Text

License Amendment Request Regarding Criticality Safety Analysis of Spent Fuel Pool Storage of Atrium 11 Fuel (TS-534)
	ML21150A022
Person / Time
Site:	Browns Ferry
Issue date:	05/29/2021
From:	Polickoski J; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML21150A021	List: CNL-21-050, License Amendment Request Regarding Criticality Safety Analysis of Spent Fuel Pool Storage of Atrium 11 Fuel (TS-534);
References
	CNL-21-050
	Download: ML21150A022 (153)
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Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Attachment 1 to the Enclosure 1101 Market Street, Chattanooga, Tennessee 37402 CNL-21-050 May 29, 2021 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:

Browns Ferry Nuclear Plant, Units 1, 2, and 3, License Amendment Request Regarding Criticality Safety Analysis of Spent Fuel Pool Storage of ATRIUM 11 Fuel (TS-534)

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, Tennessee Valley Authority (TVA) is submitting a request for an amendment to Renewed Facility Operating Licenses DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3.

The license amendment request (LAR) proposes to expand the applicability of the spent fuel pool criticality safety analysis of record for all three BFN units to include the ATRIUM 11' 1 fuel design.

The enclosure to this submittal provides a description and technical evaluation of the proposed change, a regulatory evaluation, and a discussion of environmental considerations for the proposed change. In support of the technical evaluation in the enclosure, Attachment 1 to the enclosure contains Framatome, Inc., (Framatome) Report, ANP-3910P, Revision 2, Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel. to the enclosure contains information that Framatome considers to be proprietary in nature pursuant to 10 CFR 2.390, Public inspections, exemptions, requests for withholding, paragraph (a)(4), and requests that it be withheld from public disclosure. Attachment 2 contains a non-proprietary version of Attachment 1. Attachment 3 provides the Framatome ANP-3910P 1

ATRIUM 11 is a trademark or registered trademarks of Framatome, Inc., its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited.

Other names may be trademarks of their respective owners.

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Attachment 1 to the Enclosure

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Attachment 1 to the Enclosure U.S. Nuclear Regulatory Commission CNL-21-050 Page 2 May 29, 2021 affidavit supporting this proprietary information withholding request. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390.

Accordingly, TVA requests that the information, which is proprietary to Framatome, be withheld from public disclosure in accordance with 10 CFR Section 2.390. Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Framatome affidavit should reference ANP-3910P and should be addressed to Alan Meginnis, Framatome, Manager, Product Licensing, 2101 Horn Rapids Road, Richland, Washington 99354.

To support the Unit 2 Cycle 23 fuel receipt milestone dates, TVA requests NRC approval of this LAR by May 30, 2022, with implementation within 60 days following NRC approval.

TVA determined that there are no significant hazards consideration associated with the proposed change and that the change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosure to the Alabama Department of Public Health.

There are no new regulatory commitments contained in this letter. If you have any questions regarding this submittal, please contact Kimberly D. Hulvey, Senior Manager, Fleet Licensing at 423-751-3275.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 29th day of May 2021.

Respectfully, James T. Polickoski Director, Nuclear Regulatory Affairs Enclosure cc: See Page 3 Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Attachment 1 to the Enclosure

Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Attachment 1 to the Enclosure U.S. Nuclear Regulatory Commission CNL-21-050 Page 3 May 29, 2021 Enclosure Evaluation of the Proposed Change cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Project Manager - Browns Ferry Nuclear Plant State Health Officer, Alabama Department of Public Health Proprietary Information - Withhold Under 10 CFR 2.390 This letter is decontrolled when separated from Attachment 1 to the Enclosure

Enclosure Evaluation of the Proposed Change

Subject:

Browns Ferry Nuclear Plant, Units 1, 2, and 3, License Amendment Request Regarding Criticality Safety Analysis of Spent Fuel Pool Storage of ATRIUM 11 Fuel (TS-534) 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend the Renewed Facility Operating Licenses for Browns Ferry Nuclear Plant (BFN), Unit 1 (DPR-33), Unit 2 (DPR-52), and Unit 3 (DPR-68).

The proposed amendment requests approval to expand the spent fuel pool (SFP) criticality safety analysis (CSA) of record for all three BFN units to include the ATRIUM 11' 1 fuel design.

Tennessee Valley Authority (TVA) intends to transition BFN to the ATRIUM 11 fuel design commencing with the Unit 2 Cycle 23 reload batch to be loaded in the spring of 2023. Unit 3 would transition to ATRIUM 11 fuel in the spring of 2024, followed by Unit 1 in the fall of 2024.

2.0 DESCRIPTION

OF CHANGES TVA intends to begin utilizing the ATRIUM 11 design in reload quantities in all three BFN units, beginning with Unit 2 in the spring of 2023. TVA is pursuing the ATRIUM 11 fuel type due to the improved fuel cycle economics, as well as improved fuel reliability features.

As noted in the Reference 1 meeting summary, TVA informed the NRC staff that a license amendment request (LAR) would be submitted to request approval to expand the SFP CSA of record for all three BFN units to include the ATRIUM 11 fuel design. Approval of this request grants the ability to store fresh and exposed ATRIUM 11 fuel in the BFN SFP storage racks. A separate LAR will be submitted to add new Framatome methodologies to BFN Technical Specification (TS) 5.6.5.b, in TS 5.6.5, Core Operating Limits Report, and to implement the TS changes outlined in Technical Specification Task Force (TSTF) Traveler TSTF-564. As noted in Reference 1, the review of one LAR is not dependent on the review of the other LAR, as there is no interdependence of the technical information presented in them.

3.0 TECHNICAL EVALUATION

As documented in Reference 2, the NRC evaluated the BFN CSA (Reference 3), which was docketed as a supplement (Reference 4) to the overall Extended Power Uprate LAR approved in Reference 2. That CSA included the storage of ATRIUM 10XM' 2 fuel.

Attachments 1 and 2 provide the details of the CSA analysis performed for ATRIUM 11 fuel storage in the BFN SFP racks. The evaluation demonstrates that the current BFN TS 4.3.1.1, in TS 4.3.1, Criticality, remains applicable for ATRIUM 11 fuel.

The specific document titles are shown in the table below for convenience. A brief summary of what is included in the proprietary attachment follows. The non-proprietary attachment contains the same material as the proprietary attachment with the proprietary information redacted.

1 ATRIUM 11 is a trademark or registered trademarks of Framatome, Inc., its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited.

Other names may be trademarks of their respective owners.

2 ATRIUM 10XM is a trademark or registered trademarks of Framatome, Inc., its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

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Attachment Title 1 ANP-3910P, Revision 2, Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel, Framatome Inc., April 2021 (proprietary) 2 ANP-3910NP, Revision 2, Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel, Framatome Inc., April 2021 (non-proprietary) 3 Affidavit : ANP-3910P, Revision 2, Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel ANP-3910P, Revision 2, extends the previously reviewed Reference 3 CSA to address ATRIUM 11 fuel. Per Reference 2, the NRC had explicitly stated that the criteria provided in Table 2.1 of ANP-3160 are acceptable for qualifying future ATRIUM 10XM fuel assemblies for storage in the BFN SFPs. Because the safety conclusion only applies to the ATRIUM 10XM fuel type, transitioning to ATRIUM 11 requires justification of the applicability of the previously approved criteria, or the development of new criteria specific to the ATRIUM 11 fuel type.

ANP-3910P notes that software changes were made relative to the code versions used in the ANP-3160P analysis, most notably the version of SCALE that was utilized, which incorporated several methodology changes. Section 5 of ANP-3910P, Revision 2, provides a detailed discussion of these methodology changes. ANP-3910P also provides updated versions of the analysis presented in ANP-3160P, Appendix C, to reflect benchmarking with the new version of SCALE, and the CASMO-4/KENO V.a benchmarking presented in ANP-3160P, Appendix D, to include the ATRIUM 11 fuel type.

The NRC Requests for Additional Information associated with the review of ANP-3160P were considered as part of the evaluations in ANP-3910P, Revision 2.

4.0 REGULATORY EVALUATION

4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA The BFN units were designed and constructed based on the proposed General Design Criteria (GDC) published by the Atomic Energy Commission (AEC) in the Federal Register (32 FR 10213) on July 11, 1967 (hereafter called draft GDC). The AEC published the final rule that added Appendix A to 10 CFR Part 50, GDC for Nuclear Power Plants, in the Federal Register (36 FR 3255) on February 20, 1971 (hereafter called final GDC). As discussed in Appendix A of the Updated Final Safety Analysis Report (UFSAR), the licensee has made changes to the facility over the life of the plant that may have invoked the final GDC. The extent to which the final GDC have been invoked can be found in specific sections of the UFSAR and in other design and licensing basis documentation. As noted in Reference 2, the draft GDC 66 under which BFN was licensed is equivalent to the current GDC 62. Therefore, GDC 62 is applicable to this LAR review.

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Appendix A to 10 CFR 50 states:

Criterion 62-Prevention of criticality in fuel storage and handling. Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

The CSA has been performed to demonstrate that k-eff will remain less than or equal to 0.95 for storage of ATRIUM 11 fuel.

Paragraph (b)(4) of 10 CFR 50.68, Criticality Accident Requirements, states that the k-eff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. BFN UFSAR Section 10.3 notes that BFN has chosen to comply with the criticality requirements specified in10 CFR 50.68(b). The new SFP CSA provided in Attachments 1 and 2 to this submittal demonstrates that this requirement is met for storage of ATRIUM 11 fuel at BFN.

TVA has determined that the proposed change does not require any exemptions or relief from regulatory requirements, and does not affect conformance with the intent of any GDC differently from that described in the Updated Final Safety Analysis Report. The current TS 4.3.1.1 is applicable for storage of ATRIUM 11 fuel at BFN.

4.2 PRECEDENT In Reference 2, as part of the BFN extended power uprate, the NRC approved the Framatome criticality analysis of the storage of Framatome fuel types up to and including ATRIUM 10XM at BFN. This request builds on this precedent to extend the applicability of the BFN CSA analysis to include the ATRIUM 11 fuel type.

4.3 NO SIGNIFICANT HAZARDS CONSIDERATIONS Tennessee Valley Authority (TVA) is submitting a request for an amendment to Renewed Facility Operating Licenses DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3. The license amendment request (LAR) proposes to expand the applicability of the spent fuel pool criticality safety analysis of record for all three BFN units to include the ATRIUM 11' fuel design.

TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed amendment involves a new SFP CSA. The new SFP CSA does not involve a physical change to any plant system nor does it involve a change to any of the accident mitigation features previously evaluated. The proposed amendment does not change or modify the fuel handling processes, SFP storage racks, or the SFP cooling and cleanup system. The proposed amendment will not significantly increase the probability of a fuel mispositioning event nor change margin to criticality because the new SFP CSA demonstrates that fuel assemblies that meet the TS requirements can be stored in any SFP E3 of 5

location without restriction. There is no dose consequence associated with an abnormal condition because the CSA acceptance criteria preclude criticality and does not involve a radiological release.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed amendment involves a new SFP CSA. The new SFP CSA does not involve a physical change to any plant system. The proposed amendment does not change or modify the fuel handling processes, SFP storage racks, or the SFP cooling and cleanup system.

The proposed amendment does not change the method of fuel movement or fuel storage and does not create the potential for a new accident, new fuel mispositioning event, nor modify margin to criticality.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The regulations at 10 CFR 50.68, Criticality Accident Requirements, require the SFP storage racks to maintain the effective neutron multiplication factor, k-eff, less than or equal to 0.95 when fully flooded with unborated water, which includes an allowance for uncertainties. Therefore, for criticality, the required safety margin is 5 percent, including a conservative margin to account for engineering and manufacturing uncertainties. The new SFP CSA continues to satisfy this requirement with no impact to margin.

The new SFP CSA does not affect spent fuel heat generation or the SFP cooling systems.

In addition, the radiological consequences of a dropped fuel assembly are unaffected by the implementation of the new spent fuel pool CSA.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, TVA concludes the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92 (c). Accordingly, a finding of no significant hazards consideration is justified.

4.4 CONCLUSION

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operating in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

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5.0 ENVIRONMENTAL CONSIDERATION

A review has determined the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase to individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. NRC Letter to TVA, Summary of June 8, 2020, Partially Closed Meeting with Tennessee Valley Authority to Discuss a Planned License Amendment (EPID L-2020-LRM-0023),

dated June 25, 2020 (ML20162A153)

2. NRC Letter to TVA, Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Issuance of Amendments Regarding Extended Power Uprate (CAC Nos. F6741, MF6742, AND MF6743), dated August 14, 2017 (ML17032A120)

3. ANP-3160P, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 10XM Fuel, Revision 1, December 2015 (ML15351A098, non-public)

4. TVA Letter to NRC, Proposed Technical Specification (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 1, Spent Fuel Pool Criticality Safety Analysis Information, dated December 15, 2015 (ML15351A097)

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ATTACHMENT 1 Framatome Report ANP-3910P, Revision 2 Browns Ferry Nuclear Plant, Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel (Proprietary)

CNL-21-050

ATTACHMENT 2 Framatome Report ANP-3910NP, Revision 2 Browns Ferry Nuclear Plant, Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel (Non Proprietary)

CNL-21-050

Controlled Document Browns Ferry Nuclear Plant ANP-3910NP Revision 2 Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel April 2021

0414-12-F04 (Rev. 004, 04/27/2020)

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 Page D-13 No technical changes made. A missed redaction has been corrected.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page ii Contents Page

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 ................................................................................ 5-3 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 Storage Rack Sensitivities ........................................................... 6-2 6.1.4 Reactivity Comparison of the Boral Rack Models ........................ 6-3 6.2 Boron Content......................................................................................... 6-3 6.3 Fuel Assembly Modeling ......................................................................... 6-3 6.4 Lumped Fission Products ....................................................................... 6-4 6.5 Rack to Rack Interfaces .......................................................................... 6-4 6.6 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 ........................................................... 7-4 7.4 Other Conditions ..................................................................................... 7-4 7.4.1 Assembly Rotation ....................................................................... 7-4 7.4.2 Assembly Lean............................................................................. 7-5 7.4.3 Blister Formation .......................................................................... 7-5 7.4.4 Fuel Rod Creep ............................................................................ 7-5 7.4.5 Efficacy of Boral ........................................................................... 7-6 7.4.6 Fuel Channel Growth ................................................................... 7-6 7.4.7 Spacer Growth ............................................................................. 7-6

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page iii 7.5 Normal Fuel Handling ............................................................................. 7-7 7.6 Accident Conditions ................................................................................ 7-8 7.7 Manufacturing and Other Uncertainties ................................................ 7-10 7.8 Determination of Maximum Rack Assembly k-eff (k95/95) ...................... 7-11

8.0 REFERENCES

.................................................................................................. 8-1 APPENDIX A SAMPLE CASMO-4 INPUT ................................................................ A-1 APPENDIX B REACTIVITY COMPARISON FOR ASSEMBLIES USED IN THE BROWNS FERRY REACTORS ................................................. 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 APPENDIX E CRITICALITY ANALYSIS CHECKLIST .............................................. E-1

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page iv List of Tables Table 2.1 Criticality Safety Limitations for ATRIUM 11 Fuel Assemblies Stored in the Browns Ferry Spent Fuel Storage Pool ........................................... 2-5 Table 3.1 Compliance with Interim Staff Guidance Document DSS-ISG-2010-01 Rev. 0 .......................................................................... 3-6 Table 4.1 ATRIUM 11 Fuel Assembly Parameters .................................................... 4-3 Table 4.2 Fuel Storage Rack Parameters ................................................................. 4-5 Table 6.1 Impact of Channel Thickness on In-Rack Reactivity.................................. 6-9 Table 6.2 Storage Rack Model Sensitivity ................................................................. 6-9 Table 6.3 Comparison of Modeling Options for the Boral Rack ............................... 6-10 Table 6.4 In-Rack k Sensitivity to In-core Depletion Fuel Temperature ................. 6-11 Table 6.5 In-Rack k Sensitivity to In-core Depletion Power Density ...................... 6-12 Table 6.6 In-Rack k Sensitivity to In-Core Controlled Depletion ............................ 6-13 Table 7.1 Summary of CASMO-4 Maximum In-Rack Reactivity Results ................. 7-13 Table 7.2 Summary of KENO V.a Maximum In-Rack Reactivity Results ................ 7-14 Table 7.3 Manufacturing Reactivity Uncertainties ................................................... 7-15

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page v List of Figures Figure 2.1 Overview of the Browns Ferry SFSP Criticality Safety Analysis for ATRIUM 11 Fuel ........................................................................................ 2-7 Figure 2.2 ATRIUM 11 Reference Bounding Assembly .............................................. 2-8 Figure 2.3 ATRIUM 11 Allowed Locations for the Minimum Required Gd Rods ......... 2-9 Figure 4.1 Representative ATRIUM 11 Fuel Assembly .............................................. 4-6 Figure 4.2 Browns Ferry Spent Fuel Pool Layout ....................................................... 4-7 Figure 4.3 Schematic Representation of a Section of High Density Storage Rack ..... 4-8 Figure 4.4 High Density Storage Rack Storage Cell Geometry .................................. 4-9 Figure 6.1 Single Cell CASMO-4 Model for the High Density Boral Rack ................ 6-14 Figure 6.2 Explicit 2x2 Geometry Model for High Density Boral Rack ...................... 6-15 Figure 6.3 Schematic of Rack to Rack Interfaces ..................................................... 6-16 Figure 6.4 Impact of Void History Depletion on In-Rack k-infinity ............................. 6-17

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page vi Nomenclature Acronym Definition A11B ATRIUM 11 bottom lattice geometry zone (i.e. Zone 1)

A11M ATRIUM 11 middle lattice geometry zone (i.e. Zone 2)

A11T ATRIUM 11 top lattice geometry zone (i.e. Zone 3)

AEC Atomic Energy Commission AOA area of applicability BAF bottom of active fuel BFN Browns Ferry Nuclear plant BOL beginning of life BORAL neutron absorber composed of boron dispersed within aluminum BWR boiling-water reactor CFR code of federal regulations EALF the energy of the average lethargy causing fission FPM fuel preparation machine FSAR final safety analysis report GDC general design criteria GE General Electric company GWd energy unit, giga-watt-day GWd/MTU exposure unit, giga-watt-day per metric-ton-uranium HTC Haut Taux de Combustion H/X moderating ratio, atomic ratio of hydrogen (H) to fissile isotopes (X)

ISG interim staff guidance document (Reference 9) k-eff effective neutron multiplication factor (aka k-effective) k infinite lattice neutron multiplication factor (aka k-infinity)

LWR light water reactor LUA lead use assembly PLFR part-length fuel rod NCS nuclear criticality safety NPM non-parametric margin NRC Nuclear Regulatory Commission, U.S. (also USNRC)

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page vii RBL reference bounding lattice (uniform enrichment lattice)

REBOL reactivity-equivalent at beginning of life (fresh fuel, no Gd2O3)

SFP spent fuel pool (same as SFSP)

SFSP spent fuel storage pool SRP standard review plan (NUREG-0800)

SS stainless steel TAF top of active fuel UFSAR updated final safety analysis report Z4B beta-quenched zircaloy alloy

%TD percent of theoretical density

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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 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 (BFN) 1, 2, and 3 spent fuel storage pools (SFSP). Each spent fuel pool has a similar configuration including the use of the same storage module designs. This analysis is performed on a bounding basis and is applicable to all three of the spent fuel storage pools.

The initial reviewed and approved criticality safety analysis for these racks is documented in Reference 1. The previous analysis of record has been audited and approved by the NRC (see References 2 and 3).

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 Plants (Units 1, 2 and 3). This reference bounding assembly is based on a Framatome Inc.

ATRIUM 11 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.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 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., equal to or more reactive than) all fuel designs previously used or planned to be used in Units 1, 2 and 3 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 reference bounding lattices (RBL) and reactivity equivalent at beginning of life (REBOL) lattices. CASMO-4 is also used in defining the gadolinia manufacturing uncertainty. Benchmarking against criticality experiments is included for the KENO V.a code. Multiple justifications for the use of the CASMO-4 code are provided including validation using code-to-code comparisons with the benchmarked KENO V.a code. More detail on methodology and benchmark/validation is provided in Chapter 5 and Appendices C and D.

The calculations documented herein demonstrate that the ATRIUM 11 reference bounding assembly design has been selected to be equivalent to or 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 Framatome 10x10 lattice geometries and enrichments as detailed in Appendix B *. This criticality safety analysis shows that future ATRIUM 11 assemblies meeting the storage requirements established by this criticality analysis can be safely stored with these previously manufactured assemblies.

Various LUAs were also evaluated.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 2-2 The reference bounding assembly is defined with three U-235 enrichment / gadolinia concentration zones with geometry transitions at [ ] inches. The bottom enrichment and gadolinia zone (A11B) is defined to extend from the bottom of active fuel (BAF) up to the first geometric transition boundary and contains a total of

[ ] fuel rods. The middle enrichment / gadolinia zone (A11M) extends between the two transition boundaries and contains a total of [ ] fuel rods. The top enrichment /

gadolinia zone (A11T) extends above the second geometric transition boundary to the top of active fuel (TAF) of the fuel assembly and contains a total of [ ] fuel rods.

These axial zones are illustrated in Figure 2.2.

Three REBOL lattices have been defined to represent the corresponding lattices of the reference bounding assembly in KENO calculations. [

]

This analysis includes manufacturing uncertainties for the ATRIUM 11 fuel design and the fuel pool storage racks. In addition to the manufacturing uncertainties; code modeling uncertainties, reactivity increases due to accident or abnormal 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 found in Chapter 7.

This analysis demonstrates that the reference ATRIUM 11 fuel assembly does not exceed an array k-eff of 0.95 in the Browns Ferry spent fuel storage pool. As defined in Table 2.1, ATRIUM 11 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 11 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. This can be established using the storage rack model in the CASMO-4 lattice physics code as described in Appendix A.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 2-3 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 the moderator temperature which gives the highest reactivity for the limiting rack model in 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 array is assumed to be at its lifetime maximum reactivity level. There is no assumption of a specific burnup profile for the discharged assemblies

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.

The analysis takes into account storage with or without fuel channels. *

Neutron absorption in fuel assembly structural components (spacers , tie plates, plenum springs, etc.) is neglected.

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

[

] introducing significantly more conservatism than if it had been treated as an actual uncertainty.

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

An adder has been included to account for potential future Boral blistering.

The k95/95 result is based upon a 2x2 array of fuel in the storage rack (instead of the less limiting k associated with the actual rack configurations (see Table 6.3).

The array k value is higher with a fuel channel present. This analysis supports the storage of fuel in the Browns Ferry SFSP using the Framatome advanced thick/thin channel, a uniform 100 mil channel, or unchanneled.

It is conservative to neglect the spacers because: 1) this spent fuel pool contains no soluble boron and 2) the region around the fuel rods is under-moderated; therefore, neglecting the spacers places more absorber free water within the calculational model. In addition, the Inconel present in the spacer is a stronger neutron absorber than water.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 2-4 This analysis demonstrates that all fuel assemblies delivered to the Browns Ferry Nuclear Plant can be safely stored in the spent fuel storage pools. Future ATRIUM 11 fuel designs that meet the design requirements specified in Table 2.1 or that can be shown to be less reactive than the reference bounding assembly can be safely stored in the Browns Ferry spent fuel pools. The array k-eff determined herein for the reference assembly, including all uncertainties, biases, manufacturing tolerances and worst accident or abnormal loading conditions is 0.919 (as detailed in Section 7.8 and Figure 2.1).

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 2-5 Table 2.1 Criticality Safety Limitations for ATRIUM 11 Fuel Assemblies Stored in the Browns Ferry Spent Fuel Storage Pool

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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.)

Zone Lattice Geometry Number of Fuel Rods Max. in-rack k 3 A11T [ ] 0.8825 2 A11M [ ] 0.8825 1 A11B [ ] 0.8825 Spent Fuel Storage Rack The spent fuel storage rack design parameters and dimensions are provided in Table 4.2.

[

]

It is acceptable for the minimum 9 Gd rods to be face or diagonally adjacent to other Gd rods that are not part of the minimum 9.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 2-7 Figure 2.1 Overview of the Browns Ferry SFSP Criticality Safety Analysis for ATRIUM 11 Fuel

[ ]

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 2-8 Figure 2.2 ATRIUM 11 Reference Bounding Assembly

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 2-9 Figure 2.3 ATRIUM 11 Allowed Locations for the Minimum Required Gd Rods

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 3-1 3.0 REGULATORY CRITICALITY SAFETY CRITERIA AND GUIDANCE Section 9.1.1 of the Standard Review Plan (Reference 4) 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 (GDC) 62 and portions of 10 CFR 50.68, Reference 5. 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 Unit 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 4.0. 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

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.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 3-2 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 that criticality will not occur within the basis of the existing spent fuel storage configuration for both previously manufactured (or planned) fuel and ATRIUM 11 designs to be provided in the future; therefore the intent of Criterion 66 (and GDC 62) is met.

10 CFR 50.68(a) (Reference 5) 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 SFSP 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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 3-3 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 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 not exceeding 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 SFSP. 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.

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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 SFSP.

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 since this is a criticality analysis only.

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 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 10 CFR 50.68(b) compliance statement in Section 10.3 Spent Fuel Storage" of the Browns Ferry FSAR.

This criticality safety analysis complies with items 1-7 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 issued document DSS-ISG-2010-01 Revision 0 (Reference 9) 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 evaluation checklist from Appendix C of Reference 15 has also been included as Table E.1.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 3-5 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 8).

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 7).

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 6).

NEI 12-16 Revision 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants, (Reference 15).

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 3-6 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 reactivity of the lattices of the ATRIUM Appendix B submittal to verify that it 11 reference bounding assembly is greater demonstrates that the NCS than or equal to the reactivity of the lattices analysis adequately bounds of all previously loaded fuel assemblies, all designs, including including variations due to damaged and variations within a design. modified assemblies.

the staff should verify each As discussed above, the ATRIUM 11 Section 2.0 application includes a portion reference bounding assembly is shown to Appendix B of the analysis that bound all previous designs. Compliance demonstrates that the fuel with the requirements listed in Table 2.1 assembly used in the ensures that future ATRIUM 11 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 11 reference Section 2.0 assembly design should be bounding assembly (and corresponding Appendix B assessed, lattices) is justified as described above.

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 Appendices create the isotopic number incore depletion is used to establish the B&D densities used in the inrack lifetime maximum reactivity condition 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 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.

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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. [

percent of the reactivity decrement to the burnup of interest is an acceptable

] Two independent estimates of assumption.

the depletion uncertainty were used in this should only be construed evaluation. One of these methods is as covering the uncertainty in consistent with the 5% reactivity decrement the isotopic number described in Reference 9 (except that it densities includes an additional component for the gadolinia uncertainty).

IV.2.b Reactor Parameters Sensitivity comparisons are included in Section 6.6 Section 6.6 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 (Section 6.6 depletion analysis are Assumption 2, Table 6.4) appropriate for the specific

Moderator Temperature/Void History conditions. (Section 6.6 Assumption 3, Figure 6.4);

Power Density (Section 6.6 Assumption 4, Table 6.5); and

Rodded Depletion (Section 6.6 Assumption 7, Table 6.6)

IV.2.c Burnable Absorbers Only integral burnable absorbers have been used in the Browns Ferry reactors. [

the staff should verify that Table 2.1 each application includes a Figure 2.3 portion of the analysis that demonstrates that the Section 7.1 treatment of burnable absorbers in the depletion Appendix B analysis is appropriate for the specific conditions.

]

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ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.2.d Rodded Operation Assumption 7 of Section 6.6 addresses Section 6.6 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 for portion of the analysis that the ATRIUM 11 reference bounding lattices.

demonstrates its treatment of 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 Section 6.1. Appendix D the staff should verify that Comparisons in Table 6.3 and Table 6.2 each application includes a demonstrate that the infinite 2x2 B-10 only portion of the analysis that model is more reactive than the demonstrates that the rack corresponding explicit models.

model analysis used in its submittal is appropriate for its 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.

IV.3.b.ii The efficiency of the neutron The Boral is modeled using the licensees Section absorber should be design minimum Boron-10 areal density. 7.4.5 established, especially Neutron self-shielding and streaming are considering the potential for addressed in Section 7.4.5.

self-shielding and streaming.

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ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.3.b.iii Any degradation should be A conservative blister model has been Section 7.4.3 modeled conservatively, incorporated to account for Boral blistering. & 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 Section 6.1.4 been shown to be less reactive than the 2x2 and 6.5 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.

There is no significant difference between IV.3.c.i Absent a determination of a Section 6.1.4 the 13x13 racks and the 13x17 racks.

set of biases and 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. The and 7.5 each application includes a 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.

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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 and the staff should verify that in Appendix C. Section 5.1 also provides a Appendix C 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.

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

Haut Taux de Combustion (HTC) Critical Experiment Data, issued September 2008

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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.

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ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections IV.4.d Lumped Fission Products The primary components of the Browns Section 6.4 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). Since that the lumped fission CASMO-4 includes the use of lumped products used in its validation fission products for depletion calculations, are appropriate for its specific the impact of the predicted keff is conditions.

represented in the benchmark uncertainty values (EMF-2158). Therefore, the impact of the CASMO-4 lumped fission product modeling is accounted for when the depletion uncertainties are applied in the REBOL generation process and are appropriately reflected in the KENO calculations and k95/95 result.

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 The NRC performed an audit on the Browns N/A Ferry spent fuel pool criticality work the staff should verify that performed for a previous fuel type. This is for cited precedents, the discussed in the ADAMS document application includes a portion ML17032A120 (Sections 2.8.6). (It is noted of the analysis that that the conservatism associated with demonstrates the lumped fission products has not been commonality of the precedent applied in this evaluation).

to the submittal, with any differences identified and justified with respect to the use of the precedent.

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ISG Applicable USNRC Guidance Compliance with USNRC Guidance Section Sections 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 6.6 addressed in the report.

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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 4-1 4.0 FUEL AND STORAGE ARRAY DESCRIPTION The Browns Ferry spent fuel pools contain a number of different product lines GE 7x7 to Framatome 10x10 fuel, as well as various LUAs. All of these designs are stored in one or more of the Browns Ferry spent fuel storage pools.

The ATRIUM 11 is the most recent fuel product line. For this reason, the ATRIUM 11 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 11 fuel assembly is an 11x11 fuel rod array with an internal square water channel in the center of the assembly (taking the place of nine fuel rod locations). The assembly contains two different lengths of part-length fuel rods (PLFR). Therefore, the assembly is defined by three different lattice geometry zones. The top lattice geometry will apply above the long PLFR fueled boundary. The middle lattice geometry applies to the zone above the top of the short PLFRs and below the top of the long PLFRs. Finally, the bottom lattice geometry will apply below the top of the short PLFRs.

The ATRIUM 11 mechanical design parameters are summarized in Table 4.1 and a representation of the ATRIUM 11 assembly design is provided in Figure 4.1. The ATRIUM 11 fuel in the Browns Ferry Nuclear Plants will use the Framatome advanced (i.e., thick/thin) fuel channel design.

4.2 Fuel Storage Racks Each of the Browns Ferry spent fuel pools provides the capability to store 3471 BWR 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.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 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 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.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 4-3 Table 4.1 ATRIUM 11 Fuel Assembly Parameters

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 4-4 Table 4.1 ATRIUM 11 Fuel Assembly Parameters (Continued)

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 4-5 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).

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 4-6 (Assembly length and number of spacers has been reduced for pictorial clarity.)

Figure 4.1 Representative ATRIUM 11 Fuel Assembly

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 4-7 Fuel Prep. Fuel Prep.

Sipping Can Storage Sipping Can Storage Machine Machine N Cask Pad Area (not to scale)

Figure 4.2 Browns Ferry Spent Fuel Pool Layout (Unit 2 and 3 layout shown, Unit 1 is mirror symmetric with the cask pad area on top)

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 4-8 Stainless Steel Closure Plate OUTSIDE (outside of rack)

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

Boral Tube Open Cell Boral Tube High Density Stainless Steel Boral Racks (outer box)

Multi-Purpose Storage Racks Boral Tube Open Cell Open Cell Stainless Steel Closure Plates (outside of rack)

Boral Tube Open Cell Boral Tube Open Cell Boral Tube Open Cell (not to scale)

Figure 4.3 Schematic Representation of a Section of High Density Storage Rack

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 4-9 (not to scale)

Figure 4.4 High Density Storage Rack Storage Cell Geometry

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 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 6.2 Modular Code System (Reference 10). The SCALE driver module CSAS5 uses the ENDF/B-VII.1 continuous energy data library. Framatome has benchmarked KENO V.a in accordance with NUREG/CR-6698 (Reference 11) 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 ENDF-B-VII.1 continuous energy library a KENO V.a bias magnitude of 0.001 and a standard deviation of 0.0023 will be used (Appendix C).

KENO V.a is run on the Framatome scientific computer cluster using the Linux operating system. The hardware and software configurations are governed by Framatome 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 6.2 code system is employed to:

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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 5-2 manner that is consistent with Framatomes NRC approved CASMO-4/MICROBURN-B2 methodology (Reference 12). CASMO-4 has been approved at Browns Ferry Nuclear Plant for BWR calculations and is included as a methodology reference (via Reference

12) in Section 5.6.5.b of the Browns Ferry Nuclear Plant Technical Specifications. The CASMO-4 computer code is controlled by Framatome procedures and the version used in this analysis meets the requirements of Reference 12.

In this analysis CASMO-4 is employed to:

Perform in-core isotopic depletion at characteristic void history levels, [

] for bottom geometry lattices and [ ] for the middle and top geometry lattices. Use of CASMO-4 for in-core depletion is consistent with its application in EMF-2158(P)(A) (Reference 12).

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, [

]

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.

It has been confirmed that all significant CASMO-4 cases have properly converged. For the KENO runs, it is not expected that every Shannon entropy convergence test will meet the acceptance criterion - this is especially true for cases with infinite boundary

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 5-3 conditions. This analysis has followed the industry standard that the k value will have reasonable convergence as long as the KENO standard deviation (sigma) is 0.0002.

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 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.

Parameter Benchmark Values Values in this Analysis Enrichment (wt% U-235) 2.35 to 4.74 [ ]

Moderating Ratio (H/X) 48 to 455 121 to 131 Energy of the Average Lethargy 0.073 to 0.69 0.19 to 0.25 Causing Fission (eV)

For the CASMO-4 qualification, ATRIUM 11 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.

[ ]

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 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. This rack will be modeled as an array of 2 tube cells and 2 non-tube cells. Except for the single cell model, the rack models described below were implemented in KENO V.a and the reactivity results are provided in Table 6.3. These models use infinite periodic boundary conditions in the x, y, and z directions.

6.1.1 Single Cell Model Description This is the model used in the CASMO-4 calculations. The primary simplifying assumptions are described below. Figure 6.1 provides an illustration of the geometry for the single cell model.

Boral Plate: The Boral plate is modeled as boron-10 only; i.e., the aluminum, carbon, and boron-11 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).

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.

Cell Pitch and Water Gaps: Average cell pitch and average water gap values are used in this model.

The calculated k value for this model and the response of this model to various enrichment perturbations are compared to the results for the explicit KENO model (for the same geometry zone) in Appendix D. The small standard deviation values shown

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-2 Table D.1 and in Section D.2.3 show that this model produces results that are similar to the explicit KENO model.

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. Figure 6.1 provides an example of the CASMO-4 model and Figure 6.2 provides an illustration of the geometry for the explicit KENO model. 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: An actual Boral plate is modeled in KENO V.a - with the corners and surrounding area represented as water, (the CASMO-4 model smears the Boral over this broad area).

Cases using the explicit model, moderator temperatures between 4 ºC and 100 ºC and the different fuel channel options were created and the results are shown in Table 6.1.

These results indicate that a high water density with a fuel channel produces the most reactive condition.

6.1.3 Storage Rack Sensitivities The high density Boral storage rack modules at Browns Ferry 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 each rack.

The nominal rack to rack water gap is modeled.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-3 6.1.4 Reactivity Comparison of the Boral Rack Models Table 6.3 provides KENO V.a results for the different types of geometry models. The results indicate that the explicit 2x2 model produces the most conservative result. This comparison also indicates that it is conservative to neglect the nominal water gap conditions at the edge of the storage racks.

The explicit 2x2 model with periodic boundary conditions in all directions will be represented with k=0.896 and will be used to represent the reactivity level of the Boral racks. This k value is more reactive than the 13x13 and 13x17 rack models (see Table 6.2 and Table 6.3).

6.2 Boron Content The criticality analysis is performed using the minimum Boron-10 areal density of 0.013 g/cm2. Since using the minimum areal density is conservative no manufacturing uncertainty will be applied to this parameter in Table 7.3.

6.3 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 11 fuel have been performed assuming an advanced fuel channel. Use of the advanced fuel channel is the expected condition.

Zircaloy (or Z4B) has been modeled in KENO as pure zirconium. Neglecting the neutron absorption of the alloying elements (primarily tin, iron and chromium,) is slightly conservative. Similarly, it is conservative to neglect the presence of activated corrosion and wear products because these compounds have higher neutron absorption cross sections than water.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-4

[

]

6.4 Lumped Fission Products 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 12). The adequacy and the treatment of the depletion uncertainty is addressed in Appendix D of this report. [

] It should also be noted that the basis of the k95/95 calculation in this report is a non-depleted assembly. Therefore, the lumped fission products are not translated into the KENO V.a model.

6.5 Rack to Rack Interfaces The rack interface issues are detailed in Section 6.1.3. As stated in Section 6.1.4, the comparison in Table 6.3 demonstrates that the 2x2 model is more reactive than the full storage rack options with water gaps represented. Since the reactivity of the spent fuel pool will be represented by the 2x2 model the rack interface conditions are not significant.

6.6 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 12). Input for the depletion calculation includes

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-5 the fuel assembly material and geometry. The ATRIUM 11 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 11 assembly, which contain a plenum, can be treated as a water hole in the lattice in-core depletion and in the in-rack calculations. The actual content of the plenum consists of a stainless steel spring and fill gas. Neglecting the plenum (material and fill gas) is conservative from a criticality stand point because it results in 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. Consistent fuel temperatures are used for a given geometry. 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 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 show an insignificant increase in k with increased fuel temperature.

For an increase of 100 ºF, the largest increase in the in-rack k was 0.0004 k. The small change in reactivity with respect to the in-core depletion temperature does not impact the ability to select the most reactive lattice for the actual criticality evaluations performed with CASMO-4. These results demonstrate that moderator void is 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 pressure. The more important parameter in a BWR reactor is the actual moderator density/void level. Three explicit

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-6 void conditions are used to perform the in-core depletion calculations, [

]. Figure 6.4 shows the results of a sensitivity evaluation with respect to the in-core depletion void history and its effect on the inrack lattice k. For the reference bounding lattices, the void history conditions used the maximum k condition or the difference between the void history used and the maximum is not significant.

Assumption 4: The power density used for the fuel depletion is based on the core rated power per unit volume. 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.

These results show that the power density assumed during fuel depletion has minimal impact - with only slight increases (< 0.002 k) on the in-rack k-infinity with a 50%

reduction in the assumed power density. Depletion at such a low power density from BOL to the peak reactivity exposure is not realistic. Furthermore, depletion at the core average power density is consistent with Framatomes standard NRC-approved depletion methodology and this procedure is consistent with the uncertainties defined in EMF-2158(P)(A), Reference 12. These results demonstrate that moderator void is more significant than the depletion power density.

Assumption 5: Modeling of 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. The spacer will absorb neutrons and exclude water in an under moderated region.

Grid material absorbs more neutrons than water and has a minimal impact on the neutron spectrum. Ignoring the spacer material is conservative for BWR criticality analyses in that a slightly more reactive configuration is modeled.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-7 Assumption 7: The in-core depletion is based upon uncontrolled statepoint conditions.

This is appropriate since a bundle is in an uncontrolled state (i.e. the adjacent control blade is not inserted) for the majority of its lifetime. This includes the time between BOL and when the bundle 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 the core is operated in that sequence (e.g. one chance in four in 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 rods are used in a sequence, 2) rods are typically not fully inserted (they may be in deep, intermediate or shallow positions which leaves some nodes 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 period which reduces their accumulated burnup. The net effect is that a typical bundle will experience controlled depletion for less than a quarter of its lifetime, including the time from BOL to 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 located around them will have a larger 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.

A sensitivity calculation was performed to determine the impact of in-core controlled depletion on the peak in-rack k-infinity. Table 6.6 shows that for the reference bounding lattices the uncontrolled depletion results in a higher in-rack k-infinity.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-8 Assumption 8: The CASMO-4 model uses a lumped approach for fission products that are not specifically treated. This is discussed in Section 6.4.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-9 Table 6.1 Impact of Channel Thickness on In-Rack Reactivity Fuel Channel Average KENO V.a Name Thickness k Result*

100 mil 0.100 in. 0.8953 Advanced ~0.08 in. 0.8952 No Channel -- 0.8920 Note: (All results are based upon the limiting water ( ~ 0.0002) temperature condition, 20 °C)

Table 6.2 Storage Rack Model Sensitivity KENO V.a Limiting Rack In-Rack k Temperature 13 x 13 Boral Rack k = 0.8853 4 °C 13 x 17 Boral Rack k = 0.8875 4 °C

Based on 20 °C moderator temperature with the 2x2 model (evaluated with temperatures between 4 °C and 100 °C).

Temperatures were evaluated between 4 °C and 20 °C.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-10 Table 6.3 Comparison of Modeling Options for the Boral Rack KENO V.a In-Rack k 2x2 Infinite Array Model k Base with Explicit Boral 0.8953 KENO V.a In-Rack k 13x13 Rack Model k

(closure plates and nominal rack spacing)

Base with Explicit Boral 0.8853 Base with Explicit Boral, 12 top water 0.8833 reflector and a 24 concrete bottom Note: (All results are based upon the limiting water temperature condition, 4 or 20 °C - ( ~ 0.0002) with the advanced fuel channel)

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-11 Table 6.4 In-Rack k Sensitivity to In-core Depletion Fuel Temperature

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-12 Table 6.5 In-Rack k Sensitivity to In-core Depletion Power Density

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-13 Table 6.6 In-Rack k Sensitivity to In-Core Controlled Depletion

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-14 Figure 6.1 Single Cell CASMO-4 Model for the High Density Boral Rack

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-15 Figure 6.2 Explicit 2x2 Geometry Model for High Density Boral Rack

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-16 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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 6-17 Figure 6.4 Impact of Void History Depletion on In-Rack k-infinity

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-1 7.0 CRITICALITY SAFETY ANALYSIS This criticality safety analysis is based upon an ATRIUM 11 reference bounding assembly. This reference assembly is comprised of separate top, middle, 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 11 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 11 assembly is comprised of three axial zones as described in the following table and as shown graphically in Figure 2.2.

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

Geometry Gadolinia Rods wt%

3 A11T [ ]

2 A11M [ ]

1 A11B [ ]

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 [

]

It is demonstrated in Appendix B that the ATRIUM 11 reference design in the spent fuel pool geometry is as reactive as any other fuel type used at Browns Ferry.

The CASMO-4 vs. KENO comparison in Appendix D demonstrates a stable basis for this reactivity equivalence.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-2 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, fluence induced dimension changes, and accident conditions.

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. [

] The calculation results are based upon the nominal fuel design parameters (defined in Table 4.1) and assume an advanced fuel channel. [

] Xenon 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 top, middle, and bottom zone reference bounding lattices have been developed to have a CASMO-4 in-rack k 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 slightly because low enriched rods in the corners of the lattice are replaced with rods at an average enrichment level.

In support of the KENO rack calculations, three REBOL lattices are created corresponding to the top, middle, and bottom geometries for the ATRIUM 11 design.

These lattices are defined using a water temperature of 4 °C in the spent fuel pool rack configuration. [

]

These results are also summarized in Table 7.1.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-3 As discussed in the methodology section, [

]

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 and as illustrated in Figure 6.2. The water gap between rack modules is not included in the model, but the comparison in Table 6.3 shows that the inclusion of water gaps reduces the overall k making this a conservative assumption.

[

] A periodic boundary condition is specified for both the x-y plane and for the axial direction. The KENO model assumes an advanced fuel channel which was shown in Section 6.3 to bound storage without a channel.

KENO V.a calculations were performed at various temperatures from 4 C to 100 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 0.896. 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.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-4 7.3 Arrays of Mixed BWR Fuel Types It is shown in Table B.2 that the ATRIUM 11 reference bounding lattices are equally or more reactive in the in-rack configuration than the limiting lattices of the legacy fuel. It then follows that from a reactivity perspective, the reference bounding ATRIUM 11 lattices used in this evaluation can 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 11 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 11 assemblies.

7.4 Other Conditions The unadjusted reactivity result reported in Table 7.2 is based upon a reference orientation which places the ATRIUM 11, 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. These, and other, 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 ATRIUM 11 assembly is a symmetric design with the internal water channel in the center of the assembly. The reference bounding and REBOL lattices are defined with a uniform enrichment. As such the rotation of the assembly within the storage cell would not impact the system reactivity.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-5 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 potential blistering condition*. [

]

7.4.4 Fuel Rod Creep ATRIUM 11 fuel rod measurements indicate that the fuel rod diameter will creep down (decrease) from [ ] when the reference bounding lattices achieve maximum reactivity. [

]

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.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-6 7.4.5 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. [

]

7.4.6 Fuel Channel Growth The BWR fuel channel will grow longer as the assembly is irradiated. As this occurs the thickness of the channel will decrease. The limiting k is based upon a reasonable representation of the fuel channel. Review of Table 7.1 shows that the reference bounding lattices achieve their lifetime maximum reactivity early in life. Based on this, the level of fuel channel growth will be small and the effect will be neglected.

7.4.7 Spacer Growth In the past, BWR fuel rod spacers were made of Zircaloy and had a tendency to expand with irradiation. This expansion would increase the fuel rod pitch slightly and therefore increase the reactivity of the fuel rod array. The ATRIUM 11 fuel rod spacers are made of Alloy 718 (INCONEL) which does not expand significantly with irradiation. (This is because Inconel has an isotropic crystal structure). In addition, BWR reactors operate at temperatures that tend to reduce the level of Inconel swelling. Finally, Table 7.1 shows that the reference bounding lattices achieve their lifetime maximum reactivity early in life - before much expansion can occur. For these reasons no spacer growth component of ksys will be applied for ATRIUM 11 fuel.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-7 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. 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 fuel pool. For this comparison, three ATRIUM 11 fuel assemblies with REBOL lattices at 4.0 wt% U-235 were placed together in a triangular pattern. A reactivity optimization search was performed using different assembly spacing. 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 0.932 +/- 0.0002 was calculated with unchanneled assemblies.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-8 A k95/95 result for this fuel handling condition would differ from the one in Section 7.8 as follows:

the base keff value would increase from 0.896 to 0.932

the ksys term would decrease to nearly zero because there are no applicable accident conditions

the manufacturing tolerance would be reduced by 30% because there are no applicable storage rack, or gadolinia tolerance conditions

the final k95/95 result would increase about 0.01 but remain less than the 0.95 licensing limit*.

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 considered to identify the most limiting accident condition.

Missing Boral plate in the interior of the rack. (Limiting Condition)

Missing Boral plate near the edge of the rack.

Boral Storage Racks being forced together.

Misloaded Bundle Scenarios.

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 anticipated to be the most limiting case, the moderator temperature and assembly spacing were varied to optimize the worth of the accident.

[

]

There would be substantial margin for the REBOL assembly from Table 7.1 because [

]

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-9

[

]

Investigation of a missing Boral plate near the edge of the rack was not evaluated because it is less limiting than the interior case described above. This condition was specifically evaluated in Section 2.1 of Reference 13 for ATRIUM 10XM reference bounding assemblies at the 0.8825 k limit. (This is the same reactivity level as the ATRIUM 11 reference bounding assemblies used in this evaluation). This event with minimal rack to rack spacing and optimal assembly spacing was shown to be significantly less limiting than the missing boral plate in the center of the rack.

A postulated seismic event that forces racks together with a minimal water gap between them is not meaningful for the storage racks. As shown in Table 6.3, the 13x13 rack produces a kinf of 0.8853 with nominal water gaps. This increases to 0.8906 when all racks are in contact with each other. However, the k95/95 result is based upon a condition with no water gap boundary and a more limiting kinf (0.8953). As long as this conservative 2x2 model is used to represent the spent fuel pool, this potential accident condition is not meaningful.

Review of Figure 4.2 and Figure 4.3 demonstrate that a misplaced assembly can only be placed on the outer perimeter of the storage rack array. Since the evaluation is being performed with infinitely repeating boundary conditions, the base condition is more limiting than the misloaded assembly 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 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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-10 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 Browns Ferry racks, the limiting moderator temperature will be at 4 °C or 20 °C (see Table 6.2 and Table 6.3). 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. [

]

The manufacturing tolerance values and the calculated reactivity uncertainties for the ATRIUM 11 fuel are shown in Table 7.3.* The reactivity effect of the U235 enrichment uncertainty and the gadolinia manufacturing uncertainties (gadolinia concentration and gadolinia pellet density) were evaluated at lifetime maximum reactivity conditions using CASMO-4. All other uncertainties reported in Table 7.3 were evaluated with KENO V.a at beginning of life (BOL) conditions. The ATRIUM 11 rack calculations are conservatively performed for a minimum B10 areal density; therefore, no manufacturing uncertainties are needed for this parameter.

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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-11 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 14:

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 each 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. channel thickness). 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 11 fuel design [

] the base case KENO calculated in-rack reactivity from Table 7.2 is 0.896. This k-eff value is used with the following equation to determine the upper limit 95/95 reactivity (also illustrated in Figure 2.1):

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-12 The following table provides a summary of the ksys and sys parameters applicable to this analysis. (The values are standard deviation results from KENO).

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 rounded-up to 0.919. The above determination of the upper limit 95/95 k-eff is consistent with the method documented in Reference 8 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 11 REBOL lattices meets the NRC criticality safety acceptance criterion that the array k-eff under the worst credible conditions is < 0.95.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-13 Table 7.1 Summary of CASMO-4 Maximum In-Rack Reactivity Results Reference Bounding Lattices ATRIUM 11 top, middle, and bottom lattice geometries explicitly modeled Uniform enrichment distribution as discussed in Section 7.1

[ ] as specified in Table 2.1 and Figure 2.3 Advanced fuel channel Reflective boundary condition for in-core calculations No xenon for in-rack calculations Periodic boundary condition for in-rack calculations Condition Top Lattice Middle Lattice Bottom Lattice Maximum In-Rack k, 4°C (39.2°F) 0.8825 0.8825 0.8825 Exposure, GWd/MTU 12.5 13.0 13.0 Void History [ ] [ ] [ ]

REBOL Lattices

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-14 Table 7.2 Summary of KENO V.a Maximum In-Rack Reactivity Results Fuel Assembly ATRIUM 11 top, middle, and bottom lattice geometries explicitly modeled:

No gadolinia Framatome Advanced Thick / Thin 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 20°C moderator and fuel temperatures Description k-eff In-Rack 20°C (68°F) k-eff 0.896 +/- 0.001

Maximum k95/95 Reactivity (including uncertainties, biases, manufacturing tolerances 0.919 and worst accident or abnormal loading conditions)

The standard deviation has been increased (from 0.0002) to reflect the expected uncertainty of KENO V.a calculations.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 7-15 Table 7.3 Manufacturing Reactivity Uncertainties (Based upon BOL conditions using KENO V.a except the fuel enrichment and Gd2O3 parameters used CASMO-4 to determine lifetime maximum reactivity changes.)

This calculation was performed using the minimum value so no manufacturing uncertainty is required.

Areal density calculations are not sensitive to the material thickness.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 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. 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, December 2015.

3. BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 SAFETY EVALUATION FOR EXTENDED POWER UPRATE, August 2017 (see Enclosure 5 of ADAMS #

ML17032A120).

4. 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.

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

6. 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).

7. 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.

8. 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.

9. 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).

10. ORNL/TM-2005/39, SCALE Version 6.2.2, Code System, Oak Ridge National Laboratory, February 2017.

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

12. 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.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page 8-2

13. ANP-3495P Revision 1, Response to RAI for Browns Ferry Nuclear Plant EPU Submittal

- SFSP Criticality Safety Analysis, Round 2, July 2016.

14. 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.}

15. NEI 12-16 Revision 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants, (ADAMS Accession No. ML19269E069).

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page A-1 APPENDIX A SAMPLE CASMO-4 INPUT Tables A.1, A.2, and A.3 provide the CASMO-4 spent fuel storage rack model for the reference bounding lattices defined in this analysis.

ATRIUM 11 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 Tables A.1 through A.3.

Evaluations should be performed with [ ] depletion for bottom geometry lattices, [ ] depletion for middle 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 12 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 11 fuel assembly can be safely stored in the Browns Ferry spent fuel pool storage racks.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page A-2 Table A.1 CASMO-4 Input for ATRIUM 11 Bottom Reference Bounding Lattice

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page A-3

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page A-4 Table A.2 CASMO-4 Input for ATRIUM 11 Middle Reference Bounding Lattice

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page A-5

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page A-6 Table A.3 CASMO-4 Input for ATRIUM 11 Top Reference Bounding Lattice

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page A-7

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page B-1 APPENDIX B REACTIVITY COMPARISON FOR ASSEMBLIES USED IN THE BROWNS FERRY REACTORS All assemblies used in the Browns Ferry reactors (GE 7x7, GE 8x8, GE 9x9, GE 10x10, Framatome 10x10, and various LUAs) that were manufactured prior to the issuance of the ATRIUM 10XM SFSP criticality safety analysis (Reference 2) were previously evaluated for reactivity compliance in Appendix B of that report. All assemblies manufactured since the NRC review of that report (see approval of the BFN EPU license amendment, Reference 3) have been confirmed to be in compliance with the reactivity limits specified in the Reference 2 report and BFN Technical Specifications.

As shown in Table B.1, the ATRIUM 11 reference bounding lattices have also been developed to meet the Technical Specification k limits. Since the limiting manufacturing uncertainties of the high reactivity fuel types are similar, the ATRIUM 11 reference bounding lattices can be used to represent the most reactive lattices that have been or are allowed to be loaded in the Browns Ferry spent fuel pools.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page B-2 Table B.1 Lattice Reactivity Comparisons (REBOL, Bounding, and Limiting)

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page B-3 Evaluation of Modified, Abnormal, and Damaged Assemblies The preceding discussion is based upon nominal assembly designs. For failed assemblies that have not lost fuel pellets the calculations with the nominal geometry will be representative. However, the potential exists that assemblies that have been 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 BWR lattices are under-moderated in the region around the fuel rods, 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.

Reconstitution of assemblies involves the replacement of one or more fuel rods.

Replacement rods can be from a donor assembly, inert rods (such as stainless steel pins), or newly manufactured rods (containing either natural or enriched uranium).

Browns Ferry reconstituted assemblies have used either donor rods or inert rods.

Assemblies in which a fuel rod has been replaced with either an inert rod represents a reduction in reactivity from the nominal design since the fissile material content is reduced and no change in the amount of moderator has occurred. An assembly with a donor rod represents an enrichment perturbation where the resulting assembly does not significantly deviate from the nominal design.

It will be the plant owners responsibility to ensure that any modified or damaged assemblies do not create a condition that is more reactive than the nominal assemblies examined during this evaluation. Future fuel rod failures that do not change the geometry of the fuel assembly, (e.g., no additional water in the fuel region, no change in the rod spacing, and no addition of fissile material) can continue to be represented by the applicable criticality analysis.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page B-4 It is noted that Appendix B of Reference 2 contains a thorough description of the modified, abnormal and damaged assemblies that were present in the Browns Ferry spent fuel pools at that time.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 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 6.2 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 CSAS5 driver in the SCALE 6.2 code package for use with the continuous energy cross-section library ce_v7.1 ( ENDF/B-VII.1 library) for spent fuel criticality analyses.

This calculation is performed according to the general methodology described in Reference C.1 (NUREG/CR-6698) that is also briefly described in Section C.1. The critical experiments selected to benchmark the computer code system are described in References C.2 and C.3 as 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.1 (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.1, 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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-2 factor 1/i2 (see Eq. 4, 5, and 6 in Reference C.1). The final objective is to determine the square root of the pooled variance, defined as (Eq. 7 from Reference C.1):

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 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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-3 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.1) must be included in the KL, as shown below.

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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 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 2.4 to 4.0 wt% U-235 Moderation/Moderator Heterogeneous/Water Anticipated Absorber/Materials Boron Carbide, Stainless Steel, Water Boron 10 Areal Density (g/cm2) 0.008 - 0.015 Moderating Ratio (H/X) 70 to >200 Reflection Water, Stainless Steel Neutron Energy Spectrum (Energy of the 0.1 to 0.7 Average Lethargy Causing Fission) (eV)

Lattice Geometry Square Lattice Pitch 1.1 to 1.3 cm Fuel Rod Clad and Fuel Channel Material zirconium alloys

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 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.

Seventy six (76) 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.

Detailed descriptions of the experiments are presented in Reference C.2.

In addition to the benchmark experiments evaluated an additional set of twenty three (23) experiments including actinides (i.e. HTC criticals as described in Reference C.3) was considered. These have not been included since they were designed to simulate a depleted assembly and therefore would not be representative of a REBOL based analysis. However, while not included it was verified that the statistical results would be the same if this additional data were to be included.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-6 Table C.2 Descriptions of the Critical Benchmark Experiments Neutron Experiment Case Name Brief Description Reflector Absorber CEA Critical Experiments Square pitch fuel rod arrays with varying rod LEU-COMP-THERM-007 pitch configurations. Each fuel rod is aluminum clad with UO2 fuel at 4.738 wt% None Water U235. Performed at CEA Valduc Critical Mass Laboratory.

Square pitch (pitch = 1.6 cm) fuel rod arrays adjacent to concrete slab in area filled with Water/

LEU-COMP-THERM-037 water. Each fuel rod is aluminum clad with None Concrete UO2 fuel at 4.738 wt% U-235. Performed at CEA Valduc Critical Mass Laboratory.

Square pitch (pitch = 1.26 cm) fuel rod arrays LEU-COMP-THERM-039 without fuel rods in all positions. Each fuel rod is aluminum clad with UO2 fuel at 4.738 None Water wt% U-235. Performed at CEA Valduc Critical Mass Laboratory.

Square pitch (pitch = 1.05 and 1.075 cm) fuel rod arrays separated by a cruciform-shaped area filled with water. Each fuel rod is LEU-COMP-THERM-073 None Water zircaloy clad with UO2 fuel at 4.738 wt% U-235. Performed at CEA Valduc Critical Mass Laboratory.

PNL Critical Experiments Stainless UO2 pellets enriched at 4.31 wt% U-235 clad steel, in aluminum. Three rectangular clusters of borated Stainless 1.892 cm pitch rods separated by absorber LEU-COMP-THERM-013 stainless Steel and plates with stainless steel walls on either side steel, Boral Water of the line of fuel. Performed at the Pacific and Northwest Laboratories (PNL).

Boroflex Stainless UO2 pellets enriched at 2.35 wt% U-235 clad steel, in aluminum. Three variable sized borated rectangular clusters of fuel rods on a 2.032 LEU-COMP-THERM-016 stainless Water cm pitch, separated by absorber plates with steel, Boral varying separation distances. Performed at and the Pacific Northwest Laboratories (PNL).

Zicaloy-4

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-7 Table C.2 Descriptions of the Critical Benchmark Experiments (Continued)

Neutron Experiment Case Name Brief Description Reflector Absorber Stainless UO2 pellets enriched at 2.35 wt% U-235 clad steel, in aluminum. Three rectangular clusters of borated Stainless fuel rods on a 1.684 cm pitch, separated by LEU-COMP-THERM-042 stainless Steel and absorber plates with stainless steel walls on steel, Boral Water either side of the line of fuel. Performed at and the Pacific Northwest Laboratories (PNL).

Boroflex DIMPLE Critical Experiments Large circular arrays of square pitch (pitch =

LEU-COMP-THERM-048 1.32 cm), stainless steel clad, uranium dioxide fuel rods containing 3.005 wt% U- None Water 235. Performed in the DIMPLE low-power reactor.

Large cruciform shaped arrays of square LEU-COMP-THERM-055 pitch (pitch = 1.2507 cm), stainless steel Stainless clad, uranium dioxide fuel rods containing None Steel and 3.005 wt% U-235. Performed in the DIMPLE Water low-power reactor.

Large cruciform shaped arrays of square pitch (pitch = 1.2507 cm), stainless steel Stainless Borosilicate LEU-COMP-THERM-063 clad, uranium dioxide fuel rods containing Steel and glass 3.005 wt% U-235. Performed in the DIMPLE Water low-power reactor.

Large cruciform shaped arrays of square pitch (pitch = 1.2507 cm), stainless steel Stainless Borosilicate LEU-COMP-THERM-076 clad, uranium dioxide fuel rods containing Steel and glass 3.005 wt% U-235. Performed in the DIMPLE Water low-power reactor.

IPEN/MB-01 Critical Experiments Square pitch (1.5 cm) fuel rod arrays with varying thickness of stainless steel reflector.

Each fuel rod is SS-304 clad with UO2 fuel at LEU-COMP-THERM-043 Stainless 4.3486 wt% U235. Fuel rod configuration Stainless Steel and and absorber rod location are varied. Steel, Gd Water Performed at IPEN/MB-01 research facility Sao Paulo, Brazil. (The large arrays of fuel rods only contains 4 Gd2O3 rods).

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-8 Table C.2 Descriptions of the Critical Benchmark Experiments (Continued)

Neutron Experiment Case Name Brief Description Reflector Absorber Square pitch (1.5 cm) fuel rod arrays.

Temperature was varied from 14 to 80°C.

LEU-COMP-THERM-046 Each fuel rod is SS-304 clad with UO2 fuel at None, 4.3486 wt% U235. Fuel rod configuration UO2- Water and absorber rod location are varied. GD2O3 Performed at IPEN/MB-01 research facility Sao Paulo, Brazil.

Square pitch (1.5 cm) fuel rod arrays. Each fuel rod is SS-304 clad with UO2 fuel at 4.3486 wt% U235. Fuel rod configuration UO2-LEU-COMP-THERM-054 Water and absorber rod location are varied. GD2O3 Performed at IPEN/MB-01 research facility Sao Paulo, Brazil.

Square pitch (1.5 cm) fuel rod arrays with varying void in the reflector tank. Each fuel rod is SS-304 clad with UO2 fuel at 4.3486 Water/Void UO2-LEU-COMP-THERM-058 wt% U235. Fuel rod configuration and Tank and GD2O3 absorber rod location are varied. Performed Water at IPEN/MB-01 research facility Sao Paulo, Brazil.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-9 C.4 Results of Calculations with SCALE 6.2 The critical experiments described in Section C.3 were modeled with the SCALE 6.2 computer code. 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 experimental kexp, using the following formula (Eq.9 from Reference C.1):

k calc 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.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-10 Table C.3 SCALE 6.2 Results for the Selected Benchmark Experiments Benchmark Calculated Values Benchmark Case Values kexp exp kcalc calc knorm tot 1 LEU-COMP-THERM-007-L01_CE 1.0000 0.0014 0.9976 0.0005 0.9976 0.00149 2 LEU-COMP-THERM-007-L02_CE 1.0000 0.0008 0.9986 0.0005 0.9986 0.00094 3 LEU-COMP-THERM-007-L03_CE 1.0000 0.0007 0.9982 0.0005 0.9982 0.00086 4 LEU-COMP-THERM-039-L01_CE 1.0000 0.0014 0.9969 0.0005 0.9969 0.00149 5 LEU-COMP-THERM-039-L02_CE 1.0000 0.0014 0.9983 0.0005 0.9983 0.00149 6 LEU-COMP-THERM-039-L03_CE 1.0000 0.0014 0.9980 0.0005 0.9980 0.00149 7 LEU-COMP-THERM-039-L04_CE 1.0000 0.0014 0.9959 0.0005 0.9959 0.00149 8 LEU-COMP-THERM-039-L05_CE 1.0000 0.0009 0.9968 0.0005 0.9968 0.00103 9 LEU-COMP-THERM-039-L06_CE 1.0000 0.0009 0.9977 0.0005 0.9977 0.00103 10 LEU-COMP-THERM-039-L07_CE 1.0000 0.0012 0.9967 0.0005 0.9967 0.00130 11 LEU-COMP-THERM-039-L08_CE 1.0000 0.0012 0.9966 0.0005 0.9966 0.00130 12 LEU-COMP-THERM-039-L09_CE 1.0000 0.0012 0.9961 0.0005 0.9961 0.00130 13 LEU-COMP-THERM-039-L10_CE 1.0000 0.0012 0.9980 0.0005 0.9980 0.00130 14 LEU-COMP-THERM-073-L01_CE 1.0000 0.0010 0.9966 0.0005 0.9966 0.00112 15 LEU-COMP-THERM-073-L02_CE 1.0000 0.0010 0.9969 0.0005 0.9969 0.00112 16 LEU-COMP-THERM-073-L03_CE 1.0000 0.0010 0.9968 0.0005 0.9968 0.00112 17 LEU-COMP-THERM-073-L04_CE 1.0000 0.0016 0.9993 0.0005 0.9993 0.00167 18 LEU-COMP-THERM-073-L05_CE 1.0000 0.0016 0.9969 0.0005 0.9969 0.00168 19 LEU-COMP-THERM-073-L06_CE 1.0000 0.0011 0.9990 0.0005 0.9990 0.00121 20 LEU-COMP-THERM-073-L07_CE 1.0000 0.0011 0.9971 0.0005 0.9971 0.00121 21 LEU-COMP-THERM-073-L08_CE 1.0000 0.0011 0.9977 0.0005 0.9977 0.00121 22 LEU-COMP-THERM-073-L09_CE 1.0000 0.0010 0.9990 0.0005 0.9990 0.00112 23 LEU-COMP-THERM-073-L10_CE 1.0000 0.0010 0.9981 0.0005 0.9981 0.00112 24 LEU-COMP-THERM-073-L11_CE 1.0000 0.0010 0.9972 0.0005 0.9972 0.00112 25 LEU-COMP-THERM-073-L12_CE 1.0000 0.0011 0.9988 0.0005 0.9988 0.00121 26 LEU-COMP-THERM-073-L13_CE 1.0000 0.0011 0.9982 0.0005 0.9982 0.00121 27 LEU-COMP-THERM-073-L14_CE 1.0000 0.0011 0.9978 0.0005 0.9978 0.00120 28 LEU-COMP-THERM-013-L01_CE 1.0000 0.0018 1.0011 0.0005 1.0011 0.00187 29 LEU-COMP-THERM-013-L02_CE 1.0000 0.0018 1.0014 0.0005 1.0014 0.00187 30 LEU-COMP-THERM-013-L03_CE 1.0000 0.0018 1.0002 0.0005 1.0002 0.00187 31 LEU-COMP-THERM-013-L04_CE 1.0000 0.0018 1.0018 0.0005 1.0018 0.00187 32 LEU-COMP-THERM-016-L01_CE 1.0000 0.0031 0.9989 0.0005 0.9989 0.00314

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-11 Table C.3 SCALE 6.2 Results for the Selected Benchmark Experiments (Continued)

Benchmark Calculated Values Benchmark Case Values kexp exp kcalc calc knorm tot 33 LEU-COMP-THERM-016-L02_CE 1.0000 0.0031 0.9977 0.0005 0.9977 0.00314 34 LEU-COMP-THERM-016-L03_CE 1.0000 0.0031 0.9987 0.0005 0.9987 0.00314 35 LEU-COMP-THERM-016-L04_CE 1.0000 0.0031 0.9987 0.0005 0.9987 0.00314 36 LEU-COMP-THERM-016-L05_CE 1.0000 0.0031 0.9986 0.0005 0.9986 0.00314 37 LEU-COMP-THERM-016-L06_CE 1.0000 0.0031 0.9983 0.0005 0.9983 0.00314 38 LEU-COMP-THERM-016-L07_CE 1.0000 0.0031 0.9986 0.0005 0.9986 0.00314 39 LEU-COMP-THERM-016-L08_CE 1.0000 0.0031 0.9989 0.0005 0.9989 0.00314 40 LEU-COMP-THERM-016-L09_CE 1.0000 0.0031 0.9995 0.0005 0.9995 0.00314 41 LEU-COMP-THERM-016-L10_CE 1.0000 0.0031 0.9975 0.0005 0.9975 0.00314 42 LEU-COMP-THERM-016-L11_CE 1.0000 0.0031 0.9990 0.0005 0.9990 0.00314 43 LEU-COMP-THERM-016-L12_CE 1.0000 0.0031 0.9981 0.0005 0.9981 0.00314 44 LEU-COMP-THERM-016-L13_CE 1.0000 0.0031 0.9986 0.0005 0.9986 0.00314 45 LEU-COMP-THERM-016-L14_CE 1.0000 0.0031 0.9995 0.0005 0.9995 0.00314 46 LEU-COMP-THERM-016-L31_CE 1.0000 0.0031 0.9987 0.0005 0.9987 0.00314 47 LEU-COMP-THERM-016-L32_CE 1.0000 0.0031 0.9990 0.0005 0.9990 0.00314 48 LEU-COMP-THERM-042-L01_CE 1.0000 0.0016 0.9985 0.0005 0.9985 0.00168 49 LEU-COMP-THERM-042-L02_CE 1.0000 0.0016 0.9973 0.0005 0.9973 0.00168 50 LEU-COMP-THERM-042-L03_CE 1.0000 0.0016 0.9990 0.0005 0.9990 0.00168 51 LEU-COMP-THERM-042-L04_CE 1.0000 0.0017 0.9994 0.0005 0.9994 0.00177 52 LEU-COMP-THERM-048-L01_CE 1.0000 0.0025 1.0014 0.0005 1.0014 0.00255 53 LEU-COMP-THERM-048-L02_CE 1.0000 0.0025 1.0020 0.0005 1.0020 0.00255 54 LEU-COMP-THERM-048-L03_CE 1.0000 0.0025 1.0019 0.0005 1.0019 0.00255 55 LEU-COMP-THERM-048-L04_CE 1.0000 0.0025 1.0023 0.0005 1.0023 0.00255 56 LEU-COMP-THERM-048-L05_CE 1.0000 0.0025 1.0023 0.0005 1.0023 0.00255 57 LEU-COMP-THERM-043-C01 1.0004 0.0010 1.0015 0.0005 1.0011 0.00111 58 LEU-COMP-THERM-043-C02 1.0004 0.0010 1.0008 0.0005 1.0004 0.00111 59 LEU-COMP-THERM-043-C03 1.0004 0.0010 1.0004 0.0005 1.0000 0.00111 60 LEU-COMP-THERM-043-C04 1.0006 0.0010 1.0007 0.0005 1.0001 0.00111 61 LEU-COMP-THERM-043-C05 1.0004 0.0010 1.0017 0.0005 1.0013 0.00111 62 LEU-COMP-THERM-043-C06 1.0005 0.0010 1.0013 0.0005 1.0008 0.00111 63 LEU-COMP-THERM-043-C07 1.0006 0.0010 1.0008 0.0004 1.0002 0.00109 64 LEU-COMP-THERM-043-C08 1.0004 0.0010 1.0007 0.0005 1.0003 0.00111 65 LEU-COMP-THERM-043-C09 1.0006 0.0010 1.0013 0.0005 1.0006 0.00110

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-12 Table C.3 SCALE 6.2 Results for the Selected Benchmark Experiments (Continued)

Benchmark Calculated Values Benchmark Case Values kexp exp kcalc calc knorm tot 66 LEU-COMP-THERM-046-C01 1.0004 0.0004 1.0011 0.0006 1.0007 0.00078 67 LEU-COMP-THERM-046-C02 1.0004 0.0004 1.0009 0.0006 1.0005 0.00071 68 LEU-COMP-THERM-046-C03 1.0004 0.0004 1.0011 0.0006 1.0007 0.00075 69 LEU-COMP-THERM-046-C04 1.0004 0.0004 1.0021 0.0006 1.0017 0.00071 70 LEU-COMP-THERM-046-C05 1.0004 0.0004 1.0014 0.0006 1.0010 0.00070 71 LEU-COMP-THERM-046-C06 1.0004 0.0004 1.0022 0.0005 1.0018 0.00066 72 LEU-COMP-THERM-046-C07 1.0004 0.0004 1.0034 0.0005 1.0030 0.00067 73 LEU-COMP-THERM-046-C08 1.0004 0.0004 1.0003 0.0006 0.9999 0.00078 74 LEU-COMP-THERM-046-C09 1.0004 0.0004 1.0020 0.0007 1.0016 0.00082 75 LEU-COMP-THERM-046-C10 1.0004 0.0004 1.0018 0.0006 1.0014 0.00071 76 LEU-COMP-THERM-046-C11 1.0004 0.0004 1.0012 0.0007 1.0008 0.00082

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-13 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:

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

Fuel enrichment (wt% U-235)

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

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 6.2 (kcalc) was done as suggested in Reference C.1. 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.1 suggests the use of weighting factors to reduce the importance of data with higher uncertainty. For this application, the weighted trends were evaluated.

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.1, the weighting factors were evaluated as shown below with all significant digits included in later calculations.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-14 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.1, 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 = k eff i 1

i 2 i

For the residuals, there are n - 2 degrees of freedom, where n is the number of data points in the benchmark set (76 for this evaluation). 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:

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-15 (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.4, calculated TValues are compared with the critical value of the Students T-distribution with a significance level of = 0.05 and n-2 degrees of freedom. The null hypothesis for this test (H0), is that the slope is not statistically significant; thus, a statistically significant trend may exist if: l TValue l > tcrit .

In addition, the probability of obtaining a TValue of larger magnitude from a two-tailed T-distribution with the same n - 2 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.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-16 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.4). The Anderson-Darling test described in Reference C.6 is employed for this 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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-17 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 there is no statistically valid trend. Although there is no valid trend, lower tolerance bands were calculated for all parameters. The intermediate results are listed in Table C.5.

If a statistically significant trend is identified, then a single sided lower tolerance band may be calculated. Calculational details of the single-sided lower tolerance band can be found in Reference C.1 (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).

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-18 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).

xi i

2 x= i 1

i 2

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

2 i 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.3 show the normalized keff datasets plotted as a function of the weighted EALF, U-235 enrichment and H/X, 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%. As shown in Table C.4, no statistically valid trend exists for these parameters.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-19 Table C.4 Results Summary for Weighted Trending Analysis Moderating EALF Enrichment Parameter Ratio (eV) (wt% U-235)

(H/X)

Slope -4.80E-03 -6.61E-04 5.65E-06 Intercept 1.0008 1.0026 0.9988 2

r 0.1531 0.0405 0.0655 Tcrit 1.9925 1.9925 1.9925 T-value 3.6576 1.7677 2.2767 P(T>T-value) 0.0005 0.0812 0.0257 Valid Trend? Yes No Yes A* 1.0329 --- 1.8000 Statistical Test No --- No Valid?

Table C.5 Intermediate Results for Lower Tolerance Band Evaluation Parameter x (x i

i x) 2 s 2fit (S P )fit EALF (eV) 0.2341 1.7190 2.961E-06 0.00211 Enrichment (wt% U-235) 4.352 24.0 3.355E-06 0.00220 Moderating Ratio (H/X) 158.00 5.315E+05 3.268E-06 0.00218

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-20 1.010 y=-0.0048x + 1.0008 R2 = 0.1531 1.005 1.000 Normalized k(eff) 0.995 0.990 0.985 0.980 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 EALF (eV)

Benchmark Data Lower Tolerance Band Figure C.1 Weighted EALF Trend Evaluation 1.010 y=-0.0007x + 1.0026 R2 = 0.0405 1.005 1.000 Normalized k(eff) 0.995 0.990 0.985 0.980 1 2 3 4 5 6 7 Enrichment %

Benchmark Data Lower Tolerance Band Figure C.2 Weighted U-235 Enrichment Trend Evaluation

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-21 1.010 1.005 y=5.6E-6x + 0.9988 R2 = 0.0655 1.000 Normalized k(eff) 0.995 0.990 0.985 0.980 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 450.00 500.00 Moderator to Fuel Ratio, H/X Benchmark Data Lower Tolerance Band Figure C.3 Weighted H/X Trend Evaluation

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-22 C.6 Bias and Bias Uncertainty For situations in which no significant trend in bias is identified the statistical methodology, presented in Reference C.1 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 76 case benchmark set. The resulting Anderson-Darling test statistic modified for the number of data points A* was determined to be 0.77516. A plot of the data relative to a normal distribution is provided in Figure C.4. Based on the test statistic and plot, the benchmark data cannot be considered normally distributed.

Thus, the assumption of normality was not met for the benchmark set such that the use of a single-sided lower tolerance limit may not be applicable for determining the bias and uncertainty. However, a single sided tolerance was computed for information. For n = 76, the tolerance limit multiplier (C95/95 ) is 1.974, from Reference C.5. 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:

yi 2

k eff = y =

i i

= 0.99969 1

i 2 i

Bias = keff - 1 = -0.00031

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-23 1 (y i y )2 n 1 i i2 2

s = =3.408E-06 1 1 n i i2 n

2 = = 1.478E-06 1

i 2 i

SP = s 2 + 2 = 0.00222 Therefore if the data set were normal, the bias and uncertainty based on a single sided lower tolerance would be:

Bias = -0.001 Uncertainty = (C95/95)(Sp) = (1.974)(0.00222) = 0.0044 The corresponding lower tolerance limit is:

KL = k eff - (C95/95)(Sp) = 0.999 - 0.0044 = 0.9946 When the benchmark data set cannot be determined to be a normal distribution, then a non-parametric approach can be used to determine the combination of bias and uncertainty as KL.as a lower tolerance limit.

Equations 32 and 33 of Reference C.1 is used for the determination of the following:

q 0.95 n 76 beta 0.979723 NPM 0 NUREG/CR-6698 Table 2.2 Smallest k 0.99593 experiment 0.0014 calc 0.0005 total 0.00149 KL 0.9944 Therefore, with the use of a non-parametric statistical approach, the lower tolerance limit is 0.9944.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-24 Cumulative Distribution 0.992 0.994 0.996 0.998 1 1.002 1.004 1.006 k-eff Normal Distribution k-eff Data Figure C-4 Normal Probability Plot for the keff Dataset

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-25 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 H/X 48 to 455 EALF 0.073 to 0.69 ev Absorbers Boron, Stainless Steel, Water Water Reflectors Stainless Steel

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-26 C.8 Bias Summary and Conclusions The mixed dataset of 76 criticality safety benchmarks experiments was tested against the null hypothesis of normality and was found not to be normally distributed.

Consequently, a non-parametric statistical analysis was used which resulted in a lower tolerance limit of KL = 0.9944.*

A standard trending analysis was also performed using linear regression analysis, including significance testing and goodness-of-fit evaluation. Three independent variables were examined: enrichment (wt% U-235), moderating ratio (H/X), and EALF (eV). The results of the trending analysis showed that none of these parameters exhibited a statistically valid weighted trend.

To support the lower tolerance KL value as determined with the non-parametric statistics the uncertainty of the SFSP1 dataset will be increased to 0.0023 which will result in the same calculated KL value. This results in the following bias and uncertainty:

Bias = k eff - 1 = 0.999 - 1 = -0.001 bias uncertainty = 0.0023 Uncertainty = (C95/95)(Sp) = (1.974)(0.0023) = 0.00454 KL = k eff - (C95/95)(Sp) = 0.999 - 0.0046 = 0.9944

The KL value using non-parametric statistics is 0.0002 lower that the value of 0.9946 obtained assuming a normal distribution. The more conservative value of 0.9944 will be used.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page C-27 C.9 References C.1 Nuclear Regulatory Commission, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001.

C.2 NEA No. 7328, International Handbook of Evaluated Criticality Safety Benchmark Experiments, September 2016 (www.oecd-nea.org).

C.3 NUREG/CR-6979, Evaluation of the French Haut Taux de Combustion (HTC)

Critical Experiment Data, September 2008. (Adams Accession No. ML082880452)

C.4 Rosenkrantz W.A., Introduction to Probability and Statistics for Scientists and Engineers, The McGraw-Hill Companies Inc. 1997.

C.5 Owen, D.B., Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, Sandia Corporation Monograph SRC-607, 1963.

C.6 DAgostino, R.B. and Stephans, M.A. Goodness of Fit Techniques, Statistics, Textbooks and Monographs, Volume 68, New York, NY, 1986.

C.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).

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 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 Framatomes 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 Framatome 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 of Reference 2 of the main report)

define reference lattices that have equivalent or greater reactivity 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

REBOL lattices.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-2

demonstrate that the CASMO-4 depletion uncertainty combined with a CASMO-4 calculational uncertainty [

]

D.2 Reactivity 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) and to demonstrate that a k predicted by CASMO4 is nearly identical to a k predicted by KENO V.a (Section D.2.3).

D.2.1 Comparison Methodology 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. [

]

D.2.2 Absolute k-effective Comparison The calculated multiplication factors from KENO and CASMO were tabulated. The keno term was from each individual KENO calculation and the term casmo was set to the CASMO-4 convergence criteria for the individual case. (Use of the CASMO

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-3 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.

Substituting casmo for calc and keno for exp the resulting equation for tot is:

+ keno 2 2 tot

= camo The CASMO and KENO results are provided in Table D.2. Table D.1 below provides a summary of the differences by axial geometry. The absolute difference between CASMO and KENO is not used for determining any code bias or uncertainty for the SFSP analysis.

Table D.1 k_CASMO k_KENO Comparison Summary by Geometry (Values at 4 C)

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-4 Table D.2 CASMO-4 and KENO V.a Absolute k Comparison

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-5 D.2.3 Relative (k) k-effective Comparison An additional evaluation of the capability of the CASMO-4 code to predict the change in reactivity associated with a perturbation of fuel parameters is documented in this Section based on the calculations performed in Section D.2.1. In this evaluation, the relative reactivity change is evaluated by taking the delta with respect to the initial reference reactivity. A difference is then determined between the k values obtained with KENO and the k values obtained with CASMO-4 for the same perturbation.

These results are provided in Table D.3.

[

]

A code bias is not used in the evaluation of incremental reactivity. Therefore, a trend evaluation is not performed.

Area of Applicability The fuel and rack geometry as well as fuel enrichment representing peak reactivity assemblies were evaluated consistent with the high density storage tubes.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-6 Table D.3 CASMO-4 and KENO V.a Relative k Comparison

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-7 Figure D.1 Normality Plot for kCASMO - kKENO k-infinity Comparison

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-8 D.3 CASMO-4 Depletion Uncertainty The CASMO depletion uncertainty is addressed by two different approaches in the following sections. Specifically, a value is derived based upon EMF-2158(P)(A)

(Reference D.1) and the calculation uncertainty discussed previously. A value is also provided based upon the guidance of DSS-ISG-2010-01 (Reference D.4). The results of these two approaches are compared in order to select a conservative value for this analysis.

D.3.1 EMF-2158(P)(A) Based Depletion Uncertainty This approach is based on the extensive benchmarking that is documented within the Framatome CASMO-4/MICROBURN-B2 licensing topical report (Reference D.1).

Comparisons against critical experiments were performed by Studsvik with results reported in Table 2.1 of Framatomes CASMO4/MICROBURNB2 licensing topical report {EMF-2158(P)(A) Revision 0}. In addition, [

]

Two standard deviations are used for the 95/95 confidence interval. Therefore the multiplier is 2.0.

[

]

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-9

[

]

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.3) and currently addressed in DSS-ISG-2010-01 (Reference D.4).

The reference bounding lattices used in this comparison are identified in Table B.1. All lattices are depleted in-core and then evaluated at the limiting moderator temperature (4 ºC) in the fuel storage rack configuration. Figure D.3 illustrates the two reactivity decrement values used.

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 reference bounding lattices.

This assessment produces a maximum depletion uncertainty of [ ]

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.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-10

[

]

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.

Table D.4 Depletion Uncertainty Values for Limiting Lattices

[

]

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-11 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 decrement (k) 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

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-12 Figure D.4 Gadolinia Concentration Sensitivity

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-13 D.4 Conclusions and Overall Uncertainty 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, (Section D.2.3). This was combined with uncertainty estimates from EMF-2158(P) in Section D.3.1. The estimated depletion uncertainty determined with the method from the interim staff guidance document (Reference D.4) can also be used to produce a maximum combined uncertainty 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 also results in a maximum combined uncertainty of [

]

Table D.5 CASMO-4 Combined Depletion Uncertainty

[ ]

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page D-14 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 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.4 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)

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page E-1 APPENDIX E CRITICALITY ANALYSIS CHECKLIST The following checklist is provided to help identify areas that conform or do not conform to the guidance in NEI 12-16 Revision 4. Note that items which are specific to a submittal to the NRC have TVA listed in the Included column.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page E-2 Table E.1 NEI 12-16 Checklist Subject Included Notes / Explanation 1.0 Introduction and Overview Purpose of submittal TVA Changes requested TVA Summary of physical changes TVA Summary of Tech Spec changes TVA Summary of analytical scope yes See Sections 1.0 & 2.0 2.0 Acceptance Criteria and Regulatory Guidance Summary of requirements and guidance yes See Section 3.0 Requirements documents yes referenced Guidance documents referenced yes Acceptance criteria described yes 3.0 Reactor and Fuel Design Description Describe reactor operating parameters Parameters that can affect this yes evaluation are described in Section 6.6.

Describe all fuel in pool yes Geometric dimensions (Nominal and See Tables 4.1 and 7.3 as well as Tolerances) yes Appendix B of ANP-3910P and ANP 3160P Schematic of guide tube patterns NA Material compositions See Tables 4.1 of ANP-3910P and ANP yes 3160P Describe future fuel to be covered yes Geometric dimensions (Nominal and yes See Tables 4.1 and 7.3 Tolerances)

Schematic of guide tube patterns NA Material compositions yes See Section 2.0 and Table 4.1 Describe all fuel inserts NA Geometric Dimensions (Nominal and NA Tolerances)

Schematic (axial/cross-section) NA Material compositions NA Describe non-standard fuel NA Geometric dimensions NA Describe non-fuel items in fuel cells TVA

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page E-3 Subject Included Notes / Explanation Nominal and tolerance dimensions TVA 4.0 Spent Fuel Pool/Storage Rack Description New fuel vault & Storage rack NA description Nominal and tolerance dimensions NA Schematic (axial/cross-section) NA Material compositions NA Spent fuel pool, Storage rack description yes Nominal and tolerance dimensions yes See Table 4.2 Schematic (axial/cross-section) yes See Figures 4.2, 4.3 & 4.4 Material compositions yes See Table 4.2 Other Reactivity Control Devices NA (Inserts)

Nominal and tolerance dimensions NA Schematic (axial/cross-section) NA Material compositions NA 5.0 Overview of the Method of Analysis New fuel rack analysis description NA Storage geometries NA Bounding assembly design(s) NA Integral absorber credit NA Accident analysis NA Spent fuel storage rack analysis yes See Sections 2.0 & 5.0 description Storage geometries yes See Figures 4.2, 4.3 & 4.4 and Table 6.3 Bounding assembly design(s) See Table 2.1, Figure 2.2 and Appendix yes B

Soluble boron credit NA Boron dilution analysis NA Burnup credit NA Decay/Cooling time credit NA Integral absorber credit For assembly absorber sensitivity, see yes Figure D.4. The rack absorber is described in Table 4.2.

Other credit NA Fixed neutron absorbers yes

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page E-4 Subject Included Notes / Explanation Aging management program TVA Accident analysis yes See Section 7.6 Temperature increase yes Assembly drop yes Single assembly misload yes Multiple misload NA Boron dilution NA Other yes Rack to rack spacing (See Section 7.6)

Fuel out of rack analysis yes See Section 7.5 Handling yes Movement yes Inspection yes 6.0 Computer Codes, Cross Sections and Validation Overview Code/Modules Used for Calculation of yes See Section 5.0 and Appendix D keff Cross section library yes Description of nuclides used NO This is not a depletion credit evaluation Convergence checks yes Code/Module Used for Depletion yes See Section 5.0 and Appendix D Calculation Cross section library yes Description of nuclides used yes Convergence checks yes Validation of Code and Library yes See Appendix D Major Actinides and Structural NA Materials Minor Actinides and Fission Products NA Absorbers Credited yes 7.0 Criticality Safety Analysis of the New Fuel Rack Rack model NA Boundary conditions NA Source distribution NA Geometry restrictions NA Limiting fuel design NA

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page E-5 Subject Included Notes / Explanation Fuel density NA Burnable Poisons NA Fuel dimensions NA Axial blankets NA Limiting rack model NA Storage vault dimensions and NA materials Temperature NA Multiple regions/configurations NA Flooded NA Low density moderator NA Eccentric fuel placement NA Tolerances NA Fuel geometry NA Fuel pin pitch NA Fuel pellet OD NA Fuel clad OD NA Fuel content NA Enrichment NA Density NA Integral absorber NA Rack geometry NA Rack pitch NA Cell wall thickness NA Storage vault dimensions/materials NA Code uncertainty NA Biases NA Temperature NA Code bias NA Moderator Conditions NA Fully flooded and optimum density NA moderator 8.0 Depletion Analysis for Spent Fuel Depletion Model Considerations As described in Sections 2.0, 5.0 and Appendix D, CASMO-4 is used to model yes the assembly depletion to determine the lifetime maximum kinf and a reactivity equivalent assembly is used in

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page E-6 Subject Included Notes / Explanation the KENO calculations.

Time step verification The depletion time steps for the yes reference bounding lattices are shown in Tables A.1, A.2 and A.3.

Convergence verification yes Simplifications A simplified storage rack model is used yes (see Tables A.1, A.2 and A.3).

Non-uniform enrichments yes See the second paragraph in Section 7.1 Post Depletion Nuclide Adjustment NA Cooling Time NA Depletion Parameters Burnable Absorbers yes See Table 2.1 and Figure D.4.

Integral Absorbers NA Soluble Boron NA Fuel and Moderator Temperature yes See Section 6.6 Power yes See Section 6.6 Control rod insertion yes See Section 6.6 Atypical Cycle Operating History yes See Section 6.6 9.0 Criticality Safety Analysis of Spent Fuel Pool Storage Racks Rack model yes See Tables 6.2 & 6.3 Boundary conditions yes See Table 6.3 Source distribution The final k95/95 result is based upon an infinitely repeating 2x2 model using yes 10000 neutrons per generation and more than 1500 generations.

Geometry restrictions No Design Basis Fuel Description yes Fuel density yes See Tables 4.1 & 7.3 Burnable Poisons yes See Tables 2.1, 4.1 & 7.3 and Figure D.4.

Fuel assembly inserts NA Fuel dimensions yes See Tables 4.1 and 7.3 Axial blankets yes Section 2.0 Configurations considered yes Nothing other than the nominal racks.

Borated NA Unborated NA Multiple rack designs yes See Figure 4.2 and Table 6.2 Alternate storage geometry NA

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page E-7 Subject Included Notes / Explanation Reactivity Control Devices NA Fuel Assembly Inserts NA Storage Cell Inserts NA Storage Cell Blocking Devices NA Axial burnup shapes This is a lifetime maximum reactivity NA evaluation. No other assumption is made for the axial power shape.

Uniform/Distributed NA Nodalization NA Blankets modeled NA Tolerances/Uncertainties yes The tolerance evaluation is shown in Table 7.3 Fuel geometry yes Fuel rod pin pitch yes Fuel pellet OD yes Cladding OD yes Axial fuel position [

yes

]

Fuel content yes See Table 7.3 Enrichment yes Density yes Assembly insert dimensions and NA materials Rack geometry yes Flux-trap size (width) yes Nominal Rack to rack spacing was considered. The final k95/95 calculation is based upon the 2x2 infinite array, so this water gap size becomes unimportant.

Rack cell pitch yes Rack wall thickness yes Neutron Absorber Dimensions yes Rack insert dimensions and NA materials Code validation uncertainty yes See Section 7.8 and Appendix C.

Criticality case uncertainty yes A conservative model has been used in the k95/95 calculation.

Depletion Uncertainty yes See Appendix D.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page E-8 Subject Included Notes / Explanation Burnup Uncertainty yes See the sensitivity evaluation in Section 6.6.

Biases Design Basis Fuel design No bias is applied because a reference bounding fuel design is used to conservatively represent all the other fuel types.

Code bias yes See Section 7.8 and Appendix C Temperature yes Not significant (Section 6.6)

Eccentric fuel placement yes Not significant (Sections 7.4.1 and 7.4.2)

Incore thimble depletion effect NA Not a PWR with flux thimbles NRC administrative margin No Modeling simplifications yes See Section 6.1 Identified and described yes 10.0 Interface Analysis Interface configurations analyzed yes See Sections 6.1.4, 6.5 and Figure 6.3 Between dissimilar racks NA Between storage configurations NA within a rack Interface restrictions No 11.0 Normal Conditions Fuel handling equipment yes See Section 7.5 Administrative controls No Fuel inspection equipment or processes yes See Section 7.5 Fuel reconstitution yes See Appendix B 12.0 Accident Analysis Boron dilution NA Normal conditions NA Accident conditions NA Single assembly misload yes See Section 7.6.

Fuel assembly misplacement NA Neutron Absorber Insert Misload NA Multiple fuel misload NA Dropped assembly yes See Section 7.6.

Temperature yes See Section 7.6.

Seismic event / other natural yes See Section 7.6.

Controlled Document Framatome Inc. ANP-3910NP Revision 2 Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel Page E-9 Subject Included Notes / Explanation phenomena 13.0 Analysis Results and Conclusions Summary of results yes See Section 2.0 Burnup curve(s) NA Intermediate Decay time treatment NA New administrative controls No Technical Specification markups TVA 14.0 References Appendix A: Computer Code Validation:

Code validation methodology and bases Performed using guidance from NUREG/CR-6698 in Appendix C.

New Fuel yes Depleted Fuel There is no depletion associated with the KENO benchmark shown in yes Appendix C. The fuel is depleted in the CASMO-4 validation work in Appendix D.

MOX NA HTC NA Convergence yes Trends yes Bias and uncertainty yes Range of applicability yes Analysis of Area of Applicability yes Also see Section 5.1.

coverage (Based on Appendix C of NEI 12-16 Revisions 4)

ATTACHMENT 3 Affidavit CNL-21-050

AFFIDAVIT

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for Framatome Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the pol'cies established by Framatome to ensure the proper application of these criteria.

3. I am familiar with the Framatome information contained in the report ANP-3910P, Revision 2 "Browns Ferry Nuclear Plant Units 1, 2 and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel," dated April 2021 and referred to herein as "Document." Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome 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 Framatome 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 Framatome to determine whether information should be classified as proprietary:

(a) The information reveals details of Framatome'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 Framatome.

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

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

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

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

8. Framatome 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.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on: April 30, 2021

~ j_?JiJn __.~~. ~

Alan B. Meginnis ~

v t e Letter Sequence Request
EPID:L-2021-LLA-0097, TSTF-564 (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance Supplement Administration Withholding Request Acceptance Meeting, Meeting Questions 16 February 2022 Answers 18 March 2022 Results Approval Other:
MONTHYEAR2022-03-18 [Table View]

CNL-21-050, License Amendment Request Regarding Criticality Safety Analysis of Spent Fuel Pool Storage of Atrium 11 Fuel (TS-534)

Text

Subject:

Subject:

SUMMARY

2.0 DESCRIPTION

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.4 CONCLUSION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

8.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

8.0 REFERENCES

Subject:

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