ML12100A087

ANP-2989(NP), Revision 0, Brunswick, Unit 1, Thermal-Hydraulic Design Report for Atrium 10XM Fuel Assemblies.

ML12100A087
Person / Time
Site:	Brunswick
Issue date:	05/31/2011
From:	AREVA NP
To:	Office of Nuclear Reactor Regulation
References
BSEP 12-0040 ANP-2989(NP), Rev 0
Download: ML12100A087 (32)

Text

BSEP 12-0040 Enclosure 4 ANP-2989(NP), Brunswick Unit ] Thermal-Hydraulic Design Report for ATRIUMTM 1 OXM Fuel Assemblies, Revision 0

.ument ANP-2989(NP)

Revision 0 Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies May 2011 A AREVA NP Inc. AR EVA Document AREVA NP Inc.ANP-2989(NP)

Revision 0 Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M IOXM Fuel Assemblies AREVA NP Inc.ANP-2989(NP)

Revision 0 Copyright

© 2011 AREVA NP Inc.All Rights Reserved Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page i Nature of Changes Item Page Description and Justification

1. All This is the initial issue.AREVA NP Inc.

Brunswick Unit 1 Thermal-Hydraulic ANP-2989(NP)

Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page ii Contents 1 .0 In tro d u c tio n ....................................................................................................................

1-1 2.0 Summary and Conclusions

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2-1 3.0 Thermal-Hydraulic Design Evaluation

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3-1 3.1 Hydraulic Characterization

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3-2 3 .2 H yd raulic C om patibility

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3-2 3.3 Thermal Margin Performance

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3-4 3 .4 R o d B o w .............................................................................................................

3 -5 3 .5 B y p a s s F lo w .......................................................................................................

3 -5 3 .6 S ta b ility ...............................................................................................................

3 -6 4 .0 R e fe re n c e s .....................................................................................................................

4 -1 Tables 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM 1OXM Fuel Assembly ......................................................................................

3-7 3.2 Comparative Description for Brunswick Unit 1 ATRIUM 1OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG Fuel Types ......................................................

3-9 3.3 Hydraulic Characterization Comparison for Brunswick Unit 1 ATRIUM 1OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG Fuel ....................................

3-10 3.4 Brunswick Unit 1 Thermal-Hydraulic Design Conditions

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3-11 3.5 Brunswick Unit 1 First Transition Core Thermal-Hydraulic Results at Rated Conditions (100%P / 100%F) ............................................................................

3-12 3.6 Brunswick Unit 1 First Transition Core Thermal-Hydraulic Results at Off-Rated Conditions (60%P / 45%F) ..........................................................................

3-13 3.7 Brunswick Unit 1 Thermal-Hydraulic Results at Rated Conditions (100%P / 100%F) for Transition to ATRIUM IOXM Fuel ..............................................

3-14 3.8 Brunswick Unit 1 Thermal-Hydraulic Results at Off-Rated Conditions (60%P / 45%F) for Transition to ATRIUM 1OXM Fuel ..................................................

3-15 Figures 3 .1 A xia l P ow e r S ha pe s .....................................................................................................

3-16 3.2 First Transition Core: Hydraulic Demand Curves 100%P / 100%F ..............................

3-17 3.3 First Transition Core: Hydraulic Demand Curves 60%P / 45%F ..................................

3-18 AREVA NP Inc.

ntrolled Docume Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page iii Nomenclature AOO ASME BRK1-18 BRK1-19 BWR CHF CPR CRDA IFG LOCA LTP MAPLHGR MCPR MFG NRC PLFR RPF UTP anticipated operational occurrence American Society of Mechanical Engineers Brunswick Unit 1 Cycle 18 Brunswick Unit 1 Cycle 19 boiling water reactor critical heat flux critical power ratio control rod drop accident Improved FUELGUARDTM loss-of-coolant accident lower tie plate maximum average planar linear heat generation rate minimum critical power ratio Modified FUELGUARD Nuclear Regulatory Commission, U.S.part-length fuel rod radial peaking factor upper tie plate AREVA NP Inc.

Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 1-1 1.0 Introduction The results of Brunswick Unit 1 thermal-hydraulic analyses are presented to demonstrate that AREVA NP ATRIUM T M 1OXM* fuel is hydraulically compatible with the previously loaded ATRIUM-1 0 fuel with improved FUELGUARD T M lower tie plates (designated in this report as IFG) and ATRIUM-10 fuel with modified FUELGUARD lower tie plates (designated in this report as MFG). This report also provides the hydraulic characterization of the ATRIUM 1OXM and the coresident ATRIUM-10 IFG and ATRIUM-10 MFG designs for Brunswick Unit 1.The generic thermal-hydraulic design criteria applicable to the design have been reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC) in the topical report ANF-89-98(P)(A)

Revision 1 and Supplement 1 (Reference 1). In addition, thermal-hydraulic criteria applicable to the design have also been reviewed and approved by the NRC in the topical report XN-NF-80-19(P)(A)

Volume 4 Revision 1 (Reference 2).ATRIUM and FUELGUARD are trademarks of AREVA NP.AREVA NP Inc.

Controiled Document Brunswick Unit 1 Thermal-Hydraulic ANP-2989(NP)

Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 2-1 2.0 Summary and Conclusions ATRIUM 1OXM fuel assemblies have been determined to be hydraulically compatible with the coresident ATRIUM-10 IFG and ATRIUM-10 MFG fuel designs in the Brunswick Unit 1 reactor for the entire range of the licensed power-to-flow operating map. Detailed calculation results supporting this conclusion are provided in Section 3.2 and Table 3.4 to Table 3.8.The ATRIUM 1OXM fuel design is geometrically different from the coresident ATRIUM-10 IFG and ATRIUM-10 MFG designs, but the designs are hydraulically compatible.

[I Core bypass flow (defined as leakage flow through the lower tie plate (LTP) flow holes, channel seal, core support plate, and LTP-fuel support interface) is not adversely affected by the introduction of the ATRIUM 1OXM fuel design. Analyses at rated conditions show core bypass flow varying between [ ] of rated flow for transition core configurations ranging from the BRK1-18 core loading with ATRIUM-10 IFG, ATRIUM-10 MFG, and GE14 fuel to a full ATRIUM 1OXM core, respectively.

Analyses demonstrate the thermal-hydraulic design and compatibility criteria discussed in Section 3.0 are satisfied for the Brunswick Unit 1 transition cores consisting of ATRIUM 1 OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG fuel for the expected core power distributions and core power/flow conditions encountered during operation.

AREVA NP Inc.

Brunswick Unit 1 Thermal-Hydraulic ANP-2989(NP)

Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-1 3.0 Thermal-Hydraulic Design Evaluation Thermal-hydraulic analyses are performed to verify that design criteria are satisfied and to help establish thermal operating limits with acceptable margins of safety during normal reactor operation and anticipated operational occurrences (AOOs). The design criteria that are applicable to the ATRIUM 1 OXM fuel design are described in Reference

1. To the extent possible, these analyses are performed on a generic fuel design basis. However, due to reactor and cycle operating differences, many of the analyses supporting these thermal-hydraulic operating limits are performed on a plant- and cycle-specific basis and are documented in plant- and cycle-specific reports.The thermal-hydraulic design criteria are summarized below: Hydraulic compatibility.

The hydraulic flow resistance of the reload fuel assemblies shall be sufficiently similar to the existing fuel in the reactor such that there is no significant impact on total core flow or the flow distribution among assemblies in the core. This criterion evaluation is addressed in Sections 3.1 and 3.2.* Thermal margin performance.

Fuel assembly geometry, including spacer design and rod-to-rod local power peaking, should minimize the likelihood of boiling transition during* normal reactor operation as well as during AOOs. The fuel design should fall within the bounds of the applicable empirically based boiling transition correlation approved for AREVA reload fuel. Within other applicable mechanical, nuclear, and fuel performance constraints, the fuel design should.achieve good thermal margin performance.

The thermal-hydraulic design impact on steady-state thermal margin performance is addressed in Section 3.3. Additional thermal margin performance evaluations dependent on the cycle-specific design are addressed in the reload licensing report.Fuel centerline temperature.

Fuel design and operation shall be such that fuel centerline melting is not projected for normal operation and AQOs. This criterion evaluation is addressed in the fuel rod thermal and mechanical evaluation report.Rod bow. The anticipated magnitude of fuel rod bowing under irradiation shall be accounted for in establishing thermal margin requirements.

This criterion evaluation is addressed in Section 3.4.Bypass flow. The bypass flow characteristics of the reload fuel assemblies shall not differ significantly from the existing fuel in order to provide adequate flow in the bypass region. This criterion evaluation is addressed in Section 3.5.Stability.

Reactors fueled with new fuel designs must be stable in the approved power and flow operating region. The stability performance of new fuel designs will be equivalent to, or better than, existing (approved)

AREVA fuel designs. This criterion evaluation is addressed in Section 3.6. Additional core stability evaluations dependent on the cycle-specific design are addressed in the reload licensing report.AREVA NP Inc.

Controlled Documert Brunswick Unit 1 Thermal-Hydraulic ANP-2989(NP)

Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-2 Loss-of-coolant accident (LOCA) analysis.

LOCAs are analyzed in accordance with Appendix K modeling requirements using NRC-approved models. The criteria are defined in 10 CFR 50.46. LOCA analysis results are presented in the break spectrum and MAPLHGR reports.Control rod drop accident (CRDA) analysis.

The deposited enthalpy must be less than 280 cal/gm for fuel coolability.

This criterion evaluation is addressed in the reload licensing report.ASME overpressurization analysis.

ASME pressure vessel code requirements must be satisfied.

This criterion evaluation is addressed in the reload licensing report.Seismic/LOCA liftoff. Under accident conditions, the assembly must remain engaged in the fuel support. This criterion evaluation was addressed in Reference 3.A summary of the thermal-hydraulic design evaluations is given in Table 3.1.3.1 Hydraulic Characterization Basic geometric parameters for the ATRIUM 1OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG fuel designs are summarized in Table 3.2. Component loss coefficients for the ATRIUM 10XM are based on tests and are presented in Table 3.3. These loss coefficients include modifications to the test data reduction process [I The bare rod friction, ULTRAFLOWTM*

spacer, UTP and LTP losses forATRIUM 1OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG are based on tests performed at AREVA's Portable Hydraulic Test Facility.

[The primary resistance for the leakage flow through the LTP flow holes is [] The resistances for the leakage paths are shown in Table 3.3.3.2 Hydraulic Compatibility The thermal-hydraulic analyses were performed in accordance with the AREVA thermal-hydraulic methodology for BWRs. The methodology and constitutive relationships used by AREVA for the calculation of pressure drop in BWR fuel assemblies are presented in* ULTRAFLOW is a trademark of AREVA NP.AREVA NP Inc.

Document Brunswick Unit 1 Thermal-Hydraulic ANP-2989(NP)

Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-3 Reference 4 and are implemented in the XCOBRA code. The XCOBRA code predicts steady-state thermal-hydraulic performance of the fuel assemblies of BWR cores at various operating conditions and power distributions.

XCOBRA received NRC approval in Reference 5.The NRC reviewed the information provided in Reference 6 regarding inclusion of water rod models in XCOBRA and accepted the inclusion in Reference 7.Hydraulic compatibility, as it relates to the relative performance of the ATRIUM 1OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG fuel designs, has been evaluated.

Detailed analyses were performed for the Brunswick Unit 1 Cycle 18 and full core ATRIUM 10XM configurations.

Analyses for mixed cores with ATRIUM 1OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG fuel were also performed to demonstrate that the thermal-hydraulic design criteria are satisfied for transition core configurations.

The hydraulic compatibility analysis is based on [Table 3.4 summarizes the input conditions for the analyses.

These conditions reflect two of the state points considered in the analyses:

100% power/1 00% flow and 60% power/45%

flow.Table 3.4 also defines the core loading for the transition core configurations.

Input for other core configurations is similar in that core operating conditions remain the same and the same axial power distribution is used. Evaluations were made with the bottom-, middle-, and top-peaked axial power distributions presented in Figure 3.1. Results presented in this report are for the middle-peaked power distribution.

Results for bottom- and top-peaked axial power distributions show similar trends.Table 3.5 and Table 3.6 provide a summary of calculated thermal-hydraulic results using the first transition core configuration.

Table 3.7 and Table 3.8 provide a summary of results for all core configurations evaluated.

Core average results and the differences between the ATRIUM 10XM, ATRIUM-10 IFG, and ATRIUM-10 MFG results at rated power are within the range considered compatible, as expected.

Similar agreement occurs at lower power levels.AREVA NP Inc.

Brunswick Unit 1 Thermal-Hydraulic ANP-2989(NP)

Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-4 As shown in Table 3.5, [] Table 3.6 shows that [] Differences in assembly flow between the ATRIUM 1OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG fuel designs as a function of assembly power level are shown in Figure 3.2 and Figure 3.3.Core pressure drop and core bypass flow fraction are also provided for the configurations evaluated.

Based on the reported changes in pressure drop and assembly flow caused by the transition from the BRK1-18 core loading to a full core of ATRIUM 1OXM, the ATRIUM 1OXM design is considered hydraulically compatible with the coresident fuel designs since the thermal-hydraulic design criteria are satisfied.

3.3 Thermal

Margin Performance Relative thermal margin analyses were performed in accordance with the thermal-hydraulic methodology for AREVA's XCOBRA code. The calculation of the fuel assembly critical power ratio (CPR) (thermal margin performance) is established by means of an empirical correlation based on results of boiling transition test programs.

The CPR methodology is the approach used by AREVA to determine the margin to thermal limits for BWRs.CPR values for ATRIUM 1OXM are calculated with the ACE/ATRIUM 1OXM critical power correlation (Reference

8) while the CPR values for the ATRIUM-10 IFG and ATRIUM-10 MFG fuel are calculated with the SPCB critical power correlation (Reference 9). Assembly design features are incorporated in the CPR calculation through the K-factor term in the ACE correlation and the F-eff term for the SPCB correlation.

The K-factors and F-effs are based on the local power peaking for the nuclear design and on additive constants determined in accordance with approved procedures.

The local peaking factors are a function of assembly void fraction and exposure.For the compatibility evaluation, steady-state analyses evaluated ATRIUM 1 OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG assemblies with radial peaking factors (RPFs) between[ ] The Reference 10 discussion identifies the concern that the K-factors used in the ACE correlation for the analysis of ATRIUM AREVA NP Inc.

Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-5 1OXM [] This finding does not impact the validity of the results of this evaluation.

[Table 3.5 and Table 3.6 show CPR results of the ATRIUM 1OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG fuels. Table 3.7 and Table 3.8 show similar comparisons of CPR and assembly flow for the various core configurations evaluated.

Analysis results indicate ATRIUM 1OXM fuel will not cause thermal margin problems for the coresident fuel designs.3.4 Rod Bow The bases for rod bow are discussed in the mechanical design report. Rod bow magnitude is determined during the fuel-specific mechanical design analyses.

Rod bow has been measured during, post-irradiation examinations of BWR fuel fabricated by AREVA.[ C I 3.5 Bypass Flow Total core bypass flow is defined as leakage flow through the LTP flow holes, channel seal, core support plate, and LTP-fuel support interface.

Table 3.7 shows that total core bypass flow (excluding water rod flow) fraction at rated conditions changes from [ ] of rated core flow during the transition from the BRK1-18 core loading to a full ATRIUM 1OXM core (middle-peaked power shape). In summary, adequate bypass flow will be available with the introduction of the ATRIUM 10XM fuel design and applicable design criteria are met.AREVA NP Inc.

C JDocument Brunswick Unit 1 Thermal-Hydraulic ANP-2989(NP)

Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-6 3.6 Stability Each new fuel design is analyzed to demonstrate that the stability performance is equivalent to or better than an existing (NRC-approved)

AREVA fuel design. The stability performance is a function of the core power, core flow, core power distribution, and to a lesser extent, the fuel design. [] A comparative stability analysis was performed with the NRC-approved STAIF code (Reference 11). The study shows that the ATRIUM 1OXM fuel design has decay ratios equivalent to or better than other approved AREVA fuel designs.As stated above, the stability performance of a core is strongly dependent on the core power, core flow, and power distribution in the core. Therefore, core stability is evaluated on a cycle-specific basis and addressed in the reload licensing report.AREVA NP Inc.

Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-7 Table 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM 1OXM Fuel Assembly Report Section Description Criteria Results or Disposition Thermal and Hydraulic Criteria 3.1 / 3.2 Hydraulic Hydraulic flow resistance Verified on a plant-specific basis.compatibility shall be sufficiently ATRIUM 1OXM demonstrated to be similar to existing fuel compatible with ATRIUM-10 IFG such that there is no significant impact on total core flow or flow [distribution among assemblies.

3.3 Thermal

margin Fuel design shall be ACE/ATRIUM 1OXM critical power performance within the limits of correlation is applied to the applicability of an ATRIUM 1OXM fuel.approved CHF correlation.

SPCB critical power correlation is applied to the ATRIUM-10 IFG and ATRIUM-10 MFG fuel.< 0.1% of rods in boiling Verified on cycle-specific basis for transition.

Chapter 15 analyses.Fuel centerline No centerline melting. Plant- and fuel-specific analyses temperature are performed.

3.4 Rod bow Rod bow must be The lateral displacement of the fuel accounted for in rods due to fuel rod bowing is not of establishing thermal sufficient magnitude to impact margins, thermal margins.3.5 Bypass flow Bypass flow Verified on a plant-specific basis.characteristics shall be Analysis results demonstrate that similar among adequate bypass flow is provided.assemblies to provide adequate bypass flow.AREVA NP Inc.

Controle Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-8 Table 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM 1OXM Fuel Assembly (Continued)

Report Section Description Criteria Results or Disposition Thermal and Hydraulic Criteria (Continued)

3.6 Stability

New fuel designs are ATRIUM 1OXM channel and core stable in the approved decay ratios have been power and flow operating demonstrated to be equivalent to or region, and stability better than other approved AREVA performance will be fuel designs.equivalent to (or better Core stabilit behavior is evaluated than) existing (approved) on a bc y ba sis.AREVA fuel designs. on a cycle-specific basis.LOCA analysis LOCA analyzed in Approved Appendix K LOCA accordance with model.Appendix K modeling Plant- and fuel-specific analysis requirements.

Criteria with cycle-specific verifications.

defined in 10 CFR 50.46.-CRDA analysis < 280 cal/gm for Cycle-specific analysis is coolability.

performed.

ASME over- ASME pressure vessel Cycle-specific analysis is pressurization core requirements shall performed.

analysis be satisfied.

Seismic/LOCA Assembly remains Criterion met per Reference 3.liftoff engaged in fuel support.AREVA NP Inc.

Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-9 Table 3.2 Comparative Description for Brunswick Unit 1 ATRIUM 1OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG Fuel Types Fuel Parameter ATRIUM 10XM ATRIUM-10 IFG ATRIUM-10 MFG Number of fuel rods Full-length fuel rods 79 83 83 PLFRs 12 8 8 Fuel clad OD, in 0.4047 0.3957 0.3957 Number of spacers 9 8 8 Active fuel length, ft Full-length fuel rods 12.500 12.454 12.454 PLFRs 6.25 7.5 7.5 Hydraulic resistance characteristics Table 3.3 Table 3.3 Table 3.3 Number of water rods 1 1 1 Water rod OD, in 1.378* 1.378* 1.378** Square water channel outer width.AREVA NP Inc.

Docurmen Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-10 Table 3.3 Hydraulic Characterization Comparison for Brunswick Unit I ATRIUM 1OXM, ATRIUM-10 IFG, and ATRIUM-10 MFG Fuel[I I I AREVA NP Inc.

Ccon "9 ý'ocument Brunswick Unit 1 Thermal-Hydraulic ANP-2989(NP)

Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-11 Table 3.4 Brunswick Unit 1 Thermal-Hydraulic Design Conditions Reactor Conditions 1 00%P / 1 00%F 60%P / 45%F Core power level, MWt 2923.0 1753.8 Core exit pressure, psia 1058.3 993.1 Core inlet enthalpy, Btu/lbm 528.3 505.2 Total core coolant flow, Mlbm/hr 77.0 34.7 Axial power shape Middle-peaked Middle-peaked (Figure 3.1) (Figure 3.1)Number of Assemblies Central Peripheral Region Region BRK1-18 Core Loading[ ][ ][ ]First Transition Core Loading[ ][ ][]Second Transition Core Loading[ ][ ]AREVA NP Inc.

Joeacrument Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-12 Table 3.5 Brunswick Unit 1 First Transition Core Thermal-Hydraulic Results at Rated Conditions (100%P / 100%F)[I I AREVA NP Inc.

ed D)ocument Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-13 Table 3.6 Brunswick Unit I First Transition Core Thermal-Hydraulic Results at Off-Rated Conditions (60%P / 45%F)I I I AREVA NP Inc.

Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-14 Table 3.7 Brunswick Unit 1 Thermal-Hydraulic Results at Rated Conditions (100%P / 100%F) for Transition to ATRIUM 1OXM Fuel[I AREVA NP Inc.

-ocment Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-15 Table 3.8 Brunswick Unit I Thermal-Hydraulic Results at Off-Rated Conditions (60%P / 45%F) for Transition to ATRIUM 1OXM Fuel[I AREVA NP Inc.

Contro ned t Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-16[I Figure 3.1 Axial Power Shapes AREVA NP Inc.

Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-17[I Figure 3.2 First Transition Core: Hydraulic Demand Curves 100%P / 100%F AREVA NP Inc.

Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-2989(NP)

Revision 0 Page 3-18 I I Figure 3.3 First Transition Core: Hydraulic Demand Curves 60%P / 45%F AREVA NP Inc.

~2U L c.ient Brunswick Unit 1 Thermal-Hydraulic ANP-2989(NP)

Design Report for ATRIUM T M 10XM Revision 0 Fuel Assemblies Page 4-1 4.0 References

1. ANF-89-98(P)(A)

Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.2. XN-NF-80-19(P)(A)

Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.3. ANP-2948P Revision 0, Mechanical Design Report for Brunswick ATRIUM IOXM Fuel Assemblies, AREVA NP, October 2010.4. XN-NF-79-59(P)(A), Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies, Exxon Nuclear Company, November 1983.5. XN-NF-80-19(P)(A)

Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.6. Letter, R.A. Copeland (ANF) to R.C. Jones (USNRC), "Explicit Modeling of BWR Water Rod in XCOBRA," RAC:002:90, January 9, 1990.7. Letter, R.C. Jones (USNRC) to R.A. Copeland (ANF), no subject (regarding XCOBRA water rod model), February 1, 1990.8. ANP-10298PA Revision 0, ACE/ATRIUM 1OXM Critical Power Correlation, AREVA NP, March 2010.9. EMF-2209(P)(A)

Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.10. Condition Report 2011-2274, [AREVA NP, March 2011.11. EMF-CC-074(P)(A)

Volume 1, STAIF -A Computer Program for BWR Stability Analysis in the Frequency Domain; and Volume 2, STAIF -A Computer Program for BWR Stability Analysis in the Frequency Domain -Code Qualification Report, Siemens Power Corporation, July 1994.AREVA NP Inc.

BSEP 12-0040 Enclosure 6 AREVA Affidavit Regarding Withholding ANP-3061(P), Brunswick Unit 1 Cycle 19 Reload Safety Analysis, Revision 0, from Public Disclosure AFFIDAVIT STATE OF WASHINGTON

)) ss.COUNTY OF BENTON )1. My name is Alan B. Meginnis.

I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary.

I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.3. I am familiar with the AREVA NP information contained in the report ANP-3061.(P)

Revision 0, entitled, "Brunswick Unit 1,Cycle 19 Reload Safety Analysis," dated December 2011 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

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

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information." 6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA NP's research and development plans and programs or their results.(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.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 AREVA NP'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 AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

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

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.SUBSCRIBED before me this CO day of' , 2011.Susan K. McCoy\NOTARY PUBLIC, STATE OF WAS TON MY-COMMISSION EXPIRES: 1/10/12