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Attachment 6 to W3F1-2009-0022, HI-2094376, Rev. 0, Licensing Report for Waterford, Unit 3 Spent Fuel Pool Criticality Analysis
	ML091831259
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Site:	Waterford
Issue date:	06/24/2009
From:	Brickner B; Holtec
To:	; Office of Nuclear Material Safety and Safeguards
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Attachment 6 To W3FI-2009-0022 Revised Holtec Licensing Report for SFP Criticality Analysis (Non Proprietary)

mmmEu HOLTEC INTERN ATIO NAL Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797- 0900 Fax (856) 797 - 0909 Licensing Report For Waterford Unit 3 Spent Fuel Pool Criticality Analysis FOR Entergy Holtec Report No: HI-2094376 Holtec Project No: 1712 Sponsoring Holtec Division: HTS Report Class : SAFETY RELATED COMPANY, PRIVATE This document is the propertyof Holtec International and its Client. It is to be used only in connection wvith the perform-anc&o w1.Norkj iqteIcnrtos ReprodneCtiohtb ieron e for byi H6ltec Intern:atinabor its designated subcontractors.

vot.

'Reproduction. publiation orrepresentation, i-whole orinpart, foranyother p orpose by any, party pther than-

.the-Client is expressliý fo~rbidden'.-

HOLTEC INTERNATIONAL DOCUMENT ISSUANCE AND REVISION STATUS' DOCUMENT NAME: Licensing Report for Waterford Unit 3 Spent Fuel Pool Criticality Analysis HI-2094376 11 CATEGORY:

1712 Author's T

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

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(Such as a Licensing Report)

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

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I'd 4,

Summary of Revisions:

Revision 0: Original Issue Project No. 1712 Report No. HI-2094376 Shaded areas indicate where proprietary information has been removed.

Page i

Table of Contents

1. INTRODUCTION.................................................................................

3

2.

METHODOLOGY...............................................................................

4 2.1 CRITICALITY ANALYSIS..........................................................................

4 2.2 BORON DILUTION ACCIDENT....................................................................

5

3.

ACCEPTANCE CRITERIA....................................................................

6

4.

ASSUMPTIONS...................................................................................

7

5.

INPUT DATA.....................................................................................

8 5.1 FUEL ASSEMBLY SPECIFICATION................................................................

8 5.2 CORE OPERATING PARAMETERS.................................................................

8 5.3 AXIAL BuIRNuP DISTRIBUTION..................................................................

8 5.4 BURNABLE ABSORBERS..........................................................................

9 5.5 STORAGE RACK SPECIFICATION.................................................................

9 5.5.1 Region 1 Style Storage Racks.............................................................

9 5.5.2 Region 2 Style Storage Racks..............................................................

9 5.5.3 Rack Interfaces..............................................................................

9 5.6 ADDITIONAL CALCULATIONS............................................................

...... 10 5.6.1 Fuel Transfer Carriage Criticality........................................................

10 5.6.2.

Upender Criticality.........................................................................

10 5.6.3 New Fuel Elevator Criticality............................................................

10 5.6.4 Boron Dilution Accident Evaluation.....................................................

10 5.6.5 Temporary Storage Racks.................................................................

10 5.6.6 Fuel Pin Storage Container...............................................................

11 5.6.7 New Fuel Storage Vault.......!.....................................II.......................

11

6.

COMPUTER CODES...........................................................................

12

7. ANALYSIS.......................................................................................

12 7.1 REGION I..........................'*

o.... 13 7.1.1 Identification of Reference Fuel Assembly.................................13 7.1.2 Eccentric Fuel Assembly Positioning....................................................

13 7.1.3 Uncertainties Due to Manufacturing Tolerances.......................................

13 7.1.4 Temperature and Water Density Effects......................

o...........................

14 7.1.5 Calculation of Maximum keff..............................................................

15 7.1.6' Abnormal and Accident Conditions.........................

............................. 15 7.2 REGION 2........................................................................................

16 7.2.1 Identification of Reference Fuel Assembly......... ;...........I.............17 7.2.2 Reactivity Effect of Burnable Absorbers During Depletion...........................

17 7.2.3 Reactivity Effect of Axial Burnup, Distribution.........................................

18 7.2.4 Isotopic Compositions....................................................................

18 Project No. 1712 Report No. BI-2094376 Page 1 Shaded areas indicate where proprietary information has been removed.

7.2.5 Uncertainty in Depletion Calculations..................................................................

18 7.2.6 Eccentric Fuel Assembly Positioning...................................................................

19 7.2.7 Uncertainties Due to M anufacturing Tolerances..................................................

19 7.2.8 Temperature and W ater Density Effects..............................................................

20 7.2.9 Calculation of M aximum k..

ff..........................................................................

20 7.2.10 Abnormal and Accident Conditions................... I..................................................

21 7.3 INTERFACES W ITHIN AND BETWEEN RACKS.................................................................

22 7.3.1 Gaps Between Region 1 Racks...........................................................................

22 7.3.2 Gaps Between Region 2 Racks...........................................................................

.22 7.3.3 Gaps Between Region 1 and Region 2 Racks.....................................

...................... 23 7.3.4 Patterns Within Region 2 Racks...................................

23 7.4 ADDITIONAL CALCULATIONS...............................................

............................................. 24 7.4.1 Fuel Transfer Carriage Criticality............................................................................. 24 7.4.2 Upender Criticality..............................................................................................

24 7.4.3 New Fuel Elevator Criticality.............................................................................

24 7.4.4 Boron Dilution Accident Evaluation.................................................................... 24' 7.4.4.1 Low Flow Rate Dilution........................................................................................

24 7.4.4.2 High Flow Rate Dilution.....................................................................................

25 7.4.5 Temporary Storage Racks.....................................................................................

26 7.4.6 Fuel Pin Storage Container...................................................................................

26 7.4.7 New Fuel Storage Vault........................................................................................

26 REFERENCES.............................................................................................................................

27 APPENDIX A: Benchmark Calculations....................................

A-1 Project No. 1712 Report No. HI-2094376 Page 2 Shaded areas indicate where proprietary information has been removed.

1. INTRODUCTION This report documents the criticality safety evaluation for the storage of Standard and Next Generation Fuel (NGF) assemblies in Holtec Region 1 & 2 style high-density spent fuel storage racks (SFSRs) at the Waterford Unit 3 nuclear power plant operated by Entergy Nuclear. The purpose of the present analysis is to re-perform the original criticality analysis, taking credit for soluble boron, in order to qualify the racks, etc. for the storage and handling 'of fuel assemblies having new fuel parameters.

Additional calculations are also documented such as the criticality analysis for storing fuel with an initial enrichment of up to 5.0 wt% 23"U in the Reactor Building Temporary Storage Rack (TSR) and storing fuel rods with an. initial enrichment of up to 5.0 wt% 235U in the Fuel Pin Storage Container (FPSC) in the spent fuel pool, a boron dilution analysis of the spent fuel pool, a criticality analysis of additional spent fuel pool equipment and also the New Fuel Storage Vault (NFV) (See Section 5.6).

The results of the Region 1 calculations are summarized in Table 7.1 through Table 7.6. The calculations demonstrate that maximum k.ff is less than 1.0 without credit for soluble boron and less than or equal to 0.95 with 85 ppm soluble boron. Furthermore, all reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95 with 193 ppm soluble boron present.

The results of the Region 2 calculations are summarized in Table 7.7 through Table 7.22, and Table 7.26 through Table 7.27, and Table 7.29. Under normal conditions, a soluble boron

,concentration of 524 ppm is required in the spent fuel pool. Under credible accident conditions, a

soluble boron concentration of 870 ppm is required (see Table 7.21).

Three loading patterns have been qualified for the Region 2 racks (See Tables 7.16 through Table 7.20):

" a uniform loading of spent fuel meeting the bumup versus enrichment requirements of Table 7.26,

a checkerboard of high and low reactivity fuel (i.e., spent fuel checkerboard). The high reactivity fuel assembly must have an enrichment no greater than 5.0 wt%, 23U and a bumup greater than 27 GWD/MTU and the low reactivity fuel must meet the bumup versus enrichment requirements of Table 7.27,

a checkerboard of fresh (or irradiated) fuel up to 5.0 wt% 23U and empty cell locations (i.e.,

fresh fuel checkerboard).

Within Region 2 racks, several interfaces are possible with the three loading patterns qualified for storage. The permissible interface conditions are summarized as follows:

Project No. 1712 Report No. HI-2094376 Page 3 Shaded areas indicate where proprietary information has been removed.

" No restrictions are fiecessary between the uniform loading pattern and either of the checkerboard loading patterns (fresh or spent).

" For interfaces between a fresh fuel checkerboard and spent fuel checkerboard, the high reactivity spent fuel assembly (5.0 wt% 235U, 27 GWD/MTU) may be face adjacent to no more than one fresh (or irradiated) fuel assembly. The fresh (or irradiated) fuel assembly may be face adjacent with up to 2 high reactivity spent fuel assemblies. Figure 7.4 shows one example of an acceptable 3x3 fresh fuel checkerboard within the center of a spent fuel checkerboard that meets these requirements.

2. METHODOLOGY 2.1 Criticality Analysis The principal method for the criticality analysis of the high-density storage racks is the use of the three-dimensional Monte Carlo code MCNP4a [2]. MCNP4a is a continuous energy three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory. MCNP4a was selected because it has been used previously and verified for criticality analyses and has all of the necessary features for this analysis. MCNP4a calculations' used continuous energy cross-section data predominantly based on ENDF/B-V and ENDF/B-VI. Exceptions are two lumped fission products calculated by the CASMO-4 depletion code, which do not have corresponding cross sections in MCNP4a.

For these isotopes, the CASMO-4 cross sections are used in

MCNP4a. This approach has been validated in [3] by showing that the cross sections result in the same reactivity effect in both CASMO-4 and MCNP4a.

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

=

for MCNP4a, evaluated with a 95% probability at the 95% confidence level [1]. The calculations for this analysis utilize the same computer platform and cross-section libraries used for the benchmark calculations discussed in Appendix A.

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

(1) number of histories per cycle, (2) the number of cycles skipped before averaging, (3) the total number of cycles and (4) the initial source distribution. The MCNP4a criticality output contains a great deal of useful information that may be used to determine the acceptability of the problem convergence.

This information has been used in parametric studies to develop appropriate values for the aforementioned criticality parameters to be used in storage rack criticality calculations. Based on these studies, a minimum of 10,000 histories were simulated per cycle, a minimum of 50 cycles were skipped before averaging, a minimum of 100 cycles were accumulated, and the initial source was usually specified as uniform over the fueled regions (assemblies). Further, the output was reviewed to ensure that each calculation achieved acceptable convergence. These parameters represent an acceptable compromise between calculational precision and computational time.

Fuel depletion analyses during core operation were performed with CASMO-4 (using the 70-group cross-section library), a two-dimensional multigroup transport theory code based on the Method of Project No. 1712 Report No. HI-2094376 Page 4 Shaded areas indicate where proprietary information has been removed.

Characteristics [4-6]. Detailed neutron energy spectra for each rod type are obtained in collision probability micro-group calculations for use in the condensation of the cross sections. CASMO-4 is used to determine the isotopic composition of the spent fuel. In addition, the CASMO-4 calculations are restarted in the storage rack geometry, yielding the two-dimensional infinite multiplication factor! (kinf) for the storage rack to determine the reactivity effect of fuel and rack tolerances, temperature variation, and to perform various studies. For all calculations in the spent fuel pool racks, the Xe-135 concentration in the fuel is conservatively set to zero.

Benchmark.calculations, presented in [11],

CSMO-4 evaluated with a95 prbbltatte9%cninelvl1 The maximum kff is determined from the MCNP4a calculated ker, the calculational bias, the temperature bias, and the applicable uncertainties and tolerances (bias uncertainty, calculational uncertainty, rack tolerances, fuel tolerances, depletion uncertainty) using the following formula:

Max k~fr= Calculated k'ff+ biases + [Yi (Uncertainty)2]"2 In the geometric models used for the calculations, each fuel rod and its cladding were described explicitly, and reflecting or periodic boundary conditions were used in the radial direction which has the effect of creating an infinite radial array of storage cells, except for the assessment of certain accident conditions.

2.2 Boron Dilution Accident The methodology related to the Boron Dilution accident follows the general equation for boron dilution which is, F t C, =Ce V where C,

= boron concentration at time t, C.

= initial boron concentration, V

= volume of water in the pool, and F

= flow rate of un-borated water into the pool This equation conservatively assumes the un-borated water flowing into the pool mixes instantaneously with the water in the pool.

Project No,,1712 Report No. HI-2094376 Page 5 Shaded areas indicate where proprietary information has been removed.

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

t =-In(C0 /C,)

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

3. ACCEPTANCE CRITERIA The high-density spent fuel PWR 'storage racks for Waterford Unit 3 are designed in accordance with the applicable codes and standards listed below. The objective of this evaluation is to show that the effective neutron multiplication factor, ker, is less than 1.0 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with un-borated water at a temperature corresponding to the highest reactivity. In addition, it is to be demonstrated that keff is less than or equal to 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and-flooded with borated water at a temperature corresponding to the highest reactivi-ty. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than 0.95 with a 95%

probability at a 95% confidence level [1].

Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, thereactivity will not exceed the regulatory limit of 0.95 under borated conditions..

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

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

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

USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications (GL-78-0 11),

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

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

Collins, August 19, 1998.

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

Project No. 1712 Report No. HI-2094376 Page 6 Shaded areas indicate where proprietary information has been removed.

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

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

The New Fuel Storage Vault is intended for the receipt and storage of fresh fuel under normally dry conditions where the reactivity is very low. To assure criticality safety under accident conditions and to conform to the requirements of 10 CFR 50.68, these two accident condition criteria must be met:

When fully loaded with fuel of the highest anticipated reactivity and flooded with clean unborated water, the maximum reactivity, including uncertainties,'shall not exceed a keff of 0.95.

With fuel of the highest anticipated reactivity in place and assuming the optimum hypothetical low density moderation, (i.e., fog or foam), the maximum reactivity shall not exceed a kff of 0.98.

These criteria preclude a secondary accident per ANSI 8.1 or accidents under dry conditions.

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

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

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

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

4) The neutron absorber length is modeled to be the same length as the active region of the fuel.

5) No cooling time is credited in the rack calculations.

6) The presence of burnable absorbers in fresh fuel is neglected.

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

Project No. 1712 Report No. HI-2094376 Page 7, Shaded areas indicate where proprietary information has been removed.

7) The presence of annular pellets is neglected. This is conservative as it is bounded by the solid fuel.

8) All structural materials of the new fuel storage racks are conservatively neglected and replaced with water at 'the appropriate density.

9) The concrete wall of the transfer canal is conservatively modeled as 100 cm thick.

10) The FPSC tubes holes were not modeled; however, the other steel structures of the FPSC were modeled as water. Therefore, the neglecting of the tube holes is conservative.

11) The concrete walls of the vault are conservatively modeledas 100 cm thick.

12) The two inch redwood planks in the NFV are assumed to be 1.5 inches thick.

13) In MCNP4a, the Doppler treatment and cross-sections are valid at 300K (80.33 'F);

however, in the NFV calculations no temperature bias is applied to the results to account for the actual temperature of the water.

14) In the NFV the eccentric fuel positioning condition is covered by the fuel cell spacing tolerance.

5. INPUT DATA 5.1 Fuel Assembly Specification The spent fuel storage racks are designed to accommodate various 16x16 fuel assemblies used at the Waterford Unit 3 facility. The design specifications for these fuel assemblies, which were used for this analysis, aregiven in Table 5.1.

5.2 Core Operating Parameters Core operating parameters are necessary for fuel depletion calculations performed with CASMO-4. The core parameters used for the depletion calculations are presented in Table 5.2.

Temperature and soluble boron values are taken as the upper bound (most conservative) of the core operating parameters of Waterford Unit 3. The neutron spectrum is hardened by each of these parameters, leading to a greater production of plutonium during depletion, which results in conservative reactivity values.

5.3 Axial Burnup Distribution Generic axial bumup profiles provided by the client are specified at node centers for 25 equally-spaced axial sections for burnups of less than 25 GWD/MTU and greater than 25 GWD/MTU.

The resulting profiles are presented in Table 5.3.

Project No. 1712 Report No. H]-2094376 Page 8 Shaded areas indicate where proprietary information has been removed.

5.4 Burnable Absorbers At the Waterford Unit 3 facility there is the potential for either B4C, erbia or IFBA burnable absorbers to be located in the fuel assembly as integral absorbers. In [10] it is clearlyseen that the reactivity of thefuel assembly with IFBA bound those with B4C or erbia and therefore only the IFBA is considered in this analysis. The design specifications for the IFBA rods are given in Table 5.1 and are further discussed in Section 7.2.2.

5.5 Storage Rack Specification The storage cell characteristics are summarized in Table 5.4.

5.5.1 Region 1 Style Storage Racks The Region I storage cells are composed of stainless steel boxes separated by a water gap, with fixed neutron absorber panels centered on each side. The steel walls define the storage cells, and stainless steel sheathing supports the neutron absorber panel and defines the boundary of the flux-trap water-gap used to augment reactivity control.

Stainless steel channels connect the storage cells in a rigid structure and define the flux-trap between the neutron absorber panels.

Neutron absorber panels are installed on all exterior walls facing other racks.

The calculational models consist of a single cell with reflective boundary conditions through the centerline of the water gaps, thus simulating an infinite array of Region 1 storage cells. Figure 5.1 shows the actual calculational model containing the reference 16xl 6 assembly, as drawn by the two-dimensional plotter in MCNP4a. 'The calculations are described in Section 7.1.

5.5.2 Region 2 Style Storage Racks The Region 2 storage cells are composed of stainless steel boxes with a single fixed neutron absorber panel, (attached by stainless steel sheathing) centered on each side. The stainless steel boxes are arranged in an alternating pattern such that the connection of the box corners form storage cells between those of the stainless steel boxes.

The calculational models consist of a group of four identical cells surrounded by reflective boundary conditions through the centerline of the composite of materials between the cells, thus simulating an infinite array of Region 2 storage cells. Figure 5.2 shows the actual calculational model containing the 16xl 6 assembly as drawn by the two-dimensional plotter in MCNP4a. The calculations are described in Section 7.2.

5.5.3 Rack Interfaces Project No. 1712 Report No. HI-2094376 Page 9 Shaded areas indicate where proprietary information has been removed.

Based on the layout of the spent fuel pool, there are no Region I to Region 2interfaces. The gap between adjacent Region 2 racks is conservatively neglected. The Region 2 to Region 2 rack loading pattern interfaces are analyzed in Section 7.3.

5.6 Additional Calculations 5.6.1 Fuel Transfer Carriage Criticality The fuel transfer carriage conveys the fuel assemblies through the fuel transfer tube and is capable of accommodating two fuel assemblies at a time, carried in stainless steel boxes. The results of this calculation can be found in Section 7.4.1.

5.6.2 Upender Criticality The fuel upender is a machine located at each end of the transfer tube. The criticality of this component is bounded by the fuel transfer carriage. No input required. See Section 7.4.2.

5.6.3 New Fuel Elevator Criticality The new fuel elevator has a capacity of a single fuel assembly and is utilized to lower new fuel

.from the operating level of the fuel handling building to the bottom of the spent fuel pool. See Section 7.4.3.

5.6.4 Boron Dilution Accident Evaluation The spent fuel pool at Waterford Unit 3 was conservatively assumed to have a soluble boron concentration of 1720 ppm. The spent fuel pool, volume is considered to be 38,600 ft3. Under certain abnormal conditions, un-borated water may dilute this concentration below the requirements determined in Section 7.

Makeup to the-spent fuel storage pool is from the Refueling Water Storage Pool and/or the Condensate Storage Pool. Makeup from the Refueling Water Storage Pool is provided by the refueling water pool purification pump which has a capacity of 150 gpm. The Refueling Water Storage Pool has a minimum boron concentration of 2050 ppm. The component cooling water makeup pumps provide makeup from the Condensate Storage Pool and have a capacity of 600 gpm. For the accident case a high flow rate of 600 gpm is therefore assumed. The results of these calculations are shown in Section 7.4.4.

5.6.5 Temporary Storage Racks The TSR storage cell locations are arranged in a row of 5 cells with the geometric dimensions in Table 5.5.

The design basis calculational model places 5 fresh fuel assemblies enriched to 5.0 wt% 235U in the storage rack. No steel structural material is included. For simplification, the Project No. 1712 Report No. HI-2094376 Page 10 Shaded areas indicate where proprietary information has been removed.

following tolerances are included in the design basis model: fuel density, lattice pitch and enrichment.

5.6.6 Fuel Pin Storage Container The FPSC is a square stainless steel container that fits in a fuel assembly storage rack in the spent fuel pool. It has 81 stainless steel tubes that may contain fuel rods of up to 5.0 wt% 235U (See Table 5.5).

The FPSC was modeled as 81 solid steel tubes of equal diameter, each

'containing 1 fresh fuel rod with the maximum enrichment. All other steel components of the container were neglected. The model includes 100 cm of water surrounding the FPSC or fuel assembly.

The criticality analysis of the FPSC is performed by comparing the reactivity of the FPSC loaded with the maximum number of fresh fuel pins to the reactivity of various fuel assemblies and determine which cases bound the FPSC.

These calculations are performed with the fuel assembly surrounded by 100 cm of water, meaning no storage racks, poison material or structural materials are considered (the steel tubes of the FPSC are modeled). No tolerances are included. Reflective boundary conditions are applied on all sides to maximize reactivity.

5.6.7 New Fuel Storage Vault The NGF assembly is the only fuel assembly type to be stored in the NFV. The design input

-data is tabulated in Table 5.1 and Table 5.6. The storage locations are arranged in 8 modules providing a total of 16 rows of 5 cells each for a total of 80 storage locations. The cells are located on a 21 inch pitch within each module, and on a 49 inch cell center to cell center spacing between modules in the east-west direction and a 58 inch cell center to cell center spacing between modules in the north-south direction. Normally, fuel is stored in the dry condition with very low reactivity. Graphic representations of the analytical model are shown in Figure 7.5 and 7.6. These figures were drawn (to scale) with a two-dimensional plotter.

The reactivity uncertainties associated with various manufacturing tolerances for the NFV were calculated by the difference between two MCNP4a calculations, one with the nominal value and a second independent calculation 'with the tolerance parameter changed. Based on the nominal condition results, it was determined that the 100% moderator condition, i.e. 1.0 g/cc, represented the maximum reactivity condition and therefore the tolerance calculations were performed with 100% moderator density. These tolerance effects each include the combination of statistical errors in the MCNP4a calculations due to the random nature of Monte Carlo calculations, at the 95% confidence level (Ak+(12)*2*a). In evaluating the uncertainties due to tolerances, the following tolerances were used:

Enrichment Tolerance of +/- 0.05 wt% 235U Density of +0.165 g UO2/cm 3 Project No. 1712 Report No. HI-2094376 Page 11 Shaded areas indicate where proprietary information has.been removed.

Fuel Storage Cell Spacing of The fuel storage cell spacing tolerance was only used in the 21 inch assembly pitch.

In determining the maximum keff, the effects of these manufacturing tolerances were statistically combined (square root of the sum of the squares) with the MCNP4a bias uncertainty from the benchmarking results and the MCNP4a calculational statistics (2*")

to determine the total uncertainty.

6. COMPUTER CODES The following computer codes were used during this analysis.

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

CASMO-4, Version 2.05.14 [4-6] is a two-dimensional multigroup transport theory code developed by Studsvik Scandpower, Inc. CASMO-4 performs cell criticality calculations and burnup. CASMO-4 has the capability of analytically restarting burned fuel assemblies in the rack configuration. This code was used to determine the reactivity effects of tolerances and fuel -depletion.

7. ANALYSIS This section describes the calculations that were used to determine the acceptable storage criteria for the Region 1 and Region 2 style racks.

In addition, this section discusses the possible abnormal and accident conditions.

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

As discussed in Section 2, MCNP4a was the primary code used in the PWR calculations.

CASMO-4 was used to determine the reactivity effect of tolerances and for depletion calculations. MCNP4a was used for reference cases and to perform calculations which are not possible with CASMO-4 (e.g., eccentric fuel positioning, axial burnup distributions, and fuel misloading).

Figures 5.1 and 5.2 are pictures of the basic calculational models used in MCNP4a. These pictures were created with the two-dimensional plotter in MCNP4a and clearly indicate the explicit modeling of fuel rods in each fuel assembly. In CASMO-4, a single cell is modeled, and since CASMO-4 is a two-dimensional code, the fuel assembly hardware above and below the active fuel length is not represented. The three-dimensional MCNP4a models that included axial Project No. 1712 Report No. HI-2094376 Page 12 Shaded areas indicate where proprietary information has been removed.

leakage assumed approximately 30 cm of water above and below the active fuel length.

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

7.1 Region 1 The goal of the criticality calculations for the Region I style racks is to qualify the racks for storage of fuel assemblies with design specifications as shown in Table 5.1 and a maximum nominal initial enrichment of 5.0 wt% 235U.

7.1.1 Identification of Reference Fuel Assembly CASMO-4 calculations were performed to determine which of the two assembly types in Table 5.1 is bounding in the Region 1 racks. The presence of burnable absorbers in the fuel assembly (IFBA) was neglected for determination of the reference fuel assembly. The results in Table 7.1 shows that the NGF assembly has the highest reactivity and this assembly type is therefore used in all subsequent calculations.

7.1.2 Eccentric Fuel Assembly Positioning The fuel assemblies are assumed to be normally located in the center of the storage rack cell. To investigate the potential reactivity effect of eccentric positioning of assemblies in the cells, MCNP4a calculations were performed with the fuel assemblies assumed to be in the corner of the storage ra~k cell (four-assembly cluster at closest approach).

The highest reactivity, therefore, corresponds to the reference design with the fuel assemblies positioned in the center of the storage cells. The results of this calculation is shown in Table 7.6.

7.1.3 Uncertainties Due to Manufacturing Tolerances In the calculation of the final keff, the effect of manufacturing tolerances on reactivity must be included. CASMO-4 was used to perform these calculations.

As allowed in [7], the methodology employed to calculate the tolerance effects combine both the worst-case bounding value and sensitivity study approaches. The evaluations include tolerances of the rack and fuel dimensions. As for the bounding assembly, calculations are performed at an enrichment of 5.0 wt% 235U. The reference condition is the condition with nominal dimensions and properties. To determine the Ak associated with a specific manufacturing tolerance, the kinf calculated for the reference condition is compared to the kinf from a calculation with the tolerance included. Note

'that for the individual parameters associated with a tolerance, no statistical approach is utilized.

Instead, the full tolerance value is jutilized to determine the maximum reactivity effect. All of the Ak values from the various tolerances are statistically combined (square root of the sum of the squares) to determine the final reactivity allowance for manufacturing tolerances. The fuel and Project No. 1712 Report No. HI-2094376 Page 13 Shaded areas indicate where proprietary information has been removed.

rack tolerances included in this analysis are described below; the fuel density and enrichment tolerances are typical values:

Fuel Tolerances

" Increased Fuel Density: +0.165 g/cm 3

Increased Fuel Enrichment: 0.05 wt% 235U

Fuel Rod Pitch:

Fuel Rod Cladding Outside Diameterr:

Fuel Rod Cladding Thickness

Fuel Pellet Outside Diameter:

Guide Tube Outside Diameter

Guide Tube Thickness min:

Rack Tolerances

Cell Inner Dimensionn:

Box Wall Thickness:

eCell Pitch:

Boral Width:

Poison Gap min:

Poison Loading mi:

Regarding the tolerance calculations, the following needs to be noted:

In some cases it is not obvious whether an increase or decrease of the parameter will lead to an increase in reactivity. In these cases, the reactivity effect of both increase and decrease of the parameter are calculated, and the positive reactivity effect is used when calculating the statistical combination.

" The tolerance in the flux trap is conservatively captured in the tolerances of the cell ID and cell pitch, since variations of the cell ID are evaluated for a constant cell pitch and vice versa.

Tolerance calculations were erformed for ure water and borated water. The results are 7.1.4 Temperature and Water Density Effects Pool water temperature effects on reactivity in the Region 1 racks have been calculated with CASMO-4 for an enrichment of 5.0 wt% 235U for pure water and borated water. The results are presented in Table 7.3.

The results show that the Region 1 spent fuel pool temperature coefficient of reactivity is negative for both cases,. i.e., a lower temperature results in a higher Project No. 1712 Report No. HI-2094376 Page 14 Shaded areas indicate where proprietary information has been removed.

reactivity. Consequently, the design basis calculations are evaluated at 0 'C (32 °F) for normal conditions.

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

Therefore, a Ak is determined in CASMO-4 from 32 'F to 80.33 °F, and is included in the final keff calculation as a bias. Table 7.3 shows the calculation of the bias. The temperature bias is calculated with pure water and borated water.

7.1.5 Calculation of Maximum keff Using the calculational model shown in Figure 5.1 and the reference 16x16 NGF fuel assemblies, the keff in the Region 1 storage racks has been calculated with MCNP4a.

The calculations of the maximum keff values, based on the formula in Section 2, are shown in Table 7.4 and Table 7.5. In summary, the results show that the maximum keff of the Region 1 racks is less than 1.0 at a 95% probability at a 95% confidence level with no credit for soluble boron, and by linear interpolation, less than or equal to 0.95 with 85 ppm soluble boron.

7.1.6 Abnormal and Accident Conditions The effects on reactivity of credible abnormal and accident conditions are examined in this section. This section identifies which of the credible abnormal or accident conditions will result in exceeding the limiting reactivity (keir <- 0.95). For those accident or abnormal conditions that result in exceeding the limiting reactivity, a minimum soluble boron concentration is determined to ensure that keff:< 0.95. The double contingency principal of ANS-8.1/N16.1-1975 [8] (and the USNRC letter of April 1978; see Section 3.0) specifies that it shall require at least two unlikely, independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

7.1.6.1 Abnormal Temperature All calculations for Region 1 are performed at a pool temperature of 32'F. As shown in Section 7.1.4 above, the temperature coefficient of reactivity is negative, therefore any increase in temperature above 32°F would cause a reduction in the reactivity.

Therefore, no further evaluations of abnormal temperatures are performed.

7.1.6.2 Dropped Assembly - Horizontal For the case in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the active fuel region of more than 12 inches, which is sufficient to preclude neutron coupling (i.e., an Project No. 1712.

Report No. H1-2094376 Page 15 Shaded areas indicate where proprietary information has been removed.

effectively infinite separation). Consequently, the horizontal fuel assembly drop accident will not result in a significant increase in reactivity. Furthermore, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.

7.1.6.3 Dropped Assembly - Vertical Into Fuel Cell It is also possible to vertically drop an assembly into a location that might be occupied by another assembly or that might be empty. Such a vertical impact onto another assembly has previously been shown to cause no damage to either fuel assembly. A vertical drop into an empty storage cell could result in a small deformation of the baseplate. The resultant effect would be the lowering of a single fuel assembly by the amount of the deformation. This could potentially result in further misalignment between the active fuel region and the Boral. However, the amount of deformation for this drop would be small and restricted to a localized area of the rack around the storage cell where the drop occurs. Furthermore, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.

7.1.6.4 Abnormal Location of a Fuel Assembly 7.1.6.4.1 Misloaded Fresh Fuel Assembly The Region 1 racks are qualified for the storage of fresh, unburned fuel assemblies with the maximum permissible enrichment (5.0 wt% 235U). Therefore, the abnormal location of a fuel assembly within normal Region 1 cells is of no concem.

7.1.6.4.2 Mislocated Fresh Fuel Assembly The mislocation of a fresh unburned fuel assembly could, in the'absence of soluble poison, result in exceeding the regulatory limit (keff of 0.95)Y This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt%. 235U) were to be accidentally mislocated outside of a storage rack adjacent to other fuel assemblies. The results of the analysis are shown in Table 7.6 and show by linear interpolation that a soluble boron level of 193 ppm is sufficient to ensure that the maximum kff value for this condition remains at or below 0.95 7.2 Region 2 The goal of the criticality calculations for the Region 2 style racks is to qualify the racks for storage of fuel assemblies with design specifications as shown in Table 5.1 and a maximum nominal initial enrichment of 5.0 wt% 23.5U.

Specifically, the purpose of the criticality calculations is to determine the initial enrichment and burnup combinations required for the Project No. 1712 Report No. HI-2094376 Page 16 Shaded areas indicate where proprietary infonration has been removed.

storage of spent fuel assemblies with nominal initial enrichments up to 5.0 wt%. 235U. Three loading configurations were analyzed to create burnup versus enrichment curves:

a uniform loading of spent fuel meeting the bumup versus enrichment requirements of Table 7.26,

" a checkerboard loading pattern of high and low reactivity fuel with the high reactivity fuel at an enrichment of 5.0 wt% 235U and a burnup of 27 GWD/MTU and the low reactivity fuel must meet the burnup versus enrichment requirements of Table 7.27; a checkerboard of fresh fuel up to 5.0 wt% 235U and empty cell locations (i.e., fresh fuel checkerboard). This configuration bounds a checkerboard of irradiated fuel and empty cells.

7.2.1 Identification of Reference Fuel Assembly CASMO-4 calculations were performed to determine which of the two assembly types are bounding in the Region 2 racks. In the calculations, the fuel assembly is burned in the core.

configuration and restarted in the rack configuration.

For all assemblies, the presence of burnable absorbers in the fuel assembly (BPRA, IFBA) was neglected for determination of the reference fuel assembly (see Section 7.2.2 for a discussion the-effect'of burnable poison). The results are shown in Table 7.7 (selected enrichments and burnups) and show that the NGF assembly has the highest reactivi for all enrichments and burnu s relative to the final burnu ver sus enrichmentcre 7.2.2 Reactivity Effect of Burnable Absorbers During Depletion The Waterford Unit 3 fuel makes use of burnable absorbers of either B4C,- erbia or integral fuel burnable absorber (IFBA) rods with a thin coating of ZrB2 on the U0 2 pellet.

Generic studies [10] havýe investigated the effect that integral burnable absorbers (IBAs) have on the reactivity of spent fuel assemblies.

These studies have concluded that there is a small positive reactivity effect associated with theo presence of IFBA rods, which therefore bounds the negative effects of the B4C and erbia. Therefore, only the IFBA is considered in this analysis.

To determine the reactivity effect for the Waterford Unit 3 spent fuel racks, depletion calculations were performed for' selected configurations of IFBA rods provided by Entergy. The reactivity of the fuel assembly with IFBA rods is compared to the reactivity of the respective fuel assembly without IFBA rods, for both the ure water case and the borated water case. The results are resented in Table 7.8 and

Project No. 1712 Report No. HI-2094376 Page 17 Shaded areas indicate where proprietary information has been removed.

7.2.3 Reactivity Effect of Axial Burnup Distribution Initially, fuel loaded into the reactor will burn with a slightly skewed cosine power distribution.

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

Generic analytic results of the axial burnup effect for assemblies without axial blankets have been provided by Turner [9] based upon calculated and measured axial burnup distributions.

These analyses confirm the minor and generally negative reactivity effect of the axially distributed bumup compared to a flat distribution, becoming positive at burnups greater than about 30 GWD/MTU. The trends observed in [9] suggest the possibility of a small positive reactivity effect above 30 GWD/MTU, increasing to slightly over 1% Ak at 40 GWD/MTU. The required burnup for the maximum enrichment is higher than 30 GWD/MTU. Therefore, a positive reactivity effect of the axially distributed burnup is possible. Calculations are conservatively performed with the axial burnup distribution shown in Table 5.3 (see Section 5.3) and with an axially constant burnup, and the higher reactivity is used in the analyses.

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

The CASMO-4 calculations performed to obtain. the isotopic compositions for MCNP4a were performed generically, with one calculation for each enrichment, and burnups in increments of 2.5 GWD/MTU or less. The isotopic composition for any given burnup is then determined by linear interpolation.

7.2.5 Uncertainty in Depletion Calculations Since critical experiment data with spent fuel is not available for determining the uncertainty in burnup-dependent reactivity calculations, an allowance for uncertainty in reactivity was assigned based upon other considerations.

Based on the recommendation in [7], a burnup dependent uncertainty in reactivity for burnup calculations of 5% of the reactivity decrement is used. This allowance is statistically combined with the other reactivity allowances in the determination of the maximum keff for normal conditions where assembl buru is credited. Additionally, a sensitivity stud7 was performed tP age M8 The results of this study are shown in Table 7.29.

Project No. 1712 Report No. HI-2094376 Page 18 Shaded areas indicate where proprietary information has been removed.

7.2.6. Eccentric Fuel Assembly Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell. In the absence of a fixed neutron absorber, the eccentric location of fuel assemblies in the storage cells may produce a positive reactivity effect. Therefore, the eccentric positioning is performed in a very conservative manner in MCNP4a, assuming 4 assemblies in the corners of the storage cell (four-assembly cluster at closest approach), and that these clusters of four assemblies are repeated throughout the rack. These calculations are performed with pure water and borated water. The results of these calculations are shown in Table 7.9 and indicate that eccentric fuel positioning results in a decrease in reactivity for both cases.

7.2.7 Uncertainties Due to Manufacturing Tolerances In the calculation of the final ker, the effect of manufacturing tolerances on reactivity must be included. CASMO-4 was used to perform these calculations.

As allowed in [7], the methodology employed to calculate the tolerance effects combine both the worst-case bounding value and sensitivity study approaches. The evaluations include tolerances of the rack and fuel dimensions. As for the bounding assembly, calculations are performed for different enrichments and bumups with a maximum value of 5.0 wt% 235U. The reference condition is the condition with nominal dimensions and properties. To determine the Ak associated with a. specific manufacturing tolerance, the kinf calculated for the reference condition is compared to the kinf from a calculation with the tolerance included. Note that for the individual parameters associated with a tolerance, no statistical approach is utilized. Instead, the full tolerance value is utilized to determine the maximum reactivity effect. All of the Ak values from the various tolerances are

.statistically combined (square root of the sum of the squares) to determine the final reactivity allowance for manufacturing tolerances. Only the Ak values in the positive direction (increasing reactivity) were used in the statistical combination. The fuel and rack tolerances included in this analysis are described below; the fuel density and enrichment tolerances are typical values:

Fuel Tolerances

Increased Fuel Density: +0.165 g/cm 3

Increased Fuel Enrichment: 0.05 wt% 235U Fuel Rod Pitch:

Fuel Rod Cladding Outside Diamete

Fuel Rod Cladding Thickness me:

Fuel Pellet Outside Diameter:

Guide Tube Outside Diameter:

Guide Tube Thickness min:D Rack Tolerances Cell Inner Dimension:

Box Wall Thickness:

m Project No. 1712 Report No. HI-2094376 Page 19 Shaded areas indicate where proprietary information has been removed.

Poison Width:

Poison Gap minimuum Boral B-I0 Loading min:

Regarding the tolerance calculations, the following needs to be noted:

In some cases it is not obvious whether an increase or decrease of the parameter will lead to an increase in reactivity. In these cases, the reactivity effect of both increase and decrease of the parameter are calculated, and the positive reactivity effect is used when calculating the statistical combination.

a In the CASMO-4 model used, the tolerance calculation for the Cell ID resulted in a negative reactivity for both increases and decreases in Cell ID. Conservatively, the least negative value was used as a positive reactivity effect.

Tolerance calculations were performed for ire water and borated water. The results are prsntdi Table 7.10 and Table 7.11anurwte 7.2.8 Temperature and Water Density Effects Pool water temperature effects on reactivity in the Region 2 racks have been calculated with

->CASMO-4 for various enrichments with a maximum value of 5.0 wt% 2 35U and the results are

.presented in Table 7.12. The calculations are performed with pure water and borated water. The results show that the Region 2 spent~fuel pool temperature coefficient of reactivity is negative for both cases, i.e., a higher temperature results in a lower reactivity. Consequently, all CASMO-4 calculations are evaluated at 32 "F.

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

Therefore, a Ak is determined in CASMO-4 from 32 'F to 80.33 'F, and is included in the final keff calculation as a bias. The bias is taken from the pure water cases.

7.2.9 Calculation of Maximum kff Using the calculational model shown in Figure 5.2 and the reference 16x16 NGF fuel assembly, the keff in the Region 2 storage racks has been calculated with MCNP4a for the cases discussed in Section 7.2. The determination of the maximum kcff values, based on the formula in Section 2, is shown in, for initial enrichments between 2.0 wt% 23SU and 5.0 wt% 235U, Table 7.13 for the uniform loading case, Table 7.14 for the spent fuel checkerboard loading case, and Table 7.15 for the fresh fuel checkerboard case. A summary of the calculations for non-accident conditions of the maximum keff for spent fuel of maximum nominal enrichment of 5.0 wt% 235U is shown in Table 7.16 for the uniform loading of spent fuel without soluble boron and Table 7.17 with soluble boron, Table 7.18 for the spent fuel checkerboard without soluble boron andTable 7.19 Project No. 1712 Report No. HI-2094376 Page 20 Shaded areas indicate where proprietary information has been removed.

with soluble boron, and Table 7.20 for the fresh fuel checkerboard fuel. Table 7.26 and Figure 7.1 present the bumup versus enrichment requirements for the uniform loading of spent fuel and Table 7.27 and Figure 7.2 present the burnup versus enrichment requirements for the low reactivity fuel assemblies in the spent fuel checkerboard. The results show that the maximum keff of the Region 2 racks is less than 1.0 at a 95% probability and at a 95% confidence level for the three loading patterns with no credit for soluble boron, and less than 0.95 at a 95%

probability and at a 95% confidence level with 524 ppm soluble boron.

7.2.10 Abnormal and Accident Conditions The effects on reactivity of credible abnormal and accident conditions are examined in this section. This section identifies which of the credible abnormal or accident conditions will result in exceeding the limiting reactivity (ken < 0.95). For those accident or abnormal conditions that result in exceeding the limiting reactivity, a minimum soluble boron concentration is determined to ensure that k1ff: 0.95. The double contingency principal of ANS-8.1/N16.1-1975 [8] (and the USNRC letter of April 1978; see Section 3.0) specifies that it shall require at least two unlikely, independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

7.2.10.1 Abnormal Temperature All calculations for Region 2 are performed at a pool temperature of 32 'F. As shown in Section 7.2.8 above, the temperature coefficient of reactivity is negative, therefore no additional calculations are required.

7.2.10.2 Dropped Assembly - Horizontal For the case in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a minimnum separation distance from the active fuel region of more than 12 inches, which is sufficient to preclude neutron coupling (i.e., an effectively infinite separation). Consequently, the horizontal fuel assembly drop accident will not result in a significant increase in reactivity. Furthermore, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.

7.2.10.3 Dropped Assembly - Vertical It is also possible to vertically, drop an assembly into a location that might be occupied by another assembly or that might be empty. Such a vertical impact onto another assembly has previously been shown to cause no damage to either fuel assembly. A vertical drop into an empty storage cell could result in a small deformation of the baseplate. The resultant effect would be the lowering of a single fuel assembly by the amount of the deformation. This could potentially result in further misalignment between the active fuel region and the Boral. However, the amount of deformation for this drop would be small and restricted to a localized area of the rack Project No. 1712 Report No. HI-2094376 Page 21 Shaded areas indicate where proprietary information has been removed.

around the storage cell where the drop occurs. Furthermore, the reactivity increase would be small compared to the reactivity increase created by the misloading of a fresh assembly discussed in the following section. The vertical drop is therefore bounded by this misloading accident and no separate calculation is performed for the drop accident.

7.2.10.4 Abnormal Location of a Fuel Assembly 7.2.10.4.1 Misloaded Fresh Fuel Assembly The misloading of a fresh unburned fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (klf' of 0.95). This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt 0 23U) were to be inadvertently misloaded into a storage cell intended to be used for spent fuel. The results of this accident are shown in Table 7.21.

7.2.10.4.2 Mislocated Fresh Fuel Assembly The mislocation of a fresh unburned fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (keff of 0.95). This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt% 235U were to be accidentally mislocated outside of a Region 2 storage rack adjacent to other fuel assemblies The MCNP4a model consists of an array of Region 2 fuel storage cells with a single fresh, unburned assembly placed adjacent to the rack as close to the rack faces as possible to maximize the possible reactivity effect. The results of the analysis are shown in Table 7.21.

7.3 Interfaces Within and Between Racks The calculations in Sections 7.1 and 7.2 assume laterally infinite arrangements of rack cells. This section evaluates the potential effect of the interfaces between and within rack modules.

7.3.1 Gaps Between Region 1 Racks Region I racks have poison panels on all peripheral walls facing other racks. Furthermore, the assembly distance across the gaps between Region 1 racks is larger than the assembly distance within the racks.

Under abnormal conditions, in the event of lateral rack movement, the baseplate extensions will maintain a minimum rack to rack gap that is bounded by the infinite array calculations, and no further evaluations are necessary.

7.3.2 Gaps Between Region 2 Racks Project No. 1712 Report No. HI-2094376 Page 22 Shaded areas indicate where proprietary information has been removed.

I Under normal conditions, the assembly distance across the gaps between Region 2 racks is larger than the, assembly distance within these racks. Since there is at least one Boral panel between adjacent assemblies for these rack to rack interfaces, the condition in the gap is therefore bounded by the infinite array calculations, and no further evaluations are necessary.

7.3.3 Gaps Between Region 1 and Region 2 Racks According to the data provided by Entergy, Region 1 and Region 2 are separated by distances that exceed the gaps between racks within either region, and therefore the condition is bounded by the infinite array calculations and no further evaluations are necessary.

7.3.4 Patterns Within Region 2 Racks.

The Region 2 racks are qualified for three types of fuel loading pattern: a uniform loading of spent fuel, a spent fuel checkerboard loading pattern, and a fresh (or irradiated) fuel checkerboard loading pattern with empty cells. Within the Region 2 racks, various interfaces between these patterns are qualified. To show that the selected interfaces are acceptable, the following conditions are analyzed:

A An interface between the spent fuel uniform loading pattern and the spent fuel checkerboard. The configuration was chosen so that the high. reactivity assembly in the spent fuel checkerboard pattern (5.0 wt%/27 GWD/MTU) is face adjacent to three low reactivity assemblies from the spent fuel checkerboard pattern (see Table 7.22), and face adjacent to 1 assembly meeting the uniform spent fuel requirement (see Table 7.22).

Two interfaces are evaluated between checkerboards of spent fuel and fresh fuel/empty cells. The bounding case is the case where the fresh fuel assemblies face the high reactivity assembly in the spent fuel checkerboard pattern (5.0 wt%/27 GWD/MTU) on two sides, and has an empty cell on the other two sides. This condition bounds other interfaces between fresh and spent fuel, since the spent fuel with the highest permissible reactivity is used.

The interface configuration is acceptable, when the resulting keff is equivalent to, or less than the maximum keff of the individual pattern. The results are shown in Table 7.22 and show that this requirement is fulfilled for all analyzed cases and therefore:

No restrictions are necessary between the uniform loading pattern and either of the checkerboard loading patterns (fresh or spent).

For interfaces between the fresh fuel checkerboard and spent fuel checkerboard, the high reactivity spent fuel assembly (5.0 wt% 235U, 27 GWD/MTU) may be face adjacent to no more tlhan one fresh fuel assembly. The fresh fuel assembly may be face adjacent with up to 2 high reactivity spent fuel assemblies.

Figure 7.4 shows one example of an acceptable 3x3 fresh fuel checkerboard within the center of a spent fuel checkerboard that meets these requirements.

Project No. 1712 Report No. HI-2094376 Page 23 Shaded areas indicate where proprietary information has been removed.

7.4 Additional Calculations 7.4.1 Fuel Transfer Carriage Criticality The transfer carriage is capable of accommodating two fuel assemblies at a time, carried in stainless steel boxes. The fuel transfer carriage is conservatively modeled as two fuel assemblies at 5.0 wt% 235U and zero burnup separated by 5.06 inches of water only. The calculation of the criticality of the fuel transfer carriage accounts for both the carriage and the transfer tube. The results of the MCNP4a calculations are shown in Table 7.23.

Based on the design of the fuel transfer carriage, a fuel assembly could be mislocated outside the carriage. Two additional calculations were performed with a fresh fuel'-assembly mislocated directly adjacent to one of the two fuel assemblies in the carriage. The results of the MCNP4a calculations are shown in Table 7.23.

7.4.2 UpenderCriticality

The criticality of the Upender is bounded by the calculation of the fuel transfer carriage in Section 7.4.1.

7.4.3 New.Fuel Elevator Criticality The criticality of the New Fuel Elevator is bounded by the calculation of the fuel transfer carriage in Section 7.4.1:

7.4.4 Boron Dilution Accident Evaluation The soluble boron in the spent fuel pool water is conservatively assumed to contain a minimum of 1720 ppm under operating conditions.

Significant loss or dilution of the soluble boron concentration is extremely unlikely, if not incredible. Nonetheless, an evaluation was performed based on the data provided by Entergy.

The required minimum soluble boron concentration is 524 ppm under normal conditions and 870 ppm for the most serious credible accident scenario (see Table 7.19 and Table 7.21). The volume of water in the pool is approximately 288,748 gallons. Large amounts of un-borated water would be necessary to reduce the boron concentration from 1720 ppm to 870 ppm or to 524 ppm.

'Abnormal or accident conditions are discussed below for either low dilution rates (abnormal conditions) or high dilution rates (accident conditions).

7.4.4.1 Low Flow Rate Dilution Project No. 1712 Report No. HI-2094376 Page 24 Shaded areas indicate where proprietary information has been removed.

Small dilution flow around pump seals and valve stems or mis-aligned valves could possibly occur in the normal soluble boron control system or related systems. Such failures might not be immediately detected. These flow rates would be of the order of 2 gpm maximum and the increased frequency of makeup flow might not be observed. However, an assumed loss flow-rate of 2 gpm dilution flow rate would require approximately 119 days to reduce the boron concentration to the minimum required 524 ppm under normal conditions or 68 days to reach the 870 ppm required for the most severe fuel handling accident. Routine surveillance measurements of the soluble boron concentration would readily detect the reduction in soluble boron concentration with ample time for corrective action.

Administrative controls 'require a measurement of the soluble boron concentration in the pool water at least weekly. Thus, the longest time period that a potential boron dilution might exist without a direct measurement of 'the boron concentration is 7 'days. In this time period, an undetected dilution flow rate of 34.0 gpm would be required to reduce the boron concentration to 524 ppm. No known dilution flow rate of this magnitude has been identified. Further, a total of more than 343,000 gallons of un-borated water would be associated with the dilution event and such a large flow of un-borated water would be readily evident by high-level alarms and by visual inspection on daily walk-downs of the storage pool area.

7.4.4.2 High Flow Rate Dilution Under certain accident conditions, it is conceivable that a high flow rate of un-borated water could flowinto the spent fuel pool. As discussed in Section 5.6.4, the component cooling water makeup pumps provide makeup from the Condensate Storage Pool and have a capacity of 600 gpm. Such an accident scenario could result from the continuous operation of the Condensate

'Storage Pool pump and a flow rate of up to 600 gpm which could possibly contribute large amounts of un-borated water into the spent fuel.

Conservatively assuming that all the un-borated water from the pump poured into the pool and further assuming instantaneous mixing of the un-borated water with the pool water, it would take approximately 572 minutes to dilute the' soluble boron concentration to 524 ppm, which is the minimum required concentration to maintain keff below 0.95 under normally operating conditions. In this dilution accident, some 343,000 gallons of water would be released into the spent fuel pool and multiple alarms would have alerted the control room of the accident consequences (including the fuel pool high-level alarm and the Fuel Handling Building sump high level alarm and Liquid Waste Management Trouble alarm). For this high flow rate condition, 328 minutes would be required to reach the 870 ppm required for the most severe fuel handling accident.

It is not considered credible that multiple alarms would fail or be ignored or that the spilling of large volumes of water would not be observed. Therefore, such a major failure would be detected in sufficient time for corrective action to avoid violation of an Technical Specification LCO and to assure that the health and safety of the public is protected.

Project No. 1712 Report No. HI-2094376 Page 25 Shaded areas indicate where proprietary information has been removed.

7.4.5 Temporary Storage Racks The results of the TSR are summarized in Table 724. These results show that the TSR is qualified for loading fuel assemblies with an initial enrichment of up to 5.0 wt% 235U. Based on information provided by Entergy, a fuel assembly may be mislocated on the exterior of the TSR.

The mislocated fresh fuel assembly was modeled at the closest approach (See Table 5.5). For simplification, the following tolerances are included in the design basis model: fuel density, lattice pitch and enrichment (See Table 5.5).

The results of the mislocated case and the necessary soluble boron amount are present in Table 7.24.

7.4.6 Fuel Pin Storage Container The FPSC calculation involved comparing the reactivity of the FPSC to three cases of NGF fuel assemblies under equivalent modeling conditions:

a fresh fuel assembly, a burnup of 27 GWD/MTU and a burnup of 33.4 GWD/MTU, all at 5.0 wt% 235U. These three casesmatch the most reactive fuel assembly for the three loading patterns analyzed in the main body of the report. The results of these comparisons can be seen in Table 7.25. Therefore the FPSC can be placed in any location, intended for fresh or spent fuel.

7.4.7 New Fuel Storage Vault The maximum calculated reactivity of the NFV is listed in Table 7.28. The calculated reactivity as a function of water density is also shown in Figure 7.7. The results show that the optimum moderator density occurs at 100% water density and this maximum keff is below the regulatory limit.

Project No. 1712 Report No. HI-2094376 Page 26 Shaded areas indicate where proprietary information has been removed.

REFERENCES'

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

2. J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-1 2625, Los Alamos National Laboratory (1993).

3. "Lumped Fission Product and Pml48m Cross Sections for MCNP," Holtec Report HI-203303 1, Rev 0, September 2003.

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

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

6. D. Knott, "CASMO-4 Benchmark Against MCNP," SOA-94/12, Studsvik of America, Inc.,

(proprietary).

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

Collins, August 19, 1998.

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

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

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

Note: The revision status of Holtec documents cited above is subject to updates as the project progresses. This document will be revised if a revision to any of the above-referenced Holtec work materially affects the instructions, results, conclusions or analyses contained in this document. Otherwise, a revision to this document will not be made and the latest revision of the referenced Holtec documents shall be assumed to supercede the revision numbers cited above.

The Holtee Project Manager bears the undivided responsibility to ensure that there is no intra-document conflict with respect to the information contained in all Holtec generated documents on a safety-significant project. The latest revision number of all documents produced by Holtec International in a safety significant project is readily available from the company's electronic network.

Project No. 1712 Report No. HI-2094376 Page 27 Shaded areas indicate where proprietary information has been removed.

11. HI-2094370R0, "CASMO-4 Benchmark for Spent Fuel Pool Criticality Analysis."

1' Project No. 1712 Report No. H1-2094376 Shaded areas indicate where proprietary information has been removed.

Page 28

Table 5.1 Fuel Assembly Specification Assembly Type 16x16 Standard 16x16 NGF Stack Density, g/cm 3 10.412 10.522 Fuel Rod Pitch, in 0.506 0.506 Number of Fuel Rods 236 236 Number of Guide Tubes 5

5 Fuel Rod Clad OD, in 0.382 0.374 Fuel Rod Clad ID, in 0.332 0.329 Active Length, in 149.61-150.0 150.0 Fuel Pellet Diameter, in 0.325 0.3225 Guide Tube OD, in 0.98 0.98 Guide Tube ID, in 0.9 0.9 ZrB2 Rod Coating Loading (mgm '0B/inch) 3.14 3.14 ZrB2 Rod Coating 0.0004167 0.000417 Thickhess (inches)

ZrB2 Rod Coating 136 138 Length (inches)

Fuel Assembly Width n/a 8.125

'(min), in.

Bottom of Active Fuel to Bottom of Fuel n/a

-5.402 Assembly, in.

I Project No. 1712 Report No. HI-2094376 Page 29 Shaded areas indicate where proprietary information has been removed.

Table 5.2 Core Operating Parameter for Depletion Analyses Parameter Value Soluble Boron Concentration (bounding cycle 1000 average), ppm Reactor Specific Power, MW/MTU 40.5 Core Average Fuel Temperature, OF 1041.0 Core Average Moderator Temperature at the 614.0 Top of the Active Region, °F In-Core Assembly Pitch, Inches 8.18 K

Project No. 1712 Report No. HI-2094376 Shaded areas indicate where proprietary information has been removed.

Page 30

i Table 5.3 Axial Bumup Profiles Node Center Relative Burnup Relative Burnup (cm)

< 25 GWD/MT

> 25 GWD/MT 7.62 0.54 0.593 22.86 0.773 0.819 38.1 0.921 0.961 53.34 1.013 1.028 68.58 1.055 1.051 83.82 1.065 1.057 99.06 1.064 1.058 114.3 1.061 1.058 129.54 1.058 1.057 144.78 1.056 1.056 160.02 1.054 1.055 175.26 1.053 1.054 190.5 1.052 1.054 205.74 1.051 1.053 220.98 1.05 1.051 236.22 1.047 1.049 251.46 1.046 1.048 266.7 1.044 1.046, 281.94 1.04 1.043 297.18 1.031 1.036 312.42 0.994 1.021 327.66 0.92 0.966 342.9 0.81 0.873 358.14 0.655 0.725 373.38 0.441 0.508 Project No. 1712 Report No. HI-2094376 Page 31 Shaded areas indicate where proprietary information has been removed.

Table 5.4 Storage Rack and Spent Fuel Pool Parameter Specification Region 1 Parameter Value Tolerance Cell ID, in 8.5 Cell Wall thickness, in 0.075 2Cell Pitch, in 10.185 Boundary Sheathing Thickness, in 0.075 n/a Inner Sheathing Thickness, in 0.0235 n/a 3Poison Thickness, in 0.089 n/a Poison Width, in 7.25 Poison Gap, (nominal) in 0.096 W

Flux Trap (nominal) in 1.3 B-10 Loading, (nom) g/cm 2 0.028 Region 2 Parameter Value Tolerance Cell ID, in 8.5 Cell Wall thickness, in 0.075 n/a Cell Pitch, in 8.692 Boundary Sheathing Thickness, in 0.075 n/a Inner Sheathing Thickness, in 0.035 n/a Poison Thickness, in 0.075 n/a Poison Width, in 7.25 Poison Gap, in (nominal) 0.082 B-10 Loading, (nom) g/cm2 0.0216 Additional Spent Fuel Pool Information Parameter Value Tolerance Soluble Boron Concentration, ppm 1720 n/a Spent Fuel Pool Volume, cf 38,600 n/a Fuel Transfer Carriage Gap, in 5.06 n/a Refueling Water Storage Pool (min), ppm 2050 n/a Refueling Water Pool Purification Pump, gpm 150 n/a Component Cooling Water Makeup Pumps, gpm 600 n/a 2 Note that [41 indicates a larger cell-cell pitch for the North-South direction. The value used is bounding.

3 Note that the actual model used 0.075 inches for the poison thickness for conservatism.

Project No. 1712 Report No. HI-2094376 Page 32 Shaded areas indicate where proprietary information has been removed.

Table 5.5 Reactor Building Temporary Storage Rack Parameter Value Number of Storage Cells 5

Pitch, in.

18 1

Rack Opening, in.

8.62 1 Canal Wall to Cell Center, in.

8.06 Distance from Outside Edge of Cell Wall to Outside Edge of Structural 2.25 Material of Cell, in.

Enrichment Tolerance, wt% 23 1U

+ 0.05 Fuel Density Tolerance, g UO 2/cm 3

L0.165 Rack Pitch Spacing 4 Tolerance, in.

Fuel Pin Storage Container Parameter Value Steel Tube Outer Diameter 5, in.

0.625 Steel Tube Thickness, in:

0.035 Steel Tube Pitch, in.

0.917 4 The rack pitch spacing is used to account for the possible gaps between the fuel assembly and rack inner wall. This value is used in the place of the much smaller pitch tolerance listed.

5 Note: 4 tubes have a larger outer diameter; the smaller diameter is used to conservatively model less steel.

Project No. 1712 Report No. HI-2094376 Page 33 Shaded areas indicate where proprietary information has been removed.

Table 5.6 New Fuel Vault Parameters Parameter Value Vault North-South width, ft.

27.5 Vault East-West width, ft.

29.25 Rack Cell Opening, in.

8.9375 Thickness of Redwood Planks, in.

1.5 Rack Cell Pitch, in.

21 East-West Rack Module Center-to-49 Center Cell Separation, in.

North-South Rack Module Center-to-Center Cell Separation, in.

Distance from Fuel Assembly Center 12.25 to North Wall, in.

12I Distance from Fuel Assembly Center 60 to East and West Wall, in.

Distance from Fuel Assembly Center 91.75 to South Wall, in.,

Depth of Rack Cell, in.

190

(

Project No. 1712 Report No. HI-2094376 Page 34 Shaded areas indicate where proprietary information has been removed.

Table 7.1 Results of the Region 1 Reference Fuel Assembly Calculations Assembly Type at 5.0 wt% U-235 0 ppm Soluble Boron 4-I Delta kinf kinf Delta kinr Standard 0.9164 0.0104 NGF 0.9268 d Project No. 1712 Report No. HI-2094376 Page 35 Shaded areas indicate where proprietary information has been removed.

Table 7.2 Region 1 Manufacturing Tolerances and Uncertainty Calculations 0 ppm Soluble Boron m

Parameter kl.r Delta k*.r Paraete Delta kinf Reference Case CASMO 0.9268 n/a Storage Cell ID Increase 0.9370 0.0102 Storage Cell ID Decrease 0.9205

-0.0063 Storage Cell Pitch Increase 0.9184

-0.0084 Storage Cell Pitch Decrease 0.9350 0.0082 Storage Cell Poison Width Increase 0.9250

-0.0018.

Storage Cell Poison Width Decrease 0.9289 0.0021 Storage&Cell Poison Gap Minimum 0.9263

-0.0005 Storage Cell Box Wall Decrease 0.9242

-0.0026 Storage Cell Box Wall Increase 0.9285 0.0017 Storage Cell Poison B-10 Loading Min 0.9291 0.0023 Fuel Rod Pitch Increase 0.9277 0.0009 Fuel Rod Pitch Decrease 0.9259

-0.0009 Fuel Rod Clad OD Increase 0.9248

-0.0020 Fuel Rod Clad OD Decrease 0.9288 0.0020 Fuel Rod Clad Thickness Minimum 0.9267

-0.0001 Fuel Pellet OD Increase 0.9271 0.0003 Fuel Pellet OD Decrease 0.9265

-0.0003 Guide Tube OD Increase 0.9268

.0.0000 Guide Tube OD Decrease 0.9268 0.0000 Guide Tube Thickness Minimum 0.9272 0.0004 Fuel Pellet Enrichment Increase 0.9284 0.0016 Fuel Pellet Density Increase 0.9285 0.0017 C"

Statistical Combination 0.0140 Project No. 1712 Report No. H1-2094376 Page 36 Shaded areas indicate where proprietary information has been removed.

Table 7.3 Region I Temperature and Water Density Effects Results (5.0 wt% U-235) o ppm Soluble Boron Delta Case k

1 in kinf kinr Delta kinr Ref 32 F 0.9268 n/a 39.2 F 0.9266

-0.0002 68 F 0.9253

-0.0015 80.33 F 0.9244

-0.0024 140F 0.9188

-0.0080 255 F 0% voids 0.9028

-0.0240 255F 10% voids 0.8681

-0.0587 255 F 20% voids 0.8295

-0.0973 Bias to 80.33 F 0.0024 Project No. 1712 Report No. HI-2094376 Page 37 Shaded areas indicate where proprietary information has been removed.

Table 7.4 Summary of the Criticality Safety Analysis for Region I Without Soluble Boron Uncertainties:

+-

MCNP4a Code Calculation Statistics (95%/95%,2.Oxa)

Fuel Eccentricity Manufacturing Tolerances Statistical Combination of Uncertainties

+/-

0.0014 negative

+/-

0.0140

+/-

0.0169 Reference keff (MCNP4a)

Total Uncertainty (above)

Bias to 80.33 OF 0.9354 0.0169 0.0024 0.9558 1.0000 Maximum keff Regulatory Limit keff Project No. 1712 Report No. H1-2094376 Shaded areas indicate where proprietary information has been removed.

Page 38

Table 7.5 Summary of the Criticality Safety Analysis for Region 1 with Soluble Boron Requirement Soluble Boron ppm Uncertainties:

85 MCNP4a Code Calculation Statistics MCNP4a Code Calculation Statistics (95%/95%,2.0xo)

Fuel Eccentricity Manufacturing Tolerances Statistical Combination of Uncertainties

+

0.0014 negative

+/-

0.0140

+/-

0.0169 Reference keff (MCNP4a)

Total Uncertainty (above)

Bias to 80.33 'F 0.9246 0.0169 0.0024 0.9450 0.9500 Maximum keff Regulatory Limit keff Project No. 1712 Report No. HI-2094376 Shaded areas indicate where proprietary information has been removed.

Page 39

Table 7.6 Results of Associated Region I Reactivity Calculations Eccentric Positioning Case Case keff Reference 0.9354 Eccentric 0.9332 Delta-k

-0.0022 Soluble Boron Case ppm Boron keg 0

0.9354 200 0.9099 Target keff 0.9246 Calculated ppm 85 Mislocated FA Case ppm Boron*

keff 0

0.9510 400 0.8962 Target kerr 0.9246 Calculated ppm 1

193 Project No. 1712 Report No. HI-2094376 Page 40 Shaded areas indicate where proprietary information has been removed.

Table 7.7 (1 of 2)

Region 2 Calculations for the Reference Fuel Assembly Enrichment 2.0 wt% 235U Burnup (GWD/MTU)

Standard NGF Ak 0.0 0.9568 0.9631 0.0063 0.1 0.9537 0.9600 0.0063 2.0 0.9391 0.9448 0.0057 4.0 0.9231 0.9283 0.0052 Enrichment 3.5 wt% 131U Burnup (GWD/MTU)

Standard NGF Ak 0.0 1.1113 1.1179 0.0067 0.1 1.1089 1.1156 0.0067 2.0 1.0887 1.0952 0.0064 4.0 1.0719 1.0782 0.0062 6.0 1.0547 1.0607 0.0061 8.0 1.0377 1.0435 0.0058 10.0 1.0211 1.0267 0.0055 11.0 1.0130 1.0184 0.0054 12.5 1.0012 1.0063 0.0052 15.0 0.9819 0.9867 0.0048 17.5 0.9631 0.9674 0.0043 20.0 0.9446 0.9484 0.0038 22.5 0.9265 0.9298 0.0033 25.0 0.9088 0.9115 0.0027 Project No. 1712 Report No. HI-2094376 Page 41 Shaded areas indicate where proprietary information has been removed.

Table 7.7 (2 of 2)

Enrichment 5.0 wt% U-235 Soluble 0 ppm Boron Burnup (GWD/MTU)

Standard Delta kinf NGF kinf kinf Standard Delta I

0.0 1.1933 1.1998 0.0065 0.1 1.1915 1.1980 0.0065 2.0 1.1708 1.1773 0.0064 4.0 1.1559 1.1623 0.0064 6.0 1.1406 1.1470 0.0064 8.0 1.1254 1.1317 0.0063 10.0 1.1106 1.1168 0.0062 11.0 1.1034 1.1095 0.0062 12.5 1.0927 1.0987 0.0061 15.0 1.0753 1.0812 0.0059 17.5 1.0583 1.0640 0.0056 20.0 1.0417 1.0471 0.0054 22.5 1.0254 1.0305 0.0051 25.0 1.0094 1.0141 0.0048 27.5 0.9934 0.9979 0.0044 30.0 0.9777 0.9817 0.0040 32.5 0.9620 0.9656 0.0036 35.0 0.9465 0.9497 0.0032 37.5 0.9311 0.9338 0.0027 40.0 0.9158 0.9180 0.0022 42.5 0.9006 0.9023 0.0017 45.0 0.8856 0.8868 0.0012 Project No. 1712 Report No. HI-2094376 Page 42 Shaded areas indicate where proprietary information has been removed.

Table 7.8 (1 of 2)

Region 2 Calculations for NGF Fuel IFBA Rods Reactivity Effect Soluble Boron 0 ppm wt% U235 3.5 5.0 Number of IFBA Rods 0

148 Delta k 0

148 Delta k Burnup GWDIMTU 0.0 1.1179 0.8007 -0.3172 -1.1998 0.9152 -0.2846 0.1 1.1156 0.8026 -0.3130 1.1980 0.9162 -0.2818 2.0 1.0952 0.8564 -0.2388 1.1773 0.9476 -0.2297 4.0 1.0782 0.9013 -0.1769 1.1623 0.9774 -0.1848 6.0 1.0607 0.9330 -0.1278 1.1470 1.0000 -0.1469 8.0 1.0435 0.9537 -0.0898 1.1317 1.0165 -0.1153 10.0 1.0267 0.9655 -0.0611 1.1168 1.0276 -0.0892 11.0 1.0184 0.9686 -0.0498 1.1095 1.0315 -0.0780 12.5 1.0063 0.9704 -0.0359 1.0987 1.0353 -0.0635 15.0 0.9867 0.9673 -0.0194 1.0812 1.0371

-0.0441 17.5 0.9674 0.9585 -0.0089 1.0640 1.0343 -0.0297 20.0 0.9484 0.9461

-0.0024 1.0471 1.0279 -0.0192 22.5 0.9298 0.9315 0.0017 1.0305 1.0188 -0.0117 25.0 0.9115 0.9156 0.0041 1.0141 1.0076 -0.0065 27.5 0.8935 0.8990 0.0055 0.9979 0.9951

-0.0028 30.0 0.8758 0.8821 0.0063 0.9817 0.9815 -0.0002 32.5 0.8585 0.8653 0.0067 0.9656 0.9673 0.0016 35.0 0.8417 0.8486 0.0069 0.9497 0.9525 0.0029 37.5 0.8253 0.8323 0.00706- 0.9338 0.9375 0.0037 40.0 0.8095 0.8165 0.0070 0.9180 0.9223 0.0043 42.5 0.7942 0.8011 0.0069 0.9023 0.9070 0.0047 45.0 0.7796 0.7864 0.0068 0.8868 0.8918 0.0050 Project No. 1712 Report No. HI-2094376 Page 43 Shaded areas indicate where proprietary information has been removed.

9J

Table 7.8 (2 of 2)

Soluble Boron wt% U235 3.5 5.0 Number of IFBA Rods Burnup GWD/MTU 0.0 0.1 2.0 4.0 6.0 8.0 10.0 11.0 12.5 15.0 17.5 20.0 22.5 25.0 27.5 30.0 32.5 35.0 37.5 40.0 42.5 45.0 Project No. 1712 I]

0 148 0

148 Delta k Delta k Report No. HI-2094376 Page44 Shaded areas indicate where proprietary information has been removed.

Table 7.9 (1 of 2)

Region 2 Calculations for Eccentric Fuel Positioning Soluble Boron 0 ppm Case Calculated Delta k keff Reference Uniform Loading 0.9570 Spent Fuel Uniform Loading 0.9517

-0.0053 Eccentric Positioning Reference Spent Fuel Checkerboard 0.9719 Loading

-0.0044 Spent Fuel Checkerboard Loading 0.9675 Eccentric Positioning Reference Fresh Checkerboard 0.8256 Fresh Fuel Checkerboard Eccentric 0.8224

-0.0032 Positioning 0.8224 Project No. 1712 Report No. HI-2094376 Page 45 Shaded areas indicate where proprietary information has been removed.

Table 7.9 (2 of 2)

Soluble Boron

600 ppm Case Calculated Delta k keff Reference Uniform Loading 0.8842 Spent Fuel Uniform Loading 0.8839-0.0003 Eccentric Positioning 0.8839 Reference Spent Fuel Checkerboard 0.9023 Loading

-0.0025 Spent Fuel Checkerboard Loading 0.8998 Eccentric Positioning Reference Fresh Checkerboard 0.7672 Fresh Fuel Checkerboard Eccentric

-0.0041 Positioning Project No. 1712 Report No. HI-2094376 Page 46 Shaded areas indicate where proprietary information has been removed.

Table 7. 10 (1 of 2)

Region 2 Calculations for Manufacturing Tolerance Uncertainties for Fuel Storage Cell Burnup Enrichment Ref I+

ID-Poison Poison Poison Box Box B-oadn Statistical GWD/MTU Case

+

D Width Width Gap Wall +

Wall -

Loaing Cob

_____+

Minmb 0.0 2

0.9631 1-0.0023

-0.0013

-0.0020 10.0026 0.0001 0.0001

-0.0001 10.0034 0.0045 2.0 2

0.9448

-0.0024

-0.0012

-0.0020 0.0025 0.0001 0.0001

-0.0001 0.0034 0.0043 4.0 2.5 0.9897

-0.0029

-0.0009

-0.0021 0.0026 0.0001 0.0001

-0.0001 0.0035 0.004~5 8.0 2.5 0.9534

-0.0028

-0.0008

-0.0020 0.0025 0.0001 0.0001

-0.0001 0.0034 0.0043 11.0 3

0.9769

-0.0030

-0.0006

-0.0021 0.0025 0.0001 0.0000

-0.0001 0.0035 0.0043 15.0 3

0.9443 1-0.0029

-0.0006

-0.0020 0.0024 0.0001 0.0001 1-0.0001 0.0034 0.0042 15.0 3.5 0.9867

-0.0032

-0.0004

-0.0021 0.0026 0.0001 0.0001

-0.0001 0.0035 0.0044 22.5 3.5 0.9298

-0.0029

-0.0005

-0.0020 0.0024-0.0001 0.0000

-0.0001 0.0033 0.0041 22.5 4

0.9679

-0.0032

-0.0003

-0.0020 0.0025 0.0001 0.0000

-0.0001 0.0034 0.0043 27.5 4

0.9326

-0.0030

-0.0004

-0.0020 0.0024 0.0001 0.0000

-0.0001 0.0033 0.0041 27.5 4.5 0.9673 1-0.0032

-0.0002

-0.0020 0.0025 0.0001 0.0000

-0.0001 0.0034 0.0042 32.5 4.5 0.9338

-0.003 1

-0.0003

-0.0020.0.0024 0.0001 0.0000

-0.0001 0.0033 0.0041 32.5 5

0.9656

-0.0033 1-0.0002- -0.0020 0.0025 0.000 1 0.0000

-0.000 1 0.0034 0.0042 40.0 5

0.9180

-0.0030 1-0.0002

-0.0019 0.0024 0,0001 0.0000

-0.0001 0.0032 0.0040 Project No. 1712 Report No. HT-20943 76 Page 47 Shaded areas indicate where proprietary information has been removed.

Table 7.10 (2 of 2)

Region 2 Manufacturinz Tolerance Uncertainties Soluble Boron Effect Commarison (5.0 wt% U-235)

Burnup Ref Case Poison Poison Poison Gap B-10 Statistical GWD/MTU Width +

Width -

Min Box Wall + Box Wall -

Loading Combo.

I Min 0.0 1.1185 i

m 20.0 0.9757

-0 40.0 0.8512_I 60.0 0.7366 0 ppm Soluble Boron Burnup Statistical GWD/MTU Ref Case ID +

ID -

Width +

Width -

Min Box Wall + Box Wall -

Loading Combo.

Widt

+

Wdth MinMin 0.0 1.1998

-0.0046 0.0002

-0.0025 0.0031 0.0001 0.0001

-0.0002 0.0043-0.0053 20.0 1.0471

-0.0038 0.0000

-0.0022 0.0027 0.0001 0.0000

-0.0001 0.0037 0.0046 40.0 0.9180

-0.0030

-0.0002

-0.0019 0.0024 0.0001 0.0000

-0.0001 0.0032 0.0040 60.0 0.7986

-0.0024

-0.0004

-0.0017 0.0021 0.0001 0.0000

-0.0001 0.0028 0.0035 Project No. 1712 Report No. HI-2094376 Page 48 Shaded areas indicate where proprietary information has been removed.

Table 7.11 (1 of 2)

Region 2 Calculations for Fuel Tolerance Uncertainties Bumup Clad Fuel Fuel Guide Guide Guide Fuel Fuel GW/ Er RfCa ldTubea elc Pellet Statistical MTU Case Pitch +

Pitch -

C d +

Thickness Pellet Pellet Tube Tube Th be Pellet De t Ctt bo.

GWin C

tsP OD+

OD-Min OD+

OD-OD+

OD -

Enr +

-Min

+

0.0 2.0 0.9631 0.0007

-0.0007

-0.0009 0.0009 0.0005 0.0004

-0.0004 0.0000 0.0000 0.0002 0.0074 0.0022 0.0079 2.0 2.0 0.9448 0.0007

-0.0007

-0.0008 0.0008 0.0005 0.0004

-0.0004 0,0000 0.0000 0.0002 0.0070 0.0022 0.0075 4.0 2.5 0.9897 0.0008

-0.0008

-0.0008 0.0008 0.0005 0.0004

-0.0004 0.0000 0.0000 0.0002 0.0054 0.0019 0.0059 8.0 2.5 0.9534 0.0008

-0.0008

-0.0007 0.0007 0.0005

-0.0004

-0.0004 0.0000 0.0000 0.0002 0.0054 0.0020 0.0059 11.0 3.0 0.9769 0.0009

-0.0009

-0.0008 0.0007 0.0005 0.0003

-0.0004 0,0000 0.0000 0.0002 0.0045 0.0018 0.0050 15.0 3.0 0.9443 0.0008

-0.0008

-0.0006 0.0006 0.0004 0.0004

-0.0004 0,0000 0.0000 0,0002 0.0046 0.0020 0.0051 15.0 3.5 0.9867 0.0009

-0.0009

-0.0007 0.0007 0.0004 0.0003

-0.0003 0.000 0.0000 0.0002 0.0039 0.0017 0.0044 22.5 3.5 0.9298 0.0009

-0.0008

-0.0005 0.0005 0.0004 0.0004

-0.0004 0.0000 0.0000 0.0001 0.0041 0.0020 0.0047 22.5 4.0 0.9679 0.0009

-0.0009

-0.0007 0.0006 0.0004 0.0003

-0.0004 0.0000 0.0000 0.0002 0.0035 0.0017 0.0041 27.5 4.0 0.9326 0.0009

-0.0009

-0.0005 0.0005 0.0004 0.0004

-0.0004 0.0000 0.0000 0.0001 0.0037 0.0019 0.0043 27.5 4.5 0.9673 0.0010

-0.0009

-0.0006 0.0006 0.0004 0.0003

-0.0003 0.0000 0.0000 0.0002 0.0032 0.0016 0.0038 32.5 4.5 0.9338 0.0009

-0.0009

-0.0005 0.0005 0.0004 0.0004

-0.0004 0.0000 0.0000 0.0001 0.0033 0.0018 0.0040 32.5 5.0 0.9656 0.0010

-0.0010

-0.0006 0.0006 0.0004 0.0003

-0.0003 0000 0.0000 0.0002 0.0030 0.0015 0.0036 40.0 5.0 0.9180 0.0009

-0.0009

-0.0004 0.0004 0.0004 0.0004

-0.0004 0.0000 0.0000 0.0001 0.0031 0.0019 0.0039 Project No. 1712 Report No. HI-2094376 Page 49 Shaded areas indicate where proprietary information has been removed.

Table 7.11 (2 of 2)..

Region 2 Fuel Tolerance Uncertainties Soluble Boron Effect Comparison (5.0 wt% U-235)

Burnup GWD/MT U

0.0 20.0 40.0 60.0 Ref Case F

IF-n-1 11a185 I

I III~

-0,9757 MII 0,8512 1

I II

-I III 0.7 3 6 6

]I I

0 ppm Soluble Boron G u id e F e Bu mup Clad Fuel Fuel Guide Guide Tube Fuel Fuel GWD/MT Ref Pitch Pitch -

CD C

Thicknes Pellet Pellet Tube Tube Pellet Pellet Statistical U

Case

+

OD+

OD-sMin OD+

OD -

OD+

OD-shin Enr+

s Min D

0.0 1.1998 0.0012

-0.0012

-0.00111 0.0011 0.0005 0.0002

-0.0002 0.0000 0.0000

.0.0003 0.0021 0.0012 0.0029 20.0 1.0471 0.0011

-0.0011

-0.0009 0.0009 0.0004 0.0002

-0.0002 0.0000 0.0000 0.0003 0.0027 0.0011 0.0032 40.0 0.9180 0.0009

-0.0009

-0.0004 0.0004 0.0004 0.0004

-0.0004 0.0000 0.0000 0.0001 0.0031 0.0019 0.0039 60.0 0.7986 0.0008

-0.0007 0.0003

-0.0003 0.0004 0.0008

-0.0008 0.0000 0.0000

-0.0001 0.0031 0.0033 0.0047 Project No. 1712 Report No. HI-2094376 Page 50 Shaded areas indicate where proprietary information has been removed.

Table 7.12 (1 of 2)

Region 2 Calculations for Pool Temperature Tolerance Uncertainties Burnup Enr Ref Case T=39.2F T=80.33F T=255F, T=255F, T=255F, GWD/MTU T = 32 F 0% Voids 10% Voids 20% Voids 0.0 2.0 0.9631

-0.0008

-0.0056

-0.0318

-0.0495

-0.0714 2.0 2.0 0.9448

-0.0007

-0.0051

-0.0291

-0.0462

-0.0675 4.0 2.5 0.9897

-0.0006

-0.0046

-0.0273

-0.0458

-0.0684 8.0 2.5 0.9534

-0.0005

-0.0041

-0.0248

-0.0431

-0.0655 11.0 3.0 0.9769

-0.0005

-0.0038

-0.0242

-0.0435

-0.0667 15.0 3.0 0.9443

-0.0004

-0.0035

-0.0225

-0.0414

-0.0643 15.0 3.5 0.9867

-0.0004

-0.0036

-0.0234

-0.0433

-0.0671 22.5 3.5 0.9298

-0.0004

-0.0031

-0.0208

-0.0400

-0.0631 22.5 4.0 0.9679

-0.0004

-0.0032

-0.0219

-0.0419

-0.0658 27.5 4.0 0.9326

-0.0003

-0.0029

-0.0203

-0.0399

-0.0633 27.5 4.5 0.9673

-0.0003

-0.0030

-0.0213

-0.0416

-0.0658 32.5 4.5 0.9338

-0.0003

-0.0028

-0.0199

-0.0398

-0.0635 32.5 5.0 0.9656

-0.0003

-0.0029

-0.0208

-0.0414

-0.0657 40.0 5.0 0.9180

-0.0003

-0.0025

-0.0189

-0.0388

-0.0625 Project No. 1712 Report No. HI-2094376 Page 51 Shaded areas indicate where proprietary information has been removed.

Table 7.12 (2 of 2)

Region 2 Calculations for Pool Temperature Tolerance Uncertainties Soluble Boron Effect Comparison (5.0 wt% U-235) 600 ppm Soluble Boron Burnup

=32 F T =39.2 F T =8033 F T=255F,0%

T=255F,10% T=255F,20%

GWD/MTU Enr Ref Case T Voids Voids Voids 0.0 5

1.1185

-0.0003

-0.0022

-0.0153

-0.0277

-0.0442 20.0 5

0.9757

-0.0002

-0.0020

-0.0138

-0.0268

-0.0437 60.0 5

0.7366 0.0000

-0.0003

-0.0047

-0.0153

-0.0292 0 ppm Soluble Boron Bumup RefCaseT=32 F T =39.2 F T =80.33 F T=255F, 0%

T=255F, 10%

T=255F, 20%

GWD/MTU Enr Voids Voids Voids 0.0 5

1.1998

-0.0004

-0.0034

-0.0248

-0.0462

-0.0718 20.0 5

1.0471

-0.0004

-0.0033

-0.0233

-0.0444

-0.0695 60.0 5

0.7986

-0.0001

-0.0015

-0.0129

-0.0307

-0.0518 Project No. 1712 Report No. HI-2094376 Page 52 Shaded areas indicate where proprietary information has been removed.

Table 7.13 Region 2 Results for the Spent Fuel Uniform Loading Enrichment (wt% U235) 12.0 12.5 3.0 3.5 14.0 4.5 5.0 Burnup (GWD/MTU) 0.0 6.4 12.4 18.3

24.3 28.9 34.1 CASMO Bunup for Tolerances

0.

0 0

11.0 15.0 22.5 27.5 32.5 CASMO Burup for Depletion Uncertainty n/a 8.0 12.5 20.0 25.0 30.0 35.0 Depletion Uncertainty 0.0000 0.0038 0.0057 0.0085 0.0100 0.0113 0.0125 Manufacturing Uncertainty 0.0045 0.0045 0.0043 0.0044 0.0043 0.0042 0.0042 Fuel Uncertainty 0.0079 0.0059 0.0050 0.0044 0.0041 0.0038 0.0036 CalTulational Uncertainty 0.0012 0.0012 0.0014 0.0014 0.0012 0.0012 0.0014 Total Uneertainty 0.0131 0.0125 0.0129 0.0141 0.0150 0.0158 0.0166 Temperature Bias 0.0056 0.0046 0.0038 0.0036 0.0032 0.0030 0.0029 IFBA Bias 0.0070 0.0070 0.0070 0.0070 0.0070.

0.0070 0.0070 Adjusted knra (0.945-corrections) 0.9181 0.9197 0.9201 0.9191 0.9187 0.9180 0.9173 MCNP kfrr0 ppmnBoron O

09613 0.9697 1 0.9701 0.96911 0.9687 0.9680 0.9673 MCNP slo f 600 ppm Boron 0.8560 n/a n/a 0.8901 n/a n/a 0.9003 Toa 0,,0,m oo

.95.95.9 0 0.99 0 0.905ý0 0 0.9950 Total kcar with 600 ppm Boron 08829/

n/a n/a 0.9169 n/a n/a 0.9271 Normal Conditions Interpolated Boron n/a6 489 n/a 49 n/a n/a 522 Concentration to Adjusted kefr Mislocated ker 0 ppm Boron n/a 1.0085 n/a 1.0046 n/a n/a 1.0011 Mislocated k,fr 600 ppm. Boron n/a 0.8996 n/a 0.9017 n/a n/a 0.9048 Misloeated Conditions Interpolated Boron

/

48 na 49 na na 52 Concentration to Adjusted kcfrn/

48 na 49 na na 52 Misloaded kerr 0 ppm Boron n/a 1.0105 n/a 1.0097 n/a n/a 1.0068 Misloaded kerr 800 ppm Boron n/a 0.9018 n/a 0.9103 n/a n/a 0.9156 Misloaded Conditions Interpolated Boron n/a 668 n/a 729 n/a n/a 785' Concentration to Adjusted kerr Note: For the 2.0 wt% U-235 case, the Total kerr 0 ppm Boron value was conservatively increased to 0.9950 for consistency with the other values in the same row.

Project No. 1712 Report No. HI-2094376 Page 53 Shaded areas indicate where proprietary information has been removed.

Table 7.14 Region 2 Results for the Spent Fuel Checkerboard Loading Enrichment. (wt% U235) 2.0 2.5 3.0 3.5 I4.0 4.5 5.0 Burnup (GWD/MTU) 3.7 10.7 17.9 LN24.9 31.5 36.7 43.2 CASMO Bumnup for Tolerances 2.0 '

8.0 15.0 22.5 27.5 32.5 40.0 CASMO Burnup for Depletion Uncertainty 4.0 11.0 20.0 25.0 32,5 37.5 45.0 Depletion Uncertainty 0.0017 0.0051 0.0087 0.0103 0.0126 0.0138 0.0157 Manufacturing Uncertainty 0.0043 0.0043 0.0042 0.0041 0.0041 0.0041 0.0040 Fuel Uncertainty 0.0075 0.0059 0.0051 0.914 na0.0040 0.0039 Calculational Uncertainty 0.0688 b.0012 0.0012 0.0010 0.0014 0.0014 0.0012 Total Uncertainty 0.8868 n

0.0144 0.0153 0.0168 0:0177 0.0191 Temperature Bias 0.0051 0.0041 0.0035 0.0031 0.0029 0.0028 0.0025 IFBA Bias 0.0070 0.0070 ;

.0070 0.0076 0.0070 0.0070 0.0070 Adjusted k~rr (0.995-corrections) 0.9688 ZO.9698 0.9689 0.9684 0.96-70 0.9-664 0.9652 Adjusted k~rr (0.945-corrections) 0.9188 n/a n/a 0.9184 n/a n/a 10.9152 MCNP k~rr 0 ppm Boron 0.9688 0.9698 0.9689 0.9684 0.9670 0.966j] 0.9652 MCNP kcrr 600 ppm Boron 0.8868 n/a n/a 0.8946 n/a n/a 0.9079

/

Total krr without Boron 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 Total kefr with 600 ppm Boron 0.9130 n/a n/a 0.9212 n/a n/a 0.9378 Normal Conditions Interpolated Boron 366 I /

n/a I407 I/

/

2 Concentration to Adjusted krr3 n/a n/a 524 Mislocated kerr 0 ppm Boron 1.0075 n/a n/a 1.0059 n/a n/a 1.0056 Mislocated kerr 600 ppm Boron 0.9055 n/a n/a 0.9067 n/a n/a 0.9064 Mislocated Conditions Interpolated Boron 5

Concentration to Adjusted kerr 522 n/a n/a 529 n/a n/a 547 Misloaded k~ff 0 ppm Boron 1.0194 n/a n/a 1.0125 n/a n/a 1.0114 Misloaded keel 1000 ppm Boron 0.8973 n/a n/a 0.9011 n/a n/a 0.9007 Misloaded Conditions Interpolated Boron 825 n/a n/a 844 n/a n/a Concentration to Adjusted krf Project No. 1712 Report No. HI-2094376 Page 54 Shaded areas indicate where proprietary information has been removed.

Table 7.15 Region 2 Results for the Fresh Checkerboard Loading Enrichment (wt% U235) 5.0 Burnup (GWD/MTU) 0.0 CASMO Burnup for Tolerances 0.0000 Manufacturing Uncertainty 0.0053 Fuel Uncertainty 0.0029 Calculational Uncertainty 0.0014 Total Uncertainty 0.0101 2 Temperature Bias-0.0034 IFBA Bias 0.0070 MCNP keff 0 ppm Boron 0.8256 Normal Conditions Total keff without Boron 0.8484 Adjusted keff (0.945-corrections) 0.9222 Mislocated klff 0 ppm Boron 1.0171 Mislocated klff 600 ppm Boron 0.9044 Mislocated Conditions Interpolated Boron 505 Concentration to Adjusted keff Misloaded kfr 0 ppm Boron 1.0151 Misloaded krff 1000 ppm Boron 0.9050 Normal Conditions Interpolated Boron 844 Concentration to Adjusted k._f_

Project No. 1712

'Report No. HI-2094376 Page 55 Shaded areas indicate where proprietary infornation has been removed.

Table 7.16 Summary of the Criticality Safety Analysis for Region 2, Spent Fuel Uniform Loading, 0 ppm Soluble Boron Enrichment (wt% 235U) 5.0 Burnup (GWD/MTU) 34.1 Soluble Boron ppm 0.0 Fuel Eccentricity negative Statistical Combination of Uncertainties 0.0166 Calculated keff (MCNP4a) 0.9673 IFBA Bias 0.0070 Bias to 80.33 'F 0.0029 Maximum klff 0.9950 Regulatory Limit kff 1.0000 Project No. 1712 Report No. H1-2094376 Shaded areas indicate where proprietary ififormation has been removed.

Page 56

I Table 7.17 Summary of the Criticality Safety Analysis for Region 2, Spent Fuel Uniform Loading, 448 ppm Soluble Boron Enrichment (wt% 235U) 5.0 Burnup (GWD/MTU) 34.1 Soluble Boron (ppm) 448 Statistical Combination of Uncertainties

L 0.0166 Calculated ker (MCNP4a) 0.9173 IFBA Bias 0.0070 Bias'to 80.33 'F 0.0029 Maximum kff 0.9450 Regulatory Limit k1f 0.9500 Project No. 1712 Report No. HI-2094376 Shaded areas indicate where proprietary information has been removed.

Page 57

Table 7.18 Summary of the Criticality Safety Analysis for Region 2, Spent Fuel Checkerboard Loading, 0 ppm Soluble Boron Enrichment (wt% 235

0) 5.0 Burnup (GWD/MTU) 43.2 Soluble Boron (ppm) 0.0 Fuel Eccentricity negative Statistical Combination of Uncertainties

+

0.0191 Calculated k.ff (MCNP4a) 0.9652 IFBA Bias 0.0070 Bias to 80.33 'F 0.0025 Maximum krff 0.9950 Regulatory Limit kff 1.0000 Project No. 1712 Report No. HI-2094376 Shaded areas indicate where proprietary information has been removed.

Page 58

Table 7.19 Summary of the Criticality Safety Analysis for Region 2, Spent Fuel Checkerboard Loading, 524 ppm Soluble Boron Project No. 1712 Report No. HI-2094376 Shaded areas indicate where proprietary information has been removed.

Page 59

Table 7.20 Summary of the Criticality Safety Analysis for Region 2, Fresh Fuel Checkerboard Loading, 0 ppm Soluble Boron Enrichment (wt% 235U) 5.0 Bumup (GWD/MTU) 0.0 Soluble Boron (ppm) 0.0 Fuel Eccentricity negative Statistical Combination of Uncertainties

+

0.0112 Calculated kff (MCNP4a) 0.8256 IFBA Bias 0.0070 Bias to 80.33 'F 0.0034 Maximum k t 0.8484 Regulatory Limit keff 1.0000 ProjectNo. 1712 Report No. H1-2094376 Shaded areas indicate where proprietary information has been removed.

Page 60

0 Table 7.21 Summary of Region 2 Accident Cases Case Result Dropped Fuel Assembly - Horizontal On Top of Cells Negligible Dropped Fuel Assembly - Vertical into Negligible Storage Cell Misloaded Fuel Assembly, Spent Fuel Checkerboard Loading, 5.0 wt% 235U 8706 (ppm Soluble Boron)

Mislocated Fuel Assembly, Spent Fuel Checkerboard Loading, 5.0 wt% 235U 5477 (ppm Soluble Boron) 6 This case was the maximum for the misloaded assembly in the spent fuel uniform loading, spent fuel checkerboard loading, or fresh fuel checkerboard.

7 This case was the maximum for the mislocated assembly in the spent fuel uniform loading, spent fuel checkerboard loading, or fresh fuel checkerboard.

Project No. 1712 Report No. HI-2094376 Page 61 Shaded areas indicate where proprietary information has been removed.

7

Table 7.22 Region 2 Calculation Results for the Interface Cases Ref kerr Description Axial Burnup ker (at Profile.

Er (GWD/MTU) curve)

Interface Segmented 2.0 3.7 0.9545 0.9688 between half a Spent fuel Segmented 3.5 24.9 0.9475 0.9684 rack of fresh checkerboard fuel loading, fresh Segmented 5.0 43.2 0.9425 0.9652 checkerboard FA adjacent 27 Uniform 2.0 3.7 0.9553 0.9688 and half a rack GWD/MTU, 5.0 Uniform 3.5 24.9 0.9484 0.9684 of spent fuel wt% 235U FA checkerboard Uniform 5.0 43.2 0.9436 0.9652 Interface Segmented 2.0 3.7 0.9684 0.9688 between a 3x3 set of fresh Spent fuel Segmented 3.5 24.9 0.9625 0.9684 checkerboard checkerboard Segmented 5.0 43.2 0.9570 0.9652 (fresh in loading, fresh center)

FA adjacent 27 Uniform 2.0 3.7 0.9688 0.9688 surrounded by GWD/MTU, 5.0 Uniform 3.5 24.9 0.9659 0.9684 a rack of spent wt%...

U FA fuel Uniform 5.0 43.2 0.9597 0.9652 checkerboard Segmented 2.0 3.7 0.9675 0.9688 Segmented 3.5 24.9 0.9681 0.9684 Interface between a set of spent Segmented 5.0 43.2 0.9629 0.9652 fuel checkerboard loading fuel and spent uniform loading fuel.

Uniform 2.0 3.7 0.9659 0.9688 Uniform 3.5 24.9 0.9676 0.9684 Uniform 5.0 43.2 0.9626 0.9652 Project No. 1712 Report No. HI-2094376 Page 62 Shaded areas indicate where proprietary information has been removed.

Table 7.23 Results of the Calculation of the Fuel Transfer Carriage Description Calculated keff Reference 0.9436 Case Mislocated 1.0612 Case 800 ppm 0.9209 Boron Case Project No. 1712 Report No. HI-2094376 Page 63 Shaded areas indicate where proprietary information has been removed.

Table 7.24 Results of the Criticality Analysis for the TSR si Descripation Calculated]Ik~f TSR Design Basis Model TSR Mislocated Fuel Assembly Model TSR Mislocated Fuel Assembly Model with 800 ppm Soluble Boron Extrapolated TSR Soluble Boron Requirement for Mislocated Accident, ppm Project No. 1712 Report No. HI-2094376 Page 64 Shaded areas indicate where proprietary information has been removed.

Table 7.25 Results of the Criticality Analysis for the FPSC Description Calculated kf FPSC Design Basis Model 0.6715 5.0 wt% 235U Fuel Assembly at 0.7521 33.4 GWD/MTU 5.0 wt% 235U Fuel Assembly at 27 0.7784 GWD/MTU Fresh NGF Fuel Assembly 0.9226' Project No. 1712 Report No. HI-2094376 Page 65 Shaded areas indicate where proprietary information has been removed.

Table 7.26 Region 2 Burnup Versus Enrichment Curve for Spent Fuel Uniform Loading Enrichment (wt% 235U)

Burnup (GWD/MTLJ) 2.0 0.0 2.5 6.4 3.0 12.4 3.5 18.3 4.0 24.3 4.5 28.9 5.0 34.1 I

Project No. 1712 Report No. HI-2094376 Page 66 Shaded areas indicate where proprietary information has been removed.

Table 7.27 Region 2 Bumup Versus Enrichment Curve for Spent Fuel Checkerboard Loading Enrichment (wt% 235U)

Burnup (GWD/MTU) 2.0 3.7 2.5 10.7 3.0 17.9 3.5 24.9 4.0 31.5 4.5 36.7 5.0 43.2 Project No. 1712 Report No. HI-2094376 Page 67 Shaded areas indicate where proprietary information has been removed.

/

Table 7.28 Summary of the Criticality Safety Analysis for New Fuel Vault, 100%

Moderator Density Tolerances:

Enrichment keff Enrichment Uncertainty Pellet Density ktff Pellet Density Uncertainty Storage Rack Pitch keff Storage Rack Pitch Uncertainty 0.9195 +/- 0.0008 0.9192 +/- 0.0008 0.9187 +/-

0.0007 0.0034 0.0031 0.0023

+

0.0014 Calculation Statistics (95%/95%,2xa)

Statistical Combination of Uncertainties Calculated kff (MCNP4a)

+-

0.0104 0.9184 0.9300 0.9500 Maximum keff Regulatory Limit klff Project No. 1712 Report No. HI-2094376 Shaded areas indicate where proprietary information has been removed.

Page 68

Table 7.29 Region 2 Sensitivity Calculations for Depletion Uncertainty

-with Soluble Boron ppm 0

Burnup GWD/MTU 5% Decrement 5% Decrement Delta (U-)

0.0 n/a n/a n/a 20.0 0.0076 40.0 0.0141 60.0 0.0201 Project No. 1712 Report No. HI-2094376 Page 69 Shaded areas indicate where proprietary information has been removed.

Figure Proprietary Figure 5.1 Region 1 Model Project No. 1712 Report No. HI-2094376 Shaded areas indicate where proprietary information has been removed.

Page 70

Figure Proprietary Figure 5.2 Region 2 Model Project No. 1712 Report No. HI-2094376 Shaded areas indicate where proprietary information has been removed.

Page 71

Figure 7;1 Region 2 Uniform Loading Bumup vs. Enrichment Curve 0

0-le 2.0 2.5 3.0

'3.5 4.0 4.5 wt. % U-235 Note: For Enrichments lower than 2 wt%, apply the burnup value at 2 wt%.

5.0 Project No. 1712 Report No. H11-2094376 Shaded areas indicate where proprietary information has been removed.

Page 72

Figure 7.2 Region 2 Checkerboard Loading Bumup vs. Enrichment Curve 50 I-25 2.0 2.5 3.0 3.5 4.0 4.5 wt. % U-235 Note: For Enrichments lower than 2 wt%, apply the burnup value at 2 wt%.

5.0 Project No. 1712 Report No. HI--2094376 Shaded areas indicate where proprietary information has been removed.

Page 73

Figure 7.3 Region 2 intra-rack interface between half a rack of Fresh Fuel checkerboard and half a rack of spent fuel checkerboard A

B.

A B

A A

B A

A A

A B

A

!j:B:'¸ A

A A

B 5 wt% 23U, 27 GWD/MTU Spent Fuel At Spent Fuel Checkerboard Curve 5 wt% 235U Fresh or Irradiated Fuel I

.Empty Cell Project No. 1712 Report No. HI-2094376 Page 74 Shaded areas indicate where proprietary information has been removed.

Figure 7.4 Region 2 intra-rack interface between a 3x3 set of Fresh Fuel checkerboard (fresh in center) surrounded by a rack of spent fuel checkerboard B

A B

A B:

A B

A B-A A

3B A

B B

A B

A FyA B

A B.

A B

A

.B A

B A

B, B.

A B

A 5 wt% 235U, 27 GWD/MTU B

Spent Fuel At Checkerboard Curve L'

5 wt% 235U Fresh or Irradiated Fuel Empty Cell Project No. 1712 Report No. HI-2094376 Page 75 Shaded areas indicate where proprietary information has been removed.

Figure 7.5 Two-Dimensional Representation of the Actual Calculations Model used for the New Fuel Vault as seen from above.

Figure Proprietary Project No. 1712 Report No. HI-2094376 Page 76 Shaded areas indicate where proprietary information has been removed.

Figure 7.6 Two-Dimensional Representation of the Actual Calculations Model used for the New Fuel Vault as seen from the side.

Figure Proprietary Project No. 1712 Report No. HI-2094376 Page 77 Shaded areas indicate where proprietary information has been removed.

Figure 7.7 Results of the Waterford Unit 3 New Fuel Vault Criticality Analysis As a Function of Water Density 10

% Moderator Density 100 Project No. 1712 Report No. HI-2094376 Shaded areas indicate where proprietary information has been removed.

Page 78

Appendix A Benchmark Calculations HOLTEC PROPRIETARY APPENDIX HAS BEEN REMOVED IN IT'S ENTIRETY Project No. 1712 Report No. HI-2094376 Holtec International Proprietary Information Page A-I

v t e Letter Sequence Other
TAC:MD9685, (Approved, Closed)
Initiation Request, Request, Request Acceptance Administration Withholding Request Acceptance Questions 2 January 2009 6 January 2009 14 January 2009 17 April 2009 16 September 2009 Answers 24 September 2009 26 February 2009 24 September 2009 Results Approval Other: ML091831259, ML091890538
MONTHYEAR2009-04-17 [Table View]

ML091831259

Text

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