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ANP-3108(NP), Revision 1, Applicability of Areva BWR Methods to Brunswick Extended Power Flow Operating Domain.
	ML16257A412
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Site:	Brunswick
Issue date:	07/31/2015
From:	; AREVA
To:	; Office of Nuclear Reactor Regulation
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ML16257A418	List: BSEP 16-0056, ANP-3108(NP), Revision 1, Applicability of Areva BWR Methods to Brunswick Extended Power Flow Operating Domain.; ML16257A404; ML16257A406; ML16257A407; ML16257A408; ML16257A410; ML16257A411;
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	BSEP 16-0056 ANP-3108(NP), Rev 1
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BSEP 16-0056 Enclosure 13 ANP-3108NP, Revision 1, Applicability of AREVA BWR Methods to Brunswick Extended Power Flow Operating Domain, July 2015

Controlled Document A

AREVA ANP-3108NP Applicability of AREVA BWR Revision 1 Meth.ods to Brunswick Extended Power Flow Operating Domain July 2015

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page i Nature of Changes Item Page Description and Justification

1. All Changed AREVA_ NP to AREVA. Additionally, updated format in all the tables to look the same.

2. 3-2, B-9, Updated void fraction data presented in Figure 3-1, Ta,ble B-1, Figure B-2, B-11, B-12, and Figures B B-6 to address issues identified in CR 2014-5400.

and B-13

3. F-6 Updated responses to NRC request

4. F-10 Updated Table F-1 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page ii Contents 1.0 Introduction .................................................................................................................... 1-1 2.0 Overview ........................................................................................................................2-1 3.0 Thermal-hydraulics .........................................................................................................3-1 4.0 AREVA CHF/CPR Correlations ......................................................................................4-1 5.0 Safety Limit MCPR .........................................................................................................5-1 6.0 Mechanical Limits Methodology ..................................................................................... 6-1 7.0 Core Neutronics ............................................................................................................. 7-1 8.0 Transient Analysis ..........................................................................................................8-1 9.0 LOCA Analysis ...............................................................................................................9-1 10.0 Stability Analysis .......................... .- ............................................................................... 10-1 10.1 Linear Stability .................................................................................................. 10-1 10.2 DIVOM .............................................................................................................. 10-1 10.3 ATWS-1 ............................................................................................................. 10-1 11.0 Summary ...................................................................................................................... 11-1 12.0 References ................................................................................................................... 12-1 Appendix A Application of AREVA Methodology for Mixed Cores ..................................... A-1 A.1 Discussion ................................... :..................................................................... A-1 Appendix B Void~Quality Correlations ................................................................................ B-1' B.1 AREVA Void Quality Correlations ...................................................................... B-1 B.2 Void Quality Correlation Uncertainties ............................................................... B-4 B.3 Biasing of the Void-Quality Correlation .............................................................. B-5 B.4 Void-Quality Correlation Uncertainty Summary ................................................. B-7 Appendix C Neutronic Methods .......................................................................................... C-1 C.1 Cross Section Representation ........................................................................... C-1 C.2 Applicability of Uncertainties .............................................................................. C-4 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page iii C.3 Fuel Cycle Comparisons ................................................................................... C-8 C.~.1 Bypass Voiding ........................................................................................................... C-10 C.3.2 Fuel Assembly Design ................................................................................................ C-12 Appendix D Transient Methods .......................................................................................... D-1 D.1 COTRANSA2 Cross Section Representation .................................................... D-1 Appendix E LOCA Modifications ........................................................................................ E-1 E.1 LOCA Analysis .................................................................................................. E-1 Appendix F Fuel Conductivity Degrad(;':ltion .........................................................................F-1 F.1 lntroduction .........................................................................................................F-1 F .2 Disposition of Licensing Safety Analysis for Brunswick ATRIUM 10XM Fuel ...........................................................................................F-1 F.3 Assessment of Analyses for Brunswick Operations ........................................... F-2 F.3.1 Anticipated Operational Occurrence Analyses ....................................... F-2 F.3.2 Loss of Coolant Accident Analyses ........................................................ F-4 F.3.2.1 Responses to NRG Requests ..................................................... F-5 F.3.3 Overpressurization Analyses .................................................................. F-6 F.3.3.1 Responses to NRG Requests ..................................................... F-8 F.3.4 Stability Analyses ....................................................................................F-9 F.3.5 Fire Event Analyses ................................................................................F-9 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page iv Tables Table 2-1 AREVA Licensing Topical Reports ....................................................................2-2 Table 7-1 CASM0-4/MICROBURN-B2 Operating Experience .......................................... 7-2 Table B-1 AREVA Multi-Rod Void Fraction Validation Database ...................................... B-9 Table B-2 Void Sensitivity Results .................................................................................. B-10 Table C-1 KWU-S Gamma Scan Benchmark Results from EMF-2158(P)(A) .................. C-13 Table C-2 Comparison of CASM0-4 and MCNP results for ATRIUM-10 Design ........................................................................................................... C-13 Table C-3 Fuel Enrichment Description for the Initial Brunswick EPFOD ATRIUM 10XM Fuel Cycle Design ..................................................... .'........... C-14 Table F-1 Impact of TCD on PCT ................................................................................... F-10 Table F-2 Brunswick EPFOD Overpressurization Biases and Results ........................... F-11 Figures Figure 1-1 Brunswick Power Flow Operating Map with the MELLLA+ EPFOD ................... 1-2 Figure 3-1 Comparison of Karlstein Two-Phase Pressure Drop and void fraction Test Matrices and Typical EPFOD Reactor Conditions ......................... 3-2 Figure 5-1 Assembly Power Distribution for Limiting Case in Safety Limit MCPR Analysis (Pre-EPU and EPU) .................................................................5-3 Figure 5-2 Assembly Power Distribution for Limiting Case in Safety Limit MCPR Analysis (EPU and MELLLA+) ............................................................... 5-4 Figure 8-1 Typical Hydraulic Benchmarks to Karlstein Transient Simulations .................... 8-3 Figure B-1 Validation of [ ] using FRIGG-2 and FRIGG-3 Void Data ............................................................................................................... B-11 Figure B-2 Validation of [ ] using ATRIUM-10 and ATRIUM 10XM Void Data .........................................................................................'.............. B-11 Figure B-3 Validation of Ohkawa-Lahey using FRIGG-2 and FRIGG-3 Void Data ...................................................... :........................................................ B-12 Figure B-4 Validation of Ohkawa-Lahey using ATRIUM-10 and ATRIUM 10XM Void Data ............................................................................................. B-12 Figure B-5 Modified Void Fraction Correlation Comparison to ATRI UM-1 O Test Data ....................................................................................................... B-13 Figure B-6 [ ] Void Fraction Comparison to ATRIUM-10 Test Data ....................................................................................................... B-13 AREVA Inc.

Controlled Document Applicability of AREVA BWR , ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Pagev Figure C-1 Microscopic Thermal Cross Section of U-235 from Base Depletion and Branches .................................................................................................. C-15 Figure C-2 Microscopic Fast Cross Section of U-235 from Base Depletion and Branches ......................................................................................................... C-15 Figure C-3 Microscopic Thermal Cross Section of U-235 at Beginning of Life ................. C-16 Figure C-4 Microscopic Fast Cross Section of U-235 at Beginning of Life ........................ C-16 Figure C-5 Microscopic Thermal Cross Section of U-235 Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions .................. C-17 Figure C-6 Microscopic Fast Cross Section of U-235 Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions .................. C-17 Figure C-7 Macroscopic Diffusion Coefficient (Fast and Thermal) Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions .......................................................................................................... C-18 Figure C-8 Microscopic Thermal Cross Section of U-235 at 70 GWd/MTU ...................... C-19 r

Figure C-9 Quadratic Interpolation Illustration of Microscopic Thermal Cross Section of U-235 .............................................................................................. C-20 Figure C-10 Illustration of Final Quadratic Interpolation for Microscopic Thermal Cross Section of U-235 ..................................................................... C-20 Figure C-11 Comparison of k-infinity from MICROBURN-B2 Interpolation Process with CASM0-4 Calculations at Intermediate Void Fractions of 0.2, 0.6 and 0.9 ............................................................................ C-21 Figure C-12 Comparison of k-infinity from MICROBURN-B2 Interpolation Process with CASM0-4 Calculations at 0.4 Historical Void Fractions and 0.9 Instantaneous Void Fraction ............................................... C-21 Figure C-13 Delta k-infinity from MICROBURN-B2 Interpolation Process with CASM0-4 Calculations at 0.4 Historical Void Fraction and 0.9 Instantaneous Void Fraction ............................................................................ C-22 Figure C-14 Comparison of Interpolation Process Using Void Fractions of 0.0, 0.4 and 0.8 and Void Fractions of 0.0, 0.45 and 0.9 ... :.................................... C-22 Figure C-15 EMF-2158(P)(A) TIP Statistics by Axial Level ................................................. C-23 Figure C-16 EMF-2158(P)(A) 2-D TIP Statistics for C-Lattice Plants vs. Core Power .............................................................................................................. C-23 Figure C-17 EMF-2158(P)(A) 2-D TIP Statistics for C-Lattice Plants vs. Core Average Void Fraction ..................................................................................... C-24 Figure C-18 EMF-2158(P)(A) 2-D TIP Statistics for C-Lattice Plants vs. Core ,

Power/Flow Ratio ............................................................................................ C-24 Figure C-19 EMF-2158(P)(A) 2-D TIP Statistics for D-Lattice Plants vs. Core Power .............................................................................................................. C-25

,Figure C-20 EMF-2158(P)(A) 2-D TIP Statistics for D-Lattice Plants vs. Core Average Void Fraction ..... ,............................................................................... C-25 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page vi Figure C-21 EMF-2158(P)(A) 2-D TIP Statistics for D-Lattice Plants vs. Core Power/Flow Ratio ........................................................................................... C-26 Figure C-22 EMF-2158(P)(A) 3-D Tl P Statistics for C-Lattice Plants vs. Core Power ............................................................................................................ C-26 Figure C-23 EMF-2158(P)(A) 3-D TIP Statistics for C-Lattice Plants vs. Core Average Void Fraction ................................................................................... C-27 Figure C-24 EMF-2158(P)(A) 3-D TIP Statistics for C-Lattice Plants vs. Core Power/Flow Ratio ........................................................................................... C-27 Figure C-25 EMF-2158(P)(A) 3-D TIP Statistics for D-Lattice Plants vs. Core Power ............................................................................................................ C-28 Figure C-26 EMF-2158(P)(A) 3-D TIP Statistics for D-Lattice Plants vs. Core Average Void Fraction ................................................................................... C-28 Figure C-27 EMF-2158(P)(A) 3-D TIP Statistics for D-Lattice Plants vs. Core Power/Flow Ratio ........................................................................................... C-29 Figure C-28 Quad Cities Unit 1 Pin by Pin Gamma Scan Results ...................................... C-29 Figure C-29 Maximum Assembly Power in Topical Report EMF-2158(P)(A) ..................... C-30 Figure C-30 Maximum Exit Void Fraction in Topical Report EMF-2158(P)(A) .................... C-30 Figure C-31 Maximum Assembly Power Observed from Recent Operating Experience ..................................................................................................... C-31 Figure C-32 Void Fractions Observed from Recent Operating Experience ........................ C-31 Figure C-33 Axial Power and Void Profile Observed from Recent Design Experience ...................................................................................................... C-32 Figure C-34 Nodal Void Fraction Histogram Observed from Recent Design Experience .... :...................................................................................... :......... C-32 Figure C-35 Maximum Assembly Power in an EPFOD Brunswick Design ......................... C-33 Figure C-36 Maximum Exit Void Fraction in an EPFOD Brunswick Design ........................ C-33 Figure C-37 Brunswick EPFOD Design Axial Profile of Power and Void Fraction .......................... ;............................................................................... C-34 Figure C-38 Brunswick EPFOD Design Nodal Void Fraction Histogram ............................ C-34 Figure C-39 MICROBURN-B2 Multi-Channel Average Bypass Void Distribution from a Brunswick Equilibrium Cycle Design ................................................... C-35 Figure C-40 MICROBURN-B2 Multi-Channel Exit Bypass Void Distribution from a Brunswick Equilibrium Cycle Design ................................................... C-36 Figure C-41 MICROBURN-B2 Multi-Channel Bypass Void at an LPRM Location from a Brunswick Equilibrium Cycle Design ..................................... C-37 Figure D-1 Comparison of Scram Bank Worth for [ ] ....... D-11 Figure E-1 [ ] .................. E-4 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page vii Nomenclature Acronym Definition BWR Boiling Water Reactor CHF Critical Heat Flux CPR Critical Power Ratio EPU Extended Power Uprate EPFOD Extended Power/Flow Operating Domain KATHY KArlstein Thermal HYdraulic test facility LHGR Linear Heat Generation Rate LOCA Loss of Coolant Accident LRNB Load Reject with no Bypass MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MELLLA+ Maximum Extended Load Line Limit Analysis Plus NRC Nuclear Regulatory Commission OLM CPR Operating Limit Minimum Critical Power Ratio SER Safety Evaluation Rep~rt SLMCPR Safety Limit Minimum Critical Power Ratio AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 1-1 1.0 Introduction This document reviews the AREVA* licensing methodologies to demonstrate that they are applicable to operation of the Brunswick Nuclear Power Generating Stations including Extended Power Uprate (EPU) conditions as well as the ~ELLLA+ Extended Power/Flow Operating Domain (EPFOD). The Brunswick stations are currently licensed with AREVA methods at EPU conditions (120% of the originally licensed rated thermal power). The MELLLA+ extended operating domain refers to operation at licensed EPU thermal power with reduced core flow as shown in Figure 1-1. This document addresses the generalized Extended Power Flow Operating Domain (EPFOD) of which MELLLA+ at Brunswick is only a particular example.

AREVA Inc. is an AREVA company.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 1-2 120.0 110.0 100.0 ,,

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Co re Flow Figure 1-1 Brunswick Power Flow Operating Map with the MELLLA+ EPFOD AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 2-1 2.0 Overview The first step in determining the applicability of current licensing methods to EPFOD conditions was a review of AREVA BWR topical reports listed in Table 2-1 to identify SER restrictions on the BWR methodology. This review identified that there are no SER restrictions on power level or flow for the AREVA topical reports. The review also indicated that there are no SER restrictions on the parameters most impacted by the increased power level at each core flow rate in the MELLLA+ domain: steam flow, feedwater flow, jet pump M-ratio, and core average void fraction.

The second step consisted of an evaluation of the core and reactor conditions experienced under EPFOD conditions to determine any challenges to the validity of the ,models. When the reactor power is increased and/or the core flow is decreased, the resultant ir:npact on operating margin is mitigated to a large extent by a decrease in the limiting assembly radial power factor.

This decrease in the limiting assembly radial power factor is necessary since the thermal operating limits (MCPR, MAPLHGR and LHGR) that restrict assembly power are dependent on the limiting assembly power but are fairly insensiti~e to the core thermal power. From this fundamental constraint the following observations may be made about the EPFOD operating conditions:

1. The reduction in the hot assembly radial peaking factor leads to a more uniform radial power distribution and consequently a more uniform core flow distribution. The net result being less flow starvation of the hottest assemblies.

2. With the flatter radial power distribution, more assemblies and fuel rods are near thermal limits.

3. From a system perspective, there will be higher steam flow and feedwater flow rates for a given core flow at core flows previously constrained by the MELLLA operating boundary.

4. With an increase in the avera~e assembly power in the reactor for a given core flow at core flows previously constrained by the MELLLA operating boundary, the core pressure

drop will increase slightly resulting in a decrease in the jet pump M~ratio for a given core flow rate.

5. Core average void fraction will increase.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 2-2 Based on these fundamental characteristics of power uprate and MELLLA+ operation at reduced flow, each of the major analysis domains (thermal-hydraulics, core neutronics, transient analysis, LOCA and stability) are assessed to determine any challenges to EPFOD application.

A description of the analyses performed for a transition cycle is provided in Appendix A.

Table 2-1 A~EVA Licensing :Topical Reports Document Number Document Title XN-NF-79-56(P)(A) "Gadolinia Fuel-Properties for LWR Fuel Safety Evaluation," Exxon Revision 1 and Nuclear Company, November 1981 Supplement 1 XN-75-32(P)(A) "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Supplements 1 through 4 Nuclear Company, October 1983. (Base document not approved.)

XN-N F-81-58(P)(A) "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Revision 2 and Model," Exxon Nuclear Company, March 1984 Supplements 1 and 2 XN-N F-81-51 (P)(A) "LOCA-Seismic Structural Response of an Exxon Nuclear Company BWR Jet Pump Fuel Assembly," Exxon Nuclear Company, May 1986 XN-NF-85-67(P)(A) "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Revision 1 Reload Fuel," Exxon Nuclear Company, September 1986 XN-N F-85-74(P)(A) "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Revision 0 Model" Exxon Nuclear Company, August 1986 XN-N F-85-92(P)(A) "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," Exxon Nuclear Company, November 1986 ANF-89-98(P)(A) Revision 1 "Generic Mechanical Design Criteria for BWR Fuel Designs,"

and Supplement 1 Advanced Nuclear Fuels Corporation, May 1995 ANF-90-82(P)(A) Revision 1 "Application of ANF Design Methodology for Fuel Assembly Reconstitution," Advanced Nuclear Fuels Corporation, May 1995 EMF-85-74(P) Revision 0 "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Supplement 1(P)(A) and Model," Siemens Power Corporation, February 1998 Supplement 2(P)(A)

EMF-93-177(P)(A) "Mechanical Design for BWR Fuel Channels," Framatome ANP, Revision 1 August2005 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 2-3 Table 2-1 AREVA Licensing Topical Reports (Continued)

Document Number Document Title BAW-10247PA Revision 0 '

"Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA NP, February 2008 XN-NF-80-19(P)(A) Volume 1 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic and Supplements 1 and 2 Methods for Design and Analysis," Exxon Nuclear Company, March 1983 XN-NF-80-19(P)(A) Volume 4 "Exxon Nuclear Methodology for Boiling Water Reactors:

Revision 1 Application of the ENC Methodology to BWR Reloads," Exxon

' Nuclear Company, June 1986 EMF-CC-074(P)(A) Volume 1 "STAIF -A Computer Program for BWR Stability Analysis in the Frequ.ency Domain," and Volume 2 "STAIF -A Computer Program for BWR Stability Analysis in the Frequency Domain - Code Qualification Report," Siemens Power Corporation, July 1994 EMF-2158(P)(A) Revision 0 "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASM0-4/

MICROBURN-B2," Siemens Power Corporation, October 1999 EMF-CC-074(P)(A) "BWR Stability Analysis Assessment of STAIF with Input from Volume 4, Revision 0 MICROBURN-B2," Siemens Power Corporation, August ?000 '

BAW-10255PA Revision 2 "Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code," AREVA NP, May 2008 ANP-10262PA, Revision 0 "Enhanced Option Ill Long Term Stability Solution," May 2008 EMF-3028P-A Volume 2 "RAMONA5-FA: A Computer Program for BWR Transient Analysis Revision 4 in the Time Domain Volume 2: Theory Manual," AREVA NP, May, 2013 XN-N F-79-59(P)(A) "Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies," Exxon Nuclear Company, November 1983 XN-NF-80-19(P)(A} Volume 3 "Exxon Nuclear Methodology for Boiling Water Reactors, Revision 2 THERMEX: Thermal Limits Methodology Summary Description,"

Exxon Nuclear Company, January 1987 EMF-2245(P)(A) Revision 0 "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," Siemens Power Corporation, August2000 EM F-2209(P)(A) Revision 3 "SPCB Critical Power Correlation," AREVA NP, September 2009 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended. Revision 1 Power Flow Operating Domain Page 2-4 Table 2-1 AREVA Licensing Topical Reports (Continued)

Document Number Document Title ANP-10298PA Revision 0 "ACE/ATRIUM 10XM Critical Power Correlation," AREVA NP, March 2010 ANP-10298PA Revision 0 "Improved K-Factor Model for ACE/ATRIUM 10XM Critical Power Supplement 1P Revision 0 Correlation;" AREVA NP, December 2011 ANP-10307PA Revision 0 "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," AREVA NP, June 2011 XN-CC-33(A) Revision 1 "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option Users Manual," Exxon Nuclear Company, November 1975 XN-NF-80-19(P)(A) "Exxon Nuclear Methodology for Boiling Water Reactors: EXEM Volumes 2, 2A, 28 and 2C BWR ECCS Evaluation Model," Exxon Nuclear Company, September 1982 XN-NF-82-07(P)(A) "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Revision 1 Model," Exxon Nuclear Company, November 1982 XN-NF-84-105{P)(A) "XCOBRA-T: A Computer Code for BWR Transient Thermal-Volume 1 and Volume 1 Hydraulic Core Analysis," Exxon Nuclear Company, February 1987 Supplements 1 and 2 ANF-913(P)(A) Volume 1 "COTRANSA2: A Computer Program for Boiling Water Reactor Revision 1 and Volume 1 Transient Analyses," Advanced Nuclear Fuels Corporation, August Supplements 2, 3 and 4 1990 ANF-91-0i48(P){A) "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model," Advanced Nuclear Fuels Corporation, January 1993 ANF-91-048(P)(A) "BWR Jet Pump Model Revision for RELAX," Siemens Power Supplements 1 and 2 Corporation, October 1997 EMF-2292(P)(A) Revision 0 "ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients,"

Siemens Power Corporation, September 20_QO l EMF-2361 (P)(A) Revision 0 "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP, May 2001 ANF-1358(P)(A) Revision 3 "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Framatome ANP, September 2005

./

AREVA Inc. J

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 3-1 3.0 Thermal-hydraulics AREVA assembly thermal-hydraulic methods are qualified and validated against full-scale heated bundle tests in the KATHY test facility in Karlstein, Germany. The KATHY tests are used to characterize the assembly two-phase pressure drop and CHF performance. This allows the hydraulic models to be verified for AREVA fuel designs over a wide range of hydraulic conditions prototypic of reactor conditions.

The standard matrix of test conditions for KATHY is compared to reacfor conditions in I

Figure 3-1. This figure illustrates that the test conditions bound typical EPFOD assembly conditions. The data is based upon the projected EPFOD conditions for the Brunswick reactor.

Figure 3-1 also shows that the key physical phenomena (e.g. fluid quality and assembly flows) for EPFOD conditions are equivalent to current reactor experience.

This similarity of assembly conditions is further enforced in AREVA analysis methodologies by the imposition of SPCB and ACE correlation limits and, therefore, both current core designs and uprated core designs must remai~ within the same parameter space. Since the bundle operating conditions for EPFOD are within the envelope of hydraulic test data used for model qualification and operating experience, the hydraulic models used to compute the core flow distribution and local void content remain valid for EPFOD conditions.

A more detailed discussion of the AREVA void quality correl~tions is presented in Appendix B.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 3-2 Figure 3-1 Comparison of Karlstein Two-Phase Pressure Drop and void fraction Test Matrices and Typical EPFOD Reactor Conditions AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 4-1 4.0 AREVA CHF/CPR Correlations All AREVA CHF and CPR correlations are approved by the NRC staff to be applicable over specified ranges of assembly operating conditions. The NRC staff also approved specific corrective actions when the computed conditions fall outside of the approved range to assure that conservative calculations are obtained. For both EPFOD and pre-EPFOD conditions, some analyses can predict assembly conditions to be outside the approved range of specified conditions for the CHF correlations. Consequently, the AREVA licensing methods are programmed to determine whether the computed assembly conditions fall outside of the approved range of applicability for the CHF correlation and impose approved corrective actions as appropriate to conservatively assess the critical power margin for the assembly.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 5-1 5.0 Safety Limit MCPR The safety limit MCPR (SLMCPR) methodology is used to determine the Technical Specification SLM CPR value that ensures that 99.9% of the fuel rods are expected to avoid boiling transition during normal reactor operation and anticipated operation occurrences. The SLMCPR methodology for Brunswick EPFOD is described in Reference 1. The SLMCPR is determined by statistically combining calculation uncertainties and plant measurement uncertainties that are associated with the calculation of MCPR. The thermal hydraulic, neutronic, and critical power correlation methodologies are used in the calculation of MCPR. The applicability of these methodologies for EPFOD conditions is discussed in other sections of this report. As discussed in Section 2.0, EPFOD operation will result in a flatter radial power distribution with more fuel rods operating near thermal limits. This is the most significant impact of EPFOD operation on the SLMCPR calculation and is explicitly accounted for in the methodology as discussed below.

AREVA calculates the SLMCPR on a cycle-specific basis to protect all allowed reactor operating conditions. The analysis incorporates the cycle-specific fuel and core designs. The initial MCPR distribution of the core is a major factor affecting how many rods are predicted to be in boiling transition. The MCPR distribution of the core depends on the neutronic design of the reload fuel and the fuel assembly power distributions in the core. AREVA SLMCPR methodology specifies that analyses be performed with a design basis power distribution that

" ... conservatively represents expected reactor operating states which could both exist at the MCPR operating limit and produce a MCPR equal to the MCPR safety limit during an anticipated operational occurrence." (Reference 1, Section 3.3.2).

[

]

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 5-2

[

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The impact that a flatter core power distribution may have on the SLMCPR is explicitly accounted for by the methodology. EPFOD operation will lead to a flatter core power distribution; [

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When the SLMCPR methodology (Reference 1) was applied for EPU at Brunswick a licensing condition was applied to address a concern with the application of the fuel channel bow standard deviation when the fluence gradient is computed to exceed the bound of the channel measurement database. As a consequence, the fuel channel bow standard deviation component of the channel bow model uncertainty used by ANP-10307PA to determine the Safety Limit Minimum Critical Power Ratio was increased by the ratio of channel fluence gradient to the channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty was determined. This same approach will be used to support EPFOD operations.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 5-3

~

Figure 5-1 Assembly Power Distribution for Limiting Case in Safety Limit MCPR Analysis (Pre-E~U and EPU)

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 5-4 Figure 5-2 Assembly Power Distribution for Limiting Case in Safety Limit MCPR Analysis (EPU and MELLLA+)

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 6-1 6.0 Mechanical Limits Methodology The LHGR limit is established to support plant operation while satisfying the fuel mechanical design criteria. The methodology for performing the fuel rod evaluation is described in Reference 3. Fuel rod design criteria evaluated by the methodology are contained in References 3 and 4. The evaluation addresses operation at EPFOD because the power history inputs are directly obtained from projected operation and core follow data representative of the environmental conditions for the fuel. This operation remains within the range of applicability, in terms of LHGR and burnup, of the methodology while meeting the design limits for the fuel.

Fuel rod power histories are generated as part of the methodology for equilibrium cycle conditions as well as cycle-specific operation. A comprehensive number of uncertainties are taken into account in the categories of operating power uncertainties, code model parameter uncertainties, and fuel manufacturing tolerances. In addition, adjustments are made to the power history inputs for possible differences in planned versus actual operation. Upper limits on

~

the analysis results are obtained for comparison to the design limits for fuel melt, cladding strain, rod internal pressure and other topics as described by the design criteria.

Since the power history inputs, which include LHGR, fast neutron flux, reactor coolant pressure and reactor coolant temperature, are used as input to the analysis, the results explicitly account for conditions at EPFOD such as higher coolant voiding and offsets in axial power and neutron fast flux. The resulting LHGR limit is used to monitor the fuel so it is maintained within the same maximum allowable steady-state power envelope as analyzed.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 7-1 7.0 Core Neutronics The AREVA neutronic methodologies are characterized by technically rigorous treatment of phenomena and are very well benchmarked (>100 cycles of operation plus gamma scan data for ATRIUM'-10*). Recent operating experience is tabulated in Table 7-1. These tables present the reactor operating conditions and in particular the average and hot assembly powers for both US and European applications. As can be seen from this information, the average and peak bundle powers in this experience base exceed that associated with the Brunswick EPFOD application.

The increased steam flow from increased core power, whether at rated core flows or reduced core flows in the EPFOD, comes from increased power in normally lower power assemblies in the core, operating at higher power levels. For EPFOD operation the high powered assemblies in uprated cores will be subject to the same LHGR, MAPLHGR, MCPR, and cold shutdown margin limits and restrictions as high powered assemblies in pre-EPFOD cores.

The similarity of operating conditions between current and EPFOD conditions assures that the neutronic methods used to compute the nodal reactivity and power distributions remain valid.

Furthermore, the neutronic characteristics computed by the steady-state simulator.and used in saf~ty analysis remain valid.

Detailed analysis of the neutronic methodology and it is specific applicability to EPFOD is presented in Appendix C.

ATRIUM is a trademark of AREVA Inc.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 7-2 Table 7-1 CASM0-4/MICROBURN-82 Operating Experience Ave. Approximate Bundle Peak Bundle Reactor Power, MWt Power, Power, Fuel/Cycle Uprate Reactor Size, #FA (% Uprated)* MWt/FA MWt/FA Licensing** Comments A 592 2575 (0.0) 4.4 7.2 us B 592 2575 (0.0) 4.4 7.4 us c 532 229? (0.0) 4.3 7.3 Europe Licensing only D 840 3690 (0.0) 4.4 7.5 us through Cycle 20 For3 E 500 2500 (15.7) 5.0 8.0 us operating cycles.

F 444 1800 (5.9) 4.1 7.3 us G 676 _2928 (8.0) 4.3 7.6 Europe H 700 3300 (9.3) 4.7 8.0 Europe I 784 3840 (0.0) 4.9 8.1 Europe J 624 3237 (11.9) 5.2 7.8 Europe With K 648 3600 (14.7) 5.6 8.6 Europe ATRIUM 10XM L 648 2500 (10.1) 3.9 6.9 Europe M 624 3091 (6.7) 5.0 7.7 us N 800 3898 (1.7) 4.9 7.7 us 0 764 3489 (5.0) 4.6 7.2 us P"' 560 2923 (20.0) 5.2 8.0 us Q 764 3956 (20.0) 5.2 7.7 us

Latest power uprates.

- Location of fuel licensing.

+ Brunswick AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 8-1 8.0 Transient Analysis The core phenomena of primary interest for limiting transients in BWRs are void fraction/quality relationships, determination of CHF, pressure drop, reactivity feedbacks and heat transfer correlations. One fundamental validation of the core hydraulic solution is separate effects testing against Karlstein transient CHF measurements. The transient benchmark to time of dryout for prototypic Load Reject with no Bypass (LRNB) and pump trip transients encompass the transient integration of the continuity equations (including the void-quality closure relation),

heat transfer, and determination of CHF. Typical benchmarks to Karlstein (Figure 8-1) illustrate that the transient hydraulic solution and application of ACE (AREVA Critic?ll Power Correlation) result in conservative predictions of the time of dryout. The measured data is taken from ATRIUM 10XM tests.

Outside of the core, the system simulation relies primarily on solutions of the basic conservation equations and equations of state. While there are changes to the feedwater flow rate and jet pump M-ratio associated for conditions previously constrained by the MELLLA operating domain, the most significant change is the steam flow rate and the associated impact on the steamline dynamics for pressurization events. The models associated with predicting the pressure wave are general and have no limitation within the range of variation associated with EPFOD.

The reactivity feedbacks are validated by a variety of means including initial qualification of advanced fuel design lattice calculations to Monte Carlo results as required by SER restrictions, steady-state monitoring of reactor operation (power distributions and eigenvalue), and the Peach Bottom 2 turbine trip benchmarks that exhibited a minimum of 2% conservatism in the calculation of integral power.

From these qualifications and the observation that the nodal hydraulic conditions during EPFOD are expected to be within the current operating experience, the transient analysis methods remain valid.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 8-2 Appendix D provides additional information on the transient cross section treatment in the COTRANSA2 transient simulator for both EPFOD and pre-EPFOD reactor conditions.

Appendix F provides a summary of the impact of thermal conductivity degradation on transient analysis and corrective actions taken in the Brunswick EPFOD analyses.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 8-3 Figure 8-1 Typical Hydraulic Benchmarks to Karlstein Transient Simulations

/

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 9-1 9.0 LOCA Analysis LOCA results are strongly dependent on local power and are weakly dependent on core

(

average power. As discussed in previous sections, maximum local power is not significantly changed due to the expanded operating domain because the core is still constrained by the same thermal limits. The parameters associated with the expanded domain that may impact LOCA results at each of the core flow rates in the MELLLA+ domain are: increased core average initial stored energy, decreased initial coolant inventory, relative flow distribution between highest power and average power assemblies, and increased core decay heat.

BWR LOCA analyses are not sensitive to initial stored energy. During the blowdown phase the heat transfer remains high and the stored energy is removed prior to the start of the heatup phase. Initial inventory differences may impact LOCA event timing and the minimum inventory during blowdown prior to refill of the reactor vessel. However, any impact on event timing or minimum inventory would be smaller than the impact associated with the different size breaks that are already considered in the break spectrum analyses. At the elevated powers associated with MELLLA+ EPFOD conditions, the difference in flow between the highest power assembly and the average power assembly is reduced. Therefore, these parameters do not change the range of conditions encountered or the capability of the codes to model LOCA at EPFOD conditions.

The potential impact of the EPFOD on LOCA analyses is thus primarily associated with the increase in decay heat levels in the core for reduced core flow conditions. For the EXEM BWR-2000 LOCA methodology the decay heat is conservatively modeled. The 11 decay equation curve fit to the 1971 draft ANS standard for fission product decay heat from the WREM model is used to calculate fission product decay heat during blowdown. The draft ANS standard values are used for spray cooling and reflood. The required multiplier o! 1.2 is applied to the fission product decay heat throughout the LOCA scenario. The models used for decay heat calculations are valid for EPFOD.

From the above discussion and the observation that nodal thermal-hydraulic conditions during EPFOD are expected to be within the current operating experience, the LOCA methods remain applicable for EPFOD conditions.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 9-2 Independent of EPFOD, additional modifications have been made to the approved EXEM BWR-2000 LOCA methodology to more accurately model advanced fuel designs and to address regulatory concerns with the approved methodology. These modifications are described in Appendix E. Appendix F summarizes the assessment of thermal conductivity degradation in the Brunswick LOCA analyses.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 10-1 10.0 Stability Analysis 10.1 Linear Stability The flatter radial power profile characteristic of MELLLA+ core designs will tend to decrease the first azimuthal eigenvalue separation and result in slightly higher regional decay ratios. These effects are computed by STAIF as it directly computes the channel, global, and regional decay ratio arid does not rely on a correlation to protect the regional mode.

STAIF has been benchmarked against full assembly tests (in KATHY facility) to validate the channel hydraulics from a decay ratio of approximately 0.4 to limit cycles. These tests or ,

benchmarks exceed the bounds of allowed operation. These benchmarks include prototypical ATRIUM-10 assemblies. From a reactor perspective, STAIF is benchmarked to both global and regional reactor data as late as 1998, and, therefore, includes current reactor cycle and fuel design elements. This strong benchmarking qualification and the direct computation of the regional mode assure that the impact of the MELLLA+ core designs are reflected in the stability analysis.

10.2 DIVOM RAMONAS-FA has been generically approved for MELLLA+ operation in support of the Enhanced Option Ill Long Term Stability Solution (Reference 5 and 6).

10.3 ATWS-1 A Brunswick specific assessment of the ATRIUM 1OXM fuel for the EPFOD using the NRC approved NSSS vendor methodology is outside the scope of this document and has been submitted separately.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 11-1 11.0 Summary This review concluded that there are no SE.R restrictions on AREVA methodology that are impacted by EPFOD. Since the EPFOD core and assembly conditions for the Brunswick units are equivalent to core and assembly conditions ,.of other plants for which the methodology was penchmarked, the AREVA methodology (including uncertainties) remains applicable for EPFOD conditions at the Brunswick Units.

More spedfically:

a) The steady state and transient neutronics and thermal-hydraulic analytical methods and code systems supporting EPFOD are within NRC approved applicability ranges because the conditions for EPFOD application are equivalent to existing core and assembly conditions in other plants for which the AREVA methodology was benchmarked.

b) The calculational and measurement uncertainties applied in EPFOD applications are, valid because the conditions for EPFOD application are equivalent to existing core and assembly conditions for which the AREVA methodology was b~nchmarked.

c) The assessment database and uncertainty of models used to simulate the plant response at EPFOD conditions are equivalent to core and assern,bly conditions for which ,

the AREVA methodology was benchmarked.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 12-1 12.0 References

1. ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.

2. ANP-10298PA Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, March 2010.

3. BAW-1024 ?PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.

4. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.

5. EMF-3028P-A Volume 2 Revision 4, RAMONA5-FA: A Computer Program for BWR Transient Analysis in the Time Domain Volume 2: Theory Manual, AREVA NP, March, 2013.

6. ANP-10262PA, Revision 0, Enhanced Option Ill Long Term Stability Solution, AREVA NP, May 2008.

7. ANP-10298PA Revision 0 Supplement 1P Revision 0, Improved K-Factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, December 2011.

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. ANF-913 (P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.

10. N. Zuber and J. A. Findlay, "Average Volumetric Concentration in Two-Phase Flow Systems," J. Heat Transfer, 1965.

11. P. Coddington and R. Macian, "A Study of the Performance of Void Fraction Correlations Used In the Context of Drift-Flux Two-Phase Flow Models," Nuclear Engineering and Design, 215, 199-216, June 2002.

12. [

1

13. K. Ohkawa and R. T. Lahey, Jr., "The Analysis of CCFL Using Drift-Flux Models,"

Nuclear Engineering and Design, 61, 1980.

14. S. Misu et al., "The Comprehensive Methodology for Challenging BWR Fuel Assembly and Core Design used at FANP," proceedings on CD-ROM, PHYSOR 2002, Seoul, Korea, October 7-10, 2002 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 12-2

15. XN-NF-84-105(P)(A) Volume 1 Supplement 4, XCOBRA-T: A Computer Code For BWR Transient Thermal-Hydraulic Core Analysis, Void Fraction Model Comparison to Experimental Data, Advanced Nuclear Fuels Corporation, June 1988.

16. 0. Nylund et al., "Hydrodynamic and Heat Transfer Measurements on A Full-Scale Simulated 36-Rod Marviken Fuel Element with Non-Uniform Radial Heat Flux Distribution," FRIGG-3, R-494/RL-1154, November 1969.

17. J. Skaug et al., "FT-36b, Results of void Measurements," FRIGG-PM-15, May 1968.

18. 0. Nylund et al., "Hydrodynamic and Heat Transfer Measurements on A Full-Scale Simulated 36-Rod Marviken Fuel Element with Uniform Heat Flux Distribution,"

FRIGG-2, R-447/RTL-1007, May 1968.

19. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASM0-4/MICROBURN-82, Siemens Power Corporation, October 1999.

20. XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.

~1. NED0-32047-A, "ATWS Rule Issues Relative to BWR Core Thermal-Hydraulic Stability,"

GE Nuclear Energy, June 1995.

22. W. Wolf et al., "BWR Stability Analysis with the BNL Engineering Plant Analyzer,"

NUREG/CR-5816, October 1992.

23. EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 EGGS Evaluation Model, Framatome ANP, May 2001.

24. XN-CC-33(A) Revision 1, HUXY: .f\ Geralized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option Users Manual, Exxon Nuclear Company, November 1975.

25. Letter, P. Salas (AREVA) to Document Control Desk, U.S. Nuclear Regulatory Commission, "Proprietary Viewgraphs and Meeting Summary for Closed Meeting on Application of the EXEM BWR-2000 ECCS Evaluation Methodology," NRC: 11 :096, September 22, 2011.

26. Letter, T.J. McGinty (NRC) to P. Salas (AREVA)," Response to AREVA NP, Inc.

(AREVA) Proposed Analysis Approach for Its EXEM Boiling Water Reactor (BWR)-2000 Emergency Core Cooling System (ECCS) Evaluation Model," July 5, 2012.

27. Letter, R. Gardner to NRC, "Informational Transmittal Regarding Requested White Paper on the Treatment of Exposure Dependent Fuel Thermal Conductivity Degradation in RODEX Fuel Performance Codes and Methods," NRC:09:069, ML092010160, July 14, 2009.

28. Letter, T. J. McGinty (NRC) to P. Salas (AREVA), "Nuclear Fuel Thermal Conductivity Degradation Evaluation for Light Water Reactors Using AREVA Codes and Methods (TAC No. ME5178)," March 23, 2012.

AREVA Inc.

'Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 12-3

29. Letter, P. Salas (AREVA) to NRC, "Response to NRC Letter Regarding Nuclear Fuel Thermal Conductivity Degradation Evaluation for Light Water Reactors Using AREVA Codes and Methods," NRC:12:023, April 27, 2012.

30. ANP-10300P, Revision 0, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Transient and Accident Scenarios, (ML100040158) December 2009.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page A-1 Appendix A Application of AREVA Methodology for Mixed Cores A.1 Discussion AREVA has considerable experience analyzing fuel design transition cycles and has methodology and procedures to analyze mixed cores composed of multiple fuel types. For each core design, analyses are performed to confirm that all design and licensing criteria are satisfied. The analyses performed explicitly include each fuel type in the core. The analyses consider the cycle-specific core loading and use input data appropriate for each fuel type in the core. The mixed core analyses are performed using generically approved methodology in a manner consistent with NRC approval of the methodology. Based on results from the analyses, operating limits are established for each fuel type present in the core. During operation, each fuel type _is monitored against the appropriate operating limits.

Thermal hydraulic characteristics are determined for each fuel type that will be present in the core. The thermal hydraulic characteristics used in core design, safety analysis, and core monitoring are developed on a consistent basis for both AREVA fuel and other vendor co-resident fuel to minimize variability due to methods. For Brunswick EPFOD operation, the entire core will be composed of AREVA fuel designs.

For core design and nuclear safety analyses, the neutronic cross-section data is developed for each fuel type in the core using CASM0-4. MICROBURN-B2 is used to design the core and provide input to safety analyses (core neutronic characteristics, power distributions, etc.).* Each fuel assembly is explicitly modeled in MICROBURN-B2 using cross-section data from CASM0-4 and geometric data appropriate for the fuel design.

Fuel assembly thermal mechanical limits for all fuel are verified and monitored for each mixed core designed by AREVA. AREVA performs design and licensing analyses to demonstrate that the core design meets steady-state limits and that transient limits are not exceeded during anticipated operational occurrences.

The critical power ratio (CPR) is evaluated for each fuel type in the core using calculated local fluid conditions and an appropriate critical power correlation. Fuel type specific correlation coefficients for AREVA fuel are based on data from the AREVA critical power test facility. The AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain PageA-2 SPCB critical power correlation will be used for monitoring ATRIUM-10 fuel present during the initial cycle of EPFOD operation at Brunswick. T~e critical power ratio (CPR) correlation used for the ATRIUM 1OXM fuel is the ACE/ATRIUM 1OXM critical power correlation described in Reference 2. The ACE CPR correlation uses K-factor values to account for rod local peaking, rod location and bundle geometry effects. The K-factor methodology was modified in Reference 7 in response to deficiencies found in the axial averaging process. In addition, the additive constants were revised as a result of the change to the K-factor model.

At the time of the creation of this document, Reference 7 had not yet been generically approved. Therefore, the Brunswick EPFOD analyses are performed with the approved ACE/ATRIUM 10XM correlation (Reference 2) and in accordance with the Brunswick specific licensing condition. The resultant limits are confirmed to be conservative based on the approach discussed in the licensing cqndition in the Brunswick Technical Specifications Appendix B.

In the safety limit MCPR analysis each fuel type present in the core is explicitly modeled using appropriate geometric data, thermal hydraulic characteristics, and power distribution information (from CASM0-4 and MICROBURN-B2 analyses). CPR is evaluated for each assembly using fuel type specific correlation coefficients. Plant and fuel type specific uncertainties are considered in the statistical analysis performed to determine the safety limit MCPR. The safety limit MCPR analysis is performed each cycle and uses the cycle specific core configuration.

An operating limit MCPR is established for each fuel type in the core. For fast transients the COTRANSA2 code (Reference 9) is used to determine the overall system response. The core nuclear characteristics used in COTRANSA2 are obtained from MICROBURN-B2 and reflect the actual core loading pattern. Boundary conditions from COTRANSA2 are used with an XCOBRA-T core model. In the XCOBRA-T model, a hot channel with appropriate geometric and thermal hydraulic characteristics is modeled for each fuel type present in the core. Critical power performance is ~valuated using local fluid conditions and fuel type specific CPR correlation coefficients. The transient CPR response is used to establish an operating limit MCPR for each fuel type.

For transient events that are sufficiently slow such that the heat transfer remains in phase with changes in neutron flux during the transient, evaluations are performed with steady state codes AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page A-3 such as MICROBURN-82 in accordance with NRC approval. Such slow transients are modeled by performing a series of steady state solutions with appropriate boundary conditions using the cycle specific design core loading plan. Each fuel assembly type in the core is explicitly I

modeled. The change in CPR between the initial and final condition after the transient is determined, and if the CPR change is more severe than those determined from fast transient analyses, the slow transient result is used to determine the MCPR operating limit.

Stability analyses to establish OPRM setpoints and backup stability exclusion regions are performed using the cycle-specific core loading pattern. The stability analyses performed with RAMONAS-FA and STAIF explicitly model each fuel type in the core. Each fuel type is modeled using appropriate geometric, thermal hydraulic and nuclear characteristics determined as described above. The stability OPRM setpoints and exclusion region boundaries are established based on the predicted performance of the actual core composition.

MAPLHGR operating limits are established and monitored for each fuel type in the core to ensure that 10 CFR 50.46 acceptance criteria are met during a postulated LOCA. The RELAX code is used to determine the overall system response during a postulated LOCA and provides boundary conditions for a RELAX hot channel model. While system analyses are typically performed on an equilibrium core basis, the thermal hydraulic characteristics of all fuel assemblies in the core are considered to ensure the LOCA analysis results are applicable to mixed core configurations. Results from the hot channel analysis provide boundary conditions to the HUXY computer code. The HUXY model includes fuel type specific input such as dimensions and local power peaking for each fuel rod.

The core monitoring system will monitor each fuel assembly in the ~ore. Each assembly is modeled with geometric, thermal hydraulic, neutronic, and CPR correlation input data appropriate for the specific fuel type. Each assembly in the core will be monitored relative to thermal limits that have been explicitly developed for each fuel type.

In summary, AREVA methodology is used consistent with NRC approval to perform design and licensing analyses for mixed cores. The cycle design and licensing analyses explicitly consider each fuel type in mixed core configurations. Limits are established for each fuel type and operation within these limits is verified by the monitoring system during operation.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 8-1 Appendix B Void-Quality Correlations 8.1 AREVA Void Quality Correlations The Zuber-Findlay drift flux model (Reference 10) is utilized in the AREVA nuclear and safety analysis methods for predicting vapor void fraction in the BWR system. The model has a generalized form that may be applied to two phase flow by defining an appropriate correlation for the void concentration parameter, Co, and the drift flux, Vgj. The model parameters account for the radially non-uniform distribution of velocity and density and the local relative velocity between the phases, respectively. This model has received broad acceptance in the nuclear industry and has been successfully applied to a host of different applications, geometries, and fluid conditions through the application of different parameter correlations (Reference 11).

Two different correlations are utilized at AREVA to describe the drift flux parameters for the analysis of a BWR core. The correlations and treatment of uncertainties are as follows:

The nuclear design, frequency domain stability, nuclear AOO transient and accident analysis methods use the [ ] void correlation (Reference 12) to predict nuclear parameters. Uncertainties are addressed at the overall methodology and application level rather than individually for the individual correlations of each method.

The overall uncertainties are determined statistically by comparing predictions using the methods against measured operating data for the reactors operating throughout the world.

The thermal-hydraulic design, system AOO transient and accident analysis, and loss of coolant accident (only at specified junctions) methods use the Ohkawa-Lahey void correlation (Reference 13). This correlation is not used in the direct computation of nuclear parameters in any of the methods. Uncertainties are addressed at the overall methodology level through the use of conservative assumptions and biases to assure uncertainties are bounded.

The [ ] void correlation was developed for application to multi-rod geometries operating at typical BWR operating conditions using multi-rod data and was also validated AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page B-2 against simple geometry data available in the public domain. The correlation was defined to be functionally dependent on the mass flux, hydraulic diameter, quality, and fluid properties.

The multi-rod database used in the [

]. As a result, the multi-rod database and prediction uncertainties are not available to AREVA. However, the correlation has been independently validated by AREVA against public domain multi-rod data and proprietary data collected for prototypical ATRIUM-10 and ATRIUM 10XM test assemblies. Selected results for the ATRIUM-10 test assembly are reported in the public domain in Reference 14.

The Ohkawa-Lahey void correlation was developed for application in BWR transient calculations. In particular, the correlation was carefully designed to predict the onset of counter current flow limit (CCFL) characteristics during the occurrence of a sudden inlet flow blockage.

The correlation was defined to be functionally dependent on the mass flux, quality, and fluid properties.

Independent validation of the Ohkawa-Lahey correlation was performed by AREVA at the request of the NRC during the NRC review of the XCOBRA-T code. The NRC staff subsequently reviewed and approved Reference 15, which compared the code to a selected test from the FRIGG experiments (Reference 16). More recently the correlation has been independently validated by AREVA against additional public domain multi-rod data and proprietary data collected for prototypicATRIUM-10 and ATRIUM 10XM test assemblies, as described below.

The characteristics of the AREVA multi-rod void fraction validation database are listed in Table B-1.

The FRIGG experiments have been included in the validating database because of the broad industry use of these experiments in benchmarking activities, including TRAC, RETRAN, and S-RELAP5. The experiments include a wide range of pressure, subcooling, and quality from which to validate the general applicability of a void correlation. However, the experiments do not contain features found in modern rod bundles such as part length fuel rods and mixing vane grids. The lack of such features makes the data less useful in validating correlations for modern AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 8-3 fuel designs. Also the reported instrument uncertainty for these tests is provided in Table B-1 based on mockup*testing. However, the total uncertainty of the measurements (including power and flow uncertainties) is larger than the indicated values.

Because of its prototypical geometry, the ATRIUM-10 and ATRIUM 10XM void data collected at KATHY was useful in validating void correlation performance in modern rod bundles that include part length fuel rods, mixing vane grids, and prototypic axial/radial power distributions. Void measurements were made at one of three different elevations in the bundle for each test point:

/

just before the end of the part length fuel rods, midway between the last two spacers, and just before the last spacer.

As shown in Figure 3-1, the range of conditions for the ATRIUM void data are valid for typical reactor conditions. This figure compares the equilibrium quality at the plane of measurement for the ATRIUM 10XM void data with the exit quality of bundles in the EMF-2158 benchmarks and Brunswick operating at EPFOD conditions. As seen in the figure, the data at the measurement plane covers nearly the entire range of reactor conditions. However, calculations of the exit quality from the void tests show the overall test conditions actually envelope the reactor conditions.

Figure B-1 and Figure B-2 provide comparisons of predicted versus measured void fractions for the AREVA multi-rod void fraction validation database using the [ ] correlation.

These figures show the predictions fall within +/-0.05 (predicted - measured) error bands with good reliability and with very little bias. Also, there is no observable trend of uncertainty as a function of void fraction.

Figure B-3 and Figure B-4 provide comparisons of predicted versus measured void fractions for the AREVA multi-rod void fraction validation database using the Ohkawa-Lahey correlation.

In general, the correlation predicts the void data with a scatter of about +/-0.05 (predicted -

measured), but a bias in the prediction is evident for voids between 0.5 and 0.8. The observed under prediction is consistent with the observa,tions made in Reference 17.

In conclusion, validation using the AREVA multi-rod void fraction validation database has shown that both drift flux correlations remain valid for modern fuel designs. Furthermore, there is no observable trend of uncertainty as a function of void fraction. This shows there is no increased AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page B-4 uncertainty in the prediction of nuclear parameters at EPFOD conditions within the nuclear methods when applied to the Brunswick reactor.

8.2 Void Quality Correlation Uncertainties The AREVA analyses methods and the correlations used by the methods are applicable for modern fuel designs in both pre-EPFOD and EPFOD conditions. The approach for addressing the void-quality correlation bias and uncertainties remains unchanged and is applicable for Brunswick operation at EPFOD conditions.

The OLMCPR is determined based on the safety limit MCPR (SLMCPR) methodology and the transient analysis (.LiCPR) methodology. Void-quality correlation uncertainty is not a direct input to either of these methodologies; however, the impact of void-correlation uncertainty is inherently incorporated in both methodologies as discussed below.

The SLMCPR methodology explicitly considers important uncertainties in the Monte Carlo calculation performed to determine the number of rods in boiling transition. One of the uncertainties considered in the SLMCPR methodology is the bundle power uncertainty. This uncertainty is determined through comparison of calculated to measured core power distributions. Any miscalculation of void conditions will increase the error between the calculated and measured power distributions and be reflected in the bundle power uncertainty.

Therefore, void-quality correlation uncertainty is an inherent component of the bundle power uncertainty used in the SLMCPR methodology.

The transient analysis methodology is not a statistical methodology and uncertainties are not directly input to the analyses. The transient analyses methodology is a deterministic, bounding approach that contains sufficient conservatism to offset uncertainties in individual phenomena.

Conservatism is incorporated in the methodology in two ways: (1) computer code models are developed to produce conservative results on an integral basis relative to benchmark tests, and (2) important input parameters are biased in a conservative direction in licensing calculations.

The transient analyses methodology results in predicted power increases that are bounding relative to benchmark tests. In addition, for licensing calculations a 110% multiplier is applied to the calculated integral power to provide additional conservatism to offset uncertainties in the AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page B-5 transient analyses methodology. Therefore, uncertainty in the void-quality correlation is inherently incorporated in the transient analysis methodology.

Based on the above discussions, the impact of void-quality correlation uncertainty is inherently incorporated in the analytical methods used to determine the OLMCPR. Biasing of important input parameters in licensing calculations provides additional conservatism in establishing the OLMCPR. No additional adjustments to the OLMCPR are required to address void-quality correlation uncertainty.

8.3 Biasing of the Void-Quality Correlation AREVA has performed studies to determine the OLMCPR sensitivity to biases approaching the upper and lower extremes of the data comparisons shown in Figure B-1 through i;:igure B-4.

For one of these studies, the transient L'lCPR impact was determined by propagating void-quality biases through three main computer codes: MICROBURN-B2, COTRANSA2, and XCOBRA-T.

The [ ] correlation in MICROBURN-B2 was modified to correct the mean to match the measured ATRIUM-10 void fraction data shown in Figure B-2. The modified [ ]

correlation parameters were then modified to generate two bounding correlations for the ATRIUM-10 of +/-0.05 void. The results of this modified correlation are presented in Figure B-5.

COTRANSA2 does not have the [ ] correlation. For COTRANSA2 the modified

[ ] correlations in MICROBURN-B2 were approximated in COTRANSA2 with

[ ].

Figure B-6 shows a comparison of the [ ] ratio ~esults compared to the ATRIUM-10 test data. This approach created equivalent void fractions as the [ ] correlation modifications.

The thermal hydraulic methodology in.corporates the effects of subcooled boiling through use of the Levy model. The Levy model predicts a critical subcooling that defines the onset of boiling.

The critical subcooling is used with a profile fit model to determine the total flow quality that accounts for the presence of subcooled boiling. The total flow quality is used with the void-AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page B-6 quality correlation to determine the void fraction. This void fraction explicitly includes the effects of subcooled boiling. Application of the Levy model results in a continuous void fraction distribution at the boiling boundary.

Like COTRANSA2, XCOBRA-T does not have the [ ] correlation. Unlike COTRANSA2, XCOBRA-T does not have [

]. For the other void scenarios, no correction was done in XCOBRA-T. Not modifying the void-quality correlation for the other void scenarios results in a very small difference in ~CPR.

The transient response was assessed relative to a limiting uprated BWR plant transient calculation. The impact of the change in the void correlations was also captured in the burn history of the fuel (the results are not for an instantaneous change in the void correlations). The SLMCPR response was also assessed with the new input corresponding to the three different void scenarios. The results are provided in Table B-2.

The major influence that the voia-quality models have on scram reactivity worth is through the predicted axial power shape. The void-quality models, used for ATRIUM fuel, result in a very good prediction of the axial power shape.

As seen in the results in Table B-2, modifying the void-quality correlations to correct the mean to match the measured ATRIUM-10 void fraction data results in a very small increase in ~CPR, a very small decrease in SLMCPR, and a very small increase in OLMCPR for this study; therefore, the impact of the correlation bias is insignificant.

The +0.05 void scenarios show an increase in the OLMCPR; however, as mentioned previously, the transient analysis methodology is a deterministic, bounding approach that contains sufficient conservatism to offset uncertainties in individual phenomena. Conservatism is incorporated in the models to bound results on an integral basis relative to benchmark tests. For licensing calculations, important input parameters are biased in a conservative direction. In addition, the licensing calculations include a 110% multiplier to the calculated integral power to provide additional conservatism to offset uncertainties in the transient analysis methodology (which includes the void-quality correlation). Even with an extreme bias in the void correlation of +0.05, AREVA Inc.

Controlled Document I

Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page B-7 the conservatism introduced by the 110% multiplier i's alone sufficient to offset the increase in results presented in Table B-2. For the study, the conservatism of the 110% multiplier was

[ ]. These calculations demonstrate that the overall methodology has sufficient conservatism to account for both the bias and the uncertainty in the void-quality correlation.

To provide a more accurate assessment of the impact of a +0.05 void bias, AREVA would need to re-evaluate the Peach Bottom transient benchmarks; it is likely that the +0.05 void scenario would show overconservatism in the benchmarks. Likewise, the pressure drop correlations and core m,onitoring predictions of power will likely show a bias relative to measured data.

Correcting the models t.o new benchmarks and measured data would further reduce the OLMCPR sensitivity.

8.4 Void-Quality Correlation Uncertainty Summary Integral power is a parameter obtainable from test measurements that is directly related to

.D.CPR and provides a means to assess code uncertainty. The COTRANSA transient analysis methodology was a predecessor to the COTRANSA2 methodology. The integral power figure of merit was introduced with the COTRANSA methodology as a way to assess (not account for) code uncertainty impact on b.CPR. From COTRANSA analyses of the Peach Bottom turbine trip tests, the mean of the predicted to measured integral power was 99.7% with a standard deviation of 8.1%. AREVA (Exxon Nuclear at the time) initially proposed to treat integral power as a statistical parameter. However, following discussions with the NRG, it was agreed to apply a deterministic 110% integral power multiplier (penalty) on COTRANSA calculations for licensing analyses. That increase was sufficient to make the COTRANSA predicted to

measured integral power conservative for all of the Peach Bottom turbine trip tests.

COTRANSA2 is not a statistical methodology and uncertainties are not directly input to the analyses. The methodology is a deterministic bounding approach that contains sufficient conservatism to offset uncertainties in individual phenomena. Conservatism is incorporated in the methodology in two ways: (1) computer code models are developed to produce conservative results on an integral basis relative to benchmark tests, and (2) important input parameters are biased in a conservative direction in licensing calculations. Justification that the integrated effect of all the conservatisms in COTRANSA2 licensing analyses is adequate for EPFOD operation at Brunswick is provided below.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page B-8 The COTRANSA2 methodology results in predicted power increases that are bounding

([ ] on average) relative to Peach Bottom benchmark tests. In addition, for licensing calculations a 110% multiplier is applied to the calculated integral power to provide additional conservatism. This approach adds significant conservatism to the calculated OLMCPR as discussed previously.

Biasing of important input parameters in licensing calculations provides additional conservatism in establishing the OLMCPR. The Peach Bottom turbine trips were performed assuming the measured performance of important input parameters such as control rod scram speed and turbine valve closing times. For licensing calculations, these (and other) parameters are biased in a conservative bounding direction. These conservative assumptions are not combined statistically; assuming all parameters are bounding at the same time produces very conservative results.

With the ATRIUM 10XM void fraction benchmarks presented in Figure B-2 and Figure B-4, the applicability of the void-quality correlation at high void fractions is confirmed and the uncertainty associated with the application of the correlation to the ATRIUM 1OXM design is derrionstrated to be equivalent to the data used in the bias assessment. Therefore, the sensitivity studies and conclusions drawn from the study are equally applicable to EPFOD operation of the ATRIUM 10XM at Brunswick.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page B-9 Table B-1 AREVA Multi-Rod Void Fraction Validation Database FRIGG-2 FRIGG-3 ATRIUM 10XM (Reference 18) (Reference 16 & 17) ATRIUM-10 KATHY KATHY Axial Power Shape \uniform I[ ]

[ [

Radial Power Peaking uniform mild peaking

] ]

circular array with 36 circular array with 36 [ [

Bundle Design rods + central thimble rods + central thimble

] ]

Pressure (psi) 725 725, 1000, and 1260 [ ] [ ]

Inlet Subcooling (°F) 4.3 to 40.3

4.1to54.7 [ ] [ ]

Mass Flow Rate (lbm/s)

(Based on mass flux assuming 14.3 to 31.0 10.1to42.5 [ ] [ ]

ATRIUM-10 inlet area)

Equilibrium Quality at Measurement Plane (fraction)

-0.036 to 0.203 -0.058 to 0.330 [ ] [ ]

Max Void at Measurement Plane (fraction) 0.828 0.848 [ ] [ ]

Reported Instrument Uncertainty (fraction) 0.025 0.016 [ ] [ ]

[

Number of Data 27 tests, 174 points 39 tests, 157 points [ ]

]

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 8-10 Table B-2 Void Sensitivity Results Modified Modified Modified Reference V-Q V-Q V-Q Parameter Calculation {0.0} {+0.05} {-0.05}

.LlCPR 0.305 0.307 0.321 0.305 SLM CPR 1.09 1.09 1.09 1.09

.LlSLMCPR NA -0.001 -0.002 +0.002 OLMCPR 1.395 1.396 1.409 1.397 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page B-11 Figure B-1 Validation of [ ] using FRIGG-2 and FRIGG-3 Void Data

)

Figure B-2 , Validation of [ ] using A1RIUM-10 and ATRIUM 10XM Void Data AREVA Inc.

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Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page 8-12 Figure B-3 Validation of Ohkawa-Lahey using FRIGG-2 and FRIGG-3 Void Data Figure B-4 Validation of Ohkawa-Lahey using ATRIUM-10 and ATRIUM 10XM Void Data AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page B-13 Figure B-5 Modified Void Fraction Correlation Comparison to ATRIUM-10 Test Data Figure B-6 [ ] Void Fraction Comparison to ATRIUM-10 Test Data AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-1 Appendix C Neutronic Methods C.1 Cross Section Representation CASM0-4 performs a multi-group [ ] spectrum calculation using a detailed heterogeneous description of the fuel lattice components. Fuel rods, absorber rods, water rods/channels and structural components are modeled explicitly. The library has cross sections for [ ]

materials including [ ] heavy metals. Depletion is performed with a predictor-corrector approach in each fuel or absorber rod. The two-dimensional transport solution is based upon the [ ]. The solution provides pin-by-pin power and exposure distributions, homogeneous multi-group (2) micro-scopic cross sections as well as macro-scopic cross sections. Discontinuity factors are determined from the solution. [ ]

gamma transport calculation are performed. The code has the ability to perform [

] calculations with different mesh spacings. Reflector calculations are easily performed.

MICR08URN-82 performs microscopic fuel depletion on a nodal basis. The neutron diffusion equation is solved with a full two energy group method. A modern nodal method solution using discontinuity factors is used along with a [ ]. The flux discontinuity factors are [ ]. A multilevel iteration technique is employed for efficiency.

MICR08URN-82 treats a total of [ ] heavy metal nuclides to account for the primary reactivity components. Models for nodal [ ] are used to improve the accurate representation of the in-reactor configuration. A full three-dimensional pin power reconstruction method is utilized. TIP (neutron and gamma) and LPRM response models are included to compare calculated and measured instrument responses.

Modern steady state thermal hydraulics models define the flow distribution among the assemblies. [ ] based upon CASM0-4 calculations are used for the in-channel fluid conditions as well as in the bypass and water rod regions. Modules for the calculation of CPR, LHGR and MAPLHGR are implemented for direct co~parisons to the operating limits.

MICR08URN-82 determines the nodal macroscopic cross sections by summing the contribution of the various nuclides.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-2 I

2:x(p,I1,E,R) = LNp*_:(p,I1,E,R)+~2:~(p,I1,E,R) i;l where:

2:x = nodal macroscopic cross section

~2:~ =background nodal macroscopic cross section (D, 2:r> 2:., 2:r)

N = nodal number density of nuclide "i" er: =microscopic cross section of nuclide "i" I = total number of explicitly modeled nuclides p = nodal instantaneous coolant density I1 = nodal spectral history E = nodal exposure R = control fraction The functional representations of er~ and ~2:~ come from 3 void depletion calculations with CASM0-4. Instantaneous branch calculations at alternate conditions of void and control state are also performed. The result is a multi-dimensional table of microscopic and macroscopic cross sections that is shown in Figure C-1 and Figure C-2.

At BOL the relationship is fairly simple; the cross section is only a function of void fraction (water density) and the reason for the variation is the change in the spectrum due to the water density variations. At any exposure point, a quadratic fit of the three CASM0-4 data points is used to represent the continuous cross section over instantaneous variation of void or water density.

This fit is shown in Figure C-3 and Figure C-4.

Detailed CASM0-4 calculations confirm that a quadratic fit accurately represents the cross sections as shown in Figure C-5, Figure C-6, and Figure C-7.

With depletion the isotopic changes cause other spectral changes. Cross sections change due to the spectrum changes. Cross sections also change due to self-shielding as the concentrations change. These are accounted for by the void (spectral) history and exposure parameters. Exposure variations utilize a piecewise linear interpolation over tabulated values at

[ ] exposure points. The four dimensional representation can be reduced to three dimensions (see Figure C-8) by looking at a single exposure.

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Controlled Docurnent Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-3 Quadratic interpolation is performed in each direction independently for the most accurate representation. Considering the case at 70 GWd/MTU with an instantaneous void fraction of 70% and a historical void fraction of 60%, Figure C-9 and Figure C-10 illustrate the interpolation process. The table values from the library at 0, 40 and 80 % void fractions are used to generate 3 quadratic curves representing the behavior of the cross section as a function of the historical void fraction for each of the tabular instantaneous void fractions (0, 40 and 80 %).

The intersection of the three quadratic lines with the historical void fraction of interest are then used to create another quadratic fit in order to obtain the resultant cross section as shown in Figure C-10.

The results of this process for all isotopes and all cross sections in MICROBURN-82 were compared for an independent CASM0-4 calculation with continuous operation at 20, 60 and 90% void and are presented in Figure C-11. Branch calculations at 90% void from a 40% void depletion were performed for multiple exposures. The results show very good agreement for the whole exposure range as shown in Figure C-12.

At the peak reactivity point, multiple comparisons were made (Figure C-13) to show the results for various instantaneous void fractions.

[

]

Void fraction has been used for the previous illustrations; however MICROBURN-82 uses water density rather than void fraction in order to account for pressure changes as well as sub-cooled density changes. This transformation does not change the basic behavior as water density is proportional to void fraction. MICROBURN-82 uses spectral history rather than void history in order to account for other spectral influences due to actual core conditions (fuel loading, control rod inventory, leakage, etc.) The Doppler feedback due to the fuel temperature is modeled by accumulating the Doppler broadening of microscopic cross sections of each nuclide.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-4 where:

Teff =Effective Doppler Fuel Temperature Tref =Reference Doppler Fuel Temperature CJ: = Microscopic Cross Section (fast and thermal absorption) of nuclide "i" N; =Density of nuclide "i" The partial derivatives are determined from branch calculations performed with CASM0-4 at various exposures and void fractions for each void history depletion. The tables of cross sections include data for [ ] states. The process is the same for

[ ] states. Other important feedbacks to nodal cross sections are lattice [ ] and instantaneous [ ] between lattices of different [ ]. These feedbacks are modeled in detail.

The methods used in CASM0-4 are state of the art. The methods used in MICROBURN-82 are state of the art. The methodology accurately models a wide range of thermal hydraulic conditions including EPU and extended power/flow operating map conditions.

C.2 Applicability of Uncertainties The TIPs directly measure the local neutron flux from the surrounding four fuel assemblies.

Thus, the calculated bundle power distribution uncertainty will be closely related to the calculated TIP uncertainty. However, the bundle powers in the assemblies surrounding a TIP are not independent. If a bundle is higher in power, neutronic feedback increases the power in the nearby assemblies, thus producing a positive correlation between nearby bundles. The gamma scan data provides the means to determine this correlation according to the EMF-2158(P)(A) (Reference 19) methodology. A smaller correlation coefficient implies that there is less correlation between nearby bundle powers, thus, there would be a larger bundle power distribution uncertainty.

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Controlled Docurnent Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-5 The EMF-2158(P)(A) data was re-evaluated by looking at the deviations between measured and calculated TIP response for each axial level. The standard deviation of these deviations at each axial plane are presented in Figure C-15 and demonstrate that there is no significant trend vs.

axial position, which indicates no significant trend vs. void fraction. This same data was evaluated for trends based upon the core conditions at the time of each TIP scan. The core parameters of interest that were evaluated include core thermal power, the core average void fraction and the ratio between core power and core flow. The 20 standard deviations for "C" and "O" lattice plants are presented in Figure C-16 through Figure C-21, while the 30 standard deviations are presented in Figure C-22 through Figure C-27. This evaluation of the date indicates that there is no significant trend in the data associated with these plant parameters.

To compare core physics models to the gamma scan results, the calculated pin power distribution is converted into a Ba140 density distribution. A rigorous mathematical process

,using CASM0-4 pin nuclide inventory and MICROBURN-82 nodal nuclide inventory is used.

Gamma scan comparisons for 9X9-1 and ATRIUM-10 fuel were presented in the topical report, EMF-2158(P)(A), in Section 8.2.2. Reference 19 Figures 8.18 through 8.31 showed very good comparisons between the calculated and measured relative Ba-140 density distributions for both radial and axial values.

The Quad Cities assembly gamma scan data was used to determine the correlation coefficient which accounts for the correspondence between the assembly powers of adjacent assemblies.

This correspondence is quantified by a conservative multiplier to the uncertainty in the TIP measurements. In order to conservatively account for this correspondence, the bundle power uncertainty is increased due to the radial TIP uncertainty by a multiplier based on the correlation coefficient. The correlation coefficient is statistically calculated and shown in Figure 9.1 and Figure 9.2 of EMF-2158(P)(A). It indicates a less than perfect correlation between powers of neighboring bundles. The conservative multiplier is calculated as follows:

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-6 The calculated TIP uncertainty would normally be expected to be slightly larger than the calculated power uncertainty due to the TIP model. The Quad Cities gamma scan comparison shows the 2-D radial power uncertainty of [ ] (see Section 9.6 of EMF-2158(P)(A)). The D-Lattice plant calculated radial TIP uncertainty is

[ ]. The data indicates that the calculated TIP uncertainty is indeed larger than the calculated power uncertainty. The use of the correlation coefficient to increase the calculated power uncertainty is a very conservative approach resulting from the statistical treatment. The types of fuel bundles (8x8, 9x9, or 10x10) loaded in the core has no effect on the reality of the physical model which precludes the possibility of the calculated power uncertainty to be larger than the calculated TIP uncertainty. The accuracy of the MICROBURN-B2 models is demonstrated by comparisons between measured and calculated Tl P's as well as comparison of calculated and measured Lanthanum-140 activation. The accuracy of the MICROBURN-B2 models was further validated with detailed axial pin by pin gamma scan measurements of 9X9-1 and ATRIUM-10 fuel assemblies in the reactor designated as KWU-S. These measurements demonstrated the continued accuracy of the MICROBURN-B2 models with modern fuel assemblies. Details of these measurements are provided in Section 8.2.2 of the topical report, EMF-2158(P)(A).

The AREVA SAFLIM3D code is used to calculate the number of expected rods in boiling transition (BT) for a specified value of the SLMCPR (i.e., SLMCPR is an input, not a calculated result). The extremes of the two correlation coefficients from the Quad Cities assembly gamma scan data sets [ ] discussed in section C.2 were used for a sensitivity study of the MCPR safety limit. An analysis of the safety limit was performed with SAFLIM3D using an input SLMCPR of 1.0658 and the base RPF and nodal power uncertainties. The number of boiling transition (BT) rods was calculated to be 50 from this analysis. The analysis was repeated in a series of SAFLIM3D calculations using the increased RPF and nodal power uncertainties and performed by iterating on the input value of SLMCPR. Different values for the SLMCPR input were used until the number of BT rods calculated by SAFLIM3D was the same as the base case (50 rods). A SLMCPR input value of [ ] resulted in 50 rods in BT when the increased RPF and nodal power uncertainties was input. The difference in SLMCPR input

[ ] for the two cases that resulted in the same number of BT rods is a measure of the safety limit sensitivity to the increased RPF and nodal power uncertainties.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-7 The only input parameters that changed between the two SAFLIM3D analyses were the SLMCPR, the RPF, and nodal power uncertainties. For each analysis, 1000 Monte Carlo trials were performed. To minimize statistical variations in the sensitivity study, the same random number seed was used and all bundles were analyzed for both analyses. As discussed above, 50 rods were calculated to be in BT in both analyses.

This sensitivity study was performed to quantify the sensitivity of SLMCPR to an increase in RPF and nodal power uncertainties and did not follow the standard approach used in SLMCPR licensing analyses. In standard licensing calculations, the SLMCPR is not input at a precision greater than the hundredths decimal place. As a result, the increased RPF and nodal power uncertainties would result [ ] in SLMCPR licensing analyses depending on how close the case was to the acceptance criterion prior to the increase in RPF uncertainty.

Gamma scanning provides data on the relative gamma flux from the particular spectrum associated with La140 gamma activity. The relative gamma flux corresponds to the relative La140 concentration. Based upon the time of shutdown and the time of the gamma scan the Ba140 relative distribution at the time of shutdown is determined. This Ba140 relative distribution is thus correlated to the pin or assembly power during the last few weeks of operation. The data presented in the topical report, EMF-2158(P)(A), includes both pin and assembly Ba140 relative density data. The assembly gamma scan data was taken at Quad Cities alter the operation of cycles 2, 3 and 4. Some of this data also included individual pin data. This data was from 7X7 and 8X8 fuel types. Additional fuel pin gamma scan data was taken at the Gundremingen plant for ATRIUM-9 and ATRIUM-10 fuel. This data is also presented in the topical report.

Pin-by-pin Gamma scan data is used for verification of the local peaking factor uncertainty ..

Quad Cities measurements presented in the topical report EMF-2158(P)(A) have been re-evaluated to determine any axial dependency. Figure C-28 presents the raw data including measurement uncertainty and demonstrates that there is no axial dependency. The more recent Gamma scans performed by KWU, presented in the topical report EMF-2158(P)(A) and re-arranged by axial level in Table C-1, indicate no axial dependency. Full axial scans were performed on 16 fuel rods. Comparisons to calculated data show excellent agreement at all axial levels. The dip in power associated with spacers, observed in the measured data, is not AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-8 modeled in MICROBURN-82. There is no indication of reduced accuracy at the higher void fractions.

CASM0-4 and MCNP calculations have been performed to compare the fission rate distribution statistics to Table 2-1 of the topical report EMF-2158(P)(A) which is shown in Table C-2. The fission rate differences at various void fractions demonstrate that CASM0-4 calculations have very similar uncertainties relative to the MCNP results for all void fractions. These fission rate differences also meet the criteria of the topical report EMF-2158(P)(A) for all void fractions.

Data presented in these figures and tables demonstrate that the AREVA methodology is capable of accurately predicting reactor conditions for fuel designs operated under the current operating strategies and core conditions.

C.3 Fuel Cycle Comparisons AREVA has revie\11,/ed the data presented in EMF-2158(P)(A) with regard to the maximum assembly power (Figure C-29) and maximum exit void fraction (Figure C-30) to determine the range of data previously benchmarked.

Fuel loading patterns and operating control rod patterns are constrained by the minimum critical power ratio (MCPR) limit, which consequently limits the assembly power and exit void fraction regardless of the core power level. Operating data from several recent fuel cycle designs has been evaluated and compared to that in the topical report EMF-2158(P)(A). Maximum assembly power and maximum void fraction are presented in Figure C-31 and Figure C-32.

In order to evaluate some of the details of the void distribution a current design calculation was reviewed in more detail. Figure C-33 and Figure C-~4 present the following parameters at the point of the highest exit void fraction (at 9336 MWd/MTU cycle exposure) in cycle core design for a BWR-6 reactor with ATRIUM-10 fuel. Another measure of the thermal hydraulic conditions is the population distribution of the void fractions. These are representative figures for a high power density plant.

Core average void axial profile

Axial void profile of the peak assembly

Histogram of the nodal void fractions in core AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-9 The actual core designs used for each cycle will have slightly different power distributions and reactivity characteristics than any other cycle. Conclusions from analyses that are dependent on the core design (loading pattern, control rod patterns, fuel types) are re-confirmed as part of the reload licensing analyses performed each cycle. Cycle-specific reload licensing calculations will continue to be performed for all future EPFOD cycles using NRC approved methodologies consistent with the current processes.

Brunswick operating under EPFOD conditions (Figure C-35 and Figure C-36) can be compared to the equivalent data of the topical report EMF-2158(P)(A). Comparison of Figure C-29 vs.

Figure C-35 and Figure C-30 vs. Figure C-36 shows that EPU operation in the extended power/flow map is within the range of the original methodology approval for assembly power and exit void fraction. From a neutronic perspective, moderator density (void fraction) and exposure cause the greatest variation in cross sections. NRC-approved exposure limits for ATRIUM-10 fuel evaluated with AREVA methods are unchanged for EPFOD conditions.

Reactor conditions for Brunswick with power uprate and extended power/flow map are not significantly different from that of current experience and are bounded by the experience for the important parameters.

The axial profile of the power and void fraction of the limiting assembly and core average values are presented in Figure C-37 for a Brunswick EPFOD cycle design. These profiles demonstrate that the core average void fraction and the maximum assembly power void fractions are bounded by the topical report data and are consistent with recent experience on other reactors.

Figure C-38 presents a histogram of the void fraction for EPFOD conditions. This histogram was taken at the point of maximum exit void fraction expected during the cycle. The distribution of voids is shifted slightly toward the 70 -80 % void fraction levels. The population of nodes experiencing 85 -90% voids is still small.

The neutronic and thermal hydraulic conditions predicted for the EPFOD operation are bounded by the data provided in the-*t~pical report EMF-2158(P)(A) so the isotopic validation continues to be applicable to EPFOD operation.

The AREVA methodology [ ] the reactivity coefficients that are used in the transient analysis. Conservatisms in the methodology are used to produce conservative results that bound the uncertainties in the reactivity coefficients. Data AREVA Inc.

led Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-10 presented in these referenced figures indicate that there are no significant differences between EPFOD and non- EPFOD conditions that have an impact on the reactivity coefficients.

C.3.1 Bypass Voiding The core bypass water is modeled in the AREVA steady-state core simulator, transient I

simulator, LOCA and stability codes as [ ].

The steady-state core simulator, MICR08URN-82, explicitly models the assembly specific flow paths through the lower tie-plate flow holes and the channel seals in addition to a [

] through the core support plate. The numerical solution for the individual flow paths is computed based on a general parallel channel hydraulic solution that imposes a constant pressure drop across the core fuel assemblies and the bypass region. This solution scheme incorporates [

] that is dependent on the [

].

The MICR08URN-82 state-point specific solution for bypass flow rate and [ ] ,

is then used as initial conditions in the transient and LOCA analyses. When the reactor 1 operates on high rod-lines at low flow conditions, the in-channel pressure drop decreases to a point where a solid column of water cannot be supported in the bypass region, and voiding occurs in the core bypass. For these conditions (in the region bf core stability concerns) the neutronic feedback of bypass voiding [

]

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Controlled Document Applicability of ARi::VA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-11

[

]

The level of bypass boiling for a given state-point is a direct result of the hydraulic solution. The potential for boiling increases as the power/flow ratio increases or the inlet. sub-cooling.

decreases. While the licensing methodology utilizes a [

] to estimate the potential for localized bypass boiling. This [

] to specifically determine a bounding local void distribution in the core. The I model is conservative in that it [

]. The capability of this model to predict localized bypass boiling is demonstrated in Figure C-39 through Figure C-41 for a Brunswick case that produced the maximum bypass exit void fraction. Figure C-39 presents the average void fraction for the channel bypass and Figure C-40 presents the core exit bypass channel void fraction. The most significant impact of voiding in the bypass is the impact on the LPRM reading. The average void fraction ofthe four channels surrounding any LPRM location is presented in Figure C-41. The impact of 0.3 % void fraction on the LPRM response is negligible.

Bypass voiding is of greatest concern for stability analysis due to its direct impact on the fuel channel flow rates and the axial power distributions. The reduced density head in the core bypass due to boiling results in a higher bypass flow rate and consequently a lower hot channel flow rate. This lower h.ot channel flow rate and. a more bottom-peaked ~ower distribution (due to lower reactivity in the top of the core due to boiling in the bypass region) destabilize the core through higher channel decay ratios. AREVA stability methods directly model these phenomena to assure that the core stability is accurately predicted.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-12 CASM0-4 has the capability to specify the density of the moderator in the bypass and in-channel water rods, [

].

Significant bypass voiding is not encountered during full power, steady-state EPFOD operation for Brunswick so there is no impact on steady-state analyses. For transient conditions it is conservative to ignore the density changes as additional voiding aids in shutting down the power generation.

For Brunswick, a 100% power I 85% flow statepoint (120% of the original licensed thermal power) was assessed. Even with the conservative multi-channel model, there was minimal localized bypass boiling at the EPU power level. This assessment assures that the limiting transients at the uprated thermal power are not adversely affected by bypass boiling. As the flow is reduced along the 100% power line, the decrease in flow is compensated by increased sub-cooling which compensates for the decrease in flow. When flow is further reduced along the highest rod line, more significant boiling in the bypass is calculated to begin. This is in the area of stability concerns where the boiling in the bypass is modeled explicitly. For normal operation at 100% power minimal boiling in the bypass is expected to occur, so there is no impact on the lattice local peaking or the LPRM response.

C.3.2 Fuel Assembly Design No fuel design modifications have been made for EPFOD operation, neither mechanical nor thermal hydraulic. The maximum allowed enrichment level of any fuel pellet is 4.95 wt% U-235.

A description of fuel enrichments on both a lattice basis and an assembly basis for the first reload of ATRIUM 10XM fuel in Brunswick is provided in Table C-3.

All new and spent fuel at Brunswick is stored in the Spent Fuel Storage Pool (SFSP) and in accordance with Technical Specification 4.3.1.1 must maintain a subcritical multiplication factor (keff) of less than 0.95 when flooded with non-borated water. A SFSP criticality analysis has been performed for Brunswick that confirms that this requirement is met for ATRIUM-10 fuel designs. This analysis applies to both units which have the same high density storage rack configuration, as detailed in section 9.3 of the FSAR. A reload specific evaluation is performed to verify that the specific bundle designs being loaded remain bounded by the criticality analysis.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-13 Table C-1 KWU-S Gamma Scan Benchmark Results from EMF-2158(P)(A)

Table C-2 Comparison of CASM0-4 and MCNP results for ATRIUM-10 Design AREVA Inc. /

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-14 Table C-3 Fuel Enrichment Description for the Initial Brunswick EPFOD ATRIUM 10XM Fuel Cycle Design NOTES:

1. If this limit is not met, a CASMO in-rack k-inf calculation can be performed. If k-inf is less than the base analysis, then fuel may be stored.

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Controlled Document Appl icability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-15 A10B-4 340L-15G70 U235 Thermal Absorption

-+- 0.0 VI I 0.0 VH

-+- 0.4 VI I 0.0 VH 0.8 VI I 0.0 VH

""*"" 0.0 VI I 0.4 VH

-+- 0.4 VI I 0.4 VH

-+- 0.8 VI I 0.4 VH

-+- 0.0 VI I 0.8 VH

- 0.4Vl / 0.8VH

- 0.8Vl / 0.8VH 10 20 JO 40 50 60 70 80 Ex posu re (GWd/MTU)

Figure C-1 Microscopic Thermal Cross Section of U-235 from Base Depletion and Branches A10 B-4340L-1 5G70 U235 Fast Absorpt ion

-+- 0.0 VI I 0.0 VH

-+- 0.4 VI I 0.0 VH 0.8 VI I 0.0 VH

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-+- 0.8 VI I 0.4 VH

-+- 0.0 VI I 0.8 VH

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- 0.8 VI I 0.8 VH 10 20 JO 40 50 60 70 80 Exposure {GWd/MTU)

Figure C-2 Microscopic Fast Cross Section of U-235 from Base Depletion and Branches AR EVA Inc.

Controlled Document Applicability of AREVA BWR' ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-16 SOL A10S-4340L-15G70 U235 Thermal Cross Sections l.

c 190

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Figure C-4 Microscopic Fast Cross Section of U-235 at Beginning of Life AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-17 BOL A10B-4245L-14G70 U-235 Microscopic Cross Sections (Thermal)

+ Sig*A2 (CASM0-4)

- Quadratic Ftt (0.40,80)

Sig-F2 (CASM0-4)

- Quadratic Ftt (0.40,80) 0 .1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 Void Fra ction Figure C-5 Microscopic Thermal Cross Section of U-235 Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions BOL A10B-4245L-14G70 U-235 Mic ros copic Cross Sections (Fast) 13.0 ~----------------------~

+ Sig-A1 (CAS M0-4)

- Quadratic Fit (0,40,80)

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- Quadratic Fit (0,40,80) 8.0 r=~=~====~=""""-'"'==-----------_j 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 Voi d Fra ction Figure C-6 Microscopic Fast Cross Section of U-235 Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-18 BOL A108-4254L-14G70 Macroscopic Diffusion Coefficients

+ D-1 (CASM0-4)

- Quadratic ((µ0-80) a 0-2 (CASM0 -4)

- Quadratic ((µQ-80) 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 Vo ld Fraction Figure C-7 Macroscopic Diffusion Coefficient (Fast and Thermal)

Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-19 A1 Ut:i-4J4UL-1 :>l:i fU U:.!J:> I nermal Aosorpt1on at fU l:iVVO/M I u 280 270 260 250 Cross Section (barns) 240 Instantaneous Vo id Fraction (%)

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-20 A10B-4340L-15G70 U235 Thermal Absorption at 70 GWd/MTU

+ 0.0 Inst Void (CASM0-4)

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-21 Figure C-11 Comparison of k-infinity from MICROBURN-82 Interpolation Process with CASM0-4 Calculations at Intermediate Void Fractions of 0.2, 0.6 and 0.9 Figure C-12 Comparison of k-infinity from MICROBURN-82 Interpolation Process with CASM0-4 Calculations at 0.4 Historical Void Fractions and 0.9 Instantaneous Void Fraction AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-22 Figure C-13 Delta k-infinity from MICROBURN-82 Interpolation Process with CASM0-4 Calculations at 0.4 Historical Void Fraction and 0.9 Instantaneous Void Fraction Figure C-14 Comparison of Interpolation Process Using Void Fractions of 0.0, 0.4 and 0.8 and Void Fractions of 0.0, 0.45 and 0.9 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-23 Figure C-15 EMF-2158(P)(A) TIP Statistics by Axial Level Figure C-16 EMF-2158(P)(A) 2-D TIP Statistics for C-Lattice Plants vs. Core Power AREVA Inc.

Contr611ed Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-24 Figure C-17 EMF-2158(P){A) 2-D TIP Statistics for C-Lattice Plants vs. Core Average Void Fraction Figure C-18 EMF-2158(P)(A) 2-D TIP Statistics for C-Lattice Plants vs. Core Power/Flow Ratio AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-25 Figure C-19 EMF-2158(P)(A) 2-D TIP Statisti~s for D-Lattice Plants vs. Core Power

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Figure C-20 .EMF-2158(P)(A) 2-D TIP Statistics for D-Lattice Plants vs. Core Average Void Fraction AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-26 Figure C-21 EMF-2158(P){A) 2-D TIP Statistics for D-Lattice Plants vs. Core Power/Flow Ratio Figure C-22 EMF-2158(P){A) 3-D TIP Statistics for C-Lattice Plants vs. Core Power AREVA Inc.

Controlled Document Applicability of AREVA BWR' ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-27 Figure C-23 EMF-2158(P)(A) 3-D TIP Statistics for C-Lattice Plants vs. Core Average Void Fraction Figure C-24 EMF-2158(P)(A) 3-D TIP Statistics for C-Lattice Plants vs. Core Power/Flow Ratio AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-28 Figure C-25 EMF-2158(P)(A) 3-D TIP Statistics for D-Lattice Plants vs. Core Power

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Figure C-26 EMF-2158(P)(A) 3-D TIP Statistics for D-Lattice Plants vs. Core Average Void Fraction AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-29 Figure C-27 EMF-2158(P)(A) 3-D TIP Statistics for D-Lattice Plants vs. Core Power/Flow Ratio I '

Figure C-28 Quad Cities Unit 1 Pin by Pin Gamma Scan Results AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-30 Figure C-29 Maximum Assembly Power in Topical Report EMF-2158(P)(A)

Figure C-30 Maximum Exit Void Fraction in Topical Report EMF-2158(P)(A)

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-31 Figure C-31 Maximum Assembly Power Observed from Recent Operating Experience Figure C-32 Void Fractions Observed from Recent Operating Experience AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-32 Figure C-33 Axial Po~er and Void Profile Observed from Recent Design Experience Figure*C-34 Nodal Void Fraction Histogram Observed from*Recent Design Experience AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-33 Figure C-35 Maximum Assembly Power in an EPFOD Brunswick Design Figure C-36 Maximum Exit Void Fraction in an EPFOD Brunswick Design AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-34 Figure c:.37 Brunswick EPFOD Design Axial 'Profile of Power and Void Fraction Figure C-38 Brunswick EPFOD Design Nodal Void Fraction Histogram AREVA Inc.

Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-35 EDIT OF VOID FRACTION .( %) IN BYPASS CHANNEL CD 0..

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Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-36 EDIT OF BYPASS CHANNEL EXIT VOID FRACTION (%)

0 0

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page C-37 EDIT OF LPRM REGION BYPASS VOID FRACTION (%)

Figure C-41 MICROBURN-B2 Multi-Channel Bypass Void at an LPRM Location from a Brunswick Equilibrium Cycle Design AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page D-1 Appendix D Transient Methods D.1 COTRANSA2 Cross Section Representation The COTRANSA2 transient simulator solves the one-dimensional neutron diffusion equation to predict the core average power response. In order to accurately capture the core reactivity characteristics, a series of MICROBURN-82 calculations are performed. These successive calculations are:

1) Nominal initial conditions AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page D-2 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page D-3 Th~ 1% energy group diffusion equation in steady-state can be written as The first term is a leakage. This equation is integrated over the cylindrical node depicted in the following figure.

H <l\i+1 01,i+,1 H <l\i 01.,I H <1>1,i-1 01,i-1 The leakage term is approximated as:

_ LJ201D1*(c/J1*-c/J1)

,I ,j ,I ,}

A J=I (01,i+D1,1) HV where 0 1,; = ,0 for plane of interest 0 1,j = O for the nodes adjacent to the plane of interest AREVA Inc.

Controlled Docurnent Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page D-4 W1,i = flux in the plane of interest w1,j = flux in the regions adjacent to the plan'e of interest A = surface area between nodes i and j H = distance between nodes i and nodes j V = node volume AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page D-5 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page D-6 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswicl< Extended Revision 1 Power Flow Operating Domain Page D-7 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page D-8 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page D-9 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page D-10 These final one group cross section and leakage parameters are used in a new 1-dimensional flux solution and the axial power distribution is updated for the next thermal hydraulic solution.

The process is repeated for every core solution until a converged core power, power distribution, temperature distribution and density distributions are obtained.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page D-11 Figure D-1 Comparison of Scram Bank Worth for [ ]

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page E-1 Appendix E LOCA Modifications E.1 LOCA Analysis The AREVA LOCA methodology applied for application to EPFODs differs from the approved methodology in three aspects:

Radiation View Factors In the Safety Evaluation for Reference 23 the NRC approved the AREVA EXEM BWR-2000 ECCS evaluation model. The HUXY code (Reference 24) is the part of this model that performs the heatup calculations and provides PCT and local clad oxidation at the axial plane of interest.

The code evaluates the radiation heat transfer between the fuel rod of interest and other fuels rods, the internal water canisters, and the fuel channel. AREVA has implemented an automated approach for calculating radiation view factors within the HUXY computer program.

The original approach was based on the method of cross-strings as described in Section 2.3 of Reference 24. This resulted in the derivation and programming of analytical expressions as a function of fuel rod diameters for the radiation view factors between each fuel rod and its predominant neighbors. The view factors were then internally computed throughout the HUXY heatup analyses based on these analytical expressions and the time dependent evolution of the fuel rod dimensions.

This analytical approach was well suited for the 7x7 and 8x8 designs analyzed at the time the method was originally implemented. With the evolution of design features such as larger internal water structures and part-length rods, the assembly lattice has become more heterogeneous and the derivation of the analytical expressions for the radiation view factors has become more complex. To automate the computation of the view factors for current and future fuel design conc~pts, a numerical computation of the view factors has been introduced utilizing a straight forward finite element method. The numerical computation, achieved through a ray-tracing algorithm, provides a direct accounting of the geometry to compute the view factor from each fuel rod to all other fuel rods, internal water structures, and the external fuel channel without the simplifying assumptions associated with the previous analytical approach.

Therefore, not only is the calculation of view factors automated, but the resulting values are AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page E-2 more precise. The view factors calculated with the ray-tracing algorithm are then utilized identically to the way view factors developed from the cross-string method were utilized in the HUXY heatup analysis.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page E-3 Thermal Conductivity Degradation The EXEM BWR-2000 ECCS evaluation model uses the RODEX2 fuel rod models and therefore, underpredicts the impact of thermal conductivity degradation with exposure. The evaluation of thermal conductivity degradation and impact on PCT for Brunswick EPFOD are presented in Appendix F.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page E-4 Figure E-1 [ ]

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page F-1 Appendix F Fuel Conductivity Degradation F .1 Introduction The U.S. Nuclear Regulatory Commission (NRG) issued Information Notice (IN) 2009-23 (No. ML091550527), dated October 8, 2009, for concerns regarding the use of historical fuel thermal conductivity models in the safety analysis of operating reactor plants. IN 2009-23 discusses how historical fuel thermal mechanical codes may overpredict fuel rod thermal conductivity at higher burn-ups based on new experimental data. This new experimental data showed significant degradation of fuel pellet thermal conductivity with exposure. The NRG staff concluded that the use of the older legacy fuel models will result in predicted fuel pellet conductivities that are higher than the expected values.

This appendix summarizes the impact and treatment of fuel conductivity degradation for licensing safety analyses supporting EPFOD operation at Brunswick.

F.2 Disposition of Licensing Safety Analysis for Brunswick ATRIUM 10XM Fuel RODEX2 and RODEX2A codes were approved by the NRG in the early and mid-1980's, respectively. At that time, thermal conductivity degradation (TCD) with exposure was not well characterized by irradiation tests or post-irradiation specific-effects tests at high burnups. The fuel codes developed at that time did not accurately account for this phenomenon. Analyses performed with RODEX2/2A are impacted by the lack of a~ accurate thermal conductivity degradation model. Likewise, conductivity models in the transient codes COTRANSA2 and XCOBRA-T do not account for thermal conductivity degradation.

RODEX4 (Reference 3) is a best-estimate, state-of-the-art fuel code that fully accounts for burnup degradation of fuel thermal conductivity. RODEX4, therefore, can be used to quantify the impact of burnup-dependent fuel thermal conductivity degradation and its effect on key analysis parameters.

Thermal-mechanical licensing safety analyses for Brunswick are performed with RODEX4 and therefore explicitly account for thermal conductivity degradation. No additional assessment is AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page F-2 needed for those analyses. For thermal-hydraulic and safety analyses an eval.uation is needed.

The following analysis methodologies use RODEX2 and/or include a separate U0 2 thermal conductivity correlation:

Anticipated Operational Occurrence (AOO) analysis based on COTRANSA2/RODEX2/XCOBRA-T codes;

Loss of Coolant Accidents (LOCA) analyses based on RELAX/RODEX2/HUXY codes;

Overpressurization analyses based on COTRANSA2/RODEX2 codes;

Stability analyses based on STAIF/RAMONA5-FA codes; and

Fire event (or Appendix R) analyses based on RELAX/RODEX2/HUXY codes.

F.3 Assessment of Analyses for Brunswick Operations The issues identified in IN 2009-23 were entered into the AREVA corrective action program in 2009. A summary of the investigation was provided to the NRC in a white paper (Reference 27). The white paper presented results of an extensive evaluation; for BWRs the assessments consisted primarily of ATRIUM-10 fuel. A summary of that evaluation is provided as follows for the items noted in the previous section. In addition, the assessment of EPFOD operation at Brunswick with ATRIUM 10XM fuel is discussed at thtend of each subsection.

The NRC reviewed Reference 27 and provided requests for information in Reference 28.

AREVA provided responses in Reference 29. Items relevant from References 28 and 29 are also discussed in the following subsections.

F.3.1 Anticipated Operational Occurrence Analyses The computer codes COTRANSA2 and XCOBRA-T are used in AOO analyses. Both codes use U0 2 thermal conductivity correlations that do not address TCD. In addition, the core average gap conductance used in the COTRANSA2 system calculations and the hot channel gap conductance used in XCOBRA-T hot channel calculations are obtained from RODEX2 calculations. In general, the sensitivity to conductivity and gap conductance for AOO analyses AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP '

Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page F-3 is in the opposit~ direction for the core and hot channel, i.e., putting more energy into the coolant (higher thermal conductivity/higher gap conductance) is non-conservative for the system calculation but conservative for the hot channel calculation. The competing effects between the core and hot channel calculation minimize the overall impact of thermal conductivity degradation.

The assessment of Reference 27 demonstrated that COTRANSA2 uses several conservative assumptions, which results in conservatism relative to the Peach Bottom turbine trip qualification database. The COTRANSA2 methodology-results in predicted integral power increases that are bounding relative to the Peach Bottom benchmark tests. With the 110%

integral power multiplier used in the methodology, the COTRANSA2 predicted to measured mean integral power is [ ] for the Peach Bottom turbine trip tests. The COTRANSA2 benchmark testing was performed using the same U0 2 conductivity model as used in the current licensing analyses. Therefore, the benchmarking comparisons inherently include any impact of U0 2 conductivity degradation with exposure exhibited in the Peach Bottom tests.

The prior assessment was based on fuel designs current at the time of the Peach Bottom tests.

To supplement the assessment with modern fuel, calculations were performed using the as-submitted AURORA-B code (Reference 30). AURORA-B is built from previous NRC approved methods. These methods include codes RODEX4, MICROBURN-B2, and S-RELAPS; U02 thermal conductivity degradation is correctly modeled. It is noted that the AURORA-B methodology and application have not yet been reviewed by the NRC; however, the staff accepted its use for sensitivity calculations for this assessment (Reference 28). The AURORA-B sensitivity studies show that the impact of fuel thermal conductivity degradation with exposure results in a decrease in the 8CPR of [ ] increase in the transient LHGR excursion.

Based on the inherent conservatisms associated with the transient analysis codes and the small impact of thermal conductivity degradation with exposure for the AOO analysis, it is concluded that MCPR and LHGR operating limits based on the AOO methodology are not impacted.

The application of the methodology for EPFOD operation does not change the conservatisms nor invalidate the sensitivity; therefore, the AOO methodology remains applicable for Brunswick.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page F-4 It should be noted that transient LHGR analyses are performed with the RODEX4 code for Brunswick ATRIUM 1OXM fuel, which correctly accounts for thermal conductivity degradation.

F.3.2 Loss of Coolant Accident Analyses LOCA analyses are performed using the EXEM BWR-2000 methodology and include the use of the RODEX2, RELAX and HUXY computer codes. In addition to the initial stored energy, the RODEX2 code is used to calculate fuel mechanical parameters for use in the HUXY computer

\

code that potentially impact the clad ballooning and rupture models. Clad ballooning has a small impact on Peak Cladding Temperature (PCT) and metal water reaction (MWR), but clad rupture can have a significant impact on PCT, depending on event timing.

The LOCA event is divided into two phases: the blowdown and refill/reflood phases. During the initial or blowdown portion of a LOCA, good cooling conditions exist, and the initial stored energy in the fuel is removed. While a decrease in the thermal conductivity increases the overall thermal resistance, heat transfer conditions remain sufficient to remove the initial stored energy. Numerous sensitivity studies have been performed to demonstrate that BWR LOCA analyses are insensitive to initial stored energy. After the initial phase of a LOCA, the heat transfer coefficient at the cladding surface is degraded due to the loss of coolant (low flow and high quality). As a result, the heat transfer from the fuel is primarily controlled by the surface heat flux, and the temperature profile across the pellet is very flat. When compared to the rod surface thermal resistance, the pellet thermal conductivity is not a significant portion of the fuel rod total thermal resistance. Therefore, LOCA calculations are not sensitive to the U0 2 thermal conductivity used in RELAX and HUXY.

To demonstrate that limiting LOCA calculations are not sensitive to U0 2 thermal conductivity, assessments were performed for multiple BWRs. Most LOCA analyses of record are limiting at beginning of life (BOL) conditions. Thermal conductivity degradation may impact calculated PCTs and oxidation as exposure increases; however, since the MAPLHGR limit decreases linearly at higher burnups, significant margin is gained that offsets any decrease in margin associated with thermal conductivity degradation. For these cases, increases in PCT at later exposures remained non-limiting.

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Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page F-5 Assessments of the potential impact of exposure-dependent degradation of U0 2 thermal conductivity on the fuel mechanical parameters were made using the RODEX4 computer code.

The RODEX4 code explicitly incorporates the impact of U0 2 thermal conductivity degradation with exposure. RODEX4 calculations were performed with and without the models which account for TCD. The differences in these RODEX4 results were used to adjust the RODEX2 data which is input to HUXY. The results of these evaluations were summarized to the NRG in References 27 and 29.

The impact of TCD was incorporated in the Brunswick ATRIUM 10XM HUXY analysis. For Brunswick, an input option was added to RODEX4 to allow the analyst to turn off the models for thermal conductivity degradation with exposure. The impact of T~CD was determined by running the nominal RODEX4 fuel rod depletions and then repeating them with the input option selected to turn off TCD. These RODEX4 results were used to increase the stored energy (average pellet temperature) calculated by RODEX2 prior to their input to HUXY. As shown below, this is a very conservative method for evaluating the impact of TCD since RODEX2 was developed to calculate conservatively high stored energy in support of its use as part of the AREVA Appendix K BWR LOCA methodology.

After the NRG approval of RODEX2, more Halden tests were performed with fuel centerline temperature monitoring. As with the RODEX4 submittal, this extended temperature database was used to benchmark RODEX2 over the approved burn up range. The extended temperature I

benchmarking for RODEX2 shows centerline temperature remains conservative to at least 10 GWd/MTU (Reference 27). Even so, the increase in stored energy predicted with the TCD models in RODEX4 was applied to the RODEX2 calculated stored energy for all nonzero exposures.

The ATRIUM 10XM PCT results with and without the impact of TCD are provided in Table F-1.

The highest PCT is calculated at 0.0 GWd/MTU, where there is no TCD.

F.3.2.1 Responses to NRC Requests From the NRC's review of Reference 27, additional information was requested in Reference 28.

The information requests and responses are provided as follows:

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page F-6 A detailed explanation of the source* of the heat transfer coefficients utilized in the HUXY calculation This request is answered in Reference 29 and this answer is applicable to Brunswick.

A description of how LOCA analyses are initialized in terms of power distribution; specifically, how thermal limits (such as MLHGR or OLMCPR) are considered in the initialization This request is answered in Reference 29. This response explains that AREVA's goal is to establish an MAPLHGR limit that is less restrictive than the LHGR limit and satisfies 10 CFR 50.46 acceptance criteria. When this goal is achieved, some of the rods in the HUXY heatup analysis will have LHGRs higher than the LHGR limit. For some Brunswick neutronic lattice designs, the MAPLHGR limit will be more restrictive than the LHGR limit for a small range of exposures. However, the PCT occurs at 0.0 GWd/MTU and at this exposure the MAPLHGR limit is less restrictive than the LHGR limit. For the limiting neutronic lattice, the LHGR of 54 of the 91 rods had initial LHGRs higher than the LHGR limit.

A characterization of the PCT sensitivity to fuel conductivity for plants where early boiling transition is predicted to occur during the early stages of LOCA The Brunswick EPFOD LOCA break spectrum shows boiling transition occurs at approximately 9 seconds in the limiting two-loop operation analysis and at approximately 1 second in the limiting single-loop analysis. These calculations used a conservatively high initial stored energy. The highest stored energy occurs at early exposure when the density of the pellet reaches a maximum. The maximum stored energy for Brunswick ATRIUM 10XM fuel was calculated with RODEX2 at 2 GWd/MTU. The time of boiling transition was calculated based on a conservatively high stored energy.

The change in stored energy from U02 thermal conductivity degradation is of primary concern for the LOCA analyses; however, it is important to note that maximum stored energy would occur between 0-15 GWd/MTU. Within this range, maximum stored energy occurs from pellet densification when the gap between the cladding and pellet is at its maximum. This usually occurs before 5 GWd/MTU - an exposure region that is not an issue for conductivity degradation. At later exposures where conductivity degradation is significant, the reduction in power associated with the decreasing by the MAPLHGR limit would prevent this exposure region from being limiting in terms of stored energy. The R,ELAX system and RELAX hot channel analyses are performed with stored energy determined from the earlier exposure region. As noted in Reference 27, RODEX2 has an over-prediction of fuel centerline temperature to at least 10 GWd/MTU, therefore stored energy used in the RELAX analyses is conservative.

F.3.3 Overpressurization Analyses The COTRANSA2 code is used to perform analyses to demonstrate that the reactor vessel pressure will not exceed the ASME vessel pressure limit during specified events. COTRANSA2 AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page F-7 is also used to demonstrate that the vessel pressure does not exceed the overpressure acceptance criterion for an anticipated transient without scram (ATWS) event.

Analyses using COTRANSA2 are potentially affected by U0 2 thermal conductivity degradation with exposure, as described in Section F.3.1 for AOO analyses. As discussed in Reference 27, the impact on overpressurization analysis for ATRIUM-10 fuel was assessed in two ways: using AURORA-B to assess the relative impact of using U0 2 thermal conductivity degradation with exposure; and decreasing the core average thermal conductivity input into COTRANSA2 to account for the effects of exposure. Reference 27 summarized the increase in pressure as less than a [ ] pressure rise (peak pressure - initial pressure) for the AURORA-B assessment and a pressure rise of [ ] for COTRANSA2 when the core average thermal conductivity assumed a 30% reduction. The Reference 27 evaluations concluded that the impact of U02 thermal conductivity degradation with exposure on the peak vessel pressure in overpressurization analyses was a small i~crease, the increase is less than the existing margins to the acceptance criteria.

In the ASME and ATWS overpressurization analyses analysis for Brunswick, the impact of TCD was specifically evaluated for ATRIUM 10XM fuel. This evaluation showed that if a 30%

reduction in thermal conductivity (due to increased exposure) is applied the ASME overpressure results become slightly less limiting. This is a result of the lower pellet conductivity reducing the moderator feedback and, as a consequence, the average core peak neutron flux increases faster during the event which in turn triggers an earlier scram which improves the overpressure results. No credit was taken for this slight improvement for the reported results. For ATWS analysis, the same 30% reduction in the thermal conductivity resulted in an increase of [ ]

of the pressure rise (peak pressure - initial pressure). The effects of TCD will be tracked and applied to the ASME and ATWS overpressure analyses which are performed to support each cycle of operation.

The impact c;if TCD for the EPFOD operation at Brunswick with ATRIUM 1OXM fuel is [

].

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page F-8 F.3.3.1 Responses to NRC Requests From the NRC's review of Reference 27, additional information was requested in Reference 28.

The requests and responses to the requests are provided as follows:

A comprehensive list of the identified nonconservative biases in the AREVA overpressure analysis methods The comprehensive list of items was provided in Reference 29. The biases applicable for Brunswick are summarized as follows. These biases are addressed for each cycle to ensure that the pressure limits are not exceeded.

Void-Quality Correlation: The bias is [ ] for ASME and [ ] for ATWS calculations.

Thermal Conductivity Degradation: In Reference 27 AREVA evaluated the impact of TCD for ATRIUM-10 fuel in two ways: using the AURORA-B code (Reference 30) to assess the relative impact of using U0 2 thermal conductivity with exposure qegradation; and decreasing the core average thermal conductivity input into COTRANSA2 to account for the effects of exposure degradation. It was noted that changing the U0 2 thermal conductivity model provides a conservative estimate of the impact of exposure degradation on calculated peak vessel pressure. The limiting results obtained for the plants assessed in support of Reference 27 were reported as follows. For ASME, the increase in peak reactor pressure is expected to be less than [ ] of the pressure rise (peak pressure - initial pressure). For ATWS, the increase in pressure rise was

[ ].

The effect of thermal conductivity degradation for Brunswick ASME pressurization was assessed by decreasing the core average thermal conductivity in COTRANSA2 by 30%.

The peak pressure did not increase. In the Brunswick ATWS overpressurization evaluation, the effect of TCD was accounted for by applying a [ ] increase to the calculated pressure rise. The peak pressure increased [ ].

Doppler Model Mismatch Between MICROBURN-B2 and COTRANSA2: The bias is

[ ] of the calculated pressure rise from steady-state conditions for the ASME calculation and [ ] for the A TWS calculation.

Verification that the nonconservative biases are considered in an integral sense in the safety analyses.

Reference 29 demonstrated that it is conservative to add the biases together from separate effect assessments. The integral study demonstrated a decrease in total bias pressure.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page F-9 F.3.4 Stability Analyses As summarized in Reference 27, the computer codes STAIF and RAMONAS-FA are used in stability analyses. Both of these codes have fuel models that include U0 2 thermal conductivity degradation with exposure. Therefore, there is no impact on AREVA stability analyses.

F.3.5 Fire Event Analyses The analyses to demonstrate compliance with Appendix R criteria are performed using the

LOCA analysis codes. For these analyses, *the calculated PCT is much lower than for LOCA analyses. As detailed in Section F.3.2 for the LOCA analyses, the impact of U02 thermal conductivity degradation with exposure has only a small impact on calculated PCT. Like the Brunswick LOCA analyses, the fire protection analyses are limiting at BOL. Therefore, the conclusions from these analyses would not be affected by U02 thermal conductivity degradation with exposure.

AREVA Inc.

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page F-10 Table F-1 Impact of TCD on PCT AREVA Inc. (

Controlled Document Applicability of AREVA BWR ANP-3108NP Methods to Brunswick Extended Revision 1 Power Flow Operating Domain Page F-11 Table F-2 Brunswick EPFOD Overpressurization Biases and Results AREVA Inc.

BSEP 16-0056 Enclosure 14 AREVA NP Affidavit Regarding Withholding ANP-3108P, Revision 1, Applicability of AREVA BWR Methods to Brunswick Extended Power Flow Operating Domain, July 2015

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) SS.

CITY OF LYNCHBURG )

1. My name is Gayle Elliott. I am Manager, Product Licensing, for AREVA Inc.

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

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

AREVA to ensure the proper application of these criteria.

3. I am familiar with the AREVA information contained in Licensing Report ANP.:.3108P, Revision 1, entitled, "Applicability of AREVA BWR Methods to Brunswick Extended Power Flow Operating Domain," dated May 2015 and referred to herein as "Document."

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

4. This Document contains information of a proprietary and confidential nature I

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

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

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

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

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

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

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

/

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

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

\

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

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

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

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

SUBSCRIBED before me this _.J____;;*tJ_*__

day of ~ , 2015.

41~£%)/

Danita R. Kidd NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 12/31/16 Reg.# 205569

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BSEP 16-0056, ANP-3108(NP), Revision 1, Applicability of Areva BWR Methods to Brunswick Extended Power Flow Operating Domain.

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