BSEP 07-0108, Areva Presentation Slides for Brunswick Fuel Transition License Amendment Request
| ML072950373 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/31/2007 |
| From: | AREVA NP |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| BSEP 07-0108, TAC MD4063, TAC MD4064, TSC-2006-06 | |
| Download: ML072950373 (351) | |
Text
BSEP 07-0 108 Enclosure 6 AREVA Presentation Slides for Brunswick Fuel Transition License Amendment Request (Non -proprietary Version)
AR EVA AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 I
BrunswickFuel Transition LAR NRC Meeting Richland Washington July 31 - August 2, 2007 2
Richiand, WA July31 August 2, 2007 I NRC Meeting LAR NRC Brunswick Brunswick Fuel Transition LAR Fuel Transition Meeting Richland, WA July 31 - August Z 2007 2 AREVA NP Inc. I
BrunswickFuel Transition LAR NRC Meeting
> Meeting purpose
- Provide an overview of the AREVA safety analysis methodology used to support the Brunswick transition
- Describe applicability of the AREVA methodology for Brunswick with operation at EPU conditions
- NRC review of calculation packages documenting Brunswick transition cycle analyses I AREVA NP Inc. I Brunswick Fuel TransitionLAR NRC Meeting Brunwic Richland, July 31 - August 2, 20073 FulTanstionLARNRCMeeing WNA
Brunswick Fuel Transition LAR Support Plan Requirement Product Product Details Delivery to NRC Submit LARs Letters (2) Identify TS changes January 22, 2007 Establish LAR Support Table Identify submittals April 25 Call Plan Identify schedule May 15 Meeting Identify SER Document Package existing June 21 restrictions and documentation demonstrate compliance Demonstrate Document Address mixed core and July 25 Document methodology EPU July 31 Meeting applicable for EPU Incorporate BF EPU RAI responses Submit plant-specific Fuel Transition Package Outline July 18 analyses and Report Package TH Report July 18 demonstrate LOCA Reports (2) July 18 compliance with design Neutronic Report July 25 criteria Transient Report September Support NRC Audits August 1-2 September AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August Z 2007 4 I
Agenda
" Introduction
" Safety Analysis Methodology
" Reload Core Design and Analysis Process
- Safety Analysis Methodology Overview
- Thermal-Hydraulic Methodology
" Neutronic Analysis Methodology
" Stability Analysis Methodology
- Transient Analysis Methodology
" LOCA Analysis Methodology
" EPU and Non-EPU Analysis Conditions
" Methodology Applicability for Brunswick AREVA NP Inc. L Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 5
Safety Analysis Methodology Presentation Goal
>Provide background information to facilitate discussions with NRC during LAIR review
" General licensing approach for AREVA fuel
" Reload core design and analysis process
- Overview of safety analysis methodology
- Major codes
- Calculation process
- Typical cycle-specific calculations k Brunswick Fuel Transition LAR NRC Meeting Rich/and, WA July 31 - August 2, 2007 6
Reload, Core Design and Analysis Process Mike Garrett Manager, BWR Safety Analysis 7
Transition LAR NRC Meeting LAR NRC Richland, WA July 31 - August 2, 2007 Fuel Transition Brunswick Fuel I AREVA NP Inc. I Brunswick Meeting Richland, WA July 31 - August 2, 2007 7
Reload Core LicensingApproach Transition Cycle
" AREVA currently is not an NSSS vendor (OEM) for any U.S.
" AREVA currently isthe fuel vendor for several U.S. BWRs
" Introduction of AREVA fuel requires confirmation that fuel-related and plant-related design and licensing criteria continue to be satisfied
" ARE VA licensing approach and analysis methodology was developed to support the introduction of AREVA fuel into a BWR already licensed for operation inthe U.S.
RVmPIc Brunswick Fuel Transition LAR NRC Meeting Richland, WA......-July 31 - August 2, 2007 8
Reload Core LicensingApproach Transition Cycle (continued)
" Maintain current plant licensing basis when possible
" Evaluate the introduction of AREVA fuel per the requirements of 10 CFR 50.59
" Similar to approach used for any plant change
" Similar to approach used for each reload core design (except for scope)
" Identify plant safety analyses potentially affected by a fuel or core design change
" Assess impact on potentially affected safety analyses and repeat analyses as required
- August 2, 2007 9
July31 AREVA NP Inc. Brunswick Fuel Brunswick LAR NRC Transition LAR Fuel Transition Meeting NRC Meeting Richland, WA Richiand, INA July 31 - August 2, 2007 9
Reload Core LicensingApproach Transition Cycle (continued)
" Technical Specification changes generally limited to
- References to NRC-approved methods used to determine thermal limits specified. inthe COLR
- MCPR safety limit based on AREVA methods
- Fuel design description
" COLR thermal limits are determined for the transition core based on analyses using NRC-approved methods 10 LAR NRC Meeting NRC Meeting Richlano~, WA ~July 31 - August 2, 2007 AREVA NP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition Richland, 1 31 - August Z 2007
'J uy 10 I
Major Steps in a Fuel Design Transition
" Data collection
" Develop plant and core models
" Core follow and benchmark analyses (4-5 cycles)
" Compatibility analyses
" Establish current licensing basis
" Disposition of events
" Plant transition safety analysis
" LOCA analysis
" Core monitoring system parallel operation
" Criticality analyses
" Transition cycle neutronic.d~es~ign
" Initial cycle reload licensing analyses
" Startup and core monitoring data
" Licensing support
" Training 11 Richiand, WA July31 - August 2, 2007 Meeting NRC Meeting LAR NRC I AREVA NP Inc. I Brunswick Transition LAR Fuel Transition Brunswick Fuel Richland, WA July 31 - August 2, 2007 11
Reload Core LicensingApproach
> Two steps performed as part of the transition process implement the reload core licensing approach and establish safety analysis methodology requirements
- Establish current licensing basis
" Disposition of events 12 Transition LAR Fuel Transition NRC Meeting LAR NRC Richiand, WA July31 - August 2, 2007 I AREVA NP Inc. I Brunswick Brunswick Fuel Meeting Richland, WA July 31 - August Z 2007 12
Reload Core Licensing Approach Establish Current Licensing Basis
" Licensing basis consists of all analyses performed to demonstrate that regulatory requirements are met
" Licensing basis is defined, in documents such as
- FSAR
- Technical Specifications
- Core Operating Limits Reports (COLRs)
- Technical Requirements Manual
- Cycle Reload Licensing Reports
- Extended Operating Domain (EOD) Reports (e.g. increased core flow operation)
LOCA Analysis Reports AREVA Ný-,,--I Brunswick Fuel TransitionLAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 113
Reload Core LicensingApproach Disposition of Events
" Review all event analyses in the current licensing basis
" Analyses are dispositioned as
" Not impacted by the change infuel or core design
" Bounded by the consequences of another event
- Potentially limiting -reanalyze using AREVA methodology
" Rated and off-rated conditions considered
" Results from the disposition of events define the safety analyses required for the transition cycle to address the change infuel and core design
" Disposition of events is'do'cumented in calculation notebook and QA reviewed per AREVA procedures 14 July31 - August 2, 2007 AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting NRC Meeting LAR NRC Transition LAR Richland, WA Richiand, WA July 31 - August 2, 2007 14
Transition Cycle Analyses Typical Disposition Conclusions
" Mechanical design
" Nuclear design
" Stability
" Thermal-hydraulic design
" Hydraulic compatibility
- MCPR safety limit
" MCPRf (slow flow excursion)
" ASME overpressurization
" ATWS
" Overpressurization
" Standby liquid control system AREA P nc Brunswick Fuel Transition LAR NRC Meeting Richl~nd, WA .ý July3j1 - August 2, 2007 15
Transition Cycle Analyses Typical Disposition Conclusions
" Criticality analyses
" New fuel storage
- Spent fuel storage
" Anticipated operational occurrences
" Load rejection no bypass
" Turbine trip no bypass
- Loss of feedwater heating
- Control rod withdrawal error
- Feedwater controller failure Brunswick Fuel Transition LAR NRC Meeting Rich/and, WA July 31 - August 2, 2007 16 i AREVA NP Inc.
Transition Cycle Analyses
ýTypical Disposition Conclusions
" Design basis accidents
" Control rod drop accident
" Loss-of-coolant accident
" Fuel handling accident
" Fuel assembly mislocation
" Fuel assembly misorientation
" Emergency operating procedures
- Fuel-dependent input parameters August 2, 2007 17 Rich/and, WA July 31 NRC Meeting LAR NRC -
AREVA NP 1,,. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August Z 2007 17
Reload Core Licensing Approach Follow-On Cycle
" Similar to transition core, approach but with a reduced scope
" Disposition of events for transition cycle provides basis for analyses typically performed for follow-on reload cores
" All potentially limiting events are reanalyzed or justification provided for continued applicability of previous analysis
" If plant configuration or operational changes are planned during the refueling outage, a cycle-specific disposition of events is performed and additional analyses may be required 18 Richiand, WA* July 31 August 2, 2007 NRC Meeting LAR NRC AREVA NP Inc. Brunswick Brunswick Fuel Transition LAR Fuel Transition Meeting Richland, WA. 'J 1Y ý1 7 August 2, 2007 18
Reload'Core LicensingApproach Summary
" A fuel transition is addressed as a change in the plant design basis that is evaluated relative to the current plant licensing basis
" A systematic approach (disposition of events) is used to identify the impact of the change on the plant safety analyses that constitute the current plant licensing basis
" Potentially impacted safety analyses are reanalyzed with appropriate fuel and core characteristics to ensure that all design and licensing criteria .continue to be satisfied AREVA Ný I.C. Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 19
Reload Core Design and Analysis Process Key Steps
> Several steps in the core-design and analysis process are directed towards ensuring that the planned scope, analysis methods, and input assumptions for the cycle safety analysis are valid
" Project Initialization (initial reload)
- Fuel Mechanical Design (initial, reload or design change)
" Preliminary Core Design
" Plant Parameters Document
" Fuel Design Analysis Review
" Calculation Plan
" Licensing Basis Core Design
" Safety Analyses
" Design and Licensing Reports.
" Fuel Delivery
- Startup Support AREVA NP Inc. k Brunswick Fuel Transition LAR NRC Meeting Richlond,' WA _ýJuly:31 - uut2 0720
Reload Core Design and Analysis Process Project Initialization
" A Project Initialization meeting is conducted following finalization of a new or major revision to a contact
" Purpose
" Inform Engineering and Manufacturing of contractual provisions and schedule
" Identify any unique product, material, or commercial requirements
" Establish the need for any qualification or proof-of-fa brication activities
" Any unique engineering methodology, analysis, or reporting requirements should be id~entifi~ed AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Bruswik Richla'nd,WA' Fel raniton AR RC eetng Ju-ly 31 LAugust 2, 2007 2 21
Reload Core Desi0gn and Analysis Process Plant Parameters Document
" Defines plant configuration, operating conditions, and equipment performance characteristics used inAREVA safety analyses
" Provides mechanism for utility to:
" Review and approve plant parameters used insafety analysis
" Determine when plant changes will impact safety analyses
" Notify AREVA of planned plant changes during the next refueling outage
" AREVA requests PPD updates for upcoming cycle (generally, a draft PPD with known changes is provided)
" Utility confirms or identifies PPD changes for upcoming cycle
" AREVA reviews PPD changes and performs a disposition to identify any additional analyses required
" Ensures that AREVA and :utility have a mutual agreement on the plant configuration and operation basis used in safety analyses m
RV Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 22
Reload Core Design and Analysis Process Fuel Design Analysis Review
" Primary purpose of the Fuel Design Analysis Review is to ensure that all analyses required to demonstrate compliance with design and licensing criteria are identified inthe Calculation Plan
" Review includes
" Review design and identify appropriate criteria
" Review open issues inCorrespondence Activity Tracking System
" Identify analyses require.d, to -demonstrate compliance with criteria
" Review methodology applicability-and SER restrictions
" Preliminary Calculation Plan should be available prior to Review
" For initial reload, Review should be performed after completion of licensing basis determination and disposition of events August 2, 2007 23 NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July 31 -
AREVA NP Inc. Brunswick Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 23
Reload Core Design and Analysis Process Calculation Plan
" Defines the scope of the safety analyses to be performed for a specific reload including any additional analyses required due to PPD changes
" Provides cycle-specific reference identifying analyses to be performed, associated methodology, and key assumptions
" AREVA provides draft calculation plan identifying all analyses to be performed for the cycle
" Following utility review and cpomment, final calculation plan is issued by AREVA
" Assures that the work scope and analysis bases are understood and acceptable to all parties m
RV Brunswick Fuel Transition LAR NRC Meeting Richland, WA , ,July 31 - August 2, 2007 24
Reload Core Design and Analysis Process Summary
> The AREVA core design and analysis process has procedurally controlled steps to ensure that the scope of safety analyses and applied methodology are appropriate to demonstrate that all desig~n ',andlcnigciei r satisfied for the reload ýcore .design 25 LAR NRC Transition LAR Meeting NRC Meeting Richiand, WA July 31 - August 2, 2007 Fuel Transition I AREVA NP Inc. I Brunswick Fuel Brunswick Richland, WA July 31 - August 2, 2007 25
AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August Z 2007 26 Safety Analysis Methodology Overview Mike Garrett Manager, BWR Safety Analysis 27 LAR NRC Transition LAR Meeting NRC Meeting Rich/and, WA July 31 - August 2, 2007 Fuel Transition Brunswick Fuel I AREVA NP Inc. Ik Brunswick Richland, INA July 31 - August 2, 200 7 27
Sa-fety Analysis Methodology Goals
" Perform analyses of anticipated operational occurrences (AOOs) to confirm or establish operating limits that:
- Adequately protect all fuel design criteria
- Ensure all licensing criteria are satisfied
- Promote economically efficient fuel cycles
- Provide operational flexibility
" Perform analyses of design basis accidents to confirm that results are within regulatory acceptable limits
" Perform analyses of special events to ensure regulatory requirements or industry codes are satisfied August 2, 2007 28 Richiand, WA July31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland.'WA 'July 31 - August 2, 2007 28 L
Safety Analysis Methodology
" Safety analyses include
" Anticipated operational occurrence (AQO) analyses
- Accident analyses
" Special event analyses
" Safety analysis methodology includes
- Thermal-hydraulic analysis methodology
- Neutronic analysis methodology
- Stability analysis methodology
" Transient analysis methodology
" LOCA analysis methodology July 31 August 2, 2007 29 Richiand, WA NRC Meeting LAR NRC Transition LAR -
AREVA Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 29
AQO Analyses Typical Ev-ents and Applied Methodology 0 Control rod withdrawal error Neutronic Methodology Loss of feedwater heating
- Load rejection without bypass
- Turbine trip without bypass System Transient Methodology
" Feedwater controller failure
" Recirculation flow runup Thermal-Hydraulic Methodology
" Safety limit MCPR August 2, 2007 30 Transition LAR Fuel Transition Meeting NRC Meeting Richiand, WA July 31 -
Brunswick Brunswick Fuel LAR NRC Richland, WA July 31 - August Z 2007 30 I AREVA NP Inc. I
AccidentAnalyses Typical Events and Applied Methodology
. Loss-of-coolant accident LOCA Methodology
" Control rod drop accident
" Fuel assembly loading accident Neutronic Methodology
" Fuel handling accident 31 I Transition LAR Meeting NRC Meeting LAR NRC Richland, WA July31 - August 2, 2007 AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Richland, INA July 31 - August 2, 2007 31 L
Special Analyses Typical Events and Applied Methodology
" Shutdown margin analysis
" Standby liquid control analysis Neutronics Methodology
- Stability
" ASME overpressurization analysis System Transient Methodology
- ATWS overpressurization analysis 32 LAR NRC Meeting NRC Meeting Richiand, WA July 31 - August 2, 2007 AREVA NP Inc. I Brunswick Transition LAR Fuel Transition Brunswick Fuel Richland, WA July 31 - August Z 2007 32 k
Safety Analysis Methodology Mechanical Neutronics Safety &Licensing RODEX22 CASMO-4 1 RELAX r-RAMPE IRODEX2-2A 0 1
COLAP MICROBURN-B2 HUXY STRESSIV_
Criticality Saft RODEX2 PLENU SPACE7R7 KENO / ORGE
-I COTRAN SA2 z Cn COTRAN m XCOBRA XFUEL z Core Monitoring --i XCOBRA-T:: CD, IPOWERPLEX-Ill IANNA/ STAIF Thermal-Hydraulic SAFLI M2 Stability XCOBRA STAlE I RAMONA5-FA I
- August 2, 2007 33 July31 AREVA NP Inc. Brunswick Fuel Brunswick NRC Meeting LAR NRC Transition LAR Fuel Transition Meeting Richland, WA Richiand, WA July 31 - August Z 2007 33 I,
aRVNPIc Brunswi.ck Fuel Transition LAR NRC Meeting Richland, WA. July 31 - August 2, 2007 34 Thermal-Hydraulic Methodology Darrell Carr Team Leader, BWR Safety Analysis 1
Transition LAR Fuel Transition Meeting NRC Meeting LAR NRC Richland, WA. July31 - August 2, 2007 Brunswick Fuel I AREVA NP Inc.
Brunswick Richland, WA July 31 - August 2, 2007 1
Thermal-Hydraulic Analysis Methodology
" Thermal-hydrau lic analysis methodology description
" Thermal-hydraulic compatibility
" Flow-dependent MCPR analysis
" Critical power correlation
> Safety Limit MCPR. Methodology August 2, 2007 2 Richiand, WA July 31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 2
Thermal-Hydraulic Analysis Methodology XCOBRA Computer Code Descri ptio nl XCOBRA predicts the steady-state thermal-hydraulic performance of BWR cores at various operating conditions and power distributions Use Evaluate the hydraulic compatibility of fuel designs.
Evaluate core thermal-hydraulic performance (core pressure drop, core flow distribution, bypass flow, MCPR, etc.)
Documentation XN-NIF-CC-43(P), XCOBRA Code Theory and User's Manual Acceptability XN-NIF-80-19(P)(A) Volume 3 Rev 2,Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX:
Thermal Limits Methodology Summary Description, January 1987 NRC accepts the use of XCOBRA based on the similarity of the computational models to those used in the approved code XCOBRA-T 3
Richiand, WA July 31 - August 2, 2007 AREVA NP Inc. Brunswick Brunswick Fuel Transition LAR Fuel Transition NRC Meeting LAR NRC Meeting Richland, WA July 31 - August Z 2007 3
,..XCOBRA Computer Code Major Features
" Represen ts the core as a collection of parallel flow channels
" Each flow channel can represent single or multiple fuel assemblies as well as the core bypass region
" Core flow distribution is calculated to equalize the pressure drop across each flow channel'
" Pressure drop ineach channel is determined through the use of the AREVA thermal-hydraulic methodology
" Input includes fuel assembly geometry, pressure drop coefficients, and core operating conditions
" Water rods (or channels) can be explicitly modeled
" Calculates the flow and local fluid conditions at axial locations in each channel for use in evaluating MVCPR 4
Transition LAR Meeting NRC Meeting LAR NRC Richiand, WA - - July 31 - August 2, 2007 AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Richland, WA -- ý July 31 -August 2, 200 7 4 L
XCOBRA Computer Code Physical phenomena modeled Friction pressure drop Local (irreversible) pressure drop Elevation pressure drop Acceleration pressure drop Two-phase pressure drop mult. factors Subcooled boiling Void/quality relationship Core leakage (bypass)
Local losses Inlet orifice Lower tie plate Spacers Upper tie plate Leakage (core bypass) paths Core support plate Lower tie plate flow holes Channel - lower tie plate seal Fuel support - nozzle interface August 2, 2007 5 Richland, WA July31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 5
XCOBRA Computer Code
" Empirically derived hydraulic characteristics
" Reynolds number dependent hydraulic characteristics for a number of fuel assembly pressure losses must be empirically determined
- Bare rod friction loss
- Spacer loss coefficient
- Upper tie plate loss coefficient
- Orifice/lower tie plate loss coefficient 6
Richiand, WA. July 31 August 2, 2007 NRC Meeting AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel LAR NRC Transition LAR Meeting Richland, WA,-' . Julyll - August 2, 2007 6
XCOBRA Computer Code Applications
" Thermal-hydraulic compatibility evaluations
" Flow-dependent MCPR limits
" Flow distribution
- Pressurization transient analysis
- Safety limit MCPR 7
LAR NRC Transition LAR Meeting NRC Meeting WA Richiand, WA July 31 - August 2, 2007 I AREVA NP Inc. IL Brunswick Fuel Transition Brunswick Fuel Richland, July 31 - August 2, 2007 7
ThermalI-Hydraulic Cornpa tibility
" Approved thermal-hydraulic criteria described in ANF-89-98 (P)(A) Rev 1
" Hydraulic compatibility
" Thermal margin performance
" Rod bow
" Bypass flow August 2, 2007 8 Rich/and, WA July31 NRC Meeting LAR NRC -
Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 8
Thermal-Hydraulic Cornpa tibility Criteria
> Hydraulic compatibility
- The hydraulic resistance of the reload fuel assemblies shall be sufficiently similar to the existing fuel inthe reactor such that there isno significant impact on the core flow or the flow distribution among assemblies inthe core
- (For example, the flow resistance of the AREVA fuel should not be so low as to produce an unwarranted flow penalty and associated thermal margin reduction for existing fuel inthe reactor)
Brunwic Rich/and, FulTanstionLARNRCMeeing WA July 31 - August 2, 20079 I AREVA NP Inc Brunswick Fuel Transition LAR NRC Meeting 9
Thermal-Hydraulic Cornpa tibility Criteria
> Thermal margin performance
'~Fuel assembly geometry, including spacer design and rod-to-rod local peaking, sho'.uld minimize the likelihood of boiling transition during normal reactor operation as well as during AO0s. The fuel design shall fall within the bounds of the applicable empirically based boiling transition correlation approved for AREVA reload fuel and coresident fuel. Within other applicable mechanical, neutron ic,and fuel performance constraints, the fuel design should achieve good thermal margin performance.
10 MR NRC Meeting NRC Meeting Richiand, WA July 31 - August 2, 2007 I AREVA NP Inc. I Brunswick Transition LAR Fuel Transition Brunswick Fuel Richland, WA . July 31 - August 2, 200 7 10
Thermal-Hydraulic Cornpa tibility Criteria
>Rod bow The anticipated magnitude of fuel rod bowing under irradiation shall be accounted for inestablishing thermal margin requirements
>Bypass flow The bypass flow characteristics of the reload fuel assemblies shall not differ significantly from the existing fuel in order to provide adequate flow inthe bypass region.
11 Rich/and, WA July31 August 2, 2007 NRC Meeting LAR NRC -
Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 11
Brunswick Thermal-Hydraulic CompatibilityAnalysis
>1 August 2, 2007 12 Richiand, WA* July31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA . .- July 31 - August 2, 2007 12
Thermal-Hydraulic Cornpa tibility Representative Results r
mRV Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 13
B runswick Thermal-Hydraulic Cornpa tibility Conclusion
> The approved design criteria associated with the thermal-hydraulic compatibility for Brunswick Unit 1 have been met 14 WA July 31 August 2, 2007 I AREVA NP Inc.
Brunswick Fuel Transition LAR NRC Meeting IL Brunswick Fuel Transition LAR NRC Meeting Richiand, Richland, INA July 31 - August 2, 2007 14
Flow-Dependent MCPR (MCPRf) Analysis
" MCPRf limit isestablished. to provide protection against fuel failures during a slow core flow excursion (i.e., SLMCPR is not violated during the event)
" Analysis assumes core flow increases to the maximum physically attainable value
" Limit isa function of initial core flow; a larger core flow increase (and resulting power increase) occurs from reduced core flow
" XCO BRA computer code used to calculate change inCPR 15 Richiand, WA July 31 August 2, 2007 NRC Meeting I LAR NRC Transition LAR Fuel Transition -
AREVA NP Inc. Brunswick Fuel Brunswick Meeting Richland, WA July 31 - August Z 2007 15 L
MCPRf Analysis Process r
I AREVA NP Inc.
Brunswick Fuel TransitionLAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2Z2007 1 16
MCPRf Analysis Representative Results C-AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 117
Critical Power Correlation
" Cycle-specific licensing calculations require calculation of MCPR limits
" Requirements include licensing and monitoring CPR of all assemblies inthe core, including those supplied by other fuel vendors August 2, 2007 18 Richiand, W AI July 31 NRC Meeting LAR NRC Transition LAR -~
Fuel Transition I AREVA NP Inc.
Brunswick Fuel Brunswick Meeting Richlandjyý ý __.. ý' July ý1 -7 August 2, 2007 18
SPCB Critical Power Correlation Description
" SPCB is an empirical correlation of measured critical heat flux ina fuel assembly
" The correlation predicts critical heat flux at the axial plane of interest
" The correlation is a function of:
" Pressure
- Flow
" Enthalpy
- Local peaking distribution%
" Assembly geometry (flow area, surface area, and spacer.
design)
Brunswick Fuel Transition LAR NRC Meeting Rich/and, WA July 31 - August 2, 2007 19 7- AREVA NP In
Input Requiored to Calculate MCPR Bundle Power Bundle Flow Inlet Enthalpy Core Pressure SPCB MVCPR Axial Power Shape F-eff()
()F-eff characterizes the local power peaking and flow distribution within the fuel assembly 20 WA Richland, INA July31 August 2, 2007 NRC Meeting LAR NRC -
AREVA NP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, July 31 - August 2, 2007 20
SPCB Critical Power Correlation Development Base correlation initially derived from AREVA database for ATRI UM TM -9B and ATRIUM-1O Base correlation is a function of thermodynamic parameters Non-uniform axial tests su~pport non-uniform axial correction factor Effects due to local peaking and spacer design were observed as separable and are accounted for by a correction factor 21 July 31 August 2, 2007 Richland, WA Rich/and, I AREVA NP Inc. IIL Brunswick Fuel Brunswick LAR NRC Transition LAR Fuel Transition Meeting NRC Meeting WA July 31 - August 2, 200 7 21
SPCB Critical Power Correlation Correction Factor r
J July 31 - August 2, 2007 22 AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting NRC Meeting LAR NRC Transition LAR Richland, WA Richland, WA July 31 - August 2, 2007 22
SPCB Critical Power Correlation F eff 23 AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting WA Richiand, WA Richland, 31 -- August July 31 July 2, 2007 August 2, 2007 23 L
.CPR Correlation for Coresident Fuel
" AREVA does not typically have access to CPR correlation for coresident fuel
" Approved methodology for applying AREVA critical power correlations to coresident fuel is described in EM F-2245(P)(A)
- Process can be applied to any approved AREVA critical power correlation
- Direct method
- Indirect method August 2, 2007 24 July31 Meeting NRC Meeting LAR NRC Richland, WA Rich/and, -
Brunswick Transition LAR Fuel Transition Brunswick Fuel INA July 31 - August 2, 2007 24 i AREVA NP Inc.
CPR Correlation for Coresident Fuel BrunswickApplication- Indirect Method
>1 25 LAR NRC Transition LAR Meeting NRC Meeting Richiand, WA July 31 - August 2, 2007 AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Richland, WA July 31 - August 2, 2007 25 I
MCPR Safety Limit Methodology
> NRC-Approved Topical Report
" ANF-524(P)(A) Rev 2 and Supplements 1 and 2, ANF Critical Power Methodology for Boiling Water Reactors, Advanced Nuclear Fuels Corporation, November 1990
" This report includes the MCPR calculational procedure, identifies the fuel and non-fuel related uncertainties and the statistical process used to determine a MCPR safety limit that protects 99.9% of the fuel rods inthe core from boiling transition AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland,,,WA-Ful July 31 - August 2, 2007 TanstionLARNRCMeeing 2 26
MCPR, Safety Limit Methodology
" The purpose of the safety limit MCPR"(SLMCPR) isto protect the core from boiling tran'sitio'n (B3T) during both normal operation and anticipa ted operational occurrences (AOOs)
" At least 99.9% of the fuel rods inthe core are expected to avoid BT when the minimum CPR during the transient isgreater than the SLMCPR
" The SLMCPR isdetermined by a statistical convolution of uncertainties associated with the calculation of MCPR
" The SLMCPR analysis is performed each cycle using core and fuel design specific characteristic's August 2, 2007 27 NRC Meeting Richiand, WA July31 L Transitian LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel LAR NRC Meeting Richland, INA July 31 - August 2, 2007 27
Thermal Limits Process 28 Richiand, WA July31 August 2, 2007 NRC Meeting LAR NRC -
Brunswick Fuel Transition Brunswick Fuel TransitionLAR Meeting Richland, WA July 31 - August 2, 2007 28
MCP.R Safety Limit Methodology
> Application of Methodology The cycle-specific application of the NRC-approved methodology is controlled by
" An implementing guideline' which provides instructions to the engineers who perform and review the calculation
" Automation which has been developed to perform all data manipulation between codes August 2, 2007 29 I LAR NRC Transition LAR Fuel Transition Meeting NRC Meeting Richiand, WA July 31 -
AREVA NP Inc. Brunswick Fuel Brunswick Richland, WA ' July 31 - August Z 2007 29 L
A Safety Limit MCPR Computer Codes
" MICROBURN-B2
" EMF-2158(P)(A) Rev 0
" Provides the radial peaking factor and exposure for each bundle inthe core and the core average axial power shape
" CASMO-4
" EMF-2158(P)(A) Rev 0
" Provides the local peaking factor distribution for each fuel type 0 0V
" XCOBRA
" XN-NF-80-19(P)(A) Volume 3 Rev 2
" Provides hydraulic demand curve for. each fuel type
" SLPREP
- Automation code which gathers neutronic data from MICROBURN-B32 and CASMO-4 and prepares SAFLIM2 input
" SAFLIM2
" ANF-524(P)(A) Rev 2
" EMF-2392(P), SAFLIM2 Theory, Programnmer's, and User's Manual
" Calculates the fraction of rods in boiling transition for a specified SLMCPR L 30 Richland, WA July 31 August 2, 2007 NRC Meeting LAR NRC -
Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 30
MCPR Safety Limit Methodology SLMCPR Analysis Process r
i AREVA NP Inc. Brunswick Fuel Brunswick Fuel Transition Transition LAR LAR NRC NRC Meeting Meeting Richiand, Richland, WA WA July 31 - August 2, 2007 July 31 - August 2, 2007 31 31
MCPR Safety Limit Methodology SLMCPR Analysis Process r
32 Richiand, WA July31 August 2, 2007 NRC Meeting LAR NRC IP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 32
SLMCPR Analysis Methodology Monte Carlo Technique
> A Monte Carlo analysis .is a statistical technique to determine the distribution function of a parameter that is a function of random variables
" Each random variable ischaracterized by a mean, standard deviation, and distribution function
- A random value for each input variable isselected
- The parameter of interest is calculated using the random values for the input variables
" The process is repeated a large number of times to create a probability distribution for the parameter of interest AREVA NP Inc. L Brunswick Fuel Transition LAR NRC Meeting Richland, WVA
.7 '.July.31 - August 2, 2007 33
SAFLIM2 Computer Code Major Features Convolution of uncertainties via a Monte Carlo technique Consistent with POWERPLEX CMSS calculation of MCPR Each nominal input is randomly perturbed based on its uncertainty Appropriate critical power correlation used directly to determine if a rod is in boiling transition (deterministic)
BT rods for all bundles inthe core are summed Non-parametric tolerance limits used to determine the number of BT rods with 95% confidence Explicitly accounts for channel bow New fuel designs easily accommodated Brunswick Fuel Transition LAR NRC Meeting Richland,. WAý July31l - August 2, 2007 34 i AREVA NP Inc. I
S1%AFLIM2 Computer Code Calculation Procedure
" Initialization
" Monte Carlo Trials
- Core Calculations (Outer Loop)
- Fuel Assembly Calculations (Inner Loop)
" Rods in BT Calculation July31 August 2, 2007 35 Fuel Transition Meeting NRC Meeting LAR NRC WA Richiand, WA -
AREVA NP Inc. Brunswick Fuel Brunswick TransitionLAR Richland, July 31 - August 2, 2007 35 L
S&-AFLIM2 Computer Code Initialization
" Establish initial (nominal) operating conditions at which the core MVCPR equals the desired SLMCPR
" Initial conditions are required for the following parameters
" Core flow
" Core pressure
- Feedwater temperature
" Feedwater flow
- Core inlet enthalpy
" Core power
" Assembly power (radial peaking)
" Core average axial power shape
" Assembly flow 36 LAR NRC Meeting NRC Meeting WA Rich/and, WA July 31 - August 2, 2007 AREVA NP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition Richland, July 31 - August Z 2007 36
SAFLIM2 Computer Code Initialization (continued)
August 2, 2007 37 Rich/and, WA July31 NRC Meeting LAR NRC -
Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 -August 2, 2007 37
SAFLIM2 Computer Code Core Calculations - Outer Loop August 2, 2007 38 Rich/and, WA July 31 NRC Meeting LAR NRC -
AREVA NP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 38
SAFLIM2 Computer Code Assembly Calculations - Inner Loop August 2, 2007 39 Richiand, WA July31 NRC Meeting LAR NRC -
AREVA NP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 39 L
SAWFLIM2 Computer Code Fuel Rod Calculations - Inner Loop
>1 August 2, 2007 40 LAR NRC Transition LAR Fuel Transition Brunswick Fuel Meeting Rich/and, WA July 31 -
AREVA NP Inc. Brunswick NRC Meeting Richland, WA July 31 - August 2, 2007 40
CSAFLIM2 Computer Code Number of Rods in BT
>1 41 Richiand, WA July 31 August 2, 2007 NRC Meeting LAR NRC -
AREVA NP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 41
Safety Limit Methodology 42 August 2, 2007 I AREVA NP Inc. IL Brunswick Fuel Transition Brunswick Fuel LAR NRC Transition LAR Meeting NRC Meeting Richland, WA Rich/and, WA July 31 -
July 31 - August 2, 2007 42
SPAIFLIM2 Computer Code Reactor System Uncertainties AREVA NP Inc. Brunswick Fuel Brunswick Fuel Transition Transition LAR LAR NRC NRC Meeting Rich/and, Meeting Richland, WA WA July31 - August July 31 -August 2, 2007 2, 200 7 43 43
S',AFLIM2 Computer Code Core Monitoring Uncertainties
[
August 2, 2007 44 NRC Meeting LAR NRC Transition LAR Fuel Transition Richland, WA July 31 -
AREVA NP Inc. Brunswick Fuel Brunswick Meeting Richland, WA July 31 - August 2, 2007 44
SAFLIM2 Computer Code Fuel Design Uncertainties I
45 Richiand, WA July31 August 2, 2007 I NRC Meeting LAR NRC -
Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 45 AREVA NP Inc. IL
Thermal Limits Methodology I
Average Core Conditions Design Peak Core Conditiont Allowed Operating Range Design Margin (5%-10%)
Operating LimitI 1---OLMCPR (1.19-1.45)
Transient Limit 1 1 SLMCPR (1.05-1.11)
Statistical Margin Defined Overheating _MCPR = 1.00 Cladding Damage -I - MCPR < 1.00 46 Brunswick Fuel Transition LAR NRC Meeting Richiand, WA Richland, WA 31 -- August July 31 July 2, 2007 August 2, 2007 46 I AREVA NP Inc. I11
Safety Limit Analysis RVmPIc Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 47
Safety Limit Analysi's August 2, 2007 48 Richiand, WA July 31 NRC Meeting MR NRC -
AREVA NP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, INA July 31 - August 2, 2007 48 k
Safety Limit MCPR Results Brunswick Unit 1 Cycle 17
> Analysis results support a 1.11 SLMCPR for two-loop operation
> Analysis results support a 1.12 SLMCPR for single-loop operation August 2, 2007 49 July 31 I AREVA NP Inc. IL Brunswick Fuel Brunswick NRC Meeting LAR NRC Transition LAR Fuel Transition Meeting Rich/and, WA Richland, INA July 31 - August 2, 2007 49
Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 50 I AREVA NP Inc. I
Neutronic Analysis Methodology Ken Hartley Team Leader, BWR Neutronics Bruswik Richland, Felraniton WAARRCeetngJuly 31 I AREVA Brunswick Fuel Transition LAR NRC Meeting - August 2, 2007 1
Neutronic Analysis Methodology Major Computer Codes Code Use CASMO-4 Performs fuel assembly burnup calculations and calculates nuclear data for MICROBURN-B32 MICROBURN-B32 Performs 3-dimensional steady-state reactor core neutronic analyses for assessing impact on thermal limits during localized and quasi-steady-state events 2
LAR NRC Transition LAR Meeting NRC Meeting Rich/and, WA July31 - August 2, 2007 Brunswick Fuel Transition Brunswick Fuel Richland, WA July 31 - August Z 2007 2 i AREVA NP Inc. Ik
Neutronic Analysis Methodology CASMO-4 Computer Code Description Multi-group, 2-dimensional transport theory code Use Performs fuel lattice burnup calculations and generates nuclear data for use in MICROBURN-B32 Documentation EMF-21 58(P)(A) Rev 0, Siemens Power Corporation Methodology for Boiling Water Reactors:.Evaluation and Validation of CASMO-4 /MICROBURN-B2, October 1999 Acceptability The safety evaluation by the NRC for the topical report EMF-2158(P)(A) approves the CASMO-4/
MICROBURN-B32 methodology for licensing applications 3
AREVA NP Inc. Brunswick Transition LAR Fuel Transition Brunswick Fuel Meeting NRC Meeting LAR NRC Richland, WNA July 31 - August 2, 2007 3
Neutronic Analysis Methodology MICROBURN-B2 Computer Code Description A 3-dimensi onal, two group, diffusion theory code incorporating microscopic depletion and pin power reconstruction Use Performs 3-dimensional steady-state reactor core neutronic analyses for assessing impact on thermal limits during localized and quasi-steady-state events Documentation EMF-21 58(P)(A) Rev 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation -of.CASMO-4/MICROBURN-B2, October 1b99 Acceptability The safety evaluation by the NRC for the topical report EMF-2158(P)(A) approves the CASMO-4/
MICROBURN-B32 methodology for licensing applications' 4
I Meeting NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July31 -August 2, 2007 AREVA NP Inc. Brunswick Fuel Brunswick Richland, INA July.31 -August 2, 2007 L
Neutronic Topical Report SER Restrictions
> EMF-2158(P)(A) (CASMO-4/MICROBURN-B2)
- 1. The CASMO-4/MICROBURN-B2 code systemn shall be applied in a manner that predicted results are within the range of the validation criteria (Tables 2.1 and 2.2) and measurement uncertainties (Table 2.3) presented in EMF-2158(P)
- 2. The CASMO-4/MICRO.BJRN-B code system shall be validated for analyses of any new fu'el design which departs from current orthogonal lattice designs and/or exceed gadolinia and U-235 enrichment limits
- 3. The CASMO-4/MICROBURN-B2 code system shall only be used for BWR licensing analyses and BWR core monitoring applications
- 4. The review of the CASMO-4/MICROBURN-B2 code system should not be construed as a generic review of the CASMO-4 or MICROBURN-B2 computer codes 5
LAR NRC Transition LAR Meeting NRC Meeting Richiand, WA July31 - August 2, 2007 Brunswick Fuel Transition Brunswick Fuel Richland, WA July 31 - August 2, 2007 5 i AREVA NP Inc. IL
Neutronic Topical Report SER Restrictions
>EMF-2158(P)(A) (CASMO-4/MICROBURN-B2)
- 5. The CASMO-4/MICROBURN-B2 code system is approved as a replacement for the CASMO-3/MICROBURN-B code system used in NRC-approved AREVA-BW.R licensing methodology and inAREVA BWR core monitoring applications. Such replacements shall be evaluated to ensure that each affected methodology continues to comply with its SER restrictions and/or conditions
- 6. AREVA shall notify any customer who proposes to use the CASMO-4/MICROBURN-B32 code system independent of any ARE VA fuel contract that conditions 1-4 above must be met.
AREVA's notification shall provide positive evidence to the NRC that each customer has been informed by AREVA of the applicable conditions for using the code system Conformance to No. 1 is addressed through benchmarking the code system against actual op'eratio'n of'previous cycles.
AREVA NLP Inc. Brunswick Fuel Transition LAR NRC Meeting Richiand, WA. July3.1 - August 2, 2007 6
Neutronic Code In put Flow CA SMO-4/MICROB URN-B 2 7
Richiand, WA July31 August 2, 2007 NRC Meeting LAR NRC AREVA NP Inc. Brunswick Brunswick Fuel Transition LAR Fuel Transition Meeting Richland, WA July 31 - August Z 2007 7
Neutronic Fuel Cycle Design
" Nuclear fuel assembly' design attributes
- Lattice and assembly reactivity
- Enrichment distribution
- Local peaking distribution
- CPR correlation factors
- Axial enrich mentlgadol inia zoning, natural blankets
" Lattice calculations performed with CASMVO-4
" In-core simulations performed with MICROBURN-B32 8
Richiand, WA July31 ~August 2, 2007 AREVA NP Inc. I Brunswick Brunswi.ck Fuel Transition LAR Fuel Transition NRC Meeting MR NRC Meeting Richland, ,,.-July 317 August 2, 2007 8 L
Neutronic Fuel Cycle Design
" Constraints on Nuclear Fuel Design due to Mechanical Criteria
" Limitations on LHGR for PCI, fission gas release, cladding strain
" Limitations on discharge burnup
" Licensing and safety' constraints
- MAPLHGR and MCPR limits S...A Brunswick Fuel Transition LAR NRC Meeting Rich/and, WA July 31 - August 2, 2007 9
Nuclear Design ALAADIN Graphical User Interface Assembly Browser Interface 10 July 31 August 2, 2007 I AREVA NP Inc.
Brunswick Transition LAR Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Meeting Rich/and, WA Richland, WA July 31 - August Z 2007 10
ALAADIN Graphical User Interface (continued)
Assembly Map Inspector --
F-eff I AREVA NP Inc.
Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 1 11
Gadolinia Design
" Gadolinia design isdetermilned b,,ased both on CASMO-4 results and in-core performance via MICRO.BURN-B2
" Position of gadolinia rods determined by acceptable local peaking and CPR factor results
" BOO hot excess reactivity and .cold shutdown margin calculations to determine number of gadolinia rods
" Peak reactivity calculations inthe cycle to determine concentration mPIc RV Brunswick Fuel Transition LAR NRC Meeting Richiad WA uly'31 -'August 2, 2007
Bundle Enrichment Design
> Enrichment level set based on energy requirement
> Multiple fresh sub-batches used when needed to smooth out reactivity distribution inthe core Rich/and, WA July31 - August 2, 2007 I Fuel Transition Brunswick Fuel LAR NRC Transition LAR Meeting AREVA NP Inc. Brunswick NRC Meeting Richland, WA July 31 - August 2, 200 7 L
Fuel Loading Plan Design
" Iterative process to deter mine opt~imal placement of fresh and irradiated assemblies
" Fresh batch size and enrichment determined to meet cycle energy requirement
" Constraints based on therma~l and reactivity margins, channel management, fuel conditioning from previous cycle operation, and licensed fuel burnup limits
" For mixed cores, coresident other vendor fuel limits must also be protected
" MICROBURN-B32 core simulator analyses used to set the loading plan 14 Rich/and, WA July31 -August 2, 2007 NRC Meeting I
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel LAR NRC Transition LAR Meeting R [&N6 nd,ýý4W. July,31 -':August Z 2007 14 L
CDM (Core Design Module) Graphical Interface Loading Pattern Colored by BOO Exposure (invalid Loading)
I AREVA NP Inc.
Brunswick Fuel Transition LAR NRC Meeting Brunwic Rich/and, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 15
Target Rod Pattern Design
" The step-through depleti o~nis"used for detailed analysis of the cycle operational capability
" It provides the best estimate of realistic exposure, void, and control history which is.to be -expected for the cycle-
> It provides confirmation that the core as designed can be operated with margin to assumed technical specification limits 16 Rich/and, WA July 31 -August 2, 2007 Brunswick Transition LAR Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Meeting Richland, WA July 31 2, 2007 16 i AREVA NP Inc.
CDM (Core Design Module) Graphical Interface -
-Step Through Design Thermal Margin 1 2 3 4 5 6 7 10 11 12 13 14 15 Results 0.753 0.764 0.697:'10.699
.0.759 0.709 0.732 10.606 0.749 0.478 0.46 0.539
.203 0.268 0.334 0.317
- 0. 107 0.233 0.756 0.767 0.772 0.758 4f.,i 0.763 0.779 0.674-0.64 0.276 0.371 1.10.116 z
Colored by 1,10.755 0.8 0.677 0.68 C17 0.753'ý 0. 823-, 0.558';'0.638 0.3139 00.4,12 0.3,111 H0.249 0.691 0.714 ý'O.643 0.596 0.485 0.393 0.118 Limiting 1 WýY 0.676,-
0.778 ýM 0. 6 6 7 L'M 0.75511 0 21% 0.691 0.761 U0.669 0.718 `'ý 0.5V 0.569 0.443 0.362 0.254 LHGR 0.753 [17 0.693 1 0.76 0.76
-,0.754 0.746 0.761 0.666 0.718:; 0.683 i 0.741 01.16,44
'.59 0.59 0.ýý 53 0.579
.462 0.592 -,0.661 0.502; 0.633 0.623 0.565 0.273 0.375 0.371 RO.113 0.347 0.247 Margin 0.696 0.675 -
0.76 0.768 0.686 -10.675 0.592 0.475 0.429
.5ý314 0.641 0.619 0.576 0.47 0.596 1 0.272 0 541 U0.343 0.335 0.322 0.101 0.224 0 765 0.778 -0.779 0.636 0.661 0.614 - 0.645 0.559,1ý 0.208 0.285 0.084 (Invalid 0.687 00ý1ý 0.71 0.742 '0.768 0.776 :0.634 0.589 0.642 j
0.48 L
0.61 J
0.524 1110.311 L--
0.283 0 .194 0.746 F,., . i 0.597,:0.586 0.655 0.583 0.49 0.178 0.111 0.057 Step Out EIJI EM 0.686 0.583 0.767 ' 0.512 0.495 0.551 0.628 0.524 0.464- 10.282 0.186 0.138 Pattern) 0.767 0.755 0
0+.6480..5-9 0.636,,,0.597 0.588 :0.495 0539 061 0488
-38 0.534-0.52 0.537 0.503 0.431 0.433
- 0. 146 0.243 0.079 0.186 it .0.69 0 0.656 0.581 0..447 0.445 0.422 0.348 0. 109
-0.677 ,173 .429 0 1 -. 0.58 ý54 0.642 0.555 0.'18 038+ 12 0.428 0.384 0.36 0.264
,0.736 0.675 0.714 0.593 0.639 0.612 0.65 0.532, 0.446 0.402:: 0.342 0.131 0.076 10 0.752 0.564 .0.72 0.504': 0.619 0.48 10.628 0.538 0.427 0.394 '0.367 0.309 0.181 0.61 0.64 0.6 0.655 ; 0.59 0.64 0.579 0.521 0.422 0.343, 0. 131 L 0.479 0.638 0.513 0.631 L'0.469 0.609'j 0.527 0.503 -'0.384 0.364 ý'O.316 00.44 0.132 0.434 - 0.347 5,,, 0.3924 0.567 0.562 0.538 0.523 ý;0.465 0.433' 0.361 0.3 0.555 0.49 0.592 0.615 10.588 0.207 '0.183 0.157 0.11 0.078 3 ý2067111 0.283 0.466" 0.481 0.266: 0.233
.269 0.443 ý 0.441 0.374 LO.334 0.311 'ý0.288 0.254 - 0.263 0.182
--0.335 0.374- 0.392 0.367 ý 0.334 0.286,10.116 0.077 14
ý'O.31 8 0.342 0.363 0.344 '0.323 0.284 :0.19 0.189 0.109 0.116 0.124 0.112 15 0.085 :0.059 0.239 0.25 -'0.261 0.248 0.232 0.196 Ll0.14 Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful Tanstio MRNRCMeeing WA July 31 - August 2, 2007 117
Neutronic Safety Analysis
" Once the fuel cycle design has been accepted by the customer, licensing of the core can commence
" Approved methods generally defined inXN-NF-80-19(P)(A) or subsequently submitted and approved Topical Reports XN-NF-80-19(P)(A) Vol 1 Supplements 1 and 2, and Vol 4 (No specific SER restrictions listed)
August 2, 2007 18 WA Richiand, WA July 31 NRC Meeting LAR NRC Transition LAR -
IP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, July 31 - August 2, 2007 18
Neutronic Safety Analysis Methodology Cycle-Specific Analyses
> Items Analyzed With CASMVO-4/MICROBURN-B32
" Cold shutdown margin
" Standby boron liquid control shutdown margin
- Control rod withdrawal error
- Loss of feedwater heating
- Control rod drop accident AREVA NP Inc. k Brunswick Fuel Transition LAR NRC Meeting Richland, WNA July31 - August 2, 2007 19
Neutronic Safety Analysis Methodology Cycle-Specific Analyses
> Items Analyzed With CASMO-4/MICROBURN-B2 (continued)
- Core flow increase event (.LH.GRf)
- Neutron ic input provided to BWR-SA for
- SLMCPR
- MCPRf
- Fast transient analyses
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 20 L
Neutronic..Safety:Analysis Methodology Cycle-Specific Analyses
> Other Neutronic analyses
- Fuel storage criticality (*) (CASMVO, KENO V.a)
- Fuel handling accident ()(ORIGEN-S)
- Reactor core stability (STAIF)
- Cycle-specific confirmation that analysis remains bounding August 2, 2007 21 NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July 31 -
Brunswick Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 21
Cold Shutdown Margin and Standby Liquid Control Shutdown Margin
" Cold shutdown margin (CSDM) - Evaluation of core reactivity at cold conditions with the analytically determined strongest control rod fully withdrawn, all other rods fully inserted. Must meet 0.38%
Ak/k technical specification requirement for subcriticality.
(Evaluated on a cycle-specificbasis with M/CROBURN-B2.)
Conditions analyzed are:
" Isothermal 68 TF, no Xenon
" Exposures throughout the cycle
" Standby boron liquid controlI (SLC) shutdown margin - Reactivity control by injection of boron in-the moderator (720 ppm B for Brunswick). Must be able to:. render the core subcritical in event control rods become inoperable'. (Evaluated on a cycle-specific basis with MICROBURN-B2.) Conditions analyzed are:
- Analyzed at exposures throughout the cycle July 31 August 2, 2007 22 Richiand, WA Transition LAR Fuel Transition NRC Meeting LAR NRC Brunswick Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 22
Control Rod Withdrawal Error
> Control rod withdrawal error (CRWE) - Inadvertent withdrawal of a control rod at power until it is stopped by the Rod Block Monitor (RBM.), on BWR/3-5 plants or the Rod Withdrawal Limiter (RWL.) ýon BWR/6 plants (ACPR protected by the MOPROL selected for the cycle). Conditions of the analysis:
" For BWR/3-5 plants - analyzed on a cycle-specific basis at BOG and peak reactivity exposure for the cycle with MICROBURN-B32
" Analyzed at quasi-steady-state conditions, e.g. as a series of steady solutions as the error rod iswithdrawn
" RBM setpoints corresponding to a range of MCPROLs determined
" Generic Topical approved for BWR/6 plants AREVA NP Inc. L Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 23
Stability
> For Brunswick, Long Term Stability Solution (LIS) Option Ill is used MCPR operating limits vers~us OPRM setpoints are determined
" Backup Stability Protecdtion' a~n'alyses are provided - Avoidance of core power oscillations ýby assessment of core loading (decay ratios) with the NRC-approved STAIF code. For Brunswick, scram regions and exclusion regions are determined consistent with BWROG guideline 0G02-0199-260.
" BWR core stability is sensitive to fuel rod thermal time constant, void coefficient, bundle 2- to 1-phase pressure drop, core power distribution, and operating point on the P/F map. (Methodology capable of supportinginterim corrective actions (ICAs),
exclusion Z-region, and long-term solutions, e.g., ID, EIA, and OptionIl)
AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 24
Los~s of Feedwater Heating
" Loss of feedwater he'ati'ng (LFWH) - A loss of feedwater heating capability due to the closing of a steam extraction line or the bypassing of feedwater flow around a heater, causing insertion of reduced temperature water into the core at power, i.e.,
reactivity insertion
" ACPR protected by the MOPROL selected for the core. (Generic Topical approved, AN F-I 358(P)(A)
Rev 3)
I AREVA NP Inc.
L Brunswick Fuel Transition LAR NRC Meeting Richland, WNA July 31 - August 2, 2007 25
LFWH Topical Report SER Restrictions
> ANF-1358(P)(A) Rev 3: :( Loss of Feedwater Heating)
- 1. The methodology applies to BWR/3-6 plants and the fuel types which were part of the database (GNF-8x8, 9/9B3 and 11; ANF-8x8 and 9/9; and ATRI UM-9B3 and 10) , provided that the exposure, the ratio of rated power and rated steam generation rate, rated feedwater temperature,, and.,change infeedwater temperature are within the range covered by the data points presented in ANF-1358(P)(A) Rev 3.
- 2. To confirm applicability of the correlation to fuel types outside the database, AREVA will perform additional calculations using the methodology, as described in Section 3.0 of the SER. In addition, AREVA calculations will be consistent with the methodology described in EMF-2158(P)(A) Rev 0 and comply with the guidelines and conditions identified inthe associated NRC SER.
July31 -August 2, 2007 26 Richiand; WA AREVA NP Inc. i LBrunswick Fuel Transition LAR NRC Meeting Brunswick Fuel Transition LAR NRC Meeting Richland,, INA
.".'.-:July 31 -'August 2, 2007 26
LFWH Topical Report SER Restrictions
> ANF-1358(P)(A) Rev 3'('Loss of Feedwater Heating)
- 3. The methodology applies only to the MVCPR operating limit and the LHGR for the LFWH event.
If core conditions are outside any of the SER restrictions for cycle exposure, ratio of rated power and rated steam generation rate, rated feedwater temperature, or change infeedwater temperature, cycle specific analyses are performed.
AREVA NP Inc. L Brunswick Fuel Transition LAR NRC Meeting Richland, WA July.31 - August 2, 2007 27
Control Rod Drop Accident
" The Control Rod Drop Accident (CRDA) is a postulated reactivity insertion accident (RIA). A control rod isassumed to become decoupled from the rod drive mechanism during rod withdrawal and is assumed to remain fully inserted as the drive mechanism moves to a new axial location. At some ýpoint later inthe withdrawal sequence, when the remaining control rods are in a configuration that maximizes the worth of the fully inserted decoupled blade, the decoupled blade slips free and falls to the drive mechanism location. The event results ina sudden reactivity addition causing a localized rapid increase inpower.
" ORDA acceptance criteria are::
- a deposited enthalpy less than 280 cal/gm at any axial location in any fuel rod
- maximum reactor pressure during any portion of the accident should be less than the value that will cause stresses to exceed "Service Limit C"as defined inthe ASME code
- the number of failed fuel rods shall be sufficiently small to remain within the accepted radiological consequences for the site
" The 280'cal/gm limit isthe threshold at which rapid expulsion of the fuel can occur. The fuel rod failure threshold is 170 cal/gm.
" Parametric analysis as discussed inXN-NF-80-19(P)(A) Vol 1 Supplements 1 and 2 I AREVA NP Inc.
L Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 August 2, 2007
- 28
Control Rod Drop Accident (continued)
>The severity of the event is determined based on the following parameters as determined from CASMO-4 and MICROBURN-B32 analysis:
" Doppler coefficient (aD)
" Delayed Neutron Fraction (fPeff)
- Control Rod Worth
" Four-Bundle-Local-Peaking factor (P4B3L)
" The CRDA isconservatively analyzed at hot-zero-power conditions (isothermal temperature, non-voided, and xenon free) at high reactivity cycle exposures.
" Additional conservatism is incorporated by assuming that adiabatic conditions remain during the power excursion (i.e., no direct moderator heating is credited during the analysis), and that the reactor remains at hot zero power conditions for the entire withdrawal sequence
" Deposited enthalpy for each candidate rod drop is determined by applying the XN-NF-80-19(P)(A). parameterization using the four parameters listed above AREVA NýP Inc. L Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 29
A Fuel Assembly Loading Error
" For BWRs-, there are two types of fuel assembly loading errors, the mislocation of an assembly into an unintended core location, and the misorientation of the assembly with respect to the control blade corner. It is assumed that the error goes unnoticed and the cycle operates with the misloaded assembly.
" Operating the cycle with an undiscovered fuel assembly in an 0RV improper core location results in changes in the core power distribution and potentially an in-crease in the local power density.
" An increase in the local powe 'rdensity will lower the minimum critical power ratio (MCPR) and increase the linear heat generation rate (LHGR) of the mis~located fuel assembly.
" The fuel assemblies in the cells surrounding the mislocated fuel assembly are also impacted.
I 30 I Brunswick Brunswi.ck Fuel Transition LAR Fuel Transition NRC Meeting LAR NRC Meeting Richiand, WA Richland, WA July31 - August 2, 2007 July 31 - August 2, 2007 30
Fuel Assembly Loading Error (continued)
>Per Section 15.4.7 of-NUR.EG-0800, plant operating procedures and design features minimize the likelihood of fuel assembly loading errors. Nevertheless, analyses are performed to demonstrate that inthe event-a fuel loading error is not detected and a fuel assembly isoper'at ed for the entire cycle in an improper position, the offsite dose. consequences would be no more than a small fraction (*510%) of the offsite dose (10 CFR Part 100 or 10 CFR 50.67 as applicable).
>The fuel rod failure mechanisms that are important inthe fuel assembly mislocation accident are overheating of the clad and overheating of the fuel pellets.
July 31 August 2, 2007 31 Richiand, WA NRC Meeting LAR NRC Transition LAR -
Fuel Transition Brunswick Fuel Brunswick Meeting Richland, WA July 31 - August 2, 2007 31
Fuel Assembly Loading Error (continued)
" Operating the cycle with an undiscovered fuel assembly inan improper orientation with respect to the control blade corner results in changes in the local peaking distribution inthe misrotated assembly. Misrotations of 180 degrees with respect ot the control blade corner are the most limiting.
" The misorientation results ininterference of the channel fastener at the top of the channel and the upper core support grid. The interference causes the assembly to lean toward the control blade corner. The lean of the assembly skews the out-channel water gaps leading to a skewed local peaking distribution within the assembly.
" The increase inthe out-channel water gap on two sides of the assembly leads to increased local peaking density inthe rods near the increased water gaps, which lowers the minimum critical power ratio (MC PR) and increases the linear heat generation rate (LHGR), of the misoriented fuel assembly. The fuel assemblies inthe cells surrounding the misoriented fuel assem;-bly- are also impacted (effective change in water gap between the asseamblies).
i AREVA NP Inc.
L Brunswick Fuel Transition LAR NRC Meeting Rich/and, WA -. July 31 - August 2, 2007 32
Fuel Assembly Loading Error (continued)
" Again, analyses are performed to demonstrate that inthe event a fuel loading error is not detected and a fuel assembly is operated inan improper orientation throughout the cycle, the offsite dose consequences would be no more than a small fraction (510%) of the offsite dose (10 CFR Part 100 or 10 CFR 50.67 as applicable).
" The fuel rod fai lure mechanisms that are important inthe fuel assembly misorientation accident are overheating of the clad and overheating of the fuel. pellets.
August 2, 2007 33 Richiand, WA July 31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Ill. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 33
Core Flow Excursion Event Boiling Water Reactors (BWRs) which operate within an extended operating domain of the power flow map, e.g., MEOD and MVELLA plants, or those which have implemented ARTS use a flow-dependent linear heat generation rate limnit 'multiplier, LHGRFACf or MAPFACf.
This flow-dependent limit mul1tiplier protects the fuel from exceeding mechanical criteria inthe event of an Anticipated Operational Occurrence (AOO) which results in a flow excursion. The limit set-down (enforced by multipliers < 1.0) prevents the plant from operating at an initial LHGR or MAPLHGR large enough such that at the end of the excursion, criteria would be violated.
It has been established that-a- slow recirculation flow runout produces the most limiting LHGR increase among all envisioned flow excursion scenarios.
July 31 August 2, 2007 34 Brunswick Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Transition LAR Meeting Richiand, WA Richland, VýA
.. July 31 - August 2, 2007 34
.Core Flow Excursion Event (continued)
> Based on the LHGR or MAPLHGR increase, flow-dependent multipliers are determined to protect
" Protection Against Power Transient (PAPT) criteria (AREVA fuel)
" MOP/TOP criteria (GE fuel)
> The most limiting slope of power increase versus core flow throughout the cycle is provided to BWR-SA I AREVA NP Inc.
Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 335
New and Spent Fuel Storage CriticalitySafety Analyses
" Maximum k-eff isestablished for each array assuming worst case conditions and including applicable tolerances and uncertainties
" Acceptance criteria are as defined inplant technical specifications or inNUREG-0800 (SRP), Chapters 9.1 .1 and 9.1.2 (typically array k-eff *0.95)
" Based on the analyses,. criticality safety limits are defined for each storage array which m'ust be met inorder to store fuel July 31 August 2, 2007 36 Richland, WA NRC Meeting LAR NRC Transition MR -
Fuel Transition Brunswick Fuel Brunswick Meeting Richland, WA July 31 - August 2, 2007 36
Transition Core Approach Steps in Transition
" Data collection (neutronic data includes coresident fuel geometry and neutron~ic design,ýcore geometry, as-loaded loading patterns, resultsfrom- measured TIP traces, startup conditions, cold critica'ls, -and'actual hot operating state points)
" Model setup - CASMO-4/MICROBURN-B2 (mixed core effects accounted for by explicit modeling of all fuel inthe core)
>]
" Define core hot and cold tar-get k-effs
" Establish current licensing basis (TS, FSAR, other documents)
" Disposition of events
" Provide neutronic input to transient, LOCA models
" Prepare core monitoring system input for Cycle N-I parallel operation -
Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 37
Transition Core Approach Steps in Transition (continued)
Perform criticality safety (new and spent fuel) and fuel handling accident analyses, per contract scope Develop neutronic fuel/core designs for Cycle N Obtain customer concurrence with fuel cycle design Perform Cycle N neutronic safety reload licensing analyses Issue fuel cycle design report Provide startup/core monitoring data for Cycle N Licensing support/training as: need July31 August 2, 2007 38 Richiand, WA NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA - '- July 31 - August Z 2007 38
StabilityAnalysis Methodology Dan Tinkler Engineer, BWR Safety Analysis 1
Brunswick Fuel Transition LAR NRC Meeting WA Richiand, WA July 31 -- August July 31 2, 2007 I AREVA NP Inc. II Richland, August 2, 2007 1
OPRM Setpoint Methodology Major Computer Codes Code Purpose STAlE Stability analysis code used to characterize the core and channel stability at the highest rod line at natural circulation RAMONA5-FA System analysis code used to generate the Delta over Initial CPR Versus Oscillation Magnitude (DIVOM) relationship MICROBURN-B2 Neutronics code that provides initial 3-D nodal cross-sections, core load ing/geometry, and statepoint information to both STAIF and RAMONA5-FA. Used to calculate initial core MCPR used to determine the supportable OPRM setpoi nt August 2, 2007 2 Richiand, WA July31 Brunswick Brunswick Fuel Transition LAR Fuel Transition NRC Meeting LAR NRC Meeting Richland, WA July 31 - August 2, 2007 2
OPRM Setpoint Methodology Calculation OPRM Setpoint Analysis Calculations
> Hot Channel Oscillation Magnitude (HCOM) for a given SP
" Obtained from statistical analysis by OPRMV vendor
" Plant specific / Cycle. and fuel independent
" Statistical 95/95 value calculated for a range of OPRMV setpoi nts mRVBrunswick Brunswick Fuel Transition LAR Fuel Transition NRC Meeting LAR NRC Meeting Rich/and, WA Richland, WA July 31 - August 2, 2007 July 31 - August 2, 2007 3
3
OPRM Setpoint Methodology Calculation OPRMV Setpoint Analysis Calculations
> DIVOM Relationship,:..
o Simulate regionall oscillations to determine the relationship between power oscillations and CPR response
- DIVOM slope isthe relative CPR response divided by hot bundle oscillation magnitude
- Uses a piece-wise linear interpolation in between DIVOM points to determine the slope August 2, 2007 4 Meeting WA Rich/and, WA July31 LAR NRC Transition LAR -
Brunswick Fuel Transition Brunswick Fuel NRC Meeting Richland, July 31 - August 2, 2007 4
.OPRM. Setpoint Methodology Calculation OPRMV Setpoint Analysis Calculations
> DIVOM Relationship (continued)
- DIVOM Point Definition-ý:
I 5
Richiand, WA July 31 August 2, 2007 NRC Meeting LAR NRC -
AREVA NP Inc. II Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 5
OPRM Setpoint Methodology Calculation OPRMV Setpoint Analysis Calculations
> DIVOM Relationship (continued)
"Cycle-specific DIVOM analysis is best estimate
" Best estimate RAMONA5-FA runs at prescribed exposures at the highest rod line at natural circulation
" Sensitivity cases are performed
-Limiting DIVOM exposure performed at +5%flow
" Limiting DIVOM isselected "HCOM multiplied by DIVOM slope gives the relative ACPR response for a given! "S' AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 6
Example DIVOM Curve August 2, 2007 7 Richland, WA July 31 NRC Meeting LAR NRC -
AREVA NP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 7
OPRM Setpoint Methodology Calculation OPRMV Setpoint Analysis Calculations
> Initial MVCPR
- Determine the MCPR. Margin that exists prior to the onset of oscillations
" Two scenarios
- Two-pump trip to natural circulation at the highest rod line
- Steady-state OLMCPR at 45% flow
- For the two-pump trip scenario, include the MCPR margin gained by moving down the rod line
- Calculations are cycle-specific August 2, 2007 8 NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July 31 -
Brunswick Fuel Brunswick Meeting Richland, WA July 31 - August 2, 2007 8
OPRM Setpoint Methodology Calculation OPRM Setpoint Analysis Calculations
> Two approaches for SLMCPR protection:
" Stability operating limit can then be calculated versus Sp
" Compare SLMCPR with final MCPR (initial MCPR - ACPR) 0Iterate to get the maximum SP that protects SLMCIPIR AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, FulTanstionLARNRCMeeing WA July 31 - August 2, 20079 9
OPRM Setpoint Methodology RAMONA5-FA Computer Code Description RAMVONA5-FA isa BWR system transient analysis code with models representing the reactor core, reactor vessel, and recirculation loops Use Perform core-wide and regional instability calculations to determine the relationship between the core power oscillations and the core CPR response Documentation BAW-1 0255(P) Rev 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, January 2006.
Acceptability The NRC performed an audit of the RAMIONA5-FA DIVOM methodology spring 2005. The methodology has been submitted to the NRC as part of the Enhanced Option IIl methodology.
I 10 I NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July 31 - August 2, 2007 AREVA NP Inc. Brunswick Brunswick Fuel Meeting Richland, WA-' July 31 - August 2, 200 7 10 L
RAMONA5-FA Computer Code Major Features I
August 2, 2007 11 July 31 I AREVA NP Inc. II, Brunswick Transition LAR Fuel Transition Brunswick Fuel Meeting NRC Meeting MR NRC Richland, WA Richiand, WA July 31 - August Z 2007 11
Reactor Core Stability AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 112
Transient Analysis Methodology Darrell Carr Team Leader, B WR Safety Analysis Meeting Richiand, WA July31 - August 2, 2007 Brunswick Transition MR Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Richland, WA July 31 - August 2, 2007
Transient Analysis Methodology Transient analysis method overview o COTRANSA2
- XCOBRZA-T Event descriptions Equipment out-of-service scenarios Thermal limits 2
WA July 31 August 2, 2007 I AREVA NP Inc. I Brunswick Transition LAR Fuel Transition Brunswick Fuel Meeting NRC Meeting LAR NRC Rich/and, Richland, WA July 31 - August 2, 2007 2
Transient Analysis Methodology Most transient events are classified as moderate frequency events (i.e. may occur during a calendar year to once. i~n 20 years, AOO)
Anticipated operational occurrence
- Fuel cladding integrity limits are not violated:
9 MCPR safety limit
- Reactor vessel pressure limit is not violated August 2, 2007 3 Transition LAR NRC Meeting LAR NRC Richiand, WA July31 -
I AREVA NP Inc..
Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 3
Transient Analyses Major Computer Codes Code Use MICROBURN-B2 3D cross-sections at state point of interest PRECOT2 1D cross-sections at state point of interest RODEX2 Gap conductance for core and hot channel XCOBRA Hot cha-nnel active flow COTRANSA2 System and core average transient response XCOBRA-T Hot channel 'response and ACPR calculation August 2, 2007 4 Brunswick NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July31 -
AREVA Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 4
Transient Analysis Methodology SupportingAnalyses r
August 2, 2007 5 NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July31 -
Brunswick Fuel Brunswick Meeting Richland, WA July 31 - August Z 2007 5
ACPR Calculation Flow Chart AREVA NP Inc. II Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 6
ZICPR Calculation Flow Chart (continued) r July 31 August 2, 2007 7 Richiand, WA NRC Meeting LAR NRC Transition LAR -
Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 7
ACPR Calculation Flow Chart (continued) r Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, July 31 - August 2, 20078 FulTanstionLARNRCMeeing WA 8
COTRANSA2 Computer Code Description COTRANSA2 is a BWR system transient analysis code with models representing the reactor core, reactor vessel, steam lines, recirculation loops, and control systems Evaluate key reactor system parameters such as power, flow, pressure, and temperature during core-wide BWR transient events Provide boundary conditions for hot channel analyses performed to calculate ACPR Documentation ANF-913(P)(A) Volume 1, Revision 1 and Supplements, CQTRANSA2:A Computer Program for Boiling Water Reactor Transient Analysis, August 1990 Acceptability The safety evaluation by the USNRC for the topical report ANF-913 approves COTRANSA2 for licensing applications 9
LAR NRC Transition LAR Meeting NRC Meeting WA Rich/and, INA July 31 - August 2, 2007 AREVA NP Inc. II Brunswick Fuel Transition Brunswick Fuel Richland, July 31 - August 2, 2007 9
A
- kV COTRANSA2 Computer Code Major Features
- Nodal (volume-junction) code with 1-dimensional homogeneous flow for the reactor system
- 1D neutron kinetics model for the reactor core (neutronics data obtained from MICROBURN-B32 and PRECOT2) that captures the effects of axial power shifts during the transient
- Core thermal-hydraulic model consistent with XCOBRA and XCOBRA-T
- Dynamic steam line model NOPRPRETR k Brunswick Brunswi.ck Fuel Transition LAR Fuel Transition NRC Meeting MR NRC Meeting Rich/and, WA~
Richland, _WA July~31 JuV,31
- August 2, 2007 2, 2007 10 10
COTRANSA2 Computer Code AdditionalFeatures
" Fuel rod transient heat conduction model
" Dual recirculation loops
" Dynamic pump model with homologous curves
" Steam separator model with level tracking
" Jet pump model
" Safety and relief valve models
" Turbine control, stop, and bypass valve-models
" Control system models
" Reactor protection system trip logic AREVA NP Inc. L Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 11
COTRNSA2Computer Code Typical Nodaliza tion Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 12
CO TRANSA2 Input Data
" System geometry (volumes, areas, lengths, elevations)
" Component characteristics (pumps, jet pumps, SRVs, TCVs, TSVs, TBVs)
" Reactor protection system characteristics (setpoints, delays, scram' performance, RPT performance)
" Control system characteristics (feedwater controller, pressure regulator)
" Fuel characteristics (gap conductance, neutronics data)
July31 August 2, 2007 13 Brunswick Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Transition MR Meeting Richiand, WA Richland, WA July 31 - August 2, 2007 13 AREVA NP Inc.
k
A COTRANSA2 Code Modules
" Core (COTRAN)
" Vessel and recirculation loops ARE0
" Steam line
" Control system
" Trip system I Brunswick Transition LAR Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Meeting Richiand, WA Richland, WA July 31 - August 2, 2007 July 31 - August Z 2007 14 14
Core Module - COTRAN
> Neutronics model characteristics I
I AREVA NP Inc. II Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 15
Core Module - COTRAN (continued)
" Thermal-hydraulic model
" Fuel rod model I
August 2, 2007 16 Rich/and, WA July31 NRC Meeting MR NRC -
AREVA Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 16
CO TRANSA2 System and Recirculation Loop Module
> General hydraulic models
- 1ID flow equations
- Homogeneous flow
- Conservation of mass and energy applied to control volumes; volume average density, pressure, and enthalpy defined at volume center
- Conservation of momentum applied at junctions; mass flow rate defined at junctions AREVA Brunswick Fuel Transition LAR NRC Meeting Richland, WA ,t July 31 - August 2, 2007 17
COTRANSA2 System and Recirculation Loop Module
> Phase separated volume I
> Centrifugal pump model Homologous pump curves used in dynamic model
" Jet pump model Applicable to 1 and 5 jet drive nozzle designs
" Implicit iteration scheme with core and steam line modules 18 Richiand, WA July 31 August 2, 2007 NRC Meeting I
AREVA NP Inc. Brunswick Fuel Transition Brunswi.ck Fuel Meeting Richland, WA . July 31 - August Z 2007 18 L
COTRANSA2 Steam Line Module
> Hydraulic models 1-dimensional, 3-equation (mass, energy, momentum) gas dynamics formulation
- Provisions are made to model turbine control, stop, and bypass valves as well as main steam isolation and safety/relief valves 19 WA Richiand, WA July31 August 2, 2007 NRC Meeting LAR NRC I
I Brunswick Fuel Transition Brunswick Fuel TransitionLAR Meeting Richland, July 31 - August 2, 2007 19 AREVA NP Inc.
COTRANSA2 Steam Line Module (continued)
August 2, 2007 20 Richiand, WA July 31 NRC Meeting LAR NRC Transition LAR Fuel Transition Brunswick Fuel Brunswick Meeting Richland, WA July 31 - August 2, 2007 20
COTRANSA12 Steam Line Module (continued)
Turbine flow (control and stop valve)
" Valve positions can be changed by control system or trip system actions
" If valve has not been tripped, control valve position isbased on control system demand (limited by maximum valve ramp rate)
" Turbine flow is based on control valve flow versus position characteristics Turbine bypass valve flow
" Valve position can be changed by control system or trip system actions
" Bypass valve position based on control system demand.
" Bypass flow iscalculated based on position and valve flow versus position curve AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 21
CO TRANSA2 Steam Line Module (continued)
Safety/relief valve flow
" For relief mode, valve position is calculated based on delay time and valve ramp rate
" For safety mode, valve position is pressure dependent MVSIV junction flow
" MSIV junction flow is calculated in the same way as other steam line junction flows
" Junction flow area is based valve ramp rate
" Once MSIV is fully closed, downstream nodes are dropped from the calculation Implicit iteration scheme with core and steam line modules AREVA NP Inc. Brunswick Fuel Transition MAR NRC Meeting Richland, WA July 31 - August 2, 2007 22 I
COTRANSA2 Control System Module
" Provides flexibility to model: most plant control systems - typically feedwater and pressure control systems are modeled
" Typically used to model feedwater flow, TCV position and bypass valve position
" Multiple control functions such as three-element, single-element and failed feedwater controllers can be entered m
RV Brunswick Fuel TransitionLAR NRC Meeting Richland, WA July 31 - August 2, 2007 23
COTRANSA2 Trip System Module
" Allows modeling of plant. protection functions (RIPS, ECCS, turbine, pumps, etc.)
" System valves, pumps, and control rods'can be connected to trip system
" Trip system can be used for event marking and time step control
" Comparative trips (GT, LT, EQ, GE, LE)
" Logical trips (AND,:OR),
August 2, 2007 24 Richiand, WA July 31 I NRC Meeting LAR NRC -
Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August Z 2007 24 AREVA NP Inc. IL
COTRANSA2 Benchmark Analyses
" Startup tests
" Level setpoint change
" Pressure regulator setpoint change
" Load rejection
" Recirc pump trip
" Peach Bottom turbine trip tests
"*TT2
"*TT2
- TT3
" NRC Licensing Basis Transient (LBT)
" BNL results
" GE results July 31 August 2, 2007 25 NRC Meeting Richiand, WA LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 25 I
Transient Analysis Methodology
..XCOBRA-T Computer Code Description XCOBRA-T predicts the transient thermal-hydraulic performance of BWR cores during postulated system transients Use Evaluate the transient thermal-hydraulic response of individual fuel assemblies in the core during transient events Evaluate the ACPR for the limiting fuel assemblies in the core during system transients Documentation XN-NF-84-105(P)(A) Volume 1 and Supplements, XCOBRA-T.,
A Computer Code for BWR Transient Thermal Hydraulic Core Analysis, February 1987 Acceptability The safety evaluation by the USNRC for the topical report XN-NF-84-105(P)(A) approves XCOBRA-T for licensing applications July 31 August 2, 2007 26 Richiand, WA NRC Meeting LAR NRC Transition LAR -
AREVA Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 26
XCOBRA-T Computer Code Major Features
" Hydraulic models are consistent with XCOBRA and COTRAN SA2
" Transient fuel rod model with! CHIF prediction capability.
The code iterates on' hot: channel RPF until CHIF occurs at the limiting node at the limiting time during the transient. ACPR is equal to the initial CPR minus 1.0 (CPR when CHIF occurs)
" A flow channel is used to represent the limiting assembly for each fuel type
" Nonlimiting fuel assemblies are grouped into average flow channels
" Boundary conditions (core, power, axial power shape, inlet enthalpy, upper- and lower-plenum pressure) are applied to the core August 2, 2007 27 NRC Meeting LAR NRC Transition LAR Fuel Transition Brunswick Fuel Rich/and, WA July 31 -
AREVA Brunswick Meeting Richland, WA July 31 - August Z 2007 27
XrCOBRA- T Core Representation Parallel channels representing one or more fuel assemblies
" Time-dependent total core power
" Time-dependent axial power shape 28 LAR NRC Transition LAR Meeting NRC Meeting Rich/and, WA July 31 - August 2, 2007 Fuel Transition I AREVA NP Inc. II Brunswick Fuel Brunswick Richland, INA July 31 - August 2, 2007 28
XCOBRA- T Thermal-Hydraulic Models
" 1-dimensional, 2-phase flow
" Unequal phase velocity (void-quality relationship)
" Thermal nonequilibrium (subcooled boiling)
" Fully implicit hydraulic solution includes effects of friction, local losses, elevation, area changes and density changes due to heating
" Mass and energy equations solved together using Newton-Raphson solution technique August 2, 2007 29 Rich/and, WA July 31 NRC Meeting MR NRC -
AREVA NP Inc. Brunswick Brunswick Fuel Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 29
XC11OBRA-T Fuel Rod Models
" Steady-state and transi~ent 1-dimensional (radial) heat conduction equ .ation
" Fuel rod model is consistent with the ROD EX2 fuel rod model
" Fuel-clad gap conductance isan input parameter (obtained from ROIDEX2)
" Internal heat generation based on:
" Core average power (time-de pendent)
" Axial peaking factor (time-dependent)
" Radial peaking factor (constant with time)
AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July31 - August 2, 2007 30
System Transient Analyses
" Overpressurization Analysis (COTRANSA2 only)
" ASME
" ATWS
" Load rejection no bypass (LRNB)
" Turbine trip no bypass (TTNB)
" Feedwater controller failure (FWCF)
" Pressure regulator fai~lure - downscale (PRFDS) 31 Transition LAR Fuel Transition NRC Meeting LAR NRC Richiand, WA July 31 - August 2, 2007 AREVA NP Inc. Brunswick Brunswi.ck Fuel Meeting Richland, WA July 31 - August 2, 2007 31
ASME OverpressurizationAnalyses Description
- The ASMVE overpressurization events analyzed include the closure of all main steam isolation valves (MS lVs), closure of the TSVs, and closure of the TCVs with failure of direct scram (on valve position)
Event classification
- The ASMVE event isa special analysis performed to demonstrate compliance with ASMVE pressure vessel design criteria Purpose of analysis
" Demonstrate that peak vessel pressure does not exceed 110% of the design vessel pressure
" Criteria: maximum vessel pressure (at bottom of RPV):* 1375 psig and maximum steam dome pressure <-1325 psig August 2, 2007 32 NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July 31 -
Brunswick Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 32 i AREVA
A TWS Event
]
July31 August 2, 2007 33 IP Inc. Fuel Transition Brunswick Fuel Brunswick NRC Meeting LAR NRC Transition LAR Meeting Richiand, WA Richland, -WA July 31 - August 2, 2007 33 I
Transient Analysis LRNB Description A significant grid disturbance occurs resulting in a loss of load on the generator and fast closure of the turbine control valve (TCV)
The turbine valves are required to close rapidly to prevent overspeed of the turbine generator rotor Event classification Anticipated Operation Occurrence Purpose of Calculate the ACPR for use inconfirming the analysis current OLMVCPR or inestablishing a new OLMVCPR Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 34
Transient Analysis LRNB
" Analytical methods
- COTRANSA2 and XCOBRA-T are the primary analysis codes
" Event is mitigated by contro~l rod insertion, revoiding due to increased heat flux and S'R.V.'actuation
" Special considerations
- RP.Inw -.qOO/n nnwpr thP nnwpr Inv~d Hinhqlq nn will not criiRp fi t TCV closure and the valves close in servo mode. A generator protection system produced turbine trip occurs 1 second after event initiation
" No direct scram on TCV motion or TSV position occurs below Pbypass (26% power)
" Partial arc operation for TCVs at high power Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 35
A- A Transient Analysis TTNB Description A variety of turbine or nuclear system malfunctions will initiate the closure of the turbine stop valves (TSVs) and turbine control valves (TCVs)
The turbine valves are required to close inorder to protect the turbine Event classification Anticipated Operation Occurrence Purpose of Calculate the ACPR for use inconfirming the analysis current OLMCPR or inestablishing a new OLMCPR July31 - August 2, 2007 36 Richiand, WA -
Brunswick Fuel Brunswick Transition LAR Fuel Transition NRC Meeting LAR NRC Meeting Richland, WA - .: - July 31 - August 2, 2007 36
Transient Analysis TTNB
" Analytical methods
- COTRANSA2 and XCOBRA-T are the primary analysis codes
" Event is mitigated by control rod insertion, revoiding due to increased heat flux and SRV actuation
" Special considerations
- No direct scram on TCV motion or TSV position occurs below Pbypass (26% 'power) 37 Transition LAR Fuel Transition NRC Meeting Rich/and, WA July 31 - August 2, 2007 AREVA NP Inc. Brunswick Fuel Brunswick LAR NRC Meeting Richland, WA July 31 -August Z 200 7 37
Transient Analysis FWCF Description A failure inthe feedwater control system results in a maximum demand signal to the feedwater pumps.' -The feedwater flow increases to the maximum .-capability of the feedwater pumps and results inan increase inreactor water level. A turbine trip occurs on high water level.
The high water level trip isinitiated to protect the turbine from damage due to liquid water entering the steam lines and turbine Event classification Anticipated Operation Occurrence Purpose of Calculate the ACPR for use inconfirming the analysis current OLMCPR or in establishing a new OLMCPR August 2, 2007 38 NRC Meeting LAR NRC Transition LAR Fuel Transition Brunswick Fuel Rich/and, WA July 31 -
AREVA NP Inc. Brunswick Meeting Richland, WA July 31 - August 2, 2007 38
" Analytical methods
- COTRANSA2 and XCOBRA-T are the primary analysis codes
" Event is mitigated by control rod insertion, revoiding due to increased heat flux and SRV actuation
" Special considerations
- Mismatch infeedwater flow and steam flow affects the overcooling portion of the event prior to the turbine trip
- Lower feedwater temperature can make the event more severe
- Lower initial water level results inlonger overcooling phase August 2, 2007 39 July 31 NRC Meeting LAR NRC Transition LAR Richland, WA Rich/and, -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting WA July 31 - August 2, 2007 39
Transient Analysis PRFDS Description A pressure regulator failure downscale event results inthe closure of all turbine control valves in servo mode. There is no scram on valve position or valve motion. The event isterminated by a scram on either high flux or high pressure.
When the backup pressure regulator is operational,tthe event is benign and bound by the consequences of another event Event classification Anticipated Operation Occurrence Purpose of Calculate the ACPR for use inconfirming the analysis current OLMVCPR or inestablishing a new OLMVCPR 40 LAR NRC Transition LAR Meeting NRC Meeting WA Richiand, WA July 31 - August 2, 2007 Fuel Transition Brunswick Fuel I AREVA NP Inc. I Brunswick Richland, July 31 -August 2, 2007 40
Transient Analysis PRFDS
" Analytical methods
- COTRANSA2 and XCOBRA-T are the primary analysis codes
" Event is mitigated by control rod insertion, revoiding due to increased heat flux and SRV actuation
" Special considerations
" High flux scram isattained during the event at high power
" If high flux scram does not'occur, the event continues until the high pressure scram set point isreached
" Calculations at Brunswick are performed at 90-100% power to show that operation with a pressure regulator out-of-service inthat power range issupported by the base case operating limits AREVA NP Inc. L Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 41
BrunswickPressurization Transient Analyses
> Analyses performed to support nominal scram speed (NSS) insertion times and technical specification scram speed (TSSS) insertion times
" Analyses support two individual EQOS operating limit sets and one combined EOOS limit set
" Analyses support exposure-dependent operating limits I
July 31 August 2, 2007 42 Rich/and, WA NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 42
BrunswickPressurization Transient.Analyses 43 Rich/and, WA July 31 August 2, 2007 NRC Meeting LAR NRC -
Brunswick Brunswick Fuel Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 43
Eqipent Out-of-Service
" Operation with equipment out-of-service (EOOS) can impact the consequences of an event
- Changes the system response
- Can impact events differently o EOOS scenarios are plant dependent
" Brunswick EOOS scenarios that may impact pressurization transient analysis results
" Feedwater heaters out-of-service (FHOOS)
" Turbine bypass valves out-of-service (TBVOOS)
" Safety/relief valves out-of-service (SRVOOS)
July 31 August 2, 2007 44 AREVA NP Inc. Fuel Transition Brunswick Fuel Brunswick NRC Meeting LAR NRC Transition LAR Meeting Richland, WA Richland, WA July 31 - August 2, 2007 44
Equipment Out-of-Service FHOOS
" FWCF event become more severe
" Overcooling phase is worse due to larger decrease ininlet enthalpy
" The steam flow may increase slightly making the pressurization phase more severe
" LRNB and TTNB events are typically less severe
" Lower steam flow
" Initial TCV position isfurther closed which can make the event more severe 45 Richiand, WA July 31 August 2, 2007 NRC Meeting LAR NRC -
Fuel Transition Brunswick Fuel I AREVA NP Inc. I Brunswick Transition LAR Meeting Richland, WA July 31 - August 2, 2007 45
Equipment Out-of-Service TBVOOS
" Fast opening capability of the turbine bypass valves isassumed inoperable
" For LRNB and TTNB events, the base case situation already assume TBVOOS
" For FWCF events
" Pressurization portion of the event becomes more severe
- Below Pbypass, a high.-pressure scram will occur RVmPIc Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 46
Equipment Out-of-Service SRVOOS RVOOS
- The lowest setpoint safety/relief valve isassumed inoperable
- All base case pressurization events (overpressurization and transient) are performed with at least 1 SRVOOS I AREVA NP Inc.
Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WNA July 31 - August 2, 2007 447
Thermal Limits
> MCPR limits
> LHGR limits
> Coresident fuel considerations August 2, 2007 48 WA Rich/and, WA July 31 NRC Meeting LAR NRC -
Brunswick Fuel Transition Brunswick Fuel Transition LAR Meeting Richland, July 31 - August 2, 2007 48
MCPR Limits
" Limits assembly power such- that if an AGO were to occur, less than 0.1% of the fuel rods would experience boiling transition (no violation of the SLMVC PR)
" Power-dependent MVCPR (MCPRp) limits are established to protect the sum of the limiting ACPR and the SLMVCPR
" A step change inthe MCPRp-limit is applied at Pbypass because of the loss of direct scram
" Exposure-dependent MCPRp limits are applied to provide MVCPR margin early inthe cycle
" Limits that support EGOS conditions are established
" Limiting MVCPR limit from either the power-dependent or flow-dependent MVCPR limit is applied in monitoring
" Limits are established for AREVA and coresident fuel 49 Bwnswick Brunswick Fuel Transition LAR Fuel Transition NRC Meeting LAR NRC Meeting Riphiand WA:
Rýcýlan ;.WA
- July31 August 2, 2007 July qI -, August 2, 2007 49
LHGR Limits
" LHGR limits are applied to ensure that the mechanical design criteria are satisfied. This includes the 1% plastic strain and fuel centerline temperature criteria
" To ensure that the 1% strain and centerline melt criteria are met during an AOO, a power- or flow- dependent set-down or multiplier (LHGRFACp or LHGRFACf) may be applied to the steady-state LHGR limit
" LHGR multipliers may be established on an exposure and/or EOOS-dependent basis.
" The limiting power-dependent or flow-dependent LHGR multiplier is applied in monitoring August 2, 2007 50 Transition LAR NRC Meeting LAR NRC Richiand, WA July31 -
I AREVA NP Inc. IL Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA . July 31 - August Z 2007 50
Thermal-Mechanical Limits Coresident Fuel
> For GE14 fuel, power- and flow-dependent multipliers are applied to the MAPLHGR limits to ensure that the thermal-mechanical design criteria continue to be met consistent with fuel vendor's methodology August 2, 2007 51 Richiand, WA July 31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 51
AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 52 LOCA Analysis Methodology Mike Garrett Manager, BWR Safety Analysis 1
Richiand, WA July 31 August 2, 2007 NRC Meeting LAR NRC -
Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 1
Loss-of-CoolantAccident (LOCA) Analyses
> Two types of LOCA analyses are performed
- LOCA break spectrum analysis
July 31 August 2, 2007 2 AREVA NP Inc. Fuel Transition Brunswick Fuel Brunswick NRC Meeting LAR NRC Transition LAR Meeting Richland, WA Richiand, WA -
July 31 - August 2, 2007 2
LOCA Break Spectrum Analysi's
" A LOCA may occur over a wide range of break locations and sizes
" Largest possible break isdouble-ended rupture of a recirculation pipe; however, largest break may not be the most severe challenge to the core
" Breaks inECCS piping are a special concern because the break results ina loss of an EGGS system as well as the LOCA
" Due to these complexities, a LOCA analysis must address a full range of break sizes and locations August 2, 2007 3 Richiand, WA July 31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 3
LOCA Break Spectrum Analysiffs (continued)
" The purpose of the break spectrum isto identify the characteristics of the pipe break that result inthe highest calculated POT
" Characteristics considered
" Break location
" Break type
- Break size
" Limiting ECCS single failure July 31 August 2, 2007 4 Richiand, WA NRC Meeting LAR NRC Transition LAR -
Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 4
LOCA Break Spectrum Analysi's (continued)
> Break locations
" Recirculation pump suction piping
" Recirculation pump discharge piping
- ECCS piping
- Other non-recirculation piping - generally dispositioned as being nonlimiting relative to fuel design requirements August 2, 2007 5 WA Richiand, WA July31 NRC Meeting LAR NRC Transition LAR -
Brunswick Fuel Transition Brunswick Fuel Meeting Richland, July 31 - August 2, 2007 5
LOCA Break Spectrum Analysis (continued)
> Break type (recirculation line breaks)
- Double-ended guillotine (DEG)
- Piping is assumed to be completely severed resulting in two independent flow paths to the containment
- Total break area is equal to twice full pipe cross-section area A
- Split break
- Longitudinal opening or hole in the pipe that results in a single flow path to containment
- Maximum flow area considered is equal to the cross-section area of pipe k Richiand, WA July31 August 2, 2007 6 NRC Meeting LAR NRC -
Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, INA July 31 - August Z 2007 6
LOCA Break Spectrum Analysi's (continued)
> E3reak sizes
- For recirculation DEG breaks, a range of discharge coefficients (1.0-0.4) isanalyzed to cover uncertainty in break geometry
- For recirculation split breaks, a range of break sizes is analyzed (from 0.05 ft2 to pipe cross-section area)
- For non-recirculation lines, a break size equal to the full cross-section area isassumed. unless the event is potentially limiting and requires further assessments August 2, 2007 7 NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July 31 -
AREVA Brunswick Fuel Brunswick Meeting Richland, WA July 31 - August 2, 2007 7
LOCA Break Spectrum Analysiffs (continued)
> Limiting EGGS single failure
- Regulatory requirements. specify that the most limiting single failure of ECCS equipment be 'assumed in LOCA analysis
- "Most limiting" refers to the ECCS equipment failure which produces the greatest challenge to acceptance criteria (generally PCT)
- Potentially limiting single failures are identified by the NSSS vendor and/or utility AREVA NP Inc.
L Brunswick Fuel Transition LAR NRC Meeting Richland, WNA July 31 - August 2, 2007 8
LO:CA MAPLHGR Limit Analysis
> Purpose of analysis isto demonstrate that the desired MAPLHGR limit versus exposure ensures that the LOCA-ECOS acceptance criteria are met for the limiting break identified inthe break spectrum analysis
" The break spectrum analysis is performed using BOL fuel parameters (e.g. stored energy, local peaking)
" Fuel parameters dependent on exposure have an insignificant effect on reactor system response during a LOCA; therefore, the limiting break characteristics from the break spectrum analysis are not exposure-dependent
> The thermal response of the fuel rods inthe limiting plane of the hot assembly during a.LOCAis dependent on parameters that vary with exposure August 2, 2007 9 NRC Meeting LAR NRC Transitian LAR Fuel Transition Brunswick Fuel Richiand, WA July 31 -
Brunswick Meeting Richland, WA . LI July 31 - August Z 2007 9 i AREVA
LOCA MAPLHGR Limit Analysis (continued)
" The reactor system response during a LOCA for the limiting break size and location, axial power shape, single failure, etc. is obtained from the break spectrum analysis
" Boundary conditions from the limiting LOCA are applied to each reload fuel and lattice designs in the MAPLHGR limit analysis
" Exposure-dependent POT. and MWR are calculated to confirm that the MAPLHGR limijt-prote~ctsAthe LOCA acceptance criteria
" AREVA generally establishes a MAPLHGR limit that is less restrictive than the fuel design LHGR limit 10 Richiand, WA July 31 August 2, 2007 NRC Meeting LAR NRC I
Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 - August Z 2007 10 AREVA NP Inc. IL
AR-A Loss-of-Coolant Characteristic Accident
" Blowdown phases_,.,-','.
phase
" Refill phase
" Reflood phase NONRORIEAR Brunswick Fuel Transition LAR NRC Meeting Rc/nW Richland, WA July uy1Ags 31 - August 2,,20 2007
Loss-of-CoolantAccident Blowdown Phase
> Initial phase following pipe break
" Net loss-of-coolant inventory inthe reactor vessel
" Rapid decrease insystem pressure for large breaks
" Clad temperature increase due to degraded core flow
" Depending on break size, core becomes fully or partially uncovered
" Core cooling is provided by exiting coolant and by core spray late inthe blowdown phase
>Blowdown defined to end'when LPCS reaches rated flow (or time when LPCS rated flow would have occurred in analyses with degraded LPCS 12 Transition LAR Fuel Transition NRC Meeting LAR NRC Richiand, WA July31 - August 2, 2007 Brunswick Fuel Brunswick Meeting Richland, WA July 31 - August 2, 2007 12
Loss-of-CoolantAccident Refill/ Re flood Phase
" Net increase in coolant inventory due to EGGS operation
" Core spray provides some cooling
" EGGS systems (LPCS, LPCI, and HPCS) supplies liquid to refill lower portions of the reactor vessel
" Fuel heat transfer to coolant is less than decay heat and fuel cladding temperature increases
" At the start of reflood, lower plenum mixture level reaches bottom of core
" During reflood, cooling is provided below the mixture level by pool cooling and above the mixture level by entrained liquid flow
" Eventually the entrained liquid reaching the hot node provides sufficient cooling and the cladding temperature begins to decrease (time of hot node reflood)
August 2, 2007 13 Transition LAR Fuel Transition Meeting NRC Meeting LAR NRC Richiand, WA July 31 -
AREVA NP Inc. Brunswick Brunswick Fuel Richland, WA July 31 - August 2, 2007 13
LOCA Analyses Acceptance Criteria
" The calculated maximum fuel element cladding temperature shall not exceed 2200 0F
" The calculated local oxid ation of the cladding shall not exceed 0.17 times the cladding thickness
" The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water shall not exceed 0.01 times the amount that would be generated ifall of the cladding. surrounding the fuel were to react
" Calculated changes in core geometry shall be such that the core remains coolable
" The calculated core temperature shall be maintained at an acceptable value for the extended period of time required by the long-lived radioactivity remaining inthe core 14 NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA: July31 -August 2, 2007 I AREVA NP Inc. I Brunswick Fuel Brunswick Meeting Richland,. WA;., July 317 August 2, 2007 14
LOCA Analyses Acceptance Criteria
> LOCA-ECOS criteria are commonly referred to as:
" POT criterion (<22001F)
" Local oxidation criterion (<17%)
" Hydrogen generation criterion (<1%)
" Coolable geometry criterion
- Long-term coolability criterion mRV PIc Brunswick Fuel Transition LAR NRC Meeting Richiand, WA Richland, INA July -:August 2, 31 -.August July 31 2007 2, 2007 15 15
" A limit on the maximum average planar LHGR (MAPLHGR) is established to ensure the LOCA-ECOS criteria are met
" The EXEM BWR-2000 LOCA analysis results demonstrate that the POT, local oxidation, and hydrogen generation criteria are met
" Compliance with these three criteria ensures that a coolable geometry is maintained
" Long-term coolability is confirmed for initial reloads - not fuel design dependent Brunswick Fuel Transition LAR NRC Meeting Brunwic Rihiand, Ful TanstionLARNRCMeeing WA July,31 -, August 2, 2007 116
LOCA MAPLHGR Limit Basis for Selected Limit
" Goal is to establish MAPLHGR limit that is less restrictive than the fuel design LHGR limit
" The maximum local peaking factor as a function of exposure for each void fraction isobtained from the neutronics cross-section generation calculation
" Divide the fuel design LHGR limit by the lowest maximum local peaking factor at each exposure
" Repeat for each lattice design inthe reload (ignoring blankets) 17 Richiand WA AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Transition LAR Meeting Ri6 d,ýWA Wa July 31 -August 2, 2007 17 L
Basis for MAPLHGR Limit (continued) 14-HRLii
-- MALHGR Limit 13-
' CD..
30 40 Planar Exposure (GWd/MTU)
July 31 August 2, 2007 18 Richiand, WA NRC Meeting LAR NRC -
AREVA NP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA July 31 -August 2, 200 7 18
LOCA Analysis Methodology
" The NRC-approved LOCA-ECOS Evaluation Model used by AREVA is referred to as EXEM BWR-2000
" EXEM BWR-2000 isan Evaluation Model that meets the requirements of 10 CFR 50 Appendix K
" The EXEM BWR-2000 methodology consists of three major computer codes and several auxiliary computer codes which are used to evaluate the reactor system and fuel response during a LOCA mPIc RV Brunswick Fuel Brunswick LAR NRC Transition LAR Fuel Transition Meeting NRC Meeting Rich/and, WA Richland, WA July31 - August 2, 2007 July 31 - August Z 2007 19
LOCA Analysis Methodology Major Computer Codes Code Purpose RODEX2 Fuel rod performance code used to predict the thermal-mechanical behavior of BWR fuel rods as a function of exposure RELAX BWR system analysis code used to calculate the reactor system and hot channel response during the blowdown, refill, and reflood phases of a LOCA HUXY Heat transfer code, used to calculate the heatup of a BWR fuel assembly during, all phases of a LOCA August 2, 2007 20 Richiand, WA July 31 NRC Meeting LAR NRC -
AREVA NP Inc. Brunswick Brunswick Fuel Transition LAR Fuel Transition Meeting Richland, INA July 31 - August 2, 2007 20
LOCA Analysis Methodology Auxiliary Computer Codes Code Purpose LPF Provide local peaking factors for rod groups from CASMVO tape7l file READW3 Incorporates power spikes to the MAPLHGR limit at defined exposures (typically every 5000 MWd/MTU) into the LPH RDXHXY Generates HUXY inputs from RODEX2 results PREHUXY Reads restart file from RELAX hot channel calculation and writes HUXY inputs for normalized power, fluid temperature, fluid quality and blowdown HTC SECHECK Compares HUXY initial stored energy against RODEX2 calculated stored energy 21 I NRC WALAR TransitionWA July31 Meeting August 2, 31 -- August 2007 AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Brunswick Fuel Richiand, Richland, July Z 2007 21
LOCA Analysis Methodology RODEX2 Computer Code Description Fuel rod performance code used to predict the thermal-mechanical behavior of BWR fuel rods as a function of exposure and power history Use Fuel rod stored-energy Initial fuel -rod--thermal and mechanical properties Documentation XN-NF-81-58(P)(A) Rev 2 and Supplements, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, March 1984 Acceptability The safety evaluation by the NRC for XN-NF 58(P)(A) Rev 2 and Supplements approves RODEX2 for licensing applications July 31 August 2, 2007 22 Brunswick Transition LAR Fuel Transition NRC Meeting LAR NRC Richiand, WA -
I AREVA NP Inc. I Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 22
Major RODEX2 Models
" Fission gas release
" Fuel swelling, densification and cracking
" Fuel to clad gap conductance
" Radial thermal conduction
" Free volume and internal gas pressure
" Fuel and cladding deformation
" Cladding corrosion AREVA NP Inc. L Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 23
LOCA Analysis Methodology RELAX Computer Code Description RELAX isa BWR systems analysis code used to calculate the- reactor' system and core hot channel response dur'ing-a LOCA Use Evaluate the time required to reach the end of the blowdown phase and to reach core reflood during the refill/reflood phase of the LOCA analysis Evaluate hot chan~nel fluid conditions during the blowdown phase of LOCA and time to reach hot channel reflood during the refill/reflood phase of the LOCA analysis Documentation EMF-2361 (P)(A), EXEM BWR-2000 ECCS Evaluation Model, May 2001 Acceptability The safety evaluation by the NRC for the topical report EMF-2361 (P)(A) approves RELAX for licensing applications August 2, 2007 24 Brunswick NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July 31 -
Brunswick Fuel Meeting Richidnd,'WA July 31 - August Z 2007 24 i AREVA
RELAX Computer Code Major Models
" Reactor system isnodalized into control volumes and junctions
" Mass and energy conservation equations are solved for control volumes
" Fluid momentum equation is solved at junctions to determine flow rates
" 1-dimensional, homogeneous equilibrium
" Three-equation model with drift flux model
" Complies with Appendix K requirements for EGGS analysis
>Separate models for average core and hot assembly L Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 25
RELAX System Model r
j 26 Transition LAR Fuel Transition NRC Meeting LAR NRC Richiand, WA July31 - August 2, 2007 Brunswick Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 26
RELAX Hot Channel Model AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July.31 - August 2, 2007 227
LOCA Analysis Methodology HUXY Computer Code Description Heat transfer code used to calculate the heatup of the peak power plane ina BWR fuel assembly during the blowdown., refill, and reflood phases of a LOCA Use Evaluate the peak clad temperature and metal-water reaction inthe fuel assembly resulting from a LOCA Documentation XN-CC-33(A) Rev 1, HUXY: A Generalized Multirod Heatup Code With I OCFR50 Appendix K Heatup Option
- User's Manual, December 1975 Acceptability The safety evaluation by the NRC for the topical report XN-CC-33(A) Rev 1 approves HUXY for licensing applications AREVA NP 1,,. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 28
-. JilI.4'
HUXY Computer Code Maj'o'r Features
" Models an axial plane in a fuel assembly
" Models individual rods in plane of interest
" Models assembly local power distribution and rod-to-rod radiant heat transfer
" Uses RELAX hot channel boundary conditions during blowdown
" Uses Appendix K spray .heat tran-sfer coefficients during refill
" Uses Appendix K reflood heat transfer coefficient after hot node refl ood AREVA NP Inc. Brunswi.ck Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 2 29 I
HUXY Computer Code Major Features
" Fuel rod conduction heat transfer model
" Convection heat transfer
" Radiation heat transfer (rod-to-rod, rod-to-channel)
" Fuel channel and rod quenching
" Metal-wate'r reaction (additional heat source and oxide buildup)
" Clad swelling and rupture AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 30
EXEM BWR-2000 LOCA Analysis Methodology r-August 2, 2007 31 Richiand, WA July 31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 -August 2, 2007 31
EXEM BWR-2000 Methodology Calculation Methodology Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 332
LOCA Analysis Methodology Cycle-Specific Analyses
> For each transition cycle, a complete plant-specific LOCA break spectrum analysis is performed
- Break location
- Break geometry (split, guillotine)
- Break size
- EGOS failure
- Axial power shape
- Initial core flow
> F:or each cycle, MAPLHGR limit analysis is performed
- Limiting break characteristics from break spectrum analysis
" Each lattice design incore
" Full exposure range August 2, 2007 33 NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July31 -
AREVA NP Inc. Brunswick Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 33
LOCA Analysis Methodology Break Spectrum Analyses August 2, 2007 34 Richiand, WA July 31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 34
LOCA Analysis Methodology MAPLHGR Analyses August 2, 2007 35 NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July 31 -
AREVA NP Inc. Brunswickk Fuel Meeting Richland, WA July 31 - August 2, 2007 35
I Brunswick Fuel Transition LAR NRC Meeting Rich/and, WA July 31 - August Z 2007 36 I AREVA NP Inc.
..EPU and Non-EPU Analysis Conditions Doug Pruitt Manager, Codes and Methods July 31 August 2, 2007 1 Richiand, WA NRC Meeting LAR NRC Transition MR -
Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 1
Reload Licensing Methodology
" Reload licensing analysis are performed to ensure that all fuel design and operating limits are satisfied for the limiting assembly inthe core
" Applicability of design methodology was determined by reviewing the explicit SER restrictions on the BWR methodology
" No SER restrictions on power level for the AREVA topical reports
" No SER restrictions on the parameters most impacted by the increased power level
- Core average void fraction
- Steam/Feedwater flow
- Jet Pump M-ratio
" The impact of EPU on core a'nd reactor conditions was evaluated to dtermne nycallegest6 the theoretical validity of the models 2
LAR NRC Transition LAR Meeting NRC Meeting Richiand, WA July 31 - August 2, 2007 I AREVA NP Inc. IL Brunswick Fuel Transition Brunswick Fuel Richland, WA July 31 - August 2, 2007 2
Power-Uprate Considerations
" Thermal operating limits (MCIPR, MAPLHGR, LHGR) are fairly insensitive to power uprate
" The ranges of key physical phenomena (e.g., heat flux, fluid quality, assembly flow) inlimiting assemblies during normal operation or transient events are not significantly different for uprated and non-uprated conditions
" Fuel specific determination of critical power isthe most limiting methodology for non-uprated ,and.uprated BWR operation
" AREVA analysis methodologies impose critical power correlation limits so the fundamental range of assembly conditions must remain within the same parameter space under uprate conditions August 2, 2007 3 Rich/and, WA July 31 NRC Meeting LAR NRC Transition MR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 3
Power Uprate Observations
" Maintaining the same critical,.power limits with increased core power requires flattening of the normalized radial power distributions
- Leads to a more uniform core flow distribution and slightly higher flow rates inthe hottest assemblies
" More assemblies and fuel-rod~s are near thermal limits and may result ina higher safety limit MCPR
" Higher steam flow rate and associated feedwater flow rate
" Core average void fraction will increase
" Higher core average power will lead to an increased core pressure drop and a slight decrease injet pump performance I AREVA NP Inc.
L Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 4
Repres en tative Assembly Power Distribution r
I AREVA Brunswick Fuel Transition LAR NRC Meeting Bruswik Richland, Felraniton W.A.
ARRCeetngJuly 31 - August 2, 20075 5
Plower Upra te Considerations
" Changes to the hot assemblies
" Power will be approximately the same
- Flow will slightly increase-
" Changes to the average assemblies
" Power will increase
" Flow will slightly decrease
Conclusion:
" The current parametric envelope will continue to encompass the conditions for all assembl~ies inan uprated reactor
" Therefore, the methods--us 'ed to assess assembly thermal-hydraulics are applicable to power uprate August 2, 2007 6 Richiand, WA 'July 31 NRC Meeting LAR NRC -
Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland, WA ' 'July 31 - August 2, 2007 6 i AREVA
Ther'mal-Hydraulic Core Analyses Testing Based
" AREVA tests to confirm or establish the applicability of methods
" PHTF test measurements provide assembly flow and pressure drop characteristics (e.g..., pressure loss coefficients)
- Karlstein test facility provides. both the assembly two-phase pressure drop and CHF performance characteristics
" FCTF tests confirm the conservatism of the Appendix K spray heat transfer coefficients
" Supplemental testing at Karlstein extends the validation and applicability of our methods
- Hydraulic stability
" Oscillatory dryout and rewet
" Void fractions August 2, 2007 7 NRC Meeting LAR NRC Transition LAR Fuel Transition Brunswick Fuel WA Rich/and, WA July 31 -
AREVA Brunswick Meeting Richland, July 31 - August 2, 2007 7
Karis tein Thermal ~jdraulic (KA THY) Test Loop August 2, 2007 8 Richiand, WA July31 NRC Meeting LAR NRC Transition MR AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August Z 2007 8 L
Critical Power Constraints
" SPCB fuel-specific CHF correlation based on KATHY test data
" Approved range of applicability for the SPCB correlation is enforced incodes (inlet subcooling, flow, pressure, boiling transition enthalpy) - uprate does not change this
- In some calculations, state conditions outside the limits are handled by NRC-approved conservative assumptions
" LOCA calculations fall outside the SPCB parametric envelope during the accident simulation.. In.this case, the local conditions formulation of the modified Barnett correlation isused.
August 2, 2007 9 NRC Meeting LAR NRC Transition LAR Fuel Transition Brunswick Fuel WA Richiand, WA July 31 -
Brunswick Meeting Richland, 'July 31 -August Z 2007 9 i AREVA
SPCB Out-of-Bounds Conditions CPR limited to boundary Enthalpy Conservative HL SL - All rods fail SL -All rods fail Worst flow Evaluate at higher enthalpy chosen than exists for current power Mass Flux (Mlbm/hr-ft2) mRV PIc Brunswick Fuel Transition LAR NRC Meeting Richland, WA Richiand, WA July 31 -- August July 31 2, 2007 August 2, 2007 10 10
.Critical Power Constraints
" Since the CHIF performance ischaracterized and imposed on a fuel design specific basis the assembly operating conditions must remain within the approved application range
" This fundamental restriction results inminimal differences between the benchmarked core conditions and those calculated for power uprate conditions
" This similarity isconfirmed by; comparing the assembly exit conditions
- KATHY pressure drop measurements
+ CASMO4/MICROBURN-B2 approved benchmark conditions (EMF-21 58(P)(A))
- Cycle depletion conditions for a Brunswick 120% power uprate / MELLLA+ core design August 2, 2007 11 Richiand, WA July 31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August Z 2007 11 L
Pressure Drop_ Tests Vs.
Reactor Benchmark and Design Conditions r
J AREVA NP Inc. Brunswick Fuel Transition LAR NRC Welir;g Pidd2md, W.,A July 31 - August 2, 2007 Brunwic Ful TansiionLARNRCMcel~g ichar~,. 144.Jul 31- Auust2, 00712
CA SMO-4IMICROB URN-B 2 Operating Experience Ave. Approximate Bundle. Peak Bundle Reactor Power, MWt Power, Power, Fuel/Cycle Reactor Size, #FA (% Uprated)* MWtIFA MWt/FA Licensing"* Uprate Comments A 592 2575 (0.0) 4.4 7.2 x B 592 2575 (0.0) 4.4 7.4 x 532 2292 (0.0) 4.3 7.3 (X)
Licensing D 840 3690 (0.0) 4.4 7.5 x only through Cycle 20 For 3 cycles E 500 2500 (15.7) 5.0 8.0 X oper.
F 444 1800 (5.9) 4.1 7.3 x G 676 2928 (8.0) 4.3 7.6 NX H 700 3300 (9.3) 4.7 8.0 NX 784 3840 (0.0) 4.9 8.1 WX 624 3237 (11.9) 5.2 7.8 (X)
With K 648 3600 (14.7) 5.6 8.6 NX ATRIUM-i10XM L 648 2500 (10.1) 3.9 6.9 (X
M 624 3091 (6.7) 5.0- 7.7 4.9 x
N 800 3898 (1.7) 7.7 x
0 764 3489 (5.0) 4.6 7.2 Brunswickt 560 2923 (20.0) 5.2 7.5 X Previously licensed by GNF
- Latest power uprates.
- (x)=currently fuel licensing only (Europe).
AREVA NP Inc. LBrunswick Fuel Transition LAR NRC Meeting Richland, WA. July 31 - August 2, 2007 1 13
Conclusions Thermal-Hydraulic Core Analysis Power uprate introduces changes incore design and steam flow rate Assemblies are subject to the. same LHGR, MAPLHGR, MCPR, and cold shutdown margin limits These LOOs restrict the as sembly powers, flows and void fractions typically within the ranges observed incurrent plant operation, the neutronics benchmarking database and the AREVA testing experience Therefore,
- Hydraulic models and, constitutive relationships used to compute the core flow distribution and local void content remain applicable
- Neutronic methods used to compute the nodal reactivity and power distributions remain applicable AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 14 I
Power Uprate Impact on Transient Analysis
" Phenomena of interest for BWR AOO transient analysis
- Void fraction/quality relationships
- Determination of CHF
- Pressure drop
- Reactivity feedbacks
- Heat transfer characteristics
" The dominant phenomena of interest are related to the local assembly conditions, not-the total core power
" ARE VA transient CHF measurements inKATHY are used to qualify the transient hydraulic solution
- Benchmarks capture the transient integration of the conservation equations and constitutive relations (including the void-quality closure relation.) and determination of CHF with SPCB
" AREVA benchmarks illustrate conservative predictions of time of Ldryout 15 Transition LAR Fuel Transition NRC Meeting LAR NRC Rich/and, WA July 31 - August 2, 2007 I AREVA NP Inc.
Brunswick Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 15
Transient Qualification r
Brunswick Fuel Transition LAR NRC Meeting Brunwic Rich/and, Ful WA July 31 - August 2, 2007 TanstionLARNRCMeeing 1 16
Power Uprate Impact on Transient Analysis
" Outside the core, the system simulation relies on solutions of the basic conservation equations and equations of state
- Feedwater flow and Jet Pump M-ratio changes
- Steam flow rate and steam line dynamics for pressurization events Impact of steam-flow. rate ýdependent on valve characteristics for pressurization events
- Solution of conservation equations have no limitations within the range of variation associated with power uprate
" Reactivity feedbacks are validated in a variety of ways
" Fuel lattice benchmarks to Monte Carlo results (SER restriction)
" Steady-state monitorings of reactor operation (power distributions and eigenvalue)
" Benchmark of coupled system to the Peach Bottom 2 turbine trip transients that exhibit a minimum of 5% conservatism
" Transient analysis remain valid for power uprate 17 Rich/and, WA July 31 August 2, 2007 NRC Meeting LAR NRC Transition LAR -
Fuel Transition Brunswick Fuel Brunswick Meeting Richland, WA July 31 - August 2, 2007 17
Power Uprate Impact on LOCA
> Local hot assembly parameters (POT & % MMW reaction) are determined primarily from: the. hot assembly initial stored energy, hot assembly transient decay heating and primary system liquid inventories
- Hot assembly initial stored energy, decay heating, and fluid inventory are not expected-to change significantly (same LHGR and MCPR limits)
- System inventory differences due to the increased core power have a transient feedback on the hot channel flow and fluid conditions
" Transient inventory differences due to power uprate are encompassed by the variation required to assess the entire break spectrum
- Code capabilities are not challenged by the differences
" Local hot assembly POT and % M/W reaction exhibit only small changes due to power uprate
" Core-wide parameters (core-wide M/W reaction and demands on long-term cooling) increase due to power uprate
" Current LOCA methodology- covers all phenomena for uprated Lconditions AREVA NP Inc. L Brunswick Fuel Transition LAR NRC Meeting Richland, WNA July 31 - August 2, 2007 18
Power Uprate Impact on Stability
" The flatter radial power profile induced by the power uprate will have a small impact on stability for same operating state point
- The flatter radial power :p.r'ofile may increase the core decay ratios
- Potential reduction inthe eigenvalue separation
- More assemblies operating at higher P/F ratios
" The STAIF code computes the stability characterist~ics of the core
- Frequency domain'solution of the applicable conservation and closure relationships
" Computes the regional mode directly using the actual state-point eigenvalue separation
- Bench marked against full assembly tests, as well as global and regional reactor data as late as 1998
" The impact of the "flatter" core design on stability limits will be directly computed based on-the projected operating conditions I AREVA NP Inc.
I Brunswick Fuel Transition LAR NRC Meeting Rich/land, WA. July31 - August 2, 2007 19
Power UprateImpact on Special Events
" AREVA performs ASME overpressurization analysis to demonstrate compliance with the peak pressure criteria
- System response and- se'nsitivities are essentially the same as AOO pressurization events
" AREVA performs ATWS analysis to demonstrate compliance with the peak pressurization criteria which occurs early inthe event
- Early system response and sensitivities are essentially the same as the transient simulations presented earlier
" Appendix R analysis i's performed using the approved LOCA analysis codes
" Like LOCA, the impact of power uprate isprimarily through the increase indecay heat. inthe core
" Decay heat isconservatively modeled using industry standards applied as specified by regulatory requirements
" Use of Appendix K heat transfer correlations and logic is LLBrunswickconservative for Appendix R calculations Fuel Transition LAR NRC Meeting Richland, WA July 31 August 2, 2007
- 20 AREVA NP Inc.
EPU Impact
" EPU operation does not challenge the applicability of the methods used to compute and monitor against licensing limits
" EPU operation is expected to impact the following areas:
- Safety Limit
- Transient response.. due t1o, different balance between core voids, feedwater/stearn flow rates and steamline valve characteristics
- LOCA core-wide metal-water reaction
- LOCA long term cooling
- Backup stability protection - exclusion regions 21 LAR NRC Meeting NRC Meeting Richland, WA July 31 - August 2, 2007 I AREVA NP Inc. I Brunswick Transition LAR Fuel Transition Brunswick Fuel Richland, WA July 31 - August 2, 2007 21
Power Up.ra-teApplicabilitySummary
" Maintaining margin to fuel desig n safety limits imposes restrictions on the range of operating conditions an assembly may experience during steady-state and transient conditions
" Increasing the core thermal. power isaccommodated by radial power flattening so that limiting assembly conditions deviate only slightly from current operating experience values
" The AREVA-approved licensing methods directly assess the impacts of power uprate on operating limits without modification
" The AREVA-approved licensing methods remain valid for power uprate conditions August 2, 2007 22 NRC Meeting LAR NRC Transition LAR Fuel Transition Rich/and, WA July31 -
AREVA NP Inc. Brunswick Fuel Brunswick Meeting Richland, WA .. July 31 -August Z 200 7 22 I
CA SMO-4IMICROB URN-B 2 Methodology Ralph Grummer Manager. Nuclear Technology Rich/and, WA July 31 - August 2, 2007 AREVA NP Inc. Brunswick Transition LAR Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Meeting Richland, WA . 'July 31 - August 2, 2007 1 I
BWR Methodology Applicability
> C)bjective
" Describe the cross section reconstruction process used by AREVA NP
- Demonstrate that the AREVA Methodology isaccurate for high void conditions August 2, 2007 2 Rich/and, WA July 31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 2
CASM 0-4
> CASMO-4 performs a multi-group (70) spectrum calculation using a detailed heterogeneous description of the fuel lattice components
" Explicit modeling of fuel rods, absorber rods, water rods/channels and structural components
- The library has cross sections for 108 materials including 18 heavy metals
- Depletion performed with a predictor-corrector approach in each fuel or absorber rod
" 2-dimensional transport solution is based upon the Method of Characteristics 3
LAR NRC Transition LAR Meeting NRC Meeting Rich/and, WA July31 - August 2, 2007 AREVA NP Inc. I Brunswick Fuel Transition Brunswick Fuel Richland, WA July 31 - August 2, 2007 3 L
CASM 0-4 (continued)
" Provides pin-by-pin power and exposure distributions
" Produces homogeneous mulIti-grou p (2) microscopic cross sections as well 'a'smacroscopic cross sections
" Determines discontinuity factors
" Performs 18-group gamma transport calculation
" Ability to perform colorset (2X2) calculation with different mesh spacings
" Reflector calculations are easily performed AREVA NP Inc Brunswick Fuel Transition LAR NRC Meeting Richland, WNA July31 - August 2, 2007 4
MICROB URN-B 2
" Microscopic fuel depletion
" Full two energy group neutron diffusion equation solution
" Modern nodal method solution isused
" Uses a higher order spatial method
" Water gap dependent flux discontinuity factors
" Multilevel iteration technique for efficiency
" MICROBURN-B32 treats a total of 11 heavy metal nuclides to account for the primary reactivity components 5
Richiand, WA July31 August 2, 2007 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007
MICROB URN-B 2 (continued)
" A model for nodal burnup gradient
" A model for spectral history.gradient
" Full 3-dimensional pin power reconstruction method
" TIP (neutron and gamma) and LPRMV response models
" Steady-state thermal-hydraulics model
" Direct moderator heat deposition based upon CASMVO-4 calculations
" Calculation of CPR, LHGR, and MAPLHGR August 2, 2007 6 July 31 NRC Meeting LAR NRC Richland, WA Richiand, -
AREVA NP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting WA July 31 - August 2, 2007 6
BWR Methodology
" Let us look at the cross section representation used in MICROBURN-B2
" MICROBURN-B2 determines the nodal macroscopic cross sections by summing the contribution of the various nuclides 7
Richiand; WA July 31 - August 2, 2007 Brunswick Fuel Brunswick Transition LAR Fuel Transition NRC Meeting LAR NRC Meeting Richland,ý WA :July 31 - August 2, 2007 7
MICROBURN-B2 Cross Section Representation where EX= nodal macroscopic cross section AX2= background nodal macroscopic cross section (D, Ef'Za' Z,)
i= nodal number density of nuclide "i" i= microscopic cross section of nuclide "1" I = total number of explicitly modeled nuclides p =nodal instantaneous coolant density H = nodal spectral history E nodal exposure R =control fraction AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July31 - August 2, 20078
MICR OB URN-B 2 Cross Section Repres en tation
" Functional representation of c7xi and A' comes from 3 void depletion calculations with CASMVO-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 August 2, 2007 9 Rich/and, WA July 31 Brunswick Fuel Transition LAR NRC Meeting I AREVA NP Inc. I Brunswick Fuel TransitionLAR NRC Meeting Richland, WA July 31 - August 2, 2007 9
MICR OB URN-B 2 Cross Section Repres en tation 10 Richiand, WA July31 August 2, 2007 NRC Meeting LAR NRC -
AREVA Brunswick Brunswick Fuel Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 10
MICROBURN-B2 Cross Section Representation 11 August 2, 2007 I AREVA NP Inc.
Brunswick Fuel Transition LAR NRC Meeting I Brunswick Fuel Transition LAR NRC Meeting Richiand, WA Richland, WA July 31 -
July 31 - August 2, 2007 11
MICROBURN-B2 Cross Section Representation
" At BOL the relationship, isfairly simple
" The cross section isonly a function of void fraction (water density)
" The reason for the variation isthe change inthe spectrum due to the water density variations
" At any exposure point, a quadratic fit of the three CASMVO-4 data points is used to represent the continuous cross section over instantaneous variation of void or water density August 2, 2007 12 Richiand, WA July31 NRC Meeting LAR NRC TransitionLAR Fuel Transition Brunswick Fuel Brunswick Meeting Richland, WA July 31 - August 2, 2007 12
MICR OB URN-B 2 Cross Section Repres en tation 13 I AREVA NP Inc. I Brunswick Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Transition MR Meeting Rich/and, WA Richland, WA July31 - August 2, 2007 July 31 - August Z 2007 13
MICR OB URN-B 2 Cross Section Repres en tation AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful WA July 31 - August 2, 2007 TanstionLARNRCMeeing 1 14
MICR OB URN-B 2 Cross Section Repres en tation
> Detailed CASMVO-4 calculations confirm that a quadratic fit accurately represents the cross sections 15
Richland,~
i AREVA NP Inc.
Brunswick Fuel i 6 Brunswick Meeting WA i nJuly 3 f. August 2, 2007 15
MICR OB URN-B 2 Cross Section Repres en tation July 31 August 2, 2007 16 NRC Meeting LAR NRC Transition LAR Richland, WA Richiand, -
AREVA NP Inc. Brunswick Fuel Brunswick Fuel Transition Meeting WA July 31 - August 2, 2007 16
MICROBURN-B2 Cross Section Representation July 31 - August 2, 2007 17 AREVA NP Inc. - Brunswick Fuel Brunswick NRC Meeting LAR NRC Transition LAR Fuel Transition Meeting Richland, WA Richiand, WA July 31 -August Z 2007 17 I
MICR OB URN-B 2 Cross Section Repres en tation August 2, 2007 18 I AREVA NP Inc. I Brunswick Fuel Transition LAR NRC Meeting Brunswick Fuel Transition LAR NRC Meeting Richiand, WA Richland, WA July 31 -
July 31 - August Z 2007 18
MICROBURN-B2 Cross Section Representation
" 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 100 exposure points
" The 4-dimensional representation can be reduced to 3 dimensions by looking at a single exposure August 2, 2007 19 Richiand, WA July31 NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, INA July 31 - August 2, 2007 19
MICR OB URN-B 2 Cross Section Repres en tation This is a smooth well behaved surface Brunswick Fuel Transition LAR NRC Meeting Richiand, WA July 31 - August 2, 2007 20
MICROB URN-B 2 Cross Section Representation
>Quadratic interpolation is performed ineach direction independently for the most accurate representation 21 Richiand, WA:-~. July31 - August 2, 2007 Brunswick Fuel Brunswick Transition LAR Fuel Transition NRC Meeting LAR NRC Meeting Richland, WA: % - %ý July 31 - August 2, 2007 21
MICROBURN-B2 Cross Section Representation August 2, 2007 22 Brunswick NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July31 -
AREVA NP Inc. Brunswick Fuel Meeting Richland, WA July 31 - August Z 2007 22
MICROBURN-B2 Cross Section Representation AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 223
MICROBURN--B2 Cross Section Representation
" The results of this process for all isotopes and all cross sections.
in MICROBURN-B32 were compared for an independent CASMO-4 calculation with continuous operation at 40% void (40% void history) and branch calculations at 90% void for multiple exposure
" The results show very good agreement for the whole exposure range August 2, 2007 24 Transition LAR Fuel Transition Meeting NRC Meeting LAR NRC Richiand, WA July 31 -
AREVA NP Inc. Brunswick Bruns ' k Fuel Richland, WA . July 31 - August 2, 2007 24
MICR OB URN-B 2 Cross Section Repres en tation July 31 August 2, 2007 25 Rich/and, WA NRC Meeting LAR NRC Transition LAR -
AREVA NP Inc. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA: : July 31 - August 2, 2007 25
MICR OB URN-B 2 Cross Section Repres en tation
> At the peak reactivity point multiple comparisons were made to show the results for various instantaneous void fractions July31 August 2, 2007 26 IP Inc. Fuel Transition Brunswick Fuel Brunswick NRC Meeting LAR NRC TransitionLAR Meeting Richiand, WA Richland, WA . :
A IJuly 31 ý August Z 2007 26
MICROBURN-B2 Cross Section Representation Quadratic fit using 0-40-80 provides excellent representation of data I
Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 27 I AREVA NP Inc.
MICR OB URN-B 2 Cross Section Repres en tation
" MICROBURN-B2 uses water density rather than void fraction in order to account for pressure changes as well as sub-cooled density changes
" MICROBURN-B2 uses spectral history rather than void history inorder to account for other spectral influences due to actual core conditions (fuel loading, control rod inventory, leakage, etc.)
RVmPIc runswick Fuel Transition LAR NRC Meeting Bruswik Richland, Fel raniton ARRC WA-i-eetng July 31 - August 2, 2007 228
MICROBURN-B2 Cross Section Representation
> The Doppler feedback due to the fuel temperature is modeled by accumulating the Doppler broadening of microscopic cross sections of each nuclide AE = (Je~ff- jef aFT f where:
T~f= Effective Doppler Fuel Temperature T~ef = Reference Doppler Fuel Temperature Ui= microscopic cross section (fast and thermal absorption) of nuclide i i= density of nuclide i 29 Richiand, WA......July31 - August 2, 2007 AREVA NP Inc. Brunswick Fuel Brunswick Transition LAR Fuel Transition NRC Meeting LAR NRC Meeting Richland, WA: Julý 31 - August 2, 2007 29
MICROBURN-B2 Cross Section Representation
> The partial derivatives are determined from branch calculations performed with CASMVO-4 at various exposures and void fractions for each void history depletion mRVBrunswick Transition LAR Fuel Transition Brunswick Fuel NRC Meeting MR NRC Meeting Richiand, WA Richland, WA July31 - August 2, 2007 July 31 - August 2, 2007 30 30
MICROBURN-B2 Cross Section Representation
" The tables of cross sections include data for controlled and uncontrolled states.
" Otherwise the process isthe same for controlled states
" Other important feedbacks to nodal cross sections are lattice burnup/spectral history gradient and instantaneous spectral interaction between lattices of different spectra Brunswick Fuel Transition LAR NRC Meeting Bruswik Richland, Fel raniton AR RCeetng WA, "July31 - August 2, 2007 331
32 Richland, WA July 31 - August 2, 2007 AREVA NP Inc. Brunswick Fuel Transition LAR NRC Meeting
Validation of MICROB URN-B 2 for EPU Conditions Ralph Grummer Manager, Nuclear Technology Rich/and, WA July 31 - August 2, 2007 AREVA NPINC. Brunswick LAR NRC Transition LAR Fuel Transition Brunswick Fuel Meeting NRC Meeting Richland, WA July 31 - August 2, 2007 1
B WR Methodology Applicability
> C)bjective
" Describe the validation process used by AREVA NP
" Demonstrate that the AREVA Methodology isapplicable to EPU conditions at Brunswick
- Demonstrate that data provided in the Neutronic Methodology Topical report bounds the expected conditions of EPU operation at Brunswick Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 2 j'r*'~ ~
B WR Methodology Applicability
> Validation of Steady-State Neutronic Methods for EPU conditions
- Tabulate the key parameters being validated (nodal power, pin power etc.), the type of benich marking/validation that was performed and the bundle conditions corresponding to the validation
- AREVA's neutronic method was validated by gamma scan and core follow benchmarking based upon the current fuel designs operated under the current operating strategies and core conditions July 31 August 2, 2007 3 Brunswick Fuel Transition Brunswick Fuel LAR NRC Transition LAR Meeting NRC Meeting Richiand, WA Richland, WA July 31 - August 2, 2007 3 AREVA NPINC.
EMF-,2,158(P)(A) Validation Basis
" EMF-21 58(P)(A) defined a set of criteria to demonstrate the acceptability of the Neutronic design code system
" Code system results were compared against critical experiments, higher order methods and actual commercial operating experience
" The SER states that the code system shall be applied ina manner such that results are within the range of the validation criteria (Tables 2.1, 2.2, and 2.3)
AREVA NP INC. Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, WA July 31 - August 2, 20074 FulTanstionLARNRCMeeing 4
Fuel Lattice Criteria Table 2.1 AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting RchnWA Richland, WA Jl31-ugs2,075 July 31 - August 2, 2007 5
Fuel Lattice Criteria Table 2.1 (continued) r J
August 2, 2007 6 July 31 NRC Meeting LAR NRC Transition LAR Richland, WA Richiand, -
AREVA NPINC. Brunswick Fuel Transition Brunswick Fuel Meeting WA . .. . July 31 - August 2, 2007 6
Fuel Lattice Criteria Table 2.1 (continued)
J August 2, 2007 7 Richiand, WA July 31 NRC Meeting LAR NRC Transition LAR -
AREVA NPINC. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August Z 2007 7
Core Simulator Validation Table 2.2 J
8 Richiand, WA July 31 August 2, 2007 NRC Meeting LAR NRC Transition LAR -
Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August Z 2007 8
Core Simulator Validation Table 2.2 (continued)
August 2, 2007 9 July31 NRC Meeting LAR NRC Transition LAR Richland, WA.
Rich/and, -
AREVA NPINC. Brunswick Fuel Transition Brunswick Fuel Meeting WA July 31 - August 2, 2007 9
Core Simulator Validation Table 2.2 (continued)
" TIP data taken from operating commercial power plants
" Gamma scan data taken from Quad Cities measurements on 8X8 assemblies
" Gamma scan data taken from KWU-S measurements on ATRIUM-10 assemblies Includes current fuel designs and operating strategies July31 August 2, 2007 10 Brunswick Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Transition MR Meeting Riahland, WA Richland, WA July 31 - August Z 2007 10
KWU-S Gamma Scan Benchmark Results EMF-2158(P)(A) pp. 8-8 Local power distribution uncertainty isnot axial level dependent AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 11
Ar Measured Power Distribution Uncertainty Table 2.3
-9
. Brunswick Fuel Transition LAR NRC Meeting Ful Tanstio Brunwic Richland, WA July 31 MRNRCMeeing - August 2, 2007 1 12
Continuous Validation Process
> AREVA Work Practice P1 04,129 requires evaluation for a significant fuel design change
" CASMVO-4 and MVCNP calculations are performed
" Fission rate distribution statistics are compared to Table 2.1 August 2, 2007 13 Rich/and; WA July 31 NRC Meeting LAR NRC AREVA NPINC. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland; WA July 31 ý August 2, 2007 13
ATRIUM TM -10 Lattice Validation Fission Rate Criteria Met August 2, 2007 14 Richiand, WA July 31 NRC Meeting LAR NRC Transition LAR -
AREVA NP INC. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 14 I
Continuous Validation Process
" For a new reactor, benchmark calculations are performed
" Hot operating eigenvalue statistics are compared to Table 2.2
" Cold startup eigenvalue statistics are compared to Table 2.2
" TIP statistics are compared to Table 2.2
" Local peaking comparisons are determined from the lattice calculations 15 Rich/aiki, WA~ July 31 -August 2, 2007 AREVA NPINC. i Brunswick Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Transition LAR Meeting Rich6ný,'WA-. Julý31-'August2,2007 15
Reactor Validation Results Elgen value. criteria are met RVmNIC Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful Tanstio MRNRCMeeing WA July 31 - August 2, 2007 1 16
Reactor Validation Results r
TIP comparisons include calculation and measurement uncertainties July 31 August 2, 2007 17 AREVA NPINC. Fuel Transition Brunswick Fuel Brunswick NRC Meeting LAR NRC Transition MR Meeting Richland, WA Richland, WA July 31 - August 2, 2007 17 L
Reactor Validation Results Measured power distribution uncertainties are a convolution of calculation and measurement uncertainties 613 iscalculated power uncertainty 6D issynthesized TIP uncertainty 6T iscalculated TIP uncertainty NIJ isthe number of TIPs RVmPN Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, FulTanstionLARNRCMeeing WA July 31 - August 2, 2007 18
Reactor Validation Results TIP measurements are taken with Gamma TIPs 19 Richiand, WA July 31 August 2, 2007 NRC Meeting LAR NRC Transition LAR AREVA NPINC. Brunswick Fuel Transition Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 19 k
Reactor Validation Results
" Measured and calculated TIP comparisons meet the requirements
" Measured symmetric TIP comparisons meet the requirements
" Together these indicate that the measured power uncertainty requirements are met 20 Rich/and, WA July 31 August 2, 2007 NRC Meeting LAR NRC -
AREVA NPINC. Brunswick Fuel Brunswick Transition LAR Fuel Transition Meeting Richland,;WA ý- ý July 31 - August Z 2007 20
Reactivity Coefficients - Void Coefficient
> Reactivity Coefficients - Void Coefficient
- Evaluate the AREVA methods and establish if the uncertainties and biases used in reactivity coefficients (e.g. void coefficient) are applicable or remain valid for the neutronic and thermal-hydraulic conditions expected for EPU operation.
July 31 August 2, 2007 21 Brunswick Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Transition LAR Meeting Richiand, WA Richland, WA July 31 - August Z 2007 21 AREVA NP INC.
Additional Validation
" In order to evaluate the accuracy of the void coefficient, MCNIP runs have been made
" These results indicate that CASMVO performs an accurate assessment of the void effect July 31 August 2, 2007 22 AREVA ýPINC- Brunswick Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Transition LAR Meeting Rich/and, WA Richland, WA
. July 31 - August Z 2007 22
CASMO-4 Vs. MCNP Results Casmo-4 void coefficient is nearly identical to MCNP AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 23
Void Coefficient Verification
> A measure of the quality of the simulator calculation isthe variation of the critical eigenvalue
> Observations of this behavior relative to core average void fraction indicate that there is no systematic bias
> Cycle exposure trends are accounted for by the use of target eigenvalue curves 24 Transition LAR Fuel Transition NRC Meeting LAR NRC Rich/and, WA July31 - August 2, 2007 AREVA NPINC. Brunswick Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 24
Void Coefficient Verification from Topical Report There is no trend in core elgenvalue relative to void fraction AREVA NP INC. i Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 25
Void Coefficient Verification
> The void coefficient is calculated accurately for a wide variety of core average void fractions
> The methodology retains the same accuracy for the conditions represented by EPU mRV NIC Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 2 26
Additional Validation
> Validation of Steady-State Neutronic Methods for EPU conditions SDemonstrate the current uncertainties and biases established inthe benchmarkings and presented inTables 9.8 and 9.9 of EMF-2158 (P)(A) remain valid for the neutronic and thermal-hydraulic conditions predicted for the EPU operation 27 Meeting Richl~nd, WA July 31 -August 2, 2007 AREVA NPINC. Fuel Transition Brunswick Fuel Brunswick LAR NRC TransitionLAR NRC Meeting Richl6h1d, WA " _' ýJuly 31 - August Z 2007 27
Additional Validation
> TIP measurements taken at reactors that have operated in extended power uprate con'ditions indicate that the calculation accuracy is not impacted 28 Rchlab~I, ,WA -July31. August 2, 2007 AREVA NPINC. Brunswick Fuel Transition Brunswick Fuel NRC Meeting LAR NRC Transition LAR Meeting RichOdIWA.'"
. . _V,,: . ýý ý. 4C" -. JU1Y "ji ` August 2, 2007
. -7 28
Additional Validation Power uprate experience shows that uncertainties are unchanged I Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 29 I AREVA NP INC.
Conclusion
" The neutronic methodology utilizing CASMVO-4 and MICROBURN-B2 accurately models reactor cores with a wide range of operating conditions including those anticipated for EPU at Brunswick
" The uncertainties presented in EMF-2158(P)(A) continue to be applicable for EPU operation at Brunswick August 2, 2007 30 Brunswick NRC Meeting LAR NRC Transition LAR Fuel Transition Richiand, WA July 31 -
AREVA Brunswick Fuel Meeting Richland, WA July 31 - August 2, 2007 30
CA SMO-4IMICROB URN-B 2 Methodology Experience Relative to BrunswickApplication Ralph Grummer Manager, Nuclear Technology Brunswick Fuel Transition LAR NRC Meeting RcinW Richland, WA uy1-Ags July 31 - August 2,,20 2007 1
BWR Methodology Experience
> C)bjective
" Describe the experience base for AREVA NP methodologies
- Demonstrate that the:AR EVA Methodology isapplicable to EPU conditions at Brunswi ck 2
MR NRC Meeting NRC Meeting Richiand, WA July31 - August 2, 2007 AREVA NPINC. Brunswick Fuel Brunswick Transition LAR Fuel Transition Richland, WA July 31 - August 2, 2007 2 I
AREA Topical Report Thermal-Hydraulic Conditions NOPO RITR AR AN INC..
Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 9
- August Z 2007
BWR MNethodology Experience Current Experience is consistent with the topical report I Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 4 I AREVA NPINC.
Topical Report Thermal-Hydraulic Conditions AREVA NP INC. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 5 L
BWR Methodology Experience Current Experience is consistent with the topical report AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 6
BWR Methodology Experience
> At the point of the highest exit void fraction, additional detail was evaluated
- Core average void axial profile
- Axial profile of the peak assembly
- Histogram of the nodal void fractions incore 7
Richiand, WA July31 August 2, 2007 NRC Meeting LAR NRC AREVA NPINC. Brunswick Fuel I k Brunswick TransitionLAR Fuel Transition Meeting Richland, WA July 31 - August Z 2007 7
A - .
ARrV BWR Methodology Experience 8
LAR NRC Transition LAR Meeting NRC Meeting Richiand, WA July31 - August 2, 2007 Brunswick Fuel Transition Brunswick Fuel Richland, WA July 31 - August 2, 2007 8
BWR Methodology Experience Current Experience has Similar Void Population as Expected for SQH Power Uprate Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 9 I AREVA NPINC. I
Evaluation of Power Uprate for Brunswick r
Max assembly powers are less than those presented in the topical report AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 10
Evaluation of Power Uprate for Brunswick r
Max exit voids are less than those presented in the topical report 11 Richland, WA July 31 August 2, 2007 NRC Meeting Brunswick Fuel Transition Brunswick Fuel LAR NRC Transition LAR Meeting Richland, WA July 31 - August Z 2007 11 I AREVA NPINC. II
Brunswick with Power Uprate r
12 July 31 August 2, 2007 AREVA NýP INC.
I Brunswick Fuel Brunswick LAR NRC Transition LAR Fuel Transition Meeting NRC Meeting Rich/and, WA Richland, WA July 31 - August 2, 2007 12
Brunswick with Power Uprate 13 LAR NRC Meeting NRC Meeting Richiand, WA July31 - August 2, 2007 AREVA Brunswick Fuel Transition Brunswick Fuel Transition LAR Richland, WA July 31 - August 2, 2007 13
Experience with High Void Fractions
> Conclusions
- Reactor conditions for Brunswick with power uprate are not significantly different from current experience
- The range of void fractions inthe topical report data exceeds that expected for the power uprate conditions
- The distribution of voids is nearly the same as current experience
- Cross section representation is accurate for power uprate conditions 14 Richiand, WA July 31 August 2, 2007 NRC Meeting LAR NRC -
AREVA N Brunswick Brunswick Fuel Transition LAR Fuel Transition Meeting Richland, WA July 31 - August 2, 2007 14
Power DistributionUncertainties
> C)bjective
- Describe the process used by AREVA NP to define the power distribution uncertainties
- Demonstrate that the AREVA Methodology isapplicable to EPU conditions at Brunswick 15 Transition LAR Fuel Transition Meeting NRC Meeting LAR NRC Richiand, WA July31 -August 2, 2007 Brunswick Brunswick Fuel Richland, WA July 31 - August 2, 2007 15
Power DistributionUncertainties
" First we will look at how AREVA determined the measured power distribution uncertainties
" One of the major components isthe comparison of measured and calculated TIPs
" This includes measurement uncertainty as well as calculation uncertainty 16 Meeting NRC Meeting LAR NRC WA Richiand, WA July 31 August 2, 2007 Brunswick Transition LAR Fuel Transition Brunswick Fuel Richland, July 31 - August 2, 2007 16 i AREVA NPINC.
ARrV Power Distribution Uncertain ties Brunswick Brunswick Fuel Fuel Transition Transition LAR LAR NRC NRC Meeting Meeting Richiand, Richland, WA WA July31 -
July 31 August 2~2007
- August 2, 2007
Powe'rDistribution Uncertainties 18 Richiand, WA July 31 August 2, 2007 NRC Meeting AREVA NPINC. Fuel Transition Brunswick Fuel Brunswick LAR NRC Transition LAR Meeting Richland, WA July 31 - August 2, 2007 18
Po wer Dis tribution Uncertainties I AREVA NPINC Brunswick Fuel Transition LAR NRC Meeting Brunwic Richland, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 119
Power DistributionUncertainties I July 31 August 2, 2007 20 AREVA NPINC. Brunswick Fuel Brunswick LAR NRC Transition LAR Fuel Transition Meeting NRC Meeting Richiand, WA Richland, WA July 31 - August 2, 2007 20
A
- kV Power DistributionUncertainties
" Axial power distribution uncertainties were determined by the simple relationship
" Nodal = radial
- axial
" 6Nodal 2 = 6radial 2 + 6axial 2
" Axial uncertainty was determined to be 1.81 % for C-lattice plants and 2.91 % for D-lattice plants
" Another component might be. the Iradial uncertainty at an axial level
" The EMF-2158(P)(A) data was reevaluated by looking at the deviations between measured and calculated TIP response for each axial level NOPORITR k 21 NRC Meeting LAR NRC Richland WA July 31 - August 2, 2007 Brunswick Brunswi.ck Fuel Transition LAR Fuel Transition Meeting Richland, WA' ý, ' ' 4 july 31 - August 2, 2007 21
Power DistributionUncertainties There does not appear to be any axial dependency on the standard deviation AREVA NP INC. Brunswick Fuel Transition LAR NRC Meeting Rich/and, WA July 31 - August 2, 2007 22
Purpose
" Gamma scans have bee n used to measure the assembly and individual rod power distribution
" These measurements are used to validate core physics methods and determine the associated uncertainties AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Richiand, WA Juy323ugs ,20
Gamma Scan Measurements
" Gamma scans measure the relative gamma flux resulting from isotopic decay
" Certain isotopes can be identified by gamma spectroscopy
" Power measurements target the gamma spectrum associated with La140
" La140 isa decay product of Ba140 which isdirect fission product 24 LAR NRC Transition LAR Meeting NRC Meeting Richiand, WA July:31 - August 2, 2007 AREVA Brunswick Fuel Transition Brunswick Fuel Richland,.WA ýJulyý31 August2,2007 24
Gamma Scan Measurements
" The half life of Ba140 is 12.8 days
" 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 />
" La140 activity istherefore related to the density of Ba 140
" The Ba140 density isrepresentative of the integrated fissions over the last 25 days
" Gamma scan measurements need to be taken shortly after shutdown before the Ba140 decays to undetectable levels August 2, 2007 25 Transition LAR Fuel Transition Meeting NRC Meeting MR NRC Richiand, WA July31 -
Brunswick Fuel Brunswick Richland, WA July 31 - August 2, 2007 25
Gamma Scan Equipment
> Equipment istailored to the specific application
- Assembly scans use a broad window to capture gamma particles from all of the rods
- Individual rod scans use a narrow window to isolate the rod oAn axial level measurement uses a broader (axial) window to get a higher count rate
- Axial scans use a narrow (axial) window to get a finer resolution 26 Richiand, WA July 31 August 2, 2007 NRC Meeting LAR NRC AREVA NPINC. Brunswick Fuel IL Brunswick Transition LAR Fuel Transition Meeting Richland, WA July.31 - August 2, 2007 26 I
Gamma Scan Equipment
> Gamma scan measurements are performed on individual fuel rods removed from assemblies using a high-purity germanium (HPGe) detector and an underwater collimator assembly 27 Transition LAR Fuel Transition Meeting NRC Meeting MR NRC Rich/and, WA July31 -August 2, 2007 AREVA NPINC. Brunswick Fuel Brunswick Richland, WA July 31 - August Z 2007 27 I
Gamma Scan Comparisons
> In order to compare core physics models to the gamma scan results the calculated pin power distribution isconverted into a Ba140 density distribution.
" A mathematical process using; CASMVO-4 pin nuclide inventory and MICROBURN-B52 nuclide inventory is used
" This is an additional uncertainty inthe overall comparison AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July.31 7~August 2, 2007 28
BWR Power Distribution Uncertainties
" There isvery limited data on measured power distributions
" The measured power isdetermined by modifying the calculated power distribution using the measured and calculated LPRM values SMeasured LPRM values are calibrated to the TIP measurements
" Assembly gamma scan measurements at Quad Cities were used to define the uncertainty of the correlation coefficients
" These correlation coefficients indicate the accuracy of the "UPDATE" methodology August 2, 2007 29 Transition LAR Fuel Transition Meeting NRC Meeting LAR NRC Richiand, WA July31 -
IP INC. Brunswick Brunswick Fuel Richland,: WA ý ý ý July 31 - August 2, 2007 29 L
BWR Power DistributionUncertainties
" The Bundle Correlatio'n' Coeffi~cient for QC Cycle 2 was []
" The Bundle Correlation Coefficient for QC Cycle 4 was[]
" The average value of [ ]was used inthe determination of the measured power uncertainty
" Using the minimum correlation coefficient increases the measured uncertainty by [ 1%
" Using the maximum correlation coefficient decreases the measured uncertainty by [ ]
AREVA NP INC. Brunswick Fuel Transition LAR NRC Meeting RichlanI WA> July,31 August 2,2007 30 L
Gamma Scan Data
" Pin-by-Pin Gamma sc-an.-data is'used for verification of the local peaking uncertainty
" Quad Cities Data indicated that this uncertainty was approximately [ 1%
" KWU measurements of 9X9 and ATRIUM-10 assemblies provided additional validation that this uncertainty was accurate
" Comparisons to Monte Carlo calculations indicated an uncertainty of approximately [ 1%
31 LAR NRC Meeting NRC Meeting Richland, WA 1 July 31 - August 2, 2007 AREVA N Brunswick Fuel Brunswick Transition LAR Fuel Transition ,ýich:land,'VV-"ý"-ý J 1 31 - August Z 2007 31
Quad Cities Gamma Scan Benchmark Results EMF-2158(P) (A) pp. 8-6,7 This data includes measurement uncertainty.
Local power distribution uncertainty is not axial level dependent AREVA Brunswick Fuel TransitionLAR NRC Meeting Richland, WA July 31 - August Z 2007 32
Local Peaking Uncertainty
> Recent gamma scan measurements including ATRIUM-10 show similar comparisons a~t vario~us axial levels
> These results do not indicate any trend relative to axial position 33 Transition LAR Fuel Transition Meeting NRC Meeting LAR NRC Richiand, WA. July31 -August 2, 2007 Brunswick Fuel Brunswick Richland, WA. - July 31 -August Z 2007 33
KWU-S Gamma Scan Benchmark Results EMF-2158(P) (A) pp. 8-8 Local power distribution uncertainty is not axial level dependent AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 34
KWU-S Gamma Scan Benchmark Results EMF-2 158(P) (A)
" Full axial scans were performed on 16 fuel rods
" Comparisons to calculated data show excellent agreement at all axial levels
" The dip inpower associated with spacers is not modeled in MICROBURN-B32
" There is no indication of reduced accuracy at higher void fractions AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Brunwic Rich/and, Ful TanstionLARNRCMeeing WA July 31 - August 2, 2007 335 I
KWU-S Gamma Scan Benchmark Results EMF-2 158(P) (A)
Measurements were performed for moderate void fractions AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 36
KWU-S Gamma Scan Benchmark Results EMF-2 158(P) (A)
AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 37
KWU-S Gamma Scan Benchmark Results EMF-2 158(P) (A)
Indication that the higher voids are accurately represented AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 38
Results Benchmark (A)
Gamma Scan EMF-2 158(P)
KWU-S 39 represented are accurately higher voids that the July31
- August
- 2. 2007 Indication Richland, WA Meeting MAR NRC Fuel Transition i Brunswick NP INC, REVA
KWU-S Gamma Scan Benchmark Results EMF-2 158(P) (A)'
Indication that the higher voids are accurately represented AREVA NPINC. Brunswick Fuel Transition LAR NRC Meeting Richland, WA July31 - August 2, 2007 40
KWU-S Gamma Scan Benchmark Results EMF-2 158(P) (A)
Indication that the higher voids are accurately represented Brunswick Fuel Transition LAR NRC Meeting Richland, WA July 31 - August 2, 2007 41
Power Distribution Uncertainties
" Gamma scanning provides data on relative local and radial power during last few weeks of operation
" Uncertainty in gamma scan results has small effect on measured radial power distribution uncertainty
- 50% decrease incorrelation coefficient results in0.4% increase in measured radial power distribution uncertainty
- Additional ATRIUM-10 gamma scan data would not significantly affect measured power distribution uncertainty
" Local gamma scan data available for various designs
- 11 assemblies intwo reactors
" 7x7, 8x8, 9x9, ATRI UM-l10
" Exposures include once- and twice-burned assemblies
" Various gadolinia concentrations
- Various water rod configurations
" No void dependence observed for local power uncertainties
- More ATRIUM-10 gamma scanning is not expected to change uncertainties I AREVA NPINC. I Brunswick Fuel Transition LAR NRC Meeting Richl~hd,'.WA:~-ý ' -'Jbly'3i -^'August 2, 2007 42
B WR Power Distribution Uncertainty
> Conclusion
- Recent gamma scan d-ata has confirmed the local power uncertainty
- There is no axial dependency inthe uncertainty
- There is no void dependency inthe local peaking power uncertainty
- Current uncertainties are applicable to Brunswick with power uprate conditions 43 LAR NRC Meeting NRC Meeting Richiand, WA July31 - August 2, 2007 AREVA NPINC. Brunswick Fuel Brunswick Transition LAR Fuel Transition Richland, WA July 31 - August 2, 2007 43
AREVA NPINC. I Brunswick Fuel Transition LAR NRC Meeting Richland, WA. July 31 - August 2, 2007 44 L