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ANP-3106(NP), Revision 2, Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for Atrium 10XM Fuel for Mellla+ Operation.
	ML16257A406
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Site:	Brunswick
Issue date:	12/31/2015
From:	; AREVA
To:	; Office of Nuclear Reactor Regulation
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ML16257A418	List: BSEP 16-0056, ANP-3108(NP), Revision 1, Applicability of Areva BWR Methods to Brunswick Extended Power Flow Operating Domain.; ML16257A404; ML16257A406; ML16257A407; ML16257A408; ML16257A410; ML16257A411;
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ANP-3106NP, Revision 2, Brunswick Units 1and2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel *for MELLLA+ Operation, December 2015 BSEP 16-0056 Enclosure 19 Controlled Document A AREVA Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 1 OXM Fue'I for MELLLA+ Operation December 2015 © 2015 AREVA Inc. ANP-3106NP Revision 2 Controlled Document Copyright

© 2015 AREVA Inc. All Rights Reserved ANP-3106NP Revision 2 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Nature of Changes Item Page Description and Justification

1. Table 2.1 Update the LOCA results for new limiting lattice. ANP-3106NP Revision 2 Page i 2. Page 5-1 Update limiting PCT, maximum MWR , and planar MWR for new limiting lattice. 3. Table 5.2 Update MAPLHGR analysis results for new limiting lattice. 4. Figure 5.26 Update the plot for the limiting break cladding temperature for a new limiting lattice. Changed items are further identified by yellow highlighting.

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Contents ANP-3106NP Revision 2 Page ii

1.0 INTRODUCTION

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1-1 2.0

SUMMARY

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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

................................................................................

4-1 4.1 Slowdown Analysis ...........................................................................................

4-1 4.2 Refill/Reflood Analysis ......................................................................................

4-2 4.3 Heatup Analysis ................................................................................................

4-3 4.4 Plant Parameters

..............................................................................................

4-3 4.5 ECCS Parameters

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4-4 5.0 MAPLHGR ANALYSIS DESCRIPTION AND RESUL TS ...................

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5-1 5.1 Thermal Conductivity Degradation

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5-1

6.0 CONCLUSION

S

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6-1

7.0 REFERENCES

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7-1 Tables 2.1 LOCA Results for Limiting Conditions

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2-2 4.1 Initial Conditions

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4-5 4.2 Reactor System Parameters

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4-6 4.3 ATRIUM 10XM Fuel Assembly Parameters

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4-7 4.4 High-Pressure Coolant Injection Parameters

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4-8 4.5 Low-Pressure Coolant Injection Parameters

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4-9 4.6 Low-Pressure Core Spray Parameters

...........................................................

.4-10 4.7 Automatic Depressurization System Parameters

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.4-11 4:8 Available ECCS for Recirculation Line Break LOCAs .....................................

.4-12 5.1 Event Times for Limiting Break 3.6 ft 2 Split Pump Discharge SF-LPCI

Top-Peaked Axial ..............................................................................................

5-3 5.2 ATRIUM 10XM MAPLHGR Analysis Results ....................................................

5-4 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 1 OXM Fuel for MELLLA+ Operation ANP-3106NP Revision 2 Page iii 2.1 4.1 4.2 4.3 4.4 4.5 5.1 5.2 5.3 5.4 5.5 5.6 5.7 5.8 5.9, 5.10 . 5.11 5.12 5.13 5.14 5.15 ' 5.16 5.17 5.18 5.19 5.20 5.21 5.22 5.23 5.24 5.25 5.26 Figures MAPLHGR Limit for ATRIUM 10XM Fuel.. ........................................................

2-3 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model. .....................

. .4-13 RELAX System Model ....................................................................................

4-14 RELAX Hot Channel Model Top-Peaked Axial. ..............................................

.4-15 ECCS Schematic

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4-16 Rod Average Power Distribution in RELAX Calculation

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.4-17 Limiting Break Upper Plenum Pressure ............................................................

5-5 Limiting Break Total Break Flow Rate ...............................................................

5-5 Limiting Break Core Inlet Flow Rate ...................................................................

5-6 Limiting Break Core Outlet Flow Rate ...............................................................

5-6 Limiting Break Intact Loop Jet Pump Drive Flow Rate .. : ....................................

5-7 -Limiting Break Intact Loop Jet Pump Suction Flow Rate ...................................

5-7 Limiting Break Intact Loop Jet Pump Exit Flow Rate .........................................

5-8 Limiting Break Broken Loop Jet Pump Drive Flow Rate ....................................

5-8 Limiting Break Broken Loop Jet Pump Suction Flow Rate ............................. , ... 5-9 Limiting Break Broken Loop Jet Pump Exit Flow Rate ......................................

5-9 Limiting Break ADS Flow Rate ....................

.' ...................................................

5-10 Limiting Break LPCS Flow Rate ......................................................................

5-10 Limiting Break Intact Loop LPCI Flow Rate ......................................................

5-11 Limiting Break Broken Loop LPCI Flow Rate ..................................................

5-11 Limiting Break Upper Downcomer Mixture Level. ............................................

5-12 Limiting Break Lower Downcomer Mixture Level... ..........................................

5-12 Limiting Break Intact Loop Discharge Line Liquid Mass ..................................

5-13 Limiting Break Upper Plenum Liquid Mass ......................................................

5-13 Limiting Break Lower Plenum Liquid Mass ......................................................

5-14 Limiting Break Hot Channel Inlet Flow Rate ....................................................

5-14 Limiting Break Hot Channel Outlet Flow Rate .................................................

5-15 Limiting Break Hot Channel Coolant Temperature at the Hot Node at EOB .... 5-15 Limiting Break Hot Channel Quality at the Hot Node at EOB ..........................

5-16 Limiting Break Hot Channel Heat Transfer Coeff. at the Hot Node at EOB ..... 5-16 Limiting Break Hot Channel Reflood Junction Liquid Mass Flow Rate .............

5-17 Limiting Break Cladding Temperatures

...........................................................

5-17 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation ADS ADSVOOS ANS BWR CFR CMWR DEG DG ECCS EOB HPCI LHGR LOCA LPCI. LPCS MAPLHGR MCPR MWR NRC PCT RDIV SF-BATT SF-HPCI SF-LPCI SLO TCD Nomenclature automatic depressurization system ADS valve out of service American Nuclear Society boiling-water reactor Code of Federal Regulations core average metal-water reaction double-ended guillotine diesel generator 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 metal-water reaction Nuclear Regulatory Commission, U.S. peak cladding temperature recirculation discharge isolation valve single failure of battery (DC) 'power single failure of the HPCI system . single failure of an LPCI valve single-loop operation thermal conductivity degradation ANP-3106NP Revision 2 Page iv Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation

1.0 Introduction

ANP-3106NP Revision 2 Page 1-1 The results of loss-of-coolant accident emergency core cooling system (LOCA-ECCS) analyses for Brunswick Units 1 and 2 are documented in this report. The results provide the maximum average planar linear heat generation rate (MAPLHGR) limit for ATRIUMŽ 10XM* fuel as a function of exposure for normal (two-loop) operation.

As shown in Reference 1, the MAPLHGR limit for single-loop op13ration (SLO) is equal to 0.80 times the two-loop limit. Operation in the MELLLA+ domain of the Brunswick power/flow map is supported.

The analyses documented in this report were performed with LOCA Evaluation Models developed by AREVA 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-2000 Evaluation Model for the Brunswick Units 1 and 2 LOCA break spectrum analysis to support MELLLA+ operation 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 I 3.6 ft 2 *Single failure Low-pressure coolant injection valve Axial power shape Top-peaked

ATRIUM is a trademark of AREVA Inc.

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation ANP-3106NP Revision 2 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 1 OXM fuel is shown in Figure 2.1. 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. A limiting .Brunswick ATRIUM 1 OXM neutronic design was used in the heatup analyses

performed for this report. Results for the limiting neutronic design are presented in Section 5.0. The peak cladding temperature (PCT) and metal-water reaction (MWR) results for the . ATRIUM 10XM fuel are presented in Table 2:1. The SLO analyses (Reference

1) support operation with an ATRIUM 10XM l\llAPLHGR multiplier of 0.80 applted to the normal two-loop operation MAPLHGR limit. [ ' ]

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Table 2.1 LOCA Results for Limiting Conditions Parameter ATRIUM 10XM Exposure (GWd/MTU)

Peak cladding temperature

(°F) Local cladding oxidation (max%) Total hydrogen generated

(%of total hydrogen possible) 0.0 1923 1.23 < 0.56 ANP-3106NP Revision 2 Page 2-2 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation 14.0 ,.-.,, -+-' '+-...........

12.0 3: '-" -+-' .E 10.0 =:i O::'. (!) I _J 0.. 8.0 <( 2 6.0 .0 10.0 20.0 30.0 40.0 50.0 Planar Average Exposure (GWd/MTU)

Average Planar ATRIUM 10XM Exposure MAPLHGR (GWd/MTU) (kW/ft) 0 13.1 15 13.1 67 7.7 ', Figure 2.1 MAPLHGR Limit for ATRIUM 10XM Fuer ANP-3106NP Revision 2 Page 2-3 60.0 70.0 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LO.CA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation 3.0 LOCA Description

3.1 Accident

Description ANP-3106NP Revision 2 Page 3-1 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 froni 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 primary containment before the first isolation valve. For a boiling water reactor (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 event acceptance criteria ( 10 CFR 50.46). Because of these complexities, an analysis covering the full range of break sizes and locations is required.

The results of the Brunswiqk Units 1 and 2 ATRIUM 10XM break spectrum calculations using the 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. Consistent with the discussion presented in References 3 and 4, [ 1 Controlled Document AREVA Inc. . ANP-3106NP Brunswick Units 1 and 2 LOCA-ECCS Analysis Revision 2 MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Page 3-2 [ ] The end of the blowdown phase is defined to occur when the system reaches the pressure corresponding to rated LPCS flow. In the refill phase of a LOCA, the ECCS 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 anq 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 duri.ng 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) fqr 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.

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation ANP-3106NP Revision 2 Page 3-3

Calculated changes in core geometry shall be such that the core remains amenable to cooling.

After any calculated successful initial operation of the ECCS, the calculated core, temperature shall be maintained at an acceptably low value and decay heat shall be removed 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 each fuel type to ensure that these criteria are met. 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 .

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation 4.0 LOCA Analysis Description ANP-3106NP Revision 2 Page 4-1 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 5). RODEX2 is used to determine the initial stored energy for both the blowdown analysis (RELAX 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 Slowdown

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 decay heat as required by Appendix K of 10 50. The reactor 'vessel nodalization for the system analysis is shown in Figure 4.2. This nodalization is consistent with that used in the topical report submitted to the NRC (Reference 2). The ,RELAX blowdown 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.

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 1 OXM Fuel for MELLLA+ Operation ANP-3106NP Revision 2 Page 4-2 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 blowdown calculation determines the hot channel cladding, and coolant temperatures during the blowdown phase of the LOCA. The RELAX hot channel nodalization is shown in Figure 4.3 for a top-peaked power shape. The hot channel blowdown 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 c9nservative relative to the average fuel rod temperature calculated by RODEX2 for operation of the ATRIUM 1 OXM assembly at the MAPLHGR limit. The heat transfer coefficient and fluid condition results from the RELAX hot channel calculation are used as input to the HUXY heatup analysis.

4.2 Refill/Reflood Analysis The RELAX code is also 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 the period 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. [ 1 The RELAX calculations provide HUXY Vl(ith the time of hot node reflood and the time wheri the liquid has risen in the bypass to the height of the axial plane of interest (time of bypass reflood).

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation

4.3 Heatup

Analysis ANP-3106NP Revision 2 Page 4-3 The HUXY code (Reference

6) 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 (Reference 7). The HUXY code complies with the 10 CFR 50 Appendix K criteria for LOCA Evaluation Models. HUXY uses the EOB 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 arialysis.

The principal results of a HUXY heatup analysis are the PCT and the percent local oxidation of the fuel cladding, often called the %MWR. The core average metal-water reaction (CMWR) criterion of less than 1.0% can often be satisfied by demonstrating that the maximum planar MWR calculated by HUXY is less than 1.0%. 4.4 Plant Parameters The LOCA break spectrum analysis is performed using plant parameters provided by the utility. Table 4.1 provides a summary of reactor initial conditions used in the Reference 1 limiting break analysis.

Table 4.2 lists selected reactor system parameters.

The break spectrum analysis is performed for a full core of ATRIUM 10XM fuel. Some of the key fuel parameters used in the break spectrum analysis are summarized in Table 4.3. A top-Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM toXM Fuel for MELLLA+ Operation ANP-3106NP Revision 2 Page 4-4 peaked axial power shape based on the rod average power distribution, shown in Figure 4.5, was identified as the most conservative power shape for the limiting break (Reference 1 ). 4.5 ECCS Parameters The ECCS configuration is shown in Figure 4.4, Table 4.4 -Table 4. 7 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. Although HPCI is expected to be available, no analysis mitigation credit is assumed for the HPCI system in any of the analyses discussed in this report. 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 Table 4.4 -Table 4.6. No credit for ECCS flow is assumed until the ECCS injection valves are fully open and the ECCS pumps reach rated speed. [ * ] The automatic depressurization (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. Only five ADS valves are assumed operable in the analyses to support operation with one ADSVOOS and the potential single failure of one ADS valve during the LOCA. 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 LPCS or LPCI due to high drywell pressure.

[ ] The potentially limiting single failures of the ECCS are provided in Section 5.0 of Reference

1. \ Table 4.8 shows these failures and gives the ECCS systems that are available for each assumed failure.

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation

-; Table 4.1 Initial Conditions Parameter Reactor power (% of rated) [ Reactor power (MWt) [ [ Steam flow rate (Mlb/hr) Steam dome pressure (psia) Core inlet enthalpy (Btu/lb) ATRIUM 10XM hot assembly MAPLHGR (kW/ft) [ Rod Average power distribution

[ Value 102 2981.5 13.1 1048.9 527.7 13.1 Figure 4.5 ] ] ] ] ] ANP-3106NP Revision 2 Page 4-5 Controlled Docu AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for . MELLLA+ Operation Table 4.2 Reactor System Parameters Parameter Value Vessel ID (in) 220.5 Number of fuel assemblies 560 Recirculation suction pipe area (ft 2) 3.67 1.0 DEG suction break area (ft 2)

7.33 Recirculation discharge pipe area (ft 2) 3.67 1.0 DEG discharge break area (ft 2) 7.33 ANP-3106NP Revision 2 Page 4-6 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Table 4.3 ATRIUM 10XM Fuel Assembly Parameters Parameter Fuel rod array Number bf fuel rods per assembly Non-fuel rod type Fuel rod GD (in) Active fuel length (in) (including blankets)

Water channel outside width (in) Fuel channel thickness (in) Fuel channel internal width (in) Value 10x10 79 (full-length rods) 12 rods) Water channel replaces 9 fuel rods 0.4047 150.0 (full-length rods) 75.0 (part-length rods) 1.378 -0.075 (minimum wall) 0.100 (corner) 5.278 ANP-3106NP Revision 2 Page 4-7 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation

Table 4.4 High-Pressure Coolant Injection Parameters Parameter Coolant temperature (maximum)

(°F) Value 140 Initiating Signals and Setpoints Water level (in)* High drywell pressure (psig) Time Delays Time for HPCI pump to reach rated speed and injection valve 459 Not used wide open (sec) . 60 Relative to vessel zero. Delivered Coolant Flow Rate Versus Pressure Vessel to Torus LiP (psid) 0 150 1164 Flow Rate (gpm) 0 3,825 3,825 ANP-3106NP Revision 2 Page 4-8 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation

Table 4.5 Low-Pressure Coolant Injection Parameters Parameter Value Reactor pressure permissive for opening valves -analytical (psia) 410 Coolant temperature (maximum)

(°F) 160 Initiating Signals and Setpoints Water level (in)* High drywell pressure (psig) Time Delays Time for LPCI pumps to reach 358 Not used rated speed (maximum) (sec) 31.8 LPCI injection valve stroke time (sec) 37.5 Vessel to Torus .!lP (psid) 0 20 202 Relative to vessel zero. Delivered Coolant Flow Rate Versus Pressure Flow rate for 1 pump injecting into 1 recirculation loop (gpm) 8,690 7,000 0 Flow rate for 2 pumps injecting into 1 recirculation loop (gpm) 14,420 12,000 0 ANP-3106NP Revision 2 Page 4-9 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation

Table 4.6 Low-Pressure Core Spray Parameters Parameter Reactor pressure permissive for opening valves -analytical (psia) Coolant temperature (maximum)

(°F) Water level (in)* Initiating Signals and Setpoints High drywell pressure (psig) Time Delays Time for LPCS pumps to reach Value 410 160 358 Not used rated speed (maximum) (sec) 39.7 LPCS injection valve stroke time (sec) 14.0. Relative to vessel zero. Delivered Coolant Flow Rate Versus Pressure Vessel to Flow rate Torus ilP for 1 pump (psid) (gpm) 0 5,250 113 4,000 265 0 ANP-3106NP Revision 2 Page 4-10 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Table 4. 7 Automatic Depressurization System Parameters Parameter Value Number of valves installed 7 Number of valves available*

5 Minimum flow capacity of available valves 4.15 at (Mlbm/hr at psig) 1112.4 Initiating Signals and Setpoints Water level (in)t 358 High drywell pressure (psig)+ 2 Time Delays ADS timer (delay time from initiating signal to time valves are open (sec) 121 ANP-3106NP Revision 2 Page 4-11

Only 5 valves are assumed operable in the analyses to support 1 ADSVOOS operation and the potential single failure of 1 ADS valve during the LOCA. t :i: [ Relative to vessel zero. l Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation ANP-3106NP Revision 2 Page 4-12 Assumed Failure* DC power (i) Table 4.8 Available ECCS for Recirculation Line Break LOCAs Recirculation Suction Break Systems Remaining

t. +. § Recirculation Discharge Break Systems Remaining

+. § 1 LPCS + 3LPCI + ADS 1 LPCS + 1 LPCI + ADS (SF-BATT)

t § DC power U) 2 LPCS + 2LPCI + HPCI + ADS 2LPCS + HPCI + ADS Diesel generator (i) 1 LPCS + 3LPCI + HPCI + ADS 1 LPCS + 1 LPCI + HPCI + ADS Diesel generator U) 2LPCS + 2LPCI + HPCI + ADS 2LPCS + HPCI + ADS LPCI injection valve 2LPCS + 2LPCI + HPCI + ADS 2LPCS + HPCI + ADS (SF-LPCI}

HPCI system 2LPCS + 4LPCI + ADS 2LPCS + 2LPCI + ADS (SF-HPCI}

Failure of either DC power (i) or diesel generator (i) will result in the loss of one diesel generator (DG-1 or DG-2). The loss of DC power (i) will also result in the loss of the HPCI. The loss of DC power U) or diesel generator U) will result in the loss of one diesel generator (DG-3 or DG-4). 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. 1 LPCI (1 pump into 1 loop) means one RHR pump operating in one LPCI loop, 2LPCI (2 pumps into 1 loop) means two RHR pumps operating in one loop, 3LPCI (3 pumps into 2 loops) means three RHR pumps operating in two loops, 4LPCI (4 pumps into 2 loops) means four RHR pumps operating in two loops. Although HPCI is expected to be available for some events, no accident analysis mitigation credit is assumed for this system.

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation I Fuel Data I: ... ... ..... Neutronic Data ... ...._ Core LiP, _ ... RPF,APF (CASMO, System MICROBURN)

Analysis l (RELAX) __/ Fuel Parameters (RODEX2) I Fuel Stored Energy -Boundary Conditions (power, upper & lower plenum conditions)

Gap, Fuel Stored Gap Coefficient, Fission Gas .. Hot Assembly Analysis* (RELAX) I Boundary Conditions (Pressure, Temperature, Power, Quality, Heat Transfer Coefficient)

Time of Hot Node Reflqod

ANP-3106NP Revision 2 Page 4-13 : I Plant Data I SS CoreT/H (XCOBRA) Heatup Analysis End of Slowdown, _ Time of Bypass Reflood *The hot assembly calculation may be combined with the system calculation or executed separately (HUXY) Peak Cladding Temperature, Metal Water Reaction + Figure 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Figure 4.2 RELAX System Model ANP-3106NP Revision 2 Page 4-14 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Figure 4.3 RELAX Hot Channel Model Top-Peaked Axial ANP-31Q6NP Revision 2 Page 4-15 ':;}"

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation (j) DG-3 1A LPCI Injection Valve 1A 1C (i) DG-1 Discharge Valve 1A 1A Loop-A Figure 4.4 ECCS Schematic 18 Loop-8 (i) DG-2 Discharge Valve 18 ANP-3106NP Revision 2 10 Page 4-16 (j) DG-4 18 LPCI ln1ection Valve 18 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Figure 4.5 Rod Average Power Distribution in RELAX Calculation

/ ANP-3106NP Revision 2 Page4-17 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation

5.0 MAPLHGR

Analysis Description and Results ANP-3106NP Revision 2 Page 5-1 An exposure-dependent MAPLHGR limit for ATRIUM 10XM 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 Figure 5.1 to Figure 5.26. In the MAPLHGR analysis , the 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 13.1 kW /ft for the ATRIUM 1 OXM fuel. Table 5.2 shows the MAPLHGR analysis results for the ATRIUM 1 OXM fuel. The HUXY model of the ATRIUM 10XM fuel is applied to obtain these results as described in Section 4.3. The HUXY analysis is performed at assembly average planar exposure intervals between 0 and 67 GWd/MTU. The MAPLHGR limits are provided for an assembly average planar exposure range which ensures appropriate limits are applied up to the monitored maximum assembly average and rod average exposure limits of 54 GWd/MTU and 60 GWd/MTU , respectively. The HUXY MAPLHGR input is consistent with the data in Figure 2.1. Exposure-dependent 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 10XM limiting PCT is 1923°F at 0.0 GWd/MTU exposure. The maximum local MWR of 1.23% occurred at 0.0 GWd/MTU exposure. Analysis results show that the planar average MWR at the peak power plane is 0.56%. Since all other planes in the core are at lower power , the CMWR will be significantly less than 0.56%. Figure 5.26 shows the cladding temperature of the ATRIUM 10XM PCT rod as a function of time

f or the lim i ting break. The maximum temperature of 1923°F occurs at 175. 7 seconds. These results demonstrate t he acceptability of t he ATRIUM 10XM MAPLHGR l i mit shown in Figure 2.1. 5.1 Thermal Conduc t ivity Degradation The RODEX2 code was approved by the NRC in the early 1980s. At that time , thermal conductivity degradation with burnup was not well characterized by irradiation or post-irradiation Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 1 OXM Fuel for MELLLA+ Operation ANP-3106NP Revision 2 Page 5-2 testing. As a result, all fuel codes at that time did not account for thermal conductivity degradation (TCD). In the past 20 years, requests to the NRC have been made for commercial fuel operation to increasingly higher burnup levels. This has resulted in renewed interest in the degree and nature of burn up-induced TCD. Reference 9 summarizes how TCD is addressed in the AREVA LOCA MAPLHGR analyses for Brunswick.

The impact of TCD is included in the results summar i zed in Table 5.2. The assessment for Brunswick ATRIUM 1 OXM fuel shows that the PCT calculated at 0.0 GWd/MTU, which is not affected by TCD , is the highest exposure-dependent PCT.

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Table 5.1 Event Times for Limiting Break 3.6 ft 2 Split Pump Discharge SF-LPCI Top-Peaked Axial Event Time (sec) Initiate break 0.0 Initiate scram 0.6 Low-low liquid level, L2 (459 in) 5.5 Low-low-low liquid level, L 1 (358 in) 8.2 Jet pump uncovers 9.3 Recirculation suction uncovers 15.8 Diesel generators started 15.0 LPCS high-pressure cutoff 60.5 Power at LPCS injection valves 27.8 LPCS valve pressure permissive 48.3 LPCS valve starts to open 49.3 LPCS valve open 63.3 LPCS pump at rated speed 39.7 LPCS flow starts 63.4 LPCS permissive for ADS 39.7 RDIV pressure permissive 57.2 RDIV starts to close 58.2 RDIV closed 95.2 Rated LPCS flow 90.0 ADS valves open 129.2 Slowdown ends 90.0 Bypass reflood 176.7 Core reflood 175.7 PCT 175.7 ANP-3106NP Revision 2 Page 5-3 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Table 5.2 ATRIUM 10XM MAPLHGR Analysis Results Average Planar Exposure MAPLHGR PCT (GWd/MTU) (kW/ft) (oF) 0 13.1 1923 5 13.1 1883 10 13.1 1846 15 13.1 1840 20 12.58 1775 30 11.54 1724 40 10.50 1663 50 9.47 1567 60 8.43 1490 67 7.7 1421 CMWR is <0.56% at all exposures.* Local Cladding Oxidation

(%) 1.23 1.02 0.81 0.79 0.98 0.93 0.73 0.29 0.16 0.10 ANP-3106NP Revision 2 Page 5-4

The plana r average MWR for the peak power plane is 0.56% which supports the conclusion that the CMWR is less than 0.56%.

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+, Operation Figure 5.1 Limiting Break Upper Plenum Pressure Figure 5.2 Limiting Break Total Break Flow Rate ANP-3106NP Revision 2 Page 5-5 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Figure 5.3 Limiting Break Core Inlet Flow Rate Figure 5.4 Limiting Break Core Outlet Flow Rate ANP-3106NP Revision 2 Page 5-6 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Figure 5.5 Limiting Break Intact Loop Jet Pump Drive Flow Rate *Figure 5.6 Limiting Break Intact Loop Jet Pump Suction Flow Rate ANP-3106NP Revision 2 Page 5-7 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM tOXM Fuel for MELLLA+ Operation Figure 5. 7 Limiting Break Intact Loop Jet Pump Exit Flow Rate Figure 5.8 Limiting Break Broken Loop Jet Pump Drive Flow Rate ANP-3106NP Revision 2 Page 5-8 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Figure 5.9 Limiting Break Broken Loop Jet Pump Suction Flow Rate Figure 5.10 Limiting Break Broken Loop Jet Pump Exit Flow Rate ANP-3106NP Revision 2 Page 5-9 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation 0 g 0 g 6 w ...J ;: 0 .. ...J u.. (/) 0 ..; [fl ...J 0 0 "' 0 0 0 g -o oo ...J -IL (/) ...J 0 0 0 "' 0 0 20 40 20 40 60 BO 100 120 140 160 180 200 TIME(SEC)

Figure 5.11 Limiting Break ADS Flow Rate 60 80 1 00 120 140 160 180 200 TIME(SEC)

Figure 5.12 Limiting Break LPCS Flow Rate ANP-3106NP Revision 2 Page 5-10 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation 53' ;::!. "l 0 QO ....I u. 13 n. ....I ....I IO 9 20 40 20 40 60 80 100 120 140 160 180 200 TIME(SEC)

Figure 5.13 Limiting Break Intact Loop LPCI Flow Rate 60 BO 100 120 140 160 180 200 TIME(SEC)

Figure 5.14 Limiting Break Broken Loop LPCI Flow Rate ANP-3106NP Revision 2 Page 5-11 Controlled Document AREVA Inc. Brunswick Units 1' and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation 0 ci "' f _j iO w-> w --' x a: (J a: w "! ;l: IO 9 0 ci 20 20 40 60 80 100 120 140 160 180 200 TIME(SEC)

Figure 5.15 Limiting Break Upper Downcomer Mixture Level 40 60 80 100 120 140 160 180 200 TIME(SEC)

Figure 5.16 Limiting Break Lower Downcomer Mixture Level ANP-3106NP Revision 2 Page 5-12 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation 20 40 GO 80 100 120 TIME(SEC) 140 Figure 5.17 Limiting Break 100 180 200 Intact Loop Discharge Line Liquid Mass 0 d 20 40 oO so 100 120 140 TIME(SEC)

Figure 5.18 Limiting Break Upper Plenum Liquid Mass 1 GO 180 200 ANP-3106NP Revision 2 Page 5-13 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation ci g ., 0 g d. cti 0 en ci D ,._ 50 Q ci ...J 0 mo a: g cc g UJ "' 3: Oo ...J ci g ill 0 g 20 40 60 so 100 120 140 150 TIME(SEC)

Figure 5.19 Limiting Break Lower Plenum Liquid Mass 180 200 0 20 40 60 80 100 120 140 160 TIME(SEC)

Figure 5.20 Limiting Break Hot Channel Inlet Flow Rate 180 200 ANP-3106NP Revision 2 Page 5-14 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation G' WC Ul . 00 ;;::!. u. c tu -' I-=> 0 ujq z <I: :i: () 1-0 o :i: ci 20 40 GO 80 1 00 120 140 160 180 200 TIME(SEC)

Figure 5.21 Limiting Break Hot Channel Outlet Flow Rate Figure 5.22 Limiting Break Hot Channel Coolant Temperature at the Hot Node at EOB ANP-3106NP Revision 2 Page 5-15 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Figure 5.23 Limiting Break Hot Channel Quality at the Hot Node at EOB Figure 5.24 Limiting Break Hot Channel Heat Transfer Coeff. at the Hot Node at EOB ANP-3106NP Revision 2 Page 5-16 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation Figure 5.25 Limiting Break Hot Channel Reflood Junction Liquid Mass Flow Rate 2000 E 1soo 3 E 1000 500 0 -PCT Rod -----Wate r C ha n ne l --Fue l Chan n e l 100 200 300 T i me (s ec) 400 500 Figure 5.26 Limiting Break Cladding Temperatures ANP-3106NP Revis i on 2 Page 5-17 Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation

6.0 Conclusions

ANP-3106NP Revision 2 Page 6-1 The EXEM BWR-2000 Evaluation Model was applied to confirm the acceptability of the ATRIUM 10XM MAPLHGR limit for Brunswick Units 1 and 2. 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 10XM MAPLHGR limitgiven in Figure 2.1. Peak PCT < 2200°F. Local cladding oxidation thickness

< 17%. Total hydrogen generation

< 1 %. 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 (Reference 1 ).

The MAPLHGR limit is applicable for ATRIUM 1 OXM full cores as well as transition cores containing ATRIUM 10XM fuel.

Controlled Document AREVA Inc. Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation ANP-3106NP Revision 2 Page 7-1 7.0 References

1. ANP-3105P Revision 1, Brunswick Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel forMELLLA+

Operation, AREVA, July 2015. 2. EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 EGGS Evaluation Model, Framatome ANP, May 2001. 3. Letter, P. Salas (AREVA) to Document Control Desk (USNRC), "Proprietary Viewgraphs and Meeting Summary for Closed Meeting on Application of the EXEM BWR-2000 ECCS Evaluation Methodology," NRC:11 :096, September 22, 2011. 4. Letter, T.J. McGinty (USNRC) 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. 5. XN-NF-81-58(P)(A)

Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984. 6. 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. 7. XN-NF-82-07(P)(A)

Revision 1, Exxon Nuclear Company EGGS Cladding Swelling and Rupture Model, Exxon Nuclear Company, November 1982. 8. EMF-2292(P)(A)

Revision 0, ATRIUMŽ-10:

Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000. 9. ANP-3108P Revision 1, Applicability of AREVA BWR Methods to Brunswick Extended Power Flow Operating Domain, AREVA, July 2015.

BSEP 16-0056 Enclosure 20 AREVA NP Affidavit Regarding Withholding ANP-3106P, Revision 2, Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for A TR/UM 1 OXM Fuel for MELLLA + Operation, December 2015 AFFIDAVIT STATE OF WASHINGTON ) ) SS. COUNTY OF BENTON ) 1. My name is Alan B. Meginnis.

I am Manager, Product Licensing, for AREVA Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary.

I am familiar with the policies established by AREVA to ensure the proper application of these criteria.

3. I am familiar with the AREVA information contained in the report ANP-3106P, Revision 2, "Brunswick units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation," dated December 2015 and referred to herein as "Document." Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

The request for withholding.of proprietary information is made in accordance with 1 O CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information." 6. The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA's research and development plans and programs or their results. (b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service. (c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA (d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability. (e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above. 7. In accordance with AREVA's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA policy requires that proprietary information be kept in a secured fife or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief. SUBSCRIBED before me this ----dayof Susan K. McCoy NOTARY PUBLIC, STATE OF WA INGTON MY COMMISSION EXPIRES: 1/14/2016

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