ML13070A315

ANP-3152(NP), Revision 0, Browns Ferry Unit 1, 2 and 3 LOCA Break Spectrum Analysis for Atrium 10XM Fuel

ML13070A315
Person / Time
Site:	Browns Ferry
Issue date:	10/31/2012
From:	AREVA NP
To:	Office of Nuclear Reactor Regulation
References
ANP-3152(NP), Rev 0
Download: ML13070A315 (89)

Text

ATTACHMENT 15 Browns Ferry Nuclear Plant (BFN)Units 1, 2, and 3 Technical Specifications (TS) Change 478 Addition of Analytical Methodologies to Technical Specification 5.6.5.b for Browns Ferry 1, 2, & 3, and Revision of Technical Specification 2.1.1.2 for Browns Ferry Unit 2, in Support of ATRIUM-10 XM Fuel Use at Browns Ferry LOCA Break Spectrum Analysis Report (Non-Proprietary)

Attached is the non-proprietary version of the AREVA LOCA break spectrum analysis report.

ANP-3152(NP)

Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel October 2012 A AREVA AREVA NP Inc.

AREVA NP Inc.ANP-3152(NP)

Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel AREVA NP Inc.ANP-3152(NP)

Revision 0 Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel Copyright

© 2012 AREVA NP Inc.All Rights Reserved Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 PaQe 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-3152(NP)

LOCA Break Spectrum Analysis Revision 0 forATRIUMTM 1OXM Fuel Page ii Contents 1.0 Introduction

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1-1 2.0 Sum m ary of Results .....................................................................................................

2-1 3.0 LOCA Description

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3-1 3.1 Accident Description

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3-1 3.2 Acceptance Criteria ...........................................................................................

3-2 4.0 LOCA Analysis Description

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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

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4-4 4.6 ECCS Parameters

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4-4 5.0 Break Spectrum Analysis Description

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5-1 5.1 Lim iting Single Failure ......................................................................................

5-1 5.2 Recirculation Line Breaks .................................................................................

5-2 5.3 Non-Recirculation Line Breaks ..........................................................................

5-3 5.3.1 Main Steam Line Breaks ....................................................................

5-3 5.3.2 Feedwater Line Breaks .......................................................................

5-4 5.3.3 HPCI Line Breaks ...............................................................................

5-5 5.3.4 LPCS Line Breaks ....................................

5-6 5.3.5 LPCI Line Breaks ................................................................................

5-6 5.3.6 RCIC Line Breaks ...............................................................................

5-6 5.3.7 RW CU Line Breaks .............................................................................

5-7 5.3.8 Instrument Line Breaks .......................................................................

5-7 6.0 Recirculation Line Break LOCA Analyses .....................................................................

6-1 6.1 Lim iting Break Analysis Results ........................................................................

6-1 6.2 Break Location Analysis Results .......................................................................

6-2 6.3 Break Geometry and Size Analysis Results ......................................................

6-2 6.4 Lim iting Single-Failure Analysis Results ............................................................

6-2 6.5 Axial Power Shape Analysis Results .................................................................

6-2 7.0 Non-Recirculation Line LOCA Analysis .........................................................................

7-1 7.1 Lim iting Non-Recirculation Line Break Results ..................................................

7-1 8.0 Single-Loop Operation LOCA Analysis .........................................................................

8-1 8.1 SLO Analysis Modeling Methodology

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8-1 8.2 SLO Analysis Results .......................................................................................

8-2 9.0 Long-Term Coolability

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9-1 10.0 Conclusions

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10-1 11.0 References

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11-1 Appendix A Com puter Codes ......................................................................................

A-1 Appendix B Additional Results for SF-BATTIBA TLO Recirculation Line Breaks ...........

B-1 Appendix C Additional Results for SF-BATTIBB TLO Recirculation Line Breaks ...........

C-1 Appendix D Additional Results for Non-Recirculation Line Breaks ....................................

D-1 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUM T M 1OXM Fuel Page iii Tables 4 .1 Initia l C onditio ns ...........................................................................................................

4-6 4.2 Reactor System Param eters .........................................................................................

4-7 4.3 ATRIUM 1OXM Fuel Assembly Parameters

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4-8 4.4 High-Pressure Coolant Injection Parameters

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4-9 4.5 Low-Pressure Coolant Injection Parameters

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4-10 4.6 Low-Pressure Core Spray Parameters

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4-11 4.7 Automatic Depressurization System Parameters

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4-12 4.8 Recirculation Discharge Isolation Valve Parameters

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4-13 5.1 Available ECCS for Recirculation Line Break LOCAs ...................................................

5-8 5.2 Available ECCS for Non-Recirculation Line Break LOCAs ............................................

5-9 6.1 Results for Limiting TLO Recirculation Line Break 0.2 ft 2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial 102% Power ...............................................................

6-3 6.2 Event Times for Limiting TLO Recirculation Line Break 0.2 ft 2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial 102% Power ..............................................

6-4 6.3 TLO Recirculation Line Break Spectrum Results for 102% Power SF-BATTIBA

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6-5 6.4 Results for TLO Recirculation Line Break 0.5 ft 2 Split Pump Discharge SF-BATTIBB Top-Peaked Axial 102% Power ..............................................................

6-6 6.5 Event Times for TLO Recirculation Line Break 0.5 ft 2 Split Pump Discharge SF-BATTIBB Top-Peaked Axial 102% Power ...............................................................

6-7 6.6 TLO Recirculation Line Break Spectrum Results for 102% Power SF-BATTIBB

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6-8 6.7 Summary of TLO Recirculation Line Break Results Highest PCT Cases ......................

6-9 7.1 Event Times for Limiting Non-Recirculation Line Break 0.4 ft 2 Double-Ended Guillotine SF-BATTIBA Top-Peaked Axial 102% Power ...............................................

7-2 7.2 Non-Recirculation Line Break Spectrum Results ..........................................................

7-3 8.1 Results for Limiting SLO Recirculation Line Break 0.2 ft 2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial [ ] ...........................................................

8-3 8.2 Event Times for Limiting SLO Recirculation Line Break 0.2 ft 2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial [ ] ..........................................

8-4 8.3 SLO Recirculation Line Break Spectrum Results for [ ] SF-BATTIBA

...... 8-5 8.4 Single- and Two-Loop Operation PCT Summary ..........................................................

8-6 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUM T M 1OXM Fuel Page iv Figures 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model ....................................

4-14 4 .2 [] .................................

4-15 4.3 RELAX System Blowdown Model ...............................................................................

4-16 4.4 RELAX Hot Channel Blowdown Model Top-Peaked Axial ..........................................

4-17 4.5 RELAX Hot Channel Blowdown Model Mid-Peaked Axial ...........................................

4-18 4 .6 E C C S S chem atic ........................................................................................................

4-19 4.7 Axial Power Distributions at 102% Power ] ..........................................

4-20 4.8 Axial Power Distributions at 102% Power [ ............................................

4-21 5.1 Steam, Feedwater, and HPCI Lines ...........................................................................

5-11 6.1 Limiting Recirculation Line Break Upper Plenum Pressure .........................................

6-10 6.2 Limiting Recirculation Line Break Total Break Flow Rate ............................................

6-10 6.3 Limiting Recirculation Line Break ADS Flow Rate ......................................................

6-11 6.4 Limiting Recirculation Line Break HPCI Flow Rate .....................................................

6-11 6.5 Limiting Recirculation Line Break LPCS Flow Rate .....................................................

6-12 6.6 Limiting Recirculation Line Break Intact Loop LPCI Flow Rate ...................................

6-12 6.7 Limiting Recirculation Line Break Broken Loop LPCI Flow Rate .................................

6-13 6.8 Limiting Recirculation Line Break Upper Downcomer Mixture Level ...........................

6-13 6.9 Limiting Recirculation Line Break Middle Downcomer Mixture Level ..........................

6-14 6.10 Limiting Recirculation Line Break Lower Downcomer Mixture Level ...........................

6-14 6.11 Limiting Recirculation Line Break Intact Loop Discharge Liquid Mass ........................

6-15 6.12 Limiting Recirculation Line Break Upper Plenum Liquid Mass ....................................

6-15 6.13 Limiting Recirculation Line Break Lower Plenum Liquid Mass ....................................

6-16 6.14 Limiting Recirculation Line Break Hot Channel Inlet Flow Rate ..................................

6-16 6.15 Limiting Recirculation Line Break Hot Channel Outlet Flow Rate ................................

6-17 6.16 Limiting Recirculation Line Break Hot Channel Coolant Temperature at the H ot N ode at E O B ..................................................................................................

6-17 6.17 Limiting Recirculation Line Break Hot Channel Quality at the Hot Node at E O B ...........................................................................................................................

6 -1 8 6.18 Limiting Recirculation Line Break Hot Channel Heat Transfer C oef. at the Hot Node at EO B ....................................................................................

6-18 6.19 Limiting Recirculation Line Break Cladding Temperatures

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6-19 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paae v Nomenclature ADS ADSVOOS ANS automatic depressurization system ADS valve out of service American Nuclear Society BOL BWR CFR CHF CLTP CMWR beginning of life boiling water reactor Code of Federal Regulations critical heat flux current licensed thermal power (3458 MWt)core average metal-water reaction DEG DG double-ended guillotine diesel generator ECCS EOB FFWTR FHOOS FSAR emergency core cooling system end of blowdown final feedwater temperature reduction feedwater heaters out of service Final Safety Analysis Report high-pressure coolant injection HPCI ICF increased core flow LOCA LPCI LPCS loss-of-coolant accident low-pressure coolant injection low-pressure core spray MAPLHGR MCPR MELLLA MSIV MWR maximum average planar linear heat generation rate minimum critical power ratio maximum extended load line limit analysis main steam isolation valve metal-water reaction NRC OLTP Nuclear Regulatory Commission, U.S.original licensed thermal power PCT peak cladding temperature AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Pace vi RCIC RDIV RWCU SF-ADS SF-ADSIIL SF-ADSISV SF-BATT SF-BATTIBA SF-BATTIBB SF-BATTIBC SF-DGEN SF-HPCI SF-LOCA SF-LPCI SLO Nomenclature (Continued) reactor core isolation cooling recirculation discharge isolation valve reactor water cleanup 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 single failure of the HPCI system single failure of opposite unit false LOCA signal single failure of a LPCI valve single-loop operation TLO two-loop operation AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUM T M 1OXM Fuel Page 1-1 1.0 Introduction The results of a loss-of-coolant accident (LOCA) break spectrum analysis for Browns Ferry Units 1, 2, and 3 are documented in this report. The purpose of the break spectrum analysis is to identify the parameters that result in the highest calculated peak cladding temperature (PCT)during a postulated LOCA. The LOCA parameters addressed in this report include the following:

-Break location 0 Break type (double-ended guillotine (DEG) or split)0 Break size 0 Limiting emergency core cooling system (ECCS) single failure* Axial power shape (top- or mid-peaked)

The analyses documented in this report were performed with LOCA Evaluation Models developed by AREVA NP and approved for reactor licensing analyses by the Nuclear Regulatory Commission, U.S. (NRC). The models and computer codes used by AREVA for LOCA analyses are collectively referred to as the EXEM BWR-2000 Evaluation Model (References 1 -4). The EXEM BWR-2000 Evaluation Model and NRC approval are documented in Reference

1. A summary description of the LOCA analysis methodology is provided in Section 4.0. The calculations described in this report were performed in conformance with 10 CFR 50 Appendix K requirements and satisfy the event acceptance criteria identified in 10 CFR 50.46. [The break spectrum analyses documented in this report were performed for a core composed entirely of ATRIUM T M 1OXM* fuel at beginning-of-life (BOL) conditions.

Calculations assumed an initial core power of 102% of 3458 MWt, providing an analysis licensing basis power of 3527 MWt. The 2.0% increase reflects the maximum uncertainty in monitoring reactor power, 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). The limiting assembly in the core was assumed to be at a maximum average planar heat generation rate (MAPLHGR) limit of 13.0 kW/ft. The analyses assumed a generic ATRIUM 10XM neutronic design. Other initial conditions used in the analyses are described in Section 4.0.* ATRIUM is a trademark of AREVA NP.AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUMTM 1OXM Fuel Page 1-2 This report identifies the limiting LOCA break characteristics (location, type, size, single failure, and axial power shape) that will be used in subsequent analyses to determine the LOCA-ECCS MAPLHGR limit versus exposure for ATRIUM 1OXM fuel used at Browns Ferry Units 1, 2, and 3. The value of PCT calculated for any given set of break characteristics is dependent on exposure and local fuel rod power peaking. Therefore, heatup analyses are performed to determine the PCT versus exposure for each ATRIUM 1 OXM nuclear design in the core. The heatup analyses are performed each cycle using the limiting boundary conditions determined in the break spectrum analysis.

The maximum PCT versus exposure from the heatup analyses are documented in the MAPLHGR limit report.The operating domain of the power/flow map of Reference 6 is applicable for the ATRIUM 1OXM. [] This report also presents results for single-loop operation (SLO).AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM- 10XM Fuel ANP-3152(NP)

Revision 0 Paae 2-1 2.0 Summary of Results Based on analyses presented in this report, the limiting break characteristics are identified 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 [ ]The SLO LOCA analyses support operation with an ATRIUM 1OXM MAPLHGR multiplier of 0.85 applied to the normal two-loop operation MAPLHGR limit.A more detailed discussion of results is provided in Sections 6.0 -8.0.[] The break characteristics identified in this report can be used in subsequent heatup analyses to determine the ATRIUM 1OXM MAPLHGR limit appropriate for a full core of ATRIUM 1OXM as well as transition cores.Available ADS valves are presented in Tables 5.1 and 5.2. No additional valves are assumed out-of-service (ADSVOOS).

[AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTm 10XM Fuel ANP-3152(NP)

Revision 0 PaQe 2-2 I] The conclusions of this report are applicable for operation with [ ], FHOOS, FFWTR, and SLO.While the fuel rod temperatures in the limiting plane of the hot channel during a LOCA are dependent on exposure, the factors that determine the limiting break characteristics are primarily associated with the reactor system and are not dependent on fuel-exposure characteristics.

Fuel parameters that are dependent on exposure (e.g., stored energy, local peaking) have an insignificant effect on the reactor system response during a LOCA. The limiting break characteristics determined using BOL fuel conditions are applicable for exposed fuel. Fuel exposure effects are addressed in heatup analyses performed to determine or verify MAPLHGR limits versus exposure for each ATRIUM 1OXM fuel design.The break spectrum analysis was performed using the NRC-approved AREVA EXEM BWR-2000 LOCA methodology.

All SER restrictions and ranges of applicability for the EXEM BWR-2000 methodology were reviewed prior to final documentation of the LOCA analysis to ensure compliance with NRC requirements and methodology limitations.

[AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 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.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-3152(NP)

LOCA Break Spectrum Analysis Revision 0 forATRIUM T M 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

1. 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-3152(NP)

LOCA Break Spectrum Analysis Revision 0 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 by the 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 the ATRIUM 1OXM fuel type to ensure that these criteria are met. LOCA PCT results are provided in Sections 6.0 -8.0 to determine the limiting LOCA event. LOCA analysis results demonstrating that the PCT, local oxidation, and hydrogen generation criteria are met are provided in the follow-on MAPLHGR report and cycle-specific heatup analyses.

Cycle-specific heatup analyses are performed to demonstrate that the MAPLHGR limits versus exposure for the ATRIUM 1OXM fuel remains applicable for cycle-specific nuclear designs. Compliance with these three criteria ensures that a coolable geometry is maintained.

Long-term coolability criterion is discussed in Section 9.0.AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUM T M 1OXM 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

1. 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 4). 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

1) 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 1).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 phase of the LOCA. The RELAX hot channel nodalization is shown in Figure 4.4 for a AREVA NP Inc.

Browns Ferry Units 1,2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUMTM 1OXM Fuel Page 4-2 top-peaked power shape, and in Figure 4.5 for a mid-peaked axial 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 initial average fuel rod temperature at the limiting plane of the hot channel is conservative relative to the average fuel rod temperature calculated by RODEX2 for operation of the ATRIUM 1OXM assembly at the MAPLHGR limit. The heat transfer coefficients and fluid conditions in the hot channel from the RELAX hot channel calculation are used as input to the HUXY heatup analysis.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

2) 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.AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUM T M 1OXM Fuel Page 4-3 These calculations consider thermal-mechanical interactions within the fuel rod. The clad swelling and rupture models from NUREG-0630 have been incorporated into HUXY (Reference 3). 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

1. [] 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 reaction (%MWR).4.4 [AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 10XM Fuel ANP-3152(NP)

Revision 0 Page 4-4[I 4.4.1 [I I I 4.5 Plant Parameters The LOCA break spectrum analysis is performed using the plant parameters presented in Reference

6. 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 1 OXM fuel. Some of the key ATRIUM 1OXM fuel parameters used in the break spectrum analysis are summarized in Table 4.3.4.6 ECCS Parameters The ECCS configuration is shown in Figure 4.6. Tables 4.4 -4.8 provide the important ECCS characteristics assumed in the 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 lines.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. No credit for ECCS flow is assumed until ECCS pumps reach rated speed.AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUMTM 1OXM Fuel Page 4-5 The 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.AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 4-6 Table 4.1 Initial Conditions*

Reactor power (% of rated) 102 102 Total core flow (% of rated) [ ] [ ]Reactor power (MWt) 3527 3527 Total core flow (Mlb/hr) [ ] [Steam flow rate (Mlb/hr) 14.50 14.50 Steam dome pressure (psia) 1054 1054 Core inlet enthalpy (Btu/Ib) [ ] [ATRIUM 10XM hot assembly MAPLHGR (kW/ft) 13.0 13.0[ECCS fluid temperature

('F) 120 120 Axial power shape Fig. 4.7 Fig. 4.8* The AREVA calculated heat balance is adjusted to match the 100% power/1 00% flow values given in the plant parameters document (Reference 6). The model is then rebalanced based on AREVA heat balance calculations to establish these LOCA initial conditions at 102% of rated thermal power.t C I AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 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 Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Page 4-8 Table 4.3 ATRIUM 1OXM Fuel Assembly Parameters Parameter Fuel rod array Number of fuel rods per assembly Non-fuel rod type Fuel rod OD (in)Active fuel length (in)(including blankets)Water channel outside width (in)Fuel channel thickness (in)Value 1Ox10 79 (full-length rods)12 (part-length rods)Water channel replaces 9 fuel rods 0.4047 150.0 (full-length rods)75 (part-length rods)1.378 0.075 (minimum wall)0.100 (corner)5.278 Fuel channel internal width (in)AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 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 Drywell AP Flow 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 Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(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)§ 44 LPCI injection valve stroke time (sec) 40 Delivered Flow Rate Versus Pressure Flow Rate (gpm)Vessel to 2 Pumps 4 Pumps Drywell AP Into Into (psid) 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.Analyses assume the longer delay from LPCS (40 sec) for ADS ready permissive.

Refer to Table 4.6.§ 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.** Conservative value relative to specified value in Reference 6 (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 Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 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 (gpm)Vessel to 2 Pumps 4 Pumps Drywell AP Into Into (psid) 1 Sparger 2 Spargers 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 Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(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* LI (372.5 in)Time Delays Delay time (from ADS initiating signal to time valves are opened) (sec)t 120* Relative to vessel zero.t ADS timer initiation occurs after Li setpoint 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 Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paae 4-13 Table 4.8 Recirculation Discharge Isolation Valve Parameters Parameter Value Reactor pressure permissive for closing valves -analytical (psia) 215 RDIV stroke time after pressure permissive (sec) 36 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 10XM Fuel ANP-3152(NP)

Revision 0 Paqe 4-14 Hot Assembly Analysis (RELAX)Gap, Reflood Time Hot Node Gap Coefficient, for Canister Coolant Fission Gas & Hot Node Conditions Heatup Analysis (HUXY)I* 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 Break Spectrum Analysis for ATRIUMTm 10XM Fuel ANP-3152(NP)

Revision 0 Page 4-15 Figure 4.2 [AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMNTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page 4-16 Figure 4.3 RELAX System Blowdown Model AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paae 4-17 Figure 4.4 RELAX Hot Channel Blowdown Model Top-Peaked Axial AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page 4-18 Figure 4.5 RELAX Hot Channel Blowdown Model Mid-Peaked Axial AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page 4-19 HPCI Turbine LPCI Injection Valve Discharge Shutoff Valve Recirculation Pump Recirculation Pump" Closed Valve>< Open Valve Figure 4.6 ECCS Schematic AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paae 4-20 Figure 4.7 Axial Power Distributions at 102% Power / [ I AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 PaQe 4-21 Figure 4.8 Axial Power Distributions at 102% Power / [ I AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUMTM 10XM Fuel Page 5-1 5.0 Break Spectrum Analysis Description The objective of this LOCA analyses is to ensure that the limiting break location, break type, break size, and ECCS single failure are identified.

The LOCA response scenario varies considerably over the spectrum of break locations.

Potential break locations have been separated into two groups: recirculation line breaks and non-recirculation line breaks. The basis for the break locations and potentially limiting single failures analyzed in this report is described in the following sections.5.1 Limiting Single Failure Regulatory requirements specify that the LOCA analysis be performed assuming that all offsite power supplies are lost instantaneously and that only safety grade systems and components are available.

In addition, regulatory requirements also specify that the most limiting single failure of ECCS equipment must be assumed in the LOCA analysis.

The term "most limiting" refers to the ECCS equipment failure that produces the greatest challenge to event acceptance criteria.

The limiting single failure can be a common power supply, an injection valve, a system pump, or system initiation logic. The equipment identified (Reference

6) that may produce a limiting single failure (SF) is shown below: Backup battery power (SF-BATT)-Board A (SF-BATTIBA)

-Board B (SF-BATTIBB)

-Board C (SF-BATTIBC)

• Opposite unit false LOCA signal (SF-LOCA)* Low-pressure coolant injection valve (SF-LPCI)* Diesel generator (SF-DGEN)* High-pressure coolant injection system (SF-HPCI)* Automatic depressurization system (SF-ADS)-Failure of initiation logic (SF-ADSIIL)

-Failure of single ADS valve (SF-ADSISV)

The single failures and the available ECCS for each failure assumed in this analysis are summarized in Table 5.1 for recirculation line breaks and in Table 5.2 for non-recirculation line breaks. Other potential failures are not specifically considered because they all result in as much or more ECCS capacity.

Limiting single failures of Tables 5.1 and 5.2 were determined AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUMTM 1OXM Fuel Page 5-2 by available ECCS; this was accomplished by comparison of available coolant capacity and number of ADS valves.5.2 Recirculation Line Breaks The response during a recirculation line LOCA is dependent on break size. ADS operation is an important emergency system for small breaks. The ADS assists in depressurizinrg the reactor system, and thereby reduces the time required to reach rated LPCS and LPCI flow. For large breaks, rated LPCS and LPCI flow is generally reached before or shortly after the time when the ADS valves open so the ADS system is not required to mitigate the LOCA. The availability of ADS for each single failure is presented in Table 5.1. Table 5.1 clearly demonstrates that a single failure of the battery for board A or board B results in the least amount of ECCS capacity and is therefore limiting.

It should be noted that SF-LOCA and SF-DGEN are identical.

Large recirculation line break analyses are performed for breaks in both the discharge and suction side of the recirculation pump. It is generally expected that the pump suction side break will be more severe due to the more rapid blowdown.

The two largest flow resistances in the recirculation piping are the recirculation pump and the jet pump nozzle. For breaks in the discharge piping, there is a major flow resistance in both flow paths to the break. For breaks in the suction piping, there is only one major flow resistance between the break and reactor vessel. As a result, suction side breaks allow the coolant to exit more rapidly and generally result in more severe events than discharge side breaks (if ECCS capacity is equal). Both suction and discharge recirculation pipe breaks are considered in the break spectrum analysis.Two break types (geometries) are considered for the recirculation line break. The two types are the double-ended guillotine (DEG) break and the split break.For a DEG break, the piping is assumed to be completely severed resulting in two independent flow paths to the containment.

The DEG break is modeled by setting the break area (at both ends of the pipe) equal to the full pipe cross-sectional area and varying the discharge coefficient between 1.0 and 0.4. The range of discharge coefficients is used to cover uncertainty in the actual geometry at the break. Discharge coefficients below 0.4 are unrealistic and not considered in the EXEM BWR-2000 methodology.

AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUMTM 1OXM Fuel Page 5-3 A split type break is assumed to be a longitudinal opening or hole in the piping that results in a single break flow path to the containment.

Appendix K of 10 CFR 50 defines the cross-sectional area of the piping as the maximum split break area required for analysis.The rate of reactor vessel depressurization is slower for intermediate and small breaks (break area < 1.0 ft 2) compared to large break LOCAs. The HPCI and ADS will assist in reducing the reactor vessel pressure to the pressure where the LPCI and LPCS flows start.The break spectrum analyses in the intermediate and small break region consider break sizes between 1.0 ft 2 and 0.05 ft 2.Break sizes and single failures are analyzed for both suction and discharge recirculation line breaks.Section 6.0 provides a description and result summary for large, intermediate, and small breaks in the recirculation line.5.3 Non-Recirculation Line Breaks In addition to breaks in the recirculation line, breaks in other reactor coolant system piping must be considered in the LOCA break spectrum analysis.

Although the recirculation line large break results in the largest coolant inventory loss, it does not necessarily result in the most severe challenge to event acceptance criteria.

The double-ended rupture of a main steam line is expected to result in the fastest depressurization of the reactor vessel. Special consideration is required when the postulated break occurs in ECCS piping. Although ECCS piping breaks are small relative to a recirculation pipe DEG break, this break disables an ECCS system and therefore, increases the postulated break severity.

Table 5.2 summarizes the available ECCS components of the potentially limiting single failures.

The following sections address potential LOCAs due to breaks in non-recirculation line piping.5.3.1 Main Steam Line Breaks A steam line break inside containment is assumed to occur between the reactor vessel and the inboard main steam line isolation valve (MSIV) upstream of the flow limiters.

The break results in high steam flow out of the broken line and into the containment.

Prior to MSIV closure, a steam line break also results in high steam flow in the intact steam lines as they feed the break via the steam line manifold.

A large steam line break inside containment results in a rapid depressurization of the reactor vessel. Because of the rapid depressurization, the primary AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUMTM 1OXM Fuel Page 5-4 mitigation for the event is the low-pressure ECCS. Initially the break flow will be high quality steam; however, the rapid depressurization produces a water level swell that results in liquid discharge at the break. For steam line breaks, the largest break size results in the most level swell and liquid loss out of the break. [A small steam line break will result in a slower depressurization of the reactor vessel. Because the reactor vessel does not quickly depressurize, mitigation of the event is initially by HPCI and/or ADS, followed by the low-pressure ECCS. For the smallest break sizes, HPCI and ADS are significant systems for the overall mitigation of the event. HPCI could become degraded from a steam line split break or completely disabled from a guillotine break of the HPCI steam supply line (Figure 5.1). Because a steam line break could disable HPCI, no credit for HPCI is assumed. The limiting single failure from Table 5.2 for the steam line break/HPCl steam supply line break is SF-BATTIBB.

SF-BATTIBA is nonlimiting since it has the same available ECCS as SF-BATTIBB with additional ADS valves. A break in the HPCI steam supply line is equivalent to a break of the same size in the main steam line inside containment and is nonlimiting relative to event acceptance criteria.

Additional discussion of the HPCI steam supply line break is provided in Section 5.3.3.5.3.2 Feedwater Line Breaks AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUMTM 1OXM Fuel Page 5-5 The above comparison of a feedwater line break to a recirculation line break is not applicable if one of the available ECCS systems is HPCI. Because the HPCI injection line is connected to the feedwater line, as shown in Figure 5.1, a break in the feedwater line could result in a full or partial loss of HPCI injection, which could increase the severity of the event relative to a recirculation line break. No credit for HPCI is assumed for the feedwater line break analysis (HPCI flow is not modeled).

The limiting single failure from Table 5.2 for the feedwater line break is SF-BATTIBB.

SF-BATTIBA is nonlimiting since it has the same available ECCS as SF-BATTIBB with additional ADS valves. The limiting single failure for the feedwater line break has only 4 ADS valves as compared to 6 ADS valves for SF-BATTIBA for the recirculation suction line break; otherwise, the available ECCS is the same. For the same single failure (SF-BATTIBA), the recirculation discharge line break likewise has 6 ADS valves; however, the available low-pressure systems are less than the feedwater line break (1 LPCS versus 1 LPCS and 2 LPCI). [J An analysis of the feedwater line break is presented in Section 7 and confirms that the break is nonlimiting relative to the recirculation line break.5.3.3 HPCI Line Breaks Figure 5.1 shows the location of the HPCI injection line and HPCI steam supply line.The HPCI injection line is connected to the feedwater line outside of the containment.

As discussed in Section 5.3.2, a break in the feedwater line could result in a full or partial loss of HPCI injection.

The limiting single failure from Table 5.2 for the feedwater or HPCI injection line break is SF-BATTIBB.

SF-BATTIBA is nonlimiting since it has the same available ECCS as SF-BATTIBB with additional ADS valves. The break flow is not assumed to be terminated by an isolation or check valve closure; therefore, a break in the HPCI injection line is assumed to be equivalent to a break of the same size in the feedwater line inside containment.

An analysis confirms that the HPCI injection line break via a feedwater line break is nonlimiting relative to the recirculation line breaks, refer to Section 7.The HPCI steam supply line is connected to the main steam line inside of containment.

[AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Page 5-6 I I 5.3.4 LPCS Line Breaks A break in the LPCS line is expected to have many characteristics similar to [] However, some characteristics of the LPCS line break are unique and are not addressed in other LOCA analyses.

Two important differences from other LOCA analyses are that the break flow will exit from the region inside the core shroud and the break will disable one LPCS system. The LPCS line break is assumed to occur just outside the reactor vessel. [5.3.5 LPCI Line Breaks The LPCI injection lines are connected to the larger recirculation discharge lines. A break in a LPCI line would result in the partial or full loss of LPCI injection.

As shown in Table 5.2, the limiting single failures are SF-BATTIBA and SF-BATTIBB.

Because LPCl is assumed to be fully lost from the broken line, the available ECCS is equivalent to the recirculation discharge breaks.[5.3.6 RCIC Line Breaks The RCIC discharges to the feedwater line; [The steam supply to the RCIC turbine comes from the main steam line from the reactor vessel;I AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page 5-7 5.3.7 RWCU Line Breaks The RWCU extraction line is connected to a recirculation suction line with an additional connection to the vessel bottom head. []The RWCU return line is connected to the feedwater line; [I 5.3.8 Instrument Line Breaks[I AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Pace 5-8 Table 5.1 Available ECCS for Recirculation Line Break LOCAs Systems* t Remaining Assumed Failure Recirculationt Recirculation Suction Break Discharge Break Disposition SF-BATTIBA 6 ADS, 1 LPCS, 6 ADS, 1 LPCS Analyze 2 LPCI SF-BATTIBB 4 ADS, HPCI, 4 ADS, HPCI, Analyze 1 LPCS, 2 LPCI 1 LPCS SF-BATTIBC§ 4 ADS, HPCI, 4 ADS, HPCI, Bounded by SF-BATTIBB 1 LPCS, 3 LPCI 1 LPCS, 1 LPCI SF-LOCA 6 ADS, HPCI, 6 ADS, HPCI, Bounded by SF-BATTIBA and 1 LPCS, 2 LPCI 1 LPCS SF-BATTIBB SF-LPCI 6 ADS, HPCI, 6 ADS, HPCI, Bounded by SF-BATTIBA and 2 LPCS, 2 LPCI 2 LPCS SF-BATTIBB SF-DGEN 6 ADS, HPCi 6 ADS, HPCI, Bounded by SF-BATTIBA and 1 LPCS, 2 LPCI 1 LPCS SF-BATTIBB SF-HPCI 6 ADS, 2 LPCS, 6 ADS, 2 LPCS, Bounded by SF-BATTIBA 4 LPCI 2 LPCI SF-ADSIIL 4 ADS, HPCI, 4 ADS, HPCI, Bounded by SF-BATTIBB 2 LPCS, 4 LPCI 2 LPCS, 2 LPCI SF-ADSISV 5 ADS, HPCI, 5 ADS, HPCI, Bounded by SF-BATTIBB 2 LPCS, 4 LPCI 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.1 4 ADS, 5 ADS, and 6 ADS means the number of ADS valves available for automatic activation.

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 Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 5-9 Table 5.2 Available ECCS for Non-Recirculation Line Break LOCAs Non-Recirculation Assumed Available Line Break Failure ECCS* Disposition SF-BATTIBA 6 ADS, 1 LPCS, 2 LPCI Bounded by SF-BATTIBB SF-BATTIBB 4 ADS, 1 LPCS, 2 LPCI Analyze SF-BATTIBC 4 ADS, 1 LPCS, 3 LPCI Bounded by SF-BATTIBB SF-LOCA 6 ADS, 1 LPCS, 2 LPCI Bounded by SF-BATTIBB HPClt Feedwater linet SF-LPCI 6 ADS, 2 LPCS, 2 LPCI Bounded by SF-BATTIBB Steam linet SF-DGEN 6 ADS, 1 LPCS, 2 LPCI Bounded by SF-BATTIBB SF-HPCI 6 ADS, 2 LPCS, 4 LPCI Bounded by SF-BATTIBB SF-ADSIIL 4 ADS, 2 LPCS, 4 LPCI Bounded by SF-BATTjBB SF-ADSISV 5 ADS, 2 LPCS, 4 LPCI Bounded by SF-BATTIBB SF-BATTIBA 6 ADS, 2 LPCI Analyze SF-BATTIBB 4 ADS, HPCI, 2 LPCI Analyze SF-BATTIBC 4 ADS, HPCI, 3 LPCI Bounded by SF-BATTIBB SF-LOCA 6 ADS, HPCI, 2 LPCI Bounded by SF-BATTIBA and SF-BATTIBB LPCS SF-LPCI 6 ADS, HPCI, 1 LPCS, 2 LPCI Bounded by SF-BATTIBA and SF-BATTIBB SF-DGEN 6 ADS, HPCI, 2 LPCI Bounded by SF-BATTIBA and SF-BATTIBB SF-HPCI 6 ADS, 1 LPCS, 4 LPCI Bounded by SF-BATTIBA SF-ADSIIL 4 ADS, HPCI, 1 LPCS, 4 LPCI Bounded by SF-BATTIBB SF-ADSISV 5 ADS, HPCI, 1 LPCS, 4 LPCI Bounded by SF-BATTIBB t Refer to footnotes of Table 5.1 for additional information.

Assumes no credit for HPCI flow.AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM' 1OXM Fuel ANP-3152(NP)

Revision 0 Paae 5-10 Table 5.2 Available ECCS for Non-Recirculation Line Break LOCAs (Continued)

Non-Recirculation Assumed Available Line Break Failure ECCS* Disposition SF-BATTIBA 6 ADS, 1 LPCS Equivalent to recirculation pump discharge breaks (SF-BATTIBA).

Additional analysis not required.SF-BATTIBB 4 ADS, HPCI, 1 LPCS Equivalent to recirculation pump discharge breaks (SF-BATTIBB).

Additional analysis not required.SF-BATTIBC 4 ADS, HPCI, 1 LPCS, 1 LPCI Bound by SF-BATTIBB LPCI SF-LOCA 6 ADS, HPCI, 1 LPCS Bounded by SF-BATTIBA and SF-BATTIBB SF-LPCI 6 ADS, HPCI, 2 LPCS Bounded by SF-BATTIBA and SF-BATTIBB SF-DGEN 6 ADS, HPCI, 1 LPCS Bounded by SF-BATTIBA and SF-BATTIBB SF-HPCI 6 ADS, 2 LPCS, 2 LPCI Bounded by SF-BATTIBA SF-ADSIIL 4 ADS, HPCI, 2 LPCS, 2 LPCI Bounded by SF-BATTIBB SF-ADSISV 5 ADS, HPCI, 2 LPCS, 2 LPCI Bounded by SF-BATTIBB

• Refer to footnotes of Table 5.1 for additional information.

AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page 5-11 Reactor Vessel Main Steam Line Feedwater Line HPCI Injection Line Check Valves Isolation Valves Containment Figure 5.1 Steam, Feedwater, and HPCI Lines AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUM T M 10XM Fuel Page 6-1 6.0 Recirculation Line Break LOCA Analyses The largest diameter recirculation system pipes are the suction line between the reactor vessel and the recirculation pump and the discharge line between the recirculation pump and the riser manifold ring. LOCA analyses are performed for breaks in both of these locations with consideration for both DEG and split break geometries.

The break sizes considered included DEG breaks with discharge coefficients from 1.0 to 0.4 and split breaks with areas ranging between the full pipe area to 0.05 ft 2.As shown in Table 5.1, the limiting single failures considered in the recirculation line break analyses are SF-BATTIBA and SF-BATTIBB.

6.1 Limiting

Break Analysis Results The analyses demonstrate that the limiting (highest PCT) recirculation line break is the 0.20 ft 2 split break in the pump discharge piping for SF-BATTIBA and a top-peaked axial power shape.The limiting PCT is 1909'F. The key results and event times for this limiting break are provided in Tables 6.1 and 6.2, respectively.

Figures 6.1 -6.18 provide plots of key parameters from the RELAX system and hot channel blowdown analyses.

A plot of cladding temperature versus time in the hot assembly from the HUXY heatup analysis is provided in Figure 6.19. Table 6.3 presents the detailed break spectrum PCT.The limiting PCT for SF-BATTIBB is bound by the limiting PCT for SF-BATTIBA.

For SF-BATTIBB, the highest PCT occurred for a recirculation split line break of 0.5 ft 2 in the pump discharge piping and a top-peaked axial power shape. The PCT is 1658°F. The key results and event times for this limiting break are provided in Tables 6.4 and 6.5, respectively.

Table 6.6 presents the detailed break spectrum PCTs.The results of the break analyses are discussed in the following sections.

[AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUM T M 1OXM Fuel Page 6-2 6.2 Break Location Analysis Results Table 6.7 shows that the maximum PCT calculated for a recirculation line break occursin the pump discharge piping.6.3 Break Geometry and Size Analysis Results Recirculation line break PCT results versus break geometry (DEG or split) and size are presented in Tables 6.3 and 6.6. The maximum PCT calculated for a recirculation line break occurs for a 0.20 ft 2 split break for SF-BATTIBA.

For large breaks, depressurization occurs quickly; therefore, HPCI and ADS do not significantly mitigate the event. Depressurization to activate the low-pressure ECCS occurs before ADS activation.

Both SF-BATTIBA and SF-BATTIBB have LPCS to mitigate the event. SF-BATTIBB also has HPCI; therefore, the largest break sizes are performed only for SF-BATTIBA since it has the least available ECCS.The other differences in break sizes between Table 6.3 and Table 6.6 are due to the size of the limiting break and the resolution needed to determine the limiting break size. Table 6.3 consists of 0.01 ft 2 intervals near the limiting size of 0.20 ft 2.The limiting break for Table 6.6 is 0.5 ft 2; a larger break size interval was adequate to demonstrate that SF-BATTIBB was nonlimiting.

6.4 Limiting

Single-Failure Analysis Results As previously mentioned, SF-BATTIBA and SF-BATTIBB are the limiting single failures based on available ECGS.6.5 Axial Power Shape Analysis Results The results in Table 6.3 show that the top-peaked axial power shape is limiting compared to the mid-peaked shape analyses for the limiting break size.AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM"T 1OXM Fuel ANP-3152(NP)

Revision 0 Page 6-3 Table 6.1 Results for Limiting TLO Recirculation Line Break 0.20 ft 2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial 102% Power PCT 1909°F Maximum local cladding oxidation 1.20%Maximum planar average MWR 0.69%AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 10XM Fuel ANP-3152(NP)

Revision 0 Page 6-4 Table 6.2 Event Times for Limiting TLO Recirculation Line Break 0.20 ft 2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial 102% Power Event Initiate break Initiate scram Low-low liquid level, L2 (448 in)Low-low-low liquid level, L1 (372.5 in)Jet pump uncovers Recirculation suction uncovers Lower plenum flashes LPCS high-pressure cutoff LPCS valve pressure permissive LPCS valve starts to open LPCS valve fully open LPCS permissive for ADS timer LPCS pump at rated speed LPCS flow starts RDIV pressure permissive RDIV starts to close RDIV fully closed Rated LPCS flow ADS valves open Blowdown ends Bypass reflood Core reflood PCT Time (sec)0.0 0.6 40.1 65.7 95.6 155.9 182.1 280.9 270.0 272.0 305.0 94.7 97.7 280.9 312.5 314.5 350.5 374.5 187.7 374.5 480.7 423.4 423.4 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 PaQe 6-5 Table 6.3 TLO Recirculation Line Break Spectrum Results* for 102% Power SF-BATTIBA

• Results are the limiting PCTs from analyses performed at [each flow case are presented in Appendix B.t []]. Results for AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM IOXM Fuel ANP-3152(NP)

Revision 0 Paqe 6-6 Table 6.4 Results for TLO Recirculation Line Break 0.5 ft 2 Split Pump Discharge SF-BATTIBB Top-Peaked Axial 102% Power PCT 1658 0 F Maximum local cladding oxidation 0.36%Maximum planar average MWR 0.18%AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUM T M 1OXM Fuel Page 6-7 Table 6.5 Event Times for TLO Recirculation Line Break 0.5 ft 2 Split Pump Discharge SF-BATTIBB Top-Peaked Axial 102% Power Event Time (sec)Initiate break 0.0 Initiate scram 0.6 Low-low liquid level, L2 (448 in) 17.7 Low-low-low liquid level, LI (372.5 in) 28.1 Jet pump uncovers 37.3 Recirculation suction uncovers 66.6 Lower plenum flashes 60.7 HPCI valve fully open 52.7 HPCI flow starts 52.8 LPCS high-pressure cutoff 180.0 LPCS valve pressure permissive 171.3 LPCS valve starts to open 173.3 LPCS valve fully open 206.3 LPCS permissive for ADS timer 57.1 LPCS pump at rated speed 60.1 LPCS flow starts 180.0 RDIV pressure permissive 202.1 RDIV starts to close 204.1 RDIV fully closed 240.1 Rated LPCS flow 262.8 ADS valves open 150.1 Blowdown ends 262.8 Bypass reflood 363.1 Core reflood 297.1 PCT 297.1 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paae 6-8 Table 6.6 TLO Recirculation Line Break Spectrum Results* for 102% Power SF-BATTIBB

• [I AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Page 6-9 Table 6.7 Summary of TLO Recirculation Line Break Results Highest PCT Cases AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page 6-10 Figure 6.1 Limiting Recirculation Line Break Upper Plenum Pressure Figure 6.2 Limiting Recirculation Line Break Total Break Flow Rate AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Page 6-11 Cn'n U-o CO 100 200 300 400 500 TIME (SEC)600 700 a0m Figure 6.3 Limiting Recirculation ADS Flow Rate Line Break C4)0-I- II 10 100 200 300 400 500 TIME (SEC)Figure 6.4 Limiting Recirculation HPCI Flow Rate 6W0 700 80 Line Break AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Page 6-12 Uo In o0 b.N C30 100 200 300 400 TIME (SEC)50 600 700 O0W Figure 6.5 Limiting Recirculation Line Break LPCS Flow Rate (n 0 u-)bi 10 100 200 300 400 TIME (SEC)500 600 700 am0 Figure 6.6 Limiting Recirculation Line Break Intact Loop LPCI Flow Rate AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page 6-13 I-In'0 100 200 300 400 5W0 TIME (SEC)600 700 800 Figure 6.7 Limiting Recirculation Line Break Broken Loop LPCI Flow Rate I-ýa-0-, C0 100 200 300 400 TIME (SEC)500 600 700 am Figure 6.8 Limiting Recirculation Line Break Upper Downcomer Mixture Level AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page 6-14 I.?5-._1 L" dzi_1J TIE(SC C)cO 100 200 300 400 500 500 71 TIME (SEC)Figure 6.9 Limiting Recirculation Line Break Middle Downcomer Mixture Level 30 80a I-b.bi C bJ-J 3<bi C-)C)Id C)Figure 6.10 Limiting Recirculation Line Break Lower Downcomer Mixture Level AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 10XM Fuel ANP-3152(NP)

Revision 0 Page 6-15 C,)C" LasC C-)En"0 100 200 300 400 500 600 700 80W TIME (SEC)Figure 6.11 Limiting Recirculation Line Break Intact Loop Discharge Liquid Mass m c'i~U).4: cy~ 0 id 0~J "4 0~bJ ag-O 100 200 300 400 6W0 600 TIME (SEC)700 80 Figure 6.12 Limiting Recirculation Line Break Upper Plenum Liquid Mass AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page 6-16-o a--508-0-Jo 400 TIME (SEC)am0 Figure 6.13 Limiting Recirculation Line Break Lower Plenum Liquid Mass o-o 0B 1i0 200 308 400 TIME (SEC)5W0 600 700 amo Figure 6.14 Limiting Recirculation Line Break Hot Channel Inlet Flow Rate AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paae 6-17 C).~__j C-3!?10 I I I I I I I 100 200 300 400 TIME (SEC)600 8W 700 800 Figure 6.15 Limiting Recirculation Line Break Hot Channel Outlet Flow Rate Figure 6.16 Limiting Recirculation Line Break Hot Channel Coolant Temperature at the Hot Node at EOB AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paae 6-18 Figure 6.17 Limiting Recirculation Line Break Hot Channel Quality at the Hot Node at EOB Figure 6.18 Limiting Recirculation Line Break Hot Channel Heat Transfer Coef. at the Hot Node at EOB AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 6-19 2650 2000 Iwo*a1-E 5 1000 0 0 50 100 150 200 260 300 350 400 4W0 500 Time (sec)Figure 6.19 Limiting Recirculation Line Break Cladding Temperatures AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATR UMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page 7-1 7.0 Non-Recirculation Line LOCA Analysis LOCA analyses are performed for breaks in the feedwater and LPCS lines. Breaks in other non-recirculation lines are less limiting for the reasons discussed in Section 5.3. Note that for LPCS line break cases with no core spray available, the HUXY heatup analysis is performed with the core spray heat transfer coefficients equal to 0.0.7.1 Limiting Non-Recirculation Line Break Results The results of this analysis indicate that the limiting non-recirculation line break is the 0.4 ft 2 DEG break in the LPCS line with SF-BATTIBA and a top-peaked axial power shape. The PCT for the limiting ECCS line break is 1511 °F and maximum local cladding oxidation is 0.15%.The feedwater line break analysis was performed with the limiting single failure (SF-BATTIBB)

[] The feedwater line break was confirmed to be nonlimiting.

For the LPCS line break, [] Analyses are performed for both SF-BATTIBA and SF-BATTIBB as shown in Table 5.2.The key event times for the limiting non-recirculation break are provided in Table 7.1. Table 7.2 presents PCT results for the non-recirculation line breaks. The 1511 OF maximum PCT for non-recirculation line breaks is lower than the maximum PCT for recirculation line breaks of 1909*F. Therefore, non-recirculation line breaks are nonlimiting.

AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 7-2 Table 7.1 Event Times for Limiting Non-Recirculation Line Break 0.4 ft 2 Double-Ended Guillotine SF-BATTIBA Top-Peaked Axial 102% Power Event Initiate break Initiate scram Low-low liquid level, L2 (448 in)Low-low-low liquid level, Li (372.5 in)Jet pump uncovers Recirculation suction uncovers Lower plenum flashes LPCI high-pressure cutoff LPCI valve pressure permissive LPCI valve starts to open LPCI valve fully open LPCI pump at rated speed LPCI flow starts RDIV pressure permissive RDIV starts to close RDIV fully closed ADS valves open Blowdown ends Bypass reflood Core reflood PCT Time (sec)0.0 0.6 28.5 77.7 174.7 NA 55.6 198.7 185.5 187.5 227.5 110.7 200.6 242.8 244.8 280.8 199.7 298.3 349.4 326.2 326.2 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 7-3 Table 7.2 Non-Recirculation Line Break Spectrum Results*AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUMTM 1OXM Fuel Page 8-1 8.0 Single-Loop Operation LOCA Analysis During SLO the pump in one recirculation loop is not operating.

A break may occur in either loop, but results from a break in the inactive loop would be similar to those from a two-loop operation break. If a break occurs in the inactive loop during SLO, the intact active loop flow to the reactor vessel would continue during the recirculation pump coastdown period and would provide core cooling similar to the flow occurring in breaks during two-loop operation.

System response would be similar to that resulting from an equal-sized break during two-loop operation.

A break in the active loop during SLO results in a more rapid loss of core flow and earlier degraded core conditions relative to those from a break in the inactive loop. Therefore, only breaks in the active recirculation loop are analyzed.A break in the active recirculation loop during SLO results in an earlier loss of core heat transfer relative to a similar break occurring during two-loop operation.

This occurs because there is immediate loss of jet pump drive flow. Therefore, fuel rod surface temperatures increase faster in an SLO LOCA relative to a normal operation LOCA. Also, the early loss of core heat transfer results in higher stored energy in the fuel rods at the start of the heatup. The increased severity of an SLO LOCA can be reduced by applying an SLO multiplier to the two-loop MAPLHGR limits. []8.1 SLO Analysis Modeling Methodology AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 8-2 I I 8.2 SLO Analysis Results[The SLO SF-BATTIBA analyses are performed with a 0.85 multiplier applied to the TLO MAPLHGR limit resulting in an SLO MAPLHGR limit of 11.05 kW/ft. The analyses are performed at BOL ATRIUM 1OXM fuel conditions.

The limiting SLO LOCA is the 0.2 ft 2 split pump discharge line break with SF-BATTIBA and a top-peaked axial power shape. The PCT for this case is 1747 0 F. Other key results and event times for the limiting SLO LOCA are provided in Tables 8.1 and 8.2, respectively.

Table 8.3 shows the spectrum of SLO analyses and the PCT for each case. A comparison of the limiting SLO and the limiting TLO results is provided in Table 8.4. The results in Table 8.4 show that the limiting TLO LOCA results bound the limiting SLO results when a 0.85 multiplier is applied to the TLO MAPLHGR limit.AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 8-3 Table 8.1 Results for Limiting SLO Recirculation Line Break 0.2 ft 2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial [ ]PCT 1747 0 F Maximum local cladding oxidation 0.66%Maximum planar average MWR 0.39%AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 8-4 Table 8.2 Event Times for Limiting SLO Recirculation Line Break 0.2 ft 2 Split Pump Discharge SF-BATTIBA Top-Peaked Axial [ ]Event Initiate break Initiate scram Low-low liquid level, L2 (448 in)Low-low-low liquid level, Li (372.5 in)Jet pump uncovers Recirculation suction uncovers Lower plenum flashes LPCS high-pressure cutoff LPCS valve pressure permissive LPCS valve starts to open LPCS valve fully open LPCS permissive for ADS timer LPCS pump at rated speed LPCS flow starts RDIV pressure permissive RDIV starts to close RDIV fully closed Rated LPCS flow ADS valves open Blowdown ends Bypass reflood Core reflood PCT Time (sec)0.0 0.6 39.3 64.7 94.7 155.2 180.7 280.1 269.1 271.1 304.1 93.7 96.7 280.0 311.5 313.5 349.5 373.6 186.7 373.6 483.6 421.7 421.7 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Page 8-5 Table 8.3 SLO Recirculation Line Break Spectrum Results*for [ ] SF-BATTIBA t[I AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe 8-6 Table 8.4 Single- and Two-Loop Operation PCT Summary Limiting PCT Operation Case (OF)Single-loop 0.2 ft 2 split pump discharge 1747 top-peaked SF-BATTIBA Two-loop 0.2 ft 2 split pump discharge top- 1909 peaked SF-BATTIBA AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUM T M 1OXM Fuel Page 9-1 9.0 Long-Term Coolability Long-term coolability addresses the issue of reflooding the core and maintaining a water level adequate to cool the core and remove decay heat for an extended time period following a LOCA. For non-recirculation line breaks, the core can be reflooded to the top of the active fuel and be adequately cooled indefinitely.

For recirculation line breaks, the core will initially remain covered following reflood due to the static head provided by the water filling the jet pumps to a level of approximately two-thirds core height. Eventually, the heat flux in the core will not be adequate to maintain a two-phase water level over the entire length of the core. Beyond this time, the upper third of the core will remain wetted and adequately cooled by core spray.Maintaining water level at two-thirds core height with one core spray system operating is sufficient to maintain long-term coolability as demonstrated by the NSSS vendor (Reference 7).AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA ANP-3152(NP)

Break Spectrum Analysis for ATRIUM T M Revision 0 10XM Fuel Page 10-1 10.0 Conclusions The major conclusions of this LOCA break spectrum analysis are:* The limiting recirculation line break is a 0.2 ft 2 split break in the pump discharge piping with single failure SF-BATTIBA and a top-peaked axial shape for two-loop operation.

0 The limiting .break analysis identified above satisfies all acceptance criteria specified in 10 CFR 50.46. The analysis is performed in accordance with 10 CFR 50.46 Appendix K requirements.

-Peak PCT < 2200'F (1909°F).-Local cladding oxidation thickness

< 0.17 (0.012).Total hydrogen generation

< 0.01. The break spectrum analysis determined a maximum planar average MWR of 0.0070. The core-wide metal-water reaction (CMWR) is significantly less than the maximum planar average MWR. It is concluded that CMWR would be less than 0.01 total hydrogen generation.

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.

• Breaks in the non-recirculation lines are less limiting than the most severe break in the recirculation line.0 The MAPLHGR limit multiplier for SLO is 0.85 for ATRIUM 1 OXM fuel. This multiplier ensures that a LOCA from SLO is less limiting than a LOCA from two-loop operation.

The limiting break characteristics determined in this report can be referenced and used in future Browns Ferry Units 1, 2, and 3 LOCA analyses to establish the MAPLHGR limit versus exposure for ATRIUM 1 OXM fuel.AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 ANP-3152(NP)

LOCA Break Spectrum Analysis Revision 0 for ATRIUMTM 10XM Fuel Page 11-1 11.0 References

1. EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001.2. 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 1975.3. XN-NF-82-07(P)(A)

Revision 1, Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, Exxon Nuclear Company, November 1982.4. 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 1984.5. EMF-2292(P)(A)

Revision 0, ATRIUMTm-1O:

Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.6. ANP-3014(P)

Revision 0, Browns Ferry Units 1, 2, and 3 LOCA Parameters Document, AREVA NP, July 2011.7. NEDO-20566A, Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR Appendix K, General Electric Company, September 1986.8. ANP-2908(P)

Revision 0, Browns Ferry Units 1, 2, and 3 105% OLTP LOCA Break Spectrum Analysis, AREVA NP, March 2010.9. Letter, P. Salas (AREVA) to Document Control Desk, U.S. Nuclear Regulatory Commission, Proprietary Viewgraphs and Meeting Summary for Closed Meeting on Application of the EXEM BWR-2000 ECCS Evaluation Methodology, NRC: 11:096, September 22, 2011.10. 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 Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe A-1 Appendix A Computer Codes Table A.1 Computer Codes Used for Break Spectrum Analysis AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page B-1 Appendix B Additional Results for SF-BATTIBA TLO Recirculation Line Breaks AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis forATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Page B-2 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page B-3 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paae C-1 Appendix C Additional Results for SF-BATTIBB TLO Recirculation Line Breaks AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM T M 1OXM Fuel ANP-3152(NP)

Revision 0 Page C-2 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Page C-3 AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paae D-1 Appendix D Additional Results for Non-Recirculation Line Breaks AREVA NP Inc.

Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUMTM 1OXM Fuel ANP-3152(NP)

Revision 0 Paqe D-2 AREVA NP Inc.