ML12038A038

Licensing Report for Seabrook Spent Fuel Pool and New Fuel Vault Analyses (Holtec International Document HI-2114996, Revision 2

ML12038A038
Person / Time
Site:	Seabrook
Issue date:	01/30/2012
From:	Holtec
To:	Office of Nuclear Reactor Regulation
References
HI-2114996, Rev 2
Download: ML12038A038 (178)

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Attachment 4 Licensing Report for Seabrook Spent Fuel Pool and New Fuel Vault Analyses (Holtec International document HI-2114996, Revision 2, Non-Proprietary)

Holtec Center. 555 Lincoln Drive West. Marlton. NJ 08053 Telephone (856) 797- 0900 Fax (856) 797 - 0909 INTERNATIONAL LICENSING REPORT FOR SEABROOK SPENT FUEL POOL AND NEW FUEL VAULT ANALYSES FOR NextEra Energy Seabrook Holtec Report No: HI-2114996 Holtec Project No: 2064 Sponsoring Holtec Division: HTS Report Class : SAFETY RELATED VAT~~E ional, and its CIi ci t, Itito bI use ony. i d or its desiunated subcontra'trs.. .

for any otherpurpose * !.*....*y any%party oher than

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Summary of Revisions:

Revision 0: Original Issue Revision 1: Editorial changes are made to address client's comments. All changes are denoted with revision bars.

Revision 2: All changes in Revision 1 have been accepted. Editorial changes are made to denote some proprietary information with shaded area. All new changes are denoted with revision bars.

Project No. 2064 Report No. HI-2114996 Pagei Shaded Areas Denote Holtec International Proprietary Information

TABLE OF CONTENTS TABLE OF CONTENTS ...................................................................................................... 1

1. INTRODUCTION AND

SUMMARY

........................................................................... 5

2. EFFECTS OF NEUTRON ABSORBER DEGRADATION ON SPENT FUEL POOL THERMAL-HYDRAULICS .......................................................................................... 6 2.1 INTRO D U CTIO N ...................................................................................................... 6 2.2 EFFECT OF BORALTM DEGRADATION ON BULK TEMPERATURES ................ 6 2.2.1 Steady State C onditions .......................................................................................... 6 2.2.2 T im e-to-B oil ........................................................................................................ 6 2.3 EFFECT OF BORALTM DEGRADATION ON LOCAL TEMPERATURES ............... 7 2.4 EFFECT OF THE PRESENCE OR ABSENSE OF BORAFLEXTM ............................. 7 2.5 EFFECT OF BORALTM DEGRADATION COMBINED WITH ITEMS IN FUEL A S SE MB L IE S ................................................................................................................... 7 2.6 C ON C LU SIO N S ........................................................................................................ 7

3. CRITICALITY EVALUATION OF NEW FUEL VAULT ........................................ 9 3.1 IN TRO D U CTIO N ...................................................................................................... 9 3.2 M ETH OD OLO GY ................................................................................................... 9 3.2.1 G eneral A pproach .................................................................................................. 9 3.2.2 Computer Codes and Cross Section Libraries ........................................................ 9 3.2.3 A nalysis M ethods ................................................................................................... 10 3.2.3.1 .Design Basis Fuel A ssem bly .......................................................................... 10 3.2.3.2 Reactivity Effect of Water Temperature and Density ..................................... 10 3.2.3.3 Fuel and Rack Uncertainties ........................................................................... 10 3.2.3.4 Eccentric Fuel Positioning ............................................................................. 11 3.2.3.5 R ack D eform ation ......................................................................................... 12 3.2.3.6 Calculation of Maximum keff .......................................................................... 12 3.2.3.7 Other A ccident Conditions ............................................................................. 12 3.2.3.8 Effect of the IFBA Rods ................................................................................. 12 3.2.3.9 Effect of the NFV Concrete Wall Thickness .................................................. 12 3.2.3.10 Effect of the Non-Fuel Hardware .................................................................... 12 3.3 A CCEPTAN CE CRITERIA ........................................................................................ 13 3.4 A SSU M PT ION S.............................................................................................................. 14 3.5 1N P U T D A T A .................................................................................................................. 14 3.5.1 Fuel A ssem bly Specification ................................................................................ 14 3.5.2 Burnable Absorbers Specification ........................................................................ 14 3.5.3 New Fuel Vault and Storage Rack Specification .................................................. 14 3.6 A N A L Y SIS ...................................................................................................................... 15 3.6.1 D esign Basis Fuel A ssem bly .................................................................................. 15 3.6.2 Temperature and Water Density Effects ............................................................... 15 3.6.3 Uncertainties Due to Fuel and Rack Tolerances .................................................. 15 3.6.4 Eccentric Fuel Positioning ................................................................................... 15 3.6.5 R ack D eform ation ................................................................................................ 16 3.6.6 Calculation of M axim um keff ................................................................................ 16 Project No. 2064 Report No. HI-2114996 Page I Shaded Areas Denote Holtec International Proprietary Information

3.6.7 Effect of the IFBA Rods ....................................................................................... 16 3.6.8 Effect of the NFV Concrete W all Thickness ........................................................ 16 3.7 CON CLUSION S ............................................................................................................. 16

4. CRITICALITY EVALUATION OF SPENT FUEL POOL ..................................... 27 4.1 IN TRODU CTION ....................................................................................................... 27 4.2 M ETHOD OLO GY .................................................................................................... 28 4.2.1 General A pproach ................................................................................................ 28 4.2.2 Com puter Codes and Cross Section Libraries ...................................................... 29 4.2.2.1 M CNP5 .............................................................................................................. 29 4.2.2.1.1 M CN P5 Validation ................................................................................... 29 4.2.2.2 CA SM O-4 ...................................................................................................... 31 4.2.2.2.1 Uncertainty in the Isotopic Content of Spent Fuel ..................................... 31 4.2.3 A nalysis M ethods ................................................................................................. 32 4.2.3.1 Design Basis Fuel A ssem bly .......................................................................... 32 4.2.3.2 Reactivity Effect of Spent Fuel Pool Water Temperature .............................. 32 4.2.3.3 Fuel and Storage Rack M anufacturing Tolerances ........................................ 33 4.2.3.4 Fuel Radial Positioning ................................................................................. 34 4.2.3.5 Spent Fuel Reactivity Calculations ............................................................... 34 4.2.3.5.1 Operating Parameters ..................... 34 4.2.3.5.2 Spent Fuel Isotopic Com position ............................................................... 35 4.2.3.5.3 Burnable Absorbers and Reactivity Control Devices ................................ 35 4.2.3.5.4 Burniup Uncertainty ................................................................................... 36 4.2.3.5.5 Axial Burnup Profiles and Axial Enrichment Variations ......................... 37 4.2.3.5.6 Fuel Creep and G rowth ............................................................................ 37 4.2.3.5.7 Cooling Tim e Uncertainty ........................................................................ 38 4.2.3.5.8 BORA FLEX TM G ap M aterial ................................................................... 38 4.2.3.5.9 G rid Spacers .............................................................................................. 39 4.2.3.5.10 RCCA Type and Configuration ............................................................... 39 4.2.3.5.11 M odel Sim plifications ............................................................................. 39 4.2.3.6 Synergistic Effects .......................................................................................... 41 4.2.3.7 Calculations to Determine keff Values and Loading Curves ......................... 41 4.2.3.7.1 Calculation of a Single keff Value ............................................................ 41 4.2.3.7.2 Target keff ....................................................................................................... 42 4.2.3.7.3 D eterm ination of Loading Curves ............................................................. 43 4.2.3.7.4 Soluble Boron Concentrations ................................................................. 43 4.2.4 N orm al Conditions .............................................................................................. 43 4.2.5 Accident Conditions ............................................................................................ 44 4.2.5.1 Tem perature and W ater Density Effects ........................................................ 45 4.2.5.2 Dropped A ssem bly - H orizontal .................................................................... 45 4.2.5.3 Dropped A ssem bly - Vertical into a Storage Cell ......................................... 45 4.2.5.4 M isloaded Fresh Fuel A ssem bly .................................................................... 45 4.2.5.5 M islocated Fresh Fuel A ssem bly ................................................................. 45 4.2.5.6 M issing RCCA s ............................................................................................ 46 4.2.5.7 Incorrect Loading Curve .............................................................................. 46 4.2.6 Interfaces ................................................................................................................... 46 4.2.6.1 Interfaces of D ifferent Patterns w ithin One Rack ......................................... 46 Project No. 2064 Report No. HI-2114996 Page 2 Shaded Areas Denote Holtec International Proprietary Information

4.2.6.2 Interfaces between Different Racks of the Same Type .................................. 47 4.2.6.3 Interfaces betw een different rack types ......................................................... 47 4.2.7 Guidance D SS-ISG-2010-01 ................................................................................. 47 4.3 ACCEPTAN CE CRITERIA ........................................................................................ 49 4.4 A SSUM PTION S.............................................................................................................. 50 4.5 IN PUT D ATA .................................................................................................................. 50 4.5.1 Fuel A ssem bly Specification ................................................................................ 50 4.5.2 Core Operating Param eters ................................................................................... 51 4.5.3 Burnable A bsorbers ............................................................................................... 51 4.5.4 SEABROOK Storage Rack Specification ............................................................. 51 4.5.4.1 Region I Style Storage Racks ........................................................................ 51 4.5.4.2 Region 2 Style Storage Racks ........................................................................ 52 4.5.4.3 Gaps between A djacent Racks ...................................................................... 52 4.5.4.4 Fuel Rod Storage Basket ............................................................................... 52 4.5.5 M aterial Com position ............................................................................................ 52 4.6 AN ALY SIS ...................................................................................................................... 53 4.6.1 D esign Basis Fuel A ssem bly .................................................................................. 53 4.6.2 RCCA Type and Configuration ............................................................................ 53 4.6.3 Reactivity Effect of SFP Water Temperature and Density .................................. 55 4.6.4 Fuel A ssem bly Positioning ................................................................................... 56 4.6.5 Core Operating Param eters ................................................................................... 57 4.6.6 Cooling Time Uncertainty ........................................ 57 4.6.7 Reactivity ControlbDevices ...................................................................................... 58 4.6.7.1 IFBA Rods. ........ :...................................... ................................................ 58 4.6.7.2 BPRA Rods .:....................................................................................................... 59 4.6.7.3 IFBA/BPRA Bias Used in the Loading Curve Determination ........................ 59 4.6.8 Additional Studieý and Evaluations .......................................................................... 60 4.6.8.1 BORAFLEX TM Gap M aterial ........................................................................ 60 4.6.8.2 Grid Spacers .................................................................................................. 60 4.6.8.3 Fuel Creep and G rowth ................................................................................. 60 4.6.9 Fuel and Rack Tolerances Evaluations ................................................................. 60 4.6.10 A xial Burnup Profiles ........................................................................................ 63 4.6.10.1 A pproach ............................................................................................................ 63 4.6.10.2 Profile D atabases ............................................................................................ 63 4.6.10.3 Resulting Profiles .......................................................................................... 64 4.6.11 N orm al Conditions ............................................................................................. 64 4.6.11.1 Patterns A through D w ith Pure W ater ........................................................... 64 4.6.11.2 Patterns A through D w ith Borated W ater .................................................... 65 4.6.11.3 Other N orm al Conditions ............................................................................... 65 4.6.11.3.1 Single A ssem bly in W ater ...................................................................... 65 4.6.11.3.2 The Fuel Stored on the Periphery of the Region 2 Rack ......................... 66 4.6.11.3.3 Interfaces ................................................................................................. 66 4.6.11.3.4 Fuel Rod Storage Basket ........................................................................ 67 4.6.11.3.5 Other Conditions ...................................................................................... 67 4.6.12 Accident Conditions .......................................................................................... 68 4.6.12.1 Single Fresh A ssem bly M isload .................................................................... 68 Project No. 2064 Report No. HI-2114996 Page 3 Shaded Areas Denote Holtec International Proprietary Information

4.6.12.2 Single Fresh A ssem bly M islocation ............................................................... 68 4.6.12.3 Single M issing RCCA ................................................................................... 69 4.6.12.4 M ultiple RCCA s M isload ............................................................................... 69 4.7 CON CLU SION S ............................................................................................................. 69

5. REFER EN C E S ................................................................................................................ 158 Appendix A Applicability of Criticality Benchmark Calculations A-1 Project No. 2064 Report No. HI-2114996 Page 4 Shaded Areas Denote Holtec international Proprietary Information

1. INTRODUCTION AND

SUMMARY

This report documents the safety evaluation to address the potential degradation of neutron absorbers in the Region I and Region 2 spent fuel storage racks at the Seabrook Unit I nuclear power plant operated by NextEra Energy Seabrook, LLC [21 ].

• Region 1: These racks were originally designed with BORALTM panels as the neutron absorber material in a flux-trap rack configuration. Ongoing coupon surveillance has revealed evidence of blistering and thinning/spalling of the aluminum clad on the BORALTM panels. To account for these conditions, criticality calculations use a reduced B-10 areal density in the BORALTM, and assume a voided space on one side of each panel.

• Region 2: These racks were originally designed with BORAFLEXTM as the neutron absorber material in a flux-trap rack configuration. Due to the BORAFLEXTM degradation, future credit for BORAFLEXTM in these racks is not feasible. The criticality safety evaluations of the Region 2 racks are therefore performed without credit for BORAFLEXTM.

In addition to the results of the criticality calculations, a qualitative evaluation of the thermal hydraulic impact of the neutron absorber degradation [22] is documented.

Under the degraded neutron absorber conditions, the evaluations qualify the following fuel loading configurations, with details documented in Section 4 of this report.

• Region 1: Storage of fresh and spent fuel assemblies in a checkerboard configuration of fresh fuel assemblies with a maximum nominal enrichment of 5.0 wt% 235U and spent fuel assemblies with a minimum specified burnup as a function of the initial enrichment.

• Region 2:

o Storage of spent fuel assemblies with specific burnup requirements as a function of initial enrichment between 1.5 wt% and 5.0 wt% 235U, decay time, and presence of up to 2 RCCAs in the assemblies in an 2x2 cell array.

o Storage of spent fuel assemblies in the Rows I and 2 on the periphery of Racks 3, 4 and 5 adjacent to the west side of pool wall (See Figure 4.5.7), which require unusual plant actions to reach the fuel and allow crediting higher periphery leakages for these locations.

The review of the thermal-hydraulic analyses of record and evaluation of the potential impact of changes in the physical condition of the BORALTM (or BORAFLEXTM) demonstrates that there are no potential adverse thermal-hydraulic impacts of degradation of these neutron absorbing materials.

Additionally, this report documents the fuel storage rack criticality calculations performed for the Seabrook Unit 1 New Fuel Vault (NFV) [23]. The calculations qualify the NFV to store up to 90 fresh Westinghouse 17x1 7 assemblies with enrichment up to a maximum of 5.0 wt% 235U.

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2. EFFECTS OF NEUTRON ABSORBER DEGRADATION ON SPENT FUEL POOL THERMAL-HYDRAULICS

2.1 INTRODUCTION

This Section provides an evaluation that qualitatively assesses the impact of the anticipated BORALTM blistering phenomena on each of the pertinent considerations addressed in the Plant's thermal-hydraulic analysis of record. Specifically, the following areas of concern are addressed:

1. The effect of the displacement of water due to BORALTM degradation on the bulk temperature of the spent fuel pool.

2. The effect of BORALTM degradation in reducing flow in a spent fuel cell, to ensure that local thermal-hydraulic requirements are satisfied.

3. The effect of the presence or absence of the volumetric displacement of water due to BORAFLEXTM (used in the Region 2 spent fuel racks).

4. Expected thermal-hydraulic performance issues related to BORALTM degradation in conjunction with items possibly in fuel assemblies, such as RCCAs.

2.2 EFFECT OF BORALTM DEGRADATION ON BULK TEMPERATURES 2.2.1 Steady State Conditions The steady state condition is the condition where the decay heat generation rate is equal to the heat rejection rate. The decay heat generation rate is the result of the ongoing radioactive decay of the isotopes in the spent fuel. As such it cannot be affected by the condition of the BORALTM. The heat rejection rate is a function of the design of the spent fuel pool cooling system and the temperature of the cooling water supplied to it, neither of which can be affected by the condition of the BORALTM.

As these two terms are both unaffected by the condition of the BORALTM, it is apparent that the licensing-basis spent fuel pool bulk temperature cannot be affected by any BORALTM degradation.

2.2.2 Time-to-Boil The dimensions of the BORALTM panels are 141" x 7.5" x 0.075", and the number of BORALTM-equipped rack cells is 576. The maximum BORALTM blister thickness is 45-mils (0.045").

Conservatively assuming that there are four BORALTM panels affixed to each cell, the theoretical maximum possible increase in the racks displaced volume is 475 gallons. This additional volume displaced by the blistered BORALTM would reduce the pool water volume. This corresponds to a 0.09% reduction in pool water volume. The pool water thermal capacity is proportional to the water volume, so the thermal capacity is also reduced by 0.09%. The rate of change of the bulk temperature is inversely proportional to the thermal capacity, so the rate of change would increase Project No. 2064 Report No. HI-2114996 Page 6 Shaded Areas Denote Holtec International Proprietary Information

by 0.09% and the time-to-boil would decrease by 0.09%. It is apparent that the licensing-basis spent fuel pool time-to-boil analyses would only be negligibly affected by the theoretical maximum worst-case BORALTM degradation.

2.3 EFFECT OF BORALTM DEGRADATION ON LOCAL TEMPERATURES BORALTM panels are mounted on the outsides of the boxes that form the rack storage cells, meaning the panels are in the inter-cell water gaps. As the panels are outside of the storage cells, blistering and spalling of the BORALTM surface could not reduce the flow area through the storage cells and the equivalent hydraulic diameter used in the local temperature analysis model could not be affected. Based on this observation, the licensing-basis spent fuel pool local temperature analysis is not affected by BORALTM degradation.

2.4 EFFECT OF THE PRESENCE OR ABSENSE OF BORAFLEXTM If the BORAFLEXTM were absent, the volume of water in the spent fuel pool would increase. As described in Section 2.2.1 of this report, the normal condition bulk temperatures are steady-state values and cannot be affected by the volume of water. Also, as described in Section 2.2.2, the time-to-boil analyses would only be adversely affected if the volume of water in the pool were reduced.

Thus, the absence of the BORAFLEXTM will not adversely affect the licensing-basis thermal-hydraulic analysis.

2.5 EFFECT OF BORAL TM DEGRADATION COMBINED WITH ITEMS IN FUEL ASSEMBLIES Degradation of the BORALTM could potentially release small amounts of the constituent materials of BORALTM, namely aluminum and boron carbide, into the spent fuel pool water. Boron carbide is an extremely inert material that will not react with any other material, and therefore poses no potential for adverse interactions with any items in the fuel assemblies. In the presence of water, aluminum rapidly passivates to form aluminum oxide, which is also extremely non-reactive and also poses no potential for adverse interactions with any items in the fuel assemblies.

BORALTM is manufactured, from small particles of boron carbide and aluminum, formed into flat sheets under high temperature and pressure. Any loss of materials from the BORALTM panels would therefore be expected to result in very small particles that are much smaller than the flow passages in the fuel assemblies. Any small particles would be safely collected by the spent fuel pool cleanup system.

2.6 CONCLUSION

S A review of the thermal-hydraulic analyses of record and evaluation of the potential impact of changes in the physical condition of the BORALTM (or BORAFLEXTM) has demonstrated that there Project No. 2064 Report No. HI-2114996 Page 7 Shaded Areas Denote Holtec International Proprietary Information

are no potential adverse thermal-hydraulic impacts of degradation of these neutron absorbing materials.

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3. CRITICALITY EVALUATION OF NEW FUEL VAULT

3.1 INTRODUCTION

This Section documents the fuel storage rack criticality calculations performed for NextEra Energy for the Seabrook Unit 1 New Fuel Vault (NFV). The purpose of Section 3 of this report is to qualify the capability of the Seabrook Unit I NFV to store up to 90 fresh Westinghouse 17x17 assemblies with enrichment up to a maximum of 5.0 wt% 235U.

The objective of the calculations is to demonstrate that the effective neutron multiplication factor (keff) in the NFV is less than or equal to 0.95 for the fully flooded condition with un-borated water and less than or equal to 0.98 for optimum moderation conditions, with the storage racks fully loaded with fuel of the highest anticipated reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than the regulatory limit with a 95% probability at a 95% confidence level.

3.2 METHODOLOGY 3.2.1 General Approach The analysis is performed in a manner such that the results are below. the regulatory limit with a 95%

probability at a 95% confidence level. The calculations are performed using the statistical analysis approach with respect to the various calculation parameters. Thle ,approach considered for each parameter is discussed below.

3.2.2 Computer Codes and Cross Section Libraries The principal method for the criticality analysis of the NFV is the use of the three-dimensional Monte Carlo code MCNP5-1.51 [2]. MCNP5 is a continuous energy three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory. MCNP5 was selected because it has been used previously and verified for criticality analyses and has all of the necessary features for this analysis.

In this analysis MCNP5 calculations used continuous energy cross-section data predominantly based on ENDF/B-V.

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 MCNP5 criticality Output contains a Project No. 2064 Report No. HI-2114996 Page 9 Shaded Areas Denote Holtec International Proprietary Information

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.

• . j* and the initial source was usually 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.

3.2.3 Analysis Methods 3.2.3.1 Design Basis Fuel Assembly Three types of 17x17 fuel assemblies (Standard design, Vantage 5 and RFA) have been used for Seabrook station. However, Standard design assemblies were only used during initial cycles of operation (Cycles 1-4), Vantage 5 assemblies were only used for Cycles 5-8 and further operation of these assembly types is not supposed. Therefore only RFA assemblies which are employed most recently are considered in this analysis. This way, the single set of parameters of RFA assemblies is used in all design basis calculations for NFV analysis.

3.2.3.2 Reactivity Effect of Water Temperature and Density ,

The Seabrook NFV is intended to be dry during all normal operation-condition.; However, the worst case accident, when the NFV is flooded with water is considered. The main effect of changes in the NFV water temperature is the impact of the difference in water density. Additionally there may be temperature-dependent cross section effects that need to be considered. To ensure that the appropriate temperature is used for each case, calculations at the lower end of the temperature range

- 80.33 °F (300K) and at the higher end of the temperature range - 212 OF (373K) are performed.

Note that one of the temperature adjustments, S (alpha, beta), is only available at fixed temperatures of 300K and 400K in MCNP5. Calculations are therefore performed at both temperatures for 5.0 wt% 235U enrichment cases, and the temperature identified to result in higher reactivity is determined. The calculations are all performed with unborated water.

3.2.3.3 Fuel and Rack Uncertainties In the calculation of the final keff for Seabrook NFV, the effect of manufacturing tolerances on reactivity must be included. MCNP5 was used to perform these calculations. As allowed in [3], the sensitivity study approaches is employed to calculate the tolerance effects. The evaluations include tolerances of the fuel dimensions and tolerances of the racks dimensions. The reference condition is the condition with nominal dimensions and properties. To determine the Ak associated with a specific manufacturing tolerance, the reactivity calculated for the reference condition is compared to the reactivity from a calculation with the tolerance included. The uncertainty associated with each individual calculation is statistically combined and added to the kcalc calculation according to the following equation [6]:

2 AkcaIlc = (kcal2 - kcai) + 2 * - (a2 + oT2)

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All of the Ak values from the various tolerances are statistically combined (square root of the sum of the squares) to determine the final reactivity allowance for manufacturing tolerances. In some cases it is not obvious whether an increase or decrease of the parameter will lead to an increase in reactivity. In these cases, the reactivity effect of both increase and decrease of the parameter are calculated, and the maximum value of reactivity effect is used when calculating the statistical combination. The fuel and rack tolerances included in this analysis for the NFV are described below:

Fuel Tolerances

• Increased Fuel Density: , >

• Fuel Rod Cladding Outside Diameter:

• Fuel Rod Pitch:

" Fuel Rod Cladding Inside Diameter:. .

• Fuel Pellet Outside Diameter:

" Guide Tube Outside Diameter:

• Guide Tube Inside Diameter: j*

• Increased Fuel Enrichment: +0.05 wt% 2'U Rack Tolerances

.. Cell Inside Diameter:

• ,v<:*',.

• Box Wall Thickness:

The maximum enrichment value in the Seabrook Unit I Technical Specifications is.:5.0 wt% 235U.

Therefore, NextEra Energy designs (and the fuel vendor provides) a maximum enrichment in the fresh fuel that will guarantee that no fuel assembly could exceed the Technical Specification limit with all uncertainties applied. Therefore, the highest possible enrichment of 5.0 wt% 235U is used in all calculations.

3.2.3.4 Eccentric Fuel Positioning The eccentric fuel positioning case should be included in the tolerance calculations. The MCNP5 model consists of the following eccentric fuel positioning case in NFV analyses:

• All the fuel positioned at the closest approach with the storage cell to the center point of the whole NFV model.

• All the fuel positioned within the storage cell away from the center point of the whole NFV model.

• In each 2x2 array, all the fuel positioned at the closest approach with the storage cell to the center point of the 2x2 array.

" In each 2x2 array, all the fuel positioned within the storage cell away from the center point of the 2x2 array.

The maximum positive reactivity effect among these cases is used when calculating the statistical combination of the tolerance.

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3.2.3.5 Rack Deformation Rack deformation is not considered in this calculation due to that it is not considered a credible accident scenario. See Section 3.6.5.

3.2.3.6 Calculation of Maximum keff The calculated keff are determined for fully flooded and optimum moderation conditions. The maximum keff for each case is therefore determined from the MCNP5 calculated keff, the calculation bias, and the applicable uncertainties and tolerances (bias uncertainty, calculation uncertainty, fuel and rack tolerances, and eccentric fuel positioning) using the following formula:

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

In the geometric models used for, the calculations, each fuel rod and its cladding were described explicitly.

The regulatory limit for NFV analysis is that at the optimum moderation condition, a maximum keff should be less than 0.98 at the 95/95 level; while at the full flooded condition with un-borated water, a maximum keffshould be less than 0.95 at the 95/95 level. The target keff used in this analysis retain a 0.01 delta-k margin to the regulatory limit.

3.2.3.7 Other Accident Conditions The 'consideration of two accident conditions: fully flooded and optimum moderation have been inaclutded ini the analysis of the NFV. The double contingency principal of ANS-8.I/N16.ljI975 [4]'

specifies that it shall require at least two unlikely, independent and concurrent events to iprbduce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions. No additional calculations are required for accident conditions.

3.2.3.8 Effect of the IFBA Rods Obviously the presence of the IFBA rods in the fresh fuel assemblies reduces the reactivity due to the integrated neutron absorber. However, this absorber may change the behavior of the fuel assemblies in the low temperature area. To ensure that the fresh fuel assembly without IFBA rods is bounding the criticality calculations are performed and discussed in Section 3.6.7.

3.2.3.9 Effect of the NFV Concrete Wall Thickness In the criticality analysis the wall thickness of 40 inches is assumed. The additional calculations are performed with a concrete thickness of 12 inches. See Section 3.6.8.

3.2.3.10 Effect of the Non-Fuel Hardware The non-fuel hardware such as Thimble Plugs (TP), Rod Cluster Control Assemblies (RCCAs) and non-activated sources are acceptable to storage with the fuel assemblies. Non-fuel hardware is Project No. 2064 Report No. HI-2114996 Page 12 Shaded Areas Denote Holtec International Proprietary Information

inserted in the guide tubes of the assemblies. For pure water, the reactivity of any fuel assembly with inserts is bounded by (i.e. lower than) the reactivity of the same assembly without the insert.

This is due to the fact that the insert reduces the amount of moderator in the assembly, while the amount of fissile material remains unchanged.

Therefore, from a criticality safety perspective, the presence of inserts does not impact the analysis.

3.3 ACCEPTANCE CRITERIA The PWR storage racks for Seabrook Unit 1 NFV are designed in accordance with the applicable codes and standards listed below. The objective of this analysis is to show that the effective neutron multiplication factor, kerr, is less than or equal to the target k.rr of 0.97 for the NFV at optimum moderation conditions (low density water) and less than or equal to the target keff of 0.94 with the NFV vault flooded with pure water with a density of 1.0 g/cm 3, when the NFV is fully loaded with fuel of the highest anticipated reactivity. The maximum calculated reactivity includes a margin for

-uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than the limit with a 95% probability at a 95% confidence level [1].

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

'S Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criterion 62.

"Prevention of Criticality in Fuel Storage and Handling."

05 rJ*

• USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, Criticality Safety of Fresh and Spent.Fuel *Storage and Handling, Rev. 3 - March 2007.

" L. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T. Collins, August 19, 1998.

• USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2, March 2007.

" ANSI ANS-8.17-2004, 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"

• ANSI/ANS-8.1-1998 (R2007), "Nuclear Criticality Safety in Operations with Fissionable Materials outside Reactors."

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3.4 ASSUMPTIONS To assure the true reactivity will always be less than the calculated reactivity, the following conservative design criteria and assumptions were employed:

1) A fuel pellet stack density assumed to be equal to the pellet density, and is conservatively modeled as a solid right cylinder over the entire active length, neglecting dishing and chamfering;

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

3) The concrete wall thickness of 40 inches is assumed;

4) The maximum planar average enrichment is assumed to be 5.0 wt% 235 U; 5) 3.5 INPUT DATA 3.5.1 Fuel Assembly Specification The NFV is designed to accommodate the RFA 17x17 Westinghouse fuel assembly. The design specification for RFA'fuel assembly is presented in Table 3.5,1.

3.5.2 Burnable Absorbers Specification There is the potential for burnable absorbers to be located in the assembly. The Seabrook Unit 1 fuel makes use of burnable poison rod assemblies (BPRAs) of B20 3 and integrated fuel burnable absorber (IFBA) rods with a thin coating of ZrB2 on the U0 2 pellet. The BPRAs were only utilized in Cycle 1 and then replaced with IFBA rods for the remaining cycles, therefore only IFBA rods are considered in the calculations of this section. The design specifications for the IFBA rods are given in Table 3.5.1.

3.5.3 New Fuel Vault and Storage Rack Specification The NFV and storage cell characteristics that were used in the criticality analysis are summarized in Table 3.5.2.

The NFV MCNP5 model is a finite model that consists of a storage array of five cells (East-West) by eighteen cells (South-North) with the dimensions given in Table 3.5.2. The model includes the gaps between the storage racks and the walls. The concrete walls have some variation .in the thickness (three walls at 12 inches and one at a maximum of 72 inches). In the model, the concrete wall is assumed to be 40 inches thick. Above the fuel is a 12 inch reflector with the water density of 1.0 g/cm 3 and the same temperature as within the fuel region and below the fuel is the concrete floor Project No. 2064 Report No. HI-2114996 Page 14 Shaded Areas Denote Holtec International Proprietary Information

with thickness of 40 inches. The model containing the 17 x 17 fuel assembly, as drawn by the two-dimensional plotter, is shown in Figures 3.5.1 and 3.5.2. The calculations are described in Section 3.6.

3.6 ANALYSIS This section describes the calculations that were used to determine that the storage racks meet the acceptance criteria as discussed in Section 3.3.

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

3.6.1 Design Basis Fuel Assembly The RFA 17 x 17 PWR fuel assembly used in the analysis is described in Table 3.5.1. It was evaluated at 5.0 wt% 235U enrichment and conservatively modeled without spacer grids or other non-fuel hardware.

3.6.2 Temperature and Water Density Effects Water tempdrature 'effects on reactivity in the NFV have been evaluated for the fuel with 5.0, wt%.-ý.! ,,

23'U.enfichment at'80.33 'F (300K) and 212 'F (373K). The results presented in Table 316.1show that the water 'temperature coefficient of reactivity is positive, i.e. the higher temperature of 2.12 'F' . .

results in a higher reactivity. Based on the results, it was also determined that the 100% moderator  :

condition, i.e. 1.0 g/cm 3, represent the maximum reactivity condition and 6% moderator condition,.

i.e., 0.06 g/cm 3 , represent optimum hypothetical low density moderation (i.e., fog or foam).

Therefore, all the following cases are performed with 100% and 6% moderator density at temperature of 212 'F.

3.6.3 Uncertainties Due to Fuel and Rack Tolerances In the calculation of the final keff, the effect of manufacturing tolerances on reactivity is included as discussed in Section 3.2.3.3. The evaluations include tolerances of the fuel and rack dimensions.

These tolerances are provided in Table 3.5.1. The reference condition is the condition with nominal dimensions and properties. To determine the Ak associated with a specific manufacturing tolerance, the keff calculated for the reference condition is compared to the ken- from a calculation with the tolerance included. The results are presented in Table 3.6.2.

3.6.4 Eccentric Fuel Positioning Four different eccentric fuel positioning models discussed in Section 3.2.3.4 are analyzed by the MCNP5 code. The reactivity effect of eccentric fuel positioning is included in the tolerance calculations as shown in Table 3.6.2.

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3.6.5 Rack Deformation Rack deformation is not considered in this calculation due to that it is not considered a credible accident scenario. The keff value is below 0.65 (see Table 3.6.1) for dry conditions, therefore rack deformation could not cause the reactivity of the system to exceed the regulatory limit. Note that considering rack deformation under moderated conditions would fall under the double contingency principle, as two accident scenarios would have to take place (moderator filling the NFV and rack deformation occurring).

3.6.6 Calculation of Maximum keff The maximum keff values, based on the formula in Section 3.2.3.6, were calculated for the fuel assembly described in Table 3.5.1 with enrichment of 5.0 wt% 235U. The results presented in Table 3.6.3 show the maximum keff values for different moderation conditions and confirm that keff values are well below the target keff of 0.97 at optimum moderation and the target keff of 0.94 when fully flooded with un-borated water at a 95% probability and at a 95% confidence level.

3.6.7 Effect of the IFBA Rods The criticality calculations to confirm that the presence of the IFBA rods in the fresh fuel assemblies reduces the reactivity are performed. The results presented in Table 3.6.4 show that the reference case without IFBA rods has .a maximum keff value. The minimum difference between the no IFBA . .

case and 16IFBA rods-case is about 0.011, which is sufficient margin to cover any tolerances in the IFBA rod. Therbfore,:'ll fuel assemblies .withIFBA rods are bbunded by the reference case without' IFBA and are acceptable for storage in the Seabrook INFV. Note also that for the reactivity calculation a longer IFBA length (current cycle) is bounded by the shorter length.

3.6.8 Effect of the NFV Concrete Wall Thickness As discussed in Section 3.2.3.9, the additional calculations with a concrete thickness of 12 inches are performed. The results presented in Table 3.6.5 show the reactivity effect of the NFV concrete wall thickness is statistically insignificant. Therefore, assumed reference NFV concrete thickness of 40 inches is acceptable.

3.7 CONCLUSION

S Section 3 of this report documents the criticality analysis for the storage of 17 x 17 PWR fresh fuel assemblies with an initial planar average enrichment of up to 5.0 wt% 235U in the NFV at the Seabrook Unit I Nuclear Power Plant. Calculations were made with the continuous energy MCNP5-1.51 code package, a three-dimensional Monte Carlo analytical technique, with fresh fuel assemblies enriched to 5.0 wt% 235U without IFBA rods. These calculations were made for various moderator densities, and the results shown in Figure 3.6.1 indicate that the peak reactivity (full moderation) occurs at 100% moderator density and the optimum hypothetical low density moderation (i.e., fog or foam) occurs at 6% moderator density. The effective neutron multiplication factor (keff) for the NFV is less than 0.97 for the optimum moderation condition and less than 0.94 Project No. 2064 Report No. HI-2114996 Page 16 Shaded Areas Denote Holtec International Proprietary Information

for the fully flooded with un-borated water condition. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations with a 95% probability at a 95% confidence level.

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Table 3.5.1 PWR Fuel Assembly Specifications Fuel Assembly Type 17x17 RFA Fuel Rod Data Fuel Pellet Outside Diameter, in. 0.3225 Cladding Inside Diameter, in. 0.3290 Cladding Outside Diameter, in. 0.3740 Cladding Material ZIRLO Maximum Pellet Density, g/cc 10.505 Maximum Enrichment, wt% 2"U 5.0 ZrB2 Coating Loading (mg lB/inch) 2.355 ZrB2 Coating Thickness', in. 0.000657 ZrB2 Coating Length, in. 1202 Fuel Assembly Data Fuel Rod Array 17x 17 Number of Fuel Rods 264 Fuel Rod Pitch, in. 0.496 LI Fuel Assembly Width, in. 8.426 Fuel Assembly Length, in. 159.975 Active Fuel Length, in. 144 Bottom of Active Fuel Length to Bottom of Assembly, in. 3.278 Guide/Instrument Tube Data Number of Guide Thimbles. 24 Number of Instrument Tubes I Guide Thimble Upper Region Inside Diameter, in. 0.442 Guide Thimble Upper Region Outside Diameter, in. 0.482 Guide Thimble Dashpot Region Inside Diameter, in. 0.3970 Guide Thimble Dashpot Region Outside Diameter, in. 0.4390 Instrument Tube Inside Diameter, in. 0.442 Instrument Tube Outside Diameter, in. 0.482 Guide/Instrument Tube Material ZI The coating thickness was not available. The values provided are calculated.

2 For Cycle 15, the coating length has been increased from 120 to 122 inches.

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Table 3.5.2 New Fuel Vault Storage Racks Specification Parameter Value Storage Cells Array 5 (East-West) x 18 (South-North)

Storage Cell Inside Dimension, in. 9.00 Storage Cell Wall Thickness, in. 0.093,11, Storage Cell Pitch3 , in. 21 or 33 Storage Cell Material SS304 Cell Array Width, in. 125.1875 (East-West) x 373 (South-North)

Cell Array to Wall Distance North Face, in. 4 Cell Array to Wall Distance South Face, -in. 4 Cell Array to Wall Distance East Face, i.n.. 3.375 Cell Array to Wall Distance West Face, in.. 'ý* 3.4375 Concrete Wall Thickness, in. 12 - 72 3 The cell pitch is 21 inches center to center between the cells in the (South-North) 18-cell array direction. The cell pitch for the (East-West) 5-cell array direction is 33inches center to center between the 2 d and 3rd and between the 3d and 4 th cell, and 21 inches center to center between the I" and 2 nd between the 4 th and 5 th cell.

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Table 3.6.1 Summary of the MCNP5 NFV Calculations for Different Water Density and Temperature - 5.0 wt% 235U 80.33 OF Moderator (Reference)

Cluae 212 F Cluae Density Calculated Calculated Delta kcaic kcaic cl 0 0.6263 0.6469 0.0206 3 0.8410 0.8607 0.0197 4 0.8809 0.9027 0.0218 5 0.8981 0.9254 0.0273 6 0.9022 0.9330 0.0308 7 0.8950 0.9295 0.0345 8 0.8809 0.9193 0.0384 9 0.8603 0.9034 0.0431 10 0.8396 0.8850 0.0454 15 0.7393 0.7890 0.0497 20 0.6704 0.7170 0.0466 30 00.6081

.6491 0.0410 60 0.7094 i,, 0.7309 0.0215 90 0.8603:" :, 0.8736 0.0133 -

95 ý0.8828 0.8949 0.0121 100 0.9043 0.9151 0.0108 Project No. 2064 Report No. HI-2114996 Page 20 Shaded Areas Denote Holtec International Proprietary Information

Table 3.6.2 Results of the NFV Tolerance Calculations - 5.0 wt% 235U 6 % Moderator Density 100% Moderator Density Description kcalDelta kcalc G Delta kcalc kcalc Reference keff 0.9330 0.0005 n/a 0.9151 0.0005 n/a Pellet Density max 0.9342 0.0005 0.0026 0.9155 0.0005 0.0018 Clad OD max 0.9330 0.0005 0.0014 0.9136 0.0005 -0.0029 Clad OD min 0.9342 0.0005 0.0026 0.9164 0.0006 0.0029 Pin Pitch max 0.9334 0.0005 0.0018 0.9158 0.0006 0.0023 Pin Pitch min 0.9326 0.0005 -0.0018 0.9145 0.0005 -0.0020 Clad ID max 0.9328 0.0005 -0.0016 0.9152 0.0006 0.0017 Clad ID min 0.9337 0.0005 0.0021 0.9153 0.0006 0.0018 Pellet OD max 0.9341 0.0005 0.0025 0.9153 0.0005 0.0016 Pellet OD min 0.9326 0.0004 -0.0017 0.9149 0.0006 -0.0018 GT OD max 0.9323 0.0005 -0.0021 0.9149 0.0006 -0.0018 GT OD min 0.9329 0.0005 -0.0015 0.9162 0.0005 0.0025 GT ID max 0.9340 0:0005 -0.0024 0.9152 0.0005 0.0015' GT ID min 0.9324 .0.0005 '7-0q.0020 0.9154 0.0005 0.0017 Eccentric Position #1 0.9359 0.0605 0-,:?0,'043 0.9187 0.0006 0.0052 Eccentric Position #2 0.9307 0.0004 "'-0.0036 0.9182 0.0006 0.0047 Eccentric Position #3 0.9322 0.0005 -0.0022 0.9188 0.0005 0.0051 Eccentric Position #4 0.9352 0.0005 0.0036 0.9181 0.0006 0.0046 Cell ID max 0.9324 0.0005 -0.0020 0.9155 0.0006 0.0020 Cell ID min 0.9333 0.0005 0.0017 0.9148 0.0006 -0.0019 Wall Thk max 0.9215 0.0005 -0.0129 0.9126 0.0005 -0.0039 WallThkmin 0.9446 0.0005 0.0130 0.9180 0.0006 0.0045 Square Root Sum of the Squares 0.0151 0.0091 2 Sigma (max of all cases) 0.0010 0.0012 Note: The maximum positive value of the tolerance effect for each case was used.

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Table 3.6.3 Results of the NFV MCNP5 Calculations Parameter Value Enrichment, wt% 5.0% 5.0%

Moderator Density 6.0% 100.0%

Uncertainties:

MCNP5 Code Bias Uncertainty 0.0085 0.0085 MCNP5 Calculation Statistics (95%/95%, 2a) 0.0010 0.0012 Calculated Tolerances 0.0151 0.0091 Statistical Combination of Uncertainties 0.0173 0.0125 Calculated MCNP5 kff 0.9330 0.9151 MCNP5 Code Bias 0.0043 0.0043 Maximum kff 0.9546 0.9319 Target keff 0.9700 0.9400 Regulatory Limit . . 0.9800 0.9500 Project No. 2064 Report No. HI-2114996 Page 22 Shaded Areas Denote Holtec International Proprietary Information

Table 3.6.4 Results of the IFBA Calculations 6.0 % 100%

Calculation Moderator Moderator Description Density Densi kcalc kcalc 212 OF Reference (No IFBA) 0.9330 0.0005 0.9151 0.0005 16 IFBAs 0.9215 0.0005 0.8917 0.0006 32 IFBAs 0.9062 0.0005 0.8793 0.0006 48 IFBAs 0.8902 0.0005 0.8723 0.0006 64 IFBAs 0.8794 0.0005 0.8694 0.0006 80 IFBAs 0.8633 0.0004 0.8667 0.0005 104 IFBAs 0.8490 0.0005 0.8634 0.0006 128 IFBAs 0.8155 0.0005 0.8607 0.0005 156 IFBAs 0.7901 0.0004 0.8597 0.0006 80.33 OF Reference (No IFBA) 0.9022 0.0005 0.9043 0.0005 16 IFBAs 0.8911 0.0005, 0. 8819 0.0006 32 IFBAs '0.8784 0.0005 0.8683 0.0005 48 IFBAs 0.8624 0.0005 0.8613 0.0006 64 IFBAs 0.8526 0.0005 0.8585 0.0005 80 IFBAs 0.8376 0.0004 0.8562 0.0005 104 IFBAs 0.8243 0.0005 0.8520 0.0006 128 IFBAs 0.7897 0.0004 0.8503 0.0006 156 IFBAs 0.7660 0.0005 0.8485 0.0006 Project No. 2064 Report No. HI-2114996 Page 23 Shaded Areas Denote Holtec International Proprietary Information

Table 3.6.5 Results of the NFV Calculations for Different Wall Thickness - 5.0 wt%

235u Calculation Description kcaac } Delta kcalc 6 % Moderator Density 40" (Reference) 0.9330 0.0005 n/a 12" 0.9329 0.0005 0 100% Moderator Density 40" (Reference) 0.9151 0.0005 n/a 12" 0.9145 0.0006 0 Project No. 2064 Report No. HI-2114996 Page 24 Shaded Areas Denote Holtec International Proprietary Information

Figure 3.5.1 A Two Dimensional Representation of the NFV Model (Radial Section)

Figure 3.5.2 A Two Dimensional Representation of the NFV Model (Axial Section)

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Figure 3.6.1 New Fuel Vault Reactivity for 5.0 wt% ,Fuel at Different Moderation Condition 1.10 ..... .............. .......... * .

1.00 0.90 0.80

.2 0.70 0.60 0.50 0.40 0 10 20 30 40 50 60 70 80 90 100

%Moderation 80.33 F*(300K) -"212F(373K)

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4. CRITICALITY EVALUATION OF SPENT FUEL POOL

4.1 INTRODUCTION

This chapter documents the criticality safety evaluation for the storage of PWR fresh and spent nuclear fuel in Region I & 2 spent fuel storage racks at the Seabrook nuclear power plant operated by NextEra Energy.

The Seabrook SFP (Spent Fuel Pool) has two separate rack designs (BORALTM and BORAFLEXTM) which are designated as Region 1 and Region 2 respectively. These designations do not refer to the rack geometry as both rack designs are of the flux-trap style.

Region 1 SFRs (Spent Fuel Racks) have six modules with BORALTM as the credited neutron absorber with space for 576 fuel assemblies; Region 2 SFRs contain six modules with a non-credited BORAFLEXTM absorber that allow storage of 660 fuel assemblies. RCCAs (Rod Cluster Control Assemblies) have been evaluated and may be credited for neutron absorption in selected fuel assemblies in Region 2. The maximum pool capacity is 1236 assemblies.

Criticality control in the SFP relies on the following:

• For Region 1 racks:

o Fixed Neutron Absorbers

.

• BORALTM. The BORALTM areal density is taken as lower limit value. u -.

_ Administrative controls Fuel Burnup a's a function of initial enrichment. "

o Soluble boron for. normal and accident conditions in accordance. *with 10CFR50.68(4)(b).

For Region 2 racks:

o Peripheral leakage (ONLY for the fuel in Rows I and 2 on the periphery of the Region 2 racks adjacent to the pool wall, i.e., the periphery of Racks 3, 4, and 5 adjacent to the west side of the pool shown in Figure 4.5.5, see Section 4.6.11.3.2).

o Administrative controls

" Fuel Burnup as a function of initial enrichment.

" Cooling time of fuel assemblies.

" RCCAs in guide tubes for selected patterns.

o Soluble boron for normal and accident* conditions in accordance with 10CFR50.68(4)(b)

Criticality control in the SFP does NOT rely on

• Radial neutron leakages, i.e. all configurations are considered radially infinite, except for the evaluation of the single assembly in water (see Sections 4.6.11.3.1 and 4.6.12.2) and the Project No. 2064 Report No. HI-2114996 Page 27 Shaded Areas Denote Holtec International Proprietary Information

assemblies stored on the periphery of the Region 2 racks adjacent to the pool wall (see Section 4.6.11.3.2).

BORAFLEXTM in the Region 2 racks, i.e., it is assumed for this analysis that the B4C in the BORAFLEXTM is not credited for neutron absorption.

The criticality calculations qualify fresh and spent fuel with an initial enrichment of 1.5 to 5.0 wt% 235U in the following storage rack configurations hereby denoted by Pattern names:

• Pattern A: Region I storage rack 2x2 array with a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 5.0 wt% 235U, and spent fuel assemblies with an initial enrichment of 1.5 to 5.0 wt% 235U. No credit for cooling time. All four cells contain BORALTM panels in each side of the cells. A BORALTM areal density of 0.015 gm/cm2 is considered.

• Pattern B: Region 2 storage rack 2x2 array uniformly loaded with spent fuel with an initial enrichment of 1.5 to 5.0 wt% 235U, with any two of the four cells containing RCCAs in the fuel assemblies, and the consideration of cooling times of 0, 2.5, 5, 10, 15, and 20 years. No credit for BORAFLEXTM panels. Additionally, it is acceptable to replace any cells that contain fuel assemblies with empty water cells.

• Pattern C: Region 2 storage rack 2x2 array uniformly loaded with spent fuel with an initial enrichment of 1.5 to 5.0 wt% 235U, with any one of the four cells containing RCCAs in the fuel assemblies, and the consideration of cooling times of 0, 2.5, 5, 10, 15, and 20 years.No credit, for BORAFLEXTM panels. Additionally, it is acceptable to replace any cells ,that contain fuel assemblies with empty water cells.

Pattern D: Region 2 storage rack 2x2 array uniformly loaded with spent fuel with an initial enrichment of 1.5 to 5.0 wt% 2 35U, without any RCCAs in the fuel assemblies, and, the consideration of cooling times of 0, 2.5, 5, 10, 15, and 20 years. No credit for BORAFLEXT.M panels. Additionally, it is acceptable to replace any cells that contain fuel assemblies with empty water cells.

The spent fuel assemblies stored on the periphery of the Region 2 rack adjacent to the pool wall are also qualified by taking credit of the periphery leakages for these locations.

Additionally, the fuel rod storage basket is qualified for all locations that are qualified for any fresh or spent fuel in above cases.

4.2 METHODOLOGY 4.2.1 General Approach Typically, in order to show that the results of the analyses are below the regulatory limit with a 95% probability at a 95% confidence level, independent uncertainties are considered by adding a statistical combination of their reactivity effects (square root of the sum of squares) to calculations performed at nominal conditions. However, usually only a few tolerances have a large effect, while most of them have an effect that is only of the order of the typical statistical Project No. 2064 Report No. HI-2114996 Page 28 Shaded Areas Denote Holtec International Proprietary Information

uncertainties of the reactivity. Modeling each of these small effect tolerance requires significant effort so a conservative method has been devised using a modified limit value approach. This modified approach utilizes mostly nominal parameters in design basis calculations but for each design basis case two additional calculations are performed which provides an overall conservative result. For fuel tolerances, a single calculation is conducted where the four most dominating parameters are assumed to be at the limiting condition. Likewise for rack tolerances, a single calculation is performed where the two most dominating parameters are assumed to be at the limiting condition. For each of those cases, the reactivity difference to the corresponding nominal case would be calculated, and then those two uncertainties would be statistically combined with the other uncertainties of the design basis calculation. The conservatism of this approach is large enough to allow neglecting the effect of some minor tolerances or uncertainties. To show that this approach is in fact more conservative, a comparison of this approach is performed with the approach using the statistical combination for a representative number of cases covering all aspects of the fuel and storage systems.

4.2.2 Computer Codes and Cross Section Libraries 4.2.2.1 MCNP5 MCNP5 Version 1.51 [2] is used for the criticality analyses. MCNP5 calculations use contin energy cross-section data predominantly based on ENDF/B-V and ENDF/B-VI [2].

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. I The initial source is specified as uniform over the fueled regions (assemblies) and the source distribution was confirmed to converge.

4.2.2.1.1 MCNP5 Validation 4.2.2.1.1.1 Actinides As discussed in Section 3.3.2, benchmarking of MCNP5 for criticality calculations is documented in

[5].

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4.2.2.1.1.2 Fission Products Little relevant critical experiments are publicly available for fission products. The uncertainty in the reactivity worth of those isotopes is therefore determined based on consideration of uncertainties of cross sections of fission products documented in [12]. The overall uncertainty is derived from the uncertainty associated with each individual isotope's cross section for lumped fission products (LFPs) and all the other fission product (FPs) separately. *i<>> >.

Also note that recent studies [13] indicate that the total cross section uncertainty for 16 prominent fission products is only about 1% (one standard deviation), i.e. 2% at 95% probability at a 95%

confidence level ~ i< i 1_This value is much higher, and presents a margin that may be reduced in the future based on additional information or evaluations.

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These two values are then statistically combined with the other uncertainties in the determination of the maximum kff for those cases where fuel assembly burnup is credited.

Area of Applicability: All conditions analyzed here have neutron spectra in the thermal energy range and the fission products are predominantly thermal absorbers. Additionally, fission processes are affected by the resonance integrals of the absorbers. The fission product cross section uncertainty is evaluated for the thermal neutron energy range and the resonance integral.

The uncertainties are therefore directly applicable to the calculations performed here.

4.2.2.2 CASMO-4 Fuel depletion analyses during core operation were performed with CASMO-4 (using the 70-group cross-section library), a two-dimensional multigroup transport theory code based on the Method of Characteristics [8]. Note that CASMO-4 is not used for any criticality calculations, i.e. to calculate a keff or kinf value. A validation for CASMO-4 to calculate reactivities is therefore not required here, but an uncertainty in the spent fuel isotopic composition is considered (see below).

4.2.2.2.1 Uncertainty in the Isotopic Content of Spent Fuel To account for the uncertainty in the depletion calculations performed in CASMO, a,,5%

depletion uncertainty factor as described in [3, 10] is used.

The uncertainty is applied by multiplying it with the reactivity difference (at 95%/95%) between the MCNP5 calculation with spent fuel and a corresponding MCNP5 calculation with fresh fuel:

2 Uncertaintylsotopics = [(kcalc kalc-2 ) + 2 * -/(O'calcl + cGcalc_2) ]

• 0.05 With kcalc-i =kcalc with fresh fuel kcalc-2 = kcac with spent fuel acalc-l = Standard Deviation Of kcalc-I acal,-2 Standard Deviation of kcalc-2 This value is then statistically combined with the other uncertainties in the determination of the maximum keff where fuel assembly burnup is credited.

The value of 5% depletion uncertainty is discussed in [3, 10] in the direct context of spent fuel criticality calculation, and is therefore directly applicable to the calculations performed here.

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4.2.3 Analysis Methods 4.2.3.1 Design Basis Fuel Assembly The Seabrook SFP contains three types of Westinghouse 17x17 fuel assemblies: the Standard design, Vantage 5 and RFA. The goal is to find a single type of fuel assembly that is limiting under all conditions. This way, this single set of parameters can be used in all design basis calculations, and it is not necessary to perform some design basis calculations with different sets of parameters, or to consider interfaces between areas with different fuel types. To determine this limiting set, calculations are performed for all cases, and for representative bumup and enrichment combinations. If one set shows results that either bound or are statistically equivalent to the others, then that set is used as the design basis in all further analyses. If there are minor discrepancies- in some cases or burnup and enrichment combinations, it may still be possible to use one set as a single design basis assembly if those discrepancies can be shown to be easily covered by the overall margins. If none of the actual sets of dimensions and operating parameters fulfills the stated goal, then a "hybrid" set of parameters may be established, combining those parameters that result in higher k-values, or appropriate bias values may be applied to selected analyses. Note that some parameters of fuel design, for example, the fuel pellet density has changed over time. Such variations in the fuel design are examined and the most reactive parameters are used in all design basis calculations.

4.2.3.2 Reactivity Effect of, Spent Ftuel Pool Water Temperature The main effect of changes in the SFP water temperature is the impact of the difference in water density. Additionally there may be~temperature-dependent cross section effects that need to be considered. The optimum condition may therefore be different from case to case in this analysis.

To ensure that the optimum condition is used for each case, calculations are performed for each case with representative burnup and enrichment combinations. To cover all normal and accident conditions, those calculations are performed for the following conditions, using MCNP5:

" Normal Ambient Temperature (about 300 K, which is also the standard temperature for MCNP and its cross sections). However, an upper bound water density of 1.0 g/cm 3 is used for those cases, which also covers potential lower water temperatures since this density corresponds to a water temperature of 4 'C. These calculations are performed for 0 and 500 ppm soluble boron.

• Maximum normal SFP temperature. The input files for those conditions use the appropriate water density and S (alpha, beta) according to the temperature. Note that S (alpha, beta), is only available at fixed temperatures of 300K and 400K in MCNP5. Calculations are therefore performed with the adjustment for both temperatures, and the reactivity for the maximum pool temperature is determined by interpolation. These calculations are performed for 0 and 500ppm soluble boron.

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Temperature accident condition. For this, the saturation temperature at the submergence depth of the fuel in SFP of about 260 F (400 K) is used, with and without boiling. Boiling is simulated by assuming a void of 10%, i.e. a reduction in the water density by 10%. These calculations are performed for 1000 ppm soluble boron.

After the results of those studies are determined, the design basis calculations then use the water density and temperature adjustment identified to result in higher reactivities for each pattern.

4.2.3.3 Fuel and Storage Rack Manufacturing Tolerances Traditionally, fuel and storage rack manufacturing tolerances, and other uncertainties such as positioning of fuel in the rack cells, are accounted for by evaluating the reactivity effect of each uncertainty, and then combining those statistically with the other uncertainties (e.g. from validation). In order to ensure that uncertainties are considered appropriately, those calculations need to be performed for all cases, for a bounding set of burnup, enrichment and cooling time combinations, and for water with and without soluble boron. The calculational effort of this approach is enormous, specifically when Monte Carlo codes are used to determine individual reactivity effects. However, the contribution of the tolerances to the total uncertainty is rather moderate. In order to simplify this process, a different approach is used in the current analysis.

This different approach is based on an observation that is made for many uncertainty analyses using the traditional approach stated above: there are only a few parameters that have significant reactivity effect, while the majority of the parameters have small individual reactivity effect, which is almost insignificant when statistically combined with all other larger uncertainties.

Modeling those small effect ýtoleraiicks*&witiW'"easonable accuracy would require very long computer run times. But ignoring those tolerances completely could be considered mildly non-conservative. However, it should -be !relatively easy to show that the effect of several most dominant fuel tolerances considered together (i.e. a calculation where both parameters are considered to be at its limit) would bound the statistical combination of all fuel tolerances. The same is true for the rack tolerances.

The overall process using this approach is therefore as follows:

• Design basis calculations are performed with nominal parameters (as in the traditional approach)

" For each design basis case, two additional MCNP calculations are performed:

o For fuel tolerances, a single calculation where the four most dominating parameters are assumed to be at the limiting condition. In this analysis, those parameters are fuel enrichment, fuel density, pin pitch and clad OD.

o Likewise for rack tolerances, a single calculation where the two most dominating parameters are assumed to be at the limiting condition, but the bounding assumptions are different between Region 1 and Region 2 racks. The dominating parameters are flux trap and cell ID for Region 1 racks, flux trap and wall thickness for Region 2 racks.

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• For each of those cases, the reactivity difference to the corresponding nominal case would be calculated at a 95/95 confidence level, and then those two uncertainties would be statistically combined with the other uncertainties of the calculation.

• A comparison of this approach is performed with the approach using the statistical combination for a representative number of cases covering all aspects of the fuel and storage systems, to show that those values in fact bound a "traditional" statistical combination of all uncertainties for fuel and racks. If the discrepancy between the two approaches is too large (either high or low), then the number of dominating parameters used in the two calculations would be adjusted.

• This is similar to the approach for FPs or LFPs, or for the depletion uncertainty, where additional MCNP calculations are performed for each design basis calculation to evaluate specific uncertainties.

The advantages of this approach are as follows:

• The calculated fuel and rack tolerance are still conservative.

• All uncertainty and tolerances are still considered.

• Numbers of calculations increase only moderately (only two per design basis case) compared to the traditional approach.

The uncertainty is now calculated appropriately for each design basis case, but not as a bounding value for all calculations. -

4.2.3.4 Fuel Radial Positioning Studies are performed to determine the reactivity effects of the fuel radial positioning. The studies consider both 2x2 array of cells as well as larger arrays that represent an entire rack. This way, both local and global effects of the radial positioning are accounted for. The resulting reactivity effects are then considered as uncertainties in the calculations of maximum keff to determine the loading curves.

4.2.3.5 Spent Fuel Reactivity Calculations 4.2.3.5.1 Operating Parameters Principal operating parameters of the fuel discussed here are moderator temperature, fuel temperature, soluble boron concentration in the core, and the power density. Other parameters such as axial burnup distribution and effect of burnable absorbers are discussed in some of the following sections. Generic studies [14, 15] indicate that the operating parameters that result in higher reactivities are the upper bound moderator temperature, fuel temperature, and soluble boron concentration, while the power density has a comparatively small effect with no clear trend or even a possibly higher reactivity at lower power density. Upper bound values are therefore used for the first three parameters. For the power density, a lower bound value would Project No. 2064 Report No. HI-2114996 Page 34 Shaded Areas Denote Holtec International Proprietary Information

be inconsistent with the higher fuel temperature. Therefore, consistent with the guidance in [10],

a nominal value is used for the power density. Additionally, sensitivity studies are performed to show the effect of the individual parameters, and to confirm that the selected values are in fact conservative.

4.2.3.5.2 Spent Fuel !sotopic Composition To perform the criticality evaluation for spent fuel in MCNP5, the isotopic composition of the fuel is calculated with the depletion code CASMO-4 and then specified as input data into MCNP5.

Isotopic compositions are calculated in CASMO-4 for a range of enrichments, for burnups, in increments of about 2.5 GWd/mtU, and for cooling times between 0 and 20 years Note that there may be some uncertainty associated-with the,-isotopic composition of the fuel at the specified cooling time which is not already considered by the depletion uncertainty (Section, 4.2.2.2.1), the bumup uncertainty (Section 4.2.3.5.4) or the FP/LFP uncertainty (Section 4.2.2.1.1.2). The uncertainty of the cooling time on reactivity is therefore determined by performing cooling time studies. These studies are described in Section 4.2.3.5.7.

Note that CASMO-4 has the capability to track isotopic compositions both as assembly average and for each pin ("pin-wise"). The design basis calculations use the assembly average option for simplicity and conservatism. To check if this approach is acceptable, pin-wise studies are performed where pin-specific isotopic compositions are extracted from the CASMO-4 calculations and assigned to the corresponding pins in the MCNP5 calculation. These pin-wise studies are performed in conjunction with the studies for burnable absorbers and reactivity control devices (see Section 4.2.3.5.3). Calculations are performed with the 2x2 array models and cover the range of burnups for selected fuel design. The positive reactivity effect is applied as a bias to account for the reactivity effect of the reactivity control devices and the presence of pin specific isotopic composition in determination of the maximum keff in Section 4.6.11.

4.2.3.5.3 Burnable Absorbers and Reactivity Control Devices 4.2.3.5.3. 1 Burnable Absorbers Project No. 2064 Report No. HI-2114996 Page 35 Shaded Areas Denote Holtec International Proprietary Information

Both integral burnable absorber and non-integral burnable absorber were used for Seabrook. The integrated fuel burnable absorber (IFBA) rods with a thin coating of ZrB2 on the U0 2 pellet were used to replace certain fuel rods in certain assemblies for all the cycles except for Cycle 1.

Generic studies [9] have investigated the effect of integral burnable absorbers (IFBAs). These studies have concluded that there is a small positive reactivity effect associated with the presence of IFBA rods, compared to an assembly configuration where all rods contain fuel and no neutron absorber. This is mainly because of the spectrum hardening caused by the thermal neutron absorber. To confirm the conclusion from [9] is applicable to the Seabrook fuel, studies are performed for selected cases where the absorber is explicitly modeled in the depletion analyses, so the fuel composition transferred to the MCNP criticality calculation (without any residual absorber and on a pin-by-pin basis) contains the effect of the absorber.

The burnable poison rod assemblies (BPRAs) were only utilized for low enrichment fuel in Cycle 1 (see Section 4.5.3) and then replaced with IFBA rods for the remaining cycles.

Therefore, in this analysis, their potential reactivity effect is only evaluated for Standard design fuel assembly (which is the only fuel type utilized for Cycle 1) for fuel with enrichment less than 3.6 wt% 235U.

Following these studies, all design basis calculations are still performed without any burnable absorbers and reactivity control devices. However, in order to account for the presence of the burnable absorbers, for fuel with enrichment less than 3.6 wt% 231U, the maximum positive reactivity effect associated with both IFBA and BPRA configurations is applied in the design basis calculations as a bias; for fuel with enrichment equal to and larger than 3.6 wt% 235U, only the maximum positive reactivity effect of IFBA rods is used a. awbias. 's 4.2.3.5.3.2 Reactivity ControlDevices The Seabrook reactor typically operates ARO (All Rods Out) at full power operations, where the control components are above the top of the active region. Special consideration of those control components on the reactivity of the assemblies is therefore not necessary.

4.2.3.5.4 Burnup Uncertainty In order to account for the uncertainty in the recorded burnup in reactor records, an uncertainty of 5% is used for the burnup measurement uncertainty. The reactivity effect of this uncertainty is considered and statistically combined with the other uncertainties in the determination of the maximum krr. The calculations performed at various burnups for each enrichment that are used to interpolate the burnup for the target keff (see Section 4.2.3.7) are used to determine this reactivity effect. To consider the reactivity difference at the 95/95 confidence level, the following equation is used:

Delta kcalc = ((kcalc 2 - kcalcl) +2 * (/(012+ o22))/(bul - bu 2) *( 5%

• bul)

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kcalc2, (2 Calculated keff and standard variation at (lower) burnup bu 2 kcalcl, 0"1 Calculated kefr and standard variation at (higher) burnup bul Note that in this approach, the lower burnup bu 2 is 0 GWd/MtU, and the higher bumup bu1 is higher than the bumup at the target keff.

4.2.3.5.5 Axial Burnup Profiles and Axial Enrichment Variations Two different enrichment distributions have been used at Seabrook:

• Fuel with enriched blankets. Annular fuel pellet was used for enriched blankets at the top and bottom of the fuel.

• Fuel without blankets, i.e. with an axially constant enrichment Each distribution results in different axial burnup profiles. Separate profiles are therefore determined based on a large number of plant-specific assembly information or generic profiles, and bounding profiles as a function of burnup are established. Additionally, a uniform burnup profile (constant burnup over height) is used, since this may result in a higher reactivity. This flat burnup profile is only used with an axially constant enrichment. To ensure that the bounding condition is considered in the design basis calculations, all applicable profiles and the uniform profile are analyzed and the case with the highest maximum keff is then used for each condition.

See Section 4.6.10 for more discussions for axial burnup profiles used in the analyses.

4.2.3.5.6 Fuel Creep and Growth During irradiation the fuel bundle can experience both fuel rod cladding creep down and fuel rod growth. Both of these changes have the potential to change the fuel-to-moderator ratio in both the depletion and in storage rack geometry. thus potentially increasing reactivity.

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4.2.3.5.7 Cooling Time Uncertainty 4.2.3.5.8 BORAFLEX M T Gap Material The residual B-10 (B 4C) in the BORAFLEXTM in the Region 2 racks is not credited in this analysis. The volumes initially occupied by BORAFLEX TM . are therefore filled with the BORAFLEXTM material without B 4 C. BORAFLEXTM is a polymer and replacing it entirely with water in the analysis could be justified. A comparison of configurations with an absorber gap filled with polymer (BORAFLEXTM material without B 4C) and an absorber gap filled with water is therefore performed to show which material is more reactive.

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4.2.3.5.9 Grid Spacers Grid spacers are placed in regular intervals along the length of the assembly. The current fuel design has the bottom and top grid fabricated from Inconel while the mid-span grids are made from Zircalloy. Previous designs had the same number of grids but different numbers of Inconel and Zircalloy grids. In all cases, the grid spacers reduce the amount of water within the assembly but do not provide any substantial additional neutron absorption, however they do change the local neutron spectrum during depletion. Calculations are conducted by increasing the clad thickness and clad OD of fuel rods to reduce the same amount of water that grid spacers should have reduced. The change is applied for both the depletion calculations and the SFP calculations. The evaluation is performed with and without boron to demonstrate that the effect of grid spacers on reactivity in the SFP is negligible.

4.2.3.5.10 RCCA Type and Configuration As stated in Section 4.1, RCCAs are credited for neutron absorption in selected fuel assemblies in the Region 2 racks. Analyses are therefore to consider 2x2 arrays containing 0, 1 and 2 RCCAs. Two types of RCCAs, EP-RCCA and RCCA (OEM) may be placed in those selected fuel assemblies. RCCA placement for the two RCCA case can be either adjacent or checkerboarded. Evaluation is performed and the RCCA type and configuration which provide the most limiting results is determined and used in the design basis calculations for each pattern.

Studies are also performed for the Region 2 racks to demonstrate the effect of -tolerances of 4RCCA specifications on reactivity in the SFP.

4.2.3.5.11 Model Simplifications

'While the fuel and rack models used in the analyses are very detailed, to assure the true reactivity will always be less than the calculated reactivity, a number of conservative design criteria and assumptions were still necessary to be employed. The following is a list of all those simplifications, together with a brief discussion of the possible effect, and an identification of those simplifications where additional studies may be necessary (and performed in Section 4.6) to show that they are acceptable.

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

2) Dishing and chamfering of the fuel pellets is neglected, i.e. the fuel is always modeled as solid cylinder inside the cladding. This is acceptable since the amount of fuel is maintained, and the water-to-fuel ratio and the principal location of the fuel remain unchanged.

3) For the criticality analysis, the axial noding is consistent with the cold dimensions and the thermal expanded nodal information is not utilized.

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4) All fuel cladding materials are modeled as pure zirconium, while the actual fuel cladding consists of one of several zirconium alloys. This is acceptable since the model neglects the trace elements in the alloy which provide additional neutron absorption.

5) Minor parts of the fuel and rack construction are neglected and replaced by water. Those include grid straps and similar items. Calculations and studies are performed to demonstrate that this is conservative.

6) All fuel and rack structures above and below the active region of the fuel are neglected and replaced by unborated water, even when borated water is used in the active region. This is a standard approach for spent fuel analyses, and while it may neglect some reflection from steel structures in those areas, it also neglects the absorption in those steel structures, and maximizes the axial water reflection. It is therefore considered appropriate and conservative.

7) The neutron absorber (BORAL TM ) length in the Region 1 racks is 141 inches, which is shorter than the active length of the fuel of 144 inches. In the calculations it is conservatively assumed that 3 inches at the top of the active fuel and 3 inches at the bottom of the active fuel are not covered by the neutron absorber.

8) To account for potential blistering of the BORALTM panels in the Region 1 racks, a uniform 0.045 inch void is assumed to extend outward from the BORALTM panels. The BORALTM panelss are, assumed to touch the box wall of racks. The total thickness of poison gap used in

.the analysis is therefore the sum of BORALTM thickness (0.075 inch) and BORALTM .,oid

,-thickness (0.045 inch), which is 0.12 inch. It is 0.03 inch larger than the original poisonigap thickness of 0.09 inch. Accordingly, the flux trap of Region 1 racks is reduced from 1.05-inch to 0.99 inch since the BORALTM panels are on both side of the flux trap.

9) No Boron credit is taken for the BORAFLEXTM neutron absorber panels in the Region 2 racks. The BORAFLEXTM panels are replaced by polymer (all B4C is removed in BORAFLEXTM) in the calculational models. Calculations are performed to demonstrate that this is more conservative than using water to replace the BORAFLEXTM panels.

10) The total length of absorber sections of RCCA used in the Region 2 racks is 142 inches, which is shorter than the active length of the fuel of 144 inches. In the calculations, to allow for fit-up and manufacturing flexibility, the tip of absorber rod to the bottom of active fuel is conservatively assumed as 6 inches, i.e., the total length of RCCA absorber in the active fuel region is conservatively assumed as 138 inches.

11) For the fuel assemblies that contain IFBAs, the ZrB2 coating only covers the center section of the active region. However, in this analysis it is conservatively assumed to cover the entire active length of the fuel, i.e., the coating length of IFBA is conservatively assumed as 144 inches.

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12) For the calculations with BPRAs, the BPRAs are conservatively considered to be present along the entire length of the active region.

4.2.3.6 Synergistic Effects Synergistic effects would be those where the concurrent occurrence of a specific aspect yields a different overall effect than the combination of the effects of the occurrence of each individual aspect. As an (arbitrary) example, it could be that the reactivity effect of the change in wall thickness of a rack is different when the cell ID is at its minimum, compared to a nominal cell ID. However, the statistical combination used in the traditional consideration of uncertainties reduces the impact of small changes from the individual component, while the approach used here that considers a conservative combination of tolerances already implicitly includes synergistic effects. No additional evaluations or calculations are therefore performed here to detect any synergistic effects not already explicitly included in the analyses.

4.2.3.7 Calculations to Determine kff Values and Loading Curves 4.2.3.7.1 Calculation of a Single keff Value Applying all the considerations from the previous sections, the calculation of a single keff value for the design basis calculations consists of the following steps:

The-depletion calculation is performed in CASMO-4 for the desired enrichment(s), a range of, burnups, and the desired cooling time, using the bounding set of assembly dimensions and_.,-

operating conditions. For assemblies with blankets, two CASMO-4 calculations are..

necessary, one at the enrichment of the axial blanket, and one at the enrichment of the center..

of the assembly.

• From the desired assembly-average burnup and the axial bumup profile, the burnup of each axial section of fuel is calculated.

• The average isotopic composition of the fuel is then extracted from the CASMO-4 calculation for each axial section, using interpolation between burnup steps if necessary.

• The isotopic compositions are then specified in the input to the MCNP calculation.

• The MCNP models use nominal values for all parameters of fuel, rack and inserts tolerances, for assembly locations in the rack cells, and for the water density.

• The calculated k (kcalc) is determined. As discussed in Section 4.2.3.2, the design basis calculations use the water density and temperature identified to result in higher reactivities at normal conditions. Note that one of the temperature adjustments, S (alpha, beta), is only available at fixed temperatures of 300K and 400K in MCNP5. Calculations are therefore performed with the adjustment for' both temperatures, and the reactivity for the pool temperature can be determined by interpolation.

• The maximum keff value is then calculated by adding all biases, and a statistical combination of all remaining uncertainties. The uncertainties include:

o Uncertainties from validation (criticality benchmarking) o Statistical uncertainty of the calculation (95/95)

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o 5% Depletion uncertainty. To determine this, a second MCNP calculation is performed with fresh fuel at the same enrichment.

o 5% Burnup uncertainty. As discussed in Section 4.2.3.5.4, this is determined from the calculations performed at various burnups for the interpolation process. Note that 0 GWd/MtU is used as the lower burnup in the calculation.

o Fuel uncertainty. As discussed in Section 4.2.3.3, to determine this, a fifth MCNP calculation is performed where the four most dominating fuel parameters are assumed to be at the limiting condition.

o Rack uncertainty. As discussed in Section 4.2.3.3, to determine this, a sixth MCNP calculation is performed where the two most dominating rack parameters are assumed to be at the limiting condition.

o Eccentric positioning uncertainty. As discussed in Section 4.2.3.4, to determine this, MCNP studies are performed to determine the reactivity effects of the fuel radial positioning. The maximum reactivity effect is used for each pattern.

The biases include:

0 Code bias from validation (criticality benchmarking) o Other bias, which includes pin-specific isotopic composition bias and burnable absorber rods bias. As discussed in Section 4.2.3.5.3, to determine this, MCNP studies are performed to determine the reactivity effects of pin-specific isotopic composition, the presence of the IFBA rods and/or the BPRA rods. For each pattern, for fuel with enrichment less than 3.6 wt% 235U, the maximum positive reactivity effect associated with both IFBA and BPRA configurations is applied in the design basis 235 calculations as a bias; for fuel with enrichment equal to or larger than 3.6 wt%

U, only the maximum positive reactivity effect of IFBA rods is used as a bias.

4.2.3.7.2 Target keff The approach used here takes credit for soluble boron under normal conditions (see Section 4.3).

Under this approach, the limiting condition is the non-borated condition, which needs to be shown to result in maximum kerr of less than 1.0. Many aspects of the overall approach use conservative assumptions and bounding values, which will add additional margin. However, the overall magnitude of this margin cannot always be clearly determined, and will be different between cases. To provide additional margin that is applicable to all calculations, a target kerr value of 0.99 is used, which provides an additional margin of 0.01 delta-k for each calculation Project No. 2064 Report No. HI-2114996 Page 42 Shaded Areas Denote Holtec International Proprietary Information

compared to the regulatory limit. Similarly, for the cases with credit for soluble boron, where the regulatory limit is 0.95, a target keff of 0.94 is defined to provide 0.01 delta-k additional margin.

4.2.3.7.3 Determination of Loading Curves The loading curves consist of the burnups that meet the target kerr for the given range of enrichments and cases. There is no direct way to determine this burnup, and a "trial and error" approach would use an excessive number of calculations. Therefore, an interpolation approach is used, where krff values are calculated for several burnups, and the burnup that meets the target kerr is then determined by a linear interpolation between two burnups. Burnups are selected as multiples of 5 GWd/mtU. If interpolation is not possible, i.e. if the burnup corresponding to the target keff is outside the burnup range analyzed, then additional higher or lower burnups are analyzed until an interpolation can be performed.

Since several axial profiles need to be analyzed, kerr values for all applicable profiles are calculated, and the maximum kerr value for any of those profiles is used in the interpolation process. This approach avoids the pre-determination of a "cross-over" burnup above and below which different axial profiles may be bounding.

4.2.3.7.4 Soluble Boron Concentrations Calculations for normal and accident conditions for compliance-with the regulatory limit of 0.95 are performed at fixed soluble boron levels, selected such that the limit is met for all cases and conditions. Note that the calculations with soluble boron are performed at a bounding water temperature for each cas-e.

4.2.4 Normal Conditions The normal conditions considered in the analyses are all those conditions that can normally occur with fuel in the pool and include the following:

" Normal locations of fuel in the racks according to the analyzed cases and patterns.

" Single fresh fuel assembly in water. This condition bounds all situations during normal fuel movement in the pool, since only a single assembly can be moved at a time. It also bounds the situation of a fuel assembly raised on pedestal or placed in the fresh fuel elevator during fuel inspection.

" The process of inserting and removing fuel assemblies into and from the racks. Two aspects are to be considered here, the insertion and removal of an assembly from a location that requires RCCAs, and the conditions of a partially inserted assembly, which places different sections of neighboring fuel assemblies closer to each other:

o Depending on the fuel and the pattern, it may be necessary to remove a fuel assembly that contains RCCA, while all patterns in and around the cell where the fuel assembly needs to be removed still correspond to one of the analyzed cases. This way, additional analyses are required for this operation to confirm that it is acceptable to replace any fuel assembly that contains RCCAs with a water hole (empty cell).

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Within each pattern, the reactivity is maximized by the fact that axial sections that dominate reactivity are aligned between neighboring assemblies. For example, for highly burned assemblies the dominating area is the lower burned upper section of the active region. The analyses utilize an infinite array of identical assemblies, i.e. those dominating regions are perfectly aligned. When one assembly is being removed from the rack, this alignment is locally disturbed, i.e. dominating regions are no longer aligned between the assembly that is being removed and those around it. The process of removal (or insertion) of an assembly will therefore maintain or reduce the reactivity, but cannot result in an increase in reactivity. No further analysis is therefore required for this process.

• The fuel currently stored in Rows I and 2 on the periphery of the Region 2 rack adjacent to the pool wall (the periphery of Racks 3, 4, and 5 adjacent to the west side of the pool shown in Figure 4.5.5) require unusual plant actions to be reached. They are nozzle-less handling tool assemblies and can not currently be safely moved. It is also difficult to place RCCAs in these locations. Based on the specification (burnup, enrichment and cooling time) of those assemblies, they can be not covered by the loading curve of Pattern D (Uniform loading of spent fuel assemblies without any RCCAs in Region 2) directly, therefore additional analysis is required to determine the requirement to place those fuel assemblies there.

• The debris/trash storage baskets, RCCAs may be stored in empty rack cells, no calculations are necessary since the non-fuel parts would always replace fuel in a cell, and would result in a reduction in reactivity.

• Minor damage to cells, i.e. damage that results in only minor distortion of the cell geometry, and no relocation of the poisoh (if credited), has an insignificant effect. No additional analysis is needed.

• All the following normal condition and activities are also covered:

o Ultrasonic Testing (UT) of fuel assembly to determine leaking rods.

o Reconstitution of fuel assembly o Fuel rod inspection (ECT, visuals, etc) o Storage of damaged fuel rods, fuel rod inserted in FRSB o Fuel assembly inspection o UT fuel assembly cleaning o Bottom nozzle inspections o Fuel assembly debris removal o Top Nozzle Separation visual inspection 4.2.5 Accident Conditions The credible accidents considered are the effect of temperature exceeding the normal range, dropped assemblies, misloaded and mislocated single fresh assemblies, a missing RCCA, and the multiple misload of burned assemblies. Those are briefly discussed in the following sections, and cases are identified that need to be analyzed specifically in Section 4.6.12 of this report. Note that the double contingency principle as stated in [3] specifies that "two unlikely independent and concurrent incidents or postulated accidents are beyond the scope of the required analysis."

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This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

4.2.5.1 Temperature and Water Density Effects The approach that considers temperatures exceeding the normal operating range of the SFP is discussed in Section 4.2.3.2.

4.2.5.2 Dropped Assembly - Horizontal For the case in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the active fuel region of more than 12 inches, which is sufficient to preclude neutron coupling (i.e., an effectively infinite separation). Consequently, the horizontal fuel assembly drop accident will not result in a significant increase in reactivity. Furthermore, any reactivity increase would be small compared to the reactivity increase created by the misloading of a fresh assembly discussed in one of the following sections. The horizontal drop is therefore bounded by this misloading accident and no separate calculation is performed for this drop accident.

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

4.2.5.4 Misloaded Fresh Fuel Assembly The misload of a single fresh assembly is considered the bounding condition for a single assembly misload. As a bounding approach, the misload is considered in a cell that is required to be a burned assembly which would maximize the reactivity effect of the misloaded assembly according to the analyzed cases. The various configurations for this misload are analyzed for all patterns in Region I and Region 2.

4.2.5.5 Mislocated Fresh Fuel Assembly The mislocation of a fresh unburned fuel assembly could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt% 235U) were to be accidentally mislocated outside of a storage rack adjacent to other fuel assemblies. Conservatively, this location was chosen by assuming the mislocated assembly is in the corner junction of two racks. It is important to note that even for Region 1 racks there is no BORALTM panel on the peripheral of storage cell modules, Project No. 2064 Report No. HI-2114996 Page 45 Shaded Areas Denote Holtec International Proprietary Information

except for the portion of the racks facing the other region. An assembly located outside is therefore not separated from the assemblies in the rack by BORALTM plates.

Further, an assembly could be accidentally placed next to the new fuel elevator while another assembly is in this elevator, and two assemblies could be at a close approach for fuel cleaning.

These accident conditions are conservatively evaluated by assuming two fresh assemblies in water at close distance.

4.2.5.6 Missing RCCAs Since RCCAs are credited for reactivity control, the accident of a single missing RCCA is considered for Patterns B and C. However, this condition is bounded by the misloading accident discussed in Section 4.2.5.4, since the misloading accident includes a condition where a fuel assembly that contains a RCCA is replaced by a fresh fuel assembly. Therefore no additional calculation is needed. For multiple missing RCCAs see the following section.

4.2.5.7 Incorrect Loading Curve While several independent misloads are precluded by the double contingency principle, a multiple misload could be the result of an incorrect application of the loading curves. This would be most likely between cases that are very similar, such as Patterns B and C which are both in Region :2 racks, and both contain RCCAs. The,,misload of Pattern B fuel that would.,.need 2 RCCAs in every four cells into a Pattem!Cconfiguratidn (1 insert in every four cells) is therefore considered for a radially infinite array, of cells:, This condition could also be interpreted as multiple missing absorber rods, one in each 2x2 section Calculations are therefore performed for the-Pattern C configuration with the enrichment and burnup requirements from Pattern B, and the Pattern D configuration with the enrichment and burnup requirements from Pattern C.

4.2.6 Interfaces With two different rack types and three configurations (patterns) for each rack type, there are three interface situations that need to be considered:

• Interfaces of different patterns within one rack;

• Interfaces between different racks of the same type; and

• Interfaces between different rack types The approach taken in each of those cases is discussed in the following subsections.

4.2.6.1 Interfaces of Different Patterns within One Rack Although it is desirable to use the same case over a larger area of the pool (e.g. in an entire rack),

it may also be beneficial to have transitions. between patterns within a rack. Placing case configurations directly next to each other without any further considerations may be Project No. 2064 Report No. HI-2114996 Page 46 Shaded Areas Denote Holtec International Proprietary Information

unacceptable. This problem is often approached by defining special interface patterns, which are then individually shown to be acceptable. This not only increases the calculational effort, but also the complexity of patterns in the pool.

4.2.6.2 Interfaces between Different Racks of the Same Type The interface cell pitch between racks of the same type is larger than the cell pitch within the racks for both Region 1 and Region 2. This additional assembly separation reduces the reactivity at the periphery of the racks. However, a credit for this effect would significantly increase the calculational effort and the complexity of the loading patterns in the pool. Therefore, all rack-to-rack gaps are neglected, i.e. each rack type is assumed to be a single large rack. This way, no additional calculations for assemblies placed on rack periphery are required. If there is a difference in the pattern (or case) on the two sides of a rack-to-rack interface, then it needs to be addressed in the same way as case interfaces within a single rack. See the previous section for details on this subject. ,.. ,

4.2.6.3 Interfaces between different rack types Region 1 and Region 2 racks are almost identical in design except that the BORALTM panels are placed instead of BORAFLEXTM in the Region ,I rack cells. This allows for an easy consideration of the Region I to Region 2 interface that does not require any additional patterns and analyses. This is simply achieved by requiring that the first row of the Region 1 cells next to Region 2 cells have to be loaded as Region 2 cells. This is acceptable without any further analyses since the BORALTM in the Region 1 cells will provide additional neutron absorption, so for the same fuel, the reactivity in Region 1 cells is always lower than in Region 2 cells. With this permission, all 2x2 arrays that span the Region 1 to Region 2 interface can be loaded with any of the cases qualified for Region 2. Expressed differently, this permission enforces a barrier row of assemblies, on the Region 1 side of the interface, which complies with both a Region 2 and Region I case. Note that the interface cell pitch between different types of racks is also larger than the cell pitch within the racks for both Region I and Region 2. This additional assembly separation actually reduces the reactivity at the interface of the racks.

4.2.7 Guidance DSS-ISG-2010-01 The most recently issued spent fuel pool regulatory guidance provided by the Nuclear Regulatory Commission (NRC) is ISG-2010-01. A summary of this new ISG guidance and Project No. 2064 Report No. HI-2114996 Page 47 Shaded Areas Denote Holtec International Proprietary Information

approach used here, together with a reference to the appropriate section of this report, is provided here.

1. Fuel assembly selection: A single bounding fuel assembly is used. See Section 4.2.3.1.

2. Depletion analysis: the methodology for performing depletion calculations meets the guidance as follows:

a. A reactivity decrement of 5% is only used to cover the uncertainty of the spent fuel isotopic number densities. This uncertainty is determined using fresh fuel with no integral burnable absorbers. This methodology is discussed in Section 4.2.2.2.1.

b. The reactor parameters selected are bounding values with appropriate references and are shown to be bounding by sensitivity. This methodology is discussed in Section 4.2.3.5.1.

c. BPRAs and IFBAs were used at Seabrook. Analyses are presented to address the effect of those absorbers, based on the configurations used at Seabrook. This methodology is discussed in Section 4.2.3.5.3.1.

d. Rodded operation has been considered and it has been determined that the Seabrook reactor does not typically operate with control rods inserted significantly into the active fuel region during depletion.

3. Criticality Analysis: the methodology for performing criticality calculations meets the guidance as follows:

a. Axial burnup profiles are developed for blanketed fuel from a large set of site specific axial burnup profiles, while generic prqfiles are used for non-blanketed fuel since most of plant specific profiles;are not~available. The analyses use both uniform burnup and various axial: burnup ;profiles,. depending on the axial enrichment variations, and then use, whichever is most reactive. This methodology is discussed in Section 4.2.3.5.5. ý.

b. Rack model: The storage rack models used are based on appropriate references, and the model considers nominal dimensions of parameters.

c. The analysis considers the minimum B-10 loading in the BORALTMTM panels which is below the manufacturing specifications.

d. Interfaces: the interfaces between storage rack loading configurations is discussed in Section 4.2.6 and is appropriate for regulatory compliance at the 95 percent probability at a 95 percent confidence level.

e. Normal conditions: all normal conditions such as fuel storage and fuel movement as well as fuel inspection/reconstitution and fuel in the fuel elevator have been considered. See Section 4.2.4.

f. Accident conditions: all credible accidents have been considered. See Section 4.2.5.

4. Criticality Code Validation: the methodology for performing criticality code validation demonstrates compliance by the following:

a. Area of Applicability: The area of applicability is clearly defined by the benchmark and is appropriate for use in this criticality analysis. See Section 4.2.2.1.1.1.

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i. Actinides have been validated directly using the HTC experiments. Since no direct validation for fission product reactivity worths are available, uncertainties are accounted for by applying a conservative uncertainty factor to the reactivity effect of the fission products.

ii. The selection of experiments for the benchmark is appropriate, since they include the HTC experiments that were performed specifically for storage conditions of spent fuel, and since the parameters in this analyses are consistent with those covered by the experiments.

iii. An appropriate number of experiments have been used.

b. Trend analysis: the benchmark report performs the appropriate trend analysis on each parameter used to define the area of applicability, provides justification of the criteria used to accept or reject trends (or possible trends), and all trends (if any) have been fully evaluated and appropriately applied.

c. Statistical Treatment: The statistical treatment used considers the variance of the population about the mean, uses appropriate confidence factors, and non-normal distributions (if any) are treated using appropriate statistical methods.

d. Lumped Fission Products: the use of LFP has been accounted for appropriately by applying a conservative uncertainty factor to its reactivity effect. See Section 4.2.2.1.1.2.

e. Code to code comparisons: no code to code comparisons are used to directly validate the criticality codes used in this analysis.

5. Miscellaneous: . ,

a. Precedent: no precedent is explicitly being credited iwthis'analysis.

b.

References:

all the references used are appropriate.

c. Assumptions: all assumptions are clearly explained, applicable and justified. See Section 4.2.3.5.11 and Section 4.4.

4.3 ACCEPTANCE CRITERIA The spent fuel PWR storage racks for Seabrook unit I are analyzed in accordance with the applicable codes and standards listed below. The objective of this analysis is to ensure that the effective neutron multiplication factor (keff) is less than 1.0 with the storage racks fully loaded with fuel of the highest permissible reactivity and no credit for soluble boron, i.e., assuming unborated water in the spent fuel pool. In addition, it is demonstrated that keff is less than or equal to 0.95 with the storage racks fully loaded with fuel of the highest permissible reactivity and the pool flooded with borated water at a temperature corresponding to the highest reactivity.

The maximum calculated reactivities include a margin for uncertainty in reactivity calculations, including manufacturing tolerances, and are calculated with a 95% probability at a 95% con-fidence level [1].

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

. Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criterion Project No. 2064 Report No. HI-2114996 Page 49 Shaded Areas Denote Holtec International Proprietary Information

62, "Prevention of Criticality in Fuel Storage and Handling."

• 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. [3]

" ANSI ANS-8.17-2004, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.

• Code of Federal Regulation 10CFR50.68, Criticality Accident Requirements (for soluble boron)

• ANSI ANS-8.27-2008, Burnup Credit for LWR Fuel.

• USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Rev. 3 - March 2007.

• DSS-ISG-2010-01: Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools.

_P :4.4 ASSUMPTIONS The analyses apply a number of assumptions, either for conservatism "6 to" simplify the calculation approach. Each assumption is appropriately discussed 'and justified in the text.

Important aspects of applying those assumptions are as follows:

1) Bounding or sufficiently conservative inputs and assumptions are used essentially throughout the entire analyses, and in most cases, studies are presented to show that the selected inputs and parameters are in fact conservative or bounding.

2) An evaluation is performed to estimate the overall margins of the analyses. This evaluation includes considerations for potential non-conservatisms throughout the analyses, to ensure those are covered by the margin.

4.5 INPUT DATA 4.5.1 Fuel Assembly Specification The spent fuel storage racks are designed to accommodate three types of 17x 17 fuel assemblies The standard design fuel assembly was used during initial cycles of operation, Vantage 5 was used for Cycles 5-8, and more recently, the RFA design has been employed. The design specifications for these fuel assemblies, which were used for this analysis, are given in Table 4.5.1.

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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 4.5.2.

Temperature and soluble boron values are taken as the upper bound (most conservative) of the core operating parameters of Seabrook. The neutron spectrum is hardened by each of these parameters, leading to a greater production of plutonium during depletion, which results in conservative reactivity values.

4.5.3 Burnable Absorbers The Seabrook fuel have made use of burnable poison rod assemblies (BPRAs) of B2 0 3 and integrated fuel burnable absorber (IFBA) rods with a thin coating of ZrB2 on the U0 2 pellet. The BPRAs were only utilized in Cycle 1 and then replaced with IFBA rods for the remaining cycles.

The design specifications for the IFBA rods and BPRA rods are given in Table 4.5.1 and Table 4.5.3 respectively. At Seabrook, both blanketed and non-blanketed fuels were used for fuel assemblies with IFBA, but only non-blanketed fuel was burned with BPRA rods. Table 4.5.4 shows the enrichment range and axial enrichment distributions for the fuel assemblies with IFBA, and Table 4.5.5 presents the enrichment range for the fuel assemblies that were burned with BPRA rods.

i " .45.4 SEABROOK Storage Rack Specification The storage cell characteristics that are used in the criticality evaluations is summarized in Tables 4.5.6 and 4.5.7. Additional information, including layouts, can be found in Figures.4.5.1 and 4.5.2 in this document.

4.5.4.1 Region 1 Style Storage Racks The Region 1 storage cells are composed of stainless steel boxes separated by a gap with BORALTM neutron absorber panels (attached by stainless steel sheathing), centered on each side of the storage cells. The steel walls define the storage cells and the stainless steel sheathing supports the BORALTM neutron absorber panel and defines the boundary of the flux-trap water-gap used to augment reactivity control. Stainless steel channels connect the storage cells in a rigid structure and define the flux-trap between the sheathing of the neutron absorber panels. Figure 4.5.1 provides a sketch of the Region 1 racks along with the critical dimensions.

The design basis calculational models consist of a 2x2 configuration of storage cells with periodic boundary conditions through the centerline of the water gap on the outer boundary of the cluster of four cells, thus simulating an infinite array of Region 1 storage cells. The analyses are performed for 10B loading of 0.015 g/cm 2. Note that the value is minimum value that is assumed in the analyses without any additional consideration for uncertainty. Localized variation on B-10 areal density is also analyzed in the Region 1 racks, such that the average Project No. 2064 Report No. HI-2114996 Page 51 Shaded Areas Denote Holtec International Proprietary Information

value of panels matches the specification. Figure 4.5.3 shows the MCNP calculational model of the Region 1 spent fuel storage cells, as drawn by the two-dimensional plotter in MCNP for the 2x2 configuration of storage cells model.

4.5.4.2 Region 2 Style Storage Racks The Region 2 storage cells are composed of stainless steel boxes separated by a gap with BORAFLEXTM neutron absorber panels (attached by stainless steel sheathing), centered on each side of the storage cells. The analysis takes no Boron credit for the BORAFLEXTM neutron absorber panels, i.e. the panels. are modeled as polymer without any B 4 C. The steel walls define the storage cells and the stainless steel sheathing supports the BORAFLEXTM neutron absorber panel and defines the boundary of the flux-trap water-gap previously used to augment reactivity control. Stainless steel channels connect the storage cells in a rigid structure and define the flux-trap between the sheathing of the neutron absorber panels. Figure 4.5.2 provides a sketch of the Region 2 racks along with the critical dimensions.

The calculational models consist of 2x2 configuration of storage cells with periodic boundary conditions through the centerline of the water gap on the outer boundary of the cells, thus simulating an infinite array of Region 2 storage cells. Some cases analyzed in the Region 2 racks credit the presence of RCCAs for criticality control. Table 4.5.9 shows the specification of those RCCAs. Figure 4.5.4 shows the MCNP calculational model of the Region 2 spent fuel storage cell with 1 RCCAs (Pattern C), as drawn by the two-dimensional plotter in MCNP for, the 2x2

-configuration of storage cells model.

4.5.4.3 Gaps between Adjacent Racks Figure 4.5.5 shows a diagram of the spent fuel pool layout with the locations of the Region, 1 and Region 2 racks. The cell-to-cell pitches within racks, and between racks across the gaps, are listed in Table 4.5.8.

4.5.4.4 Fuel Rod Storage Basket The Seabrook SFP has a Fuel Rod Storage Basket (FRSB) with the dimensions shown in Table 4.5.10 and presented in Figure 4.5.6. The FRSB is a basket that contains fuel rods or fuel rod debris and that is stored in a fuel storage rack cell in the SFP. For the purposes of this analysis the 235U.

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.

4.5.5 Material Composition Table 4.5.11 shows the material compositions of all materials except fuel that is used in the analyses.

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4.6 ANALYSIS This section describes the calculations that were used to determine the acceptable storage criteria for the Region I and Region 2 style racks. In addition, this section discusses the possible abnormal and accident conditions.

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

4.6.1 Design Basis Fuel Assembly For the Region 1 and Region 2 racks (Patterns A through D), the three assembly types shown in Table 4.5.1 were evaluated for a low and high enrichment each. The calculations are listed in Tables 4.6.1 and 4.6.2. For all assemblies, the presence of burnable absorbers in the fuel assembly (BPRA, IFBA) was neglected for determination of the design basis fuel assembly (see Section 4.6.7 for a discussion of the effect of burnable poison). Also note that only non-blanketed fuel was used for Standard design, only enriched blanketed fuel was used for RFA, but both non-blanketed and enriched blanketed fuel were used for Vantage 5. Calculations were therefore performed to compare Vantage 5 and Standard design with the axial profile for the non-blanketed assemblies, and compare Vantage 5 and RFA with the axial profile for the non-blanketed assemblies.

The resultý 'sfihoWthat the Vantage 5 assembly has the highest reactivity or statistically equivalent-reactivity for all cases when it is compared with Standard design and RFA assembly. Hence, the':

Vantage. 5 assembly type is used as the single design basis assembly in all subsequent calculations.

4.6.2 RCCA Type and Configuration For Pattern B and Pattern C with RCCAs in the Region 2 racks, calculations were performed to evaluate the reactivity of the two types of RCCA shown in Table 4.5.9. The results presented in Table 4.6.3 show that reactivity with RCCA (OEM) is statistically identical to that with EP-RCCA. RCCA (OEM) is therefore used in the RCCA models for design basis cases.

Further note that for Pattern C, the following configurations are considered:

• Storage rack 2x2 array uniformly loaded, with one of the four cells containing RCCAs in the fuel assemblies.

" Storage rack 2x2 array is uniformly loaded, with one of the four cells is empty.

For Pattern B where two RCCAs are credited, the following configurations are considered:

• Storage rack 2x2 array uniformly loaded, with two adjacent of the four cells containing RCCAs in the fuel assemblies.

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" Storage rack 2x2 array uniformly loaded, with two diagonal cells of the four cells containing RCCAs in the fuel assemblies.

• Storage rack 2x2 array loaded with three assemblies, with one of the three cells containing an RCCA in the fuel assemblies. An empty cell is adjacent with the cell containing an RCCA.

• Storage rack 2x2 array loaded with three assemblies, with one of the three cells containing an RCCA in the fuel assemblies. The empty cell and the cell containing an RCCA are diagonal.

Based on the results in Table 4.6.3, the configuration with one RCCA shows the highest reactivity for Pattern C, and the configuration with two RCCAs in adjacent cells shows the highest reactivity for Pattern B, therefore, they are selected for all subsequent calculations unless otherwise stated.

As described in Section 4.2.3.5.10, in order to identify the effect of RCCA tolerances on reactivity, calculations were therefore performed with high and low burnup conditions for the configuration with one RCCA (OEM) for Pattern C, and the configuration with two RCCAs (OEM) in adjacent cells for Pattern B. Results of RCCA tolerances are included in Tables 4.6.56 and 4.6.57, and rows with the following information are presented:

• Reference (All Nominal Parameters) - Reference case includes all nominal parameters of RCCA specifications. The same parameters are used in all further cases except for changes that are.described below;

• RCCA Tolerances - Individual uncertainties are determined by changing selected parameters and then comparing the result to the that of the reference calculation; J 4 RCCA- Absorber Density Increased: absorber (Ag-In-Cd) density is increased o RCCAA - Absorber Density Decreased: absorber (Ag-In-Cd) density is decreased o RCCA - Absorber OD Increased: increasing of the absorber outside diameter is analyzed; o RCCA - Absorber OD Decreased: decreasing of the absorber outside diameter is analyzed; o RCCA - Cladding Thickness Increased: maximum absorber rod thickness is analyzed.

o RCCA - Cladding Thickness Decreased: minimum absorber rod thickness is analyzed.

o RCCA - Cladding OD Increased: maximum fuel rod cladding outside diameter is analyzed.

o RCCA - Cladding OD Decreased: minimum fuel rod cladding outside diameter is analyzed.

o RCCA - Silver Content Increased: silver content in the absorber is increased(

, accordingly, both indium and cadmium contents are decreased o RCCA - Silver Content Decreased: silver content in the absorber is decreased *

',accordingly, both indium and cadmium contents are increased Project No. 2064 Report No. HI-2114996 Page 54 Shaded Areas Denote Holtec International Proprietary Information

o RCCA - Indium Content Increased: indium content in the absorber is increased accordingly, both silver and cadmium contents are decreased o RCCA - Indium Content Decreased: indium content in the absorber is decreased accordingly, both silver and cadmium contents are increased o RCCA - Cadmium Content Increased: cadmium content in the absorber is increased accordingly, both indium and silver contents are decreased o RCCA - Cadmium Content Decreased: cadmium content in the absorber is decreased accordingly, both indium and silver contents are increased 0 Total, RCCA Tolerances: This is the statistical combination of all RCCA tolerances. Note that the maximum value of tolerance for each parameter (absorber OD, cladding thickness, etc.) was used.

Calculation and Display of delta-k Values: Note that in Tables 4.6.56 and 4.6.57, and in other

-tables as indicated in the analyses, if a difference (delta-k) is less than or equal to its uncertainty it is listed as 0. For example, 0.0005 +/- 0.0009 would be listed as 0 to indicate that the difference is statistically insignificant. If a difference is positive and larger than the uncertainty, the uncertainty is added. For example, +0.0025 +/- 0.0009 would be listed as +0.0034. And if a difference is negative and the absolute value is larger than the uncertainty, then the uncertainty is subtracted. For 'example, -0.0025 - 0.0009 would be listed as -0.0034.

• L.. -

Regarding the toleranc*6 calculations, it also needs to be noted the reactivity effect of both increase and decrease of the parameter are calculated, anid only the maximum positive value of reactivity effect for each parameter is used when calculating the statistical combination.

The results show that the reactivity effects of RCCA tolerances are statistically insignificant for Patterns B and C. Therefore, they are not considered for the final determination of maximum keff in Section 4.6.11.

4.6.3 Reactivity Effect of SFP Water Temperature and Density As described in Section 4.2.3.2, calculations are performed for the various SFP water temperature conditions with and without soluble boron. Note that for the maximum normal operating temperature of the SFP of 185 F, the corresponding water density is 0.968 g/cm 3.

Results of the studies are presented in Table 4.6.4 for pure unborated water and in Table 4.6.5 for borated water, which support the following conclusions.

Unborated Water:

For unborated water, the lower temperature is bounding for Region I racks, while the higher water temperature is more reactive for the Region 2 racks, with a reactivity difference up to around 0.01 delta-k. This is generally expected, since Region 1 racks has the higher amount of neutron absorber in the form of BORALTM panels, which is known to result in a negative Project No. 2064 Report No. HI-2114996 Page 55 Shaded Areas Denote Holtec International Proprietary Information

temperature coefficient. To account for this in a simple way, for the design basis calculations, all patterns with unborated water are performed at the lower temperature for Region 1 racks, and at the higher temperature for Region 2 racks. The difference between upper and lower temperature including uncertainty of the difference, listed in Table 4.6.4 for each case.

• Normal Conditions (500 ppm):

For borated water with 500 ppm soluble boron, similarly, the lower temperature results in higher reactivities for Region I racks, while the maximum normal temperature results in a higher reactivity in Region 2 racks. The results are shown in Table 4.6.5. All design basis calculations at that soluble boron level are performed at the lower temperature and the corresponding water density for Region 1 racks, and at the maximum normal temperature and the corresponding water density for Region 2 racks.

" Accident Conditions, except temperature accident:

Comparison of the calculations at 0 and 500 ppm soluble boron shows that the bounding temperature for Region 2 racks is still the maximum pool temperature with the soluble boron.

This would also be true at the higher soluble boron level required for the accident conditions.

All misloading accident conditions for Region 2 racks are therefore evaluated at the maximum normal spent fuel pool temperature and corresponding water density. For Region 1 racks, all misloading accident conditions for are still evaluated at the minimum normal spent fuel pool temperature and corresponding water density since it is bounding.

• Temperature Accident Condition (1000 ppm):

To evaluate temperature accident condition, i.e. exceeding the maximum normal pool operating conditions, calculations were performed at 1000 ppm soluble boron and a temperature of 400K, which corresponds approximately to the boiling temperature of water at the submergence ,depth of the iacks. Calculations are performed at 100% water density corresponding to that temperature,. and at 90% to simulate boiling with 10% void. In Table 4.6.5, the condition. (100% or 90% water density) that results inthe highest reactivity is highlighted in bold. In all cases the calculated reactivities are well below the value for the normal condition at 500 ppm soluble boron. The soluble boron amount of 1000 ppm credited for the temperature accident condition therefore more than offsets the effect of the increase temperature including boiling.

4.6.4 Fuel Assembly Positioning As described in Section 4.2.3.4, calculations with various fuel and insert positioning were performed. The results are listed in Tables 4.6.6 through 4.6.9 for Patterns A through D respectively. The following conditions are analyzed:

• 2x2 array surrounded by periodic boundary conditions, which is the same configuration that is used in the design basis calculations for each case.

o Reference (Design Basis): Assemblies centered in the cells.

o All assemblies moved closest to the center of the 2x2 array. With the boundary condition, this creates a laterally infinite arrangement of 2x2 arrays where the assemblies are close together in each 2x2 array. Note that a configuration with Project No. 2064 Report No. HI-2114996 Page 56 Shaded Areas Denote Holtec International Proprietary Information

assemblies moved away from the center in each 2x2 array would be equivalent due to the boundary condition and is therefore not separately considered o All assemblies moved towards the same comer of the cell. This creates a laterally infinite configuration with assemblies moved towards the same corner of each cell.

Large array where all assemblies are moved closest to the center of that array. This essentially represents entire racks and therefore captures global positioning effects, where the 2x2 arrays evaluate more local effects. A periodic boundary condition is also used on the periphery of the model. This neglects the gap between adjacent racks, and is therefore conservative and simplifies model generation. Calculations were performed for two types of arrays:

o 8x8 array o 1Oxl 0 array For Pattern A in Region 1 racks, it shows that the condition where the fuel assemblies centered in the cells (reference case) is the most bounding condition. For Patterns B, C and D used in the Region 2 racks, the condition where all assemblies are moved closest to the center of large array is the bounding condition. It is also shown in the tables that for each pattern, the results for 8x8 array and l0xl0 array are essentially the same. The maximum reactivity effects of these conditions for each pattern are then considered as uncertainties to determine the maximum ken- in Section 4.6.11.

4.6.5 Core Operating Parameters As described in Section 4.2.3.25.1, a, Sehsitivlity study is performed on the effect of the core operating parameters. The results are-listed in Tables 4.6.10 through 4.6.13 for patterns A through D respectively. They show.',that for Region 1 racks, all parameters have very small effects because of low fuel burnup; while for Region 2 racks that contain fuel with a higher burnup, higher moderator and fuel temperature and higher soluble boron concentration result in higher reactivity, while the power density has a small effect. Therefore, conservative high values are selected for all parameters except the power density. The values used, and listed in Table 4.5.2, are based on the following considerations:

" The moderator temperature is taken as the peak power assembly exit temperature.

• Fuel temperature: A conservatively high value was utilized

• Soluble boron concentration: the value utilized in the analysis is very conservative compared to the past and expected future cycle average values listed in Table 4.6.58.

• Power Density: Core average values are used for power density, consistent with the discussion in Section 4.2.3.5.1 and the small effect of this parameter.

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Since cooling time credit is only taken for Region 2 racks, the calculations are only performed for Patterns B, C and D.

4.6.7 Reactivity Control Devices As discussed in Section 4.2.3.5.3, The Seabrook fuel makes use of burnable poison rod assemblies (BPRAs) of B 2 0 3 and integrated fuel burnable absorber (IFBA) rods with a thin coating of ZrB 2 on the U0 2 pellet.

4.6.7.1 IFBA Rods IFBA rods were used for all the cycles except for Cycle 1. As shown in Table 4.5.4, various patterns of IFBA rods have been used for three types of fuel assemblies.

• For the Standard design fuel assemblies, all fuel with IFBA rods is non-blanketed. IFBA rods calculations were therefore performed with, Standard design fuel assemblies to determine the appropriate bias for non-blanketed fuel. The, 1FBA bias for the non-blanketed fuel is then calculated from the IFBA rods calculations (with Standard design)- compared to design basis calculations (with Vantage 5). The. enrichment: range for non-blanketed fuel with IFBA rods is from 3.4 wt% to 4.4 wt% 2 35U. -

. For the Vantage 5 and RFA fuel assemblies, all fuel with IFBA rods has enriched blanket.

Since it is confirmed in Tables 4.6.1 and 4.6.2 that Vantage 5 assembly has the higher reactivity than RFA assembly, IFBA rods calculations were performed only with Vantage 5 assembly for enriched blanketed fuel. Calculations were also performed for enriched blanketed fuel without IFBA (with Vantage 5) as the base cases, and then the IFBA bias for the enriched blanketed fuel was determined. The enrichment range for enriched-blanketed fuel with IFBA rods is from 3.6 wt% to 5.0 wt% 2 35U except for the top and bottom axial blanket (for which the enrichment is 2.6 wt% 235U).

In all the calculations, cooling time of both 0 and 20 years are used for upper limit of enrichment range for Pattern B, C and D. The values compared are those of the MCNP calculations, using the isotopic compositions from depletion calculations with and without the IFBA rods.

Note that as discussed in Section 4.2.3.5.2, planar average rather than pin-specific isotopic fuel compositions are used in the MCNP calculations for the base cases without any Integral Reactivity Control Devices. To determine the reactivity effect of pin-wise isotopic compositions, for all the cases with IFBA rods, isotopic compositions are assigned on a pin-by-pin basis.

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The calculations are performed with and without boron in the water. The results are listed in Tables 4.6.17 through 4.6.20 for non-blanketed fuel (Standard design), and Tables 4.6.22 through 4.6.25 for enriched blanketed fuel (Vantage 5), and the conclusions are as follows:

" For non-blanketed fuel, the IFBA rods calculations are compared with the design basis calculations. It is shown that IFBA rods have a negligible effect for fuel in the Region 1 racks, but have significant positive reactivity effect for fuel in the Region 2 racks.

• For enriched blanketed fuel, comparisons are made between the cases with IFBA and the reference cases (enriched blanketed fuel without IFBA). It is shown that the reactivity effect of IFBA rods for enriched blanketed fuel is positive and larger than the IFBA effect for non-blanketed fuel for all patterns. However, it should be noted that based on the conclusion in Section 4.6.11, for Region 2 racks it is the non-blanketed fuel not enriched blanketed fuel that is bounding and used in the design basis cases to determine loading curves. The reactivity difference between design basis cases and reference cases is large enough to offset all the reactivity effect of IFBA rods for enriched blanketed fuel, therefore, the IFBA bias from the non-blanketed cases will be utilized.

4.6.7.2 BPRA Rods To evaluate the reactivity effects of BPRA rods, the case with the maximum number (24) of BPRA rods is modeled. Since BPRAs were just utilized in Cycle 1, only non-blanketed Standard design fuel assembly is evaluated with BPRAs, within enrichment range of 2.4 wt% to 3.1 wt%

U which is shown in Table 4.5.5. The BPRA rods calculations are performed by using pin-wise ,isotopic compositions, with and without boron-invthe water, and compared with the design 2 racks, BPRA rods have basis calculations. Results listed in Table 4.6.21 show*,that for Region significant positive reactivity effect, which is~larger than the maximum reactivity effect of IFBA rods. However, for Region 1 racks, BPRA: rods have a negligible effect.

4.6.7.3 IFBA/BPRA Bias Used in the Loading Curve Determination Based on the enrichment range of the fuel with specified type of burnable absorber rods and the results shown in Tables 4.6.17 through 4.6.25, to account for the reactivity effects of IFBA and BPRA rods, bias is applied in the loading curve determination as follows:

• For Region 1 racks, since BPRA and IFBA rods with non-blanketed fuel have negligible effects on reactivity, the only positive reactivity effect is of IFBA rods for enriched blanketed fuel, so it is used as a bias to account for the presence of the poison rods in order to determine the maximum keff in Section 4.6.11.1. All design basis calculations are still performed without any burnable absorber rods.

• For Region 2 racks, o For fuel with the enrichment less than 3.6 wt% 235U, the reactivity effect of BPRA rods is bounding, therefore the maximum positive reactivity effect of BPRA rods is applied as a bias.

o For fuel with enrichment equal to or larger than 3.6 wt% 23 5U, BPRA rods were not used..

The IFBA penalty is taken from the maximum positive reactivity effect of IFBA rods for the non-blanketed fuel. This is appropriate, since the 'loading curves are based on the Project No. 2064. Report No. HI-2114996 Page 59 Shaded Areas Denote Holtec International Proprietary Information

non-blanketed results and Tables 4.6.22 through 4.6.25 clearly show that even considering the blanketed IFBA penalty the non-blanketed cases are bounding.

Also note that only the most limiting effect for the case without boron in the water is used as bias in the development of the loading curves; while the effect for the case with boron in the water is only considered to determine the maximum keff for calculations with soluble boron content discussed in Section 4.6.11.2 and 4.6.12.

4.6.8 Additional Studies and Evaluations 4.6.8.1 BORAFLEX TM Gap Material As described in Section 4.2.3.5.8, for Region 2 racks, a comparison of configurations with an absorber gap filled with polymer (BORAFLEXTM material without B4C) and an absorber gap filled with water is performed, results shown in Table 4.6.26 indicate that using an absorber gap filled with polymer is more conservative. Therefore, it is used in the design basis calculations.

4.6.8.2 Grid Spacers As described in Section 4.2.3.5.9, to find the reactivity effects of grid spacers additional calculations are conducted by increasing the clad thickness and clad OD of fuel rods to reduce the same amount of water that grid spacers should have reduced. "The,evaluation was performed with and without boron in the water. The results presented in Table 4.6.27 show that the change in the amount of water by grid spacers has a negligible, effect on reactivity. The grid spacers are, therefore not considered in the geometric models utilized in the analyses.

4.6.8.3 Fuel Creep and Growth 4.6.9 Fuel and Rack Tolerances Evaluations This section presents the calculations of uncertainties from fuel and rack tolerances, and on the basis of those, it provides estimates of the margin from the use of selected bounding tolerances and modeling assumptions in the design basis calculations.

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As described in Section 4.2.3.3, calculations were performed to identify fuel and rack tolerances with high and low impact on reactivity, in the following termed major and minor tolerances.

Results of fuel tolerances are included in Tables 4.6.29 through 4.6.32 for Patterns A through D, respectively, while rack tolerances are presented in Tables 4.6.33 through 4.6.36 for Patterns A through D, separately. For all the cases calculations were performed for a high and low burnup condition. For the Region 1 racks that include the fresh fuel, calculations were performed with fresh fuel at 5.0 wt%.

In Tables 4.6.29 through 4.6.32, rows with the following information are presented:

" Reference (Design Basis) - Design basis case includes all nominal parameters. The same parameters are used in all further cases except for changes that are described below;

• Fuel Uncertainties - Individual uncertainties are determined by changing selected parameters and then comparing the result to the that of the reference calculation o Fuel - Enrichment Increased: enrichment uncertainty effect was estimated simply by increasing 0.05 wt% enrichment; o Fuel - Density Increased: fuel density is increased o Fuel - Pellet OD Increased: increasing of the pellet outside diameter is analyzed; o Fuel - Pellet OD Decreased: decreasing of the pellet outside diameter is analyzed; o Fuel - Rod Pitch Increased: maximum fuel rod pitch is analyzed; o Fuel - Rod Pitch Decreased: minimum fuel rod pitch is analyzed; o Fuel - Cladding OD Increased: maximum fuel rod cladding outside diameter is analyzed. . .

" Fuel - Cladding OD Decreased: minimum .fuetlrod:,claddirig outside diameter is analyzed.. .

o Fuel - Cladding ID Increased: maximum 'fuel rod cladding inside diameter is analyzed.

o Fuel - Cladding ID Decreased: minimum fuel rod cladding inside diameter is analyzed.

o Fuel - GT/IT OD Increased: maximum outside diameter of the guide/instrument tube is used.

o Fuel - GT/IT OD Decreased: minimum outside diameter of the guide/instrument tube is used.

o Fuel - GT/IT ID Increased: maximum inside diameter of the guide/instrument tube is used.

o Fuel - GT/IT ID Decreased: minimum inside diameter of the guide/instrument tube is used.

• Total, Fuel Uncertainties: This is the statistical combination of all fuel uncertainties. Note that the maximum value of tolerance for each parameter (rod pitch, cladding ID, etc.) was used.

" Fuel - Dominating Parameters - Four most dominating fuel parameters are assumed to be at the limiting condition. Maximum fuel enrichment, maximum fuel density, maximum rod pitch and minimum clad OD are selected in this analysis.

• Difference from Dominant Fuel Parameters: This is the difference between the values in the two previous rows.

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In Tables 4.6.33 through 4.6.36, rows with the following information are presented:

• Reference (Design Basis) - Design basis case includes all nominal parameters. The same parameters are used in all further cases except for changes that are described below;

• Rack Uncertainties - Individual uncertainties are determined by changing selected parameters and then comparing the result to the that of the reference calculation o Rack - Flux Trap Increased: maximum cell pitch is analyzed.

o Rack - Flux Trap Decreased: minimum cell pitch is analyzed.

o Rack - Cell ID Increased: maximum cell inner diameter is analyzed.

o Rack - Cell ID Decreased: minimum cell inner diameter is analyzed.

o Rack -Wall Thickness Increased: maximum box wall thickness is analyzed.

o Rack -Wall Thickness Decreased: minimum box wall thickness is analyzed.

o Rack- Sheathing Thickness Increased: maximum sheathing thickness is analyzed.

o Rack- Sheathing Thickness Decreased: minimum sheathing thickness is analyzed.

o Rack - Poison Gap Increased: increasing of the poison gap are analyzed.

o Rack - Poison Gap Decreased: decreasing of the poison gap are analyzed.

o Rack - Poison Thickness Increased: maximum thickness of the poison (BORALTM for Region 1 racks, BORAFLEXTM without B4C for Region 2 racks) was used; o Rack - Poison Thickness Decreased: minimum thickness of the poison was used; o Rack - Poison Width Increased: maximum width of the poison was used; o Rack - Poison Width Decreased: minimum width of the poison was used; o Rack - Localized Variation on B-10: Localized Variation on B-10 Areal Density of BORALTM is analyzed in the Region 1 racks. To account for local-variations of B-10 areal density, calculations were performed for the 2x2,array where the%8 panel on the inside of the array have an areal density reduced by 20%, Whereas the 8 panels on the outside of the array have an areal density increased by 20%.

• Total, Rack Uncertainties: This is the statistical combination of all rack uncertainties. Note that the maximum value of tolerance for each parameter (cell pitch, cell ID, etc.) was used.

• Dominating Rack Parameters - two most dominating rack parameters are assumed to be at the limiting condition.

o Maximum cell ID and minimum flux trap are used for the Region I racks; o Minimum box wall thickness and minimum flux trap are used for the Region 2 racks.

• Difference from Dominant Rack Parameters: This is the difference between the values in the two previous rows.

Regarding the tolerance calculations for traditional approach, it needs to be noted the reactivity effect of both increase and decrease of the parameter are calculated, and only the maximum positive value of reactivity effect for each parameter is used when calculating the statistical combination.

As discussed in Section 4.2.3.3, reactivity difference of the new approach using dominating parameters in comparison to the traditional approach (statistical combination of uncertainties) are calculated and listed in Tables 4.6.29 through 4.6.36 to indicate the level of conservatism introduced by the new approach. Depending on the case, the additional level of conservatism Project No. 2064 Report No. HI-21 14996 Page 62 Shaded Areas Denote Holtec International Proprietary Information

using the new approach is up to 0.005 delta-k for fuel tolerance, and 0.0039 delta-k for rack tolerance. Only for one case in Table 4.6.35, a negative difference of -0.0002 is shown.

However, this value is much smaller than calculation uncertainty of delta-k values, and it is more than covered by the margin of the analyses and therefore does not affect the validity of the results.

4.6.10 Axial Burnup Profiles Fuel at Seabrook has used two different axial enrichment variations over time. Initially, the fuel without any axial blankets was used, i.e. the enrichment was axially constant. After a few cycles, blanketed fuel with 2.6 wt% enriched blankets was used. Axial burnup profiles were determined separately for both axial enrichment distributions, so that calculations can be performed with the appropriate burnup profile for each enrichment profile (however, note that loading curves are determined in a bounding fashion so that they cover all axial distributions with a single curve rather than having separate curves for assemblies with different axial profiles.). The profiles are determined in a bounding fashion, based on a large number of plant-specific and generic burnup profiles. The approach is described below, followed by a description of the profile database used, and the resulting profiles. Note that the term profile always refers to the profile of relative burnups, i.e. the nodal burnups divided by the assembly average burnup.

4.6,10.1 Approach.

,I-:For: each axial node in each profile type, the nodal values are considered as a;:function.,of the 1 assembly average burnup, and a linear expression is determined that bounds the:.values.-in all profiles for this node. This way profiles are determined that;bound every node ofevery profile in the database at the assembly average burnup, and excessive over-conservatism is avoided since the burnup-dependence of the profile form is considered. Also, since a linear function is selected instead of a step function (as in [18]), this approach avoids steps in the loading curves. Note that the individual profiles are not re-normalized (i.e. the average of all nodes is less than 1.0). This adds an unspecified conservatism, although this would be very small. Since there are only few profiles available above a burnup of about 45 GWd/mtU, the value in each node is considered constant and at the value of 45 GWd/mtU for all burnups above that limit. This is conservative since it neglects the further flattening of the profiles with higher burnups.

4.6.10.2 Profile Databases For the blanketed assemblies, the profile database consists of a total of 3474 profiles, from Cycle 6 and Cycles 8 through 15 of the plant. Note that because many of the profiles are from radial nodes of a single fuel assembly, where some fuel assemblies give 4 radial nodes and some only give I radial node, therefore, for many of the 3474 profiles, the average burnup is not the actual fuel assembly average burnup, but only the radial node average burnup. Those profiles cover all profiles for fuel with enriched blankets as well as some non-blanketed fuel from Cycle 6. It is noted that the profiles from Cycle 15 have 26 nodes which is different from 24 nodes that provided by other cycles. However, the axial segments for 24 node and 26 node profiles are Project No. 2064 Report No. HI-2114996 Page 63 Shaded Areas Denote Holtec International Proprietary Information

actually very close. The 26 nodes profiles are therefore converted into 24 nodes profiles and a bounding profile is found.

For the non-blanketed assemblies, most of the plant-specific profiles were not available, since this fuel was only used in earlier cycles of the plant. Therefore, those profiles were determined based on the generic profile database [17], which is also used as the basis for [18]. From this database, assembly profile for the WE-type 17x17 assemblies which is identical or similar to that used at Seabrook was developed in [19] and selected for determining the bounding profile in this analysis. All the generic profiles used for the non-blanketed assemblies are with 18 nodes.

4.6.10.3 Resulting Profiles The profiles are listed in Tables 4.6.37 and 4.6.38. In each case, the linear function in each node is specified by listing the relative burnup at 0 and 45 GWd/mtU. As an illustration, profiles at 30 and 45 GWd/mtU are shown for each axial enrichment distribution in Figures 4.6.1 and 4.6.2.

Note that the lengths of the axial sections were selected based on the cold condition and the thermal expanded nodal information is not utilized.

4.6.11 Normal Conditions 46.411.1 Patterns A through D with Pure Water

-Using:the- calculational model shown in Figures 4.5.3 and 4.5.4, and the reference;qq7x17;.

,,....Yantage 5 fuel assemblies, the keff values in the spent fuel storage racks have been calculated.by1 MCNP5 for all applicable profiles with different enrichments. Since Pattern A through Pattern D.

all: credit, fuel burnup, loading, curves, (minimum burnup as a function of enrichment) are determined by using the maximum keff value for any of those profiles in the interpolation process for, those cases. For pattern A, a single curve is determined for 0 cooling time. For each of patterns B, C, and D several curves are determined, for various cooling times. The approach to determine the individual points of the loading curves follow the process outlined in Section 4.2.3.7.3. Note that based on the results in Section 4.6.3, for Region 1 racks, all the design basis calculations are performed at the water temperature of 4 'C, and the reactivities with S(alpha, beta) of 300K are used to determine the maximum keff value; for Region 2 racks, all the design basis calculations are performed at the maximum pool temperature with the S(alpha, beta) adjustment for both 300K and 400K, and the reactivity for the maximum pool temperature is determined by interpolation. One calculation for each case is listed as an example in Table 4.6.39. Note that in Table 4.6.39 only results for the axial burnup/enrichment profile that result in the highest keff and therefore establish the minimum burnup are listed. The calculation results show that for most of cases it is the non-blanketed fuel not enriched blanketed fuel that is limiting. The full specifications of the loading curves, i.e. the minimum burnup as a function of various initial enrichments are summarized in Table 4.6.40 for Region I racks, Table 4.6.41 for fuel with enrichment from less than 3.6 wt% 235U in Region 2 racks, and Table 4.6.42 for fuel with enrichment no less than 3.6 wt% 235U in Region 2 racks. Note that calculations are also performed for the step of 3.6 wt% 235U in Table 4.6.41, so that the burnup can be calculated from Project No. 2064 Report No. HI-2114996 Page 64 Shaded Areas Denote Holtec International Proprietary Information

the polynomial functions with BPRA requirements for the fuel within the enrichment range of 2.6 -3.6 wt% 231U. Data was 2then fitted to second order polynomial functions in the form of BU=A+B*E+C*E With BU Minimum Burnup in GWd/mtU E Enrichment in wt%

A, B, C Coefficients The coefficients of those polynomial functions are listed in Table 4.6.43, 4.6.44 and 4.6.45, and the corresponding burnups at the enrichments used in the analyses are listed in Table 4.6.46, 4.6.47 and 4.6.48 accordingly. Note that the coefficients were selected so that the burnup calculated from the polynomial functions and shown in Tables 4.6.43 through 4.6.45 are always equal to or higher than the values listed in Tables 4.6.40 through 4.6.42. The loading curves are also shown graphically in Figures 4.6.3 through 4.6.6, which show the results from MCNP calculations and the polynomial functions.

Additionally, to confirm that it is acceptable to replace either of the assemblies required to have an RCCA with an empty cell (water hole) for Region 2 racks, the following additional MCNP calculations are performed for Patterns B and C by utilizing the loading curves in Tables 4.6.47 and 4.6.48.

• with enrichment of 2.6 wt% 235U and cooling time of 0 years at the minimum required buriiup*

2

• withenrirhment of 5.0 wt% 11U and cooling time of 0 an 20 years at the minimum required

• burnup..

The detailed results are all listed in Table 4.6.49, which show the maximum keff values for thie additional'cases are well below the target keff Of 0.99 specified in Section 4.2.3.7.2.

4.6.11.2 Patterns A through D with Borated Water To ensure that the effective neutron multiplication factor (keff) is less than the regulatory limit with the storage racks fully loaded with fuel of the highest permissible reactivity and the pool flooded with borated water at a temperature corresponding to the highest reactivity, the calculations with soluble boron content of 500 ppm for lower and higher enrichment and corresponding burnup/cooling time from loading curve are performed. The bounding profiles for appropriate cases were used. Note that these profiles were also used for all the subsequent normal and accident conditions. The results of the calculations are presented in Table 4.6.50, and show that the maximum keff values for all cases are less than the target keff of 0.94 specified in Section 4.2.3.7.2.

4.6.11.3 Other Normal Conditions 4.6.11.3.1 Single Assembly in Water Project No. 2064 Report No. HI-2114996 Page 65 Shaded Areas Denote Holtec International Proprietary Information

As described in Section 4.2.4, a single fresh fuel assembly in water was analyzed. This bounds any conditions during movements of a single assembly in the pool, including an assembly located in the fuel elevator. Result is presented in Table 4.6.51 and show that the reactivity for this condition is well below the regulatory limit.

4.6.11.3.2 The Fuel Stored on the Periphery of the Region 2 Rack As described in Section 4.2.4, the fuel currently stored in Rows 1 and 2" on the periphery of the Region 2 rack adjacent to the pool wall (the periphery of Racks 3, 4, and 5 adjacent to the west side of the pool shown in Figure 4.5.5) require unusual plant actions to be reached. They can not currently be safely moved and it is also difficult to place RCCAs in these locations. All those assemblies are with enrichment of 3.1 wt% and cooling time of more than 15 years, the minimum burnup of those assemblies is 24.727 GWD/MTU, which is slightly less than the value of burnup requirement determined by the loading curve of Pattern D (Uniform loading of spent fuel assemblies without any RCCAs in Region 2). However, Additional analysis is performed to confirm that the water next to the two outer rows should offset the slightly reduced burnups in those fuel assemblies. The analyzed configuration is shown in Figure 4.6.7, and a more conservative burnup value of 24 GWD/MTU for those assemblies is used in the calculation. The results are presented in Table 4.6.52 and show that the reactivity for the analyzed configuration is below the reactivity of reference case (infinite array calculation) of Pattern D, and therefore the fuel assemblies stored on the periphery of the Region 2 rack meet the regulatory requirements.

4.6.11.3.3 Interfaces Interfaces between racks, as described in Section 4.2.6.3,, were considered. Figure 4.5.5 is, a layout of the existing Seabrook spent fuel pool, with the gaps between racks detailed for each interface. Further details on the gaps are shown in Table 4.5.8, which shows the assembly center-to-center distance across the rack interfaces in comparison to the corresponding value within the racks. The minimum distances across the gaps are defined by the base plates extensions of the racks, and are considered in the evaluations in this section. During seismic events, the base plate extensions prevent the racks to be any closer. Therefore, seismic event are bounded by the evaluations in this section and do not require any additional calculations.

4.6.11.3.3.1 Gaps between Region 1 Racks Region 1 racks have poison panels on all peripheral walls facing other racks. Further, the assembly distance across the gaps between Region 1 racks is larger than the assembly distance within the racks. Since all the normal patterns must be followed between Region I racks, the condition in the gap is therefore bounded by the infinite array calculations. In this analysis, the gaps are conservatively ignored and no further evaluations are necessary.

4.6.11.3.3.2 Gaps between Region 2 Racks Project No. 2064 Report No. HI-2114996 Page 66 Shaded Areas Denote Holtec International Proprietary Information

The assembly distance across the gaps between Region 2 racks is larger than the assembly distance within these racks. Since all the normal patterns must be followed between Region I racks, the condition in the gap is therefore bounded by the infinite array calculations. In this analysis, the gaps are conservatively ignored and no further evaluations are necessary.

4.6.11.3.3.3 Gaps between Region 1 and Region 2 Racks As described in Section 4.2.6.3, the Region 1 to Region 2 interface does not require any additional patterns and analyses.

4.6.11.3.3.4 Interface between Row I and 2 and the Remainder of the Pool For the fuel currently stored on the periphery of the Region 2 rack adjacent to the pool wall discussed in Section 4.6.11.3.2, the interface between Row I and 2 and the remainder of the Pool does not require any additional patterns and analyses. ..

4.6.11.3.4 Fuel Rod Storage Basket The FRSB is modeled as described in Section 4.5.4.4 and shown in Figure 4.5.6, as bare fuel pins with fresh'fuel enriched to 5.0 wt % 2 3 5 U. The FRSB was modeled to replace the least . :

reactive fuel assembly in each pattern, i.e. the spent fuel assembly in Patterns A and D, or the spent fuel assembly with an RCCA in Patterns B and C. The' results of these calculations are presented in Table 4.6.53 and show that any of the fuel assembly in any pattern may be replaced with a FRSB.

4.6.11.3.5 Other Conditions There are a number of other conditions in the spent fuel pool that is considered acceptable without explicit evaluation based on the following discussions:

• Cells that are required to be empty as part of the respective specification for a given case are generally not permitted to contain any non-fuel hardware. However, insertion of non-fuel hardware in the cells allowed to contain fresh or spent fuel is permitted. Also, storage of inserts or control rods in the empty cells is considered acceptable, since those devices only replace a very small amount of water, while at the same time adding a substantial amount of neutron absorber.

• As already discussed in Section 4.2.4, insertion and removal of fuel assemblies does not require any further evaluation. This is then also applicable to conditions where fuel assemblies are placed in cells on pedestals for inspections or repair.

• As discussed in Section 4.2.4, the minor damage that results in only minor distortion of the cell geometry, and no relocation of the poison (if credited), has an insignificant reactivity effect.

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All the following normal condition and activities have been already covered by the analyzed normal conditions in this report.

o Ultrasonic Testing (UT) of fuel assembly to determine leaking rods.

o Reconstitution of fuel assembly o Fuel rod inspection (ECT, visuals, etc) o Storage of damaged fuel rods, fuel rod inserted in FRSB o Fuel assembly inspection o UT fuel assembly cleaning o Bottom nozzle inspections o Fuel assembly debris removal o Top Nozzle Separation visual inspection 4.6.12 Accident Conditions Accidents considered are the single fresh assembly misload or mislocation, single missing RCCAs, and accidents with multiple missing RCCAs. For mislocated assembly accident conditions in Region 1 racks, a soluble boron level of 1600 ppm is used. For all the other accident conditions, a soluble boron level of 1400 ppm is used. All results are summarized in Table 4.6.54, and are all below the regulatory limit of 0.95 with 0.01 delta-k additional margin.

All calculations except for the mislocation are performed using an 8x8 array of cells surrounded by periodic boundary conditions, which creates a laterally infinite array with an accident in each 8W8 section. Further-details are presented in the following subsection.

4.6.12. 1iýSingle Fresh Assembly Misload The misload of a single fresh, assembly in a cell could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5 wt%) were to be inadvertently misloaded into a storage cell intended to be used for spent fuel. This is evaluated for all patterns, as described in Section 4.2.5.4.

4.6.12.2 Single Fresh Assembly Mislocation The MCNP model consists of an array of Region 1 and Region 2 fuel storage cells with a single fresh, unburned assembly placed adjacent to the rack. The example of configuration is shown in Figure 4.6.8. The mislocated fuel assembly is placed as close to the rack faces as possible to maximize the possible reactivity effect. Conservatively, it is assumed that the mislocated assembly faces rack cells on two sides, and all the fuel assemblies in the rack are pushed to the corner closest to the mislocated assembly outside.

The condition of two fresh assemblies in close proximity in water which bounds the accidental placement of an assembly next to the loaded fuel elevator is analyzed.

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4.6.12.3 Single Missing RCCA As described in Section 4.2.5.6, the accident of a single missing RCCA for Patterns B and C is bounded by the accident of the misload of a single fresh assembly discussed in Section 4.6.12.1.

4.6.12.4 Multiple RCCAs Misload Multiple assembly misload which also could be interpreted as multiple missing RCCAs is considered. The Pattern C configuration with the enrichment and burnup requirements from Pattern B, and the Pattern D configuration with the enrichment and burnup requirements from Pattern C, as described in Section 4.2.5.7, were analyzed.

4.7 CONCLUSION

S Section 4 of this report documents the criticality analysis for the storage of PWR spent nuclear fuel in the Region I and Region 2 spent fuel storage racks for both fresh and spent fuel with an initial enrichment of 1.5 to 5.0 wt% 235U at the Seabrook nuclear power plant. The analysis demonstrates that the effective neutron multiplication factor (keff) is less than 1.0 with unborated water and less than or equal to 0.95 with soluble boron credit at a 95% probability with a 95%

confidence level. Further, the reactivity effects of abnormal and accident conditions have been evaluated to assure that under credible abnormal and accident conditions, the reactivity will not exceed 0.95 with soluble boron credit at a 95% probability with a 95% confidence level. The maximum calculated reactivityt includes a margin for uncertainty in reactivity calculations including manufacturing tolerances' and is shown to be less than the regulatory limit with a 95%

probability at a 95% confidence level. The minimum required burnups as a function of the initial enrichment is listed in Tables '4.6.46 through 4.6.48.

The following patterns were qualified:

" Pattern A: Region 1 storage rack 2x2 array with a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 5.0 wt% 235U, and spent fuel assemblies with an initial enrichment of 1.5 to 5.0 wt% 235U. All four cells contain BORALTM panels with an areal density of 0.015 gm/cm 2 in each side of the cells. No credit for cooling time.

" Pattern B: Region 2 storage rack 2x2 array uniformly loaded with spent fuel with an initial enrichment of 1.5 to 5.0 wt% 235U, with any two of the four cells containing RCCAs in the fuel assemblies, and the consideration of cooling times of 0, 2.5, 5, 10, 15, and 20 years. The B4C in the BORAFLEXTM is not credited for neutron absorption. Additionally, it is acceptable to replace any cells that contain fuel assemblies with empty water cells.

• Pattern C: Region 2 storage rack 2x2 array uniformly loaded with spent fuel with an initial enrichment of 1.5 to 5.0 wt% 235U, with any one of the four cells containing RCCAs in the fuel assemblies, and the consideration of cooling times of 0, 2.5, 5, 10, 15, and 20 years. The B4C in the BORAFLEXTM is not credited for neutron absorption. Additionally, it is acceptable to replace any cells that contain fuel assemblies with empty water cells.

• Pattern D: Region 2 storage rack 2x2 array uniformly loaded with spent fuel with an initial enrichment of 1.5 to 5.0 wt% 215U, without any RCCAs in the fuel assemblies, and the Project No. 2064 Report No. HI-2114996 Page 69 Shaded Areas Denote Holtec International Proprietary Information

consideration of cooling times of 0, 2.5, 5, 10, 15, and 20 years. The B4C in the BORAFLEXTM is not credited for neutron absorption. Additionally, it is acceptable to replace any cells that contain fuel assemblies with empty water cells.

For all cases with spent fuel, loading curves were determined as polynomial functions in the form. 2 BU=A+B*E+C*E With BU Minimum Burnup in GWd/mtU E Enrichment in wt%

A, B, C Coefficients The coefficients of these functions are listed in Tables 4.6.43 through 4.6.45. When curves for several cooling times are specified, no interpolation for intermediate cooling times are permitted, i.e. the curve for a cooling time at or below the actual cooling time of the assembly must be used.

Additionally, all the normal conditions discussed in Section 4.2.4 were analyzed and qualified.

Particularly,

" The fuel rod storage basket is qualified for all locations that are qualified for any fresh or spent fuel in above cases.

• The Rows 1 and 2 on the periphery, of Region 2 racks adjacent to the pool wall, i.e., the periphery of Racks 3, 4; and 5. adjacent to the west side of the pool shown in Figure 4.5.5, are analyzed and qualified for higher :reactive fuel, crediting the higher neutron leakage in this area.

The interface situations have also been considered. By requiring that that not only neighboring patterns, but also all overlapping patterns meet one of the analyzed cases for every 2x2 array in the pool, and the first row of the Region 1 cells next to Region 2 cells have to be loaded as Region 2 cells, the reactivity will not exceed the regulatory limit for all interface situations.

A soluble boron level of 500 ppm is sufficient to ensure that the maximum keff is below the regulatory limit under all qualified normal conditions.

For accident conditions, a soluble boron level of 1600 ppm is sufficient to ensure that the maximum kefr is below regulatory limit.

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Table 4.2.1 I ,# Ii%" I5>;

• [......: +

..... .... " .. ii '

.... ,. ... , z , 1

-I- 4

' . . . .. _ _ _ _ _ ~

_ ... ~__I..

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Table 4.5.1: PWR Fuel Assembly Specifications 17x17 Standard 17x17 Vantage 5 17x17 RFA design Fuel Rod Data Fuel pellet outside diameter, in. 0.3225 0.3225 * ,0.3225 1 Cladding inside diameter 4 , in. 0.3290 0.3290 0.3290 L, Cladding outside diameter, in. 0.3740 0.3740 0.3740 ,

Cladding material Zr-4 ZIRLO ZIRLO Maximum Pellet density, g/cc 10.4621--, 10.517 10.505 i Maximum enrichment, wt% 23'U 4.402 5.0 5.0 ZrB 2 Coating Loading (mg 1.57' 2.355 2.355

' 0B/inch)

ZrB 2 Coating Thickness", in. 0.000437 0.000657 0.000657 ZrB 2 Coating Length', in. 132 120 1208 Fuel Assembly Data Fuel rod array 17x17 17x17 17x17 Number of fuel rods 264 264 264 Fuel rod pitch, in. 0.496 0.496 .... 0.496.

Fuel Assembly Width, in. 8.426 8.426 8.426 Fuel Assembly Length, in. 159.975 159.975 159.975 Active fuel Length, in. 144 144 144 Bottom of Active Fuel Length to 3.168 to 4.168 3.278 Bottom of Assembly, in. , _____. _____'_"___-_

Guide/InstrumenitTube Data Number of Guide Thimbles 24 24 24 Number of Instrument Tubes 1 -

Guide Thimble Upper Region 0.448 .20.442 .

Inside Diameter, in.

Guide Thimble Upper Region 0.484 3/4, 0.474 V 0 . ...................

0.482 r Outside Diameter, in. 0.484 Guide Thimble Dashpot Region 0.395 0.3970 0.3970 Inside Diameter, in.

Guide Thimble Dashpot Region 0.431 0.4300

• 0.4390 ..

Outside Diameter, in.

Instrument Tube Inside Diameter, 0.448, 0.448 3/4> 0.442 in.

Instrument Tube Outside 0.484 0.484 0.482 Diameter, in. J Guide/Instrument Tube Material Zr-4 ZIRLO ZIRLO 4 The tolerances of cladding inside diameter for Standard design and. Vantage 5 assemblies were not available; the tolerance for RFA assembly was therefore used in analyses for all three types of assemblies.

5 This does not represent Cycle 1. Cycle I included Borosilicate Glass with 12.5 w/o B 2 0 3 .

6 The coating thickness was not available. The values provided are calculated. See Appendix C of

[211.

7 In this analysis, for all cases that contain IFBA, the coating length is conservatively assumed as 144 inches. See section 4.2.3.5.11.

8 For Cycle 15, the coating length has been increased from 120 to 122 inches.

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Table 4.5.2: Core Operating Parameter for Depletion Analyses Parameter Value Soluble Boron Concentration (cycle average), 1100 ppm Uprated Core Nominal Power, MW 3648 Average Core Weight (MTU) 88.04064 Reactor Specific Power9 , MW/MTU 41.435 Conservative Fuel Temperature"°, 'F 1656 Peak Power Assembly Exit Temperature, 'F 646 In-Core Assembly Pitch, Inches 8.466 9 This value is calculated as the uprated core nominal power divided by the average core weight.

10 ~--

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Table 4.5.3: BPRA Specifications Parameter Value Number of BPRA Rods per Assembly 24 (Maximum)

Boric Oxide Content 12.5 Boron- 10 Atom Percent 19.9 Burnable Absorber Material Borosilicate Glass Burnable Absorber Composition (wt %) 40.9006% Silicon, 55.2173% Oxygen, 0.698784% B-10, 3.18335% B-111I Burnable Absorber Density (g/cm 3) 2.299 Burnable Absorber Inner Diameter (in.) 0.19 Burnable Absorber Outer Diameter(in.) 0.336 Burnable Absorber Clad Material Stainless Steel 304 Burnable Absorber Inner Clad Thickness (in.) 0.007 Burnable Absorber Inner Clad O.D. (in.) 0.181 Burnable Absorber Outer Clad Thickness (in.) 0.0185 Burnable Absorber Outer Clad O.D. (in.) 0.381 Bottom of BA Referenced to Bottom of Fuel 12 (in.) +1.987 The Borosilicate Glass is assumed as SiO 2 -B 20 3. The density and composition of SiO 2-B2 0 3 is taken directly from

[19].

12 In this analysis, the BPRAs are conservatively considered to be present along the entire length of the active region. See section 4.2.3.5.11.

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Table 4.5.4: Number of IFBA Rods vs. Enrichment Range for Various Fuel Types Number of IFBA Enrichment Range Axial Blanket (top and bottom)

Fuel Type Rods (wt %) Enrichment Length Fuel Pellet (wt %) (inch) Type 0 1.6-4.4 NA NA NA 64 3.4-4 NA NA NA Standard 80 3.4 - 4.4 NA NA NA design 104 3.4-4 NA NA NA 128 3.6-4.4 NA NA NA 156 4.4 NA NA NA NA or 2.4 - 4.2 NA or 2.6 NA or 6 Anur 0

Annular 32 3.8-4.95 2.6 6 Annular 48 4.2 2.6 6 Annular Vantage 5 80 3.8 - 4.95 2.6 6 Annular

.104 3.8 -4.95 2.6 6 Annular 128 4.8 2.6 6 Annular 156 4-4.95 2.6 6 Annular 0 4-4.2 2.6 6 Annular 16 4-4.7 2.6 6 Annular 32 4.8 - 4.95 2.6;

• 6 Annular 48 3.6-4.8 2'.6.. 6 Annular 80 3.6 - 4.7 2.6 - 6 Annular 104 3.6-4.95 2.6 6 Annular 128 4 - 4.4 2.6 6 Annular 156 4.2-4.95 2.6 6 Annular Project No. 2064 Report No. HI-2114996 Page 75 Shaded Areas Denote Holtec International Proprietary Information

Table 4.5.5: Number of BPRA Rods vs. Enrichment Range for Various Fuel Types Fuel Type Number of BPRA Rods Enrichment Range (wt %)

0 1.6-3.1 3 3.1 4 2.4-3.1 Standard design' 3 8 2.4 9 3.1 12 2.4 2314 3.1 24 2.4-3.1 13 All the Standard design fuel assemblies are non-blanketed.

14 The design with 23 BPRA rods also included a primary source rod.

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Table 4.5.6: Fuel Rack Specifications - Region 1 Racks Parameter Value Storage Cell Inside Dimension, in. 8.9 Storage Cell Steel Thickness, in. 0.09 Storage Cell Sheathing Thickness, in. 0.02 Storage Cell Sheathing Width, in. 7.70 Storage Cell Poison Gap Thickness, in 15 0.09 Storage Cell Flux Trap Width, in 16 1.05 (Nominal)

Storage Cell Pitch, in 10.35

• Storage Cell Material SS304 BORAL TM Thickness, in. 0.075 BORAL TM Blistering Void Thickness, mil 45 BORAL TM 10B Loading (min), g/cm 2 0.015 BORAL TM Width, in. 7.5 Poison Height, in. 17 141.00 15 To account for BORALTM blistering, the total thickness of the poison gap of 0.012 inch is actually used in the analysis. This value is the sum of poison thickness (0.075 inch) and BORAL TM void thickness (0.045 inch). See Section 4.2.3.5.11.

16 The flux trap of 0.99 inch is actually used in all cases except for the peripheral cells to account for BORALTM blistering, see Section 4.2.3.5.11.

17 The poison height of 138 inch is actually used in analysis, see Section 4.2.3.5.11.

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Table 4.5.7: Fuel Rack Specifications - Region 2 Racks Parameter Value Storage Cell Inside Dimension, in. 8.9 _ .;J.j Storage Cell Steel Thickness, in. 0.09 Storage Cell Sheathing Thickness, in. 0.02 Storage Cell Sheathing Width, in. 7.63

• Storage Cell Poison Gap Thickness, in 0.09 Storage Cell Flux Trap Width, in 1.05 (Nominal)

Storage Cell Pitch, in 10.35, Storage Cell Material SS304 BORAFLEX TM Thickness, in. 0.071 7 BORAFLEX TM 1°B Loading (min), g/cm 2 Neglected BORAFLEX TM Width, in. 7.46 <7 Poison Height, in. 141.25 1, Project No. 2064 Report No. HI-2114996 Page 78 Shaded Areas Denote Holtec International Proprietary Information

Table 4.5.8: Interface Dimensions for Region 1 and Region 2 Racks' 8 Interface Rack Cell-to- Cell Centerline Cell-to-Cell Cell Pitch (in) to Baseplate Pitch9 (in) Periphery (in)

Region 1-Region 1 11.06 10.35 5.53 Ref.

Region 2-Region 2 11.04 10.35 5.52 Region 1-Region 2 11.05 10.35 --

18 The values listed in this table are not used in calculations but are provided for reference only.

19 The "Interface Cell-to-Cell Pitch" is determined by combining the respective distances provided in the "Cell Centerline to Baseplate Periphery" Column.

20 The "Cell Centerline to Baseplate Periphery" is taken directly from the referenced drawings and identifies the location of the centerline of the outermost rack cell with regards to the periphery of the rack baseplate.

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Table 4.5.9: RCCAs Specifications Value21 Parameter RCCA (OEM) EP-RCCA Number of Control Rods per 24 24 Cluster Absorber OD, Zone 1, in. 0.341 0.341 Absorb Length, Zone 1, in. 142 130 Absorber OD, Zone 2, in. 0.336 Absorb Length, Zone 2, in. 12 Cladding Thickness, in. 0.0185 0.0185 Cladding OD, in. 0.385 0.381 Total Length of Absorber 124 221414 Sections, in.

Material Silver-Indium-Cadmium Silver-Indium-Cadmium Neutron Absorber (AgInCd) (AgInCd)

Poison Density, g/cm 3 10.17 1.

Silver Content, wt% 80 *' 80 Indium Content, wt% 15 .i 1512 Cadmium Content, wt% 5 5 Cladding Material SS304 SS304 21 All the tolerances shown in this table are assumed values.

22 The total length of absorber sections in the active fuel region is assumed as 138 inches, see Section 4.2.3.5.11.

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Table 4.5.10: Fuel Rod Storage Basket Specifications Parameter Value Number of Cells 52 Cell Pitch, in. 0.937 Array Type 8x8 Basket Wall Thickness, in. 23 0.035 Storage Material SS 304 23 The cell walls are not modeled in the analysis and conservatively assumed as water.

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Table 4.5.11: Material Composition Parameter Value MCNP ZAID Stainless Steel 304 (Density 7.84 g/cc, minimum)

Cr (atoms/barn*cm) 0.0172 24000.50c Mn (atoms/bam*cm) 0.00172 25055.51c Fe (atoms/bam*cm) 0.058377 26000.55c Ni (atoms/bam*cm) 0.008047 28000.50c Zircaloy (Density = 6.5 g/cc)

Zr-40 (atoms/bam*cm) 0.043239 40000.56c Pure Water (Density = 1.0 g/cc)

H-1 (atoms/bam*cm) 0.066854 1001.50c 0-16 (atoms/bam*cm) 0.03343 8016.50c RCCA Absorber Rod (Density =10.17 g/cc)

Ag (wt %) 80 47000.50c In (wt %) 15 49000.60c Cd (wt %) 5 48000.5 1c 500 ppm Soluble Boron in Water (Density = 1.0 g/cc)

H-1 (weight fraction) 0.111840 1001.50c 0-16 (weight fraction) 0.88766 8016.50c B-10 (weight fraction) 9.OOE-05 5010.50c B- II (weight fraction) 4.1000E-04 5011.55c 1000 ppm Soluble Boron in Water (Density = 1.0 g/cc)

H-I (weight fraction) 0.11179 1001.50c 0-16 (weight fraction) 0.88721 8016.50c B-10 (weight fraction) 1.800E-04 5010.50c B-11 (weight fraction) 8.200E-04 5011.55c T

M 2 BORAL with Areal Density = 0.01 5g/cm (Density= 2.665 g/cc)

B-10 (wt %) 2.95 5010.50c B-I1 (wt %) 13.46 5011.55c C (wt %) 4.55 6000.50c Al (wt %) 79.03 13027.50c BORAFLEXTM without B 4 C (Density 1.38 g/cc)

Si (wt %) 41 14000.51c 0 (wt %) 37 1001.50c H (wt %) 4.5 8016.50c C (wt %) 17.5 6000.50c Project No. 2064 Report No. HI-2114996 Page 82 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.1 Effect of Different Type of Fuel Assembly for Fresh Unborated Pool Water Pattern24 Enrichment Burnup Non-Blanketed Fuel Assembly Enriched-Blanketed Fuel Assembly Vantage 5 Standard design Vantage 5 RFA (wt%) (GWd/mtU) k-calc k-calc Deltaik-calc' 95/95 unc k-calc k-calc Delta k-calc 95/95 unc A 3.6 5 0.9593 0.9582 -0:0011- 0.0014 0.9586 0.9566 -0.0020 0.0014 B 2.6 5 0.9457 0.9445 -0.0012 0.0011 0.9433 0.9428 -0.0005 0.0011 C 2.6 15 0.9296 0.9272 -0.0024 0.0011 0.9198 0.9191 -0.0007 0.0011 D 2.6 20 0.9433 0.9424 -0.0009 0.0011 0.9301 0.9294 -0.0007 0.0011 A 5 15 0.9628 0.9612 -0.0016 0.0014 0.9603 0.9590 -0.0013 0.0014 B 5 35 0.9441 0.9422 -0.0019 0.0014 0.9287 0.9267 -0.0020 0.0011 C 5 45 0.9311 0.9299 -0.0012 0.0011 0.9187 0.9171 -0.0016 0.0013 D 5 55 0.9305 0.9291 -0.0014 0.0011 0.9129 0.9125 -0.0004 0.0011 24 Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 5.0 wt% 235U, the enrichments and burnups presented in the table are only for spent fuel.

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Table 4.6.2 Effect of Different Type of Fuel Assembly for Pool Water with 500 ppm Soluble Boron PEnrichment Burnup Vantage Non-Blanketed Fuel Assembly Enriched-Blanketed Fuel Assembly 5 5 Standard design Vantage 5 RFA (wt%) (GWd/mtU) k-calc k-calc Delta k-calc 95/95 unc k-calc k-calc Delta k-calc 95/95 unc A 3.6 5 0.9034 0.9026 -0.0008 0.0014 0.9025 0.9016 -0.0009 0.0014 B 2.6 5 0.8468 0.8464 -0.0004 0.0011 0.8449 0.8443 -0.0006 0.0011 C 2.6 15 0.8312 0.8295 -0.0017 0.0011 0.8237 0.8224 -0.0013 0.0013 D 2.6 20 0.8430 0.8403 -0.0027 0.0011 0.8318 0.8316 -0.0002 0.0011 A 5 15 0.9098 0.9091 -0.0007 0.0014 0.9078 0.9079 0.0001 0.0014 B 5 35 0.8622 0.8610 -0.0012 0.0013 0.8478 0.8465 -0.0013 0.0011 C 5 45 0.8481 0.8473 -0.0008 0.0013 0.8363 0.8366 0.0003 0.0013 D 5 55 0.8451 0.8440 -0.0011 0.0011 0.8304 0.8303 -0.0001 0.0013 21 Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 5.0 wt% 23 5 U, the enrichments and burnups presented in the table are only for spent fuel.

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Table 4.6.3 Reactivity Effect of RCCAs Type/Configuration K-calc 95/95 unc K-calc 95/95 unc Pattern B 2.6 wt% / 5 5.0 wt% / 35 GWd/MTU GWd/MTU 2 RCCAs (OEM) - Row 0.9457 0.0008 0.9441 0.0010 2 EP-RCCAs - Row 0.9459 0.0008 0.9446 0.0008 2 RCCAs (OEM) - Diagonal 0.9277 0.0008 0.9326 0.0008 2 EP-RCCAs - Diagonal 0.9277 0.0008 0.9324 0.0008 1 RCCA (OEM) & I Empty Cell - Row 0.9011 0.0010 0.8947 0.0010 I EP-RCCA & 1 Empty Cell - Row 0.9008 0.0010 0.8947 0.0010 1 RCCA (OEM) & 1 Empty Cell - Diagonal 0.8569 0.0008 0.8567 0.0010 1 EP-RCCA & I Empty Cell - Diagonal 0.8568 0.0010 0.8562 0.0008 Pattern C 2.6 wt% / 15 5.0 wt% / 45 GWd/MTU GWd/MTU 1 RCCA (OEM) 0.9296 0.0008 0.9311 0.0008 1 EP-RCCA 0.9291 0.0008 0.9318 0.0008 1 Empty Cell 0.8815 0.0008 0.8795 0.0010 Project No. 2064 Report No. HI-2114996 Page 85 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.4 Reactivity Effect of Water Density for Fresh Water Unborated Water Water Density, g/cc 1.00 0.9680 S(cP3) 300K 300K 400K 358.15K Delta k-calc 26

______(I (85F)

Pattern2 7 Enr Bu k-calc2 8 k-calc Pattern A 3.6 5 0.9593 0.9513 0.9469 0.9487 -0.0120 Pattern B 2.6 5 0.9464 0.9450 0.9560 0.9514 0.0061 Pattern C 2.6 15 0.9289 0.9284 0.9420 0.9363 0.0085 Pattern D 2.6 20 0.9424 0.9422 0.9588 0.9519 0.0106 Pattern A 5.0 15 0.9628 0.9540 0.9519' 0.9528 -0.0114 Pattern B 5.0 35 0.9454 0.9435 0.9582 0.9520 0.0079 Pattern C 5.0 45 0.9319 0.9319 0.9496 0.9422 0.0114 Pattern D 5.0 55 0.9307 0.9297 0.9485 0.9406 0.0113 26 The delta k-calc is calculated by the results of 1.00g/cc, 300K and 0.968g/cc, 358.15K.

27 Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 5.0 wt% 235U, the enrichments and burnups presented in the table are only for spent fuel.

28 The standard deviation of k-calc is around 0.0005 for all calculations.

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Table 4.6.5 Reactivity Effect of Water Density for Borated Water 500 ppm 1000 ppm Water Density, g/cc 1.00 0.9680 0.9390 0.8450 I l~358.15K 2 S(a,3) 300K 300K 400K (185F) Delta k-calc29 400K 400K Pattern30 Enr Bu k-calc 3t k-calc k-calc k-calc Pattern A 3.6 5 0.9034 0.8987 0.8971 0.8978 -0.0070 0.8465 0.8338 Pattern B 2.6 5 0.8457 0.8468 0.8562 0.8523 0.0078 0.7845 0.7915 Pattern C 2.6 15 0.8290 0.8301 0.8443 0.8384 0.0105 0.7733 0.7837 PatternD 2.6 20 0.8405 0.8431 0.8565 0.8509 0.0115 0.7856 0.7964 Pattern A 5.0 15 0.9098 0.9047 0.9027 0.9035 -0.0077 0.8567 0.8431 Pattern B 5.0 35 0.8619 0.8623 0.8759 0.8702 0.0096 0.8155 0.8171 Pattern C 5.0 45 0.8482 0.8486 0.8643 0.8577 0.0109 0.8023 0.8076 Pattern D 5.0 55 0.8438 0.8454 0.8624 0.8553 0.0126 0.8013 0.8072 29 The delta k-calc is calculated by the results of 1.00g/cc, 300K and 0.968g/cc, 358.15K.

30 Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 235 5.0 wt% U, the enrichments and burnups presented in the table are only for spent fuel.

31 The standard deviation of k-calc is around 0.0005 for all calculations.

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Table 4.6.6 Reactivity Effect of Eccentric Positioning, Pattern A Pattern A A (500 ppm)

Enrichment, wt% 5.0/3.6 5.0/5.0 5.0/3.6 5.0/5.0 Burnup, GWd/MTU 0/5 0/15 0/5 0/15 Reference (centered assembly) 0.9593 - 0.9628 0.9034 - 0.9098 FA - Eccentric In 0.9561 -0.0046 0.9592 -0.0050 0.8991 -0.0057 0.9061 -0.0051 FA - Eccentric Corner 0.9529 -0.0078 0.9580 -0.0062 0.8965 -0.0083 0.9035 -0.0077 8x8 Rack - Eccentric In 0.9563 -0.0044 0.9597 -0.0045 0.8994 -0.0054 0.9060 -0.0052 10x10 Rack - Eccentric In 0.9557 -0.0050 0.9585 -0.0057 0.8982 -0.0066 0.9051 -0.0061 Project No. 2064 Report No. HI-2114996 Page 88 Shfaded Areas Denote Holtec International Proprietary Information

Table 4.6.7 Reactivity Effect of Eccentric Positioning, Pattern B Pattern B B (500 ppm)

Enrichment, wt% 2.6 5.0 2.6 5.0 Burnup, GWd/MTU 05 35 05 35 Reference (centered assembly) 0.9457 - 0.9441 - 0.8468 - 0.8622 -

FA - Eccentric In 0.9519 0.0073 0.9509 0.0081 0.8538 0.0084 0.8690 0.0082 FA - Eccentric Corner 0.9482 0.0036 0.9491 0.0064 0.8466 0.0000 0.8627 0.0000 8x8 Rack - Eccentric In 0.9577 0.0131 0.9564 0.0136 0.8579 0.0124 0.8736 0.0127 10xlORack-EccentricIn 0.9578 0.0132 0.9562 0.0134 0.8575 0.0120 0.8743 0.0134 Project No. 2064 Report No. HI-2114996 Page 89 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.8 Reactivity Effect of Eccentric Positioning, Pattern C Pattern C C (500 ppm)

Enrichment, wt% 2.6 5.0 2.6 5.0 Burnup, GWd/MTU 15 45 15 45 Reference (centered assembly) 0.9296 - 0.9311 - 0.8312 - 0.8481 FA - Eccentric In 0.9367 0.0082 0.9390 0.0092 0.8385 0.0087 0.8569 0.0102 FA - Eccentric Corner 0.9330 0.0045 0.9356 0.0056 0.8307 0.0000 0.8496 0.0028 8xWRack-EccentricIn 0.9403 0.0118 0.9450 0.0150 0.8424 0.0126 0.8603 0.0136 10x10 Rack-EEccentric In 0.9410 0.0125 0.9442 0.0144 0.8428 0.0130 0.8612 0.0145 Table 4.6.9 Reactivity Effect of Eccentric Positioning, Pattern D Pattern __ D ... D D (500 ppm)

Enrichment, wt% 2.6 5.0 2.6 5.0 Burnup, GWd/MTU 20 55 20 55 Reference (centered assembly) 0.9433 - 0.9305 - 0.8430 - 0.8451 -

FA - Eccentric In 0.9517 0.0095 0.9381 0.0089 0.8501 0.0082 0.8537 0.0097 FA - Eccentric Corner 0.9456 0.0034 0.9345 0.0051 0.8418 -0.0023 0.8461 0.0000 8x8Rack-Eccentricln 0.9569 0.0147 0.9430 0.0136 0.8544 0.0127 0.8570 0.0132 10xlO Rack-Eccentric In 0.9569 0.0147 0.9436 0.0142 0.8543 0.0124 0.8570 0.0130 Project No. 2064 Report No. HI-2114996 Page 90 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.10 Reactivity Effect of the Core Operating Parameters, Pattern A Pattern A Enrichment, wt% 5.0/3.6 5.0/5.0 Burnup, GWd/MTU 0/5 0/15 K-calc Delta k-calc 95/95 unc K-calc Delta k-calc 95/95 unc Reference 0.9593 - - 0.9628 - -

Fuel Temp Decreased by 300F 0.9578 -0.0015 0.0014 0.9618 -0.0010 0.0014 Fuel Temp Increased by 300F 0.9607 0.0014 0.0016 0.9621 -0.0007 0.0014 Mod Temp Decreased by 25F 0.9599 0.0006 0.0014 0.9619 -0.0009 0.0014 Mod Temp Increased by 25F 0.9591 -0.0002 0.0014 0.9635 0.0007 0.0014 Soluble Boron Decreased by 300 ppm 0.9599 0.0006 0.0014 0.9629 0.0001 0.0014 Soluble Boron Increased by 300 ppm 0.9599 0.0006 0.0014 0.9622 -0.0006 0.0014 Specific Power Decreased by 5MW/MTU 0.9591 -0.0002 0.0014 0.9627 -0.0001 0.0016 Specific Power Increased by 5MW/MTU 0.9593 0.0000 0.0014 0.9628 0.0000 0.0014 Project No. 2064 Report No. HI-2114996 Page 91 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.11 Reactivity Effect of the Core Operating Parameters, Pattern B Pattern B Enrichment, wt% 2.6 5.0 Burnup, GWd/MTU 5 35 K-calc Delta k-calc 95/95 unc K-calc Delta k-calc 95/95 unc Reference 0.9457 - .. 0.9441 - -

Fuel Temp Decreased by 300F 0.9450 -0.0007 0.0013 0.9423 -0.0018 0.0013 Fuel Temp Increased by 300F 0.9460 0.0003 0.0013 0.9451 0.0010 0.0013 Mod Temp Decreased by 25F 0.9441 -0.0016 0.0013 0.9404 -0.0037 0.0013 Mod Temp Increased by 25F 0.9493 0.0036 0.0013 0.9536 0.0095 0.0013 Soluble Boron Decreased by 300 ppm 0.9441 -0.0016 0.0013 0.9428 -0.0013 0.0013 Soluble Boron Increased by 300 ppm 0.9453 -0.0004 0.0013 0.9454 0.0013 0.0014 Specific Power Decreased by 5MWiMTU 0.9455 - -0.0002 0.0013 0.9449 0.0008 0.0013 Specific Power Increased by 5MW/MTU 0.9449 -0.0008 0.0013 0.9442 0.0001 0.0014 Project No. 2064 Report No. HI-2114996 Page 92 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.12 Reactivity Effect of the Core Operating Parameters, Pattern C Pattern C Enrichment, wt% .2.6 5.0 Burnup, GWd/MTU 15 45 K-calc Delta k-calc 95/95 unc K-calc Delta k-calc 95/95 unc Reference 0.9296 - - 0.9311 -

Fuel Temp Decreased by 300F 0.9269 -0.0027 0.0013 0.9285 -0.0026 0.0013 Fuel Temp Increased by 300F 0.9289 -0.0007 0.0013 0.9332 0.0021 0.0014 Mod Temp Decreased by 25F 0.9253 -0.0043 0.0013 0.9248 -0.0063 0.0013 Mod Temp Increased by 25F 0.9411 0.0115 0.0013 0.9487 0.0176 0.0013 Soluble Boron Decreased by 300 ppm 0.9275 -0.0021 0.0013 0.9296 -0.0015 0.0013 Soluble Boron Increased by 300 ppm 0.9298 0.0002 0.0014 0.9333 0.0022 0.0013 Specific Power Decreased by 5MW/MTU 0.9289 -0.0007 0.0013 0.9317 0.0006 0.0014 Specific Power Increased by 5MW/MTU 0.9285 -0.0011 0.0013 0.9305 -0.0006 0.0013 Project No. 2064 Report No. HI-2114996 Page 93 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.l3 Reactivity Effect of the Core Opeirating Parameters, Pattern D Pattern D Enrichment, wt% 2.6 5.0 Burnup, GWd/MTU 20 55 K-calc Delta k-calc 95/95 unc K-calc Delta k-calc 95/95 unc Reference 0.9433 - - 0.9305 - -

Fuel Temp Decreased by 300F 0.9414 -0.0019 0.0013 0.9271 -0.0034 0.0013 Fuel Temp Increased by 300F 0.9445 0.0012 0.0013 0.9326 0.0021 0.0013 Mod Temp Decreased by 25F 0.9381 -0.0052 0.0013 0.9214 -0.0091 0.0013 Mod Temp Increased by 25F 0.9571 0.0138 0.0013 0.9532 0.0227 0.0013 Soluble Boron Decreased by_300 ppm 0.9409 -0.0024 0.0013 0.9285 -0.0020 0.0013 Soluble Boron Increased by 300 ppm 0.9456 0.0023 0.0013 0.9323 0.0018 0.0013 Specific Power Decreased by 5MW/MTU 0.9437 0.0004 0.0013 0.9300 -0.0005 0.0013 Specific Power Increased by 5MW/MTU 0.9430 -0.0003 0.0013 0.9287 -0.0018 0.0013 Project No. 2064 Report No. HI-2114996 Page 94 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.14 Reactivity Effect of the Cooling Time Uncertainty for Pattern B Enrichment, wt% 1.5 2.6- 3.6 4.5 5 Burnup (GWD/MTU) 5 15 30 40 45 Cooling Time Case K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc (years)

Reference 0.8085 - 0.9441 - 0.9354 - 0.9380 - 0.9375 -

Cooling Time Bounding 0.8077 0.0000 0.9434 0.0000 0.9366 0.0023 0.9369 0.0000 0.9377 0.0000 Reference 0.8053 - 0.9437 - 0.9336 - 0.9349 - 0.9348 Cooling Time Bounding 0.8068 0.0026 0.9429 0.0000 0.9327 0.0000 0.9343 0.0000 0.9335 -0.0024 Reference 0.8046 - 0.9432 - 0.9297 - 0.9295 - 0.9280 -

Cooling Time Bounding 0.8042 0.0000 0.9420 -0.0023 0.9305 0.0000 0.9296 0.0000 0.9276 0.0000 Reference 0.8040 - 0.9417 - 0.9264 - 0.9264 - 0.9235 -

Cooling Time Bounding 0.8028 -0.0022 0.9414 0.0000 0.9269 0.0000 0.9258 0.0000 0.9218 -0.0028 Reference 0.8025 - 0.9411 - 0.9253 - 0.9226 - 0.9203 -

Cooling Time Bounding 0.8013 -0.0023 0.9415 0.0000 - 0.9264 0.0000 0.9234 0.0000 0.9197 0.0000 Maximum Delta k-calc 0.0026 0.0000 0.0000 0.0000 0.0000 Average Value of Delta k-calc -0.0004 -0.0005 0.0000

' 0.0000 -0.0011 Project No. 2064 Report No. HI-2 114996 Page 95 Shaded Areas Denote Holtec International Pioprietary Information

Table 4.6.15 Reactivity Effect of the Cooling Time Uncertainty for Pattern C Enrichment, wt% 1.5 2.6' 3.6 4.5 5 Burnup (GWD/MTU) 5 15.. 30 40 45 Cooling -

Time Case K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc (years) .

2.5 Reference 0.8485 - 0.9252 -- ... 0.9230 - 0.9232 - 0.9220 -

Cooling Time Bounding 0.8489 0.0000 0.9258 0.0000 0.9237 0.0000 0.9243 0.0000 0.9217 0.0000 5 Reference 0.8476 - 0.9216 - 0.9192 - 0.9189 - 0.9149 -

Cooling Time Bounding 0.8473 0.0000 0.9225 0.0000 0.9194 0.0000 0.9182 0.0000 0.9151 0.0000 Reference 0.8446 - 0.9185 - 0.9118 - 0.9081 - 0.9036 -

Cooling Time Bounding 0.8450 0.0000 0.9192 0.0000 0.9123 0.0000 0.9083 0.0000 0.9046 0.0000 Reference 0.8430 - 0.9168 -. 0.9063 - 0.9005 - 0.8975 -

15 Cooling Time Bounding 0.8434 0.0000 0.9161 0.0000 0.9073 0.0000 0.9020 0.0026 0.8976 0.0000 20 Reference 0.8423 - 0.9150 - 0.9032 - 0.8962 - 0.8915 -

Cooling Time Bounding 0.8424 0.0000 0.9136 -0.0025 0.9032 0.0000 0.8961 0.0000 0.8910 0.0000 Maximum Delta k-calc 0.0000 0.0000 0.0000 0.0026 0.0000 Average Value of Delta k-calc 0.0000 -0.0005 - 0.0000 0.0005 0.0000 Project No. 2064 Report No. HI-2114996 Page 96 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.16 Reactivity Effect of the Cooling Time Uncertainty for Pattern D Enrichment, wt% 1.5 2.6 3.6 4.5 5 Burnup (GWD/MTU) 5 20 35 45 55 Cooling Time Case K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc (years)

Reference 0.8894 - 0.9377 - 0.9326 - 0.9300 - 0.9176 -

Cooling Time Bounding 0.8888 0.0000 0.9371 -0.0000. 0.9339 0.0024 0.9302 0.0000 0.9183 0.0000 Reference 0.8877 - 0.9336 . .0.9272 - 0.9228 - 0.9100 -

Cooling Time Bounding 0.8877 0.0000 0.9325 0.0000 0.9258 -0.0025 0.9222 0.0000 0.9103 0.0000 0.8855 - 0.9280 - 0.9168 - 0.9098 - 0.8966 -

10 Reference Cooling Time Bounding 0.8857 0.0000 0.9272 0.0000 0.9161 0.0000 0.9107 0.0000 0.8964 0.0000 0.8842 - 0.9234 - 0.9089 - 0.9008 - 0.8862 -

15 Reference CoolingTime Bounding 0.8839 0.0000 0.9233 0.0000 0.9102 0.0024 0.9012 0.0000 0.8865 0.0000 20 Reference 0.8825 - 0.9202 - 0.9042 - 0.8944 - 0.8796 -

Cooling Time Bounding 0.8826 0.0000 0.92 5 0.0024 0.9043 0.0000 0.8944 0.0000 0.8790 0.0000 Maximum Delta k-calc 0.0000 0.0024 0.0024 0.0000 0.0000 Average Value of Delta k-calc 0.0000 0.0005 0.0005 0.0000 0.0000 Project No. 2064 Report No. HI-2114996 Page 97 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.17 Reactivity Effect of the IFBA Rods for Non-blanketed Fuel, Pattern A Pattern A A (500 ppm)

Cooling Time, years 0 0 Enrichment, wt% 3.4 .A4 3.4 4.4 Burnup, GWd/MTU 1 10 1 10 K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc 1 Reference (Design Basis) 0.9645 - 0.9620 - 0.9081 - 0.9098 -

2 Bias _ '_

3 64 IFBAs 0.9641 0.0000 0.9620 0.0000 0.9064 -0.0031 0.9083 0.0000 4 80 IFBAs 0.9649 0.0000 0.9615 0.0000 0.9066 -0.0029 0.9084 0.0000 5 104 IFBAs 0.9646 0.0000 0.9628 0.0000 0.9066 -0.0029 0.9084 0.0000 6 128 IFBAs 0.9640 0.0000 0.9622 0.0000 0.9075 0.0000 0.9082 -0.0030 7 156 IFBAs 0.9653 0.0000 0.9630 0.0000 0.9063 -0.0032 0.9090 0.0000 8 Maximum Delta-k 0.0000 0.0000 0.0000 0.0000 Project No. 2064 Report No. HI-2114996 Page 98 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.18 Reactivity Effect of the IFBA Rods for Non-blanketed Fuel, Pattern B Pattern B B (500 ppm)

Cooling Time, years 0 -20.. 0 20 Enrichment, wt% 3.4 4.4 . 4.4 3.4 4.4 4.4 Burnup, GWd/MTU 15 30 -. 30 15 30 30 Delta k- Delta k-SK-calcK-cal

- - Deltd k- 1 Delta k- Delta k- Delta k-Item calc caic .. ...... c calc K-calc calc K-calc calc K-calc calc Reference 0.917,7 -. 0.8589 0.8543 - 0.8334 -

(Design Basis) 0.9508 _0.9388 _ 0.8589 0.85 2 Bias 3 64 IFBAs 0.9515 0.0000 0.9389 0.0000 0.9170 0.0000 0.8595 0.0000 0.8541 0.0000 0.8335 0.0000 4 80 IFBAs 0.9508 0.0000 0.9376 0.0000 0.9182. 0.0000 0.8585 0.0000 0.8545 0.0000 0.8344 0.0000 5 104 IFBAs 0.9512 0.0000 0.9392 0.0000 0.9176 0.0000 0.8595 0.0000 0.8551 0.0000 0.8345 0.0000 6 128 IFBAs 0.9525 0.0030 0.9387 0.0000 0.9187 0.0000 0.8605 0.0029 0.8550 0.0000 0.8343 0.0000 7 156 IFBAs 0.9513 0.0000 0.9403 0.0028 0.9204 0.0038 0.8608 0.0032 0.8557 0.0027 0.8343 0.0000 8 Maximum 0.0030 0.0028 0.0038 0.0032 0.0027 0.0000 Delta-k I II Project No. 2064 Report No. HI-2114996 Page 99 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.19 Reactivity Effect of the IFBA Rods for Non-blanketed Fuel, Pattern C Pattern C C (500 ppm)

Cooling Time, 0 20 0 20 years Enrichment, wt% 3.4 4.4 4.4 3.4 4.4 4.4 Burnup, GurnMpU GWd/MTU 25 40 40-- 25 40 40 K-calc Delta K-calc Delta K-calc Delta Delta K-calc Delta Delta K -calc calc..ca k-ca K-calc k-calc K-calc k-calc K-calc k-calc Reference 1 (Design 0.9418 - 0.9266 - 0.8898 - 0.8505 0.8413 0.8054 -

Basis) _ . ...

2 Bias 3 64 IFBAs 0.9422 0.0000 0.9268 0.0000 0.8900 0.0000 0.8509 0.0000 0.8415 0.0000 0.8067 0.0024 4 80 IFBAs 0.9440 0.0033 0.9276 0.0000 0.8909 0.0000 0.8507 0.0000 0.8416 0.0000 0.8082 0.0041 5 104 IFBAs 0.9444 0.0037 0.9277 0.0000 0.8918 0.0031 0.8511 0.0000 0.8428 0.0028 0.8079 0.0036 6 128 IFBAs 0.9450 0.0043 0.9280 0.0025 0.8902 0.0000 0.8527 0.0035 0.8431 0.0029 0.8086 0.0043 7 156 IFBAs 0.9459 0.0052 0.9290 0.0035 0.8913 0.0026 0.8527 0.0033 0.843 1 0.0031 0.8085 0.0042 8 Maximum 0.0052 0.0035 0.0031 0.0035 0.0031 0.0043 8 Delta-k 0.0035 0.0031 0.0035 T.0052 .0.0031 0.0043 Project No. 2064 Report No. HI-2114996 Page 100 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.20 Reactivity Effect of the IFBA Rods for Non-blanketed Fuel, Pattern D Pattern D D (500 ppm)

Cooling Time, 0 20 0 20 years Enrichment, wt% 3.4 4.4 4.4 3.4 4.4 4.4 Burnup, 35 45 45 35 45 45 GWd/MTU Delta Delta Delta Delta Delta Delta Item Reernc k-calc K-calc k-calc K-calc k-calc k-calc K-calc k-calc K-calc k-calc Reference I (Design 0.9281 0.9356 0.8887 .- 0.8355 0.8481 - 0.8034 -

B asis) .... ........... __ __ __.

2 Bias ,

3 64 IFBAs 0.9290 0.0000 0.9355 0.0000 0.8885 0.0000 0.8373 0.0029 0.8489 0.0000 0.8040 0.0000 4 80 IFBAs 0.9302 0.0032 0.9361 0.0000 0.8893' 0.0000 0.8362 0.0000 0.8489 0.0000 0.8042 0.0000 5 104 IFBAs 0.9306 0.0036 0.9363 0.0000 0.8897 0.0000 0.8367 0.0023 0.8504 0.0036 0.8047 0.0024 6 128 IFBAs 0.9301 0.0031 0.9387 0.0042 0.8901 0.0025 0.8383 0.0039 0.8503 0.0033 0.8054 0.0031 7 156 IFBAs 0.9311 0.0041 0.9382 0.0037 0.8898 0.0000 0.8387 0.0043 0.8500 0.0030 0.8053 0.0030 8 Maximum 0.0041 0.0042 0.0025-- 0.0043 0.0036 0.0031

___ Delta-k__ _ _ _ _ _ _ __ _ _ _ _ _ _ __ _ _ _ _ _ _ __ _ _ _ _ _ _ __ _ _ _ _ _ _

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Table 4.6.21 Reactivity Effect of BPRA Rods for Non-blanketed Fuel Boron Concentration 0 ppm 500 ppm Cooling Time, years 0 20 0 20 Enrichment, wt% 2.4 3.1 3.1 2.4 3.1 3.1 Burnup, GWd/MTU 15 30 30 15 30 30 Delta k-K-acK-calc Delta k- K-calc Delta k- K-calc Delta k- K-calc Delta c-cack- Delta k-Pattern32 calc calc calc calc calc calc Reference 0.9571 - - 0.8787 - 0.8993 -

A (Design 0.9369 -

Basis) 24 BPRAs 0.9367 0.0000 0.9571 0.0000 " . ' 0.8781 0.0000 0.8988 0.0000 -

Reference (Design 0.9259 - 0.9290 - 0.9173 - 0.8266 - 0.8360 - 0.8260 Basis) . ....

24BPRAs 0.9317 0.0069 0.9369 0.0090 0.9220. 0.0058 0.8331 0.0076 0.8462 0.0115 0.8330 0.0081 Reference (Design 0.9354 - 0.9457 - 0.9275 0.8361 - 0.8509 - 0.8344 -

C B asis) I ..... _

24BPRAs 0.9468 0.0125 0.9542 0.0096 0.9350 0.0086 0.8471 0.0121 0.8609 0.0113 0.8418 0.0087 Reference (Design 0.9249 - 0.9322 - 0.8988 - 0.8232 - 0.8374 - 0.8058 -

D Basis) I _

24BPRAs 0.9381 0.0143 0.9449 0.0138 0.9091 0.0114 0.8390 0.0169 0.8516 0.0153 0.8177 0.0130 235 32 Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 5.0 wt% U, the enrichments and burnups presented in the table are only for spent fuel.

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Table 4.622

• Reactivity Effect of IFBA Rods for Enriched Blanketed Fuel, Pattern A Pattern A A (500 ppm)

Cooling Time, years 0 0 Enrichment, wt% 5.0/3.6 5.0/5.0 5.0/3.6 5.0/5.0 Burnup, GWdiMTU 0/5 0/15 0/5 0/15 Delta Delta Delta Delta k-Itmk-calc ItmK-calc K-ac k-calc K-calc k-calc K-calc cac cl 1

Reference:

Enriched-blanketed Fuel w/o WFBA 0.9579 - 0.9597 - 0.9029 - 0.9075 -

2 Bias 3 16 TFBAs 0.9591 0.0000 0.9605 0.0000 0.9033 0.0000 0.9078 0.0000 4 32 IFBAs 0.9582 0.0000 0.9604 0.0000 0.9030 0.0000 0.9084 0.0000 5 48 IFBAs 0.9597 0.0034 0.9608 0.0000 0.9028 0.0000 0.9076 0.0000 6 64 IFBAs 0.9585 0.0000 0.9606 0.0000 0.9037 0.0000 0.9096 0.0035 7 80 IFBAs 0ý9586.', 10.000 0.9612 0.0029 0.9036 0.0000 0.9093 0.0032 8 104 IFBAs 0.9598,-- ,'0.0033 0.9599 0.0000 0.9046 0.0031 0.9088 0.0000 9 128 IFBAs 0.9588 0.0000 0.9603 0.0000 0.9040 0.0000 0.9093 0.0032 10 156 IFBAs 0.95971 10.0032 0.9602 0.0000 0.9050 0.0035 0.9091 0.0030 11 Maximum Delta-k 0.0034 0.0029 0.0035 0.0035 12 Design Basis Case: Non-blanketed Fuel w/o IFBA 0.9593 0.0000 0.9628 0.0045 0.9034 0.0000 0.9098 0.0037 Project No. 2064 Report No. HI-2114996 Page 103 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.23 Reactivity Effect of IFBA Rods for Enriched Blanketed Fuel, Pattern B Pattern B B (500 ppm)

Cooling Time, years 0 20 0 20 Enrichment, wt% 3.6 5 5 3.6 5 5 Burnup, GWd/MTU 20 35 35 20 35 35 calc Delta Delta Delta Delta Delta Iteca l k-calc K-calc

-cacck-ackal Delt K-calc kkal K-calc K-calc klk-caac k-calc k-calc

Reference:

1 Enriched-blanketed 0.9251 - 0.9280 - 0.8925 - 0.8374 - 0.8478 - 0.8142 -

Fuel w/o IFBA 2 Bias 3 16 IFBAs 0.9269 0.0029 0.9274 0 0.8926 0 0.8398 0.0035 0.8473 0 0.8146 0 4 32 IFBAs 0.9269 0.0031 0.9295 0.0028 0.8930 0 0.8391 0.0028 0.8480 0 0.8148 0 5 48 IFBAs 0.9278 0.0038 0.9298 0.0031 0.8938 7.0.0024 0.8407 0.0044 0.8491 0.0024 0.8146 0 6 64 IFBAs 0.9283 0.0043 0.9294 0.0027 0.8934"' 0 0.8410 0.0047 0.8483 0 0.8146 0 7 80IFBAs 0.9289 0.0049 0.9294 0.0027 0.8935 0 0.8423 0.0062 0.8491 0.0024 0.8160 0.0029 8 104 IFBAs 0.9291 0.0051 0.9313 0.0046' 0'.8945 0.0031 0.8426 0.0065 0.8499 0.0032 0.8166 0.0035 9 128 IFBAs 0.9292 0.0052 0.9322 0.0055' 0.8947t3 'ý0.0035 0.8428 0.0065 0.8508 0.0041 0.8164 0.0033 10 156 IFBAs 0.9319 0.0079 0.9318 0.0051 0.8953 0.0039 0.8445 0.0082 0.8511 0.0046 0.8171 0.0040 11 Maximum Delta-k 0.0079 0.0055 - 0.0039 0.0082 0.0046 0.0040 Design Basis Case:

12 Non-blanketed Fuel 0.9393 0.0155 0.9441 0.0175 0.9203 0.0291 0.8509 0.0148 0.8614 0.0147 0.8386 0.0258 w/o IFBA Project No. 2064 Report No. HI-2114996 Page 104 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.24 Reactivity Effect of IFBA Rods for Enriched Blanketed Fuel, Pattern C Pattern C C (500 ppm)

Cooling Time, years 0 20 0 20 Enrichment, wt% 3.6 5 5 3.6 5 5 Burnup, GWdiMTU 30 45 45 30 45 45 Delta Delta Delta Delta Delta Delta k-K-calc kl K-calc K-calc Dl K-calc Dela K-calc caic Item K-calc Deca k-calc k-calc k-calc k-calc k-calc calc

Reference:

1 Enriched-blanketed 0.9137 - 0.9183 0.8677 0.8243 - 0.8375 - 0.7899 Fuel w/o IFBA _

2 Bias 3 16 IFBAs 0.9140 0 0.9182 0 0.8677 0 0.8253 0 0.8373 0 0.7896 0 4 32 IFBAs 0.9143 0 0.9191 0 0.8691 0.0025 0.8261 0.0031 0.8378 0 0.7887 -0.0023 5 48 IFBAs 0.9145 0 0.9203 0.0031 0.8686, 0 0.8275 0.0045 0.8377 0 0.7906 0 6 64 IFBAs 0.9159 0.0035 0.9206 0.0034 0.8690 0.'0024 0.8282 0.0052 0.8391 0.0027 0.7895 0 7 80 IFBAs 0.9155 0.0031 0.9207 0.0035 0.8695 -0.0029 0.8274 0.0044 0.8395 0.0031 0.7906 0 8 104 IFBAs 0.9165 0.0041 0.9214 0.0042fi'0.87Y.'00'i:*0.034 0.8298 0.0068 0.8382 0 0.7914 0.0026-9 128 IFBAs 0.9177 0.0053 0.9227 0.0055- 0'87.,*-. :0:0046 0.8304 0.0075 0.8398 0.0034 0.7921 0.0033 10 156 IFBAs 0.9189 0.0065 0.9231 0.0059 0.8718 'o0.0,052 0.8313 0.0083 0.8415 0.0051 0.7929 0.0041 11 Maximum Delta-k 0.0065 0.0059 0.0Q52 0.0083 0.0051 0.0041 Design Basis Case:

12 Non-blanketed 0.9311 0.0187 0.9311 0.0139 0.8915 0.0251 0.8407 0.0177 0.8475 0.0111 0.8105 0.0217 Fuel w/o IFBA I Project No. 2064 Report No. HI-2114996 Page 105 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.25 Reactivity Effect of IFBA Rods for Enriched Blanketed Fuel, Pattern D Pattern D D (500 ppm)

Cooling Time, years 0 20 0 20 Enrichment, wt% 3.6 5 '5 3.6 5 5 Burnup, GWd/MTU 35 55 55 35 55 55 Deltac k-cal K-calc Delta Dla K-calc Delta K-calc Delta K-calc Delta k-calc Delta Item K-cal k-calc k-calc k--calc kcalc K-calc k-calc

Reference:

1 Enriched-blanketed 0.9244 - 0.9129 - 0.8480 - 0.8351 - 0.8301 - 0.7697 -

Fuel w/o IFBA 2 Bias 3 16 IFBAs 0.9257 0.0024 0.9126 0 0.8481 0 0.8346 0 0.8290 0 0.7689 0 4 32 lFBAs 0.9252 0 0.9120 0 0.8477 0 0.8356 0 0.8291 0 0.7697 0 5 48 IFBAs 0.9272 0.0039 0.9136 0 0.8479 0 0.8362 0 0.8305 0 0.7705 0 6 64 LFBAs 0.9277 0.0044 0.9129 0 0.8481 0 0.8364 0.0024 0.8303 0 0.7708 0 7 80 IFBAs 0.9280 0.0047 0.9131 0 0.8495 ,.0.0026 0.8377 0.0037 0.8313 0.0023 0.7706 0 8 104 IFBAs 0.9291 0.0058 0.9146 0.0028 0U8497:-,0.0028 0.8377 0.0037 0.8304 0 0.7718 0.0032 9 128 IFBAs 0.9288 0.0055 0.9158 0.0040 0.8504" 06.0035 0.8382 0.0042 0.8319 0.0029 0.7718 0.0032 10 156 IFBAs 0,9304 0.0071 0.9158 0.0040 0":85.,2 ,:-020043 0.8405 0.0065 0.8332 0.0042 0.7735 0.0049 11 Maximum Delta-k 0.0071 0.0040 .. 0 O43 0.0065 0.0042 0.0049 Design Basis Case: . .. 9 ..

12 Non-blanketed Fuel 0.9413 0.0180 0.9305 0.0187 0.8796. %:0-0327 0.8504 0.0164 0.8451 0.0161 0.7980 0.0296 w/o IFBA I _ I Project No. 2064 Report No. HI-2114996 Page 106 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.26 Modeling of the BORAFLEX TM Material Enrichment, wt% 2.6 5 Soluble Boron Content, ppm Bu 0 500 Bu 0 500 Pattern B K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc Filled Absorber Gap-Polymer 5 0.9457 - 0.8468 - 35 0.9441 - 0.8622 -

Filled Absorber Gap -Water 5 0.9415 -0.0053 0.8400 -0.0079 35 0.9402 -0.0053 0.8564 -0.0071 Pattern C K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc Filled Absorber Gap-Polymer 15 0.9296 - 0.8312 - 45 0.9311 - 0.8481 -

Filled Absorber Gap -Water 15 0.9246 -0.0061 0.8247 -0.0076 45 0.9276 -0.0046 0.8417 -0.0077 Pattern D K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc Filled Absorber Gap -Polymer 20 0.9433 - 0.8430 - 55 .93 05 - 0.8451 -

Filled Absorber Gap -Water 20 0.9377 -0.0067 0.8361 -0.0080 55 0.9252 -0.0064 0.8385. -0.0079 Project No. 2064 Report No. HI-2114996 Page 107 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.27 Reactivity Effect of the Grid Spacers Enrichment, wt% 5.0/3.6 5.0/5.0 Soluble Boron Content, Bu 0 500 Bu 0 500 ppm ___

Pattern A K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc Reference 0/5 0.9593 - 0.9034 - 0/15 0.9628 - 0.9098 -

Grid Spacers 0/5 0.9563 -0.0044 0.9021 0 0/15 0.961 -0.0032 0.9088 0 Enrichment, wt% 2.6 5 Soluble Boron Content, Bu 0 -- 500 Bu 0 500 ppm Pattern B K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc Reference 5 0.9457 - 0.8468 - 35 0.9441 - 0.8622 -

Grid Spacers 5 0.9443 -0.0025 0.8453 -0.0028 35 0.9437 0 0.8610 0 Pattern C K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc Reference 15 0.9296 - 0.8312 - 45 0.9311] - 0.8481 -

Grid Spacers 15 0.9292 0 0.8306 0 45 0.9313] 0 0.8482 0 Pattern D K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc Reference 20 0.9433 - 0.8430 - 55 0.9305 - 0.8451 -

Grid Spacers 20 0.9422 0 0.8423 0 55 0.9286 -0.0030 0.8456 0 Project No. 2064 Report No. HI-2114996 Page 108 Shaded Areas Denote Holtec International Proprietary Information

Table 4'6.28 Reactivity Effect of the Fuel Creep and Growth Enrichment, wt% 5.0/3.6 5.0/5.0 Soluble Boron Content, Bu 0 500 Bu 0 500 ppm Pattern A K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc Reference 0/5 0.9593 - 0.9034 - 0/15 0.9628 - 0.9098 -

Fuel Creep and Growth 0/5 0.9618 0.0039 0.9047 0 0/15 0.964 0 0.9111 0 Enrichment, wt% 2.6 5 Soluble Boron Content, Bu 0 500 Bu 0 500 ppm Pattern B K-calc Delta k-calc K-calc. Delta-k-calc K-calc Delta k-calc K-calc Delta k-calc Reference 5 0.9457 - 0.8468 - 35 0.9441 - 0.8622 -

Fuel Creep and Growth 5 0.9488 0.0042 0.8482 0.0025 35 0.9468 0.0040 0.8641 0.0032 Pattern C K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc Reference 15 0.9296 - 0.8312 - 45 0.9311 - 0.8481 -

Fuel Creep and Growth 15 0.9313 0.0028 0.8321 0 45 0.9334 0.0034 0.8486 0 Pattern D K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc K-calc Delta k-calc Reference 20 0.9433 - 0.8430 - 55 0.9305 - 0.8451 -

Fuel Creep and Growth 20 0.9449 0.0027 0.8438 0 55 0.9322 0.0028 0.8462 0 Project No. 2064 Report No. HI-2114996 Page 109 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.29 Reactivity Effect of the Fuel Tolerances, Pattern A Pattern A A (500 ppm)

Enrichment, wt% 5.0/3.6 . 5.0/5.0 5.0/3.6 5.0/5.0 Burnup, GWdiMTU 0/5 ....... 0/15 0/5 0/15 Itm RfrneK-calc T Delta k-K-calc Delta k-I K-calc Delta k-1 K-calc Delta k-cl calc Item calc calc calc I Reference (Design Basis) 0.9593 - 0.9628 - 0.9034 - 0.9098 -

2 Fuel Uncertainties:

3 Fuel - Enrichment Increased 0.9612 0.0033 0.9645 0.0031 0.9059 0.0039 0.9125 0.0041 4 Fuel - Density Increased 0.9611 0.0032 0.9638 0.0000 0.9063 0.0043 0.9115 0.0031 5 Fuel - Pellet OD Increased 0.9594 0.0000 0.9635 0.0000 0.9046 0.0000 0.9106 0.0000 6 Fuel - Pellet OD Decreased 0.9582 0.0000 0.9619 0.0000 0.9032 0.0000 0.9099 0.0000 7 Fuel - Rod Pitch Increased 0.9620 0.Q041 0.9641 0.0000 0.9049 0.0029 0.9117 0.0033 8 Fuel - Rod Pitch Decreased 0.9571 -0.0036 0.9618 0.0000 0.9021 0.0000 0.9078 -0.0034 9 Fuel - Clad OD Increased 0.9572 -0.0035 0.9603 -0.0039 0.9036 0.0000 0.9084 0.0000 10 Fuel - Clad OD Decreased 0.9612 0.0033 0.9651 0.0037 0.9047 0.0000 0.9107 0.0000 11 Fuel - Clad ID Increased 0.9600 0.0000 0.9625 0.0000 0.9034 0.0000 0.9100 0.0000 12 Fuel - Clad ID Decreased 0.9606 0.0000 0.9618 0.0000 0.9035 0.0000 0.9087 0.0000 13 Fuel - GT/IT OD Increased 0.9594 0.0000 0.9623 0.0000 0.9037 0.0000 0.9097 0.0000 14 Fuel - GT/IT OD Decreased 0.9604 0.0000 0.9633 0.0000 0.9038 0.0000 0.9096 0.0000 15 Fuel - GT/IT ID Increased 0.9590 0.0000 0.9628 0.0000 0.9030 0.0000 0.9099 0.0000 16 Fuel - GT/IT ID Decreased 0.9585 0.0000 0.9624 0.0000 0.9029 0.0000 0.9091 0.0000 17 Total, Fuel Uncertainties 0.0070 0.0048 0.0065 0.0061 18 Fuel - Dominating Parameters 0.9671 0.0092 0.9693 0.0079 0.9098 0.0078 0.9165 0.0081 19 Difference from Dominant Fuel Parameters 0.0022 0.0031 0.0013 0.0020 (18-17)

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Table 4.6.30 Reactivity Effect of the Fuel Tolerances, Pattern B Pattern - B B (500 ppm)

Enrichment, wt% 2.6 5 2.6 5 Burnup, GWd/MTU 5 35 5 35 Delta k- Delta k- Delta k- Delta k-Item Itm RfrneK-calc calc1 K-calc calc K-calc calc1 K-calc ci calc I Reference (Design Basis) 0.9457 - 0.9441 - 0.8468 - 0.8622 2 Fuel Uncertainties:

3 Fuel - Enrichment Increased 0.9504 0.0058 0.9472 0.0044 0.8514 0.0057 0.8636 0.0000 4 Fuel - Density Increased 0.9476 0.0030 0.9451 0.0000 0.8491 0.0034 0.8640 0.0031 5 Fuel - Pellet OD Increased 0.9463 0.0000 0.9434 0.0000 0.8469 0.0000 0;8625 0.0000 6 Fuel - Pellet OD Decreased 0.9445 -0.0023 0.9433 0.0000 0.8450 -0.0029 0.8612 0.0000 7 Fuel - Rod Pitch Increased 0.9481 0.0035 - 0.9454 0.0026 0.8485 0.0028 0.8637 0.0028 8 Fuel - Rod Pitch Decreased 0.9437 -0.0031 0.9400 -0.0054 0.8449 -0.0030 0.8601 -0.0035 9 Fuel - Clad OD Increased 0.9443 -0.0025 0.9441 0.0000 0.8459 0.0000 0.8611 0.0000 10 Fuel - Clad OD Decreased 0.9456 0.0000 0.9451 0.0000 0.8468 0.0000 0.8614 0.0000 11 Fuel - Clad ID Increased 0.9457 0.0000 0.9435 0.0000 0.8465 0.0000 0.8619 0.0000 12 Fuel - Clad ID Decreased 0.9455 0.0000 0.9431 0.0000 0.8466 0.0000 0.8612 0.0000 13 Fuel - GT/IT OD Increased 0.9440 -0.0028 0.9442 0.0000 0.8458 0.0000 0.8620 0.0000 14 Fuel - GT/IT OD Decreased 0.9462 0.0000 0.9429 0.0000 0.8469 0.0000 0.8612 0.0000 15 Fuel - GT/IT ID Increased 0.9464 0.0000 0.9442 0.0000 0.8477 0.0000 0.8629 0.0000 16 Fuel - GT/IT ID Decreased 0.9445 -0.0023 0.9442 0.0000 0.8465 0.0000 0.8617 0.0000 17 Total, Fuel Uncertainties 0.0075 0.0051 0.0073 0,0041 18 Fuel - Dominating Parameters 0.9557T 0.0111 0.9510I 0.0083 0.8560 1 0.0103 0.8689 0.0080 19 Difference from Dominant Fuel Parameters 0.0037 0.0032 0.0031 0.0038 (18-17) 1 -] 11 1 Project No. 2064 Report No. HI-21i 4909 Page 11 Shaded Areas Denote Holtec International- Proprietary Information

Table 4.6.31 Reactivity Effect of the Fuel Tolerances, Pattern C 2 1Fuel Uncertainties:

3 Fuel - Enrichment Increased 0.9340 0.0055 0.9331 0.0031 0.8349 0.0050 0,8507 0.0040 4 Fuel - Density Increased 0.9306 0.0000 0.9333 0.0033 0.8337 0.0036 0.8494 0.0026 5 Fuel - Pellet OD Increased 0.9286 0.0000 0.9320 0.0000 0.8317 0.0000 0.8483 0.0000 6 Fuel - Pellet OD Decreased 0.9297 0.0000 0.9307 0.0000 0.8309 0.0000 0.8468 -0.0026 7 Fuel - Rod Pitch Increased 0.9308 0.0023 0.9331 0.0031 0.8328 0.0027 0.8499 0.0031 8 Fuel - Rod Pitch Decreased 0.9265 -010042 0.9293 -0.0029 0.8303 0.0000 0.8469 0.0000 9 Fuel - Clad OD Increased 0.9274 -0.0033- 0.9310 0.0000 0.8313 0.0000 0.8477 0.0000 10 Fuel - Clad OD Decreased 0.9301 0.0000 0.9314 0.0000 0.8311 0.0000 0.8487 0.0000 11 Fuel - Clad ID Increased 0.9288 0.0000 0.9302 0.0000 0.8311 0.0000 0,8488 0.0000 12 Fuel - Clad ID Decreased 0.9289 0.0000 0.9319 0.0000 0.8303 0.0000 0.8478 0.0000 13 Fuel - GT/IT OD Increased 0.9289 0.0000 0.9310 0.0000 0.8311 0.0000 0.8486 0.0000 14 Fuel - GT/LT OD Decreased 0.9296 0.0000 0.9309 0.0000 0.8308 0.0000 0.8477 0.0000 15 Fuel - GT/IT ID Increased 0.9286 0.0000 0.9319 0.0000 0.8312 0.0000 0.8480 0.0000 16 Fuel - GT/IT ID Decreased 0.9278 -0.0029 0.9296 -0.0026 0.8308 0.0000 0,8477 0.0000 17 Total, Fuel Uncertainties 0.0060 0.0055 0.0067 0.0057 18 Fuel - Dominating Parameters 0.9385 1 0.0100 0.9377 0.0077 0.8404 1 0.0103 0.8566 0.0099 19 Difference from Dominant Fuel Parameters 0.0040 0.0022 0.0036 0.0042 (18-17)

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Table 4.6.32 Reactivity Effect of the Fuel Tolerances, Pattern D Pattern D D (500 ppm)

Enrichment, wt% 2.6 5 2.6 5 Burnup, GWd/MTU 15 .. _... 50 15 50 K-clc Delta-k-.- ac Delta k- K- Delta k- K- Delta k-t.calc .K-calc alc calc calc Item I________

1 Reference (Design Basis) 0.9433 - -0.9305 - 0.8430 0.8451 2 Fuel Uncertainties.

3 Fuel - Enrichment Increased 0.9476 0.0054 0.9321 0.0027 0.8462 0.0045 0.8482 0.0042 4 Fuel - Density Increased 0.9452 0.0030 0.9314 0.0000 0.8458 0.0039 0.8473 0.0033 5 Fuel - Pellet OD Increased 0.9429 0.0000 0.9310 0.0000 0.8437 0.0000 0.8455 0.0000 6 Fuel - Pellet OD Decreased 0.9439 0.0000 0.9297 0.0000 0.8411 -0.0030 0.8439 -0.0023 7 Fuel - Rod Pitch Increased 0.9453 0.0031 0.9326 0.0034 0.8433 0.0000 0.8468 0.0028 8 Fuel - Rod Pitch Decreased 0.9400 -0.0044 0.9272 -0.0044 0.8400 -0.0041 0.8426 -0.0036 9 Fuel - Clad OD Increased 0.9423 0.0000 0.9300 0.0000 0.8424 0.0000 0.8452 0.0000 10 Fuel - Clad OD Decreased 0.9432 0.0000 0.9296 0.0000 0.8425 0.0000 0.8454 0.0000 11 Fuel - Clad ID Increased 0.9424 0.0000 0.9300 0.0000 0.8422 0.0000 0.8442 0.0000 12 Fuel - Clad ID Decreased 0.9430 0.0000 0.9289 -0.0027 0.8429 0.0000 0.8444 0.0000 13 Fuel - GT/IT OD Increased 0.9426 0.0000 0.9291 -0.0025 0.8424 0.0000 0.8457 0.0000 14 Fuel - GT/IT OD Decreased 0.9432 0.0000 0.9290 -0.0026 0.8417 -0.0024 0.8450 0.0000 15 Fuel - GT/IT ID Increased 0.9433 0.0000 0.9307 0.0000 0.8429 0.0000 0.8450 0.0000 16 Fuel - GT/IT ID Decreased 0.9424 0.0000 0.9295 0.0000 0.8430 0.0000 0.8452 0.0000 17 Total, Fuel Uncertainties 0.0070 0.0050 0.0060 0.0061 18 Fuel - Dominating Parameters 0.9523 0.0101 0.9363 0.0069 0.8528 0.0109 0.8524 0.0086 Difference from Dominant Fuel Parameters 19 0.0032 0.0019 0.0050 0.0025 (18-17)

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Table 4.6.33 Reactivit, Effect of the Rack Tolerances, Pattern A Pattern A A (500 ppm)

Enrichment, wt% 5.0/3.6 5.0/5.0 5.0/3.6 5.0/5.0 Burnup, GWd/MTU 0/5 0/15 0/5 0/15 Item Itm RfrneK-calc. "

Delta k-cc calc*-,t. .. K-calc Delta k-cac calc K-calc I Delta k-cac calc K-calc Delta k-ci calc 1 Reference (Design Basis) 0.9593 . 0.9628 0.9034 - 0.9098 2 Rack Uncertainties.;

3 Rack - Flux Trap Increased 0.9525, -0.0082" 0.9549 -0.0093 0.8964 -0.0084 0.9027 -0.0085 4 Rack - Flux Trap Decreased 0.9681, 0.0102- .0.9695 0.0081 0.9112 0.0092 0.9174 0.0090 5 Rack - Cell ID Increased 0.9610 0.0031t 0.9636 0.0000 0.9039 0.0000 0.9091 0.0000 6 Rack - Cell ID Decreased 0.9588 0.0000 0.9624 0.0000 0.9028 0.0000 0.9102 0.0000 7 Rack - Wall Thickness Increased 0.9594 0.0000 0.9622 0.0000 0.9040 0.0000 0.9106 0.0000 8 Rack - Wall Thickness Decreased 0.9593 0.0000 0.9628 0.0000 0.9035 0.0000 0.9097 0.0000 9 Rack - Sheathing Increased . 0.9583 0.0000 0.9630 0.0000 0.9032 0.0000 0.9107 0.0000 10 Rack - Sheathing Decreased 0.9597 0.0000 0.9631 0.0000 0.9039 0.0000 0.9097 0.0000 11 Rack - Poison Gap Increased 0.9588 0.0000 0.9622 0.0000 0.9031 0.0000 0.9094 0.0000 12 Rack - PoisonGap Decreased 0.9608 0.0029 .0.9637 0.0000 0.9040 0.0000 0.9103 0.0000 13 Rack - BORALTM Thickness Increased 0.9584 0.0000 0.9601 -0.0043 0.9007 -0.0041 0.9074 -0.0038 14 Rack - BORALTM Thickness Decreased 0.9619 0.0040 0.9655 0.0041 0.9070 0.0050 0.9113 0.0029 15 Rack - BORALTM Width Increased 0.95850'. 0000- -0.9611 -0.0031 0.9025 0.0000 0.9096 0.0000 16 Rack - BORALTM Width Decreased 0.9602 0.0000 0.9636 0.0000 0.9044 0.0000 0.9106 0.0000 17 Rack - Localized Variation on B-10 0.9591 0.0000 A0.9628 0.0000 0.9046 0.0000 0.9101 0.0000 18 Total, Rack Uncertainties 0.0118 0.0091 0.0105 0.0095 19 Rack-Dominating Parameters 0.9714 0.0135 0.9744 0.0130 0.9138 0.0118 0.92021 0.0118 20 Difference from Dominant Rack Parameters 0.0017 0.0039 0.0013 0.0023 (19-18)

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Table 4.6.34 Reactivity Effect of the Rack Tolerances, Pattern B Pattern B B (500 ppm)

Enrichment, wt% 2.6 5 2.6 5 Burnup, GWd/MTU 5 35 5 35 Delta k- Delta k- Delta k- Delta k-Item ItmRfrneK-calc cac calc K-calc calc K-calc calc1 K-calc cl calc 1 Reference (Design Basis) 0.9457 - 0.9441 - 0.8468 - 0.8622 -

2 Rack Uncertainties:

3 Rack - Flux Trap Increased 0.9392 -0.0076 0.9372 -0.0082 0.8399 -0.0080 0.8545 -0.0091 4 Rack - Flux Trap Decreased 0.9514 6.0068-0.:9492 0.0064 0.8526 0.0069 0.8678 0.0069 5 Rack - Cell ID Increased 0.9414 -0.0,054, 40.9379 -0.0075 0.8423 -0.0058 0.8561 -0.0074 6 Rack - Cell ID Decreased 0.9476 0.003,0, 0:9465 0.0037 0.8488 0.0031 0.8639 0.0031 7 Rack - Wall Thickness Increased 0.9406 -0.0062--.. 0;9384 -0.0070 0.8433 -0.0046 0.8578 -0.0058 8 Rack - Wall Thickness Decreased 0.9511 0.0065 0.9498 0.0070 0.8501 0.0046 0.8661 0.0052 9 Rack - Sheathing Increased 0.9446 0.0000 0.9428 -0.0026 0.8458 0.0000 0.8608 -0.0027 10 Rack - Sheathing Decreased 0.9473 0.0027 0.9459 0.0031 0.8480 0.0023 0.8629 0.0000 11 Rack - Poison Gap Increased 0.9438 -0.0032 0.9422 -0.0032 0.8440 -0.0039 0.8587 -0.0048 12 Rack - Poison Gap Decreased 0.9480 0.0034 0.9459 0.0031 0.8492 0.0035 0.8641 0.0032 13 Rack - Polymer Thickness Increased 0.9464 0.0000 0.9438 0.0000 0.8468 0.0000 0.8640 0.0031 14 Rack - Polymer Thickness Decreased 0.9450 0.0000 0.9429 0.0000 0.8456 -0.0023 0.8611 0.0000 15 Rack - Polymer Width Increased 0.9444 -0.0024 0.9441 0.0000 0.8467 0.0000 0.8627 0.0000 16 Rack - Polymer Width Decreased 0.9453 0.0000 0.9447 0.0000 0.8459 0.0000 0.8613 0.0000 17 Total, Rack Uncertainties 0.0109 0.0110 0.0098 0.0102 18 Rack-Dominating Parameters 0.9577 0.0131 0.9540 0.0112 0.8572 0.0115 0.8729_[" 0.0121 19 Difference from Dominant Rack Parameters 0.0023 0.0001 0.0017 0.0019 (18-17)

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Table 4.6.35,.

Reactivity Effect of the Rack Tolerances, Pattern C Pattern C C (500 ppm)

Enrichment, wt% 2.6 5 2.6 5 Burnup, GWd/MTU 10 45 10 45 Delta k- Delta k- Delta k- Delta k-Item Itm RfrneK-calc cac calc K-calc 1 calc K-calc talc1 K-calc ci talc I Reference (Design Basis) 0.9296 0.9311 - 0.8312 - 0.8481 -

2 Rack Uncertainties:

3 Rack - Flux Trap Increased 0.9224 -0.0083 0.9240 -0.0082 0.8240 -0.0083 0.8413 -0.0081 4 Rack - Flux Trap Decreased 0.9344 0.0059 0.9380 0.0080 0.8383 0.0082 0.8551 0.0083 5 Rack - Cell ID Increased 0.9243 ---0.0064: . 0.9263 -0.0059 0.8256 -0.0067 0.8431 -0.0063 6 Rack - Cell ID Decreased 0.9323 0.00381 . 0:9337 0.0037 0.8336 0.0035 0.8510 0.0042 7 Rack - Wall Thickness Increased 0.9240 0.0067T 7709264 -0.0058 0.8276 -0.0047 0.8441 -0.0053 8 Rack - Wall Thickness Decreased 0.9346 0.0061 09367 0.0067 0.8347 0.0048 0.8528 0.0060 9 Rack - Sheathing Increased 0.9272 -0.0035:. .0'9299 -0.0023 0.8291 -0.0032 0.8470 0.0000 10 Rack - Sheathing Decreased 0.9308 0.0023

• 0:9330 0.0030 0.8310 0.0000 0.8499 0.0031 11 Rack - Poison Gap Increased 0.9267 -0.0040 0.9290 -0.0032 0.8286 -0.0037 0.8460 -0.0034 12 Rack - Poison Gap Decreased 0.9300 0.0000 0.9333 0.0033 0.8329 0.0028 0.8498 0.0030 13 Rack - Polymer Thickness Increased 0.9295 0.0000 0.9311 0.0000 0.8310 0.0000 0.8496 0.0028 14 Rack - Polymer Thickness Decreased 0.9296 0.0000 0.9309 0.0000 0.8295 -0.0028 0.8467 0.0000 15 Rack - Polymer Width Increased 0.9287 0.0000 0.9313 0.0000 0.8308 0.0000 0.8493 0.0000 16 Rack - Polymer Width Decreased 0.9293 0.0000 0.9313 0.0000 0.8306 0.0000 0.8485 0.0000 17 Total, Rack Uncertainties 0.0096 0.0120 0.0105 0.0122 18 Rack-Dominating Parameters 0.94071 0.0122. 0.9429T 0.0129 0.8418 0.0117 0.8588 0.0120 19 Difference from Dominant Rack Parameters 0.0026 0.0009 0.0012 -0.0002 (18-17)

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Table 4.6.36 Reactivity Effect of the Rack Tolerances, Pattern D Pattern D. D (500 ppm)

Enrichment, wt% 2.6 5 2.6 5 Burnup, GWdIMTU 15 _ 50 15 50 Delta k- Delta k- Delta k- Delta k-Item RfrneK-calc

.calc cac K-calc calc K-calc calc K-calc ci calc 1 Reference (Design Basis) 0.9433 - 0.9305 - 0.8430 - 0.8451 -

2 Rack Uncertainties:

3 Rack - Flux Trap Increased 0.9368 -0.0076 0.9237 -0.0079 0.8354 -0.0087 0.8388 -0.0074 4 Rack - Flux Trap Decreased 0.9488 0.0066 0.9369 0.0075 0.8502 0.0083 0.8519 0.0081 5 Rack - Cell ID Increased 0.9373 -,0.0071 09248 -0.0070 0.8370 -0.0071 0.8392 -0.0070 6 Rack - Cell ID Decreased 0.9464 0.0042 -.0.9321 0.0027 0.8447 0.0028 0.8479 0.0041 7 Rack - Wall Thickness Increased 0.9381 -0.0063 .0.9249 -0.0067 0.8377 -0.0064 0.8406 -0.0056 8 Rack - Wall Thickness Decreased 0.9497 0.00715 - -.9356 0.0062 0.8464 0.0045 0.8486 0.0046 9 Rack - Sheathing Increased 0.9421 -0.0023 0.9283 -0.0033 0.8421 0.0000 0.8445 0.0000 10 Rack - Sheathing Decreased 0.9439 0.0000 0.9315 0.0000 0.8437 0.0000 0.8455 0.0000 11 Rack - Poison Gap Increased 0.9409 -0.0035 0.9274 -0.0042 0.8399 -0.0042 0.8426 -0.0036 12 Rack - Poison Gap Decreased 0.9452 0.0030 0.9329 0.0035 0.8457 0.0038 0.8477 0.0039 13 Rack - Polymer Thickness Increased 0.9432 0.0000 0.9301 0.0000 0.8430 0.0000 0.8461 0.0000 14 Rack - Polymer Thickness Decreased 0.9430 0.0000 0.9296 0.0000 0.8420 0.0000 0.8437 -0.0025 15 Rack - Polymer Width Increased 0.9430 0.0000 0.9297 0.0000 0.8430 0.0000 0.8445 0.0000 16 Rack - Polymer Width Decreased 0.9426 0.0000 0.9299 0.0000 0.8423 0.0000 0.8453 0.0000 17 Total, Rack Uncertainties 0.0113. 0.0107 0.0106 0.0109 18 Rack-Dominating Parameters 0.9563F 0.0141 0.9409T 0.0115 0.8535 0.0116 0.8569 0.0129 19 Difference from Dominant Rack Parameters 0.0028 0.0008 0.0010 0.0020

_ (18-17)

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Table 4.6.37 Axial Burnup Distribution, Enriched Blankets Axial(1 =

Section Size, Burnup, Relative Burnup, Relative boStom) inches GWd/MTU Burnup GWd/MTU Burnup 1 6 0 0.3886 45 0.4238 2 6 0 0.6888 45 0.7983 3 6 0 0.9285 45 0.9738 4 6 0 1.0399 45 1.0470.

5 6 0 1.0881 45 1.0744 6 6 0 1.1056 45 1.0840 7 6 0 1.1009 45 1.0847 8 6 0 1.0951 45 1.0824 9 6 0 1.0904 45 1.0798 10 6 0 1.0852 45 1.0780 11 6 0 1.0858 45 1.0783 12 6 0 1.0875 45 1.0824 13 6 0 1.0805 45 1.0804 14 6 0 1,0762 45 1.0760 15 6 0 1 0726 45 1.0718 16 6 0  :.1.0641 45 1.0670 17 6 0 1*.0427 45 1.0618 18 6 0 1.0167 45 1.0545 19 6 0 0.9887 45 1.0397 20 6 0 0.9508 45 1.0129 21 6 0 0.8888 45 0.9648 22 6 0 0.7770 45 0.8774 23 6 0 0.5602 45 0.7253 24 6 0 0.3397 45 0.4278 Project No. 2064 Report No. HI-2114996 Page 118 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.38 Axial Burnup Distribution, Non-Blanketed Axial Section Size, Burnup, Relative Burnup, Relative (1 = bottom) inches GWd/MTU Burnup GWd/MTU Burnup 1 8 0 0.4327 45 0.5735 2 8 0 0.7329 45 0.9174 3 8 0 0.9155 45 1.0332 4 8 0 1.0196 45 1.0412 5 8 0 1.0805 45 1.0609 6 8 0 1.0381 45 1.0453 7 8 0 1.0572 45 1.0656 8 8 0 1.0515 45 1.0646 9 8 0 1.0191 45 1.0465 10 8 0 1.0486 45 1.0635 11 8 0 1.0221 45 1.0497 12 8 0 1.0302 45 1.0594 13 8 0 1.0207 45 1.0549 14 8 ,0  %',1.0029 45 1-0320 15 8 0. _-

-0.9852 45 1.0318 16 8 0' ,048386 45 0.9862 17 8 0 0.4455 45 0.8318 18 8 0 0.0661 45 0.5121 Project No. 2064 Report No. HI-2114996 Page 119 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.39 Examples of Calculation for Burnup Versus Enrichment Requirement Pattern 33 A B C D 235 Enrichment (wt% U) 5 5 5 5 Burnup (GWD/MTU) 15 10 35 30 45 40 55 50 Cooling time (yr) 0 0 0 0 0 0 0 0 Axial Profile 4 1 4 4 4 4 4 4 kcalc for Fresh Fuel 1.0061 1.0061 1.1417 1.1417 1.1901 1.1901 1.2369 1.2369 Standard deviation (Y) 0.0006 0.0006 0.0005 0.0005 0.0004 0.0004 0.0004 0.0004 kcalc for S(alpha, beta)=300K 0.9628 0.9725 0.9441 0.9684 0.9311 0.9571 0.9305 0.9475 Calculation Uncertainty (a) 0.0005 0.0005 0.0005 0.0005 0.0004 0.0004 0.0004 0.0004 Depletion Uncertainty 0.0022 0.0018 0.01 0.0087 0.013 0.0117 0.0154 0.0145 Burnup Uncertainty 0.0022 0.0018 0.01 0.0087 0.013 0.0117 0.0154 0.0145 kcalc for S(alpha, beta)=400K 0.959 0.9698 0.9583 0.9833 0.9475 0.9744 0.9477 0.966 Standard deviation (F) 0.0005 0.0005 0.0005 0.0004 0.0005 0.0004 0.0004 0.0005 kcalc Used to Determine Max. keff34 0.9628 0.9725 0.9524 0.9771 0.9406 0.9672 0.9405 0.9583 kcalc for FP Uncertainty 0.9795 0.987 1.0068- 1.0246 1.0098 1.0291 1.0181 1.0323 Standard deviation (a) 0.0005 0.0005' 0.0004 0.0004 0.0005 0.0005 0.0004 0.0004 FP Uncertainty 0.0009 "0.0008 ! 0.00 32' 0.0029 0.004 0.0037 0.0044 0.0043 kcalc for LFP Uncertainty 0.9643. 0.9743. _::0.953-.* 0.974 0.9431 0.9671 0.9439 0.9611 Standard deviation (a) 0.0005 6.6005'. -. 66604 0.0005 0.0004 0.0004 0.0004 0.0004 LFP Uncertainty 0.0004 0.0005 0.0i615 0.0011 0.002 0.0017 0.0022 0.0022 33 Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 5.0 wt% 235U, the enrichments and burnups presented in the table are only for spent fuel.

For Pattern A, kcalc for S(alpha, beta)=300K are used, for Patterns B, C and D, the values are interpolated for the temperature of 358.15K using kcalc for S(alpha, beta)=300K and kcalc for S(alpha, beta)=400K.

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Table 4.6.39 Examples of Calculation for Bumup Versus Enrichment Requirement (Continued)

Pattern A B C D 235 Enrichment (wt% U) 5 5 5 5 Bumup (GWD/MTU) 15 .10 35 30 45 40 55 50 Cooling time (yr) 0 0 0 0 0 0 0 0 Axial Profile 4- I 4 4 4 4 4 4 kcalc for Fuel Uncertainty 0.9693 0.9802 0.951 0.975 0.9377 0.9648 0.9363 0.9546 Standard deviation (a) 0.0005 0.0006 0.0005 0.0005 0.0004 0.0004 0.0004 0.0004 Fuel Uncertainty 0.0079 0.0093 0.0083 0.008 0.0077 0.0088 0.0069 0.0082 kcalc for Rack Uncertainty 0.9701 0,9812 0.954 0.98 0,9429 0.9695 0.9409 0.9609 Standard deviation (a) 0.0005 0.0005 0.0004 0.0005 0.0004 0.0005 0.0004 0.0004 Rack Uncertainty 0.0087 0.0101 0.0112 0.013 0.0129 0.0137 0.0115 0.0145 MCNP Code Uncertainty 0.0085 0.0085 0.0085 0.0085 0.0085 0.0085 0.0085 0.0085 Calculation Uncertainty (2a) 0.001 0.001 0.001 0.001 0.0008 0.0008 0.0008 0.0008 Fuel Creep Uncertainty 0.0039 0.0039 0.0042 0.0042 0.0034 0.0034 0.0028 0.0028 Eccentric Positioning Uncertainty 0 0 _0.0136 0.0136 0.015 0.015 0.0147 0.0147 Total Uncertainty (statistical combination) 0.0154 0.0l16"9,. jO.#0261 0.0259 0.0299 0.0294 0.0312 0.0319 Code Bias 0.0043 0.00.0 0.0043 0.0043 0.0043 0.0043 0.0043 0.0043 Other Bias (Pin Specific, Burnable Absorber) 0.0034, 0:0034. :0.0038 0.0038 0.0052 0.0052 0.0042 0.0042 Maximum k 0.9859 0.9971: .0.9866. 1.0111 0.98 1.0061 0.9802 0.9987 Target ker 0.99 ' 0.99 0.99 0.99 Calculated Bumup (GWD/MTU) 13.17 34.31 43.08 52.35 Project No. 2064 Report No. HI-2114996 Page 121 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.40 35 Minimum Burnup Requirements (GWd/MTU) from MCNP Calculations for Region I Racks Enrichment 1.5 2.6 3.6 4.5 5 Pattern A 0.00 0.00 2.72 9.33 13.17 35 Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 5.0 wt% 235U. The minimum burnup requirements in the table are only applicable for spent fuel. In addition, the results in this table are calculated by MCNP directly and do not include the polynomial adjustment.

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Table 4.6.41 Minimum Burnup Requirements (GWd/MTU) from MCNP Calculations 235 for Fuel in Region 2 Racks with Enrichment Range up to 3.6 wt% U 36 Enrichment 37 1.5 2.6 Pattern B, 0 years 0.00 4.85 Pattern B, 2.5 years 0.00 4.63 Pattern B, 5 years 0.00 4.43 Pattern B, 10years 0.00 4.38 Pattern B, 15 years 0.00 4.32 Pattern B, 20 years 0.00 4.29 Enrichment 1.5 2.6 Pattern C, 0years 0.00 12.81 Pattern C, 2.5 years 0.00 12.16 Pattern C, 5 years 0.00 11.78 Pattern C, 10 years 0.00 11.37 Pattern C, 15 years 0.00 10.96 Pattern C, 20 years 0.00 10.81 Enrichment 1.5 2.6 Pattern D, 0 years 0.00 21.42 r Pattern D, 2.5 years 0.00 ,20.13 Pattern D, 5 years 0.00 19.49 PatternD, 10 years 0.00 18.57 Pattern D, 15 years 0.00 18.02 Pattern D, 20 years 0.00 17.60 36 The results in this table are calculated by MCNP directly and do not include the polynomial adjustment.

37 Minimum Burnup Requirements in this table are applied for fuel with enrichment up to 3.6 wt%, but not including 3.6 wt%.

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Table 4.6.42 Minimum Burnup Requirements (GWd/MTU) from MCNP Calculations 235 for Fuel in Region 2 Racks with Enrichment Range of 3.6 - 5.0 wt% U 31 Enrichment 3.6 4.5 5 Pattern B, 0 years 17.94 29.22 34.31 Pattern B, 2.5 years 17.08 28.24 33.21 Pattern B, 5 years 16.88 27.46 32.46 PatternB, 10 years 16.28 26.56 31.64 Pattern B, 15 years 16.02 26.04 30.84 Pattern B, 20 years 15.74 25.76 30.56 Enrichment 3.6 4.5 5 Pattern C, 0 years 27.01 38.03 43.08 Pattern C, 2.5 years 26.09 36.60 41.42 Pattern C, 5 years 25.23 35.64 40.47 Pattern C, 10 years 24.37 34.43 38.95 Pattern C, 15 years 23.77 33.38 38.12 Pattern C, 20 years 23.34 32.82 37.47 Enrichment 3.6 4.5 5 Pattern D, 0 years 34.71 45.25 52.35 Pattern D, 2.5 years 33.13 43.29 49.59 PatternD, 5 years 32.16 42.12 47.55 PatternD, 10 years 30.81 40.30 44.84 Pattern D, 15 years 29.98 39.22 43.66 Pattern D, 20 years 29.31 38.54 42.85 38 The results in this table are calculated by MCNP directly and do not include the polynomial adjustment.

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Table 4.6.43 Bounding Polynomial Fits of Minimum Burnup Requirements for Region I Racks39 Coefficient A B C Pattern A -24.1514 7.4643 0.0000 39 Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 235 5.0 wt% U. The polynomial fits of minimum burnup requirements are only applicable for spent fuel.

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Table 4.6.44 Bounding Polynomial Fits of Minimum Burnup Requirements for Fuel in Region 2 Racks with Enrichment Range of up to 3.6 wt% 235U 40 Coefficient A B C Pattern B, 0 years -46.6454 23.4858 -1.4154 Pattern B, 2.5 years -45.0454 22.6125 -1.3487 Pattern B, 5 years -41.2633 20.4058 -1.0890 Pattern B, 10 years -38.5149 18.9692 -0.9504 Pattern B, 15 years -37.8530 18.7075 -0.9566 Pattern B, 20 years -37.3476 18.4617 -0.9412 Coefficient A B C Pattern C, 0 years -44.6593 26.7356 -1.7815 Pattern C, 2.5 years -45.2071 26.8990 -1.8595 Pattern C, 5 years -42.4764 25.2213 -1.6744 Pattern C, 10 years -41.0757 24.4119 -1.6310 Pattern C, 15 years -40.4829 23.9210 -1.5905 Pattern C, 20 years -39.4907 23.3811 -1.5518 Coefficient A B C Pattern D, 0 years -26.8286 20.9095 -0.9048 PatternD, 2.5 years -28.1650 21.2508 -1.0292 PatternD, 5 years -29.2879 21.8885 -1.2030 PatternD, 10 years -30.6536 22.5737 -1.4006 PatternD, 15 years -31.8257 23.2386 -1.5643 Pattern D, 20 years -30.5557 22.3502 -1.4726 40 Minimum Burnup Requirements in this table are applied for fuel with enrichment up to 3.6 wt%, but not including 3.6 wt%.

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Table 4.6.45 Bounding Polynomial Fits of Minimum Burnup Requirements for Fuel in Region 2 Racks with Enrichment Range of 3.6 - 5.0 wt% 235U Coefficient A B C Pattern B, 0 years -45.1986 22.2592 -1.2715 Pattern B, 2.5 years -43.8453 21.4900 -1.2158 Pattern B, 5 years -41.1255 19.9583 -1.0482 Pattern B, 10 years -36.9034 17.6517 -0.7886 PatternB, 15 years -37.7850 18.3083 -0.9167 Pattern B, 20 years -36.8808 17.8083 -0.8640 Coefficient A B C Pattern C, 0years -41.8843 24.6516 -1.5317 Pattern C, 2.5 years -39.5300 23.4678 -1.4556 Pattern C, 5 years -38.4729 22.5981 -1.3619 PatternC, 10years -40.6071 23.5463 -1.5270 Pattern C, 15 years -28.5300 17.6078 -0.8556 Pattern C, 20 years -28.8514 17.6690 -0.8810 Coefficient A B C Pattern D, 0 years 21.3500 -2.6889 1.7778 Pattern D, 2.5 years 7.6614. 3.7032 .0.9365 Pattern D, 5 years -10.0714 12.2624 -0.1476 Pattern D, 10 years -24.0957 19.0173 -1.0460 Pattern D, 15 years -23.0257 18.2895 -0.9905 PatternD, 20 years -26.5357 19.7184 -1.1683 Project No. 2064 Report No. HI-2114996 Page 127 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.46 Minimum Burnup Requirements (GWd/MTU) from Polynomial Fits for Region I Rack 4" Enrichment 1.5 2.6 3.6 4.5 5 Pattern A 0.00 0.00 2.72 9.44 13.17 41 Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 235 5.0 wt% U. The polynomial fits of minimum burnup requirements are only applicable for spent fuel.

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Table 4.6.47 Minimum Burnup Requirements (GWd/MTU) from Polynomial Fits for Fuel in Region 2 Racks with Enrichment Range of up to 3.6 wt% 235U Enrichment42 1.5 2.6 Pattern B, 0 years 0.00 4.85 Pattern B, 2.5 years 0.00 4.63 Pattern B, 5 years 0.00 4.43 Pattern B, 10 years 0.00 4.38 Pattern B, 15 years 0.00 4.32 Pattern B, 20 years 0.00 4.29 Enrichment 1.5 2.6 Pattern C, 0 years 0.00 12.81 Pattern C, 2.5 years 0.00 12.16 Pattern C, 5 years 0.00 11.78 Pattern C, 10 years 0.00 11.37 Pattern C, 15 years 0.00 10.96 PattemrC, 20 years 0.00 10.81

. Enrichment 1.5 2.6 Pattern D, 0 years 2.50 21.42 Pattern D, 2.5 years 1.40 20.13 Pattern D, 5 years 0.84 19.49 Pattern D, 10 years 0.06 18.57 Pattern D, 15 years 0.00 18.02 Pattern D, 20 years 0.00 17.60 42 Minimum Burnup Requirements in this table are applied for fuel with enrichment up to 3.6 wt%, but not including 3.6 wt%.

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Table 4.6.48 Minimum Burnup Requirements (GWd/MTU) from Polynomial Fits for Fuel in Region 2 Racks with Enrichment Range of 3.6 - 5.0 wt% 235U Enrichment 3.6 4.5 5 Pattern B, 0 years 18.46 29.22 34.31 PatternB, 2.5 years 17.76 28.24 33.21 Pattern B, 5 years 17.14 27.46 32.46 Pattern B, 10 years 16.42 26.56 31.64 Pattern B, 15 years 16.25 26.04 30.84 Pattern B, 20 years 16.03 25.76 30.56 Enrichment 3.6 4.5 5 Pattern C, 0 years 27.01 38.03 43.08 Pattern C, 2.5 years 26.09 36.60 41.42 Pattern C, 5 years 25.23 35.64 40.47 Pattern C, 10 years 24.37 34.43 38.95 PatternC, 15 years 23.77 33.38 38.12 Pattern C;.20 years 23.34 32.82 37.47 Enrichnfent, 3.6 4.5 5 Pattern D,O years 34.71 45.25 52.35 Pattern D, 2.5 years 33.13 43.29 49.59 Pattern D, 5 years 32.16 42.12 47.55 Pattern D, 10 years 30.81 40.30 44.84 Pattern D, 15 years 29.98 39.22 43.66 Pattern D, 20 years 29.31 38.54 42.85 Project No. 2064 Report No. HI-2114996 Page 130 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.49 Summary of Additional Configuration for Pattern B and Pattern C Enr Cooling Bu K-calc Unc 43, K-calc +

time Unc 2.6 0 4.85 0.9083 0.0381 0.9464 Pattern B, but 1 RCCA is replaced by Empty Cell 5.0 0 34.31 0.9052 0.0357 0.9409 5.0 20 30.56 0.9050 0.0371 0.9421 2.6 0 12.81 0.8998 0.0424 0.9422 Pattern C, but 1 RCCA is 5.0 0 43.08 0.8981 0.0406 0.9387 replaced by Empty Cell 5.0 20 37.47 0.8987 0.0407 0.9394 43 The Unc is the total uncertainty and bias determined by the calculation of loading curve.

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Table 4.6.50 44 Summary of Soluble Boron Calculations, 500 ppm Enr Cooling time Bu K-calc Unc 45 K-calc + Unc Pattern A 3.4 0 0.00 0.9130 0.0248 0.9378 5.0, 0 13.17 0.9136 0.0252 0.9388 2.3 0 0.00 0.8516 0.0405 0.8921 Pattern B 5.0 0 34.31 0.8780 0.0357 0.9137 5.0 20 30.56 0.8764 0.0371 0.9135 1.9 0 0.00 0.8338 0.0434 0.8772 Pattern C 5.0 0 43.08 0.8743 0.0406 0.9149 5.0 20 37.47 0.8722 0.0407 0.9129 1.5 0 0.00 0.7961 0.0480 0.8441 Pattern D 5.0 0 52.35 0.8705 0.0423 0.9128 5.0 20 42.85 0.8702 0.0433 0.9135 44Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 5.0 23 5 wt% U. The enrichments and burnups presented in the table are only for spent fuel.

45 For some cases, the Unc is larger than the total uncertainty and bias determined by the loading curve calculation, since the reactivity effect of the BPRA or IFBA rods for the 500 ppm case is considered. It may be larger than the reactivity effect for the 0 ppm case that applied in the loading curve calculation.

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Table 4.6.51 Summary of Other Normal Conditions I

Enr Bu K-calc IUnc" K-calc + Unc Single Fuel Assembly (normal temperature) 5 0 0.9357 0.0422 0.9779 Single Fuel Assembly (off-normal temperature) 5 0 0.9379 0.0422 0.9801 46 The Unc is the total uncertainty and bias determined by the calculation of loading curve.

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)

Table 4.6.52 Summary of Pattern D with Two Peripheral Row 7

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Table 4.6.53 Summary of Calculations for the Fuel Rod Storage Basket47 Case Enr Bu K-calc Unc 48 K-calc +

Unc Pattern A with a FRSB Full 5 13.17 0.9285 0.025 0.9535 with Fresh Fuel Pins Pattern D with a FRSB Full 5 52.35 0.9149 0.0422 0.9571 with Fresh Fuel Pins Pattern B with a FRSB Full 5 43.08 0.9461 0.0406 0.9867 with Fresh Fuel Pins Pattern C with a FRSB Full 34.31 0.9465 0.0357 0.9822 with Fresh Fuel Pins _

47 Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 5.0 wt% 235U. The enrichments and burnups presented in the table are only for spent fuel.

48 The Unc is the total uncertainty and bias determined by the calculation of loading curve.

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Table 4.6.54 Summary of Accident Conditions4 9 Enr Bu Boron K-calc Unc 50 K-calc + Unc Misload Accidents Single Fuel Assembly 5 0 1400 0.8707 0.0423 0.9130 Misloaded Assembly Accidents Pattern A 5 13.17 1400 0.8383 0.0252 0.8635 Pattern B - RCCAs-Row 5 34.31 1400 0.8302 0.0357 0.8659 Pattern B - RCCAs-Diagonal 5 34.31 1400 0.8383 0.0357 0.8740 Pattern C 5 43.08 1400 0.8315 0.0406 0.8721 Pattern D 5 52.35 1400 0.8214 0.0423 0.8637 Multiple Assembly/Insert Misload Pattern C 5 34.31 1400 0.8081 0.0406 0.8487 Pattern D 5 43.08 1400 0.7987 0.0423 0.8410 Mislocated Assembly Accidents Pattern A 5 13.17 1600 0.9067 0.0252 0.9319 Pattern B 5 34.31 1400 0.8640 0.0357 .0.8997 Pattern C 5 43.08 1400 0.8677 0.0406 0.9083 Pattern D 5 52.35 1400 0.8528 0.0423 0.8951 49 Pattern A is a checkerboard of fresh fuel and spent fuel, with fresh fuel at the maximum nominal enrichment of 5.0 wt% 235 U. The enrichments and burnups presented in the table are only for spent fuel.

50 For some cases, the Unc is larger than the total uncertainty and bias determined by the loading curve calculation, since the reactivity effect of the BPRA or IFBA rods for the 500 ppm case is considered. It may be larger than the reactivity effect for the 0 ppm case that applied in the loading curve calculation.

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Table 4.6.55 Area of Applicability Parameter Seabrook Fuel Assemblies Fresh and Spent U0 2 fuel Initial Fuel Enrichments 1.5 to 5 wt% 2,J Fuel Density up to 10.517 g/cc Burnup Range up to about 55GWD/MTU for 5% enrichment Moderator Material H 20 Credited Soluble Boron 0 to 1600 ppm Temperature 39 F to 185 F Neutron Poison 13B-IO (BORALTM Panel),

Ag-In-Cd (RCCA)

Interstitial Material Steel Fuel Cladding Zirconium Alloy Reflective or Periodic Reflector Boundary Neutron Energy Thermal spectrum Project No. 2064 Report No. HI-2114996 Page 137 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.56 Reactivity Effect of the RCCA Tolerances, Pattern B Pattern B 2.6 wt% /5 GWd/MTU 5.0 wt% / 35 GWd/MTU k-calc delta-k k-calc delta-k Reference - All Nominal Parameters 0.9457 - 0.9441 RCCA Tolerances RCCA - Absorber Density Increased 0.9458 0.0000 0.9440 0.0000 RCCA - Absorber Density Decreased 0.9458 0.0000 0.9439 0.0000 RCCA - Absorber OD Increased 0.9457 0.0000 0.9438 0.0000 RCCA - Absorber OD Decreased 0.9456 0.0000 0.9443 0.0000 RCCA - Cladding Thickness Increased 0.9457 0.0000 0.9431 0.0000 RCCA - Cladding Thickness Decreased 0.9454 0.0000 0.9431 0.0000 RCCA - Cladding OD Increased 0.9460 0.0000 0.9441 0.0000 RCCA - Cladding OD Decreased 0.9451 0.0000 0.9443 0.0000 RCCA - Silver Content Increased 0.9450 0.0000 0.9434 0.0000 RCCA - Silver Content Decreased 0.9455 0.0000 0.9438 0.0000 RCCA - Indium Content Increased 0.9456 0.0000 0.9435 0.0000 RCCA - Indium Content Decreased 0.9457 0.0000 -.0.9450 -00000 RCCA - Cadmium Content Increased 0.9455 0.0000 0.9445 0.0000 RCCA - Cadmium Content Decreased 0.9456 0.0000 0.9439 0.0000 Total, RCCA Tolerances 0.0000 0.0000 Project No. 2064 Report No. HI-2114996 Page 138 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.57 Reactivity Effect of the RCCA Tolerances, Pattern C Pattern C 2.6 wt% / 15 GWd/MTU 5.0 wt% / 45 GWd/MTU k-calc delta-k k-calc delta-k Reference - All Nominal Parameters 0.9296 - 0.9311 -

RCCA Tolerances RCCA - Absorber Density Increased 0.9295 0.0000 0.9308 0.0000 RCCA - Absorber Density Decreased 0.9283 -0.0024 0.9311 0.0000 RCCA - Absorber OD Increased 0.9290 0.0000 0.9307 0.0000 RCCA - Absorber OD Decreased 0.9288 0.0000 0.9308 0.0000 RCCA - Cladding Thickness Increased 0.9289 0.0000 0.9313 .0.0000 RCCA - Cladding Thickness Decreased 0.9294 0.0000 0.9306 0.0000 RCCA - Cladding OD Increased 0.9289 0.0000 0.9314 0.0000 RCCA - Cladding OD Decreased 0.9297 0.0000 0.9314 0.0000 RCCA - Silver Content Increased 0.9289 0.0000 0.9317 0.0000 RCCA - Silver Content Decreased 0.9289 0.0000 0.9314 0.0000 RCCA ,Indium Content Increased 0.9287 0.0000 0.9305 _ 0:0000 RCCA L-Indium Content Decreased 0.9285 0.0000 0.9312;: , 0:0000 RCCA L-Cadmium Content Increased 0.9285 0.0000 0.9305 "0.0000 RCCA - Cadmium Content Decreased 0.9289 0.0000- 0.9318 ' 0.00000 Total, RCCA Tolerances 0.0000 0.0000 Project No. 2064 Report No. HI-2114996 Page 139 Shaded Areas Denote Holtec International Proprietary Information

Table 4.6.58 Average Boron Concentration for Each Cycle Cycle Average Boron Concentration (ppm) 1 449.98 2 485.91 3 671.42 4 747.72 5 881.30 6 971.91 7 893.46 8 808.05 9 727.02 10 778.34 11 900.43 12 878.79 13 912.99 14 875.16

• r.

Ir Project No. 2064 Report No. HI-2114996 Page 140 Shaded Areas Denote Holtec International Proprietary Information

BORAL PANEL 7.5~ vý -75t 0~1015 AN .U 2B-toWLAUiNG gkm r

!/

• I /,*

T---  ?.

0.525 Figure 4.5.1: Sketch of Region I Racks, Detailing Important Dimensions SNiL(NOT

~;%7 TO SCALE, all dimensions in inches)

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BORA FLEX PANEL (NOT CREDITED[

7.46 .x 0.0-7' Figure 4.5.2: Sketch of Region 2 Racks, Detailing Important Dimensions . .. .(NOT TO SCALE, all dimensions in inches)

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Figure 4.5.3: A Two-Dimensional Representation of the Actual Calculational Model Used for the Region I Rack Analysis (Pattern A)

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Figure 4.5.4: A Two-Dimensional Representation of the Actual Calculational Model Used for the Region 2 Rack Analyses (Pattern C that Contains I RCCA)

Project No. 2064 Report No. HI-2114996 Page 144 Shaded Areas Denote Holtec International Proprietary Information

N 4VKe I

-5O0RVrý 1 i --

Ratt lox 10 1Ox9 O9 44.40 Rdf.

-4 27A40 Re(.

D7 @ a 198.60 Ref.

+1I +

f- 7t+

+ +1+ +

1 x 10 lOx0O

• .--22.40 Ref. 1 4- *--

91

, 16.80 Ref.

_LL-1

+/- ++/- +1+ + 450.00 Ref.

lOx 11 lOxll 1oxli 0 a

+1+- -+

11.85 Tmp,

+

hg..

+ 4+ -I 10 lOxll 1 1 10350 .Ty 1117p 7es1. 2 O Ref.

i$R15 324.' Re Rot Figure 4.5.5: A Two-Dimensional Representation of the Existing Seabrook Spent Fuel Pool Layout with Rack Region Layout Project No. 2064 Report No. HI-2114996 Page 145 Shaded Areas Denote Holtec International Proprietary Information

Figure 4.5.6: A Two-Dimensional Representation of the Actual Calculational Model Used for the Fuel Rod Basket Rack Analysis Project No. 2064 Report No. HI-2114996 Page 146 Shaded Areas Denote Holtec International Proprietary Information

Non-Blanketed Fuel 30GWd/mtU --- 45 GWd/mtU 1.2000 1.0000 0.8000 C.

C-0.6000 Z

0.4000 0.2000 0.0000 0 30 60 90 120 150 Axial Position (inch)

Figure 4.6.1 Axial Burnup Profiles for Non-Blanketed Fuel Project No. 2064 Report No. HI-21 14996 Page 147 Shaded Areas Denote Holtec International Proprietary Information

Fuel with Enriched Blankets

-30 GWd/mtU 45 GWd/mtU 1.2000 1.0000 0.8000 C.

C

.a 0.6000 0.4000 0.2000 0.0000 -r- - -------- -.15o 1

0 30 60 90 120 150 Axial Position (inch)

Figure 4.6.2: Axial Bumup Profiles for Fuel with Enriched Uranium Blankets Project No. 2064 Report No. HI-2114996 Page 148 Shaded Areas Denote Holtec International Proprietary Information

Pattern A 25.00 20.00 Loading Permited E

15.00 C.

m Pattern A

-Fit Pattern A 10.00 5.00 I .oading Not Permited 0.00 1.5 2 2.5 3 3.5 4 4.5 5 Enrichment, wt%

Figure 4.6.3: Loading Curve for Pattern A Project No. 2064 Report No. HI-2114996 Page 149 Shaded Areas Denote Holtec International Proprietary Information

Pattern B 20.00 Loading Permited 15.00 + Pattern B, 0 years 0 Pattern B, 2.5 years A Pattern B, 5 years X Pattern B, 10 years E

X Pattern B, 15 years 0 Pattern B, 20 years E -Fit Pattern B, 0 years

.5

.- - Fit Pattern B, 2.5 years Loading Not Loadinget - Fit Pattern B, 5 years

-Fit Pattern B, 10 years 5.00

- Fit Pattern B, 15 years

- Fit Pattern B, 20 years 0.00 ----.. .

1.5 1.8 2.1 2.4 2.7 3 3.3 3.6 Enrichment, wt%

5 Figure 4.6.4.a: Loading Curves for Pattern B, Enrichment up to 3.6 wt% 1

"' Loading curves are applied for fuel with enrichment up to 3.6 wt%, but not including 3.6 wt%.

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Pattern B 40.00 35.00 30.00

• Pattern B, 0 years M Pattern B, 2.5 years 25.00

• Pattern B, 5 years X Pattern B, 10 years

• Pattern B, 15 years E 20.00

• Pattern B, 20 years E - Fit Pattern B, 0 years "E 15.00 -Fit Pattern B, 2.5 years

- Fit Pattern B, 5 years 10.00 - Fit Pattern B, 10 years

- Fit Pattern B, 15 years

- Fit Pattern B, 20 years 5.00 0.00 . ..... .........................

L. .. . .. ....... ... .. I 3.6 3.8 4 4.2 4.4 4.6 4.8 5 Enrichment, wt%

Figure 4.6.4.b: Loading Curves for Pattern B, Enrichment Range 3.6 -5.0 wt%

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Pattern C 30.00 25.00 Loading Permited

• Pattern C, 0 years

• Pattern C, 2.5 years M20.00 E A Pattern C, 5 years X Pattern C, 10 years

• Pattern C, 15 years c 15.00

• Pattern C, 20 years E - Fit Pattern C, 0 years E - Fit Pattern C, 2.5 years 10.00 -

.- Fit Pattern C, 5 years Loading Not Permited - Fit Pattern C, 10 years

- Fit Pattern C, 15 years 5.00 1

-Fit Pattern C, 20 years 0.00 1.5 1.8 2.1 2.4 2.7 3 3.3 3.6 Enrichment, wt%

Figure 4.6.5.a: Loading Curves for Pattern C, Enrichment up to 3.6 wt%5 2 52 Loading curves are applied for fuel with enrichment up to 3.6 wt%, but not including 3.6 wt%.

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Pattern C 50.00 45.00 40.(00 I n~dinu Pprmitpd

. Pattern C, 0 years 35.(000 . Pattern C, 2.5 years E

A Pattern C, 5 years 30.0 X Pattern C, 10 years X# Pattern C, C2y15 years P0r 25.o

- Pattern C, 20 years E

- Fit Pattern C, 0 years 20.(

'E Loading Not --- Fit Pattern C, 2.5 years Permited -- Fit Pattern C, 5 years 15.(

- Fit Pattern C, 10 years 10.( _0__- Fit Pattern C, 15 years i Fit Pattern C, 20 years 5.0 0o 0 .. . . a L. ........................ ... . ......................... .....

0.0 3.6 3.8 4 4.2 4.4 4.6 4.8 5 Enrichment, wt%

Figure 4.6.5.b: Loading Curves for Pattern C, Enrichment Range 3.6 -5.0 wt%

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Pattern D 40.00 35.00 30.00

• Pattern D, 0 years

• Pattern D, 2.5 years 25.00 A Pattern D, 5 years

31. X Pattern D, 10 years c=20.00

• Pattern D, 15 years

• Pattern D, 20 years E - Fit Pattern D, 0 years

._E 15.00 15.00--Fit Pattern D, 2.5 years

-. Fit Pattern D, 5 years 10.00 10.00

- Fit Pattern D, 10 years

-Fit Pattern D, 15 years

- Fit Pattern D, 20 years 5.00 0.00 1.5 1.8 2.1 2.4 2.7 3 3.3 3.6 Enrichment, wt%

53 Figure 4.6.6.a: Loading Curves for Pattern D, Enrichment up to 3.6 wt%

13 Loading curves are applied for fuel with enrichment up to 3.6 wt%, but not including 3.6 wt%.

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Pattern D 60.00 55.00 -

50.00 Loading Permited 45.00

• Pattern D, 0 years

• Pattern D, 2.5 years 4 40.00 E A Pattern D, 5 years 3 35.00 4 X Pattern D, 10 years

• Pattern D, 15 years 30.00

• Pattern D, 20 years E 25.00 -Fit Pattern D, 0 years

.E - Fit Pattern D, 2.5 years 220.00*

- Fit Pattern D, 5 years

- Fit Pattern D, 10 years 15.00

- Fit Pattern D, 15 years 10.00 -Fit Pattern D, 20 years 5.00 0.00 L .....* .....................

. ............... L .........................

I...... ..............-- ...... . . . .....L

.........................I 3.6 3.8 4 4.2 4.4 4.6 4.8 5 Enrichment, wt%

Figure 4.6.6.b: Loading Curves for Pattern D, Enrichment Range 3.6 -5.0 wt%

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Figure 4.6.7: A Two-Dimensional Representation of the Actual Calculational Model Used for the Analysis of the Fuel Stored on the Periphery of the Region 2 Rack Project No. 2064 Report No. HI-2114996 Page 156 Shaded Areas Denote Holtec International Proprietary Information

Figure 4.6.8: A Two-Dimensional Representation of the Actual Calculational Model Used for the Analysis for a Mislocated Fuel Assembly54 14 This figure only presents the partial of the model which shows the position of the mislocated assembly.

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5. REFERENCES

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

2. "MCNP - A General Monte Carlo N-Particle Transport Code, Version 5," Los Alamos National Laboratory, LA-UR-03-1987, April 24, 2003 (Revised 2/1/2008).

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

5. "Nuclear Group Computer Code Benchmark Calculations", Holtec Report HI-2104790 Rev.

0, January 2011.

6. "Sensitivity Studies to Support Criticality Analysis Methodology", Holtec Report HI-2104598 Rev. 1, October 2010.

7. "Lumped FissionProductand P'Ml48m Crss'-Sections for MCNP," Holtec Report HI-2033031, Rev 2, Januaiy201 1.,

8. 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).

9. "Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit," NUREG/CR-6760, ORNL/TM-2000-321, March 2002.

10. DSS-ISG-2010-01: Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools.

11. Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001.

12. "Atlas of Neutron Resonances", S.F. Mughabghab, 5th Edition, National Nuclear Data Center, Brookhaven National Laboratory, Upton, USA.

13. "Spent Nuclear Fuel Burnup Credit Analysis Validation", ORNL Presentation to NRC, September 21, 2010.

Project No. 2064 Report No. HI-2114996 Page 158 Shaded Areas Denote Holtec International Proprietary Information

14. "Assessment of Reactivity Margin and Loading Curves for PWR Burnup-Credit Cask Designs", NTJREG/CR-6800 / ORNL/TM 2002/6, March 2003.

15. M. D. DeHart, "Sensitivity and Parametric Evaulations of Significant Aspects of Burnup Credit for PWR Spent Fuel Packages", ORNL/TM-12973, May 1996.

16. "HI-STAR 100 Safety Analysis Report", Holtec Report HI-951251 Rev. 15, USNRC Docket 71-9261.

17. R. J. Cacciapouti and S. Van Volkinburg, "Axial Burnup Profile Database for Pressurized Water Reactors", Yankee Atomic Electric Company Report YAEC-1937, 1997.

18. "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses",

ORNLTM-2001/273, NUREG/CR-6801, USNRC Office of Nuclear Regulatory Research, March 2003.

19. "Burnup Credit for the MPC-32", Holtec Report HI-2012630, Rev. 2, October 2006.

20. ORIGEN-S Decay Data Library and Half-Life Uncertainties, 0. W. Hermann, P. R. Daniel, and J. C. Ryman, ORNL/TM-13624, September 1998.

21. "Criticality Analysis .of ihe Seabrook Unit. 1 Spent Fuel Pool Racks", Holtec Report HI-.

2114904;Rev. 1;;Qctobei'ri201 1.

22. "Effects of Neutron Absorber Degradation 6n Seabrook Spent Fuel Pool Thermal-Hydraulics", Holtec Report HI-2073799 Rev. 1, November 2007.

23. "Criticality Analysis of the Seabrook Unit 1 New Fuel Vault", Holtec Report HI-2114912 Rev. 1, October 2011.

Project No. 2064 Report No. HI-2114996 Page 159 Shaded Areas Denote Holtec International Proprietary Information

Appendix A Applicability of Criticality Benchmark Calculations Benchmarking of MCNP5-1.51 is documented in [5], based on a total set of 291 critical experiments. The bias and bias uncertainty is established in [5] for the entire set of those 291 experiments, and for various subsets of experiments.

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Attachment 5 Marked-up Technical Specification Bases Pages

INDEX BASES SECTION PAGE 3/4.9.8 RESIDUAL HE EMOVAL AND COOLANT CIRCULATION B 3/4 9-3 314.9.9 (THIS SPECIFI ATION NUMBER IS NOT USED) -------------------------------------- B 3/4 9-4 3/4,9.10 and 3/4.9.11 ATER LEVEL - REACTOR VESSEL and

• SI RAGE POOL ................................... B 3/4 9-4 3/4.9.12 FUEL.OTRAGE BUILDING EMERGENCY AIR CLEANING SYSTEM B 3/4 9-4 3/4.9. SYSPETEM...................... ..... .................................... B3/4 9-4 3/4.9.13 SPENTFUEL AS BLY STOR GE B 3/4 9-4f5 O 3/4.10 SPECIAL TEST EXCEPTINS 3/4.10.1 SHUTDOWN MARGIN .......................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS ......... B 3/4 10-1 3/4.10.3 PHYSICS TESTS .............................................. B 3/4 10-1 3/4.10.4 (THIS SPECIFICATION NUMBER IS NOT USED) ............................................ B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN ..................... B 3/4 10-1 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 (THIS SPECIFICATION NUMBER IS NOT USED) ............................................ B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS .................................... ... ;. .B 3/4 11-2 3/.4,11.3 ,(THIS SPECIFICATION NUMBER IS NOT USED) ..................... B 3/4 11-5 3/4.11.4 :(THIS SPECIFICATION NUMBER IS NOT USED) 1:3/411-5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 (THIS SPECIFICATION NUMBER IS NOT USED) ................. . B 3/4 12-1 3/4.12.2 (THIS SPECIFICATION NUMBER IS NOT USED) ..................... B 3/4 12-1 3/4.12.3 (THIS SPECIFICATION NUMBER IS NOT USED)..................... B 3/4 12-2 3/1jq.S. P SP:r Furi.L PO L61 13 6 .@0J cooi c.. A-r'ran,. -3. I q SEABROOK - UNIT 1 iii BOR Ne. 00-02, BC 04,-04,47-81

3/4.9 REFUELING OPERATIONS (Continued)

BASES 3/4.9.9 (THIS SPECIFICATION NUMBER IS NOT USED.)

3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR. VESSEL and STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis. Suspending fuel movement or crane operation does not preclude moving a component to a safe location.

3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM The limitations on the Fuel Storage Building Emergency Air Cleaning System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with-,

the assum lmtion s'of the safety analyses'. ANSI N510-1980 will be used as a procedural guide forsurveillance testing. Suspending fuel movement or crane operation does not preclude.moving a component to a safe location.

Ione train of the Fuel Storage Building Emergency Air Cleaning System must be in:

operation during fuel movement. This requirement, however, does not apply to movement of a spent fuel cask containing irradiated fuel in preparation for transfer to dry storage.

Movement of fuel after it has been inserted into a spent fuel cask and unlatched from the lifting tool is no longer a consideration with regard to this specification.

3/4.9.13 SPENT FUEL ASSEMBLY STORAGE WXP-+  :&-'

J'r 1 Retitoso lcment of fuel assemblies fcranercmnswti h pn

('Fuel Pool is dictated by Figure 3.9-1. These restrictions ensure that the Kef of the Spent ,

Fulool will always remain less than 0.95 assuming tepo ob loed with unborated /

wae Th . es~trictions delineated in Figure 3.9-1 andteato ttment are consistent wihth rticality safety analysis performed for the Spent Fuel Pool as documented in the FSAR. --.

SEABROOK - UNIT 1 B 3/4 9-4 A,mcndmont 6, 85, BC 0-5-07,

jq r,)D TAtS E SEABROOK - UNIT 1 B 3/4 9-5 BC 08-10 I

Technical Specification Bases 3/4.9.13 SPENT FUEL ASSEMBLY STORAGE :r*l."

Teslimtions on placement of fuel assemblies of certain enrichments within the Spent cFuel Pool is dictated by Specification 5.6.1.3. These restrictions ensure that the keff of the moeSpent Fuel Pool will always remain less than 1.0 assuming the pool to be flooded with

'* unborated water and less than or equal 0.95 when flooded with water borated to 500 ppm. The restrictions delineated in Specification 5.6.1.3 and the action statement are consistent with the crtcality safety analysis performed for the Spent Fuel Pool as documented in the FSAR. ..

3/4.9.15 SPENT FUEL POOL BORON CONCENTRATION.*/ " -

Th iiato nte Spent Fuel Pool boron concentration ensure that sufficient boron is present""

Sto maintain criticality margin during any potential spent fuel pool accident. The required boron

')concentration is also sufficient to ensure that no boron dilution event could reduce the spent fuel kconcentration below 500 ppm. The action statement requires immediately suspending /

"[movementof_ fuel until the boron concentration has been restored. This does not preclude .,.***,,;;'"'i, ovmn f ulasembly to a safe position..