ML090970248

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Enclosure 4 and 6 to BSEP 09-0034 - ANP-2729(NP), Rev. 0, Brunswick, Unit 2, Thermal-Hydraulic Design Report for ATRIUM-10 Fuel Assemblies, and Areva Affidavit Withholding ANP-2727(P), Rev. 0 Form Public Disclosure
ML090970248
Person / Time
Site: Brunswick Duke Energy icon.png
Issue date: 07/31/2008
From:
AREVA, AREVA NP
To:
Office of Nuclear Reactor Regulation
References
BSEP 09-0034 ANP-2727(P), Rev 0, ANP-2729(NP), Rev 0
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BSEP 09-0034 Enclosure 4 AREVA Report ANP-2729(NP), Revision 0, Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T r-1]O Fuel Assemblies, dated July 2008 ANP-2729(NP)

Revision 0 Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies July 2008 ARE VA AREVA NP Inc.ANP-2729(NP)

Revision 0 Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies AREVA NP Inc.ANP-2729(NP)

Revision 0 Copyright

© 2008 AREVA NP Inc.All Rights Reserved paj Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies ANP-2729(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 2 ANP-2729(NP)

Thermal-Hydraulic Design Report Revision 0 for ATRIUM T M-1 0 Fuel Assemblies Page ii Contents 1 .0 Intro d u ctio n ....................................................................................................................

1-1 2.0 S um m ary and C onclusions

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2-1 3.0 Therm al-Hydraulic Design Evaluation

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3-1 3.1 H ydraulic C haracterization

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3-2 3.2 H ydraulic C om patibility

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3-3 3.3 Thermal Margin Performance

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

3 -5 3 .5 B y pa ss F low ................................................................................

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

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3 -5 4 .0 R e fe re n ce s .....................................................................................................................

4 -1 Tables 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM-10 Fuel Assembly ........ .........................

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3-7 3.2 Comparative Description of Brunswick Unit 2 ATRIUM-10 and GE14 Fuel ...................

3-9 3.3 Hydraulic Characterization Comparison Between Brunswick Unit 2 ATRIUM-10 and G E 14 Fuel A ssem blies ..........................................................................................

3-10 3.4 Brunswick Unit 2 Thermal-Hydraulic Design Conditions

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3-11 3.5 Brunswick Unit 2 First Transition Core Thermal-Hydraulic Results at Rated C onditions (100% P / 100% F) ......................................................................................

3-12 3.6 Brunswick Unit 2 First Transition Core Thermal-Hydraulic Results at Off-Rated Conditions (60%P / 45%F) ......................

....... ..........

3-13 3.7 Brunswick Unit 2 Thermal-Hydraulic Results at Rated Conditions (100%P I 100%F) for Transition to ATRIUM-10 Fuel ..................................................................

3-14 3.8 Brunswick Unit 2 Thermal-Hydraulic Results at Off-Rated Conditions (60%P /45%F) for Transition to ATRIUM-10 Fuel .....................................................................

3-15 Figures 3.1 Axial Power Shapes ................................................

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

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

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3-18 AREVA NP Inc.

Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRI UM T M-1 0 Fuel Assemblies ANP-2729(NP)

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

Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies ANP-2729(NP)

Revision 0 Page 1-1 1.0 Introduction The results of Brunswick Unit 2 thermal-hydraulic analyses are presented to demonstrate that AREVA NP* ATRIUM TM_1 0 t fuel is hydraulically compatible with coresident GE14 fuel. This report also provides the hydraulic characterization of the ATRIUM-10 and coresident GE14 fuel designs for Brunswick Unit 2. The ATRIUM-10 fuel design includes the modified high-pressure drop FUELGUARD~t LTP and the snap-in channel seal spring.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).* AREVA NP Inc. is an AREVA and Siemens company.t ATRIUM and FUELGUARD are trademarks of AREVA NP.AREVA NP Inc.

Brunswick Unit 2 ANP-2729(NP)

Thermal-Hydraulic Design Report Revision 0 for ATRIUM T M-10 Fuel Assemblies Page 2-1 2.0 Summary and Conclusions ATRIUM-10 fuel assemblies have been determined to be hydraulically compatible with GE14 fuel coresident in the 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 Tables 3.4 to 3.8.The ATRIUM-10 fuel design is geometrically different from the coresident GE14 design, but hydraulically the two designs are compatible.

[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-10 fuel design. Analyses at rated conditions show core bypass flow varying between [ ] of rated flow for transitioncore configurations ranging from a full GE14 fuel core to a full ATRIUM-10 core, respectively.

Analyses demonstrate the thermal-hydraulic design and compatibility criteria discussed in Section 3.0 are satisfied for the Brunswick Unit 2 transition core consisting of ATRIUM-10 and GE14 fuel for the expected core power distributions and core power/flow conditions encountered during operation.

AREVA NP Inc.

Brunswick Unit 2 ANP-2729(NP)

Thermal-Hydraulic Design Report Revision 0 for ATRIUM T M-10 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-10 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 AOOs. This criterion evaluation is addressed in the mechanical design 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.

Brunswick Unit 2 ANP-2729(NP)

Thermal-Hydraulic Design Report Revision 0 for ATRIUM T M-1 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 is addressed in the mechanical design report.A summary of the thermal-hydraulic design evaluations is given in Table 3.1.3.1 Hydraulic Characterization Basic geometric parameters for ATRIUM-10 and GE14 fuel designs are summarized in Table 3.2. Component loss coefficients for the ATRIUM-10 are based on tests and are presented in Table 3.3. These loss coefficients include modifications to the test data reduction process [] The bare rod friction, ULTRAFLOW T M* spacer, and UTP losses for ATRIUM-10 are based on flow tests. The local losses for the Brunswick ATRIUM-10 modified high-pressure drop FUELGUARD LTP and snap-in channel seal spring (including the effects of a reduced LTP envelope) are based on pressure drop tests performed at AREVA's Portable Hydraulic Test Facility.

[] The local component (LTP, spacer, and UTP) loss coefficients for the GE14 fuel are based on flow test results.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.* ULTRAFLOW is a trademark of AREVA NP.AREVA NP Inc.

Brunswick Unit 2 ANP-2729(NP)

Thermal-Hydraulic Design Report Revision 0 for ATRIUM T M-10 Fuel Assemblies Page 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 Reference 3 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 4.The NRC reviewed the information provided in Reference 5 regarding inclusion of water rod.models in XCOBRA and accepted the inclusion in Reference 6.Hydraulic compatibility, as it relates to the relative performance of the ATRIUM-10 and GE14 fuel designs, has been evaluated.

Detailed analyses were performed for full core GE14 and full core ATRIUM-10 configurations.

Analyses for mixed ATRIUM-10 and GE14 cores 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 ATRIUM-10 and GE14 fuel rated power results are within the range considered compatible, as expected AREVA NP Inc.

Brunswick Unit 2 ANP-2729(NP)

Thermal-Hydraulic Design Report Revision 0 for ATRIUM T M-10 Fuel Assemblies Page 3-4 based on previous transitions involving GE14 fuel. Similar agreement occurs at lower power levels. As shown in Table 3.5, [] Table 3.6 shows that, [] Differences in assembly flow between the ATRIUM-10 and GE14 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 GE14 to ATRIUM-10, the ATRIUM-10 design is considered hydraulically compatible with the GE14 design 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-10 and GE14 fuel are calculated with the SPCB critical power correlation (Reference 7). The NRC-approved methodology to demonstrate the acceptability of using the SPCB correlation for computing GE14 fuel CPR is presented in Reference 8.Assembly design features are incorporated in the CPR calculation through the F-eff term. The F-eff is based on the local power peaking for the nuclear design and on additive constants determined in accordance with approved procedures.

The localpeaking factors are a function of assembly void fraction and exposure.For the compatibility evaluation, steady-state analyses evaluated ATRIUM-10 and GE14 assemblies with radial peaking factors (RPFs) between [AREVA NP Inc.

Brunswick Unit 2 ANP-2729(NP)

Thermal-Hydraulic Design Report Revision 0 for ATRIUM T M-10 Fuel Assemblies Page 3-5] Table 3.5 and Table 3.6 show CPR results of the ATRIUM-10 and GE14 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-10 fuel will not cause thermal margin problems for the coresident GE14 fuel.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.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 a full GE14 core to a full ATRIUM-10 core (middle-peaked power shape). [] In summary, adequate bypass flow will be available with the introduction of the ATRIUM-10 fuel design and applicable design criteria are met.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 9). The study AREVA NP Inc.

Brunswick Unit 2 ANP-2729(NP)

Thermal-Hydraulic Design Report Revision 0 for ATRIUM T M-10 Fuel Assemblies Page 3-6 shows that the ATRIUM-10 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 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies ANP-2729(NP), Revision 0'Page 3-7 Table 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM-10 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 similar to existing fuel ATRIUM-10 demonstrated to be such that there is no compatible with GE14.significant impact on total core flow or flow distribution among assemblies.

3.3 Thermal

margin Fuel design shall be SPCB is applied to both the performance within the limits of ATRIUM-10 and GE14 fuel.applicability of an approved CHF correlation.

< 0.1% of rods in boiling Verified on cycle-specific basis for transition.

Chapter 15 analyses.Fuel centerline No centerline melting. Refer to the mechanical design temperature report.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 similar among Analysis results demonstrate that assemblies to provide adequate bypass flow is provided.adequate bypass flow.AREVA NP Inc.

Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies ANP-2729(NP)

Revision 0 Page 3-8 Table 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM-10 Fuel Assembly (Continued)

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

3.6 Stability

New fuel designs are ATRIUM-10 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 than) existing (approved)

Core stability behavior is evaluated AREVA fuel designs. on a cycle-specific basis.LOCA analysis LOCA analyzed in Approved Appendix K LOCA accordance with model.Appendix K modeling requirements.

Criteria Plant- and fuel-specific analysis defined in 10 CFR 50.46. with cycle-specific verifications.

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 Refer to the mechanical design liftoff engaged in fuel support. report.AREVA NP Inc.

Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M_-10 Fuel Assemblies ANP-2729(NP)

Revision 0 Page 3-9 Table 3.2 Comparative Description of Brunswick Unit 2 ATRIUM-10 and GE14 Fuel Fuel Parameter ATRIUM-10 GE14 Number of fuel rods Full-length fuel rods 83 78 PLFRs 8 14 Fuel clad OD, in 0.3957 0.404 Number of spacers 8 8 Active fuel length, ft Full-length fuel rods 12.454 12.500 PLFRs 7.5 7.0 Hydraulic resistance characteristics Table 3.3 Table 3.3 Number of water rods 1 2 Water rod OD, in 1.378*' 0.980* Square water channel outer width.AREVA NP Inc.

Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies ANP-2729(NP)

Revision 0 Page 3-10 Table 3.3 Hydraulic Characterization Comparison Between Brunswick Unit 2 ATRIUM-10 and GE14 Fuel Assemblies

[I[I AREVA NP Inc.

  • Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRI UM T M-10 Fuel Assemblies ANP-2729(NP)

Revision 0 Page 3-11 I Table 3.4 Brunswick Unit 2 Thermal-Hydraulic Design Conditions Reactor conditions 100%P / 100%F 60%P / 45%F Core power level, MWt 2923.0 1753.8 Core exit pressure, ,psia 1054.5 986.0 Core inlet enthalpy, Btu/lbm 528.3 504.3 Total core coolant flow, Mlbm/hr 77 34.65 Axial power shape Middle-peaked Middle-peaked (Figure 3.1) (Figure 3.1)Number of Assemblies Central Peripheral Region Region First Transition Core Loading[ i[ ]Second Transition Core Loading[ ][ ]AREVA NP Inc.

Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies ANP-2729(NP)

Revision 0 Page 3-12 Table 3.5 Brunswick Unit 2 First Transition Core Thermal-Hydraulic Results at Rated Conditions (100%P I 100%F)I I AREVA NP Inc.

Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies ANP-2729(NP)

Revision 0 Page 3-13 Table 3.6 Brunswick Unit 2 First Transition Core Thermal-Hydraulic Results at Off-Rated Conditions (60%P / 45%F)I I I AREVA NP Inc.

Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies ANP-2729(NP)

Revision 0 Page 3-14 Table 3.7 Brunswick Unit 2 Thermal-Hydraulic Results at Rated Conditions (100%P I 100%F) for Transition to ATRIUM-10 Fuel I AREVA NP Inc.

Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies ANP-2729(NP)

  • Revision 0 Page 3-15 Table 3.8 Brunswick Unit 2 Thermal-Hydraulic Results at Off-Rated Conditions (60%P / 45%F) for Transition to ATRIUM-10 Fuel I AREVA NP Inc.

Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies ANP-2729(NP)

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

Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRI UM T M-1 0 Fuel Assemblies ANP-2729(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 2 Thermal-Hydraulic Design Report for ATRIUMTM-1 0 Fuel Assemblies ANP-2729(NP)

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

Brunswick Unit 2 ANP-2729(NP)

Thermal-Hydraulic Design Report Revision 0 for ATRIUM T M-10 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. XN-NF-79-59(P)(A), Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies, Exxon Nuclear Company, November 1983.4. 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 1987.5. Letter, R.A. Copeland (ANF) to R.C. Jones (USNRC), "Explicit Modeling of BWR Water Rod in XCOBRA," RAC:002:90, January 9, 1990.6. Letter, R.C. Jones (USNRC) to R.A. Copeland (ANF), no subject (regarding XCOBRA water rod model), February 1, 1990.7. EMF-2209(P)(A)

Revision 2, SPCB Critical Power Correlation, Framatome ANP, September 2003.8. EMF-2245(P)(A)

Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.9. 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 09-0034 Enclosure 6 AREVA Affidavit Regarding Withholding ANP-2727(P), Revision 0, from Public Disclosure AFFIDAVIT COMMONWEALTH OF VIRGINIA )) ss.CITY OF LYNCHBURG

)1. My name is Gayle F. Elliott. 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 familiarwith 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 ANP-2727(P), Revision 0, entitled "Brunswick Unit 2 Cycle 19 Fuel Cycle Design," dated June 2008 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) and 6(c) 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 I /H day of JY. ,2008.Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/10 Reg. # 7079129 SHERRY L. MCFA-,-N Notary Public Commonwealth of Wtglvdo 7079129 MY Commission Explres Oat 31. 1010
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