BSEP 10-0112, ANP-2943(NP), Revision 0, Brunswick Units 1 and 2, LOCA-ECCS Analysis MAPLHGR Limit for Atrium 10XM Fuel
| ML102780681 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/29/2010 |
| From: | AREVA, AREVA NP |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| BSEP 10-0112, TSC-2010-01, TSC-2010-02 ANP-2943(NP), Rev 0 | |
| Download: ML102780681 (58) | |
Text
BSEP 10-0112 Enclosure 6 AREVA Report ANP-2943(NP), Revision 0 Brunswick Units 1 and 2 LOCA-ECCS Analysis M
MAPLHGR Limitfor ATRIUMTF JOXM Fuel dated September 2010
Controlled Document ANP-2943(NP)
Revision 0 Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM TM 1OXM Fuel September 2010 AREVA NP Inc. AR E VA
ControIled Document AREVA NP Inc.
ANP-2943(NP)
Revision 0 Brunswick Units I and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM TM 1OXM Fuel
Controlled Document AREVA NP Inc.
ANP-2943(NP)
Revision 0 Copyright © 2010 AREVA NP Inc.
All Rights Reserved
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page i Nature of Changes Item Page Description and Justification
- 1. All This is the initial issue.
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Controled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 1OXM Fuel M
T Page ii Contents 1.0 Introduction .................................................................................................................... 1-1 2.0 Sum m ary ...................................................................................................................... 2-1 3.0 LOCA Description .......................................................................................................... 3-1 3.1 Accident Description ........................................................................................... 3-1 3.2 Acceptance Criteria ............................................................................................ 3-2 4.0 LOCA Analysis Description ............................................................................................ 4-1 4.1 Blowdown Analysis ............................................................................................. 4-1 4.2 Refill/Reflood Analysis ........................................................................................ 4-2 4.3 Heatup Analysis ................................................................................................. 4-2 4.4 Plant Param eters ................................................................................................ 4-3 4.5 ECCS Param eters .............................................................................................. 4-3 5.0 MAPLHG R Analysis Desuript-6rI and Results ................................................................ 5-1 6.0 Conclusions .................................................................................................................... 6-1 7.0 References ..................................................................................................................... 7-1 Tables 2.1 LOCA Results for Lim iting Conditions ............................................................................ 2-2 4.1 Initial Conditions ............................................................................................................. 4-5 4.2 Reactor System Param eters .......................................................................................... 4-6 4.3 ATRIUM 1OXM Fuel Assem bly Param eters ................................................................... 4-7 4.4 High-Pressure Coolant Injection Param eters ................................................................. 4-8 4.5 Low-Pressure Coolant Injection Param eters .................................................................. 4-9 4.6 Low-Pressure Core Spray Param eters ........................................................................ 4-10 4.7 Autom atic Depressurization System Param eters ......................................................... 4-11 4.8 Available ECCS for Recirculation Line Break LOCAs .................................................. 4-12 5.1 Event Times for Limiting Break 0.8 DEG Pump Discharge SF-LPCI Top-Peaked Axial ........................................................................................................... 5-2 5.2 ATRIUM 1OXM MAPLHG R Analysis Results ................................................................. 5-3 AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM MT 1OXM Fuel Page iii Figures 2.1 MAPLHGR Limit for ATRIUM 1OXM Fuel ...................................................................... 2-3 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model ................................... 4-13 4.2 RELAX System Model .......................................... 4-14 4.3 RELAX Hot Channel Model Top-Peaked Axial ............................................................ 4-15 4 .4 EC C S S che matic ......................................................................................................... 4-16 4.5 Rod Average Power Distribution in RELAX Calculation ............................................... 4-17 5.1 Limiting Break Upper Plenum Pressure ......................................................................... 5-4 5.2 Limiting Break Total Break Flow Rate ............................................................................ 5-4 5.3 Limiting Break Core Inlet Flow Rate ............................................................................... 5-5 5.4 Limiting Break Core Outlet Flow Rate ............................................................................ 5-5 5.5 Limiting Break Intact Loop Jet Pump Drive Flow Rate ................................................... 5-6 5.6 Limiting Break Intact Loop Jet Pump Suction Flow Rate ............................................... 5-6 5.7 Limiting Break Intact Loop Jet Pump Exit Flow Rate ..................................................... 5-7 5.8 Limiting Break Broken Loop Jet Pump DAiiV Plow6W ' - .......................................... 5-7 5.9 Limiting Break Broken Loop Jet Pump Suction Flow Rate ............................................. 5-8 5.10 Limiting Break Broken Loop Jet Pump Exit Flow Rate ................................................... 5-8 5.11 Limiting Break ADS Flow Rate ....................................................................................... 5-9 5.12 Limiting Break LPCS Flow Rate ..................................................................................... 5-9 5.13 Limiting Break Intact Loop LPCI Flow Rate ................................................................. 5-10 5.14 Limiting Break Broken Loop LPCI Flow Rate ............................................................... 5-10 5.15 Limiting Break Upper Downcomer Mixture Level ......................................................... 5-11 5.16 Limiting Break Lower Downcomer Mixture Level ......................................................... 5-11 5.17 Limiting Break Intact Loop Discharge Line Liquid Mass .............................................. 5-12 5.18 Limiting Break Upper Plenum Liquid Mass .................................................................. 5-12 5.19 Limiting Break Lower Plenum Liquid Mý)szt ................................................................... 5-13 5.20 Limiting Break Hot Channel Inlet Flow Rate ................................................................ 5-13 5.21 Limiting Break Hot Channel Outlet Flow Rate .............................................................. 5-14 5.22 Limiting Break Hot Channel Coolant Temperature at the Hot Node at EOB ................ 5-14 5.23 Limiting Break Hot Channel Quality at the Hot Node at EOB ....................................... 5-15 5.24 Limiting Break Hot Channel Heat Transfer Coeff. at the Hot Node at EOB ................. 5-15 5.25 Limiting Break Hot Channel Reflood Junction Liquid Mass Flow Rate ........... 5-16 5.26 Limiting Break Cladding Temperatures ........................................................................ 5-16 AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 1OXM Fuel M
T Page iv Nomenclature ADS automatic depressurization system ADSVOOS ADS valve out of service ANS American Nuclear Society BWR boiling-water reactor CFR Code of Federal Regulations CMWR core average metal-water reaction DEG double-ended guillotine DG diesel generator ECCS emergency core cooling system EOB end of blowdown HPCI high-pressure coolant injection LHGR linear heat generation rate LOCA loss-of-coolant accident LPCI low-pressure coolant injection LPCS low-pressure core spray MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio MWR metal-water reaction NRC Nuclear Regulatory Commission, U.S.
PCT peak cladding temperature RDIV recirculation discharge isolation valve SF-BATT single failure of battery (DC) power SF-HPCI single failure of the HPCI system SF-LPCI single failure of an LPCI valve SLO single-loop operation AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 1OXM Fuel M
T Page 1-1 1.0 Introduction 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 ATRIUM TM 1OXM* fuel as a function of exposure for normal (two-loop) operation. As shown in Reference 1, the MAPLHGR limit for single-loop operation (SLO) is equal to 0.80 times the two-loop limit.
The analyses documented in this report were performed with LOCA Evaluation Models developed by AREVA NPt 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 descriptirduoftthe 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 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 Double-ended guillotine
/ 0.8 discharge coefficient Single failure Low-pressure coolant injection valve Axial power shape Top-peaked ATRIUM is a trademark of AREVA NP.
t AREVA NP Inc. is an AREVA and Siemens company AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 2-1 2.0 Summary The MAPLHGR limit was determined by applying the EXEM BWR-2000 Evaluation Model for the analysis of the limiting LOCA event. The exposure-dependent MAPLHGR limit for ATRIUM 1OXM fuel is shown in Figure 2.1. The 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.
Local power distributions for all the Brunswick Unit 2 Cycle 20 ATRIUM 1OXM neutronic designs were 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 1OXM fuel are presented in Table 2.1.
The SLO analyses (Reference 1) support operation with an ATRIUM 1OXM MAPLHGR multiplier of 0.80 applied to the normal two-loop operation MAPLHGR limit.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 10XM Fuel M
T Page 2-2 Table 2.1 LOCA Results for Limiting Conditions Parameter ATRIUM 1OXM Exposure (GWd/MTU) 0.0 Peak cladding ternprature (OF)',-:. 1871 Local cladding oxidation (max %) 0.99 Total hydrogen generated
(% of total hydrogen possible) < 0.46 AREVA NP Inc.
Con-'oled Docu--e t Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 2-3 16.0 14.0
-" 12.0
-- 10.0
_J
-j
< 8.0 6.0 4.0
.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 Planar Average Exposure (GWd/MTU)
Average Planar ATRIUM 1OXM Exposure MAPLHGR (GWd/MTU) (kW/ft) 0 13.1 15 13.1 67 7.7 Figure 2.1 MAPLHGR Limit for ATRIUM IOXM Fuel AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM M T
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 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 Brunswick Units 1 and 2 ATRIUM 1OXM 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. Low-pressure core spray (LPCS) also provides some heat removal. The-end of the blowdown (EOB) phase is defined to occur when the system reaches the pressure corresponding to rated LPCS flow.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 3-2 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 (i0 CPFk'50 Akppien K). The conformance 6f 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 22000 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 dlodding surrounding the plenum volume, were to react.
- Calculated changes in core geometry shall be such that the core remains amenable to cooling.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 1OXM Fuel M
T Page 3-3 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.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM M T
10XM Fuel Page 4-1 4.0 LOCA Analysis Description The Evaluation Model used for the break spectrum analysis is the EXEM BWR-2000 LOCA analysis methodology described in Reference 2. The EXEM BWR-2000 methodology employs three major computer codes to evaluate the system and fuel response during all phases of a LOCA. These are the RELAX, HUXY, and RODEX2 computer codes. RELAX is used to calculate the system and hot channel response during the blowdown, refill, and reflood phases of the LOCA. The HUXY code is used to perform heatup calculations for the entire LOCA, and calculates the PCT and local clad oxidation at the axial plane of interest. RODEX2 is used to determine fuel parameters (such as stored energy) for input to the other LOCA codes. The code interfaces for the LOCA methodology are illustrated in Figure 4.1.
A complete analysis for a given break size starts with the specification of fuel parameters using RODEX2 (Reference 3). RODEX2 is used to determine the initial stored energy foi 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 Blowdown Analysis The RELAX code (Reference 2) is used to calculate the system thermal-hydraulic response during the blowdown phase of the LOCA. For the system blowdvown.analysis, the r-ore 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 CFR 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 ania!ysis.. .
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 fuel, cladding, and coolant temperatures during the AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 10XM Fuel Page 4-2 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 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 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 contirfuebeyo*rthý-Iend 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. [
The RELAX calculations provide HUXY with the time of hot node reflood and the time when the liquid has risen in the bypass to the height of the axial plane of interest (time of bypass reflood).
4.3 Heatup Analysis The HUXY code (Reference 4) is used to perform heatup calculations for the entire LOCA transient and provides PCT and local clad oxidation at the axial plane of interest. The heat generated by metal-water reaction (MWR) is included in the HUXY analysis. HUXY is used to
.cafcULiaf6e the thermal response of each fuel rod in one axial plane of the hot channel assembly.
These calculations consider thermal-mechanical interactions within the fuel rod. The clad swelling and rupture models from NUREG-0630 have been incorporated into HUXY AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 4-3 (Reference 5). The HUXY code complies with the 10 CFR 50 Appendix K criteria for LOCA Evaluation Models.
HUXY uses the 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 analysis. The principal results of a HUXY heatup analysiS are-TirdPCGTand 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 r?-cton-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-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 assurned in the arwlysis.- 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.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 1OXM Fuel M
T Page 4-4 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. Also, no credit for ECCS flow is assumed until 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. " ' " .
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 forATRIUM TM 1OXM Fuel Page 4-5 Table 4.1 Initial Conditions Parameter Value Reactor power (% of rated) 102 Reactor power (MWt) 2981.5 C S]
]
Steam flow rate (Mlb/hr) 13.1 Steam dome pressure (psia) 1048.9 Core inlet enthalpy (Btu/Ib) 527.7 ATRIUM 1OXM hot assembly MAPLHGR (kW/ft) 13.1 Rod Average power distribution Figure 4.5
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 4-6 Table 4.2 Reactor System Parameters Parameter Value Vessel ID (in) 220.5 Number of fuel assemblies 560 Recirculation suction pipe area (ft2) 3.67 1.0 DEG suction break area (ft2) 7.33 Recirculation discharge pipe area (ft) 3.67 1.0 DEG discharge break area (ft) 7.33 AREVA NP Inc.
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T Page 4-7 Table 4.3 ATRIUM 1OXM Fuel Assembly Parameters Parameter Value Fuel rod array 10x10 Number of fuel rods per 79 (full-length rods) assembly 12 (part-length rods)
Non-fuel rod type Water channel replaces 9 fuel rods Fuel rod OD (in) 0.4047 Active fuel length (in) 150.0 (full-length rods)
(including blankets) 75.0 (part-length rods)
Water channel outside width (in) 1.378 Fuel channel thickness (in) 0.075 (minimum wall) 0.100 (corner)
Fuel channel internsi Wijth (in) 5.278 AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 1OXM Fuel M
T Page 4-8 Table 4.4 High-Pressure Coolant Injection Parameters Parameter Value Coolant temperature (maximum) (OF) 140 Initiating Signals and Setpoints Water level (in)* . 459 High drywell pressure (psig) Not used Time Delays Time for HPCI pump to reach rated speed and injection valve wide open (sec) 60 Delivered Coolant Flow Rate Versus Pressure Vessel to Torus AP - Flow P~t,--- -.
(psid) (gpm) 0 0 150 3,825 1164 3,825
- Relative to vessel zero.
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T Page 4-9 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)* 358 High drywell pressure (psig) Not used Time Delays Time for LPCI pumps to reach rated speed (maximum) (sec) 31.8 LPCI injection valve stroke time (sec) 37.5 Delivered Coolant Flow Rate Versus Pressure Flow rate for Flow rate for 1 pump 2 pumps injecting into injecting into Vessel to 1 recirculation 1 recirculation Torus AP loop loop (psid) (gpm) (gpm) 0 8,690 14,420 20 7,000 12,000 202 0 0
- Relative to vessel zero.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 4-10 Table 4.6 Low-Pressure Core Spray Parameters Parameter Value Reactor pressure permissive for opening valves - analytical (psia) 410 Coolant temperature (maximum) (OF) 160 Initiating Signals and Setpoints Water level (in)* 358 High drywell pressure (psig) Not used Time Delays Time for LPCS pumps to reach rated speed (maximum) (sec) 39.7 LPCS injection valve stroke time (sec) 14.0 Delivered Coolant Flow Rate Versus Pressure Vessel to Flow rate Torus AP for 1 pump (psid) (gpm) 0 5,250 113 4,000 265 0
- Relative to vessel zero.
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 10XM Fuel M
T Page 4-11 Table 4.7 Automatic Depressurization System Parameters Parameter Value Number of valves installed 7 Number of valves available* 5 Minimum flow capacity of avai!able 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 tirnei (deiay time froi*i initiating signal to time valves are open (sec) 121
- 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 Relative to vessel zero.
,[
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LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 4-12 Table 4.8 Available ECCS for Recirculation Line Break LOCAs Recirculation Recirculation Assumed Suction Break Discharge Break Failure
- Systems Systems Remaining t, t' § Remaining * § DC power (i) 1LPCS + 3LPCI + ADS 1LPCS + 1LPCI + ADS (SF-BATT) . __..... _......
DC power (j) 2 LPCS + 2LPCI + HPCI + ADS 2LPCS + HPCI + ADS Diesel generator (i) 1LPCS + 3LPCI + HPCI + ADS 1LPCS + 1LPCI + HPCI + ADS Diesel generator (j) 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 (j) or diesel generator (j) will result in the loss of one diesel generator (DG-3 or DG-4).
t Systems remaining, as identified in this table for recirculation suction !ine 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.
t 1LPCI (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.
AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 4-13
- The hot assembly calculation may be combined with the system calculation or executed Peak Cladding Temperature, separately Metal Water Reaction Figure 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model AREVA NP Inc.
Conti oed Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 4-14
[
Figure 4.2 RELAX System Model AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 10XM Fuel Page 4-15
[
Figure 4.3 RELAX Hot Channel Model Top-Peaked Axial AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 10XM Fuel M
T Page 4-16 (i) (i)
D6-2 0)
DG-3 DG-1 DG-4 Loop-A Loop-B Figure 4.4 ECCS Schematic AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 4-17 Figure 4.5 Rod Average Power Distribution in RELAX Calculation AREVA NP Inc.
Controlled kDcument Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 5-1 5.0 MAPLHGR Analysis Description and Results 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 5.25. 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 1OXM fuel.
Table 5.2 shows the MAPLHGR analysis results for the ATRIUM 1OXM fuel. The HUXY model of the ATRIUM 1OXM fuel is applied to obtain these results as described in Section 4.3. The HUXY analysis is performed at 5 GWd/MTU exposure intervals for assembly average planar exposures bEiween 0 and 65 GWd/MTU and an ending exposure of 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 1871°F at 0.0 GWd/MTU exposure. The maximum local MWR of 0.99% occurred at 0.0 GWd/MTU exposure. Analysis results show that the planar average MWR at the peak power plane is 0.46%. Since all other planes in the core are at lower power, the CMWR will be significantly less than 0.46%.
Figure 5.26 shows the cladding temperature of the ATRIUM 1OXM PCT rod as a function of time for the limiting break. The maximum temperature of 1871OF occurs at 171.9 seconds. These results demonstrate the acceptability of the ATRIUM 1OXM MAPLHGR limit shown in Figure 2.1.
AREVA NP Inc.
Controlled Documient Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 5-2 Table 5.1 Event Times for Limiting Break 0.8 DEG 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, L1 (358 in) 8.2 Jet pump uncovers 9.4 Reci'cu;ation. su.ioruncovers 16.4 Lower plenum flashes 16.6 Diesel generators started 15.0 LPCS high-pressure cutoff 59.5 Power at LPCS injection valves 27.8 LPCS valve pressure permissive 47.8 LPCS valve starts to open 48.8 LPCS valve open 62.8 LPCS pump at rated speed 39.7 LPCS flow starts 62.9 LPCS permissive for ADS 39.7 RDIV pressure permissive 56.3 RDIV starts to close 57.3 RDIV closed 94.3 Rated LPCS flow 87.3 Blowdown ends 87.3 ADS valves open 129.2 bypass reflood 175.1 Core reflood 171.9 PCT 171.9 AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 10XM Fuel M
T Page 5-3 Table 5.2 ATRIUM 1OXM MAPLHGR Analysis Results Average Local Planar Cladding Exposure MAPLHGR PCT Oxidation (GWd/MTU) (kW/ft) (OF) (%)
0.0 13.1 1871 0.99 5.0 13.1 1832 0.82 10.0 13.1 1777 0.65 15.0 13.1 1758 0.59 20.0 12.58 1710 0.49 25.0 12.06 1676 0.41 30.0 11.54 1646 0.35 35.0 11.02 1644 0.35 40.0 10.50 1577 0.25 45.0 9.98 1543 0.20 50.0 9.47 1499 0.16 55.0 8.95 1451 0.12 60.0 8.43 1400 0.09 65.0 7.91 1343 0.06 67.0 7.7 1314 0.05 CMWR is <0.46% at all exposures.*
- The planar average MWR for the peak power plane is 0.46% which supports the conclusion that the CMWR is less than 0.46%.
AREVA NP Inc.
Cont.-,-e Do, .met Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM M 1OXM Fuel T
Page 5-4 Figure 5.1 Limiting Break Upper Plenum Pressure Figure 5.2 Limiting Break Total Break Flow Rate AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 5-5 Figure 5.3 Limiting Break Core Inlet Flow Rate JI Figure 5.4 Limiting Break Core Outlet Flow Rate AREVA NP Inc.
Con~trole Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 forATRIUM TM 1OXM Fuel Page 5-6
[
Figure 5.5 Limiting Break Intact Loop Jet Pump Drive Flow Rate Figure 5.6 Limiting Break Intact Loop Jet Pump Suction Flow Rate AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 1OXM Fuel M
T Page 5-7 I
Figure 5.7 Limiting Break Intact Loop Jet Pump Exit Flow Rate I
Figure 5.8 Limiting Break Broken Loop Jet Pump Drive Flow Rate AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 1OXM Fuel M
T Page 5-8 I
1.
Figure 5.9 Limiting Break Broken Loop Jet Pump Suction Flow Rate I
Figure 5.10 Limiting Break Broken Loop Jet Pump Exit Flow Rate AREVA NP Inc.
Controled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM M 1OXM Fuel T
Page 5-9
-D 40 80 120 160 200 240 280 320 TIME (SEC)
Figure 5.11 Limiting Break ADS Flow Rate BRIUNSWICK UNII 2 0.8 DEG/PD JOP AT-1OXM SF S I i i i C')
0 0~
00 40 80 120 160 200 240 250 320 TIME (SEC)
Figure 5.12 Limiting Break LPCS Flow Rate AREVA NP Inc.
Con rolied Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 10XM Fuel M
T Page 5-10 BRUNSWICK UNIT 2 0.8 DEG/PD TOP AT-10XM SF-LPCI 102P/105F ia)
I I ~ j 0
-j U_
-j 0 40 so 120 160 200 240 280 320 TIME (SEC)
Figure 5.13 Limiting Break Intact Loop LPCI Flow Rate BRUNSWICK UNIT 2 0.8 DEG/PD TOP AT-10XM SF-LPCI 102P/105F If) N 0~
LI OF)7
_j
_j 00 0o 40 80 120 160 200 240 280 320 TIME (SEC)
Figure 5.14 Limiting Break Broken Loop LPCI Flow Rate AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM M 10XM Fuel T
Page 5-11 BRUNSWICK UNIT 2 0.8 DEG/PD TOF AT-10XM SF-LPCI 102P/lo5F
-J
--- T 1-
_J X
x 0
z 0
0 cr 0 4 80 120 160 200 240 280 320 TIME (SEC)
Figure 5.15 Limiting Break Upper Downcomer Mixture Level (0
U-0 0J 0
0 C-)
00 40 80 12u "SO
- 200 240 280 320 TIME (SEC)
Figure 5.16 Limiting Break Lower Downcomer Mixture Level AREVA NP Inc.
Controlled Docum nt Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 10XM Fuel M
T Page 5-12 BRUNSWICK UNIT 2 0.8 DEC/PD TOP AT-1OXM SF-LPCI 1 02
<0
-i 00 40 80 120 160 200 240 28' 20 320 TIME (SEC)
Figure 5.17 Limiting Break Intact Loop Discharge Line Liquid Mass V;
-o U)-
of a-00 40 80 120 160
- 200 240 - 280 - 320 TIME (SEC)
Figure 5.18 Limiting Break Upper Plenum Liquid Mass AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 5-13 If) 0 Z o Lj Oo LJ U EDl 40 80 120 160 200 240 280 322 TIME (SEC)
Figure 5.19 Limiting Break Lower Plenum Liquid Mass BRUNSWICK UNIT 2 0.8 DEG/PD TOP AT-10XM SF-LPCI 102P/105F g
0
-j z
L0 00 Oo I I I I I I I 10 40 60 120 160 200 240 280 320 TIME (SEC)
Figure 5.20 Limiting Break Hot Channel Inlet Flow Rate AREVA NP Inc.
ontrol.ed Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRI UMTM 10XM Fuel Page 5-14 BRUNSWICK UNIT 2 0.8 DEG/PD TOP AT-1OXM SF-LPCI 102P/105F 00 C-)
1I1 m
I. 0 C-O 0
Ed zz 0
I0 40 80 120 160 200 240 280 320 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 AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 5-15 I
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 AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 1OXM Fuel M
T Page 5-16 I
Figure 5.25 Limiting Break Hot Channel Reflood Junction Liquid Mass Flow Rate BRK2 CY20 0 GWd/MTU (XMLCB-4348L-13C65 lattice) 2500 2000 1500 Q) a E
0) 1000 500 0 25 50 75 100 125 150 175 200 Time (sec)
Pf*
Run, OB/22/203O14t&3 5Q42127 Figure 5.26 Limiting Break Cladding Temperatures AREVA NP Inc.
Con~trolled Docu~ment Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM 1OXM Fuel M
T Page 6-1 6.0 Conclusions 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 1OXM MAPLHGR limit given 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 1OXM full cores as well as transition cores containing ATRIUM 10XM fuel.
AREVA NP Inc.
Controlled Document Brunswick Units 1 and 2 ANP-2943(NP)
LOCA-ECCS Analysis MAPLHGR Limit Revision 0 for ATRIUM TM 1OXM Fuel Page 7-1 7.0 References
- 1. ANP-2941 (P) Revision 0, Brunswick Units I and 2 LOCA Break Spectrum Analysis for ATRIUM TM IOXM Fuel, AREVA NP, September 2010.
- 3. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-MechanicalResponse Evaluation Model, Exxon Nuclear Company, March 1984.
- 4. XN-CC-33(A) Revision 1, HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option Users Manual, Exxon Nuclear Company, November 1975.
- 5. XN-NF-82-07(P)(A) Revision 1, Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, Exxon Nuclear Company, November 1982.
- 6. EMF-2292(P)(A) Revision 0, A TRIUM'm-10: Appendix K Spray Heat Transfer Coefficients, Siemens-Power Corporation, September 2000.
AREVA NP Inc.
BSEP 10-0112 Enclosure 3 AREVA Affidavit Regarding Withholding ANP-2941 (P), Revision 0, from Public Disclosure
AFFIDAVIT STATE OF WASHINGTON )
) ss.
COUNTY OF BENTON )
- 1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.
- 2. I am familiar with the criteria applied by AREVA NP to determine whether
,cei'1airr;AF'VAt.-NP information is proprietary. I am familiar with the policies established- by. S ~ -.
AREVA NP to ensure the proper application of these criteria.
- 3. I am familiar with the AREVA NP information contained in the report ANP-2941 (P), Revision 0, entitled "Brunswick Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 1OXM Fuel," dated September 2010 and referred to herein as "Document."
Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP 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 NP 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 10 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 NP to determine whether information should be classified as proprietary:
(a) The information reveals details of AREVA NP'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.
(6) 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 NP.
(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.
(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to coin*6etit6rs to AREVA NP, and would likely cause-substantial harm to the competitive position of AREVA NP.
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 NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basisi to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.
- 8. AREVA NP policy requires that proprietary information be kept in a secured file 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.
'~1 SUBSCRIBED before me this o -
day of'5 , 2010.
Susan K. McCoy ý_
NOTARY PUBLIC, STATE OF WASHINGTON MY COMMISSION EXPIRES: 1/10/12
BSEP 10-0112 Enclosure 4 AREVA Affidavit Regarding Withholding ANP-2943(P), Revision 0, from Public Disclosure
AFFIDAVIT STATE OF WASHINGTON )
) ss.
COUNTY OF BENTON )
- 1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.
- 2. I am familiar with the criteria applied by AREVA NP to determine whether certain -ARE.VA N1P information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.
- 3. I am familiar with the AREVA NP information contained in the report ANP-2943(P), Revision 0, entitled "Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM I OXM Fuel," dated September 2010 and referred to herein as "Document."
Information contained in this Document has been classified by AREVA NP as proprietary in
.crcordance vith the policies established by AREVA NP 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 NP 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 10 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 NP to determine whether information should be classified as proprietary:
(a) The information reveals details of AREVA NP'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 NP.
(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.
(e) The information is vital to a competitive advantage held by AREVA NP, would be I.eIpfi1to comp~tiýfors to ARCEVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.
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 NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to otherm outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.
- 8. AREVA NP policy requires that proprietary information be kept in a secured file 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 *- 0 day of -C, 2010.
Susan K. McCoy NOTARY PUBLIC, STATE OF WASHIIP MY COMMISSION EXPIRES: 1/10/12