ML103500365

Holtec Report No. HI-2012621 (Non-Proprietary), Criticality Safety Evaluation for the Nine Mile Point 2 Rack Installation Project

ML103500365
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Issue date:	12/13/2010
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HI-2012621
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ATTACHMENT 2 HOLTEC REPORT NO. HI-2012621(NON-PROPRIETARY)

Nine Mile Point Nuclear Station, LLC December 13,2010

"Eu'.M. Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797-0900 HOLTEC INTERNATIONAL Fax (856) 797-0909 CRITICALITY SAFETY EVALUATION FOR THE NINE MILE POINT 2 RACK INSTALLATION PROJECT FOR NIAGARA MOHAWK POWER CORPORA TION by Holtec International 555 Lincoln Drive West Marlton, New Jersey 08053 Holtec Project No. 1101 Holtec Report No. 2012621 Report Category: Safety Related This document version has all proprietaryinformation removed and has replaced those sections,figures, and tables with highlightingand/or notes to designate the removal ofsuch information. This document is to be used only in connection with the performance of work by Holtec International or its designatedsubcontractors.Reproduction,publication or presentation,in whole or in part,for any other purpose by any party other than the Client is expressly forbidden.

I Holtec Report HI-2012621 Holtec Project 1101

Summary of Revisions Revision 1:

This revision incorporates the comments made by Carl Lepine of Nine Mile Point Unit 2 regarding bundle manufacturing tolerances and interaction between new and old racks.

Project No. I110 1 Holtec Report No. HI-2012621 Pagei

Table of Contents 1.0 Introduction ....................................................................................................................... 1 2.0 M ethodology ...................................................................................................................... 1 2.1 Code Validation ...................................................................................................... 2 3.0 Acceptance C riteria ..................................................................................................... 2 4.0 Assum ptions ...................................................................................................................... 3 5.0 Input D ata .......................................................................................................................... 4 5.1 Fuel A ssem bly Specifications ................................................................................... 4 5.2 Storage Rack Cell Specifications ............................................................................... 5 6.0 Com puter Codes ......................................................................................................... 5 7.0 Calculations ....................................................................................................................... 6 7.1 Reference Fuel A ssem bly .......................................................................................... 7 7.2 M anufacturing Tolerances and Burnup Uncertainty .................................................. 8 7.3 CA SMO -4 to M CN P-4A Com parison ........................................................................ 9 7.4 Tem perature Effect .................................................................................................. 10 7.5 Effect of the Channel and Eccentric Fuel Positioning .............................................. 10 7.5.1 Channel Removal and Channel Thickness....................................................... 10 7.5.2 Channel Bulging .............................................................................................. 10 7.5.3 Eccentric Positioning........................................................................................ 11 7.6 Comparison to Vendor Calculations ................................... 11 7.7 Boral Height Reduction ............................................................................................ 11 7.8 Final Calculations ..................................................................................................... 12 7.8.1 Maximum knf in the StandardCold Core Geometry.......................................... 12 7.8.2 CriteriaFor Minim um Gadolinium Loading..................................................... 12 7.9 Long Term Reactivity Changes ............................................................................... 13 7.10 A bnorm al and A ccident Conditions ......................................................................... 13 7.10.1 Dropped Fuel Assem bly .................................................................................... 13 7.10.2 Fuel Rack LateralMovement ..... .................................................. 14 7.10.3 Abnormal Location of a Fuel Assembly ........................................................... 14 7.11 Interaction w ith Boraflex Racks .............................................................................. 15 8.0 Com puter Files ................................................................................................................ 15 9.0 R esults and Conclusions ............................................................................................ 15 10.0 R eferences ........................................................................................................................ 16 A ppendix A ................................................................................................................................ A-1 A ppendix B ................................................................................................................................. B-I A ppendix C ........................................................................................ o....................................... C-1 A ppendix D ................................................................................................................................ D-1 A ppendix E ................................................................................................................................ E-1 Project No. 1101 Holtec Report No. HI-2012621 Page ii

1.0 INTRODUCTION

This report is the calculation package to support the criticality analysis in the 50.59 report for Nine Mile Point Unit 2 (NMP2) rack installation project. This report describes in detail the criticality analysis performed to ensure the acceptability of the NMP2 fuel to be stored in the new Holtec spent fuel storage racks.

2.0 METHODOLOGY The analytical methodology used in this report consists primarily of using two computer codes to perform the calculations, CASMO-4 [1-4] and MCNP-4A [5]. CASMO-4 was used to perform the in-core burnup calculations and then used to restart the burned fuel assemblies in the standard cold core geometry (SCCG) and in the storage rack. The core multiplication factors calculated in the rack configuration were used to determine the acceptable storage criteria. CASMO-4 was also used to calculate the reactivity effect of manufacturing tolerances. MCNP-4A was used to verify CASMO-4 results for reference cases and to perform large three-dimensional rack calculations. MCNP-4A was also used to determine the effect of eccentric fuel positioning with the rack.

Since CASMO-4 is a two-dimensional deterministic code, the models are infinite in the axial direction. Therefore the core multiplication factor being calculated is actually the infinite multiplication factor, kinf (k~f is used in this report when axial leakage is included in the model).

For a reference case, the CASMO-4 results were compared to MCNP-4A results for verification.

This was necessary for two reasons:

1. Since it is not possible to fully benchmark CASMO-4 against critical experiments, it had to be compared to MCNP-4A, which has been fully benchmarked, to verify the CASMO calculations and determine if a calculational correction factor was necessary.

2. It is standard practice at Holtec to compare criticality results from one code to results from another ifidependent code for the purpose of validation.

A detailed discussion of the results from the MCNP-4A to CASMO-4 comparison is provided in Section 7.3.

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2.1 Code Validation As stated, CASMO-4 was used for burnup calculations during in-core operations and criticality calculations of rack configurations. As proof of its acceptability in this application, CASMO-4 has been verified [3,4] against Monte Carlo calculations and a few critical experiments.

Benchmarking of MCNP-4A against critical experiments has been performed at Holtec. The results of the benchmark calculations, presented in Appendix A, indicate a bias of 0.0009 -

0.00 11 for MCNP-4A over a wide range of compositions and geometries, evaluated at the 95%

probability, 95% confidence level [6]. The MCNP-4A bias and calculational statistics were included in the MCNP-4A to CASMO-4A comparison which is discussed in Section 7.3.

3.0 ACCEPTANCE CRITERIA The Holtec high-density spent fuel storage racks for the Nine Mile Point 2 Nuclear Power Station are designed to assure that the neutron multiplication factor (keff) is equal or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with unborated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations and in manufacturing tolerances, statistically combined, giving assurance that the true keff will be equal to or less than 0.95 with a 95% probability at a 95% confidence level. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under credible abnormal and accident conditions, the reactivity will be maintained less than 0.95. The purpose of the present analysis is to confirm the acceptability of the rack design for the designated fuel assembly designs.

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

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

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

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

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

" ANSI/ANS-8.17-1974, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.

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

USNRC guidelines and the applicable ANSI standards specify that the maximum effective multiplication factor, kfr, including uncertainties, shall be less than or equal to 0.95. The infinite multiplication factor, kinf, is calculated for a radially and axially infinite array, neglecting neutron loss due to leakage from the actual storage rack, and therefore is a higher and more conservative value.

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

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

• Moderator in the spent fuel pool rack is pure, unborated water at a temperature within the design basis range corresponding to the highest reactivity.

• Criticality safety analyses are based upon the infinite multiplication factor (kinf), i.e.,

lattice of storage racks is assumed infinite in all directions. No credit is taken for axial or radial neutron leakage, except in the assessment of certain abnormal or accident conditions where neutron leakage is inherent.

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

• In-core depletion calculations assume conservative operating conditions and include voids due to boiling during in-core BWR operations.

Uniform average enrichments were used for all fuel pins in the fuel assembly, which has been found to be conservative compared to calculations with distributed enrichments.

This is a conservative assumption.

5.0 INPUT DATA 5.1 Fuel Assembly Specifications The design basis fuel assembly, chosen from the fuel assembly designs present at the NMP2 site

[9], is a standard GE-14 assembly with U0 2 rods clad in Zircaloy.

Other designs were also evaluated, as listed below.

" A GE-6/6B design with 62 fuel rods and 2 water rods

• A GE-9B design 8x8 assembly with 60 fuel rods and 1 large central water rod.

• A GE- Il design 9x9 assembly with 74 fuel rods and 2 large water rods.

" A GE-13 design 9x9 assembly with 74 fuel rods and 2 large water rods.

As the GE-Il and GE-13 are identical in regards to physical characteristics that are important to criticality, they are not modeled separately. The GE-I /GE-13 and the GE-14 fuel designs have partial length fuel rods, which would result in a planar region of higher reactivity above the partial length rods, if all of the parameters were the same. For conservatism, this region was used for the design basis calculations for the GE-I /GE-13 and GE-14 assemblies. Design parameters for the five fuel assembly design types considered in this evaluation are summarized in Table 1.

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5.2 Storage Rack Cell Specifications The storage cell characteristics were taken from the following Holtec drawings. The information on these drawings is summarized below.

• Holtec Drawing 3307, Sheet 2, Rev. 2

• Holtec Drawing 3307, Sheet 3, Rev. 1

• Holtec Drawing 3307, Sheet 4, Rev. 1 The tolerances on the box wall stainless steel thickness are taken from ASME standard for sheet material.

The high-density storage rack cells consist of an egg-crate structure, a cell of which is illustrated in Figure 1, with fixed neutron absorber material (Boral) of 0.0216 g/cm 2 nominal dl I IZZ) boron-10 areal density positioned between the fuel assembly storage cells in a 0.085-inch channel. This arrangement provides a nominal center-to-center lattice spacing of 6.18 ]

= inches. Manufacturing tolerances used in evaluating uncertainties in reactivity are indicated in Section 7.2. The 0.075-inch stainless steel box that defines the fuel assembly storage cell has a nominal inside dimension of 5.985 inches. This allows adequate clearance for inserting/removing the fuel assemblies, with or without the Zircaloy flow channel. The Boral panels are 145 inches long, 4.75 inches wide and 0.075 inches thick. Boral panels are used on only one exterior surface of the modules that face each other across the small water gap between the modules.

6.0 COMPUTER CODES In the fuel-rack evaluation, criticality analysis of the high-density spent fuel storage racks were performed with the CASMO-4 [1] code, a two-dimensional multi-group transport code.

Independent verification calculations were made with the MCNP-4A code [5] (a continuous energy Monte Carlo code developed by the Los Alamos National Laboratory).

Benchmark calculations are presented in Appendix A of this report and indicate a bias of 0.0009

+ 0.0011 for MCNP. In the geometric model used in the calculations, each fuel rod and its Project No. 1101 Holtec Report No. HI-2012621 Page 5 Shaded Areas Denote Proprietary Information

cladding were explicitly described and reflecting boundary conditions (zero neutron current) were used in the axial direction and at the equivalent centerline of the Boral and steel plate between storage cells. These boundary conditions have the effect of conservatively creating an infinite array of storage cells in all directions.

The CASMO4 computer code was used as the primary method of analysis as well as a means of evaluating small reactivity increments associated with manufacturing tolerances. Burnup calculations were also performed with CASMO4, using the restart option to describe spent fuel in the storage cell. MCNP-4A was used to assess the reactivity consequences of eccentric fuel positioning and abnormal locations of fuel assemblies that required a three-dimensional model.

7.0 CALCULATIONS This section will describe the calculations that were used to determine the acceptable storage criteria for the BWR racks provided by Holtec International. 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 described below.

As discussed in Section 2.0, CASMO-4 was the primary code used in the calculations. MCNP-4A was used to check CASMO-4 for reference cases and to perform certain calculations which are not possible with CASMO-4 (e.g., eccentric fuel positioning).

Since CASMO-4 is a two-dimensional code, the fuel assembly hardware above and below the active fuel length is not represented. The three-dimensional MCNP models that included axial leakage did not include the fuel assembly hardware in the model. Instead, 30 cm of water was conservatively modeled above and below the active fuel length.

Figure 2 is a plot of the calculational model used in MCNP-4A. Figure 2 was created with the two dimensional plotter in MCNP-4A and clearly indicates the explicit modeling of the fuel rods in each assembly.

The goal of the BWR calculations was to verify that the fuel assemblies listed in Table 1 are acceptable for storage with maximum planar average enrichments less than or equal 4.95 wt%

235U.

Additionally, the limiting kinf in the standard cold core geometry (SCCG) for the Project No. 1101 Holtec Report No. HI-2012621 Page 6 Shaded Areas Denote Proprietary Information

assemblies in Table I must be determined to ensure that the kinf in the rack is less than or equal to 0.95. SCCG is defined as an infinite array of fuel assemblies on a 6 inch lattice spacing at 20' C, without any control absorber or voids.

The general methodology was to perform in-core depletion calculations for these assemblies using CASMO-4 and then analytically restart the assemblies in the rack geometry and in the SCCG geometry. The burnup at the limiting kinf in the SCCG was determined and then the kinf in the rack was calculated at this burnup. Reactivity allowances for manufacturing tolerances, depletion uncertainty and for possible differences with vendor calculations were added to the rack kinf and the result compared to the regulatory limit of 0.95. Since the final kinf was less than the regulatory limit, storage of the assemblies in Table 1, with planar average enrichments less than or equal to 4.95 wt% 235U and with a maximum SCCG kinf less than or equal to 1.32, is acceptable.

7.1 Reference Fuel Assembly The GE-i 1/GE-13 and GE 14 assemblies both contain some partial length fuel rods. Since CASMO-4 is a two-dimensional code it is not possible to properly represent the three-dimensional nature of a fuel assembly which contains partial length rods. Therefore two separate calculations were performed for the GE- 11/GE-13 and the GE-14 assembly types. The first calculation replaces the partial length fuel rods with full length fuel rods. The second calculation removes the partial length fuel rods and replaces them with water.

Even though all the fuel assemblies listed in Table 1 were evaluated for storage acceptability, a reference fuel assembly was determined for miscellaneous calculations (e.g. tolerance calculations). In order to determine the reference assembly, all fuel assemblies were analyzed in the rack configuration at 0 burnup (fresh) and at the burnup corresponding to a kinf in the SCCG of 1.32 for an enrichment of 4.95 wt% 235U. The results of this comparison are presented on Worksheet C.2 in Appendix C. These results indicate that the GE-14 assembly with partial length rods replaced by water is the most reactive assembly either when fresh or at a burnup corresponding to a kinf of 1.32 in the SCCG. Therefore, the GE-14 assembly with the partial length fuel rods replaced by water, was chosen as the reference assembly for the various CASMO-4 analyses.

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7.2 Manufacturing Tolerances and Burnup Uncertainty In the calculation of the final kinf, the effect of manufacturing tolerances on reactivity must be included. CASMO-4 was used to perform these calculations. The reference fuel assembly, with an initial nominal enrichment of 4.95 wt% 235U was used for these studies. To determine the Ak associated with a specific manufacturing tolerance, the reference kinf was compared to the kinf from a calculation with the tolerance included. 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. Only the Ak values in the positive direction (increasing reactivity) were used in the statistical combination.

The following is a list of the manufacturing tolerances that were included.

• U0 2 density - +/- R% of nominal density.

" Cell box ID and Pitch' - +/- - inches.

• Box wall thickness - inches.

• Boral width - modeled at the minimum width (f .

235

• Enrichment - +/---] wt% U.

Other manufacturing tolerances of the fuel assembly such as fuel pin pitch, pellet O.D., clad thickness, etc. have a negligible impact on reactivity and therefore are not included. Worksheet C.3 in Appendix C shows the kinf from the reference case as compared to the kinffrom the cases with the increased and decreased U0 2 and 235U enrichment. Worksheet C.3 also shows the kinf from the reference case as compared to the kinf from the cases with the manufacturing tolerances included. The Ak was calculated for a fresh assembly and at a burnup corresponding to a kinf in the SCCG of 1.32. Conservatively, the largest uncertainty from either the fresh or burned condition was used in the statistical combination of uncertainties.

CASMO-4 was used to perform the depletion calculations. Since there are no depleted fuel critical experiments with which to benchmark CASMO-4's depletion calculations, a reactivity

' As the Cell Box I.D. and the Pitch are interrelated, a change in one of these parameters will necessarily change the other parameter. Therefore, a change in the Box I.D. results in a= change in the pitch.

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allowance for uncertainty in depletion is needed. This uncertainty is statistically combined with the uncertainties from the manufacturing tolerances. Conservatively, 5% of the reactivity decrement from 0 burnup (fresh fuel) to the burnup corresponding to a ki"f in the SCCG of 1.32 is used as the reactivity allowance for depletion. Worksheet C.2 in Appendix C shows the reactivity decrement for the Holtec racks. The largest reactivity decrement for the fuel assemblies analyzed was used in the calculation of the total uncertainty.

Worksheet C.8 in Appendix C shows the final statistical combination of the reactivity uncertainties from the consideration of the manufacturing tolerances and depletion.

7.3 CASMO-4 to MCNP-4A Comparison Since CASMO-4 cannot be benchmarked against a majority of the benchmarked critical experiments, the reference CASMO-4 calculations were compared to the MCNP-4A calculations for validation and to determine if any code to code correction factor is necessary. MCNP-4A has been thoroughly benchmarked against critical experiments as described in Appendix A and is therefore considered the reference code to which reference CASMO-4 'calculations are compared. The CASMO-4 calculations are two-dimensional models, infinite in height with reflecting boundary conditions that simulate an infinite array of assemblies. Additionally, the MCNP-4A models differ slightly from the CASMO-4 model in the representation of the Boral and the rack cell wall, due to the limited geometry capabilities of CASMO-4. For purposes of this comparison, all fuel assembly types were assumed to have a nominal bundle average enrichment of 4.95 wt% 235U with no gadolinium.

The MCNP-4A calculated kinfvalues were combined with the bias (from Appendix A) and the calculational statistics for the comparison. The MCNP calculational statistics, for MCNP-4A, were statistically combined with the bias uncertainty from Appendix A.

The results of the comparison are presented on Worksheet C.5 in Appendix C. These results show that the CASMO-4 models are slightly non-conservative when compared to the MCNP-4a models. Therefore, the maximum positive correction factor is applied to determine the maximum krff on Worksheet C.8 and C.9 in Appendix C.

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7.4 Temperature Effect The effect on reactivity of increasing the spent fuel pool temperature was evaluated using CASMO-4. The results are presented on Worksheet C.4 in Appendix C. The highest temperature evaluated was 1200 C (2480 F). A case including 10% void was also evaluated at this temperature in order to simulate boiling at the bottom of the spent fuel pool. These results clearly indicate that the spent fuel pool temperature coefficient of reactivity is always negative.

7.5 Effect of the Channel and Eccentric Fuel Positioning 7.5.1 Channel Removal and Channel Thickness The BWR fuel assemblies usually have a zircaloy channel attached to the fuel bundle. However, it can not be guaranteed that this channel will be present during storage. Therefore, MCNP-4A calculations were performed to verify that including the channel in the final analysis in conservative. The results of this study are presented on Worksheet C.I in Appendix C.

Additionally, the channels do not have a uniform thickness around the entire assembly. Rather, the channels are typically thinner on the sides and thicker on the corners. To reduce the complexity of the model, the MCNP-4a and CASMO-4 models assume that the channels are uniformly thick and the corners are square (rather than rounded). To ensure that the models are conservative, the effect of the channel thickness on reactivity was determined. The results of this study are on Worksheet C.1 in Appendix C and show that by modeling the channel with a uniform maximum thickness, the results are conservative.

7.5.2 Channel Bulging Another possible reactivity effect results from the potential bulging of the zircaloy channel, which moves the channel wall outward toward the Boral absorber. It was conservatively assumed that the maximum bulging that could occur wouldresult in the bowed channel touching the cell walls. Since this would not occur over the entire length of the channel, the model assumed that the entire channel was enlarged so that the mid-point of the channel wall was placed equidistant between the channel outer dimension and the cell wall. The calculations to determine the channel I.D. and O.D., while preserving the volume of Zircaloy, are shown on Worksheet C.7 in Appendix C. This reactivity effect was evaluated using MCNP-4a and comparison with the reference case (no channel bulging) is shown on Worksheet C. 1 in Project No. 1101 Holtec Report No. HI-2012621 Page 10 Shaded Areas Denote Proprietary Information

Appendix C. The MCNP-4A calculation was performed using the GE-14 assembly with the partial length fuel rods replaced by water. This positive reactivity allowance is conservatively added to the calculated kinf instead of being statistically combined with the other reactivity uncertainties.

7.5.3 Eccentric Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell and in the BWR rack there are bottom fittings and spacers that mechanically restrict lateral movement of the fuel assemblies. Nevertheless, MCNP-4A calculations were made with the fuel assemblies assumed to be in the corner of the storage rack cell. These calculations indicate that eccentric positioning results in a negligible increase in reactivity as shown on Worksheet C.I in Appendix C. The highest reactivity, therefore, corresponds to the reference design with the fuel assemblies positioned at the center of the storage cells.

7.6 Comparison to Vendor Calculations CASMO-4 was used to perform depletion calculations and to calculate the kinf in the SCCG.

Since the fuel vendor also calculates the kinf in the SCCG with potentially different analytical techniques, a reactivity allowance for comparisons between vendor calculations and Holtec calculations is applied. Conservatively, a flat reactivity allowance of 0.01 Ak is used. This value is not statistically combined with the other uncertainties but rather added directly to the calculated kinf. The allowance also is used to encompass any potential differences between the SCCG calculations performed here and by the vendor.

7.7 Boral Height Reduction MCNP-4A calculations were performed in which the Boral height was lowered

• inches below the top of the active fuel (which is identical to raising the Boral height above the bottom of the active fuel). These calculation are infinite in the lateral direction and therefore are highly conservative. The results of these calculations are presented on Worksheet C. 1 in Appendix C and show that extending the active fuel region of a fuel assembly above or below the Boral (up to

] inches) has a negligible impact on the reactivity of the rack.

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7.8 Final Calculations 7.8.1 Maximum kinf in the Standard Cold Core Geometry It is conventional practice for the fuel vendor to, in developing a specific assembly design, to provide values for kinf in the SCCG for each planar (axial).region of significantly different compositions or arrangements. These kinf values are provided at 0% void (core inlet), 40% void (core average), and 70% void (exit condition). The 40% core average is the most meaningful since the 0% and 70% void calculations are applicable only to the ends of the assemblies (small volume and high leakage). The initial design Gd 2 0 3 loading enters into the fuel vendor's calculations of the burnup at which the peak reactivity occurs. At this burnup, the gadolinium is essentially depleted. Consequently, calculations of the reactivity in the storage rack do not need to include gadolinium, but only the average enrichment. Calculations are provided herein illustrating this fact and correlating the kinf in the storage rack to the vendor supplied kinf in the SCCG. Figure 3 illustrates the variation in reactivity in the storage rack with values of the kinf in the SCCG.

In order to confirm that the NMP2 fuel is acceptable to be stored in the new Holtec spent fuel racks, the maximum kinf in the SCCG that will limit the kinf in the rack to less than or equal to 0.95 had to be determined. This was performed for each fuel assembly listed in Table 1 and the results are provided on Worksheet C.8 in Appendix C. As discussed earlier, the GE-I l/GE-13 and the GE-14 assemblies were analyzed twice to account for the partial length rods.

The statistically combined reactivity allowances for manufacturing tolerances and depletion, the effect of channel bulging, the correction factor from the MCNP-4A and CASMO-4 comparison and the 0.01 Ak allowance for comparison to vendor calculations are included. Table 2 summarizes the results from Worksheet C.8 of Appendix C and demonstrates that by limiting the kinf in the SCCG to 1.32 for the Nine Mile Point 2 fuel assemblies the kinf in the new Holtec spent fuel storage racks will be less then 0.95.

7.8.2 Criteria For Minimum Gadolinium Loading Gadolinia (Gd 2 0 3 ) is normally used in BWR fuel to augment reactivity control during in-core operation. A very wide variety of Gd2 0 3 loading are commonly used - often differing in planar Project No. 1101 Holtec Report No. HI-2012621 Page 12 Shaded Areas Denote Proprietary Information

(axial) regions. Furthermore, the Gd 20 3 loadings for fuel of 4.95 wt% nominal initial enrichment have not yet been developed. However, it is possible to develop and define criteria for the minimum Gd 2 0 3 loadings required to assure that the peak reactivity over burnup is always less than the regulatory limit of 0.95.

Gadolinium has a higher cross-section than 235U and the reactivity of an assembly increases with burnup, reaching a maximum at some point in burnup where the Gadolinium is virtually depleted. For fuel with an initial nominal enrichment of 4.95 wt%, Figure 4 illustrates the reactivity variation with burnup for several illustrative gadolinia loadings, with the reference fuel assembly, evaluated in the spent fuel storage racks (without bias or uncertainties). Some of these example fuel assemblies would be acceptable for storage and some would not. Also shown in Figure 4 is the calculation without any Gd 20 3. The maximum reactivity fromeach of the calculations in Figure 4 is determined and plotted versus the initial gadolinia loading in Figure 5.

The minimum loading is calculated assuming a total of either 6 or 8 rods contain Gd 20 3. The resulting minimum loading is determined by a linear interpolation on Worksheet C.6 in Appendix C. These results are summarized in Table 3, and indicate that a minimum loading of 4.2% in 6 rods is required to ensure that kinf in the storage rack is less than or equal to 0.95.

7.9 Long Term Reactivity Changes At reactor shutdown, the reactivity of the fuel initially-decreases due to the growth of 135 Xe, from 1351 decay. Subsequently, the Xenon decays and the reactivity increases to a maximum at several hundred hours when the Xenon is gone. Over the next 30 years, the reactivity continuously 24 1 24 1 decreases due primarily to Pu decay and Am growth. At lower burnup, the reactivity decrease will be less pronounced since less 24 1 pu would have been produced. No credit is taken for this long-term decrease in reactivity other than to indicate additional and increasing conservatism in the design criticality analysis.

7.10 Abnormal and Accident Conditions 7.10.1 Dropped Fuel Assembly For a drop on top of the rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the active fuel region of more then 12 inches, which is sufficient to preclude neutron coupling (i.e. an effectively infinite separation). Maximum Project No. 1101 Holtec Report No. HI-2012621 Page 13 Shaded Areas Denote Proprietary Information

expected deformation under seismic or accident conditions will not reduce the minimum spacing to less than 12 inches.

It is also possible to vertically drop an assembly into a location occupied by another assembly.

Such a vertical impact would at most cause a small compression of the stored assembly, reducing the water-to-fuel ratio and thereby reducing reactivity. In addition, the distance between the active fuel regions of both assemblies will be more than sufficient to ensure no neutron interaction between the two assemblies.

The last scenario is the drop of a fuel assembly into an open storage cell. The dropped assembly would impact the baseplate and could result in a localized deformation of the baseplate that would affect that storage cell and the cells immediately surrounding it. A conservatively bounding analysis in Reference [8] has shown that the deformation could be approximately 2.5 inches. The consequence of this drop accident on criticality is that the active fuel length of that fuel assembly, and possibly the surrounding assemblies, could extend below the Boral panels.

The conservative analysis presented in Section 7.7 has demonstrated that the consequences of this accident are negligible.

7.10.2 Fuel Rack Lateral Movement Boral panels are installed in the rack wall along one side of the water gap between adjacent racks. With this configuration, the maximum reactivity of the storage rack is not dependent upon the water gap spacing between modules. Thus, misalignment of the racks or seismically induced movement will not affect the reactivity of the rack. The reactivity effects of abnormal and accident conditions are summarized in Table 2.2.

7.10.3 Abnormal Location of a Fuel Assembly It is hypothetically possible to suspend a fuel assembly of the highest allowable reactivity outside and adjacent to the fuel rack, although such an accident condition is highly unlikely. The exterior walls of the rack modules facing the outside (where such an accident condition might be conceivable) is a region of high neutron leakage. Additionally, all outer rack surfaces where such an assembly could be suspended, with the exception of the west wall of Rack J, have Boral panels. For comparison to the reference kinf, calculations were performed for the condition of a fuel assembly suspended directly west of Rack J, near the gap between Rack I and Rack J.

Models were created both with and without an extraneous fuel assembly present. Calculations Project No. 1101 Holtec Report No. HI-2012621 Page 14 Shaded Areas Denote Proprietary Information

were performed with the GE-14 fuel assembly at both 3.2 wt% and 4.95 wt%. Descriptions of the models and the reactivity effect of a misplaced assembly is detailed on Worksheet C. 1 in Appendix C. With neutron leakage included, the kff with an extraneous fuel assembly of the maximum reactivity, located outside and adjacent to the fuel rack, is significantly less than the MCNP-4a kinf. Additionally, comparison of the kff of an extraneous fuel assembly outside and adjacent to the fuel rack, to the rffof the same model, without the extraneous fuel assembly indicates that such an accident has a negligible effect on the reactivity.

7.11 Interactionwith Boraflex Racks There is no direct interaction between the Boraflex and Boral racks that would effect the limitations of either type of rack. Boral panels are installed along all Holtec rack walls facing the existing Boraflex racks and the water gap between the Boraflex and Boral racks are at least 2.5" to preclude neutron coupling between the racks. With this configuration, the maximum reactivity of the storage rack is not dependent upon the interaction of the Boraflex and Boral racks.

8.0 COMPUTER FILES A list of the file names and a brief description of the calculations that were performed for this analysis is provided in Appendix D. All related computer files are stored on the computer server at the Holtec International office in Marlton, NJ. The files are stored in the following directory:

G:\PROJECTS\1 101\KWC.

9.0 RESULTS AND CONCLUSIONS The fuel assembly used as the principal design basis for the racks is the GE-14 (lOxlO),

containing U0 2 fuel rods clad in Zircaloy, and using uniform nominal initial enrichments up to 4.95 wt% 235U. Explicit analyses of all other fuel assembly types were performed to confirm their acceptability for storage in the high-density racks. The effects of calculational and manufacturing tolerances were evaluated and added in determining the maximum kerr in the storage rack.

Project No. 1101 Holtec Report No. HI-2012621 Page 15 Shaded Areas Denote Proprietary Information

In BWR fuel, there is a wide variety of designs, including enrichment distribution and gadolinia loading, which often vary in both the axial and radial directions. Two different criteria which bound fuel acceptable for safe storage are defined below.

1. A maximum nominal enrichment of 4.95 wt% 235U, with a maximum planar kinf in the standard cold core geometry (SCCG) of 1.32, where the SCCG is defined as the multiplication factor (kinf) for an infinite array of fuel assemblies on a 6-inch lattice spacing, at 200 C without voids or control rods.

2. A maximum nominal enrichment for the GE-14 assembly of 4.95 wt% 235U with a minimum enrichment of 4.2 wt% Gd 20 3 in at least 6 rods These criteria are discussed more fully in subsequent paragraphs of this report. Either of these criteria is sufficient to determine the acceptability of fuel for safe storage in the spent fuel racks.

These criteria should be applied to the axial (planar) region of highest reactivity. Each planar region should be separately evaluated to assure that the planar region of highest reactivity is assessed.

The basic calculations supporting the criticality safety of the Nine Mile Point 2 fuel storage racks are summarized in Tables 2 and 3.

Abnormal and accident conditions were also evaluated. None of the abnormal or accident conditions that have been identified as credible will result in exceeding the limiting reactivity (k~ff of 0.95). The effects on reactivity of credible abnormal and accident conditions are summarized on Worksheet C. 1 in Appendix C. The double contingency principle of ANSI 16.1-1975 (and the USNRC letter of April 1978) specifies that it shall require at least two unlikely independent and concurrent events to produce a criticality accident. This principle precludes consideration of the simultaneous occurrence of multiple accident conditions. Other hypothetical events were considered and no credible occurrences or configurations have been identified that might have any adverse effect on the storage rack criticality safety.

10.0 REFERENCES

Project No. 1101 Holtec Report No. HI-2012621 Page 16 Shaded Areas Denote Proprietary Information

[1] M. Edenius, et al., "CASMO-4, A Fuel Assembly Burnup Program, User Manual",

Studsvik/SOA-95/1, Studsvik of America, Inc., and Studsvik Core Analysis AB (proprietary).

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

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

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

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

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

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

[8] "Analysis of Drop Accidents for the Nine Mile Point Nuclear Station Spent Fuel Storage Racks.", Holtec Report No. HI-2002584, Rev. 0, January 2, 2001.

[9] "Fwd: SFP Data", Email correspondence from Christopher Gass to Pankaj Chaudhary, February 15, 2001.

Project No. 1101 Holtec Report No. HI-2012621 Page 17 Shaded Areas Denote Proprietary Information

Table I BWR Fuel Characteristics Project No. 1101 Holtec Report No. HI-2012621 Page 18 Shaded Areas Denote Proprietary Information

Table 2 Summary of Criticality Safety Calculations For Limiting kinf in the SCCG Fuel Assembly GE-6/6B GE-9B GE- 11/GE-13 GE-14 Temperature 4 0C 40 C 40 C 4 0C SCCG kinf < 1.32 < 1.32 < 1.32 < 1.32 Enrichment (235U) 4.95 wt% 4.95 wt% 4.95 wt% 4.95 wt%

CASMO-4 kinf 0.9090 0.9107 0.9172 0.9202 Uncertainties B-10 Loading 0.0052 U0 2 density 0.0027 Enrichment 0.0023 Box I.D./Pitch 0.0027 Wall Thickness 0.0003 Total Uncertainty 0.0069 0.0069 0.0069 0.0069 Depletion 0.0052 0.0052 0.0052 0.0052 Channel Bulging 0.0052 0.0052 0.0052 0.0052 Vendor Comparison 0.0100 0.0100 0.0100 0.0100 Correction Factor 0.0028 0.0028 0.0028 0.0028 Maximum kinf 0.9356 0.9373 0.9438 0.9468 Regulatory Limit 0.9500 0.9500 0.9500 0.9500 Project No. 1101 Holtec Report No. HI-2012621 Page 19 Shaded Areas Denote Proprietary Information

Table 3 Summary of Criticality Safety Calculations For Minimum Gd 203 Fuel Assembly GE-14 Temperature 4 0C Enrichment (235U) 4.95 wt%

Min. Gd20 3 Loading 4.2 wt%

Reference 0.9210 CASMO-4 kinf Uncertainties B-10 Loading 0.0052 U0 2 density 0.0027 Enrichment 0.0023 Box I.D./Pitch 0.0027 Wall Thickness 0.0003 Total Uncertainty 0.0069 Depletion 0.0052 Channel Bulging 0.0052 Vendor Comparison 0.0100 Correction Factor 0.0028 Maximum kinf 0.9476 Regulatory Limit 0.9500 Project No. 1101 Holtec Report No. HI-2012621 Page 20 Shaded Areas Denote Proprietary Information

.B" A So.

REFLECTIVE B[IUNOARY CONDITION I--

z Li z ui u

LU

-J l-z U- uJ LU U-

-J 5,gB5" SO REFLECTIVE BOUNDARY CONDITION DETAIL "All (BOX WALL THICKNESS + SHEATHING THICKNESS)/2

=(0,075 + 0,035)/2 = 0.055" I

fl""'-""" I I I-------------

I II-------- i I I--- I (BORAL CHANNEL THICKNESS - BORAL THICKNESS)/2 (BORAL THICKNESS )/2 =(0.085 - 0.075)/2 = 0,005"

=(0.075)/2 = 0.0375' DETAIL "A" Project No. 1101 Holtec Report No. HI-2012621 Page 21 Shaded Areas Denote Proprietary Information

Figure 1: A Cross-Sectional View of the Calculational Model Used For BWR Rack Analysis (Not to Scale)

Figure 2: A Two Dimensional Representation of the Actual Calculational Model Used For the BWR Rack Analysis. This Figure was Drawn (To Scale) with the Two-Dimensional Plotter in MCNP4A.

Project No. 1101 Holtec Report No. HI-2012621 Page 22 Shaded Areas Denote Proprietary Information

I

' 0.9 GE..6i6B

-'*-GE9B forpart.

GE 11/13 w/h2o or pa . rt 0.85--#-- / 13 w/f uh lf

- * - GE 1 forpart.

GEI4 w/fuel

---

1.45

-- 1.4

-- L 13 ---

1.25 Geometry 1.2 0.83 Cold Core k-inf in Standard Designs Rack for NMp2 Fuel and kinf in the Storage of the k10 f in the SCCG Figure 3: Correlation Project No.I0 Inorato HotcR pToritt Information Proprietay Areas Denote Shaded

or~

0 0 2 4 6 8 10 12 14 16 18 20 Bumup (GWD/MTU)

Figure 4: Kinf in the Storage Rack at Various Fuel Burnups and Gadolinia Loadings Project No. 1101 Holtec Report No. HI-2012621 Page 24 Shaded Areas Denote Proprietary Information

0.9600 0.9500 0.9400

• 0.9300 J 0.9200 U

0.9100 0.9000 0.8900 2.0% 2.5% 3.0% 3.5% 4.0% 4.5% 5.0%

Gadolina Content per Rod wt%

Figure 5: Correlation of the kinf in the Storage Rack and the kinf in the SCCG Project No. 1101 Holtec Report No. HI-2012621 Page 25 Shaded Areas Denote Proprietary Information

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

(this appendix was taken from a different report and because of this the next page is labeled Appendix 4A, Page 1)

Project No. 1101 Holtec Report No. HI-2012621 Page A- I Shaded Areas Denote Proprietary Information

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

SUMMARY

Benchmark calculations have been made on selected critical experiments, chosen, in so far as possible, to bound the range of variables in the rack designs. Two independent methods of analysis were used, differing in cross section libraries and in the treatment of the cross sections. MCNP4a [4A. 1] is a continuous energy Monte Carlo code and KENO5a [4A.2]

uses group-dependent cross sections. For the KENO5a analyses reported here, the 238-group library was chosen, processed through the NITAWL-II [4A.2] program to create a working library and to account for resonance self-shielding in uranium-238 (Nordheim integral treatment). The 238 group library was chosen to avoid or minimize the errorst (trends) that have been reported (e.g., [4A.3 through 4A.5]) for calculations with collapsed cross section sets.

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

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

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

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

Holtec International Appendix 4A, Page I

Figures 4A. 1 and 4A.2 show the calculated kfn for the benchmark critical experiments as a function of the EALF for MCNP4a and KENO5a, respectively (U0 2 fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental effort in performing the critical experiments within each laboratory, as well as between the various testing laboratories. The B&W critical experiments show a larger experimental error than the PNL criticals. This would be expected since the B&W criticals encompass a greater range of critical parameters than the PNL criticals.

Linear regression analysis of the data in Figures 4A. 1 and 4A.2 show that there are no trends, as evidenced by very low values of the correlation coefficient (0.13 for MCNP4a and 0.21 for KENO5a). The total bias (systematic error, or mean of the deviation from a kff of exactly 1.000) for the two methods of analysis are shown in the table below.

Calculational Bias of MCNP4a and KENO5a MCNP4a 0.0009+/-=0.0011 KENO5a 0.0030+/-0.0012 The bias and standard error of the bias were derived directly from the calculated ke.r values in Table 4A. 1 using the following equations tt , with the standard error multiplied by the one-sided K-factor for 95 % probability at the 95 % confidence level from NBS Handbook 91 [4A. 18] (for the number of cases analyzed, the K-factor is -2.05 or slightly more than 2).

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

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

[4A.6] (or the MCNP4a manual) and is the same methodology used in MCNP4a and in KENO5a.

Appendix 4A, Page 2 International Holtec Intemational Appendix 4A, Page 2

E 1- _(E k,)2 /n 2 t=i i=1 (4A.2)

S-=n (n-1)

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

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

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

The bias values are used to evaluate Ihe maximum krff values for the rack designs.

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

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

Appendix 4A, Page 3 International Holtec International Appendix 4A, Page 3

As further confirmation of the absence of any trends with enrichment, a typical configuration was calculated with both MCNP4a and KENO5a for various enrichments.

The cross-comparison of calculations with codes of comparable sophistication is suggested in Reg. Guide 3.41. Results of this comparison, shown in Table 4A.2 and Figure 4A.5, confirm no significant difference in the calculated values of k. for the two independent codes as evidenced by the 45' slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence of an enrichment effect (trend) in the bias.

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

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

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

To further confirm the absence of a significant trend with 10B concentration in the absorber, a cross-comparison was made with MCNP4a and KENO5a (as suggested in Reg.

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

These data substantiate the absence of any error (trend) in either of the two codes for the conditions analyzed (data points fall on a 45

• line, within an expected 95 % probability limit).

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

Holtec International Appendix 4A, Page 4

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

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

4A.4.3 Soluble Boron Concentration Effects Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KENO5a) calculations are shown in Table 4A.6. Analyses of the very high boron concentration experiments (> 1300 ppm) show a tendency to slightly overpredict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the racks with higher soluble boron concentrations could be slightly conservative.

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

Holtcc International Appendix 4A, Page 5

4A.5 MQXFul The number of critical experiments with PuO2 bearing fuel (MOX) is more limited than for UO2 fuel. However, a number of MOX critical experiments have been analyzed and the results are shown in Table 4A.7. Results of these analyses are generally above a Klq of 1.00, indicating that when Pu is present, both MCNP4a and KENO5a overpredict the reactivity. This may indicate that calculation for MOX fuel will be expected to be conservative, especially with MCNP4a. It may be noted that for the larger lattice spacings, the KENO5a calculated reactivities are below 1.00, suggesting that a small trend may exist with KENO5a. It is also possible that the overprediction in k, for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and Am-241 growth. This possibility is supported by the consistency in calculated kfn over a wide range of the spectral index (energy of the average lethargy causing fission).

Appendix 4A, Page 6 Holtec International Intemational Appendix 4A, Page 6

4A.6 Reference

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

[4A.2] SCALE 4.3, "A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", NUREG-0200 (ORNL-NUREG-CSD-2/U2/R5, Revision 5, Oak Ridge National Laboratory, September 1995.

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

Oak Ridge National Laboratory, September 1994.

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

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

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

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

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

[4A.9] L.W. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-1810, Babcock and Wilcox Company, April 1984.

Appendix 4A, Page 7 Holtec International Intemational Appendix 4A, Page 7

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

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

[4A. 11] S.R. Bierman and E.D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o 235 U Enriched UO 2 Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981.

[4A. 12] S.R. Bierman et al., Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o 23U Enriched U0 2 Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December, 1981.

[4A. 13] S.R. Bierman et al., Critical Separation Between Subcritical Clusters of 4.31 w/o 235U Enriched U0 2 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977.

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

[4A. 15] B.M. Durst et al., Critical Experiments with 4.31 wt % 23 `U Enriched U0 2 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982.

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

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

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

Appendix 4A, Page 8 International Holtec International Appendix 4A, Page 8

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k ., --EALIE t (MV Reference Identification Enrich. MCNP4a KENO5a MCNP4a IKENO5a 1 B&W-1484 (4A.7) Core I 2.46 0.9964 +/- 0.0010 0.9898+/- 0.0006 0.1759 0.1753 2 B&W-1484 (4A.7) Core II 2.46 1.0008 +/- 0.0011 1.0015 +/- 0.0005 0.2553 0.2446 3 B&W-1484 (4A.7) Core In 2.46 1.0010 +/- 0.0012 1.0005 +/- 0.0005 0.1999 0.1939 4 B&W-1484 (4A.7) Core IX 2.46 0.9956 +/- 0.0012 0.9901 +/- 0.0006 0.1422 0.1426 5 B&W-1484 (4A.7) Core X 2.46 0.9980 +/- 0.0014 0.9922 +/- 0.0006 0.1513 0.1499 6 B&W-1484 (4A.7) Core XI 2.46 0.9978 +/- 0.0012 1.0005 +/- 0.0005 0.2031 0.1947 7 B&W-1484 (4A.7) Core XII 2.46 0.9988 +/- 0.0011 0.9978 +/- 0.0006 0.1718 0.1662 8 B&W-1484 (4A.7) Core XIII 2.46 1.0020 +/- 0.0010 0.9952 +/- 0.0006 0.1988 0.1965 9 B&W-1484 (4A.7) Core XIV 2.46 0.9953 +/- 0.0011 0.9928 +/- 0.0006 0.2022 0.1986 10 B&W-1484 (4A.7) Core XV" 2.46 0.9910*+/- 0.0011 0.9909 +/- 0.0006 0.2092 0.2014 11 B&W-1484 (4A.7) Core XVI t 2.46 0.9935 +/- 0.0010 0.9889 +/- 0.0006 0.1757 0.1713 12 B&W-1484 (4A.7) Core XVII 2.46 0.9962 +/- 0.0012 0.9942 +/- 0.0005 0.2083 0.2021 13 B&W-1484 (4A.7) Core XVIII 2.46 1.0036 +/- 0.0012 0.9931 +/- 0.0006 0.1705 0.1708 Holtec Intemational Appendix 4A, Page 9

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k._ YALF ' (eV1 Reference Identification Enrich. MCNP4a KEN05a MCNP4a KENO5a 14 B&W-1484 (4A.7) Core XIX 2.46 0.9961 +/- 0.0012 0.9971 +/- 0.0005 0.2103 0.2011 15 B&W-1484 (4A.7) Core XX 2.46 1.0008 +/- 0.0011 0.9932 +/- 0.0006 0.1724 0.1701 16 B&W-1484 (4A.7) Core XXI 2.46 0.9994 +/- 0.0010 0.9918 +/- 0.0006 0.1544 0.1536 17 B&W-1645 (4A.8) S-type Fuel, w1886 ppm B 2.46 0.9970 +/- 0.0010 0.9924 +/- 0.0006 1.4475 1.4680 18 B&W-1645 (4A.8) S-type Fuel, w/746 ppm B 2.46 0.9990 +/- 0.0010 0.9913 +/- 0.0006 1.5463 1.5660 19 B&W-1645 (4A.8) SO-type Fuel, w/1156 ppm B 2.46 0.9972 +/- 0.0009 0.9949 +/- 0.0005 0.4241 0.4331 20 B&W-1810 (4A.9) Case 1 1337 ppm B 2.46 1.0023 +/- 0.0010 NC 0.1531 NC 21 B&W-1810 (4A.9) Case 12 1899 ppm B 2.46/4.02 1.0060 +/- 0.0009 NC 0.4493 NC 22 French (4A.10) Water Moderator 0 gap 4.75 0.9966 +/- 0.0013 NC 0.2172 NC 23 French (4A.10) Water Moderator 2.5 cm gap 4.75 0.9952 +/- 0.0012 NC 0.1778 NC 24 French (4A.10) Water Moderator 5 cm gap 4.75 0.9943 +/- 0.0010 NC 0.1677 NC 25 French (4A.10) Water Moderator 10 cm gap 4.75 0.9979 +/- 0.0010 NC 0.1736 NC 26 PNL-3602 (4A.11) Steel Reflector, 0 separation 2.35 NC 1.0004 +/- 0.0006 NC 0.1018 Holtec Intemational Appendix 4A, Page 10

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k-- EALFt 03 Enrich. MCNP4a KENO5a MCNP4a KENO5a Reference Identification 2.35 0.9980 +/- 0.0009 0.9992 +/- 0.0006 0.1000 0.0909 27 PNL-3602 (4A.11) SteelReflector, 1.321 cm sepn.

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

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

4.306 NC 1.0003 +/- 0.0007 NC 0.3282 31 PNL-3602 (4A.11) Steel Reflector, 0 cm sepn.

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

4.306 0.9994 +/- 0.0012 0.9974 +/- 0.0007 0.2911 0.2927 33 PNL-3602 (4A.11) Steel Reflector, 2.616 cm sepn.

4.306 0.9969 +/- 0.0011 0.9951 +/- 0.0007 0.2828 0.2860 34 PNL-3602 (4A.11) Steel Reflector, 5.405 cm sepu.

4.306 0.9910 +/- 0.0020 0.9947 + 0.0007 0.2851 0.2864 35 PNL-3602 (4A.11) Steel Reflector, Infinite sepn.

• 4.306 0.9941 +/- 0.0011 0.9970 +/- 0.0007 0.3135 0.3150 36 PNL-3602 (4A.11) Steel Reflector, with Boral Sheets 4.306 NC 1.0003 +/- 0.0007 NC 0.3159 37 PNL-3926 (4A.12) Lead Reflector, 0 cm sepn.

4.306 1.0025 +/- 0.0011 0.9997 +/- 0.0007 0.3030 0.3044 38 PNI,3926 (4A.12) Lead Reflector, 0.55 cm sepn.

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

Appendix 4A, Page I I Holtec intemational

Table 4A.1 Summary of Criticality Benchmark Calculations 1241 ed kin EALF (eV Reference Identification Enrich. MCNP4a KENO*a MCNP4a KENOSa 40 PNL-3926 (4A.12) Lead Reflector, 5.405 cm sepn. 4.306 0.9971 +/- 0.0012 0.9946 + 0.0007 0.2831 0.2854 41 PNL-2615 (4A.13) Experiment 0041032 - no absorber 4.306 0.9925 +/- 0.0012 0.9950 + 0.0007 0.1155 0.1159 42 PNL-2615 (4A.13) Experiment 030 - Zr plates 4.306 NC 0.9971 +/- 0.0007 NC 0.1154 43 PNL-2615 (4A.13) Experiment 013 - Steel plates 4.306 NC 0.9965 +/- 0.0007 NC 0.1164 44 PNL-2615 (4A.13) Experiment 014 - Steel plates 4.306 NC 0.9972 +/- 0.0007 NC 0.1164 45 PNL-2615 (4A.13) Exp. 009 1.05% Boron-Steel plates 4.306 0.9982 +/- 0.0010 0.9981 +/- 0.0007 0.1172 0.1162 46 PNL-2615 (4A.13) Exp. 012 1.62% Boron-Steel plates 4.306 0.9996 +/- 0.0012 0.9982 +/- 0.0007 0.1161 0.1173 47 PNL-2615 (4A.13) Exp. 031 - Boral plates 4.306 0.9994 +/- 0.0012 0.9969 +/- 0.0007 0.1165 0.1171 48 PNL-7167 (4A.14) Experiment 214R - with flux trap 4.306 0.9991 +/- 0.0011 0.9956 +/- 0.0007 0.3722 0.3812 49 PNL-7167 (4A.14) Experiment 214V3 - with flux trap 4.306 0.9969 +/- 0.0011 0.9963 +/- 0.0007 0.3742 0.3826 50 PNL-4267 (4A.15) Case 173 - 0 ppm B 4.306 0.9974 + 0.0012 NC 0.2893 NC 51 PNL-4267 (4A.15) Case 177 - 2550 ppm B 4.306 1.0057 +/- 0.0010 NC 0.5509 NC 52 PNL-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 21 20% Pu 1.0041 +/- 0.0011 1.0046 +/- 0.0006 0.9171 0.8868 Holtec International Appendix 4A, Page 12

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

Reference Identification Enrich. MCNP4a KENO5a MCNP4a KENOSa 53 PNL-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 43 20% Pu 1.0058 + 0.0012 1.0036 - 0.0006 0.2968 0.2944 54 PNL-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 13 20% Pu 1.0083 0.0011 0.99899 0.0006 0.1665 0.1706 55 PNL,5803 (4A.16) MOX Fuel - Type 3.2 Exp. 32 20% Pu 1.0079 +/- 0.0011 0.9966 +/- 0.0006 0.1139 0.1165 56 WCAP-3385 (4A.17) Saxton Case 52 PuO2 0.52' pitch 6.6% Pu 0.9996,+/- 0.0011 1.0005 +/- 0.0006 0.8665 0.8417 57 WCAP-3385 (4A.17) Saxton Case 52 U 0.52" pitch 5.74 1.0000 +/- 0.0010 0.9956 +/- 0.0007 0.4476 0.4580 58 WCAP-3385 (4A.17) Saxton Case 56 PuO2 0.56" pitch 6.6% Pu 1.0036 +/- 0.0011 1.0047 +/- 0.0006 0.5289 0.5197 59 WCAP-3385 (4A.17) Saxton Case 56 borated PuO2 6.6% Pu 1.0008 +/- 0.0010 NC 0.6389 NC 60 WCAP-3385 (4A.17) Saxton Case 56 U 0.56" pitch 5.74 0.9994 +/- 0.0011 0.9967 +/- 0.0007 0.2923 0.2954 61 WCAP-3385 (4A.17) Saxton Case 79 PuO2 0.79"1 pitch 6.6% Pu 1.0063 +/- 0.0011 1.0133 +/- 0.0006 0.1520 0.1555 62 WCAP-3385 (4A.17) Saxton Case 79 U 0.79" pitch 5.74 1.0039 +/- 0.0011 1.0008 +/- 0.0006 0.1036 0.1047 Notes: NC stands for not calculated.

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

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

Holtc Inerntionl Apendx 4, Pae 1 Holtee International Appendix 4A, Page 13

Table 4A.2 COMPARISON OF MCNP4a AND KENO5a CALCULATED REACTIM1TIESt FOR VARIOUS ENRICHMENTS Calculated keff +/- lo Enrichment MCNP4a KENO5a 3.0 0.8465 +/- 0.0011 0.8478 +/- 0.0004 3.5 0.8820 + 0.0011 0.8841 +/-:0.0004 3.75 0.9019 + 0.0011 0.8987 +/- 0.0004 4.0 0.9132 +/- 0.0010 0.9140 +/- 0.0004 4.2 0.9276 +/- 0.0011 0.9237 +/- 0.0004 4.5 0.9400 +/- 0.0011 0.9388 +/-:0.0004 t Based on the GE x8xR fuel assembly.

Holtec International Appendix 4A, Page 14

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

Holtec International Appendix 4A, Page 15

Table 4A.4 COMPARISON OF MCNP4a AND KENO5a CALCULATED REACTIVITIESt FOR VARIOUS ' 1B LOADINGS Calculated kf +/- lo 10B, g/cm2 MCNP4a KENO5a 0.005 1.0381 +/- 0.0012 1.0340 - 0.0004 0.010 0.9960 +/- 0.0010 0.9941 +/- 0.0004 0.015 0.9727 +/- 0.0009 0.9713 +/- 0.0004 0.020 0.9541 +/- 0.0012 0.9560 +/- 0.0004 0.025 0.9433 + 0.0011 0.9428 + 0.0004 0.03 0.9325 +/- 0.0011 0.9338 +/- 0.0004 0.035 0.9234 +/- 0.0011 0.9251 + 0.0004 0.04 0.9173 +/- 0.0011 0.9179 +/- 0.0004 Based on a 4.5% enriched GE 8x8R fuel assembly.

Holtec International Appendix 4A, Page 16

Table 4A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH THICK LEAD AND STEEL REFLECTORSt Separation, Ref. Case E, wt% cm MCNP4a kff KENO5a k.

4A.11 Steel 2.35 1.321 0.9980+/-0.0009 0.9992+/-0.0006 Reflector 2.35 2.616 0.9968+/-0.0009 0.9964+/-0.0006 2.35 3.912 0.9974+/-0.0010 0.9980+/-0.0006 2.35 00 0.9962+/-0.0008 0.9939+/-0.0006 4A. 11 Steel 4.306 1.321 0.9997+/-0.0010 1.0012+/-0.0007 Reflector 4.306 2.616 0.9994+/-0.0012 0.9974+/-0.0007 4.306 3.405 0.9969+/-0.0011 0.9951+/-0.0007 4.306 0.9910+/-0.0020 0.9947+/-0.0007 4A. 12 Lead 4.306 0.55 1.0025+/-0.0011 0.9997+/-0.0007 Reflector 4.306 1.956 1.0000+/-0.0012 0.9985+/-0.0007 4.306 5.405 0.9971+/-0.0012 0.9946+/-0.0007 t Arranged in order of increasing reflector-fuel spacing.

Holtec Intemational Appendix 4A, Page 17

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

Appendix 4A, Page 18 International Holtec Intemational Appendix 4A, Page 18

Table 4A.7 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNP4a KENO5a Reference Caset lff EALF" kt EALF" PNL-5803 MOX Fuel - Exp. No. 21 1.0041+/-0.0011 0.9171 1.0046+/-0.0006 0.8868

[4A. 161 MOX Fuel - Exp. No. 43 1.0058+/-0.0012 0.2968 1.0036+/-0.0006 0.2944 MOX Fuel - Exp. No. 13 1.0083+/-0.0011 0.1665 0.9989+/-0.0006 0.1706 MOX Fuel - Exp. No. 32 1.0079+/-0.0011 0.1139 0.9966+/-0.0006 0.1165 WCAP- Saxton @ 0.52" pitch 0.9996+/-0.0011 0.8665 1.0005+/-0.0006 0.8417 3385-54

[4A. 171 Saxton @ 0.56" pitch 1.0036+/-0.0011 0.5289 1.0047+/-0.0006 0.5197 Saxton @ 0.56" pitch borated 1.0008+/-0.0010 0.6389 NC NC Saxton @ 0.79" pitch 1.0063+/-0.0011 0.1520 1.0133+/-0.0006 0.1555 Note: NC stands for not calculated t Arranged in order of increasing lattice spacing.

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

Holtec Intemational Appendix 4A, Page 19

-- - - Linear Regression with Correlation Coefficient of 0.13 1.010 1.005 4) 1000 4-4e-

• -:3 1.000 - -

-a)J U

0 0.995 0.990 0.1 Energy of Average Lethargy Causing Fission (Log Scale)

FIGURE 4A.1 MCNP CALCULATED k-eff VALUES for VARIOUS VALUES OF THE SPECTRAL INDEX

Linear Regression with Correlation Coefficient of 0.21 1.010 1.005 a,

a, 1.000 4-W

"-f

-o

"* 0.995 0.990 0.985 0.1 1 Energy of Average Lethargy Causing Fission (Log Scale)

FIGURE 4A.2 KENO5a CALCULATED k-eff VALUES FOR VARIOUS VALUES OF THE SPECTRAL INDEX

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

1.005 a,

U a) 4-a, 1.000 -

-o a,

0

3 U

0 C-)

0.995 -

0.990 m 2.0 2.!

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

Linear Regression with Correlation Coefficient of 0.38 1.010 1.005 0 1.000

-*0.995 0

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

0.94 E 0.92 0 0.90 4-0.88 0.86 0.84 -

0.

MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KENO5A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS

1.04 1.03 1 .02 1.01

'3 z

0 1.00 4m C--

0.99 0.98

-o 0(D 0.

0.96 (D

[I-

-4.- 0.

0 0.94 0.93 0.92 Reactivity Calculated with KENO5a FIGURE 4A.6 COMPARISON OF MCNP AND KENO5a CALCULATIONS FOR VARIOUS BORON-10 AREAL DENSITIES

Appendix B Boral Composition PROPRIETARY Project No. 1101 Holtec Report No. HI-2012621 Page B-1 Shaded Areas Denote Proprietary Information

Appendix C BWR Fuel Calculations (total number of pages: 11 including this page)

Project No. 1101 Holtec Report No. HI-2012621 Page C- I Shaded Areas Denote Proprietary Information

Index of EXCEL Worksheets (EXCEL file Appendix C.xls)

EXCEL SHEET Page Number Worksheet C. 1: Reactivity Effect of Various Conditions C-3 Worksheet C.2: Determination of Reference Assembly, C-4 Depletion Uncertainty and kinf in the SCCG Worksheet C.3: Manufacturing Tolerances C-5 Worksheet C.4: Temperature Effects C-6 Worksheet C.5 MCNP to CASMO Code Comparison C-7 Worksheet C.6: Determination of Minimum Gd 203 C-8 Loading Worksheet C.7: Channel Bulging Calculations C-9 Worksheet C.8: Summary of Criticality Analysis - For C-10 Limiting kinf in the SCCG.

Worksheet C.9: Summary of Criticality Analysis - For C-11 Minimum Gd 20 3 Loading.

Project No. 1101 Holtec Report No. HI-2012621 Page C-2 Shaded Areas Denote Proprietary Information

Worksheet C. 1: Reactivity Effect of Various Conditions Reference GE-14 assembly w. partial rods renlaced by water at 4.95 wt% enrichment and density of 10.631 P/cc MCNP k-inf delta k Fuel Channel Reactivity Effect - 4.95 wt% run in rack Reference Assembly nmp206 1.0219 reference Fuel Channel Removed nmp208 1.0159 -0.0060 Thinner Fuel Channel nmp209 1.0212 -0.0007 MCNP k-inf delta k Channel Bulging Reactivity Effect - 4.95 wt% run in rack Reference GE- 14 nmp206 1.0219 reference With Channel Bulging nmp210 1.0271 0.0052 MCNP k-inf delta k Eccentric positioning - 4.95 wt% run in rack reference fuel assembly 1 nmp206 1.0219 reference reference fuel assembly - eccentric positioning nmp207 1.0221 0.0002 MCNP k-inf delta k Boral Height Reduction - 3.2 wt% run in rack Reference Assembly nmp2l6 0.9152 reference nmp2l1 0.9134 -0.0018 nmp2l2 0.9142 -0.0010 nmp2l3 0.9154 0.0002 MCNP k-inf delta k Misplaced Assembly run in rack from ref.

Infinite Array - Reference Assembly 4.95% _ nmp206 1.0219 reference Rack I and Rack J - Reference (no misplaced assembly) - 4.95% nmp2l5 1.0010 -0.0209 Rack I and Rack J - Reference (with misplaced assembly, with channel) -

4.95% in rack nmp2l4 0.9997 -0.0222 Rack I and Rack J - Reference (with misplaced assembly, no channel) -

4.95% in rack nmp2l9 1.0004 -0.0215 Infinite Array - Reference Assembly 3.2% nmp2l6 0.9152 reference Rack I and Rack J - Reference (no misplaced assembly) - 3.2% in rack nmp2l8 0.8946 -0.0206 Rack I and Rack J - Reference (with misplaced assemblyof 4.95 wt%

enrichment) - 3.2% in rack nmp2l7 0.8955 -0.0196 Holtec Report: HI-2012621 Shaded Areas Denote Proprietary Information Page C-3

Worksheet C.2: Determination of Reference Assembly, Depletion Uncertainty and lf in the SCCG GE-6/6B GE9B GEl 1/13 w/h2o for part. GEl 1/13 w/fuel for part. r GE14 w/h2o for part. GE14 w/fuel for part.

Burnup SCCG Rack SCCG Rack SCCG Rack SCCG Rack SCCG Rack SCCG Rack 0 1.46558 1.00934 1.46346 1.0103 1.46505 1.01978 1.46197 1.01253 1.46716 1.02431 1.46385 1.01602 0.5 1.45339 1.00189 1.46101 1.00874 1.46261 1.0182 1.45952 1.01093 1.46472 1.02274 1.4614 1.01441 1.5 1.44139 0.99422 1.43962 0.99537 1.44162 1.00483 1.43795 0.99749 1.44372 1.00934 1.43948 1.00087 5 1.40904 0.9722 1.40797 0.97365 1.41066 0.98325 1.40598 0.97585 1.41268 0.98773 1.40672 0.97895 7.5 1.3859 0.95595 1.38526 0.95767 1.38829 0.96722 1.38316 0.95989 1.39024 0.97165 1.38345 0.96283 10 1.36307 0.93977 1.3628 0.94156 1.366 0.95106 1.36075 0.94402 1.36786 0.9554 1.36072 0.94684 12.5 1.3406 0.92373 1.34062 0.9256 1.34382 0.93481 1.33877 0.92828 1.3456 0.93908 1.33858 0.93113 15 1.31843 0.90788 1.31866 0.90973 1.32174 0.91851 1.31717 0.91268 1.32344 0.92273 1.31696 0.91565 17.5 1.29645 0.89202 1.29681 0.8939 1.29967 0.9022 1.29585 0.89724 1.30132 0.90636 1.29574 0.90045 20 1.27456 0.87632 1.27499 0.87803 1.2775 0.88579 1.27467 0.88189 1.27914 0.88986 1.2748 0.88536 25 1.23051 0.8448 1.2309 0.84608 1.2325 0.85248 1.23237 0.85115 1.23413 0.85649 1.2333 0.85547 30 1.18592 0.81302 1.1859 0.81367 :1.18607 0.8183 1.18963 0.8203 1.18783 0.82223 1.1919 0.82573 35 1.1406 0.78099 1.13989 0.78077 1.13815 0.78334 1.14637 0.78918 1.14011 0.7873 1.1505 0.79615 40 1.0948 0.74884 1.09299 0.74745 1.08885 0.74754 1.10274 0.75804 1.09112 0.75163 1.10921 0.76676 K-inf in rack 1.32 0.9090 1.32 0.9107 1.32 0.9172 1.32 0.9147 1.32 0.9202 1.32 0.9178 Burnup 14.792 14.824 15.185 14.635 15.375 14.598 Required Design Basis Fuel Assembly Array Type Burnup k-inf k-inf Array Type k-inf k-inf Bumup 5% of GWD/MTU in rack in SCCG in rack in SCCG GWD/MTU reac. Dec.

GE-6/6B 0 1.0093 1.4656 GE-6/6B 0.9090 1.32 14.792 0.0050 GE9B 0 1.0103 1.4635 GE9B 0.9107 1.32 14.824 0.0050 GE 11/13 0 1.0125 1.4620 GEl1/13 0.9147 1.32 14.635 0.0049 GEl 1/13* 0 1.0198 1.4651 GEl 1/13* 0.9172 1.32 15.185 0.0051 GE14 0 1.0160 1.4639 GE14 0.9178 1.32 14.598 0.0049 GE14* 0 1.0243 1.4672 GEl4* 0.9202 1.32 15.375 0.0052 0.0052 J*Note: partial length fuel rods are replaced by waterl Holtec Report: HI-2012621 Shaded Areas Denote Proprietary Information Page C-4

Worksheet C.3: Manufacturing Tolerances NMP2 Racks Enrichment =4.95% no Gadolinium present k-inf in the rack Burnup k-inf max min max min min min GWD/MTU SCCG reference box-id box-id box-wall box-wall boral width boral loading 0 1.4671 1.0256 1.02324 1.02834 1.02593 1.02461 1.032 1.03082 0.1 1.46466 1.02402 1.02168 1.02676 1.02436 1.02312 1.0304 1.02921 0.5 1.4552 1.01818 1.01583 1.02092 1.01853 1.01729 1.02452 1.02334 1 1.44853 1.01383 1.01151 1.01658 1.0142 1.01292 1.02016 1.01901 2 1.43923 1.00756 1.00521 1.01031 1.00794 1.00667 1.01383 1.01269 3 1.43047 1.00152 0.99917 1.00426 1.00189 1.00061 1.00773 1.00659 4 1.42164 0.99531 0.99299 0.99808 0.99567 0.99442 1.00149 1.00036 5 1.41273 0.98904 0.98675 0.99169 0.9894 0.98815 0.99512 0.99401 6 1.40376 0.98261 0.98036 0.98528 0.98298 0.98176 0.98868 0.98761 7 1.39478 0.97616 0.97395 0.97882 0.97651 0.97531 0.98218 0.98111 8 1.3858 0.96969 0.9675 0.97227 0.97005 0.96882 0.97566 0.97459 9 1.37684 0.96319 0.96101 0.96572 0.96353 0.96231 0.96909 0.96806 10 1.36789 0.95667 0.95449 0.95918 0.957 0.95579 0.96255 0.9615 11 1.35897 0.95009 0.94798 0.9526 0.95042 0.94924 0.95597 0.95489 12 1.35006 0.94354 0.94145 0.94602 0.94387 0.94271 0.94933 0.9483 13 1.34116 0.93698 0.93492 0.93944 0.93731 0.93614 0.94272 0.94171 14 1.33228 0.93042 0.92837 0.93286 0.93074 0.9296 0.93611 0.93511 15.384 1.32 0.9213 0.9193 0.9237 0.9217 0.9205 0.9270 0.9260 15 1.32341 0.92386 0.92183 0.92627 0.92419 0.92303 0.92951 0.92851 17.5 1.30136 0.90751 0.90551 0.90986 0.90781 0.9067 0.91304 0.91208 20 1.27918 0.89102 0.88906 0.89328 0.8913 0.89024 0.89643 0.89549 Delta-k values manufacturing tolerance min min min min box-id box wall boral width boral loading Burnup 0 0.0027 0.0003 0.0064 0.0052 15.384 0.0024 0.0003 0.0056 0.0046 Fresh Assemblies k-inf in _

Array Type IDescription rack GE14* reference 1 1.0256 reference (E=4.95% D=10.631 g/cc)

GEI4* D=10.412 1.0228 -0.0028 GEI4* D=10.850 1.0283 0.0027 GE14* E=4.90% 1.0233 -0.0023 GEl4* E=5.00% 1.0279 0.0023 Summary B-10 Loading 0.0052 Lattice Spacing 0.0027 SS Thickness 0.0003 Fuel Enrichment 0.0023 Fuel Density 0.0027 Statistical Combination of tolerance uncertainties 0.0069 1 1 Holtec Report: HI-2012621 Shaded Areas Denote Proprietary Information Page C-5

Worksheet C.4: Temperature Effects Reference GE-14 assembly w. partial rods replaced by water at 4.95 wt% enrichment and density of 10.631 g/cc I I I I III The reactivity effect of temperature and void was determined for fresh fuel with No Gadolinia CASMO degrees degrees fresh delta-k run C F void % k-inf rel to 4C nmp2t01 4 1.0256 reference nmp2t01 20 1.0235 -0.0021 nmp2t01 40 1.0201 -0.0055 nmp2tO0 60 1.0168 -0.0088 nmp2t01 80 1.0125 -0.0131 nmp2t01 100 1.0072 -0.0184 nmp2t01 120 1.0029 -0.0228 nmp2t01 120 10 0.9819 -0.0437 Holtec Report: HI-2012621 Shaded Areas Denote Proprietary Information Page C-6

Worksheet C.5: MCNP to CASMO Code Comparison Comparison of CASMO-4 to MCNP-4A Adjustments for MCNP values Bias from Appendix A 0.0009 Bias uncertainty from Appendix A 0.0011 MCNP Statistics 0.0017 Total Calculational Statistics 0.0020 Fresh Fuel, No Burnup, No Gad MCNP CASMO MCNP CASMO Array Type run run result GE-6/6B nmp201 nmp2d0l 1.0102 0.0008 1.0093 0.0028 GE-9B nmp202 nmp2d02 1.0103 0.0008 1.0103 0.0020 GEl l/GE13 nmp203 nmp2d05 1.0114 0.0007 1.0125 0.0008 GE-14 nmp204 nmp2d06 1.0136 0.0008 1.0160 -0.0004 GEI I/GEI3 nmp205 nmp2dO3 1.0183 0.0008 1.0198 0.0005 GE-14* nmp206 nmp2d04 1.0219 0.0008 1.0243 -0.0004 0.0008 0.0028

• partial rods replaced by water Holtec Report: HI-2012621 Shaded Areas Denote Proprietary Information Page C-7

Worksheet C.6: Determination of Minimum Gd 2O3 Loading 6@2% 6@3% 6@4% 6@5% 8@2% 8@3% 8@4% 8@5%

Burnup 0 0.9208 0.9151 0.9110 0.9077 0.8882 0.8808 0.8754 0.8711 0.1 0.9200 0.9141 0.9099 0.9066 0.8877 0.8800 0.8746 0.8702 1 0.9186 0.9107 0.9054 0.9014 0.8888 0.8785 0.8717 0.8666 2 0.9230 0.9121 0.9054 0.9005 0.8962 0.8822 0.8735 0.8672 3 0.9286 0.9145 0.9061 0.9002 0.9052 0.8870 0.8761 0.8685 4 0.9351 0.9172 0.9070 0.9000 0.9151 0.8922 0.8791 0.8701 5 0.9422 0.9204 0.9082 0.9000 0.9260 0.8980 0.8823 0.8719 6 0.9492 0.9241 0.9097 0.9003 0.9367 0.9044 0.8860 0.8739 7 0.9541 0.9284 0.9115 0.9007 0.9451 0.9116 0.8901 0.8762 8 0.9559 0.9329 0.9137 0.9013 0.9498 0.9193 0.8946 0.8788 9 0.9549 0.9368 0.9163 0.9022 0.9509 0.9262 0.8997 0.8817 10 0.9516 0.9389 0.9192 0.9034 0.9490 0.9311 0.9052 0.8850 11 0.9468 0.9387 0.9219 0.9049 0.9451 0.9332 0.9105 0.8887 12 0.9411 0.9364 0.9236 0.9067 0.9399 0.9326 0.9147 0.8928 13 0.9349 0.9323 0.9236 0.9084 0.9341 0.9298 0.9170 0.8968 14 0.9286 0.9270 0.9216 0.9095 0.9278 0.9253 0.9169 0.9001 15 0.9221 0.9211 0.9180 0.9093 0.9215 0.9199 0.9148 0.9020 16 0.9156 0.9149 0.9131 0.9075 0.9150 0.9139 0.9110 0.9021 17 0.9091 0.9085 0.9074 0.9041 0.9085 0.9076 0.9058 0.9003 18 0.9026 0.9020 0.9013 0.8994 0.9020 0.9011 0.9000 0.8968 19 0.8960 0.8954 0.8949 0.8938 0.8954 0.8946 0.8938 0.8919 20 0.8894 0.8889 0.8884 0.8877 0.8888 0.8881 0.8874 0.8862 22.5 0.8729 0.8724 0.8720 0.8716 0.8724 0.8717 0.8711 0.8704 25 0.8562 0.8557 0.8553 0.8549 0.8556 0.8550 0.8544 0.8539 MAX 0.9559 0.9389 0.9236 0.9095 0.9509 0.9332 0.9170 0.9021 MAX SCCG 1.3659 1.3420 1.3208 1.3013 1.3594 1.3348 1.3123 1.2917 Summary 6 rods Gadolinia Rack content k-inf 2.0% 0.9559 3.0% 0.9389 4.0% 0.9236 5.0% 0.9095 Target: 0.9210 4.17% Min. Loading 8 rods Gadolinia Rack content k-inf 2.0% 0.9509 3.0% 0.9332 4.0% 0.9170 5.0% 0.9021 Target: 0.9210 3.75% Min. Loading_

Holtec Report: HI-2012621 Shaded Areas Denote Proprietary Information Page C-8

Worksheet C.7: Channel Bulging Calculation Channel ID 5.278 Channel OD 5.518 BOX ID 5.985 Midpoint of Channel 5.398 Midpoint between Channel and Box ID 5.6915 Channel Area 1 2.59104 Equation to determine half thickness (5.6915+t)-2-(5.6915It)A2=2.59104 Simplified Equation 22.764t=2.59104 Result t= 0.1138 New Channel ID 5.5777 New Channel OD 5.8053 Confirm Area 2.5913 MCNP Input New Channel ID 7.08365 New Channel OD 1 7.37276 Holtec Report: HI-2012621 Shaded Areas Denote Proprietary Information Page C-9

Worksheet C.8: Summary of Criticality Analysis - For Limiting 'Snf in the SCCG Summary of Criticality Safety Analyses I I I For limiting keff in the Standard Cold Core Geometry 1.32 Temperature used for this analysis 4 C Initial Fuel Enrichment 1 5 Maximum k-inf in SCCG 1.32 Reference CASMO4 k-inf CASMO MCNP GE-6/6B Fuel 0.9090 GE9B Fuel _ 0.9107 GEl I/GEl3 Fuel 0.9147 GE I !/GE 13 Fuel w/ partial rods replaced by water 0.9172 GE14 Fuel 1 0.9178 GE14 Fuel with partial rods replaced by water 0.9202 Uncertainties Removal of flow channel negative Eccentric assembly location negative Tolerances 1 0.0069 Uncertainty in depletion calculations 0.0052 Statistical Combination 0.0086 Effect of Channel Bulge 0.0052 Comparison to Vendor Calculations 0.0100 CASMO to MCNP Correction 0.0028 Total Adjustment 0.0266 Maximum Reactivity BP8x8 Fuel 0.9356 GE9B Fuel 0.9373 GEl l/GEl3 Fuel 0.9413 GEl 1/GE 13 Fuel w/ partial rods replaced by water 0.9438 GE14 FuelI I I 0.9444 GEl4 Fuel with partial rods replaced by water I 1 0.9468 Holtec Report: HI-2012621 Shaded Areas Denote Proprietary Information Page C- 10

Worksheet C.9: Summary of Criticality Analysis - For Minimum Gd20 3 Loading Summary of Criticality Safety Analyses For minimum Gd203 4.17%

Temperature used for this analysis 4C Initial Fuel Enrichment 4.95 GD203 loading % 4.17%

Reference CASMO4 k-inf CASMO GE14 Fuel with partial rods replaced by water 0.9210 Uncertainties Removal of flow channel negative Eccentric assembly location negative Tolerances 0.0069 Uncertainty in depletion calculations 0.0052 Statistical Combination 0.0086 Effect of Channel Bulge 0.0052 Comparison to Vendor Calculations 0.0100 CASMO to MCNP Correction 0.0028 Total Adjustment 0.0266 Maximum Reactivity GE14 Fuel with partial rods replaced by water 0.9476 Holtec Report: HI-2012621 Shaded Areas Denote Proprietary Information Page C-11

Appendix D List of Computer Runs (total number of pages: 3 including this page)

Project No. 1101 Holtec Report No. HI-2012621 Page D-1 Shaded Areas Denote Proprietary Information

Worksheet D. 1: List of Calculations Project: 1101 Report: 2012621 Input ID Code Computer Description BWR CALCULATIONS nmp201 MCNP4a PC 4.95% BP8x8 - reference nmp202 MCNP4a PC 4.95% GE9B - reference nmp203 MCNP4a PC 4.95% GEl I & GE13 - reference partial rods replaced with fuel nmp204 MCNP4a PC 4.95% GEl4 - reference partial rods replaced with fuel nmp205 MCNP4a PC 4.95% GEl 1 & GE13 - reference partial rods replaced with water nmp206 MCNP4a PC 4.95% GE14 - reference, partial rods replaced with water nmp207 MCNP4a PC 4.95% GE14 - reference, eccentric fuel positioning nmp208 MCNP4a PC 4.95% GE14 - reference, fuel channel removed nmp209 MCNP4a PC 4.95% GE14 - reference, thiner channel 0.100" nmp2lO MCNP4a PC 4.95% GE14 - reference, fuel channel bulge included nmp2l1 MCNP4a PC nmp212 MCNP4a PC nmp2l3 MCNP4a PC nmp214 MCNP4a PC 4.95% GE14 - misplaced assembly between Rack I &J, with misplaced assembly nmp2l5 MCNP4a PC 4.95% GE14 - misplaced assembly between Rack I &J, without misplaced assembly nmp2l6 MCNP4a PC 3.2% GEl4 - reference, partial rods replaced with water nmp2l7 MCNP4a PC 3.2% GEl4 - misplaced assembly between Rack I &J, with misplaced assembly nmp218 MCNP4a PC 3.2% GE14 - misplaced assembly between Rack I &J, without misplaced assembly nmp219 MCNP4a PC 4.95% GEl4 - misplaced assembly between Rack I &J, with misplaced assembly - no Channel nmp220 MCNP4a PC 4.95% GEl4 - misplaced assembly between Rack I &J, with misplaced assembly source redistributed nmp221 MCNP4a PC 3.2% GE14 - misplaced assembly between Rack I &J, with misplaced assembly of, 4.95% enrichment nmp2d01 CASMO Dec 500 4.95% BP8x8 depletion, SCCG and rack, 0.100" channel nmp2d02 CASMO Dec 500 4.95% GE9B depletion, SCCG and rack, 0.100" channel nmp2d03 CASMO Dec 500 4.95% GEl 1/13 depletion, SCCG and rack - partial rods replaced by water, 0.120" channel nmp2d04 CASMO Dec 500 4.95% GEl4 depletion, SCCG and rack - partial rods replaced by water, 0.120" channel nmp2d05 CASMO Dec 500 4.95% GEl 1/13 depletion, SCCG and rack - partial rods replaced by fuel, 0.120" channel nmp2d06 CASMO Dec 500 4.95% GE14 depletion, SCCG and rack - partial rods replaced by fuel, 0.120" channel nmp2d07 CASMO Dec 500 4.95% GE14 depletion, SCCG and rack - partial rods replaced by water 6 Gad rods of 2%

nmp2d08 CASMO Dec 500 4.95% GE14 depletion, SCCG and rack - partial rods replaced by water 6 Gad rods of 3%

nmp2d09 CASMO Dec 500 4.95% GE14 depletion, SCCG and rack - partial rods replaced by water 6 Gad rods of 4%

nmp2dl0 CASMO Dec 500 4.95% GE14 depletion, SCCG and rack - partial rods replaced by water 6 Gad rods of 5%

nmp2dl 1 CASMO Dec 500 4.95% GE14 depletion, SCCG and rack - partial rods replaced by water 8 Gad rods of 1 12%

Holtec Report: HI-2012621 Shaded Areas Denote Proprietary Information Page D-2

Worksheet D.1: List of Calculations nmp2dl2 CASMO Dec 500 4.95% GE14 depletion, SCCG and rack - partial rods replaced by water 8 Gad rods of 3%

nmp2dl3 CASMO Dec 500 4.95% GE14 depletion, SCCG and rack - partial rods replaced by water 8 Gad rods of 4%

nmp2dl4 CASMO Dec 500 4.95% GEl4 depletion, SCCG and rack - partial rods replaced by water 8 Gad rods of 5%

nmp2tOl CASMO Dec 500 4.95% GE14 depletion, SCCG and rack - partial rods replaced by water, 0.120" 1 1_

_ channel calculation of manufacturing tolerances and temperature effects Holtec Report: HI-2012621 Shaded Areas Denote Proprietary Information Page D-3

Appendix E Email Correspondence PROPRIETARY Project No. 1101 Holtec Report No. HI-2012621 Page E- 1 Shaded Areas Denote Proprietary Information

ATTACHMENT 3 AFFIDAVIT FROM HOLTEC JUSTIFYING WITHHOLDING PROPRIETARY INFORMATION Nine Mile Point Nuclear Station, LLC December 13, 2010

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 I, Savit Sinha, being duly sworn, depose and state as follows:

(1) I have reviewed the information described in paragraph (2) which is sought to be withheld, and am authorized to apply for its withholding.

(2) The information sought to be withheld is the Attachment 4 and 5 of Letter from Nine Mile Point to Document Control Desk (NRC) dated December 13, 2010:

"Response to Request for Additional Information Regarding Nine Mile Point Nuclear' Station, Unit No. 2 - Re: The License Amendment Request for Extended Power Uprate Operation (TAC No. ME1476) - Spent Fuel Pool Criticality Analysis" (3) In making this application for withholding of proprietary information of which it is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10CFR Part 9.17(a)(4), 2.390(a)(4), and 2.390(b)(1) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential commercial information", and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992),

and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir.

1983).

1 of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Holtec's competitors without license from Holtec International constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies of Holtec International, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holtec International;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4.a and 4.b above.

(5) The information sought to be withheld is being submitted to the NRC in confidence. The information (including that compiled from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary 2 of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge.

Access to such documents within Holtec International is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation.

Disclosures outside Holtec International are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information classified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is classified as proprietary because it contains detailed descriptions of analytical approaches and methodologies not available elsewhere. This information would provide other parties, including competitors, with information from Holtec International's technical database and the results of evaluations performed by Holtec International. A substantial effort has been expended by Holtec International to develop this information. Release of this information would improve a competitor's position because it would enable Holtec's competitor to copy our technology and offer it for sale in competition with our company, causing us financial injury.

3 of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Holtec International's comprehensive spent fuel storage technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology, and includes development of the expertise to determine and apply the appropriate evaluation process.

The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

Holtec International's competitive advantage will be lost if its competitors are able to use the results of the Holtec International experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to Holtec International would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

4 of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 STATE OF NEW JERSEY )

) SS:

COUNTY OF BURLINGTON )

Mr. Savit Sinha, being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.

Executed at Marlton, New Jersey, this 07 th day of December, 2010.

Savit Sinha Holtec International Subscribed and sworn before me this 7 day of D t , 2010.

MARIA C MASSI OF NEW t40OTAR~y pUBLICExpires AprilJESISEY 25.,2015 My commission 5 of 5

ATTACHMENT 4 AFFIDAVIT FROM GLOBAL NUCLEAR FUEL JUSTIFYING WITHHOLDING PROPRIETARY INFORMATION Nine Mile Point Nuclear Station, LLC December 13,2010

Global Nuclear Fuel - Americas LLC AFFIDAVIT I, Anthony P. Reese, state as follows:

(1) I am the Manager, Reload Design and Analysis, Global Nuclear Fuel - Americas, LLC (GNF-A), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in Table 1 of Appendix E of Holtec Report No. HI-2012621, Criticality Safety Evaluation for the Nine Mile Point 2 Rack Installation Project. Table 1 is entitled NMP2 SPENT FUEL POOL RERACK PROJECT.

Table 1 is proprietary in its entirety. The header of the page containing Table 1 carries the notation "GEH Proprietary Information33 1." The superscript notation f } refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GNF-A relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F2d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F2d 1280 (DC Cir. 1983).

(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GNF-A's competitors without license from GNF-A constitutes a competitive economic advantage over GNF-A and/or other companies.

b. Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. Information that reveals aspects of past, present, or future GNF-A customer-funded development plans and programs, that may include potential products of GNF-A.

d. Information that discloses trade secret and/or potentially patentable subject matter for which it may be desirable to obtain patent protection.

GNF Proprietary Information in Holtec Report No. HI-2012621 Affidavit Page I of 3

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to the NRC in confidence. The information is of a sort customarily held in confidence by GNF-A, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GNF-A, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary and/or confidentiality agreements that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure are as set forth in the following paragraphs (6) and (7).

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GNF-A. Access to such documents within GNF-A is limited to a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GNF-A are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary and/or confidentiality agreements.

(8) The information identified in paragraph (2) above is classified as proprietary because it contains details of GNF-A's fuel design and licensing methodology for the Boiling Water Reactor (BWR.). Development of these methods, techniques, and information and their application for the design, modification, and analyses methodologies and processes was achieved at a significant cost to GNF-A. The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GNF-A asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit-making opportunities. The fuel design and licensing methodology is part of GNF-A's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GNF-A. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to GNF Proprietary Information in Holtec Report No. HI-2012621 Affidavit Page 2 of 3

quantify, but it clearly is substantial. GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GNF-A would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GNF-A of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 9 th day of December, 2010.

Anthony P. Reese Manager, Reload Design and Analysis Global Nuclear Fuel - Americas LLC 3901 Castle Hayne Rd.

Wilmington, NC 28401 GNF Proprietary Information in Holtec Report No. HI-2012621 Affidavit Page 3 of 3