -0.0002 0.0001 -0.0001 0.0018 0.0015 0.0032 10.0 3.0 0 0.9897 0.0006 0.0016 0.0012 0.0002 -0.0002 0.0001 -0.0001 0.0017 0.0014 0.0030 20.0 3.0 0 0.9201 0.0006 0.0014 0.0011 0.0001 -0.0002 0.0001 -0.0001 0.0016 0.0013 0.0027 30.0 3.0 0 0.8602 0.0006 0.0013 0.0010 0.0001 -0.0002 0.0001 -0.0001 0.0015 0.0012 0.0025 0.0 3.0 2000 0.8346 -0.0008 0.0023 0.0008 0.0001 -0.0001 0.0001 -0.0001 0.0013 0.0012 0.0030 10.0 3.0 2000 0.7810 -0.0009 0.0022 0.0007 0.0001 -0.0001 0.0001 -0.0001 0.0012 0.0011 0.0028 20.0 3.0 2000 0.7293 -0.0008 0.0020 0.0007 0.0001 -0.0001 0.0001 -0.0001 0.0011 0.0010 0.0026 30.0 3.0 2000 0.6830 -0.0008 0.0019 0.0006 0.0001 -0.0001 0.0001 -0.0001 0.0011 0.0010 0.0024 0.0 4.0 0 1.1471 0.0004 0.0020 0.0013 0.0002 -0.0002 0.0001 -0.0001 0.0019 0.0017 0.0035 10.0 4.0 0 1.0607 0.0003 0.0019 0.0012 0.0002 -0.0002 0.0001 -0.0001 0.0018 0.0015 0.0033 20.0 4.0 0 0.9922 0.0003 0.0017 0.0012 0.0002 -0.0002 0.0001 -0.0001 0.0017 0.0014 0.0030 30.0 4.0 0 0.9315 0.0003 0.0015 0.0011 0.0002 -0.0002 0.0001 -0.0001 0.0016 0.0013 0.0028 40.0 4.0 0 0.8763 0.0004 0.0014 '0.0011 0.0001 -0.0002 0.0001 -0.0001 0.0015 0.0013 0.0026 50.0 4.0 0 0.8275 0.0004 0.0013 0.0010 0.0001 -0.0001 0.0001 -0.0001 0.0014 0.0012 0.0025 0.0. 4.0 2000 0.9229 -0.0012 0.0026 0.0008 0.0001 -0.0001 0.0001 -0.0001 0.0014 0.0013 0.0034 10.0 4.0 2000 0.8625 -0.0011 0.0025 0.0008 0.0001 -0.0001 0.0001 -0.0001 0.0013 0.0012 0.0032 20.0 4.0 2000 0.8076 -0.0011 0.0023 0.0008 0.0001 -0.0001 0.0001 -0.0001 0.0013 0.0011 0.0029 30.0 4.0 2000 0.7574 -0.0010 0.0021 0.0007 0.0001 -0.0001 0.0001 -0.0001 0.0012 0.0011 0.0027 40.0 4.0 2000 0.7112 -0.0009 0.0020 0.0007 0.0001 -0.0001 0.0001 -0.0001 0.0011 0.0010 0.0026 50.0 4.0 2000 0.6705 -0.0008 0.0018 0.0006 0.0001 -0.0001 0.0001 -0.0001 0.0011 0.0010 0.0024 0.0 5.0 0 1.1954 0.0000 0.0023 0.0014 0.0002 -0.0002 0.0001 -0.0001 0.0020 0.0017 0.0037 10.0 5.0 0 1.1129 -0.0001 0.0021 0.0013 0.0002 -0.0002 0.0001 -0.0001 0.0018 0.0016 0.0035 20.0 5.0 0 1.0476 0.0000 0.0019 0.0012 0.0002 -0.0002 0.0001 -0.0001 0.0017 0.0015 0.0032 30.0 5.0 0 0.9893 0.0001 0.0017 0.0012 0.0002 -0.0002 0.0001 -0.0001 0.0016 0.0014 0.0030 40.0 5.0 0 0.9348 0.0002 0.0016 0.0011 0.0001 -0.0002 0.0001 -0.0001 0.0016 0.0013 0.0028 50.0 5.0 0 0.8840 0.0002 0.0014 0.0011 0.0001 -0.0001 0.0001 -0.0001 0.0015 0.0013 0.0027 60.0 5.0 0 0.8378 0.0003 0.0013 0.0010 0.0001 -0.0001 0.0001 -0.0001 0.0014 0.0012 0.0025 0.0 5.0 2000 0.9877 -0.0014 0.0029 0.0009 0.0001 -0.0001 0.0001 -0.0001 0.0015 0.0014 0.0037 10.0 5.0 2000 0.9260 -0.0014 0.0027 0.0008 0.0001 -0.0001 0.0001 -0.0001 0.0014 0.0013 0.0034 20.0 5.0 2000 0.8715 -0.0013 0.0025 0.0008 0.0001 -0.0001 0.0001 -0.0001 0.0013 0.0012 0.0032 30.0 5.0 2000 0.8212 -0.0012 0.0023 0.0008 0.0001 -0.0001 0.0001 -0.0001 0.0013 0.0012 0.0030 40.0 5.0 2000 0.7736 -0.0010 0.0021 0.0007 0.0001 -0.0001 0.0001 -0.0001 0.0012 0.0011 0.0028 50.0 5.0 2000 0.7292 -0.0009 0.0020 0.0007 0.0001 -0.0001 0.0001 -0.0001 0.0011 0.0010 0.0026 60.0 5.0 2000 0.6890 -0.0009 0.0019 0.0007 0.0001 -0.0001 0.0001 -0.0001 0.0011 0.0010 0.0024 Holtec Report HI-2084175 4-65 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7. 15a CASMO Calculations for Pool Temperature Tolerance Uncertainties Case T=39.2F(ref) T=32F T=O .33F T= 150F T=255F, 0% Voids T=255F,10% Voids T 255 F, 20% Voids Buep SlbeBrnEnr kinf De lta kinf Delta kinf Delta konf Delta kinf Delta kinf Delia kinf GWDAITU -ppm) 0.0 0 2.0 0.9648 0.0007 -0.0049 -00145 -00312 -0,0515 -0,0762 10.0 0 2.0 08889 0.0005 -0.0032 -0.0101 -0.0219 -0,0414 -0.0648 20.0 0 2.0 0.8277 0.0003 -00022 -0.0072 -0.0163 -0.0351 -0.0575 30.0 0 2.0 0.7802 0.0002 -0.0015 -0.0052 -0.0123 -0.0304 -0.0519 0.0 2000 2.0 0.7056 0.0004 -0.0017 -0.0038 -0.0063 -0.0021 -0.0010 10.0 2000 2.0 0.6753 0.0001 -0,0003 -0.0001 0.0010 0.0029 0.0018 20.0 2000 2.0 0.6366 0.0000 0.0004 0.0015 00041 0.0050 0.0030 30.0 2000 2.0 0.6049 -0.0001 0.0008 0.0026 0.0060 0.0063 0.0038 0.0 0 3.0 1.0773 0.0006 -0,0041 -0.0130 -0.0292 -0.0520 -0.0793 10.0 0 3.0 0.9897 0.0005 -0.0035 -0.0111 -0.0248 -0.0470 -0.0732 20.0 0 3.0 0.9201 0.0004 -0.0028 -0.0092 -0.0210 -0.0424 -0.0677 30.0 0 3.0 0.8602 0.0003 -0.0022 -0.0074 -0.0172 -0.0377 -0.0618 40.0 0 3.0 0.8098 0.0002 -0.0016 -0.0058 -0.0139 -0.0334 -0.0565 0.0 2000 3.0 0.8346 0.0003 -0.0015 -0.0034 -0.0061 -0.0047 -0.0068 10.0 2000 3.0 0.7810 0.0001 -0.0007 -0,0015 -0.0024 -0.0029 -0.0068 20.0 2000 3.0 0.7293 0.0001 -0.0002 -0.0002 0.0000 -0.0012 -0.0057 30.0 2000 3.0 0.6830 0.0000 0.0002 0.0009 0.0022 0.0007 -0.0039 40.0 2000 3.0 0.6438 -0.0001 0.0006 0.0018 00040 0.0024 -0.0022 0.0 0 4.0 1.1471 0.0005 -0.0036 -00118 -0.0273 -0.0513 -0.0797 10.0 0 4.0 1.0607 0.0005 -0.0034 -0.0111 -0,0254 -0.0489 -0.0765 20.0 0 4.0 0.9922 0.0004 -0.0030 -0.0101 -0.0231 -0.0462 -0,0732 30.0 0 4.0 0.9315 0.0003 -0.0026 -0.0088 -0.0204 -0.0427 -00689 40.0 0 4.0 0.8763 0.0002 -0.0021 -0.0074 -0.0175 -0.0390 -0.0641 50.0 0 4.0 0,8275 0.0002 -0.0016 -0,0061 -0.0147 -0.0352 -0.0593 0.0 2000 4.0 0.9229 0.0002 -00012 -0,0030 -0.0059 -00066 -0.0111 10.0 2000 4.0 0.8625 0.0002 -0.0008 -0.0021 -0,0042 -0.0065 -0.0123 20.0 2000 4.0 0.8076 0.0001 -0.0005 -0.0013 -0,0026 -0.0057 -0.0123 30.0 2000 4.0 0.7574 0.0000 -0.0002 -0. 0004 -0.0009 -0.0041 -0.0109 40.0 2000 4.0 0.7112 0.0000 0.0002 0.0004 0.0009 -0.0023 -0.0089 50.0 2000 4.0 0.6705 -0.0001 0.0005 0.0012 0.0025 -0.0006 -0.0069 0.0 0 5.0 1.1954 0.0004 -0.0032 -0.0109 -0,0257 -0.0501 -0.0789 10.0 0 5.0 1.1129 0.0004 -0.0032 -0.0108 -0.0251 -0.0492 -0.0774 20.0 0 5.0 1.0476 0.0004 -0.0030 -0.0103 -0.0239 -0.0478 -0.0756 30.0 0 5.0 0.9893 0.0003 -0.0027 -0.0095 -0.0221 -0.0456 -0.0729 40.0 0 5.0 0.9348 0.0003 -0.0024 -0.0084 -0.0200 -00428 -0.0694 50.0 0 5.0 0.8840 0.0002 -0.0020 -0.0073 -0.0176 -0.0396 -0.0653 60.0 0 5.0 0.8378 0.0002 -0.0016 -0.0062 -0.0152 -0.363 -0.0610 0.0 2000 5.0 0.9877 0.0002 -0.0010 -0.0027 -0.057 -0.0079 -0.0139 10.0 2000 5.0 0.9260 0.0002 -0.0009 -0.0024 -00051 -00088 -0.0162 20.0 2000 50 0.8715 0.0001 -0.0007 -0.0020 -0.0042 -0.0086 -0.0166 30.0 2000 5.0 0.8212 0.0001 -00004 -0 0013 -0.0030 -0.0078 -0.0161 40.0 2000 5.0 0.7736 0.0000 -0.0002 -0.0006 -0.0016 -0.0064 -0.0146 50.0 2000 5.0 0.7292 0.0000 0.0001 0.0001 -0.0001 -0.0047 -0.0126 600 2000 5.0 0.6890 -0.0001 00004 0.0008 0.0014 -0.0030 -0.0105 Holtec Report 1HI-2084175 4-66 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.15b Zalculation of the Temperature Bias CASMO-4 (water density = 1 g/cc parameter T = 39.2 F T = 80.33 F wt% U235 ppm gwd/mtu kinf delta kinf 2 0 0.0 0.9648 -0.0043 2 0 10.0 0.8889 -0.0027 2 0 20.0 0.8277 -0.0017 2 0 30.0 0.7803 -0.0010 3 0 0.0 1.0773 -0.0035 3 0 10.0 0.9897 -0.0028 3 0 20.0 0.9201 -0.0022 3 0 30.0 0.8602 -0.0016 3 0 40.0 0.8098 -0.0010 4 0 0.0 1.1471 -0.0029 4 0 10.0 1.0608 -0.0027 4 0 20.0 0.9922 -0.0023 4 0 30.0 0.9315 -0.0019 4 0 40.0 0.8763 -0.0014 5 0 0.0 1,1954 -0.0025 5 0 10.0 1.1129 -0.0025 5 0 20.0 1.0476 -0.0023 5 0 30.0 0.9893 -0.0020 5 0 40.0 0.9348 -0.0017 5 0 50.0 0.8840 -0.0014 5 0 60.0 0.8378 -0.0010 2 2000 0.0 0.7056 -0.00201 2 2000 10.0 0.6753 -0.0005 2 2000 20.0 0.6366 0.0002 2 2000 30.0 0.6049 0.0007 3 2000 0.0 0.8346 -0.0016 3 2000 10.0 0.7810 -0.0008 3 2000 20.0 0.7293 -0.0003 3 2000 30.0 0.6830 0.0002 3 2000 40.0 0.6438 0.0005 4 2000 0.0 0.9229 -0.0013 4 2000 10.0 0.8625 -0.0009 4 2000 20.0 0.8076 -0.0005 4 2000 30.0 0.7574 -0.0002 4 2000 40.0 0.7112 0.0001 5 2000 0.0 0.9877 -0.0011 5 2000 10.0 0.9260 -0.0009 5 2000 20.0 0.8715 -0.0007 5 2000 30.0 0.8212 -0.0004 5 2000 40.0 0.7736 -0.0001 5 2000 50.0 0.7292 0.0001 5 2000 60.0 0.6890 0.0004 Holtec Report HI-2084175 4-67 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.16 Verification of the Initial Enrichment and Burnup Combinations and Calculation of Soluble Boron Requirements Region 3 Enrichment (wto U-235) 2.0 2.0 2.0 2.0 2.0 2.0 2.0 Region 3 Burnup (GWD/MTU) 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Region 2 Enrichment (wV/o U-235) 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Region 2 Burnup (GWD/MTU) 9.65 18.16 25.83 32.51 38.85 44.31 51.67 Reference Reactivity (Calculated keff) 0.9697 0.9689 0.9683 0.9677 0.9671 0.9666 0.9660 Curve Verification Reactivity, 0 ppm 0.9672 0.9683 0.9666 0.9687 0.9683 0.9663 0.9641 Delta Reactivity -0.0025 -0.0006 -0.0017 0.0010 0.0012 -0.0003 -0.0019 Calculational Uncertainty +/- 2*sqrt(sumrsq(a)) 0.0017 0.0017 0.0017 0.0017 0.0020 0.0020 0.0020 Soluble Boron Reactivity, 800 ppm 0.8735 0.8782 0.8787 0.8808 0.8820 0.8825 0.8828 Soluble Boron Target Reactivity 0.9197 0.9189 0.9183 0.9177 0.9171 0.9166 0.9160 Interpolated Soluble Boron Requirement 406 438 439 464 475 474 473 Reactivity, Interpolated Soluble Boron 0.9162 0.9150 0.9163 0.9144 0.9141 0.9153 0.9128 Region 3 Enrichment (wt%/o U-235) 2.5 2.5 2.5 2.5 2.5 2.5 2.5 Region 3 Burnup (GWD/MTU) 7.78 7.78 7.78 7.78 7.78 7.78 7.78 Region 2 Enrichment (wt/o U-235) 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Region 2 Burnup (GWD/MTU) 9.65 18.16 25.83 32.51 38.85 44.31 51.67 Reference Reactivity (Calculated keff) 0.9697 0.9689 0.9683 0.9677 0.9671 0.9666 0.9660 Curve Verification Reactivity, 0 ppm 0.9692 0.9677 0.9665 0.9673 0.9686 0.9670 0.9643 Delta Reactivity -0.0005 -0.0012 -0.0018 -0.0004 0.0015 0.0004 -0.0017 Calculational Uncertainty + 2*sqrt(sumsq(o)) 0.0020 0.0020 0.0020 0.0020 0.0020 0.0020 0.0020 Soluble Boron Reactivity, 800 ppm 0.8756 0.8787 0.8798 0.8820 0.8827 0.8835 0.8825 Soluble Boron Target Reactivity 0.9197 0.9189 0.9183 0.9177 0.9171 0.9166 0.9160 Interpolated Soluble Boron Requirement 423 438 444 465 480 482 472 Reactivity, Interpolated Soluble Boron 0.9171 0.9148 0.9161 0.9141 0.9150 0.9142 0.9137 Region 3 Enrichment (wtl/o U-235) 3 3 3 3 3 3 3 Region 3 Burnup (GWD/MTU) 14.03 14.03 14.03 14.03 14.03 14.03 14.03 Region 2 Enrichment (wt/o U-235) 2 2.5 3 3.5 4 4.5 5 Region 2 Burnup (GWD/MTU) 9.65 18.16 25.83 32.51 38.85 44.31 51.67 Reference Reactivity (Calculated keff) 0.9697 0.9689 0.9683 0.9677 0.9671 0.9666 0.9660 Curve Verification Reactivity, 0 ppm 0.9683 0.9691 0.9678 0.9674 0.9668 0.9684 0.9646 Delta Reactivity -0.0014 0.0002 -0.0005 -0.0003 -0.0003 0.0018 -0.0014 Calculational Uncertainty+ 2*sqrt(sumsq(;)) 0.0017 0.0020 0.0020 0.0020 0.0017 0.0020 0.0017 Soluble Boron Reactivity, 800 ppm 0.8754 0.8791 0.8811 0.8818 0.8846 0.8845 0.8836 Soluble Boron Target Reactivity 0.9197 0.9189 0.9183 0.9177 0.9171 0.9166 0.9160 Interpolated Soluble Boron Requirement 419 446 456 465 484 493 480 Reactivity, Interpolated Soluble Boron 0.9168 0.9139 0.9152 0.9155 0.9141 0.9152 0.9133 Note: The reference reactivity is the target keff from Table 4.7.6c for the Region 2 enrichment and bumup combination for each case.

4-68 Holtec Project 1702 HI-20841 75 Holtec Report HI-2084175 4-68 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.16 Continued Region 3 Enrichment (wt% U-235) 3.5 3.5 3.5 3.5 3.5 3.5 3.5 Region 3 Bturnup (GWD/MTU) 20.35 20.35 20.35 20.35 20.35 20.35 20.35 Region 2 Enrichment (wt%/. U-235) 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Region 2 Burnup (GWD/MTU) 9.65 18.16 25.83 32.51 38.85 44.31 51.67 Reference Reactivity (Calculated keff) 0.9697 0.9689 0.9683 0.9677 0.9671 0.9666 0.9660 Curve Verification Reactivity, 0 ppm 0.9680 0.9681 0.9667 0.9687 0.9676 0.9676 0.9658 Delta Reactivity -0.0017 -0.0008 -0.0016 0.0010 0.0005 0.0010 -0.0002 Calculational Uncertainty +/- 2*sqrt(sunsq(a)) 0.0020 0.0020 0.0017 0.0017 0.0020 0.0020 0.0017 Soluble Boron Reactivity, 800 ppm 0.8757 0.8797 0.8810 0.8824 0.8839 0.8852 0.8841 Soluble Boron Target Reactivity 0.9197 0.9189 0.9183 0.9177 0.9171 0.9166 0.9160 Interpolated Soluble Boron Requirement 419 445 451 473 483 495 488 Reactivity, Interpolated Soluble Boron 0.9179 0.9150 0.9164 0.9144 0.9145 0.9149 0.9141 Region 3 Enrichment (wt% U-235) 4 4 4 4 4 4 4 Region 3 Burnup (GWD/MTU) 26.52 26.52 26.52 26.52 26.52 26.52 26.52 Region 2 Enrichment (wt% U-235) 2 2.5 3 3.5 4 4.5 5 Region 2 Burnup (GWD/MTU) 9.65 18.16 25.83 32.51 38.85 44.31 51.67 Reference Reactivity (Calculated keff) 0.9697 0.9689 0.9683 0.9677 0.9671 0.9666 0.9660 Curve Verification Reactivity, 0 ppm 0.9677 0.9677 0.9682 0.9667 0.9672 0.9676 0.9651 Delta Reactivity -0.0020 -0.0012 -0.0001 -0.0010 0.0001 0.0010 -0.0009 Calculational Uncertainty + 2*sqrt(sumsq(o)) 0.0017 0.0017 0.0020 0.0020 0.0020 0.0020 0.0017 Soluble Boron Reactivity, 800 ppm 0.8763 0.8786 0.8804 0.8817 0.8840 0.8852 0.8852 Soluble Boron Target Reactivity 0.9197 0.9189 0.9183 0.9177 0.9171 0.9166 0.9160 Interpolated Soluble Boron Requirement 420 438 454 461 482 495 492 Reactivity, Interpolated Soluble Boron 0.9162 0.9164 0.9137 0.9147 0.9138 0.9133 0.9129 Region 3 Enrichment (wt% U-235) 4.5 4.5 4.5 4.5 4.5 4.5 4.5 Region 3 Burnup (GWD/MTU) 33.15 33.15 33.15 33.15 33.15 33.15 33.15 Region 2 Enrichment (wt%/o U-235) 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Region 2 Burnup(GWD/MTU) 9.65 18.16 25.83 32.51 38.85 44.31 51.67 Reference Reactivity (Calculated keff) 0.9697 0.9689 0.9683 0.9677 0.9671 0.9666 0.9660 Curve Verification Reactivity, 0 ppm 0.9672 0.9683 0.9673 0.9677 0.9658 0.9663 0.9646 Delta Reactivity -0.0025 -0.0006 -0.0010 0.0000 -0.0013 -0.0003 -0.0014 Calculational Uncertainty +/- 2*sqrt(sumsq(o)) 0.0017 0.0020 0.0020 0.0020 0.0017 0.0023 0.0020 Soluble Boron Reactivity, 800 ppm 0.8758 0.8772 0.8792 0.8814 0.8847 0.8857 0.8853 Soluble Boron Target Reactivity 0.9197 0.9189 0.9183 0.9177 0.9171 0.9166 0.9160 Interpolated Soluble Boron Requirement 416 434 445 464 481 493 490 Reactivity, Interpolated Soluble Boron 0.9159 0.9164 0.9167 0.9146 0.9151 0.9139 0.9152 Region 3 Enrichment (wt0 /o U-235) 5.0 5.0 5.0 5.0 5.0 5.0 5.0 Region 3 Burnup (GWD/MTU) 39.58 39.58 39.58 39.58 39.58 39.58 39.58 Region 2 Enrichment (wt%/o U-235) 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Region 2 Burnup (GWD/MTU) 9.65 18.16 25.83 32.51 38.85 44.31 51.67 Reference Reactivity (Calculated keff) 0.9697 0.9689 0.9683 0.9677 0.9671 0.9666 0.9660 Curve Verification Reactivity, 0 ppm 0.9676 0.9681 0.9649 0.9652 0.9640 0.9669 0.9638 Delta Reactivity -0.0021 -0.0008 -0.0034 -0.0025 -0.0031 0.0003 -0.0022 Calculational Uncertainty +/- 2*sqrt(sumsq(o)) 0.0020 0.0020 0.0017 0.0023 0.0017 0.0020 0.0020 Soluble Boron Reactivity, 800 ppm 0.8752 0.8783 0.8789 0.8823 0.8844 0.8856 0.8855 Soluble Boron Target Reactivity 0.9197 0.9189 0.9183 0.9177 0.9171 0.9166 0.9160 Interpolated Soluble Boron Requirement 415 438 433 459 472 495 488 Reactivity, Interpolated Soluble Boron 0.9170 0.9153 0.9168 0.9154 0.9157 0.9161 0.9147 Note: The reference reactivity is the target keff from Table 4.7.6c for the Region 2 enrichment and burnup combination for each case.

Holtec Report HI-2084175 4-69 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.17 Calculation of the Misloaded Fresh Fuel Assembly Accident Region 2 Enrichment (wt% U-235) 2.0 3.5 5.0 Region 2 Burnup (GWD/MTU) 9 32 50 Region 3 Enrichment (wt% U-235) 5.0 5.0 5.0 Region 3 Burnup (GWD/MTU) 39.00 39.00 39.00 Misloaded Fresh Fuel Assembly in Region 2 (outer row)

Reactivity, 0 ppm 1.0202 1.0175 1.0187 Reactivity, 2500 ppm 0.7972 0.8022 0.8068 Target keff (0.945-corrections)t 0.9196 0.9177 0.9160 Interpolated Soluble Boron Requirement, ppm 1127 1159 1212 Reactivity, Interpolated Soluble Boron 0.9015 0.9023 0.9001 Misloaded Fresh Fuel Assembly in Region 2 (inner row)

Reactivity, 0 ppm 1.0024 0.9989 1.0017 Reactivity, 2500 ppm 0.7786 0.7856 0.7917 Target keff (0.945-corrections)t 0.9196 0.9177 0.9160 Interpolated Soluble Boron Requirement, ppm 924 952 1020 Reactivity, Interpolated Soluble Boron 0.8986 0.9042 0.8998 Misloaded Fresh Fuel Assembly in Region 3 Reactivity, 0 ppm 0.9882 0.9874 0.9882 Reactivity, 2500 ppm 0.7809 0.7790 0.7800 Target keff (0.945-corrections)t 0.9196 0.9177 0.9160 Interpolated Soluble Boron Requirement, ppm 827 836 867 Reactivity, Interpolated Soluble Boron 0.9011 0.9043 0.9019 f Maximum of total corrections for the Region 2 and Region 3 enrichment and burnup combination for each case shown.

Holtec Report HI-2084175 4-70 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.18 Calculation of the Mislocated Fresh Fuel Assembly Accident Region 2 Enrichment (wt% U-235) 2.0 3.5 5.0 Region 2 Burnup (GWD/MTU) 9 32 50 Region 3 Enrichment (wt% U-235) 5.0 5.0 5.0 Region 3 Burnup (GWD/MTU) 39.00 39.00 39.00 Mislocated Case Reactivity, 0 ppm 1.0251 1.0250 1.0240 Mislocated Case Reactivity, 2500 ppm 0.7856 0.7946 0.7963 Target k-eff (0.945-corrections)'f 0.9196 0.9177 0.9160 Interpolated Soluble Boron Requirement, ppm 1101 1164 1186 Reactivity, Interpolated Soluble Boron 0.8956 0.8949 0.8967 t Maximum of total corrections for the Region 2 and Region 3 enrichment and bumup combination for each case shown.

Holtec Report HI-2084175 4-71 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.19 deleted Holtec Report HI-2084175 4-72 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.20 MCNP4a Calculations for the Fuel Rod Storage Basket Case keff 2 35 Fresh Fuel Assembly, 5.0 wt% U 0.9382 5.0 wt% 2 35 U at 39.58 GWD/MTU 0.7639 5.0 wt% 235U at 51.67 GWD/MTU 0.7305 FRSB Full with 5.0 wt% 235 U Fuel Pins 0.7263 (I

Holtec Report HI-2084175 4-73 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.21 Beaver Valley Power Station Unit No. 2 Spent Fuel Pool Boron Dilution Accident Analysis Parameters Value Technical Specification Soluble Boron Concentration 2000 ppm Soluble Boron Concentration to Maintain keff 0.9500 495 ppm Credited Volume of 2000 ppm Water in SFP 182,472 gallons High Flow Rate Accident Evaluation High Flow Rate (Service Water System) 3000 gpm Volume Needed to High Level Alarm Setpoint 7850 gallons Time to High Level Alarm Setpoint 2.62 minutes Time Needed to Dilute to 495 ppm 85 minutes Volume Needed to Dilute to 495 ppm 254901 gallons Time to Respond to High Level Alarm Setpoint 82.4 minutes Low Flow Rate Accident Evaluation Low Flow Rate 2 gpm Volume Needed to High Level Alarm Setpoint 7850 gallons Time to High Level Alarm Setpoint 2.73 days Time Needed to Dilute to 495 ppm 89 days Volume Needed to Dilute to 495 ppm 254901 gallons Time to Respond to High Level Alarm Setpoint 85.8 days Report HI-2084175 Holtec Report 4-74 Holtec Project 1702 Holtec HI-2084175 4-74 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.22 CASMO Calculations for Metamic Measurement Uncertainty Case Metamic Density Ref Case -5%

Burnup Enr Soluble Boron kinf Delta kinf GWD/MTU (ppm) 0.0 2 0 0.9648 0.0022 10.0 2 0 0.8889 0.0020 20.0 2 0 0.8277 0.0019 0.0 2 2000 0.7056 0.0016 10.0 2 2000 0.6753 0.0015 20.0 2 2000 0.6366 0.0014 0.0 3.0 0 1.0773 0.0025 10.0 3.0 0 0.9897 0.0023 20.0 3.0 0 0.9201 0.0021 30.0 3.0 0 0.8602 0.0020 0.0 3.0 2000 0.8346 0.0019 10.0 3.0 2000 0.7810 0.0018 20.0 3.0 2000 '0.7293 0.0017 30.0 3.0 2000 0.6830 0.0015 0.0 4.0 0 1.1471 0.0027 10.0 4.0 0 1.0607 0.0025 20.0 4.0 0 0.9922 0.0023 30.0 4.0 0 0.9315 0.0022 40.0 4.0 0 0.8763 0.0020 50.0 4.0 0 0.8275 0.0019 0.0 4.0 2000 0.9229 0.0021 10.0 4.0 2000 0.8625 0.0020 20.0 4.0 2000 0.8076 0.0018 30.0 4.0 2000 0.7574 0.0017 40.0 4.0 2000 0.7112 0.0016 50.0 4.0 2000 0.6705 0.0015 0.0 5.0 0 1.1954 0.0028 10.0 5.0 0 1.1129 0.0026 20.0 5.0 0 1.0476 0.0024 30.0 5.0 0 0.9893 0.0023 40.0 5.0 0 0.9348 0.0022 50.0 5.0 0 0.8840 0.0020 60.0 5.0 0 0.8378 0.0019 0.0 5.0 2000 0.9877 0.0022 10.0 5.0 2000 0.9260 0.0021 20.0 5.0 2000 0.8715 0.0020 30.0 5.0 2000 0.8212 0.0019 40.0 5.0 2000 0.7736 0.0018 50.0 5.0 2000 0.7292 0.0017 60.0 5.0 2000 0.6890 0.0016 HI-2084175 4-75 Holtec Project 1702 Holtec Report HI-2084175 Holtec, 4-75 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.23 Calculation of the Rack Displacement Accident Region 2 Enrichment (wt% U-235) 2.0 3.5 5.0 Region 2 Burnup (GWD/MTU) 9 32 50 Region 3 Enrichment (wt% U-235) 5.0 5.0 5.0 Region 3 Bumup (GWD/MTU) 39.00 39.00 39.00 Rack Displacement Case Reactivity, 0 ppm 0.9951 0.9943 0.9947 Rack Displacement Case Reactivity, 2500 ppm 0.7680 0.7785 0.7854 Target k-eff (0.945-corrections) f 0.9196 0.9177 0.9160 Interpolated Soluble Boron Requirement, ppm 831 888 940 Reactivity, Interpolated Soluble Boron 0.9007 0.9009 0.9003 f Maximum of total corrections for the Region 2 and Region 3 enrichment and bumup combination for each case shown.

Holtec Report HI-2084175 4-76 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Table 4.7.24 Calculations for the Fuel Elevator Case Parameter Result Fuel Elevator Mislocated Case Reactivity, 5.0 wt% Fresh, 0 ppm 1.0734 Fuel Elevator Mislocated Case Reactivity, 5.0 wt% Fresh, 2000 ppm 0.8654 Soluble Boron Target Reactivity 0.9132 Interpolated Soluble Boron Requirement, ppm 924 Reactivity, Interpolated Soluble Boron 0.9019 Holtec Report HI-2084175 4-77 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.5.1 128 IFBA Pin Loading Pattern for 17x 17 Fuel Assemblies

-. 0!

- * *6 S Si

101S ~90 S(

0 0o 0 0 et -N LE-GE NJD W IFBA ROD DI= -NON IFBA ROD TH-IM=L & 'NSTIRUMENTý T!~~LocTfrJ Holtec Report HI-2084175 4-78 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.5.2 r7 ~

'7 <44'

>44>

'44

' '7' 4 4 ~,'7*- '444(~*' ~743/4"~

'4 "'7 "'7

~

'4, 'p

'4444

<4444,44414444 vj'~ s >44 I

~ 44 1 24 4' 474 444 4 4 4 44~44 (444(4(<'4444444444>

A "n44414t44" ~ 3/4' <A '747744'

'7.t&4 4 4444y<A44%44 't~'#>&~ 444444' 44444 4' 4'.. '744,47 444" 41 4 4,..s4, 4~44444 'A'4 4 4 ' 44444,

~%'744~'S444;v4s'p4~4~'w" ~74447 I> A' t'4144~t~& 444444 '7~ ~.

444 44 444~~'~ ~ 4744'~ ' 4; 44~ < '7 44< 4~4\ 4$ t N 4' 444 *2>444 :~t't< 4 4 47444'7j

~24~tr> 4'~~>4&~' (4 (4

+4 4(74 4 4444 (444414 4424; 4<~44 '44444~44 i~

774

'744 44~14 '4' 44, 4 4 4,

~744~.24 1/241441

'4444 4444(44>44 '7t444~~ ,,.44~A~4' Holtec Report HI-2084175 4-79 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.5.3 MZTR 9 X 12 Rack Permissible Loading Configuration Region 1 / Gap Between Racks_

Note: This figure is shown for additional clarification of what the allowable Region 1 loading configurations are if the rack comers are not used as Region 1 cells. The requirement is to have at least one Region 2 cell separating Region 1 cells, even on the diagonal.

111-2084175 Report HI-2084175 4-80 Holtec Project 1702 Holtec Report 4-80 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.5.4a MZTR 9 X 13 Rack Permissible Loading Configuration Region 1 Gap Between Racks/

Note: This figure is shown for additional clarification of what the allowable Region 1 loading configurations are if the rack comers are not used as Region 1 cells. The requirement is to have at least one Region 2 cell separating Region 1 cells, even on the diagonal.

HI-2084 175 Report HI-2084175 4-81 Holtec Project 1702 Holtec Report 4-81 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.5.4b Example ofa MZTR 9 X 13 Rack With a Cut-Out Section Permissible Loading Configuration Region 1 / Gap Between Racks" Note: This figure is shown to demonstrate how racks with cut-out sections are handled. Note that this rack does not actually exist. For any rack with a cut-out, the cell loading is assumed to remain as if the cut-out was not present. Also, this figure is shown for additional clarification of what the allowable Region 1 loading configurations are if the rack comers are not used as Region 1 cells. The requirement is to have at least one Region 2 cell separating Region 1 cells, even on the diagonal.

175 4-82 Holtec Project 1702 Report HI-2084 Holtec Report HI-2084175 4-82 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.5.5 200 IFBA Pin Loading Pattern for 17x17 Fuel Assembly LEGEND IFBA ROD WL NON IFBA ROD THIMBLE & INSTRUMENT TUBE LOCATIONS Holtec Report HI-2084175 4-83 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.5.6 A Two Dimensional Representation of the MCNP4a FRSB Model 175 4-84 Holtec Project 1702 Holtec Report HI-2084 HI-2084175 4-84 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.5.7 Spent Fuel Pool Mixed Zone Three Region Layout 1ý 11 .1 RI R2 R3 221212121212 2212 21212 m22 18 33 0 122222222221 8 43 0 22121212122 12 222222 221 22 Fi2 1 2 21 o 16 35 12 21 8 8 35 22 22g 2r2 2121212I112k122 2 2 0 16 35 12 21n 1 2 2 1 8 8 35 "1005"1000 "l2 n 22 22 2 2 2 2 0 14 35 an inn: n2 21 Io211l12112 an1121 1 2 2 1 7 7 35 222222222222 0 49 0 Ul 22 222 "n22 121121121121121 11211211 25 24 0 2222 222 222 an 2 2 212121212121a 26 25 0 1212 212 212 22 22 n2 1111212111121 OR] 2la222 22 " Hnn2n2111m 2:1121112111211121 11211 1 21 11 0 8

51 8

0 35 222222222222 o 16 35 12o2 2 2 2 2 8 8 35 0 16 35 22122 1112 211 8 8 35 21 1 12 2 2 ma112 ninn 11 0 51 0 222222222222n 212 1201 22 22 22l .21 21222 1 212 1 26 25 0 12"Hunaa 12111211121 1121 1121 1 1 211 2 11 26 25 0 2L22 22 22 22 22 222 0 51 0 nIoaI 1.2 IN"e 21 an I

222222222222a 21212121212 1 2 2 2 1 2 2

2 2

1 2

1 8

0 8

8 16 8

35 35 35 2 2 2 2 0 16 35 n22o 7112 271 7 7 35 ailo 2 2 2 unmtnn 2 22 22 22 22 22 22 2 0 49 0 U2112112112 12 12 12 1 25 24 0 1121 1121 1 121112 1 121 112 16 16 0 212 2 22 22 2 2 0 32 0 no 21 6 6 20 2o 0 10 20 21 5 5 20 cutot nn 0 10 20 21n 5 5 20 22222222n 0 30 0 15 15 0 Totals 271 759 660 1690 Holtec Project 1702 HI-2084 175 Holtec Report HI-2084175 4-85 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.5.8 A Two Dimensional Representation of the Region 3 MCNP4a Single Cell Model HI-2084 175 4-86 Report HI-2084175 Holtec Report 4-86

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.7.1 a Region 2 Initial Enrichment and Burnup Combinaitons For Enriched Blankets Profile (Includes 5% bumup uncertainty, and polynomial fit adjustment) 60 50 40 30 I. 20 10 0

2.0 2.5 3.0 3.5 4.0 4.5 5.0 wt% U-235 4-87 Holtec Project 1702 Holtec Report HL-2084 Holtec 175 HI-2084175 4-87 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.7.1b Region 2 Initial Enrichment and Bumup Combinaitons For Natural Blankets Profile (Includes 5%burnup uncertainty, and polynomial fit adjustment) 60 50 48.14 41.56*. *t 40Acceptable for Placement in Region 2

,.40 30 27.13

" 19.2 7 ",

"20 10 9Not Acceptable for Placement in Region 2 10 2.0 2.5 3.0 3.5 4.0 4.5 5.0 wt% U-235, 4-88 Holtec Project 1702 Holtec Report Holtec HI-2084 175 Report FH-2084175 4-88 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.7.1c Region 2Initial Enrichment and Bumup Combinaitons For No Blankets Profile (Includes 5%burnup uncertainty, and polynomial fit adjustment) 60 50 " 48.17 Acceptable for Placement inRegion 2416

40o

30 -27.29--

20 -

1119Not Acceptable for Placement in Region 2 10 0

2.0 2.5 3.0 3.5 4.0 4.5 5.0 wt% U-235 4-89 Holtec Project 1702 Holtec Report HI-2084 175 FH-2084175 4-89 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.7.2a Region 3 Initial Enrichment and Burnup Combinaitons For Enriched Blankets Profile (Includes 5%bumup uncertainty, and polynomial fit adjustment) 50 40 30 I-20 10 C

.0 2.0 2.5 3.0 3.5 4.0 4.5 5.0 wt% U-235 4-90 Holtec Project 1702 Holtec Report HI-2084175 HI-2084175 4-90 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.7.2b Region 3 Initial Enrichment and Bumup Combinaitons For Natural Blankets Profile (Includes 5%burnup uncertainty, and polynomial fit adjustment) 50 40 30 I-.

20 10 0

2.0 2.5 3.0 3.5 4.0 4.5 5.0 wt% U-235 Holtec Report HI-2084175 4-91 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.7.2c Region 3Initial Enrichment and Burnup Combinaitons For No Blankets Profile (Includes 5%bumup uncertainty, and polynomial fit adjustment) 50 41.97 40 40*'35.47

,-, Acceptable for Placement inRegion 3

" 30

' 20 E 15.19 10 R Not Acceptable for Placement in Region 3 1.)3 0

2.0 2.5 3.0 3.5 4.0 4.5 5.0 wt% U-235 HI-2084 175 4-92 Holtec Project 1702 Holtec Report 1-H-2084175 4-92 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.7.3 A Two Dimensional Representation of the MCNP4a Mislocated Fuel Assembly Accident Model Holtec Report HI-2084175 4-93 Holtec Project 1702

SHADED AREAS DENOTE PROPRIETARY INFORMATION Figure 4.7.4 Representation of Misloading Accident MCNP Model with Reflective Boundary Conditions Holtec Report HI-2084175 4-94 Holtec Project 1702

6.0 THERMAL-HYDRAULIC EVALUATION 6.1 Introduction Beaver Valley Power Station (BVPS) Unit No. 2 is a Pressurized Water Reactor (PWR) nuclear power plant operated by FirstEnergy Nuclear Operating Company (FENOC). The plant contains a Fuel Storage Pool (FSP) for holding offloaded spent fuel assemblies, with an attendant FSP cooling system (FSPCS). The FSP is currently equipped with spent, fuel storage racks containing the neutron absorber Boraflex, which must be replaced. To maximize storage capacity, FENOC proposes to replace all of the existing racks with maximum density, non-flux-trap storage racks.

In the reracked configuration, 15 new fuel racks are added to replace the removed racks. The total number of storage locations in the FSP will increase from 1088 to 1690.

This section provides a summary of the methods, models, analyses and numerical results to demonstrate that the BVPS Unit No. 2 fuel storage pool meets the thermal-hydraulic requirements for safe storage of spent fuel set forth in Section 6.2 herein. Similar thermal-hydraulic analyses have been used in fuel storage pool licensing applications at many nuclear plants worldwide (see Table 6.1.1 for a partial list).

The following specific thermal-hydraulic analyses for the BVPS Unit No. 2 FSP are performed:

1. Calculation of the spent fuel decay heat. The decay heat contributions from both previously stored fuel assemblies and recently discharged fuel assemblies are considered.

2. Determination of the FSP bulk thermal response versus time in accordance with each discharge scenario. For normal refueling conditions the maximum allowable cooling water temperature is determined as a function of the refueling start time (i.e., in-core hold time or decay time).

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

4. A rigorous Computational Fluid Dynamics (CFD) based study to conservatively quantify the peak local water temperature in the FSP.

5. Determination of a bounding maximum fuel cladding temperature.

Holtec Report HI-2084175 6-1 Holtec Project 1702

These analyses are described in detail in Sections 6.4 through 6.7. The following discharge scenarios are postulated and analyzed:

" Normal Full Core Offload [Case I]:

A full core (157 fuel assemblies) is transferred to the FSP at the end of a normal operating cycle at the rate of 6 assemblies per hour. The heat load from this recently discharged batch and the background heat load from previous discharges (1541 assemblies from 23 previous operating cycles, for a total of 1698 assemblies from 24 operating cycles) are removed by one FSPCS pump (the second pump is assumed a single active failure) and both FSPCS heat exchangers.

For this scenario, the FSP bulk temperature is evaluated for refueling start times ranging from 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months after reactor shutdown and the component cooling water inlet temperature necessary to meet acceptance criterion 1 in Subsection 6.2 is determined through a predict-and-correct approach. Three different refueling start times (100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , 125 hours0.00145 days 0.0347 hours 2.066799e-4 weeks 4.75625e-5 months and 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months ) are evaluated and these are labeled as Cases Ia through Ic.

Abnormal Full Core Offload [Case II]:

A full core (157 fuel assemblies) is transferred to the FSP starting after 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months of decay time at the rate of 6 assemblies per hour. Additionally, 72 fuel assemblies have 36 days decay and another 72 assemblies stored in the pool have 400 days decay. Both FSPCS pumps and both FSPCS heat exchangers are operating.

" Loss of Cooling Scenario [Case III]:

Following each full core discharge, a loss of FSP cooling occurs at the maximum FSP bulk temperature. The time required for the fuel storage pool water surface to reach the saturation temperature is evaluated.

In the sections that follow, analysis methods are described, results are presented and discharge scenarios are evaluated.

Holtec Report HI-2084175 6-2 Holtec Project 1702

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

a. NUREG-0800, Standard Review Plan, Section 9.1.3, Revision 1.

b. USNRC OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Application, 4/78 [6.2.1].

The design of the rack modules must ensure that fuel assemblies are adequately cooled by natural circulation of water for the bounding discharge conditions. The BVPS Unit No. 2 FSP storage is evaluated to the following criteria:

1. Under a normal full core offload with a single active FSPCS failure (Case I, refer to Section 6.1), the bulk pool temperature shall be limited to 170 0 F. Local temperatures in the rack cells shall be demonstrated to be below the local saturation temperature.

2. Under an abnormal full core offload without any FSPCS failures (Case II, refer to Section 6.1), the bulk pool temperature shall be limited'to 173TF. Local temperatures in the rack cells shall be demonstrated to be below the local saturation temperature.

3. Under a loss of cooling scenario and a full core offload (Case III, refer to Section 6.1), the pool surface is allowed to reach saturation. For a loss of cooling scenario, sufficient time must be available to implement corrective measures.

6.3 Assumptions and Design Data 6.3.1 Assumptions The following assumptions are applied to render a conservative portrayal of thermal-hydraulic conditions in the BVPS Unit No. 2 FSP.

Holtec Report HI-2084175 6-3 Holtec Project 1702

6.3.1.1 Maximum Bulk FSP Temperature and Minimum Time-To-Boil Calculation

1. On the basis of maximum storage capacity (1690 cells), a total of 1698 fuel assemblies from 24 operating cycles are assumed. This conservatively overstates the number of fuel assemblies in the pool under all analyzed scenarios (Case I through Case III).

2. Heat loss by natural convection and mass diffusion from the water surface, conduction through the pool walls and by radiation cooling from the pool water surface is neglected.

Thus, all SFP heat loads, being decay heat and FSPCS pump heat, in the temperature cases are considered to be removed by the FSP heat exchangers, maximizing temperature.

3. The thermal capacity of the FSP is based on the net water volume only, completely neglecting the thermal capacity of the fuel assemblies, racks, liner and concrete FSP structure. The water in the cask area, which although separate can be in communication with the FSP, is also neglected. Since this assumption understates the FSP thermal capacity, it results in faster computed heat-up rates and shorter times-to-boil.

4. The decay heat load contribution of previously discharged fuel assemblies is assumed constant during all discharge scenarios. This assumption is conservative because it neglects the exponential decay of the heat generation from these old fuel assemblies.

5. All postulated (i.e., future) refueling batches are assumed to have a bounding exposure (62 GWD/MTU). This conservatively maximizes the decay heat load associated with these fuel assemblies.

6. The postulated full core offloads are assumed to have two regions. The first region, equivalent in size to a maximum refueling batch (e.g., 72 assemblies), is assumed to have a bounding exposure (62 GWD/MTU). The second region, comprising the remainder of the assemblies in the full core, has an exposure that bounds the average of once and twice burned assemblies (42.6 GWD/MTU). This conservatively maximizes the decay heat load associated with these fuel assemblies.

7. The loss of forced cooling of the FSP is assumed to occur when the maximum bulk FSP temperature for each discharge scenario occurs. This minimizes the time-to-boil.

6.3.1.2 Maximum Local Water and Fuel Clad Temperature Calculation

1. All passive losses (i.e., conduction through walls and slab or losses from the surface) are completely neglected. This conservatively maximizes the pool heat load, maximizing both global and local temperatures.

Holtec Report HI-2084175 6-4 Holtec Project 1702

2. All calculations are performed with the hydraulic resistance of the most resistive spent fuel assembly type in the most resistive storage rack type assigned to every rack storage location.

This conservatively maximizes the overall hydraulic resistance of the rack.

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

4. The large flow holes in the rack base plate are not credited, and all rack cells are assumed to have the side flow hole geometry of the pedestal cells. This conservatively reduces the water flow area into the storage cells, thereby increasing the hydraulic resistance.

5. The hydraulic resistance of every rack cell in each spent fuel storage rack includes the inertial resistance that would result from a dropped fuel assembly lying across the top of the rack. This conservatively increases the total rack cell hydraulic resistance and bounds the thermal-hydraulic effects of a fuel assembly dropped anywhere in the spent fuel storage area.

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

7. Instead of explicitly modeling the pipes that supply and remove water from the FSP (and direct it into the FSP cooling system), the inlets and outlets are modeled as 4" high slots along the length of north and south FSP walls just below the water surface. As the mixing of relatively warmer and cooler water within the FSP is dominated by buoyancy effects, this geometric simplification does not have a significant effect on the calculated results.

8. An additional heat transfer resistance of 0.0005 (hrxft2x°F)/Btu is conservatively imposed on the outside of the fuel rods, to account for any crud layer, thereby increasing the calculated fuel clad superheat.

9. The maximum local water temperature (at the fuel rack cell exit) and peak heat flux (typically near the mid-height of the active fuel region) are considered to occur co-incidentally. 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.

6.3.2 Design Data 6.3.2.1 Background Decay Heat Calculation The decay heat calculations for both old and recently discharged fuel assemblies are performed using the ORIGEN2 computer program from Oak Ridge National Laboratory [6.2.2]. The Holtec Report HI-2084175 6-5 Holtec Project 1702

cumulative background decay heat load from 1541 assemblies total from 23 operating cycles is conservatively assumed, at which time the ability to accommodate a full core discharge will be lost. To maximize decay heat, a bounding exposure is assumed for all fuel assemblies.

6.3.2.2 Maximum Bulk FSP Temperatures Calculation The principal input data employed to determine the maximum FSP bulk temperature of the BVPS Unit No. 2 FSP is summarized in Table 6.3.1.

6.3.2.3 Maximum Local Water and Fuel Clad Temperature In addition to the input data listed in Table 6.3.1, the principal input data employed for the local thermal-hydraulic analyses are presented in Table 6.3.2.

6.4 Fuel Rod Cladding Temperatures The peak fuel rod cladding temperature is computed by following a series of calculation steps as outlined below:

Step 1: Compute the maximum local water temperature in the FSP for the stipulated discharge scenarios. This procedure is described in Section 6.7.

Step 2: Compute the maximum cladding to local water temperature difference (ATe).

Step 3: Compute the peak fuel rod cladding temperature by adding ATc to the maximum local water temperature.

The final results are discussed in Section 6.7. The procedure to perform Step 2 is presented next.

The maximum specific decay power of a single fuel assembly among the recently discharged batch of assemblies is denoted by QA. A fuel rod can produce f, times the average heat emission rate over a small length, where f, is the axial peaking factor. The axial heat distribution in a fuel Holtec Report HI-2084175 6-6 Holtec Project 1702

rod is maximum in the central region, and tapers off at its two extremities. Thus, peak cladding heat flux per unit heat transfer area of fuel assembly is given by the equation:

_QAxfz A=QA qpeak where:

Arod is the total external heat transfer 2

area of the cladding in the active fuel region of a single fuel assembly, ft Within each fuel assembly sub-channel, water is continuously heated by the cladding as it moves axially upwards from bottom to top under laminar flow conditions. Rohsenow and Hartnett

[6.4.1 ] report a Nusselt number, Nu, for heat transfer in a laminar flow situation through a heated channel. Nu is defined as follows:

h Nu= h*c Dh =4.364 kwater he= 4.364 x water Dh where:

kwater is the water thermal conductivity, Btu/(hr-ft-°F) h, ' is the laminar flow convective heat transfer coefficient, Btu/(hr-ft2-°F)

Dh is the sub-channel hydraulic diameter, ft In order to introduce some additional conservatism in the analysis, we assume that the fuel cladding has a crud deposit thermal resistance, Rcrud, that covers the entire surface. Therefore, the overall heat transfer coefficient U, considering a crud deposit resistance Rcrud, can be defined by the following:

U=

The temperature drop, AT,, between the outer surface of the fuel cladding and the water flowing up through the assembly at the peak cladding flux location is computed by the following:

Holtec Report HI-2084175 6-7 Holtec Project 1702

ATq peak U

Finally, the maximum fuel rod temperature is defined by the following:

Trod = TlocaI + AT, where:

Trod is the maximum fuel clad temperature Tiocal is the maximum local'water temperature in the hottest cell calculated by the local temperature analysis (Section 6.7) 6.5 Description of Fuel Storage Pool Cooling System The BVPS Unit No. 2 FSPCS is designed to remove the decay heat produced by spent fuel assemblies stored in the pool following a unit refueling and accumulated fuel from previous discharges. The system incorporates two 100% capacity cooling trains designed to maintain the FSP at or below 140'F with a component cooling water supply at 100°F under non-refueling conditions.

Each cooling train of the FSPCS incorporates a heat exchanger, pump, associated piping, valves, and instrumentation. Each cooling train is designed to service the FSP for the design-basis heat loads and to maintain the bulk temperature of the pool water below 140'F during non-refueling operation. The FSPCS transfers decay heat from stored fuel via the fuel pool heat exchanger to the component cooling water system.

The FSPCS heat exchangers are shell-and-tube construction. Fuel storage pool water circulates through the tubes while component cooling water is circulated on the shell side. Even with the sngle active failure of an FSPCS pump, both heat exchangers are available. The design flow rates and inlet temperatures are provided in Table 6.3.1.

Holtec Project 1702 Holtec Report HI-2084175 Holtec Report HI-20841756-8 6-8 Holtec Project 1702

6.6 Heat Loads and Bulk Pool Temperatures 6.6.1 Background Decay Heat Load Calculation An analysis is performed to calculate the accumulated decay heat of all previously discharged fuel assemblies (i.e., assemblies with at least one year of decay time) stored in the FSP prior to recently discharged fuel assemblies. The decay heat is calculated using Oak Ridge National Laboratory's ORIGEN2 program [6.2.2]. The cumulative decay heat from previous discharges (Peons) irradiated to a bounding exposure is computed as approximately 5.6 million Btu/hr for a normal full core offload and 6.0 million Btulhr for an abnormal full core offload. The heat added to the flowing pool water by the FSPCS pump (conservatively assumed to slightly exceed the rated pump motor power) is then added to these values, yielding a total pool heat of approximately 5.7 million Btu/hr for a normal full core offload and 6.1 million Btuihr for an abnormal full core offload. This total heat, Pco, 0 s, is included in the bulk pool temperature calculations as a constant background heat input as described next.

6.6.2 Maximum Bulk FSP Temperature Calculation This analysis is performed to determine the transient FSP bulk temperature and decay heat profiles for the postulated discharge of fuel assemblies into the FSP.

The mathematical formulation for this analysis can be explained with reference to the simplified heat exchanger alignment shown in Figure 6.6.1. The governing differential equation for bulk pool temperature can be written by utilizing conservation of energy as:

dT C x-= Pcons + Q(r)-QHx(T)-QEv(T, TA) dr where:

C is the thermal capacity of water in the pool, Btu/°F Pcons is the heat generation rate from "old" fuel + pump heat, Btu/hr Q(T) is the heat generation rate from recently discharged fuel vs. time, Btulhr QHX(T) is the FSP cooling system heat rejection, Btu/hr QEv(T,TA) is the evaporative and passive sensible heat losses, Btu/hr Holtec Report HI-2084175 6-9 Holtec Project 1702

T is the bulk pool temperature, OF TA is the fuel building ambient temperature, OF The FSPCS heat rejection, QHX(T) is defined by the following governing equation.

QHx (T)=Wc xcp x px (T - T) where:

Wc is the cooling water flow rate, lb/hr cp is the cooling water specific heat, Btu/(lb-°F) p is the temperature effectiveness of heat exchanger T is the bulk pool temperature, °F Tc is the cooling water inlet temperature, OF The equation used to determine the temperature effectiveness, p of the FSPCS heat exchanger is as follows:

P TCo - TC Tpi -TC where:

Tc is the cooling water inlet temperature, OF Tco is the cooling water outlet temperature, OF Tpi is the pool water inlet temperature, OF A computer program used in the nuclear industry for heat exchanger thermal performance rating is used to calculate the temperature values for this equation. In performing thermal performance prediction evaluations, the inlet temperatures of each flow stream are inputs to the program. Thus, the only calculated value that is input to this equation is the cooling water outlet temperature (Tco).

The FSP decay heat contribution from all previously stored fuel assemblies is held constant during the entire analysis because its decrease with decay time after shutdown can be conservatively neglected. The decay heat generation, Q(c), of the recently discharged fuel will decay exponentially with elapsed time after reactor shutdown. The decay heat generation Q(T) is a function of the elapsed time after reactor shutdown, number of fuel assemblies discharged, and in-core exposure time. For all discharge scenarios, the FSP decay heat contribution from the recently discharged fuel equals Q(T).

6-10 Holtec Project 1702 Holtec Report HI-2084175 Holtec Report HI-2084 1756-10 Holtec Project 1702

The evaporative and passive sensible heat losses, QEv(TTA), are a nonlinear function of pool temperature (T) and ambient temperature (TA), and include cooling by evaporation and natural convection heat transfer from the pool surface. For conservatism, these cooling mechanisms are completely neglected.

The results of the maximum bulk pool temperature analyses for the discharge scenarios are summarized in Table 6.6.1. The results demonstrate that bulk pool water temperatures remain below the prescribed limits during fuel discharges. These results comply with the acceptance criteria set forth in Section 6.2.

The decay heat load profiles and the bulk pool temperatures in the fuel storage pool as functions of time (after reactor shutdown) are shown in Figures 6.6.2 through 6.6.4 for the normal full-core discharge scenarios and in Figure 6.6.5 for the abnormal full-core discharge scenario. For the normal full-core offload scenarios, the maximum allowable cooling water temperature is plotted as a function of refueling start time in Figure 6.6.6. The maximum pool decay heat is reached upon completion of fuel transfer. The thermal inertia of the pool water delays the bulk pool temperature maximum, the lag being a direct result of the system thermal capacitance. The coincident time to the maximum temperature is the summation of. refueling start time, fuel transfer time and the lag time.

6.6.3 Minimum Time-To-Boil Calculation This analysis is to determine the time that it takes for the pool water to boil, in case all forced cooling becomes unavailable. Clearly, the most critical instant of loss-of-cooling is when the pool water temperature has reached its maximum value. Although the probability of a loss-of-cooling event occurring when the pool water is hottest is low, the calculations are performed based on the hottest possible initial temperature. The following differential equation governs the thermal response of the water in the fuel storage pool without including the heat rejection terms.

Holtec Report HI-2084175 6-11 Holtec Project 1702

dT C(r)x dr- = Pcons + Q(r) where:

C(t) is the time-varying thermal capacity of the pool, Btu/°F Pos 0 is the heat generation rate from "old" fuel, Btu/hr Q(t) is the heat generation rate from recently discharged fuel vs. time, Btu/hr T is the bulk pool temperature, 'F The time-to-boil calculations are conservatively performed assuming no makeup water is available. The time-to-boil results are summarized in Table 6.6.2. The results show that a, minimum of nearly 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is required for the pool water to start boiling. For a scenario of loss of all fuel storage pool cooling, sufficient time would be available for necessary repairs or to implement alternate cooling or makeup water.

6.7 Local Water and Fuel Cladding Temperatures The objective of the local temperature analysis is to demonstrate that the principal thermal-hydraulic criteria of ensuring local subcooled conditions in the pool are met for all scenarios.

Adequate cooling of recently discharged fuel in the fuel pool is demonstrated by performing a rigorous evaluation of the velocity and temperature fields in the pool created by the interaction of buoyancy driven flows and water injection/egress. A 3-Dimensional Computational Fluid Dynamics (CFD) analysis for this demonstration is implemented.

There are several significant geometric and thermal-hydraulic features of the BVPS Unit No. 2 FSP 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 bulk pool region where the classical Navier-Stokes equations are solved with turbulence effects included. The other region is the heat-generating zone of spent fuel storage racks loaded with fuel assemblies, located near the FSP bottom. In this region, water flow is directed vertically upwards by the buoyancy forces through relatively small flow channels formed by the fuel assembly rod arrays in each rack cell.

This situation is modeled as a porous medium region in which Darcy's Law [6.7.1 ] governs fluid flow.

Holtec Project 1702 Holtec Report HI-2084175 Holtec Report HI-20841756-12 6-12 Holtec Project 1702

The distributed heat sources in the fuel storage pool racks are modeled by identifying distinct heat generation zones considering full-core discharge, peaking effects, and presence of background decay heat from old discharges. Three heat generating zones were modeled. The first consists of background fuel assemblies from previous discharges, while the remaining two zones are for fuel assemblies recently discharged from the reactor. This is a conservative model, since all recently discharged fuel assemblies with higher than average decay heats are postulated to be placed in a contiguous area.

The CFD analysis is performed using the FLUENT [6.7.2] fluid flow and heat transfer modeling program. The FLUENT code enables buoyancy flow and turbulence effects to be included in the CFD analysis. Turbulence effects are modeled by relating time-varying "Reynolds' Stresses" to the mean bulk flow quantities by the standard k-c ELturbulence model.

A single local temperature scenario that bounds both the normal and abnormal full core offload conditions defined in Section 6.2 is considered. A solution of the CFD model is performed to obtain the FSP flow and temperature fields. Temperature contours in a vertical plane through the hottest fuel assemblies are shown in Figure 6.7.1. The plot confirms that hot fuel is safely and reliably cooled by natural convection action. Local hot spots induced by water circulation in the racks are rapidly dissipated in the pool water resulting in a nearly uniform temperature distribution away from the hot racks. The bounding local water temperature in the FSP racks and the fuel clad superheat for the bounding scenario are summarized in Table 6.7.1. At the top of the active fuel, the local saturation temperature is approximately 240'FK From the local water and fuel cladding temperature results, it is concluded that local water and fuel cladding temperatures remain below saturation.

During installation of the new racks in the FSP, one new rack will temporarily be placed in the cask pit to provide additional fuel storage space. This is needed to provide enough fuel storage space to permit emptying existing racks for removal. Only fuel assemblies with at least 18 months of cooling time may be placed in the rack in the cask pit. As part of the new rack installation sequence, all fuel in the rack in the cask pit will be shuffled into the FSP and the rack moved to its final position in the FSP. An additional CFD model of the rack in the cask pit was Holtec Report HI-2084175 6-13 Holtec Project 1702

constructed and analyzed, and demonstrated that the rack in cask pit condition is bounded by the normal condition with all racks in the FSP.

6.8 References

[6.2.1] "OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14, 1978.

[6.2.2] A, G. Croff, "ORIGEN 2 - A Revised and Updated Version of the Oak Ridge Isotope Generation and Depletion Code," ORNL-5621, Oak Ridge National Laboratory, 1980.

[6.4.1] Rohsenow, W.M. and J.P. Hartnett, "Handbook of Heat Transfer," McGraw Hill Book Company, NY, 1973.

[6.7.1] "Flow of Fluids Through Valves, Fittings, and Pipe," Crane Technical Paper No. 410, Crane Valve Company, Twenty-Second Printing, 1985.

[6.7.2] FLUENT Computational Fluid Dynamics Software, Fluent Inc., Centerra Resource Park, 10 Cavendish Court, Lebanon, NH 03766.

Holtec Report HI-2084175 6-14 Holtec Project 1702

Table 6.1.1 PARTIAL LISTING OF RERACK APPLICATIONS USING SIMILAR METHODS OF THERMAL-HYDRAULIC ANALYSIS, PLANT DOCKET NO.

Enrico Fermi Unit 2 USNRC 50-341 Quad Cities 1 and 2 USNRC 50-254, 50-265 Rancho Seco USNRC 50-312 Grand Gulf Unit 1 USNRC 50-416 Oyster Creek USNRC 50-219 Pilgrim USNRC 50-293 V.C. Summer USNRC 50-395 Diablo Canyon Units 1 and 2 USNRC 50-275, 50-455 Byron Units 1 and 2 USNRC 50-454, 50-455 Braidwood Units 1 and 2 USNRC 50-456, 50-457 Vogtle Unit 2 USNRC 50-425 St. Lucie Unit 1 USNRC 50-425 Millstone Point Unit 1 USNRC 50-245 D.C. Cook Units 1 and 2 USNRC 50-315, 50-316 Indian Point Unit 2 USNRC 50-247 Three Mile Island Unit 1 USNRC 50-289 J.A. FitzPatrick USNRC 50-333 Shearon Harris Unit 2 USNRC 50-401 Hope Creek USNRC 50-354 Kuosheng Units 1 and 2 Taiwan Power Company Ulchin Unit 2 Korea Electric Power Corp.

Laguna Verde Units 1 and 2 Comision Federal de Electricidad Zion Station Units 1 and 2 USNRC 50-295, 50-304 Sequoyah USNRC 50-327, 50-328 La Salle Unit One USNRC 50-373 HI-2084 175 6-15 Holtec Project 1702 Holtec Report HI-2084175 6-15 Holtec Project 1702

Table 6.1.1 (continued)

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

Duane Arnold USNRC 50-331 Chin Shan Units 1 and 2 Taiwan Power Company Fort Calhoun USNRC 50-285 Nine Mile Point Unit One USNRC 50-220 BVPS Unit No. 1 USNRC 50-334 Limerick Unit 2 USNRC 50-353 Ulchin Unit I Korea Electric Power Corp.

J.A. Fitzpatrick USNRC 50-333 Callaway USNRC 50-483 Wolf Creek USNRC 50-482 Hatch Units 1 & 2 USNRC 50-321, 50-366 Harris Pools C and D USNRC 50-401 Waterford 3 USNRC 50-382 Holtec Report HI-2084175 6-16 Holtec Project 1702

Table 6.3.1

SUMMARY

OF INPUTS FOR BULK TEMPERATURE ANALYSIS INPUT DATA VALUE FSP Heat Exchanger Cooling Water Flow Rate (total for two exchangers) 1,094,020 lb/hr FSP Heat Exchanger Inlet Temperatures CCW Inlet Temperature - Normal Various (see Figure 6.6.6)

CCW Inlet Temperature - Abnormal 97.1 OF FSP Water Temperature - Initial 140°F North Wall: 471 in South Wall: 304 in Pool Nominal Dimensions East Wall: 260 in West Wall: 351 in Minimum Water Depth of FSP 38.5 ft Nominal Height of Spent Fuel Racks 180 3/4 in Reactor Thermal Power 2918 MW (including uncertainty)

Reactor Core Size 157 assemblies Refueling Start Time Normal Offload 100 hr, 125 hr, 150 hr Abnormal Offload 100 hr Fuel Transfer Rate 6 per hour Maximum Refueling Batch Size 72 assemblies Cycle Length 18 months Number of Fuel Storage Cells 1690 Holtec Report HI-2084175 6-17 Holtec Project 1702

Table 6.3.2

SUMMARY

OF INPUTS FOR LOCAL TEMPERATURE ANALYSIS INPUT DATA VALUE Rack-to-Wall Gaps North Wall 4 in South Wall 9 in East Wall 2 in West Wall 4 in Rack Bottom Plenum Height 11 3/8" (conservatively lower than actual)

Rack Baseplate + Cell Height 169 3/8" Active Fuel Length 144 in Fuel Assembly Array Size 17x 17 Rack Cell Pitch 9.03 in Minimum Rack Cell Nominal ID 8.8 in Number of Side Holes per Cell 4 Pedestal Cell Side Hole Diameter 1 1/2" Assembly Axial Peaking Factor 1.55 Holtec Report HI-2084175 6-18 Holtec Project 1702

Table 6.6.1

SUMMARY

OF BULK POOL TEMPERATURE RESULTS Component Maximum FSP Coincident Time Coincident FSP Discharge Cooling Water Bulk After Shutdown Heat Load Scenario Inlet Temperature Temperature (hr) (BtSdhe)

(OF) (OF)_(hr_(Bthr Normal Full Core, 100 Hour 97.1 170.0 136 36.11x10 6 Refueling Start (Case Ia)

Normal Full Core,125Hour 101.3 169.9 161 34.04x106 Refueling Start (Case lb)

Normal Full Core, 150 Hour 104.8 170.0 187 32.30x 106 Refueling Start (Case Ic)

Abnormal Full 97.1 170.3 134 42.96x10 6 Core (Case II) _

Note: For the normal full core offload scenarios, the component cooling water inlet temperature is an output of the analysis, while for the abnormal full core offload scenario it is an input, set equal to the minimum value obtained for the normal offload. For the normal full core offload scenarios, a single active failure of a FSPCS pump is assumed, and for the abnormal full core offload scenario no FSPCS pump failure is included.

Holtec Report HI-2084175 6-19 Holtec Project 1702

Table 6.6.2

SUMMARY

OF TIME-TO-BOIL RESULTS Discharge Time-to-Boil Time-to- lO' Above Bounding Maximum Racks Boil-Off Rate Scenario (hrs) (hrs) (gpm)

Normal Full Core (Case I) 2.24 25.3 77.0 Abnormal Full 1.87 21.1 91.8 Core (Case II)

Holtec Report HI-2084 175 6-20 Holtec Project 1702

Table 6.7.1

SUMMARY

OF LOCAL TEMPERATURE RESULTS Parameter Calculated Value Maximum Water Temperature (OF) 202.0 Cladding Superheat (OF) 24.9 Peak Cladding Temperature (IF) 226.9 Holtec Report HI-2084175 6-21 Holtec Project 1702

EVAPORATION LOSS QEv(T,TA) (conservatively neglected)

SFP THERMAL CAPACITY dr HEAT LOAD (PREVIOUSLY AND RECENTLY DISCHARGED FUEL)

Pcons + Q('[)

SFP COOLING SYSTEM HEAT EXCHANGER QHx(T)

T /00/M 14C T.

I Figure 6.6.1: Simplified Heat Exchanger Alignment Holtec Report HI-2084175 6-22 Holtec Project 1702

[- SFP Heat Load - -SFP Bulk Temperature 4.00E+07 172 3.60E+07 165 3.20E+07 158 2.80E+07 .................

.. 15 1 2.40E+07 144 0

0 137 -2 2.OOE+07 0

E 1.60E+07 130 I 1.20E+07 123 8.OOE+06 ------. ..... . ..... .1 16 4.OOE+06 -- - - -. JI. 109 0.OOE+00 102 0 50 100 150 200 250 300 Time After Reactor Shutdown (hr)

Figure 6.6.2: FSP Heat Load and FSP Bulk Temperature Profiles - Normal Full Core Offload - 100 Hour Refueling Start Time 6-23 Holtec Project 1702 HI-2084 175 Holtec Report HI-2084175 6-23 Holtec: Project 1702

-- SFP Heat Load - - SFP Bulk Temperature 3.60E+07 172 3.20E+07 165 2.80E+07 158

-~2.40E+07 151 w2.00E+07 144 0

0 0.

S1 .60E+07 137 E 0

x1.20E+07 130 8.OOE+06 123 4.OOE+06 116 0.00E+00 J 109 0 50 100 150 200 250 300 350 400 Time After Reactor Shutdown (hr)

Figure 6.6.3: FSP Heat Load and FSP Bulk Temperature Profiles - Normal Full Core Offload - 125 Hour Refueling Start Time HI-2084 175 Report HI-2084175 Holtec Report 6-24 Holtec Project 1702 Holtec 6-24 Holtec Project 1702

- SFP Heat Load - -SFP Bulk Temperaturej 3.60E+07 172 3.20E+07 165 2.80E+07 158

.~2.40E+07 151 (A2.00E+07 144 0

-J 0

S1 .60E+07 137 E 0

I-x 1.20E+07 130 8.00E+06 123 4.OOE+06 - 116 0.00E+00 109 0 50 100 150 200 250 300 350 400 450 500 Time After Reactor Shutdown (hr)

Figure 6.6.4: FSP Heat Load and FSP Bulk Temperature Profiles - Normal Full Core Offload - 150 Hour Refueling Start Time Holtec Report HI-2084175 6-25 Holtec Project 1702

- SFP Heat Load - - SFP Bulk Temperature 4.80E+07 - 172 4.40E+07 167 4.00E+07 162 3.60E+07 157

3.20E+07 152 2.80E+07 147 0 0 I.-

o

. 2.40E+07 142 0 02.

0 2.00E+07 -- E 137 12 x 1.60E+07-132 1.20E+07 127 8.OOE+06 122 4.OOE+06 117 0.OOE+00 112 0 50 100 150 200 250 300 Time After Reactor Shutdown (hr)

Figure 6.6.5: FSP Heat Load and FSP Bulk Temperature Profiles - Abnormal Full Core - 100 Hour Refueling Start Time Holtec Report HI-2084175 6-26 Holtec Project 1702

Beaver Valley Unit 2 CCW Inlet Temperature vs. Refueling Start Time 105 104 103

. 102

a. 101 E

I-4)

.E 100 C.)

C.)

99 98 977-100 105 110 115 120 125 130 135 140 145 150 Refueling Start Time (hr)

Figure 6.6.6: Maximum Allowable Component Cooling Water Inlet Temperature vs. Refueling Start Time - Normal Full Core Offload Holtec Report HI-2084175 6-27 Holtec Project 1702

2.01 e+02 2.00e+02 1.98e+02 1.97e+02 1.95e+02 1.94e+02 1.92e-02 1.90e+02 1.89e+02 1.87e+02 1.86e+02 1.84e+02 1.83e+02 1.81 e+02 1.79e+02 1.78e+02 1.76e+02 1.75e+02 1.73e+02 1.71 e÷02 YNIX 1.70e+02 Contours of Static Temperature (f) Jul 13, 2010 FLUENT 6.3 (3d, dp, pbns, ske)

Figure 6.7.1: Contours of Static Temperature in a Vertical Plane through the Center of the FSP Holtec Report HI-2084175 6-28 Holtec Project 1702

8.0 RADIOLOGICAL EVALUATION 8.1 Introduction The new maximum-density fuel storage racks for the plant are capable of storing a greater number of spent fuel assemblies than the racks that are to be replaced. Consequently, an evaluation of the radiological effect of the increased number of fuel assemblies on the gamma dose rate at the surface of the fuel pool water is prudent. In addition, the radiological consequences of a fuel-handling accident must be addressed. Finally, the person-rem exposure resulting from the removal of the old racks and the installation of the new ones must be considered.

This section provides a summary of the radiological evaluations undertaken in support of the use, of the new maximum-density storage racks.

8.2 Acceptance Criteria The radiation dose incurred in the removal of the old racks and their preparation for shipment, as well as the installation of the new racks, must be consistent with operations carried out in full observance of ALARA principles.

8.3 Assumptions Certain basic parameter values, all of which are consistent with values specified in the licensing basis for the station, or are anticipated for the future, were assumed to apply to the radiological evaluation. The basic parameter values assumed to apply to the radiological evaluations are given in Table 8.3.1.

175 8-I Holtec Project 1702 Holtec Report HI-2084 Holtec Report HI-2084175 8-1 Holtec Project 1702

8.4 Dose Rate at the Surface of the Pool The calculated gamma dose rate at the surface of the fuel pool is the sum of the dose rate from radionuclides in the pool water, the dose rate from a fuel assembly in transit, and the dose rate from the fuel stored in the racks. These surface dose rates are not significantly affected by the increase in spent fuel storage capability, since the relative contribution from the fuel stored in the racks to the total surface dose rate is small.

8.4.1 Dose Due to Radionuclides in the Water Normally, the sources of radionuclides in the spent fuel pool water are crud and some fission products plated on the fuel surfaces, and small amounts of primary system water carried over in the spent fuel pool during refueling operations. This is currently estimated to be 23.5 mrem/hr.

These sources result in an increase in the contamination level of the pool during refueling and are soon reduced to normal levels after refueling is completed. Since quantities of radionuclides transferred from the core during future refuelings will be similar to that now experienced, there will be no significant change after rack installation is completed. Therefore, the existing pool cleanup system will continue to be adequate to maintain the radionuclide concentrations in the pool water similar to what is now being experienced. New rack installation will not increase the burden to the pool cleanup system, except perhaps for some crud temporarily stirred up during installation of the new racks.

8.4.2 Dose Due to Spent Fuel The dose rate at the pool surface (and consequent radiation dose rate to personnel) is not significantly increased with an increased number of spent fuel assemblies stored in the pool. The estimated dose rate at the SFP surface from stored spent fuel is << 1 mrem/hr. The pool surface dose rate is determined by the fuel being unloaded from the core in refueling and this is the same as that experienced before installation of the racks. The dose at the spent fuel pool surface from an assembly in transit from the core during refueling is dependant upon the operating specific power in the core, fuel burnup, and the time after shutdown when the spent fuel is moved, and Holtec Report HI-2084175 8-2 Holtec Project 1702

the depth of the water above the spent fuel. The estimated dose rate at the surface from a freshly discharged fuel assembly is approximately 4.0 mrem/hr. Fuel assemblies that are in the spent fuel racks are shielded such that their contribution to the pool surface dose rate is negligible compared to the assembly in transit. Therefore, the spent fuel in storage does not contribute significantly to the radiation dose above the pool.

8.5 Fuel Handling Accident For Beaver Valley Power Station (BVPS) Unit No. 2, the doses from the ,fuel-handling accident were calculated for the currently installed racks. The factors affecting the calculation, such as depth of the pool, the exposure of the fuel assembly considered when the accident occurs, the maximum number of rods ruptured in a fuel handling accident, and the cooling time of the assembly, have not changed. Consequently, the doses from the fuel-handling accident remain the values determined in the earlier calculations.

8.6 Person-Rem Exposure The person-rem exposure from re-racking one plant is shown in Table 8.6.1. The values are estimates of exposures associated with different aspects of the job, and they are based upon practical experience with re-racking projects at many nuclear plants in the U.S. and overseas.

8.7 Conclusions The total refueling outage dose rate at the surface of the pool water following installation of the new racks is below 50 mrem/hr and should not require changes in the radiation zoning of the plant.

The overall personnel (person-rem) dose from the operations necessary for the removal of the old racks and the installation of the new ones is low and should be acceptable as part of the annual dose incurred at the plant.

Holtec Report HI-2084175 8-3¸ Holtec Project 1702

Table 8.3.1. Parameter Values Utilized in the Radiological Evaluations PARAMETER VALUE Core power 2918 MWt 235 Initial fuel enrichment 5.0 Wt % U Fuel exposure 62,000 MWd/MtU Fuel cooling time 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Table 8.6.1. Person-rem Exposure from Re-racking One Plant Activity Number of Total Man- Dose Rate Estimated Person-Personnel Hours Rem Exposure Remove & Package 17 Empty 8 N/A 100 1.7 Racks mrem/rack Clean and Vacuum Pool 4 300 2 mrem/hr 0.6 Install 15 New Rack Modules 8, 500 (in 2 mrem/hr 1.0 SFP area)

Move Fuel to New Racks 3 450 4 mremihr 1.8 Total Exposure, Person-Rem 1 1 5 Holtec Report HI-2084175 8-4 Holtec Project 1702

9.0 INSTALLATION 9.1 Introduction The reracking at Beaver Valley Power Station (BVPS) Unit No. 2 involves the removal of all 17 existing spent fuel racks and the installation of 15 new maximum density spent fuel racks in the spent fuel pool (SFP). All installation work at the plant will be performed in compliance with NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," and applicable Holtec and plant procedures.

Crane operators will be trained in the operation of overhead cranes per the requirements of ANSI/ASME B30.2 and the plant's specific training program. Consistent with past practices, a videotape-aided training session will be presented to the installation team, all of whom will be required to successfully complete a written examination prior to the commencement of work.

To move the old and new fuel racks out and into the spent fuel pool, a temporary crane will be assembled on site and will share the same rails with the existing movable platform. The existing movable platform and its electric hoists will be used for all of the fuel shuffling involved in the reracking, and will also be the motive power for the new temporary crane used solely to carry the fuel racks. The temporary crane will be qualified to meet a 10:1 lift ratio for the "new" and "old" racks. Allowing two tons for any lift rig, the temporary crane will be designed to support a total vertical load of 150 tons without suffering plastic collapse. A hoist system with an ultimate load rating of at least 150 tons will be used to effect the load movements.

The lifting device designed for removal of the existing racks engages the racks near the bottom of the rack cells. The lifting device designed for installation of the new racks engages the rack baseplate. Both of these lifting devices are engaged to and disengaged from the racks utilizing a long handled tool. The lifting devices comply with the provisions of ANSI N14.6-1992 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.

Holtec Report HI-2084175 9-1 Holtee PrQject 1702

A surveillance and inspection program will be maintained throughout the installation phase of the project. A set of inspection points, which have been proven to eliminate any incidence of rework or erroneous installation in previous rack projects, will be implemented by the installer.

The Holtec procedures cover the scope of activities for the rack removal and installation effort.

Similar procedures have been utilized and successfully implemented on many previous rack installation projects. These procedures are written to include ALARA practices. and provide requirements to assure equipment, personnel, and plant safety. These procedures will be reviewed and approved in accordance with the plant administrative procedures prior to use on site. The following is a list of the procedures, which will be used to implement the installation phase of the project.

A. Removal and Handling/Installation Procedure:

This procedure provides overall direction for the handling and removal of the existing rack modules and the handling and installation of the new storage rack modules. This procedure delineates the steps necessary to receive the new racks on site, the proper method for unloading and uprighting the new racks, staging the new racks prior to installation, removal of the existing racks, and installation of the new racks. The procedure also provides for the installation of new rack bearing pads, adjustment of the new rack pedestals and verification of the as-built field configuration to ensure compliance with design documents. With regard to the removal of the existing racks, this procedure also delineates steps necessary to their handling after removal from the spent fuel pool.

B. Receipt Inspection Procedure:

This procedure delineates the steps necessary to perform a thorough receipt inspection of a new rack module after its arrival on site. The receipt inspection includes dimensional measurements, cleanliness inspection, visual weld examination, and verticality measurements.

C. Cleaning Procedure:

Holtec Report HI-2084175 9-2 Holtec Project 17.02

This procedure provides for the cleaning of a new rack module if required. The modules are to meet the requirements of ANSI N45.2.1, Level B prior to placement in the pool. Methods and limitations on materials to be utilized are provided.

D. Pre- and Post-Installation Drag Test Procedures:

These two procedures stipulate the requirements for performing a functional test on a new rack module prior to and following installation into the pool. The procedures provide direction for inserting and withdrawing an inspection gage into designated cell locations, and establish an acceptance criterion in terms of maximum drag force.

E. ALARA Procedure:

The plant ALARA Program and associated procedures provide guidance to minimize the total man-rem received during the rack installation project, by accounting for time, distance, and shielding.

F. Liner Inspection Procedure:

In the event that a visual inspection of any submerged portion of the spent fuel 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 will be applied to any suspect area of the pool liner.

Holtec Report HI-2084175 9-3 Holtec Project 1702

H. Liner Repair and Underwater Welding Procedure:

In the event of a positive leak test result, underwater welding procedures will be implemented which will 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 AWS D3.6-93, Specification for Underwater Welding.

9.2 Rack Arrangement The existing rack arrangement contains 17 racks bolted to an array of subbase beams on the spent fuel pool floor providing a total of 1088 storage locations. All of the existing racks are of a flux-trap design. Upon completion of the reracking project, the plant SFP will have enough storage spaces for a total capacity of 1690 assemblies in 15 freestanding racks (the subbase beams will remain, but the racks will not be bolted to them). A schematic plan view depicting the new rack arrangement can be seen in Figure 1.1.

9.3 Installation Sequence Installation of new racks involves the following activities. Necessary fuel movements will be performed prior to rack removal and installation activities. Fuel movement operations will be conducted in accordance with plant procedures. Fuel movements will happen several times over the project duration in order to support the planned sequence of existing rack removal and new rack installation.

Existing rack modules, after they have been emptied of spent fuel and surveyed by Health Physics, will be removed from the SFP using a temporary crane and a remotely engaged lift rig.

The rack will be cleaned using a pressure washer and then removed from the SFP following a safe load path. The rack will then be downended into a horizontal orientation and be ready for packaging and transport to a processing location.

Holtec Report HI-2084175 9-4 Holtec Project 1702

The new racks will be delivered in the horizontal position. A new rack module will be removed from the shipping trailer using a suitably rated crane and a spreader beam, while maintaining the horizontal configuration. The rack will be placed on an upender frame and secured. Using two independent overhead hooks, or a single overhead hook and a spreader beam, the module will be up-righted into a vertical position.

The new rack lifting device will be engaged in the lift points at the bottom of the rack. The rack will then be transported to a pre-leveled surface where, after leveling the rack, the appropriate quality control receipt inspection will be performed.

To address ALARA considerations, fuel in the SFP will be moved in preparation for rack installation. Additionally, the pool floor will be inspected and any debris that might inhibit the installation of bearing pads will be removed.

After pool floor preparation, new rack bearing pads will be positioned on the floor before the module is lowered into the SFP. The new rack module, the heaviest of which weighs less than eleven tons, will be lifted with a temporary crane installed specifically for this purpose and transported along the pre-established safe load path. The rack modules will be cautiously lowered into the pool and onto the bearing pads.

Elevation readings will be taken to confirm that the module is level. In addition, rack-to-rack and rack-to-wall offset distances will be measured. Adjustments will be made as necessary to ensure compliance with design documents. The lifting device will then be disengaged and removed from the pool under Health Physics direction. Post-installation free path verification will be performed using an inspection gage in order to ensure that no cell location poses excessive resistance to the insertion or withdrawal of a fuel assembly. This test confirms final acceptability of the installed rack module.

Holtec Report HI-2084175 9-5 Holtec Project 1702

v t e Letter Sequence Other
TAC:ME1079, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance Meeting, Meeting Questions 29 December 2010 4 December 2009 16 February 2010 19 March 2010 11 June 2010 24 January 2011 Answers 3 May 2010 21 May 2010 9 August 2010 Results Approval Other: ML100130345, ML101720365, ML102940458, ML110750478
MONTHYEAR2010-05-21 [Table View]

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