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Attachment 8 - ANP-3338NP, Revision 1, Applicability of Areva BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate. (Non-Proprietary)
	ML15251A383
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Site:	Dresden, Quad Cities
Issue date:	08/31/2015
From:	; AREVA
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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ML15251A394	List: ML15251A383; ML15251A384; ML15251A385; ML15251A386; ML15251A387; ML15251A406; RS-15-237, /Quad Cities Nuclear Power Station, Units 1 and 2 - Supplemental Information in Support of Request for License Amendment Regarding Transition to Areva Fuel; RS-15-237, Dresden Nuclear Power Station, Units 2 and 3/Quad Cities Nuclear Power Station, Units 1 and 2 - Supplemental Information in Support of Request for License Amendment Regarding Transition to Areva Fuel;
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	RS-15-237 ANP-3338NP, Rev 1
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Attachment 8 Applicability of AREVA Methodology (Non-Proprietary Version)

Conrholled Document A

Controlled Document AREVA Inc.

Controlled Document AN P-3338N P Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Pagei Nature of Changes Item Page Description and Justification 1.Pages iv, B-5, Adjustment to Proprietary marks.

D-14, E-l, E-2, F-4, F-5, F-6, and G-1 AREVA Inc.

Controlled Document AN P-3338N P Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page ii Contents 1 Introduction.................................................................................... 1-1 2 Overview ...................................................................................... 2-1 3 Thermal Hydraulic Analysis .................................................................. 3-1 4 AREVA CHFICPR Correlations.............................................................. 4-1 5 Safety Limit MCPR ........................................................................... 5-1 6 Mechanical Limits Methodology ............................................................. 6-I 7 Core Neutronics .............................................................................. 7-1 7.1 Shutdown Margin .................................................................... 7-I 7.2 LHGR Monitoring of Advanced Fuel Designs ...................................... 7-3 7.3 Bypass Boiling........................................................................ 7-3 7.4 Normal Operation .................................................................... 7-4 8 Transient Analysis ............................................................................ 8-1 9 LOCA Analysis................................................................................ 9-1 10 Stability Analysis ............................................................................ 10-1 10.1 Linear Stability ...................................................................... 10-1 10.2 DIVOM ............................................................................... 10-1 11 ATWS ........................................................................................ 11-1 11.1 ATWS General...................................................................... 11-1 11.2 Void Quality Correlation Bias ...................................................... 11-1 11.3 ATWS Containment Heatup / Long-Term Evaluation ............................ 11-2 12 Summary..................................................................................... 12-I 13 References................................................................................... 13-I Appendix A. Application of AREVA Methodology for Mixed Cores........................... A-I A. 1 Discussion.................................................................................... A-I Appendix B. Void-Quality Correlations.......................................................... B-i B.1 ARE VA Void Quality Correlations.......................................................... B-I B.2 Void Quality CorrelationUncertainties..................................................... B-4 B.3 Biasing of the Void-Quality Correlation..................................................... B-5 AREVA Inc.

ControIled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page iii B.4 Void-Quality Correlation Uncertainty Summary ........................................... B-7 Appendix C. Neutronic Methods ................................................................ C-I

0. 1 Cross Section Representation ............................................................. C-1 C.2 Applicability of Uncertainties................................................................0C-4

0. 3 Fuel Cycle Comparisons................................................................... 0C-8 0.3.1 Bypass Voidingq..............................................................................0C-9 0.3.2 Fuel Assembly Desigqn..................................................................... C-Il 0.4 Gamma Scans ............................................................................. C-Il Appendix D. Transient Methods ................................................................ D-1

0. 1 COTRANSA2................................................................................ D-1 D.1.1 Conservatism ................................................................................ D-1 0.1.2 COTRANSA2 Cross Section Representation.............................................. D-2 D.2 XCOBRA-T................................................................................. D-10 0.2.1 Axial Geometry Changqes..................................................................0D-10 0.2.2 Power.......................................................................................0D-12 0.2.3 Default Models............................................................................. D-13 0.2.4 Bounds Checkingq..........................................................................0D-14 0.2.5 Heat Transfer Correlations................................................................0D-14 0.2.6 Axial Power Shape ........................................................................ 0D-15 0.2.7 Thermal Mechanical Performance........................................................0D-16 Appendix E. LOCA Modifications............................................................... E-1 E. 1 L OCA Analysis .............................................................................. E-1 E.1.1. Radiation View Factors ..................................................................... E-1 E.1.2. [ ]..................... E-2 E.1.3. Thermal Conductivity Degqradation ......................................................... E-3 AREVA Inc.

ControIled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page iv Appendix F. Fuel Conductivity Degradation ................................................... F-I F. I Introduction.................................................................................. F-I F.2 Disposition of Licensing Safety Analysis for Dresden and Quad Cities ATRIUM IOXM Fuel ........................................................................ F-I F.3 Assessment of Analyses for Dresden and Quad Cities Operations...................... F-2 F.3.I Anticipated Operational Occurrence Analyses ............................................ F-2 F.3.2 Loss of Coolant Accident Analyses ........................................................ F-4 F.3.2.1 Resp~onses to NRC Requests .............................................................. F-6 F.3.30Overpressurization Analyses ............................................................... F-7 F.3.3.1 Responses to NRC Requests .............................................................. F-7 F.3.4 Stability Analyses......................... ................................................... F-8 Appendix G. [

.................................................................. G-1 AREVA Inc.

Controfled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page v Tables Table 2-I ARE VA Licensing Topical Reports ................................................. 2-2 Table 4-I SPCB Application to OPTIMA2 Statistics.......................................... 4-3 Table 7-I CASMO-4/MICROBURN-B2 Operating Experience .............................. 7-5 Table 7-2 Dresden and Quad Cities Target Cold Critical Eigenvalue ....................... 7-5 Table 11-1 [

] ............................................................... 11-3 Table B-i ARE VA Multi-Rod Void Fraction Validation Database............................ B-9 Table B-2 Void Sensitivity Results ........................................................... B-l0 Table C-I KWU-S Gamma Scan Benchmark Results from EMF-2158(P)(A) ............. C-13 Table C-2 Comparison of CASMO-4 and MCNP results for ATRIUM-I10 Design ............................................................................. C-13 Table D-I Bounds Checking................................................................... D-17 Figures Figure 1-1 Dresden and Quad Cities Power Flow Operating Map............................ 1-2 Figure 3-1 Comparison of KATHY Two-Phase Pressure Drop and Void Fraction Test Matrices and Typical Dresden and Quad Cities Reactor Conditions.................................................................. 3-2 Figure 4-1 [ ] ........... 4-2 Figure 4-2 SPCB Critical Power Versus Westinghouse Critical Power for OPTIMA2 ............................................................................ 4-3 Figure 5-1 Assembly Power Distribution for Limiting Case in Safety Limit MCPR Analysis...................................................................... 5-4 Figure 7-1 Dresden and Quad Cities Cold Critical Data ...................................... 7-6 Figure 7-2 Quad Cities Cold Critical Data Used For Cold Target Determination........................................................................ 7-6 Figure 7-3 MICROBURN-B2 Multi-Channel Average Bypass Void Distribution from a Dresden and Quad Cities Equilibrium Cycle Design ...................... 7-7 Figure 7-4 MICROBURN-B2 Multi-Channel Exit Bypass Void Distribution from a Dresden and Quad Cities Equilibrium Cycle Design ...................... 7-8 Figure 7-5 MICROBURN-B2 Multi-Channel Bypass Void at an LPRM Location from a Dresden and Quad Cities Equilibrium Cycle Design................................................................................ 7-9 Figure 7-6 Maximum Assembly Power in a Dresden and Quad Cities Design ............. 7-10 AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page vi Figure 7-7 Maximum Exit Void Fraction in a Dresden and Quad Cities Design............ 7-10 Figure 7-8 Dresden and Quad Cities Design Axial Profile of Power and Void Fraction ............................................................................. 7-11 Figure 7-9 Dresden and Quad Cities Design Nodal Void Fraction Histogram.............. 7-11 Figure 8-1 Typical Hydraulic Benchmarks to KATHY Transient Simulations (time to dryout)....................................................................... 8-2 Figure B-I Validation of [ ] using FRIGG-2 and FRIGG-3 Void Data ................................................................................ B-Il Figure B-2 Validation of [ ] using ATRIUM-I0 and ATRIUM I0XM Void Data .......................................................................... B-Il Figure B-3 Validation of Ohkawa-Lahey using FRIGG-2 and FRIGG-3 Void Data ................................................................................ B-12 Figure 8-4 Validation of Ohkawa-Lahey using ATRIUM-10 and ATRIUM 10XM Void Data ................................................................... B-12 Figure 8-5 Modified Void Fraction Correlation Comparison to ATRIUM-I0 Test Data........................................................................... B-I13 Figure B-6 [ ] Void Fraction Comparison to ATRIUM-Ia Test Data........................................................................... B-I13 Figure C-I Microscopic Thermal Cross Section of U-235 from Base Depletion and Branches ........................... ,........................................... C-14 Figure C-2 Microscopic Fast Cross Section of U-235 from Base Depletion and Branches........................................................................... C-14 Figure C-3 Microscopic Thermal Cross Section of U-235 at Beginning of Life............. C-I5 Figure C-4 Microscopic Fast Cross Section of U-235 at Beginning of Life................. C-15 Figure C-5 Microscopic Thermal Cross Section of U-235 Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions ............. C-16 Figure C-6 Microscopic Fast Cross Section of U-235 Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions ............. C-16 Figure C-7 Macroscopic Diffusion Coefficient (Fast and Thermal) Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions ........................................................................... C-1 7 Figure C-8 Microscopic Thermal Cross Section of U-235 at 70 GWd/MTU ................ C-I18 Figure C-9 Quadratic Interpolation Illustration of Microscopic Thermal Cross Section of U-235................................................................. .. C-I19 Figure C-I10 Illustration of Final Quadratic Interpolation for Microscopic Thermal Cross Section of U-235 ................................................. C-I19 Figure C-I I Comparison of k-infinity from MICROBURN-B2 Interpolation Process with CASMO-4 Calculations at Intermediate Void Fractions of 0.2, 0.6 and 0.9...................................................... C-20 AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page vii Figure C-12 Comparison of k-infinity from MICROBURN-B2 Interpolation Process with CASMO-4 Calculations at 0.4 Historical Void Fractions and 0.9 Instantaneous Void Fraction..................................0C-20 Figure C-13 Delta k-infinity from MICROBURN-B2 Interpolation Process with CASMO-4 Calculations at 0.4 Historical Void Fraction and 0.9 Instantaneous Void Fraction......................................................0C-21 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 ............................ 0C-21 Figure 0-15 EMF-21 58(P)(A) TIP Statistics by Axial Level ................................... 0C-22 Figure C-17 EMF-2158(P)(A) 2-D TIP Statistics for C-Lattice Plants vs. Core Average Void Fraction ............................................................ 0C-23 Figure C-18 EMF-2158(P)(A) 2-D TIP Statistics for C-Lattice Plants vs. Core Power/Flow Ratio..................................................................0C-23 Figure 0-19 EMF-2158(P)(A) 2-0 TIP Statistics for 0-Lattice Plants vs. Core Power .............................................................................. 0C-24 Figure 0-20 EMF-21 58(P)(A) 2-0 TIP Statistics for 0-Lattice Plants vs. Core Average Void Fraction ............................................................ 0C-24 Figure 0-21 EMF-21 58(P)(A) 2-0 TIP Statistics for 0-Lattice Plants vs. Core Power/Flow Ratio ......................... ........................................ 0C-25 Figure 0-22 EMF-2158(P)(A) 3-0 TIP Statistics for C-Lattice Plants vs. Core Power .............................................................................. 0C-25 Figure 0-23 EMF-21 58(P)(A) 3-D TIP Statistics for C-Lattice Plants vs. Core Average Void Fraction ............................................................ 0C-26 Figure 0-24 EMF-2158(P)(A) 3-D TIP Statistics for C-Lattice Plants vs. Core Power/Flow Ratio..................................................................0C-26 Figure 0-25 EMF-21 58(P)(A) 3-0 TIP Statistics for 0-Lattice Plants vs. Core Power...............................................................................0C-27 Figure 0-26 EMF-21 58(P)(A) 3-D TIP Statistics for 0-Lattice Plants vs. Core Average Void Fraction ............................................................ 0C-27 Figure C-27 EMF-21 58(P)(A) 3-0 TIP Statistics for 0-Lattice Plants vs. Core Power/Flow Ratio..................................................................0C-28 Figure 0-28 Quad Cities Unit 1 Pin by Pin Gamma Scan Results............................0C-28 Figure C-29 Maximum Assembly Power in Topical Report EMF-21 58(P)(A)................0C-29 Figure 0-30 Maximum Exit Void Fraction in Topical Report EMF-21 58(P)(A) ............... 0C-29 Figure C-31 Maximum Assembly Power Observed from Recent Operating Experience.........................................................................0C-30 Figure C-32 Void Fractions Observed from Recent Operating Experience..................0C-30 Figure 0-33 Axial Power and Void Profile Observed from Recent Design Experience.........................................................................0C-31 AREVA Inc.

Cuontrolled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page viii Figure C-34 Nodal Void Fraction Histogram Observed from Recent Design Experience........................................................................... C-31 Figure D-I Comparison of Scram Bank Worth for [

J ..................................................................... D-18 Figure E-I [ ............... E-4 AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods toPower the Dresden Revision 1 Cities Reactors Operating at Extended Uprate and Quad Page ix Nomenclature Definition Acronym ACE ARE VAs advanced critical power correlation [

]

BWR Boiling Water Reactor CHF Critical Heat Flux CPR Critical Power Ratio DIVOM Delta-over-Initial CPR Versus Oscillation Magnitude EPU Extended Power Uprate 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 NRC Nuclear Regulatory Commission OLMCPR Operating Limit Minimum Critical Power Ratio SLMCPR Safety Limit Minimum Critical Power Ratio SPCB AREVA (formerly Siemens Power Corporation) critical power correlation WREM Water Reactor Evaluation Model AREVA Inc.

ControlUed Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page 1-1 I Introduction This document reviews the AREVA approved licensing methodologies to demonstrate that they are applicable to licensing and operation of the Dresden and Quad Cities Nuclear Power Generating Stations with the introduction of ATRIUMTM* 10XM fuel. This confirmation of the applicability of AREVA methods to these plants includes the current operating domain as defined by the power/flow operating map in Figure 1-1.

All four BWR/3s (Dresden Units 2 and 3 and Quad Cities Units 1 and 2) are essentially the same since the core operational conditions (number of assemblies, rated thermal power, rated core flow), modeled geometryt, safety system performance and ECCS parameters are either identical or have minor differences. The most significant difference between the units is the core loadings and corresponding core designs. The impact of the differences in core designs between units and cycles is addressed in the cycle specific reload report for each unit. Minor differences between the plants and units do not impact the application of AREVA's methodology as presented in this document.

ATRIUM is a trademark of AREVA.

t Dresden Units 2 and 3 have an Isolation Condenser, but this is not credited in any licensing calculations.

AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 1-2 ID Drosdoe *nd ouad CitiCs 100% E9U Po.-er = 2957 12t1 110 1000 Core Flow, - 98.0 lHlbe/he A: .13.0% F 23.0% F C, 300.08 F 95.30 F D

0; 100.0% F 100.0% F E: 100.0% F 1I08.08 F 9, 27.01 F I108.0% F Dresden 0; 54.3% 9 35.5% F 00: 18.8% P /33.7% F Qusd Citise 0: 54.20 F 35.01 F 70 0: 38.8% F 30.64 F I,

25O 3O 0 10

30 J3 50 60 70 GO 90 iO* 10 120 Core E'!ow (5)

Figure 1-1 Dresden and Quad Cities Power Flow Operating Map AREVA Inc.

Controlled Document AN P-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page 2-1 2 Overview The first step in determining the applicability of current licensing methods to Dresden and Quad Cities operating conditions was a review of AREVA BWR topical reports listed in Table 2-1 to identify SER restrictions. This review identified that there are no SER restrictions on core power level or core flow for the AREVA topical reports. The review also indicated that there are no SER restrictions on the parameters most impacted by operation at EPU power level at any core flow rate: 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 Dresden and Quad Cities operating conditions to determine any challenges to the validity of the models. Operating margin for variations in the reactor power within the constraints of the power/flow map is mitigated to a large extent by variations in the limiting assembly radial power factor. A 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 insensitive to the core thermal power.

Based on these fundamental characteristics each of the major analysis domains (thermal-hydraulics, mechanics, core neutronics, transient analysis, LOCA and stability) are assessed to determine any challenges to application. A description of the application of AREVA methodology to a mixed core is provided in Appendix A.

AREVA Inc.

Controfled Document ANP-3338NP Revision 1 Applicability of AREVA Cities Reactors BWR Operating Methods toPower at Extended the Dresden Uprate and Quad Page 2-2 Table 2-1 ARE VA Licensing Topical Reports Document Number Document Title XN-NF-79-56(P)(A) Revision 1 "Gadolinia Fuel Properties for LWR Fuel Safety Evaluation,"

and Supplement I Exxon Nuclear Company, November 1981 XN-75-32(P)(A) Supplements 1 "Computational Procedure for Evaluating Fuel Rod Bowing,"

through 4 Exxon Nuclear Company, October 1983. (Base document not approved.)

XN-NF-81-58(P)(A) Revision 2 "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation and Supplements 1 and 2 Model," Exxon Nuclear Company, March 1984 XN-NF-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) Revision I "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company, September 1986 XN-NF-85-74(P)(A) "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model" Exxon Nuclear Company, August 1986 XN-NF-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 I "Generic Mechanical Design Criteria for BWR Fuel Designs,"

and Supplement I Advanced Nuclear Fuels Corporation, May 1995 ANF-90-82(P)(A) Revision I "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 I (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 August 2005 BAW-10247PA Revision 0 "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA NP, February 2008 XN-NF-80-1 9(P)(A) Volume I "Exxon Nuclear Methodology for Boiling Water Reactors -

and Supplements 1 and 2 Neutronic Methods for Design and Analysis," Exxon Nuclear Company, March 1983 AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Ooeratina at Extended Power Uorate Paae 2-3 Table 2-1 AREVA Licensing Topical Reports (Continued)

Document Number Document Title XN-NF-80-1 9(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 Frequency Domain," and Volume 2 "STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain - Code Qualification Report," Siemens Power Corporation, July 1994 EMF-2158(P)(A) Revision 0 "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/

MICROBURN-B2," Siemens Power Corporation, October 1999 EMF-CC-074(P)(A) Volume 4, "BWR Stability Analysis Assessment of STAIF with Input from Revision 0 MICROBURN-B2," Siemens Power Corporation, August 2000 BAW-1 0255PA Revision 2 "Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code," AREVA NP, May 2008 EMF-3028P-A Volume 2 "RAMONA5-FA: A Computer Program for BWR Transient Revision 4 Analysis in the Time Domain Volume 2: Theory Manual," AREVA NP, May, 2013 XN-NF-79-59(P)(A) "Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies," Exxon Nuclear Company, November 1983 XN-NF-80-1 9(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, August 2000 EMF-2209(P)(A) Revision 3 "SPCB Critical Power Correlation," AREVA NP, September 2009 ANP-10298PA Revision 1 "ACE/ATRIUM 10XM Critical Power Correlation," AREVA, March 2014 ANP-1 0307PA Revision 0 "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," AREVA NP, June 2011 AREVA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Ooeratina at Extended Power Uorate Paae 2-4 Table 2-1 AREVA Licensing Topical Reports (Continued)

Document Number Document Title 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-80-1 9(P)(A) Volumes 2, "Exxon Nuclear Methodology for Boiling Water Reactors: EXEM 2A, 2B and 2C BWR ECCS Evaluation Model," Exxon Nuclear Company, September 1982 XN-NF-82-07(P)(A) Revision 1 "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, November 1982 XN-NF-84-1 05(P)(A) Volume 1 "XCOBRA-T: A Computer Code for BWR Transient Thermal-and Volume 1 Supplements 1 Hydraulic Core Analysis," Exxon Nuclear Company, February and 2 1987 ANF-91 3(P)(A) Volume 1 "COTRANSA2: A Computer Program for Boiling Water Reactor Revision I and Volume 1 Transient Analyses," Advanced Nuclear Fuels Corporation, Supplements 2, 3 and 4 August 1990 ANF-91-048(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 TM-10: Appendix K Spray Heat Transfer Coefficients,"

Siemens Power Corporation, September 2000 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.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 3-1 3 Thermal Hydraulic Analysis 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 reactor conditions in Figure 3-1.

This figure illustrates that the test conditions bound typical assembly conditions as well as anticipated operation for Dresden and Quad Cities. The data is based upon the projected operating conditions for the Dresden and Quad Cities reactors. Figure 3-1 also shows that the key physical phenomena (e.g. fluid quality and assembly flows) for Dresden and Quad Cities operating conditions are consistent with current reactor experience.

This similarity of assembly conditions is further enforced in AREVA analysis methodologies by the imposition of SPCB and ACE critical power correlation limits and, therefore, core designs must remain within the same parameter space. Since the bundle operating conditions for Dresden and Quad Cities 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 Dresden and Quad Cities operating conditions.

A more detailed discussion of the ARE VA void quality correlations is presented in Appendix B.

ARE VA Inc.

Con'rotec: Documnent ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 3-2 Figure 3-1 Comparison of KATHY Two-Phase Pressure Drop and Void Fraction Test Matrices and Typical Dresden and Quad Cities Reactor Conditions AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 4-1 4 AREVA CHF/CPR Correlations All ARE VA 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 Dresden and Quad Cities operating 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. The CPR correlation used for the ATRIUM 1OXM fuel is the ACE/ATRIUM IOXM critical power correlation and the corrective actions for when the computed conditions fall outside the approved range are provided in Reference 2.

The application of AREVA critical power correlations to co-resident fuel is governed by Reference 7. [

AREVA Inc.

Contro~ted Docum ent ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operatingi at Extended Power Uprate Page 4-2 Figure 4-1 [ I AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Paae 4-3 The overall statistics for the SPCB application to OPTIMA2 fuel is given in Table 4-I. [

Figure 4-2 shows the SPCB predicted critical power versus the Westinghouse critical power.

The data shows a reasonable prediction of critical power without obvious trends.

Table 4-1 SPCB Application to OPTIMA2 Statistics Figure 4-2 SPCB Critical Power Versus Westinghouse Critical Power for OPTIMA2 AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors. Operating at Extended Power Uprate Page 5-1 5 Safety Limit MCPR The safety limit MCPR (SLMCPR) methodology is used to determine the Technical Specification SLMCPR 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 Dresden and Quad Cities 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 Dresden and Quad Cities operating conditions is discussed in other sections of this report.

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 ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Page 5-2 Cities Reactors Operating at Extended Power Uprate

[

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

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 5-3

]

AREVA Inc.

Contro~Ied Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 5-4 7

Figure 5-1 Assembly Power Distribution for Limiting Case in Safety Limit MCPR Analysis ARE VA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 6-1 6 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.

Fuel rod power histories are generated as part of the methodology for equilibrium cycle conditions as well as cycle-specific operation. These power histories include the impact of channel bow using the same model and limitations as previously described in Section 5 (Safety Limit MCPR). 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 EPU 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 ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 7-1 7 Core Neutronics The AREVA neutronic methodologies (Reference 19) 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 Dresden and Quad Cities application.

For Dresden and Quad Cities 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 all other cores.

Detailed analysis of the neutronic methodology is presented in Appendix C. Specific applicability to Dresden and Quad Cities is addressed below.

7.1 Shutdown Margin In order to accurately determine shutdown margins during transition cycles, AREVA typically performs detailed benchmarking analyses of the three to five cycles previous to insertion of AREVA fuel in that reactor. This benchmarking is performed with the CASMO-4/MICROBURN-B2 3-D core simulator code system. Hot depletions are performed using actually operated state conditions including as-loaded core configurations, as-operated control rod patterns, and operating power, pressure, flow, and inlet subcooling. To confirm the validity of the hot depletions, comparison of eigenvalue trends and predicted versus measured TIP distributions for the benchmark cycles are performed. These results are used to establish the hot-operating target k-eff for design of the first transition cycle. All cold critical measurements taken during the benchmarking cycles are also modeled in MICROBURN-B2 by restarting from the hot cycle depletions discussed above. The results from the cold critical benchmarks are used to define a cold critical k-eff target. A typical design target is 1% Ak/k which ensures that the transition loading fuel design will support the 0.38% Ak/k technical specification cold shutdown margin requirement with additional margin to cover the uncertainty in the design target chosen based on the benchmarking results. Past ARE VA experience AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 7-2 indicates that the variation in the target cold critical k-eff when transitioning to AREVA fuel is small.

During the design of each transition cycle, shutdown margin is computed by performing restart solutions based on a shuffled core from a short window previous cycle condition. This means that the previous cycle is assumed to shutdown earlier than the nominal planned shutdown for the cycle. The short window shutdown of the previous cycle results in additional carryover reactivity for the shutdown margin analysis of the cycle being designed. Setting the gadolinia design of the fresh fuel and the loading plan to meet the design shutdown margin based on the assumed short window shutdown of the previous cycle assures that adequate shutdown margin exists for the entire cycle at the design stage. At startup, when each designed cycle reaches cold critical conditions, comparison to the predicted point of criticality to the actual point of criticality is made. High accuracy of the predicted versus actual critical eigenvalue demonstrates the validity of the shutdown margin design for that cycle.

The initial critical and any subsequent cold critical data points achieved in each transition and follow-on cycle are fed back into the cold critical eigenvalue database for the reactor unit, and the target is revised as needed for the design of the subsequent cycle. This method assures continued accuracy in predicting the cold shutdown margin as new fuel is transitioned into the reactor core during the first and second transition cycles and all subsequent cycles.

As part of the design process for developing the fuel/core design for Dresden and Quad Cities it is necessary to establish a target cold critical eigenvalue. Benchmarking of the previous Dresden and Quad Cities cycles that contained Westinghouse OPTIMA2" fuel resulted in the cold critical data presented in Figure 7-1. The target cold critical eigenvalues used for the reference cycle design focused on a subset of this data. Specifically, only the Quad Cities benchmark data was used since the reference cycle is based upon Quad Cities Unit 2 Cycle 24.

Furthermore, the eigenvalue selection concentrated more heavily on the later operating cycles which is typical of the selection process. The resulting target cold critical eigenvalues used for the Quad Cities reference design is shown in Table 7-2 and Figure 7-2. This determination of the target, together with a conservatively chosen design goal, ensures conservative Early cycles of the benchmark analysis include mixed cores containing AREVA 9X9 and GNF GE-14 fuel.

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Controlte.d Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 7-3 determination of shutdown margin for the design. A similar approach will be used for cold target determination for the initial Dresden transition cycles.

7.2 LHGR Monitoring of Advanced Fuel Designs Through various interactions between AREVA and the NRC, the NRC has requested verification that certain detailed models available with MICROBURN-B2 are utilized in the modeling of advanced fuel designs. These models include the impact of LPRM detectors (instrument tube) on the surrounding fuel rods and the impact of modeling the plenum region above the end of the heated portion of the part-length rods. The explicit LPRM model is used in the core monitoring to account for perturbations to the local peaking factors of rods surrounding the LPRM, hence rod power biases due to the presence of LPRM detectors are accounted for in the monitoring of LHGR limits. Monitoring for conformance with the operating limit LHGR will include explicit modeling of the fission gas plena in the node directly above the top of PLFR active fuel length.

This provides the confirmation that the NRC has requested.

7.3 Bypass Boiling 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 subcooling 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 bypass.

The model is conservative in that it [

]. Review of the edit of bypass channel exit void for a Dresden and Quad Cities equilibrium cycle case identified a few assemblies with minimal (< 0.005 void fraction) bypass channel exit void at a cycle exposure of 13,500 MWd/MTU. To force more boiling in the bypass the inlet subcooling was set to a value of 20.1 BTU/Ibm (compared to the typical value of 25.463 for this statepoint) at the 13,500 MWd/MTU statepoint to demonstrate the capability of this model to predict localized bypass boiling. The results are presented in Figure 7-3 through Figure 7-5.

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ControIled Document AN P-3338N P Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 7-4 Figure 7-3 presents the average void fraction for the channel bypass and Figure 7-4 presents the core exit bypass channel void fraction. One of the more significant impacts of voiding in the bypass is the impact on the LPRM reading. The average void fraction of the four channels surrounding any LPRM location is presented in Figure 7-5. Since no boiling is observed at any LPRM location for normal operating conditions, there is no impact on LPRM readings.

7.4 Normal Operation From a neutronic perspective, moderator density (void fraction) and exposure cause the greatest variation in cross sections. Reactor conditions for Dresden and Quad Cities are not significantly different from that of current experience and are bounded by the experience for the important parameters. Dresden and Quad Cities operating conditions (Figure 7-6, Figure 7-7, and Figure 7-8) can be compared to the equivalent data of the topical report EMF-2158(P)(A).

Comparison of Figure 7-6 vs. Figure C-29 and Figure 7-7 vs. Figure C-30 shows that Dresden and Quad Cities operation is within the range of the original methodology approval for assembly power and exit void fraction.

The axial profile of the power and void fraction of the limiting assembly and core average values are presented in Figure 7-8 for a Dresden and Quad Cities 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 7-9 presents a histogram of the void fraction for Dresden and Quad Cities 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 Dresden and Quad Cities operation are bounded by the data provided in the topical report EMF-2158(P)(A). Concerns about Pu production with high voids are not relevant as the isotopic validation presented in the topical report continues to be applicable to Dresden and Quad Cities operation.

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Controlled Document AN P-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors O~eratinci at Extended Power U~rate Paaie 7-5 I Y ! Y Table 7-1 CASMO-4/MICROBURN-B2 Operating Experience Ave. Approximate Bundle Peak Bundle Reactor Power, MWt (% Power, Power, Reactor Size, #FA Uprated)* MWt/FA MWtIFA A. 592 2575 (0.0) 4.4 7.2 B 592 2575 (0.0) 4.4 7.4 C 532 2292 (0.0) 4.3 7.3 D 840 3690 (0.0) 4.4 7.5 E 500 2500 (15.7) 5.0 8.0 F 444 1800 (5.9) 4.1 7.3 G 676 2928 (8.0) 4.3 7.6 H 700 3300 (9.3) 4.7 8.0 1 784 3840 (0.0) 4.9 8.1 J 624 3237 (11.9) 5.2 7.8 K 648 3600 (14.7) 5.6 8.6 L 648 2500 (10.1) 3.9 6.9 M 624 3091 (6.7) 5.0 7.7 N 800 3898 (1.7) 4.9 7.7 0 764 3489 (5.0) 4.6 7.2 P 560 2923 (20.0) 5.2 8.0 Q 764 3952 (20.0) 5.2 7.7 R*724 2957 (17.8) 4.1 6.6

Latest power uprates.

Dresden and Quad Cities Table 7-2 Dresden and Quad Cities Target Cold Critical Eigenvalue Cycle Exposure (MWdlMTU) k-eff 0.0 0.9970 6,000.0 0.9950 20,000.0 0.9950 AREVA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Pa~ae 7-6 Figure 7-1 Dresden and Quad Cities Cold Critical Data Figure 7-2 Quad Cities Cold Critical Data Used For Cold Target Determination ARE VA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors ODeratina at Extended Power Uprate Pane 7-7 Figure 7-3 MICROBURN-B2 Multi-Channel Average Bypass Void Distribution from a Dresden and Quad Cities Equilibrium Cycle Design K- J"*

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Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Paoe 7-8 Figure 7-4 MICROBURN-B2 Multi-Channel Exit Bypass Void Distribution from a Dresden and Quad Cities Equilibrium Cycle Design K ]1 AREVA Inc.

Contro~led Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page 7-9 p v I Figure 7-5 MICROBURN-B2 Multi-Channel Bypass Void at an LPRM Location from a Dresden and Quad Cities Equilibrium Cycle Design AREVA Inc.

Controlled Document ANP-3338NP Revision I Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page 7-10 Figure 7-6 Maximum Assembly Power in a Dresden and Quad Cities Design Figure 7-7 Maximum Exit Void Fraction in a Dresden and Quad Cities Design AREVA Inc,

C; ontrc1/2d Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Paqe 7-11

, . v Figure 7-8 Dresden and Quad Cities Design Axial Profile of Power and Void Fraction Figure 7-9 Dresden and Quad Cities Design Nodal Void Fraction Histogram AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 8-1 8 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 KATHY 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 KATHY (Figure 8-1) illustrate that the transient hydraulic solution and application of ACE (AREVA Critical 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. The models associated with predicting the pressure wave are general and have no limitation within the range of variation associated with Dresden and Quad Cities EPU operation.

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 EPU operation are expected to be within the current operating experience, the transient analysis methods remain valid.

Appendix 0 provides additional information on the transient cross section treatment in the COTRANSA2 transient simulator for both EPU and pre-EPU reactor conditions.

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

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Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page 8-2 Figure 8-1 Typical Hydraulic Benchmarks to KATHY Transient Simulations (time to dryout)

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Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 9-1 9 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 EPU because the core is still constrained by the same thermal limits. The parameters associated with EPU that may impact LOCA results at each of the core flow rates in the operating 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 EPU 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 EPU conditions.

The potential impact of the EPU on LOCA analyses is thus primarily associated with the increase in decay heat levels in the core. 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 of 1.2 is applied to the fission product decay heat throughout the LOCA scenario. The models used for decay heat calculations are valid for EPU.

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

Independent of EPU, additional modifications have been made to the approved EXEM BWR-2000 LOCA methodology to more accurately model advanced fuel designs and to address AREVA Inc.

Controlled Docu ment ANP-3338NP Revision 1 Applicability of AREVA BWR Methods toPower the Dresden Cities Reactors Op)eratinci at Extended Uprate and Quad Pacie 9-2 regulatory concerns with the approved methodology. These modifications are described in Appendix E. Appendix F summarizes the assessment of thermal conductivity degradation in the Dresden and Quad Cities LOCA analyses.

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Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 10-1 10 Stability Analysis 10.1 Linear Stability The flatter radial power profile characteristic of EPU 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 and does not rely on a correlation to protect the regional mode.

STAlF has been bench marked 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, STAlF is benchmarked to both global and regional reactor data, and 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 EPU core designs are reflected in the stability analysis.

10.2 DIVOM RAMONA5-FA has been generically approved to calculate DIVOM for EPU operation (Reference 5 and 6).

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Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 11-1 11 ATWS 11.1 A TWS General The COTRANSA2 computer code is the primary code used for the ATWS overpressurization analysis. The ATWS overpressurization event is not used to establish operating limits for critical power; therefore, the critical power correlation(s) pressure limit is not a factor in the analysis.

Dryout conditions are not expected to occur for the core average channel that is modeled in COTRANSA2 for the ATWS overpressurization analysis. Dryout might occur in the limiting (high power) channels of the core during the ATWS event; however, these channels are not modeled in COTRANSA2 analyses. For the ATWS overpressurization analysis, ignoring dryout for the hot channels is conservative in that it maximizes the heat transferred to the coolant and results in a higher calculated pressure.

The ATWS event is not limiting relative to acceptance criteria identified in 10 CFR 50.46. The core remains covered and adequately cooled during the event. Following the initial power increase during the pressurization phase, the core returns to natural circulation conditions after the recirculation pumps trip and fuel cladding temperatures are maintained at acceptably low levels. The ATWS event is significantly less limiting than the loss of coolant accident relative to 10 CFR 50.46 acceptance criteria.

11.2 Void Quality Correlation Bias AREVA performs cycle-specific ATWS analyses of the short-term reactor vessel peak pressure using the COTRANSA2 computer code. The ATWS peak pressure calculation is a core-wide pressurization event that is sensitive to similar phenomenon as other pressurization transients.

Bundle design is included in the development of input for the coupled neutronic and thermal-hydraulic COTRANSA2 core model. Important inputs to the COTRANSA2 system model are biased in a conservative direction.

The AREVA analysis methods and the correlations used by the methods are applicable for both pre-EPU and EPU conditions. The transient analysis methodology is a deterministic, bounding approach that contains sufficient conservatism to offset biases and uncertainties in individual phenomena. The void-quality correlation is robust as discussed in Appendix B for past and ARE VA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 11-2 present fuel designs. For future fuel designs the void-quality correlation would be reviewed for applicability, which may involve additional verification and validation.

A sensitivity study was performed for the limiting ATWS pressurization event for a proposed BWR/4 cycle with EPU to assess the bias between the ATRIUM-I10 test data and the void-quality correlation. The event was a pressure regulator failure-open (PRFO), which is a depressurization event, followed by pressurization due to main steam line isolation valve (MSIV) closure. The neutronics input included the impact of the fuel depleted with the changes in the void-quality correlation. To remove the bias in the M~ICROBURN-B2 neutronics input, the [

] void-quality correlation was modified. To address the bias in the Ohkawa-Lahey void-quality correlation for the COTRANSA2 code, the void-quality relationship was changed to a

[ ]. Additionally, the sensitivity study was repeated without depleting the fuel with the changes in the void-quality correlation (the change in the void-quality correlation was instantaneous at the exposure of interest).

The reference ATWS case had a peak vessel pressure of 1477 psig. The change in the void-quality correlations resulted in a [ ] increase in the peak vessel pressure. The results for an instantaneous change in the void-quality correlation showed the same impact.

A study was also performed for the ASME overpressure event for the same BVVR/4 cycle with EPU. The event was the MSIV closure. The change in the void-quality correlations resulted in a[ ] increase in the peak vessel pressure.

The impact of a change in the bias of the void-quality correlations on peak pressure is expected to be more than offset by the model conservatisms. Until quantitative values of the conservatisms can be demonstrated, AREVA has imposed that a [ ] increase to the peak vessel pressure for the ATIWS overpressure analysis and a [ ] increase to the peak vessel pressure for the ASME overpressure analysis be included in analyses results.

11.3 A TW4S Containment Heatup /Long-Term Evaluation Fuel design differences may impact the power and pressure excursion experienced during the ATWS event. This in turn may impact the amount of steam discharged to the suppression pool and containment. [

]

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Controfled Document AN P-3338N P Revision 1 Applicability of AREVA Cities Reactors BWR Operating Methods toPower at Extended the Dresden Uprate and Quad Page 11-3 C

Table 11-1 [ ]

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Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 12-1 12 Summary This review concluded that there are no SER restrictions on ARE VA methodology that impact the transition to AREVA fuel at Dresden and Quad Cities. Since the core and assembly conditions for the Dresden and Quad Cities units are bounded by core and assembly conditions of other plants for which the methodology was benchmarked, the AREVA methodology (including uncertainties) remains applicable for conditions at the Dresden and Quad Cities Units.

More specifically:

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

b) The calculational and measurement uncertainties applied in Dresden and Quad Cities applications are valid because the conditions for Dresden and Quad Cities application are bounded by existing core and assembly conditions for which the AREVA methodology was bench marked.

c) The assessment database and uncertainty of models used to simulate the plant response at Dresden and Quad Cities conditions are bounded by core and assembly conditions for which the AREVA methodology was benchmarked.

Sections 4, 5, 7, 9, and 11 summarize methodology or application enhancements specifically for Dresden and Quad Cities:

a) SPCB CPR correlation applied to OPTIMA2 b) [

c) [

d) LOCA radiation view factors and [

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Controlled Documre-nt ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page 12-2 e) Thermal conductivity degradation f) Void quality correlation biases AREVA Inc.

Controfled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 13-1 13 References

1. ANP-1 0307PA Revision 0, ARE VA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.

2. ANP-10298PA Revision 1, ACE/ATRIUM IOXM CriticalPower Correlation,AREVA, March 2014.

3. BAW-1 0247PA Revision 0, Realistic Thermal-MechanicalFuel Rod Methodology for Boiling Water Reactors, ARE VA NP, February 2008.

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

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

6. BAW-1 0255PA Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008.

7. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation'sCriticalPower Correlationsto Co-Resident Fuel, Siemens Power Corporation, August 2000.

8. 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.

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

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

Nuclear Engineering and Design, 61, 1980.

13. 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.

14. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.

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Controlled Document ANP-3338N P Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 13-2

15. XN-NF-84-1 05(P)(A) Volume 1 Supplement 4, XCOBRA-T: A Computer Code ForBWR 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-1 154, 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-1 007, May 1968.

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

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

21. 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.

22. 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:1 1:096, September 22, 2011.

23. Letter, T.J. McGinty (NRC) to P. Salas (ARE VA)," Response to AREVA, Inc. (AREVA)

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

24. 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.

25. 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.

26. 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.

27. ANP-1 0300P, Revision 0, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Transientand Accident Scenarios, (ML100040158) December 2009.

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Controfled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 13-3

28. XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.

29. Correspondence, R.A. Copeland (Siemens) to R.C. Jones (NRC), "ATRIUM-10 Presentations," RAC:95:080, May 4, 1995 (38-9091703-000).

30. EMF-2209(P)(A) Revision 3, SPCB CriticalPower Correlation, AREVA NP, September 2009.

31. R.T. Lahey Jr. and F.J. Moody, The Thermal-Hydraulics of a Boiling Water Nuclear Reactor, Second Edition, American Nuclear Society, Hinsdale, IL, 1993.

32. Y. Aounallah, "Boiling Suppression in Convective Flow," Proceedings of ICAPP '04, Pittsburgh, PA, June 13-17, 2004, Paper 4251.

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Controlled Document AN P-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page A-I Appendix A. Application of ARE VA Methodology for Mixed Cores A.I 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

[ ] ina 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 core design and nuclear safety analyses, the neutronic cross-section data is developed for each fuel type in the core using CASMO-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 CASMO-4 and geometric data appropriate for the fuel design.

Fuel assembly thermal mechanical limits for both AREVA and co-resident fuel are verified and monitored for each mixed core designed by AREVA. The thermal mechanical limits established by the co-resident fuel vendor continue to be applicable for mixed (transition) cores. The thermal mechanical limits (steady-state and transient) for the co-resident fuel are provided to AREVA by the utility. 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.

AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page A-2 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 ARE VA critical power test facility.

Consistent with Reference 7 [ I the SPCB critical power correlation will be used for monitoring OPTIMA2 fuel present in transition cycles of operation at Dresden and Quad Cities. The critical power ratio (CPR) correlation used for the ATRIUM 10XM fuel is the ACE/ATRIUM I0XM 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.

Analyses performed to determine the safety limit MCPR explicitly address mixed core effects.

Each fuel type present in the core is explicitly modeled using appropriate geometric data, thermal hydraulic characteristics, and power distribution information (from CASMO-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 8) 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 evaluated 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 such as MICROBURN-B2 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 modeled. The change in CPR between the initial and final condition after the transient is AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page A-3 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 RAMONA5-FA and STAlF 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. MAPLHGR limits are established using each fuel vendor's LOCA methodology. For ATRIUM 10XM fuel 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 core. 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.

AREVA Inc.

Controfled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page B-I Appendix B. Void-Quality Correlations B. I A REVA Void Quality Correlations The Zuber-Findlay drift flux model (Reference 9) 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 10).

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 1 1) 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 AGO transient and accident analysis, and loss of coolant accident (only at specified junctions) methods use the Ohkawa-Lahey void correlation (Reference 12). 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 ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate 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 [

1. 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-lO test assembly are reported in the public domain in Reference 13.

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 prototypic ATRIUM-10 and ATRIUM 10XM test assemblies, as described below.

The characteristics of the ARE VA multi-rod void fraction validation database are listed in Table B-I.

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.

Controfled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page B-3 fuel designs. Also the reported instrument uncertainty for these tests is provided in Table B-I 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 Quad Cities operating 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-I and Figure B-2 provide comparisons of predicted versus measured void fractions for both the FRIGG data and 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 both the FRIGG data and 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 observations 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 AN P-3338N P Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page B-4 uncertainty in the prediction of nuclear parameters at EPU conditions within the nuclear methods when applied to the Dresden and Quad Cities.

B.2 Void Quality Correlation Uncertainties The AREVA analysis methods and the correlations used by the methods are applicable for modern fuel designs in both pre-EPU and EPU conditions. The approach for addressing the void-quality correlation bias and uncertainties remains unchanged and is applicable for EPU conditions.

The OLMCPR is determined based on the safety limit MCPR (SLMCPR) methodology and the transient analysis (ACPR) 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 analysis 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 analysis 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.

Controlted Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate 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.

B.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-I through Figure B-4.

For one of these studies, the transient ACPR 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-i10 of +_0.05void. 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 [ ] results compared to the ATRIUM-10 test data. This approach created equivalent void fractions as the [ ] correlation modifications.

The thermal hydraulic methodology incorporates 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.

ControIled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate 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 ACPR.

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 void-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-la void fraction data results in a very small increase in ACPR, 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 OLMGPR; 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 ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page B-7 the conservatism introduced by the 110% multiplier is 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 monitoring predictions of power will likely show a bias relative to measured data.

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

B.4 Void-Quality Correlation Uncertainty Summary Integral power is a parameter obtainable from test measurements that is directly related to ACPR 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 ACPR. 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 NRC, 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 EPU operation at Dresden and Quad Cities is provided below.

AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page B-8 The COTRANSA2 methodology results in predicted power increases that are bounding

( J 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 10OXM design is demonstrated 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 EPU operation of the ATRIUM I0XM fuel at Dresden and Quad Cities.

AREVA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operatingq at Extended Power Uprate Page B-9 Table B-I AREVA Multi-Rod Void Fraction Validation Database S FRIGG-2 FRIGG-3 ATRIUM I0XM (Reference 18) (Reference 16 & 17) ARU-0KTY KATHY Axial Power Shape ]uniform ~ uniform[][

Radial Power Peaking uniform mild peaking((

Bundle Design circular array with 36 circular array with 36 ((

rods + central thimble rods + central thimble ))

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

Inlet Subcooling (0 F) 4.3 to 40.3 4.1 to 54.7 [ ]J [

Mass Flow Rate (Ibm/s) [ [

(Based on mass flux assuming 14.3 to 31.0 10.1 to 42.5 ATRIUM-b1 inlet area)

[. [

Equilibrium Quality at 006t .0 008t .3 Measurement Plane (fraction) -006t0.3 -. 58o030

[

Max Void at Measurement 082084 Plane (fraction)0.2084 Reported Instrument Uncertainty (fraction) 0.2L.1

[ [

Number of Data 27 tests, 174 points 39 tests, 157 points AREVA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page B-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)

AC PR 0.305 0.307 0.321 0.305 SLMCPR 1.09 1.09 1.09 1.09 ASLMCPR NA -0.001 -0.002 +0.002 OLMCPR 1.395 1.396 1.409 1.397 AREVA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page B-11 Figure B Validation of [. ] using FRIGG-2 and FRIGG-3 Void Data Figure B-2 Validation of [: ] using ATRIUM-10 and ATRIUM 10XM Void Data AREVA Inc.

AN P-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page B-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.

(..,ul I O Dcu*, rneni':J'*"'h' ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page B-13 Figure B-5 Modified Void Fraction Correlation Comparison to ATRIUM-l0 Test Data Figure B-6

[ ] Void Fraction Comparison to ATRIUM-bO Test Data ARE VA Inc.

Controlled Document AN P-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page C-I Appendix C. Neutronic Methods C.1 Cross Section Representation CASMO-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.

MICROBURN-B2 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.

MICROBURN-B2 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 CASMO-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 comparisons to the operating limits.

AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page 0-2 MICROBURN-B2 determines the nodal macroscopic cross sections by summing the contribution of the various nuclides.

ZxCo,fl,E, R) =XZNi §(p,l,E, R)+/-+A~b(p,fl, E,R) j=1 where:

Ex = nodal macroscopic cross section A~x= background nodal macroscopic cross section (D, El, Za, E,)

Ni = nodal number density of nuclide "1"

-x = microscopic cross section of nuclide "i" I = total number of explicitly modeled nuclides p = nodal instantaneous coolant density LI = nodal spectral history E = nodal exposure R = control fraction The functional representations of o-x and A comefrm3videltocauainswh CASMO-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 are shown in Figure C-I and Figure 0-2 for a representative lattice and each lattice will have specific cross section data.

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 CASMO-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 0-3 and Figure 0-4.

Detailed CASMO-4 calculations confirm that a quadratic fit accurately represents the cross sections and diffusion coefficient 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 AREVA Inc.

Controlled Document AN P-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page C-3

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

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-I0 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 FigureC-I0.

The results of this process for all isotopes and all cross sections in MICROBURN-B2 were compared for an independent CASMO-4 calculation with continuous operation at 20, 60 and 90% void and are presented in Figure C-I11. 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-I13) to show the results for various instantaneous void fractions.

Void fraction has been used for the previous illustrations; however MICROBURN-B2 uses water density rather than void fraction in order to account for pressure changes as well as subcooled density changes. This transformation does not change the basic behavior as water density is proportional to void fraction. MICROBURN-B2 uses spectra! history rather than void history in order to account for other spectral influences due to actual core conditions (fuel loading, control AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page C-4 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.

(, 'ef where:

Teff = Effective Doppler Fuel Temperature Tref = Reference Doppler Fuel Temperature o-x = Microscopic Cross Section (fast and thermal absorption) of nuclide "i" N1 = Density of nuclide "i" The partial derivatives are determined from branch calculations performed with CASMO-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 CASMO-4 are state of the art. The methods used in MICROBURN-B2 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 AREVA Inc.

Control~ed Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page C-S there is less correlation between nearby bundle powers, thus, there would be a larger bundle power distribution uncertainty.

The EMF-21 58(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-IS 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 2D standard deviations for "C" and "D" lattice plants are presented in Figure C-16 through Figure C-21, while the 3D standard deviations are presented in Figure C-22 through Figure C-27. This evaluation of the data indicates that there is no significant trend in the data associated with these plant parameters.

Parameters such as fuel density, part length rods, active fuel length, fuel pellet diameter and fuel cladding diameter are all inputs to the methodology. The methodology explicitly accounts for such changes in design parameters. The changes in these parameters for the ATRIUM 10 XM fuel are insignificant relative to the changes that have been included in the validation of the methodology that demonstrate the methodology's capability to evaluate these parameters. Fuel designs including 7X7, 8X8, 9X9 and I0X10 with corresponding changes in pellet and cladding diameters were presented in Reference 19.

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:

AREVA Inc.

ControIled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate 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 I 0x10) 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 TIP's as well as comparison of calculated and measured Ba-140 density distribution. The accuracy of the MICROBURN-B2 models was further validated with detailed axial pin by pin gamma scan measurements of 9X9-1 and ATRIUM-l0 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). Reference 19 Figures 8.18 through 8.31 showed very good comparisons between the calculated and measured relative Ba-I140 density distributions of actual irradiated assemblies for both radial and axial values.

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 above 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 AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page C-7 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.

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 [ J 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 Lal40 concentration. Based upon the time of shutdown and the time of the gamma scan the Bal40 relative distribution at the time of shutdown is determined. This Bal40 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 Ba 140 relative density data. The assembly gamma scan data was taken at Quad Cities after 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 AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page C-8 re-arranged by axial level in Table C-I, 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 modeled in MICROBURN-B2. There is no indication of reduced accuracy at the higher void fractions. Details of the gamma scan process is described in Section C.4.

CASMO-4 and MCNP calculations have been performed to compare the fission rate distribution statistics to Table 2-I 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 CASMO-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-21 58(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.

0.3 Fuel Cycle Comparisons AREVA has reviewed 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 have 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-34 present the parameters shown below at the point of the highest exit void fraction (at 9336 MWd/MTU cycle exposure) in cycle core design for a BWR-6 reactor (high power density plant) with ATRIUM-I10 fuel. Additional details 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 and peak assembly axial power profile ARE VA Inc.

Controgled Docu ment ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page C-9

Core average void axial profile

Axial void profile of the peak assembly

Histogram of the nodal void fractions in core 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 cycles using NRC approved methodologies consistent with the current processes.

The AREVA methodology [ ] the reactivity coefficients that are used in the transient analysis. Conservatisms in the methodology are used to produce results that bound the uncertainties in the reactivity coefficients. Data presented in these referenced figures indicate that there are no significant differences between EPU and non- EPU 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 simulator, LOCA and stability codes as [ ].

The steady-state core simulator, MICROBURN-B2, 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 MICROBURN-B2 state-point specific solution for bypass flow rate and [

is then used as initial conditions in the transient and LOCA analyses. When the reactor operates on high rod-lines at low flow conditions, the in-channel pressure drop decreases to a AREVA Inc.

Controlled Document AN P-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page C-10 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 of core stability concerns) the neutronic feedback of bypass voiding [

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 hot channel flow rate and a more bottom-peaked power 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. These effects are small compared to the general conditions of low flow and high power that dominate the stability regime. Never the less, AREVA stability methods directly model these phenomena to assure that the core stability is accurately predicted.

CASMO-4 has the capability to specify the density of the moderator in the bypass and in-channel water rods, [

I.

AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page C-11 Significant bypass voiding is not encountered during full power, steady-state EPU operation for Dresden and Quad Cities 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 Dresden and Quad Cities, a 100% power!/85% flow statepoint (120% of the original licensed thermal power) was assessed even though this statepoint is outside of the power/flow operating map to cover a wider range of flow. 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 subcooling 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 EPU operation, neither mechanical nor thermal hydraulic. The maximum allowed enrichment level of any fuel pellet is 4.95 wt% U-235.

All new and spent fuel at Dresden and Quad Cities 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. The customer has chosen another vendor to perform the required criticality safety analyses so AREVA methods are not utilized.

C,4 Gamma Scans The gamma scanning process for irradiated fuel uses germanium semi-conductor detectors for gamma radiation energy spectral analysis. Gamma rays deposit their energy in the germanium and produce free electrons and holes (vacancies where the electrons were located in the crystalline germanium). The amount of charge collected is correlated with the amount of energy AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 0-12 deposited in the detector and therefore with the energy of the gamma ray that caused it. The detectors are used with single channel analyzers to sort the pulses according to pulse height.

This means that if multiple gamma-ray energies are being analyzed simultaneously, the germanium detector will separate them cleanly. A single-channel analyzer (SCA) uses two discriminators. The discriminators are called upper and lower level discriminators. Pulses from the amplifier are fed to the analyzer, and if the pulse height falls between the lower and upper discriminators the usual logic is to allow such a pulse to be recorded (counted). The voltage levels of the two discriminators are adjustable so that the gap between them corresponds to a group of pulse heights within a fixed energy interval. Even though the gamma rays from a specific decay transition are of a discrete energy, there is a statistical spread of pulses coming from the detector and associated electronics so that the gap between the discriminators must be large enough to include most of such pulses. By varying the voltage levels of each of the discriminators, it is possible to measure gamma rays of different energies.

Power measurements for irradiated fuel target the gamma spectrum associated with Lanthanum (La) 140. Lal40 is a decay product of Barium (Ba) 140 which is a direct fission product. The half-life of 8a140 is 12.8 days and the half-life of La140 is 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. La14O activity is, therefore, directly related to the density of Bal40. The Ba140 density is representative of the integrated fissions over the last 25 days due to its short half-life. Gamma scan measurements are taken shortly after reactor shutdown (within 25 days) before the Bal40 decays to undetectable levels.

Gamma scan measurements may be performed on individual fuel rods removed from assemblies using a high-purity germanium (HPGe) detector and an underwater collimator assembly or on entire fuel assemblies where the collimator has a broad opening to capture the gamma radiation from all of the pins in the assembly.

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 CASMO-4 pin nuclide inventory and MICROBURN-B2 nodal nuclide inventory is used.

AREVA Inc.

Controlled Document ANP-3338NP Revision I Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page C-13 Table C-I KWU-S Gamma Scan Benchmark Results from EMF-2158(P)(A)

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

ControI~ed Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page C-14 A10B-4340L-15G70 U235 Thermal Absorption 280270 S270 -*

-'e--0.0 Vl / 0.0 VH

--U- 0.4 VI/ 0.0 VH 260 -,

-0.8 Vl/0.0 VH

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220

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--I-- 0.0 VI / 0.8 VH

--0.4 VI /0.8 VH-

-- 0.8 VI/ 0.8 VH 180 0 10 20 30 40 50 60 0 80 Exposure (GWdlMTU)

Figure C-I Microscopic Thermal Cross Section of U-235 from Base Depletion and Branches A10B-4340L-15G70 U235 Fast Absorption

--0.4 VI / 0.0 VH 0.8 VIIO .0OVH

--)-- 0.0 VI/10. VHI

°= 12.0

~-4---0.0 --0.4 V / 0,4VH V I /0.0 VI-

-0.8 V / 0.4 VH

-I-0.0VL /0.8 VH-

-- 0.4 VI / 0.8 VH 11.5 --0.8 VI/0.8 VI-11.0 10.5 10 20 30 40 50 60 70 80 Exposure (GWdIMTU)

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

Controfled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors O~eratin~i at Extended Power Uprate Pa~qe C-15 I v

w BOL A10B-4340L-15G70 U235 Thermal Cross Sections 200 E

CASMO-4 Data

-- Quadratic Fit

.,

CASMO-4 Data

-- Quadratic Fit 0- 180 170

Fission 160 0 0 20 30 40 50 50 70 80 50 100 Veid Fraction (%)

Figure C-3 Microscopic Thermal Cross Section of U-235 at Beginning of Life BOL A10B-4340L-15G70 U235 Fast Cross Sections 12.0 -- - - - - -

11.0

CASMO-4 Data

--Quadratic Fit S10.0.

& CASMD-4 Data anl --- Quadratic Fit 2

(0 9.0.

-*-------_Fission 8.0.

7.0.

0 10 20 30 40 50 60 70 80 80 100 Veld Fraction (%)

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

Controlled Documenit ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operatin~a at Extended Power Uprate Pane 0-16 BOL A10B-4245L-14G70 U-235 MIcroscopic Cross Sections (Thermal) 200*

195 '* ' 6 t90 *k-*Absorption

185.

o_ 180-

Sig-A2. (CASMO-4)

-- Quadratic Fit (0,40,80) aI75

Sig-F2 (CASMO-4)

,0

-- = Quadratic Fit (0,40,50)

'.170 Fission *-*

155.

150 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.9 0.9 Void Frction 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 Microscopic Cross Sections (Fast) 12.0 -Z==---

" *" *, . Absorption E t 11.0

Sig-A1 (CASMO-4) ]

--Quadratic Fit (0,40,80)/

o 1. A- Sb-Fl (CASMO-4) l 2 --Quadratic Fit (0,40,80)1 8.0-Fission 7.0-0 0.1 0.2 0.3 5.4 0.5 0.6 0.7 0.8 0.5 Void Fraction Figure C-6 Microscopic Fast Cross Section of U-235 Comparison of Quadratic Fit with Explicit Calculations at Various Void Fractions AREVA Inc.

Controfled DocLument ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 0-17 BOL A10B-4254L-14G70 Macroscopic Diffusion Coefficients 2,5o000-2.0000) 1.5000

.

D3-1(CASMO-4)

- - Quadratic(0-40-80)

-*

D-2 (CASMO-4)

-- Quadratic (0-40-80)

- 1,0000*

0.5000 -I1---*"

0.0000 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 Void 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 ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors ODeratinci at Extended Power Uorate Pa~ae C-18 r v i v Cross Section (barns) 7060 50 90 80io*

4 Instantaneous VdFrcon(%) 201 * .' I ""o :* ;l-istorical Void Fraction

, .,9 .* - V

-o "

0 o0 > d

.5 6 Figure C-8 Microscopic Thermal Cross Section of U-235 at 70 GWdlMTU AREVA Inc.

Controfled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Op)eratino at Extended Power Uorate Paae 0-19 v D A10B-4340L-15G70 U235 Thermal Absorption at 70 GWd/MTU

o270 * * "

c

0.0 inst Void (CASMO-4)

-0.0 inst Void (quadi)

N 0.4 Inst Void (CASMO-4)

--0.4 Inst Void (quad)

0.8 Inst Void (CASMO-4

--0.8 Inst Void (quad) 230 -- Lookup at 0.00 VH 220 "

210 ______________

I 0

0 20 30 40 50 I

60 70 80 90 100 Histarioal Void Fraction Figure C-9 Quadratic Interpolation Illustration of Microscopic Thermal Cross Section of U-235 At0B-4340L-tSG70 U235 Thermal Absorption at 70 GWdIMTU 250 ____________

aE

Intersocted points 5-245 --0.6 Void History (quad)r

-40.7 Inst Void o= -UResultant Value 235 230 , ,

0 10 20 30 40 50 60 70 80 90 100 Instantaneous Void Fraction Figure C-1O Illustration of Final Quadratic Interpolation for Microscopic Thermal Cross Section of U-235 AREVA Inc.

Controlled Doc~ ien~t ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page C-20 Figure C-lI Comparison of k-infinity from MICROBURN-B2 Interpolation Process with CASMO-4 Calculations at Intermediate Void Fractions of 0.2, 0.6 and 0.9 Figure C-12 Comparison of k-infinity from MICROBURN-B2 Interpolation Processwi CASMVO-4 Calculations at 0.4 Historical Void Fractions and 0.9 Instantaneous Void Fraction ARE VA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page C-21 Figure C-13 Delta k-infinity from MICROBURN-B2 Interpolation Process with CASMO-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 AN P-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page C-22 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.

ControIled Doc iment AN P-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operatinq at Extended Power U*)rate Page 0-23

v W 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 Documenit ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page C-24 Figure 0-19 EMF-2158(P)(A) 2-D TIP Statistics for D-Lattice Plants vs. Core Power Figure C-20 EMF-2158(P)(A) 2-D TIP Statistics for D-Lattice Plants vs.

Core Average Void Fraction AREVA Inc.

CorntroledFDocument ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page C-25 Figure C-21 EIMF-2158(P)(A) 2-0 TIP Statistics for D-Lattice Plants vs.

Core PowerlFlow Ratio Figure C-22 EMF-2158(P)(A) 3-D TIP Statistics for C-Lattice Plants vs. Core Power AREVA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page C-26 Figure C-23 EMF-2158(P)(A) 3-D TIP Statistics for C-Lattice Plants vs.

Core Average Void Fraction Figure C-24 EMF-21 58(P)(A) 3-D TIP Statistics for C-Lattice Plants vs.

Core PowerlFlow Ratio AREVA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page 0-27 Figure C-25 EMF-2158(P)(A) 3-D TIP Statistics for D-Lattice Plants vs. Core Power Figure C-26 EMF-2158(P)(A) 3-D TIP Statistics for D-Lattice Plants vs.

Core Average Void Fraction AREVA Inc.

Contro~led Dock ment ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Page 0-28 Cities Reactors Ooeratina at Extended Power Uprate I v ,

Core PowerlFlow Ratio Figure C-28 Quad Cities Unit I Pin by Pin Gamma Scan Results AREVA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page C-29 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)

AREVA Inc.

Controiled Docu~ment AN P-3338N P Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page C-30 Figure C-31 Maximum Assembly Power Observed from Recent Operating Experience Figure C-32 Void Fractions Observed from Recent Operating Experience AREVA Inc.

C ontrolled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page C-31 Figure C-33 Axial Power and Void Profile Observed from Recent Design Experience Figure C-34 Nodal Void Fraction Histogram Observed from Recent Design Experience AREVA Inc.

Controlled Docu ment ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page D-1 Appendix D. Transient Methods D.1 CO TRANSA2 D.1.1 Conservatism Integral power is a parameter obtainable from test measurements that is directly related to ACPR and provides a means to assess code uncertainty by increasing heat flux during the event. 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 ACPR. 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 NRC, 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 measure integral power conservative for all of the Peach Bottom turbine trip tests.

COTRANSA2 (Reference 8) was developed and approved as a replacement for COTRANSA in the AREVA thermal limits methodology (Reference 28). Initially it was not planned to use the 110% integral power multiplier with the COTRANSA2 methodology. COTRANSA2 predictions of integral power were conservative for all Peach Bottom turbine trip tests. The minimum conservatism was [ ] and the mean of the predicted to measured integral power was

[ ]1. The comparisons to the Peach Bottom turbine trip tests demonstrated that the 110% integral power multiplier was not needed for COTRANSA2. However, because the thermal limits methodology that was approved independently of COTRANSA2 included discussion of the 110% integral power multiplier, the use of the multiplier was retained for COTRANSA2 licensing calculations. With the 110% multiplier, the COTRANSA2 predicted to measured mean integral power is [ ] for the Peach Bottom turbine trip tests. Applying a [ ] integral power multiplier provides an OLMCPR conservatism of AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page D-2

[ ]. The 110% integral power multiplier is just one part of the conservatism in the COTRANSA2 methodology and application process that covers methodology uncertainties.

The 110% integral thermal power multiplier is applied to the output of COTRANSA2 that is used as the input to XCOBRA-T; therefore, the 110% integral power multiplier is included in the transient analyses. Important input parameters are biased in a conservative direction in licensing calculations. For Technical Specification (TS) controlled input parameters, the biasing is either the limiting value allowable by TS, or an analytical limit that is beyond the limiting value allowable by TS. If a particular equipment out-of-service is applicable to a particular transient event, the transient analysis is performed with the limiting plant configuration for the allowable equipment out-of-service.

D.1.2 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-B2 calculations are performed. These successive calculations are:

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Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Page D-3 Cities Reactors Operatingq at Extended Power Uprate AREVA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page D-4 The 1% energy group diffusion equation in steady-state can be written as V. D1V'P1 -Ia

+ 1-j =0 The first term is a leakage. This equation is integrated over the cylindrical node depicted in the following figure.

H H

H AREVA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability Cities Reactorsof AREVA BWR Operatincq Methods toPower at Extended the Dresden Uprate and Quad Pa~ie D-5 r w p v The leakage term is approximated as:

3 2D1 iD 1 1~j(d~,i-d 1 ,1 ) A j~ D + D1,j ) HV where D1,, = D for plane of interest D1i = D for the nodes adjacent to the plane of interest

.*1,i = flux in the plane of interest

lj = flux in the regions adjacent to the plane of interest A = surface area between nodes iand j H = distance between nodes i and nodes j V = node volume AREVA Inc.

Controlled Document ANP-3338NP Revision I Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page D-6 ARE VA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page D-7 ARE VA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page D-8 AREVA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page D-9 7

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Controlled Document ANP-3338NP Revision 1 Applicability of AREVA Cities Reactors BWR Operating Methods toPower at Extended the Dresden Uprate and Quad Page 0-10 D.2 XCOBRA- T D.2.1 Axial Geometry Changes XCOBRA-T calculates the fuel rod surface heat flux using a fuel rod heat conduction model, the power generated in the fuel rod, and the fluid conditions at the surface of the rod. The power generated in the fuel rod is described in Reference 14 Section 2.5.5. The power generated in each axial section of a fuel rod is calculated using Equation 2.130 from Reference 14. Although Reference 14 states that Equation 2.130 is calculated for each axial node, the equation itself does not denote which variables are axially dependent. Because the equation is for each axial AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 0-11 node, the variables for heat generation rate, axial peaking factor, and number of rods are axial dependent. At the time Reference 14 was prepared, the number of rods at each axial plane was a constant for the fuel designs being supplied. For the ATRIUM-10 and ATRIUM 10XM fuel design with part-length fuel rods (PLFRs), the number of rods became axial dependent and the code was modified to make application of Equation 2.130 correct and consistent with the N RC-approved Reference 14. For application to current fuel designs, a better definition of the variable Nr in Equation 2.130 would be "number of heated rods per assembly at the axial plane" (italic indicates added text).

For bundles with part-length fuel rods (PLFRs), the rod heat flux calculation begins by computing the time-dependent heat flux generation rate at each axial section in the fuel rod.

The updated equation, corresponding to Equation 2.130 of Reference 14 is:

q"(t) - P(t) 1 (ff + fc)FriFiF 2Z7rDndi LNaNn where P(t) = transient reactor power ff = fraction of power produced in the fuel fc = fraction of power produced in the cladding Na = total number of assemblies in the core N, = total number of heated rods for type i assembly at the axial plane F* = radial peaking factor of type i assembly F1, = local peaking factor of type i assembly Fa = axial peaking factor at the axial plane Do = fuel rod diameter of type i assembly L = axial heated length This equation differs from that in Reference 14 by replacing the initial reactor power in the denominator with *. In addition, the variable definitions have been modified to identify that the total number of heated rods is dependent on both the assembly type and axial elevation and the definition of L has been corrected to the axial heated length of the assembly. This equation is substituted into equations 2.129a and 2.129b in Section 2.5.5 of Reference 14 to define the volumetric heat deposition rate for the fuel pellet and cladding, respectively. This volumetric heat deposition rate is used in the right-hand side of equation 2.85 of Reference 14 to iteratively solve the transient heat conduction equation and the hydraulic conservation equations for the AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page D-12 new time step temperatures and surface heat flux. The heat flux is introduced into the channel energy equation (2.2 of Reference 14) through the term q'. This linear heat deposition rate is a summation of the energy added by direct energy deposition and surface heat flux:

= P(t)faa01 FtriFa + Hsrf ( TNadesT--Tfluid )"2. Drod,i* Nri }Ni where fo = fraction of power produced in the coolant Hsurf = film heat transfer coefficient at the axial plane TNodesT = cladding surface temperature at the axial plane Tf,,*d = fluid temperature at the axial plane N, = number of fuel assemblies in channeli In addition to axially varying number of heated rods, proper modeling of PLFRs also requires axial variations in the active flow area, the heated perimeter, and the wetted perimeter and these parameters are now defined as axially dependent quantities in AREVA methods.

Consequently, all references to these parameters or parameters derived from the basic geometry data in the approved topical reports should be interpreted as being axially dependent variables. The pressure drop due to the area expansion at the end of the PLFRs (or anywhere in the active flow path) is modeled using the specific volume for momentum as expressed in Equations 2.78 and 2.79 of Reference 14. For current designs, area contractions occur in the single phase region, but the coding was generalized to address area contractions in the two-phase region based on a solution of the two-phase Bernoulli equation.

D.2.2 Power The decay heat is calculated by COTRANSA2 and is included in the total core power versus time provided as a boundary condition to XCOBRA-T. The decay heat model used in COTRANSA2 is a curve fit (11 groups) to the 1973 ANS standard decay heat model. The COTRANSA2 core power boundary condition includes the decay heat contribution based on the core average power density. The decay heat power remains essentially constant during the transient. Therefore, the decay heat during the transient is primarily a function of initial power density. Application of power peaking factors (axial, radial, local) to the COTRANSA2 average power properly accounts for local decay heat in the XCOBRA-T hot channel analysis.

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C;ontrolled Document AN P-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page D-13 Gamma smearing does not affect the XCOBRA-T hot channel calculation. The hot channel calculation models an average fuel rod (the average is not affected by flattening of the distribution). The calculation process for determining the peak transient LHGR is equivalent [

] and is not dependent of the actual rod local peaking factor.

The total core power calculated by COTRANSA2 is distributed between the fuel rod, the active channel coolant, and the core bypass coolant. The fraction deposited in each component is based on fuel type specific calculations performed with the CASMO computer code. Power fractions calculated by CASMO are used in XCOBRA-T.

D.2.3 Default Models The Dresden and Quad Cities transient analyses used the default models of XCOBRA-T. The default models include Levy subcooled boiling model, the Martinelli-Nelson two phase friction multipliers, the two phase component loss multiplier, and the heated wall viscosity correction model. [

] as discussed with the NRC on May 4, 1995 (Reference 29).

Thermo-dynamic properties from the ASME steam tables were used. The code provides a message if the default models are not used. Per AREVA's licensing analyses requirements, use of default models is required.

The Martinelli-Nelson two phase friction multiplier has been confirmed to be applicable in the annular flow regime by verifying the AREVA hydraulic models against two-phase full-scale heated bundle tests in the KATHY test facility in Karlstein, Germany. The range of assembly conditions at EPU are bounded by the tested two-phase flow conditions. Many of the tested conditions are in the annular flow regime.

The Levy subcooled boiling model does not directly predict void fraction in subcooled boiling.

Instead, the model predicts a critical subcooling that defines the onset of boiling. The critical subcooling is used in conjunction with a profile fit model to determine the local flow quality that accounts for the presence of subcooled boiling. The local flow quality is then used in the Ohkawa-Lahey correlation to predict the void fraction in subcooled boiling.

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Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page 0-14 D.2.4 Bounds Checking Bounds checking is provided in the XCOBRA-T coding to ensure the conditions provided to the CPR correlations are within the correlation limits. Should any of the condition limits be violated, the behavior will be as specified in Section 2.6 of Reference 30 and Section 5.13 of Reference

2. [

With respect to the remaining parameters, the behavior for transient calculations is summarized in Table D-1.

The out-of-bounds corrections affect the [ ] used in the evaluation of the transient LHGR. Therefore, the corrections do not impact the evaluation of the thermal-mechanical performance.

The critical power calculations for Dresden and Quad Cities fuel are made with the ACE and SPCB critical power correlations. The range of applicability of these parameters is sufficiently

broad to cover the ranges of conditions encountered during the licensing calculations.

Correlation bounds checking is incorporated in the XCOBRA-T critical power calculations. The bounds checking routine does not allow a calculation outside the range of applicability of these parameters except as described in References 2 and 30.

The transient code, XCOBRA-T, evaluates the Reynolds number for each node for each step of the calculation. If the flow becomes negative at any node, the code stops the calculation.

D.2.5 Heat Transfer Correlations The thermocouples used for measuring temperature data in full scale critical power tests are

[

] measure heat transfer coefficients associated with pre-CHF heat transfer in the range of mass and heat fluxes associated with BWRs. As a result, no relevant qualification studies of the Dittus-Boelter and Thom heat transfer correlations can be performed from the test data.

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uontro°Ied Docuiment ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page D-15 As noted in Reference 31, fully developed nucleate boiling is relatively insensitive to flow rate and quality. However, "boiling suppression" may occur in high quality annular flow that provides very high heat transfer coefficients, resulting in decreasing wall temperature as the heat flux increases.

Extracted heat transfer information from experiments in a tube for a range of pressure, flow, and quality that is relevant to BWRs is reported in Reference 32. This reference shows the relative insensitivity of heat transfer coefficient to flow rate and quality and that boiling suppression does not become significant until quality reaches approximately 0.47, which is well above the range of interest to BWRs. [

1 Therefore, it is concluded that liquid entrainment and droplet redeposition does not have an impact on boiling heat transfer for flow conditions that are applicable to an operating BWR at EPU.

0.2.6 Axial Power Shape The initial axial power shape is determined from the COTRANSA2 calculation based on the cross section data for the core exposure considered in the analysis. The cross section data is obtained from the MICROBURN-B2 computer code. The MI~CROBURN-B2 calculations used to generate cross section data for COTRANSA2 licensing calculations are typically performed assuming that all control rods are fully withdrawn. Assuming all control rods are fully withdrawn results in a significant conservatism in calculated scram reactivity for exposure conditions with some control rods partially inserted.

During pressurization transients, the axial power shape shifts due to the void collapse in the top of core (void reactivity), the core flow increase in the bottom of core (void reactivity), and control rod insertion in the bottom of the core (scram reactivity). The coupled 1-D neutronic and thermal-hydraulic core model in COTRANSA2 determines a [' ]

including the impact of void collapse and scram. This [ ] is used in XCOBRA-T. The assembly [

] and the assembly radial peaking factor input to XCOBRA-T.

AREVA Inc.

ControI~ed Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page D-16 The change in axial nodal power, [

] results in a change in fuel rod surface heat flux and the energy transferred to the coolant at each axial node. Both COTRANSA2 and XCOBRA-T have fuel rod heat transfer models that determine the fuel rod surface heat flux based on the nodal power history and the coolant conditions at each axial node.

Both COTRANSA2 and XCOBRA-T have thermal-hydraulic models that are used to calculate the flow at each axial node in the core and the hot channel during the pressurization transient.

The energy equation captures the effect of changes in fuel rod surface heat flux on coolant conditions. The mass and momentum equations, with applicable correlations, are used to determine the local coolant flow rate during the pressurization transient. During the initial phase of the pressurization transient, these models predict a decrease in flow near the top of the core and an increase in flow near the bottom of the core. Note, although the flow decreases in the upper portion of the hot assembly, the assembly flow does not stagnate during the pressurization phase of an AOO or ATWS. Local fluid conditions (enthalpy and flow) calculated from the thermal-hydraulic model are used to determine local dryout conditions.

D.2.7 Thermal Mechanical Performance XCOBRA-T was used to demonstrate acceptable fuel rod thermal-mechanical performance during transients (AQOs). The fuel rod models in XCOBRA-T are consistent with RODEX2 and the fuel rod gap conductance values input to XCOBRA-T are obtained from RODEX2 analyses.

The gap conductance includes the effect of pellet geometry changes (densification, swelling, etc.).

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Controlled Document AN P-3338N P Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page D-17 Table D-1 Bounds Checking ARE VA Inc.

Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page D-18 7

m Figure D-1 Comparison of Scram Bank Worth for [

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Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page E-1 Appendix E. LOCA Modifications E.1 LOCA Analysis The AREVA LOCA methodology applied at Dresden and Quad Cities differs from the approved methodology in three aspects:

E.1.1. Radiation View Factors In the Safety Evaluation for Reference 20 the NRC approved the AREVA EXEM BWR-2000 ECCS evaluation model. The HUXY code (Reference 21) 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 21. 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.

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Contro~led Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page E-2 E.1.2. [ ]

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ControR~ed DocCument ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page E-3 I

E.1.3. 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 Dresden and Quad Cities are presented in F.3.2.

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uon'(o*c uocJen L~t ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page E-.4 r v f v Figure E-1 [ I AREVA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page F-I Appendix F. Fuel Conductivity Degradation F.1 Introduction The U.S. Nuclear Regulatory Commission (NRC) 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 NRC staff concluded that the use of the older legacy fuel models will result in predicted fuel pellet conductivties that are higher than the expected values.

This appendix summarizes the impact and treatment of fuel conductivity degradation for licensing safety analyses supporting operation at Dresden and Quad Cities.

F.2 Disposition of Licensing Safety Analysis for Dresden and Quad Cities A TRIUM IOXM Fuel RODEX2 and RODEX2A codes were approved by the NRC in the early and mid-I1980'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 an 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 Dresden and Quad Cities are performed with RODEX4 and therefore explicitly account for thermal conductivity degradation. No additional AREVA Inc.

Cont rolled Docurment ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page F-2 assessment is needed for those analyses. For thermal-hydraulic and safety analyses an evaluation is needed. The following analysis methodologies use RODEX2 and/or include a separate UO 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.

F.3 Assessment of Analyses for Dresden and Quad Cities 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 24). The white paper presented results of an extensive evaluation; for BWRs the assessments consisted primarily of ATRIUM-10 fuel.

The NRC reviewed Reference 24 and provided requests for information in Reference 25.

AREVA provided responses in Reference 26. Items relevant from References 25 and 26 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 UO2 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 AQO analyses is in the opposite 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.

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Controiled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision 1 Cities Reactors Operating at Extended Power Uprate Page F-3 The assessment of Reference 24 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 UO 2 conductivity model as used in the current licensing analyses. Therefore, the benchmarking comparisons inherently include any impact of UO2 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 27). AURORA-B is built from previous NRC approved methods. These methods include codes RODEX4, MICROBURN-B2, and S-RELAP5; UO 2 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 25). The AURORA-B sensitivity studies show that the impact of fuel thermal conductivity degradation with exposure results in a decrease in the ACPR 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 EPU operation does not change the conservatisms nor invalidate the sensitivity; therefore, the AOO methodology remains applicable for Dresden and Quad Cities. It should be noted that transient LHGR analyses are performed with the RODEX4 code for Dresden and Quad Cities ATRIUM I0XM fuel, which correctly accounts for thermal conductivity degradation.

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ControlIed Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page F-4 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 (POT) and metal water reaction (MWR), but clad rupture can have a significant impact on POT, 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 UO 2 thermal conductivity used in RELAX and HUXY.

To demonstrate that limiting LOCA calculations are not sensitive to UO 2 thermal conductivity, assessments were performed for multiple BWRs. [

Assessments of the potential impact of exposure-dependent degradation of UO 2 thermal conductivity on the fuel mechanical parameters were made using the RODEX4 computer code.

[

]

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Controlled Document ANP-3338NP Revision 1 Applicability of AREVA Cities Reactors BWR Operatinci Methods toPower at Extended the Dresden Uprate and Quad Pacqe F-5

v

y

[

] The results of these evaluations were summarized to the NRC in References 24 and 26.

The impact of TCD was incorporated in the Dresden and Quad Cities ATRIUM I0XM HUXY analysis. [

]

After the NRC approval of RODEX2, more Halden tests were performed with fuel centerline temperature monitoring. As with the RODEX4 submittal, [

The ATRIUM 10OXM PCT results with the impact of TCD will be presented in the MAPLHGR report that will be included in the Exelon Licensing Amendment Request to transition to AREVA fuel. Each cycle the MAPLHGR limit will be analyzed for any new neutronic lattice designs.

The impact of TCD will be analyzed ifwarranted by the exposure dependent PCT results for the new lattice.

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Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page F-6 F.3.2.1 .Responses to NRC Requests From the NRC's review of Reference 24, additional information was requested in Reference 25.

The information requests and responses are provided as follows:

A detailed explanation of the source of the heat transfer coefficients utilized in the HUXY calculation This request is answered in Reference 26 and this answer is applicable to Dresden and Quad Cities.

A description of how LOCA analyses are initializedin terms of power distribution; specifically, how thermal limits (such as MLHGR or OLMCPR) are considered in the initialization This request is answered in Reference 26. [

A characterizationof the PCT sensitivity to fuel conductivity for plants where early boiling transition is predicted to occur during the early stages of L OCA LOCA break spectrum analyses for ATRIUM I0XM fuel show boiling transition occurring after [ ] inthe limiting two-loop operation analysis and [

] in the limiting single-loop analysis. [

I ARE VA Inc.

Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page F-7 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 is also used to demonstrate that the vessel pressure does not exceed the overpressure acceptance criterion for an anticipated transient without scram (ATIWS) 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 24, the impact on overpressurization analysis 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 24 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 24 evaluations concluded that the impact of U0 2 thermal conductivity degradation with exposure on the peak vessel pressure in overpressurization analyses was a small increase, the increase is less than the existing margins to the acceptance criteria.

The impact of TCD will be accounted for in ASME and ATWS overpressurization analyses performed for Dresden and Quad Cities by reducing the core average thermal conductivity in COTRANSA2 to account for the effects of exposure. The reduction will be calculated based on the exposure of the fuel in the core.

F.3.3.1 Responses to NRC Requests From the NRC's review of Reference 24, additional information was requested in Reference 25.

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

A comprehensive list of the identified nonconservative biases in the ARE VA overpressure analysismethods The comprehensive list of items was provided in Reference 26. The biases applicable for Dresden and Quad Cities 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 ATrWS calculations.

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Controlled Document ANP-3338NP Applicability of AREVA BWR Methods to the Dresden and Quad Revision I Cities Reactors Operating at Extended Power Uprate Page F-8 Thermal Conductivity Degqradation: In Reference 24 AREVA evaluated the impact of TCD for ATRIUM-I10 fuel in two ways: using the AURORA-B code (Reference 27) to assess the relative impact of using UO2 thermal conductivity with exposure degradation; and decreasing the core average thermal conductivity input into COTRANSA2 to account for the effects of exposure degradation. It was noted that changing the UO 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 24 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

[ ]1.

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 ATWS calculation.

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

Reference 26 demonstrated that it is conservative to add the biases together from separate effect assessments. The integral study demonstrated a decrease in total bias pressure.

F.3.4 Stability Analyses As summarized in Reference 24, the computer codes STAIF and RAMONA5-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.

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Controlled Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page G-1 Appendix G. [ ]

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ControB~ed Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors O~eratin~i at Extended Power Uprate Pa~qe G-2 p v

w AREVA Inc.

C~ontrote.. Document ANP-3338NP Revision 1 Applicability of AREVA BWR Methods to the Dresden and Quad Cities Reactors Operating at Extended Power Uprate Page G-3 AREVA Inc.

ML15251A383

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