ML072950369

Areva Report, ANP-2674(NP), Revision 0, Brunswick, Unit 1, Cycle 17 Reload Safety Analysis, September 2007

ML072950369
Person / Time
Site:	Brunswick
Issue date:	09/30/2007
From:	AREVA NP, Siemens
To:	Office of Nuclear Reactor Regulation
References
BSEP 07-0108, TAC MD4063, TAC MD4064, TSC-2006-06 ANP-2674(NP), Rev 0
Download: ML072950369 (147)

Text

BSEP 07-0108 Enclosure 3 AREVA Report ANP-2674(NP), Revision 0, Brunswick Unit 1 Cycle 17 Reload Safety Analysis, dated September 2007 An AREA. an Simn compan ANP-2674(NP)

Revision 0 Brunswick Unit 1 Cycle 17 Reload Safety Analysis September 2007 A.AR EVA AREVA NP Inc.ANP-2674(NP)

Revision 0 Brunswick Unit 1 Cycle 17 Reload Safety Analysis AREVA NP Inc.AN P-2674(N P)Revision 0 Copyright

@ 2007 ARE VA NP Inc.All Rights Reserved paj Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page Nature of Changes Item Page Description and Justification

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

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page ii Contents 1.0 Introduction

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

1-1 2.0 Disposition of Events..........................................................................

2-1 3.0 Mechanical Design Analysis..................................................................

3-1 4.0 Thermal-Hydraulic Design Analysis .........................................................

4-1 4.1 Thermal-Hydraulic Design and Compatibility.......................................

4-1 4.2 Safety Limit MCPR Analysis .........................................................

4-1 4.3 Core Hydrodynamic Stability.........................................................

4-2 5.0 Anticipated Operational Occurrences

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

5-1 5.1 System Transients....................................................................

5-1 5.1.1 Load Rejection No Bypass (LRNB) .......................................

5-3 5.1.2 Turbine Trip No Bypass (TTNB) ...........................................

5-4 5.1.3 Feedwater Controller Failure (FWCF) ....................................

5-4 5.1.4 Pressure Regulator Failure Downscale (PREDS) .......................

5-5 5.1.5 Loss of Feedwater Heating................................................

5-5 5.1.6 Control Rod Withdrawal Error .............................................

5-6 5.2 Slow Flow Runup Analysis...........................................................

5-6 5.3 Equipment Out-of-Service Scenarios

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

5-8 5.3.1 FHOOS ......................................................................

5-8 5.3.2 TBVOOS ....................................................................

5-8 5.3.3 Combined FHOOS and TBVOOS.........................................

5-9 5.3.4 One SRVOOS ..............................................................

5-9 5.3.5 One MSIVOOS..............................................................

5-9 5.3.6 Single-Loop Operation

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

5-9 5.4 Licensing Power Shape.............................................................

5-10 6.0 Postulated Accidents..........................................................................

6-1 6.1 Loss-of-Coolant-Accident (LOCA)...................................................

6-1 6.2 Control Rod Drop Accident (CRDA).................................................

6-1 6.3 Fuel and Equipment Handling Accident ............

6................................

6-1 6.4 Fuel Loading Error (Infrequent Event) ..............................................

6-2 6.4.1 Mislocated Fuel Bundle ....................................................

6-2 6.4.2 Misoriented Fuel Bundle ...................................................

6-2 7.0 Special Analyses...............................................................................

7-1 7.1 ASME Overpressurization Analysis .................................................

7-1 7.2 ATWS Event Evaluation..............................................................

7-2 7.2.1 ATWS Overpressurization Analysis .......................................

7-2 7.2.2 Long-Term Evaluation......................................................

7-2 7.3 Standby Liquid Control System......................................................

7-3 7.4 Fuel Criticality

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

7-4 8.0 Operating Limits and COLR Input ...........................................................

8-1 8.1 MCPR Limits..........................................................................

8-1 8.2 LHGR Limits...........................................................................

8-1 8.3 MAPLHGR Limits.....................................................................

8-2 9.0 References.....................................................................................

9-1 Appendix A Operating Limits and Results Comparisons......................................

A-i AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page iii Tables 1.1 EQOS Operating Conditions

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

1-2 2.1 Disposition of Events Summary for Brunswick Steam Electric Plant .....................

2-3 2.2 Disposition of Operating Flexibility and EQOS Options on Limiting Events..............

2-6 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses.................

4-4 4.2 Results Summary for Safety Limit MCPR Analyses........................................

4-5 4.3 OPRM Setpoints

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

4-6 4.4 BSP Endpoints for Brunswick Unit 1 Cycle 17..............................................

4-7 5.1 Exposure Basis for Brunswick Unit 1 Cycle 17 Transient Analysis......................

5-11 5.2 Scram Speed Insertion Times ..............................................................

5-12 5.3 NEOC Base Case LRNB Transient Results ...............................................

5-13 5.4 EOCLB Base Case LRNB Transient Results..............................................

5-14 5.5 NEOC Base Case TTNB Transient Results ...............................................

5-15 5.6 EOCLB3 Base Case TTNB Transient Results..............................................

5-16 5.7 NEOC Base Case FWCF Transient Results ..............................................

5-17 5.8 EOCLB3 Base Case FWCF Transient Results.............................................

5-18 5.9 Loss of Feedwater Heating Transient Analysis Results ..................................

5-19 5.10 Control Rod Withdrawal Error ACPR Results..............................................

5-20 5.11 RBM Operability Requirements.............................................................

5-21 5.12 Flow-Dependent MCPR Results ...........................................................

5-22 5.13 Licensing Basis Core Average Axial Power Profile .......................................

5-23 7.1 ASME Overpressurization Analysis Results ................................................

7-5 7.2 ATWS Overpressurization Analysis Results................................................

7-6 7.3 [ .................

7-7 8.1 MCPRP Limits for NSS Insertion Times BOC to < NEOC (16,500 MWd/MTU) ..........

8-4 8.2 MCPRp Limits for TSSS Insertion Times BOG to < NEOC (16,500 MWd/MTU).........

8-5 8.3 MCPRp Limits for NSS Insertion Times BOC to < EOCLB (18,909 MWd/MTU).........

8-6 8.4 MCPRp Limits for TSSS Insertion Times BOO to < EOCLB3 (18,909 MWd/MTU) ......8-7 8.5 MCPRp Limits for NSS Insertion Times FFTRlCoastdown

(: 18,909 MWd/MTU)......8-8

8.6 MCPRP

Limits for TSSS Insertion Times FFTRlCoastdown

(ý 18,909 MWd/MTU).....8-9 8.7 Flow-Dependent MCPR Limits ATRIUM-10 and GE14 Fuel .............................

8-10 8.8 ATRIUM-10 Steady-State LHGR Limits ...................................................

8-11 8.9 ATRIUM-10 LHGRFACP Multipliers for NSS Insertion Times BOO to < EOCLB3 (18,909 MWd/MTU)..........................................................................

8-12 8.10 ATRIUM-1 LHGRFACp Multipliers for TSSS Insertion Times BOC to < EOCLB (18,909 MWd/MTU)..........................................................................

8-13 AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page iv Tables (Continued) 8.11 ATRIUM-10 LHGRFACp Multipliers for NSS Insertion Times FFTRlCoastdown (t 18,909 MWd/MTU)........................................................................

8-14 8.12 ATRIUM-10 LHGRFACp Multipliers for TSSS Insertion Times FFTIR/Coastdown

( 18,909 MWd/MTU)........................................................................

8-15 8.13 ATRIUM-10 LHGRFACf Multipliers All Cycle 17 Exposures

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

8-16 8.14 ATRIUM-10 MAPLHGR Limits..............................................................

8-17 8.15 GE14 MAPFACp Multipliers for NSS and TSSS Insertion Times All Cycle 17 Exposures....................................................................................

8-18 8.16 GE14 MAPFACf Multipliers All Cycle 17 Exposures......................................

8-19 Figures 1.1 Brunswick Unit 1 Power/Flow Map...........................................................

1-3 5.1 EOCLB LRN B at 100OP/1 04.5F- TSSS Key Parameters

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

5-24 5.2 EOCLB LRNB at l oop/I 04.5F -TSSS Sensed Water Level............................

5-25 5.3 EOCLB LRNB at 1 GOP/i 04.5F -TSSS Vessel Pressures...............................

5-26 5.4 EOCLB FWCF at 1 GOP/i 04.5F -TSSS Key Parameters................................

5-27 5.5 EOCLB FWCF at 1lOOP/i 04.5F -TSSS Sensed Water Level ...........................

5-28 5.6 EOCLB FWCF at 1lOOP/I 04.5F -TSSS Vessel Pressures

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

5-29 7.1 MSIV Closure Overpressurization Event at 102P/1 04.5F -Key Parameters............

7-8 7.2 MSIV Closure Overpressurization Event at 102P/1 04.5F -Sensed Water Level ......7-9 7.3 MSIV Closure Overpressurization Event at 1 02P3/1 04.5F -Vessel Pressures.........

7-10 7.4 MSIV Closure Overpressurization Event at 102P/104.5F

-Safety/Relief Valve Flow Rates ...................................................................................

7-1 1 7.5 PRFO ATWS Overpressurization Event at 1lOOP/i 04.5F -Key Parameters

...........

7-12 7.6 PRFO ATWS Overpressurization Event at 1lOOP/i 04.5F -Sensed Water Level......7-13 7.7 PRFO ATWS Overpressurization Event at 1lOOP/i 04.5F -Vessel Pressures..........

7-14 7.8 PRFO ATWS Overpressurization Event at 1 GOP/i 04.5F -Safety/Relief Valve Flow Rates ...................................................................................

7-15 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page v Nomenclature AOO ARO APRM ASME AST ATWS ATWS-RPT BOO BPWS BSEP BSP BWROG CFR COLR CPR ORDA CRWE EFPD EFPH EQO EOCLB EQEP EOOS FFTR FHOOS FWCF GE GNF HOOM HFR HPCI IOF LFWH LHGR LHGRFACf LHGRFACp LOCA LPRM LRNB anticipated operational occurrence all control rods out average power range monitor American Society of Mechanical Engineers alternative source term anticipated transient without scram anticipated transient without scram recirculation pump trip beginning-of-cycle banked position withdrawal sequence Brunswick Steam Electric Plant backup stability protection Boiling Water Reactor Owners Group Code of Federal Regulations core operating limits report critical power ratio control rod drop accident control rod withdrawal error effective full-power days effective full-power hours end-of-cycle end-of-cycle licensing basis end of full power equipment out-of-service final feedwater temperature reduction feedwater heaters out-of-service feedwater controller failure General Electric Global Nuclear Fuels hot channel oscillation magnitude heat flux ratio high pressure coolant injection increased core flow loss of feedwater heating linear heat generation rate flow-dependent linear heat generation rate multipliers power-dependent linear heat generation rate multipliers loss-of-coolant accident local power range monitor generator load rejection with no bypass AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page vi Nomenclature (Continued)

MAPFACf MAPFACp MAPLHGR MCPR MCPRf MCPRP MELLLA MSIV MSIVOOS NEOC NSS NRC flow-dependent maximum average planar linear heat generation rate multipliers power-dependent maximum average planar linear heat generation rate multipliers maximum average planar linear heat generation rate minimum critical power ratio flow-dependent minimum critical power ratio power-dependent minimum critical power ratio maximum extended load line limit analysis main steam isolation valve main steam isolation valve out-of-service near end-of-cycle nominal scram speed Nuclear Regulatory Commission, U.S.OLMCPR operating limit minimum critical power ratio OPRM oscillation power range monitor Pbypass power below which direct scram on TSVITCV closure is bypassed PCT peak cladding temperature PLU power load unbalance PRFDS pressure regulator failure downscale PRFO pressure regulator failure open RBM (control) rod block monitor ROIC reactor core isolation cooling RHR residual heat removal RPT recirculation pump trip SAT startup auxiliary transformer SLC standby liquid control SLMCPR safety limit minimum critical power ratio SLO single-loop operation SRV safety/relief valve S RVOOS safety/relief valve out-of-service TBVOOS turbine bypass valves out-of-service TCV turbine control valve TIP traversing incore probe TLO two-loop operation TSSS technical specifications scram speed TSV turbine stop valve TTNB3 turbine trip with no bypass UAT unit auxiliary transformer UFSAR updated final safety analysis report ACPR change in critical power ratio AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 1-1 1.0 Introduction This report presents the results of the reload licensing analyses performed by ARE VA NP* in support of Brunswick Unit 1 Cycle 17. The analyses reported in this document were performed using methodologies previously approved for generic application to boiling water reactors.

The NRC technical limitations associated with the application of the approved methodologies have been satisfied by these analyses.The Cycle 17 core consists of a total of 560 fuel assemblies, including 248 fresh ATRI UM TM-10 I assemblies and 312 irradiated GE14 assemblies.

The licensing analysis supports the core design presented in Reference 1.The Cycle 17 reload licensing analysis consists of the calculation of the potentially limiting events and analyses that were identified in the disposition of events. The results of the analyses are used to establish the Technical Specifications/COLR limits and ensure that the design and licensing criteria are met. The design and safety analyses are based on the design and operational assumptions and plant parameters provided by the utility. The results of the reload licensing analysis support operation for the power/flow map presented in Figure 1.1 and also -support operation with the equipment out-of-service (EQOS) scenarios presented in Table 1.1.t AREVA NP Inc. is an AREVA and Siemens company.ATRIUM is a trademark of AREVA NP.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 1-2 Table 1.1 EQOS Operating Conditions

• Single-loop operation (SLO)Turbine bypass valves out-of-service (TBVOOS)Feedwater heaters out-of-service (EHOOS)One safety relief valve out-of-service (SRVOOS)One main steam isolation valve out-of-service (MSIVOOS)Up to 40% of the TIP channels out-of-service (100%available at startup)Up to 50% of the LPRMs out-of-service

• Each EOOS condition is supported in combination with 1 SRVOOS, up to 40% of the TIP channels out-of-service, and/or up to 50% of the LPRMs out-of-service.

AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 1-3 120.0 110.0 100.0 90.0 80.0 70.0 o1 60.0 50.0 40.0 30.0 20.0 10.0-0.0 R F i-F--________ I _______ ________ ______ ________ ______n Natural Circulation Line I"I 1-15% Minimum Pump MniuPoe I I II_______~~~

~~ I 4 --- -4 4 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr 0 10 20 30 40 50 60 70 80 90 100 110 120(%Core Flow Figure 1.1 Brunswick Unit 1 Power/Flow Map AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 2-1 2.0 Disposition of Events The objective of the disposition of events is to identify the limiting events which must be analyzed to support operation at the Brunswick Steam Electric Plant (BSEP) with the introduction of ATRIUM-10 fuel. Events and analyses identified as potentially limiting are either evaluated generically for the introduction of AREVA fuel or on a cycle-specific basis.The first step in the disposition of events is to identify the licensing basis of the plant. Included in the licensing basis are descriptions of the postulated events/analyses and the associated criteria.

Fuel-related system design criteria which must be met to ensure regulatory compliance and safe operation are also included.

The BSEP licensing basis is contained in the Updated Final Safety Analysis Report (UFSAR), the Technical Specifications, Core Operating Limits Reports (COLR), and other reload analysis reports.AREVA reviewed all the fuel-related design criteria, events and analyses identified in the licensing basis. In many cases, when the operating limits are established to ensure acceptable consequences of an anticipated operational occurrence or accident, the fuel-related aspects of the system design criteria are met. All the fuel-related events were reviewed and dispositioned into one of the following categories:

1 .No further analysis required.

This classification may result from one of the following:

a. The consequences of the event are bound by consequences of a different event.b. The consequences of the event are benign, i.e., the event causes no significant change in margins to the operating limits.C. The event is not affected by the introduction of a new fuel design and/or the current analysis of record remains applicable.

2. Address event each reload. The consequences of the event are potentially limiting and need to be addressed each reload.3. Address for initial reload. This classification may result from one of the following:

a. The analysis is performed using conservative bounding assumptions and inputs such that the initial reload results will remain applicable for future reloads of the same fuel design.b. Results from the first reload will be used to quantitatively demonstrate that the results remain applicable for future reloads of the same fuel design because the consequences are benign or bound by those of another event.The impact of operation in the EOOS scenarios presented in Table 1.1 was also considered in the disposition of events.AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 2-2 A summary of the disposition of events results is presented in Tables 2.1 and 2.2. Table 2.1 presents a list of the events and analyses, the corresponding UFSAR section, the disposition status and any applicable comments.

Table 2.2 presents a summary of the disposition of events for the EQOS scenarios.

Note that operation in the ICE and MELILLA regions of the power/flow map are included in the disposition results presented in Table 2.1.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page 2-3 Table 2.1 Disposition of Events Summary for Brunswick Steam Electric Plant UFSAR Section 4.4 4.4 4.A 5.2.2 6.2.5 Event /Analysis Thermal-Hydraulic Compatibility Safety Limit MCPR Stability

-OPRM setpoints ASME Overpressurization Combustible Gas Control in Containment New Fuel Storage Criticality Spent Fuel Storage Criticality Standby Liquid Control (SLC)System Disposition Status Address initial reload.Address each reload.Address each reload.Address each reload.No further analysis required.Address for initial reload.Address for initial reload.Address each reload.Comments Demonstrate design criteria are met.Required to establish operating limits.Required to establish OPRM setpoint.Potentially limiting event.Fuel design change has a negligible impact.Evaluate for new fuel storage vault.Confirm applicability each reload.Evaluate for spent fuel pool. Confirm applicability each reload.Analysis performed to verify adequate SLC system shutdown capacity.9.1.1 9.1.2 9.3.4 9.5.1 Appendix R Address initial reload. Ensure that the Appendix R criteria are Evaluation met for ATRI UM-1 fuel. This issue is addressed in a separate report.15.1.1 Loss of Feedwater Address each reload. Potentially limiting AQO.Heater (LFWH)15.1.2 Feedwater Controller Address each reload. Potentially limiting AQO.Failure (FWCF) -Maximum Demand 15.1.3 Inadvertent HPCI or No further analysis Consequences bound by the LFWH.RCIC Pump Start required.15.1.4 Pressure Regulator No further analysis Benign event.Failure Open required.(PRFO)15.1.5 Inadvertent Opening No further analysis Benign event.of a Relief Valve or required.Safety Valve 15.1.6 Inadvertent RHR No further analysis Benign event.Shutdown Cooling required.Operation AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 2-4 Table 2.1 Disposition of Events Summary for Brunswick Steam Electric Plant (Continued)

UFSAR Section 15.2.1 15.2.2 Event /Analysis Generator Load Rejection Turbine Trip 15.2.3 Main Steam Isolation Valve (MSIV)Closure 15.2.4 Loss of Condenser Vacuum 15.2.5 Loss of Auxiliary Power 15.2.6 Loss of Feedwater Flow 15.2.7 Loss of RHR Shutdown Cooling 15.2.8 Pressure Regulator Failure-Closed 15.3.1 Recirculation Pump Trip 15.3.2 Recirculation Flow Control Failure -Decreasing Flow 15.3.3 Recirculation Pump Seizure 15.4.1 Rod Withdrawal Error During Low Power Operation 15.4.2 Rod Withdrawal Error at Power 15.4.3 Startup of Idle Recirculation Loop Disposition Status Address each reload.Address for initial reload.No further analysis required.No further analysis required.No further analysis required.No further analysis required.No further analysis required.No further analysis required.No further analysis required.No further analysis required.No further analysis required.No further analysis required.Address each reload.No further analysis required.Comments Potentially limiting AQO with bypass failure.Potentially limiting AQO. It is expected that results will show that this event is bound by LRNB.Consequences bound by the LRNB and TTNB.Consequences bound by the LRNB.Consequences bound by the loss of feedwater flow and the LRNB.Benign event.Benign event.Benign event with backup pressure regulator in service.Benign event.Benign event and bound by the trip of one recirculation pump.Current analysis of record remains applicable as long as MCPR operating limit is greater then 1.40 at 73.3% power.Benign event.Potentially limiting AQO.Benign event.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 2-5 Table 2.1 Disposition of Events Summary for Brunswick Steam Electric Plant (Continued)

UFSAR Section Event /Analysis Disposition Status Comments 15.4.4 Recirculation Flow No further analysis Benign event.Control Failure -required.Increasing Flow 15.4.5 Fuel Assembly Error No further analysis Fuel loading errors during refueling with all During Refueling required.

rods fully inserted cannot result in core criticality.

15.4.6 Control Rod Drop Address each reload. Potentially limiting accident.Accident 15.6.3 Main Steam Line No further analysis Fuel-related consequences bound by Break Accident required.

other LOCA events. Results of the current radiological release evaluation remain applicable.

15.6.4 Loss of Coolant Address each reload. Potentially limiting accident.

Only heatup also 6.3 Accident (LOCA) analysis is expected for follow-on reloads to address changes in neutronic design.15.7.1 Refueling Accident Address for initial Potentially limiting accident.reload.15.8 Anticipated Transient Address each reload. Potentially limiting event. Only over-Without Scram pressurization analysis portion of the event is expected for follow-on reloads.15.9 Analytical Methods Address for initial Analyses to demonstrate that the nuclide for Evaluating reload, inventories for ATRIUM-la0 fuel are bound Radiological Effects by the AST analysis of record.With Alternative Source Term Slow Flow Runup Address each reload. Analysis results are used to establish the flow-dependent operating limits.Backup Stability Address each reload. Required to establish exclusion regions.Protection Mislocated or Address each reload.--Misoriented Fuel Assembly AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 2-6 Table 2.2 Disposition of Operating Flexibility and EQOS Options on Limiting Events Affected Limiting Option Events/Analyses .Comments MSIV Out-of-Service Slow flow runup The impact of the increase in steam line pressure drop on the slow flow runup analysis will be evaluated each reload.One SRV Out-of- ASME Overpressurization This scenario will be included as part of Service FWFthe base case condition for the FWCF events/analyses identified.

LRNB TTN B ATWS One ADS Valve LOCA The scenario will be included in the Out-of-Service break spectrum analyses for the initial cycle.FFTIRlFeedwater Option Ill Stability Solution This scenario will be examined each Heater Out-of- FWFreload for each of these events FWCFic /analyses.

Backup Stability Protection (BSP)Single-Loop Operation LOCA The impact of SLO on LOCA (including (SLO) SLCRthe main steam line break) will be SLMCPRaddressed for the initial cycle in the break spectrum analyses.The SLO SLMCPR will be addressed each reload.Turbine Bypass FWCF The FWCF event with TBVOOS will be Valves Out-of- evaluated each reload.Service Unit Auxiliary LIRNB The effect of the generator overspeed is Transformer (UAT)/ included in the base case LRNB, Site Auxiliary analysis.Transformer (SAT)AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 3-1 3.0 Mechanical Design Analysis The mechanical design analysis is presented in the mechanical design report. The maximum exposure limits for the ATRIUM-10 reload fuel are: 54.0 GWd/MTU average assembly exposure 62.0 GWd/MTU rod average exposure (full-length fuel rods)Even though the ATRIUM-la0 design is licensed for operation to a peak rod average exposure of 62 GWd/MTU, it will be limited to 60 GWd/MTU as prescribed in Brunswick Unit 1 license amendment 124 (Reference 2).The ATRIUM-10 LHGR limits are presented in Section 8.0. The GE14 MAPLHGR limits discussed in Section 8.0 ensure that the thermal-mechanical design criteria for GE14 fuel are satisfied.

The fuel cycle design analyses (Reference

1) have verified that all GE fuel assemblies remain within licensed burnup limits.AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 4-1 4.0 Thermal-Hydraulic Design Analysis 4.1 Thermal-Hydraulic Design and Compatibility The results of the thermal-hydraulic characterization and compatibility analyses are presented in the thermal-hydraulic design report (Reference 3). The analysis results demonstrate that the thermal-hydraulic design and compatibility criteria are satisfied for the Brunswick Unit 1 transition core consisting of ATRIUM-10 and GE14 fuel.4.2 Safety Limit MCPR Analysis The safety limit MCPR (SLMCPR) is defined as the minimum value of the critical power ratio which ensures that less than 0. 1% of the fuel rods in the core are expected to experience boiling transition during normal operation or an anticipated operational occurrence (AQO). The safety limit MCPR for all fuel in the Brunswick Unit 1 Cycle 17 core was determined using the methodology described in Reference

4. The analysis is performed with a power distribution that conservatively represents expected reactor operating states that could both exist at the MCPR operating limit and produce a MCPR equal to the safety limit MCPR during an AOO.The Brunswick Unit 1 Cycle 17 safety limit MCPR analysis used the SPC1B critical power'correlation additive constants and additive constant uncertainty for ATRIUM-i 0 fuel described in Reference

5. The SPC1B additive constants and additive constant uncertainty for the coresident GE14 fuel were developed using the indirect approach described in Reference 6.The determination of the safety limit MCPR explicitly includes the effects of channel bow relying on the following assumptions:

Cycle 17 will not contain fuel channels used for more than one fuel bundle lifetime, the average assembly burnup in Cycle 17 is less than 45 GWd/MTU for ATRIUM-10 fuel and 55 GWd/MTU for GE14 fuel. The channel bow local peaking uncertainty is a function of the nominal and bowed local peaking factors and the standard deviation of the channel bow.The fuel- and plant-related uncertainties used in the safety limit MCPR analysis are presented in Table 4.1. The radial power uncertainty used in the analysis includes the effects of up to 40% of the TIP channels out-of-service, up to 50% of the LPRMs out-of-service, and a 2500 EFPH LPRM calibration interval.The analysis results support a two-loop operation (TLO) SLMCPR of 1.09 and a single-loop operation (SLO) SLMCPR of 1. 10. However, per direction from Progress Energy, the Cycle 17 AREVA NP Inc.

AN P-2674(N P)Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 4-2 MCPR operating limits are based on SLMCPR values of 1.11 for TLO and 1.12 for SLO, the values currently in the plant Technical Specifications.

Table 4.2 presents a summary of the analysis results including the SLMCPR and the percentage of rods expected to experience boiling transition.

4.3 Core Hydro dynamic Stability Brunswick has implemented BWROG Long Term Stability Solution Option Ill (Oscillation Power Range Monitor-OPRM).

Reload validation has been performed in accordance with Reference 7.The stability based Operating Limit MCPR (OLMCPR) is provided for two conditions as a function of OPRM amplitude setpoint in Table 4.3. The two conditions evaluated are for a postulated oscillation at 45% core flow steady state operation (SS) and following a two recirculation pump trip (2PT) from the limiting full power operation state point. The Cycle 17 power- and flow-dependent limits provide adequate protection against violation of the SLMCPR for postulated reactor instability as long as the operating limit is greater than or equal to the specified value for the selected OPRM setpoint.

The results in Table 4.3 are valid for normal and reduced feedwater temperature (including EHOOS and FFTR) operation.

Evaluatio ns by General Electric (GE) have shown that the generic DIVOM curves specified in Reference 7 may not be conservative for current plant operating conditions for plants which have implemented Stability Option Ill. To address this issue, ARE VA has performed calculations for the relative change in CPR as a function of the calculated hot channel oscillation magnitude (HCOM). These calculations were performed with the RAMONA5-FA code. This code is a coupled neutron ic-thermal-hyd rau lic three-dimensional transient model for the purpose of determining the relationship between the relative change in ACPR and the HCOM on a plant specific basis. This method was developed consistent with the recommendations of the BWROG in Reference

8. The generation of the plant-specific DIVOM data with this model is consistent with the BWROG resolution of this nonconservatism as provided in Reference

9. The stability-based OLMCPRs were calculated for Cycle 17 using the most limiting calculated change in relative ACPR for a given oscillation magnitude.

In cases where the OPRM system is declared inoperable for Brunswick Unit 1 Cycle 17, Backup Stability Protection (BSP) in accordance with Reference 10 is provided.

BSP curves have been evaluated using STAIF (Reference

11) to determine endpoints that meet decay ratio criteria for the BSP Base Minimal Region I (scram region) and Base Minimal Region 11 (controlled entry AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 4-3 region). Stability boundaries based on these endpoints can then be determined using the generic shape generating function from Reference

10. Analyses have been performed to support operation with nominal feedwater temperature conditions and reduced feedwater temperature conditions (both FFTR and FHOOS).The STAlE acceptance criteria for the BSP endpoints are global decay ratios:!

0.85, and regional and channel decay ratios ý 0.80. The BSP analysis performed in support of Brunswick Unit 1 Cycle 17 was prepared to include conservatism in the supported BSP regions. The endpoints for the BSP regions provided in Table 4.4 have global decay ratios! <0.80, and regional and channel decay ratios ! 0.75. The intent of the conservatism is to provide regions that can be validated for future cycles using the STAIF criteria without the need to change the endpoints.

AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 4-4 Table 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses Parameter Fuel-Related Uncertainties Uncertainty Plant-Related Uncertainties Feedwater flow rate 1.8%Feedwater ternperature 0.8%Core pressure 0.8%Total core flow rate TLO 2.5%SLO 6%* [I AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 4-5 Table 4.2 Results Summary for Safety Limit MCPR Analyses Percentage SLMCPR* of Rods in Boiling Transition TLO -1.09 0.055 SLO -1.10 0.065*Note that per direction from Progress Energy, the Cycle 17 MCPR operating limits are based on SLMPCR values of 1.11 for TLO and 1.12 for SLO, the values currently in the plant Technical Specifications.

AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page 4-6 Table 4.3 OPRM Setpoints OPIRM OLMCPR OLMCPR Setpoint (SS) (2PT)1.05 1.21 1.20 1.06 1.23 1.22 1.07 1.24 1.24 1.08 1.26 1.26 1.09 1.29 1.28 1.10 1.31 1.30 1.11 1.33 1.32 1.12 1.36 1.35 1.13 1.38 1.37 1.14 1.41 1.40 1.15 1.43 1.42 Less than or Less than or equal to the equal to the Rated Power Off-Rated OLMCPR as Acceptance OLMCPR described in Criteria at 45% Flow Section 8.0 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 4-7 Table 4.4 BSP Endpoints for Brunswick Unit 1 Cycle 17 Feedwater Temperature Operation End Point Power Flow Mode Region Designation

(% rated) (% rated)Nominal Scram IA 56.6 40.0 Nominal Scram l13 40.7 31.0 Nominal Controlled 11A 64.5 50.0 entry Nominal Controlled 18B 28.5 31.0 entry FFTR/ Scram IA 64.9 50.5 EHOOS FFTRI Scram lB 37.3 31.0 EHOOS FFTRI Controlled 111A 66.1 52.0 FHOOS entry FFTRJ Controlled I11B 28.5 31.0 EHOOS entry AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 5-1 5.0 Anticipated Operational Occurrences This section describes the analyses performed to determine the power- and flow-dependent MCPR operating limits for base case operation at Brunswick Unit 1 Cycle 17.COTRANSA2 (Reference 12), XCOBRA-T (Reference 13), XCOBRA (Reference 14), and CASMO-4IMICROBURN-B2 (Reference

15) are the major codes used in the thermal limits analyses as described in the ARE VA THERMEX methodology report (Reference

14) and neutronics methodology report (Reference 15). COTRANSA2 is a system transient simulation code, which includes an axial one-dimensional neutronics model that captures the effects of axial power shifts associated with the system transients.

XCOBRA-T is a transient thermal-hydraulics code used in the analysis of thermal margins for the limiting fuel assembly.

XCOBRA is used in steady-state analyses.

The SPCB critical power correlation (Reference

5) is used to evaluate the thermal margin of the ATRIUM-10 and GE14 fuel. The application of the SPCB correlation to GE14 fuel follows the indirect process described in Reference

6. Fuel pellet-to-cladding gap conductance values are based on RODEX2 (Reference

16) calculations for the Brunswick Unit 1 Cycle 17 core.5.1 System Transients The reactor plant parameters for the system transient analyses were provided by the utility.Analyses have been performed to determine power-dependent MCPR limits that protect operation throughout the power/flow domain depicted in Figure 1.1.At Brunswick, direct scram on turbine stop valve (TSV) position and turbine control valve (TCV)fast closure are bypassed at power levels less than 26% of rated (Pbypass).

Scram will occur when the high pressure or high neutron flux scram setpoint is reached. Reference 17 indicates that MCPR limits only need to be monitored at power levels greater than or equal to 23% of rated, which is the lowest power analyzed for this report.The limiting exposure for rated power pressurization transients is typically at end of full power (EOFP) when the control rods are fully withdrawn.

To provide additional margin to the operating limits earlier in the cycle, analyses were also performed to establish operating limits at a near end-of-cycle (NEOC) exposure of 16,500 MWd/MTU. Analyses were performed at cycle exposures prior to NEOC to ensure that the operating limits provide the necessary protection.

The end-of-cycle licensing basis (EOCLB) analysis was performed at EOFP + 14 EFPD AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 5-2 (18,909 MWdIMTU).

Analyses were also performed to support extended cycle operation with final feedwater temperature reduction (FFTR) and power coastdown.

The Brunswick Unit 1 Cycle 17 licensing basis exposures used to develop the neutronics inputs to the transient analyses are presented in Table 5.1.All pressurization transients assumed that one of the lowest setpoint safety relief valves (SRV)was inoperable.

This basis supports operation with 1 SRV out-of-service.

The Brunswick Unit 1 turbine bypass system includes 4 bypass valves. However, for base case analyses in which credit is taken for turbine bypass operation, only 3 of the turbine bypass valves are assumed operable.Reductions in feedwater temperature of less than 10 0 OF from the nominal feedwater temperature are considered base cafse operation, not an EQOS condition.

This decrease in feedwater temperature causes a small increase in the core inlet subcooling which changes the axial power shape and core void fraction.

In addition, the steam flow for a given power level decreases since more power is used to increase the coolant enthalpy to saturated conditions.

The consequences of the FWCF event are more severe as a result of the increase in corei inlet subcooling during the overcooling phase of the event. Analyses were performed to demonstrate that reduced feedwater temperature is limiting for the FWCF event. While a decrease in steam flow tends to make the LRNB event less severe, the TCV initial position is further closed which tends to make the event more severe, especially at higher power levels. The LRNB events for base case operation were evaluated for both nominal and 1 0 0 F reduced feedwater temperatures.

FFTR is used to extend rated power operation by decreasing the feedwater temperature.

The amount of feedwater temperature reduction is a function of power with the maximum decrease of 11 0.3 0 F at rated power. Analyses were performed to support constant rated dome pressure combined FFTRlCoastdown operation to a cycle exposure of 20,674 MWd/MTU. The FWCF analyses were performed with the lowest feedwater temperature associated with the initial power level. LRNB analyses were performed for nominal and reduced feedwater temperatures.

The results of the system pressurization transients are sensitive to the scram speed used in the calculations.

To take advantage of average scram speeds faster than those associated with the Technical Specifications requirements, scram speed-dependent MCPRp limits are provided.

The nominal scram speed (NSS) insertion times and the Technical Specifications scram speed AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 5-3 (TSSS) insertion times used in the analyses are presented in Table 5.2. The NSS MCPRP limits can only be applied if the scram speed test results meet the NSS insertion times. System transient analyses were performed to establish MCPRp limits for both NSS and TSSS insertion times. The Brunswick Unit 1 Technical Specifications (Reference

17) allow for operation with up to 10 "slow" and 1 stuck control rod. One additional control rod is assumed to fail to scram.Conservative adjustments to the NSS and TSSS scram speeds were made to the analysis inputs to appropriately account for these effects on scram reactivity.

For cases below 26%power, the results are relatively insensitive to scram speed, and only TSSS analyses are per-formed.

At 26% power (Pbypass), FWCF analyses were performed both with and without bypass of the direct scram function which results in a step change in the operating limits.5.1.1 load Reiection No Bypass (LRNB)The load rejection causes a fast closure of the turbine control valves. The resulting compression wave travels through the steam lines into the vessel and creates a rapid pressurization.

The increase in pressure causes a decrease in core voids, which in turn causes a rapid increase in power. The fast closure of the turbine control valves also causes a reactor scram. Turbine bypass system operation, which also mitigates the consequences of the event, is not credited.The excursion of the core power due to the void collapse is terminated primarily by the reactor scram and revoiding of the core.For power levels less than 50% of rated, the LRN13 analyses assume that the power load unbalance (PLU) is inoperable.

With the PLU inoperable, the LRNB3 sequence of events is different than the standard event. Instead of a fast closure, the TCVs close in servo mode and there is no direct scram on TCV closure. The power and pressure excursion continues until the high pressure scram occurs.Operation with the recirculation pump power supplied by the Unit Auxiliary Transformer (UAT)results in a turbine overspeed and a subsequent recirculation pump overspeed during the LRNB event. The increase in pump speed causes an increase in core flow and a corresponding increase in power resulting in a slightly more severe event. All LRNB analyses were performed assuming the UAT supplies power to the recirculation pumps.LRNB analyses were performed for a range of power/flow conditions to support generation of the thermal limits. Tables 5.3 and 5.4 present the base case limiting LRNB3 transient analysis AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 5-4 results used to generate the NEOC and EOCLB operating limits for both TSSS and NSS insertion times. Figures 5.1 -5.3 show the responses of various reactor and plant parameters during the LRNB event initiated at 100% of rated power and 104.5% of rated core flow with TSSS insertion times.5.1.2 Turbine Trip No Bypass (TTNB)The turbine trip causes a closure of the turbine stop valves. The resulting compression wave travels through the steam lines into the vessel and creates a rapid pressurization.

The increase in pressure causes a decrease in core voids, which in turn causes a rapid increase in power.The closure of the turbine stop valves also causes a reactor scram. Turbine bypass system operation, which also mitigates the consequences of the event, is not credited.

The excursion of the core power due to the void collapse is terminated primarily by the reactor scram and revoiding of the core.At near rated conditions, the closure time for the TSV in the TTNB3 event is faster than the closure time of the TCV during the LRNB event. In addition, the scram delay for the TSV closure is shorter than the scram delay on TCV fast closure. At lower power levels (between 26% and 90% of rated), the TCV closure time is faster than the TSV closure ti me and the scram delay on valve motion is longer for the LRNB event than for the TTNB event. For power levels between 26% and 80% power, the consequences of the TTNB event are clearly bound by those of the LRNB event. TTNB analyses at 80% and 90% power were performed to ensure that the LRNB event is bounding when the closure times are similar.TTNB analyses were performed for power/flow conditions above 80% power and below Pbypass to support generation of the thermal limits. Tables 5.5 and 5.6 present the base case TTNB transient analysis results for both TSSS and NSS insertion times for Cycle 17. The system response to a TTNB event is very similar to the response during an LRNB event.5.1.3 Feedwater Controller Failure (FWCF)The increase in feedwater flow due to a failure of the feedwater control system to maximum demand results in an increase in the water level and a decrease in the coolant temperature at the core inlet. The increase in core inlet subcooling causes an increase in core power. As the feedwater flow continues at maximum demand, the water level continues to rise and eventually reaches the high water level trip setpoint.

The initial water level is conservatively assumed to be AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 5-5 at the low level normal operating range to delay the high-level trip and maximize the core inlet subcooling that results from the FWCF. The high water level trip causes the turbine stop valves to close in order to prevent damage to the turbine from excessive liquid inventory in the steam line. The valve closures create a compression wave that travels to the core causing a void collapse and subsequent rapid power excursion.

The closure of the turbine stop valves also initiates a reactor scram. Three of the four installed turbine bypass valves are assumed operable and provide some pressure relief. The core power excursion is mitigated in part by the pressure relief, but the primary mechanism for termination of the event is reactor scram.FWCF analyses were performed for a range of power/flow conditions to support generation of the thermal limits. Tables 5.7 and 5.8 present the base case limiting FWCF transient analysis results used to generate the NEOC and EOCLB operating limits for both TSSS and NSS insertion times. Figures 5.4 -5.6 show the responses of various reactor and plant parameters during the FWCF event initiated at 100% of rated power and 104.5% of rated core flow with TSSS insertion times.5.1.4 Pressure Regulator Failure Downscale (PRFIDS)The pressure regulator failure downscale event occurs when the pressure regulator fails and sends a signal to close all four turbine control valves in control mode. Normally, the backup pressure regulator would take control and maintain the setpoint pressure, resulting in a mild pressure excursion and a benign event. If one of the pressure regulators were out-of-service, there would be no backup pressure regulator and the event would be more severe. The core would pressurize resulting in void collapse and a subsequent power increase.

The event would be terminated by scram when either the high-neutron flux or a high-pressure setpoint is reached. Operation with one pressure regulator out-of-service is not supported for Brunswick over the entire power/flow map. However, Progress Energy requested analyses to demonstrate that the consequences of the PRFIDS event with one pressure regulator out-of-service are bound by the LRNB3 event at power levels greater than 90% of rated. Cycle 17 analysis results demonstrate that the LRNB is more limiting in this power range.5.1.5 Loss of Feedwater Heating The loss of feedwater heating (LFWH) event analysis supports an assumed 100OF decrease in the feedwater temperature.

The result is an increase in core inlet subcooling, which reduces AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 5-6 voids thereby increasing the core power and shifting the axial power distribution toward the bottom of the core. As a result of the axial power shift and increased core power, voids begin to build up in the bottom region of the core, acting as negative feedback to the increased subcooling effect. The negative feedback moderates the core power increase.

Although there is a substantial increase in core thermal power during the event, the increase in steam flow is much less because a large part of the added power is used to overcome the increase in inlet subcooling.

The increase in steam flow is accommodated by the pressure control system via the TCVs or the turbine bypass valves, so no pressurization occurs. For Brunswick Unit 1 Cycle 17, a cycle-specific analysis was performed in accordance with the Reference 18 methodology to determine the change in MCPR for the event. The LFWH results are presented in Table 5.9.5.1.6 Control Rod Withdrawal Error The control rod withdrawal error (CRWE) transient is an inadvertent reactor operator initiated withdrawal of a control rod. This withdrawal increases local power and core thermal power, lowering the core MCPR. The CRWE transient is typically terminated by control rod blocks initiated by the rod block monitor (RBM). The ORWE event was analyzed assuming no xenon and allowing credible instrumentation out-of-service in the rod block monitor (RBM) system. The analysis further assumes that the plant could be operating in either an A or B sequence control rod pattern. The rated power ORWE results are shown in Table 5.10 for the analytical RBM high power setpoint values of 108% to 117%. At all intermediate and lower power setpoint values, the MCPRP values for ATRIUM-1 and GE14 fuel bound or are equal to the CRWE MCPR values. AREVA analyses show that standard filtered RBM setpoint reductions are supported.

Analyses demonstrate that the 1 % strain and centerline melt criteria are met for both ATRIUM-11 and GEl 4 fuel with the LHGR and MAPLHGR limits and their associated multipliers presented in Sections 8.2 and 8.3. The recommended operability requirements based on the unblocked CRWE results are shown in Table 5.11 based on the SLMCPR values presented in Section 4.2.5.2 Slow Flow Runup Analysis Flow-dependent MCPR and LHGR limits are established to support operation at off-rated core flow conditions.

The limits are based on the CPR and heat flux changes experienced by the fuel during slow flow excursions.

The slow flow excursion event assumes a failure of the recirculation flow control system such that the core flow increases slowly to the maximum flow AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 5-7 physically attainable by the equipment (1107% of rated core flow). An uncontrolled increase in flow creates the potential for a significant increase in core power and heat flux. For initial core flows less than 60% of rated, operation with One MSIVOOS causes a larger increase in pressure and power during the flow excursion.

The result is a steeper flow runup path. A conservatively steep flow runup path was used in the analysis.

The slow flow runup analyses were performed to support operation in all the EOOS scenarios.

MCPRf limits are determined for both ATRIUM-10 and GE14 fuel. XCOBRA is used to calculate the change in critical power ratio during a two-loop flow runup to the maximum flow rate. The MCPRf limit is set so that the increase in core power resulting from the maximum increase in core flow is such that the TLO safety limit MCPR is not violated.

Calculations were performed for a range of initial flow rates to determine the corresponding MCPR values that put the limiting assembly on the safety limit MCPR at the high flow condition at the end of the flow excursion.

Results of the flow runup analysis are presented in Table 5.12. MCPRf limits that provide the required protection are presented in Table 8.7. The Cycle 17 MCPRf limits are applicable for all Cycle 17 exposures.

Flow runup analyses were performed with CASMO-4/MICROBURN-B2 to determine flow-dependent LHGR multipliers (LHGRFACf) for ATRIUM-10 fuel. The analysis assumes that the recirculation flow increases slowly along the limiting rod line to the maximum flow physically attainable by the equipment.

A series of flow excursion analyses were performed at several exposures throughout the cycle starting from different initial power/flow conditions.

Xenon is assumed to remain constant during the event. The LHGRFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a flow runup. The Cycle 17 LHGRFACf multipliers are presented in Table 8.13. A process consistent with the GNF thermal-mechanical methodology was used to determine flow-dependent MAPLHGR multipliers (MAPFACf) for GE14 fuel. These MAPFACf multipliers, presented in Table 8.16, provide protection against fuel centerline melt and overstraining of the cladding for GE14 fuel during operation at off-rated core flow conditions.

The maximum flow during a flow excursion in single-loop operation is much less than the maximum flow during two-loop operation.

Therefore, the flow-dependent MCPR limits and LHGRIMAPLHGR multipliers for two-loop operation are applicable for SLO.AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 5-8 5.3 Equipment Out-of-Service Scenarios The following equipment out-of-service (EQOS) scenarios are supported for Brunswick Unit 1 Cycle 17 operation:

• Feedwater heater out-of-service (EHOOS) -up to 11 0.3'F feedwater temperature reduction* Turbine bypass valves out-of-service (TB VOOS)* Combined EHOOS and TBVOOS* One safety/relief valve out-of-service (One SRVOOS)* One main steam isolation valve out-of-service (One MSIVOOS)* Single-loop operation (SLO)5.3.1 FHOOS The FHOOS scenario assumes a feedwater temperature reduction of 11 0.3 0 F at rated power and steam flow. The effect of the reduced feedwater temperature is an increase in the core inlet subcooling which can change the axial power shape and core void fraction.

In addition, the steam flow for a given power level decreases since more power is used to increase the enthalpy of the coolant to saturated conditions.

The consequences of the FWCF event are more severe as a result of the increase in core inlet subcooling during the overcooling phase of the event.While the decrease in steam flow tends to make the LRNB event less severe, the TCV initial position is further closed which tends to make the event more severe, especially at higher power levels. FWCF and LRNB events were analyzed to establish the appropriate FHOOS operating limits.5.3.2 TBVOOS For this EOOS scenario, operation with TBVOOS means that the fast opening capability of two or more of the turbine bypass valves cannot be assured, thereby reducing the pressure relief capacity during fast pressurization transients.

While the base case LRNB and TTNB events are analyzed assuming the turbine bypass valves out-of-service, operation with TBVOOS has an adverse effect on the FWCF event. Analyses of the FWCF event with TBVOOS were performed to establish the TBVOOS operating limits.AREVA NP Inc.

AN P-2674(N P)Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 5-9 5.3.3 Combined FHOOS and TBVOOS FWCF analyses with both FHOOS and TBVOOS were performed to support Cycle 17 operation.

Operating limits for this combined EQOS scenario were established using these FWCF results.5.3.4 One SRVOOS As noted earlier, all pressurization transient analyses were performed with one of the lowest setpoint SRVs assumed inoperable.

Therefore, the base case operating limits support operation with one SRVOOS. The EOOS operating limits also support operation with one SRVOOS.5.3.5 One MSIVOOS Operation with One MVSIVOOS is supported for operation less than 70% of rated power. At these reduced power levels, the flow through any one steam line will not be greater than the flow at rated power when all MSIVs are available.

Since all four turbine control valves are available, adequate pressure control can be maintained.

The main difference in operation with One MSIVOOS is that the steam line pressure drop between the steam dome and the turbine valves is higher than if all MVSIVs are available.

Since low steam line pressure drop is limiting for pressurization transients, the results of the pressurization events with all MVSIVs in service bound the results with One MVSIVOOS.

In addition, operation with One MSIVOOS has no impact on the other non-pressurization events evaluated to establish power-dependent operating limits.Therefore, the power-dependent operating limits applicable to operation with all MVSIVs in service remain applicable for operation with One MVSIVOOS for power levels less than or equal to 70% of rated. As noted earlier, slow flow runup analyses were performed to support operation with One MSIVOOS.5.3.6 Sin-gle-LooD OPeration In SLO, the two-loop operation ACPRs and LHGRFAC/MAPFAC multipliers remain applicable.

The only impacts on the MCPR, LHGR, and MAPLHGR limits for SLO are an increase of 0.01 in the SLMCPR as discussed in Section 4.2, and the application of an SLO MAPLHGR multiplier discussed in Section 8.3. The net result is a 0.01 increase in the base case MCPRp limits and a decrease in the MAPLHGR limit. The same situation is true for the EOOS scenarios.

Adding 0.01 to the corresponding two-loop operation EOOS MCPRp limits results in SLO MCPRp limits for the EOOS conditions.

The TLO EOOS LHGRFAC and MAPFAC multipliers limits remain applicable in SLO.AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 5-10 5.4 Licensing Power Shape The licensing axial power profile used by ARE VA for the plant transient analyses bounds the projected end of full power axial power profile. The conservative licensing axial power profile generated at EOCLB (18,909 MWd/MTU) is given in Table 5.13. Cycle 17 operation is considered to be in compliance when the normalized power generated in the bottom 7 nodes from the projected EOFP axial power shape at the given state conditions is greater than the normalized power generated in the bottom 7 nodes in the licensing basis axial power profile. If the criteria cannot be fully met (i.e., not all 7 nodes are at a higher power than the licensing profile), the licensing basis may nevertheless remain valid but further assessment will be required.The licensing basis power profile in Table 5.13 was calculated using the MICROBURN-B32 code.Compliance analyses must also be performed using MICROBURN-B32.

Note that the power profile comparison should be done without incorporating instrument updates to the axial profile because the updated power is not used in the core monitoring system to accumulate assembly burnups.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-11 Table 5.1 Exposure Basis for Brunswick Unit 1 Cycle 17 Transient Analysis Cycle Core Exposure at Average End of Interval Exposure (MWd/MTU) (MWd/MTU)

Comments 0 13,166 Beginning of cycle 16,500 29,667 Break point for exposure-dependent MCPRp limits (NEOC)18,909 32,076 Design basis rod patterns to EOFP + 14 EFPD (EOCLB)20,674 33,841 End of reactivity for FFTR/Coastdown

-maximum core exposure AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page 5-12 Table 5.2 Scram Speed Insertion Times Control Rod TSSS NSS Position Time Time (notch) (sec) (sec)48 (full-out) 0.000 0.000 48 0.200 0.200 46 0.440 0.304 36 1.080 0.822 26 1.830 1.353 6 3.350 2.468 0 (full-in) 3.806 2.804 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page 5-13 Table 5.3 NEOC Base Case LRNB Transient Results Power ATRIUM-10 ATRIUM-10 GE14/Flow ACPR HER ACPR T 100 /104.5 100 /99 90/104.5 901/84 80/I106 80 /70 70 /108 60/110 50/110 50 I1110 PLU inoperable 26 / 106 PLU inoperable 26 / 106 below Pbypass 26 / 50 below Pbypass 23 1 103 below Pbypass'SSS Insertion 0.35 0.32 0.36 0.27 0.38 0.27 0.37 0.36 0.39 0.93 1.27 1.27 1.06 1.34 Times 1.33 1.33 1.33 1.31 1.34 1.26 1.33 1.33 1.39 1.96 2.25 2.25 1.90 2.30 0.32 0.31 0.34 0.27 0.36 0.24 0.36 0.35 0.36 0.91 1.27 1.27 1.03 1.34 23 / 50below Pbypass 1.15 1.96 1.12 NSS Insertion Times 100 /104.5 0.23 1.25 0.22 100/99 0.20 1.22 0.19 90 /104.5 0.27 1.26 0.25 80/106 0.29 1.27 0.28 70/108 0.30 1.27 0.29 60/110 0.31 1.27 0.29 50/110 0.35 1.36 0.32 50 1 110 PLU inoperable 0.89 1.92 0.89 26 1 106 PLU inoperable 1.25 2.23 1.25 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-14 Table 5.4 EOCLB Base Case LRNB Transient Results Power IFlow ATRIUM-b A 100 /104.5 100 /99 90 /104.5 90 /84 80/106 80 /70 70 /108 60/110 50 /110 50 / 110 PLU inoperablE 26 / 106 PLU inoperablE 26 / 106 below Pbypass 26 / 50 below Pbypass 23 / 103 below Pbypas, 23 / 50 below Pbypass ACPR TSSS Insertion Times 0.35 0.34 0.36 0.32 0.38 0.32 0.38 0.38 0.39 0.93 1.27 1.27 1.06 1.34 1.15 NSS Insertion Times 0.29 0.27 0.32 0.34 0.35 0.35 0.37 0.89 1.25 JTRIUM-1O GE14 HFR ACPR 1.40 1.41 1.39 1.38 1.39 1.34 1.39 1.38 1.40 1.99 2.25 2.25 1.90 2.30 1.97 0.35 0.34 0.36 0.31 0.37 0.31 0.37 0.37 0.39 0.92 1.27 1.27 1.03 1.34 1.12 100 /104.5 100 /99 90 /104.5 80/106 70/108 60/110 50/110 50 / 110 PLU inoperablE 26 / 106 PLU inoperablE 1.35 1.35 1.35 1.35 1.35 1.35 1.39 1.93 2.23 0.30 0.28 0.32 0.34 0.35 0.35 0.37 0.89 1.25 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-15 Table 5.5 NEOC Base Case TTNB Transient Results Power/ Flow ATRIUM-b ATRIUIV 100 /104.5 90 /104.5 80/I106 26 / 106 below Pbypass 26 /150 below PbypasS 23 /103 below Pbypass 23 / 50 below Pbypa~s ACPR HFR TSSS Insertion Times 0.33 1.31 0.34 1.31 0.35 1.31 1.15 2.03 1.00 1.73 1.23 2.08 1.09 1.77 NSS Insertion Times 0.22 1.23 0.24 1.23 0.25 1.24 1-10 GE14 ACPR 0.31 0.32 0.33 1.16 0.99 1.23 1.07 100 /104.5 90/104.5 80 /106 0.21 0.23 0.24 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-16 Table 5.6 EOCLB Base Case TTNB Transient Results Power/ Flow ATRIUM-b ATF 100 /104.5 90/104.5 80 /106 26 1 106 below Pbypass 26 1 50 below Pbypass 23 1 103 below Pbypass 23 1 50 below Pbypass ACPR TSSS Insertion Times 0.34 0.35 0.36 1.15 1.00 1.23 1.09 NSS Insertion Times 0.28 0.301 0.321 1.38 1.37 1.36 2.03 1.73 2.08 1.78 0.33 0.34 0.35 1.16 0.99 1.23 1.07 IUM-10 GE14-IFR ACPR 100 /104.5 90 /104.5 80/I106 1.34 1.32 1.32 0.29 0.30 0.31 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-17 Table 5.7 NEOC Base Case FWCF Transient Results Power ATRIUM-l0 ATRIUM-10 GE14/Flow ACPR HER ACPR TSSS Insertion Times 100/104.5 0.31 1.29 0.29 100/99 0.29 1.28 0.27 90 /104.5 0.33 1.30 0.32 90/84 0.24 1.26 0.24 80/106 0.36 1.32 0.35 80 /70 70 /108 60/110 50 /110 26 /106 26 1 106 below Pbypass_26 1 50 below Pbypass 23 1 103 below Pbypass_23 / 50 below Pbypass 0.23 0.39 0.41 0.42 0.60 1.29 1.23 1.38 1.27 1.22 1.36 1.38 1.40 1.55 2.19 2.00 2.25 2.02 0.21 0.37 0.40 0.41 0.63 1.33 1.22 1.41 1.26 100 /104.5 100 /99 90 /104.5 80 /106 70 /108 60 /110 50 /110 26 /106 NSS Insertion Times 0.21 0.18 0.24 0.27 0.31 0.34 0.37 0.56 1.20 1.19 1.23 1.26 1.30 1.33 1.36 1.53 0.20 0.18 0.23 0.27 0.30 0.34 0.36 0.56 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page 5-18 Table 5.8 EOCLB Base Case FWCF Transient Results Power/ Flow ATRIUM-b ATF 100 /104.5 100 /99 90 /104.5 90/I84 80 /106 80/I70 70 /108 60 /110 50 /110 26 /106 26 1 106 below Pbypa~s 26 1 50 below Pbypass 23 / 103 below Pbypass 23 / 50 below Pbypass ACPR TSSS Insertion Times 0.31 0.30 0.33 0.28 0.36 0.28 0.39 0.41 0.42 0.60 1.29 1.23 1.38 1.27 1.33 1.33 1.34 1.31 1.36 1.27 1.39 1.41 1.42 1.55 2.19 2.00 2.25 2.02 0.30 0.29 0.32 0.27 0.35 0.27 0.37 0.40 0.41 0.63 1.33 1.22 1.41 1.26 IlUM-10 GE14 HER ACPR NSS Insertion Times 100 /104.5 0.26 1.28 0.26 100/99 0.24 1.27 0.24 90 /104.5 0.29 1.30 0.28 80/106 0.32 1.32 0.31 70 /108 60 /110 50/I110 26 /106 0.35 0.37 0.39 0.56 1.36 1.39 1.40 1.53 0.34 0.37 0.38 0.56 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-19 Table 5.9 Loss of Feedwater Heating Transient Analysis Results Power ATRIUM-i O/GE14 (% rated) ACPR 100 0.10 90 0.11 80 0.12 70 0.13 60 0.14 50 0.16 40 0.19 30 0.24 25 0.28 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-20 Table 5.10 Control Rod Withdrawal Error ACPR Results Analytical RBM Setpoint (without filter) ACPR*M%108 0.18 ill 0.21 114 0.24 117 0.28 Results are for the most limiting of the ATRIUM-l0 or GE14 fuel in the core.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-21 Table 5.11 RBM Operability Requirements Applicable Thermal Power ATRIUM-i O/GEI4 (% rated) MCPR 29% nd <90%1.64 TLO29%ad<90%1.65 SLO>90% 1.39 TLO AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 5-22 Table 5.12 Flow-Dependent MCPR Results Core ATRIUM-10 GE14 Flow Limiting Limiting (% rated) MCPR MCPR 31 1.43 1.44 40 1.41 1.40 50 1.38 1.38 60 1.36 1.36 70 1.27 1.28 80 1.23 1.23 90 1.20 1.20 100 1.16 1.16 107 1.11 1.11 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-23 Table 5.13 Licensing Basis Core Average Axial Power Profile State Conditions for Power Shape Evaluation Power, MWt 2923.0 Core pressure, psia 1055.0 Inlet subcooling, Btu/lbm 21.58 Flow, Mlb/hr 80.47 Control state ARO Core average exposure 32,076 (EOCLB), MWd/MT1U Licensing Axial Power Profile (Normalized)

Node Top 25 24 23 22 21 20 19 18 17 16 13 12 11 10 9 8 7 6 5 4 3 2 Bottom 1 Power 0.195 0.621 0.807 0.908 0.982 1.04 1 1.092 1.141 1.188 1.246 1.288 1.397 1.420 1.422 1.401 1.360 1.297 1.210 1.099 0.972 0.847 0.742 0.654 0.522 0.149 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0-vlrv Quu.u I~400.0 -300.0 -0 CL Core Power Heat Flux-CoreT FlO w SteamFlow

_Feed Flow7 200.0 -100.0-.0 -I (I~NJ 1/-IOlA.0 1.0 2.0 3!0 Time, (seconds)Figure 5.1 EOCLB LRNB at lOOP/I 04.5F -TSSS Key Parameters 4.~0 5.0 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page 5-25 E 0 C 0 U1)Figure 5.2 EOCLB LRNB at 100P1104.5F -TSSS Sensed Water Level AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-26 U)9-Figure 5.3 EOCLB LRNB at 100P1104.5F -TSSS Vessel Pressures AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-27 ci, a 0 ci)C.)ci)0~15.0 Time, (seconds)Figure 5.4 EOCLB FWCF at lOOP/I 04.5 F -TSSS Key Parameters AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-28 r E n U)CL V)1 ...Time, (seconds)Figure 5.5 EOCLB FWCF at 100P1104.5F

-TSSS Sensed Water Level AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 5-29 (L Time, (seconds)Figure 5.6 EOCLB FWCF at lOOP/I 04.5F -TSSS Vessel Pressures AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 6-1 6.0 Postulated Accidents 6.1 Loss-of-Coolant-Accident (LOCA)The results of the ATRIUM-i10 LOCA analysis are presented in References 19 and 20. The ATRIUM-10 licensing PCT is 1900OF and the peak local metal water reaction is 1.16%. The maximum core wide metal water reaction (for hydrogen generation) for a full core of ATRIUM-i1 0 fuel is < 0.50%. The SLO MAPLHGR multiplier is 0.85.The GE14 LOCA analysis results are presented in Reference 21 (UFSAR).6.2 Control Rod Drop Accident (CRDA)Brunswick Unit 1 uses a bank position withdrawal sequence (BPWS) including reduced notch worth rod pull to limit high worth control rod movements.

A CRDA evaluation was performed for both A and B sequence startups consistent with the withdrawal sequence specified by Progress Energy. Reference 22 describes the approved AREVA generic CRDA methodology.

Subsequent calculations have shown that the methodology is applicable to fuel modeled with the CASMO4/MICROBURN-B2 code system. The CRDA analysis was performed with the approved methodology described in Reference 22..The ORDA analysis results demonstrate that the maximum deposited fuel rod enthalpy is less than the NRC threshold of 280 cal/g and that the estimated number of fuel rods that exceed the fuel damage threshold of 170 cal/g is less than the number of failed rods assumed in the Brunswick Unit 1 UFSAR radiological assessment (1200 rods).Maximum dropped control rod worth, mk 12.1 Core average Doppler coefficient, Ak/kI 0 F -1 1.0 X 10-6 Effective delayed neutron fraction 0.0054 Four-bundle local peaking factor 1.388 Maximum deposited fuel rod enthalpy, cal/g 209 Maximum number of rods exceeding 170 cal/g 366 6.3 Fuel and Equipment Handling Accident The Brunswick fuel handling accident radiological analysis implementing the alternative source term (AST) methodology was performed with consideration of GE fuel in the core inventory source terms. Progress Energy has subsequently shown that the current BNP source term is AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 6-2 applicable to cores with ATRIUM-10 fuel. AREVA has performed an analysis that shows that the number of failed fuel rods due to a fuel handling accident impacting the ATRIUM-11 fuel is 163.This is less than the number of rods analyzed in the Brunswick AST analysis (172 rods). The analysis also shows that the slightly higher mass of the ATRIUM-i1 0 fuel does not result in an increase in rod failures when dropped onto GE14 fuel. Therefore, the AST analysis remains applicable for either an ATRIUM-10/GE14 mixed core or a full core of ATRIUM-10 fuel.6.4 Fuel Loading Error (in frequent Event)There are two types of fuel loading errors possible in a BWR -the mislocation of a fuel assembly in a core position prescribed to be loaded with another fuel assembly, and the misorientation of a fuel assembly with respect to the control blade. As described in Reference 23, the fuel loading error is characterized as an infrequent event. The acceptance criteria is that the offsite dose consequences due to the event shall not exceed a small fraction of the 10 CFR 50.67 limits.6.4.1 Mislocated Fuel Bundle ARE VA has performed a bounding fuel mislocation error analysis and has demonstrated continued applicability of the bounding results to Brunswick.

This analysis evaluated the impact of a mislocated assembly against potential fuel rod failure mechanisms due to increased LHGR and reduced CPR. Based on these analyses, the offsite dose criteria (a small fraction of 10 CFR 50.67) is conservatively satisfied.

Since no rod approached the fuel centerline melt or 1 % strain limits, and less than 0. 1% of the fuel rods are expected to experience boiling transition which could result in a dryout induced failure, a dose consequence evaluation is not necessary.

6.4.2 Misoriented

Fuel Bundle AREVA has performed a bounding fuel assembly misorientation analysis.

The analysis was performed assuming that the limiting assembly was loaded in the worst orientation (rotated 1800) while producing sufficient power to be on the MCPR limit if it had been oriented correctly.

The analysis demonstrates that the small fraction of 10 CFR 50.67 offsite dose criteria is conservatively satisfied.

A dose consequence evaluation is not necessary since no rod approached the fuel centerline melt or 1 % strain limits and less than 0.1 % of the fuel rods are expected to experience boiling transition.

AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 7-1 7.0 Special Analyses 7.1 A SME Overpressurization Analysis This section describes the maximum overpressurization analyses performed to demonstrate compliance with the ASME Boiler and Pressure Vessel Code. The analysis shows that the safety/relief valves at Brunswick Unit 1 have sufficient capacity and performance to prevent the reactor vessel pressure from reaching the safety limit of 110% of the design pressure.MSIV closure and TSV closure (without bypass) analyses were performed with the ARE VA plant simulator code COTRANSA2 (Reference

12) for 102% power and both 99% and 104.5% flow at the highest Cycle 17 exposure where rated power operation can be attained.

The MSIV closure event is similar to the other steam line valve closure events in that the valve closure results in a rapid pressurization of the core. The increase in pressure causes a decrease in void which in turn causes a rapid increase in power. The turbine bypass valves do not impact the system response and are not modeled in the analysis.

The following assumptions were made in the analysis.* The most critical active component (direct scram on valve position) was assumed to fail.However, scram on high neutron flux and high dome pressure is available.

• To maintain consistency with the bases discussion in Reference 17, the plant configuration analyzed assumed that two of the lowest setpoint SRVs were inoperable.

• TSSS insertion times were used.* The initial dome pressure was set at the maximum allowed by the Technical Specifications, 1059.7 psia (1045 psig).* A fast MSIV closure time of 2.7 seconds was used.Results of the MSIV closure and TSV closure overpressurization analyses are presented in Table 7.1. Figures 7.1 -7.4 show the response of various reactor plant parameters during the MSIV closure event, the event which results in the maximum vessel pressure.

The maximum pressure of 1352 psig occurs in the lower plenum. The maximum dome pressure for the same event is 1317 psig. The results demonstrate that the maximum vessel pressure limit of 1375 psig and dome pressure limit of 1325 psig are not exceeded for all analyses.AREVA NP Inc.

AN P-2674(N P)Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 7-2 7.2 ATWS Event Evaluation 7.2.1 ATWS overpressurization Analysis This section describes the analyses performed to demonstrate that the peak vessel pressure for the limiting ATWS event is less than the ASME Service Level C limit of 120% of the design pressure (1500 psig). The ATWS overpressurization analyses were performed at 100% power at both 99% and 104.5% flow over the Cycle 17 exposure range for both the MSIV closure event and the pressure regulator failure open (PRFO) event. The pressure regulator failure open event assumes a step decrease in pressure demand such that the pressure control system opens the turbine control and turbine bypass valves such that 115% of rated steam flow (maximum combined steam flow limit) is attained.

The system pressure decreases until the low pressure setpoint is reached resulting in the closure of the MSIVs. The resulting pressurization wave causes a decrease in core voids, an increase in core power, and an increase in core pressure.The following assumptions were made in the analyses.* The analytical limit ATWS-RPT setpoint and function were assumed.* To support operation with 1 SRVOOS, the plant configuration analyzed assumed that one of the lowest setpoint SRVs was inoperable.

0 All scram functions were disabled.0 The initial dome pressure was set to the nominal pressure of 1045 psia.0 A nominal MSIV closure time of 4.0 seconds was used for both events.Results of analyses for the ATWS overpressurization analyses are presented in Table 7.2.Figures 7.5 -7.8 show the response of various reactor plant parameters during the limiting PRFO event, the event which results in the maximum vessel pressure.

The maximum lower plenum pressure is 1465 psig and the maximum dome pressure is 1447 psig. The results demonstrate that the ATWS maximum vessel pressure limit of 1500 psig is not exceeded.7.2.2 Long-Term Evaluation Fuel design differences may impact the power and pressure excursion experienced during the ATWS event. This in turn may impact the amount of steam discharged to the suppression pool and containment.

[AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-3 I Relative the 10 CFR 50.46 acceptance criteria (i.e., POT and cladding oxidation), the consequences of an ATWS event are bound by those of the limiting LOCA event.7.3 Standby Liquid Control System In the event that the control rod scram function becomes incapable of rendering the core in a shutdown state, the standby liquid control (SLC) system is required to be capable of bringing the reactor from full power to a cold shutdown condition at any time in the core life. The Brunswick Unit 1 SLO system is required to be able to inject 720 ppm natural boron equivalent at 70OF into the reactor coolant (including a 25% allowance for imperfect mixing, leakage, and volume of other piping connected to the reactor).

ARE VA has performed an analysis that demonstrates that the SLC system meets the required shutdown capability for Cycle 17. The analysis was AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 7-4 performed at a coolant temperature of 360OF with a boron concentration equivalent to 720 ppmn at 70 0 F. The temperature of 360OF corresponds to the low pressure permissive for the RHR shutdown cooling suction valves, and represents the maximum reactivity condition with soluble boron in the coolant. The analysis shows the core to be subcritical throughout the cycle by at least 1.31% Ak/k.7.4 Fuel Criticality The new fuel storage vault criticality analysis for ATRIUM-I1 0 fuel is presented in Reference 24.The spent fuel pool criticality analysis for ATRIUM-i1 0 fuel is presented in Reference

25. The ATRIUM-i1 0 fuel assemblies identified for loading in Cycle 17 meet both the new and spent fuel storage requirements.

AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-5 Table 7.1 ASME Overpressurization Analysis Results Maximum Peak Peak Vessel Maximum Neutron Heat Pressure Dome Flux Flux Lower-Plenum Pressure Event (% rated) (% rated) (psig) (psig)MSIV closure (102P/104.5F) 338 127 1352 1317 MSIV closure (102P/99F) 321 128 1350 1317 TSV closure without bypass (1102P/1 04.5F) 513 135 1330 1294 TSV closure without bypass (102P/99F) 496 136 1327 1293 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-6 Table 7.2 ATWS Overpressurization Analysis Results Maximum Peak Peak Vessel Maximum Neutron Heat Pressure Dome Flux Flux Lower-Plenum Pressure Event (% rated) (% rated) (psig) (psig)MSIV closure (100P/104.5F) 254 136 1440 1423 MSIV closure (11OQP/99F) 258 135 1442 1425 PRFO (110OP/1104.5F) 258 144 1465 1447 PRFO (100P/99F) 252 142 1464 1447 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-7 Table 7.3 1 I I t I AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-8 a 0 4, U 4)0~4.0 6.*0 Time, (seconds)Figure 7.1 MSIV Closure Overpressurization Event at 102P1104.5F

-Key Parameters AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-9 190.0 0 N E (D, CL U)a)0D C (D (n Figure 7.2 MSIV Closure Overpressurization Event at 102 P/104.5 F -Sensed Water Level AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-10 0 C,, 0~C,, U)U)C,)0~4.0 6.'0 Time, (seconds)Figure 7.3 MSIV Closure Overpressurization Event at 102P/1104.5F

-Vessel Pressures AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-11 E V)4.0 6.0 Time, (seconds)Figure 7.4 MSIV Closure Overpressurization Event at 102 P/1 04.51F -Safety/Relief Valve Flow Rates AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-12 0 (D 20.0 30.0 Time, (seconds)Figure 7.5 PRFO ATWS Overpressurization Event at 100 P/104.5F -Key Parameters AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-13 0 N 4)E Figure 7.6 PRFO ATWS; Overpressurization Event at 10 OP/1104.5F

-Sensed Water Level AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-14 CIn (L 10.0 20.0 30.0 40.0 Time, (seconds)Figure 7.7 PRFO ATWS Overpressurization Event at 10 OP/1 04.5 F -Vessel Pressures AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 7-15 (n E-o 0 (n Figure 7.8 PRFO ATWS Overpressurization Event at 1 OOPII 04.5F -Safety/Relief Valve Flow Rates AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 8-1 8.0 Operating Limits and COLR Input 8.1 MCPR Limits The determination of the MCPR limits for Brunswick Unit 1 Cycle 17 is based on the analyses of the limiting anticipated operational occurrences (AQOs). The MCPR operating limits are established so that less than 0. 1% of the fuel rods in the core are expected to experience boiling transition during an AQO initiated from rated or off-rated conditions and are based on the Technical Specifications two-loop operation SLMCPR of 1.11 and a single-loop operation SLMCPR of 1.12. Exposure-dependent MCPR limits were established to support operation from BOO to near end-of-cycle (NEOC), NEOC to end-of-cycle licensing basis (EOCLB) and combined FFTRlCoastdown.

MCPR limits are established to support base case operation and the EOOS scenarios presented in Table 1.1.Cycle 17 two-loop operation MCPRp limits for ATRIUM-1 and GE14 fuel are presented in Tables 8.1 -8.6 for base case operation and the EQOS conditions.

Limits are presented for nominal scram speed (NSS) and Technical Specification scram speed (TSSS) insertion times for the exposure ranges considered.

Tables 8.1 and 8.2 present the MCPRp limits for the BOO to NEOC exposure range. Tables 8.3 and 8.4 present the MCPRp limits applicable for the BOO to EOCLB exposure range. Tables 8.5 and 8.6 present the MCPRp limits for FFTRlCoastdown operation.

The FFTR/Coastdown limits (both base case and TBVOOS) support constant rated dome pressure operation with feedwater temperatures consistent with reduction of up to 11 0.3 0 F at rated power. MCPRP limits for single-loop operation are 0.01 higher for all cases.Comparisons of the limiting AQO analysis results and the MCPRP limits for ATRIUM-i 0 and GE14 fuel are presented in Appendix A.MCPRf limits that protect against fuel failures during a postulated slow flow excursion for ATRIUM-10 and GE14 fuel are presented in Table 8.7 and are applicable for all Cycle 17 exposures and the EOOS conditions identified in Table 1.1.8.2 LHGR Limits The LHGR limits for ATRIUM-10 fuel are presented in Table 8.8. Power- and flow-dependent multipliers (LHGRFACp and LHGRFACf) are applied directly to the LHGR limits to protect against fuel melting and overstraining of the cladding during an AQO.AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 8-2 LHGRFACp multipliers are determined using. the heat flux ratio results from the transient analyses.

Exposure-dependent LHGRFACP multipliers were established to support operation from BOC to EOCLB and combined FFTRlCoastdown for both NSS and TSSS insertion times and for the EOOS conditions identified in Table 1.1. The ATRIUM-10 Cycle 17 LHGRFACp multipliers for the BOO to EOCLB exposure range are presented in Tables 8.9 and 8.10. The FFTRlCoastdown LHGRFACP multipliers are presented in Tables 8.11 and 8.12. The FFTRlCoastdown limits (both base case and TBVOOS) support constant rated dome pressure operation with feedwater temperatures consistent with reduction of up to 11 0.3 0 F at rated power. Comparisons of the limiting analysis results and the LHGRFACP limits for ATRIUM-lO fuel are presented in Appendix A.LHGRFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a postulated slow flow excursion.

For ATRIUM-i1 0 fuel, the multipliers are presented in Table 8.13 and are applicable for all Cycle 17 exposures and the EQOS conditions identified in Table 1.1.Note that LHGR limits are not applied to the GE1 4 fuel so there are no GE1 4 power- or flow-dependent LHGR multipliers.

The fuel centerline melt and overstraining of the cladding for GE14 fuel are ensured by applying power- and flow-dependent MAPLHGR limits as discussed below.8.3 MAPLHGR Limits The ATRIUM-la0 MAPLHGR limits are discussed in Reference

20. The TLO operation limits are presented in Table 8.14. For operation in SLO, a multiplier of 0.85 must be applied to the TLO MAPLHGR limits.The MAPLHGR limits for GE14 fuel are presented in Reference

26. Power- and flow-dependent multipliers are applied to the GE14 MAPLHGR limits. Application of the MAPFACp and MAPFACf multipliers to the GE14 fuel ensures that the fuel centerline melt and overstraining of the cladding criteria are met during AQ0s. The MAPFACp and MAPFACf multipliers were developed in a manner consistent with the GNF thermal-mechanical methodology for GE14 fuel.MAPFACp multipliers were determined using the transient analysis results. Exposure-dependent MAPFACp multipliers were established to support operation for all Cycle 17 exposures, both NSS and TSSS insertion times and all the EQOS conditions identified in Table 1.1. The GE14 MAPFACp multipliers for all Cycle 17 exposures are presented in Table 8.15.AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 8-3 MAPFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a postulated slow flow excursion for GE14 fuel. The GE14 MAPFACf multipliers are presented in Table 8.16 and are applicable for all Cycle 17 exposures and the EQOS conditions identified in Table 1.1.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-4 Table 8.1 MCPRP Limits for NSS Insertion Times BOC to < N EOC (116,500 MWdIMTU)*EQOS Power ATRIUM-l0 GE14 Condition

(% rated) MCPRP MCPRP 100.0 1.35 1.35 90.0 1.38 1.37 50.0 1.53 1.53 Base 50.0 2.02 2.02 case 26.0 2.39 2.39 operation 26.0 at > 50%F 2.44 2.48 23.0 at > 50%F 2.53 2.56 26.0 at:5 50%F 2.38 2.37 23.0 at S 50%F 2.42 2.41 100.0 1.38 1.37 90.0 1.42 1.40 50.0 1.59 1.58 50.0 2.02 2.02 TBVOOS 26.0 2.39 2.39 26.0 at > 50%F 3.00 3.01 23.0 at > 50%F 3.18 3.17 26.0 at:5 50%F 2.66 2.62 23.0 at S 50%F 2.88 2.84 100.0 1.36 1.36 90.0 1.40 1.38 50.0 1.57 1.57 50.0 2.02 2.02 EHOOS 26.0 2.39 2.39 26.0 at > 50%F 2.60 2.63 23.0 at > 50%F 2.73 2.76 26.0 at 5 50%F 2.47 2.45 23.0 at 5 50%F 2.55 2.54 100.0 1.42 1.39 90.0 1.45 1.42 50.0 1.64 1.63 TBVOOS 50.0 2.02 2.02 and FHOOS 26.0 2.39 2.39 26.0Oat >50%F 3.14 3.13 23.0 at > 50%F 3.31 3.30 26.0 at:5 50%F 2.76 2.72__________23.0 at S 50%F 3.00 2.95*Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRP limits will be 0.01 higher.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-5 Table 8.2 MCPRp Limits for TSSS Insertion Times BOC to < N EOC (16,500 MWd/MTU)*EQOS Power ATRIUM-10 GE14 Condition

(% rated) MCPRP MCPRP 100.0 1.46 1.43 90.0 1.47 1.45 50.0 1.55 1.54 Base 50.0 2.06 2.04 case 26.0 2.41 2.41 operation 26.0 at > 50%F 2.44 2.48 23.0 at > 50%F 2.53 2.56 26.0 at:5 50%F 2.38 2.37 23.0 at:5 50%F 2.42 2.41 100.0 1.49 1.46 90.0 1.52 1.49 50.0 1.65 1.64 50.0 2.06 2.04 TBVOOS 26.0 2.41 2.41 26.0 at > 50%F 3.00 3.01 23.0 at > 50%F 3.18 3.17 26.0 at S 50%F 2.66 2.62__________23.0 at S 50%F 2.88 2.84 100.0 1.47 1.44 90.0 1.48 1.46 50.0 1.62 1.62 50.0 2.06 2.04 EHOOS 26.0 2.41 2.41 26.0 at > 50%F 2.60 2.63 23.0 at > 50%F 2.73 2.76 26.0 at:5 50%F 2.47 2.45 23.0 at S 50%F 2.55 2.54 100.0 1.50 1.47 90.0 1.53 1.50 50.0 1.70 1.69 TBVOOS 50.0 2.06 2.04 and FHOOS 26.0 2.41 2.41 26.0 at > 50%F 3.14 3.13 23.0 at > 50%F 3.31 3.30 26.0 at S 50%F 2.76 2.72__________23.0 at:5 50%F 3.00 2.95*Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRP limits will be 0.01 higher.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-6 Table 8.3 MCPRp Limits for NSS Insertion Times BOC to < EOCLB (18,909 MWd/MTU)*EQOS Power ATRIUM-10 GE14 Condition

(% rated) MCPRP MCPRP 100.0 1.40 1.41 90.0 1.43 1.43 50.0 1.53 1.53 Base 50.0 2.02 2.02 case 26.0 2.39 2.39 operation 26.0 at > 50%F 2.44 2.48 23.0 at > 50%F 2.53 2.56 26.0 at 5 50%F 2.38 2.37_________23.0 at 5 50%F 2.42 2.41 100.0 1.43 1.43 90.0 1.46 1.45 50.0 1.62 1.61 50.0 2.02 2.02 TBVOOS 26.0 2.39 2.39 26.0 at > 50%F 3.00 3.01 23.0 at > 50%F 3.18 3.17 26.0 at 5 50%F 2.66 2.62_________23.0 at:5 50%F 2.88 2.84 100.0 1.41 1.42 90.0 1.44 1.44 50.0 1.57 1.57 50.0 2.02 2.02 EHOOS 26.0 2.39 2.39 26.0 at > 50%F 2.60 2.63 23.0 at > 50%F 2.73 2.76 26.0 at 5 50%F 2.47 2.45_________23.0 at:5 50%F 2.55 2.54 100.0 1.45 1.48 90.0 1.48 1.51 50.0 1.64 1.63 TBVOOS 50.0 2.02 2.02 and EHOOS 26.0 2.39 2.39 26.0 at > 50%F 3.14 3.13 23.0 at > 50%F 3.31 3.30 26.0 at:5 50%F 2.76 2.72__________23.0 at 50%F 3.00 2.95*Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRP limits will be 0.01 higher.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page 8-7 Table 8.4 MCPRP Limits for.TSSS Insertion Times BOC to < EOCLB (18,909 MWdIMTU)*EQOS Power ATRIUM-10 GE14 Condition

(% rated) MCPRP MCPRP 100.0 1.46 1.46 90.0 1.47 1.47 50.0 1.55 1.54 Base 50.0 2.06 2.05 case 26.0 2.41 2.41 operation 26.0 at > 50%F 2.44 2.48 23.0 at > 50%F 2.53 2.56 26.0 at 50%F 2.38 2.37_________23.0 at 5 50%F 2.42 2.41 100.0 1.49 1.48 90.0 1.52 1.50 50.0 1.65 1.64 50.0 2.06 2.05 TBVOOS 26.0 2.41 2.41 26.0 at > 50%F 3.00 3.01 23.0 at > 50%F 3.18 3.17 26.0 at:5 50%F 2.66 2.62_________23.0 at:5 50%F 2.88 2.84 100.0 1.47 1.47 90.0 1.48 1.49 50.0 1.62 1.62 50.0 2.06 2.05 EHOOS 26.0 2.41 2.41 26.0 at > 50%F 2.60 2.63 23.0 at > 50%F 2.73 2.76 26.0 at:5 50%F 2.47 2.45_________23.0 at S 50%F 2.55 2.54 100.0 1.50 1.54 90.0 1.53 1.55 50.0 1.70 1.69 TBVOOS 50.0 2.06 2.05 and FHOOS 26.0 2.41 2.41 26.0 at > 50%F 3.14 3.13 23.0 at > 50%F 3.31 3.30 26.0 at:5 50%F 2.76 2.72__________23.0 at S 50%F 3.00 2.95*Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRP limits will be 0.01 higher.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-8 Table 8.5 MCPRP Limits for NSS Insertion Times FFTRlCoastdown

(ý: 18,909 MWdIMTU )*EQOS Power ATRIUM-10 GE14 Condition

(% rated) MCPRP MCPRP 100.0 1.45 1.44 90.0 1.45 1.44 50.0 1.57 1.57 Base 50.0 2.02 2.02 case 26.0 2.39 2.39 operation 26.0 at > 50%F 2.60 2.63 23.0 at > 50%F 2.73 2.76 26.0 at 5 50%F 2.47 2.45 23.0 at:5 50%F 2.55 2.54 100.0 1.46 1.48 90.0 1.48 1.51 50.0 1.70 1.65 50.0 2.02 2.02 TBVOOS 26.0 2.39 2.39 26.0 at >50%F 3.14 3.13 23.0 at > 50%F 3.31 3.30 26.0 at:5 50%F 2.76 2.72__________23.0 at:5 50%F 3.00 2.95*Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRP limits will be 0.01 higher.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-9 Table 8.6 MCPRP Limits for TSSS Insertion Times FFTRlCoastdown (2:18,909 MWdIMTU)*EQOS Power ATRIUM-10 GE14 Condition

(% rated) MCPRP MCPRP 100.0 1.48 1.47 90.0 1.48 1.49 50.0 1.62 1.62 Base 50.0 2.06 2.05 case 26.0 2.41 2.41 operation 26.0 at > 50%F 2.60 2.63 23.0 at > 50%F 2.73 2.76 26.0 at 5 50%F 2.47 2.45 23.0 at S 50%F 2.55 2.54 100.0 1.50 1.54 90.0 1.53 1.55 50.0 1.70 1.69 50.0 2.06 2.05 TBVOOS 26.0 2.41 2.41 26.0 at > 50%F 3.14 3.13 23.0 at > 50%F 3.31 3.30 26.0 at S 50%F 2.76 2.72__________23.0 at:5 50%F 3.00 2.95*Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRP limits will be 0.01 higher.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-10 Table 8.7 Flow-Dependent MCPR Limits ATRIUM-10 and GE14 Fuel Core Flow (%of rated) MCPRf 0.0 1.55 31.0 1.55 100.0 1.20 107.0 1.20 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-11 Table 8.8 ATRIUM-10 Steady-State LHGR Limits Peak Pellet Exposure LHGR (GWd/MTU) (kW/ft)0.0 13.4 18.9 13.4 74.4 7.1 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-12 Table 8.9 ATRIUM-10 LHGRFACp Multipliers for NSS Insertion Times BOC to < EOCLB3 (18,909 MWd/MTU)EQOS Power ATRIUM-10 Condition

(% rated) LHGRFACP 100.0 1.00 80.0 1.00 50.0 0.94 Base 50.0 0.67 case 26.0 0.57 operation 26.0 at > 50%F 0.56 23.0 at > 50%F 0.54 26.0 at 5 50%F 0.63 23.0 at S 50%F 0.62 100.0 0.99 80.0 0.96 50.0 0.87 50.0 0.67 TBVOOS 26.0 0.57 26.0 at > 50%F 0.46 23.0 at > 50%F 0.44 26.0 at S 50%F 0.58 23.0 at:5 50%F 0.54 100.0 1.00 80.0 0.99 50.0 0.91 50.0 0.67 EHOOS 26.0 0.57 26.0 at > 50%F 0.53 23.0 at > 50%F 0.51 26.0 at:5 50%F 0.60__________23.0 at:5 50%F 0.59 100.0 0.99 80.0 0.95 50.0 0.86 TBVOOS 50.0 0.67 and EHOOS 26.0 0.57 26.0 at > 50%F 0.44 23.0 at > 50%F 0.42 26.0 at:5 50%F 0.55__________23.0 at S 50%F 0.50 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page 8-13 Table 8.10 ATRIUM-10 LHGRFACp Multipliers for TSSS Insertion Times BOC to < EOCLB (18,909 MWd/MTU)EQOS Power ATRIUM-10 Condition

(% rated) LHGRFACp 100.0 1.00 80.0 0.94 50.0 0.93 Base 50.0 0.65 case 26.0 0.57 operation 26.0 at > 50%F 0.56 23.0 at > 50%F 0.54 26.0 at:5 50%F 0.63 23.0 at:5 50%F 0.62 100.0 0.95 80.0 0.93 50.0 0.87 50.0 0.65 TBVOOS 26.0 0.57 26.0 at > 50%F 0.46 23.0 at > 50%F 0.44 26.0 at S 50%F 0.58 23.0 at:5 50%F 0.54 100.0 0.95 80.0 0.94 50.0 0.90 50.0 0.65 EHOOS 26.0 0.57 26.0 at > 50%F 0.53 23.0 at > 50%F 0.51 26.0 at:5 50%F 0.60 23.0 at:5 50%F 0.59 100.0 0.95 80.0 0.93 50.0 0.85 TBVOOS 50.0 0.65 and EHOOS 26.0 0.57 26.0 at > 50%F 0.44 23.0 at > 50%F 0.42 26.0 at:5 50%F 0.55___________23.0 at S 50%F 0.50 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-14 Table 8.11 ATRIUM-10 LHGRFACp Multipliers for NSS Insertion Times FFTRlCoastdown (2:18,909 MWdIMTU)EQOS Power ATRIUM-10 Condition

(% rated) LHGRFACP 100.0 1.00 80.0 0.97 50.0 0.91 Base 50.0 0.67 case 26.0 0.57 operation 26.0 at > 50%F 0.53 23.0 at > 50%F 0.51 26.0 at:5 50%F 0.60___________23.0 at 5 50%F 0.59 100.0 0.97 80.0 0.92 50.0 0.86 50.0 0.67 TBVOOS 26.0 0.57 26.0 at > 50%F 0.44 23.0 at > 50%F 0.42 26.0 at:5 50%F 0.55___________23.0 at 50%F 0.50 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-15 Table 8.12 ATRIUM-lU LHGRFACp Multipliers for TSSS Insertion Times FFTRlCoastdown (2:18,909 MWd/MTU)EQOS Power ATRIUM-l0 Condition

(% rated) LHGRFACp 100.0 1.00 80.0 0.94 50.0 0.89 Base 50.0 0.65 case 26.0 0.57 operation 26.0 at > 50%F 0.53 23.0 at > 50%F 0.51 26.0 at 5 50%F 0.60 23.0 at S 50%F 0.59 100.0 0.95 80.0 0.91 50.0 0.85 50.0 0.65 TBVOOS 26.0 0.57 26.0 at > 50%F 0.44 23.0 at > 50%F 0.42 26.0 at 50%F 0.55___________23.0 at:5 50%F 0.50 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page 8-16 Table 8.13 ATRIUM-10 LHGRFACf Multipliers All Cycle 17 Exposures Core Flow (% of rated) LHGRFACf 0.0 0.90 31.0 0.90 49.12 1.00 107.0 1.00 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-17 Table 8.14 ATRIUM-10 MAPLHGR Limits Average Planar Exposure MAPLHGR (GWd/MTU) (kW/ft)0.0 12.5 15.0 12.5 67.0 7.3 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page 8-18 Table 8.15 GE14 MAPFACp Multipliers for NSS and TSSS Insertion Times All Cycle 17 Exposures EQOS Power GE14 Condition

(% rated) MAPFACP 100.0 1.00 50.0 0.73 50.0 0.46 Base 26.0 0.43 caseBeo bps operation and Beo bas EHOOS 26 at > 50%F 0.43 23 at > 50%F 0.42 26 at 5 50%F 0.46 23 at 5 50%F 0.42 100.0 1.00 50.0 0.73 50.0 0.46 TBVOOS and 26.0 0.43 CombinedBeo bps TBVOOS and Beo bas FHOOS 26 at > 50%F 0.36 23 at > 50%F 0.34 26 at 5 50%F 0.42__________23 at 5 50%F 0.40 AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page 8-19 Table 8.16 GE14 MAPFACf Multipliers All Cycle 17 Exposures Core Flow GE14 (% rated) MAPFACf 0.0 0.56 31.0 0.56 80.0 1.00 107.0 1.00 AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 9-1 9.0 References 1 .ANP-2658(P)

Revision 0, Brunswick Unit 1 Cycle 17 Fuel Cycle Design, AREVA NP, July 2007.2. Letter, Edmond G. Tourigny (NRC) to E. E. Utley (CP&L), "Issuance of Amendment No.124 to Facility Operating License No. DPR-71 -Brunswick Steam Electric Plant, Unit 1 Regarding Cycle 7 Reload (TAC No. 69200)," February 6, 1989.3. ANP-2646(P)

Revision 0, Brunswick Unit 1 Thermal-Hydraulic Design Report for ATRIUM T M-10 Fuel Assemblies, AREVA NP, June 2007.4. AN F-524(P)(A)

Revision 2 and Supplements 1 and 2, ANF Critical Power Methodology for Boiling Water Reactors, Advanced Nuclear Fuels Corporation, November 1990.5. EMF-2209(P)(A)

Revision 2, SPCB Critical Power Correlation, Framatome ANP, September 2003.6. EMF-2245(P)(A)

Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.7. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications, GE Nuclear Energy, August 1996.8. 0G04-0153-260, Plant-Specific Regional Mode DIVOM Procedure Guideline, June 15, 2004.9. BWROG-03047, Resolution of Reportable Condition for Stability Reload Licensing Calculations Using Generic Regional Mode DIVOM Curve, September 30, 2003.10. 0G02-01 19-260, Backup Stability Protection (BSP) for Inoperable Option Ill Solution, GE Nuclear Energy, July 17, 2002.11. EMF-CC-074(P)(A)

Volume 4 Revision 0, BWR Stability Analysis -Assessment of STAlE with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.12. ANF-913(P)(A)

Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2:

A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.13. XN-NF-84-1 05(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T:

A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.14. XN-NF-80-19(P)(A)

Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.15. EMF-21 58(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors:

Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.AREVA NP Inc.

AN P-2674(N P)Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page 9-2 16. XN-NF-81-58(P)(A)

Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.17. Operating License and Technical Specifications, Brunswick Steam Electric Plant, Unit No 1, Progress Energy, as amended.18. AN F-1 358(P)(A)

Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, September 2005.19. ANP-2625(P)

Revision 0, Brunswick Units I and 2 LOCA Break Spectrum Analysis for ATRIUM T M -10 Fuel, AREVA NP, June 2007.20. ANP-2624(P)

Revision 0, Brunswick Units I and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM~m-1O Fuel, AREVA NP, June 2007.21. Updated FSAR Brunswick Steam Electric Plant, Units I and 2, Revision 20.22. XN-NF-80-1 9(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors -Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.23. XN-N F-80-1 9(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.24. ANP-2661 (P) Revision 0, Brunswick Nuclear Plant New Fuel Storage Vault Criticality.

Safety Analysis for A TRIUM T M -10 Fuel, AREVA NP, September 2007.25. ANP-2642(P)

Revision 0, Brunswick Nuclear Plant Spent Fuel Storage Pool Criticality Safety Analysis for A TRIUM T M-_10 Fuel, AREVA NP, September 2007.26. Brunswick Unit 1, Cycle 16 Core Operating Limits Report Revision 1, Progress Energy, May 2007.AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page A-i Appendix A Operating Limits and Results Comparisons The figures and tables presented in this appendix show comparisons of the Brunswick Unit 1 Cycle 17 operating limits and the transient analysis results. Comparisons are presented for the ATRIUM-10 and GE14 MCPRp limits and the ATRIUM-10 L HGRFACp multipliers.

AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page A-2 3.0 2.5 F-0t E-2.0 o3 FWCF 0 LRNB3 A PRFDS+ TTNB3 o3 x CRWE 0 0' LFWH 0++ 0+0 x 00 8x 0 00 0 0 1.5 1.0 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.35 90.0 1.38 50.0 1.53 50.0 2.02 26.0 2.39 26.0 > 50%F 2.44 23.0 > 50%F 2.53 26.0:5:,50%F 2.38 23.0:5 50%F 2.42 Figure A.1 BOC to NEOC Power-Dependent MCPR Limits for ATRIUM-lO0 Fuel -NSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-3 3.0 2.5 l--zj 0_ 2.0 1Y_o3 FWCF 0 LRN13 a PRFDS+ TTNB 0x CRWE 0 0 LW Y FW+ 0 0 0 I 00 0 0 0 0 o0 0 0 1.5 [F 1.0 i i i i i i i i 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.35 90.0 1.37 50.0 1.53 50.0 2.02 26.0 2.39 26.0 > 50%F 2.48 23.0 > 50%F 2.56 26.0:5 50%F 2.37 23.0:5 50%F 2.41 Figure A.2 BOC to NEOC Power-Dependent MCPR Limits for GE14 Fuel -NSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-4 3.5 3.0_j 0~C-)2.5* FWCF* LRNB* PRFDS* TTNB x CRWE* LFWH 0 0++ 0+x 00 0 0~~ 0 2.0 H-1.5 H 1.0 i 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.38 90.0 1.42 50.0 1.59 50.0 2.02 26.0 2.39 26.0 > 50%F 3.00 23.0 > 50%F 3.18 26.05 50%F 2.66 23.05~ 50%F 2.88 Figure A.3 BOC to NEOC Power-Dependent MCPR Limits for ATRIUM-lU0 Fuel -NSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-5 3.5 3.0-j CL a-)2.5 2.0* FWCF* LRNB* PRFDS* TTNB x CRWE 0 LFWH 0+ 0 0+ 0 0 x 000 x OX x 0 0 0 0 0 0 0 0 1.5 1.0 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)Power MCPRP (% of rated) Limit 100.0 1.37 90.0 1.40 50.0 1.58 50.0 2.02 26.0 2.39 26.0 > 50%F 3.01 23.0 > 50%F 3.17 26.05~ 50%F 2.62 23.0:5 50%F 2.84 Figure AA4 BOC to NEOC Power-Dependent MCPR Limits for GE14 Fuel -NSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-6 3.0 2.5 I-ZJ 0_ 2.0 0~C-)* FWCF* LRNB* PRFDS* TTNB 0 x CRWE 0 0 LFWH 0++ 0+3 0 x I00 0 X X x 00 0 0 1.5 I-1.0 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.36 90.0 1.40 50.0 1.57 50.0 2.02 26.0 2.39 26.0 > 50%F 2.60 23.0 > 50%F 2.73 26.05~ 50%F 2.47 23.05~ 50%F 2.55 Figure A.5 BOC to NEOC Power-Dependent MCPR Limits for ATRIUM-la0 Fuel -NSS Insertion Times -FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-7 3.0 2.5 1-:t c.2.0 n~O FWCF* LRNB* PRFDS* TTNB x CRWE 0* LFWH+ 0 0 0 0 00 00 0 0 0 0 0 0 0 00 1.5 F-1.0 i i i i i j i i i i 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.36 90.0 1.38 50.0 1.57 50.0 2.02 26.0 2.39 26.0 > 50%F 2.63 23.0 > 50%F 2.76 26.05~ 50%F 2.45 23.05~ 50%F 2.54 Figure A.6 BOC to NEOC Power-Dependent MCPR Limits for GE14 Fuel -NSS Insertion Times -FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page A-8 3.5 I I I I I I I I I I 3.0 0 0 x 0 FWCF LRNB PRFDS TTNB ORWE LFWH_j CL 2.5 2.0 0 0+0 0 0 x 1.5 x 0 6 0 0?x9 0 0 1.0 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRp (% ofrated) Limit 100.0 1.42 90.0 1.45 50.0 1.64 50.0 2.02 26.0 2.39 26.0 > 50%F 3.14 23.0 > 50%F 3.31 26.0:5 50%F 2.76 23.0:5 50%F 3.00 Figure A.7 BOC to NEOC Power-Dependent MCPR Limits for ATRIUM-lO0 Fuel -NSS Insertion Times -TBVOOS FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-9 3.5 3.0-j OL 2.5 2.0 E3 FWCF 0 LRNB a PRFDS+ TTNB x CRWE 0 LFWH 0+ 0 0 x Co 0 0X 0 0 0 0 000 1.5 1.0 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.39 90.0 1.42 50.0 1.63 50.0 2.02 26.0 2.39 26.0 > 50%F 3.13 23.0 > 50%F 3.30 26.0:5 50%F 2.72 23.0~5O5%F 2.95 Figure A.8 BOC to NEOC Power-Dependent MCPR Limits for GE14 Fuel -NSS Insertion Times -TBVOOS FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-1 3.0 2.5 l--CL2.0* FWCF* LRNB A PRFDS+ TTNB 0 x ORWE 0 0 0 LFWH 0 0 ++ 0+0 0 x 0 0o 0 0 0 0 1.5 1.0 0 10 20 30 40 50 60 70 50 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.46 90.0 1.47 50.0 1.55 50.0 2.06 26.0 2.41 26.0 > 50%F 2.44 23.0 > 50%F 2.53 26.0:5 50%F 2.38 23.05~ 50%F 2.42 Figure A.9 BOC to NEOC Power-Dependent MCPR Limits for ATRIUM-lO0 Fuel -TSSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-1 1 3.0 2.5 l--J 0-2.0 k-*l FWCF o LRNB* PRFDS* TTNB 0 )< CRWE 0 Y LFWH 0 0 0+1 0 0 x 00 x x x 000 000 0 0 4 1.5 1.0 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRp (% of rated) Limit 100.0 1.43 90.0 1.45 50.0 1.54 50.0 2.04 26.0 2.41 26.0 > 500/F 2.48 23.0 > 50%F 2.56 26.0:5 50%F 2.37 23.0:5 50%F 2.41 Figure A.10 BOC to NEOC Power-Dependent MCPR Limits for GE14 Fuel -TSSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page A-1 2 3.5 3.0 C-)2.5 o FWCF* LRNB* PRFDS* TTNB x GRWE* LFWH 0 0 0+0+ 0 0 x C 0 0 X xx, 0 00 0 0 0 0 2.0 1-1.5 1.0 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRp (% of rated) Limit 100.0 1.49 90.0 1.52 50.0 1.65 50.0 2.06 26.0 2.41 26.0 > 50%F 3.00 23.0 > 50%F 3.18 26.05~ 50%F 2.66 23.05~ 50%F 2.88 Figure A.11 BOC to NEOC Power-Dependent MCPR Limits for ATRIUM-lO Fuel -TSSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-1 3 3.5 3.0 a C-)2.5 2.0 o FWCF 0 LRNB3& PRFDS+ TTN13 x ORWE 0 LFWH 0 0 0+0 0 0 x 0000 x OX x x 0 0 000 0 0 0 1.5 1.0 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.46 90.0 1.49 50.0 1.64 50.0 2.04 26.0 2.41 26.0 > 50%F 3.01 23.0 > 50%F 3.17 26.05 50%F 2.62 23.0:5 50%F 2.84 Figure A.12 BOC to NEOC Power-Dependent MCPR Limits for GE14 Fuel -TSSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-14 3.0 2.5 [-2.0 O FWCF 0 LRNB A PRFDS+ TTNB 0x ORWE 0+0 0 LFWH 0 +0+ 0 0 0 0 xx x X X x 0 00 0 0 0 0 8 1.5 [-1.0 i i i i i i i 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.47 90.0 1.48 50.0 1.62 50.0 2.06 26.0 2.41 26.0 > 50%F 2.60 23.0 > 50%F 2.73 26.05~ 50%F 2.47 23.05~ 50%F 2.55 Figure A.1 3 BOC to NEOC Power-Dependent MCPR Limits for ATRIUM-10 Fuel -TSSS Insertion Times -FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-1i5 3.0 2.5 l--Q_ 2.0 a-)1.5 1.0 o3 FWCF* LRNB* PRFDS* TTNB x ORWE 0 0+ 0 00 0 0 0 10 20 30 40 50 Power 60 (% Rated)70 80 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.44 90.0 1.46 50.0 1.62 50.0 2.04 26.0 2.41 26.0 > 50%F 2.63 23.0 > 50%F 2.76 26.05 50%F 2.45 23.0:5 50%F 2.54 Figure A.14 BOC to NEOC Power-Dependent MCPR Limits for GE14 Fuel -TSSS Insertion Times -FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-1 6-j:3.5 3.0 2.5 2.0 1.5 1.0 o FWCF o LRNB3* PRFDS* TTNB x CRWE 0 LFWH 0 0+0 0+0+0 0 03 x 0 0 0 0 A 0 0 00 0 0 0 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)Power MCPRP (% ofrated) Limit 100.0 1.50 90.0 1.53 50.0 1.70 50.0 2.06 26.0 2.41 26.0 > 50%F 3.14 23.0 > 50%F 3.31 26.05 50%F 2.76 23.0:5 50%F 3.00 Figure A.1 5 BOC to NEOC Power-Dependent MCPR Limits for ATRIUM-lO0 Fuel -TSSS Insertion Times -TBVOOS FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-1 7 3.5 3.0 a_2.5 o FWCF 0 LRNB A PRFDS+ TTN13 x ORWE 0 LFWH 0++0 0+0+00 00 0+00 x C,0x0 x x X 0,< , ~ C , A 2.0 I-1.5 1.0 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% ofrated) Limit 100.0 1.47 90.0 1.50 50.0 1.69 50.0 2.04 26.0 2.41 26.0 > 50%F 3.13 23.0 > 50%F 3.30 26.0:5 50%F 2.72 23.0:5 50%F 2.95 Figure A.1 6 BOC to NEOC Power-Dependent MCPR Limits for GE14 Fuel -TSSS Insertion Times -TBVOOS FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-1 8 3.0 2.5 H--J E~_j 2.0 o FWCF 0 LRNB a PRFDS+ TTNB ox CRWE 0 0 LFWH 0++ 0 D+0 x 0C* 0 C, C, A 1.5 1.0 0 10 20 30 40 50 60 Power (% Rated)70 50 90 100 110 Power MCPRp (% of rated) Limit 100.0 1.40 90.0 1.43 50.0 1.53 50.0 2.02 26.0 2.39 26.0 > 50%F 2.44 23.0 > 50%F 2.53 26.0:550%F 2.38 23.05~ 50%F 2.42 Figure A.17 BOC to EOCLB Power-Dependent MCPR Limits for ATRIUM-l0 Fuel -NSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-1 9 3.0 2.5 l--2.0 0~C-)D FWCF 0 LRNB a PRFDS+ TTNB ox ORWE 0 0 LFWH+ 0 0 0 0 0 1.5 [-1.0 i i i i i i i j 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.41 90.0 1.43 50.0 1.53 50.0 2.02 26.0 2.39 26.0 > 50%F 2.48 23.0 > 50%F 2.56 26.05~ 50%F 2.37 23.05~ 50%F 2.41 Figure A.18 BOC to EOCLB Power-Dependent MCPR Limits for GE14 Fuel -NSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page A-20 3.5 3.0_j o-af a-2.5 2.0 O3 FWCF 0 LRNB A PRFDS+ TTN13 x CRWE 0 LFWH 0 0++ 0+x xx 00 0 1.5 1.0 i i i i i i i i i i 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.43 90.0 1.46 50.0 1.62 50.0 2.02 26.0 2.39 26.0 > 50%F 3.00 23.0 > 50%F 3.18 26.0:5 50%F 2.66 23.05~ 50%F 2.88 Figure A.19 BOC to EOCLB Power-Dependent MCPR Limits for ATRIUM-1 Fuel -NSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-21 3.5 3.0 41 a-)2.5 2.0 o FWCF* LRNB* PRFDS* TTNB x ORWE* LFWH 0+ 0 0+ 0 130 00 0 X 0 0 I~ 0 0 1.5 1.0 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.43 90.0 1.45 50.0 1.61 50.0 2.02 26.0 2.39 26.0 > 50%F 3.01 23.0 > 50%F 3.17 26.05 50%F 2.62 23.05 50%F 2.84 Figure A.20 BOC to EOCLB Power-Dependent MCPR Limits for GE14 Fuel -NSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-22 3.0 2.5 [-I]0_ 2.0 Of a-)* FWCF 0 LRNB* PRFDS* TTNB3 o CRWE 0 0 LFWH 0++ 0 x O x x 0 0 A 0 00 0 0 0 0 1.5 l--1.0 0 10 20 30 40 50 60 Power (%~ Rated)70 80 90 100 110 Power MCPRp (% of rated) Limit 100.0 1.41 90.0 1.44 50.0 1.57 50.0 2.02 26.0 2.39 26.0 > 50%F 2.60 23.0 > 50%F 2.73 26.0:5 50%F 2.47 23.05~ 50%F 2.55 Figure A.21 BOC to EOCLB3 Power-Dependent MCPR Limits for ATRIUM-la Fuel -NSS Insertion Times -FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-23 3.0 2.5 P-:t 0_ 2.0 o3 FWCF* LRNB* PRFDS* TTNB x CRWE 0*' LFWH+0 0 I x 0 0 0 0 0 0 0 0 0 1.5 H-1.0 i i i i 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.42 90.0 1.44 50.0 1.57 50.0 2.02 26.0 2.39 26.0 > 50%F 2.63 23.0 > 50%F 2.76 26.05~ 50%F 2.45 23.05~ 50%F 2.54 Figure A.22 BOC to EOCLB3 Power-Dependent MCPR Limits for GE14 Fuel -NSS Insertion Times -FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page A-24 3.5 3.0 0~0~C-)2.5 2.0 o FWCF 0 LRNB& PRFDS+ TTNB x CRWE 0 LFWH 0 0+0+ 0 0 0 0 0 0 00 0 0 0 1.5 1.0 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.45 90.0 1.48 50.0 1.64 50.0 2.02 26.0 2.39 26.0 > 50%F 3.14 23.0 > 50%F 3.31 26.05~ 50%F 2.76 23.0:5 50%F 3.00 Figure A.23 BOC to EOCLB Power-Dependent MCPR Limits for ATRIUM-10 Fuel -NSS Insertion Times -TBVOOS FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-25 3.5 3.0_j CY-2.5 o FWCF o LRNB* PRFDS* TTNB* CRWE*' LFWH 0+ 0 0+ 0+Q 4O0 00 2.0I 1.5 1.0 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.48 90.0 1.51 50.0 1.63 50.0 2.02 26.0 2.39 26.0 > 50%F 3.13 23.0 > 50%F 3.30 26.05~ 50%F 2.72 23.0:5 50%F 2.95 Figure A.24 BOC to EOCLB3 Power-Dependent MCPR Limits for GE14 Fuel -NSS Insertion Times -TBVOOS FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-26 3.0 I I I I I I III o3 FWCF* LRNB* PRFDS* TTNB3 x CRWE* LFWH 2.5 F-03 0 0 0 ++ 0 I+Q_2.0 W~n-03 01 1.5 1-x x x x 0 0 0 0 0 0 0 0 1.0 III I I I I I I I 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.46 90.0 1.47 50.0 1.55 50.0 2.06 26.0 2.41 26.0 > 50%F 2.44 23.0 > 50%F 2.53 26.0:5 50%F 2.38 23.05~ 50%F 2.42 Figure A.25 BOC to EOCLB Power-Dependent MCPR Limits for ATRIUM-lO0 Fuel -TSSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-27 3.0 2.5 1-2.0 O3 FWCF o LRNB3 A~ PRFDS+ TTNB3 0x CRWE 0 0 LFWH 0 0 0+x xx 0A 0 0 1.5 F-1.0 0 10 20 30 40 50 60 Power (% Rated)70 50 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.46 90.0 1.47 50.0 1.54 50.0 2.05 26.0 2.41 26.0 > 50%F 2.48 23.0 > 50%F 2.56 26.05~ 50%F 2.37 23.05~ 50%F 2.41 Figure A.26 BOC to EOCLB Power-Dependent MCPR Limits for GE14 Fuel -TSSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-28 3.5 3.0F_j a-)2.5 o FWCF* LRN13* PRFDS* TTNB x ORWE 0 LFWH 0 0 0+ 0+ 0 00 0 00 0 0 0 0 0 2.0 1--1.5 1.0 0 10 20 30 40 50 60 70 s0 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.49 90.0 1.52 50.0 1.65 50.0 2.06 26.0 2.41 26.0 > 50%F 3.00 23.0 > 50%F 3.18 26.05 50%F 2.66 23.05~ 50%F 2.88 Figure A.27 BOC to EOCLB3 Power-Dependent MCPR Limits for ATRIUM-lU0 Fuel -TSSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-29 3.5 3.0 0-2.5 F-O FWCF* LRNB* PRFDS* TTNB x CRWE* LFWH 0+0+0 0+0+0 x 0 0 0 0 0 2.0 1.5 1.0 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.48 90.0 1.50 50.0 1.64 50.0 2.05 26.0 2.41 26.0 > 50%F 3.01 23.0 > 50%F -3.17 26.05 50%F 2.62 23.05S 50%F 2.84 Figure A.28 BOC to EOCLB3 Power-Dependent MCPR Limits for GE14 Fuel -TSSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-30 3.0 2.5 1--0_ 2.0 0-)O FWCF 0 LRNB A~ PRFDS+ TTNB 0 x ORWE 0+0 0 LFWH 0 +0+ 0+ 0 0 0 0 0 x 0 x x x 00000 0 0 1.5 l-1.01 C 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.47 90.0 1.48 50.0 1.62 50.0 2.06 26.0 2.41 26.0 > 50%F 2.60 23.0 > 50%F 2.73 26.0:5 50%F 2.47 23.05~ 50%F 2.55 Figure A.29 BOC to EOCLB3 Power-Dependent MCPR Limits for ATRIUM-lO0 Fuel -TSSS Insertion Times -FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-31 3.0 2.5:t E~0j C0 2.0 F-O3 FWCF 0 LRNB A PRFDS+ TTNB x CRWE 0 0 LFWH 0 0 0 x 00 x x X x 0a 00 0 0 0 0 0 1.5 I.1.0 i i i i 0 10 20 30 40 50 60 Power (% Rated)70 s0 90 100 110 Power MCPRP (% of rated) Limit 100.0 1.47 90.0 1.49 50.0 1.62 50.0 2.05 26.0 2.41 26.0 > 50%F 2.63 23.0 > 50%F 2.76 26.05~ 50%F 2.45 23.05~ 50%F 2.54 Figure A.30 BOC to EOCLB3 Power-Dependent MCPR Limits for GE14 Fuel -TSSS Insertion Times -FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-32 3.5 3.0-j n~2.5 o FWCF o LRNB* PRFDS* TTNB x CRWE* LFWH 0 0+0+0 0++0+x 00 x 0 0 0 0 0 0 0 0 2.0 1-1.5 1.0 i i i i 0 10 20 30 40 50 60 70 60 90 100 Power (%~ Rated)110 Power MCPRP (% of rated) Limit 100.0 1.50 90.0 1.53 50.0 1.70 50.0 2.06 26.0 2.41 26.0 > 50%F 3.14 23.0 > 50%F 3.31 26.05~ 50%F 2.76 23.0:5 50%F 3.00 Figure A.31 BOC to EOCLB3 Power-Dependent MCPR Limits for ATRIUM-lO Fuel -TSSS Insertion Times -TBVOOS FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-33 3.5 3.0 L-)2.5 2.0 O FWCF 0 LRNB a PRFDS+ TTNB x CRWE 0 LFWH 0++0 0+0+0 9 0 9 Q 9 0 & 0 x 0 00 0 0 0 0 1.5 1.0 i i i i i i i i i i 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power MCPRP (% ofrated) Limit 100.0 1.54 90.0 1.55 50.0 1.69 50.0 2.05 26.0 2.41 26.0 > 50%F 3.13 23.0 > 50%F 3.30 26.05~ 50%F 2.72 23.05~ 50%F 2.95 Figure A.32 BOC to EOCLB3 Power-Dependent MCPR Limits for GE14 Fuel -TSSS Insertion Times -TBVOOS FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-34 3.0 2.5 1-2.0 a-)o FWCF* LRNB* PRFDS* TTNB o CRWE 0 0 LFWH 0++ 0 0 0 00 1.5 F-1.0 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRp (% of rated) Limit 100.0 1.45 90.0 1.45 50.0 1.57 50.0 2.02 26.0 2.39 26.0 > 50%F 2.60 23.0 > 50%F 2.73 26.0:5 50%F 2.47 23.0:5 50%F 2.55 Figure A.33 FFTRlCoastdown Power-Dependent MCPR Limits for ATRIUM-lO0 Fuel -NSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-35 3.0 2.5 I-a 2.0 o FWCF 0 LRNB* PRFDS* TTNB o x CRWE 0 LFWH+0 0+0 x 00x~ X X x 00 0 0 0 0 0 1.5 1--1.0 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.44 90.0 1.44 50.0 1.57 50.0 2.02 26.0 2.39 26.0 > 50%F 2.63 23.0 > 50%F 2.76 26.0:5 50%F 2.45 23.05~ 50%F 2.54 Figure A.34 FFTRlCoastdown Power-Dependent MCPR Limits for GE14 Fuel -NSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-36 3.5 3.0-J 0~0~C-)2.5 2.0 o3 FWCF* LRNB* PRFDS+ TTNB x CRWE* LFWH 0 0+0+0 00 0 00 0 0 0 0 0 0 0 0 1.5 1.0 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.46 90.0 1.48 50.0 1.70 50.0 2.02 26.0 2.39 26.0 > 50%F 3.14 23.0 > 50%F 3.31 26.05~ 50%F 2.76 23.05~ 50%F 3.00 Figure A.35 FFTRlCoastdown Power-Dependent MCPR Limits for ATRIUM-10 Fuel -NSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-37 3.5 3.0_j 0-2.5 2.0 1.5 1.0 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)Power MCPRP (% of rated) Limit 100.0 1.48 90.0 1.51 50.0 1.65 50.0 2.02 26.0 2.39 26.0 > 50%F 3.13 23.0 > 50%F 3.30 26.0:5 50%F 2.72 23.05~ 50%F 2.95 Figure A.36 FFTRlCoastdown Power-Dependent MCPR Limits for GE14 Fuel -NSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page A-38 3.0 2.5 I-C,2.0 C-)O FWCF 0 LRNB A PRFDS+ TTNB 0 x ORWE 0 0 0 LFWH 0 0 ++ 0 0 00 A~ x 0 0 0 0 0 0 0 II I I I I I 1.5 F--1.0 0 10 20 30 40 50 60 70 80 90 100 Power (%~ Rated)110 Power MCPRP (% of rated) Limit 100.0 1.48 90.0 1.48 50.0 1.62 50.0 2.06 26.0 2.41 26.0 > 50%F 2.60 23.0 > 50%F 2.73 26.05~ 50%F 2.47 23.0:5 50%F 2.55 Figure A.37 FFTRlCoastdown Power-Dependent MCPR Limits for ATRIUM-10 Fuel -TSSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-39 3.0 2.5 1-2.0 0 O3 FWCF 0 LRNB A PRFDS 0+ TTNB x CRWE+FW 0+ 0' L W 0 0 x : 0 x X x 0 0 0 1.5 I-1.0 0 10 20 30 40 50 60 Power (% Rated)70 80 .90 100 110 Power MCPRP (% of rated) Limit 100.0 1.47 90.0 1.49 50.0 1.62 50.0 2.05 26.0 2.41 26.0 > 50%F 2.63 23.0 > 50%F 2.76 26.0:5 50%F 2.45 23.05~ 50%F 2.54 Figure A.38 FFTRlCoastdown Power-Dependent MCPR Limits for GE14 Fuel -TSSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis AN P-2674(N P)Revision 0 Page A-40 3.5 3.0-j a-2.5 2.0 O FWCF* LRNB* PRFDS* TTNB x ORWE 0 LFWH 0 0 0+0+ 0 x 00 0 0 00 0 0 0 0 0 0 1.5 1.0 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power MCPRP (% of rated) Limit 100.0 1.50 90.0 1.53 50.0 1.70 50.0 2.06 26.0 2.41 26.0 > 50%F 3.14 23.0 > 50%F 3.31 26.0:550%F 2.76 23.0:5 50%F 3.00 Figure A.39 FFTRlCoastdown Power-Dependent MCPR Limits for ATRIUM-10 Fuel -TSSS Insertion Times -T13VOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-41 3.5 3.0 a-0~C-)2.5 2.0 1.5 1.0 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)Power MCPRP (% of rated) Limit 100.0 1.54 90.0 1.55 50.0 1.69 50.0 2.05 26.0 2.41 26.0 > 50%F 3.13 23.0 > 50%F 3.30 26.05~ 50%F 2.72 23.0:5 50%F 2.95 Figure A.40 FFTRlCoastdown Power-Dependent MCPR Limits for GE14 Fuel -TSSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-42 1.4 1.2-j C-,_j~1.0.8 0 FWCF 0 LRNB A PRFDS+ TTN13 0 0 0 n 0.6.4 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power LHGRFACp (% of rated) Limit 100.0 1.00 80.0 1.00 50.0 0.94 50.0 0.67 26.0 0.57 26.0 > 50%F 0.56 23.0 > 50%F 0.54 26.05~ 50%F 0.63 23.0:5 50%F 0.62 Figure A.41 BOC to EQCLB3 Power-Dependent LHGRFAC Multipliers for ATRIUM-10 Fuel -NSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-43 1.4 1.2 C-,_j 1.0 o3 FWCF 0 LRNB A PRFDS+ TTNB 00 00+0+0 0.8 I-.6.4 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power LHGRFACp (% of rated) Limit 100.0 0.99 80.0 0.96 50.0 0.87 50.0 0.67 26.0 0.57 26.0 > 50%F 0.46 23.0 > 50%F 0.44 26.05~ 50%F 0.58 23.0:5 50%F 0.54 Figure A.42 BOC to EOCLB Power-Dependent LHGRFAC Multipliers for ATRIUM-10 Fuel -NSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-44 1.4 1.2_j 1.0.8.6.4 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)Power LHGRFACp (% of rated) Limit 100.0 1.00 80.0 0.99 50.0 0.91 50.0 0.67 26.0 0.57 26.0 > 50%F 0.53 23.0 > 50%F 0.51 26.0 5 50%F 0.60 23.05 50%F 0.59 Figure A.43 BOC to EOCLB Power-Dependent LHGRFAC Multipliers for ATRIUM-1 Fuel -NSS Insertion Times -FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-45 1.4 1.2 0~0-J 1.0.8 O3 FWCF 0 LRNB A PRFOS+ TTNB 00 0 0 0 E.6.4 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power LHGRFACp (% of rated) Limit 100.0 0.99 80.0 0.95 50.0 0.86 50.0 0.67 26.0 0.57 26.0 > 50%F 0.44 23.0 > 50%F 0.42 26.0:5 50%F 0.55 23.0:5 50%F 0.50 Figure A.44 BOC to EOCLB3 Power-Dependent LHGRFAC Multipliers for ATRIUM-lO0 Fuel -NSS Insertion Times -TBVOOS FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-46 1.4 1.2 0~C-)0 1.0 O FWCF 0 LRNB A PRFDSA+ TTNB 7 +I II II I0.8.6.4 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)Power LHGRFACp (% ofrated) Limit 100.0 1.00 80.0 0.94 50.0 0.93 50.0 0.65 26.0 0.57 26.0 > 50%F 0.56 23.0 > 50%F 0.54 26.05~ 50%F 0.63 23.05~ 50%F 0.62 Figure A.45 BOC to EOCLB Power-Dependent LHGRFAC Multipliers for ATRIUM-lO0 Fuel -TSSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-47 1.4 1.2 0~1.0.8 0 FWCF 0 LRNB A PRFDS+ TTNB 0 0 000 0 0 0+I~ ~~1 0.6.4 0 10 20 30 40 50 60 70 80 Power (% Rated)90 100 110 Power LHGRFACp (% of rated) Limit 100.0 0.95 80.0 0.93 50.0 0.87 50.0 0.65 26.0 0.57 26.0 > 50%F 0.46 23.0 > 50%F 0.44 26.0:5 50%F 0.58 23.0:5 50%F 0.54 Figure A.46 BOC to EOCLB Power-Dependent LHGRFAC Multipliers for ATRIUM-10 Fuel -TSSS Insertion Times -T13VOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-48 1.4 1.2 01.0.8 o FWCF 0 LRNB A PRFDS+ TTNB 0 0 0 0 0 0 0 03.6.4 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power LHGRFACp (% of rated) Limit 100.0 0.95 80.0 0.94 50.0 0.90 50.0 0.65 26.0 0.57 26.0 > 50%F 0.53 23.0 > 50%F 0.51 26.0:5 50%F 0.60 23.05~ 50%F 0.59 Figure A.47 BOC to EOCLB Power-Dependent LHGRFAC Multipliers for ATRIUM-11 Fuel -TSSS Insertion Times -FHOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-49 1.4 1.2 I IIIII o FWCF o LRNB A PRFDS+ TTNB-j 0 r I 1.0.8 0 0 0 0 0 0 0 0 03 0.6.4 I I I I I I I I I I 0 10 20 30 40 50 60 70 80 90 100 Power (%~ Rated)110 Power LHGRFACP (% of rated) Limit 100.0 0.95 80.0 0.93 50.0 0.85 50.0 0.65 26.0 0.57 26.0 > 50%F 0.44 23.0 > 50%F 0.42 26.0:550%F 0.55 23.0:5 50%F 0.50 Figure A.48 BOC to EOCLB Power-Dependent LHGRFAC Multipliers for ATRIUM-lO0 Fuel -TSSS Insertion Times -TBVOOS FHOOS AREVA NP Inc.

ANP-2674(NP)

Brunswick Unit 1 Cycle 17 Revision 0 Reload Safety Analysis Page A-50 1.4 1.2 0~C-)_j 1.0.8 o3 FWCF 0 LRNB A PRFDS+ TTNBA 0 0 00 0 0++0 0.6.4 i i i j i i 0 10 20 30 40 50 60 Power (% Rated)70 80 90 100 110 Power LHGRFACP (% of rated) Limit 100.0 1.00 80.0 0.97 50.0 0.91 50.0 0.67 26.0 0.57 26.0 > 50%F 0.53 23.0 > 50%F 0.51 26.0:5 50%F 0.60 23.0:5 50%F 0.59 Figure A.49 FFTRiCoastdown Power-Dependent LHGRFAC Multipliers for ATRIUM-1 Fuel -NSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-51 1.4 1.2 a-C.)_j 1.0.8.6.4 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)Power LHGRFACP (% of rated) Limit 100.0 0.97 80.0 0.92 50.0 0.86 50.0 0.67 26.0 0.57 26.0 > 50%F 0.44 23.0 > 50%F 0.42 26.05~ 50%F 0.55 23.05 50%F 0.50 Figure A.50 FFTRlCoastdown Power-Dependent LHGRFAC Multipliers for ATRIUM-lO0 Fuel -NSS Insertion Times -TBVOOS AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-52 1.4 1.2 0~LL_1.0.8.6.4 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)Power LHGRFACp (% of rated) Limit 100.0 1.00 80.0 0.94 50.0 0.89 50.0 0.65 26.0 0.57 26.0 > 50%F 0.53 23.0 > 50%F 0.51 26.05~ 50%F 0.60 23.05 50%F 0.59 Figure A.51 FFTRlCoastdown Power-Dependent LHGRFAC Multipliers for ATRIUM-1 Fuel -TSSS Insertion Times -Base Case AREVA NP Inc.

Brunswick Unit 1 Cycle 17 Reload Safety Analysis ANP-2674(NP)

Revision 0 Page A-53 1.4 1.2-j a C-)W 0 I-J 1.0 0 FWGF 0 LRNB A PRFDS+ TTNBA 00 00++0 00 0.8.6.4.I I I I I I I I 0 10 20 30 40 50 60 70 80 90 100 Power (% Rated)110 Power LHGRFACP (% of rated) Limit 100.0 0.95 80.0 0.91 50.0 0.85 50.0 0.65 26.0 0.57 26.0 > 50%F 0.44 23.0 > 50%F 0.42 26.0 5 50%F 0.55 23.0:5 50%F 0.50 Figure A.52 FFTRlCoastdown Power-Dependent LHGRFAC Multipliers for ATRIUM-10 Fuel -TSSS Insertion Times -TBVOOS AREVA NP Inc.