ML13276A066

ANP-3153(NP), Revision 0, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for Atrium Tm 1OXM Fuel, Enclosure 8

ML13276A066
Person / Time
Site:	Browns Ferry
Issue date:	08/31/2013
From:	AREVA NP
To:	Office of Nuclear Reactor Regulation
References
ANP-3153(NP), Rev 0
Download: ML13276A066 (55)

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{{#Wiki_filter:Enclosure 8ANP-3153NP, "Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits forATRIUMTM 1OXM Fuel"- Non Proprietary ANP-3153(NP) Revision 0Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM1OXM FuelAugust 2013AAREVAAREVA NP Inc. AREVA NP Inc.ANP-3153(NP) Revision 0Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM1OXM Fuel AREVA NP Inc.ANP-3153(NP) Revision 0Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelCopyright © 2013AREVA NP Inc.All Right Reserved Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM1OXM.Fuel ANP-3153(NP) Revision 0Page iNature of ChangesItemPage Description and Justification

1. All This is the initial releaseAREVA NP Inc.

Browns Ferry Units 1,2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 10XM Fuel Page iiContents1.0 Introduction ................................................... 1-12 .0 S u m m a ry ...................................................................................................................... 2-13.0 LOCA Description ......................................................................................................... 3-13.1 Accident Description ......................................................................................... 3-13.2 Acceptance Criteria ........................................................................................... 3-24.0 LOCA Analysis Description ........................................................................................... 4-14.1 Blowdown Analysis ........................................................................................... 4-14.2 Refill / Reflood Analysis .................................................................................... 4-24.3 Heatup Analysis ................................................................................................ 4-24.4 [] .................................................. 4-34 .4 .1 [ ...................................................................... 4 -44.5 Plant Parameters .............................................................................................. 4-44.6 ECCS Parameters ............................................................................................ 4-45.0 MAPLHGR Analysis Results ......................................................................................... 5-16.0 Conclusions .................................................................................................................. 6-17.0 References ................................................................................................................... 7-1Appendix A Supplemental Information .......................................................................... A-1TablesTable 2.1 LOCA Results for PCT Limiting Conditions ........................... 2-3Table 4.1 Initial Conditions ..................................................................................................... 4-6Table 4.2 Reactor System Parameters ................................................................................... 4-7Table 4.3 ATRIUM 1OXM Fuel Assembly Parameters ............................................................ 4-8Table 4.4 High-Pressure Coolant Injection Parameters .......................................................... 4-9Table 4.5 Low-Pressure Coolant Injection Parameters ......................................................... 4-10Table 4.6 Low-Pressure Core Spray Parameters ................................................................. 4-11Table 4.7 Automatic Depressurization System Parameters .................................................. 4-12Table 4.8 Recirculation Discharge Isolation Valve Parameters ............................................. 4-13Table 4.9 ECCS Single Failure ............................................................................ ......4-14Table 5.1 Event Times for Limiting Break 0.20 ft2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial 102% Power ......................................................... 5-2Table 5.2 ATRIUM 1OXM MAPLHGR Analysis Results .......................................................... 5-3Table A.1 Computer Codes Used for MAPLHGR Limit Analysis ............................................ A-2AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 10XM Fuel Page iiiFiguresFigure 2.1 MAPLHGR Limit for ATRIUM 10XM Fuel ............................................................... 2-4Figure 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model .............................. 4-15Figure 4.2 [ ] ........................ 4-16Figure 4.3 RELAX System Blowdown Model ......................................................................... 4-17Figure 4.4 RELAX Hot Channel Blowdown Model Top-Peaked Axial .................................... 4-18Figure 4.5 E C C S S chem atic ................................................................................................. 4-19Figure 4.6 Axial Power Distribution for Limiting LOCA Case in RELAX Calculation ............... 4-20Figure 5.1 Limiting Break Upper Plenum Pressure .................................................................. 5-4Figure 5.2 Limiting Break Total Break Flow Rate .................................................................... 5-4Figure 5.3 Limiting Break ADS Flow Rate ............................................................................... 5-5Figure 5.4 Limiting Break HPCI Flow Rate .............................................................................. 5-5Figure 5.5 Limiting Break LPCS Flow Rate ............................................................................. 5-6Figure 5.6 Limiting Break Intact Loop LPCI Flow Rate ............................................................ 5-6Figure 5.7 Limiting Break Broken Loop LPCI Flow Rate .......................................................... 5-7Figure 5.8 Limiting Break Upper Downcomer Mixture Level .................................................... 5-7Figure 5.9 Limiting Break Middle Downcomer Mixture Level ................................................... 5-8Figure 5.10 Limiting Break Lower Downcomer Mixture Level .................................................. 5-8Figure 5.11 Limiting Break Intact Loop Discharge Line Liquid Mass ........................................ 5-9Figure 5.12 Limiting Break Upper Plenum Liquid Mass ........................................................... 5-9Figure 5.13 Limiting Break Lower Plenum Liquid Mass ......................................................... 5-10Figure 5.14 Limiting Break Hot Channel Inlet Flow Rate ........................ 5-10Figure 5.15 Limiting Break Hot Channel Outlet Flow Rate .................................................... 5-11Figure 5.16 Limiting Break Hot Channel Coolant Temperature at the Limiting Node ............. 5-11Figure 5.17 Limiting Break Hot Channel Quality at the Limiting Node ................................... 5-12Figure 5.18 Limiting Break Hot Channel Heat Transfer Coefficient at the Limiting Node ....... 5-12Figure 5.19 Limiting Break Cladding Temperatures .............................................................. 5-13AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page ivADSANSBWRCFRCLTPCMWRDEGNomenclature automatic depressurization systemAmerican Nuclear Societyboiling water reactorCode of Federal Regulations current licensed thermal power (3458 MWt)core average metal-water reactiondouble-ended guillotine emergency core cooling systemend of blowdownhigh-pressure coolant injection linear heat generation rateloss-of-coolant accidentlow-pressure coolant injection low-pressure core spraymaximum average planar linear heat generation rateminimum critical power ratiomaximum extended load line limit analysismetal-water reactionECCSEOBHPCILHGRLOCALPCILPCSMAPLHGRMCPRMELLLAMWRNRCOLTPPCTRDIVNuclear Regulatory Commission, U.S.original licensed thermal powerpeak cladding temperature recirculation discharge isolation valveSF-ADSSF-ADSIIL SF-ADSISV SF-BATTSF-BATTIBA SF-BATTIBB SF-BATTIBC SF-DGENsingle failure of ADSsingle failure of ADS, initiation logicsingle failure of ADS, single valvesingle failure of battery (DC) powersingle failure of battery (DC) power, board Asingle failure of battery (DC) power, board Bsingle failure of battery (DC) power, board Csingle failure of a diesel generator AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page vSF-HPCISF-LOCASF-LPCISLONomenclature (Continued) single failure of the HPCI systemsingle failure of opposite unit false LOCA signalsingle failure of a LPCI valvesingle-loop operation AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 1-11.0 Introduction The results of the loss-of-coolant accident emergency core cooling system (LOCA-ECCS) analyses for Browns Ferry Units 1, 2 and 3 are documented in this report. The purpose of theLOCA-ECCS analysis is to specify the maximum average planar linear heat generation rate(MAPLHGR) limit versus exposure for ATRIUMTM 1OXM* fuel and to demonstrate that theMAPLHGR limit is adequate to ensure that the LOCA-ECCS criteria in 10 CFR 50.46 aresatisfied for operation at or below the limit. The report also documents the licensing basis peakcladding temperature (PCT) and corresponding local cladding oxidation from the metal waterreaction (MWR) for ATRIUM 1 OXM fuel used at Browns Ferry Units 1, 2, and 3.The analyses documented in this report were performed with LOCA Evaluation Modelsdeveloped by AREVA NP and approved for reactor licensing analyses by the U.S. NuclearRegulatory Commission (NRC). The models and computer codes used by AREVA for LOCAanalyses are collectively referred to as the EXEM BWR-2000 Evaluation Model. The EXEMBWR-2000 Evaluation Model and NRC approval are documented in Reference

2. A summarydescription of the LOCA analysis methodology is provided in Section 4.0.The application of the EXEM BWR Evaluation Model for the Browns Ferry LOCA breakspectrum analysis is documented in Reference

1. The LOCA conditions evaluated inReference 1 include break size, type, location, axial power shape, and ECCS single failure.

Thelimiting LOCA break characteristics identified in Reference 1 are presented below:Limiting LOCABreak Characteristics Location recirculation discharge pipeType / size split / 0.20 ft2Single failure battery (DC) power, board AAxial power shape top-peaked Initial State [ I* ATRIUM is a trademark of AREVA NP.AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 10XM Fuel Page 1-2The LOCA break spectrum analysis documented in Reference 1 was based on a genericATRIUM 1OXM neutronic design at beginning of life conditions. The PCT and MWR calculated for a fuel rod experiencing the fluid conditions during the limiting LOCA are affected by fuelcharacteristics that depend on the fuel assembly neutronic design and exposure (e.g., local rodpower, stored energy). The fuel assembly heatup analysis results presented in this report arefor the limiting (minimum margin to acceptance criteria) ATRIUM 1OXM neutronic designcurrently designed for use at Browns Ferry Units 1, 2, and 3. The heatup analyses wereperformed using the fluid conditions from the limiting LOCA identified in Reference

1. Cyclespecific heatup analyses are performed to confirm that the results in this report remain boundingfor nuclear designs used in each core design.Calculations assumed an initial core power of 102% of 3458 MWt as per NRC requirements.

3458 MWt corresponds to 105% of the original licensed thermal power (OLTP) and is referred toas the current licensed thermal power (CLTP).AREVA NP Inc. Browns Ferry Units 1,2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 10XM Fuel Page 2-12.0 SummaryThe MAPLHGR limit was determined by applying the EXEM BWR-2000 Evaluation Model for theanalysis of the limiting LOCA event. The exposure-dependent MAPLHGR limit for ATRIUM1OXM fuel is shown in Figure 2.1. [The response of the reactor system and hot channel during the limiting LOCA analysis fromReference 1 are presented in Section 5.0. The MAPLHGR analysis results for the limiting latticedesign are presented in Section 5.0. The peak cladding temperature (PCT) and metal-water reaction (MWR) results for the ATRIUM 1OXM limiting lattice design are presented in Table 2.1.The SLO analyses (Reference

1) support operation with an ATRIUM 1OXM MAPLHGR multiplier of 0.85 applied to the normal two-loop operation MAPLHGR limit. The results of thesecalculations confirm that the LOCA acceptance criteria in the Code of Federal Regulations (10CFR 50.46) are met for operation at or below these limits.Note that the analysis PCT documented in this report is lower than the limiting PCT given in theLOCA break spectrum report (Reference 1). [AREVA NP Inc.

Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 2-2.[IAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 2-3Table 2.1 LOCA Results forPCT Limiting Conditions Parameter Exposure (GWd/MTU) Peak cladding temperature (*F)Local cladding oxidation (max %)Planar average oxidation (max %)Total hydrogen generated (% of total hydrogen possible) ATRIUM 1OXM0.019031.160.65<1.0AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM1OXM FuelANP-3153(NP) Revision 0Page 2-416.014.0CL12.010.08.06.04.0.0 10.0 20.0 30.0 40.0 50.0 60.0Planar Average Exposure (GWd/MTU) 70.0Average Planar ATRIUM 1OXMExposure MAPLHGR(GWd/MTU) (kW/ft)0 13.015 13.067 7.6Figure 2.1 MAPLHGR Limitfor ATRIUM 1OXM FuelAREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 1OXM Fuel Page 3-13.0 LOCA Description 3.1 Accident Description The LOCA is described in the Code of Federal Regulations 10 CFR 50.46 as a hypothetical accident that results in a loss of reactor coolant from breaks in reactor coolant pressureboundary piping up to and including a break equivalent in size to a double-ended rupture of thelargest pipe in the reactor coolant system. There is not a specifically identified cause thatresults in the pipe break. However, for the purpose of identifying a design basis accident, thepipe break is postulated to occur inside the primary containment before the first isolation valve.For a BWR, a LOCA may occur over a wide spectrum of break locations and sizes. Responses to the break vary significantly over the break spectrum. The largest possible break is a double-ended rupture of a recirculation pipe; however, this is not necessarily the most severe challenge to the emergency core cooling system (ECCS). A double-ended rupture of a main steam linecauses the most rapid primary system depressurization, but because of other phenomena, steam line breaks are seldom limiting with respect to the criteria of 10 CFR 50.46. Specialanalysis considerations are required when the break is postulated to occur in a pipe that is usedas the injection path for an ECCS (e.g. core spray line). Although these breaks are relatively small, their existence disables the function of an ECCS. In addition to break locationdependence, different break sizes in the same pipe produce quite different event responses, and the largest break area is not necessarily the most severe challenge to the event acceptance criteria. Because of these complexities, an analysis covering the full range of break sizes andlocations is required. The results of the Browns Ferry ATRIUM 1OXM break spectrumcalculations using EXEM BWR-2000 LOCA methodology are summarized in Reference 1.Regardless of the initiating break characteristics, the event response is conveniently separated into three phases: the blowdown phase, the refill phase, and the reflood phase. The relativeduration of each phase is strongly dependent upon the break size and location. The last twophases are often combined and will be discussed together in this report.During the blowdown phase of a LOCA, there is a net loss-of-coolant inventory, an increase infuel cladding temperature due to core flow degradation, and for the larger breaks, the corebecomes fully or partially uncovered. There is a rapid decrease in pressure during theblowdown phase. During the early phase of the depressurization, the exiting coolant providescore cooling. Later in the blowdown, core cooling is provided by lower plenum flashing as theAREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 1OXM Fuel Page 3-2system continues to depressurize. The blowdown phase is defined to end when LPCS reachesrated flow.In the refill phase of a LOCA, the ECCS is functioning and there is a net increase of coolantinventory. During this phase the core sprays provide core cooling and, along with low-pressure and high-pressure coolant injection (LPCI and HPCI), supply liquid to refill the lower portion ofthe reactor vessel. In general, the core heat transfer to the coolant is less than the fuel decayheat rate and the fuel cladding temperature continues to increase during the refill phase.In the reflood phase, the coolant inventory has increased to the point where the mixture levelreenters the core region. During the core reflood phase, cooling is provided above the mixturelevel by entrained reflood liquid and below the mixture level by pool boiling. Sufficient coolanteventually reaches the core hot node and the fuel cladding temperature decreases. 3.2 Acceptance CriteriaA LOCA is a potentially limiting event that may place constraints on fuel design, local powerpeaking, and in some cases, acceptable core power level. During a LOCA, the normal transferof heat from the fuel to the coolant is disrupted. As the liquid inventory in the reactor decreases, the decay heat and stored energy of the fuel cause a heatup of the undercooled fuel assembly. In order to limit the amount of heat that can contribute to the heatup of the fuel assembly duringa LOCA, an operating limit on the MAPLHGR is applied to each fuel assembly in the core.The Code of Federal Regulations prescribes specific acceptance criteria (10 CFR 50.46) for aLOCA event as well as specific requirements and acceptable features for Evaluation Models(10 CFR 50 Appendix K). The conformance of the EXEM BWR-2000 LOCA Evaluation Modelsto Appendix K is described in Reference

2. The ECCS must be designed such that the plantresponse to a LOCA meets the following acceptance criteria specified in 10 CFR 50.46:* The calculated maximum fuel element cladding temperature shall not exceed 2200'F.* The calculated local oxidation of the cladding shall nowhere exceed 0.17 times the localcladding thickness.

The calculated total amount of hydrogen generated from the chemical reaction of thecladding with water or steam shall not exceed 0.01 times the hypothetical amount thatwould be generated if all of the metal in the cladding cylinders surrounding the fuel,except the cladding surrounding the plenum volume, were to react.AREVA NP Inc. Browns Ferry Units 1,2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 10XM Fuel Page 3-3Calculated changes in core geometry shall be such that the core remains amenable tocooling.After any calculated successful operation of the ECCS, the calculated core temperature shall be maintained for the extended period of time required bythe long-lived radioactivity remaining in the core.These criteria are commonly referred to as the peak cladding temperature (PCT) criterion, thelocal oxidation criterion, the hydrogen generation criterion, the coolable geometry criterion, andthe long-term cooling criterion. A MAPLHGR limit is established for each ATRIUM IOXM fueltype to ensure that these criteria are met. For jet pump BWRs, the most challenging criterion isthat PCT must not exceed 2200'F.LOCA analysis results demonstrating that the PCT, local oxidation, and hydrogen generation criteria are met are provided in Section 5.0. Compliance with these three criteria ensures that acoolable geometry is maintained. Compliance with the long-term coolability criterion isdiscussed in Reference 1.AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 10XM Fuel Page 4-14.0 LOCA Analysis Description The Evaluation Model used for the break spectrum analysis is the EXEM BWR-2000 LOCAanalysis methodology described in Reference

2. The EXEM BWR-2000 methodology employsthree major computer codes to evaluate the system and fuel response during all phases of aLOCA. These are the RELAX, HUXY, and RODEX2 computer codes. RELAX is used tocalculate the system and hot channel response during the blowdown, refill, and reflood phasesof the LOCA. The HUXY code is used to perform heatup calculations for the entire LOCA, andcalculates the PCT and local clad oxidation at the axial plane of interest.

RODEX2 is used todetermine fuel parameters (such as stored energy) for input to the other LOCA codes. Thecode interfaces for the LOCA methodology are illustrated in Figure 4.1.A complete analysis for a given break size starts with the specification of fuel parameters usingRODEX2 (Reference 3). RODEX2 is used to determine the initial stored energy for both theblowdown analysis (RELAX system and hot channel) and the heatup analysis (HUXY). This isaccomplished by ensuring that the initial stored energy in RELAX and HUXY is the same orhigher than that calculated by RODEX2 for the power, exposure, and fuel design beingconsidered. 4.1 Blowdown AnalysisThe RELAX code (Reference

2) is used to calculate the system thermal-hydraulic responseduring the blowdown phase of the LOCA. For the system blowdown

analysis, the core isrepresented by an average core channel.

The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and with decay heatingas required by Appendix K of 10 CFR 50. The reactor vessel nodalization for the systemblowdown analysis is shown in Figure 4.3. This nodalization is consistent with that used in thetopical report submitted to the NRC (Reference 2).The RELAX analysis is performed from the time of the break initiation through the end ofblowdown (EOB). The system blowdown calculation provides the upper and lower plenumtransient boundary conditions for the hot channel analysis. Following the system blowdown calculation, another RELAX analysis is performed to analyzethe maximum power assembly (hot channel) of the core. The RELAX hot channel calculation isused to calculate hot channel fuel, cladding, and coolant temperatures during the blowdownAREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 1OXM Fuel Page 4-2phase of the LOCA. The RELAX hot channel nodalization is shown in Figure 4.4 for a top-peaked power shape. The hot channel analysis is performed using the system blowdownresults to supply the core power and the system boundary conditions at the core inlet and exit.The results from the RELAX hot channel calculation used as input to the HUXY heatup analysisare heat transfer coefficients and fluid conditions in the hot channel.4.2 Refill / Reflood AnalysisThe RELAX code is used to compute the system and hot channel hydraulic response during therefill/reflood phase of the LOCA. The RELAX system and RELAX hot channel analysescontinue beyond the end of blowdown to analyze system and hot channel responses during therefill and reflood phases. The refill phase is the period when the lower plenum is filling due toECCS injection. The reflood phase is when some portions of the core and hot assembly arebeing cooled with ECCS water entering from the lower plenum. The purpose of the RELAXcalculations beyond blowdown is to determine the time when the liquid flow via upwardentrainment from the bottom of the core becomes high enough at the hot node in the hotassembly to end the temperature increase of the fuel rod cladding. This event time is called thetime of hot node reflood. [] The time when the core bypass mixture level rises to the elevation ofthe hot node in the hot assembly is also determined. RELAX provides a prediction of fluid inventory during the ECCS injection period. Allowing forcountercurrent flow through the core and bypass, RELAX determines the refill rate of the lowerplenum due to ECCS water and the subsequent reflood times for the core, hot assembly, andthe core bypass. The RELAX calculations provide HUXY with the time of hot node reflood andthe time when the liquid has risen in the bypass to the height of the axial plane of interest (timeof bypass reflood). 4.3 Heatup AnalysisThe HUXY code (Reference

4) is used to perform heatup calculations for the entire LOCAtransient and provides PCT and local clad oxidation at the axial plane of interest.

The heatgenerated by metal-water reaction (MWR) is included in the HUXY analysis. HUXY is used tocalculate the thermal response of each fuel rod in one axial plane of the hot channel assembly. These calculations consider thermal-mechanical interactions within the fuel rod. The cladswelling and rupture models from NUREG-0630 have been incorporated into HUXYAREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 10XM Fuel Page 4-3(Reference 5). The HUXY code complies with the 10 CFR 50 Appendix K criteria for LOCAEvaluation Models.HUXY uses the end of blowdown time and the times of core bypass reflood and core reflood atthe axial plane of interest from the RELAX analysis. []Throughout the calculations, decay power is determined based on the ANS 1971 decay heatcurve plus 20% as described in Reference

2. [] are used in the HUXY analysis.

The principal results of a HUXY heatupanalysis are the PCT and the percent local oxidation of the fuel cladding, often called thepercent maximum local metal water reactor (%MWR). The core average metal-water reaction(CMWR) criterion of less than 1.0% can often be satisfied by demonstrating that the maximumplanar average MWR calculated by HUXY is less than 1.0%.4.4 [AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUM- 1OXM FuelANP-3153(NP) Revision 0Page 4-44A4.1III4.5 Plant Parameters The LOCA break spectrum analysis is performed using the plant parameters presented inReference

7. Table 4.1 provides a summary of reactor initial conditions used in the breakspectrum analysis.

Table 4.2 lists selected reactor system parameters. The break spectrum analysis is performed for a full core of ATRIUM 1OXM fuel. Some of thekey fuel parameters used in the break spectrum analysis are summarized in Table 4.3. A top-peaked axial power shape, shown in Figure 4.6, was identified as the most conservative powershape for the limiting break (Reference 1).4.6 ECCS Parameters The ECCS configuration is shown in Figure 4.5. Tables 4.4 -4.8 provide the important ECCScharacteristics assumed in the LOCA break spectrum analysis. The ECCS is modeled as filljunctions connected to the appropriate reactor locations: LPCS injects into the upper plenum,HPCI injects into the upper downcomer, and LPCI injects into the recirculation line.The flow through each ECCS valve is determined based on system pressure and valve position. Flow versus pressure for a fully open valve is obtained by linearly interpolating the pumpcapacity data provided in Tables 4.4 -4.6. For the break spectrum

analyses, no credit forECCS flow is assumed until ECCS pumps reach rated speed.AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 1OXM Fuel Page 4-5The automatic depressurization system (ADS) valves are modeled as a junction connecting thereactor steam line to the suppression pool. The flow through the ADS valves is calculated based on pressure and valve flow characteristics. The valve flow characteristics are determined such that the calculated flow is equal to the rated capacity at the reference pressure shown inTable 4.7.In the AREVA LOCA analysis model, ECCS initiation is assumed to occur when the water leveldrops to the applicable level setpoint. No credit is assumed for the start of HPCI, LPCS, orLPCI due to high drywell pressure. []The recirculation discharge isolation valve (RDIV) parameters are shown in Table 4.8.The potentially limiting single failures of the ECCS are provided in Section 5.0 of Reference 1.Table 4.9 shows these failures and gives the ECCS systems that are available for eachassumed failure.AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 4-6Table 4.1 Initial Conditions* Reactor power (% of rated) 102Total core flow (% of rated) [ ]Reactor power (MWt) 3527Total core flow (Mlb/hr) [[ ]Steam flow rate (Mlb/hr) 14.50Steam dome pressure (psia) 1054Core inlet enthalpy (Btu/Ib) [ ]ATRIUM 1OXM hot assemblyMAPLHGR (kW/ft) 13.0[ ]ECCS fluid temperature ('F) 120Axial power shape Fig. 4.6The AREVA calculated heat balance is adjusted to match the 100% power/100% flow values given inthe plant parameters document (Reference 7). The model is then rebalanced based on AREVA heatbalance calculations to establish these LOCA initial conditions at 102% of rated thermal power.IAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM1OXM FuelANP-3153(NP) Revision 0Page 4-7Table 4.2 Reactor SystemParameters Parameter ValueVessel ID (in) 251Number of fuel assemblies 764Recirculation suction pipearea (ft2) 3.5071.0 DEG suction breakarea (ft2) 7.013Recirculation discharge pipe area (ft2) 3.5071.0 DEG discharge breakarea (ft2) 7.013AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 10XM FuelANP-3153(NP) Revision 0Page 4-8Table 4.3 ATRIUM IOXM FuelAssembly Parameters Parameter ValueFuel rod array 10x10Number of fuel rods per 79 (full-length rods)assembly 12 (part-length rods)Non-fuel rod type Water channel replaces9 fuel rodsFuel rod OD (in) 0.4047Active fuel length (in) 150.0 (full-length rods)(including blankets) 75 (part-length rods)Water channel outsidewidth (in) 1.378Fuel channel thickness (in) 0.075 (minimum wall)0.100 (corner)Fuel channel internal width (in) 5.278AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 4-9Table 4.4 High-Pressure CoolantInjection Parameters Parameter ValueCoolant temperature (OF) 120Initiating Signalsand Setpoints Water level* L2 (448 in)High drywell pressure (psig) 2.6 (Not Used)TimeDelaysTime for HPCI pump to reachrated speed and injection valvewide open (sec) 35Delivered Flow RateVersus PressureVessel to FlowDrywell AP Rate(psid) (gpm)0 0150 50001120 50001174 3600* Relative to vessel zero.AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 4-10Table 4.5 Low-Pressure CoolantInjection Parameters Parameter ValueReactor pressure permissive foropening valves (psia) 350Coolant temperature (OF) 120Initiating Signalsand Setpoints Water level* Li (372.5 in)High drywell pressure (psig) 2.6 (Not Used)TimeDelaysTime for LPCI pumps to reachADS permissive (max) (sec)t 32*Time for LPCI pumps to reachrated speed (max) (sec)t 44LPCI injection valve stroke time(sec) 40Delivered Flow RateVersus PressureFlow RateVessel to (gpm)Drywell AP(psid) 2 Pumps Into 4 Pumps Into1 Loop 2 Loops0 17,240 34,48020 16,540 33,080§319.5 0 0Relative to vessel zero.t Includes 13-second delay for diesel generator start. 2-second signal processing delay for water leveltrip Li is assumed in parallel with diesel generator delay.tl Analyses assume the larger delay from LPCS (40 sec) for ADS permissive. Refer to Table 4.6.§ Conservative value relative to specified value in Reference 7 (33,240 gpm). Modeling limitations require the more conservative value of either the specified 4 pumps into 2 loops flow or twice thespecified 2 pumps into 1 loop flow be used.AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 4-11Table 4.6 Low-Pressure CoreSpray Parameters Parameter ValueReactor pressure permissive foropening valves (psia) 350Coolant temperature (OF) 120Initiating Signalsand Setpoints Water level* Li (372.5 in)High drywell pressure (psig) 2.6 (not used)TimeDelaysTime for LPCS pumps to reachADS permissive (max) (sec)t 40Time for LPCS pumps to reachrated speed (max) (sec)t 43LPCS injection valve stroketime (sec) 33Delivered Flow RateVersus PressureFlow RateVessel (gpm)toDrywell AP 2 Pumps 4 Pumps(psid) Into Into1 Sparger 2 Sparger0 6,935 13,870105 5,435 10,870200 3,835 7,670289 0 0* Relative to vessel zero.t Includes 13-second delay for diesel generator start. 2-second signal processing delay for water leveltrip Li is assumed in parallel with diesel generator delay.AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 10XM FuelANP-3153(NP) Revision 0Page 4-12Table 4.7 Automatic Depressurization System Parameters Parameter ValueNumber of valves installed 6Number of valves available 6Minimum flow capacity ofavailable valves(Mlbm/hr at psig) 4.8 at 1125Initiating Signalsand Setpoints Water level* L1 (372.5 in)TimeDelaysDelay time (from ADS initiating signal to time valves are opened)t(sec) 120* Relative to vessel zero.t ADS timer initiation occurs after Li set point trip. ADS valves are opened after the timer has elapsedand LPCS or LPCI pumps reach the ADS ready permissive. Analyses assume the longer delay fromLPCS for ADS ready permissive (see Table 4.6).AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 4-13Table 4.8 Recirculation Discharge Isolation Valve Parameters Parameter ValueReactor pressure permissive forclosing valves -analytical (psia) 215RDIV stroke time (sec) 36AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 4-14Table 4.9 ECCS Single FailureSystems ,tAssumed Remaining Failure Recirculationt Recirculation Suction Break Discharge BreakSF-BATTIBA 6 ADS, 1 LPCS, 2 LPCI 6 ADS, 1 LPCSSF-BATTIBB 4 ADS, HPCI, 1 LPCS, 2 LPCI 4 ADS, HPCI, 1 LPCSSF-BATTIBC§ 4 ADS, HPCI, 1 LPCS, 3 LPCI 4 ADS, HPCI, 1 LPCS, 1 LPCISF-LOCA 6 ADS, HPCI, 1 LPCS, 2 LPCI 6 ADS, HPCI, 1 LPCSSF-LPCI 6 ADS, HPCI, 2 LPCS, 2 LPCI 6 ADS, HPCI, 2 LPCSSF-DGEN 6 ADS, HPCI, 1 LPCS, 2 LPCI 6 ADS, HPCI, 1 LPCSSF-HPCI 6 ADS, 2 LPCS, 4 LPCi 6 ADS, 2 LPCS, 2 LPCISF-ADSIIL 4 ADS, HPCI, 2 LPCS, 4 LPCI 4 ADS, HPCI, 2 LPCS, 2 LPCISF-ADSISV 5 ADS, HPCI, 2 LPCS, 4 LPCI 5 ADS, HPCI, 2 LPCS, 2 LPCIEach LPCS means operation of two core spray pumps in a system. It is assumed that both pumps ina system must operate to take credit for core spray cooling or inventory makeup. Furthermore, 2LPCI refers to two LPCI pumps into one loop, 3 LPCI refers to two LPCI pumps into one loop and oneLPCI pump into one loop. 4 LPCI refers to four LPCI pumps into two loops, two per loop.t 4 ADS, 5 ADS and 6 ADS means the number of ADS values available for automatic activation. t Systems remaining, as identified in this table for recirculation suction line breaks, are applicable toother non-ECCS line breaks. For a LOCA from an ECCS line break, the systems remaining are thoselisted for recirculation suction breaks, less the ECCS in which the break is assumed.§ Unit 3 systems remaining. Conservative for Units 1 and 2.AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMM 10XM FuelANP-3153(NP) Revision 0Page 4-15Gap, Reflood Time Hot NodeGap Coefficient, for Canister CoolantPeak Cladding Temperature, Metal Water ReactionFigure 4.1 Flow Diagram forEXEM BWR-2000 ECCS Evaluation ModelAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 4-16Figure 4.2 [AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 4-17Figure 4.3 RELAX System Blowdown ModelAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTm 1OXM FuelANP-3153(NP) Revision 0Page 4-18Figure 4.4 RELAX Hot Channel Blowdown ModelTop-Peaked AxialAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM1OXM FuelANP-3153(NP) Revision 0Page 4-19HPCITurbineLPCI Injection ValveDischarge ShutoffValveRecirculation Pump Recirculation Pump"4 Closed Valve>K Open ValveFigure 4.5 ECCS Schematic AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM1OXM FuelANP-3153(NP) Revision 0Page 4-20Figure 4.6 Axial Power Distribution for Limiting LOCA Case in RELAX Calculation AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 1OXM Fuel Page 5-15.0 MAPLHGR Analysis ResultsAn exposure-dependent MAPLHGR limit for ATRIUM 1OXM fuel is obtained by performing HUXY heatup analyses using results from the limiting LOCA analysis case identified inReference

1. The break characteristics for the limiting analysis are summarized in Section 1.0.Table 5.1 shows event times for the analysis.

The response of the reactor system is shown inFigures 5.1 -5.18. In the MAPLHGR analysis, the ATRIUM 1OXM fuel rod stored energy is setto be bounding at all exposures and the RELAX hot channel peak power node is modeled at thehighest MAPLHGR, which is 102% of 13.0 kW/ft for the ATRIUM 1OXM fuel.Table 5.2 shows the MAPLHGR analysis results for the ATRIUM 1OXM fuel. The HUXY modelof the ATRIUM 1OXM fuel is applied to obtain these results as described in Section 4.3. TheHUXY analysis is performed at 5 GWd/MTU exposure intervals for exposures between 0 and65 GWd/MTU and an ending exposure of 67 GWd/MTU. The HUXY MAPLHGR input isconsistent with the data in Figure 2.1. Exposure-dependent ATRIUM 1OXM fuel rod data isprovided from RODEX2 results and includes gap coefficient, hot gap thickness, cold gapthickness, gas moles, fuel rod plenum length, and spring relaxation time. This data is providedas a function of linear heat generation rate at each exposure analyzed. The ATRIUM 1OXM limiting PCT is 1903°F at the 0.0 GWd/MTU exposure. The corresponding maximum local cladding oxidation at the PCT limiting exposure is 1.16%. Analysis results showthe CMWR is less than 1.0% total hydrogen generated. Figure 5.19 shows the rod surface temperature of the ATRIUM 1OXM PCT rod as a function oftime for the limiting break. The maximum temperature of 1903*F occurs at 423.4 seconds.These results demonstrate the acceptability of the ATRIUM 1OXM MAPLHGR limit shown inFigure 2.1.AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 10XM FuelANP-3153(NP) Revision 0Page 5-2Table 5.1 Event Times for Limiting Break0.20 ft2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial 102% PowerEvent Time (sec)Initiate break 0.0Initiate scram 0.6Low-Low liquid level, L2 (448 in) 40.1Low-Low-Low liquid level, Li (372.5 in) 65.7Jet pump uncovers 95.6Recirculation suction uncovers 155.9Lower plenum flashes 182.1LPCS high-pressure cutoff 280.9LPCS valve pressure permissive 270.0LPCS valve starts to open 272.0LPCS valve fully open 305.0LPCS permissive for ADS timer 94.7LPCS pump at rated speed 97.7LPCS flow starts 280.9RDIV pressure permissive 312.5RDIV starts to close 314.5RDIV fully closed 350.5Rated LPCS flow 374.5ADS valves open 187.7Blowdown ends 374.5Bypass reflood 480.7Core reflood 423.4PCT 423.4AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 110XVXM FuelANP-3153(NP) Revision 0Page 5-3Table 5.2 ATRIUM 1OXM MAPLHGRAnalysis ResultsAverage LocalPlanar CladdingExposure MAPLHGR PCT Oxidation (GWd/MTU) (kW/ft) (OF) (%)0.0 13.00 1903 1.165.0 13.00 1869 1.0210.0 13.00 1842 0.9015.0 13.00 1828 0.8520.0 12.48 1779 0.7025.0 11.96 1744 0.6130.0 11.44 1714 0.5535.0 10.92 1684 0.4840.0 10.40 1657 0.4345.0 9.89 1629 0.3850.0 9.37 1601 0.3355.0 8.84 1569 0.2860.0 8.33 1564 0.2765.0 7.81 1542 0.2567.0 7.60 1531 0.23CMWR is <1.0% at all exposures. AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0.Page 5-4Figure 5.1 Limiting BreakUpper Plenum PressureFigure 5.2 Limiting BreakTotal Break Flow RateAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 5-5liiU)faFigure 5.3 Limiting BreakADS Flow RateC):r70 100 200 300400TE (SEC)0 6G00 700800Figure 5.4 Limiting BreakHPCI Flow RateAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 5-63:C)L.J(I)000(j,~C)0~-J400TIM (SEC)Figure 5.5 Limiting BreakLPCS Flow RateC)bJ00003:0b.C)'0~-J~00 200 300400TIM (SEC)50W000 700 BooFigure 5.6 Limiting BreakIntact Loop LPCI Flow RateAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 5-7C)bJa,gD-J~00-J-Ja,Figure 5.7 Limiting BreakBroken Loop LPCI Flow Rate400TIME~ Figure 5.8 Limiting BreakUpper Downcomer Mixture LevelAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 5-8LJo__j25MFigure 5.9 Limiting BreakMiddle Downcomer Mixture LevelU0U]U]C)U]-3-100 200 300 400TWIE (SEC)6SW 600 700 B60Figure 5.10 Limiting BreakLower Downcomer Mixture LevelAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 5-9Figure 5.11 Limiting BreakIntact Loop Discharge Line Liquid Mass4(STIME (SEC)Figure 5.12 Limiting BreakUpper Plenum Liquid MassAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM10XM FuelANP-3153(NP) Revision 0Page 5-10400liME (SEC)Figure 5.13 Limiting BreakLower Plenum Liquid MassC)bien000-JF- '~4liiL.J_C)F-oTm1Isa 20 388 400lIME (SEC)500 0W 700amFigure 5.14 Limiting BreakHot Channel Inlet Flow RateAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTMIOXM FuelANP-3153(NP) Revision 0Page 5-11C-,bJ(N,NDI-bi ~I-U.S -C)I-_0Fo100 200 300 400TE (SEC)500 600 7008amFigure 5.15 Limiting BreakHot Channel Outlet Flow RateFigure 5.16 Limiting BreakHot Channel Coolant Temperature at the Limiting NodeAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page 5-12Figure 5.17 Limiting BreakHot Channel Quality at the Limiting NodeFigure 5.18 Limiting BreakHot Channel Heat Transfer Coefficient at the Limiting NodeAREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTm 1OXM FuelANP-3153(NP) Revision 0Page 5-1325002000L.-L.'J)EC7001500100050000 50 100 150 200 250 300Time (sec)350 400 450 500Figure 5.19 Limiting BreakCladding Temperatures AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 1OXM Fuel Page 6-16.0 Conclusions The EXEM BWR-2000 Evaluation Model was applied to determine the ATRIUM 10XMMAPLHGR limit for Browns Ferry. The following conclusions were made from the analysespresented. The acceptance criteria of the Code of Federal Regulations (10 CFR 50.46) are met foroperation at or below the ATRIUM 1OXM MAPLHGR limit given in Figure 2.1.-Peak PCT < 2200°F.-Local cladding oxidation thickness < 0.17.-Total hydrogen generation < 0.01.-Coolable

geometry, satisfied by meeting peak PCT, local cladding oxidation, andtotal hydrogen generation criteria.

-Core long-term

cooling, satisfied by concluding core flooded to top of active fuelor core flooded to the jet pump suction elevation with one core spray operating (Reference 1).The MAPLHGR limit is applicable for ATRIUM 1OXM full cores as well as transition corescontaining ATRIUM 1OXM fuel.AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0Limit for ATRIUMTM 10XM Fuel Page 7-17.0 References

1. ANP-3152(P)

Revision 0, Browns Ferry Units 1, 2, and 3 LOCA Break SpectrumAnalysis for A TRIUM IOXM Fuel, AREVA NP, October 2012.2. EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001.3. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.4. XN-CC-33(A) Revision 1, HUXY: A Generalized Multirod Heatup Code with 10 CFR 50Appendix K Heatup Option Users Manual, Exxon Nuclear Company, November 1975.5. XN-NF-82-07(P)(A) Revision 1, Exxon Nuclear Company ECCS Cladding Swelling andRupture Model, Exxon Nuclear Company, November 1982.6. EMF-2292(P)(A) Revision 0, ATRIUMTM-1O: Appendix K Spray Heat TransferCoefficients, Siemens Power Corporation, September 2000.7. ANP-3014(P) Revision 0, Browns Ferry Units 1, 2, and 3 LOCA Parameters

Document, AREVA NP, July 2011.8. Letter, P. Salas (AREVA) to Document Control Desk, U.S. Nuclear Regulatory Commission, Proprietary Viewgraphs and Meeting Summary for Closed Meeting onApplication of the EXEM BWR-2000 ECCS Evaluation Methodology, NRC:1 1:096,September 22, 2011.9. Letter, T.J. McGinty (NRC) to P. Salas (AREVA),

Response to AREVA NP, Inc.(AREVA) Proposed Analysis Approach for Its EXEM Boiling Water Reactor (BWR)-2000 Emergency Core Cooling System (ECCS) Evaluation Model, July 5, 2012.AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS. Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page A-1APPENDIX A SUPPLEMENTAL INFORMATION AREVA NP Inc. Browns Ferry Units 1, 2, and 3LOCA-ECCS Analysis MAPLHGRLimit for ATRIUMTM 1OXM FuelANP-3153(NP) Revision 0Page A-2Table A.1 Computer Codes Used forMAPLHGR Limit AnalysisAREVA NP Inc.}}