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 8 ANP-3153NP, "Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUMTM 1OXM Fuel"- Non Proprietary ANP-3153(NP) Revision 0 Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM T M 1OXM Fuel August 2013 A AREVA AREVA NP Inc. AREVA NP Inc.ANP-3153(NP) Revision 0 Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM T M 1OXM Fuel AREVA NP Inc.ANP-3153(NP) Revision 0 Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM T M 1OXM Fuel Copyright © 2013 AREVA NP Inc.All Right Reserved Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM T M 1OXM.Fuel ANP-3153(NP) Revision 0 Page i Nature of Changes Item Page Description and Justification

1. All This is the initial release AREVA NP Inc.

Browns Ferry Units 1,2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0 Limit for ATRIUMTM 10XM Fuel Page ii Contents 1.0 Introduction ................................................... 1-1 2 .0 S u m m a ry ...................................................................................................................... 2-1 3.0 LOCA Description ......................................................................................................... 3-1 3.1 Accident Description ......................................................................................... 3-1 3.2 Acceptance Criteria ........................................................................................... 3-2 4.0 LOCA Analysis Description ........................................................................................... 4-1 4.1 Blowdown Analysis ........................................................................................... 4-1 4.2 Refill / Reflood Analysis .................................................................................... 4-2 4.3 Heatup Analysis ................................................................................................ 4-2 4.4 [] .................................................. 4-3 4 .4 .1 [ ...................................................................... 4 -4 4.5 Plant Parameters .............................................................................................. 4-4 4.6 ECCS Parameters ............................................................................................ 4-4 5.0 MAPLHGR Analysis Results ......................................................................................... 5-1 6.0 Conclusions .................................................................................................................. 6-1 7.0 References ................................................................................................................... 7-1 Appendix A Supplemental Information .......................................................................... A-1 Tables Table 2.1 LOCA Results for PCT Limiting Conditions ........................... 2-3 Table 4.1 Initial Conditions ..................................................................................................... 4-6 Table 4.2 Reactor System Parameters ................................................................................... 4-7 Table 4.3 ATRIUM 1OXM Fuel Assembly Parameters ............................................................ 4-8 Table 4.4 High-Pressure Coolant Injection Parameters .......................................................... 4-9 Table 4.5 Low-Pressure Coolant Injection Parameters ......................................................... 4-10 Table 4.6 Low-Pressure Core Spray Parameters ................................................................. 4-11 Table 4.7 Automatic Depressurization System Parameters .................................................. 4-12 Table 4.8 Recirculation Discharge Isolation Valve Parameters ............................................. 4-13 Table 4.9 ECCS Single Failure ............................................................................ ......4-14 Table 5.1 Event Times for Limiting Break 0.20 ft2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial 102% Power ......................................................... 5-2 Table 5.2 ATRIUM 1OXM MAPLHGR Analysis Results .......................................................... 5-3 Table A.1 Computer Codes Used for MAPLHGR Limit Analysis ............................................ A-2 AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0 Limit for ATRIUM T M 10XM Fuel Page iii Figures Figure 2.1 MAPLHGR Limit for ATRIUM 10XM Fuel ............................................................... 2-4 Figure 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model .............................. 4-15 Figure 4.2 [ ] ........................ 4-16 Figure 4.3 RELAX System Blowdown Model ......................................................................... 4-17 Figure 4.4 RELAX Hot Channel Blowdown Model Top-Peaked Axial .................................... 4-18 Figure 4.5 E C C S S chem atic ................................................................................................. 4-19 Figure 4.6 Axial Power Distribution for Limiting LOCA Case in RELAX Calculation ............... 4-20 Figure 5.1 Limiting Break Upper Plenum Pressure .................................................................. 5-4 Figure 5.2 Limiting Break Total Break Flow Rate .................................................................... 5-4 Figure 5.3 Limiting Break ADS Flow Rate ............................................................................... 5-5 Figure 5.4 Limiting Break HPCI Flow Rate .............................................................................. 5-5 Figure 5.5 Limiting Break LPCS Flow Rate ............................................................................. 5-6 Figure 5.6 Limiting Break Intact Loop LPCI Flow Rate ............................................................ 5-6 Figure 5.7 Limiting Break Broken Loop LPCI Flow Rate .......................................................... 5-7 Figure 5.8 Limiting Break Upper Downcomer Mixture Level .................................................... 5-7 Figure 5.9 Limiting Break Middle Downcomer Mixture Level ................................................... 5-8 Figure 5.10 Limiting Break Lower Downcomer Mixture Level .................................................. 5-8 Figure 5.11 Limiting Break Intact Loop Discharge Line Liquid Mass ........................................ 5-9 Figure 5.12 Limiting Break Upper Plenum Liquid Mass ........................................................... 5-9 Figure 5.13 Limiting Break Lower Plenum Liquid Mass ......................................................... 5-10 Figure 5.14 Limiting Break Hot Channel Inlet Flow Rate ........................ 5-10 Figure 5.15 Limiting Break Hot Channel Outlet Flow Rate .................................................... 5-11 Figure 5.16 Limiting Break Hot Channel Coolant Temperature at the Limiting Node ............. 5-11 Figure 5.17 Limiting Break Hot Channel Quality at the Limiting Node ................................... 5-12 Figure 5.18 Limiting Break Hot Channel Heat Transfer Coefficient at the Limiting Node ....... 5-12 Figure 5.19 Limiting Break Cladding Temperatures .............................................................. 5-13 AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page iv ADS ANS BWR CFR CLTP CMWR DEG Nomenclature automatic depressurization system American Nuclear Society boiling water reactor Code of Federal Regulations current licensed thermal power (3458 MWt)core average metal-water reaction double-ended guillotine emergency core cooling system end of blowdown high-pressure coolant injection linear heat generation rate loss-of-coolant accident low-pressure coolant injection low-pressure core spray maximum average planar linear heat generation rate minimum critical power ratio maximum extended load line limit analysis metal-water reaction ECCS EOB HPCI LHGR LOCA LPCI LPCS MAPLHGR MCPR MELLLA MWR NRC OLTP PCT RDIV Nuclear Regulatory Commission, U.S.original licensed thermal power peak cladding temperature recirculation discharge isolation valve SF-ADS SF-ADSIIL SF-ADSISV SF-BATT SF-BATTIBA SF-BATTIBB SF-BATTIBC SF-DGEN single failure of ADS single failure of ADS, initiation logic single failure of ADS, single valve single failure of battery (DC) power single failure of battery (DC) power, board A single failure of battery (DC) power, board B single failure of battery (DC) power, board C single failure of a diesel generator AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page v SF-HPCI SF-LOCA SF-LPCI SLO Nomenclature (Continued) single failure of the HPCI system single failure of opposite unit false LOCA signal single failure of a LPCI valve single-loop operation AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 1-1 1.0 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 the LOCA-ECCS analysis is to specify the maximum average planar linear heat generation rate (MAPLHGR) limit versus exposure for ATRIUM T M 1OXM* fuel and to demonstrate that the MAPLHGR limit is adequate to ensure that the LOCA-ECCS criteria in 10 CFR 50.46 are satisfied for operation at or below the limit. The report also documents the licensing basis peak cladding temperature (PCT) and corresponding local cladding oxidation from the metal water reaction (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 Models developed by AREVA NP and approved for reactor licensing analyses by the U.S. Nuclear Regulatory Commission (NRC). The models and computer codes used by AREVA for LOCA analyses are collectively referred to as the EXEM BWR-2000 Evaluation Model. The EXEM BWR-2000 Evaluation Model and NRC approval are documented in Reference

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

1. The LOCA conditions evaluated in Reference 1 include break size, type, location, axial power shape, and ECCS single failure. The limiting LOCA break characteristics identified in Reference 1 are presented below: Limiting LOCA Break Characteristics Location recirculation discharge pipe Type / size split / 0.20 ft 2 Single failure battery (DC) power, board A Axial 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 0 Limit for ATRIUM T M 10XM Fuel Page 1-2 The LOCA break spectrum analysis documented in Reference 1 was based on a generic ATRIUM 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 fuel characteristics that depend on the fuel assembly neutronic design and exposure (e.g., local rod power, stored energy). The fuel assembly heatup analysis results presented in this report are for the limiting (minimum margin to acceptance criteria) ATRIUM 1OXM neutronic design currently designed for use at Browns Ferry Units 1, 2, and 3. The heatup analyses were performed using the fluid conditions from the limiting LOCA identified in Reference

1. Cycle specific heatup analyses are performed to confirm that the results in this report remain bounding for 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 to as the current licensed thermal power (CLTP).AREVA NP Inc. Browns Ferry Units 1,2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0 Limit for ATRIUMTM 10XM Fuel Page 2-1 2.0 Summary The MAPLHGR limit was determined by applying the EXEM BWR-2000 Evaluation Model for the analysis of the limiting LOCA event. The exposure-dependent MAPLHGR limit for ATRIUM 1OXM fuel is shown in Figure 2.1. [The response of the reactor system and hot channel during the limiting LOCA analysis from Reference 1 are presented in Section 5.0. The MAPLHGR analysis results for the limiting lattice design 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 these calculations confirm that the LOCA acceptance criteria in the Code of Federal Regulations (10 CFR 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 the LOCA break spectrum report (Reference 1). [AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 2-2.[I AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 2-3 Table 2.1 LOCA Results for PCT 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 1OXM 0.0 1903 1.16 0.65<1.0 AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM T M 1OXM Fuel ANP-3153(NP) Revision 0 Page 2-4 16.0 14.0 CL 12.0 10.0 8.0 6.0 4.0.0 10.0 20.0 30.0 40.0 50.0 60.0 Planar Average Exposure (GWd/MTU)70.0 Average Planar ATRIUM 1OXM Exposure MAPLHGR (GWd/MTU) (kW/ft)0 13.0 15 13.0 67 7.6 Figure 2.1 MAPLHGR Limit for ATRIUM 1OXM Fuel AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0 Limit for ATRIUMTM 1OXM Fuel Page 3-1 3.0 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 pressure boundary piping up to and including a break equivalent in size to a double-ended rupture of the largest pipe in the reactor coolant system. There is not a specifically identified cause that results in the pipe break. However, for the purpose of identifying a design basis accident, the pipe 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 line causes 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. Special analysis considerations are required when the break is postulated to occur in a pipe that is used as 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 location dependence, 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 and locations is required. The results of the Browns Ferry ATRIUM 1OXM break spectrum calculations 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 relative duration of each phase is strongly dependent upon the break size and location. The last two phases 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 in fuel cladding temperature due to core flow degradation, and for the larger breaks, the core becomes fully or partially uncovered. There is a rapid decrease in pressure during the blowdown phase. During the early phase of the depressurization, the exiting coolant provides core cooling. Later in the blowdown, core cooling is provided by lower plenum flashing as the AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0 Limit for ATRIUMTM 1OXM Fuel Page 3-2 system continues to depressurize. The blowdown phase is defined to end when LPCS reaches rated flow.In the refill phase of a LOCA, the ECCS is functioning and there is a net increase of coolant inventory. 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 of the reactor vessel. In general, the core heat transfer to the coolant is less than the fuel decay heat 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 level reenters the core region. During the core reflood phase, cooling is provided above the mixture level by entrained reflood liquid and below the mixture level by pool boiling. Sufficient coolant eventually reaches the core hot node and the fuel cladding temperature decreases. 3.2 Acceptance Criteria A LOCA is a potentially limiting event that may place constraints on fuel design, local power peaking, and in some cases, acceptable core power level. During a LOCA, the normal transfer of 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 during a 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 a LOCA 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 Models to Appendix K is described in Reference

2. The ECCS must be designed such that the plant response 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 local cladding thickness.

The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would 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 0 Limit for ATRIUMTM 10XM Fuel Page 3-3 Calculated changes in core geometry shall be such that the core remains amenable to cooling.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, the local oxidation criterion, the hydrogen generation criterion, the coolable geometry criterion, and the long-term cooling criterion. A MAPLHGR limit is established for each ATRIUM IOXM fuel type to ensure that these criteria are met. For jet pump BWRs, the most challenging criterion is that 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 a coolable geometry is maintained. Compliance with the long-term coolability criterion is discussed in Reference 1.AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0 Limit for ATRIUMTM 10XM Fuel Page 4-1 4.0 LOCA Analysis Description The Evaluation Model used for the break spectrum analysis is the EXEM BWR-2000 LOCA analysis methodology described in Reference

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

RODEX2 is used to determine fuel parameters (such as stored energy) for input to the other LOCA codes. The code 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 using RODEX2 (Reference 3). RODEX2 is used to determine the initial stored energy for both the blowdown analysis (RELAX system and hot channel) and the heatup analysis (HUXY). This is accomplished by ensuring that the initial stored energy in RELAX and HUXY is the same or higher than that calculated by RODEX2 for the power, exposure, and fuel design being considered. 4.1 Blowdown Analysis The RELAX code (Reference

2) is used to calculate the system thermal-hydraulic response during the blowdown phase of the LOCA. For the system blowdown analysis, the core is represented 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 heating as required by Appendix K of 10 CFR 50. The reactor vessel nodalization for the system blowdown analysis is shown in Figure 4.3. This nodalization is consistent with that used in the topical report submitted to the NRC (Reference 2).The RELAX analysis is performed from the time of the break initiation through the end of blowdown (EOB). The system blowdown calculation provides the upper and lower plenum transient boundary conditions for the hot channel analysis.Following the system blowdown calculation, another RELAX analysis is performed to analyze the maximum power assembly (hot channel) of the core. The RELAX hot channel calculation is used to calculate hot channel fuel, cladding, and coolant temperatures during the blowdown AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0 Limit for ATRIUM T M 1OXM Fuel Page 4-2 phase 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 blowdown results 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 analysis are heat transfer coefficients and fluid conditions in the hot channel.4.2 Refill / Reflood Analysis The RELAX code is used to compute the system and hot channel hydraulic response during the refill/reflood phase of the LOCA. The RELAX system and RELAX hot channel analyses continue beyond the end of blowdown to analyze system and hot channel responses during the refill and reflood phases. The refill phase is the period when the lower plenum is filling due to ECCS injection. The reflood phase is when some portions of the core and hot assembly are being cooled with ECCS water entering from the lower plenum. The purpose of the RELAX calculations beyond blowdown is to determine the time when the liquid flow via upward entrainment from the bottom of the core becomes high enough at the hot node in the hot assembly to end the temperature increase of the fuel rod cladding. This event time is called the time of hot node reflood. [] The time when the core bypass mixture level rises to the elevation of the hot node in the hot assembly is also determined. RELAX provides a prediction of fluid inventory during the ECCS injection period. Allowing for countercurrent flow through the core and bypass, RELAX determines the refill rate of the lower plenum due to ECCS water and the subsequent reflood times for the core, hot assembly, and the core bypass. The RELAX calculations provide HUXY with the time of hot node reflood and the time when the liquid has risen in the bypass to the height of the axial plane of interest (time of bypass reflood).4.3 Heatup Analysis The HUXY code (Reference

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

The heat generated by metal-water reaction (MWR) is included in the HUXY analysis. HUXY is used to calculate 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 clad swelling and rupture models from NUREG-0630 have been incorporated into HUXY AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0 Limit for ATRIUM T M 10XM Fuel Page 4-3 (Reference 5). The HUXY code complies with the 10 CFR 50 Appendix K criteria for LOCA Evaluation Models.HUXY uses the end of blowdown time and the times of core bypass reflood and core reflood at the axial plane of interest from the RELAX analysis. []Throughout the calculations, decay power is determined based on the ANS 1971 decay heat curve plus 20% as described in Reference

2. [] are used in the HUXY analysis.

The principal results of a HUXY heatup analysis are the PCT and the percent local oxidation of the fuel cladding, often called the percent 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 maximum planar average MWR calculated by HUXY is less than 1.0%.4.4 [AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM- 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-4 4A4.1 I I I 4.5 Plant Parameters The LOCA break spectrum analysis is performed using the plant parameters presented in Reference

7. Table 4.1 provides a summary of reactor initial conditions used in the break spectrum 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 the key 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 power shape 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 ECCS characteristics assumed in the LOCA break spectrum analysis. The ECCS is modeled as fill junctions 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 pump capacity data provided in Tables 4.4 -4.6. For the break spectrum analyses, no credit for ECCS 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 0 Limit for ATRIUMTM 1OXM Fuel Page 4-5 The automatic depressurization system (ADS) valves are modeled as a junction connecting the reactor 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 in Table 4.7.In the AREVA LOCA analysis model, ECCS initiation is assumed to occur when the water level drops to the applicable level setpoint. No credit is assumed for the start of HPCI, LPCS, or LPCI 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 each assumed failure.AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-6 Table 4.1 Initial Conditions* Reactor power (% of rated) 102 Total core flow (% of rated) [ ]Reactor power (MWt) 3527 Total core flow (Mlb/hr) [[ ]Steam flow rate (Mlb/hr) 14.50 Steam dome pressure (psia) 1054 Core inlet enthalpy (Btu/Ib) [ ]ATRIUM 1OXM hot assembly MAPLHGR (kW/ft) 13.0[ ]ECCS fluid temperature ('F) 120 Axial power shape Fig. 4.6 The AREVA calculated heat balance is adjusted to match the 100% power/100% flow values given in the plant parameters document (Reference 7). The model is then rebalanced based on AREVA heat balance calculations to establish these LOCA initial conditions at 102% of rated thermal power.I AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM T M 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-7 Table 4.2 Reactor System Parameters Parameter Value Vessel ID (in) 251 Number of fuel assemblies 764 Recirculation suction pipe area (ft 2) 3.507 1.0 DEG suction break area (ft 2) 7.013 Recirculation discharge pipe area (ft 2) 3.507 1.0 DEG discharge break area (ft 2) 7.013 AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 10XM Fuel ANP-3153(NP) Revision 0 Page 4-8 Table 4.3 ATRIUM IOXM Fuel Assembly Parameters Parameter Value Fuel rod array 10x10 Number of fuel rods per 79 (full-length rods)assembly 12 (part-length rods)Non-fuel rod type Water channel replaces 9 fuel rods Fuel rod OD (in) 0.4047 Active fuel length (in) 150.0 (full-length rods)(including blankets) 75 (part-length rods)Water channel outside width (in) 1.378 Fuel channel thickness (in) 0.075 (minimum wall)0.100 (corner)Fuel channel internal width (in) 5.278 AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-9 Table 4.4 High-Pressure Coolant Injection Parameters Parameter Value Coolant temperature (OF) 120 Initiating Signals and Setpoints Water level* L2 (448 in)High drywell pressure (psig) 2.6 (Not Used)Time Delays Time for HPCI pump to reach rated speed and injection valve wide open (sec) 35 Delivered Flow Rate Versus Pressure Vessel to Flow Drywell AP Rate (psid) (gpm)0 0 150 5000 1120 5000 1174 3600* Relative to vessel zero.AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-10 Table 4.5 Low-Pressure Coolant Injection Parameters Parameter Value Reactor pressure permissive for opening valves (psia) 350 Coolant temperature (OF) 120 Initiating Signals and Setpoints Water level* Li (372.5 in)High drywell pressure (psig) 2.6 (Not Used)Time Delays Time for LPCI pumps to reach ADS permissive (max) (sec)t 32*Time for LPCI pumps to reach rated speed (max) (sec)t 44 LPCI injection valve stroke time (sec) 40 Delivered Flow Rate Versus Pressure Flow Rate Vessel to (gpm)Drywell AP (psid) 2 Pumps Into 4 Pumps Into 1 Loop 2 Loops 0 17,240 34,480 20 16,540 33,080§319.5 0 0 Relative to vessel zero.t Includes 13-second delay for diesel generator start. 2-second signal processing delay for water level trip 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 the specified 2 pumps into 1 loop flow be used.AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-11 Table 4.6 Low-Pressure Core Spray Parameters Parameter Value Reactor pressure permissive for opening valves (psia) 350 Coolant temperature (OF) 120 Initiating Signals and Setpoints Water level* Li (372.5 in)High drywell pressure (psig) 2.6 (not used)Time Delays Time for LPCS pumps to reach ADS permissive (max) (sec)t 40 Time for LPCS pumps to reach rated speed (max) (sec)t 43 LPCS injection valve stroke time (sec) 33 Delivered Flow Rate Versus Pressure Flow Rate Vessel (gpm)to Drywell AP 2 Pumps 4 Pumps (psid) Into Into 1 Sparger 2 Sparger 0 6,935 13,870 105 5,435 10,870 200 3,835 7,670 289 0 0* Relative to vessel zero.t Includes 13-second delay for diesel generator start. 2-second signal processing delay for water level trip Li is assumed in parallel with diesel generator delay.AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 10XM Fuel ANP-3153(NP) Revision 0 Page 4-12 Table 4.7 Automatic Depressurization System Parameters Parameter Value Number of valves installed 6 Number of valves available 6 Minimum flow capacity of available valves (Mlbm/hr at psig) 4.8 at 1125 Initiating Signals and Setpoints Water level* L1 (372.5 in)Time Delays Delay 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 elapsed and LPCS or LPCI pumps reach the ADS ready permissive. Analyses assume the longer delay from LPCS for ADS ready permissive (see Table 4.6).AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-13 Table 4.8 Recirculation Discharge Isolation Valve Parameters Parameter Value Reactor pressure permissive for closing valves -analytical (psia) 215 RDIV stroke time (sec) 36 AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-14 Table 4.9 ECCS Single Failure Systems ,t Assumed Remaining Failure Recirculationt Recirculation Suction Break Discharge Break SF-BATTIBA 6 ADS, 1 LPCS, 2 LPCI 6 ADS, 1 LPCS SF-BATTIBB 4 ADS, HPCI, 1 LPCS, 2 LPCI 4 ADS, HPCI, 1 LPCS SF-BATTIBC§ 4 ADS, HPCI, 1 LPCS, 3 LPCI 4 ADS, HPCI, 1 LPCS, 1 LPCI SF-LOCA 6 ADS, HPCI, 1 LPCS, 2 LPCI 6 ADS, HPCI, 1 LPCS SF-LPCI 6 ADS, HPCI, 2 LPCS, 2 LPCI 6 ADS, HPCI, 2 LPCS SF-DGEN 6 ADS, HPCI, 1 LPCS, 2 LPCI 6 ADS, HPCI, 1 LPCS SF-HPCI 6 ADS, 2 LPCS, 4 LPCi 6 ADS, 2 LPCS, 2 LPCI SF-ADSIIL 4 ADS, HPCI, 2 LPCS, 4 LPCI 4 ADS, HPCI, 2 LPCS, 2 LPCI SF-ADSISV 5 ADS, HPCI, 2 LPCS, 4 LPCI 5 ADS, HPCI, 2 LPCS, 2 LPCI Each LPCS means operation of two core spray pumps in a system. It is assumed that both pumps in a system must operate to take credit for core spray cooling or inventory makeup. Furthermore, 2 LPCI refers to two LPCI pumps into one loop, 3 LPCI refers to two LPCI pumps into one loop and one LPCI 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 to other non-ECCS line breaks. For a LOCA from an ECCS line break, the systems remaining are those listed 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 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMM 10XM Fuel ANP-3153(NP) Revision 0 Page 4-15 Gap, Reflood Time Hot Node Gap Coefficient, for Canister Coolant Peak Cladding Temperature, Metal Water Reaction Figure 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-16 Figure 4.2 [AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-17 Figure 4.3 RELAX System Blowdown Model AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTm 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-18 Figure 4.4 RELAX Hot Channel Blowdown Model Top-Peaked Axial AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM T M 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-19 HPCI Turbine LPCI Injection Valve Discharge Shutoff Valve Recirculation Pump Recirculation Pump"4 Closed Valve>K Open Valve Figure 4.5 ECCS Schematic AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM T M 1OXM Fuel ANP-3153(NP) Revision 0 Page 4-20 Figure 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 0 Limit for ATRIUM T M 1OXM Fuel Page 5-1 5.0 MAPLHGR Analysis Results An exposure-dependent MAPLHGR limit for ATRIUM 1OXM fuel is obtained by performing HUXY heatup analyses using results from the limiting LOCA analysis case identified in Reference

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 in Figures 5.1 -5.18. In the MAPLHGR analysis, the ATRIUM 1OXM fuel rod stored energy is set to be bounding at all exposures and the RELAX hot channel peak power node is modeled at the highest 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 model of the ATRIUM 1OXM fuel is applied to obtain these results as described in Section 4.3. The HUXY analysis is performed at 5 GWd/MTU exposure intervals for exposures between 0 and 65 GWd/MTU and an ending exposure of 67 GWd/MTU. The HUXY MAPLHGR input is consistent with the data in Figure 2.1. Exposure-dependent ATRIUM 1OXM fuel rod data is provided from RODEX2 results and includes gap coefficient, hot gap thickness, cold gap thickness, gas moles, fuel rod plenum length, and spring relaxation time. This data is provided as 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 show the 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 of time 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 in Figure 2.1.AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 10XM Fuel ANP-3153(NP) Revision 0 Page 5-2 Table 5.1 Event Times for Limiting Break 0.20 ft 2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial 102% Power Event Time (sec)Initiate break 0.0 Initiate scram 0.6 Low-Low liquid level, L2 (448 in) 40.1 Low-Low-Low liquid level, Li (372.5 in) 65.7 Jet pump uncovers 95.6 Recirculation suction uncovers 155.9 Lower plenum flashes 182.1 LPCS high-pressure cutoff 280.9 LPCS valve pressure permissive 270.0 LPCS valve starts to open 272.0 LPCS valve fully open 305.0 LPCS permissive for ADS timer 94.7 LPCS pump at rated speed 97.7 LPCS flow starts 280.9 RDIV pressure permissive 312.5 RDIV starts to close 314.5 RDIV fully closed 350.5 Rated LPCS flow 374.5 ADS valves open 187.7 Blowdown ends 374.5 Bypass reflood 480.7 Core reflood 423.4 PCT 423.4 AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 110XVXM Fuel ANP-3153(NP) Revision 0 Page 5-3 Table 5.2 ATRIUM 1OXM MAPLHGR Analysis Results Average Local Planar Cladding Exposure MAPLHGR PCT Oxidation (GWd/MTU) (kW/ft) (OF) (%)0.0 13.00 1903 1.16 5.0 13.00 1869 1.02 10.0 13.00 1842 0.90 15.0 13.00 1828 0.85 20.0 12.48 1779 0.70 25.0 11.96 1744 0.61 30.0 11.44 1714 0.55 35.0 10.92 1684 0.48 40.0 10.40 1657 0.43 45.0 9.89 1629 0.38 50.0 9.37 1601 0.33 55.0 8.84 1569 0.28 60.0 8.33 1564 0.27 65.0 7.81 1542 0.25 67.0 7.60 1531 0.23 CMWR is <1.0% at all exposures. AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0.Page 5-4 Figure 5.1 Limiting Break Upper Plenum Pressure Figure 5.2 Limiting Break Total Break Flow Rate AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 5-5 lii U)fa Figure 5.3 Limiting Break ADS Flow Rate C):r 70 100 200 300 400 TE (SEC)0 6G00 700 800 Figure 5.4 Limiting Break HPCI Flow Rate AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 5-6 3: C)L.J (I)00 0 (j,~C)0~-J 400 TIM (SEC)Figure 5.5 Limiting Break LPCS Flow Rate C)bJ 00 00 3: 0 b.C)'0~-J~00 200 300 400 TIM (SEC)50W 000 700 Boo Figure 5.6 Limiting Break Intact Loop LPCI Flow Rate AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 5-7 C)bJ a, gD-J~0 0-J-J a, Figure 5.7 Limiting Break Broken Loop LPCI Flow Rate 400 TIME~ Figure 5.8 Limiting Break Upper Downcomer Mixture Level AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 5-8 LJo__j 25 M Figure 5.9 Limiting Break Middle Downcomer Mixture Level U0 U]U]C)U]-3-100 200 300 400 TWIE (SEC)6SW 600 700 B60 Figure 5.10 Limiting Break Lower Downcomer Mixture Level AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 5-9 Figure 5.11 Limiting Break Intact Loop Discharge Line Liquid Mass 4(S TIME (SEC)Figure 5.12 Limiting Break Upper Plenum Liquid Mass AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM T M 10XM Fuel ANP-3153(NP) Revision 0 Page 5-10 400 liME (SEC)Figure 5.13 Limiting Break Lower Plenum Liquid Mass C)bi en 00 0-J F- '~4 lii L.J_C)F-o Tm 1 Isa 20 388 400 lIME (SEC)500 0W 700 am Figure 5.14 Limiting Break Hot Channel Inlet Flow Rate AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM T M IOXM Fuel ANP-3153(NP) Revision 0 Page 5-11 C-, bJ (N, ND I-bi ~I-U.S -C)I-_0 Fo 100 200 300 400 TE (SEC)500 600 700 8am Figure 5.15 Limiting Break Hot Channel Outlet Flow Rate Figure 5.16 Limiting Break Hot Channel Coolant Temperature at the Limiting Node AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page 5-12 Figure 5.17 Limiting Break Hot Channel Quality at the Limiting Node Figure 5.18 Limiting Break Hot Channel Heat Transfer Coefficient at the Limiting Node AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTm 1OXM Fuel ANP-3153(NP) Revision 0 Page 5-13 2500 2000 L.-L.'J)E C 70 0 1500 1000 500 0 0 50 100 150 200 250 300 Time (sec)350 400 450 500 Figure 5.19 Limiting Break Cladding Temperatures AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0 Limit for ATRIUMTM 1OXM Fuel Page 6-1 6.0 Conclusions The EXEM BWR-2000 Evaluation Model was applied to determine the ATRIUM 10XM MAPLHGR limit for Browns Ferry. The following conclusions were made from the analyses presented. The acceptance criteria of the Code of Federal Regulations (10 CFR 50.46) are met for operation 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, and total hydrogen generation criteria.-Core long-term cooling, satisfied by concluding core flooded to top of active fuel or 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 cores containing ATRIUM 1OXM fuel.AREVA NP Inc. Browns Ferry Units 1, 2, and 3 ANP-3153(NP) LOCA-ECCS Analysis MAPLHGR Revision 0 Limit for ATRIUMTM 10XM Fuel Page 7-1 7.0 References

1. ANP-3152(P)

Revision 0, Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis 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 50 Appendix 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 and Rupture Model, Exxon Nuclear Company, November 1982.6. EMF-2292(P)(A) Revision 0, ATRIUM T M-1O: Appendix K Spray Heat Transfer Coefficients, 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 on Application 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 3 LOCA-ECCS. Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page A-1 APPENDIX A SUPPLEMENTAL INFORMATION AREVA NP Inc. Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTM 1OXM Fuel ANP-3153(NP) Revision 0 Page A-2 Table A.1 Computer Codes Used for MAPLHGR Limit Analysis AREVA NP Inc.}}