CNL-15-249, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 1, Spent Fuel Pool Criticality Safety Analysis Information

Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 1, Spent Fuel Pool Criticality Safety Analysis Information

ML15351A097
Person / Time
Site:	Browns Ferry
Issue date:	12/15/2015
From:	James Shea Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15351A096	List: CNL-15-249, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 1, Spent Fuel Pool Criticality Safety Analysis Information
References
CNL-15-249 ANP-3160(NP), Rev. 1
Download: ML15351A097 (170)

Text

Proprietary Information Withhold from Public Disclosure Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 1 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-15-249 December 15, 2015 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296

Subject:

Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) -

Supplement 1, Spent Fuel Pool Criticality Safety Analysis Information

Reference:

Letter from TVA to NRC, CNL-15-169, "Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments -

Extended Power Uprate (EPU)," dated September 21, 2015 By the reference letter dated September 21, 2015, Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) for the Extended Power Uprate (EPU) of Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3. The proposed LAR modifies the renewed operating licenses to increase the maximum authorized core thermal power level from the current licensed thermal power of 3458 megawatts to 3952 megawatts.

During a November 10, 2015, Nuclear Regulatory Commission (NRC) public meeting with TVA representatives regarding the EPU LAR, the NRC requested the Spent Fuel Pool (SFP) Criticality Safety Analysis (CSA) be submitted for review. Enclosure 1 of this letter provides the BFN Units 1, 2, and 3 SFP CSA for ATRIUM 10XM Fuel.

U.S. Nuclear Regulatory Commission CNL-15-249 Page 2 December 15, 2015 AREVA considers portions of the information provided in Enclosure 1 of this letter to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390, public inspections, exemptions, requests for withholding . An affidavit for withholding information, executed by AREVA, is provided in Enclosure 3. A non-proprietary version of the document is provided in Enclosure 2. Therefore, on behalf of AREVA, TVA requests that Enclosure 1 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390.

During the November 10, 2015, NRC public meeting with TVA representatives regarding the EPU LAR, the NRC also requested information be submitted regarding the impact of fuel assembly operation with suppression blade on the SFP CSA and the aging management program for the Boral neutron absorbers used in the SFP storage racks. This information is provided in Enclosures 4 and 5, respectively.

TVA has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the NRC in the reference letter. The supplemental information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration . In addition, the supplemental information in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed license amendment. Additionally, in accordance with 10 CFR 50.91 (b)(1 ), TVA is sending a copy of this letter and the non-proprietary enclosures to the Alabama State Department of Public Health.

There are no new regulatory commitments associated with this submittal. If there are any questions or if additional information is needed, please contact Mr. Edward D.

Schrull at (423) 751-3850.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 15th day of December 2015.

ully,

.~

Shea resident, Nuclear Licensing Enclosures cc: See Page 2

U.S. Nuclear Regulatory Commission CNL-15-249 Page 3 December 15, 2015

Enclosures:

1. ANP-3160(P) Revision 1, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 10XM Fuel (Proprietary)

2. ANP-3160(NP) Revision 1, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 10XM Fuel (Non-Proprietary)

3. AREVA Affidavit

4. Suppressed Fuel Assembly Impact on Criticality Safety Analysis

5. Boral Neutron Absorber Aging Management Program cc:

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant State Health Officer, Alabama Department of Public Health (w/o Enclosure 1)

ENCLOSURE 2 ANP-3160(NP) Revision 1, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 10XM Fuel (Non-Proprietary)

ANP-3160(NP)

Revision 1 Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM' 10XM Fuel December 2015 Proprietary

AREVA Inc.

ANP-3160(NP)

Revision 1 Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM' 10XM Fuel sja

AREVA Inc.

ANP-3160(NP)

Revision 1 Copyright © 2015 AREVA Inc.

All Rights Reserved

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page i Nature of Changes Item Page Description and Justification

1. 3-12 Table 3.1, IV.5.b corrected typo.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page ii Contents 1.0 Introduction ..................................................................................................................1-1 2.0 Summary and Conclusions ...........................................................................................2-1 3.0 Regulatory Criticality Safety Criteria and Guidance ......................................................3-1 4.0 Fuel and Storage Array Description ..............................................................................4-1 4.1 Fuel Assembly Design ......................................................................................4-1 4.2 Fuel Storage Racks ..........................................................................................4-1 5.0 Calculation Methodology ..............................................................................................5-1 5.1 Area of Applicability ..........................................................................................5-2 6.0 Modeling Options and Assumptions .............................................................................6-1 6.1 Geometric Modeling of the High Density Boral Rack .........................................6-1 6.1.1 Single Cell Model Description .............................................................6-1 6.1.2 Explicit Storage Cell Model Description ...............................................6-2 6.1.3 Explicit Rack Model Description ..........................................................6-2 6.1.4 Reactivity Comparison of the Boral Rack Models ................................6-2 6.2 Fuel Assembly Modeling ...................................................................................6-3 6.3 Co-Resident Fuel Racks ...................................................................................6-3 6.4 BLEU versus Commercial Grade Uranium ........................................................6-4 6.5 General CASMO-4 Modeling Assumptions .......................................................6-4 7.0 Criticality Safety Analysis..............................................................................................7-1 7.1 Definition of the Reference Bounding and REBOL Lattices ...............................7-2 7.2 Storage Array Reactivity ...................................................................................7-3 7.3 Arrays of Mixed BWR Fuel Types .....................................................................7-3 7.4 Other Conditions ...............................................................................................7-4 7.4.1 Assembly Rotation ..............................................................................7-4 7.4.2 Assembly Lean ...................................................................................7-4 7.4.3 Blister Formation .................................................................................7-4 7.5 Normal Fuel Handling .......................................................................................7-5 7.6 Accident Conditions ..........................................................................................7-6 7.7 Manufacturing and Other Uncertainties .............................................................7-8 7.8 Determination of Maximum Rack Assembly k-eff (k95/95) ...................................7-9 8.0 References ...................................................................................................................8-1 Appendix A Sample CASMO-4 Input ................................................................................ A-1 Appendix B Reactivity Comparison for Assemblies Used at Browns Ferry........................ B-1 Appendix C KENO V.a Bias and Bias Uncertainty Evaluation ........................................... C-1 Appendix D CASMO-4 Qualification for In-Rack Modeling ................................................ D-1 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page iii Tables 2.1 Criticality Safety Limitations for ATRIUM 10XM Fuel Assemblies Stored in the Browns Ferry Plant Spent Fuel Storage Pools ........................................................2-4 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 ...........................................................................................................................3-5 4.1 ATRIUM 10XM Fuel Assembly Parameters ..................................................................4-3 4.2 Fuel Storage Rack Parameters.....................................................................................4-4 6.1 Comparison of Modeling Options for the Boral Rack ....................................................6-8 6.2 Impact of Channel Thickness on In-Rack Reactivity .....................................................6-9 6.3 Co-Resident Storage Rack Comparison .......................................................................6-9 6.4 In-Rack k Sensitivity to In-core Depletion Fuel Temperature .....................................6-10 6.5 In-Rack k Sensitivity to In-core Depletion Power Density ..........................................6-11 6.6 In-Rack k Sensitivity to In-Core Controlled Depletion ................................................6-12 7.1 Summary of CASMO-4 Maximum In-Rack Reactivity Results ....................................7-11 7.2 Summary of KENO V.a Maximum In-Rack Reactivity Results ....................................7-12 7.3 Manufacturing Reactivity Uncertainties .......................................................................7-13 Figures 2.1 Overview of the Browns Ferry SFP Criticality Safety Analysis ......................................2-6 2.2 ATRIUM 10XM Reference Bounding Assembly ............................................................2-7 4.1 Representative ATRIUM 10XM Fuel Assembly ............................................................4-5 4.2 Browns Ferry Spent Fuel Pool Layout ..........................................................................4-6 4.3 Schematic Representation of a Section of High Density Storage Rack .........................4-7 4.4 High Density Boral Storage Rack Geometry .................................................................4-8 6.1 Single Cell Model for the High Density Boral Rack .....................................................6-13 6.2 Explicit Geometry Model for High Density Boral Rack ................................................6-14 6.3 Schematic of Rack to Rack Interfaces ........................................................................6-15 6.4 BLEU versus Commercial Grade Uranium Reactivity Comparison .............................6-16 6.5 Impact of Void History Depletion on In-Rack k-infinity.................................................6-17 7.1 Evaluated Assembly Rotation Cases ..........................................................................7-14 7.2 Limiting Accident (Missing Boral Plate) .......................................................................7-15 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page iv Nomenclature AEC Atomic Energy Commission BAF bottom of active fuel BLEU blended low enriched uranium BOL beginning of life BORAL neutron absorber composed of boron dispersed within aluminum BWR boiling-water reactor CGU commercial grade uranium EALF the energy of the average lethargy causing fission FPM fuel preparation machine GDC general design criteria GWd energy unit, giga-watt-day H/X moderating ratio, atomic ratio of hydrogen (H) to fissile isotopes (X)

ISG interim staff guidance document (Reference 7) k-eff effective neutron multiplication factor (aka k-effective) k infinite lattice neutron multiplication factor (aka k-infinity)

LUA lead use assembly PLR part-length fuel rod NCS nuclear criticality safety NRC Nuclear Regulatory Commission, U.S. (also USNRC)

RAI request for additional information REBOL reactivity-equivalent at beginning of life (fresh fuel, no Gd2O3)

SFP spent fuel pool TAF top of active fuel

%TD percent of theoretical density

[ ] Square brackets enclose information that is proprietary to AREVA.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 1-1 1.0 Introduction This report presents the results of a criticality safety analysis performed for the Browns Ferry Nuclear Plant Units 1, 2, and 3 spent fuel storage pools. Each spent fuel pool has the same configuration including rack design and number of storage modules. This analysis is performed on a bounding basis and is applicable to all three spent fuel storage pools. The previous Nuclear Regulatory Commission (NRC) approved criticality safety evaluation is identified as Reference 1.

In this report, a reference bounding assembly has been defined to bound the reactivity of all past and current fuel assembly types delivered to the Browns Ferry Nuclear Plant. This reference bounding assembly is based on an AREVA Inc. (AREVA) ATRIUM'* 10XM fuel assembly. This analysis demonstrates that with the reference bounding assembly the pool k-eff remains below the 0.95 k-effective acceptance criterion established by the NRC.

• ATRIUM is a trademark of AREVA NP.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-1 2.0 Summary and Conclusions Criticality safety calculations have been performed and are documented herein for the Browns Ferry Nuclear Plant spent fuel storage pools. Figure 2.1 provides an overview of the various steps involved in this criticality safety analysis. The analysis flow in this figure begins at the bottom with the evaluation of the existing fuel inventory and ends at the top with the calculation of an array keff that meets the regulatory acceptance criterion of 0.95.

This criticality safety analysis is based on the use of a reference fuel assembly design that is bounding for (i.e., more reactive than) all fuel designs previously used or planned to be used at the Browns Ferry Nuclear Plant. The KENO V.a code was used for all calculations that do not require fuel depletion. The CASMO-4 code is used to compare lattice k values at peak reactivity conditions. The results of these comparisons are used to define the reference bounding lattices and the reactivity-equivalent at beginning of life (REBOL) lattices that are used in KENO V.a.

CASMO-4 is also used in defining a portion of the gadolinia manufacturing uncertainty.

Benchmarking against criticality experiments is included for the KENO V.a code and justification for the use of the CASMO4 code is also provided. More detail on methodology and code benchmark / justification is provided in Chapter 5 and Appendices C and D.

The calculations documented herein demonstrate that the ATRIUM 10XM reference bounding assembly design has been selected to be more reactive in an in-rack configuration than any of the current or past fuel assembly designs used in the Browns Ferry reactors. These comparisons are based upon actual GE 7x7, GE 8x8, GE 9x9, GE 10x10, and AREVA 10x10 (ATRIUM-10) lattice geometries and enrichments as detailed in Appendix B.* This criticality safety analysis shows that future ATRIUM 10XM assemblies meeting the storage requirements established in Table 2.1 can be safely stored with these previously manufactured assemblies.

The reference bounding assembly is defined with two U-235 enrichment / gadolinia concentration zones separated by the ATRIUM 10XM geometry transition at [ ] inches. The bottom enrichment and gadolinia zone is defined to extend up to this transition boundary and contains

[ ] fuel rods. The top enrichment / gadolinia zone extends from this geometric transition boundary to the top of the fuel assembly and contains [ ] fuel rods. These axial zones are

• Various LUAs were also evaluated.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-2 illustrated in Figure 2.2. Two REBOL lattices have been defined to represent the lattices of the reference bounding assembly in KENO calculations. The neutron multiplication factors of the REBOL lattices have been increased by greater than or equal to 0.010 k to address all uncertainties associated with defining these reactivity equivalent lattices.

This analysis includes manufacturing uncertainties for the ATRIUM 10XM fuel design and the fuel storage racks. In addition to the manufacturing uncertainties; code modeling uncertainties, reactivity increases due to accident or other conditions, and a one-sided tolerance multiplier are used to determine the 95/95 upper limit k-eff. The conditions and uncertainties assumed in this analysis are described in the various sections of Chapter 7.

This analysis demonstrates that the reference ATRIUM 10XM fuel assembly does not exceed an array k-eff of 0.95 in the Browns Ferry spent fuel storage pools. As defined in Table 2.1, ATRIUM 10XM fuel that contains equivalent or less enrichment and equivalent or higher Gd2O3 concentrations in the fuel zones depicted in Figure 2.2 can be safely stored in the Browns Ferry spent fuel storage pools. In addition, ATRIUM 10XM fuel that contains more enrichment and/or lower Gd2O3 concentrations than the reference assembly design can be safely stored provided each zone of the assembly is less reactive than the corresponding zone of the reference bounding assembly design (i.e., less than 0.8825 in-rack k-infinity for both zones in accordance with Table 2.1). This can be established using the storage rack model of the CASMO-4 lattice physics code as described in Appendix A.

This analysis supports the storage of channeled and unchanneled fuel assemblies including assemblies with the AREVA advanced fuel channel. Additionally, there is no limitation for bundle orientation or position in the storage cell since these are accounted for in the analysis.

To assure that the actual reactivity will always be less than the calculated reactivity, the following conservatisms have been included:

• The results are based on a moderator temperature of 4 °C (39.2 °F), which gives the highest reactivity for the fuel storage pool.

• Fuel assemblies are assumed to contain the high reactivity reference bounding lattices for the entire length of the assembly (i.e., natural uranium blankets are not modeled).

• Each lattice in each fuel assembly in the storage rack is assumed to be at its lifetime maximum reactivity level. There is no assumption of a specific burnup profile for the discharged assemblies. In other words, this is a peak reactivity analysis that does AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-3 not take credit for lower reactivity conditions associated with burnup past the maximum reactivity.

• The minimum Boron-10 areal density is used when modeling the Boral.

• The most limiting orientation or position of each assembly in its rack cell is accounted for in the analysis.

• Neutron absorption in fuel assembly structural components (i.e., spacers, tie plates, etc.) is neglected.

• The maximum reactivity value includes all significant manufacturing and calculational uncertainties.

• The 0.010 k uncertainty value applied when the REBOL lattice is defined is treated as a bias - introducing significantly more conservatism than if it had been treated as an uncertainty.*

• The fuel array is modeled as being infinite in all dimensions.

• An adder has been included to account for Boral blistering.

• The bias from the KENO V.a benchmark (Appendix C) has been increased to also bound trending conditions that were shown to be statistically insignificant.

This analysis demonstrates that all fuel assemblies previously delivered to the Browns Ferry Nuclear Plant can be safely stored in the spent fuel storage pools. Future ATRIUM 10XM fuel designs that meet the design requirements specified in Table 2.1 or that can be shown to be less reactive (on a lattice basis) than the reference bounding assembly can be safely stored in the Browns Ferry spent fuel pools. The k-eff determined herein for the reference assembly, including all uncertainties, biases, manufacturing tolerances and worst accident or other loading conditions is 0.928 (as detailed in Section 7.8 and Figure 2.1).

• As applied in this evaluation a k95/95 value of 0.928 is produced. If the 0.010 k uncertainty were not applied to the REBOL lattices and then treated as an additional uncertainty term in Section 7.8, the k95/95 value would decrease to 0.922.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-4 Table 2.1 Criticality Safety Limitations for ATRIUM 10XM Fuel Assemblies Stored in the Browns Ferry Plant Spent Fuel Storage Pools ATRIUM 10XM Fuel Configuration The ATRIUM 10XM fuel configuration is provided in Table 4.1.

Fuel Channels Fuel may be stored with or without fuel channels.

Fuel Design Limitations for Enriched Lattices*

The fuel may be stored in the spent fuel storage pool provided the enriched lattices are not more reactive than the reference bounding lattices. This can be demonstrated by meeting either of the following two requirements:

1. The U-235 enrichment and gadolinia loading levels must meet the requirements specified below and shown graphically in Figure 2.2. The dimensions represent fuel column height above the bottom of active fuel (BAF) and below the top of active fuel (TAF).

Above [ ] Maximum Lattice Average Enrichment, wt% U-235 4.70 Minimum Number of Rods containing Gd2O3 8 Minimum wt% Gd2O3 in these Gd Rod 3.5 At and below [ ] Maximum Lattice Average Enrichment, wt% U-235 4.70 Minimum Number of Rods containing Gd2O3 8 Minimum wt% Gd2O3 in these Gd Rod 3.919 These eight gadolinia rods cannot be loaded on the perimeter of the lattice or adjacent to the water channel. An equivalent of 2 gadolinia rods must be loaded along each side.

Gadolinia is not required in natural Uranium blankets and there are no restrictions on the number, concentration, or placement of any additional gadolinia rods.

Or,

2. The lattice average enrichment is less than 5.0 wt% U-235, and the k of each enriched lattice does not exceed the following in-rack k values at any point during its lifetime. (The CASMO-4 storage rack model that must be used for this calculation is defined in Appendix A and the transition between top and bottom lattice geometries occurs at [ ] inches from the bottom of the fueled length.)

Zone Lattice Geometry Distance from BAF Max. in-rack k 2 10XMLCT [ ] [ ] to TAF 0.8825 1 10XMLCB [ ] 0" to [ ] 0.8825

• These requirements describe the reference bounding lattices shown in Figure 2.2 and Table 7.1.

Two face adjacent gadolinia rods count as a single rod.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-5 Table 2.1 Criticality Safety Limitations for ATRIUM 10XM Fuel Assemblies Stored in the Browns Ferry Plant Spent Fuel Storage Pools (Continued)

Spent Fuel Storage Rack The spent fuel storage rack design parameters and dimensions are provided in Table 4.2.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-6

[

]

Figure 2.1 Overview of the Browns Ferry SFP Criticality Safety Analysis AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-7 TAF (Top of Active Fuel)

Zone 2 XMLCT-470UL-8G35

[ ] PLR fueled Boundary Zone 1 XMLCB-470UL-8G3919 0.0" BAF (Bottom of Active Fuel)

Figure 2.2 ATRIUM 10XM Reference Bounding Assembly AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-1 3.0 Regulatory Criticality Safety Criteria and Guidance Section 9.1.1 of the Standard Review Plan (Reference 2) identifies the regulatory requirements and associated acceptance criteria considered to be applicable to criticality safety analyses.*

Since this analysis does not support a change in the facility only the requirements specific to the criticality safety analysis apply. The primary requirements relevant to this analysis are General Design Criteria 62 and portions of 10 CFR 50.68, Reference 3. Although it is not specifically cited by SRP 9.1.1, General Design Criteria 5 is potentially of interest in a spent fuel criticality analysis.

The Browns Ferry units were not designed or licensed to the General Design Criteria provided in 10 CFR 50 Appendix A. Instead Appendix A of the Browns Ferry FSAR (UFSAR) provides a description of conformance to the AEC Proposed General Design Criteria. For Browns Ferry, the corresponding licensing basis applicable criteria are Criterion 4, Sharing of Systems, and Criterion 66, Prevention of Fuel Storage Criticality.

Criterion 4 (similar to GDC 5) addresses the sharing of systems important to safety specifically to ensure that the ability to perform their safety function is not significantly impaired. The existence of a transfer canal allows for the transfer of fuel bundles between the Units 1 and 2 spent fuel pools (i.e., the only shared components). All three of the spent fuel pools at the Browns Ferry plant have essentially the same configuration and use the same rack designs, as described in Section 4. The previously manufactured fuel evaluation identifies the most limiting fuel from all three pools and conservatively applies this in the definition of common reference bounding lattices to be used for all three pools. This bounding treatment helps to ensure that the ability of the spent fuel pool racks to maintain subcriticality is not impaired when fuel transfer between pools occurs and the intent of AEC Proposed Criterion 4 (and GDC 5) is therefore met.

Criterion 66 (similar to GDC 62) specifies that criticality of fuel in handling or storage will be prevented by physical systems or processes with the preference for geometrically safe configurations. There is no physical change being implemented that affects the configuration of the spent fuel storage system (i.e. no change to the systems, components, or structures that comprise the spent fuel storage system). The purpose of this analysis is to provide assurance

• SRP 9.1.1 is used as the basis for discussion of general requirements for criticality safety analyses in this report. This context does not represent a commitment on the part of the licensee in regard to conformance with this section of the Standard Review Plan.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-2 that criticality will not occur within the basis of the existing spent fuel storage configuration for both previously manufactured (or planned) fuel and ATRIUM 10XM designs to be provided in the future; therefore the intent of Criterion 66 (and GDC 62) is met.

10 CFR 50.68*(a) requires that a licensee must either: 1) maintain monitoring systems in accordance with 10 CFR 70.24 to reduce the consequences of a criticality accident, or 2) adhere to the requirements of 10 CFR 50.68(b) to reduce the likelihood that a criticality accident will occur. Browns Ferry complies with the requirements of part (b) of 10 CFR 50.68. The role of this criticality safety analysis in meeting the specific requirements for each of the 10 CFR 50.68(b) requirements is discussed below:

1) Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

Technical Specification 4.3.1.1(a) requires that a k-effective of less than or equal to 0.95 must be maintained with unborated water. This analysis establishes the SFP storage requirements that meet this licensing requirement. Fuel handling has also been addressed by this analysis.

2) The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

This requirement does not apply because this is not a fresh fuel storage rack criticality analysis.

3) If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum

• Reference 3.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-3 moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.

This requirement does not apply because this is not a fresh fuel storage rack criticality analysis.

4) If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

This criticality safety analysis is being performed specifically to show that this requirement has been met. The applicable requirement is a k-effective of 0.95 at a 95 percent probability with a 95 percent confidence level because Browns Ferry is a BWR site with unborated water in the SFP. This requirement is also enforced in section 4.3.1.1(a) in the Technical Specifications for each Browns Ferry unit. The analysis described in this report demonstrates that the calculated k95/95 value meets this requirement.

5) The quantity of SNM, other than nuclear fuel stored onsite, is less than the quantity necessary for a critical mass.

This requirement does not apply because this analysis only addresses nuclear fuel storage in the SFP.

6) Radiation monitors are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions.

This requirement does not apply because this is a criticality analysis only.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-4

7) The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0) percent by weight.

This criticality safety analysis establishes maximum allowable enrichments below the regulatory requirement and therefore complies with the intent of this requirement.

8) The FSAR is amended no later than the next update which § 50.71(e) of this part requires, indicating that the licensee has chosen to comply with § 50.68(b).

The licensee has included the required 10CFR 50.68(b) compliance statement in section 10.3 Spent Fuel Storage of the Browns Ferry FSAR.

This criticality safety analysis complies with the intent of all of the applicable sections of 10 CFR 50.68(b).

Based upon the discussion above, this analysis complies with the intent of the Proposed AEC General Design Criteria 4 and 66 as well as 10 CFR 50.68(b).

The USNRC has recently issued document DSS-ISG-2010-01 Revision 0 (Reference 7) that provides interim staff guidance (ISG) for the review of spent fuel criticality safety analyses.

Table 3.1 provides a top level summary discussion regarding the compliance of this criticality safety analysis to the ISG document. Where possible, this discussion includes a cross-reference to where specific items identified in the ISG are addressed within this criticality safety analysis report.

The following sources provide additional guidance in meeting the aforementioned regulatory requirements:

• Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, also known as the Kopp letter this was issued by the NRC in 1998 (Reference 6).

• OT Position for the Review and Acceptance of Spent Fuel Storage and Handling Applications, issued by the NRC in 1978 and amended in 1979 (Reference 5).

• ANSI/ANS American National Standard 8.17-1984 (Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors) issued by the American Nuclear Society, January 1984 (Reference 4).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-5 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.1 Fuel Assembly Selection the staff should review the The lattices of the ATRIUM 10XM reference Appendix B submittal to verify that it bounding assembly are demonstrated to be demonstrates that the NCS more reactive than the lattices of any analysis adequately bounds previously loaded fuel assembly, including all designs, including variations due to damaged and modified variations within a design. assemblies.

the staff should verify each As discussed above, the ATRIUM 10XM Section 2.0 application includes a portion reference bounding assembly is shown to of the analysis that bound all previous designs. Compliance with Appendix B demonstrates that the fuel the requirements listed in Table 2.1 ensures assembly used in the that future ATRIUM 10XM assemblies analysis is appropriate for the remain bounded by this evaluation.

specific conditions.

IV.1.a Use of a single limiting fuel The use of the ATRIUM 10XM reference Section 2.0 assembly design should be bounding assembly (and corresponding assessed, lattices) is justified as described above. Appendix B IV.2 Depletion Analysis simulates the use of fuel in This evaluation does not directly use the Sections 7.0 a reactor. These depletion depletion based isotopic number density and 7.1 simulations are used to values in KENO. The CASMO-4 based create the isotopic number incore depletion is used to establish the Appendices densities used in the inrack lifetime maximum reactivity condition B&D criticality analysis. of the reference bounding lattices. Reactivity equivalent at beginning of life (REBOL) lattices are then defined for use in the KENO calculations. The REBOL lattices are defined with a conservative bias to address the uncertainty in the CASMO-4 depletion process and reactivity equivalence method.

The definition of the reference bounding and REBOL lattices are described in more detail in Sections 7.0, 7.1, and Appendix B.

Appendix D provides details on the treatment of the depletion uncertainty.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-6 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.2.a Depletion Uncertainty An overall CASMO-4 uncertainty reflecting Appendix D calculational and depletion based isotopic an uncertainty equal to 5 uncertainties is defined in Section D.4. This percent of the reactivity value is bounded by the 0.010 k bias term decrement to the burnup of applied during the reactivity equivalence interest is an acceptable calculation. Two independent estimates of assumption. the depletion uncertainty were used in this evaluation. One of these methods is should only be construed consistent with the 5% reactivity decrement as covering the uncertainty in described in Reference 7 (except that it the isotopic number includes an additional component for the densities gadolinia uncertainty).

IV.2.b Reactor Parameters Sensitivity comparisons are included in Section 6.5 Section 6.5 to show that reasonable the staff should verify that parameters have been used in the depletion each application includes a calculations. The parameters evaluated portion of the analysis that include:

demonstrates that the reactor parameters used in the Fuel Temperature (Assumption 2, Table depletion analysis are 6.4); Moderator Temperature/Void appropriate for the specific History ( Assumption 3, Figure 6.5);

conditions. Power Density (Assumption 4, Table 6.5); and Rodded Depletion (Assumption 7, Table 6.6)

IV.2.c Burnable Absorbers Only integral burnable absorbers have been Table 2.1 used in the Browns Ferry reactor and they the staff should verify that have been modeled appropriately in Section 7.1 each application includes a Appendix B. The placement of the 8 portion of the analysis that gadolinia rods in the reference bounding Appendix B demonstrates that the lattices have been selected to produce a treatment of burnable high reactivity condition. Table 2.1 requires absorbers in the depletion that all enriched lattices of future ATRIUM analysis is appropriate for the 10XM assemblies contain a minimum specific conditions. number of absorber rods with a minimum concentration level or that a CASMO-4 k less than the applicable reference bounding lattice be demonstrated.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-7 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.2.d Rodded Operation Assumption 7 of Section 6.5 addresses Section 6.5 rodded depletion. The use of uncontrolled the staff should verify that depletion at rated power conditions is shown each application includes a to bound depletion at controlled conditions portion of the analysis that for the ATRIUM 10XM reference bounding demonstrates its treatment of lattices.

rodded operation is appropriate for its specific conditions.

IV.3 Criticality Analysis IV.3.a Axial Burnup Profile This evaluation uses the lifetime maximum Section 7.0 reactivity of each lattice of the reference the staff should verify that bounding assembly as discussed in Section each application includes a 7.0. Therefore, there is no burn-up profile portion of the analysis that assumption.

demonstrates its treatment of axial burnup profile is appropriate for its specific conditions.

IV.3.b Rack Model The modeling of the spent fuel racks have Section 6.1 been explicitly addressed in Table 6.3. Appendix D the staff should verify that Comparisons in Table 6.1 demonstrate that each application includes a the infinite 2x2 model is more reactive than portion of the analysis that the explicit model. Comparisons in Table demonstrates that the rack 6.1 and Appendix D show that the 2x2 model analysis used in its model agrees well with the single cell model submittal is appropriate for its used in CASMO-4.

specific conditions.

IV.3.b.i The dimensions and The rack dimensions and materials in Table Section 4.2 materials of construction 4.2 were derived from the licensees design should be traceable to documents.

licensee design documents.

The Boral is modeled using the licensees IV.3.b.ii The efficiency of the neutron Section 4.2 design minimum Boron-10 areal density.

absorber should be Neutron self shielding and streaming are established, especially addressed in Section 4.2.

considering the potential for self-shielding and streaming.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-8 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.3.b.iii Any degradation should be A conservative blister model has been Sections modeled conservatively, incorporated to account for Boral blistering. 7.4.3 & 7.8 consistent with the certainty with which the material condition can be established.

IV.3.c Interfaces The 13x13 and 13x17 Boral racks have Sections 4.2 been shown to be less reactive than the 2x2 & 6.3 the staff should verify that infinite array model.

each application includes a portion of the analysis that demonstrates that the interface analysis used is appropriate for its specific conditions.

IV.3.c.i Absent a determination of a There is no significant difference between Section 6.3 set of biases and the 13x13 racks and the 13x17 racks.

uncertainties specifically for the combined interface model, use of the maximum biases and uncertainties from the individual storage configurations should be acceptable in determining whether the keff of the combined interface model meets the regulatory requirements.

IV.3.d Normal Conditions Translation and orientation variations of the Sections assemblies within the storage racks are 7.4.1, 7.4.2, the staff should verify that considered in Sections 7.4.1 and 7.4.2. and 7.5 each application includes a The fuel handling considerations for normal portion of the analysis that conditions are addressed in Section 7.5.

demonstrates that the NCS analysis considers all appropriate normal conditions for its specific conditions.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-9 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.3.e Accident Conditions The accident conditions have been Section 7.6 evaluated in Section 7.6.

The reviewer should verify all credible accident conditions are addressed.

IV.4 Criticality Code Validation IV.4 The proposed analysis The criticality benchmark is shown in Appendix C methods and neutron cross- Appendix C. Since this is a fresh fuel section data should be equivalent evaluation, only critical benchmarked, by the analyst experiments for fresh fuel have been or organization performing included in the benchmark data set.

the analysis, by comparison with critical experiments. ...

The critical experiments should include configurations having neutronic and geometric characteristics as nearly comparable to those of the proposed storage facility as possible.

IV.4.a Area of Applicability The area of applicability is defined by the Section 5.1 criticality benchmark comparisons provided Appendix C the staff should verify that in Appendix C. Section 5.1 also provides a applications demonstrate that summary of this validation and addresses the validation fully covers the the area of applicability for this Browns Ferry area of applicability for their spent fuel storage pool criticality safety specific SFP; analysis.

HTC benchmarks are not included in the IV.4.a.i The reviewer should verify Appendix C validation set since this is a fresh fuel any validation used for SNF reactivity equivalent evaluation. The appropriately considers treatment of actinides and fission products is actinides and fission part of the CASMO-4 depletion uncertainty products. NUREG/CR-6979, addressed in Appendix D.

Evaluation of the French Haut Taux de Combustion However, the addendum to Appendix C (HTC) Critical Experiment compares the impact on the KENO bias and Data, issued September uncertainties if appropriate benchmark 2008 experiments from the HTC criticals were included. This comparison shows that a more conservative result is obtained without inclusion of the HTC criticals.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-10 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.4.a.ii Experiments should be The criticality benchmark data shown in Appendix C appropriate to the system Appendix C meets the requirements being analyzed. expressed in the ISG.

IV.4.a.iii The reviewer should The criticality benchmark dataset has been Appendix C

.{review the selection of selected to provide a balanced benchmark data} representation of the spent fuel pool environment. It is shown in Appendix C.

IV.4.a.iv The reviewer should ensure The criticality benchmark is shown in Appendix C that the experiments are not Appendix C.

all highly correlated, e.g.

critical configurations performed with the same fuel rods at the same facility.

IV.4.b Trend Analysis The trending analysis is performed in Appendix C Appendix C.

the staff should verify that each application includes a portion of the analysis that demonstrates that the trend analysis used in its validation is appropriate for its specific conditions.

IV.4.c Statistical Treatment The benchmark validation suite in Appendix Appendix C C follows the guidance given in NUREG/CR-the staff should verify that 6698 with respect to using the variance each application includes a about the mean, confidence factors, and the portion of the analysis that treatment of non-normal distributions.

demonstrates that the statistical treatment used in its validation is appropriate for its specific conditions.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-11 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.4.d Lumped Fission Products The primary components of the Browns Section 7.1 Ferry nuclear criticality safety (NCS) the staff should verify that analysis include the use of the CASMO-4 each application that includes code in the definition of the reference lumped fission products bounding and REBOL lattices followed by includes a portion of the the actual NCS calculations with KENO V.a analysis that demonstrates (using the defined REBOL lattices). While that the lumped fission CASMO-4 does include the use of lumped products used in its validation fission products, they are not credited in the are appropriate for its specific definition of the reference bounding lattices.

conditions. Therefore, the KENO calculations and k95/95 result are conservative since the lumped fission products have been removed.

IV.4.e Code-to-Code Comparisons Code-to-code comparisons are not used in Appendix C the validation of KENO V.a - the code used Appendix D the use of a code-to-code for the criticality analysis.

comparison for validating criticality codes is outside the The only use of code-to-code comparisons scope of this ISG. is for the depletion code, CASMO-4. This use is limited to perturbation calculations used to quantify the CASMO-4 calculational uncertainty relative to KENO V.a.

IV.5 Miscellaneous IV.5.a Precedents Although not specifically cited, the approach N/A taken in this spent fuel pool criticality safety the staff should verify that analysis is similar to a previous SFP for cited precedents, the criticality analysis recently reviewed and application includes a portion approved by the USNRC (Reference of the analysis that accession numbers ML092810281 and demonstrates the ML101650230).

commonality of the precedent to the submittal, with any Some changes were incorporated to directly differences identified and address USNRC concerns identified in the justified with respect to the SER accepting the above submittal use of the precedent. (ML110250051) and to provide closer compliance to the staff guidance document.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-12 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.5.a Specifically, Continued

• The criticality benchmark data suite has been modified to remove soluble boron and MOX benchmarks.

• The previous submittal content was split between a submitted report and additional answers to USNRC requests for additional information. This information content has been reformatted into a single report.

• Differences in modeling, primarily to address differences in the plant specific rack designs.

• The CASMO-4 lumped fission products are not credited in the in-rack k values for the reference bounding lattices when the reactivity equivalence comparison is being performed.

IV.5.b References The analysis uses references N/A appropriately.

the NRC reviewer should verify that references cited in the application are used in context and within the bounds and limitations of the references. Any extrapolation outside the context or bounds of the reference should be demonstrated as appropriate.

IV.5.c Assumptions Modeling assumptions have been explicitly Section 6.0 addressed in the report.

applications should explicitly identify and justify all assumptions used in their applications.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-1 4.0 Fuel and Storage Array Description A number of different assembly types have previously been loaded in the Browns Ferry spent fuel pools with lattice geometries ranging from 7x7 to 10x10. This includes variations in the type and number of water rods and part length fuel rods. The AREVA ATRIUM 10XM fuel product line is planned for use in future reloads. For this reason, the ATRIUM 10XM reference bounding assembly design forms the basis for demonstrating that the maximum k-eff of the spent fuel pool storage array remains less than 0.95.

4.1 Fuel Assembly Design The ATRIUM 10XM fuel assembly is a 10x10 fuel rod array with an internal square water channel offset in the center of the assembly (taking the place of nine fuel rod locations). The assembly contains part-length fuel rods (PLR); therefore, the top lattice geometry will apply above the PLR fueled boundary and the bottom lattice geometry will apply below the PLR fueled boundary. The ATRIUM 10XM mechanical design parameters are summarized in Table 4.1 and a representation of the ATRIUM 10XM assembly design is provided in Figure 4.1. The ATRIUM 10XM fuel in the Browns Ferry Nuclear Plant uses the AREVA advanced (i.e.,

thick/thin) fuel channel design.

4.2 Fuel Storage Racks Each of the Browns Ferry spent fuel pools provide the capability of storing 3471 fuel assemblies. Each pool contains 14 - 13x13 high density Boral storage rack modules and 5 13x17 high density Boral storage rack modules. The dimensional parameters for these racks are given in Table 4.2 and the pool arrangement is shown in Figure 4.2. The layout in all three pools is essentially the same except that the Unit 1 pool has a mirror symmetric layout when compared to the Unit 2 or Unit 3 pool.

A transfer canal is provided to join the Unit 1 and 2 pools. This transfer canal is the same depth as the transfer slot between the reactor well and the fuel pool. The transfer canal has a gate at each end so that the fuel pools can be isolated. There is no corresponding transfer canal for the Unit 3 pool.

Each high density Boral rack module is composed of alternating or staggered stainless-steel square container tubes. This arrangement results in only one container-tube wall between AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-2 adjacent fuel assemblies, as illustrated in Figure 4.3 and Figure 4.4. Each container-tube wall has a core of Boral sandwiched between inner and outer surfaces of stainless steel. The Boral core is made up of a central segment composed of a dispersion of boron carbide in aluminum.

This central segment is clad on both sides with aluminum. These stainless steel container tubes are closure welded with vent holes to prevent the buildup of hydrogen gas. The completed storage tubes are fastened together by angles welded along the corners and attached to a base plate to form storage modules. These modules are designed to be free standing with low-friction between the module support and pool floor liner.

Note on the Efficacy of Boral: In a water environment, neutron scattering ensures that neutrons approach the Boral from a full range of incident angles. This minimizes the potential for neutron streaming and reduces the significance of self-shielding.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-3 Table 4.1 ATRIUM 10XM Fuel Assembly Parameters Parameter Value Fuel Assembly Fuel Rod Array 10x10 Fuel Rod Pitch, in. [ ]

Number of Full Length Fuel Rods [ ]

Number of Part Length Fuel Rods [ ]

Location of Part Length Fuel Rods See Figure 6.1 Water Channel 1 Fuel Rods Fuel Material UO2 Pellet Density, % of Theoretical Density (%TD) [ ]

Pellet Diameter, in. [ ]

Pellet Void Volume, % [ ]

Cladding Material Zircaloy Cladding OD, in. [ ]

Cladding ID, in. [ ]

Internal Water Channel Outside Dimension, in. [ ]

Inside Dimension, in. [ ]

Channel Material Zircaloy Fuel Channel (standard 100 mil)

Outside Dimension, in. [ ]

Inside Dimension, in. [ ]

Channel Material Zircaloy Fuel Column Lengths Distance from the bottom of the fuel to the top of the fuel in the part length fuel rods, in. [ ]

Total Fueled Length, in. [ ]

• Criticality safety analysis is also valid for lower fuel densities. The analysis uses the effective stack density which is a combination of the pellet density and the pellet void volume.

The conclusions in this report are also valid for advanced fuel channels (see Section 6.2).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-4 Table 4.2 Fuel Storage Rack Parameters Parameter Value High Density Boral Racks Boral B-10 areal density, g/cm2 0.013 minimum Rack Box OD, in. 6.653 +/- 0.04 Box material Stainless steel Inner rack box wall thickness, in. 0.0355 +/- 0.004 Box material Stainless steel B4C plate thickness, in. 0.076 +/- 0.005 plate material B4C and aluminum clad in two 0.010" aluminum sheets width, in 6.20* +/- 0.03 height, in 152.00 Outer rack box wall thickness, in. 0.090 +/- 0.008 Box material Stainless steel Rack cell pitch, in. 6.563 +/- 0.03 Closure plate thickness, in. 0.125 material Stainless steel Rack to Rack Spacing, in. 2.33

• 6.20 has previously been represented as a minimum width value. For consistency with a similar storage rack design, 6.20 will be treated as a nominal value with the indicated uncertainty.

Rack to Rack spacing is the distance from the outside surface of adjacent closure plates. (This value is derived from a 1.875" spacing at the rack module base).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-5

[

]

Figure 4.1 Representative ATRIUM 10XM Fuel Assembly (Assembly length and number of spacers has been reduced for pictorial clarity)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-6 Fuel Prep. Fuel Prep.

Sipping Can Storage Sipping Can Storage Machine Machine N Cask Pad Area Figure 4.2 Browns Ferry Spent Fuel Pool Layout (not to scale)

(Unit 2 and 3 layout shown, Unit 1 is mirror symmetric with the cask pad area on top)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-7 Stainless Steel Closure Plate (outside of rack)

Boral (Boron Carbide in OUTSIDE Aluminum) clad with aluminum CORNER OF RACK Stainless Steel (inner box)

Boral Tube Open Cell Boral Tube Stainless Steel (outer box)

Open Cell Boral Tube Open Cell Stainless Steel Closure Plates (outside of rack)

Boral Tube Open Cell Boral Tube Open Cell Boral Tube Open Cell Figure 4.3 Schematic Representation of a Section of High Density Storage Rack (not to scale)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-8 Section A-A 0.010" Al clad outside inside 0.090" +/- 0.008" 0.056" SS Boral 0.0355" +/- 0.004" core SS 6.563" +/- 0.03" cell pitch A A 6.653" +/- 0.04" outside dimension 6.563" +/- 0.03" cell pitch Figure 4.4 High Density Boral Storage Rack Geometry (not to scale)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 5-1 5.0 Calculation Methodology The spent fuel storage criticality safety evaluation is performed with the KENO V.a Monte Carlo code, which is part of the SCALE 4.4a Modular Code System (Reference 8). The SCALE driver module CSAS25 uses the ENDF/B-V 44 energy group data library. It also uses modules BONAMI-2 and NITAWL to perform spatial and energy self-shielding adjustments of the cross sections for use in KENO V.a. AREVA has benchmarked KENO V.a in accordance with NUREG/CR-6698 (Reference 9) using critical experiments related to the storage of fuel assemblies in water - including neutron absorbing materials such as stainless steel and Boral.

For applications using the 44 energy group data libraries, a KENO V.a bias magnitude of 0.0075 and a standard deviation of 0.0027 will be used (see Appendix C).

KENO V.a is run on the AREVA scientific computer cluster using the Linux operating system.

The hardware and software configurations are governed by AREVA procedures to ensure calculational consistency in licensing applications. The code modules are installed on the system and the installation check cases are run to ensure the results are consistent with the installation check cases that are provided with the code. The binary executable files are put under configuration control so that any changes in the software will require re-certification. The hardware configuration of each machine in the cluster is documented so that any significant change in hardware or operating system that could result in a change in results is controlled. In the event of such a change in hardware or operating system, the hardware validation suite is rerun to confirm that the system still performs as it did when the code certification was performed.

In this analysis the SCALE 4.4a code system is employed to:

• Calculate Dancoff coefficients.

• Calculate absolute k-effective results.

• Evaluate accident conditions, alternate loading conditions, and manufacturing tolerance conditions.

The CASMO-4 code is used when conditions require fuel and gadolinia depletion. CASMO-4 is a multigroup, two-dimensional transport theory code with a rack geometry option that allows typical storage rack geometries to be defined on an infinite lattice basis. This code is used for fuel depletion and relative reactivity comparisons in a manner that is consistent with AREVAs NRC approved CASMO-4 / MICROBURN-B2 methodology (Reference 10). CASMO-4 has been approved at Browns Ferry Nuclear Plant for BWR calculations and is included as a AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 5-2 methodology reference (via Reference 10) in Section 5.6.5.B of the Browns Ferry Nuclear Plant Technical Specifications. The CASMO-4 computer code is controlled by AREVA procedures and the version used in this analysis meets the requirements of Reference 10.

In this analysis CASMO-4 is employed to:

• Perform in-core isotopic depletion at characteristic void history levels, [ ] for bottom geometry lattices and [ ] for top geometry lattices. Use of CASMO-4 for in-core depletion is consistent with its application in EMF-2158(P)(A) (Reference 10).

• Perform in-rack k assessments to identify the most reactive lattices.

• Define lattices for a reference bounding assembly that represent the maximum reactivity condition supported by the analysis.

• Define the reactivity equivalent at beginning-of-life (REBOL) lattices with fresh fuel and no gadolinia for the subsequent KENO V.a base case criticality calculations. Note that for the REBOL lattices, the U-235 content is manually adjusted upward until the REBOL k is at least 0.010 k greater than the lattices of the reference bounding assembly. This 0.010 k is used to account for all uncertainties associated with defining the REBOL lattices - including calculational and depletion uncertainties of the CASMO-4 code as discussed in Appendix D.

• Evaluate a component of the manufacturing uncertainty for gadolinia content (i.e., the depletion component). This evaluation is needed because changes in gadolinia content affect reactivity more near peak reactivity than at beginning of life.

5.1 Area of Applicability Table C.6 in Appendix C shows the ranges of key parameters represented in the KENO V.a benchmark analysis. Parameters such as rectangular lattices of zircaloy clad UO2 fuel rods in a pool of water with stainless steel and boron are sufficiently general to not require comparison.

The remaining parameters are compared in the following table and show that the KENO V.a portion of this analysis has been performed within the range of experimental conditions used in the KENO V.a benchmark.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 5-3 Parameter Benchmark Values Values in this Analysis Enrichment (wt% U-235) 2.35 to 4.74 3.2 to 3.4 Fuel Rod Pitch (cm) 1.26 to 2.54 1.295 Moderating Ratio (H/X) 110 to >400 115 to 122 Energy of the Average 0.060 to 0.247 0.148 to 0.245 Lethargy Causing Fission (eV)

For the CASMO-4 qualification, ATRIUM 10XM fuel lattices were modeled using the Browns Ferry Nuclear Plant limiting storage rack geometry. Therefore, the CASMO-4 calculations performed for this evaluation are within the area of applicability of the comparisons shown in Appendix D.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-1 6.0 Modeling Options and Assumptions The following sections describe the primary modeling simplifications and assumptions used in this analysis including discussion of impact on in-rack reactivity.

6.1 Geometric Modeling of the High Density Boral Rack The geometry of the high density spent fuel storage racks includes an arrangement of staggered or alternating Boral tubes, as shown in Figure 4.3 and Figure 4.4. As a minimum, this rack requires an array of 2 tube cells and 2 non-tube cells for explicit modeling. The rack models described below were implemented in KENO V.a and the reactivity results are provided in Table 6.1. These models use infinite periodic boundary conditions in the x, y, and z directions.

6.1.1 Single Cell Model Description The primary simplifying assumptions can be generally described as follows:

• Boral Plate: The Boral plate is modeled as boron-10 only; i.e., the aluminum, carbon, and boron11 in the core of the plate and the aluminum clad on the outside of the plate are not included in this model. The location of the Boral is shifted to be between storage cells so that half of the actual thickness is assigned to each cell wall. The plate is assumed to extend to the corners of the storage cell (i.e., the water gap in the corners of the Boral tube is not modeled). Neglecting the non- boron-10 components of the Boral is slightly conservative because the neglected materials are relatively weak neutron absorbers. Extending the Boral plate to the corners is expected to have the opposite effect since it introduces a small additional amount of a strong neutron absorber.

• Stainless Steel Channels: One half of the total inner and outer stainless steel channels were combined in the model and assumed to make up the inside surface of the storage cell. The impact of this modeling simplification is expected to be minor since the amount of stainless steel is conserved and it still surrounds the Boral plate.

• Cell Pitch and Water Gaps: Average cell pitch and average water gap values are used in this model. This helps maintain the accuracy of this simplified model.

This is the model used in the CASMO-4 calculations. Figure 6.1 provides an illustration of the geometry for the single cell model.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-2 6.1.2 Explicit Storage Cell Model Description KENO V.a allows for more detailed modeling of the storage rack geometry than is possible with CASMO-4. The primary modeling changes in comparison to the single cell model are:

• Storage Geometry: The explicit model for KENO V.a is composed of a 2x2 array with two Boral tube storage cells and two open or non-tube cells.

• Boral Plate: The plate is modeled using the nominal width and the corner region is modeled as water. For comparison purposes one solution is provided using Boron10 only and the second solution models all components of the Boral plate; i.e., Boron, Carbon, and Aluminum.

• Cell Pitch: The average assembly pitch is modeled.

Figure 6.2 provides an illustration of the geometry for the KENO V.a explicit model.

6.1.3 Explicit Rack Model Description The high density Boral storage rack modules have an odd number of rows and columns. For this reason, each module has a Boral tube in each corner. When the racks are placed together, the cells in the adjacent rack have the same geometric configuration (i.e., a Boral cell is face adjacent to another Boral cell and an open cell is face adjacent to another open cell). As shown in Figure 6.3, some cells have two Boral plates between adjacent assemblies and some cells have no Boral material between assemblies. Details associated with the individual storage racks were modeled as described below.

• Storage Geometry: The explicit model from Section 6.1.2 is expanded to a 13x13 array with tube cells in each corner.

• Stainless steel closure plates are approximated for non-tube cells along the perimeter of the rack.

• The nominal rack to rack water gap* is modeled.

6.1.4 Reactivity Comparison of the Boral Rack Models Table 6.1 provides KENO V.a results for the single cell and more explicit geometry models.

Neglecting the single cell model, these results indicate that the explicit 2x2 storage cell model with the Boral modeled as boron-10 only produces the most conservative result. Other conclusions from this comparison are also listed below:

• The single cell model provides a good representation of the reactivity of the Boral rack.

• It is conservative to use a boron-10 only model for the Boral.

• 2.33 inches between the outer surfaces of the closure plates.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-3

• It is conservative to neglect the water gaps and closure plates between the storage rack arrays, i.e., the infinite cell model is more conservative than the 13x13 rack model.

The boron-10 only, explicit 2x2 model with periodic boundary conditions in all directions is represented as 0.897 and will be used to represent the reactivity level of the Boral racks. This k value is about 0.015 k more reactive than the result from the actual storage rack model (the 13x13 rack model with explicit B4C and finite axial boundary conditions in Table 6.1).

6.2 Fuel Assembly Modeling The CASMO-4 modeling of the previously manufactured fuel is performed using the actual lattice dimensions, enrichment, gadolinia loading, and channel type for each specific fuel product line. The KENO V.a in-rack calculations for the limiting ATRIUM 10XM fuel have been performed assuming a uniform 100 mil fuel channel.

A sensitivity calculation was performed with various channel thicknesses with the results summarized in Table 6.2. This analysis shows that in-rack reactivity generally increases with increasing fuel channel wall thickness. The increase in wall thickness results in an increase in channel mass and wall cross-sectional area which in turn results in larger water displacement.

The AREVA advanced channel design for ATRIUM 10XM fuel is thicker at the corners with a thinner wall along the sides and has a cross-sectional area that falls between the 80 mil and 100 mil channels. Consequently, an ATRIUM 10XM assembly modeled with a uniform 100 mil fuel channel is more reactive than an assembly without a fuel channel, an assembly with a uniform 80 mil fuel channel, and an assembly with the advanced fuel channel.

Zircaloy has been modeled in KENO as pure zirconium. Neglecting the neutron absorption of the alloying elements (primarily tin, iron, chromium, and nickel) is slightly conservative. In addition, the presence of activated corrosion and wear products (CRUD) is neglected because most of these compounds have higher neutron absorption cross sections than water.

6.3 Co-Resident Fuel Racks As shown in Figure 4.2 , the Browns Ferry spent fuel pools contain a combination of high density Boral racks (13x13 and 13x17). The in-rack k values for these storage rack types are compared in Table 6.3 with the limiting water temperature specified. This comparison shows that the high density Boral racks have similar reactivity characteristics.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-4 Review of Section 6.1.4 confirms that the 2x2 Boral infinite cell model is the most limiting overall; therefore, it will be used as the bounding representation of all possible rack configurations in the Browns Ferry spent fuel pool. Based on these comparisons, the 0.897 k result will be used as the primary basis for the k95/95 calculation.

6.4 BLEU versus Commercial Grade Uranium ATRIUM 10XM reloads for Browns Ferry may contain both commercial grade uranium and BLEU* fuel. The reference bounding and REBOL lattices used in this analysis are conservatively based upon commercial grade uranium. The previously manufactured BLEU lattices have been explicitly modeled using the U-234 and U-236 content corresponding to the feed material used in each reload batch.

The U-236 content in the BLEU fuel acts as a neutron absorber and reduces the lattice reactivity compared to an equivalent lattice composed of commercial grade uranium. This is illustrated in Figure 6.4 which compares the in-rack reactivity for the reference bounding lattices with and without BLEU fuel. As can be seen in this figure, the use of BLEU fuel significantly reduces lattice reactivity. The use of commercial grade uranium in the reference bounding and REBOL lattices is therefore conservative. With this level of conservatism, no BLEU specific manufacturing uncertainties will be applied to address application of these results to fuel containing BLEU material.

6.5 General CASMO-4 Modeling Assumptions The application of CASMO-4 for in-core fuel depletion is consistent with the NRC approval of EMF-2158(P)(A) (Reference 10). Input for the depletion calculation includes the fuel assembly material and geometry. The ATRIUM 10XM fuel assembly parameters are given in Table 4.1.

The key fuel pool storage rack parameters are given in Table 4.2. The following general assumptions have been made in regard to CASMO-4 modeling.

Assumption 1: The top of the part length rods in the ATRIUM 10XM assembly, which contain a 6 inch plenum, can be treated as water in the lattice in-core depletion and in the in-rack

• Blended Low Enriched Uranium (BLEU) is surplus Department of Energy material that has been down-blended to commercially acceptable enrichment levels (i.e. < 5% U-235 by weight). The primary difference of the BLEU material compared to commercial grade uranium is the existence of higher levels of U-234 and U-236 - due to previous in-reactor use prior to reprocessing. The presence of the U-236 isotope has the primary impact on lattice reactivity since it is a strong neutron absorber.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-5 calculations. The actual content of the 6 inch plenum consists of a stainless steel spring and fill gas. Neglecting the 6 inch plenum is conservative from a criticality stand point because it models a more reactive condition by adding more moderator and neglecting the neutron absorption of the plenum spring material.

Assumption 2: A fuel temperature is assumed for the fuel depletion based on the core average linear heat generation rate. Therefore, consistent fuel temperatures are used for each geometry type. Sensitivity studies were performed to determine the impact of the fuel temperature used in the fuel depletion on the in-rack storage reactivity. The fuel temperature was varied plus and minus 100 ºF relative to the base depletion temperature for the reference bounding and limiting lattices. Table 6.4 provides the in-rack results based on in-core depletion at the different temperatures (i.e., the cold in-rack calculations were repeated for the in-core depletions performed at the different temperatures). These results demonstrate that moderator void is much more significant than the depletion fuel temperature.

Assumption 3: The moderator temperature used for in-core depletion is assumed to be at saturated conditions corresponding to the rated dome pressure. The more important parameter in a BWR reactor is the actual moderator density/void level. The in-core depletion calculations are performed at [ ] void history conditions for bottom geometry lattices and [

] void history conditions for top geometry lattices. Figure 6.5 shows the results of a sensitivity evaluation with respect to the in-core depletion void history and its effect on the maximum inrack lattice k. For the reference bounding and limiting lattices the discrete void history conditions evaluated produced (or exceeded)* the maximum credible k result.

Assumption 4: The power density used for the fuel depletion is based on the core rated power per unit volume which is consistent with AREVAs standard NRC-approved depletion methodology, Reference 10. Table 6.5 provides the reactivity effect as a function of power density where 100% power density represents the core average power density at rated power.

This sensitivity analysis was performed for the reference bounding lattices and the limiting lattices listed in Table B.1 of Appendix B. These results show a small effect on inrack lattice k

• The k value reported for the top geometry GE14 lattice is based upon [ ] void history (see Tables B.2 and B.5. Given that [ ] void history is not credible for full power operation in the top of an assembly it would be acceptable to use the [ ] void history value (0.861). It is conservative to use of the 0% void history value (0.862) for this lattice.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-6 over very large changes in depletion power density. These results also demonstrate that moderator void is much more significant than the depletion power density.

Assumption 5: Modeling the pellet deformation with respect to burnup can be ignored for the in-core depletion and in-rack calculations. Modeling of the pellet deformation does not significantly change the neutronic characteristics of the fuel since the material content is unchanged.

Assumption 6: The spacer (i.e., spacer grid) material can be ignored in the in-core depletion and in-rack calculations. There is no soluble boron in this BWR spent fuel pool, and the spacers will absorb more neutrons than water. Therefore, a more reactive configuration is modeled when the spacer material is neglected.

Assumption 7: The in-core depletion is based upon uncontrolled statepoint conditions. This is appropriate because a bundle is in an uncontrolled state (i.e., the adjacent control blade is not inserted) for the majority of its lifetime, including the time from beginning of life (BOL) to the time when it reaches its lifetime maximum reactivity.

Bundles are physically located in a control cell that is associated with a specific control rod sequence (i.e., A1, A2, B1, or B2); therefore, the potential for controlled operation is limited to the times when the core is operated in that sequence (e.g., one control period in four for a core operated with the typical four control rod sequence strategy). Furthermore, the following factors tend to mitigate the amount of controlled depletion: 1) not all available in-sequence control rods are used during a sequence, 2) control rods are typically not fully inserted (they may be in deep, intermediate or shallow positions which leaves the upper lattices in an uncontrolled state), 3) bundles in peripheral and near peripheral core locations are usually not controlled, and 4) the bundles are at a reduced power during the controlled time period which reduces their accumulated burnup while in the controlled state (i.e., they experience a lower burnup rate). The net effect is that a typical bundle will experience controlled depletion for only a fraction of its time from BOL to the exposure that produces its lifetime maximum reactivity.

Potential exceptions to this behavior are: 1) bundles in a power suppression cell, and 2) bundles in a control cell in which the control rod has been declared inoperable. Power suppression is the practice of inserting a control rod to reduce power in suspected leaking fuel bundles. The control rod is typically fully inserted in an inoperable control cell. In either case, the control rod may be inserted for a significant period of time and the bundles around them will have a larger AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-7 fraction of their lifetime spent in a controlled state. However, this only affects a small population of bundles four bundles for each affected control cell.

Rodded depletion introduces impacts due to the power gradient imposed by the inserted blade as well as to the assumed power density due to the associated power reduction in the bundle.

A sensitivity calculation was performed to determine the impact of in-core controlled depletion on the peak in-rack k-infinity in which both of these parameters were varied. Table 6.6 compares the lifetime maximum k results for the ATRIUM 10XM reference bounding lattices and the limiting fuel lattices. The results in Table 6.6 are very conservative because no significant number of fuel assemblies will be controlled from BOL to the peak reactivity exposure. These results indicate a reactivity increase for the limiting GE lattices; however, as shown in Table B.1 there is substantial margin between the reactivity of the legacy fuel lattices and the ATRIUM 10XM reference bounding lattices. The results in Table 6.6 also show that uncontrolled depletion results in a higher in-rack k-infinity for the reference bounding lattices.

Assumption 8: The CASMO-4 model uses a lumped approach for fission products that are not specifically treated. CASMO-4 creates two pseudo nuclides to represent the general behavior of two fission product groups - one non-saturating and one slowly saturating. Any errors in the treatment of these pseudo nuclides becomes part of the depletion uncertainty and is included in the benchmarking and qualification of the CASMO-4 code for in-core depletion, as described in the approved topical report EMF-2158(P)(A) (Reference 10). For this evaluation, the lumped fission products were removed from the reference bounding lattices when the REBOL lattices were defined*. This modeling adds additional conservatism to the evaluation.

• Note that the lumped fission products were not removed for relative comparison calculations such as Table 6.4, Table 6.5, Table 6.6, Figure 6.4, Figure 6.5, and comparisons in the Appendices.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-8 Table 6.1 Comparison of Modeling Options for the Boral Rack KENO V.a In-Rack k Single Cell Model k Rack Model Developed for CASMO-4 0.8969 (boron-10 only*)

2x2 Infinite Array Model k Base (boron-10 only*) 0.8964 Base with Explicit Boral 0.8936 13x13 Rack Model k

(closure plates and nominal rack spacing)

Base (boron-10 only*) k = 0.8870 Base with 12" water reflector on the top and keff = 0.8847 a 24" concrete reflector on the bottom Base with Explicit Boral, 12" top water keff = 0.8819 reflector and a 24" concrete bottom NOTE: The neutron multiplication values are based upon the limiting water temperature condition, (4 °C for infinite cell conditions or 20 °C for finite rack conditions). These cases produce a KENO standard deviation of about 0.0008.

• All non-boron-10 materials in the Boral plate are neglected (i.e., modeled as void).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-9 Table 6.2 Impact of Channel Thickness on In-Rack Reactivity Fuel Channel Thickness KENO V.a (mil) (inch) k Result*

100 0.100 0.8950 80 0.080 0.8939 0 0.000 0.8921 Table 6.3 Co-Resident Storage Rack Comparison KENO V.a Limiting In-Rack k Temperature 13 x 13 Boral Rack k = 0.8870 20 °C 13 x 17 Boral Rack keff = 0.8876 20 °C

• Based on 20 °C moderator temperature.

Cases were evaluated between 4 °C and 60 °C.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-10 Table 6.4 In-Rack k Sensitivity to In-core Depletion Fuel Temperature

[

]

• Includes lumped fission products.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-11 Table 6.5 In-Rack k Sensitivity to In-core Depletion Power Density

[

]

• Includes lumped fission products.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-12 Table 6.6 In-Rack k Sensitivity to In-Core Controlled Depletion

[

]

• Includes lumped fission products.

PD refers to Power Density.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-13

[

]

Figure 6.1 Single Cell Model for the High Density Boral Rack (not to scale - top zone geometry)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-14

[

]

Figure 6.2 Explicit Geometry Model for High Density Boral Rack (not to scale - top zone geometry)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-15 Open Cell Boral Tube Boral Tube Open Cell Boral Tube Open Cell Open Cell Boral Tube Open Cell Boral Tube Boral Tube Open Cell No Boral Plate between cells in adjacent racks Two Boral Plates between cells in adjacent racks Figure 6.3 Schematic of Rack to Rack Interfaces AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-16

[

]

Figure 6.4 BLEU versus Commercial Grade Uranium Reactivity Comparison AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-17

[

]

Figure 6.5 Impact of Void History Depletion on In-Rack k-infinity AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-1 7.0 Criticality Safety Analysis This criticality safety analysis is based upon an ATRIUM 10XM reference bounding assembly.

This reference assembly is comprised of separate top and bottom geometry reference bounding lattices* and they have been defined to be more reactive than all previously manufactured lattices - as well as future ATRIUM 10XM lattices. The evaluation of the previously manufactured fuel and comparisons to these reference bounding lattices are detailed in Appendix B of this report. The reference bounding ATRIUM 10XM assembly is comprised of two axial zones as described in the following table and as shown graphically in Figure 2.2.

Lattice 235 No. of Gadolinia Gadolinia Zone Distance from BAF U wt%

Geometry Rods wt%

2 10XMLCT [ ] to TAF 4.70 8 3.5 1 10XMLCB 0" to [ ] 4.70 8 3.919 The reference bounding lattices are depleted in the reactor core environment to establish the lifetime maximum k of these lattices in the storage pool environment. The resulting k values are mainly dependent upon the lattice geometry, the U-235 enrichment level, and the gadolinia concentration; therefore, there is no axial burn-up profile assumption associated with this method.

The actual KENO V.a calculations are based upon reactivity equivalent at beginning of life (REBOL) lattices that have been designed to be more reactive than the reference bounding lattices and their calculational uncertainties. For this evaluation, a U-235 enrichment level of 3.38 wt% is applied for the top (10XMLCT) geometry and 3.21 wt% is applied for the bottom (10XMLCB) geometry.

The final k95/95 evaluation is based upon a number of factors that include the worst credible conditions and uncertainties. Items considered include assembly placement within the storage cell, assembly orientation, manufacturing uncertainties, and accident conditions.

• It is demonstrated in Appendix B that the ATRIUM 10XM reference design in the spent fuel pool geometry is more reactive than the other fuel types used at Browns Ferry.

The CASMO-4 vs. KENO comparison in Appendix D demonstrates a stable basis for this reactivity equivalence. The 2x2 KENO model used in Appendix D was also established as the maximum k case in Section 6.1.4.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-2 7.1 Definition of the Reference Bounding and REBOL Lattices The CASMO-4 lattice depletion calculations are performed at hot operating, uncontrolled, characteristic void history conditions. These void history conditions are [ ] for top geometry lattices and [ ] for the bottom geometry lattices. The calculation results are based upon the nominal fuel design parameters (defined in Table 4.1) and assume a standard 100 mil fuel channel. The location of the 8 gadolinia rods in the reference bounding lattices have been selected to maximize the reactivity of the lattices. Xenon and lumped fission product free restart calculations are performed as a function of exposure and void history to establish the highest in-rack reactivity (k) at any time throughout the life of these fuel lattices.

The CASMO-4 in-rack k of the top and bottom zone reference bounding lattices are both 0.8825. These results are summarized in Table 7.1.

The reference bounding and REBOL lattices are based upon a uniform enrichment distribution.

A uniform enrichment distribution increases the BWR lattice reactivity because low enriched rods in the corners of the lattice are replaced with rods at an average enrichment level. Relative to a representative top and bottom ATRIUM 10XM lattice design, a uniform enrichment distribution was determined to be more reactive by 0.002 to 0.004 k. Consequently, the use of these lattices with uniform enrichment distributions conservatively bound the distributed enrichment distributions of expected future lattice designs.

In support of the KENO rack calculations, two REBOL lattices are created corresponding to the top and bottom geometries for the ATRIUM 10XM design. These lattices are defined using a water temperature of 4 °C in the spent fuel pool rack configuration. The top REBOL lattice is defined with a uniform 3.38 wt% U-235 enrichment level, and the bottom REBOL lattice is defined with a uniform 3.21 wt% U-235 enrichment level. These results are also summarized in Table 7.1.

As discussed in the methodology section, an adder of at least 0.010 k is included in the generation of the REBOL lattices to address CASMO-4 code, geometry, material, and depletion uncertainties. The adequacy of this adder is the primary subject of Appendix D.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-3 7.2 Storage Array Reactivity The base storage array reactivity is calculated using KENO V.a as an infinite array of fuel storage cells using the explicit storage cell model as described in Section 6.1.2 and as illustrated in Figure 6.2. This model was shown to be conservative in Section 6.1.4. (KENO V.a results using this model were also shown to trend well with CASMO-4 results in Appendix D)

Each cell is assumed to contain an assembly composed of 3.38 wt% U-235 (top) and 3.21 wt%

U-235 (bottom), uniformly enriched REBOL lattices without gadolinia. As discussed earlier, each REBOL lattice is defined to be at least 0.010 k more reactive than its corresponding reference bounding lattice. A periodic boundary condition is specified for both the x-y plane and for the axial direction. The KENO model assumes a standard 100 mil fuel channel which was shown in Section 6.2 to bound storage with no channel, an 80 mil channel, and the advanced thick-thin channel.

KENO V.a calculations were performed at various temperatures from 4 C to 60 C that confirmed that the REBOL assembly is bounded by the 4 C results. As shown in Table 7.2, the limiting base case KENO k-eff is 0.897. Except as specifically noted, the reactivity values presented in Table 7.1 and Table 7.2 do not include adjustments for uncertainties or KENO V.a code biases. Section 7.8 presents the determination of the upper limit 95/95 reactivity for the storage rack array.

7.3 Arrays of Mixed BWR Fuel Types It is shown in Tables B.1 and B.2 that the ATRIUM 10XM reference bounding lattices are more reactive in the in-rack configuration than the limiting lattices of the legacy fuel. Additionally, it is also shown in Appendix B that the other legacy fuel types have significant margin relative to the limiting lattices. It then follows that from a reactivity perspective, the reference bounding ATRIUM 10XM lattices used in this evaluation can conservatively represent past assembly fuel types.

The assembly reactivity limits (either enrichment and gadolinia limitations or direct k values) defined in Table 2.1 are applicable to all future ATRIUM 10XM fuel assemblies that will be built for the Browns Ferry reactors. Therefore, there will not be a more reactive assembly to consider in an accident scenario and an array composed of a mixture of these fuel types will not exceed the reactivity calculated for an array of limiting ATRIUM 10XM assemblies.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-4 7.4 Other Conditions The unadjusted reactivity result reported in Table 7.2 (0.897) is based upon a reference orientation which places the ATRIUM 10XM internal water channel toward the bottom right corner of the storage cell with the assembly centered within the cell as shown in Figure 6.2.

The actual position of assemblies in the storage racks will include assembly rotation and lean.

In addition, it is possible that blisters could form on the surface of the Boral plates. This deformation of the Boral plate will exclude water and therefore affect the reactivity of the storage racks. These conditions will be evaluated in this section and their worth will be included as a direct adder in the k95/95 equation.

7.4.1 Assembly Rotation The rotational combinations shown in Figure 7.1 and the simple 90°, 180°, and 270° cases were evaluated to determine if the asymmetric nature of the ATRIUM 10XM fuel assembly will produce a more reactive condition than the base case shown in Figure 6.2. The simple 90° rotation case was the most limiting with a k increase of 0.001 +/- 0.001 k. This effect will be included in the ksys parameters in the calculation of k95/95 in Section 7.8.

7.4.2 Assembly Lean Each storage cell has a hole in the bottom where the lower tie plate nose piece fits to center the assembly. There is no corresponding mechanism to keep the upper part of the assembly centered; therefore, each assembly has the ability to lean toward a side or corner of the storage cell. The impact of this lean condition was evaluated by assuming the entire bundle can be positioned anywhere within the storage cell. Between 1 and 4 assemblies were moved relative to one another within their cells. The result of this evaluation showed no statistically significant increase relative to the centered position.

7.4.3 Blister Formation Under certain conditions, corrosion gases can be trapped within a Boral plate and the aluminum cladding can be deformed to create blisters on the surface of the plate. These blister regions exclude water and can therefore affect the neutron absorption of the Boral storage rack. For this analysis a uniform 0.055" void region has been used as a conservative model of this AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-5 potential blistering condition*. Calculations indicate that this level of void on all the Boral plates in the pool would increase reactivity by 0.004 +/- 0.001 k . This effect will be included in the ksys parameters in the calculation of k95/95 in Section 7.8.

7.5 Normal Fuel Handling Normal fuel assembly movement is generally described as those movements required to load and unload assemblies into allowable storage locations within the spent fuel pool. The allowed storage locations include the spent fuel pool storage racks and the fuel preparation machines (FPMs).

Fuel movements are accomplished with the use of a refueling bridge with a mast and grapple assembly. Fuel assemblies are grasped and suspended from the mast/grapple assembly with normal lateral movements occurring above the top of the storage cell locations. The base storage array reactivity model assumes an infinite lattice array in both radial and axial dimensions using a periodic boundary condition as addressed in Section 7.2. This infinite array of fuel lattices bounds the case for suspending a single bundle over the rack during normal fuel movements. Loading or unloading an assembly into a storage location requires the raising or lowering of the fuel into the storage cell. This operation is also bound by the base storage array reactivity, which assumes the racks are fully loaded.

The spent fuel storage pool contains two FPMs that allow for the storage of a single assembly within each. Each FPM is neutronically isolated from the other so interaction between them is not considered. It is feasible that an assembly suspended from the refuel bridge can be brought into close proximity to an assembly already located in an FPM. An analysis was performed that considered the additional potential for a misplaced assembly for a total of three (3) assemblies in close proximity. These assemblies are isolated from all other fuel assemblies in the spent

• A uniform void with a 0.055 inch height bounds the condition of having a 1/8 inch high blister with a spherical cross section on every 1.25x1.25 unit cell on one side of a Boral plate (i.e., 1.25 diameter blisters with a height of 1/8 inch packed edge to edge). This in turn would be equivalent to each side of the Boral plate having blisters of this size with 50% area coverage. This conservatively bounds the results from the stainless steel clad coupon surveillances performed at Browns Ferry, on an average basis.

An additional sensitivity calculation has been performed assuming a 0.080 inch uniform void condition that results in a total reactivity increase of 0.006 +/- 0.001 k, only 0.002 k higher than the condition assumed in the k95/95 calculation.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-6 fuel pool. For this comparison, three ATRIUM 10XM fuel assemblies with REBOL lattices at 3.5 wt% U-235 were placed together in a triangular pattern. A reactivity optimization search was performed using different assembly spacing and different assembly orientations. In addition, calculations were performed with and without fuel channels and the water temperature was varied from 4 °C to 60 °C. A maximum keff of 0.897 +/- 0.001 was calculated with unchanneled assemblies at 4 °C.

By coincidence the resulting keff for this configuration is equivalent to the base array reactivity identified in Section 7.2 and used in the k95/95 calculation in Section 7.8. If a k95/95 result were calculated for the fuel handling condition using the REBOL lattices from the main calculation it would be less than the value for the limiting rack (Section 7.8) because:

• this configuration is based on 3.5 wt% U235 lattices where the REBOL lattices use a lower U-235 enrichment level (3.21 wt% U235 (Bottom) and 3.38 wt% U235 (top))

• there are no applicable accident conditions for this configuration

• the manufacturing tolerance value is lower for this application because there are no applicable storage rack, fuel channel, or gadolinia tolerance conditions Both the misloading of an assembly into a location adjacent to a loaded rack (i.e., a non-allowed storage location) and the dropping of an assembly during fuel movements (i.e., fuel handing accident) are accident conditions which are evaluated in Section 7.6.

7.6 Accident Conditions In addition to the nominal storage cell arrangement, accident conditions have also been considered. All k values provided in this section are based upon comparative KENO V.a calculations, i.e., only the most limiting scenario will be reflected in the k95/95 calculation in Section 7.8. The following scenarios were evaluated to identify the most limiting accident condition.

• Missing Boral plate in the interior of the rack. (limiting condition)

• Boral Storage Racks being forced together.

• Misloaded Bundle Scenarios.

o Assembly misloaded between the pool wall and storage rack adjacent to an open cell (no Boral between assemblies) o Assembly misloaded into the corner region adjacent to 3 racks.

o Assembly misloaded between the fuel preparation machine adjacent to an open cell (no Boral between assemblies)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-7

• Dropped assembly lying horizontally across the top of the spent fuel pool.

• Loss of Spent Fuel Pool Cooling.

The situation where a single Boral plate is missing from an interior storage rack location was evaluated. Since this was the most limiting case, the moderator temperature and assembly position were varied to optimize the worth of the accident. (The use of unchanneled assemblies was considered but was not evaluated because the calculation results indicated an optimum condition occurs when water is between the assemblies and because removal of the fuel channel tends to reduce the reactivity). The most limiting condition occurred at 4 °C with one assembly moved to the edge of the storage cell and the adjacent assembly moved half the distance to the edge of the cell as shown in Figure 7.2. This accident condition has a reactivity worth of 0.006 +/- 0.001 k. This will be included in the ksys parameters in the calculation of k95/95 in Section 7.8.

It is postulated that 2 or more Boral racks could be forced together during a seismic event. For this situation, the spacing between racks is reduced from 2 or more inches to less than 1/2 inch.

Should this occur, the pool k is calculated to increase by about 0.005 k. This accident scenario is less limiting than the optimized missing Boral plate scenario.

The case of a misloaded assembly was investigated by assuming that an assembly was placed on the edge of a Boral storage rack adjacent to an assembly in a non-tube or open storage cell.

This misloaded assembly was moved to a location very near the adjacent assembly. The results confirm that this accident scenario increases the system k by less than 0.001 k.

As shown in Figure 4.2, a misloaded assembly could be placed in a location where 3 racks meet together. With this geometry, the corner storage cells are all Boral tube cells. As expected, no significant reactivity increase is produced.

It is also possible for an assembly to be in the fuel preparation machine while a second assembly is moved between the fuel preparation machine and the fuel storage rack. This is conservatively modeled as two assemblies placed against each other adjacent to an open cell of the Boral storage rack. The results show a reactivity increase of less than 0.001 k and confirm that this accident scenario is less limiting than the missing Boral plate scenario.

For the case of dropping a fuel assembly onto an assembly in the storage rack (i.e., a fuel handling accident in the spent fuel pool), the potential exists for damaging the dropped AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-8 assembly as well as any other assemblies it contacts. This event has the potential to cause deformation to the affected assemblies, however; the reactivity impact of this deformation on rack reactivity is minimal since it involves only 2-3 assemblies in a localized area. There will also be no significant effect on the array reactivity when the dropped assembly comes to rest in a horizontal or inclined position on top of the storage rack because the dropped assembly will be neutronically isolated from the fuel in the storage cells (greater than 12 inches of water between the dropped assembly and the top of the active fuel zone of the fuel in the storage rack). Finally, similar to the previous discussion for normal fuel handling it is noted that the axial boundary condition used in the KENO model provides an infinitely repeating fuel column.

Consequently, the base model conservatively bounds the potential impact of a dropped assembly and no increase in reactivity applies for this event.

For the infinite Boral rack model, the limiting moderator temperature is 4 °C (39.2 °F).

Therefore, an increase in the pool water temperature (a loss of spent fuel pool cooling event) will not increase the reactivity of these racks.

7.7 Manufacturing and Other Uncertainties Uncertainties associated with defining bounding REBOL lattices are addressed in Appendix D.

Specifically, uncertainties associated with CASMO-4 depletion and modeling capabilities are included within the REBOL definition process (through the requirement for the lattice to have a 0.010 k higher reactivity when compared to the corresponding reference bounding lattice).

Table 7.1 demonstrates that the requirement for this adder has been met with a minimum difference of 0.0109 k for the top lattice.

The manufacturing tolerance values and the calculated reactivity uncertainties for the ATRIUM 10XM fuel are shown in Table 7.3.* The gadolinia manufacturing uncertainty (gadolinia concentration and gadolinia pellet density) effect on reactivity was evaluated with a combination of KENO V.a and CASMO-4. All other uncertainties reported in Table 7.3 were evaluated with KENO V.a. The ATRIUM 10XM rack calculations are conservatively performed for a minimum B10 areal density, therefore no manufacturing uncertainty is needed for this parameter. BOL

• The manufacturing uncertainties for other fuel types in the SFP are not explicitly addressed in this analysis due to the reactivity margin between all existing fuel and the reference bounding lattices.

See Appendix B for more detail.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-9 dimensions have been assumed, except the fuel rod pitch and channel growth results are based upon conservative spacer and channel growth dimensions.

For the various tolerances which are evaluated with KENO, the k and the standard deviation (s) values are combined consistent with the variance equation listed in Section 4.1.5 of Reference 11:

k2 = (u2/x2)((k - kref)2 +/- (sMC2 + sMC,ref2))

where: (k - kref) change in keff induced by change x on parameter x u standard uncertainty of parameter x x change in parameter x sMC Monte Carlo standard deviation values The manufacturing tolerance results have been evaluated using the upper and lower bounds of the full tolerance range; therefore, x represents a range greater than 2u. Rather than define a single uncertainty interval for this calculation and then multiply it by 2 to reestablish a 95/95 bounding interval, u2/ x2 is conservatively treated as unity in this calculation.

The Monte Carlo uncertainty values have been added to the limiting case and where (k - kref) is negative for both the upper and lower bounds of the tolerance interval, a zero value has been used (e.g., the channel thickness, pellet diameter, and Boral sheet width). The adjusted k values are the square root of the variance for that particular case. The statistically combined result is the square root of the sum of the variance values.

7.8 Determination of Maximum Rack Assembly k-eff (k95/95)

For the ATRIUM 10XM fuel design with REBOL lattice enrichments of 3.21 and 3.38 wt%

U235, the base case KENO calculated in-rack reactivity from Table 7.2 is 0.897. This k-eff value is used with the following equation to determine the upper limit 95/95 reactivity (also illustrated in Figure 2.1):

k95/95 = keff + biasm + ksys + (C2k2 + Cm2m2 + C2sys2 + ktol2)1/2, where:

keff = Base in-rack reactivity from KENO V.a, (0.897, Table 7.2)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-10 biasm = KENO V.a validation methodology bias (0.0075, Appendix C - Section C.8) ksys = Summation of applicable system variables (See the following table and Sections 7.4.1, 7.4.3, and 7.6)

C = 95% confidence level consistent with KENO V.a (2.0)

Cm = 95/95 one-sided tolerance multiplier for a sample size of 68 (1.996) k = k-eff standard deviation from KENO V.a, (0.001, Table 7.2) m = KENO V.a methodology uncertainty (0.0027, Appendix C - Section C.8) sys = (sys12 + sys22 + sys_n2)1/2, for ksys uncertainties. (See the following table and Sections 7.4.1, 7.4.3, and 7.6) ktol = Statistical combination of manufacturing reactivity uncertainties ( [ ],

Table 7.3)

The following table provides a summary of the ksys and sys parameters applicable to this analysis. (The values are standard deviation results from KENO).

Description ksys sys Assembly Rotation Effects (Section 7.4.1) 0.001 0.001 Boral Blisters (Section 7.4.3) 0.004 0.001 Limiting Accident (Missing Insert, Section 7.6) 0.006 0.001 Combined Values 0.011 0.0017 The standard deviations and tolerance uncertainties are included as the square root of the sum of the squares since they represent independent events. Solving for k95/95 yields a 95/95 upper limit k-eff that is larger than 0.927 so it is rounded-up to 0.928. The above determination of the upper limit 95/95 k-eff is consistent with the method documented in Reference 6 and allows one to state that at least 95% of the normal population is less than the 95/95 k-eff value calculated with a 95% confidence.

The results demonstrate the postulated configuration with the ATRIUM 10XM REBOL lattices meets the NRC criticality safety acceptance criterion that the array k-eff under the worst credible conditions is < 0.95.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-11 Table 7.1 Summary of CASMO-4 Maximum In-Rack Reactivity Results Reference Bounding Lattices ATRIUM 10XM geometry, top and bottom lattice geometries explicitly modeled 4.70 wt% U-235 uniform enrichment distribution above [ ]

4.70 wt% U-235 uniform enrichment distribution at and below [ ]

8 gadolinia rods with 3.5 wt% Gd2O3 above [ ]

8 gadolinia rods with 3.919 wt% Gd2O3 at and below [ ]

Standard 100 mil fuel channel Reflective boundary condition for in-core calculations No xenon or lumped fission products for in-rack calculations Periodic boundary condition for in-rack calculations Condition Top Lattice Bottom Lattice Maximum In-Rack k, 4°C (39.2°F) 0.8825 0.8825 Exposure, GWd/MTU 10.5 11.5 Void History [ ] [ ]

REBOL Lattices ATRIUM 10XM geometry, top and bottom lattice geometries explicitly modeled 3.38 wt% U-235 uniform enrichment distribution above [ ]

3.21 wt% U-235 uniform enrichment distribution at and below [ ]

No gadolinia Standard 100 mil fuel channel BOL (zero exposure, no xenon, no fission products)

Periodic boundary condition Condition Top Lattice Bottom Lattice Maximum In-Rack k, 4°C (39.2°F) 0.8934 0.8935 Margin to Reference Bounding Lattice 0.0109 0.0110 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-12 Table 7.2 Summary of KENO V.a Maximum In-Rack Reactivity Results Fuel Assembly ATRIUM 10XM geometry, top and bottom lattice geometries explicitly modeled 3.38 wt% U-235 uniform enrichment distribution above [ ]

3.21 wt% U-235 uniform enrichment distribution at and below [ ]

No gadolinia Standard 100 mil Channel BOL (zero exposure, no xenon, no fission products)

Periodic boundary conditions for in-rack x-y plane and the axial direction Storage Array Configuration Explicit 2x2 rack model with infinite periodic boundary conditions Assembly centered in cell water volume 4°C moderator and fuel temperatures Description k-eff In-Rack 4°C (39.2°F) k-eff 0.897 +/- 0.001 Maximum k95/95 Reactivity (including uncertainties, biases, manufacturing tolerances 0.928 and worst accident or abnormal loading conditions)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-13 Table 7.3 Manufacturing Reactivity Uncertainties (Based upon BOL conditions using KENO V.a except as noted.)

[

]

• This is a conservative approximation of the spacer growth at peak reactivity exposures.

This is a lifetime maximum value and is assigned to each side of the fuel channel, [

].

Depletion based adders of [ ] have been added to the gad concentration and gad density cases, respectively.

§ This calculation was performed using the minimum value so no manufacturing uncertainty is required, see discussion in Section 7.3.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-14 Figure 7.1 Evaluated Assembly Rotation Cases AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-15 Figure 7.2 Limiting Accident (Missing Boral Plate)

(Note that assembly positions have been shifted to maximize the worth of this accident condition. The missing Boral plate is located between the center-top and center cell in the figure.)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 8-1 8.0 References

1. Tennessee Valley Authority Docket Nos. 50-259, 50-260 and 50-296 Tennessee Valley Authority Notice of Issuance of Amendment, Browns Ferry Units 1, 2, and 3 License Amendments 42, 39, and 16, Authorizing You to Increase Storage Capacity of Each Spent Fuel Pool, September 21, 1978. (Adams # ML020040269)

2. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 9.1.1 Revision 3 (Criticality Safety of Fresh and Spent Fuel Storage and Handling), U.S. Nuclear Regulatory Commission, March 2007.

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

4. Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors, ANSI/ANS American National Standard 8.17-1984, American Nuclear Society, January 1984, (withdrawn 2004).

5. Letter, Brian K. Grimes, Assistant Director for Engineering and Projects Division of Operating Reactors, U.S. Nuclear Regulatory Commission, to All Power Reactor Licensees, OT Position for the Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14, 1978, as amended by letter January 18, 1979.

6. Letter, Laurence Kopp (Reactor Systems Branch, NRC) to Timothy Collins, Chief (Reactor Systems Branch-NRC),

Subject:

Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, August 19, 1998.

7. Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01 Revision 0, Staff Guidance Regarding The Nuclear Criticality Safety Analysis For Spent Fuel Pools, (ADAMS # ML110620086).

8. NUREG/CR-0200 Revision 6, SCALE Version 4.4 A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, Oak Ridge National Laboratory, May 2000.

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

10. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.

11. ICSBEP Guide to the Expression of Uncertainties, Revision 5, V. F. Dean, September 30, 2008. {Distributed with the International Handbook of Evaluated Criticality Safety Benchmark Experiments, Nuclear Energy Agency, NEA/NSC/DOC(95)03, September 2009 Edition.}

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page A-1 Appendix A Sample CASMO-4 Input Tables A.1 and A.2 provide the CASMO-4 spent fuel storage rack model for the reference bounding lattices defined in this analysis.

ATRIUM 10XM fuel which does not conform to the enrichment and gadolinia requirements described in Table 2.1 can be analyzed for storage in the spent fuel storage racks by adapting the CASMO-4 sample inputs presented in Table A.1 or A.2. Evaluations should be performed with [ ] depletion for bottom geometry lattices and [ ]

depletion for top geometry lattices. These calculations will be performed with the NRC approved CASMO-4 code described in EMF-2158(P)(A), (Reference 10 of the main report).

If the lifetime maximum in-rack k of the new lattices are less than the k of the corresponding reference bounding lattices (0.8825), the ATRIUM 10XM fuel assembly can be safely stored in the Browns Ferry Nuclear Plant spent fuel storage racks.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page A-2 Table A.1 CASMO-4 Input for ATRIUM 10XM Top Reference Bounding Lattice

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page A-3 Table A.2 CASMO-4 Input for ATRIUM 10XM Bottom Reference Bounding Lattice

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-1 Appendix B Reactivity Comparison for Assemblies Used in the Browns Ferry Reactors All previously manufactured assemblies used in the Browns Ferry Units 1, 2, and 3 reactors have been evaluated to determine the most limiting lattices on the basis of highest in-rack k.

This reactivity evaluation included an initial screening that resulted in the identification of a set of the most limiting lattices which is detailed in Section B.2 of this Appendix. The resulting limiting lattices were then used to establish a reference bounding lattice for each geometry zone as well as a corresponding REBOL lattice that is used as the basis for the KENO V.a criticality analysis.

Section B.1 provides a comparison of the resulting limiting lattices and their corresponding reference bounding lattices and REBOL lattices.

B.1: Summary of Lattice In-Rack Reactivity Comparisons The screening detailed in Section B.2 of this appendix resulted in the selection of the highest reactivity previously manufactured (or as-fabricated, or legacy) lattices based upon calculated CASMO-4 in-rack k values. These limiting as-fabricated lattices are compared to the corresponding ATRIUM 10XM reference bounding lattice and REBOL lattice in Table B.1. This comparison shows that the ATRIUM 10XM reference bounding lattices described in Table 7.1 are more reactive than any of the previously manufactured lattices used in the Browns Ferry reactors. It also shows that the REBOL lattices defined for use in the KENO V.a calculations are more reactive than the reference bounding lattices.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-2 Table B.1 Lattice Reactivity Comparisons (REBOL, Bounding, and Limiting)

Maximum In-Rack k Case Description Lattice Description (CASMO-4 @ 4 °C)

Top Lattices REBOL, Top Lattice XMLCT-3.38 (no Gad) 0.8934

[ ]

Reference Bounding Top Lattice XMLCT-470UL-8G35 0.8797*

[ ]

Limiting As-Fabricated Top Lattice GE14 (From Table B.2) 0.8619

[ ]

Margin to Reference Bounding Lattice (Reference - Limiting) 0.0178 k Bottom Lattices REBOL, Bottom Lattice XMLCB-3.21 (no Gad) 0.8935

[ ]

Reference Bounding Bottom XMLCB-470UL-8G3919 0.8790*

Lattice [ ]

Limiting As-Fabricated Bottom GE13 (From Table B.2) 0.8227 Lattice [ ]

Margin to Reference Bounding Lattice (Reference - Limiting) 0.0563 k

• For direct comparison with the legacy fuel, this value also includes the effects of lumped fission products. Without lumped fission products the k value increases to 0.8825 as reported in Table 7.1.

Lattice descriptions for non-AREVA supplied fuel are not provided in this document since they have been identified as proprietary by that vendor.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-3 Table B.2 Limiting Lattices In-Rack Kinf Comparison (By Product Line)

Maximum Margin to In-Rack k 10XM Case Array (CASMO-4 Bounding

@ 4 °C) Lattice (k)

Limiting Top Lattices by Product Line [ ]

ATRIUM 10XM TRBL 10x10 0.8797 ---

ATRIUM-10 TRBL 10x10 0.8747 0.0050 ATRIUM-10 10x10 0.8520 0.0277 GE14 10x10 0.8619 0.0178 GE13 9x9 0.8254 0.0543 GE11 9x9 0.8237 0.0560 GE9B 8x8 0.8114 0.0683 GE7B 8x8 0.7915 0.0882 Limiting Bottom Lattices by Product Line [ ]

ATRIUM 10XM BRBL 10x10 0.8790 ---

ATRIUM-10 BRBL 10x10 0.8738 0.0052 ATRIUM -10 10x10 0.8121 0.0669 GE14 10x10 0.8086 0.0704 GE13 9x9 0.8227 0.0563 GE11 9x9 0.8214 0.0576 GE9B 8x8 0.8114 0.0676 GE7B 8x8 0.7915 0.0875 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-4 B.2: Previously Manufactured Lattices Browns Ferry Units 1, 2, and 3 have loaded a number of different product lines including GE 7x7 (GE2 and GE3), GE 8x8 (GE4, GE5, GE6, GE7/7B, and GE9B), GE 9x9 (GE11 and GE13), GE 10x10 (GE14), and AREVA 10x10 (ATRIUM-10) fuel. Several variations of LUAs have also been loaded which include variants of older designs and a set of 4 Westinghouse QUAD+

assemblies.

Initial operation of the Browns Ferry reactors transitioned from initial 12 month nominal cycle lengths to 18 month cycles and finally to the current 24 month cycles. The later cycles have operated at 105% of the original licensed thermal power level. As a consequence of the movement towards longer cycles and higher operating power levels, the fuel designs have transitioned to designs with higher U-235 enrichments and Gadolinia loading. The general trend is that the later higher enrichment fuel designs bound the earlier low enrichment designs.*

The Unit 1 and 2 spent fuel pools at the Brown Ferry Nuclear Plant include a transfer canal that allows assemblies from one pool to be moved to the other. The Unit 3 spent fuel pool is not connected to either of the other pools. To simplify this analysis the limiting lattices will be based upon the inventory of all three spent fuel pools.

B.2.1: Current Inventory and Initial Screening The previously manufactured fuel inventory is summarized in Table B.3 and Table B.4. An initial screening of the previously manufactured fuel assemblies was performed based upon U235 enrichment and gadolinia loading. It was determined that explicit calculation of in-rack k is not required for lattices with gadolinia and with initial peak average enrichment at or below the lowest REBOL lattice enrichment of 3.21 wt% U-235. These fuel assemblies are listed in standard font in Table B.3. This criterion is based upon the recognition that a lattice without gadolinia (such as the REBOL lattices) will always exhibit a higher reactivity than a lattice having the same enrichment and also containing gadolinia (this condition is illustrated in Figure D.4 of Appendix D). While it is noted that the application of enrichment only screening does not specifically address changes in lattice geometry (i.e., 7x7 or 8x8 versus later 9x9 and 10x10

• One exception to this trend is encountered with the introduction of BLEU fuel since the presence of U236 acts as a neutron poison and reduces lattice reactivity when compared to an equivalent enrichment commercial grade uranium lattice. See section 6.4 in the main body of the report for more detail.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-5 designs), these lattice geometry impacts are small compared to the conservatism established with the non-gadolinia REBOL lattice, as discussed above. Furthermore, Table B.5 shows that similar lattices with higher U235 enrichment levels have significant reactivity margin to the limiting lattices. Application of this criterion immediately screens* out the GE 7x7 fuel and many of the older GE 8x8 assemblies. All bundles in bold type in Table B.3 are analyzed explicitly.

As summarized in Table B.4, the previously loaded ATRIUM-10 fuel includes both commercial grade uranium (CGU) and blended low enriched uranium (BLEU) assemblies. These previously manufactured lattices have been explicitly modeled with the U-234 and U-236 content associated with the fabrication of the specific reload batch.

B.2.2: Explicit Reactivity Evaluations Tables B.5 and B.6 provide a comparison of the peak in-rack CASMO-4 k values of the lattices in the bundles selected for explicit evaluation in the previous section. The axial zones are separated at [ ].

The ATRIUM-10 fuel product line is the fuel currently being loaded in reload quantities in Units 2 and 3 and planned for use in Unit 1. These and other previously used designs are stored in the Browns Ferry spent fuel storage pools. The reference bounding ATRIUM 10XM assembly has a higher in-rack reactivity than all previously loaded fuel assembly designs, including the ATRIUM-10 reference bounding lattice developed in Reference B.1 . As such, the ATRIUM 10XM reference bounding assembly design forms the basis for demonstrating that the maximum k-eff of the spent fuel pool storage array remains less than 0.95.

The calculations documented herein demonstrate that the ATRIUM 10XM reference bounding assembly design has been selected to be more reactive in an in-rack configuration than any of the current or past fuel assembly designs used in the Browns Ferry reactors. From Table B.2,

• A conservative screening of 2.99% U-235 was actually used. An exception to the screening was a group of four (4) LUAs slightly above the 3.1% U-235 criteria. This design was screened out since only four of these LUAs exist and the enrichment is slightly below the 3.21% enrichment screening criteria. Also, as noted in the text above, the presence of gadolinia in these lattices will significantly reduce the peak in-rack reactivity when compared to the REBOL non-gad lattices.

[ ]

Therefore, future ATRIUM-10 assemblies that meet the storage requirements of Reference B.1 will be less reactive than the ATRIUM 10XM reference bounding lattices.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-6 the limiting case is a GE14 top lattice that has been demonstrated to be 0.0178 k less reactive than the top zone reference bounding lattice. Table B.5 shows that this limiting lattice is much more reactive than the other lattices in the same bundle - indicating that there is more margin available on an assembly basis than is apparent with this lattice comparison.

These in-rack lattice k-infinity comparisons are based upon actual GE 8x8, GE 9x9, GE 10x10, and ATRIUM-10 lattice geometries and enrichment distributions. This evaluation establishes that the fuel assemblies previously manufactured for use in the Browns Ferry reactors can be safely stored in the Browns Ferry spent fuel storage pools. This evaluation also shows that future ATRIUM 10XM assemblies meeting the storage requirements established by this criticality analysis can be safely stored with these previously manufactured assemblies.

B.2.3: Evaluation of Modified, Abnormal, and Damaged Assemblies The preceding evaluation of previously supplied fuel is based upon nominal assembly designs.

The potential exists that assemblies that have been previously damaged or modified may have a configuration that is more reactive than the nominal design.

One of the primary issues that could affect in-rack reactivity is the removal of one or more fuel rods from an assembly. Since the lattices are under moderated in the in-rack configuration, removal of a fuel rod without replacement would introduce additional moderator which could result in an increase in the lattice reactivity. This would also apply to the case where a rod has been broken and a portion of the broken fuel rod removed from the bundle. Assemblies in which a fuel rod has been replaced with either an inert rod or a natural uranium rod would represent a reduction in reactivity from the nominal design since the fissile material content is reduced and no change in the amount of moderator would occur.

The existing inventory of fuel in the Browns Ferry spent fuel pools includes a number of bundles that have experienced damage of one kind or another. The following is a summary of the status of this fuel:

• Missing Fuel Rods: No fuel assemblies in the Unit 1, 2, or 3 spent fuel pools have a missing fuel rod (i.e. without an inert or natural rod replacement).

• Replacement Fuel Rods: A total of eight fuel assemblies have had either one or two pins that have been removed and have been replaced with an inert stainless steel rod as detailed in Table B.7.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-7

• Broken Fuel Rods: Four bundles have been identified with broken fuel rods as detailed in Table B.7. The affected rods were either not disturbed or were returned to the bundle. Consequently, the amount of moderator and fissile material is maintained and there is no adverse impact on in-rack reactivity versus the nominal design.

Another type of modification is related to reconstitution of a bundle using replacement rods from other donor assemblies. A reconstitution campaign at Browns Ferry was completed for Unit 2 Cycle 6 that affected a number of twice and thrice burnt assemblies.

204 8DRB284L / P8DRB284L (U2 Cycle 3 / U2 Cycle 4) assemblies reconstituted 60 8DRB284L / P8DRB284L (U2 Cycle 3 / U2 Cycle 4) donor assemblies Therefore, a total of 264 high burnup assemblies were modified including a combination of the reconstituted and donor assemblies. The following items of interest are noted regarding these assemblies:

• The nominal enrichment of 2.84 wt% U-235 and corresponding planar enrichment of 3.08 wt% U-235 of the affected assemblies is below the previously applied screening criteria. Table B.2 provides the results for a GE7 lattice with a similar geometry and higher enrichment that still shows >0.080 k margin to the reference bounding lattices.

• The rod replacements were based upon swaps with rods of similar burnup and reactivity between the donor and reconstituted assemblies. Criteria for the selection of replacement rods were established to minimize the impact on hot in-core k-infinity and local peaking within the assembly lattices; this would also tend to minimize impact on in-rack reactivity.

• None of the reconstituted or donor assemblies were left with open fuel rod positions (i.e. all rod positions contained a fuel rod in the modified assemblies).

It is also noted that all of the reconstituted or donor assemblies are high burnup (twice or thrice burnt) and are therefore well past their peak reactivity condition. Of the reconstituted assemblies, 212 were then loaded and operated for another cycle which further reduces their actual reactivity.

Based upon the above discussion, all of the assemblies identified as modified or damaged are appropriately evaluated at nominal conditions. That is the assembly modifications do not provide a significant adverse impact on in-rack reactivity that could potentially make them limiting reactivity assemblies.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-8 B.3: References B.1 ANP-2945(P) Revision 1, Browns Ferry Nuclear Plants Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis, AREVA NP, July 2011.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-9 Table B.3 Non-AREVA Supplied Fuel Inventory Nominal Product Bundle Design Name Array enrichment Line Unit 1: Spent Fuel Pool GE2-7DB110 1.1 GE2 7x7 GE14-P10DNAB147-NOG-1T-150-T6-2893 1.47 GE14 10x10 GE13-P9DTB156-NOG-100T-146-T6 1.56 GE13 9x9 GE14-P10DNAB157-NOG-1T-150-T6-2889 1.57 GE14 10x10 GE13-P9DTB163-NOG-100T-146-T6 1.63 GE13 9x9 GE14-P10DNAB200-3GZ-1T-150-T-2603 2.0 GE14 10x10 GE3-7DB250-3G3/2G4 2.5 GE3 7x7 GE3-7DB250-4G3 2.5 GE3 7x7 GE5-8DRB265-6G2 2.65 GE5 8x8 GE5-8DRB265-6G3 2.65 GE5 8x8 GE6-P8DRB265-6G2 2.65 GE6 8x8 GE6-P8DRB265-6G3 2.65 GE6 8x8 GE4-8DB274-5G2 2.74 GE4 8x8 GE4-8DB274-5G3 2.74 GE4 8x8 GE6-P8DRB284-4G4/3G2 2.84 GE6 8x8 GE6-P8DRB284-7GZ 2.84 GE6 8x8 GE7-P8DRB284-4G4/3G2 2.84 GE7 8x8 LTA-P8DRB284-GZLTA1 2.84 LTA 8x8 LTA-P8DRB284-GZLTA2 2.84 LTA 8x8 GE14-P10DNAB377-16GZ-1T-150-T6-2890 3.77 GE14 10x10 GE13-P9DTB391-13GZ-100T-146-T 3.91 GE13 9x9 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-10 Table B.3 Non-AREVA Supplied Fuel Inventory (Continued)

Nominal Product Bundle Design Name Array enrichment Line Unit 1: In-Core GE14-P10DNAB157-NOG-100T-150-T6 1.57 GE14 10x10 GE14-P10DNAB350-16GZ-100T-150-T6 3.50 GE14 10x10 GE14-P10DNAB368-15GZ-100T-150-T6 3.68 GE14 10x10 GE14-P10DNAB377-16GZ-100T-150-T6 3.77 GE14 10x10 GE14-P10DNAB377-17GZ-100T-150-T6 3.77 GE14 10x10 GE14-P10DNAB400-17GZ-100T-150-T6 4.00 GE14 10x10 GE14-P10DNAB402-16GZ-100T-150-T6 4.02 GE14 10x10 GE14-P10DNAB402-19GZ-100T-150-T6 4.02 GE14 10x10 GE14-P10DNAB404-15GZ-100T-150-T6 4.04 GE14 10x10 GE14-P10DNAB406-16GZ-100T-150-T6 4.06 GE14 10x10 GE14-P10DNAB406-15GZ-100T-150-T6 4.06 GE14 10x10 GE14-P10DNAB408-16GZ-100T-150-T6 4.08 GE14 10x10 GE14-P10DNAB408-17GZ-100T-150-T6 4.08 GE14 10x10 GE14-P10DNAB412-16GZ-100T-150-T6 4.12 GE14 10x10 GE14-P10DNAB417-16GZ-100T-150-T6 4.17 GE14 10x10 GE14-P10DNAB418-16GZ-100T-150-T6 4.18 GE14 10x10 GE14-P10DNAB419-16GZ-100T-150-T6 4.19 GE14 10x10 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-11 Table B.3 Non-AREVA Supplied Fuel Inventory (Continued)

Nominal Product Bundle Design Name Array enrichment Line Unit 2: Spent Fuel Pool GE2-7DB110 1.1 GE2 7x7 GE3-7DB250-3G3/2G4 2.5 GE3 7x7 GE3-7DB250-4G3 2.5 GE3 7x7 GE6-P8DRB265-6G3 2.65 GE6 8x8 LTA-QUAD+270-7G5 2.7 LTA 4-4x4 GE4-8DB274-5G2 2.74 GE4 8x8 GE4-8DB274-5G3 2.74 GE4 8x8 GE5-8DRB284-4G4/3G2 2.84 GE5 8x8 GE6-P8DRB284-4G4/3G2 2.84 GE6 8x8 GE6-P8DRB284-7GZ 2.84 GE6 8x8 GE7-P8DRB284-4G4/3G2 2.84 GE7 8x8 GE7B-BP8DRB299-7G4

• 2.99 GE7B 8x8 GE7B-BP8DRB301-6G4

• 3.01 GE7B 8x8 GE9B-P8DWB319-9GZ-80M-150-T 3.19 GE9B 8x8 GE9B-P8DWB325-10GZ-80M-150-T 3.25 GE9B 8x8 GE9B-P8DWB326-7GZ-80M-150-T 3.26 GE9B 8x8 GE11-P9HUB366-12G4.0-100T-146-T 3.66 GE11 9x9 GE11-P9HUB367-14GZ-100T-146-T 3.67 GE11 9x9 GE14-P10DNAB367-14GZ-1T-150-T-2602 3.67 GE14 10x10 GE13-P9HTB384-12G4.0-100T-146-T 3.84 GE13 9x9 GE13-P9DTB391-13GZ-100T-146-T 3.91 GE13 9x9 GE13-P9DTB401-14GZ-100T-146-T 4.01 GE13 9x9 GE13-P9DTB406-13GZ-100T-146-T 4.06 GE13 9x9 GE13-P9DTB412-2G7.0/11G5.0-1T-146-T 4.12 GE13 9x9 GE14-P10DNAB416-16GZ-1T-150-T-2600 4.16 GE14 10x10 GE14-P10DNAB416-16GZ-1T-150-T-2601 4.16 GE14 10x10 GE14-P10DNAB416-18GZ-1T-150-T-2627 4.16 GE14 10x10 GE14-P10DNAB417-18GZ-1T-150-T-2628 4.17 GE14 10x10

• The GE7 geometry provides a representation of earlier 8x8 designs which used multiple water rods unlike the large single water rod in the GE9B design.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-12 Table B.3 Non-AREVA Supplied Fuel Inventory (Continued)

Nominal Product Bundle Design Name Array enrichment Line Unit 3: Spent Fuel Pool GE4-8DB219-4G4/2G2.5/1G1.5 2.19 GE4 8x8 GE4-8DB219-2G2.5/1G1.5 2.19 GE4 8x8 GE6-P8DRB265-6G2 2.65 GE6 8x8 GE5-8DRB265-6G2 2.65 GE5 8x8 LTA-P8DRB284-7GZLTA 2.84 LTA 8x8 GE7B-BP8DRB284-4G4/3G2 2.84 GE7B 8x8 GE6-P8DRB284-7GZ 2.84 GE6 8x8 GE6-P8DRB299-7G4 2.99 GE6 8x8 LTA-P8DRB314-GZLTA

• 3.14 LTA 8x8 GE11-P9HUB323-8G4.0-100T-146-T 3.23 GE11 9x9 GE11-P9HUB323-5G5/4G4-100T-146-T 3.23 GE11 9x9 GE11-P9HUB325-14GZ-100T-146-T 3.25 GE11 9x9 GE13-P9HTB372-11GZ-100T-146-T 3.72 GE13 9x9 GE13-P9HTB386-12GZ-100T-146-T 3.86 GE13 9x9 GE13-P9DTB400-13GZ1-100T-146-T 4.00 GE13 9x9 GE14-P10DNAB401-14GZ-1T-150-T-2514 4.01 GE14 10x10 GE14-P10DNAB402-15GZ-1T-150-T-2513 4.02 GE14 10x10 GE13-P9DTB414-15GZ-100T-146-T 4.14 GE13 9x9

• This design was screened out since only four of these LUAs exist and the enrichment is slightly below the 3.21% enrichment screening criteria. It is also noted that the presence of gadolinia in these lattices will significantly reduce the peak in-rack reactivity when compared to the REBOL non-gad lattices.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-13 Table B.4 AREVA Supplied Fuel Inventory Max Planar Reload Bundle Design Name Type Enrichment Unit 3 A10-3813B-13GV80 4.157 A10-4077B-15GV80 4.381 BFE3-12 CGU A10-4088B-13GV80 4.393 A10-1623B-5GV80 1.739 A10-4171B-14GV80-FCB 4.483 BFE3-13 A10-4163B-16GV80-FCB 4.477 BLEU A10-4181B-13GV80-FCB 4.495 A10-4218B-15GV80-FCC 4.557 BFE3-14 BLEU A10-4218B-13GV80-FCC 4.543 A10-3831B-15GV80-FCD 4.184 A10-3403B-9GV80-FCD 3.739 A10-3392B-10GV80-FCD 3.724 BFE3-15 BLEU A10-4218B-15GV80-FCC* 4.557 A10-4218B-13GV80-FCC* 4.543 A10-3757B-10GV80-FCC* 3.997 A10-3440B-11GV80-FCE 3.668 A10-3826B-13GV80-FCE 4.111 BFE3-16 BLEU A10-4075B-13GV80-FCE 4.384 A10-4081B-12GV80-FCE 4.391

• This design was originally fabricated as part of the BFE3-14 reload but was used in both Unit 2 and Unit 3.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-14 Table B.4 AREVA Supplied Fuel Inventory (Continued)

Max Planar Reload Bundle Design Name Type Enrichment Unit 2 BFE2-14 A10-3920B-14GV70 4.245 BLEU A10-4227B-15GV80-FBB 4.555 BFE2-15 A10-4239B-15GV80-FBB 4.555 BLEU A10-3552B-10GV80-FBB 3.937 A10-4019B-14GV80-FBC 4.318 A10-3841B-14GV80-FBC 4.121 BFE2-16 BLEU A10-4218B-13GV80-FCC* 4.543 A10-3757B-10GV80-FCC* 3.997 A10-3799B-14GV80-FBD 4.157 BFE2-17 BLEU A10-4004B-15GV80-FBD 4.303 Unit 1 A10-3562B-14GV80 FAA 4.059 BFE1-10 A10-3676B-10GV80 FAA 4.109 BLEU A10-4111B-15GV80 FAA 4.424 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-15 Table B.5 Non-AREVA Supplied Fuel Limiting Bundles In-Rack Reactivity Comparison (by product line)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Bundle Bottom Top Depletion Depletion Depletion GE7 - 8x8 Array GE7B-BP8DRB299-7G4 1 X X 0.7838 0.7853 0.7869 GE7B-BP8DRB301-6G4 1 X X 0.7853 0.7878 0.7915 Max. Top and Bottom Lattice 0.7853 0.7878 0.7915 GE9 - 8x8 Array 1 X 0.7853 0.7877 0.7906 2 X X 0.8085 0.8094 0.8100 GE9B-P8DWB325-10GZ-80M-150-T 3 X 0.8079 0.8086 0.8086 4 X 0.8004 0.8018 0.8031 1 X 0.7855 0.7879 0.7908 GE9B-P8DWB326-7GZ-80M-150-T 2 X X 0.8101 0.8110 0.8114 3 X 0.8007 0.8021 0.8034 1 X 0.7843 0.7866 0.7891 GE9B-P8DWB319-9GZ-80M-150-T 2 X X 0.8019 0.8036 0.8047 3 X 0.7930 0.7953 0.7972 Max. Bottom Lattice 0.8101 0.8110 0.8114 Max. Top Lattice 0.8101 0.8110 0.8114 GE11 - 9x9 Array 1 X X 0.8214 0.8205 0.8185 GE11-P9HUB366-12G4.0-100T-146-T 2 X 0.8237 0.8233 0.8218 1 X X 0.8214 0.8205 0.8185 GE11-P9HUB367-14GZ-100T-146-T 2 X 0.8136 0.8125 0.8105 3 X 0.8237 0.8233 0.8218 1 X X 0.7997 0.7999 0.8001 GE11-P9HUB325-14GZ-100T-146-T 2 X 0.7905 0.7910 0.7914 3 X 0.8004 0.8012 0.8020 1 X X 0.7939 0.7963 0.7988 GE11-P9HUB323-5G5/4G4-100T-146-T 2 X 0.7947 0.7982 0.8016 1 X X 0.8050 0.8064 0.8079 GE11-P9HUB323-8G4.0-100T-146-T 2 X 0.8074 0.8084 0.8094 Max. Bottom Lattice 0.8214 0.8205 0.8185 Max. Top Lattice 0.8237 0.8233 0.8218 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-16 Table B.5 Non-AREVA Supplied Fuel Limiting Bundles In-Rack Reactivity Comparison (by product line) - (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Bundle Bottom Top Depletion Depletion Depletion GE13 - 9x9 Array 1 X X 0.8227 0.8210 0.8182 GE13-P9HTB384-12G40-100T-146-T 2 X 0.8254 0.8246 0.8231 1 X 0.8051 0.8041 0.8007 GE13-P9DTB406-13GZ-100T-146-T 2 X X 0.8163 0.8142 0.8099 3 X 0.8199 0.8185 0.8154 1 X 0.7849 0.7840 0.7814 GE13-P9DTB401-14GZ-100T-146-T 2 X X 0.7980 0.7966 0.7933 3 X 0.8004 0.8000 0.7980 1 X 0.7816 0.7819 0.7815 GE13-P9DTB391-13GZ-100T-146-T 2 X X 0.8029 0.8025 0.8009 3 X 0.8056 0.8059 0.8053 1 X X 0.8132 0.8114 0.8075 GE13-P9DTB412-2G7l11G5-1T-146-T 2 X 0.8156 0.8144 0.8114 1 X 0.8019 0.8020 0.8014 GE13-P9HTB372-11GZ-100T-146-T 2 X X 0.8021 0.8021 0.8016 3 X 0.8027 0.8038 0.8048 1 X 0.8193 0.8184 0.8158 GE13-P9HTB386-12GZ-100T-146-T 2 X X 0.8198 0.8187 0.8160 3 X 0.8217 0.8216 0.8204 1 X 0.7965 0.7956 0.7931 GE13-P9DTB400-13GZ1-100T-146-T 2 X X 0.8043 0.8031 0.8003 3 X 0.8070 0.8063 0.8042 1 X 0.8011 0.7990 0.7945 GE13-P9DTB414-15GZ-100T-146-T 2 X X 0.8135 0.8117 0.8073 3 X 0.8178 0.8163 0.8123 Max. Bottom Lattice 0.8227 0.8210 0.8182 Max. Top Lattice 0.8254 0.8246 0.8231 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-17 Table B.5 Non-AREVA Supplied Fuel Limiting Bundles In-Rack Reactivity Comparison (by product line) - (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Bundle Bottom Top Depletion Depletion Depletion GE14 - 10x10 Array 1 X 0.7691 0.7707 0.7720 2 X X 0.7816 0.7852 0.7885 GE14-P10DNAB377-16GZ-100T-150-T6 3 X 0.7613 0.7652 0.7696 4 X 0.7825 0.7862 0.7904 5 X 0.8578 0.8583 0.8579 1 X 0.7829 0.7817 0.7790 2 X X 0.7873 0.7874 0.7866 GE14-P10DNAB402-16GZ-100T-150-T6 3 X 0.7663 0.7680 0.7692 4 X 0.7879 0.7896 0.7906 5 X 0.8605 0.8596 0.8574 1 X 0.7592 0.7643 0.7703 2 X X 0.7718 0.7776 0.7842 GE14-P10DNAB350-16GZ-100T-150-T6 3 X 0.7504 0.7579 0.7663 4 X 0.7698 0.7776 0.7871 5 X 0.8466 0.8470 0.8467 1 X 0.7923 0.7904 0.7863 2 X X 0.7969 0.7963 0.7942 GE14-P10DNAB419-16GZ-100T-150-T6 3 X 0.7786 0.7796 0.7792 4 X 0.8009 0.8020 0.8013 5 X 0.8539 0.8530 0.8505 1 X 0.7633 0.7657 0.7680 2 X X 0.7763 0.7805 0.7846 GE14-P10DNAB368-15GZ-100T-150-T6 3 X 0.7551 0.7598 0.7650 4 X 0.7752 0.7798 0.7854 5 X 0.8502 0.8508 0.8509 1 X 0.7851 0.7831 0.7793 2 X X 0.7839 0.7828 0.7803 GE14-P10DNAB402-19GZ-100T-150-T6 3 X 0.7637 0.7647 0.7648 4 X 0.7854 0.7863 0.7863 5 X 0.8619 0.8608 0.8580 1 X 0.7692 0.7707 0.7717 2 X X 0.7816 0.7852 0.7885 GE14-P10DNAB377-17GZ-100T-150-T6 3 X 0.7613 0.7652 0.7696 4 X 0.7825 0.7862 0.7904 5 X 0.8578 0.8583 0.8579 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-18 Table B.5 Non-AREVA Supplied Fuel Limiting Bundles In-Rack Reactivity Comparison (by product line) - (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Bundle Bottom Top Depletion Depletion Depletion GE14 - 10x10 Array (Continued) 1 X 0.7650 0.7644 0.7618 2 X X 0.7649 0.7648 0.7633 GE14-P10DNAB406-16GZ-100T-150-T6 3 X 0.7441 0.7463 0.7481 4 X 0.7646 0.7669 0.7689 1 X 0.7635 0.7632 0.7608 2 X X 0.7633 0.7637 0.7626 GE14-P10DNAB406-15GZ-100T-150-T6 3 X 0.7414 0.7442 0.7466 4 X 0.7614 0.7645 0.7673 1 X 0.7723 0.7711 0.7677 2 X X 0.7705 0.7701 0.7677 GE14-P10DNAB418-16GZ-100T-150-T6 3 X 0.7532 0.7548 0.7552 4 X 0.7741 0.7760 0.7764 1 X 0.7694 0.7694 0.7686 2 X X 0.7687 0.7697 0.7708 GE14-P10DNAB400-17GZ-100T-150-T6 3 X 0.7407 0.7435 0.7464 4 X 0.7612 0.7639 0.7669 1 X 0.7721 0.7709 0.7676 2 X X 0.7703 0.7699 0.7676 GE14-P10DNAB417-16GZ-100T-150-T6 3 X 0.7838 0.7835 0.7816 4 X 0.8061 0.8059 0.8040 1 X 0.7667 0.7660 0.7636 2 X X 0.7664 0.7662 0.7647 GE14-P10DNAB408-16GZ-100T-150-T6 3 X 0.7461 0.7481 0.7494 4 X 0.7668 0.7689 0.7703 1 X 0.7684 0.7675 0.7649 2 X X 0.7670 0.7667 0.7650 GE14-P10DNAB412-16GZ-100T-150-T6 3 X 0.7472 0.7492 0.7502 4 X 0.7678 0.7700 0.7712 1 X 0.7632 0.7631 0.7615 2 X X 0.7629 0.7633 0.7626 GE14-P10DNAB404-15GZ-100T-150-T6 3 X 0.7427 0.7453 0.7474 4 X 0.7629 0.7658 0.7682 1 X 0.7670 0.7658 0.7630 2 X X 0.7666 0.7660 0.7640 GE14-P10DNAB408-17GZ-100T-150-T6 3 X 0.7469 0.7484 0.7492 4 X 0.7677 0.7694 0.7702 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-19 Table B.5 Non-AREVA Supplied Fuel Limiting Bundles In-Rack Reactivity Comparison (by product line) - (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Bundle Bottom Top Depletion Depletion Depletion GE14 - 10x10 Array (Continued) 1 X 0.7985 0.7961 0.7914 2 X X 0.8045 0.8027 0.7990 GE14-P10DNAB416-16GZ-1T-150-T-2600 3 X 0.7879 0.7876 0.7859 4 X 0.8106 0.8104 0.8085 1 X 0.8027 0.8014 0.7986 2 X X 0.8063 0.8052 0.8027 GE14-P10DNAB416-16GZ-1T-150-T-2601 3 X 0.7904 0.7910 0.7907 4 X 0.8129 0.8137 0.8132 1 X 0.7718 0.7735 0.7758 2 X X 0.7774 0.7791 0.7812 GE14-P10DNAB367-14GZ-1T-150-T-2602 3 X 0.7570 0.7593 0.7622 4 X 0.7786 0.7807 0.7833 1 X 0.7981 0.7955 0.7905 2 X X 0.8042 0.8022 0.7981 GE14-P10DNAB416-18GZ-1T-150-T-2627 3 X 0.7874 0.7869 0.7849 4 X 0.8104 0.8099 0.8075 1 X 0.8027 0.8012 0.7979 2 X X 0.8062 0.8048 0.8018 GE14-P10DNAB417-18GZ-1T-150-T-2628 3 X 0.7904 0.7908 0.7899 4 X 0.8131 0.8135 0.8124 1 X 0.8045 0.8032 0.8006 2 X X 0.8086 0.8074 0.8050 GE14-P10DNAB402-15GZ-1T-150-T-2513 3 X 0.7928 0.7931 0.7925 4 X 0.8156 0.8160 0.8152 1 X 0.8040 0.8025 0.7991 2 X X 0.8071 0.8054 0.8018 GE14-P10DNAB401-14GZ-1T-150-T-2514 3 X 0.7905 0.7902 0.7886 4 X 0.8134 0.8131 0.8111 Max. Bottom Lattice 0.8086 0.8074 0.8050 Max. Top Lattice 0.8619 0.8608 0.8580 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-20 Table B.6 AREVA Supplied Fuel Limiting Lattices In-Rack Reactivity Comparison Peak In-Rack k-infinity

[ ] [ ] [ ]

Lattice Bottom Top Depletion Depletion Depletion A10B-4059L-14GV80-FAA X 0.7764 0.7789 0.7811 A10B-4058L-12G80-FAA X X 0.7765 0.7800 0.7829 A10T-3235L-8G80-FAA X 0.7534 0.7596 0.7665 A10T-3236L-8G50-FAA X 0.7888 0.7919 0.7954 A10B-4109L-10G80-FAA X X 0.7835 0.7871 0.7909 A10T-3501L-8G80-FAA X 0.7660 0.7715 0.7774 A10T-3502L-8G50-FAA X 0.8018 0.8045 0.8070 A10B-4424L-15GV80-FAA X 0.7964 0.7974 0.7980 A10B-4423L-13GV80-FAA X X 0.7965 0.7976 0.7984 A10T-4311L-11GV80-FAA X 0.8028 0.8048 0.8065 A10T-4241L-11G50-FAA X 0.8308 0.8312 0.8306 A10B-4245L-14G70 X 0.7981 0.7999 0.8007 A10B-4236L-12G70 X X 0.7986 0.8002 0.8007 A10T-4040L-13G70 X 0.7982 0.7999 0.8002 A10T-4030L-11G50 X 0.8218 0.8225 0.8220 A10B-4545L-15G80-FBB X 0.7902 0.7910 0.7914 A10B-4555L-13G80-2CGU495-FBB X X 0.7953 0.7964 0.7970 A10T-4414L-12G80-3CGU495-FBB X 0.7961 0.7986 0.8008 A10T-4415L-12G50-3CGU495-FBB X 0.8363 0.8362 0.8355 A10B-4545L-15G80-FBB X 0.7902 0.7910 0.7914 A10B-4555L-13G80-2CGU495-FBB X X 0.7953 0.7964 0.7970 A10T-4454L-11G80-4CGU495-FBB X 0.7941 0.7979 0.8023 A10T-4454L-11G50-4CGU495-FBB X 0.8352 0.8362 0.8370 A10B-3698L-10G80-FBB X X 0.7677 0.7719 0.7760 A10T-3937L-8G80-2CGU495-FBB X 0.7850 0.7884 0.7916 A10B-4306L-14G80-FBC X 0.7873 0.7886 0.7893 A10B-4318L-12G80-FBC X X 0.7905 0.7922 0.7939 A10T-4214L-12G75-FBC X 0.7899 0.7943 0.7993 A10T-4213L-12G50-FBC X 0.8250 0.8266 0.8279 A10B-4115L-14G80-FBC X 0.7713 0.7746 0.7787 A10B-4121L-13G80-FBC X X 0.7774 0.7802 0.7832 A10T-4029L-13G75-FBC X 0.7805 0.7846 0.7898 A10T-4024L-13G50-FBC X 0.8143 0.8159 0.8177 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-21 Table B.6 AREVA Supplied Fuel Limiting Lattices In-Rack Reactivity Comparison (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Lattice Bottom Top Depletion Depletion Depletion A10B-4157L-14G80-FBD X 0.7700 0.7737 0.7783 A10B-4157L-13G80-FBD X X 0.7748 0.7781 0.7817 A10T-3812L-13G70-FBD X 0.7773 0.7817 0.7869 A10T-3812L-13G50-FBD X 0.8049 0.8071 0.8094 A10B-4303L-15G80-FBD X 0.7776 0.7798 0.7818 A10B-4303L-14G80-FBD X X 0.7776 0.7804 0.7833 A10T-4189L-14G70-FBD X 0.7876 0.7915 0.7959 A10T-4188L-14G50-FBD X 0.8168 0.8184 0.8199 A10B-4148L-13G80 X 0.7869 0.7896 0.7926 A10B-4157L-12G80 X X 0.7888 0.7919 0.7947 A10T-3863L-12G70 X 0.7901 0.7942 0.7983 A10T-3874L-10G50 X 0.8252 0.8265 0.8271 A10B-4375L-15G80 X 0.7905 0.7923 0.7943 A10B-4381L-14G80 X X 0.7929 0.7949 0.7966 A10T-4267L-14G80 X 0.7884 0.7926 0.7970 A10T-4295L-10G50 X 0.8426 0.8429 0.8420 A10B-4387L-13G80 X 0.7952 0.7973 0.7992 A10B-4393L-12G80 X X 0.7970 0.7994 0.8013 A10T-4281L-12G70 X 0.8066 0.8094 0.8119 A10T-4295L-10G50 X 0.8426 0.8429 0.8420 A10B-1739L-5G80 X X 0.6537 0.6648 0.6799 A10T-1728L-4G80 X 0.6455 0.6568 0.6715 A10B-4477L-14G80-1CGU495-FCB X 0.7902 0.7914 0.7922 A10B-4483L-13G80-2CGU495-FCB X X 0.7926 0.7940 0.7952 A10T-4374L-13G70-2CGU495-FCB X 0.8033 0.8056 0.8079 A10T-4374L-13G40-2CGU495-FCB X 0.8466 0.8463 0.8451 A10B-4465L-16G80-FCB X 0.7877 0.7887 0.7892 A10B-4477L-14G80-2CGU495-FCB X X 0.7902 0.7915 0.7923 A10T-4367L-14G80-2CGU495-FCB X 0.7864 0.7899 0.7935 A10T-4367L-14G40-2CGU495-FCB X 0.8442 0.8438 0.8427 A10B-4483L-13G80-FCB X 0.7923 0.7937 0.7949 A10B-4495L-11G70-2CGU495-FCB X X 0.8112 0.8120 0.8121 A10T-4387L-11G70-2CGU495-FCB X 0.8107 0.8122 0.8132 A10T-4387L-11G40-2CGU495-FCB X 0.8520 0.8514 0.8496 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-22 Table B.6 AREVA Supplied Fuel Limiting Lattices In-Rack Reactivity Comparison (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Lattice Bottom Top Depletion Depletion Depletion A10B-4545L-15G80-FCC X 0.7906 0.7913 0.7916 A10B-4557L-13G80-2CGU495-FCC X X 0.7956 0.7967 0.7973 A10T-4386L-13G80-2CGU495-FCC X 0.7909 0.7939 0.7972 A10T-4386L-12G50-2CGU495-FCC X 0.8355 0.8356 0.8348 A10B-4543L-13G80-FCC X X 0.7948 0.7959 0.7966 A10T-4399L-11G80-2CGU495-FCC X 0.7988 0.8010 0.8029 A10T-4399L-11G50-2CGU495-FCC X 0.8382 0.8381 0.8371 A10B-3997L-10G80-FCC X X 0.7797 0.7831 0.7861 A10T-3997L-8G80-FCC X 0.7870 0.7909 0.7941 A10T-3997L-8G50-FCC X 0.8237 0.8250 0.8254 A10B-4184L-15GV80-FCD X 0.7798 0.7818 0.7838 A10B-4103L-13G80-FCD X X 0.7794 0.7821 0.7850 A10T-3961L-11G80-FCD X 0.7818 0.7853 0.7889 A10T-3962L-11G50-FCD X 0.8205 0.8214 0.8220 A10B-3739L-9G80-FCD X X 0.7734 0.7775 0.7821 A10T-3360L-9GV70-FCD X 0.7748 0.7790 0.7828 A10T-3360L-9G40-FCD X 0.8070 0.8091 0.8113 A10B-3724L-9G80-FCD X 0.7730 0.7771 0.7818 A10B-3724L-10GV80-FCD X 0.7721 0.7762 0.7808 A10T-3356L-10GV70-FCD X 0.7737 0.7777 0.7814 A10T-3356L-10G40-FCD X 0.8050 0.8070 0.8089 A10B-3668L-11GV80-FCE X X 0.7690 0.7728 0.7768 A10T-3632L-10G80-FCE X 0.7681 0.7722 0.7764 A10T-3632L-10G50-FCE X 0.8069 0.8083 0.8092 A10B-4111L-13GV80-FCE X 0.7876 0.7896 0.7912 A10B-4111L-11GV80-FCE X X 0.7871 0.7896 0.7919 A10T-3994L-11GV70-FCE X 0.8136 0.8148 0.8158 A10T-3994L-11GV50-FCE X 0.8243 0.8248 0.8246 A10B-4384L-13G80-FCE X X 0.7850 0.7869 0.7894 A10T-4253L-11GV80-FCE X 0.7955 0.7991 0.8034 A10T-4254L-10G50-FCE X 0.8294 0.8307 0.8316 A10B-4391L-12G80-FCE X X 0.7881 0.7902 0.7923 A10T-4260L-10G70-FCE X 0.8029 0.8058 0.8088 A10T-4260L-10G50-FCE X 0.8295 0.8307 0.8316 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-23 Table B.6 AREVA Supplied Fuel Limiting Lattices In-Rack Reactivity Comparison (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Lattice Bottom Top Depletion Depletion Depletion Max. Bottom Lattice [ ] 0.8112 0.8120 0.8121 Max. Top Lattice [ ] 0.8520 0.8514 0.8496 ATRIUM 10 Reference Bounding Lattices*

A10B-460L-8G40 X X 0.8738 0.8734 0.8715 A10T-450L-8G40 X 0.8747 0.8747 0.8733

• Reference B.1.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-24 Table B.7 Summary of Damaged and Modified Bundles Product Nominal Bundle ID Description Line Enrichment Replaced Fuel Rods YJK363 GE13 3.84 pin location B3 replaced with SS rod YJS776 GE13 4.01 pin location A5 replaced with SS rod YJS614 GE13 4.06 pin locations G9 and H2 replaced with SS rods YJS734 GE13 4.06 pin locations B8 and H2 replaced with SS rods YJS616 GE13 4.06 pin location B8 replaced with SS rod YJN587 GE13 3.72 pin location E9 replaced with SS rod JLB602 GE14 4.01 pin locations A4 and D1 replaced with SS rods FCA199 ATRIUM-10 4.077 pin location L4 replaced with SS rod Broken Fuel Rods Rod F7 broke into two pieces as it was withdrawn from the bundle. The two halves of the rod were restored to the bundle and the upper tie plate LJD968 GE5 2.84 installed.

Rod C7 was identified as broken within the fuel assembly. The rod was not disturbed (no attempt made to withdraw).

Rod G3 was identified as broken within the fuel LJE002 GE5 2.84 assembly. The rod was not disturbed (no attempt made to withdraw).

Rod J5 broke into two pieces (severed at a prior identified 91 inch defect location of a YJN587 GE13 3.72 circumferential crack and fracture) as it was rotated during a fuel inspection. The failed rod was stabilized within the bundle.

Rod A5 was noted as having a circumferential crack at approximately 127 inches. Although not JLB435 GE14 4.02 separated, rod A5 is conservatively classified as broken using visual information of this cracking.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-1 Appendix C KENO V.a Bias and Bias Uncertainty Evaluation The purpose of this Appendix is to determine the bias of the keff calculated with the SCALE 4.4a computer code for spent fuel pool criticality analysis. A statistical methodology is used to evaluate criticality benchmark experiments that are appropriate for the expected range of parameters. The scope of this report is limited to the validation of the KENO V.a module and CSAS25 driver in the SCALE 4.4a code package for use with the 44 energy group cross-section library 44GROUPNDF5 for spent fuel criticality analyses.

This calculation is performed according to the general methodology described in Reference C.2 (NUREG/CR-6698) that is also briefly described in Section C.1. The critical experiments selected to benchmark the computer code system are discussed in Section C.3. The results of the criticality benchmark calculations, the trending analysis, the basis for the statistical technique chosen, the bias, and the bias uncertainty are presented in Sections C.4 - C.7. Final results are summarized in Section C.8.

C.1 Statistical Method for Determining the Code Bias As presented in Reference C.2 (NUREG/CR-6698), the validation of the criticality code must use a statistical analysis to determine the bias and bias uncertainty in the calculation of keff. The approach involves determining a weighted mean of keff that incorporates the uncertainty from both the measurement (exp) and the calculation method (calc). A combined uncertainty can be determined using Equation 3 from Reference C.2, for each critical experiment:

2 t = calc + 2exp The weighted mean keff, the variance about the mean (s2), and the average total uncertainty of the benchmark experiments ( 2 ) can be calculated using the weighting factor 1/i2 (see Eq. 4, 5, and 6 in Reference C.2). The final objective is to determine the square root of the pooled variance, defined as (Eq. 7 from Reference C.2):

Sp = s2 + 2 Determination of the keff bias and uncertainty requires evaluation of the distribution of data and investigation of possible trends. Trends are identified by regression analysis to determine key AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-2 parameters including the slope, intercept, coefficient of determination, the T-value associated with the Students T-distribution, and a check for normality of the distribution of residuals in order to evaluate goodness-of-fit. These key parameters are used to establish the statistical significance of the calculated trend. If a trend is found to have statistical significance, then a one-sided lower tolerance band may be used to determine the bias and uncertainty. This method provides a fitted curve (KL(x)), above which 95% of the true population of keff is expected to lie, with a 95% confidence level.

If no trends of statistical significance are found and the data is normally distributed, then the bias and uncertainty can be based on a single-sided lower tolerance limit technique. This method defines a lower tolerance limit (KL) above which 95% of the true population of keff is expected to lie, with a 95% confidence level. The KL is defined in terms of the weighted-average of the data ( k eff ), the 95/95 single-sided lower tolerance factor (C95/95 - dependent on the size of the observed population), and the square-root of the pooled variance (Sp), as shown below.

K=L k eff C95 / 95 Sp In this case, the statistical bias and uncertainty are defined as shown below.

Bias =k eff 1, for k eff < 1, otherwise, Bias =

0 Uncertainty = C95/95 SP Finally, if the data is not normally distributed, then a nonparametric analysis can be employed.

This method considers the size of the observed population and determines the mth lowest value (keffm < 1) and the associated uncertainty (m) to determine a limiting value (KL), above which 95% of the true population of keff is expected to lie, with a 95% confidence level. Here, the sample size must exceed 59 in order to attain a 95/95 confidence interval, otherwise additional Non-Parametric Margin (NPM - defined by NUREG/CR-6698, see Reference C.2) must be included in the KL, as shown below.

K L =k eff m - m -NPM Bias = k eff m -1 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-3 Uncertainty = m + NPM Regardless of the method employed, the Area of Applicability (AOA) must also be defined based on evaluation of key parameters of the criticality experiments that are included in the validation. Key parameters fall into three categories: materials, geometry, and neutron energy spectrum. In general, use of the criticality evaluation is restricted to the range of parameters identified in the AOA.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-4 C.2 Area of Applicability Required for the Benchmark Experiments Commercial reactor spent fuel pools will primarily contain nuclear fuel in metal rods in a square array. This fuel is characterized by the parameter values provided in Table C.1. These typical values were used as primary tools in selecting the benchmark experiments appropriate for determining the code bias.

Benchmark calculations have been made on selected critical experiments, chosen, in so far as possible, to bound the range of variables in the spent fuel rack analysis. In rack designs, the most significant parameters affecting criticality are: (1) the fuel enrichment, (2) the neutron absorbing material, and (3) the lattice spacing. Other parameters have a smaller effect but have also been included in the analysis.

One possible way of representing the data is through a spectral parameter that incorporates influences from the variations in other parameters. The energy of the average lethargy causing fission (EALF) is this type of parameter and it is computed by KENO V.a. The range for this parameter is also included in Table C.1.

Table C.1 Range of Values for Key Spent Fuel Pool Parameters Parameter Range of Values Fissile material - Physical/Chemical Form UO2 rods Enrichment 2.35 to 4.74 wt% U-235 Moderation/Moderator Heterogeneous/Water Lattice Square, Rectangular Pitch 1.26 to 2.54 cm Clad Zircaloy, Aluminum Anticipated Absorber/Materials Boron, Stainless Steel, Water Moderating Ratio (H/X) 110 to >400 Reflection Water, Stainless Steel Neutron Energy Spectrum (Energy of the 0.060 to 0.247 eV Average Lethargy Causing Fission)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-5 C.3 Description of the Criticality Experiments Selected The set of criticality benchmark experiments has been constructed to accommodate large variations in the range of parameters of the rack configurations and also to provide adequate statistics for the evaluation of the code bias.

Sixty eight (68) critical configurations were selected from various sources. These benchmarks include configurations performed with lattices of UO2 fuel rods in water having various enrichments and moderating ratios (H/X). The area of applicability (AOA) is established within this range of benchmark experiment parameter values.

A brief description of the selected benchmark experiments is presented in Table C.2. The table includes the references where detailed descriptions of the experiments are presented.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-6 Table C.2 Descriptions of the Critical Benchmark Experiments Experiment Measured Neutron exp Brief Description Reflector Case Name keff Absorber LEU-COMP-THERM-007 (Reference C.1)

CEA-001-001 1.0000 0.0014 Square pitch fuel rod arrays with varying rod pitch configurations.

CEA-001-002 1.0000 0.0008 Each fuel rod is aluminum clad with UO2 fuel at 4.738 wt%

None Water CEA-001-003 1.0000 0.0007 U235. Performed at CEA CEA-001-004 1.0000 0.0008 Valduc Critical Mass Laboratory.

LEU-COMP-THERM-0034 (Reference C.1)

CEA-003-003 1.0000 0.0039 CEA-003-004 1.0000 0.0039 CEA-003-005 1.0000 0.0039 Borated A 2x2 array UO2 fuel rod clusters CEA-003-006 1.0000 0.0039 with 4.738 wt% U-235 Stainless Steel CEA-003-007 1.0000 0.0039 surrounded by plates of neutron CEA-003-008 1.0000 0.0039 absorbing material. Each fuel rod cluster is comprised of an 18x18 Water CEA-003-010 1.0000 0.0048 array of aluminum clad fuel rods CEA-003-011 1.0000 0.0048 with a square lattice pitch of 1.6 CEA-003-012 1.0000 0.0048 cm. Performed at CEA Valduc Critical Mass Laboratory. Boral CEA-003-013 1.0000 0.0048 CEA-003-014 1.0000 0.0043 CEA-003-015 1.0000 0.0043 LEU-COMP-THERM-039 (Reference C.1)

CEA-005-001 1.0000 0.0014 CEA-005-002 1.0000 0.0014 CEA-005-003 1.0000 0.0014 CEA-005-004 1.0000 0.0014 CEA-005-005 1.0000 0.0009 CEA-005-006 1.0000 0.0009 CEA-005-007 1.0000 0.0012 Square pitch (pitch = 1.26 cm) fuel rod arrays without fuel rods CEA-005-008 1.0000 0.0012 in all positions. Each fuel rod is CEA-005-009 1.0000 0.0012 aluminum clad with UO2 fuel at None Water CEA-005-010 1.0000 0.0012 4.738 wt% U-235. Performed at CEA Valduc Critical Mass CEA-005-011 1.0000 0.0013 Laboratory.

CEA-005-012 1.0000 0.0013 CEA-005-013 1.0000 0.0013 CEA-005-014 1.0000 0.0013 CEA-005-015 1.0000 0.0013 CEA-005-016 1.0000 0.0013 CEA-005-017 1.0000 0.0013 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-7 Table C.2 Descriptions of the Critical Benchmark Experiments (Continued)

Experiment Measured Neutron exp Brief Description Reflector Case Name keff Absorber LEU-COMP-THERM-001 (Reference C.1)

PNL-001-001 0.9998 0.0031 PNL-001-002 0.9998 0.0031 UO2 pellets enriched at 2.35 wt%

PNL-001-003 0.9998 0.0031 U-235 clad in aluminum.

PNL-001-004 0.9998 0.0031 Varying clusters of fuel rods on a 2.032 cm pitch, moderated by None Water PNL-001-005 0.9998 0.0031 water. Single cluster or multiple PNL-001-006 0.9998 0.0031 clusters with varying separation distances.

PNL-001-007 0.9998 0.0031 PNL-001-008 0.9998 0.0031 LEU-COMP-THERM-002 (Reference C.1)

PNL-002-001 0.9997 0.0020 UO2 pellets enriched at 4.31 wt%

PNL-002-002 0.9997 0.0020 U-235. Varying clusters of fuel rods on a 2.54 cm pitch, PNL-002-003 0.9997 0.0020 moderated by water. Single None Water PNL-002-004 0.9997 0.0018 cluster or multiple clusters with PNL-002-005 0.9997 0.0019 varying separation distances.

LEU-COMP-THERM-009 (Reference C.1)

PNL-009-001 1.0000 0.0021 PNL-009-002 1.0000 0.0021 Steel PNL-009-003 1.0000 0.0021 PNL-009-004 1.0000 0.0021 PNL-009-005 1.0000 0.0021 UO2 pellets enriched at 4.31 wt%

PNL-009-006 1.0000 0.0021 U-235 clad in aluminum. Three Stainless Steel 15x8 clusters of fuel rods on a (1.05 - 1.62 PNL-009-007 1.0000 0.0021 2.54 cm pitch, separated by wt% Boron) Water PNL-009-008 1.0000 0.0021 different absorber plates.

PNL-009-009 1.0000 0.0021 Varying separation distances. Boral PNL-009-024 1.0000 0.0021 Aluminum PNL-009-025 1.0000 0.0021 PNL-009-026 1.0000 0.0021 Zircaloy-4 PNL-009-027 1.0000 0.0021 LEU-COMP-THERM-016 (Reference C.1)

PNL-016-008 1.0000 0.0031 PNL-016-009 1.0000 0.0031 Stainless Steel UO2 pellets enriched at 2.35 wt% (1.05 - 1.62 PNL-016-010 1.0000 0.0031 wt% Boron)

U-235 clad in aluminum. Three PNL-016-011 1.0000 0.0031 variable sized clusters of fuel PNL-016-012 1.0000 0.0031 rods on a 2.032 cm pitch, Water PNL-016-013 1.0000 0.0031 separated by absorber plates Boral with varying separation PNL-016-014 1.0000 0.0031 distances.

PNL-016-031 1.0000 0.0031 Zircaloy-4 PNL-016-032 1.0000 0.0031 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-8 C.4 Results of Calculations with SCALE 4.4a The critical experiments described in Section C.3 were modeled with the SCALE 4.4a computer system. The resulting keff and calculational uncertainty, along with the experimental keff and experimental uncertainty are tabulated in Table C.3. The parameters of interest in performing a trending analysis of the bias are also included in the table.

In order to address situations in which the critical experiment being modeled was at other than a critical state (i.e., slightly super or subcritical), the calculated keff is normalized to the k calc experimental kexp, using the following formula (Eq.9 from Reference C.2): k norm =

k exp In the following, the normalized values of the keff were used in the determination of the code bias and bias uncertainty.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-9 Table C.3 SCALE 4.4a Results for the Selected Benchmark Experiments Benchmark SCALE 4.4a Rod Values Calculated Values Enrichment EALF No. Case Name Pitch H/X keff exp keff calc (wt% U-235) (eV)

(cm) 1 CEA-001-001 1.0000 0.0014 0.9928 0.0009 4.74 1.26 110 0.247 2 CEA-001-002 1.0000 0.0008 0.9952 0.0008 4.74 1.6 229 0.110 3 CEA-001-003 1.0000 0.0007 0.9976 0.0009 4.74 2.1 455 0.070 4 CEA-001-004 1.0000 0.0008 0.9977 0.0008 4.74 2.52 693 0.060 5 CEA-003-003 1.0000 0.0039 1.0020 0.0009 4.74 1.6 229 0.143 6 CEA-003-004 1.0000 0.0039 0.9987 0.0009 4.74 1.6 229 0.139 7 CEA-003-005 1.0000 0.0039 0.9982 0.0009 4.74 1.6 229 0.135 8 CEA-003-006 1.0000 0.0039 1.0003 0.0009 4.74 1.6 229 0.131 9 CEA-003-007 1.0000 0.0039 0.9988 0.0008 4.74 1.6 229 0.129 10 CEA-003-008 1.0000 0.0039 0.9986 0.0008 4.74 1.6 229 0.127 11 CEA-003-010 1.0000 0.0048 0.9994 0.0008 4.74 1.6 229 0.149 12 CEA-003-011 1.0000 0.0048 1.0001 0.0009 4.74 1.6 229 0.147 13 CEA-003-012 1.0000 0.0048 0.9967 0.0008 4.74 1.6 229 0.145 14 CEA-003-013 1.0000 0.0048 0.9961 0.0009 4.74 1.6 229 0.142 15 CEA-003-014 1.0000 0.0043 0.9924 0.0009 4.74 1.6 229 0.140 16 CEA-003-015 1.0000 0.0043 0.9954 0.0008 4.74 1.6 229 0.137 17 CEA-005-001 1.0000 0.0014 0.9951 0.0009 4.74 1.26 110 0.227 18 CEA-005-002 1.0000 0.0014 0.9963 0.0010 4.74 1.26 110 0.216 19 CEA-005-003 1.0000 0.0014 0.9978 0.0009 4.74 1.26 110 0.197 20 CEA-005-004 1.0000 0.0014 0.9947 0.0008 4.74 1.26 110 0.186 21 CEA-005-005 1.0000 0.0009 0.9963 0.0008 4.74 1.26 110 0.141 22 CEA-005-006 1.0000 0.0009 0.9986 0.0009 4.74 1.26 110 0.146 23 CEA-005-007 1.0000 0.0012 0.9952 0.0009 4.74 1.26 110 0.217 24 CEA-005-008 1.0000 0.0012 0.9934 0.0010 4.74 1.26 110 0.208 25 CEA-005-009 1.0000 0.0012 0.9957 0.0009 4.74 1.26 110 0.202 26 CEA-005-010 1.0000 0.0012 0.9973 0.0008 4.74 1.26 110 0.176 27 CEA-005-011 1.0000 0.0013 0.9922 0.0008 4.74 1.26 110 0.227 28 CEA-005-012 1.0000 0.0013 0.9937 0.0009 4.74 1.26 110 0.222 29 CEA-005-013 1.0000 0.0013 0.9926 0.0009 4.74 1.26 110 0.219 30 CEA-005-014 1.0000 0.0013 0.9934 0.0009 4.74 1.26 110 0.218 31 CEA-005-015 1.0000 0.0013 0.9944 0.0009 4.74 1.26 110 0.217 32 CEA-005-016 1.0000 0.0013 0.9951 0.0009 4.74 1.26 110 0.214 33 CEA-005-017 1.0000 0.0013 0.9954 0.0009 4.74 1.26 110 0.215 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-10 Table C.3 SCALE 4.4a Results for the Selected Benchmark Experiments (Continued)

Benchmark SCALE 4.4a Rod Values Calculated Values Enrichment EALF No. Case Name Pitch H/X keff exp keff calc (wt% U-235) (eV)

(cm) 34 PNL-001-001 0.9998 0.0031 0.9970 0.0008 2.35 2.032 399 0.096 35 PNL-001-002 0.9998 0.0031 0.9958 0.0009 2.35 2.032 399 0.096 36 PNL-001-003 0.9998 0.0031 0.9947 0.0008 2.35 2.032 399 0.095 37 PNL-001-004 0.9998 0.0031 0.9955 0.0007 2.35 2.032 399 0.095 38 PNL-001-005 0.9998 0.0031 0.9950 0.0007 2.35 2.032 399 0.094 39 PNL-001-006 0.9998 0.0031 0.9963 0.0007 2.35 2.032 399 0.095 40 PNL-001-007 0.9998 0.0031 0.9959 0.0007 2.35 2.032 399 0.093 41 PNL-001-008 0.9998 0.0031 0.9936 0.0007 2.35 2.032 399 0.094 42 PNL-002-001 0.9997 0.0020 0.9963 0.0009 4.31 2.54 256 0.114 43 PNL-002-002 0.9997 0.0020 0.9956 0.0009 4.31 2.54 256 0.114 44 PNL-002-003 0.9997 0.0020 0.9961 0.0008 4.31 2.54 256 0.114 45 PNL-002-004 0.9997 0.0018 0.9951 0.0008 4.31 2.54 256 0.113 46 PNL-002-005 0.9997 0.0019 0.9945 0.0008 4.31 2.54 256 0.111 47 PNL-009-001 1.0000 0.0021 0.9979 0.001 4.31 2.54 256 0.114 48 PNL-009-002 1.0000 0.0021 0.9958 0.0007 4.31 2.54 256 0.113 49 PNL-009-003 1.0000 0.0021 0.9977 0.0008 4.31 2.54 256 0.114 50 PNL-009-004 1.0000 0.0021 0.9968 0.0008 4.31 2.54 256 0.114 51 PNL-009-005 1.0000 0.0021 0.9975 0.0008 4.31 2.54 256 0.115 51 PNL-009-006 1.0000 0.0021 0.9973 0.0009 4.31 2.54 256 0.114 53 PNL-009-007 1.0000 0.0021 0.9961 0.0009 4.31 2.54 256 0.115 54 PNL-009-008 1.0000 0.0021 0.9972 0.0008 4.31 2.54 256 0.114 55 PNL-009-009 1.0000 0.0021 0.9967 0.0008 4.31 2.54 256 0.115 56 PNL-009-024 1.0000 0.0021 0.9964 0.0007 4.31 2.54 256 0.114 57 PNL-009-025 1.0000 0.0021 0.9970 0.0009 4.31 2.54 256 0.114 58 PNL-009-026 1.0000 0.0021 0.9950 0.0008 4.31 2.54 256 0.113 59 PNL-009-027 1.0000 0.0021 0.9957 0.0008 4.31 2.54 256 0.113 60 PNL-016-008 1.0000 0.0031 0.9952 0.0007 2.35 2.032 399 0.097 61 PNL-016-009 1.0000 0.0031 0.9965 0.0008 2.35 2.032 399 0.096 62 PNL-016-010 1.0000 0.0031 0.9946 0.0006 2.35 2.032 399 0.097 63 PNL-016-011 1.0000 0.0031 0.9954 0.0007 2.35 2.032 399 0.096 64 PNL-016-012 1.0000 0.0031 0.9954 0.0007 2.35 2.032 399 0.097 65 PNL-016-013 1.0000 0.0031 0.9960 0.0007 2.35 2.032 399 0.096 66 PNL-016-014 1.0000 0.0031 0.9943 0.0007 2.35 2.032 399 0.097 67 PNL-016-031 1.0000 0.0031 0.9949 0.0008 2.35 2.032 399 0.095 68 PNL-016-032 1.0000 0.0031 0.9965 0.0007 2.35 2.032 399 0.095 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-11 C.5 Trending Analysis The next step of the statistical methodology used to evaluate the code bias for the pool of experiments selected is to identify any trend in the bias. This is done by using the following trending parameters:

• Fuel enrichment (wt% U-235)

• Fuel rod pitch

• Atom ratio of the moderator to fuel (H/X)

• Energy of the Average Lethargy causing Fission, EALF (eV)

The first step in calculating the bias uncertainty limit is to apply regression-based methods to identify any correlation of the calculated values of keff with the trending parameters. The trends show the results of systematic errors or bias inherent in the calculational method used to estimate criticality.

For the critical benchmark experiments that were slightly super or subcritical, an adjustment to the keff value calculated with SCALE 4.4a (kcalc) was done as suggested in Reference C.2. This adjustment is done by normalizing the calculated (kcalc) value to the experimental value (kexp).

This normalization does not affect the inherent bias in the calculation due to very small differences in keff. Unless otherwise mentioned, the normalized keff values (knorm) have been used in all subsequent calculations.

The regression analysis employs the normalized keff values (knorm) and corresponding total uncertainty values (t), which are the values of the dependent variable and the corresponding weighting factors defined by 1/i2, where i = t for the ith data point. Data points consist of the ordered pairs (xi,yi), where yi = keff for the ith data point. Reference C.2 suggests the use of weighting factors to reduce the importance of data with higher uncertainty. For this application, the weighted trends were evaluated and the results were verified by comparison to the non-weighted trending results.

Note that t values are an intermediate calculational result and all downstream calculations should include all significant digits resulting from the intermediate calculation. Therefore, to be consistent with the guidance from Reference C.2, the weighting factors were evaluated as shown below with all significant digits included in later calculations.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-12 1 1

= 2 i2 2 exp + calc i The linear fitting function is defined as: yi = mxi + b, where m and b are the fitting coefficients, slope and intercept, respectively. The slope (m) and intercept (b) are determined by application of the following equations (from Reference C.2, page 8):

1 1 xi yi xi y i m=

2 i2 2

i2 i i i i i i 1 x i2 yi xi x i y i b=

2 2 2

i2 i i i i i i i 2

1 x i2 x i

= 2 2 i2 i i i i i The weighted-average value of the dependent variable ( k eff ) is calculated as follows:

yi i

2

= y = i k eff 1

i 2 i

For the residuals, there are n - 2 = 66 degrees of freedom, since there are n = 68 data points.

The ith value of the regression is expressed as y i = mx i + b and the weighted sums of the squares for the residuals (SSResidual), for the regression (SSRegression), and for the total (SSTotal) are calculated as follows:

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-13 (y y i )2 i

i i2 SS Residual =

1 i

2 i

(y y ) 2 i

i i2 SS Regression =

1 i

2 i

SS Total = SS Re sidual + SS Re gression These, in turn, allow calculation of the goodness-of-fit parameters: coefficient of correlation (r2),

and the TValue corresponding to the Students T-distribution:

SS Re gression r2 =

SS Total (n 2) SS Regression TValue =

SS Re sidual The r2 value represents the proportion of the sum of the squares for the y-values about their mean that can be attributed to a linear relation between x and y. The closer that r2 approaches a value of 1, the better the fit of the data to the linear equation. As described in Section 10.3.2 of Reference C.3, calculated TValues are compared with the critical value of the Students T-distribution with a significance level of = 0.05/2 = 0.025 and n - 2 = 66 degrees of freedom (i.e., a critical value of 1.996). The null hypothesis for this test (H0), is that the slope is not statistically significant; thus, a statistically significant trend may exist if: TValue > 1.996 .

Alternatively, the probability of obtaining a TValue of larger magnitude from a two-tailed T-distribution with the same n - 2 = 66 degrees of freedom is calculated. In general, a low probability (e.g., p < 0.05) is necessary to confirm that a statistically significant trend exists.

In cases where a statistically significant trend is indicated by the Students T-test, then the residuals of the regression are tested to determine if the error component is normally distributed with mean zero, which confirms that the statistical test for significance is valid (Section 10.4 of Reference C.3). The Anderson-Darling test described in Reference C.5 is employed for this AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-14 purpose; calculation of the test statistic (A2) proceeds by first sorting the sample into ascending order.

X1 X n Calculate the sample average X , and standard deviation X :

1 n X=

n X i i =1 n

(X i X)

X = i 1 n 1 Xi X Then, compute standardized values: Yi = .

X Now, the Anderson-Darling test statistic can be calculated, as shown:

n ln ( Pi ) + ln (1 Pn +1i )

( 2i 1) 2 A = n i =1 n Here, Pi is the cumulative normal probability corresponding to the standard score of Yi, defined above. Finally, the calculated A2 value is adjusted for the size of the sample (n):

0.75 2.25 A*= A 2 1.0 + +

n n2 The null hypothesis of normality is rejected if the value of A* exceeds the critical value of 0.752, at a significance level of 0.05. Therefore, if A* 0.752, then the residuals are distributed normally and the statistical test for significance is valid.

Results of the weighted regression analysis and statistical tests are summarized in Table C.4 for all key parameters. This table shows that only H/X produces a valid trend. Therefore, a single-sided lower tolerance band will be used to establish the bias and uncertainty as a function of H/X. Although there is no trend for U-235 enrichment and the residuals for rod pitch and EALF are not normally distributed (indicating an invalid trend), lower tolerance bands have also been calculated for these parameters. The intermediate results are listed in Table C.5.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-15 Calculational details of the single-sided lower tolerance band can be found in Reference C.2 (pages 12 - 13); some details will be repeated here for the sake of convenience, clarity, and for verification of intermediate values used in the calculations. The equation for the single-sided lower tolerance band is as follows:

1 K L ( x ) = K fit ( x ) (S P )fit 2Fa( fit,n2) +

(x x ) + z 2

(n 2) 2P 1 n (x i x )

2 12 ,n2 i

Kfit(x) is the function derived in the trending analysis for independent variable x. Because a positive bias may be non-conservative, the value Kfit = 1.00 is substituted for all x where Kfit(x) >

1.00. Other symbols not previously introduced are defined below:

= p the desired confidence level

= 0.95 Ffit,n-2 = the F distribution percentile with degree of fit (2, for linear) and n-2 degrees of freedom, based on the Excel function FINV with arguments (1-0.95, 2, n-2).

z 2P-1 = the symmetric percentile of the Gaussian (normal) distribution that contains the P fraction, based on the Excel function NORMSINV with argument (0.95).

1-p 1 - 0.95

= g = = 0.025 2 2 c1-2 g ,n-2 = the upper Chi-square percentile

= based on the Excel function CHIINV with arguments (1-0.025, n-2).

In addition to the constants defined above, the equations listed below are quantities that are dependent upon the type of fit and the specific independent variable (except that 2 is constant, as shown).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-16 xi i

2 x= i 1

i 2

i (x i x )2 2I (x i x )

2 i

=

1 1 i

n I 2I n

2 =

1 i

2 i

1 (y i y i )2 n2 i 2

2 s fit =

i 1 1 n i i2 (SP )fit = s fit2

+ 2 Figures C.1 to C.4 show the normalized keff datasets plotted as a function of the weighted U-235 enrichment, rod pitch, H/X, and EALF data, respectively. The plotted data is overlaid with the linear trend line and the lower tolerance band. This lower tolerance band bounds 95% of the population with a confidence level of 95%.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-17 Table C.4 Results Summary for Weighted Trending Analysis Enrichment Rod Pitch Moderating Parameter EALF (eV)

(wt% U-235) (cm) Ratio (H/X)

Slope 4.19E-05 1.14E-03 4.30E-06 -1.75E-02 Intercept 0.9957 0.9838 0.9949 0.9985 2

r 0.0003 0.1331 0.1429 0.3170 Tcrit 1.996 1.996 1.996 1.996 T-value 0.1422 3.1838 3.3175 5.5343 P(T>T-value) 0.8874 0.0022 0.0015 5.802E-07 Valid Trend? NO YES YES YES A* --- 1.2471 0.5916 0.8537 Statistical Test Valid? --- NO YES NO AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-18 Table C.5 Intermediate Results for Lower Tolerance Band Evaluation x (x i

i x) 2 s 2fit (S P )fit Weighted Fit Enrichment 4.390 35.126 3.0529E-06 0.00270 (wt% U-235)

Rod Pitch 1.766 20.620 2.6473E-06 0.00262 (cm)

Moderating Ratio 229.11 1.5584E+06 2.6174E-06 0.00262 H/X EALF 0.1498 0.2082 2.0859E-06 0.00251 (eV)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-19 1.010 1.005 1.000 y = 4.192E-05x + 0.9957 R2 = 0.0003 Normalized k(eff) 0.995 0.990 0.985 0.980 1 2 3 4 5 6 U-235 Enrichment (%)

Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.1 Weighted U-235 Enrichment Trend Evaluation AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-20 1.010 1.005 y = 1.141E-03x + 0.9938 R2 = 0.1331 1.000 Normalized k(eff) 0.995 0.990 0.985 0.980 1.0 1.5 2.0 2.5 3.0 Fuel Rod Pitch (cm)

Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.2 Weighted Fuel Rod Pitch Trend Evaluation AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-21 1.010 1.005 y = 4.299E-06x + 0.9949 1.000 R2 = 0.1429 Normalized k(eff) 0.995 0.990 0.985 0.980 0 100 200 300 400 500 600 700 800 Moderating Ratio, H/X Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.3 Weighted H/X Trend Evaluation AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-22 1.010 1.005 y = -0.01752x + 0.9985 R2 = 0.3170 1.000 Normalized k(eff) 0.995 0.990 0.985 0.980 0.00 0.05 0.10 0.15 0.20 0.25 0.30 EALF (eV)

Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.4 Weighted EALF Trend Evaluation AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-23 C.6 Bias and Bias Uncertainty For situations in which no significant trend in bias is identified the statistical methodology, presented in Reference C.2 and summarized in Section C.1 of this appendix, suggests to first check the distribution of the normalized keff dataset. The Anderson-Darling test statistic is calculated consistent with the description presented in Section C.5. The null hypothesis of normality is rejected if the value of A* exceeds the critical value of 0.752, (based upon a significance level of 0.05). Therefore, if A* 0.752, then the data are distributed normally.

The Anderson-Darling test was completed for the 68 case benchmark set. The resulting Anderson-Darling test statistic modified from the number of data points A* was determined to be 0.4186. A plot of the data relative to a normal distribution is provided in Figure C.5. Based on the test statistic and plot, the benchmark data can be considered normally distributed.

With the assumption of normality being validated, a single-sided lower tolerance limit can be used to determine the bias and uncertainty. For n = 68, the tolerance limit is C95/95 = 1.996, from Reference C.4. Results obtained for the weighted average keff ( k eff ), the variance about the mean (s2), the average total uncertainty ( 2 ), and the square-root of the pooled variance (Sp),

are shown below.

Weighted yi i

2 k eff = y = i

= 0.99585 1

i 2 i

Bias = k eff - 1 = -0.00415 1 (y i y )2 n 1 i i2 2

s = =3.0083-06 1 1 n i i2 n

2 = = 4.2355E-06 1

i 2 i

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-24 SP = s 2 + 2 = 0.00269 The bias and bias uncertainty are:

Bias = -0.00415 Uncertainty = (C95/95)(Sp) = (1.996)(0.00269) = 0.00537 The corresponding lower tolerance limit is:

KL = k eff - (C95/95)(Sp) = 0.99585 - 0.00537 = 0.99048 When this lower tolerance limit, KL = 0.99048, is compared with the lower tolerance bands of the trended data in Figures C.1 through C.4, the lower tolerance limit is not sufficiently conservative to bound all the trended parameters. A minimum keff of 0.98761 is projected for the EALF* trend evaluation (see Figure C.4). Based upon this minimum value, a trend corrected bias can be calculated as follows:

BiasCorr = -0.00415 - (0.99048-0.98761) = -0.00702 If the magnitude of this corrected bias is conservatively adjusted to 0.0075 (with the pooled uncertainty rounded to 0.0027) a bounding limit is established as shown:

kL = (1 - 0.0075) - (1.996)(0.0027) = 0.9871 These adjusted values will be used to represent this benchmark data, i.e., lBiasl = 0.0075 and Sp = 0.0027.

• This is a conservative treatment because the EALF trend was shown to not be statistically valid in Table C.4.

Including the trend correction in the bias term will result in a more conservative k95/95 than treating it as an increased uncertainty.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-25 Normal Probability Distribution Actual Data Probability 0.991 0.993 0.995 0.997 0.999 1.001 Data Figure C.5 Normal Probability Plot for the keff Dataset AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-26 C.7 Area of Applicability A brief description of the spectral and physical parameters characterizing the set of selected benchmark experiments is provided in Table C.6.

Table C.6 Range of Values for Key Benchmark Experiment Parameters Parameter Range of Values Heterogeneous lattices, Geometrical Shape with Square and Rectangular pitch Fuel type UO2 fuel rods Enrichment (for UO2 fuel) 2.35 to 4.74 wt% U-235 Fuel rod pitch 1.26 to 2.54 cm H/X 110 to >400 EALF 0.060 to 0.247 eV Stainless steel, borated stainless Absorbers steel, aluminum, Zircaloy-4, and Boral Water Reflectors Stainless Steel AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-27 C.8 Bias Summary and Conclusions The mixed dataset of 68 criticality safety benchmarks experiments was tested against the null hypothesis of normality and was found to be normally distributed. Thus, a parametric analysis was used to determine the bias and bias uncertainty, which resulted in a lower tolerance limit of KL = 0.99048.

A standard trending analysis was also performed using linear regression analysis, including significance testing and goodness-of-fit evaluation. Four independent variables were examined:

enrichment (wt% U-235), rod pitch, moderating ratio (H/X), and EALF (eV). The results of the trending analysis showed that the weighted trend for H/X met the criteria for statistical validity.

Although most trends for the other parameters were deemed statistically insignificant, lower tolerance bands were calculated for all variables and then overlaid on the data plots to illustrate the effect.

When the lower tolerance limit, KL = 0.99048, was compared with the lower tolerance bands of the trended data, the lower tolerance limit (KL) was not conservative for all trended parameters.

Thus, the bias term was increased as shown below.

lAdjusted biasl = 0.0075 Adjusted KL = (1 - 0.0075) - (1.996)(0.0027) = 0.9871 (The following adjusted values are referenced in Section 5.0 and applied in Section 7.8 of the report: lBiasl = 0.0075 and Sp = 0.0027).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-28 C.9 References C.1 Nuclear Energy Agency, International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03, Nuclear Energy Agency, Organization for Co-operation and Development, September 2009.

C.2 Nuclear Regulatory Commission, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001.

C.3 Rosenkrantz W.A., Introduction to Probability and Statistics for Scientists and Engineers, The McGraw-Hill Companies Inc. 1997.

C.4 Owen, D.B., Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, Sandia Corporation Monograph SRC-607, 1963.

C.5 DAgostino, R.B. and Stephans, M.A. Goodness of Fit Techniques, Statistics, Textbooks and Monographs, Volume 68, New York, NY, 1986.

C.6 Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01 Revision 0, Staff Guidance Regarding The Nuclear Criticality Safety Analysis For Spent Fuel Pools, (ADAMS # ML110620086).

C.7 NUREG/CR-6979, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data, September 2008, (ADAMS # ML082880452).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-29 Addendum to Appendix C Benchmark Extension with HTC Critical Experiments Critical experiments with Plutonium and other actinides are outside of the area of applicability for BOL reactivity equivalent evaluations such as this one. However, item IV.4.a.i of the Reference C.6 guidance document indicates that the HTC critical experiments should be considered. This section has been created to demonstrate that it is reasonable to exclude the HTC critical experiments from the SCALE 4.4a benchmarking evaluation.

Twenty-three of the twenty-six cases from the phase 3 experiments (Reference C.7) were added to the 68 cases shown in Tables C.2 and C.3 (producing a total of 91 cases). These cases were selected because they are similar to BWR spent fuel pool conditions and because they do not contain soluble boron or soluble gadolinia. The SCALE 4.4a results for these cases are shown in Table C.7. A statistical evaluation performed per Reference C.2 indicates that this expanded benchmark set is normally distributed with an average keff of 0.99765 with a pooled uncertainty of 0.002536. Therefore the Bias, the total uncertainty, and the parametric lower tolerance limit (see below) are less limiting for this expanded dataset than for the 68 case dataset (see Section C.6).

Bias = k eff - 1 = 0.99765 - 1 = -0.00235 Uncertainty = (C95/95)(Sp) = (1.942)(0.002536) = 0.00492 KL = k eff - (C95/95)(Sp) = 0. 99765 - 0. 00492 = 0.99273 This extended dataset produced statistically significant trends for H/X and EALF. Therefore, lower tolerance bands were determined for these parameters (per Reference C.2) and the resulting comparison plots are included as Figures C.6 and C.7. These trend results show that the minimum overall value remains unchanged (about 0.988 for EALF at 0.247 eV).

From this comparison the recommended HTC critical benchmark cases can be excluded without creating non-conservative results. Since these benchmark cases are outside of the area of applicability for the BOL reactivity equivalent KENO calculations, the k95/95 evaluation in the main body of this report will be based upon the 68 case dataset summarized in Section C.8.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-30 Table C.7 SCALE 4.4a Results for the HTC Critical Benchmark Experiments SCALE 4.4a Rod Benchmark Values Enrichment EALF No. Case Name Calculated Values Pitch H/X keff exp keff calc (wt% U-235) (eV)

(cm) 1 to 68 see Table C.3 69 HTC-2518 1.0000 0.0011 0.9973 0.0002 1.57 1.6 466 0.125 70 HTC-2521 1.0000 0.0011 0.9974 0.0002 1.57 1.6 466 0.131 71 HTC-2522 1.0000 0.0011 0.9975 0.0002 1.57 1.6 466 0.126 72 HTC-2523 1.0000 0.0011 0.9967 0.0002 1.57 1.6 466 0.137 73 HTC-2511 1.0000 0.0011 0.995 0.0002 1.57 1.6 466 0.131 74 HTC-2525 1.0000 0.0011 0.9955 0.0002 1.57 1.6 466 0.135 75 HTC-2526 1.0000 0.0011 0.9972 0.0003 1.57 1.6 466 0.131 76 HTC-2527 1.0000 0.0011 0.9942 0.0002 1.57 1.6 466 0.139 77 HTC-2509 1.0000 0.0008 0.999 0.0002 1.57 1.6 466 0.114 78 HTC-2531 1.0000 0.0008 0.9989 0.0002 1.57 1.6 466 0.113 79 HTC-2532 1.0000 0.0008 0.9995 0.0002 1.57 1.6 466 0.113 80 HTC-2532 1.0000 0.0008 0.999 0.0002 1.57 1.6 466 0.112 81 HTC-2533 1.0000 0.0008 0.9989 0.0002 1.57 1.6 466 0.112 82 HTC-2534 1.0000 0.0008 0.9978 0.0002 1.57 1.6 466 0.11 83 HTC-2536 1.0000 0.0008 0.9995 0.0002 1.57 1.6 466 0.107 84 HTC-2537 1.0000 0.0008 1.0003 0.0002 1.57 1.6 466 0.105 85 HTC-2538 1.0000 0.0008 1.0001 0.0002 1.57 1.6 466 0.103 86 HTC-2539 1.0000 0.0008 0.9998 0.0002 1.57 1.6 466 0.106 87 HTC-2541 1.0000 0.0008 0.9999 0.0002 1.57 1.6 466 0.108 88 HTC-2544 1.0000 0.0008 0.9985 0.0002 1.57 1.6 466 0.116 89 HTC-2547 1.0000 0.0008 0.9998 0.0002 1.57 1.6 466 0.154 90 HTC-2548 1.0000 0.0008 1.0001 0.0002 1.57 1.6 466 0.129 91 HTC-2549 1.0000 0.0008 0.9991 0.0002 1.57 1.6 466 0.117 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-31 1.010 1.005 y = 8.956E-06x + 0.9942 R2 = 0.3925 1.000 Normalized k(eff) 0.995 0.990 0.985 0.980 0 100 200 300 400 500 600 700 800 Moderating Ratio, H/X Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.6 Weighted H/X Trend (HTC Extended Benchmark)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-32 1.010 1.005 y = -0.03175x + 1.0018 R2 = 0.3348 1.000 Normalized k(eff) 0.995 0.990 0.985 0.980 0.00 0.05 0.10 0.15 0.20 0.25 0.30 EALF (eV)

Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.7 Weighted EALF Trend (HTC Extended Benchmark)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-1 Appendix D CASMO-4 Qualification for In-Rack Modeling D.1 Introduction The criticality safety analysis provided in this report is primarily a KENO V.a based analysis.

However, KENO V.a does not have depletion capability so the CASMO-4 code is used for a subset of calculations that require fuel depletion. Since CASMO-4 is a two-dimensional code, it cannot provide stand-alone benchmark results of finite criticality experiments.

CASMO-4 has demonstrated acceptable isotopic depletion and nuclear library capability for reactor core related calculations in Reference D.1. It is a multi-group, two-dimensional transport theory code which also has an in-rack geometry option where typical storage rack geometries can be modeled on an infinite lattice basis. This code is used for fuel depletion in a manner that is consistent with AREVAs NRC approved CASMO-4 / MICROBURN-B2 methodology (Reference D.1). The library files used in this evaluation are the standard CASMO-4 70 group library based on ENDFB-IV. The CASMO-4 computer code and data library are controlled by AREVA procedures and the version used in this analysis meets the requirements of Reference D.1.

Within this criticality evaluation, CASMO-4 is used to:

• perform a k ranking of fuel lattices at peak in-rack reactivity conditions (see Appendix B)

• define reference lattices that are more reactive than all past and expected future fuel lattices (the lattices of the reference bounding assembly)

• define fresh fuel reactivity equivalent lattices* for use in KENO V.a.

In support of this usage, this appendix will:

• compare CASMO-4 k results with KENO V.a to demonstrate that the fuel storage rack option in CASMO-4 also produces reasonable results

• estimate the CASMO-4 depletion uncertainty

• demonstrate that the CASMO-4 depletion uncertainty combined with a CASMO-4 calculational uncertainty is smaller than the 0.010 k uncertainty adder that is applied when the REBOL lattice is defined.

• REBOL lattices.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-2 D.2 k Comparisons These comparisons are performed in accordance with the guidance provided in References D.2.

They are performed to quantify the differences in predicted k between CASMO4 and KENO V.a (Section D.2.2.1) and to demonstrate that a k predicted by CASMO4 is nearly identical to a k predicted by KENO V.a (Section D.2.2.2).

D.2.1 Comparison Methodology The evaluation in this appendix will compare the k values produced by the CASMO-4 code to the SCALE 4.4a KENO V.a code for different geometries and U-235 enrichment levels.

The validation of the CASMO-4 code in this Appendix is performed in two steps to demonstrate its acceptability for the two different ways that CASMO-4 is used in this analysis.

• Identify the relative reactivity of a lattice with the use of the storage rack geometry option. This is addressed by determining the CASMO-4 uncertainty relative to KENO V.a by comparison of calculated k-infinities from the two codes.

• Evaluate relative changes in reactivity associated with changes in geometry and U-235 enrichment. For this evaluation, the differential k-infinities from the two codes are compared based upon the same input perturbations.

These different approaches are described in more detail in the following sections.

D.2.1.1 CASMO-4 Uncertainty for Absolute k Relative to KENO The approach taken is to perform a series of calculations with varied enrichments and geometries with the two codes and then to compare the k results. The validation guidance of NUREG/CR-6698 (Reference D.2) is followed to determine a code uncertainty for CASMO-4 relative to KENO V.a. The KENO V.a calculations are treated as the critical experiments in this comparison. GE8x8 fuel as well as top and bottom lattices from the GE9x9, GE10x10, ATRIUM-10 (10x10), and ATRIUM 10XM (10x10) product lines are used.

D.2.1.2 CASMO-4 Uncertainty for k Relative to KENO The capability of the CASMO-4 code to predict the change in reactivity associated with a perturbation of fuel parameters is demonstrated by comparison of k values obtained with KENO V.a to those obtained with CASMO-4. The approach taken is to evaluate small AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-3 perturbations in reactivity by varying the enrichment relative to a base case. The same cases used in the evaluation of the uncertainty of the absolute multiplication factor are used in this evaluation. The k values will be determined for both KENO V.a and CASMO-4 for enrichment perturbations from the reference case.

The k values are compared between the two codes and a statistical evaluation similar to that identified in Reference D.2 is used to establish an uncertainty for the determination of k values with CASMO-4 relative to k values with KENO V.a.

D.2.1.3 Experiment Descriptions As noted, KENO calculations are used as the reference experiments. The evaluations are based on the Boral storage racks in the Browns Ferry spent fuel pool. The validation is performed using GE8x8 lattices and both bottom and top lattices from the GE9x9, GE10x10, ATRIUM-10 (10x10), and ATRIUM 10XM (10x10) product lines. These lattices represent the limiting past and current fuel types for the Browns Ferry Nuclear plant. Enrichment is varied in 0.05 increments around a base of 3.35% U-235 by weight. A total of eleven (11) enrichment levels from a minimum of 3.1 wt% to a maximum of 3.6 wt% are evaluated.

The calculations are reported for 4 ºC since it represents the limiting in-rack reactivity condition for the Boral storage racks (see Table 6.1). The fuel assembly data and rack geometry are consistent with the inventory and configuration of the Browns Ferry spent fuel pool.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-4 D.2.2 Analysis of Validation Results D.2.2.1 CASMO-4 Uncertainty for Absolute k-effective Relative to KENO The calculated multiplication factors from KENO and CASMO were tabulated. The keno terms are taken from each individual KENO calculation and the casmo terms are set to the CASMO-4 convergence criterion for the individual case. (Use of the CASMO convergence is consistent with footnote 1 on page 6 of Reference D.2.) A combined uncertainty tot was determined consistent with equation 3 of Reference D.2.

+ casmo 2 2 tot

= keno The tabulated results are provided in Table D.1 for variations of geometry and enrichment. The geometry is specified by product line. A suffix of B or T is used to describe bottom or top lattice geometry, respectively. For example, A10XMT specifies ATRIUM 10XM top lattice geometry. The GE8x8 fuel contains only one geometry configuration and therefore does not have this suffix. The differences of the calculated multiplication factor values along with the components used in the statistical evaluation are provided in Table D.2.

The weighted average difference (kbar), the variance about the mean (s2), and the average total uncertainty (2) are calculated using the weighting factor 1/t2 . The square root of the pooled variance is determined per Equation 7 of Reference D.2 as shown. These results are listed below.

S s +

2 2 p

=

[

]

The simple average and standard deviation values were also tabulated by lattice geometry type:

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-5

[

]

A data normality test was completed using the Anderson-Darling test (see section 9.5.4.1 of Reference D.3). (The Anderson-Darling test is described in Section C.5 of Appendix C). The AD test statistic was calculated to be 0.8143 and the criterion is 0.746 *. Since the AD test statistic is greater than the test criterion one can conclude that the data is not from a normal distribution.

A distribution free one sided tolerance limit evaluation was also performed for this data set of 99 values. This was performed for both the upper and lower bounds. This evaluation indicated that on a 95/95 basis the more limiting k difference boundary is [ ]. For the weighted mean difference of [ ] and the limiting boundary value (above), the limiting effective uncertainty term is [ ].

Area of Applicability The fuel and rack geometries as well as representative fuel enrichments were selected to be consistent with the Browns Ferry GE High density Boral storage racks. It is recognized that

• In Appendix C the AD test statistic was adjusted for the number of data points and compared to the criteria of 0.752. In this appendix, the criterion was adjusted for the number of data points.

As indicated by equation 20 of Reference D.2, the uncertainty component is effectively the difference between the limiting boundary and the mean value.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-6 spent fuel pool storage tube modeling simplifications are included in the CASMO model relative to the more explicit model used with KENO, see Section 6.1. This difference in the modeling technique is included in this comparison. The REBOL lattice enrichment and geometries used in the k95/95 determination for the Browns Ferry Spent Fuel Pool are within the area of applicability of this comparison.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-7 Table D.1 CASMO-4 and KENO V.a Validation Case Information

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-8 Table D.1 CASMO4 and KENO Validation Case Information (Continued)

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-9 Table D.1 CASMO4 and KENO Validation Case Information (Continued)

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-10 Table D.2 CASMO - KENO Difference and Statistical Parameters

[

]

• k is kCASMO - kKENO.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-11 Table D.2 CASMO - KENO Difference and Statistical Parameters (Continued)

[

]

• k is kCASMO - kKENO.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-12 Table D.2 CASMO - KENO Difference and Statistical Parameters (Continued)

[

]

• k is kCASMO - kKENO.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-13

[

]

Figure D.1 Normality Plot for CASMO-KENO k-infinity Comparison AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-14 D.2.2.2 CASMO-4 Uncertainty for k-effective The actual KENO and CASMO calculations used in this k evaluation are those used in Section D.2.2.1. In this evaluation, the relative reactivity change is evaluated by taking the delta with respect to the reference case. A difference is then determined between the k values obtained with KENO and the k values obtained with CASMO-4 for the same perturbation.

The Anderson-Darling goodness of fit for normality test was also completed with the AD test statistic calculated to be 0.7425 with the criterion of 0.7456. Based on these results and the comparison in Figure D.2, it is determined that the data is normally distributed.

The magnitude of the average difference between the k values was [ ] with a standard deviation of [ ]. For the data sample of 50 the single sided tolerance factor is 2.065 from Table 2.1 of Reference D.2. This is conservatively applied for 90 data samples. Therefore the 95/95 bias uncertainty is: [ ] when rounded to four decimal places.

Area of Applicability The fuel and rack geometries as well as representative fuel enrichments were selected to be consistent with the Browns Ferry GE High density Boral storage racks. It is recognized that spent fuel pool storage tube modeling simplifications are included in the CASMO model relative to the more explicit model used with KENO, see Section 6.1. This difference in the modeling technique is included in this comparison. The REBOL lattice enrichment and geometries used in the k95/95 determination for the Browns Ferry Spent Fuel Pool are within the area of applicability of this comparison.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-15 Table D.3 CASMO versus KENO Relative Reactivity Differences at 4ºC

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-16 Table D.3 CASMO versus KENO Relative Reactivity Differences at 4ºC (Continued)

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-17 Table D.3 CASMO versus KENO Relative Reactivity Differences at 4ºC (Continued)

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-18

[

]

Figure D.2 Normality Plot for kCASMO - kKENO k-infinity Comparison AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-19 D.3 Depletion Uncertainty Estimates Depletion uncertainty estimates from EMF-2158(P) (Reference D.1) and from the interim staff guidance document (Reference D.5) will be described in this section.

D.3.1 EMF-2158 Based Depletion Uncertainty The CASMO-4 depletion uncertainty is derived from the AREVA licensing topical report based on the extensive benchmarking that is documented within Reference D.1. Comparisons against critical experiments were performed by Studsvik with results reported in Table 2.1 of AREVAs CASMO4/MICROBURNB2 licensing topical report (Reference D.1). In addition, the beginning of cycle cold critical calculations reported in Table 2.2 of this same licensing topical report also provide comparisons to critical data. Results of these comparisons indicate that CASMO-4 results will have a standard deviation of [ ] k (Table 2.1 of Reference D.1) without depletion and a standard deviation of [ ] k (Table 2.2 of Reference D.1) when the majority of assemblies have been depleted*.

D.3.2 ISG Based Depletion Uncertainty Five percent of the reactivity difference from BOL (without gadolinia) to peak reactivity is used to estimate the isotopic uncertainty associated with depletion to peak reactivity, (i.e., the uncertainty in the uranium depletion, fission product production, and actinide production). The approach presented here is a conservative application of the 5% reactivity decrement approach originally suggested in Section 5.A.5.d of the Kopp memo (Reference D.4) and currently addressed in DSS-ISG-2010-01 (Reference D.5).

The reference bounding and limiting lattices used in this comparison are identified in Table B.1.

All lattices are depleted in-core and then evaluated at the limiting moderater temperature (4 ºC) in the fuel storage rack configuration. Figure D.3 illustrates the two reactivity decrement values used.

• The uncertainty of cold critical benchmarks effectively includes a depletion uncertainty since the majority of the bundles in the core have some depletion. It is noted, that an in-sequence critical has significant similarities to an in-rack calculation since the majority of the control blades remain inserted effectively surrounding the majority of the fuel with a strong neutron absorber on two sides.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-20 A BOL no gad solution for each lattice was completed by removing the gadolinium and maintaining the same uranium number density in the lattice.* The depletion reactivity decrement is determined by subtracting the peak in-rack k from the BOL no gad in-rack k. A second reactivity decrement representing the uncertainty in gadolinia content was also determined by subtracting the peak in-rack k from a value similar to the gadolinia free k at the peak reactivity exposure .

Based on the calculation process illustrated in Figure D.3, five percent of the burn-up reactivity decrement (kbu=0.05*k) and five percent of the residual gadolinia reactivity change (kgd=0.05*kg) are tabulated in Table D.4 for the limiting lattices. This assessment produces a maximum depletion uncertainty of 0.0055 k for the reference bounding lattices.

It is noted that this process will produce a larger penalty as the gadolinia content increases (either the number of rods or the concentration). However, increasing the gadolinia content within a given lattice will substantially decrease the peak in-rack k of the lattice as shown in Figure D.4.

• This is accomplished by setting the gadolinia number densities to zero with the CASMO CNU input.

The peak k-infinity values with no gadolinia use an in-core depletion with gadolinia to the maximum reactivity exposure, all gadolinia is then removed and an in-rack calculation is performed.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-21 Table D.4 Depletion Uncertainty Values for Limiting Lattices Peak BOL Peak kbu kgd kbu +

k k nogad k nogad * (0.05*k) (0.05*kg) kgd Top Zone Limiting Legacy Lattice 0.8619 0.9491 0.8777 0.0044 0.0008 0.0052 Bounding Lattice 0.8797 0.9638 0.8931 0.0042 0.0007 0.0049 Bottom Zone Limiting Legacy Lattice 0.8227 0.9385 0.8424 0.0058 0.0010 0.0068 Bounding Lattice 0.8790 0.9733 0.8954 0.0047 0.0008 0.0055

• The Peak k-infinity values with no gadolinia assume in-core depletion with gadolinia to the maximum reactivity exposure, all gadolinia is then removed and an in-rack calculation is performed.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-22 1.05 1.00 k-inf, In-Rack with Residual Gd k-inf, In-Rack without Gd 0.95 k-inf, In-Rack without Gd - BOL k-infinity (in-rack)

Burnup (k) decrement 0.90 Residual gadolina (kg) 0.85 0.80 0 5 10 15 20 25 30 Burnup (GWd/MT)

Figure D.3 Representation of the ISG Depletion Uncertainty Assessment AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-23

[

]

Figure D.4 Gadolinia Concentration Sensitivity AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-24 D.4 Conclusions and Overall Uncertainty The evaluation of GE8x8 fuel lattices as well as top and bottom lattices from the GE9x9, GE10x10, ATRIUM-10 (10x10), and ATRIUM 10XM (10x10) product lines demonstrated that the CASMO4 fuel storage rack calculations will produce reasonable results for these types of geometries. In addition, it has been demonstrated that reasonable results can be obtained for U-235 enrichment levels between 3.1 and 3.6 wt% U-235. These comparisons also indicate that [

].

When applied on a differential basis a k predicted by CASMO-4 agrees with the KENO V.a based k with a standard deviation of [ ] k, (see Section D.2.2.2). This can be combined with uncertainty estimates from EMF-2158(P) (Section D.3.1) or the estimated depletion uncertainty determined with the method from the interim staff guidance document (Section D.3.2) to produce a maximum combined value. A 95/95 uncertainty result is obtained by multiplying these uncertainty values by an appropriate multiplier. Since these values are independent they will be combined using the square root of the sum of the squares as shown below. This process results in a maximum combined uncertainty of [ ]. The 0.010 k adder used when defining the REBOL lattices conservatively bounds this CASMO-4 uncertainty value.

95/95 Combined Uncertainty Value 95/95 Uncertainty Multiplier Uncertainty Calculational [ ] 2.065 [ ]

(k based)

EMF-2158 Depletion [ ] 2.0 [ ] [ ]

Calculational [ ] 2.065 [ ]

(k based)

ISG Depletion --- --- 0.0055* [ ]

• This is not necessarily a 95/95 value; however, it is acceptable per Section IV.2.a of Reference D.5.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-25 D.5 References D.1 EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.

D.2 NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, USNRC, January 2001.

D.3 MIL-HDBK-5J, Metallic Materials and Elements for Aerospace Vehicle Structures, Department of Defense Handbook, January 2003.

D.4 Memorandum L. Kopp to T. Collins, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, NRC, August 19, 1998. (NRC -ADAMS Accession Number ML072710248)

D.5 Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01 Revision 0, Staff Guidance Regarding The Nuclear Criticality Safety Analysis For Spent Fuel Pools, (NRC - ADAMS Accession Number ML110620086).

AREVA Inc.

ENCLOSURE 3 AREVA Affidavit

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) SS.

CITY OF LYNCHBURG )

1. My name is Morris Byram. I am Manager, Product Licensing, for AREVA Inc.

(AREVA) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary. I am familiar with the policies established by AREVA to ensure the proper application of these criteria.

3. I am familiar with the AREVA information contained in the AREVA document ANP-3160(P), Revision 1, "Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM' 10XM Fuel," and referred to herein as "Document."

Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA Inc. for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(d) above.

7. In accordance with AREVA's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this *~

dayof ~ ,2015.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg.# 7079129 SHERRYL. MCFADEN Notery PUDllC Comm~nwellth of Vlrglnl1

' 7071129 My Comm1111on Expfret Ocl 31, 2018

ENCLOSURE 4 Suppressed Fuel Assembly Impact on Criticality Safety Analysis

Suppressed Fuel Assembly Impact on Criticality Safety Analysis During an NRC public meeting on November 10, 2015, with TVA representatives, the NRC asked about the applicability of the Browns Ferry Nuclear Plant (BFN) Spent Fuel Pool (SFP)

Criticality Safety Analysis (CSA) for a fuel assembly that had been suppressed for an entire 24 month cycle.

Response

The impact of operation with a suppression blade is addressed as part of the BFN ATRIUM 10XM CSA report ANP-3160(P), Revision 1, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM' 10XM Fuel, AREVA Inc., December 2015 (Included in Enclosure 1 of this submittal). Specifically, part of Assumption 7 of ANP-3160(P), Section 6.5, General CASMO-4 Modeling Assumptions, addresses power suppression from the perspective of the following: the effects are limited to only a small population of fuel bundles (i.e., the four fuel bundles in each affected control cell),

and the effect on lattice reactivity by the power gradient and the reduced power density. The reactivity effects are supported with a sensitivity analysis with the results provided in Table 6.6 of ANP-3160(P). The conclusion is reached that the uncontrolled depletion results are bounding for the ATRIUM 10XM reference bounding lattices.

E41

ENCLOSURE 5 Boral Neutron Absorber Aging Management Program

Boral Neutron Absorber Aging Management Program During an NRC public meeting on November 10, 2015, with TVA representatives, the NRC asked if Browns Ferry Nuclear Plant (BFN) had an aging management program for the Boral neutron absorbers used in the Spent Fuel Pool (SFP) storage racks and credited in the BFN SFP Criticality Safety Analysis (CSA).

Response

BFN does have an aging management program for Boral neutron absorbers used in the SFP storage racks. The topic of an aging management program for the Boral neutron absorbers in the SFP storage racks was addressed during the NRC review of the BFN License Renewal Application (LRA). The NRC review and acceptance of the BFN aging management program for Boral neutron absorbers used in the SFP storage racks is documented in NUREG-1843, Safety Evaluation Report Related to the License Renewal of the Browns Ferry Nuclear Plant, Units 1, 2, and 3, dated April 30, 2006 (ADAMS No. ML061030027). Applicable portions of NUREG-1843 describing the NRC review and acceptance of the BFN aging management program for SFP storage rack Boral neutron absorbers are as follows.

NUREG-1843, Section 2.4.2.1.2 (page 2-161) contains the following NRC evaluation of Request for Additional Information (RAI) 2.4-4.

In RAI 2.4-4, dated December 20, 2004, the staff stated that LRA Table 2.4.2.1 presents a list of component types that are part of the reactor building, the auxiliary and emergency systems of the NSSS, the biological shield, the spent fuel pool, the steam dryer/moisture separator storage pool, the reactor cavity reactor auxiliary equipment, the steel superstructure with metal siding and the built-up roof. Therefore, the staff requested the applicant to provide a description of the "Neutron-Absorbing Sheets" used for the spent fuel storage racks and confirm that they are part of the spent fuel storage racks listed in LRA Table 2.4.2.1.

In its response, by letter dated January 24, 2005, the applicant stated:

NUREG 1801,Section VII.A2.1-b, identifies "Spent Fuel Storage Racks -

neutron absorbing sheets" as a component type. In BFN LRA Section 2.3.3.27 Fuel Handling and Storage System (079)," it states that the spent fuel pool components are evaluated as structural components in Section 2.4.2.1 "Reactor Building Structure." BFN LRA Table 2.4.2.1 "Reactor Building Structure" identifies "Spent Fuel Storage Racks (includes new fuel storage racks)" as a component requiring aging management. The "Neutron Absorbing Sheet" is a component of the BFN spent fuel storage rack container tube wall and is comprised of Boral sandwiched within the stainless steel wall of each container tube.

The staff found the above response acceptable. Therefore, the staff's concern described in RAI 2.4-4 is resolved.

NUREG-1843, Table 3.3-1, (page 3-205) contains a row which identifies the Chemistry Control Program as the Aging Management Program (AMP) for the Boral neutron absorbing sheets.

E51

Boral Neutron Absorber Aging Management Program NUREG-1843, Section 3.3.2.2.10, (page 3-217) provides the following discussion.

3.3.2.2.10 Reduction of Neutron-Absorbing Capacity and Loss of Material due to General Corrosion The staff reviewed LRA Section 3.3.2.2.10 against the criteria in SRP-LR 3.3.2.2.10.

In LRA Section 3.3.2.2.10, the applicant addressed the further evaluation of programs to manage reduction of neutron-absorbing capacity and loss of material due to general corrosion, which could occur in the neutron absorbing sheets of the spent fuel storage rack in the spent fuel storage.

SRP-LR Section 3.3.2.2.10 states that reduction of neutron-absorbing capacity and loss of material due to general corrosion could occur in the neutron-absorbing sheets of the spent fuel storage rack in the spent fuel storage. The [Generic Aging Lessons Learned Report (GALL)] Report recommends further evaluation to ensure that these aging effects are adequately managed.

The applicant stated that boral is used as a neutron absorbing material in the spent fuel pools. Reduction of neutron absorbing capacity and loss of material due to general corrosion could occur in the boral neutron absorbing material in spent fuel storage racks.

The Chemistry Control Program manages general corrosion. An inspection of boral coupon test specimens was performed that confirmed no significant aging degradation had occurred and the neutron absorbing capability of the boral had not been reduced.

Reduction of neutron absorbing capacity and loss of material due to general corrosion will be managed by the Chemistry Control Program.

The staff reviewed the Chemistry Control Program and found that the program will adequately manage the effects of aging so that the intended functions will be maintained.

NUREG-1843, Section 3.5.2.1, (page 3-288) provides the following discussion regarding the Aging Management Review (AMR) for Boral.

In reference to LRA Table 3.5.2.2, the staff also requested the applicant to describe the AMR for Boral and to clarify whether stainless steel components are used to support the Boral. If the AMR supports the conclusion that Boral does not require aging management, but the stainless steel supports do, then the Chemistry Control Program would be an acceptable AMP for this item. If not, the applicant was requested to provide the technical basis for crediting the Chemistry Control Program as the appropriate AMP for Boral.

By letter dated October 8, 2004, the applicant submitted its formal response to the staff, stating that the Boral core is made up of a central segment of a dispersion of boron carbide in aluminum. This central segment is clad on both sides with aluminum to form a plate. The Boral plates are sandwiched between two stainless steel plates which are closure-welded form the container. Vent holes have been added to prevent the buildup of hydrogen gas between the stainless steel containers. If the stainless steel containers remain intact, the Boral core will be unaffected and will retain its neutron-absorbing capacity. The Chemistry Control Program will manage aging of the stainless steel E52

Boral Neutron Absorber Aging Management Program containers. With these clarifications, the staff concluded that this item is consistent with the GALL Report.

NUREG-1843, Section 3.5.2.3.2, (page 3-323) provides the following discussion regarding RAI 3.5-14.

In RAI 3.5-14, dated December 10, 2004, the staff stated that, with respect to the neutron-absorbing sheets in spent fuel storage racks, as described in LRA Section 3.3.2.2, the applicant stated that the Chemistry Control Program manages general corrosion and that an inspection of Boral coupon test specimens was performed at BFN that confirmed that no significant aging degradation had occurred and that the neutron-absorbing capacity of the Boral had not been reduced. Since it is implied that some Boral aging degradations had occurred at the time of inspection of the test specimens, the staff requested the applicant to discuss the basis for the above assertion that the neutron-absorbing capacity of the Boral will be maintained at an adequate level during the extended period of plant operation.

In its response, by letter dated January 31, 2005, the applicant stated:

A total of 16 boral coupons were placed in the Unit 3 spent fuel storage pool (SFSP) in October 1983. The coupons supplied by the rack manufacturer are of the same metallurgical condition as the high density fuel storage racks (HDFSR) in thickness, chemistry, finish, and temper. For the first six years of the planned fifteen year surveillance program, examination was to have taken place at two-year intervals. Accordingly, two coupons were removed in October 1985. Blisters were found upon examination, and because of this unexpected anomaly, three additional coupons were analyzed not finding any blisters. As a result of blisters found on the coupons removed in 1985, the surveillance program has been expanded to include monitoring the formation and behavior of these blisters.

These boral coupons are periodically removed from the fuel pool for testing and are evaluated for corrosion or other degradation of the neutron absorber plates by comparing various physical characteristics of the test coupons to baseline measurements taken when the coupons were installed. Also, a metallurgical engineer examines the coupons for general corrosion, local pitting, and bonding.

No further blisters, corrosion, or degradation has been identified in coupons evaluated through 2003.

The above response states that these Boral coupons are periodically removed from the fuel pool for testing and are evaluated for corrosion or other degradation of the neutron absorber plates by comparing various physical characteristics of the test coupons to baseline measurements taken when the coupons were installed. The response also implies that a metallurgical engineer periodically examines the coupons for general corrosion, local pitting, and bonding. Also, no further blisters, corrosion, or degradation have been identified in coupons evaluated through 2003; however, it was not clear to the staff whether these periodic inspections are ongoing activities that are an extension of the 1983 Boral Coupon Inspection Program covering Boral coupon test specimens or a separate AMP in addition to the Chemistry Control Program mentioned above. The applicant was requested to clarify the key parameters of this periodic inspection program or activity including the objective, scope, frequency, and inspection approach of the program.

E53

Boral Neutron Absorber Aging Management Program In its response, by letter May 24, 2005, the applicant stated that:

The Boral coupon inspection program was initiated in 1983 to implement the inspection and testing requirements of UFSAR Section 10.3.6; this checks the long-term behavior of the material of the high density spent fuel racks. The inspection is performed per BFN Technical Instruction (TI) TI-116, "High Density Fuel Storage System Surveillance Program." When the TI is performed, Boral coupons are removed from the spent fuel storage pool and examined by the Metallurgical Engineer in their original condition to determine if sampling of surface corrosion products is appropriate. Thickness measurements are obtained of each coupon and documented in accordance with the TI. If degradation is such that further investigation is warranted, a minimum of one coupon is selected to be unsheathed or opened. Prior to the unsheathing process, a dye penetrant test for indications on the outer surfaces of the coupon will be performed and is examined by the Metallurgical Engineer. The Metallurgical Engineer decides if further unsheathing of the coupons is required. The visual examination by the Metallurgical Engineer is documented on the appropriate forms of the TI. The current frequency for performing this TI is two years. The surveillance frequency is re-evaluated each time the surveillance is performed and can be changed based on the trend of the historical data results. The inspection of the Boral coupons will continue until such time as the trend of the historical data results collected provides a basis to discontinue the inspections.

Based on its review, the staff found the applicant's response to RAI 3.5-14 acceptable.

Therefore, the staffs concern described in RAI 3.5-14 is resolved.

E54

Proprietary Information Withhold from Public Disclosure Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 1 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-15-249 December 15, 2015 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296

Subject:

Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) -

Supplement 1, Spent Fuel Pool Criticality Safety Analysis Information

Reference:

Letter from TVA to NRC, CNL-15-169, "Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments -

Extended Power Uprate (EPU)," dated September 21, 2015 By the reference letter dated September 21, 2015, Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) for the Extended Power Uprate (EPU) of Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3. The proposed LAR modifies the renewed operating licenses to increase the maximum authorized core thermal power level from the current licensed thermal power of 3458 megawatts to 3952 megawatts.

During a November 10, 2015, Nuclear Regulatory Commission (NRC) public meeting with TVA representatives regarding the EPU LAR, the NRC requested the Spent Fuel Pool (SFP) Criticality Safety Analysis (CSA) be submitted for review. Enclosure 1 of this letter provides the BFN Units 1, 2, and 3 SFP CSA for ATRIUM 10XM Fuel.

U.S. Nuclear Regulatory Commission CNL-15-249 Page 2 December 15, 2015 AREVA considers portions of the information provided in Enclosure 1 of this letter to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390, public inspections, exemptions, requests for withholding . An affidavit for withholding information, executed by AREVA, is provided in Enclosure 3. A non-proprietary version of the document is provided in Enclosure 2. Therefore, on behalf of AREVA, TVA requests that Enclosure 1 be withheld from public disclosure in accordance with the AREVA affidavit and the provisions of 10 CFR 2.390.

During the November 10, 2015, NRC public meeting with TVA representatives regarding the EPU LAR, the NRC also requested information be submitted regarding the impact of fuel assembly operation with suppression blade on the SFP CSA and the aging management program for the Boral neutron absorbers used in the SFP storage racks. This information is provided in Enclosures 4 and 5, respectively.

TVA has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the NRC in the reference letter. The supplemental information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration . In addition, the supplemental information in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed license amendment. Additionally, in accordance with 10 CFR 50.91 (b)(1 ), TVA is sending a copy of this letter and the non-proprietary enclosures to the Alabama State Department of Public Health.

There are no new regulatory commitments associated with this submittal. If there are any questions or if additional information is needed, please contact Mr. Edward D.

Schrull at (423) 751-3850.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 15th day of December 2015.

ully,

.~

Shea resident, Nuclear Licensing Enclosures cc: See Page 2

U.S. Nuclear Regulatory Commission CNL-15-249 Page 3 December 15, 2015

Enclosures:

1. ANP-3160(P) Revision 1, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 10XM Fuel (Proprietary)

2. ANP-3160(NP) Revision 1, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 10XM Fuel (Non-Proprietary)

3. AREVA Affidavit

4. Suppressed Fuel Assembly Impact on Criticality Safety Analysis

5. Boral Neutron Absorber Aging Management Program cc:

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant State Health Officer, Alabama Department of Public Health (w/o Enclosure 1)

ENCLOSURE 2 ANP-3160(NP) Revision 1, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 10XM Fuel (Non-Proprietary)

ANP-3160(NP)

Revision 1 Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM' 10XM Fuel December 2015 Proprietary

AREVA Inc.

ANP-3160(NP)

Revision 1 Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM' 10XM Fuel sja

AREVA Inc.

ANP-3160(NP)

Revision 1 Copyright © 2015 AREVA Inc.

All Rights Reserved

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page i Nature of Changes Item Page Description and Justification

1. 3-12 Table 3.1, IV.5.b corrected typo.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page ii Contents 1.0 Introduction ..................................................................................................................1-1 2.0 Summary and Conclusions ...........................................................................................2-1 3.0 Regulatory Criticality Safety Criteria and Guidance ......................................................3-1 4.0 Fuel and Storage Array Description ..............................................................................4-1 4.1 Fuel Assembly Design ......................................................................................4-1 4.2 Fuel Storage Racks ..........................................................................................4-1 5.0 Calculation Methodology ..............................................................................................5-1 5.1 Area of Applicability ..........................................................................................5-2 6.0 Modeling Options and Assumptions .............................................................................6-1 6.1 Geometric Modeling of the High Density Boral Rack .........................................6-1 6.1.1 Single Cell Model Description .............................................................6-1 6.1.2 Explicit Storage Cell Model Description ...............................................6-2 6.1.3 Explicit Rack Model Description ..........................................................6-2 6.1.4 Reactivity Comparison of the Boral Rack Models ................................6-2 6.2 Fuel Assembly Modeling ...................................................................................6-3 6.3 Co-Resident Fuel Racks ...................................................................................6-3 6.4 BLEU versus Commercial Grade Uranium ........................................................6-4 6.5 General CASMO-4 Modeling Assumptions .......................................................6-4 7.0 Criticality Safety Analysis..............................................................................................7-1 7.1 Definition of the Reference Bounding and REBOL Lattices ...............................7-2 7.2 Storage Array Reactivity ...................................................................................7-3 7.3 Arrays of Mixed BWR Fuel Types .....................................................................7-3 7.4 Other Conditions ...............................................................................................7-4 7.4.1 Assembly Rotation ..............................................................................7-4 7.4.2 Assembly Lean ...................................................................................7-4 7.4.3 Blister Formation .................................................................................7-4 7.5 Normal Fuel Handling .......................................................................................7-5 7.6 Accident Conditions ..........................................................................................7-6 7.7 Manufacturing and Other Uncertainties .............................................................7-8 7.8 Determination of Maximum Rack Assembly k-eff (k95/95) ...................................7-9 8.0 References ...................................................................................................................8-1 Appendix A Sample CASMO-4 Input ................................................................................ A-1 Appendix B Reactivity Comparison for Assemblies Used at Browns Ferry........................ B-1 Appendix C KENO V.a Bias and Bias Uncertainty Evaluation ........................................... C-1 Appendix D CASMO-4 Qualification for In-Rack Modeling ................................................ D-1 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page iii Tables 2.1 Criticality Safety Limitations for ATRIUM 10XM Fuel Assemblies Stored in the Browns Ferry Plant Spent Fuel Storage Pools ........................................................2-4 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 ...........................................................................................................................3-5 4.1 ATRIUM 10XM Fuel Assembly Parameters ..................................................................4-3 4.2 Fuel Storage Rack Parameters.....................................................................................4-4 6.1 Comparison of Modeling Options for the Boral Rack ....................................................6-8 6.2 Impact of Channel Thickness on In-Rack Reactivity .....................................................6-9 6.3 Co-Resident Storage Rack Comparison .......................................................................6-9 6.4 In-Rack k Sensitivity to In-core Depletion Fuel Temperature .....................................6-10 6.5 In-Rack k Sensitivity to In-core Depletion Power Density ..........................................6-11 6.6 In-Rack k Sensitivity to In-Core Controlled Depletion ................................................6-12 7.1 Summary of CASMO-4 Maximum In-Rack Reactivity Results ....................................7-11 7.2 Summary of KENO V.a Maximum In-Rack Reactivity Results ....................................7-12 7.3 Manufacturing Reactivity Uncertainties .......................................................................7-13 Figures 2.1 Overview of the Browns Ferry SFP Criticality Safety Analysis ......................................2-6 2.2 ATRIUM 10XM Reference Bounding Assembly ............................................................2-7 4.1 Representative ATRIUM 10XM Fuel Assembly ............................................................4-5 4.2 Browns Ferry Spent Fuel Pool Layout ..........................................................................4-6 4.3 Schematic Representation of a Section of High Density Storage Rack .........................4-7 4.4 High Density Boral Storage Rack Geometry .................................................................4-8 6.1 Single Cell Model for the High Density Boral Rack .....................................................6-13 6.2 Explicit Geometry Model for High Density Boral Rack ................................................6-14 6.3 Schematic of Rack to Rack Interfaces ........................................................................6-15 6.4 BLEU versus Commercial Grade Uranium Reactivity Comparison .............................6-16 6.5 Impact of Void History Depletion on In-Rack k-infinity.................................................6-17 7.1 Evaluated Assembly Rotation Cases ..........................................................................7-14 7.2 Limiting Accident (Missing Boral Plate) .......................................................................7-15 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page iv Nomenclature AEC Atomic Energy Commission BAF bottom of active fuel BLEU blended low enriched uranium BOL beginning of life BORAL neutron absorber composed of boron dispersed within aluminum BWR boiling-water reactor CGU commercial grade uranium EALF the energy of the average lethargy causing fission FPM fuel preparation machine GDC general design criteria GWd energy unit, giga-watt-day H/X moderating ratio, atomic ratio of hydrogen (H) to fissile isotopes (X)

ISG interim staff guidance document (Reference 7) k-eff effective neutron multiplication factor (aka k-effective) k infinite lattice neutron multiplication factor (aka k-infinity)

LUA lead use assembly PLR part-length fuel rod NCS nuclear criticality safety NRC Nuclear Regulatory Commission, U.S. (also USNRC)

RAI request for additional information REBOL reactivity-equivalent at beginning of life (fresh fuel, no Gd2O3)

SFP spent fuel pool TAF top of active fuel

%TD percent of theoretical density

[ ] Square brackets enclose information that is proprietary to AREVA.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 1-1 1.0 Introduction This report presents the results of a criticality safety analysis performed for the Browns Ferry Nuclear Plant Units 1, 2, and 3 spent fuel storage pools. Each spent fuel pool has the same configuration including rack design and number of storage modules. This analysis is performed on a bounding basis and is applicable to all three spent fuel storage pools. The previous Nuclear Regulatory Commission (NRC) approved criticality safety evaluation is identified as Reference 1.

In this report, a reference bounding assembly has been defined to bound the reactivity of all past and current fuel assembly types delivered to the Browns Ferry Nuclear Plant. This reference bounding assembly is based on an AREVA Inc. (AREVA) ATRIUM'* 10XM fuel assembly. This analysis demonstrates that with the reference bounding assembly the pool k-eff remains below the 0.95 k-effective acceptance criterion established by the NRC.

• ATRIUM is a trademark of AREVA NP.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-1 2.0 Summary and Conclusions Criticality safety calculations have been performed and are documented herein for the Browns Ferry Nuclear Plant spent fuel storage pools. Figure 2.1 provides an overview of the various steps involved in this criticality safety analysis. The analysis flow in this figure begins at the bottom with the evaluation of the existing fuel inventory and ends at the top with the calculation of an array keff that meets the regulatory acceptance criterion of 0.95.

This criticality safety analysis is based on the use of a reference fuel assembly design that is bounding for (i.e., more reactive than) all fuel designs previously used or planned to be used at the Browns Ferry Nuclear Plant. The KENO V.a code was used for all calculations that do not require fuel depletion. The CASMO-4 code is used to compare lattice k values at peak reactivity conditions. The results of these comparisons are used to define the reference bounding lattices and the reactivity-equivalent at beginning of life (REBOL) lattices that are used in KENO V.a.

CASMO-4 is also used in defining a portion of the gadolinia manufacturing uncertainty.

Benchmarking against criticality experiments is included for the KENO V.a code and justification for the use of the CASMO4 code is also provided. More detail on methodology and code benchmark / justification is provided in Chapter 5 and Appendices C and D.

The calculations documented herein demonstrate that the ATRIUM 10XM reference bounding assembly design has been selected to be more reactive in an in-rack configuration than any of the current or past fuel assembly designs used in the Browns Ferry reactors. These comparisons are based upon actual GE 7x7, GE 8x8, GE 9x9, GE 10x10, and AREVA 10x10 (ATRIUM-10) lattice geometries and enrichments as detailed in Appendix B.* This criticality safety analysis shows that future ATRIUM 10XM assemblies meeting the storage requirements established in Table 2.1 can be safely stored with these previously manufactured assemblies.

The reference bounding assembly is defined with two U-235 enrichment / gadolinia concentration zones separated by the ATRIUM 10XM geometry transition at [ ] inches. The bottom enrichment and gadolinia zone is defined to extend up to this transition boundary and contains

[ ] fuel rods. The top enrichment / gadolinia zone extends from this geometric transition boundary to the top of the fuel assembly and contains [ ] fuel rods. These axial zones are

• Various LUAs were also evaluated.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-2 illustrated in Figure 2.2. Two REBOL lattices have been defined to represent the lattices of the reference bounding assembly in KENO calculations. The neutron multiplication factors of the REBOL lattices have been increased by greater than or equal to 0.010 k to address all uncertainties associated with defining these reactivity equivalent lattices.

This analysis includes manufacturing uncertainties for the ATRIUM 10XM fuel design and the fuel storage racks. In addition to the manufacturing uncertainties; code modeling uncertainties, reactivity increases due to accident or other conditions, and a one-sided tolerance multiplier are used to determine the 95/95 upper limit k-eff. The conditions and uncertainties assumed in this analysis are described in the various sections of Chapter 7.

This analysis demonstrates that the reference ATRIUM 10XM fuel assembly does not exceed an array k-eff of 0.95 in the Browns Ferry spent fuel storage pools. As defined in Table 2.1, ATRIUM 10XM fuel that contains equivalent or less enrichment and equivalent or higher Gd2O3 concentrations in the fuel zones depicted in Figure 2.2 can be safely stored in the Browns Ferry spent fuel storage pools. In addition, ATRIUM 10XM fuel that contains more enrichment and/or lower Gd2O3 concentrations than the reference assembly design can be safely stored provided each zone of the assembly is less reactive than the corresponding zone of the reference bounding assembly design (i.e., less than 0.8825 in-rack k-infinity for both zones in accordance with Table 2.1). This can be established using the storage rack model of the CASMO-4 lattice physics code as described in Appendix A.

This analysis supports the storage of channeled and unchanneled fuel assemblies including assemblies with the AREVA advanced fuel channel. Additionally, there is no limitation for bundle orientation or position in the storage cell since these are accounted for in the analysis.

To assure that the actual reactivity will always be less than the calculated reactivity, the following conservatisms have been included:

• The results are based on a moderator temperature of 4 °C (39.2 °F), which gives the highest reactivity for the fuel storage pool.

• Fuel assemblies are assumed to contain the high reactivity reference bounding lattices for the entire length of the assembly (i.e., natural uranium blankets are not modeled).

• Each lattice in each fuel assembly in the storage rack is assumed to be at its lifetime maximum reactivity level. There is no assumption of a specific burnup profile for the discharged assemblies. In other words, this is a peak reactivity analysis that does AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-3 not take credit for lower reactivity conditions associated with burnup past the maximum reactivity.

• The minimum Boron-10 areal density is used when modeling the Boral.

• The most limiting orientation or position of each assembly in its rack cell is accounted for in the analysis.

• Neutron absorption in fuel assembly structural components (i.e., spacers, tie plates, etc.) is neglected.

• The maximum reactivity value includes all significant manufacturing and calculational uncertainties.

• The 0.010 k uncertainty value applied when the REBOL lattice is defined is treated as a bias - introducing significantly more conservatism than if it had been treated as an uncertainty.*

• The fuel array is modeled as being infinite in all dimensions.

• An adder has been included to account for Boral blistering.

• The bias from the KENO V.a benchmark (Appendix C) has been increased to also bound trending conditions that were shown to be statistically insignificant.

This analysis demonstrates that all fuel assemblies previously delivered to the Browns Ferry Nuclear Plant can be safely stored in the spent fuel storage pools. Future ATRIUM 10XM fuel designs that meet the design requirements specified in Table 2.1 or that can be shown to be less reactive (on a lattice basis) than the reference bounding assembly can be safely stored in the Browns Ferry spent fuel pools. The k-eff determined herein for the reference assembly, including all uncertainties, biases, manufacturing tolerances and worst accident or other loading conditions is 0.928 (as detailed in Section 7.8 and Figure 2.1).

• As applied in this evaluation a k95/95 value of 0.928 is produced. If the 0.010 k uncertainty were not applied to the REBOL lattices and then treated as an additional uncertainty term in Section 7.8, the k95/95 value would decrease to 0.922.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-4 Table 2.1 Criticality Safety Limitations for ATRIUM 10XM Fuel Assemblies Stored in the Browns Ferry Plant Spent Fuel Storage Pools ATRIUM 10XM Fuel Configuration The ATRIUM 10XM fuel configuration is provided in Table 4.1.

Fuel Channels Fuel may be stored with or without fuel channels.

Fuel Design Limitations for Enriched Lattices*

The fuel may be stored in the spent fuel storage pool provided the enriched lattices are not more reactive than the reference bounding lattices. This can be demonstrated by meeting either of the following two requirements:

1. The U-235 enrichment and gadolinia loading levels must meet the requirements specified below and shown graphically in Figure 2.2. The dimensions represent fuel column height above the bottom of active fuel (BAF) and below the top of active fuel (TAF).

Above [ ] Maximum Lattice Average Enrichment, wt% U-235 4.70 Minimum Number of Rods containing Gd2O3 8 Minimum wt% Gd2O3 in these Gd Rod 3.5 At and below [ ] Maximum Lattice Average Enrichment, wt% U-235 4.70 Minimum Number of Rods containing Gd2O3 8 Minimum wt% Gd2O3 in these Gd Rod 3.919 These eight gadolinia rods cannot be loaded on the perimeter of the lattice or adjacent to the water channel. An equivalent of 2 gadolinia rods must be loaded along each side.

Gadolinia is not required in natural Uranium blankets and there are no restrictions on the number, concentration, or placement of any additional gadolinia rods.

Or,

2. The lattice average enrichment is less than 5.0 wt% U-235, and the k of each enriched lattice does not exceed the following in-rack k values at any point during its lifetime. (The CASMO-4 storage rack model that must be used for this calculation is defined in Appendix A and the transition between top and bottom lattice geometries occurs at [ ] inches from the bottom of the fueled length.)

Zone Lattice Geometry Distance from BAF Max. in-rack k 2 10XMLCT [ ] [ ] to TAF 0.8825 1 10XMLCB [ ] 0" to [ ] 0.8825

• These requirements describe the reference bounding lattices shown in Figure 2.2 and Table 7.1.

Two face adjacent gadolinia rods count as a single rod.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-5 Table 2.1 Criticality Safety Limitations for ATRIUM 10XM Fuel Assemblies Stored in the Browns Ferry Plant Spent Fuel Storage Pools (Continued)

Spent Fuel Storage Rack The spent fuel storage rack design parameters and dimensions are provided in Table 4.2.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-6

[

]

Figure 2.1 Overview of the Browns Ferry SFP Criticality Safety Analysis AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 2-7 TAF (Top of Active Fuel)

Zone 2 XMLCT-470UL-8G35

[ ] PLR fueled Boundary Zone 1 XMLCB-470UL-8G3919 0.0" BAF (Bottom of Active Fuel)

Figure 2.2 ATRIUM 10XM Reference Bounding Assembly AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-1 3.0 Regulatory Criticality Safety Criteria and Guidance Section 9.1.1 of the Standard Review Plan (Reference 2) identifies the regulatory requirements and associated acceptance criteria considered to be applicable to criticality safety analyses.*

Since this analysis does not support a change in the facility only the requirements specific to the criticality safety analysis apply. The primary requirements relevant to this analysis are General Design Criteria 62 and portions of 10 CFR 50.68, Reference 3. Although it is not specifically cited by SRP 9.1.1, General Design Criteria 5 is potentially of interest in a spent fuel criticality analysis.

The Browns Ferry units were not designed or licensed to the General Design Criteria provided in 10 CFR 50 Appendix A. Instead Appendix A of the Browns Ferry FSAR (UFSAR) provides a description of conformance to the AEC Proposed General Design Criteria. For Browns Ferry, the corresponding licensing basis applicable criteria are Criterion 4, Sharing of Systems, and Criterion 66, Prevention of Fuel Storage Criticality.

Criterion 4 (similar to GDC 5) addresses the sharing of systems important to safety specifically to ensure that the ability to perform their safety function is not significantly impaired. The existence of a transfer canal allows for the transfer of fuel bundles between the Units 1 and 2 spent fuel pools (i.e., the only shared components). All three of the spent fuel pools at the Browns Ferry plant have essentially the same configuration and use the same rack designs, as described in Section 4. The previously manufactured fuel evaluation identifies the most limiting fuel from all three pools and conservatively applies this in the definition of common reference bounding lattices to be used for all three pools. This bounding treatment helps to ensure that the ability of the spent fuel pool racks to maintain subcriticality is not impaired when fuel transfer between pools occurs and the intent of AEC Proposed Criterion 4 (and GDC 5) is therefore met.

Criterion 66 (similar to GDC 62) specifies that criticality of fuel in handling or storage will be prevented by physical systems or processes with the preference for geometrically safe configurations. There is no physical change being implemented that affects the configuration of the spent fuel storage system (i.e. no change to the systems, components, or structures that comprise the spent fuel storage system). The purpose of this analysis is to provide assurance

• SRP 9.1.1 is used as the basis for discussion of general requirements for criticality safety analyses in this report. This context does not represent a commitment on the part of the licensee in regard to conformance with this section of the Standard Review Plan.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-2 that criticality will not occur within the basis of the existing spent fuel storage configuration for both previously manufactured (or planned) fuel and ATRIUM 10XM designs to be provided in the future; therefore the intent of Criterion 66 (and GDC 62) is met.

10 CFR 50.68*(a) requires that a licensee must either: 1) maintain monitoring systems in accordance with 10 CFR 70.24 to reduce the consequences of a criticality accident, or 2) adhere to the requirements of 10 CFR 50.68(b) to reduce the likelihood that a criticality accident will occur. Browns Ferry complies with the requirements of part (b) of 10 CFR 50.68. The role of this criticality safety analysis in meeting the specific requirements for each of the 10 CFR 50.68(b) requirements is discussed below:

1) Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

Technical Specification 4.3.1.1(a) requires that a k-effective of less than or equal to 0.95 must be maintained with unborated water. This analysis establishes the SFP storage requirements that meet this licensing requirement. Fuel handling has also been addressed by this analysis.

2) The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

This requirement does not apply because this is not a fresh fuel storage rack criticality analysis.

3) If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum

• Reference 3.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-3 moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.

This requirement does not apply because this is not a fresh fuel storage rack criticality analysis.

4) If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

This criticality safety analysis is being performed specifically to show that this requirement has been met. The applicable requirement is a k-effective of 0.95 at a 95 percent probability with a 95 percent confidence level because Browns Ferry is a BWR site with unborated water in the SFP. This requirement is also enforced in section 4.3.1.1(a) in the Technical Specifications for each Browns Ferry unit. The analysis described in this report demonstrates that the calculated k95/95 value meets this requirement.

5) The quantity of SNM, other than nuclear fuel stored onsite, is less than the quantity necessary for a critical mass.

This requirement does not apply because this analysis only addresses nuclear fuel storage in the SFP.

6) Radiation monitors are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions.

This requirement does not apply because this is a criticality analysis only.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-4

7) The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0) percent by weight.

This criticality safety analysis establishes maximum allowable enrichments below the regulatory requirement and therefore complies with the intent of this requirement.

8) The FSAR is amended no later than the next update which § 50.71(e) of this part requires, indicating that the licensee has chosen to comply with § 50.68(b).

The licensee has included the required 10CFR 50.68(b) compliance statement in section 10.3 Spent Fuel Storage of the Browns Ferry FSAR.

This criticality safety analysis complies with the intent of all of the applicable sections of 10 CFR 50.68(b).

Based upon the discussion above, this analysis complies with the intent of the Proposed AEC General Design Criteria 4 and 66 as well as 10 CFR 50.68(b).

The USNRC has recently issued document DSS-ISG-2010-01 Revision 0 (Reference 7) that provides interim staff guidance (ISG) for the review of spent fuel criticality safety analyses.

Table 3.1 provides a top level summary discussion regarding the compliance of this criticality safety analysis to the ISG document. Where possible, this discussion includes a cross-reference to where specific items identified in the ISG are addressed within this criticality safety analysis report.

The following sources provide additional guidance in meeting the aforementioned regulatory requirements:

• Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, also known as the Kopp letter this was issued by the NRC in 1998 (Reference 6).

• OT Position for the Review and Acceptance of Spent Fuel Storage and Handling Applications, issued by the NRC in 1978 and amended in 1979 (Reference 5).

• ANSI/ANS American National Standard 8.17-1984 (Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors) issued by the American Nuclear Society, January 1984 (Reference 4).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-5 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.1 Fuel Assembly Selection the staff should review the The lattices of the ATRIUM 10XM reference Appendix B submittal to verify that it bounding assembly are demonstrated to be demonstrates that the NCS more reactive than the lattices of any analysis adequately bounds previously loaded fuel assembly, including all designs, including variations due to damaged and modified variations within a design. assemblies.

the staff should verify each As discussed above, the ATRIUM 10XM Section 2.0 application includes a portion reference bounding assembly is shown to of the analysis that bound all previous designs. Compliance with Appendix B demonstrates that the fuel the requirements listed in Table 2.1 ensures assembly used in the that future ATRIUM 10XM assemblies analysis is appropriate for the remain bounded by this evaluation.

specific conditions.

IV.1.a Use of a single limiting fuel The use of the ATRIUM 10XM reference Section 2.0 assembly design should be bounding assembly (and corresponding assessed, lattices) is justified as described above. Appendix B IV.2 Depletion Analysis simulates the use of fuel in This evaluation does not directly use the Sections 7.0 a reactor. These depletion depletion based isotopic number density and 7.1 simulations are used to values in KENO. The CASMO-4 based create the isotopic number incore depletion is used to establish the Appendices densities used in the inrack lifetime maximum reactivity condition B&D criticality analysis. of the reference bounding lattices. Reactivity equivalent at beginning of life (REBOL) lattices are then defined for use in the KENO calculations. The REBOL lattices are defined with a conservative bias to address the uncertainty in the CASMO-4 depletion process and reactivity equivalence method.

The definition of the reference bounding and REBOL lattices are described in more detail in Sections 7.0, 7.1, and Appendix B.

Appendix D provides details on the treatment of the depletion uncertainty.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-6 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.2.a Depletion Uncertainty An overall CASMO-4 uncertainty reflecting Appendix D calculational and depletion based isotopic an uncertainty equal to 5 uncertainties is defined in Section D.4. This percent of the reactivity value is bounded by the 0.010 k bias term decrement to the burnup of applied during the reactivity equivalence interest is an acceptable calculation. Two independent estimates of assumption. the depletion uncertainty were used in this evaluation. One of these methods is should only be construed consistent with the 5% reactivity decrement as covering the uncertainty in described in Reference 7 (except that it the isotopic number includes an additional component for the densities gadolinia uncertainty).

IV.2.b Reactor Parameters Sensitivity comparisons are included in Section 6.5 Section 6.5 to show that reasonable the staff should verify that parameters have been used in the depletion each application includes a calculations. The parameters evaluated portion of the analysis that include:

demonstrates that the reactor parameters used in the Fuel Temperature (Assumption 2, Table depletion analysis are 6.4); Moderator Temperature/Void appropriate for the specific History ( Assumption 3, Figure 6.5);

conditions. Power Density (Assumption 4, Table 6.5); and Rodded Depletion (Assumption 7, Table 6.6)

IV.2.c Burnable Absorbers Only integral burnable absorbers have been Table 2.1 used in the Browns Ferry reactor and they the staff should verify that have been modeled appropriately in Section 7.1 each application includes a Appendix B. The placement of the 8 portion of the analysis that gadolinia rods in the reference bounding Appendix B demonstrates that the lattices have been selected to produce a treatment of burnable high reactivity condition. Table 2.1 requires absorbers in the depletion that all enriched lattices of future ATRIUM analysis is appropriate for the 10XM assemblies contain a minimum specific conditions. number of absorber rods with a minimum concentration level or that a CASMO-4 k less than the applicable reference bounding lattice be demonstrated.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-7 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.2.d Rodded Operation Assumption 7 of Section 6.5 addresses Section 6.5 rodded depletion. The use of uncontrolled the staff should verify that depletion at rated power conditions is shown each application includes a to bound depletion at controlled conditions portion of the analysis that for the ATRIUM 10XM reference bounding demonstrates its treatment of lattices.

rodded operation is appropriate for its specific conditions.

IV.3 Criticality Analysis IV.3.a Axial Burnup Profile This evaluation uses the lifetime maximum Section 7.0 reactivity of each lattice of the reference the staff should verify that bounding assembly as discussed in Section each application includes a 7.0. Therefore, there is no burn-up profile portion of the analysis that assumption.

demonstrates its treatment of axial burnup profile is appropriate for its specific conditions.

IV.3.b Rack Model The modeling of the spent fuel racks have Section 6.1 been explicitly addressed in Table 6.3. Appendix D the staff should verify that Comparisons in Table 6.1 demonstrate that each application includes a the infinite 2x2 model is more reactive than portion of the analysis that the explicit model. Comparisons in Table demonstrates that the rack 6.1 and Appendix D show that the 2x2 model analysis used in its model agrees well with the single cell model submittal is appropriate for its used in CASMO-4.

specific conditions.

IV.3.b.i The dimensions and The rack dimensions and materials in Table Section 4.2 materials of construction 4.2 were derived from the licensees design should be traceable to documents.

licensee design documents.

The Boral is modeled using the licensees IV.3.b.ii The efficiency of the neutron Section 4.2 design minimum Boron-10 areal density.

absorber should be Neutron self shielding and streaming are established, especially addressed in Section 4.2.

considering the potential for self-shielding and streaming.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-8 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.3.b.iii Any degradation should be A conservative blister model has been Sections modeled conservatively, incorporated to account for Boral blistering. 7.4.3 & 7.8 consistent with the certainty with which the material condition can be established.

IV.3.c Interfaces The 13x13 and 13x17 Boral racks have Sections 4.2 been shown to be less reactive than the 2x2 & 6.3 the staff should verify that infinite array model.

each application includes a portion of the analysis that demonstrates that the interface analysis used is appropriate for its specific conditions.

IV.3.c.i Absent a determination of a There is no significant difference between Section 6.3 set of biases and the 13x13 racks and the 13x17 racks.

uncertainties specifically for the combined interface model, use of the maximum biases and uncertainties from the individual storage configurations should be acceptable in determining whether the keff of the combined interface model meets the regulatory requirements.

IV.3.d Normal Conditions Translation and orientation variations of the Sections assemblies within the storage racks are 7.4.1, 7.4.2, the staff should verify that considered in Sections 7.4.1 and 7.4.2. and 7.5 each application includes a The fuel handling considerations for normal portion of the analysis that conditions are addressed in Section 7.5.

demonstrates that the NCS analysis considers all appropriate normal conditions for its specific conditions.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-9 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.3.e Accident Conditions The accident conditions have been Section 7.6 evaluated in Section 7.6.

The reviewer should verify all credible accident conditions are addressed.

IV.4 Criticality Code Validation IV.4 The proposed analysis The criticality benchmark is shown in Appendix C methods and neutron cross- Appendix C. Since this is a fresh fuel section data should be equivalent evaluation, only critical benchmarked, by the analyst experiments for fresh fuel have been or organization performing included in the benchmark data set.

the analysis, by comparison with critical experiments. ...

The critical experiments should include configurations having neutronic and geometric characteristics as nearly comparable to those of the proposed storage facility as possible.

IV.4.a Area of Applicability The area of applicability is defined by the Section 5.1 criticality benchmark comparisons provided Appendix C the staff should verify that in Appendix C. Section 5.1 also provides a applications demonstrate that summary of this validation and addresses the validation fully covers the the area of applicability for this Browns Ferry area of applicability for their spent fuel storage pool criticality safety specific SFP; analysis.

HTC benchmarks are not included in the IV.4.a.i The reviewer should verify Appendix C validation set since this is a fresh fuel any validation used for SNF reactivity equivalent evaluation. The appropriately considers treatment of actinides and fission products is actinides and fission part of the CASMO-4 depletion uncertainty products. NUREG/CR-6979, addressed in Appendix D.

Evaluation of the French Haut Taux de Combustion However, the addendum to Appendix C (HTC) Critical Experiment compares the impact on the KENO bias and Data, issued September uncertainties if appropriate benchmark 2008 experiments from the HTC criticals were included. This comparison shows that a more conservative result is obtained without inclusion of the HTC criticals.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-10 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.4.a.ii Experiments should be The criticality benchmark data shown in Appendix C appropriate to the system Appendix C meets the requirements being analyzed. expressed in the ISG.

IV.4.a.iii The reviewer should The criticality benchmark dataset has been Appendix C

.{review the selection of selected to provide a balanced benchmark data} representation of the spent fuel pool environment. It is shown in Appendix C.

IV.4.a.iv The reviewer should ensure The criticality benchmark is shown in Appendix C that the experiments are not Appendix C.

all highly correlated, e.g.

critical configurations performed with the same fuel rods at the same facility.

IV.4.b Trend Analysis The trending analysis is performed in Appendix C Appendix C.

the staff should verify that each application includes a portion of the analysis that demonstrates that the trend analysis used in its validation is appropriate for its specific conditions.

IV.4.c Statistical Treatment The benchmark validation suite in Appendix Appendix C C follows the guidance given in NUREG/CR-the staff should verify that 6698 with respect to using the variance each application includes a about the mean, confidence factors, and the portion of the analysis that treatment of non-normal distributions.

demonstrates that the statistical treatment used in its validation is appropriate for its specific conditions.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-11 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.4.d Lumped Fission Products The primary components of the Browns Section 7.1 Ferry nuclear criticality safety (NCS) the staff should verify that analysis include the use of the CASMO-4 each application that includes code in the definition of the reference lumped fission products bounding and REBOL lattices followed by includes a portion of the the actual NCS calculations with KENO V.a analysis that demonstrates (using the defined REBOL lattices). While that the lumped fission CASMO-4 does include the use of lumped products used in its validation fission products, they are not credited in the are appropriate for its specific definition of the reference bounding lattices.

conditions. Therefore, the KENO calculations and k95/95 result are conservative since the lumped fission products have been removed.

IV.4.e Code-to-Code Comparisons Code-to-code comparisons are not used in Appendix C the validation of KENO V.a - the code used Appendix D the use of a code-to-code for the criticality analysis.

comparison for validating criticality codes is outside the The only use of code-to-code comparisons scope of this ISG. is for the depletion code, CASMO-4. This use is limited to perturbation calculations used to quantify the CASMO-4 calculational uncertainty relative to KENO V.a.

IV.5 Miscellaneous IV.5.a Precedents Although not specifically cited, the approach N/A taken in this spent fuel pool criticality safety the staff should verify that analysis is similar to a previous SFP for cited precedents, the criticality analysis recently reviewed and application includes a portion approved by the USNRC (Reference of the analysis that accession numbers ML092810281 and demonstrates the ML101650230).

commonality of the precedent to the submittal, with any Some changes were incorporated to directly differences identified and address USNRC concerns identified in the justified with respect to the SER accepting the above submittal use of the precedent. (ML110250051) and to provide closer compliance to the staff guidance document.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 3-12 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 (Continued)

ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.5.a Specifically, Continued

• The criticality benchmark data suite has been modified to remove soluble boron and MOX benchmarks.

• The previous submittal content was split between a submitted report and additional answers to USNRC requests for additional information. This information content has been reformatted into a single report.

• Differences in modeling, primarily to address differences in the plant specific rack designs.

• The CASMO-4 lumped fission products are not credited in the in-rack k values for the reference bounding lattices when the reactivity equivalence comparison is being performed.

IV.5.b References The analysis uses references N/A appropriately.

the NRC reviewer should verify that references cited in the application are used in context and within the bounds and limitations of the references. Any extrapolation outside the context or bounds of the reference should be demonstrated as appropriate.

IV.5.c Assumptions Modeling assumptions have been explicitly Section 6.0 addressed in the report.

applications should explicitly identify and justify all assumptions used in their applications.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-1 4.0 Fuel and Storage Array Description A number of different assembly types have previously been loaded in the Browns Ferry spent fuel pools with lattice geometries ranging from 7x7 to 10x10. This includes variations in the type and number of water rods and part length fuel rods. The AREVA ATRIUM 10XM fuel product line is planned for use in future reloads. For this reason, the ATRIUM 10XM reference bounding assembly design forms the basis for demonstrating that the maximum k-eff of the spent fuel pool storage array remains less than 0.95.

4.1 Fuel Assembly Design The ATRIUM 10XM fuel assembly is a 10x10 fuel rod array with an internal square water channel offset in the center of the assembly (taking the place of nine fuel rod locations). The assembly contains part-length fuel rods (PLR); therefore, the top lattice geometry will apply above the PLR fueled boundary and the bottom lattice geometry will apply below the PLR fueled boundary. The ATRIUM 10XM mechanical design parameters are summarized in Table 4.1 and a representation of the ATRIUM 10XM assembly design is provided in Figure 4.1. The ATRIUM 10XM fuel in the Browns Ferry Nuclear Plant uses the AREVA advanced (i.e.,

thick/thin) fuel channel design.

4.2 Fuel Storage Racks Each of the Browns Ferry spent fuel pools provide the capability of storing 3471 fuel assemblies. Each pool contains 14 - 13x13 high density Boral storage rack modules and 5 13x17 high density Boral storage rack modules. The dimensional parameters for these racks are given in Table 4.2 and the pool arrangement is shown in Figure 4.2. The layout in all three pools is essentially the same except that the Unit 1 pool has a mirror symmetric layout when compared to the Unit 2 or Unit 3 pool.

A transfer canal is provided to join the Unit 1 and 2 pools. This transfer canal is the same depth as the transfer slot between the reactor well and the fuel pool. The transfer canal has a gate at each end so that the fuel pools can be isolated. There is no corresponding transfer canal for the Unit 3 pool.

Each high density Boral rack module is composed of alternating or staggered stainless-steel square container tubes. This arrangement results in only one container-tube wall between AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-2 adjacent fuel assemblies, as illustrated in Figure 4.3 and Figure 4.4. Each container-tube wall has a core of Boral sandwiched between inner and outer surfaces of stainless steel. The Boral core is made up of a central segment composed of a dispersion of boron carbide in aluminum.

This central segment is clad on both sides with aluminum. These stainless steel container tubes are closure welded with vent holes to prevent the buildup of hydrogen gas. The completed storage tubes are fastened together by angles welded along the corners and attached to a base plate to form storage modules. These modules are designed to be free standing with low-friction between the module support and pool floor liner.

Note on the Efficacy of Boral: In a water environment, neutron scattering ensures that neutrons approach the Boral from a full range of incident angles. This minimizes the potential for neutron streaming and reduces the significance of self-shielding.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-3 Table 4.1 ATRIUM 10XM Fuel Assembly Parameters Parameter Value Fuel Assembly Fuel Rod Array 10x10 Fuel Rod Pitch, in. [ ]

Number of Full Length Fuel Rods [ ]

Number of Part Length Fuel Rods [ ]

Location of Part Length Fuel Rods See Figure 6.1 Water Channel 1 Fuel Rods Fuel Material UO2 Pellet Density, % of Theoretical Density (%TD) [ ]

Pellet Diameter, in. [ ]

Pellet Void Volume, % [ ]

Cladding Material Zircaloy Cladding OD, in. [ ]

Cladding ID, in. [ ]

Internal Water Channel Outside Dimension, in. [ ]

Inside Dimension, in. [ ]

Channel Material Zircaloy Fuel Channel (standard 100 mil)

Outside Dimension, in. [ ]

Inside Dimension, in. [ ]

Channel Material Zircaloy Fuel Column Lengths Distance from the bottom of the fuel to the top of the fuel in the part length fuel rods, in. [ ]

Total Fueled Length, in. [ ]

• Criticality safety analysis is also valid for lower fuel densities. The analysis uses the effective stack density which is a combination of the pellet density and the pellet void volume.

The conclusions in this report are also valid for advanced fuel channels (see Section 6.2).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-4 Table 4.2 Fuel Storage Rack Parameters Parameter Value High Density Boral Racks Boral B-10 areal density, g/cm2 0.013 minimum Rack Box OD, in. 6.653 +/- 0.04 Box material Stainless steel Inner rack box wall thickness, in. 0.0355 +/- 0.004 Box material Stainless steel B4C plate thickness, in. 0.076 +/- 0.005 plate material B4C and aluminum clad in two 0.010" aluminum sheets width, in 6.20* +/- 0.03 height, in 152.00 Outer rack box wall thickness, in. 0.090 +/- 0.008 Box material Stainless steel Rack cell pitch, in. 6.563 +/- 0.03 Closure plate thickness, in. 0.125 material Stainless steel Rack to Rack Spacing, in. 2.33

• 6.20 has previously been represented as a minimum width value. For consistency with a similar storage rack design, 6.20 will be treated as a nominal value with the indicated uncertainty.

Rack to Rack spacing is the distance from the outside surface of adjacent closure plates. (This value is derived from a 1.875" spacing at the rack module base).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-5

[

]

Figure 4.1 Representative ATRIUM 10XM Fuel Assembly (Assembly length and number of spacers has been reduced for pictorial clarity)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-6 Fuel Prep. Fuel Prep.

Sipping Can Storage Sipping Can Storage Machine Machine N Cask Pad Area Figure 4.2 Browns Ferry Spent Fuel Pool Layout (not to scale)

(Unit 2 and 3 layout shown, Unit 1 is mirror symmetric with the cask pad area on top)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-7 Stainless Steel Closure Plate (outside of rack)

Boral (Boron Carbide in OUTSIDE Aluminum) clad with aluminum CORNER OF RACK Stainless Steel (inner box)

Boral Tube Open Cell Boral Tube Stainless Steel (outer box)

Open Cell Boral Tube Open Cell Stainless Steel Closure Plates (outside of rack)

Boral Tube Open Cell Boral Tube Open Cell Boral Tube Open Cell Figure 4.3 Schematic Representation of a Section of High Density Storage Rack (not to scale)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 4-8 Section A-A 0.010" Al clad outside inside 0.090" +/- 0.008" 0.056" SS Boral 0.0355" +/- 0.004" core SS 6.563" +/- 0.03" cell pitch A A 6.653" +/- 0.04" outside dimension 6.563" +/- 0.03" cell pitch Figure 4.4 High Density Boral Storage Rack Geometry (not to scale)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 5-1 5.0 Calculation Methodology The spent fuel storage criticality safety evaluation is performed with the KENO V.a Monte Carlo code, which is part of the SCALE 4.4a Modular Code System (Reference 8). The SCALE driver module CSAS25 uses the ENDF/B-V 44 energy group data library. It also uses modules BONAMI-2 and NITAWL to perform spatial and energy self-shielding adjustments of the cross sections for use in KENO V.a. AREVA has benchmarked KENO V.a in accordance with NUREG/CR-6698 (Reference 9) using critical experiments related to the storage of fuel assemblies in water - including neutron absorbing materials such as stainless steel and Boral.

For applications using the 44 energy group data libraries, a KENO V.a bias magnitude of 0.0075 and a standard deviation of 0.0027 will be used (see Appendix C).

KENO V.a is run on the AREVA scientific computer cluster using the Linux operating system.

The hardware and software configurations are governed by AREVA procedures to ensure calculational consistency in licensing applications. The code modules are installed on the system and the installation check cases are run to ensure the results are consistent with the installation check cases that are provided with the code. The binary executable files are put under configuration control so that any changes in the software will require re-certification. The hardware configuration of each machine in the cluster is documented so that any significant change in hardware or operating system that could result in a change in results is controlled. In the event of such a change in hardware or operating system, the hardware validation suite is rerun to confirm that the system still performs as it did when the code certification was performed.

In this analysis the SCALE 4.4a code system is employed to:

• Calculate Dancoff coefficients.

• Calculate absolute k-effective results.

• Evaluate accident conditions, alternate loading conditions, and manufacturing tolerance conditions.

The CASMO-4 code is used when conditions require fuel and gadolinia depletion. CASMO-4 is a multigroup, two-dimensional transport theory code with a rack geometry option that allows typical storage rack geometries to be defined on an infinite lattice basis. This code is used for fuel depletion and relative reactivity comparisons in a manner that is consistent with AREVAs NRC approved CASMO-4 / MICROBURN-B2 methodology (Reference 10). CASMO-4 has been approved at Browns Ferry Nuclear Plant for BWR calculations and is included as a AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 5-2 methodology reference (via Reference 10) in Section 5.6.5.B of the Browns Ferry Nuclear Plant Technical Specifications. The CASMO-4 computer code is controlled by AREVA procedures and the version used in this analysis meets the requirements of Reference 10.

In this analysis CASMO-4 is employed to:

• Perform in-core isotopic depletion at characteristic void history levels, [ ] for bottom geometry lattices and [ ] for top geometry lattices. Use of CASMO-4 for in-core depletion is consistent with its application in EMF-2158(P)(A) (Reference 10).

• Perform in-rack k assessments to identify the most reactive lattices.

• Define lattices for a reference bounding assembly that represent the maximum reactivity condition supported by the analysis.

• Define the reactivity equivalent at beginning-of-life (REBOL) lattices with fresh fuel and no gadolinia for the subsequent KENO V.a base case criticality calculations. Note that for the REBOL lattices, the U-235 content is manually adjusted upward until the REBOL k is at least 0.010 k greater than the lattices of the reference bounding assembly. This 0.010 k is used to account for all uncertainties associated with defining the REBOL lattices - including calculational and depletion uncertainties of the CASMO-4 code as discussed in Appendix D.

• Evaluate a component of the manufacturing uncertainty for gadolinia content (i.e., the depletion component). This evaluation is needed because changes in gadolinia content affect reactivity more near peak reactivity than at beginning of life.

5.1 Area of Applicability Table C.6 in Appendix C shows the ranges of key parameters represented in the KENO V.a benchmark analysis. Parameters such as rectangular lattices of zircaloy clad UO2 fuel rods in a pool of water with stainless steel and boron are sufficiently general to not require comparison.

The remaining parameters are compared in the following table and show that the KENO V.a portion of this analysis has been performed within the range of experimental conditions used in the KENO V.a benchmark.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 5-3 Parameter Benchmark Values Values in this Analysis Enrichment (wt% U-235) 2.35 to 4.74 3.2 to 3.4 Fuel Rod Pitch (cm) 1.26 to 2.54 1.295 Moderating Ratio (H/X) 110 to >400 115 to 122 Energy of the Average 0.060 to 0.247 0.148 to 0.245 Lethargy Causing Fission (eV)

For the CASMO-4 qualification, ATRIUM 10XM fuel lattices were modeled using the Browns Ferry Nuclear Plant limiting storage rack geometry. Therefore, the CASMO-4 calculations performed for this evaluation are within the area of applicability of the comparisons shown in Appendix D.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-1 6.0 Modeling Options and Assumptions The following sections describe the primary modeling simplifications and assumptions used in this analysis including discussion of impact on in-rack reactivity.

6.1 Geometric Modeling of the High Density Boral Rack The geometry of the high density spent fuel storage racks includes an arrangement of staggered or alternating Boral tubes, as shown in Figure 4.3 and Figure 4.4. As a minimum, this rack requires an array of 2 tube cells and 2 non-tube cells for explicit modeling. The rack models described below were implemented in KENO V.a and the reactivity results are provided in Table 6.1. These models use infinite periodic boundary conditions in the x, y, and z directions.

6.1.1 Single Cell Model Description The primary simplifying assumptions can be generally described as follows:

• Boral Plate: The Boral plate is modeled as boron-10 only; i.e., the aluminum, carbon, and boron11 in the core of the plate and the aluminum clad on the outside of the plate are not included in this model. The location of the Boral is shifted to be between storage cells so that half of the actual thickness is assigned to each cell wall. The plate is assumed to extend to the corners of the storage cell (i.e., the water gap in the corners of the Boral tube is not modeled). Neglecting the non- boron-10 components of the Boral is slightly conservative because the neglected materials are relatively weak neutron absorbers. Extending the Boral plate to the corners is expected to have the opposite effect since it introduces a small additional amount of a strong neutron absorber.

• Stainless Steel Channels: One half of the total inner and outer stainless steel channels were combined in the model and assumed to make up the inside surface of the storage cell. The impact of this modeling simplification is expected to be minor since the amount of stainless steel is conserved and it still surrounds the Boral plate.

• Cell Pitch and Water Gaps: Average cell pitch and average water gap values are used in this model. This helps maintain the accuracy of this simplified model.

This is the model used in the CASMO-4 calculations. Figure 6.1 provides an illustration of the geometry for the single cell model.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-2 6.1.2 Explicit Storage Cell Model Description KENO V.a allows for more detailed modeling of the storage rack geometry than is possible with CASMO-4. The primary modeling changes in comparison to the single cell model are:

• Storage Geometry: The explicit model for KENO V.a is composed of a 2x2 array with two Boral tube storage cells and two open or non-tube cells.

• Boral Plate: The plate is modeled using the nominal width and the corner region is modeled as water. For comparison purposes one solution is provided using Boron10 only and the second solution models all components of the Boral plate; i.e., Boron, Carbon, and Aluminum.

• Cell Pitch: The average assembly pitch is modeled.

Figure 6.2 provides an illustration of the geometry for the KENO V.a explicit model.

6.1.3 Explicit Rack Model Description The high density Boral storage rack modules have an odd number of rows and columns. For this reason, each module has a Boral tube in each corner. When the racks are placed together, the cells in the adjacent rack have the same geometric configuration (i.e., a Boral cell is face adjacent to another Boral cell and an open cell is face adjacent to another open cell). As shown in Figure 6.3, some cells have two Boral plates between adjacent assemblies and some cells have no Boral material between assemblies. Details associated with the individual storage racks were modeled as described below.

• Storage Geometry: The explicit model from Section 6.1.2 is expanded to a 13x13 array with tube cells in each corner.

• Stainless steel closure plates are approximated for non-tube cells along the perimeter of the rack.

• The nominal rack to rack water gap* is modeled.

6.1.4 Reactivity Comparison of the Boral Rack Models Table 6.1 provides KENO V.a results for the single cell and more explicit geometry models.

Neglecting the single cell model, these results indicate that the explicit 2x2 storage cell model with the Boral modeled as boron-10 only produces the most conservative result. Other conclusions from this comparison are also listed below:

• The single cell model provides a good representation of the reactivity of the Boral rack.

• It is conservative to use a boron-10 only model for the Boral.

• 2.33 inches between the outer surfaces of the closure plates.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-3

• It is conservative to neglect the water gaps and closure plates between the storage rack arrays, i.e., the infinite cell model is more conservative than the 13x13 rack model.

The boron-10 only, explicit 2x2 model with periodic boundary conditions in all directions is represented as 0.897 and will be used to represent the reactivity level of the Boral racks. This k value is about 0.015 k more reactive than the result from the actual storage rack model (the 13x13 rack model with explicit B4C and finite axial boundary conditions in Table 6.1).

6.2 Fuel Assembly Modeling The CASMO-4 modeling of the previously manufactured fuel is performed using the actual lattice dimensions, enrichment, gadolinia loading, and channel type for each specific fuel product line. The KENO V.a in-rack calculations for the limiting ATRIUM 10XM fuel have been performed assuming a uniform 100 mil fuel channel.

A sensitivity calculation was performed with various channel thicknesses with the results summarized in Table 6.2. This analysis shows that in-rack reactivity generally increases with increasing fuel channel wall thickness. The increase in wall thickness results in an increase in channel mass and wall cross-sectional area which in turn results in larger water displacement.

The AREVA advanced channel design for ATRIUM 10XM fuel is thicker at the corners with a thinner wall along the sides and has a cross-sectional area that falls between the 80 mil and 100 mil channels. Consequently, an ATRIUM 10XM assembly modeled with a uniform 100 mil fuel channel is more reactive than an assembly without a fuel channel, an assembly with a uniform 80 mil fuel channel, and an assembly with the advanced fuel channel.

Zircaloy has been modeled in KENO as pure zirconium. Neglecting the neutron absorption of the alloying elements (primarily tin, iron, chromium, and nickel) is slightly conservative. In addition, the presence of activated corrosion and wear products (CRUD) is neglected because most of these compounds have higher neutron absorption cross sections than water.

6.3 Co-Resident Fuel Racks As shown in Figure 4.2 , the Browns Ferry spent fuel pools contain a combination of high density Boral racks (13x13 and 13x17). The in-rack k values for these storage rack types are compared in Table 6.3 with the limiting water temperature specified. This comparison shows that the high density Boral racks have similar reactivity characteristics.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-4 Review of Section 6.1.4 confirms that the 2x2 Boral infinite cell model is the most limiting overall; therefore, it will be used as the bounding representation of all possible rack configurations in the Browns Ferry spent fuel pool. Based on these comparisons, the 0.897 k result will be used as the primary basis for the k95/95 calculation.

6.4 BLEU versus Commercial Grade Uranium ATRIUM 10XM reloads for Browns Ferry may contain both commercial grade uranium and BLEU* fuel. The reference bounding and REBOL lattices used in this analysis are conservatively based upon commercial grade uranium. The previously manufactured BLEU lattices have been explicitly modeled using the U-234 and U-236 content corresponding to the feed material used in each reload batch.

The U-236 content in the BLEU fuel acts as a neutron absorber and reduces the lattice reactivity compared to an equivalent lattice composed of commercial grade uranium. This is illustrated in Figure 6.4 which compares the in-rack reactivity for the reference bounding lattices with and without BLEU fuel. As can be seen in this figure, the use of BLEU fuel significantly reduces lattice reactivity. The use of commercial grade uranium in the reference bounding and REBOL lattices is therefore conservative. With this level of conservatism, no BLEU specific manufacturing uncertainties will be applied to address application of these results to fuel containing BLEU material.

6.5 General CASMO-4 Modeling Assumptions The application of CASMO-4 for in-core fuel depletion is consistent with the NRC approval of EMF-2158(P)(A) (Reference 10). Input for the depletion calculation includes the fuel assembly material and geometry. The ATRIUM 10XM fuel assembly parameters are given in Table 4.1.

The key fuel pool storage rack parameters are given in Table 4.2. The following general assumptions have been made in regard to CASMO-4 modeling.

Assumption 1: The top of the part length rods in the ATRIUM 10XM assembly, which contain a 6 inch plenum, can be treated as water in the lattice in-core depletion and in the in-rack

• Blended Low Enriched Uranium (BLEU) is surplus Department of Energy material that has been down-blended to commercially acceptable enrichment levels (i.e. < 5% U-235 by weight). The primary difference of the BLEU material compared to commercial grade uranium is the existence of higher levels of U-234 and U-236 - due to previous in-reactor use prior to reprocessing. The presence of the U-236 isotope has the primary impact on lattice reactivity since it is a strong neutron absorber.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-5 calculations. The actual content of the 6 inch plenum consists of a stainless steel spring and fill gas. Neglecting the 6 inch plenum is conservative from a criticality stand point because it models a more reactive condition by adding more moderator and neglecting the neutron absorption of the plenum spring material.

Assumption 2: A fuel temperature is assumed for the fuel depletion based on the core average linear heat generation rate. Therefore, consistent fuel temperatures are used for each geometry type. Sensitivity studies were performed to determine the impact of the fuel temperature used in the fuel depletion on the in-rack storage reactivity. The fuel temperature was varied plus and minus 100 ºF relative to the base depletion temperature for the reference bounding and limiting lattices. Table 6.4 provides the in-rack results based on in-core depletion at the different temperatures (i.e., the cold in-rack calculations were repeated for the in-core depletions performed at the different temperatures). These results demonstrate that moderator void is much more significant than the depletion fuel temperature.

Assumption 3: The moderator temperature used for in-core depletion is assumed to be at saturated conditions corresponding to the rated dome pressure. The more important parameter in a BWR reactor is the actual moderator density/void level. The in-core depletion calculations are performed at [ ] void history conditions for bottom geometry lattices and [

] void history conditions for top geometry lattices. Figure 6.5 shows the results of a sensitivity evaluation with respect to the in-core depletion void history and its effect on the maximum inrack lattice k. For the reference bounding and limiting lattices the discrete void history conditions evaluated produced (or exceeded)* the maximum credible k result.

Assumption 4: The power density used for the fuel depletion is based on the core rated power per unit volume which is consistent with AREVAs standard NRC-approved depletion methodology, Reference 10. Table 6.5 provides the reactivity effect as a function of power density where 100% power density represents the core average power density at rated power.

This sensitivity analysis was performed for the reference bounding lattices and the limiting lattices listed in Table B.1 of Appendix B. These results show a small effect on inrack lattice k

• The k value reported for the top geometry GE14 lattice is based upon [ ] void history (see Tables B.2 and B.5. Given that [ ] void history is not credible for full power operation in the top of an assembly it would be acceptable to use the [ ] void history value (0.861). It is conservative to use of the 0% void history value (0.862) for this lattice.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-6 over very large changes in depletion power density. These results also demonstrate that moderator void is much more significant than the depletion power density.

Assumption 5: Modeling the pellet deformation with respect to burnup can be ignored for the in-core depletion and in-rack calculations. Modeling of the pellet deformation does not significantly change the neutronic characteristics of the fuel since the material content is unchanged.

Assumption 6: The spacer (i.e., spacer grid) material can be ignored in the in-core depletion and in-rack calculations. There is no soluble boron in this BWR spent fuel pool, and the spacers will absorb more neutrons than water. Therefore, a more reactive configuration is modeled when the spacer material is neglected.

Assumption 7: The in-core depletion is based upon uncontrolled statepoint conditions. This is appropriate because a bundle is in an uncontrolled state (i.e., the adjacent control blade is not inserted) for the majority of its lifetime, including the time from beginning of life (BOL) to the time when it reaches its lifetime maximum reactivity.

Bundles are physically located in a control cell that is associated with a specific control rod sequence (i.e., A1, A2, B1, or B2); therefore, the potential for controlled operation is limited to the times when the core is operated in that sequence (e.g., one control period in four for a core operated with the typical four control rod sequence strategy). Furthermore, the following factors tend to mitigate the amount of controlled depletion: 1) not all available in-sequence control rods are used during a sequence, 2) control rods are typically not fully inserted (they may be in deep, intermediate or shallow positions which leaves the upper lattices in an uncontrolled state), 3) bundles in peripheral and near peripheral core locations are usually not controlled, and 4) the bundles are at a reduced power during the controlled time period which reduces their accumulated burnup while in the controlled state (i.e., they experience a lower burnup rate). The net effect is that a typical bundle will experience controlled depletion for only a fraction of its time from BOL to the exposure that produces its lifetime maximum reactivity.

Potential exceptions to this behavior are: 1) bundles in a power suppression cell, and 2) bundles in a control cell in which the control rod has been declared inoperable. Power suppression is the practice of inserting a control rod to reduce power in suspected leaking fuel bundles. The control rod is typically fully inserted in an inoperable control cell. In either case, the control rod may be inserted for a significant period of time and the bundles around them will have a larger AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-7 fraction of their lifetime spent in a controlled state. However, this only affects a small population of bundles four bundles for each affected control cell.

Rodded depletion introduces impacts due to the power gradient imposed by the inserted blade as well as to the assumed power density due to the associated power reduction in the bundle.

A sensitivity calculation was performed to determine the impact of in-core controlled depletion on the peak in-rack k-infinity in which both of these parameters were varied. Table 6.6 compares the lifetime maximum k results for the ATRIUM 10XM reference bounding lattices and the limiting fuel lattices. The results in Table 6.6 are very conservative because no significant number of fuel assemblies will be controlled from BOL to the peak reactivity exposure. These results indicate a reactivity increase for the limiting GE lattices; however, as shown in Table B.1 there is substantial margin between the reactivity of the legacy fuel lattices and the ATRIUM 10XM reference bounding lattices. The results in Table 6.6 also show that uncontrolled depletion results in a higher in-rack k-infinity for the reference bounding lattices.

Assumption 8: The CASMO-4 model uses a lumped approach for fission products that are not specifically treated. CASMO-4 creates two pseudo nuclides to represent the general behavior of two fission product groups - one non-saturating and one slowly saturating. Any errors in the treatment of these pseudo nuclides becomes part of the depletion uncertainty and is included in the benchmarking and qualification of the CASMO-4 code for in-core depletion, as described in the approved topical report EMF-2158(P)(A) (Reference 10). For this evaluation, the lumped fission products were removed from the reference bounding lattices when the REBOL lattices were defined*. This modeling adds additional conservatism to the evaluation.

• Note that the lumped fission products were not removed for relative comparison calculations such as Table 6.4, Table 6.5, Table 6.6, Figure 6.4, Figure 6.5, and comparisons in the Appendices.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-8 Table 6.1 Comparison of Modeling Options for the Boral Rack KENO V.a In-Rack k Single Cell Model k Rack Model Developed for CASMO-4 0.8969 (boron-10 only*)

2x2 Infinite Array Model k Base (boron-10 only*) 0.8964 Base with Explicit Boral 0.8936 13x13 Rack Model k

(closure plates and nominal rack spacing)

Base (boron-10 only*) k = 0.8870 Base with 12" water reflector on the top and keff = 0.8847 a 24" concrete reflector on the bottom Base with Explicit Boral, 12" top water keff = 0.8819 reflector and a 24" concrete bottom NOTE: The neutron multiplication values are based upon the limiting water temperature condition, (4 °C for infinite cell conditions or 20 °C for finite rack conditions). These cases produce a KENO standard deviation of about 0.0008.

• All non-boron-10 materials in the Boral plate are neglected (i.e., modeled as void).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-9 Table 6.2 Impact of Channel Thickness on In-Rack Reactivity Fuel Channel Thickness KENO V.a (mil) (inch) k Result*

100 0.100 0.8950 80 0.080 0.8939 0 0.000 0.8921 Table 6.3 Co-Resident Storage Rack Comparison KENO V.a Limiting In-Rack k Temperature 13 x 13 Boral Rack k = 0.8870 20 °C 13 x 17 Boral Rack keff = 0.8876 20 °C

• Based on 20 °C moderator temperature.

Cases were evaluated between 4 °C and 60 °C.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-10 Table 6.4 In-Rack k Sensitivity to In-core Depletion Fuel Temperature

[

]

• Includes lumped fission products.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-11 Table 6.5 In-Rack k Sensitivity to In-core Depletion Power Density

[

]

• Includes lumped fission products.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-12 Table 6.6 In-Rack k Sensitivity to In-Core Controlled Depletion

[

]

• Includes lumped fission products.

PD refers to Power Density.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-13

[

]

Figure 6.1 Single Cell Model for the High Density Boral Rack (not to scale - top zone geometry)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-14

[

]

Figure 6.2 Explicit Geometry Model for High Density Boral Rack (not to scale - top zone geometry)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-15 Open Cell Boral Tube Boral Tube Open Cell Boral Tube Open Cell Open Cell Boral Tube Open Cell Boral Tube Boral Tube Open Cell No Boral Plate between cells in adjacent racks Two Boral Plates between cells in adjacent racks Figure 6.3 Schematic of Rack to Rack Interfaces AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-16

[

]

Figure 6.4 BLEU versus Commercial Grade Uranium Reactivity Comparison AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 6-17

[

]

Figure 6.5 Impact of Void History Depletion on In-Rack k-infinity AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-1 7.0 Criticality Safety Analysis This criticality safety analysis is based upon an ATRIUM 10XM reference bounding assembly.

This reference assembly is comprised of separate top and bottom geometry reference bounding lattices* and they have been defined to be more reactive than all previously manufactured lattices - as well as future ATRIUM 10XM lattices. The evaluation of the previously manufactured fuel and comparisons to these reference bounding lattices are detailed in Appendix B of this report. The reference bounding ATRIUM 10XM assembly is comprised of two axial zones as described in the following table and as shown graphically in Figure 2.2.

Lattice 235 No. of Gadolinia Gadolinia Zone Distance from BAF U wt%

Geometry Rods wt%

2 10XMLCT [ ] to TAF 4.70 8 3.5 1 10XMLCB 0" to [ ] 4.70 8 3.919 The reference bounding lattices are depleted in the reactor core environment to establish the lifetime maximum k of these lattices in the storage pool environment. The resulting k values are mainly dependent upon the lattice geometry, the U-235 enrichment level, and the gadolinia concentration; therefore, there is no axial burn-up profile assumption associated with this method.

The actual KENO V.a calculations are based upon reactivity equivalent at beginning of life (REBOL) lattices that have been designed to be more reactive than the reference bounding lattices and their calculational uncertainties. For this evaluation, a U-235 enrichment level of 3.38 wt% is applied for the top (10XMLCT) geometry and 3.21 wt% is applied for the bottom (10XMLCB) geometry.

The final k95/95 evaluation is based upon a number of factors that include the worst credible conditions and uncertainties. Items considered include assembly placement within the storage cell, assembly orientation, manufacturing uncertainties, and accident conditions.

• It is demonstrated in Appendix B that the ATRIUM 10XM reference design in the spent fuel pool geometry is more reactive than the other fuel types used at Browns Ferry.

The CASMO-4 vs. KENO comparison in Appendix D demonstrates a stable basis for this reactivity equivalence. The 2x2 KENO model used in Appendix D was also established as the maximum k case in Section 6.1.4.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-2 7.1 Definition of the Reference Bounding and REBOL Lattices The CASMO-4 lattice depletion calculations are performed at hot operating, uncontrolled, characteristic void history conditions. These void history conditions are [ ] for top geometry lattices and [ ] for the bottom geometry lattices. The calculation results are based upon the nominal fuel design parameters (defined in Table 4.1) and assume a standard 100 mil fuel channel. The location of the 8 gadolinia rods in the reference bounding lattices have been selected to maximize the reactivity of the lattices. Xenon and lumped fission product free restart calculations are performed as a function of exposure and void history to establish the highest in-rack reactivity (k) at any time throughout the life of these fuel lattices.

The CASMO-4 in-rack k of the top and bottom zone reference bounding lattices are both 0.8825. These results are summarized in Table 7.1.

The reference bounding and REBOL lattices are based upon a uniform enrichment distribution.

A uniform enrichment distribution increases the BWR lattice reactivity because low enriched rods in the corners of the lattice are replaced with rods at an average enrichment level. Relative to a representative top and bottom ATRIUM 10XM lattice design, a uniform enrichment distribution was determined to be more reactive by 0.002 to 0.004 k. Consequently, the use of these lattices with uniform enrichment distributions conservatively bound the distributed enrichment distributions of expected future lattice designs.

In support of the KENO rack calculations, two REBOL lattices are created corresponding to the top and bottom geometries for the ATRIUM 10XM design. These lattices are defined using a water temperature of 4 °C in the spent fuel pool rack configuration. The top REBOL lattice is defined with a uniform 3.38 wt% U-235 enrichment level, and the bottom REBOL lattice is defined with a uniform 3.21 wt% U-235 enrichment level. These results are also summarized in Table 7.1.

As discussed in the methodology section, an adder of at least 0.010 k is included in the generation of the REBOL lattices to address CASMO-4 code, geometry, material, and depletion uncertainties. The adequacy of this adder is the primary subject of Appendix D.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-3 7.2 Storage Array Reactivity The base storage array reactivity is calculated using KENO V.a as an infinite array of fuel storage cells using the explicit storage cell model as described in Section 6.1.2 and as illustrated in Figure 6.2. This model was shown to be conservative in Section 6.1.4. (KENO V.a results using this model were also shown to trend well with CASMO-4 results in Appendix D)

Each cell is assumed to contain an assembly composed of 3.38 wt% U-235 (top) and 3.21 wt%

U-235 (bottom), uniformly enriched REBOL lattices without gadolinia. As discussed earlier, each REBOL lattice is defined to be at least 0.010 k more reactive than its corresponding reference bounding lattice. A periodic boundary condition is specified for both the x-y plane and for the axial direction. The KENO model assumes a standard 100 mil fuel channel which was shown in Section 6.2 to bound storage with no channel, an 80 mil channel, and the advanced thick-thin channel.

KENO V.a calculations were performed at various temperatures from 4 C to 60 C that confirmed that the REBOL assembly is bounded by the 4 C results. As shown in Table 7.2, the limiting base case KENO k-eff is 0.897. Except as specifically noted, the reactivity values presented in Table 7.1 and Table 7.2 do not include adjustments for uncertainties or KENO V.a code biases. Section 7.8 presents the determination of the upper limit 95/95 reactivity for the storage rack array.

7.3 Arrays of Mixed BWR Fuel Types It is shown in Tables B.1 and B.2 that the ATRIUM 10XM reference bounding lattices are more reactive in the in-rack configuration than the limiting lattices of the legacy fuel. Additionally, it is also shown in Appendix B that the other legacy fuel types have significant margin relative to the limiting lattices. It then follows that from a reactivity perspective, the reference bounding ATRIUM 10XM lattices used in this evaluation can conservatively represent past assembly fuel types.

The assembly reactivity limits (either enrichment and gadolinia limitations or direct k values) defined in Table 2.1 are applicable to all future ATRIUM 10XM fuel assemblies that will be built for the Browns Ferry reactors. Therefore, there will not be a more reactive assembly to consider in an accident scenario and an array composed of a mixture of these fuel types will not exceed the reactivity calculated for an array of limiting ATRIUM 10XM assemblies.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-4 7.4 Other Conditions The unadjusted reactivity result reported in Table 7.2 (0.897) is based upon a reference orientation which places the ATRIUM 10XM internal water channel toward the bottom right corner of the storage cell with the assembly centered within the cell as shown in Figure 6.2.

The actual position of assemblies in the storage racks will include assembly rotation and lean.

In addition, it is possible that blisters could form on the surface of the Boral plates. This deformation of the Boral plate will exclude water and therefore affect the reactivity of the storage racks. These conditions will be evaluated in this section and their worth will be included as a direct adder in the k95/95 equation.

7.4.1 Assembly Rotation The rotational combinations shown in Figure 7.1 and the simple 90°, 180°, and 270° cases were evaluated to determine if the asymmetric nature of the ATRIUM 10XM fuel assembly will produce a more reactive condition than the base case shown in Figure 6.2. The simple 90° rotation case was the most limiting with a k increase of 0.001 +/- 0.001 k. This effect will be included in the ksys parameters in the calculation of k95/95 in Section 7.8.

7.4.2 Assembly Lean Each storage cell has a hole in the bottom where the lower tie plate nose piece fits to center the assembly. There is no corresponding mechanism to keep the upper part of the assembly centered; therefore, each assembly has the ability to lean toward a side or corner of the storage cell. The impact of this lean condition was evaluated by assuming the entire bundle can be positioned anywhere within the storage cell. Between 1 and 4 assemblies were moved relative to one another within their cells. The result of this evaluation showed no statistically significant increase relative to the centered position.

7.4.3 Blister Formation Under certain conditions, corrosion gases can be trapped within a Boral plate and the aluminum cladding can be deformed to create blisters on the surface of the plate. These blister regions exclude water and can therefore affect the neutron absorption of the Boral storage rack. For this analysis a uniform 0.055" void region has been used as a conservative model of this AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-5 potential blistering condition*. Calculations indicate that this level of void on all the Boral plates in the pool would increase reactivity by 0.004 +/- 0.001 k . This effect will be included in the ksys parameters in the calculation of k95/95 in Section 7.8.

7.5 Normal Fuel Handling Normal fuel assembly movement is generally described as those movements required to load and unload assemblies into allowable storage locations within the spent fuel pool. The allowed storage locations include the spent fuel pool storage racks and the fuel preparation machines (FPMs).

Fuel movements are accomplished with the use of a refueling bridge with a mast and grapple assembly. Fuel assemblies are grasped and suspended from the mast/grapple assembly with normal lateral movements occurring above the top of the storage cell locations. The base storage array reactivity model assumes an infinite lattice array in both radial and axial dimensions using a periodic boundary condition as addressed in Section 7.2. This infinite array of fuel lattices bounds the case for suspending a single bundle over the rack during normal fuel movements. Loading or unloading an assembly into a storage location requires the raising or lowering of the fuel into the storage cell. This operation is also bound by the base storage array reactivity, which assumes the racks are fully loaded.

The spent fuel storage pool contains two FPMs that allow for the storage of a single assembly within each. Each FPM is neutronically isolated from the other so interaction between them is not considered. It is feasible that an assembly suspended from the refuel bridge can be brought into close proximity to an assembly already located in an FPM. An analysis was performed that considered the additional potential for a misplaced assembly for a total of three (3) assemblies in close proximity. These assemblies are isolated from all other fuel assemblies in the spent

• A uniform void with a 0.055 inch height bounds the condition of having a 1/8 inch high blister with a spherical cross section on every 1.25x1.25 unit cell on one side of a Boral plate (i.e., 1.25 diameter blisters with a height of 1/8 inch packed edge to edge). This in turn would be equivalent to each side of the Boral plate having blisters of this size with 50% area coverage. This conservatively bounds the results from the stainless steel clad coupon surveillances performed at Browns Ferry, on an average basis.

An additional sensitivity calculation has been performed assuming a 0.080 inch uniform void condition that results in a total reactivity increase of 0.006 +/- 0.001 k, only 0.002 k higher than the condition assumed in the k95/95 calculation.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-6 fuel pool. For this comparison, three ATRIUM 10XM fuel assemblies with REBOL lattices at 3.5 wt% U-235 were placed together in a triangular pattern. A reactivity optimization search was performed using different assembly spacing and different assembly orientations. In addition, calculations were performed with and without fuel channels and the water temperature was varied from 4 °C to 60 °C. A maximum keff of 0.897 +/- 0.001 was calculated with unchanneled assemblies at 4 °C.

By coincidence the resulting keff for this configuration is equivalent to the base array reactivity identified in Section 7.2 and used in the k95/95 calculation in Section 7.8. If a k95/95 result were calculated for the fuel handling condition using the REBOL lattices from the main calculation it would be less than the value for the limiting rack (Section 7.8) because:

• this configuration is based on 3.5 wt% U235 lattices where the REBOL lattices use a lower U-235 enrichment level (3.21 wt% U235 (Bottom) and 3.38 wt% U235 (top))

• there are no applicable accident conditions for this configuration

• the manufacturing tolerance value is lower for this application because there are no applicable storage rack, fuel channel, or gadolinia tolerance conditions Both the misloading of an assembly into a location adjacent to a loaded rack (i.e., a non-allowed storage location) and the dropping of an assembly during fuel movements (i.e., fuel handing accident) are accident conditions which are evaluated in Section 7.6.

7.6 Accident Conditions In addition to the nominal storage cell arrangement, accident conditions have also been considered. All k values provided in this section are based upon comparative KENO V.a calculations, i.e., only the most limiting scenario will be reflected in the k95/95 calculation in Section 7.8. The following scenarios were evaluated to identify the most limiting accident condition.

• Missing Boral plate in the interior of the rack. (limiting condition)

• Boral Storage Racks being forced together.

• Misloaded Bundle Scenarios.

o Assembly misloaded between the pool wall and storage rack adjacent to an open cell (no Boral between assemblies) o Assembly misloaded into the corner region adjacent to 3 racks.

o Assembly misloaded between the fuel preparation machine adjacent to an open cell (no Boral between assemblies)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-7

• Dropped assembly lying horizontally across the top of the spent fuel pool.

• Loss of Spent Fuel Pool Cooling.

The situation where a single Boral plate is missing from an interior storage rack location was evaluated. Since this was the most limiting case, the moderator temperature and assembly position were varied to optimize the worth of the accident. (The use of unchanneled assemblies was considered but was not evaluated because the calculation results indicated an optimum condition occurs when water is between the assemblies and because removal of the fuel channel tends to reduce the reactivity). The most limiting condition occurred at 4 °C with one assembly moved to the edge of the storage cell and the adjacent assembly moved half the distance to the edge of the cell as shown in Figure 7.2. This accident condition has a reactivity worth of 0.006 +/- 0.001 k. This will be included in the ksys parameters in the calculation of k95/95 in Section 7.8.

It is postulated that 2 or more Boral racks could be forced together during a seismic event. For this situation, the spacing between racks is reduced from 2 or more inches to less than 1/2 inch.

Should this occur, the pool k is calculated to increase by about 0.005 k. This accident scenario is less limiting than the optimized missing Boral plate scenario.

The case of a misloaded assembly was investigated by assuming that an assembly was placed on the edge of a Boral storage rack adjacent to an assembly in a non-tube or open storage cell.

This misloaded assembly was moved to a location very near the adjacent assembly. The results confirm that this accident scenario increases the system k by less than 0.001 k.

As shown in Figure 4.2, a misloaded assembly could be placed in a location where 3 racks meet together. With this geometry, the corner storage cells are all Boral tube cells. As expected, no significant reactivity increase is produced.

It is also possible for an assembly to be in the fuel preparation machine while a second assembly is moved between the fuel preparation machine and the fuel storage rack. This is conservatively modeled as two assemblies placed against each other adjacent to an open cell of the Boral storage rack. The results show a reactivity increase of less than 0.001 k and confirm that this accident scenario is less limiting than the missing Boral plate scenario.

For the case of dropping a fuel assembly onto an assembly in the storage rack (i.e., a fuel handling accident in the spent fuel pool), the potential exists for damaging the dropped AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-8 assembly as well as any other assemblies it contacts. This event has the potential to cause deformation to the affected assemblies, however; the reactivity impact of this deformation on rack reactivity is minimal since it involves only 2-3 assemblies in a localized area. There will also be no significant effect on the array reactivity when the dropped assembly comes to rest in a horizontal or inclined position on top of the storage rack because the dropped assembly will be neutronically isolated from the fuel in the storage cells (greater than 12 inches of water between the dropped assembly and the top of the active fuel zone of the fuel in the storage rack). Finally, similar to the previous discussion for normal fuel handling it is noted that the axial boundary condition used in the KENO model provides an infinitely repeating fuel column.

Consequently, the base model conservatively bounds the potential impact of a dropped assembly and no increase in reactivity applies for this event.

For the infinite Boral rack model, the limiting moderator temperature is 4 °C (39.2 °F).

Therefore, an increase in the pool water temperature (a loss of spent fuel pool cooling event) will not increase the reactivity of these racks.

7.7 Manufacturing and Other Uncertainties Uncertainties associated with defining bounding REBOL lattices are addressed in Appendix D.

Specifically, uncertainties associated with CASMO-4 depletion and modeling capabilities are included within the REBOL definition process (through the requirement for the lattice to have a 0.010 k higher reactivity when compared to the corresponding reference bounding lattice).

Table 7.1 demonstrates that the requirement for this adder has been met with a minimum difference of 0.0109 k for the top lattice.

The manufacturing tolerance values and the calculated reactivity uncertainties for the ATRIUM 10XM fuel are shown in Table 7.3.* The gadolinia manufacturing uncertainty (gadolinia concentration and gadolinia pellet density) effect on reactivity was evaluated with a combination of KENO V.a and CASMO-4. All other uncertainties reported in Table 7.3 were evaluated with KENO V.a. The ATRIUM 10XM rack calculations are conservatively performed for a minimum B10 areal density, therefore no manufacturing uncertainty is needed for this parameter. BOL

• The manufacturing uncertainties for other fuel types in the SFP are not explicitly addressed in this analysis due to the reactivity margin between all existing fuel and the reference bounding lattices.

See Appendix B for more detail.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-9 dimensions have been assumed, except the fuel rod pitch and channel growth results are based upon conservative spacer and channel growth dimensions.

For the various tolerances which are evaluated with KENO, the k and the standard deviation (s) values are combined consistent with the variance equation listed in Section 4.1.5 of Reference 11:

k2 = (u2/x2)((k - kref)2 +/- (sMC2 + sMC,ref2))

where: (k - kref) change in keff induced by change x on parameter x u standard uncertainty of parameter x x change in parameter x sMC Monte Carlo standard deviation values The manufacturing tolerance results have been evaluated using the upper and lower bounds of the full tolerance range; therefore, x represents a range greater than 2u. Rather than define a single uncertainty interval for this calculation and then multiply it by 2 to reestablish a 95/95 bounding interval, u2/ x2 is conservatively treated as unity in this calculation.

The Monte Carlo uncertainty values have been added to the limiting case and where (k - kref) is negative for both the upper and lower bounds of the tolerance interval, a zero value has been used (e.g., the channel thickness, pellet diameter, and Boral sheet width). The adjusted k values are the square root of the variance for that particular case. The statistically combined result is the square root of the sum of the variance values.

7.8 Determination of Maximum Rack Assembly k-eff (k95/95)

For the ATRIUM 10XM fuel design with REBOL lattice enrichments of 3.21 and 3.38 wt%

U235, the base case KENO calculated in-rack reactivity from Table 7.2 is 0.897. This k-eff value is used with the following equation to determine the upper limit 95/95 reactivity (also illustrated in Figure 2.1):

k95/95 = keff + biasm + ksys + (C2k2 + Cm2m2 + C2sys2 + ktol2)1/2, where:

keff = Base in-rack reactivity from KENO V.a, (0.897, Table 7.2)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-10 biasm = KENO V.a validation methodology bias (0.0075, Appendix C - Section C.8) ksys = Summation of applicable system variables (See the following table and Sections 7.4.1, 7.4.3, and 7.6)

C = 95% confidence level consistent with KENO V.a (2.0)

Cm = 95/95 one-sided tolerance multiplier for a sample size of 68 (1.996) k = k-eff standard deviation from KENO V.a, (0.001, Table 7.2) m = KENO V.a methodology uncertainty (0.0027, Appendix C - Section C.8) sys = (sys12 + sys22 + sys_n2)1/2, for ksys uncertainties. (See the following table and Sections 7.4.1, 7.4.3, and 7.6) ktol = Statistical combination of manufacturing reactivity uncertainties ( [ ],

Table 7.3)

The following table provides a summary of the ksys and sys parameters applicable to this analysis. (The values are standard deviation results from KENO).

Description ksys sys Assembly Rotation Effects (Section 7.4.1) 0.001 0.001 Boral Blisters (Section 7.4.3) 0.004 0.001 Limiting Accident (Missing Insert, Section 7.6) 0.006 0.001 Combined Values 0.011 0.0017 The standard deviations and tolerance uncertainties are included as the square root of the sum of the squares since they represent independent events. Solving for k95/95 yields a 95/95 upper limit k-eff that is larger than 0.927 so it is rounded-up to 0.928. The above determination of the upper limit 95/95 k-eff is consistent with the method documented in Reference 6 and allows one to state that at least 95% of the normal population is less than the 95/95 k-eff value calculated with a 95% confidence.

The results demonstrate the postulated configuration with the ATRIUM 10XM REBOL lattices meets the NRC criticality safety acceptance criterion that the array k-eff under the worst credible conditions is < 0.95.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-11 Table 7.1 Summary of CASMO-4 Maximum In-Rack Reactivity Results Reference Bounding Lattices ATRIUM 10XM geometry, top and bottom lattice geometries explicitly modeled 4.70 wt% U-235 uniform enrichment distribution above [ ]

4.70 wt% U-235 uniform enrichment distribution at and below [ ]

8 gadolinia rods with 3.5 wt% Gd2O3 above [ ]

8 gadolinia rods with 3.919 wt% Gd2O3 at and below [ ]

Standard 100 mil fuel channel Reflective boundary condition for in-core calculations No xenon or lumped fission products for in-rack calculations Periodic boundary condition for in-rack calculations Condition Top Lattice Bottom Lattice Maximum In-Rack k, 4°C (39.2°F) 0.8825 0.8825 Exposure, GWd/MTU 10.5 11.5 Void History [ ] [ ]

REBOL Lattices ATRIUM 10XM geometry, top and bottom lattice geometries explicitly modeled 3.38 wt% U-235 uniform enrichment distribution above [ ]

3.21 wt% U-235 uniform enrichment distribution at and below [ ]

No gadolinia Standard 100 mil fuel channel BOL (zero exposure, no xenon, no fission products)

Periodic boundary condition Condition Top Lattice Bottom Lattice Maximum In-Rack k, 4°C (39.2°F) 0.8934 0.8935 Margin to Reference Bounding Lattice 0.0109 0.0110 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-12 Table 7.2 Summary of KENO V.a Maximum In-Rack Reactivity Results Fuel Assembly ATRIUM 10XM geometry, top and bottom lattice geometries explicitly modeled 3.38 wt% U-235 uniform enrichment distribution above [ ]

3.21 wt% U-235 uniform enrichment distribution at and below [ ]

No gadolinia Standard 100 mil Channel BOL (zero exposure, no xenon, no fission products)

Periodic boundary conditions for in-rack x-y plane and the axial direction Storage Array Configuration Explicit 2x2 rack model with infinite periodic boundary conditions Assembly centered in cell water volume 4°C moderator and fuel temperatures Description k-eff In-Rack 4°C (39.2°F) k-eff 0.897 +/- 0.001 Maximum k95/95 Reactivity (including uncertainties, biases, manufacturing tolerances 0.928 and worst accident or abnormal loading conditions)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-13 Table 7.3 Manufacturing Reactivity Uncertainties (Based upon BOL conditions using KENO V.a except as noted.)

[

]

• This is a conservative approximation of the spacer growth at peak reactivity exposures.

This is a lifetime maximum value and is assigned to each side of the fuel channel, [

].

Depletion based adders of [ ] have been added to the gad concentration and gad density cases, respectively.

§ This calculation was performed using the minimum value so no manufacturing uncertainty is required, see discussion in Section 7.3.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-14 Figure 7.1 Evaluated Assembly Rotation Cases AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 7-15 Figure 7.2 Limiting Accident (Missing Boral Plate)

(Note that assembly positions have been shifted to maximize the worth of this accident condition. The missing Boral plate is located between the center-top and center cell in the figure.)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page 8-1 8.0 References

1. Tennessee Valley Authority Docket Nos. 50-259, 50-260 and 50-296 Tennessee Valley Authority Notice of Issuance of Amendment, Browns Ferry Units 1, 2, and 3 License Amendments 42, 39, and 16, Authorizing You to Increase Storage Capacity of Each Spent Fuel Pool, September 21, 1978. (Adams # ML020040269)

2. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 9.1.1 Revision 3 (Criticality Safety of Fresh and Spent Fuel Storage and Handling), U.S. Nuclear Regulatory Commission, March 2007.

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

4. Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors, ANSI/ANS American National Standard 8.17-1984, American Nuclear Society, January 1984, (withdrawn 2004).

5. Letter, Brian K. Grimes, Assistant Director for Engineering and Projects Division of Operating Reactors, U.S. Nuclear Regulatory Commission, to All Power Reactor Licensees, OT Position for the Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14, 1978, as amended by letter January 18, 1979.

6. Letter, Laurence Kopp (Reactor Systems Branch, NRC) to Timothy Collins, Chief (Reactor Systems Branch-NRC),

Subject:

Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, August 19, 1998.

7. Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01 Revision 0, Staff Guidance Regarding The Nuclear Criticality Safety Analysis For Spent Fuel Pools, (ADAMS # ML110620086).

8. NUREG/CR-0200 Revision 6, SCALE Version 4.4 A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, Oak Ridge National Laboratory, May 2000.

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

10. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.

11. ICSBEP Guide to the Expression of Uncertainties, Revision 5, V. F. Dean, September 30, 2008. {Distributed with the International Handbook of Evaluated Criticality Safety Benchmark Experiments, Nuclear Energy Agency, NEA/NSC/DOC(95)03, September 2009 Edition.}

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page A-1 Appendix A Sample CASMO-4 Input Tables A.1 and A.2 provide the CASMO-4 spent fuel storage rack model for the reference bounding lattices defined in this analysis.

ATRIUM 10XM fuel which does not conform to the enrichment and gadolinia requirements described in Table 2.1 can be analyzed for storage in the spent fuel storage racks by adapting the CASMO-4 sample inputs presented in Table A.1 or A.2. Evaluations should be performed with [ ] depletion for bottom geometry lattices and [ ]

depletion for top geometry lattices. These calculations will be performed with the NRC approved CASMO-4 code described in EMF-2158(P)(A), (Reference 10 of the main report).

If the lifetime maximum in-rack k of the new lattices are less than the k of the corresponding reference bounding lattices (0.8825), the ATRIUM 10XM fuel assembly can be safely stored in the Browns Ferry Nuclear Plant spent fuel storage racks.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page A-2 Table A.1 CASMO-4 Input for ATRIUM 10XM Top Reference Bounding Lattice

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page A-3 Table A.2 CASMO-4 Input for ATRIUM 10XM Bottom Reference Bounding Lattice

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-1 Appendix B Reactivity Comparison for Assemblies Used in the Browns Ferry Reactors All previously manufactured assemblies used in the Browns Ferry Units 1, 2, and 3 reactors have been evaluated to determine the most limiting lattices on the basis of highest in-rack k.

This reactivity evaluation included an initial screening that resulted in the identification of a set of the most limiting lattices which is detailed in Section B.2 of this Appendix. The resulting limiting lattices were then used to establish a reference bounding lattice for each geometry zone as well as a corresponding REBOL lattice that is used as the basis for the KENO V.a criticality analysis.

Section B.1 provides a comparison of the resulting limiting lattices and their corresponding reference bounding lattices and REBOL lattices.

B.1: Summary of Lattice In-Rack Reactivity Comparisons The screening detailed in Section B.2 of this appendix resulted in the selection of the highest reactivity previously manufactured (or as-fabricated, or legacy) lattices based upon calculated CASMO-4 in-rack k values. These limiting as-fabricated lattices are compared to the corresponding ATRIUM 10XM reference bounding lattice and REBOL lattice in Table B.1. This comparison shows that the ATRIUM 10XM reference bounding lattices described in Table 7.1 are more reactive than any of the previously manufactured lattices used in the Browns Ferry reactors. It also shows that the REBOL lattices defined for use in the KENO V.a calculations are more reactive than the reference bounding lattices.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-2 Table B.1 Lattice Reactivity Comparisons (REBOL, Bounding, and Limiting)

Maximum In-Rack k Case Description Lattice Description (CASMO-4 @ 4 °C)

Top Lattices REBOL, Top Lattice XMLCT-3.38 (no Gad) 0.8934

[ ]

Reference Bounding Top Lattice XMLCT-470UL-8G35 0.8797*

[ ]

Limiting As-Fabricated Top Lattice GE14 (From Table B.2) 0.8619

[ ]

Margin to Reference Bounding Lattice (Reference - Limiting) 0.0178 k Bottom Lattices REBOL, Bottom Lattice XMLCB-3.21 (no Gad) 0.8935

[ ]

Reference Bounding Bottom XMLCB-470UL-8G3919 0.8790*

Lattice [ ]

Limiting As-Fabricated Bottom GE13 (From Table B.2) 0.8227 Lattice [ ]

Margin to Reference Bounding Lattice (Reference - Limiting) 0.0563 k

• For direct comparison with the legacy fuel, this value also includes the effects of lumped fission products. Without lumped fission products the k value increases to 0.8825 as reported in Table 7.1.

Lattice descriptions for non-AREVA supplied fuel are not provided in this document since they have been identified as proprietary by that vendor.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-3 Table B.2 Limiting Lattices In-Rack Kinf Comparison (By Product Line)

Maximum Margin to In-Rack k 10XM Case Array (CASMO-4 Bounding

@ 4 °C) Lattice (k)

Limiting Top Lattices by Product Line [ ]

ATRIUM 10XM TRBL 10x10 0.8797 ---

ATRIUM-10 TRBL 10x10 0.8747 0.0050 ATRIUM-10 10x10 0.8520 0.0277 GE14 10x10 0.8619 0.0178 GE13 9x9 0.8254 0.0543 GE11 9x9 0.8237 0.0560 GE9B 8x8 0.8114 0.0683 GE7B 8x8 0.7915 0.0882 Limiting Bottom Lattices by Product Line [ ]

ATRIUM 10XM BRBL 10x10 0.8790 ---

ATRIUM-10 BRBL 10x10 0.8738 0.0052 ATRIUM -10 10x10 0.8121 0.0669 GE14 10x10 0.8086 0.0704 GE13 9x9 0.8227 0.0563 GE11 9x9 0.8214 0.0576 GE9B 8x8 0.8114 0.0676 GE7B 8x8 0.7915 0.0875 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-4 B.2: Previously Manufactured Lattices Browns Ferry Units 1, 2, and 3 have loaded a number of different product lines including GE 7x7 (GE2 and GE3), GE 8x8 (GE4, GE5, GE6, GE7/7B, and GE9B), GE 9x9 (GE11 and GE13), GE 10x10 (GE14), and AREVA 10x10 (ATRIUM-10) fuel. Several variations of LUAs have also been loaded which include variants of older designs and a set of 4 Westinghouse QUAD+

assemblies.

Initial operation of the Browns Ferry reactors transitioned from initial 12 month nominal cycle lengths to 18 month cycles and finally to the current 24 month cycles. The later cycles have operated at 105% of the original licensed thermal power level. As a consequence of the movement towards longer cycles and higher operating power levels, the fuel designs have transitioned to designs with higher U-235 enrichments and Gadolinia loading. The general trend is that the later higher enrichment fuel designs bound the earlier low enrichment designs.*

The Unit 1 and 2 spent fuel pools at the Brown Ferry Nuclear Plant include a transfer canal that allows assemblies from one pool to be moved to the other. The Unit 3 spent fuel pool is not connected to either of the other pools. To simplify this analysis the limiting lattices will be based upon the inventory of all three spent fuel pools.

B.2.1: Current Inventory and Initial Screening The previously manufactured fuel inventory is summarized in Table B.3 and Table B.4. An initial screening of the previously manufactured fuel assemblies was performed based upon U235 enrichment and gadolinia loading. It was determined that explicit calculation of in-rack k is not required for lattices with gadolinia and with initial peak average enrichment at or below the lowest REBOL lattice enrichment of 3.21 wt% U-235. These fuel assemblies are listed in standard font in Table B.3. This criterion is based upon the recognition that a lattice without gadolinia (such as the REBOL lattices) will always exhibit a higher reactivity than a lattice having the same enrichment and also containing gadolinia (this condition is illustrated in Figure D.4 of Appendix D). While it is noted that the application of enrichment only screening does not specifically address changes in lattice geometry (i.e., 7x7 or 8x8 versus later 9x9 and 10x10

• One exception to this trend is encountered with the introduction of BLEU fuel since the presence of U236 acts as a neutron poison and reduces lattice reactivity when compared to an equivalent enrichment commercial grade uranium lattice. See section 6.4 in the main body of the report for more detail.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-5 designs), these lattice geometry impacts are small compared to the conservatism established with the non-gadolinia REBOL lattice, as discussed above. Furthermore, Table B.5 shows that similar lattices with higher U235 enrichment levels have significant reactivity margin to the limiting lattices. Application of this criterion immediately screens* out the GE 7x7 fuel and many of the older GE 8x8 assemblies. All bundles in bold type in Table B.3 are analyzed explicitly.

As summarized in Table B.4, the previously loaded ATRIUM-10 fuel includes both commercial grade uranium (CGU) and blended low enriched uranium (BLEU) assemblies. These previously manufactured lattices have been explicitly modeled with the U-234 and U-236 content associated with the fabrication of the specific reload batch.

B.2.2: Explicit Reactivity Evaluations Tables B.5 and B.6 provide a comparison of the peak in-rack CASMO-4 k values of the lattices in the bundles selected for explicit evaluation in the previous section. The axial zones are separated at [ ].

The ATRIUM-10 fuel product line is the fuel currently being loaded in reload quantities in Units 2 and 3 and planned for use in Unit 1. These and other previously used designs are stored in the Browns Ferry spent fuel storage pools. The reference bounding ATRIUM 10XM assembly has a higher in-rack reactivity than all previously loaded fuel assembly designs, including the ATRIUM-10 reference bounding lattice developed in Reference B.1 . As such, the ATRIUM 10XM reference bounding assembly design forms the basis for demonstrating that the maximum k-eff of the spent fuel pool storage array remains less than 0.95.

The calculations documented herein demonstrate that the ATRIUM 10XM reference bounding assembly design has been selected to be more reactive in an in-rack configuration than any of the current or past fuel assembly designs used in the Browns Ferry reactors. From Table B.2,

• A conservative screening of 2.99% U-235 was actually used. An exception to the screening was a group of four (4) LUAs slightly above the 3.1% U-235 criteria. This design was screened out since only four of these LUAs exist and the enrichment is slightly below the 3.21% enrichment screening criteria. Also, as noted in the text above, the presence of gadolinia in these lattices will significantly reduce the peak in-rack reactivity when compared to the REBOL non-gad lattices.

[ ]

Therefore, future ATRIUM-10 assemblies that meet the storage requirements of Reference B.1 will be less reactive than the ATRIUM 10XM reference bounding lattices.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-6 the limiting case is a GE14 top lattice that has been demonstrated to be 0.0178 k less reactive than the top zone reference bounding lattice. Table B.5 shows that this limiting lattice is much more reactive than the other lattices in the same bundle - indicating that there is more margin available on an assembly basis than is apparent with this lattice comparison.

These in-rack lattice k-infinity comparisons are based upon actual GE 8x8, GE 9x9, GE 10x10, and ATRIUM-10 lattice geometries and enrichment distributions. This evaluation establishes that the fuel assemblies previously manufactured for use in the Browns Ferry reactors can be safely stored in the Browns Ferry spent fuel storage pools. This evaluation also shows that future ATRIUM 10XM assemblies meeting the storage requirements established by this criticality analysis can be safely stored with these previously manufactured assemblies.

B.2.3: Evaluation of Modified, Abnormal, and Damaged Assemblies The preceding evaluation of previously supplied fuel is based upon nominal assembly designs.

The potential exists that assemblies that have been previously damaged or modified may have a configuration that is more reactive than the nominal design.

One of the primary issues that could affect in-rack reactivity is the removal of one or more fuel rods from an assembly. Since the lattices are under moderated in the in-rack configuration, removal of a fuel rod without replacement would introduce additional moderator which could result in an increase in the lattice reactivity. This would also apply to the case where a rod has been broken and a portion of the broken fuel rod removed from the bundle. Assemblies in which a fuel rod has been replaced with either an inert rod or a natural uranium rod would represent a reduction in reactivity from the nominal design since the fissile material content is reduced and no change in the amount of moderator would occur.

The existing inventory of fuel in the Browns Ferry spent fuel pools includes a number of bundles that have experienced damage of one kind or another. The following is a summary of the status of this fuel:

• Missing Fuel Rods: No fuel assemblies in the Unit 1, 2, or 3 spent fuel pools have a missing fuel rod (i.e. without an inert or natural rod replacement).

• Replacement Fuel Rods: A total of eight fuel assemblies have had either one or two pins that have been removed and have been replaced with an inert stainless steel rod as detailed in Table B.7.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-7

• Broken Fuel Rods: Four bundles have been identified with broken fuel rods as detailed in Table B.7. The affected rods were either not disturbed or were returned to the bundle. Consequently, the amount of moderator and fissile material is maintained and there is no adverse impact on in-rack reactivity versus the nominal design.

Another type of modification is related to reconstitution of a bundle using replacement rods from other donor assemblies. A reconstitution campaign at Browns Ferry was completed for Unit 2 Cycle 6 that affected a number of twice and thrice burnt assemblies.

204 8DRB284L / P8DRB284L (U2 Cycle 3 / U2 Cycle 4) assemblies reconstituted 60 8DRB284L / P8DRB284L (U2 Cycle 3 / U2 Cycle 4) donor assemblies Therefore, a total of 264 high burnup assemblies were modified including a combination of the reconstituted and donor assemblies. The following items of interest are noted regarding these assemblies:

• The nominal enrichment of 2.84 wt% U-235 and corresponding planar enrichment of 3.08 wt% U-235 of the affected assemblies is below the previously applied screening criteria. Table B.2 provides the results for a GE7 lattice with a similar geometry and higher enrichment that still shows >0.080 k margin to the reference bounding lattices.

• The rod replacements were based upon swaps with rods of similar burnup and reactivity between the donor and reconstituted assemblies. Criteria for the selection of replacement rods were established to minimize the impact on hot in-core k-infinity and local peaking within the assembly lattices; this would also tend to minimize impact on in-rack reactivity.

• None of the reconstituted or donor assemblies were left with open fuel rod positions (i.e. all rod positions contained a fuel rod in the modified assemblies).

It is also noted that all of the reconstituted or donor assemblies are high burnup (twice or thrice burnt) and are therefore well past their peak reactivity condition. Of the reconstituted assemblies, 212 were then loaded and operated for another cycle which further reduces their actual reactivity.

Based upon the above discussion, all of the assemblies identified as modified or damaged are appropriately evaluated at nominal conditions. That is the assembly modifications do not provide a significant adverse impact on in-rack reactivity that could potentially make them limiting reactivity assemblies.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-8 B.3: References B.1 ANP-2945(P) Revision 1, Browns Ferry Nuclear Plants Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis, AREVA NP, July 2011.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-9 Table B.3 Non-AREVA Supplied Fuel Inventory Nominal Product Bundle Design Name Array enrichment Line Unit 1: Spent Fuel Pool GE2-7DB110 1.1 GE2 7x7 GE14-P10DNAB147-NOG-1T-150-T6-2893 1.47 GE14 10x10 GE13-P9DTB156-NOG-100T-146-T6 1.56 GE13 9x9 GE14-P10DNAB157-NOG-1T-150-T6-2889 1.57 GE14 10x10 GE13-P9DTB163-NOG-100T-146-T6 1.63 GE13 9x9 GE14-P10DNAB200-3GZ-1T-150-T-2603 2.0 GE14 10x10 GE3-7DB250-3G3/2G4 2.5 GE3 7x7 GE3-7DB250-4G3 2.5 GE3 7x7 GE5-8DRB265-6G2 2.65 GE5 8x8 GE5-8DRB265-6G3 2.65 GE5 8x8 GE6-P8DRB265-6G2 2.65 GE6 8x8 GE6-P8DRB265-6G3 2.65 GE6 8x8 GE4-8DB274-5G2 2.74 GE4 8x8 GE4-8DB274-5G3 2.74 GE4 8x8 GE6-P8DRB284-4G4/3G2 2.84 GE6 8x8 GE6-P8DRB284-7GZ 2.84 GE6 8x8 GE7-P8DRB284-4G4/3G2 2.84 GE7 8x8 LTA-P8DRB284-GZLTA1 2.84 LTA 8x8 LTA-P8DRB284-GZLTA2 2.84 LTA 8x8 GE14-P10DNAB377-16GZ-1T-150-T6-2890 3.77 GE14 10x10 GE13-P9DTB391-13GZ-100T-146-T 3.91 GE13 9x9 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-10 Table B.3 Non-AREVA Supplied Fuel Inventory (Continued)

Nominal Product Bundle Design Name Array enrichment Line Unit 1: In-Core GE14-P10DNAB157-NOG-100T-150-T6 1.57 GE14 10x10 GE14-P10DNAB350-16GZ-100T-150-T6 3.50 GE14 10x10 GE14-P10DNAB368-15GZ-100T-150-T6 3.68 GE14 10x10 GE14-P10DNAB377-16GZ-100T-150-T6 3.77 GE14 10x10 GE14-P10DNAB377-17GZ-100T-150-T6 3.77 GE14 10x10 GE14-P10DNAB400-17GZ-100T-150-T6 4.00 GE14 10x10 GE14-P10DNAB402-16GZ-100T-150-T6 4.02 GE14 10x10 GE14-P10DNAB402-19GZ-100T-150-T6 4.02 GE14 10x10 GE14-P10DNAB404-15GZ-100T-150-T6 4.04 GE14 10x10 GE14-P10DNAB406-16GZ-100T-150-T6 4.06 GE14 10x10 GE14-P10DNAB406-15GZ-100T-150-T6 4.06 GE14 10x10 GE14-P10DNAB408-16GZ-100T-150-T6 4.08 GE14 10x10 GE14-P10DNAB408-17GZ-100T-150-T6 4.08 GE14 10x10 GE14-P10DNAB412-16GZ-100T-150-T6 4.12 GE14 10x10 GE14-P10DNAB417-16GZ-100T-150-T6 4.17 GE14 10x10 GE14-P10DNAB418-16GZ-100T-150-T6 4.18 GE14 10x10 GE14-P10DNAB419-16GZ-100T-150-T6 4.19 GE14 10x10 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-11 Table B.3 Non-AREVA Supplied Fuel Inventory (Continued)

Nominal Product Bundle Design Name Array enrichment Line Unit 2: Spent Fuel Pool GE2-7DB110 1.1 GE2 7x7 GE3-7DB250-3G3/2G4 2.5 GE3 7x7 GE3-7DB250-4G3 2.5 GE3 7x7 GE6-P8DRB265-6G3 2.65 GE6 8x8 LTA-QUAD+270-7G5 2.7 LTA 4-4x4 GE4-8DB274-5G2 2.74 GE4 8x8 GE4-8DB274-5G3 2.74 GE4 8x8 GE5-8DRB284-4G4/3G2 2.84 GE5 8x8 GE6-P8DRB284-4G4/3G2 2.84 GE6 8x8 GE6-P8DRB284-7GZ 2.84 GE6 8x8 GE7-P8DRB284-4G4/3G2 2.84 GE7 8x8 GE7B-BP8DRB299-7G4

• 2.99 GE7B 8x8 GE7B-BP8DRB301-6G4

• 3.01 GE7B 8x8 GE9B-P8DWB319-9GZ-80M-150-T 3.19 GE9B 8x8 GE9B-P8DWB325-10GZ-80M-150-T 3.25 GE9B 8x8 GE9B-P8DWB326-7GZ-80M-150-T 3.26 GE9B 8x8 GE11-P9HUB366-12G4.0-100T-146-T 3.66 GE11 9x9 GE11-P9HUB367-14GZ-100T-146-T 3.67 GE11 9x9 GE14-P10DNAB367-14GZ-1T-150-T-2602 3.67 GE14 10x10 GE13-P9HTB384-12G4.0-100T-146-T 3.84 GE13 9x9 GE13-P9DTB391-13GZ-100T-146-T 3.91 GE13 9x9 GE13-P9DTB401-14GZ-100T-146-T 4.01 GE13 9x9 GE13-P9DTB406-13GZ-100T-146-T 4.06 GE13 9x9 GE13-P9DTB412-2G7.0/11G5.0-1T-146-T 4.12 GE13 9x9 GE14-P10DNAB416-16GZ-1T-150-T-2600 4.16 GE14 10x10 GE14-P10DNAB416-16GZ-1T-150-T-2601 4.16 GE14 10x10 GE14-P10DNAB416-18GZ-1T-150-T-2627 4.16 GE14 10x10 GE14-P10DNAB417-18GZ-1T-150-T-2628 4.17 GE14 10x10

• The GE7 geometry provides a representation of earlier 8x8 designs which used multiple water rods unlike the large single water rod in the GE9B design.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-12 Table B.3 Non-AREVA Supplied Fuel Inventory (Continued)

Nominal Product Bundle Design Name Array enrichment Line Unit 3: Spent Fuel Pool GE4-8DB219-4G4/2G2.5/1G1.5 2.19 GE4 8x8 GE4-8DB219-2G2.5/1G1.5 2.19 GE4 8x8 GE6-P8DRB265-6G2 2.65 GE6 8x8 GE5-8DRB265-6G2 2.65 GE5 8x8 LTA-P8DRB284-7GZLTA 2.84 LTA 8x8 GE7B-BP8DRB284-4G4/3G2 2.84 GE7B 8x8 GE6-P8DRB284-7GZ 2.84 GE6 8x8 GE6-P8DRB299-7G4 2.99 GE6 8x8 LTA-P8DRB314-GZLTA

• 3.14 LTA 8x8 GE11-P9HUB323-8G4.0-100T-146-T 3.23 GE11 9x9 GE11-P9HUB323-5G5/4G4-100T-146-T 3.23 GE11 9x9 GE11-P9HUB325-14GZ-100T-146-T 3.25 GE11 9x9 GE13-P9HTB372-11GZ-100T-146-T 3.72 GE13 9x9 GE13-P9HTB386-12GZ-100T-146-T 3.86 GE13 9x9 GE13-P9DTB400-13GZ1-100T-146-T 4.00 GE13 9x9 GE14-P10DNAB401-14GZ-1T-150-T-2514 4.01 GE14 10x10 GE14-P10DNAB402-15GZ-1T-150-T-2513 4.02 GE14 10x10 GE13-P9DTB414-15GZ-100T-146-T 4.14 GE13 9x9

• This design was screened out since only four of these LUAs exist and the enrichment is slightly below the 3.21% enrichment screening criteria. It is also noted that the presence of gadolinia in these lattices will significantly reduce the peak in-rack reactivity when compared to the REBOL non-gad lattices.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-13 Table B.4 AREVA Supplied Fuel Inventory Max Planar Reload Bundle Design Name Type Enrichment Unit 3 A10-3813B-13GV80 4.157 A10-4077B-15GV80 4.381 BFE3-12 CGU A10-4088B-13GV80 4.393 A10-1623B-5GV80 1.739 A10-4171B-14GV80-FCB 4.483 BFE3-13 A10-4163B-16GV80-FCB 4.477 BLEU A10-4181B-13GV80-FCB 4.495 A10-4218B-15GV80-FCC 4.557 BFE3-14 BLEU A10-4218B-13GV80-FCC 4.543 A10-3831B-15GV80-FCD 4.184 A10-3403B-9GV80-FCD 3.739 A10-3392B-10GV80-FCD 3.724 BFE3-15 BLEU A10-4218B-15GV80-FCC* 4.557 A10-4218B-13GV80-FCC* 4.543 A10-3757B-10GV80-FCC* 3.997 A10-3440B-11GV80-FCE 3.668 A10-3826B-13GV80-FCE 4.111 BFE3-16 BLEU A10-4075B-13GV80-FCE 4.384 A10-4081B-12GV80-FCE 4.391

• This design was originally fabricated as part of the BFE3-14 reload but was used in both Unit 2 and Unit 3.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-14 Table B.4 AREVA Supplied Fuel Inventory (Continued)

Max Planar Reload Bundle Design Name Type Enrichment Unit 2 BFE2-14 A10-3920B-14GV70 4.245 BLEU A10-4227B-15GV80-FBB 4.555 BFE2-15 A10-4239B-15GV80-FBB 4.555 BLEU A10-3552B-10GV80-FBB 3.937 A10-4019B-14GV80-FBC 4.318 A10-3841B-14GV80-FBC 4.121 BFE2-16 BLEU A10-4218B-13GV80-FCC* 4.543 A10-3757B-10GV80-FCC* 3.997 A10-3799B-14GV80-FBD 4.157 BFE2-17 BLEU A10-4004B-15GV80-FBD 4.303 Unit 1 A10-3562B-14GV80 FAA 4.059 BFE1-10 A10-3676B-10GV80 FAA 4.109 BLEU A10-4111B-15GV80 FAA 4.424 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-15 Table B.5 Non-AREVA Supplied Fuel Limiting Bundles In-Rack Reactivity Comparison (by product line)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Bundle Bottom Top Depletion Depletion Depletion GE7 - 8x8 Array GE7B-BP8DRB299-7G4 1 X X 0.7838 0.7853 0.7869 GE7B-BP8DRB301-6G4 1 X X 0.7853 0.7878 0.7915 Max. Top and Bottom Lattice 0.7853 0.7878 0.7915 GE9 - 8x8 Array 1 X 0.7853 0.7877 0.7906 2 X X 0.8085 0.8094 0.8100 GE9B-P8DWB325-10GZ-80M-150-T 3 X 0.8079 0.8086 0.8086 4 X 0.8004 0.8018 0.8031 1 X 0.7855 0.7879 0.7908 GE9B-P8DWB326-7GZ-80M-150-T 2 X X 0.8101 0.8110 0.8114 3 X 0.8007 0.8021 0.8034 1 X 0.7843 0.7866 0.7891 GE9B-P8DWB319-9GZ-80M-150-T 2 X X 0.8019 0.8036 0.8047 3 X 0.7930 0.7953 0.7972 Max. Bottom Lattice 0.8101 0.8110 0.8114 Max. Top Lattice 0.8101 0.8110 0.8114 GE11 - 9x9 Array 1 X X 0.8214 0.8205 0.8185 GE11-P9HUB366-12G4.0-100T-146-T 2 X 0.8237 0.8233 0.8218 1 X X 0.8214 0.8205 0.8185 GE11-P9HUB367-14GZ-100T-146-T 2 X 0.8136 0.8125 0.8105 3 X 0.8237 0.8233 0.8218 1 X X 0.7997 0.7999 0.8001 GE11-P9HUB325-14GZ-100T-146-T 2 X 0.7905 0.7910 0.7914 3 X 0.8004 0.8012 0.8020 1 X X 0.7939 0.7963 0.7988 GE11-P9HUB323-5G5/4G4-100T-146-T 2 X 0.7947 0.7982 0.8016 1 X X 0.8050 0.8064 0.8079 GE11-P9HUB323-8G4.0-100T-146-T 2 X 0.8074 0.8084 0.8094 Max. Bottom Lattice 0.8214 0.8205 0.8185 Max. Top Lattice 0.8237 0.8233 0.8218 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-16 Table B.5 Non-AREVA Supplied Fuel Limiting Bundles In-Rack Reactivity Comparison (by product line) - (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Bundle Bottom Top Depletion Depletion Depletion GE13 - 9x9 Array 1 X X 0.8227 0.8210 0.8182 GE13-P9HTB384-12G40-100T-146-T 2 X 0.8254 0.8246 0.8231 1 X 0.8051 0.8041 0.8007 GE13-P9DTB406-13GZ-100T-146-T 2 X X 0.8163 0.8142 0.8099 3 X 0.8199 0.8185 0.8154 1 X 0.7849 0.7840 0.7814 GE13-P9DTB401-14GZ-100T-146-T 2 X X 0.7980 0.7966 0.7933 3 X 0.8004 0.8000 0.7980 1 X 0.7816 0.7819 0.7815 GE13-P9DTB391-13GZ-100T-146-T 2 X X 0.8029 0.8025 0.8009 3 X 0.8056 0.8059 0.8053 1 X X 0.8132 0.8114 0.8075 GE13-P9DTB412-2G7l11G5-1T-146-T 2 X 0.8156 0.8144 0.8114 1 X 0.8019 0.8020 0.8014 GE13-P9HTB372-11GZ-100T-146-T 2 X X 0.8021 0.8021 0.8016 3 X 0.8027 0.8038 0.8048 1 X 0.8193 0.8184 0.8158 GE13-P9HTB386-12GZ-100T-146-T 2 X X 0.8198 0.8187 0.8160 3 X 0.8217 0.8216 0.8204 1 X 0.7965 0.7956 0.7931 GE13-P9DTB400-13GZ1-100T-146-T 2 X X 0.8043 0.8031 0.8003 3 X 0.8070 0.8063 0.8042 1 X 0.8011 0.7990 0.7945 GE13-P9DTB414-15GZ-100T-146-T 2 X X 0.8135 0.8117 0.8073 3 X 0.8178 0.8163 0.8123 Max. Bottom Lattice 0.8227 0.8210 0.8182 Max. Top Lattice 0.8254 0.8246 0.8231 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-17 Table B.5 Non-AREVA Supplied Fuel Limiting Bundles In-Rack Reactivity Comparison (by product line) - (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Bundle Bottom Top Depletion Depletion Depletion GE14 - 10x10 Array 1 X 0.7691 0.7707 0.7720 2 X X 0.7816 0.7852 0.7885 GE14-P10DNAB377-16GZ-100T-150-T6 3 X 0.7613 0.7652 0.7696 4 X 0.7825 0.7862 0.7904 5 X 0.8578 0.8583 0.8579 1 X 0.7829 0.7817 0.7790 2 X X 0.7873 0.7874 0.7866 GE14-P10DNAB402-16GZ-100T-150-T6 3 X 0.7663 0.7680 0.7692 4 X 0.7879 0.7896 0.7906 5 X 0.8605 0.8596 0.8574 1 X 0.7592 0.7643 0.7703 2 X X 0.7718 0.7776 0.7842 GE14-P10DNAB350-16GZ-100T-150-T6 3 X 0.7504 0.7579 0.7663 4 X 0.7698 0.7776 0.7871 5 X 0.8466 0.8470 0.8467 1 X 0.7923 0.7904 0.7863 2 X X 0.7969 0.7963 0.7942 GE14-P10DNAB419-16GZ-100T-150-T6 3 X 0.7786 0.7796 0.7792 4 X 0.8009 0.8020 0.8013 5 X 0.8539 0.8530 0.8505 1 X 0.7633 0.7657 0.7680 2 X X 0.7763 0.7805 0.7846 GE14-P10DNAB368-15GZ-100T-150-T6 3 X 0.7551 0.7598 0.7650 4 X 0.7752 0.7798 0.7854 5 X 0.8502 0.8508 0.8509 1 X 0.7851 0.7831 0.7793 2 X X 0.7839 0.7828 0.7803 GE14-P10DNAB402-19GZ-100T-150-T6 3 X 0.7637 0.7647 0.7648 4 X 0.7854 0.7863 0.7863 5 X 0.8619 0.8608 0.8580 1 X 0.7692 0.7707 0.7717 2 X X 0.7816 0.7852 0.7885 GE14-P10DNAB377-17GZ-100T-150-T6 3 X 0.7613 0.7652 0.7696 4 X 0.7825 0.7862 0.7904 5 X 0.8578 0.8583 0.8579 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-18 Table B.5 Non-AREVA Supplied Fuel Limiting Bundles In-Rack Reactivity Comparison (by product line) - (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Bundle Bottom Top Depletion Depletion Depletion GE14 - 10x10 Array (Continued) 1 X 0.7650 0.7644 0.7618 2 X X 0.7649 0.7648 0.7633 GE14-P10DNAB406-16GZ-100T-150-T6 3 X 0.7441 0.7463 0.7481 4 X 0.7646 0.7669 0.7689 1 X 0.7635 0.7632 0.7608 2 X X 0.7633 0.7637 0.7626 GE14-P10DNAB406-15GZ-100T-150-T6 3 X 0.7414 0.7442 0.7466 4 X 0.7614 0.7645 0.7673 1 X 0.7723 0.7711 0.7677 2 X X 0.7705 0.7701 0.7677 GE14-P10DNAB418-16GZ-100T-150-T6 3 X 0.7532 0.7548 0.7552 4 X 0.7741 0.7760 0.7764 1 X 0.7694 0.7694 0.7686 2 X X 0.7687 0.7697 0.7708 GE14-P10DNAB400-17GZ-100T-150-T6 3 X 0.7407 0.7435 0.7464 4 X 0.7612 0.7639 0.7669 1 X 0.7721 0.7709 0.7676 2 X X 0.7703 0.7699 0.7676 GE14-P10DNAB417-16GZ-100T-150-T6 3 X 0.7838 0.7835 0.7816 4 X 0.8061 0.8059 0.8040 1 X 0.7667 0.7660 0.7636 2 X X 0.7664 0.7662 0.7647 GE14-P10DNAB408-16GZ-100T-150-T6 3 X 0.7461 0.7481 0.7494 4 X 0.7668 0.7689 0.7703 1 X 0.7684 0.7675 0.7649 2 X X 0.7670 0.7667 0.7650 GE14-P10DNAB412-16GZ-100T-150-T6 3 X 0.7472 0.7492 0.7502 4 X 0.7678 0.7700 0.7712 1 X 0.7632 0.7631 0.7615 2 X X 0.7629 0.7633 0.7626 GE14-P10DNAB404-15GZ-100T-150-T6 3 X 0.7427 0.7453 0.7474 4 X 0.7629 0.7658 0.7682 1 X 0.7670 0.7658 0.7630 2 X X 0.7666 0.7660 0.7640 GE14-P10DNAB408-17GZ-100T-150-T6 3 X 0.7469 0.7484 0.7492 4 X 0.7677 0.7694 0.7702 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-19 Table B.5 Non-AREVA Supplied Fuel Limiting Bundles In-Rack Reactivity Comparison (by product line) - (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Bundle Bottom Top Depletion Depletion Depletion GE14 - 10x10 Array (Continued) 1 X 0.7985 0.7961 0.7914 2 X X 0.8045 0.8027 0.7990 GE14-P10DNAB416-16GZ-1T-150-T-2600 3 X 0.7879 0.7876 0.7859 4 X 0.8106 0.8104 0.8085 1 X 0.8027 0.8014 0.7986 2 X X 0.8063 0.8052 0.8027 GE14-P10DNAB416-16GZ-1T-150-T-2601 3 X 0.7904 0.7910 0.7907 4 X 0.8129 0.8137 0.8132 1 X 0.7718 0.7735 0.7758 2 X X 0.7774 0.7791 0.7812 GE14-P10DNAB367-14GZ-1T-150-T-2602 3 X 0.7570 0.7593 0.7622 4 X 0.7786 0.7807 0.7833 1 X 0.7981 0.7955 0.7905 2 X X 0.8042 0.8022 0.7981 GE14-P10DNAB416-18GZ-1T-150-T-2627 3 X 0.7874 0.7869 0.7849 4 X 0.8104 0.8099 0.8075 1 X 0.8027 0.8012 0.7979 2 X X 0.8062 0.8048 0.8018 GE14-P10DNAB417-18GZ-1T-150-T-2628 3 X 0.7904 0.7908 0.7899 4 X 0.8131 0.8135 0.8124 1 X 0.8045 0.8032 0.8006 2 X X 0.8086 0.8074 0.8050 GE14-P10DNAB402-15GZ-1T-150-T-2513 3 X 0.7928 0.7931 0.7925 4 X 0.8156 0.8160 0.8152 1 X 0.8040 0.8025 0.7991 2 X X 0.8071 0.8054 0.8018 GE14-P10DNAB401-14GZ-1T-150-T-2514 3 X 0.7905 0.7902 0.7886 4 X 0.8134 0.8131 0.8111 Max. Bottom Lattice 0.8086 0.8074 0.8050 Max. Top Lattice 0.8619 0.8608 0.8580 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-20 Table B.6 AREVA Supplied Fuel Limiting Lattices In-Rack Reactivity Comparison Peak In-Rack k-infinity

[ ] [ ] [ ]

Lattice Bottom Top Depletion Depletion Depletion A10B-4059L-14GV80-FAA X 0.7764 0.7789 0.7811 A10B-4058L-12G80-FAA X X 0.7765 0.7800 0.7829 A10T-3235L-8G80-FAA X 0.7534 0.7596 0.7665 A10T-3236L-8G50-FAA X 0.7888 0.7919 0.7954 A10B-4109L-10G80-FAA X X 0.7835 0.7871 0.7909 A10T-3501L-8G80-FAA X 0.7660 0.7715 0.7774 A10T-3502L-8G50-FAA X 0.8018 0.8045 0.8070 A10B-4424L-15GV80-FAA X 0.7964 0.7974 0.7980 A10B-4423L-13GV80-FAA X X 0.7965 0.7976 0.7984 A10T-4311L-11GV80-FAA X 0.8028 0.8048 0.8065 A10T-4241L-11G50-FAA X 0.8308 0.8312 0.8306 A10B-4245L-14G70 X 0.7981 0.7999 0.8007 A10B-4236L-12G70 X X 0.7986 0.8002 0.8007 A10T-4040L-13G70 X 0.7982 0.7999 0.8002 A10T-4030L-11G50 X 0.8218 0.8225 0.8220 A10B-4545L-15G80-FBB X 0.7902 0.7910 0.7914 A10B-4555L-13G80-2CGU495-FBB X X 0.7953 0.7964 0.7970 A10T-4414L-12G80-3CGU495-FBB X 0.7961 0.7986 0.8008 A10T-4415L-12G50-3CGU495-FBB X 0.8363 0.8362 0.8355 A10B-4545L-15G80-FBB X 0.7902 0.7910 0.7914 A10B-4555L-13G80-2CGU495-FBB X X 0.7953 0.7964 0.7970 A10T-4454L-11G80-4CGU495-FBB X 0.7941 0.7979 0.8023 A10T-4454L-11G50-4CGU495-FBB X 0.8352 0.8362 0.8370 A10B-3698L-10G80-FBB X X 0.7677 0.7719 0.7760 A10T-3937L-8G80-2CGU495-FBB X 0.7850 0.7884 0.7916 A10B-4306L-14G80-FBC X 0.7873 0.7886 0.7893 A10B-4318L-12G80-FBC X X 0.7905 0.7922 0.7939 A10T-4214L-12G75-FBC X 0.7899 0.7943 0.7993 A10T-4213L-12G50-FBC X 0.8250 0.8266 0.8279 A10B-4115L-14G80-FBC X 0.7713 0.7746 0.7787 A10B-4121L-13G80-FBC X X 0.7774 0.7802 0.7832 A10T-4029L-13G75-FBC X 0.7805 0.7846 0.7898 A10T-4024L-13G50-FBC X 0.8143 0.8159 0.8177 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-21 Table B.6 AREVA Supplied Fuel Limiting Lattices In-Rack Reactivity Comparison (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Lattice Bottom Top Depletion Depletion Depletion A10B-4157L-14G80-FBD X 0.7700 0.7737 0.7783 A10B-4157L-13G80-FBD X X 0.7748 0.7781 0.7817 A10T-3812L-13G70-FBD X 0.7773 0.7817 0.7869 A10T-3812L-13G50-FBD X 0.8049 0.8071 0.8094 A10B-4303L-15G80-FBD X 0.7776 0.7798 0.7818 A10B-4303L-14G80-FBD X X 0.7776 0.7804 0.7833 A10T-4189L-14G70-FBD X 0.7876 0.7915 0.7959 A10T-4188L-14G50-FBD X 0.8168 0.8184 0.8199 A10B-4148L-13G80 X 0.7869 0.7896 0.7926 A10B-4157L-12G80 X X 0.7888 0.7919 0.7947 A10T-3863L-12G70 X 0.7901 0.7942 0.7983 A10T-3874L-10G50 X 0.8252 0.8265 0.8271 A10B-4375L-15G80 X 0.7905 0.7923 0.7943 A10B-4381L-14G80 X X 0.7929 0.7949 0.7966 A10T-4267L-14G80 X 0.7884 0.7926 0.7970 A10T-4295L-10G50 X 0.8426 0.8429 0.8420 A10B-4387L-13G80 X 0.7952 0.7973 0.7992 A10B-4393L-12G80 X X 0.7970 0.7994 0.8013 A10T-4281L-12G70 X 0.8066 0.8094 0.8119 A10T-4295L-10G50 X 0.8426 0.8429 0.8420 A10B-1739L-5G80 X X 0.6537 0.6648 0.6799 A10T-1728L-4G80 X 0.6455 0.6568 0.6715 A10B-4477L-14G80-1CGU495-FCB X 0.7902 0.7914 0.7922 A10B-4483L-13G80-2CGU495-FCB X X 0.7926 0.7940 0.7952 A10T-4374L-13G70-2CGU495-FCB X 0.8033 0.8056 0.8079 A10T-4374L-13G40-2CGU495-FCB X 0.8466 0.8463 0.8451 A10B-4465L-16G80-FCB X 0.7877 0.7887 0.7892 A10B-4477L-14G80-2CGU495-FCB X X 0.7902 0.7915 0.7923 A10T-4367L-14G80-2CGU495-FCB X 0.7864 0.7899 0.7935 A10T-4367L-14G40-2CGU495-FCB X 0.8442 0.8438 0.8427 A10B-4483L-13G80-FCB X 0.7923 0.7937 0.7949 A10B-4495L-11G70-2CGU495-FCB X X 0.8112 0.8120 0.8121 A10T-4387L-11G70-2CGU495-FCB X 0.8107 0.8122 0.8132 A10T-4387L-11G40-2CGU495-FCB X 0.8520 0.8514 0.8496 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-22 Table B.6 AREVA Supplied Fuel Limiting Lattices In-Rack Reactivity Comparison (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Lattice Bottom Top Depletion Depletion Depletion A10B-4545L-15G80-FCC X 0.7906 0.7913 0.7916 A10B-4557L-13G80-2CGU495-FCC X X 0.7956 0.7967 0.7973 A10T-4386L-13G80-2CGU495-FCC X 0.7909 0.7939 0.7972 A10T-4386L-12G50-2CGU495-FCC X 0.8355 0.8356 0.8348 A10B-4543L-13G80-FCC X X 0.7948 0.7959 0.7966 A10T-4399L-11G80-2CGU495-FCC X 0.7988 0.8010 0.8029 A10T-4399L-11G50-2CGU495-FCC X 0.8382 0.8381 0.8371 A10B-3997L-10G80-FCC X X 0.7797 0.7831 0.7861 A10T-3997L-8G80-FCC X 0.7870 0.7909 0.7941 A10T-3997L-8G50-FCC X 0.8237 0.8250 0.8254 A10B-4184L-15GV80-FCD X 0.7798 0.7818 0.7838 A10B-4103L-13G80-FCD X X 0.7794 0.7821 0.7850 A10T-3961L-11G80-FCD X 0.7818 0.7853 0.7889 A10T-3962L-11G50-FCD X 0.8205 0.8214 0.8220 A10B-3739L-9G80-FCD X X 0.7734 0.7775 0.7821 A10T-3360L-9GV70-FCD X 0.7748 0.7790 0.7828 A10T-3360L-9G40-FCD X 0.8070 0.8091 0.8113 A10B-3724L-9G80-FCD X 0.7730 0.7771 0.7818 A10B-3724L-10GV80-FCD X 0.7721 0.7762 0.7808 A10T-3356L-10GV70-FCD X 0.7737 0.7777 0.7814 A10T-3356L-10G40-FCD X 0.8050 0.8070 0.8089 A10B-3668L-11GV80-FCE X X 0.7690 0.7728 0.7768 A10T-3632L-10G80-FCE X 0.7681 0.7722 0.7764 A10T-3632L-10G50-FCE X 0.8069 0.8083 0.8092 A10B-4111L-13GV80-FCE X 0.7876 0.7896 0.7912 A10B-4111L-11GV80-FCE X X 0.7871 0.7896 0.7919 A10T-3994L-11GV70-FCE X 0.8136 0.8148 0.8158 A10T-3994L-11GV50-FCE X 0.8243 0.8248 0.8246 A10B-4384L-13G80-FCE X X 0.7850 0.7869 0.7894 A10T-4253L-11GV80-FCE X 0.7955 0.7991 0.8034 A10T-4254L-10G50-FCE X 0.8294 0.8307 0.8316 A10B-4391L-12G80-FCE X X 0.7881 0.7902 0.7923 A10T-4260L-10G70-FCE X 0.8029 0.8058 0.8088 A10T-4260L-10G50-FCE X 0.8295 0.8307 0.8316 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-23 Table B.6 AREVA Supplied Fuel Limiting Lattices In-Rack Reactivity Comparison (Continued)

Peak In-Rack k-infinity

[ ] [ ] [ ]

Lattice Bottom Top Depletion Depletion Depletion Max. Bottom Lattice [ ] 0.8112 0.8120 0.8121 Max. Top Lattice [ ] 0.8520 0.8514 0.8496 ATRIUM 10 Reference Bounding Lattices*

A10B-460L-8G40 X X 0.8738 0.8734 0.8715 A10T-450L-8G40 X 0.8747 0.8747 0.8733

• Reference B.1.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page B-24 Table B.7 Summary of Damaged and Modified Bundles Product Nominal Bundle ID Description Line Enrichment Replaced Fuel Rods YJK363 GE13 3.84 pin location B3 replaced with SS rod YJS776 GE13 4.01 pin location A5 replaced with SS rod YJS614 GE13 4.06 pin locations G9 and H2 replaced with SS rods YJS734 GE13 4.06 pin locations B8 and H2 replaced with SS rods YJS616 GE13 4.06 pin location B8 replaced with SS rod YJN587 GE13 3.72 pin location E9 replaced with SS rod JLB602 GE14 4.01 pin locations A4 and D1 replaced with SS rods FCA199 ATRIUM-10 4.077 pin location L4 replaced with SS rod Broken Fuel Rods Rod F7 broke into two pieces as it was withdrawn from the bundle. The two halves of the rod were restored to the bundle and the upper tie plate LJD968 GE5 2.84 installed.

Rod C7 was identified as broken within the fuel assembly. The rod was not disturbed (no attempt made to withdraw).

Rod G3 was identified as broken within the fuel LJE002 GE5 2.84 assembly. The rod was not disturbed (no attempt made to withdraw).

Rod J5 broke into two pieces (severed at a prior identified 91 inch defect location of a YJN587 GE13 3.72 circumferential crack and fracture) as it was rotated during a fuel inspection. The failed rod was stabilized within the bundle.

Rod A5 was noted as having a circumferential crack at approximately 127 inches. Although not JLB435 GE14 4.02 separated, rod A5 is conservatively classified as broken using visual information of this cracking.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-1 Appendix C KENO V.a Bias and Bias Uncertainty Evaluation The purpose of this Appendix is to determine the bias of the keff calculated with the SCALE 4.4a computer code for spent fuel pool criticality analysis. A statistical methodology is used to evaluate criticality benchmark experiments that are appropriate for the expected range of parameters. The scope of this report is limited to the validation of the KENO V.a module and CSAS25 driver in the SCALE 4.4a code package for use with the 44 energy group cross-section library 44GROUPNDF5 for spent fuel criticality analyses.

This calculation is performed according to the general methodology described in Reference C.2 (NUREG/CR-6698) that is also briefly described in Section C.1. The critical experiments selected to benchmark the computer code system are discussed in Section C.3. The results of the criticality benchmark calculations, the trending analysis, the basis for the statistical technique chosen, the bias, and the bias uncertainty are presented in Sections C.4 - C.7. Final results are summarized in Section C.8.

C.1 Statistical Method for Determining the Code Bias As presented in Reference C.2 (NUREG/CR-6698), the validation of the criticality code must use a statistical analysis to determine the bias and bias uncertainty in the calculation of keff. The approach involves determining a weighted mean of keff that incorporates the uncertainty from both the measurement (exp) and the calculation method (calc). A combined uncertainty can be determined using Equation 3 from Reference C.2, for each critical experiment:

2 t = calc + 2exp The weighted mean keff, the variance about the mean (s2), and the average total uncertainty of the benchmark experiments ( 2 ) can be calculated using the weighting factor 1/i2 (see Eq. 4, 5, and 6 in Reference C.2). The final objective is to determine the square root of the pooled variance, defined as (Eq. 7 from Reference C.2):

Sp = s2 + 2 Determination of the keff bias and uncertainty requires evaluation of the distribution of data and investigation of possible trends. Trends are identified by regression analysis to determine key AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-2 parameters including the slope, intercept, coefficient of determination, the T-value associated with the Students T-distribution, and a check for normality of the distribution of residuals in order to evaluate goodness-of-fit. These key parameters are used to establish the statistical significance of the calculated trend. If a trend is found to have statistical significance, then a one-sided lower tolerance band may be used to determine the bias and uncertainty. This method provides a fitted curve (KL(x)), above which 95% of the true population of keff is expected to lie, with a 95% confidence level.

If no trends of statistical significance are found and the data is normally distributed, then the bias and uncertainty can be based on a single-sided lower tolerance limit technique. This method defines a lower tolerance limit (KL) above which 95% of the true population of keff is expected to lie, with a 95% confidence level. The KL is defined in terms of the weighted-average of the data ( k eff ), the 95/95 single-sided lower tolerance factor (C95/95 - dependent on the size of the observed population), and the square-root of the pooled variance (Sp), as shown below.

K=L k eff C95 / 95 Sp In this case, the statistical bias and uncertainty are defined as shown below.

Bias =k eff 1, for k eff < 1, otherwise, Bias =

0 Uncertainty = C95/95 SP Finally, if the data is not normally distributed, then a nonparametric analysis can be employed.

This method considers the size of the observed population and determines the mth lowest value (keffm < 1) and the associated uncertainty (m) to determine a limiting value (KL), above which 95% of the true population of keff is expected to lie, with a 95% confidence level. Here, the sample size must exceed 59 in order to attain a 95/95 confidence interval, otherwise additional Non-Parametric Margin (NPM - defined by NUREG/CR-6698, see Reference C.2) must be included in the KL, as shown below.

K L =k eff m - m -NPM Bias = k eff m -1 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-3 Uncertainty = m + NPM Regardless of the method employed, the Area of Applicability (AOA) must also be defined based on evaluation of key parameters of the criticality experiments that are included in the validation. Key parameters fall into three categories: materials, geometry, and neutron energy spectrum. In general, use of the criticality evaluation is restricted to the range of parameters identified in the AOA.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-4 C.2 Area of Applicability Required for the Benchmark Experiments Commercial reactor spent fuel pools will primarily contain nuclear fuel in metal rods in a square array. This fuel is characterized by the parameter values provided in Table C.1. These typical values were used as primary tools in selecting the benchmark experiments appropriate for determining the code bias.

Benchmark calculations have been made on selected critical experiments, chosen, in so far as possible, to bound the range of variables in the spent fuel rack analysis. In rack designs, the most significant parameters affecting criticality are: (1) the fuel enrichment, (2) the neutron absorbing material, and (3) the lattice spacing. Other parameters have a smaller effect but have also been included in the analysis.

One possible way of representing the data is through a spectral parameter that incorporates influences from the variations in other parameters. The energy of the average lethargy causing fission (EALF) is this type of parameter and it is computed by KENO V.a. The range for this parameter is also included in Table C.1.

Table C.1 Range of Values for Key Spent Fuel Pool Parameters Parameter Range of Values Fissile material - Physical/Chemical Form UO2 rods Enrichment 2.35 to 4.74 wt% U-235 Moderation/Moderator Heterogeneous/Water Lattice Square, Rectangular Pitch 1.26 to 2.54 cm Clad Zircaloy, Aluminum Anticipated Absorber/Materials Boron, Stainless Steel, Water Moderating Ratio (H/X) 110 to >400 Reflection Water, Stainless Steel Neutron Energy Spectrum (Energy of the 0.060 to 0.247 eV Average Lethargy Causing Fission)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-5 C.3 Description of the Criticality Experiments Selected The set of criticality benchmark experiments has been constructed to accommodate large variations in the range of parameters of the rack configurations and also to provide adequate statistics for the evaluation of the code bias.

Sixty eight (68) critical configurations were selected from various sources. These benchmarks include configurations performed with lattices of UO2 fuel rods in water having various enrichments and moderating ratios (H/X). The area of applicability (AOA) is established within this range of benchmark experiment parameter values.

A brief description of the selected benchmark experiments is presented in Table C.2. The table includes the references where detailed descriptions of the experiments are presented.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-6 Table C.2 Descriptions of the Critical Benchmark Experiments Experiment Measured Neutron exp Brief Description Reflector Case Name keff Absorber LEU-COMP-THERM-007 (Reference C.1)

CEA-001-001 1.0000 0.0014 Square pitch fuel rod arrays with varying rod pitch configurations.

CEA-001-002 1.0000 0.0008 Each fuel rod is aluminum clad with UO2 fuel at 4.738 wt%

None Water CEA-001-003 1.0000 0.0007 U235. Performed at CEA CEA-001-004 1.0000 0.0008 Valduc Critical Mass Laboratory.

LEU-COMP-THERM-0034 (Reference C.1)

CEA-003-003 1.0000 0.0039 CEA-003-004 1.0000 0.0039 CEA-003-005 1.0000 0.0039 Borated A 2x2 array UO2 fuel rod clusters CEA-003-006 1.0000 0.0039 with 4.738 wt% U-235 Stainless Steel CEA-003-007 1.0000 0.0039 surrounded by plates of neutron CEA-003-008 1.0000 0.0039 absorbing material. Each fuel rod cluster is comprised of an 18x18 Water CEA-003-010 1.0000 0.0048 array of aluminum clad fuel rods CEA-003-011 1.0000 0.0048 with a square lattice pitch of 1.6 CEA-003-012 1.0000 0.0048 cm. Performed at CEA Valduc Critical Mass Laboratory. Boral CEA-003-013 1.0000 0.0048 CEA-003-014 1.0000 0.0043 CEA-003-015 1.0000 0.0043 LEU-COMP-THERM-039 (Reference C.1)

CEA-005-001 1.0000 0.0014 CEA-005-002 1.0000 0.0014 CEA-005-003 1.0000 0.0014 CEA-005-004 1.0000 0.0014 CEA-005-005 1.0000 0.0009 CEA-005-006 1.0000 0.0009 CEA-005-007 1.0000 0.0012 Square pitch (pitch = 1.26 cm) fuel rod arrays without fuel rods CEA-005-008 1.0000 0.0012 in all positions. Each fuel rod is CEA-005-009 1.0000 0.0012 aluminum clad with UO2 fuel at None Water CEA-005-010 1.0000 0.0012 4.738 wt% U-235. Performed at CEA Valduc Critical Mass CEA-005-011 1.0000 0.0013 Laboratory.

CEA-005-012 1.0000 0.0013 CEA-005-013 1.0000 0.0013 CEA-005-014 1.0000 0.0013 CEA-005-015 1.0000 0.0013 CEA-005-016 1.0000 0.0013 CEA-005-017 1.0000 0.0013 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-7 Table C.2 Descriptions of the Critical Benchmark Experiments (Continued)

Experiment Measured Neutron exp Brief Description Reflector Case Name keff Absorber LEU-COMP-THERM-001 (Reference C.1)

PNL-001-001 0.9998 0.0031 PNL-001-002 0.9998 0.0031 UO2 pellets enriched at 2.35 wt%

PNL-001-003 0.9998 0.0031 U-235 clad in aluminum.

PNL-001-004 0.9998 0.0031 Varying clusters of fuel rods on a 2.032 cm pitch, moderated by None Water PNL-001-005 0.9998 0.0031 water. Single cluster or multiple PNL-001-006 0.9998 0.0031 clusters with varying separation distances.

PNL-001-007 0.9998 0.0031 PNL-001-008 0.9998 0.0031 LEU-COMP-THERM-002 (Reference C.1)

PNL-002-001 0.9997 0.0020 UO2 pellets enriched at 4.31 wt%

PNL-002-002 0.9997 0.0020 U-235. Varying clusters of fuel rods on a 2.54 cm pitch, PNL-002-003 0.9997 0.0020 moderated by water. Single None Water PNL-002-004 0.9997 0.0018 cluster or multiple clusters with PNL-002-005 0.9997 0.0019 varying separation distances.

LEU-COMP-THERM-009 (Reference C.1)

PNL-009-001 1.0000 0.0021 PNL-009-002 1.0000 0.0021 Steel PNL-009-003 1.0000 0.0021 PNL-009-004 1.0000 0.0021 PNL-009-005 1.0000 0.0021 UO2 pellets enriched at 4.31 wt%

PNL-009-006 1.0000 0.0021 U-235 clad in aluminum. Three Stainless Steel 15x8 clusters of fuel rods on a (1.05 - 1.62 PNL-009-007 1.0000 0.0021 2.54 cm pitch, separated by wt% Boron) Water PNL-009-008 1.0000 0.0021 different absorber plates.

PNL-009-009 1.0000 0.0021 Varying separation distances. Boral PNL-009-024 1.0000 0.0021 Aluminum PNL-009-025 1.0000 0.0021 PNL-009-026 1.0000 0.0021 Zircaloy-4 PNL-009-027 1.0000 0.0021 LEU-COMP-THERM-016 (Reference C.1)

PNL-016-008 1.0000 0.0031 PNL-016-009 1.0000 0.0031 Stainless Steel UO2 pellets enriched at 2.35 wt% (1.05 - 1.62 PNL-016-010 1.0000 0.0031 wt% Boron)

U-235 clad in aluminum. Three PNL-016-011 1.0000 0.0031 variable sized clusters of fuel PNL-016-012 1.0000 0.0031 rods on a 2.032 cm pitch, Water PNL-016-013 1.0000 0.0031 separated by absorber plates Boral with varying separation PNL-016-014 1.0000 0.0031 distances.

PNL-016-031 1.0000 0.0031 Zircaloy-4 PNL-016-032 1.0000 0.0031 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-8 C.4 Results of Calculations with SCALE 4.4a The critical experiments described in Section C.3 were modeled with the SCALE 4.4a computer system. The resulting keff and calculational uncertainty, along with the experimental keff and experimental uncertainty are tabulated in Table C.3. The parameters of interest in performing a trending analysis of the bias are also included in the table.

In order to address situations in which the critical experiment being modeled was at other than a critical state (i.e., slightly super or subcritical), the calculated keff is normalized to the k calc experimental kexp, using the following formula (Eq.9 from Reference C.2): k norm =

k exp In the following, the normalized values of the keff were used in the determination of the code bias and bias uncertainty.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-9 Table C.3 SCALE 4.4a Results for the Selected Benchmark Experiments Benchmark SCALE 4.4a Rod Values Calculated Values Enrichment EALF No. Case Name Pitch H/X keff exp keff calc (wt% U-235) (eV)

(cm) 1 CEA-001-001 1.0000 0.0014 0.9928 0.0009 4.74 1.26 110 0.247 2 CEA-001-002 1.0000 0.0008 0.9952 0.0008 4.74 1.6 229 0.110 3 CEA-001-003 1.0000 0.0007 0.9976 0.0009 4.74 2.1 455 0.070 4 CEA-001-004 1.0000 0.0008 0.9977 0.0008 4.74 2.52 693 0.060 5 CEA-003-003 1.0000 0.0039 1.0020 0.0009 4.74 1.6 229 0.143 6 CEA-003-004 1.0000 0.0039 0.9987 0.0009 4.74 1.6 229 0.139 7 CEA-003-005 1.0000 0.0039 0.9982 0.0009 4.74 1.6 229 0.135 8 CEA-003-006 1.0000 0.0039 1.0003 0.0009 4.74 1.6 229 0.131 9 CEA-003-007 1.0000 0.0039 0.9988 0.0008 4.74 1.6 229 0.129 10 CEA-003-008 1.0000 0.0039 0.9986 0.0008 4.74 1.6 229 0.127 11 CEA-003-010 1.0000 0.0048 0.9994 0.0008 4.74 1.6 229 0.149 12 CEA-003-011 1.0000 0.0048 1.0001 0.0009 4.74 1.6 229 0.147 13 CEA-003-012 1.0000 0.0048 0.9967 0.0008 4.74 1.6 229 0.145 14 CEA-003-013 1.0000 0.0048 0.9961 0.0009 4.74 1.6 229 0.142 15 CEA-003-014 1.0000 0.0043 0.9924 0.0009 4.74 1.6 229 0.140 16 CEA-003-015 1.0000 0.0043 0.9954 0.0008 4.74 1.6 229 0.137 17 CEA-005-001 1.0000 0.0014 0.9951 0.0009 4.74 1.26 110 0.227 18 CEA-005-002 1.0000 0.0014 0.9963 0.0010 4.74 1.26 110 0.216 19 CEA-005-003 1.0000 0.0014 0.9978 0.0009 4.74 1.26 110 0.197 20 CEA-005-004 1.0000 0.0014 0.9947 0.0008 4.74 1.26 110 0.186 21 CEA-005-005 1.0000 0.0009 0.9963 0.0008 4.74 1.26 110 0.141 22 CEA-005-006 1.0000 0.0009 0.9986 0.0009 4.74 1.26 110 0.146 23 CEA-005-007 1.0000 0.0012 0.9952 0.0009 4.74 1.26 110 0.217 24 CEA-005-008 1.0000 0.0012 0.9934 0.0010 4.74 1.26 110 0.208 25 CEA-005-009 1.0000 0.0012 0.9957 0.0009 4.74 1.26 110 0.202 26 CEA-005-010 1.0000 0.0012 0.9973 0.0008 4.74 1.26 110 0.176 27 CEA-005-011 1.0000 0.0013 0.9922 0.0008 4.74 1.26 110 0.227 28 CEA-005-012 1.0000 0.0013 0.9937 0.0009 4.74 1.26 110 0.222 29 CEA-005-013 1.0000 0.0013 0.9926 0.0009 4.74 1.26 110 0.219 30 CEA-005-014 1.0000 0.0013 0.9934 0.0009 4.74 1.26 110 0.218 31 CEA-005-015 1.0000 0.0013 0.9944 0.0009 4.74 1.26 110 0.217 32 CEA-005-016 1.0000 0.0013 0.9951 0.0009 4.74 1.26 110 0.214 33 CEA-005-017 1.0000 0.0013 0.9954 0.0009 4.74 1.26 110 0.215 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-10 Table C.3 SCALE 4.4a Results for the Selected Benchmark Experiments (Continued)

Benchmark SCALE 4.4a Rod Values Calculated Values Enrichment EALF No. Case Name Pitch H/X keff exp keff calc (wt% U-235) (eV)

(cm) 34 PNL-001-001 0.9998 0.0031 0.9970 0.0008 2.35 2.032 399 0.096 35 PNL-001-002 0.9998 0.0031 0.9958 0.0009 2.35 2.032 399 0.096 36 PNL-001-003 0.9998 0.0031 0.9947 0.0008 2.35 2.032 399 0.095 37 PNL-001-004 0.9998 0.0031 0.9955 0.0007 2.35 2.032 399 0.095 38 PNL-001-005 0.9998 0.0031 0.9950 0.0007 2.35 2.032 399 0.094 39 PNL-001-006 0.9998 0.0031 0.9963 0.0007 2.35 2.032 399 0.095 40 PNL-001-007 0.9998 0.0031 0.9959 0.0007 2.35 2.032 399 0.093 41 PNL-001-008 0.9998 0.0031 0.9936 0.0007 2.35 2.032 399 0.094 42 PNL-002-001 0.9997 0.0020 0.9963 0.0009 4.31 2.54 256 0.114 43 PNL-002-002 0.9997 0.0020 0.9956 0.0009 4.31 2.54 256 0.114 44 PNL-002-003 0.9997 0.0020 0.9961 0.0008 4.31 2.54 256 0.114 45 PNL-002-004 0.9997 0.0018 0.9951 0.0008 4.31 2.54 256 0.113 46 PNL-002-005 0.9997 0.0019 0.9945 0.0008 4.31 2.54 256 0.111 47 PNL-009-001 1.0000 0.0021 0.9979 0.001 4.31 2.54 256 0.114 48 PNL-009-002 1.0000 0.0021 0.9958 0.0007 4.31 2.54 256 0.113 49 PNL-009-003 1.0000 0.0021 0.9977 0.0008 4.31 2.54 256 0.114 50 PNL-009-004 1.0000 0.0021 0.9968 0.0008 4.31 2.54 256 0.114 51 PNL-009-005 1.0000 0.0021 0.9975 0.0008 4.31 2.54 256 0.115 51 PNL-009-006 1.0000 0.0021 0.9973 0.0009 4.31 2.54 256 0.114 53 PNL-009-007 1.0000 0.0021 0.9961 0.0009 4.31 2.54 256 0.115 54 PNL-009-008 1.0000 0.0021 0.9972 0.0008 4.31 2.54 256 0.114 55 PNL-009-009 1.0000 0.0021 0.9967 0.0008 4.31 2.54 256 0.115 56 PNL-009-024 1.0000 0.0021 0.9964 0.0007 4.31 2.54 256 0.114 57 PNL-009-025 1.0000 0.0021 0.9970 0.0009 4.31 2.54 256 0.114 58 PNL-009-026 1.0000 0.0021 0.9950 0.0008 4.31 2.54 256 0.113 59 PNL-009-027 1.0000 0.0021 0.9957 0.0008 4.31 2.54 256 0.113 60 PNL-016-008 1.0000 0.0031 0.9952 0.0007 2.35 2.032 399 0.097 61 PNL-016-009 1.0000 0.0031 0.9965 0.0008 2.35 2.032 399 0.096 62 PNL-016-010 1.0000 0.0031 0.9946 0.0006 2.35 2.032 399 0.097 63 PNL-016-011 1.0000 0.0031 0.9954 0.0007 2.35 2.032 399 0.096 64 PNL-016-012 1.0000 0.0031 0.9954 0.0007 2.35 2.032 399 0.097 65 PNL-016-013 1.0000 0.0031 0.9960 0.0007 2.35 2.032 399 0.096 66 PNL-016-014 1.0000 0.0031 0.9943 0.0007 2.35 2.032 399 0.097 67 PNL-016-031 1.0000 0.0031 0.9949 0.0008 2.35 2.032 399 0.095 68 PNL-016-032 1.0000 0.0031 0.9965 0.0007 2.35 2.032 399 0.095 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-11 C.5 Trending Analysis The next step of the statistical methodology used to evaluate the code bias for the pool of experiments selected is to identify any trend in the bias. This is done by using the following trending parameters:

• Fuel enrichment (wt% U-235)

• Fuel rod pitch

• Atom ratio of the moderator to fuel (H/X)

• Energy of the Average Lethargy causing Fission, EALF (eV)

The first step in calculating the bias uncertainty limit is to apply regression-based methods to identify any correlation of the calculated values of keff with the trending parameters. The trends show the results of systematic errors or bias inherent in the calculational method used to estimate criticality.

For the critical benchmark experiments that were slightly super or subcritical, an adjustment to the keff value calculated with SCALE 4.4a (kcalc) was done as suggested in Reference C.2. This adjustment is done by normalizing the calculated (kcalc) value to the experimental value (kexp).

This normalization does not affect the inherent bias in the calculation due to very small differences in keff. Unless otherwise mentioned, the normalized keff values (knorm) have been used in all subsequent calculations.

The regression analysis employs the normalized keff values (knorm) and corresponding total uncertainty values (t), which are the values of the dependent variable and the corresponding weighting factors defined by 1/i2, where i = t for the ith data point. Data points consist of the ordered pairs (xi,yi), where yi = keff for the ith data point. Reference C.2 suggests the use of weighting factors to reduce the importance of data with higher uncertainty. For this application, the weighted trends were evaluated and the results were verified by comparison to the non-weighted trending results.

Note that t values are an intermediate calculational result and all downstream calculations should include all significant digits resulting from the intermediate calculation. Therefore, to be consistent with the guidance from Reference C.2, the weighting factors were evaluated as shown below with all significant digits included in later calculations.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-12 1 1

= 2 i2 2 exp + calc i The linear fitting function is defined as: yi = mxi + b, where m and b are the fitting coefficients, slope and intercept, respectively. The slope (m) and intercept (b) are determined by application of the following equations (from Reference C.2, page 8):

1 1 xi yi xi y i m=

2 i2 2

i2 i i i i i i 1 x i2 yi xi x i y i b=

2 2 2

i2 i i i i i i i 2

1 x i2 x i

= 2 2 i2 i i i i i The weighted-average value of the dependent variable ( k eff ) is calculated as follows:

yi i

2

= y = i k eff 1

i 2 i

For the residuals, there are n - 2 = 66 degrees of freedom, since there are n = 68 data points.

The ith value of the regression is expressed as y i = mx i + b and the weighted sums of the squares for the residuals (SSResidual), for the regression (SSRegression), and for the total (SSTotal) are calculated as follows:

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-13 (y y i )2 i

i i2 SS Residual =

1 i

2 i

(y y ) 2 i

i i2 SS Regression =

1 i

2 i

SS Total = SS Re sidual + SS Re gression These, in turn, allow calculation of the goodness-of-fit parameters: coefficient of correlation (r2),

and the TValue corresponding to the Students T-distribution:

SS Re gression r2 =

SS Total (n 2) SS Regression TValue =

SS Re sidual The r2 value represents the proportion of the sum of the squares for the y-values about their mean that can be attributed to a linear relation between x and y. The closer that r2 approaches a value of 1, the better the fit of the data to the linear equation. As described in Section 10.3.2 of Reference C.3, calculated TValues are compared with the critical value of the Students T-distribution with a significance level of = 0.05/2 = 0.025 and n - 2 = 66 degrees of freedom (i.e., a critical value of 1.996). The null hypothesis for this test (H0), is that the slope is not statistically significant; thus, a statistically significant trend may exist if: TValue > 1.996 .

Alternatively, the probability of obtaining a TValue of larger magnitude from a two-tailed T-distribution with the same n - 2 = 66 degrees of freedom is calculated. In general, a low probability (e.g., p < 0.05) is necessary to confirm that a statistically significant trend exists.

In cases where a statistically significant trend is indicated by the Students T-test, then the residuals of the regression are tested to determine if the error component is normally distributed with mean zero, which confirms that the statistical test for significance is valid (Section 10.4 of Reference C.3). The Anderson-Darling test described in Reference C.5 is employed for this AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-14 purpose; calculation of the test statistic (A2) proceeds by first sorting the sample into ascending order.

X1 X n Calculate the sample average X , and standard deviation X :

1 n X=

n X i i =1 n

(X i X)

X = i 1 n 1 Xi X Then, compute standardized values: Yi = .

X Now, the Anderson-Darling test statistic can be calculated, as shown:

n ln ( Pi ) + ln (1 Pn +1i )

( 2i 1) 2 A = n i =1 n Here, Pi is the cumulative normal probability corresponding to the standard score of Yi, defined above. Finally, the calculated A2 value is adjusted for the size of the sample (n):

0.75 2.25 A*= A 2 1.0 + +

n n2 The null hypothesis of normality is rejected if the value of A* exceeds the critical value of 0.752, at a significance level of 0.05. Therefore, if A* 0.752, then the residuals are distributed normally and the statistical test for significance is valid.

Results of the weighted regression analysis and statistical tests are summarized in Table C.4 for all key parameters. This table shows that only H/X produces a valid trend. Therefore, a single-sided lower tolerance band will be used to establish the bias and uncertainty as a function of H/X. Although there is no trend for U-235 enrichment and the residuals for rod pitch and EALF are not normally distributed (indicating an invalid trend), lower tolerance bands have also been calculated for these parameters. The intermediate results are listed in Table C.5.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-15 Calculational details of the single-sided lower tolerance band can be found in Reference C.2 (pages 12 - 13); some details will be repeated here for the sake of convenience, clarity, and for verification of intermediate values used in the calculations. The equation for the single-sided lower tolerance band is as follows:

1 K L ( x ) = K fit ( x ) (S P )fit 2Fa( fit,n2) +

(x x ) + z 2

(n 2) 2P 1 n (x i x )

2 12 ,n2 i

Kfit(x) is the function derived in the trending analysis for independent variable x. Because a positive bias may be non-conservative, the value Kfit = 1.00 is substituted for all x where Kfit(x) >

1.00. Other symbols not previously introduced are defined below:

= p the desired confidence level

= 0.95 Ffit,n-2 = the F distribution percentile with degree of fit (2, for linear) and n-2 degrees of freedom, based on the Excel function FINV with arguments (1-0.95, 2, n-2).

z 2P-1 = the symmetric percentile of the Gaussian (normal) distribution that contains the P fraction, based on the Excel function NORMSINV with argument (0.95).

1-p 1 - 0.95

= g = = 0.025 2 2 c1-2 g ,n-2 = the upper Chi-square percentile

= based on the Excel function CHIINV with arguments (1-0.025, n-2).

In addition to the constants defined above, the equations listed below are quantities that are dependent upon the type of fit and the specific independent variable (except that 2 is constant, as shown).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-16 xi i

2 x= i 1

i 2

i (x i x )2 2I (x i x )

2 i

=

1 1 i

n I 2I n

2 =

1 i

2 i

1 (y i y i )2 n2 i 2

2 s fit =

i 1 1 n i i2 (SP )fit = s fit2

+ 2 Figures C.1 to C.4 show the normalized keff datasets plotted as a function of the weighted U-235 enrichment, rod pitch, H/X, and EALF data, respectively. The plotted data is overlaid with the linear trend line and the lower tolerance band. This lower tolerance band bounds 95% of the population with a confidence level of 95%.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-17 Table C.4 Results Summary for Weighted Trending Analysis Enrichment Rod Pitch Moderating Parameter EALF (eV)

(wt% U-235) (cm) Ratio (H/X)

Slope 4.19E-05 1.14E-03 4.30E-06 -1.75E-02 Intercept 0.9957 0.9838 0.9949 0.9985 2

r 0.0003 0.1331 0.1429 0.3170 Tcrit 1.996 1.996 1.996 1.996 T-value 0.1422 3.1838 3.3175 5.5343 P(T>T-value) 0.8874 0.0022 0.0015 5.802E-07 Valid Trend? NO YES YES YES A* --- 1.2471 0.5916 0.8537 Statistical Test Valid? --- NO YES NO AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-18 Table C.5 Intermediate Results for Lower Tolerance Band Evaluation x (x i

i x) 2 s 2fit (S P )fit Weighted Fit Enrichment 4.390 35.126 3.0529E-06 0.00270 (wt% U-235)

Rod Pitch 1.766 20.620 2.6473E-06 0.00262 (cm)

Moderating Ratio 229.11 1.5584E+06 2.6174E-06 0.00262 H/X EALF 0.1498 0.2082 2.0859E-06 0.00251 (eV)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-19 1.010 1.005 1.000 y = 4.192E-05x + 0.9957 R2 = 0.0003 Normalized k(eff) 0.995 0.990 0.985 0.980 1 2 3 4 5 6 U-235 Enrichment (%)

Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.1 Weighted U-235 Enrichment Trend Evaluation AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-20 1.010 1.005 y = 1.141E-03x + 0.9938 R2 = 0.1331 1.000 Normalized k(eff) 0.995 0.990 0.985 0.980 1.0 1.5 2.0 2.5 3.0 Fuel Rod Pitch (cm)

Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.2 Weighted Fuel Rod Pitch Trend Evaluation AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-21 1.010 1.005 y = 4.299E-06x + 0.9949 1.000 R2 = 0.1429 Normalized k(eff) 0.995 0.990 0.985 0.980 0 100 200 300 400 500 600 700 800 Moderating Ratio, H/X Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.3 Weighted H/X Trend Evaluation AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-22 1.010 1.005 y = -0.01752x + 0.9985 R2 = 0.3170 1.000 Normalized k(eff) 0.995 0.990 0.985 0.980 0.00 0.05 0.10 0.15 0.20 0.25 0.30 EALF (eV)

Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.4 Weighted EALF Trend Evaluation AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-23 C.6 Bias and Bias Uncertainty For situations in which no significant trend in bias is identified the statistical methodology, presented in Reference C.2 and summarized in Section C.1 of this appendix, suggests to first check the distribution of the normalized keff dataset. The Anderson-Darling test statistic is calculated consistent with the description presented in Section C.5. The null hypothesis of normality is rejected if the value of A* exceeds the critical value of 0.752, (based upon a significance level of 0.05). Therefore, if A* 0.752, then the data are distributed normally.

The Anderson-Darling test was completed for the 68 case benchmark set. The resulting Anderson-Darling test statistic modified from the number of data points A* was determined to be 0.4186. A plot of the data relative to a normal distribution is provided in Figure C.5. Based on the test statistic and plot, the benchmark data can be considered normally distributed.

With the assumption of normality being validated, a single-sided lower tolerance limit can be used to determine the bias and uncertainty. For n = 68, the tolerance limit is C95/95 = 1.996, from Reference C.4. Results obtained for the weighted average keff ( k eff ), the variance about the mean (s2), the average total uncertainty ( 2 ), and the square-root of the pooled variance (Sp),

are shown below.

Weighted yi i

2 k eff = y = i

= 0.99585 1

i 2 i

Bias = k eff - 1 = -0.00415 1 (y i y )2 n 1 i i2 2

s = =3.0083-06 1 1 n i i2 n

2 = = 4.2355E-06 1

i 2 i

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-24 SP = s 2 + 2 = 0.00269 The bias and bias uncertainty are:

Bias = -0.00415 Uncertainty = (C95/95)(Sp) = (1.996)(0.00269) = 0.00537 The corresponding lower tolerance limit is:

KL = k eff - (C95/95)(Sp) = 0.99585 - 0.00537 = 0.99048 When this lower tolerance limit, KL = 0.99048, is compared with the lower tolerance bands of the trended data in Figures C.1 through C.4, the lower tolerance limit is not sufficiently conservative to bound all the trended parameters. A minimum keff of 0.98761 is projected for the EALF* trend evaluation (see Figure C.4). Based upon this minimum value, a trend corrected bias can be calculated as follows:

BiasCorr = -0.00415 - (0.99048-0.98761) = -0.00702 If the magnitude of this corrected bias is conservatively adjusted to 0.0075 (with the pooled uncertainty rounded to 0.0027) a bounding limit is established as shown:

kL = (1 - 0.0075) - (1.996)(0.0027) = 0.9871 These adjusted values will be used to represent this benchmark data, i.e., lBiasl = 0.0075 and Sp = 0.0027.

• This is a conservative treatment because the EALF trend was shown to not be statistically valid in Table C.4.

Including the trend correction in the bias term will result in a more conservative k95/95 than treating it as an increased uncertainty.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-25 Normal Probability Distribution Actual Data Probability 0.991 0.993 0.995 0.997 0.999 1.001 Data Figure C.5 Normal Probability Plot for the keff Dataset AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-26 C.7 Area of Applicability A brief description of the spectral and physical parameters characterizing the set of selected benchmark experiments is provided in Table C.6.

Table C.6 Range of Values for Key Benchmark Experiment Parameters Parameter Range of Values Heterogeneous lattices, Geometrical Shape with Square and Rectangular pitch Fuel type UO2 fuel rods Enrichment (for UO2 fuel) 2.35 to 4.74 wt% U-235 Fuel rod pitch 1.26 to 2.54 cm H/X 110 to >400 EALF 0.060 to 0.247 eV Stainless steel, borated stainless Absorbers steel, aluminum, Zircaloy-4, and Boral Water Reflectors Stainless Steel AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-27 C.8 Bias Summary and Conclusions The mixed dataset of 68 criticality safety benchmarks experiments was tested against the null hypothesis of normality and was found to be normally distributed. Thus, a parametric analysis was used to determine the bias and bias uncertainty, which resulted in a lower tolerance limit of KL = 0.99048.

A standard trending analysis was also performed using linear regression analysis, including significance testing and goodness-of-fit evaluation. Four independent variables were examined:

enrichment (wt% U-235), rod pitch, moderating ratio (H/X), and EALF (eV). The results of the trending analysis showed that the weighted trend for H/X met the criteria for statistical validity.

Although most trends for the other parameters were deemed statistically insignificant, lower tolerance bands were calculated for all variables and then overlaid on the data plots to illustrate the effect.

When the lower tolerance limit, KL = 0.99048, was compared with the lower tolerance bands of the trended data, the lower tolerance limit (KL) was not conservative for all trended parameters.

Thus, the bias term was increased as shown below.

lAdjusted biasl = 0.0075 Adjusted KL = (1 - 0.0075) - (1.996)(0.0027) = 0.9871 (The following adjusted values are referenced in Section 5.0 and applied in Section 7.8 of the report: lBiasl = 0.0075 and Sp = 0.0027).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-28 C.9 References C.1 Nuclear Energy Agency, International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03, Nuclear Energy Agency, Organization for Co-operation and Development, September 2009.

C.2 Nuclear Regulatory Commission, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001.

C.3 Rosenkrantz W.A., Introduction to Probability and Statistics for Scientists and Engineers, The McGraw-Hill Companies Inc. 1997.

C.4 Owen, D.B., Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, Sandia Corporation Monograph SRC-607, 1963.

C.5 DAgostino, R.B. and Stephans, M.A. Goodness of Fit Techniques, Statistics, Textbooks and Monographs, Volume 68, New York, NY, 1986.

C.6 Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01 Revision 0, Staff Guidance Regarding The Nuclear Criticality Safety Analysis For Spent Fuel Pools, (ADAMS # ML110620086).

C.7 NUREG/CR-6979, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data, September 2008, (ADAMS # ML082880452).

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-29 Addendum to Appendix C Benchmark Extension with HTC Critical Experiments Critical experiments with Plutonium and other actinides are outside of the area of applicability for BOL reactivity equivalent evaluations such as this one. However, item IV.4.a.i of the Reference C.6 guidance document indicates that the HTC critical experiments should be considered. This section has been created to demonstrate that it is reasonable to exclude the HTC critical experiments from the SCALE 4.4a benchmarking evaluation.

Twenty-three of the twenty-six cases from the phase 3 experiments (Reference C.7) were added to the 68 cases shown in Tables C.2 and C.3 (producing a total of 91 cases). These cases were selected because they are similar to BWR spent fuel pool conditions and because they do not contain soluble boron or soluble gadolinia. The SCALE 4.4a results for these cases are shown in Table C.7. A statistical evaluation performed per Reference C.2 indicates that this expanded benchmark set is normally distributed with an average keff of 0.99765 with a pooled uncertainty of 0.002536. Therefore the Bias, the total uncertainty, and the parametric lower tolerance limit (see below) are less limiting for this expanded dataset than for the 68 case dataset (see Section C.6).

Bias = k eff - 1 = 0.99765 - 1 = -0.00235 Uncertainty = (C95/95)(Sp) = (1.942)(0.002536) = 0.00492 KL = k eff - (C95/95)(Sp) = 0. 99765 - 0. 00492 = 0.99273 This extended dataset produced statistically significant trends for H/X and EALF. Therefore, lower tolerance bands were determined for these parameters (per Reference C.2) and the resulting comparison plots are included as Figures C.6 and C.7. These trend results show that the minimum overall value remains unchanged (about 0.988 for EALF at 0.247 eV).

From this comparison the recommended HTC critical benchmark cases can be excluded without creating non-conservative results. Since these benchmark cases are outside of the area of applicability for the BOL reactivity equivalent KENO calculations, the k95/95 evaluation in the main body of this report will be based upon the 68 case dataset summarized in Section C.8.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-30 Table C.7 SCALE 4.4a Results for the HTC Critical Benchmark Experiments SCALE 4.4a Rod Benchmark Values Enrichment EALF No. Case Name Calculated Values Pitch H/X keff exp keff calc (wt% U-235) (eV)

(cm) 1 to 68 see Table C.3 69 HTC-2518 1.0000 0.0011 0.9973 0.0002 1.57 1.6 466 0.125 70 HTC-2521 1.0000 0.0011 0.9974 0.0002 1.57 1.6 466 0.131 71 HTC-2522 1.0000 0.0011 0.9975 0.0002 1.57 1.6 466 0.126 72 HTC-2523 1.0000 0.0011 0.9967 0.0002 1.57 1.6 466 0.137 73 HTC-2511 1.0000 0.0011 0.995 0.0002 1.57 1.6 466 0.131 74 HTC-2525 1.0000 0.0011 0.9955 0.0002 1.57 1.6 466 0.135 75 HTC-2526 1.0000 0.0011 0.9972 0.0003 1.57 1.6 466 0.131 76 HTC-2527 1.0000 0.0011 0.9942 0.0002 1.57 1.6 466 0.139 77 HTC-2509 1.0000 0.0008 0.999 0.0002 1.57 1.6 466 0.114 78 HTC-2531 1.0000 0.0008 0.9989 0.0002 1.57 1.6 466 0.113 79 HTC-2532 1.0000 0.0008 0.9995 0.0002 1.57 1.6 466 0.113 80 HTC-2532 1.0000 0.0008 0.999 0.0002 1.57 1.6 466 0.112 81 HTC-2533 1.0000 0.0008 0.9989 0.0002 1.57 1.6 466 0.112 82 HTC-2534 1.0000 0.0008 0.9978 0.0002 1.57 1.6 466 0.11 83 HTC-2536 1.0000 0.0008 0.9995 0.0002 1.57 1.6 466 0.107 84 HTC-2537 1.0000 0.0008 1.0003 0.0002 1.57 1.6 466 0.105 85 HTC-2538 1.0000 0.0008 1.0001 0.0002 1.57 1.6 466 0.103 86 HTC-2539 1.0000 0.0008 0.9998 0.0002 1.57 1.6 466 0.106 87 HTC-2541 1.0000 0.0008 0.9999 0.0002 1.57 1.6 466 0.108 88 HTC-2544 1.0000 0.0008 0.9985 0.0002 1.57 1.6 466 0.116 89 HTC-2547 1.0000 0.0008 0.9998 0.0002 1.57 1.6 466 0.154 90 HTC-2548 1.0000 0.0008 1.0001 0.0002 1.57 1.6 466 0.129 91 HTC-2549 1.0000 0.0008 0.9991 0.0002 1.57 1.6 466 0.117 AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-31 1.010 1.005 y = 8.956E-06x + 0.9942 R2 = 0.3925 1.000 Normalized k(eff) 0.995 0.990 0.985 0.980 0 100 200 300 400 500 600 700 800 Moderating Ratio, H/X Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.6 Weighted H/X Trend (HTC Extended Benchmark)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page C-32 1.010 1.005 y = -0.03175x + 1.0018 R2 = 0.3348 1.000 Normalized k(eff) 0.995 0.990 0.985 0.980 0.00 0.05 0.10 0.15 0.20 0.25 0.30 EALF (eV)

Benchmark Data Lower Tolerance Band Weighted Linear Trend Figure C.7 Weighted EALF Trend (HTC Extended Benchmark)

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-1 Appendix D CASMO-4 Qualification for In-Rack Modeling D.1 Introduction The criticality safety analysis provided in this report is primarily a KENO V.a based analysis.

However, KENO V.a does not have depletion capability so the CASMO-4 code is used for a subset of calculations that require fuel depletion. Since CASMO-4 is a two-dimensional code, it cannot provide stand-alone benchmark results of finite criticality experiments.

CASMO-4 has demonstrated acceptable isotopic depletion and nuclear library capability for reactor core related calculations in Reference D.1. It is a multi-group, two-dimensional transport theory code which also has an in-rack geometry option where typical storage rack geometries can be modeled on an infinite lattice basis. This code is used for fuel depletion in a manner that is consistent with AREVAs NRC approved CASMO-4 / MICROBURN-B2 methodology (Reference D.1). The library files used in this evaluation are the standard CASMO-4 70 group library based on ENDFB-IV. The CASMO-4 computer code and data library are controlled by AREVA procedures and the version used in this analysis meets the requirements of Reference D.1.

Within this criticality evaluation, CASMO-4 is used to:

• perform a k ranking of fuel lattices at peak in-rack reactivity conditions (see Appendix B)

• define reference lattices that are more reactive than all past and expected future fuel lattices (the lattices of the reference bounding assembly)

• define fresh fuel reactivity equivalent lattices* for use in KENO V.a.

In support of this usage, this appendix will:

• compare CASMO-4 k results with KENO V.a to demonstrate that the fuel storage rack option in CASMO-4 also produces reasonable results

• estimate the CASMO-4 depletion uncertainty

• demonstrate that the CASMO-4 depletion uncertainty combined with a CASMO-4 calculational uncertainty is smaller than the 0.010 k uncertainty adder that is applied when the REBOL lattice is defined.

• REBOL lattices.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-2 D.2 k Comparisons These comparisons are performed in accordance with the guidance provided in References D.2.

They are performed to quantify the differences in predicted k between CASMO4 and KENO V.a (Section D.2.2.1) and to demonstrate that a k predicted by CASMO4 is nearly identical to a k predicted by KENO V.a (Section D.2.2.2).

D.2.1 Comparison Methodology The evaluation in this appendix will compare the k values produced by the CASMO-4 code to the SCALE 4.4a KENO V.a code for different geometries and U-235 enrichment levels.

The validation of the CASMO-4 code in this Appendix is performed in two steps to demonstrate its acceptability for the two different ways that CASMO-4 is used in this analysis.

• Identify the relative reactivity of a lattice with the use of the storage rack geometry option. This is addressed by determining the CASMO-4 uncertainty relative to KENO V.a by comparison of calculated k-infinities from the two codes.

• Evaluate relative changes in reactivity associated with changes in geometry and U-235 enrichment. For this evaluation, the differential k-infinities from the two codes are compared based upon the same input perturbations.

These different approaches are described in more detail in the following sections.

D.2.1.1 CASMO-4 Uncertainty for Absolute k Relative to KENO The approach taken is to perform a series of calculations with varied enrichments and geometries with the two codes and then to compare the k results. The validation guidance of NUREG/CR-6698 (Reference D.2) is followed to determine a code uncertainty for CASMO-4 relative to KENO V.a. The KENO V.a calculations are treated as the critical experiments in this comparison. GE8x8 fuel as well as top and bottom lattices from the GE9x9, GE10x10, ATRIUM-10 (10x10), and ATRIUM 10XM (10x10) product lines are used.

D.2.1.2 CASMO-4 Uncertainty for k Relative to KENO The capability of the CASMO-4 code to predict the change in reactivity associated with a perturbation of fuel parameters is demonstrated by comparison of k values obtained with KENO V.a to those obtained with CASMO-4. The approach taken is to evaluate small AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-3 perturbations in reactivity by varying the enrichment relative to a base case. The same cases used in the evaluation of the uncertainty of the absolute multiplication factor are used in this evaluation. The k values will be determined for both KENO V.a and CASMO-4 for enrichment perturbations from the reference case.

The k values are compared between the two codes and a statistical evaluation similar to that identified in Reference D.2 is used to establish an uncertainty for the determination of k values with CASMO-4 relative to k values with KENO V.a.

D.2.1.3 Experiment Descriptions As noted, KENO calculations are used as the reference experiments. The evaluations are based on the Boral storage racks in the Browns Ferry spent fuel pool. The validation is performed using GE8x8 lattices and both bottom and top lattices from the GE9x9, GE10x10, ATRIUM-10 (10x10), and ATRIUM 10XM (10x10) product lines. These lattices represent the limiting past and current fuel types for the Browns Ferry Nuclear plant. Enrichment is varied in 0.05 increments around a base of 3.35% U-235 by weight. A total of eleven (11) enrichment levels from a minimum of 3.1 wt% to a maximum of 3.6 wt% are evaluated.

The calculations are reported for 4 ºC since it represents the limiting in-rack reactivity condition for the Boral storage racks (see Table 6.1). The fuel assembly data and rack geometry are consistent with the inventory and configuration of the Browns Ferry spent fuel pool.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-4 D.2.2 Analysis of Validation Results D.2.2.1 CASMO-4 Uncertainty for Absolute k-effective Relative to KENO The calculated multiplication factors from KENO and CASMO were tabulated. The keno terms are taken from each individual KENO calculation and the casmo terms are set to the CASMO-4 convergence criterion for the individual case. (Use of the CASMO convergence is consistent with footnote 1 on page 6 of Reference D.2.) A combined uncertainty tot was determined consistent with equation 3 of Reference D.2.

+ casmo 2 2 tot

= keno The tabulated results are provided in Table D.1 for variations of geometry and enrichment. The geometry is specified by product line. A suffix of B or T is used to describe bottom or top lattice geometry, respectively. For example, A10XMT specifies ATRIUM 10XM top lattice geometry. The GE8x8 fuel contains only one geometry configuration and therefore does not have this suffix. The differences of the calculated multiplication factor values along with the components used in the statistical evaluation are provided in Table D.2.

The weighted average difference (kbar), the variance about the mean (s2), and the average total uncertainty (2) are calculated using the weighting factor 1/t2 . The square root of the pooled variance is determined per Equation 7 of Reference D.2 as shown. These results are listed below.

S s +

2 2 p

=

[

]

The simple average and standard deviation values were also tabulated by lattice geometry type:

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-5

[

]

A data normality test was completed using the Anderson-Darling test (see section 9.5.4.1 of Reference D.3). (The Anderson-Darling test is described in Section C.5 of Appendix C). The AD test statistic was calculated to be 0.8143 and the criterion is 0.746 *. Since the AD test statistic is greater than the test criterion one can conclude that the data is not from a normal distribution.

A distribution free one sided tolerance limit evaluation was also performed for this data set of 99 values. This was performed for both the upper and lower bounds. This evaluation indicated that on a 95/95 basis the more limiting k difference boundary is [ ]. For the weighted mean difference of [ ] and the limiting boundary value (above), the limiting effective uncertainty term is [ ].

Area of Applicability The fuel and rack geometries as well as representative fuel enrichments were selected to be consistent with the Browns Ferry GE High density Boral storage racks. It is recognized that

• In Appendix C the AD test statistic was adjusted for the number of data points and compared to the criteria of 0.752. In this appendix, the criterion was adjusted for the number of data points.

As indicated by equation 20 of Reference D.2, the uncertainty component is effectively the difference between the limiting boundary and the mean value.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-6 spent fuel pool storage tube modeling simplifications are included in the CASMO model relative to the more explicit model used with KENO, see Section 6.1. This difference in the modeling technique is included in this comparison. The REBOL lattice enrichment and geometries used in the k95/95 determination for the Browns Ferry Spent Fuel Pool are within the area of applicability of this comparison.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-7 Table D.1 CASMO-4 and KENO V.a Validation Case Information

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-8 Table D.1 CASMO4 and KENO Validation Case Information (Continued)

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-9 Table D.1 CASMO4 and KENO Validation Case Information (Continued)

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-10 Table D.2 CASMO - KENO Difference and Statistical Parameters

[

]

• k is kCASMO - kKENO.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-11 Table D.2 CASMO - KENO Difference and Statistical Parameters (Continued)

[

]

• k is kCASMO - kKENO.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-12 Table D.2 CASMO - KENO Difference and Statistical Parameters (Continued)

[

]

• k is kCASMO - kKENO.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-13

[

]

Figure D.1 Normality Plot for CASMO-KENO k-infinity Comparison AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-14 D.2.2.2 CASMO-4 Uncertainty for k-effective The actual KENO and CASMO calculations used in this k evaluation are those used in Section D.2.2.1. In this evaluation, the relative reactivity change is evaluated by taking the delta with respect to the reference case. A difference is then determined between the k values obtained with KENO and the k values obtained with CASMO-4 for the same perturbation.

The Anderson-Darling goodness of fit for normality test was also completed with the AD test statistic calculated to be 0.7425 with the criterion of 0.7456. Based on these results and the comparison in Figure D.2, it is determined that the data is normally distributed.

The magnitude of the average difference between the k values was [ ] with a standard deviation of [ ]. For the data sample of 50 the single sided tolerance factor is 2.065 from Table 2.1 of Reference D.2. This is conservatively applied for 90 data samples. Therefore the 95/95 bias uncertainty is: [ ] when rounded to four decimal places.

Area of Applicability The fuel and rack geometries as well as representative fuel enrichments were selected to be consistent with the Browns Ferry GE High density Boral storage racks. It is recognized that spent fuel pool storage tube modeling simplifications are included in the CASMO model relative to the more explicit model used with KENO, see Section 6.1. This difference in the modeling technique is included in this comparison. The REBOL lattice enrichment and geometries used in the k95/95 determination for the Browns Ferry Spent Fuel Pool are within the area of applicability of this comparison.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-15 Table D.3 CASMO versus KENO Relative Reactivity Differences at 4ºC

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-16 Table D.3 CASMO versus KENO Relative Reactivity Differences at 4ºC (Continued)

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-17 Table D.3 CASMO versus KENO Relative Reactivity Differences at 4ºC (Continued)

[

]

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-18

[

]

Figure D.2 Normality Plot for kCASMO - kKENO k-infinity Comparison AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-19 D.3 Depletion Uncertainty Estimates Depletion uncertainty estimates from EMF-2158(P) (Reference D.1) and from the interim staff guidance document (Reference D.5) will be described in this section.

D.3.1 EMF-2158 Based Depletion Uncertainty The CASMO-4 depletion uncertainty is derived from the AREVA licensing topical report based on the extensive benchmarking that is documented within Reference D.1. Comparisons against critical experiments were performed by Studsvik with results reported in Table 2.1 of AREVAs CASMO4/MICROBURNB2 licensing topical report (Reference D.1). In addition, the beginning of cycle cold critical calculations reported in Table 2.2 of this same licensing topical report also provide comparisons to critical data. Results of these comparisons indicate that CASMO-4 results will have a standard deviation of [ ] k (Table 2.1 of Reference D.1) without depletion and a standard deviation of [ ] k (Table 2.2 of Reference D.1) when the majority of assemblies have been depleted*.

D.3.2 ISG Based Depletion Uncertainty Five percent of the reactivity difference from BOL (without gadolinia) to peak reactivity is used to estimate the isotopic uncertainty associated with depletion to peak reactivity, (i.e., the uncertainty in the uranium depletion, fission product production, and actinide production). The approach presented here is a conservative application of the 5% reactivity decrement approach originally suggested in Section 5.A.5.d of the Kopp memo (Reference D.4) and currently addressed in DSS-ISG-2010-01 (Reference D.5).

The reference bounding and limiting lattices used in this comparison are identified in Table B.1.

All lattices are depleted in-core and then evaluated at the limiting moderater temperature (4 ºC) in the fuel storage rack configuration. Figure D.3 illustrates the two reactivity decrement values used.

• The uncertainty of cold critical benchmarks effectively includes a depletion uncertainty since the majority of the bundles in the core have some depletion. It is noted, that an in-sequence critical has significant similarities to an in-rack calculation since the majority of the control blades remain inserted effectively surrounding the majority of the fuel with a strong neutron absorber on two sides.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-20 A BOL no gad solution for each lattice was completed by removing the gadolinium and maintaining the same uranium number density in the lattice.* The depletion reactivity decrement is determined by subtracting the peak in-rack k from the BOL no gad in-rack k. A second reactivity decrement representing the uncertainty in gadolinia content was also determined by subtracting the peak in-rack k from a value similar to the gadolinia free k at the peak reactivity exposure .

Based on the calculation process illustrated in Figure D.3, five percent of the burn-up reactivity decrement (kbu=0.05*k) and five percent of the residual gadolinia reactivity change (kgd=0.05*kg) are tabulated in Table D.4 for the limiting lattices. This assessment produces a maximum depletion uncertainty of 0.0055 k for the reference bounding lattices.

It is noted that this process will produce a larger penalty as the gadolinia content increases (either the number of rods or the concentration). However, increasing the gadolinia content within a given lattice will substantially decrease the peak in-rack k of the lattice as shown in Figure D.4.

• This is accomplished by setting the gadolinia number densities to zero with the CASMO CNU input.

The peak k-infinity values with no gadolinia use an in-core depletion with gadolinia to the maximum reactivity exposure, all gadolinia is then removed and an in-rack calculation is performed.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-21 Table D.4 Depletion Uncertainty Values for Limiting Lattices Peak BOL Peak kbu kgd kbu +

k k nogad k nogad * (0.05*k) (0.05*kg) kgd Top Zone Limiting Legacy Lattice 0.8619 0.9491 0.8777 0.0044 0.0008 0.0052 Bounding Lattice 0.8797 0.9638 0.8931 0.0042 0.0007 0.0049 Bottom Zone Limiting Legacy Lattice 0.8227 0.9385 0.8424 0.0058 0.0010 0.0068 Bounding Lattice 0.8790 0.9733 0.8954 0.0047 0.0008 0.0055

• The Peak k-infinity values with no gadolinia assume in-core depletion with gadolinia to the maximum reactivity exposure, all gadolinia is then removed and an in-rack calculation is performed.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-22 1.05 1.00 k-inf, In-Rack with Residual Gd k-inf, In-Rack without Gd 0.95 k-inf, In-Rack without Gd - BOL k-infinity (in-rack)

Burnup (k) decrement 0.90 Residual gadolina (kg) 0.85 0.80 0 5 10 15 20 25 30 Burnup (GWd/MT)

Figure D.3 Representation of the ISG Depletion Uncertainty Assessment AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-23

[

]

Figure D.4 Gadolinia Concentration Sensitivity AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-24 D.4 Conclusions and Overall Uncertainty The evaluation of GE8x8 fuel lattices as well as top and bottom lattices from the GE9x9, GE10x10, ATRIUM-10 (10x10), and ATRIUM 10XM (10x10) product lines demonstrated that the CASMO4 fuel storage rack calculations will produce reasonable results for these types of geometries. In addition, it has been demonstrated that reasonable results can be obtained for U-235 enrichment levels between 3.1 and 3.6 wt% U-235. These comparisons also indicate that [

].

When applied on a differential basis a k predicted by CASMO-4 agrees with the KENO V.a based k with a standard deviation of [ ] k, (see Section D.2.2.2). This can be combined with uncertainty estimates from EMF-2158(P) (Section D.3.1) or the estimated depletion uncertainty determined with the method from the interim staff guidance document (Section D.3.2) to produce a maximum combined value. A 95/95 uncertainty result is obtained by multiplying these uncertainty values by an appropriate multiplier. Since these values are independent they will be combined using the square root of the sum of the squares as shown below. This process results in a maximum combined uncertainty of [ ]. The 0.010 k adder used when defining the REBOL lattices conservatively bounds this CASMO-4 uncertainty value.

95/95 Combined Uncertainty Value 95/95 Uncertainty Multiplier Uncertainty Calculational [ ] 2.065 [ ]

(k based)

EMF-2158 Depletion [ ] 2.0 [ ] [ ]

Calculational [ ] 2.065 [ ]

(k based)

ISG Depletion --- --- 0.0055* [ ]

• This is not necessarily a 95/95 value; however, it is acceptable per Section IV.2.a of Reference D.5.

AREVA Inc.

ANP-3160(NP)

Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Revision 1 Criticality Safety Analysis for ATRIUM' 10XM Fuel Page D-25 D.5 References D.1 EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.

D.2 NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, USNRC, January 2001.

D.3 MIL-HDBK-5J, Metallic Materials and Elements for Aerospace Vehicle Structures, Department of Defense Handbook, January 2003.

D.4 Memorandum L. Kopp to T. Collins, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, NRC, August 19, 1998. (NRC -ADAMS Accession Number ML072710248)

D.5 Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01 Revision 0, Staff Guidance Regarding The Nuclear Criticality Safety Analysis For Spent Fuel Pools, (NRC - ADAMS Accession Number ML110620086).

AREVA Inc.

ENCLOSURE 3 AREVA Affidavit

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) SS.

CITY OF LYNCHBURG )

1. My name is Morris Byram. I am Manager, Product Licensing, for AREVA Inc.

(AREVA) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary. I am familiar with the policies established by AREVA to ensure the proper application of these criteria.

3. I am familiar with the AREVA information contained in the AREVA document ANP-3160(P), Revision 1, "Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM' 10XM Fuel," and referred to herein as "Document."

Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA Inc. for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(d) above.

7. In accordance with AREVA's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this *~

dayof ~ ,2015.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg.# 7079129 SHERRYL. MCFADEN Notery PUDllC Comm~nwellth of Vlrglnl1

' 7071129 My Comm1111on Expfret Ocl 31, 2018

ENCLOSURE 4 Suppressed Fuel Assembly Impact on Criticality Safety Analysis

Suppressed Fuel Assembly Impact on Criticality Safety Analysis During an NRC public meeting on November 10, 2015, with TVA representatives, the NRC asked about the applicability of the Browns Ferry Nuclear Plant (BFN) Spent Fuel Pool (SFP)

Criticality Safety Analysis (CSA) for a fuel assembly that had been suppressed for an entire 24 month cycle.

Response

The impact of operation with a suppression blade is addressed as part of the BFN ATRIUM 10XM CSA report ANP-3160(P), Revision 1, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM' 10XM Fuel, AREVA Inc., December 2015 (Included in Enclosure 1 of this submittal). Specifically, part of Assumption 7 of ANP-3160(P), Section 6.5, General CASMO-4 Modeling Assumptions, addresses power suppression from the perspective of the following: the effects are limited to only a small population of fuel bundles (i.e., the four fuel bundles in each affected control cell),

and the effect on lattice reactivity by the power gradient and the reduced power density. The reactivity effects are supported with a sensitivity analysis with the results provided in Table 6.6 of ANP-3160(P). The conclusion is reached that the uncontrolled depletion results are bounding for the ATRIUM 10XM reference bounding lattices.

E41

ENCLOSURE 5 Boral Neutron Absorber Aging Management Program

Boral Neutron Absorber Aging Management Program During an NRC public meeting on November 10, 2015, with TVA representatives, the NRC asked if Browns Ferry Nuclear Plant (BFN) had an aging management program for the Boral neutron absorbers used in the Spent Fuel Pool (SFP) storage racks and credited in the BFN SFP Criticality Safety Analysis (CSA).

Response

BFN does have an aging management program for Boral neutron absorbers used in the SFP storage racks. The topic of an aging management program for the Boral neutron absorbers in the SFP storage racks was addressed during the NRC review of the BFN License Renewal Application (LRA). The NRC review and acceptance of the BFN aging management program for Boral neutron absorbers used in the SFP storage racks is documented in NUREG-1843, Safety Evaluation Report Related to the License Renewal of the Browns Ferry Nuclear Plant, Units 1, 2, and 3, dated April 30, 2006 (ADAMS No. ML061030027). Applicable portions of NUREG-1843 describing the NRC review and acceptance of the BFN aging management program for SFP storage rack Boral neutron absorbers are as follows.

NUREG-1843, Section 2.4.2.1.2 (page 2-161) contains the following NRC evaluation of Request for Additional Information (RAI) 2.4-4.

In RAI 2.4-4, dated December 20, 2004, the staff stated that LRA Table 2.4.2.1 presents a list of component types that are part of the reactor building, the auxiliary and emergency systems of the NSSS, the biological shield, the spent fuel pool, the steam dryer/moisture separator storage pool, the reactor cavity reactor auxiliary equipment, the steel superstructure with metal siding and the built-up roof. Therefore, the staff requested the applicant to provide a description of the "Neutron-Absorbing Sheets" used for the spent fuel storage racks and confirm that they are part of the spent fuel storage racks listed in LRA Table 2.4.2.1.

In its response, by letter dated January 24, 2005, the applicant stated:

NUREG 1801,Section VII.A2.1-b, identifies "Spent Fuel Storage Racks -

neutron absorbing sheets" as a component type. In BFN LRA Section 2.3.3.27 Fuel Handling and Storage System (079)," it states that the spent fuel pool components are evaluated as structural components in Section 2.4.2.1 "Reactor Building Structure." BFN LRA Table 2.4.2.1 "Reactor Building Structure" identifies "Spent Fuel Storage Racks (includes new fuel storage racks)" as a component requiring aging management. The "Neutron Absorbing Sheet" is a component of the BFN spent fuel storage rack container tube wall and is comprised of Boral sandwiched within the stainless steel wall of each container tube.

The staff found the above response acceptable. Therefore, the staff's concern described in RAI 2.4-4 is resolved.

NUREG-1843, Table 3.3-1, (page 3-205) contains a row which identifies the Chemistry Control Program as the Aging Management Program (AMP) for the Boral neutron absorbing sheets.

E51

Boral Neutron Absorber Aging Management Program NUREG-1843, Section 3.3.2.2.10, (page 3-217) provides the following discussion.

3.3.2.2.10 Reduction of Neutron-Absorbing Capacity and Loss of Material due to General Corrosion The staff reviewed LRA Section 3.3.2.2.10 against the criteria in SRP-LR 3.3.2.2.10.

In LRA Section 3.3.2.2.10, the applicant addressed the further evaluation of programs to manage reduction of neutron-absorbing capacity and loss of material due to general corrosion, which could occur in the neutron absorbing sheets of the spent fuel storage rack in the spent fuel storage.

SRP-LR Section 3.3.2.2.10 states that reduction of neutron-absorbing capacity and loss of material due to general corrosion could occur in the neutron-absorbing sheets of the spent fuel storage rack in the spent fuel storage. The [Generic Aging Lessons Learned Report (GALL)] Report recommends further evaluation to ensure that these aging effects are adequately managed.

The applicant stated that boral is used as a neutron absorbing material in the spent fuel pools. Reduction of neutron absorbing capacity and loss of material due to general corrosion could occur in the boral neutron absorbing material in spent fuel storage racks.

The Chemistry Control Program manages general corrosion. An inspection of boral coupon test specimens was performed that confirmed no significant aging degradation had occurred and the neutron absorbing capability of the boral had not been reduced.

Reduction of neutron absorbing capacity and loss of material due to general corrosion will be managed by the Chemistry Control Program.

The staff reviewed the Chemistry Control Program and found that the program will adequately manage the effects of aging so that the intended functions will be maintained.

NUREG-1843, Section 3.5.2.1, (page 3-288) provides the following discussion regarding the Aging Management Review (AMR) for Boral.

In reference to LRA Table 3.5.2.2, the staff also requested the applicant to describe the AMR for Boral and to clarify whether stainless steel components are used to support the Boral. If the AMR supports the conclusion that Boral does not require aging management, but the stainless steel supports do, then the Chemistry Control Program would be an acceptable AMP for this item. If not, the applicant was requested to provide the technical basis for crediting the Chemistry Control Program as the appropriate AMP for Boral.

By letter dated October 8, 2004, the applicant submitted its formal response to the staff, stating that the Boral core is made up of a central segment of a dispersion of boron carbide in aluminum. This central segment is clad on both sides with aluminum to form a plate. The Boral plates are sandwiched between two stainless steel plates which are closure-welded form the container. Vent holes have been added to prevent the buildup of hydrogen gas between the stainless steel containers. If the stainless steel containers remain intact, the Boral core will be unaffected and will retain its neutron-absorbing capacity. The Chemistry Control Program will manage aging of the stainless steel E52

Boral Neutron Absorber Aging Management Program containers. With these clarifications, the staff concluded that this item is consistent with the GALL Report.

NUREG-1843, Section 3.5.2.3.2, (page 3-323) provides the following discussion regarding RAI 3.5-14.

In RAI 3.5-14, dated December 10, 2004, the staff stated that, with respect to the neutron-absorbing sheets in spent fuel storage racks, as described in LRA Section 3.3.2.2, the applicant stated that the Chemistry Control Program manages general corrosion and that an inspection of Boral coupon test specimens was performed at BFN that confirmed that no significant aging degradation had occurred and that the neutron-absorbing capacity of the Boral had not been reduced. Since it is implied that some Boral aging degradations had occurred at the time of inspection of the test specimens, the staff requested the applicant to discuss the basis for the above assertion that the neutron-absorbing capacity of the Boral will be maintained at an adequate level during the extended period of plant operation.

In its response, by letter dated January 31, 2005, the applicant stated:

A total of 16 boral coupons were placed in the Unit 3 spent fuel storage pool (SFSP) in October 1983. The coupons supplied by the rack manufacturer are of the same metallurgical condition as the high density fuel storage racks (HDFSR) in thickness, chemistry, finish, and temper. For the first six years of the planned fifteen year surveillance program, examination was to have taken place at two-year intervals. Accordingly, two coupons were removed in October 1985. Blisters were found upon examination, and because of this unexpected anomaly, three additional coupons were analyzed not finding any blisters. As a result of blisters found on the coupons removed in 1985, the surveillance program has been expanded to include monitoring the formation and behavior of these blisters.

These boral coupons are periodically removed from the fuel pool for testing and are evaluated for corrosion or other degradation of the neutron absorber plates by comparing various physical characteristics of the test coupons to baseline measurements taken when the coupons were installed. Also, a metallurgical engineer examines the coupons for general corrosion, local pitting, and bonding.

No further blisters, corrosion, or degradation has been identified in coupons evaluated through 2003.

The above response states that these Boral coupons are periodically removed from the fuel pool for testing and are evaluated for corrosion or other degradation of the neutron absorber plates by comparing various physical characteristics of the test coupons to baseline measurements taken when the coupons were installed. The response also implies that a metallurgical engineer periodically examines the coupons for general corrosion, local pitting, and bonding. Also, no further blisters, corrosion, or degradation have been identified in coupons evaluated through 2003; however, it was not clear to the staff whether these periodic inspections are ongoing activities that are an extension of the 1983 Boral Coupon Inspection Program covering Boral coupon test specimens or a separate AMP in addition to the Chemistry Control Program mentioned above. The applicant was requested to clarify the key parameters of this periodic inspection program or activity including the objective, scope, frequency, and inspection approach of the program.

E53

Boral Neutron Absorber Aging Management Program In its response, by letter May 24, 2005, the applicant stated that:

The Boral coupon inspection program was initiated in 1983 to implement the inspection and testing requirements of UFSAR Section 10.3.6; this checks the long-term behavior of the material of the high density spent fuel racks. The inspection is performed per BFN Technical Instruction (TI) TI-116, "High Density Fuel Storage System Surveillance Program." When the TI is performed, Boral coupons are removed from the spent fuel storage pool and examined by the Metallurgical Engineer in their original condition to determine if sampling of surface corrosion products is appropriate. Thickness measurements are obtained of each coupon and documented in accordance with the TI. If degradation is such that further investigation is warranted, a minimum of one coupon is selected to be unsheathed or opened. Prior to the unsheathing process, a dye penetrant test for indications on the outer surfaces of the coupon will be performed and is examined by the Metallurgical Engineer. The Metallurgical Engineer decides if further unsheathing of the coupons is required. The visual examination by the Metallurgical Engineer is documented on the appropriate forms of the TI. The current frequency for performing this TI is two years. The surveillance frequency is re-evaluated each time the surveillance is performed and can be changed based on the trend of the historical data results. The inspection of the Boral coupons will continue until such time as the trend of the historical data results collected provides a basis to discontinue the inspections.

Based on its review, the staff found the applicant's response to RAI 3.5-14 acceptable.

Therefore, the staffs concern described in RAI 3.5-14 is resolved.

E54