BSEP 09-0034, Enclosure 10 to BSEP 09-0034 - ANP-2771(NP), Rev. 0, Brunswick, Unit 2 Cycle 19 Reload Safety Analysis, Dated January 2009

Enclosure 10 to BSEP 09-0034 - ANP-2771(NP), Rev. 0, Brunswick, Unit 2 Cycle 19 Reload Safety Analysis, Dated January 2009

ML090970249
Person / Time
Site:	Brunswick
Issue date:	01/31/2009
From:	AREVA, AREVA NP
To:	Office of Nuclear Reactor Regulation
References
BSEP 09-0034 ANP-2771(NP), Rev 0
Download: ML090970249 (97)

Text

BSEP 09-0034 Enclosure 10 ANP-2771 (NP), Revision 0, Brunswick Unit 2 Cycle] 9 Reload Safety Analysis, dated January 2009 y

0ANP-2771(NP)

Revision 0 Brunswick Unit 2 Cycle 19 Reload Safety Analysis January 2009 SARE AVA

AREVA NP Inc.

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Revision 0 Brunswick Unit 2 Cycle 19 Reload Safety Analysis

AREVA NP Inc.

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Revision 0 Copyright © 2009 AREVA NP Inc.

All Rights Reserved paj

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Pagei Nature of Changes Item Page Description and Justification

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

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page ii Contents 1.0 Introduction .................................................................................................................... 1-1 2.0 Disposition of Events ....................................................................................................... 2-1 2.1 Variable Frequency Drive Im plem entation ......................................................... 2-1 2.1.1 VFD Im pact on the Disposition of Events ............................................. 2-1 2.1.2 VFD Im pact on Cycle 19 Licensing Analyses ....................................... 2-2 3.0 Mechanical Design Analysis .......................................................................................... 3-1 4.0 Therm al-Hydraulic Design Analysis ............................................................................... 4-1 4.1 Therm al-Hydraulic Design and Com patibility ..................................................... 4-1 4.2 Safety Lim it MCPR Analysis ............................................................................... 4-1 4.3 Core Hydrodynam ic 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 (FW CF) .............................................  :.... 5-4 5.1.4 Pressure Regulator Failure Dow nscale (PRFDS) ................................ 5-5 5.1.5 Loss of Feedwater Heating .................................................................. 5-5 5.1.6 Control Rod W ithdrawal Error .............................................................. 5-6 5.2 Slow Flow Runup Analysis ................................................................................. 5-6 5.3 Equipm ent O ut-of-Service Scenarios ................................................................. 5-8 5.3.1 FHO OS ................................................................................................. 5-8 5.3.2 TBVO OS .............................................................................................. 5-8 5.3.3 Com bined FHO OS and TBVOO S ........................................................ 5-9 5.3.4 One SRVO OS ...................................................................................... 5-9 5.3.5 One MSIVO OS .................................................................................... 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 Equipm ent Handling Accident ............................................................. 6-2 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 ASM E Overpressurization Analysis .................................................................... 7-1 7.2 ATW S Event Evaluation ..................................................................................... 7-1 7.2.1 ATW S Overpressurization Analysis ..................................................... 7-1 7.2.2 Long-Term Evaluation .......................................................................... 7-2 7.3 Standby Liquid Control System .......................................................................... 7-3 7.4 Fuel Criticality ..................................................................................................... 7-3 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page iii 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 MELLLA+ O peration .................................................................................. A-1 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page iv Tables 1.1 E O O S O perating C onditions ......................................................................................... 1-2 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses ....................... 4-3 4.2 Results Summary for Safety Limit MCPR Analyses ....................................................... 4-4 4 .3 O P R M S etpoints ............................................................................................................ 4-5 4.4 BSP Endpoints for Brunswick Unit 2 Cycle 19 ............................................................... 4-6 5.1 Exposure Basis for Brunswick Unit 2 Cycle 19 Transient Analysis .............................. 5-11 5.2 Scram S peed Insertion Tim es ...................................................................................... 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 EOCLB Base Case TTNB Transient Results ............................................................... 5-16 5.7 NEOC Base Case FWCF Transient Results ................................................................ 5-17 5.8 EOCLB 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 R BM O perability R equirem ents .................................................................................... 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 ................................................................... 7-4 7.2 ATWS Overpressurization Analysis Results .................................................................. 7-5 8.1 MCPRp Limits for NSS Insertion Times BOC to < NEOC ............................................... 8-4 8.2 MCPRP Limits for TSSS Insertion Times BOC to < NEOC ............................................. 8-5 8.3 MCPRp Limits for NSS Insertion Times BOC to < EOCLB ............................................. 8-6 8.4 MCPRP Limits for TSSS Insertion Times BOC to < EOCLB ........................................... 8-7 8.5 MCPRP Limits for NSS Insertion Times FFTR/Coastdown ............................................. 8-8 8.6 MCPRP Limits for TSSS Insertion Times FFTR/Coastdown ........................................... 8-9 8.7 Flow-Dependent MCPR Limits ATRIUM-10 and GE14 Fuel ................. I.......................8-10 8.8 ATRIUM-10 Steady-State LHGR Limits ...................................................................... 8-11 8.9 ATRIUM-10 LHGRFACP Multipliers for NSS Insertion Times BOC to < EOCLB ......... 8-12 8.10 ATRIUM-10 LHGRFACp Multipliers for TSSS Insertion Times BOC to < EOCLB ....... 8-13 8.11 ATRIUM-10 LHGRFACP Multipliers for NSS Insertion Times FFTR/Coastdown ......... 8-14 8.12 ATRIUM-10 LHGRFACP Multipliers for TSSS Insertion Times FFTR/Coastdown ....... 8-15 8.13 ATRIUM-10 LHGRFACf Multipliers All Cycle 19 Exposures ........................................ 8-16 8.14 ATR IU M-10 M A PLHG R Lim its ..................................................................................... 8-17 8.15 GE14 MAPFACp Multipliers for NSS and TSSS Insertion Times All Cycle 19 Ex p os u re s .................................................................................................................... 8-1 8 8.16-- GE14 MAPFACf Multipliers All Cycle 19 Exposures .................................................... 8-19 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page v Figures 1.1 Brunsw ick U nit 2 Pow er/Flow Map ................................................................................. 1-3 5.1 EOCLB LRNB at 1OOP/1 04.5F - TSSS Key Parameters ............................................. 5-24 5.2 EOCLB LRNB at 1OOP/1 04.5F - TSSS Sensed Water Level ...................................... 5-25 5.3 EOCLB LRNB at 100P/104.5F - TSSS Vessel Pressures .......................................... 5-26 5.4 EOCLB FWCF at 100P/104.5F - TSSS Key Parameters ............... 5-27 5.5 EOCLB FWCF at 1OOP/1 04.5F - TSSS Sensed Water Level ..................................... 5-28 5.6 EOCLB FWCF at 1OOP/1 04.5F - TSSS Vessel Pressures .......................................... 5-29 7.1 MSIV Closure Overpressurization Event at 102P/104.5F - Key Parameters ................ 7-6 7.2 MSIV Closure Overpressurization Event at 102P/104.5F - Sensed Water Level .......... 7-7 7.3 MSIV Closure Overpressurization Event at 102P/104.5F - Vessel Pressures .............. 7-8 7.4 MSIV Closure Overpressurization Event at 102P/104.5F - Safety/Relief Valve F low R a tes ..................................................................................................................... 7-9 7.5 PRFO ATWS Overpressurization Event at 100P/99F - Key Parameters .................... 7-10 7.6 PRFO ATWS Overpressurization Event at 100P/99F - Sensed Water Level .............. 7-11 7.7 PRFO ATWS Overpressurization Event at 1OOP/99F - Vessel Pressures .................. 7-12 7.8 PRFO ATWS Overpressurization Event at 10OP/99F - Safety/Relief Valve F low R a te s ................................................................................................................... 7 -1 3 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page vi Nomenclature AOO anticipated operational occurrence ARO all control rods out ASME American Society of Mechanical Engineers AST alternative source term ATWS anticipated transient without scram ATWS-RPT anticipated transient without scram recirculation pump trip BOC beginning-of-cycle BPWS banked position withdrawal sequence BSEP Brunswick Steam Electric Plant BSP backup stability protection BWROG Boiling Water Reactor Owners Group CFR Code of Federal Regulations COLR core operating limits report CPR critical power ratio CRDA control rod drop accident CRWE control rod withdrawal error EFPD effective full-power days EFPH effective full-power hours EOC end-of-cycle EOCLB end-of-cycle licensing basis EOFP end of full power EOOS equipment out-of-service FFTR final feedwater temperature reduction FHOOS feedwater heaters out-of-service FWCF feedwater controller failure GE General Electric GNF Global Nuclear Fuels HCOM hot channel oscillation magnitude HFR heat flux ratio ICF increased core flow LFWH loss of feedwater heating LHGR linear heat generation rate LHGRFACf flow-dependent linear heat generation rate multipliers LHGRFACP power-dependent linear heat generation rate multipliers LOCA loss-of-coolant accident LPRM local power range monitor LRNB generator load rejection with no bypass AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page vii Nomenclature (Continued)

MAPFACf flow-dependent maximum average planar linear heat generation rate multipliers MAPFACP power-dependent maximum average planar linear heat generation rate multipliers MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio MCPRf flow-dependent minimum critical power ratio MCPRp power-dependent minimum critical power ratio MELLLA maximum extended load line limit analysis MG motor generator MSIV main steam isolation valve MSIVOOS main steam isolation valve out-of-service NEOC near end-of-cycle NSS nominal scram speed NRC Nuclear Regulatory Commission, U.S.

OLMCPR operating limit minimum critical power ratio OPRM oscillation power range monitor Pbypass power below which direct scram on TSV/TCV 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 RHR residual heat removal RPT recirculation pump trip SLC standby liquid control SLMCPR safety limit minimum critical power ratio SLO single-loop operation SRV safety/relief valve SRVOOS 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 TTNB turbine trip with no bypass UAT unit auxiliary transformer UFSAR updated final safety analysis report VFD variable frequency drive ACPR change in critical power ratio AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 1-1 1.0 Introduction This report presents the results of the reload licensing analyses performed by AREVA NP* in support of Brunswick Unit 2 Cycle 19. 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 19 core consists of a total of 560 fuel assemblies, including 238 fresh ATRIUMTM-10t assemblies and 322 irradiated GE14 assemblies. The licensing analysis supports the core design presented in Reference 1.

The Cycle 19 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 in Reference 2 and augmented by Reference 32 . The results of the reload licensing analysis support operation in the MELLLA region of the power/flow map presented in Figure 1.1 and also support operation with the equipment out-of-service (EOOS) scenarios presented in Table 1.1. A discussion of the analyses for the MELLLA+ region is presented in Appendix A.

• AREVA NP Inc. is an AREVA and Siemens company.

t ATRIUM is a trademark of AREVA NP.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 1-2 Table 1.1 EOOS Operating Conditions*

Single-loop operation (SLO)

Turbine bypass valves out-of-service (TBVOOS)

Feedwater heaters out-of-service (FHOOS)

One safety relief valve out-of-service (SRVOOS)

One main steam isolation valve out-of-servicet (MSIVOOS)

One pressure regulator out of service$

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.

t Operation with one MSIVOOS is only supported at power levels less than 70% of rated.

Operation with one pressure regulator out of service is only supported at power levels greater than 90% of rated and less than 50% of rated.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 1-3 120.0 110.0 100.0 90.0 80.0 70.0 o 60.0 0.

50.0 40.0 30.0 20.0 10.0 0.0 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 2 PowerlFlow Map AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 2-1 2.0 Disposition of Events A disposition of events to identify the limiting events which need to be analyzed to support operation at the Brunswick Steam Electric Plant (BSEP) was performed for the introduction of ATRIUM-10 fuel. Events and analyses identified as potentially limiting were either evaluated generically for the introduction of ATRIUM-10 fuel or are performed on a cycle-specific bases.

The results of the disposition of events are presented in Reference 3. The Reference 4 calculation plan for the Brunswick Unit 2 Cycle 19 reload licensing analyses was based on the disposition of events.

The parameter differences between those used in the initial Brunswick ATRIUM-10 licensing analyses and the planned analyses for the Brunswick Unit 2 Cycle 19 reload were reviewed to determine if the conclusions remain applicable. The review concluded that analyses affected by the differences were included in the Reference 4 calculation plan.

2.1 Variable FrequencyDrive Implementation Progress Energy is planning on replacing the recirculation pump motor generator (MG) sets with variable frequency drive (VFD) systems. While a firm implementation schedule for the VFDs is not yet available, Progress Energy has indicated that the licensing analyses for Brunswick Unit 2 Cycle 19 should support operation with either the MG sets or the VFDs. The disposition of events was reviewed to evaluate the impact of the VFD implementation. In addition, a review of the Brunswick Unit 2 Cycle 19 licensing analyses was performed to ensure that the analyses remain applicable and/or bounding for operation with the VFDs. Reference 33 provides the VFD parameters used in the disposition of events evaluation.

2.1.1 VFD Impact on the Disposition of Events The AOOs, accident and other fuel design and fuel related events and analyses were reviewed to determine if any event or analysis that was previously identified as non-limiting has the potential to become limiting with the VFD installation. Any new potentially limiting event would need to be evaluated to ensure that appropriate operating limits are established. The review concluded that while some of the events will be affected by the VFD implementation, in most cases the relative severity of the events will not change. The two events which were previously identified as either non-limiting or needing to be addressed for the initial reload that required additional evaluation are the trip of two recirculation pumps and the turbine trip without bypass AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 2-2 (TTNB). Both of these events were analyzed using AREVA's approved transient analysis methodology to determine if they could become limiting. The results are discussed below.

Trip of Two Recirculation Pumps. With the VFDs, the effective recirculation pump inertia decreases which results in a faster. pump coastdown and a more severe event. While the event becomes more severe, analysis results show that the event remains a benign event and no further analysis is required.

Turbine Trip No Bypass. Analyses performed with the MG driven recirculation pumps showed that the consequences of the TTNB event are bound by the generator load rejection without bypass (LRNB) event at all power levels. With the VFD implementation, there will no longer be a recirculation pump overspeed during a LRNB event so the consequences will become slightly less severe and may no longer bound the consequences of the TTNB event at all power levels. Analysis results demonstrate that with the VFD implementation, the consequences of the LRNB event will no longer bound those of the TTNB event at rated power and below Pbypass. Therefore, with the VFD implementation, the TTNB event should be considered a potentially limiting event at high powers and below Pbypass.

It is also noted that the lower effective recirculation pump inertia with the VFD implementation will result in a faster recirculation pump coastdown during a LOCA. While the LOCA is identified as a potentially limiting event, only the heatup portion of the analysis is addressed each reload.

For a given plant, the system analysis is a one-time analysis performed for each fuel design.

The impact of the VFD implementation on the system analysis should be reviewed.

2.1.2 VFD Impact on Cycle 19 Licensing Analyses The Brunswick Unit 2 Cycle 19 licensing analyses presented in this report are based on operation with the MG sets. As a result, the LRNB analysis results include the impact of the recirculation pump overspeed and are conservative for operation with the VFDs. Cycle 19 analysis results also show that the consequences of the LRNB event (with the pump overspeed) bound those of the TTNB event. Cycle 19 analyses for the trip of two recirculation pumps with the lower effective recirculation pump inertia associated with the VFDs show that the event remains non-limiting.

As noted above, the faster recirculation pump coastdown with the VFDs will impact the LOCA analyses. The ATRIUM-10 LOCA analysis presented in References 24 and 25 conservatively did not include the inertia of the MG sets. Therefore the recirculation pump inertia used is consistent with the VFDs. The results therefore support operation with either the MG sets or the VFDs. It is recommended that the current GE14 LOCA analysis be evaluated for continued applicability since the VFDs can make the event more severe.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 2-3 The disposition and/or analysis for the other events are either unaffected by the VFD implementation or the evaluation performed assuming operation with the MG sets remains applicable and/or bounding.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 3-1 3.0 Mechanical Design Analysis The mechanical design analysis is presented in the applicable mechanical design report (Reference 5). 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-10 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 2 license amendment 153 (Reference 6).

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 and ATRIUM-10 fuel assemblies remain within licensed burnup limits.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 4-1 4.0 Thermal-Hydraulic Design Analysis 4.1 Thermal-HydraulicDesign'andCompatibility The results of the thermal-hydraulic characterization and compatibility analyses are presented in the thermal-hydraulic design report (Reference 7). The analysis results demonstrate that the thermal-hydraulic design and compatibility criteria are satisfied for the Brunswick Unit 2 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 (AOO). The SLMCPR for all fuel in the Brunswick Unit 2 Cycle 19 core was determined using the methodology described in Reference 8. 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 SLMCPR during an AOO.

The Brunswick Unit 2 Cycle 19 SLMCPR analysis used the SPCB critical power correlation additive constants and additive constant uncertainty for ATRIUM-10 fuel described in References 9 and 10. The SPCB additive constants and additive constant uncertainty for the coresident GE14 fuel were developed using the indirect approach described in Reference 11.

The determination of the SLMCPR explicitly includes the effects of channel bow relying on the following assumptions: Cycle 19 will not contain fuel channels used for more than one fuel bundle lifetime, and the average assembly burnup in Cycle 19 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 SLMCPR 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'.11 and a single-loop operation (SLO) SLMCPR of 1.12. The Cycle 19 MCPR operating limits are based on SLMCPR AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 4-2 values of 1.11 for TLO and 1.13 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 Hydrodynamic Stability Brunswick has implemented BWROG Long Term Stability Solution Option III (Oscillation Power Range Monitor-OPRM). Reload validation has been performed in accordance with Reference

12. 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 19 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 FHOOS and FFTR) operation.

AREVA 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 in accordance with Reference 13. This code is a coupled neutronic-thermal-hydraulic 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. The stability-based OLMCPRs are calculated using the most limiting of the calculated change in relative ACPR for a given oscillation magnitude or the generic value provided in Reference 12.

The generic value was determined to be limiting for Cycle 19.

In cases where the OPRM system is declared inoperable for Brunswick Unit 2 Cycle 19, Backup Stability Protection (BSP) in accordance with Reference 14 is provided. BSP curves have been evaluated using STAIF (Reference 15) to determine endpoints that meet decay ratio criteria for the BSP Base Minimal Region I (scram region) and Base Minimal Region II (controlled entry region). Stability boundaries based on these endpoints are then determined using the generic shape generating function from Reference 14. Analyses have been performed to support operation with nominal feedwater temperature conditions and reduced feedwater temperature conditions (both FFTR and FHOOS). The endpoints for the BSP regions are provided in Table 4.4 and are the same as the regions presented in Reference 3.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 4-3 Table 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses Parameter Uncertainty Fuel-Related Uncertainties Plant-RelatedUncertainties Feedwater flow rate 1.8%

Feedwater temperature 0.8%

Core pressure 0.8%

Total core flow rate TLO 2.5%

SLO 6%

• [

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 4-4 Table 4.2 Results Summary for Safety Limit MCPR Analyses Percentage SLMCPR* of Rods in Boiling Transition TLO - 1.11 0.089 SLO - 1.12 0.057

• Note that the Cycle 19 MCPR operating limits are based on SLMPCR values of 1.11 for TLO and 1.13 for SLO, the Unit 2 values currently in the plant Technical Specifications.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 4-5 Table 4.3 OPRM Setpoints OPRM OLMCPR OLMCPR OPRM OLMCPR OLMCPR Setpoint (2PT) 1.05 1.20 1.18 1.06 1.22 1.20 1.07 1.24 1.21 1.08 1.26 1.23 1.09 1.28 1.25 1.10 1.30 1.27 1.11 1.32 1.29 1.12 1.34 1.31 1.13 1.36 1.33 1.14 1.38 1.35 1.15 1.40 1.38 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.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 4-6 Table 4.4 BSP Endpoints for Brunswick Unit 2 Cycle 19 Feedwater Temperature Operation End Point Power Flow Mode Region Designation (% rated) (% rated)

Nominal Scram IA 56.6 40.0 Nominal Scram lB 40.7 31.0 Nominal Controlled 11A 64.5 50.0 entry Nominal Controlled lIB 28.5 31.0 entry FFTR/ Scram IA 64.9 50.5 FHOOS FFTR/ Scram lB 37.3 31.0 FHOOS FFTR/ Controlled IIA 66.1 52.0 FHOOS entry FFTR/ Controlled 1iB 28.5 31.0 FHOOS entry AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 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 2 Cycle 19.

COTRANSA2 (Reference 16), XCOBRA-T (Reference 17), XCOBRA (Reference 18), and CASMO-4/MICROBURN-B2 (Reference 19) are the major codes used in the thermal limits analyses as described in the AREVA THERMEX methodology report (Reference 18) and neutronics methodology report (Reference 19). 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 (References 9 and 10) 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 11. Fuel pellet-to-cladding gap conductance values are based on RODEX2 (Reference 20) calculations for the Brunswick Unit 2 Cycle 19 core.

5.1 System Transients The reactor plant parameters for the system transient analyses are presented in Reference 2.

Analyses have been performed to determine power-dependent MCPR limits that protect operation in the MELLLA region of the power/flow domain.

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 2 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 16,300 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.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-2 (18,760 MWd/MTU). Analyses were also performed to support extended cycle operation with final feedwater temperature reduction (FFTR) and power coastdown. The Brunswick Unit 2 Cycle 19 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 2 turbine bypass system includes 10 bypass valves. However, for base case analyses in which credit is taken for turbine bypass operation, only 8 of the turbine bypass valves are assumed operable.

Reductions in feedwater temperature of less than 10°F from the nominal feedwater temperature are considered base case operation, not an EOOS 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 core 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. LRNB events for base case operation were evaluated for both nominal and 10°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 110.3°F at rated power. Analyses were performed to support both nominal and constant rated dome pressure with combined FFTR/Coastdown operation to a cycle exposure of 20,655 MWd/MTU. The FWCF analyses were performed with the lowest feedwater temperature associated with the initial power level.

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.

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Brunswick Unit 2 Cycle 19 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 2 Technical Specifications (Reference 21) 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 performed. At 26% power (Pbypass), FWCF analyses were performed both with and without bypass of the direct scram function which can result 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 LRNB analyses assume that the power load unbalance (PLU) is inoperable. With the PLU inoperable, the LRNB 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. Given that there is no direct scram when the PLU is inoperable, the above and below Pbypass results at 26% power will be identical.

During an LRNB event with the recirculation pump power supplied by the Unit Auxiliary Transformer (UAT) the recirculation pump speed increases causing an increase in core flow and a corresponding increase in power. The result is a slightly more severe event. All LRNB analyses were performed assuming the UAT supplies power to the recirculation pumps.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-4 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 LRNB transient analysis 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.

Results presented in Reference 3 demonstrate that the consequences of the TTNB event are bound by those of the LRNB event. Analyses were performed to demonstrate that the LRNB is also bounding for Unit 2. TTNB analyses were performed for power/flow conditions at 100%

power and below Pbypass. Tables 5.5 and 5.6 present the base case TTNB transient analysis results for both TSSS and NSS insertion times for Cycle 19.

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 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. Eight of the ten installed turbine bypass valves are assumed operable AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-5 and provide 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 (PRFDS)

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 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 that AREVA review the PRFDS event with one pressure regulator out-of-service to determine if it is bound by the LRNB event at power levels greater than 90% of rated and less than 50% or rated. Analysis results demonstrate that the LRNB is more limiting at power levels greater than 90% of rated. Since LRNB analyses assume the PLU is inoperable below 50% of rated power, the TCVs close in servo or control mode without a direct scram on fast closure. Therefore, the consequences of the PRFDS event with one pressure regulator out of service are no more severe than the LRNB event at power levels less than 50% of rated.

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 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 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-6 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 2 Cycle 19, a cycle-specific analysis was performed in accordance with the Reference 22 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 CRWE 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 CRWE results are shown in Table 5.10 for the analytical RBM high power setpoint values of 108% to 117%. An assumed RBM high power setpoint of 108% was used to develop the MCPRP limits consistent with the direction provided by Progress Energy in Reference 23. At all intermediate and lower power setpoint values, the MCPRP values for ATRIUM-10 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-10 and GE14 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 physically attainable by the equipment (107% of rated core flow). An uncontrolled increase in AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-7 flow creates the potential for a significant increase in core power and heat flux. Operation with One MSIVOOS causes a larger increase in pressure and power during the flow excursion which results in 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 19 MCPRf limits are applicable for all Cycle 19 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 19 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 LHGR/MAPLHGR multipliers for two-loop operation are applicable for SLO.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-8 5.3 Equipment Out-of-Service Scenarios The following equipment out-of-service (EOOS) scenarios are supported for Brunswick Unit 2 Cycle 19 operation:

• Feedwater heater out-of-service (FHOOS) - up to 11 0.3 0 F feedwater temperature reduction

• Turbine bypass valves out-of-service (TBVOOS)

• Combined FHOOS 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 analyses protect operation with the feedwater temperature reduction characteristic presented in Reference 2. This results in a feedwater temperature reduction of 110.3 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 events were analyzed to ensure that appropriate FHOOS operating limits are established.

5.3.2 TBVOOS For this EOOS scenario, operation with TBVOOS means that the fast opening capability of three 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.

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Brunswick Unit 2 Cycle 19 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 19 operation.

Operating limits for this combined EOOS 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 MSIVOOS 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 MSIVs are available. Since low steam line pressure drop is limiting for pressurization transients, the results of the pressurization events with all MSIVs in service bound the results with One MSIVOOS. 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 base case operation with all MSIVs in service remain applicable for operation with One MSIVOOS 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-Loop 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 0.02 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 0.02 increase in the base case MCPRp limits and a decrease in the MAPLHGR limit. The same situation is true for the EOOS scenarios. Adding 0.02 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.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-10 5.4 Licensing Power Shape The licensing axial power profile used by AREVA for the plant transient analyses bounds the projected end of full power axial power profile. The conservative licensing axial power profile generated at the EOCLB core average exposure of 32,881 MWd/MTU is given in Table 5.13.

Cycle 19 operation is considered to be in compliance when:

The normalized power generated in the bottom 7 nodes from the projected EOFP solution at the state conditions provided in Table 5.13 is greater than the normalized power generated in the bottom 7 nodes in the licensing basis axial power profile.

The projected EOFP condition occurs at a core average exposure less than or equal to EOCLB.

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-B2 code.

Compliance analyses must also be performed using MICROBURN-B2. 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.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-11 Table 5.1 Exposure Basis for Brunswick Unit 2 Cycle 19 Transient Analysis Cycle Core Exposure at Average End of Interval Exposure (MWd/MTU) (MWd/MTU)* Comments 0 14,121 Beginning of cycle 16,300 30,421 Break point for exposure-dependent MCPRp limits (NEOC) 18,760 32,881 Design basis rod patterns to EOFP + 14 EFPD (EOCLB) 20,655 34,776 Maximum licensing core exposure - including FFTR/

Coastdown

• Note that the limits presented in Tables 8.1 - 8.6 and Tables 8.9 - 8.12 are based on core average exposure.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis 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 0.322 36 1.080 0.862 26 1.830 1.422 6 3.350 2.593 0 (full-in) 3.806 2.944 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-13 Table 5.3 NEOC Base Case LRNB Transient Results ATRIUM-I 0 ATRIUM-10 ATRIUM-10 GE14 Power ACPR HFR ACPR TSSS Insertion Times 100 0.34 1.35 0.36 90 0.35 1.36 0.37 80 0.36 1.37 0.38 70 0.36 1.36 0.38 60 0.35 1.34 0.37 50 0.33 1.31 0.34 50 at > 65%F PLU inoper able 0.84 1.87 0.90 50 at < 65%F PLU inoperýable 0.62 1.66 0.66 26 at > 65%F PLU inoper;able 1.14 2.10 1.10 26 at < 65%F PLU inoperý able 0.88 1.92 0.89 26 at > 65%F below Pbypa. S 1.14 2.10 1.10 26 at < 65%F below Pbypa,. S 0.88 1.92 0.89 23 at > 65%F below Pbypa, S 1.19 2.14 1.14 23 at < 65%F below Pbypa. 0.96 1.98 0.96 NSS Insertion Times 100 0.24 1.26 0.26 90 0.27 1.29 0.29-80 0.29 1.31 0.31 70 0.30 1.31 0.32 60 0.30 1.30 0.32 50 0.28 1.27 0.29 50 at > 65%F PLU inoperuable 0.81 1.84 0.87 50 at < 65%F PLU inoperzable 0.60 1.63 0.64 26 at > 65%F PLU inoper, able 1.12 2.08 1.08 26 at < 65%F PLU inoperEable 0.87 1.89 0.88 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-14 Table 5.4 EOCLB Base Case LRNB Transient Results ATRIUM-10 ATRIUM-10 GE14 Power ACPR HFR ACPR TSSS Insertion Times 100 0.37 1.35 0.38 90 0.38 1.35 0.38 80 0.39 1.45 0.39 70 0.39 1.44 0.39 60 0.38 1.41 0.38 50 0.35 1.36 0.35 50 at > 65%F PLU inoper able 0.86 1.94 0.90 50 at < 65%F PLU inoper.able 0.64 1.69 0.66 26 at > 65%F PLU inoper able 1.14 2.10 1.10 26 at < 65%F PLU inoper able 0.88 1.96 0.89 26 at > 65%F below Pbypa,5S 1.14 2.10 1.10 26 at < 65%F below Pbypa,5S 0.88 1.96 0.89 23 at > 65%F below Pbypa 5 .1.19 2.14 1.14 23 at < 65%F below Pbypa,;s 0.96 2.00 0.96 NSS Insertion Times 100 0.32 1.35 0.33 90 0.34 1.35 0.34 80 0.36 1.35 0.36 70 0.36 1.41 0.36 60 0.36 1.39 0.36 50 0.33 1.34 0.33 50 at > 65%F PLU inoper able 0.82 1.90 0.87 50 at < 65%F PLU inoper able 0.61 1.67 0.64 26 at > 65%F PLU inoper able 1.12 2.08 1.08 26 at < 65%F PLU inoper.able 0.87 1.94 0.88 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-15 Table 5.5 NEOC Base Case TTNB Transient Results ATRIUM-10 ATRIUM-10 GE14 Power ACPR HFR ACPR TSSS Insertion Times 100 0.33 1.33 0.35 26 at > 65%F below Pbypass 1.00 1.89 0.98 26 at < 65%F below Pbypass 0.83 1.72 0.83 23 at > 65%F below Pbypass 1.05 1.93 1.02 23 at < 65%F below Pbypas 0.89 1.78 0.89 NSS Insertion Times 100 / 104.5 0.24 1.25 0.25 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-16 Table 5.6 EOCLB Base Case TTNB Transient Results ATRIUM-10 ATRIUM-10 GE14 Power ACPR HFR ACPR TSSS Insertion Times 100 0.36 1.35 0.36 26 at > 65%F below Pbypass 1.00 1.89 0.98 26 at < 65%F below Pbypass 0.83 1.75 0.83 23 at > 65%F below Pbypa,, 1.05 1.93 1.02 23 at < 65%F below Pbypass 0.89 1.79 0.89 NSS Insertion Times 100/ 104.5 0.31 1.35 0.32 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-17 Table 5.7 NEOC Base Case FWCF Transient Results ATRIUM-10 ATRIUM-10 GE14 Power ACPR HFR ACPR TSSS Insertion Times 100 0.21 1.21 0.22 90 0.23 1.23 0.24 80 0.25 1.25 0.26 70 0.28 1.27 0.29 60 0.30 1.29 0.32 50 0.33 1.32 0.35 26 0.51 1.50 0.53 26 at > 65%F below Pbypa SS 0.58 1.58 0.61 26 at < 65%F below Pbypa ss 0.37 1.35 0.39 23 at > 65%F below Pbypa SS 0.63 1.63 0.66 23 at < 65%F below Pbypa SS 0.41 1.37 0.43 NSS Insertion Times 100 0.15 1.15 0.15 90 0.17 1.18 0.18 80 0.20 1.21 0.21 70 0.23 1.23 0.24 60 0.27 1.26 0.28 50 0.31 1.30 0.32 26 0.47 1.49 0.49 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-18 Table 5.8 EOCLB Base Case FWCF Transient Results ATRIUM-i 0 ATRIUM-10 ATRIUM-10 GE14 Power ACPR HFR ACPR TSSS Insertion Times 100 0.24 1.26 0.24 90 0.25 1.27 0.24 80 0.26 1.29 0.26 70 0.28 1.31 0.29 60 0.30 1.33 0.32 50 0.33 1.34 0.35 26 0.51 1.50 0.53 26 at > 65%F below Pbypa *SS 0.58 1.58 0.61 26 at - 65%F below Pbypa SS 0.37 1.35 0.39 23 at > 65%F below Pbypa *ss 0.63 1.63 0.66 23 at < 65%F below Pbypa sS 0.41 1.37 0.43 NSS Insertion Times 100 0.20 1.22 0.20 90 0.21 1.24 0.21 80 0.23 1.26 0.23 70 0.25 1.29 0.25 60 0.27 1.31 0.28 50 0.31 1.33 0.32 26 0.47 1.49 0.49 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-19 Table 5.9 Loss of Feedwater Heating Transient Analysis Results Power ATRIUM-10/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 23 0.30 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-20 Table 5.10 Control Rod Withdrawal Error ACPR Results Analytical RBM Setpoint (without filter) ACPR*

108 0.18 111 0.22 114 0.27 117 0.33

• Results are for the most limiting of the ATRIUM-1 0 or GE14 fuel in the core.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-21 Table 5.11 RBM Operability Requirements Applicable Thermal Power ATRIUM-10/GE14

(% rated) MCPR 1.53 15 TLO L

_>29% and < 90% 1.55 SLO

> 90% 1.47 TLO AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 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 1.59 1.57 40 1.55 1.50 50 1.49 1.45 60 1.43 1.42 70 1.36 1.35 80 1.30 1.29 90 1.24 1.23 100 1.18 1.17 107 1.11 1.11 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-23 Table 5.13 Licensing Basis Core Average Axial Power Profile State Conditions for Power Shape Evaluation Power, MWt 2923.0 MICROBURN-B2 1044.5 pressure, psia Inlet subcooling, Btu/Ibm 20.61 Flow, Mlb/hr 80.46 Control state ARO Core average exposure 32,881 (EOCLB), MWd/MTU Licensing Axial Power Profile (Normalized)

Node Power Top 25 0.194 24 0.617 23 0.803 22 0.923 21 1.010 20 1.080 19 1.140 18 1.196 17 1.245 16 1.297 15 1.332 14 1.435 13 1.445 12 1.433 11 1.398 10 1.342 9 1.262 8 1.158 7 1.037 6 0.909 5 0.790 4 0.697 3 0.619 2 0.497 Bottom 1 0.143 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-24 Core Power Heat Flux 500.0- -.- Core Flow

--- Steam Flow Feed Flow_

400.0 -

~0 ci) a 300.0 -

0 ci)

C-) 200.0 -

ci) 0~

100.0-

.0-

-, tflJAJ +

.0 1.0 2.0 3.0 4.0 5.0 Time, (seconds)

Figure 5.1 EOCLB LRNB at 100P/104.5F - TSSS Key Parameters AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-25 E

0 Ul) 0)

Q)

CL 0

Q)

(I)

V)

Figure 5.2 EOCLB LRNB at 1OOP/104.5F- TSSS Sensed Water Level AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-26 rj U) 0~

U)

3 U)

U)

U) 0~

Figure 5.3 EOCLB LRNB at 100PI104.5F - TSSS Vessel Pressures AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-27 0

a1) 41) 0O Figure 5.4 EOCLB FWCF at 100PI104.5F - TSSS Key Parameters AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-28 C:

0 C:

E 0

Ofl 5.0 10.0 15.0 20.0 Time, (seconds)

Figure 5.5 EOCLB FWCF at 100PI104.5F - TSSS Sensed Water Level AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 5-29 0

(I) 0~

J U)

(I) 0~

5.0 10.0 15.0 20.0 Time, (seconds)

Figure 5.6 EOCLB FWCF at 100PI104.5F - TSSS Vessel Pressures AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 6-1 6.0 Postulated Accidents 6.1 Loss-of-Coolant-Accident (LOCA)

The results of the ATRIUM-10 LOCA analysis are presented in References 24 and 25. The ATRIUM-10 PCT is 1907'F which is higher than the PCT reported in Reference 3 for Brunswick Unit 1 Cycle 17. The increase in the licensing basis PCT is a result of a design change and not a change in the calculation process. The peak local metal water reaction is 1.15% and the core wide metal water reaction is < 0.50%. The SLO MAPLHGR multiplier is 0.85.

For operation with the MG sets, the GE14 LOCA analysis results are presented in Reference 26 (UFSAR). As discussed in Section 2.1.2, Progress Energy should evaluate the continued applicability of the GE14 LOCA analysis for operation with the VFDs.

6.2 Control Rod Drop Accident (CRDA)

Brunswick Unit 2 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 27 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 27.

The CRDA 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 2 UFSAR radiological assessment (1200 rods).

Maximum dropped control rod worth, mk 10.9 Core average Doppler coefficient, Ak/k/°F -10.0 x 10.6 Effective delayed neutron fraction 0.0052 Four-bundle local peaking factor 1.407 Maximum deposited fuel rod enthalpy, cal/g 201.4 Maximum number of rods exceeding 170 cal/g 366 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 6-2 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 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-10 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-10 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(Infrequent 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 28, 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 AREVA 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 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 6-3 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.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 7-1 7.0 Special Analyses 7.1 ASME OverpressurizationAnalysis 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 2 have sufficient capacity and performance to prevent the reactor vessel pressure from reaching the safety limit of 110% of the design pressure.

An MSIV closure analysis was performed with the AREVA plant simulator code COTRANSA2 (Reference 16) for 102% power and 104.5% flow at the highest Cycle 19 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 21, the plant configuration analyzed assumed that two of the lowest setpoint SRVs were inoperable.

0 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 maximum pressure of 1362 psig occurs in the lower plenum. The maximum dome pressure for the same event is 1318 psig. The results demonstrate that the maximum vessel pressure limit of 1375 psig and dome pressure limit of 1325 psig are not exceeded.

7.2 A TWS 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 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 7-2 pressure (1500 psig). The ATWS overpressurization analyses were performed at 100% power at 99% and 104.5% flow over the Cycle 19 exposure range. The MSIV closure and pressure regulator failure open (PRFO) events were evaluated. Failure of the pressure regulator in the open position causes the turbine control and turbine bypass valves to open 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.

0 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 for the 99% and 104.5% flow cases.

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 1485 psig and the maximum dome pressure is 1463 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. For Unit 1 Cycle 17 (Reference 3) an evaluation was previously performed that concluded that the introduction of ATRIUM-10 fuel will not significantly impact the long term ATWS response (suppression pool temperature and containment pressure) and the current analysis remains applicable. This conclusion was confirmed by analysis for Unit 2 Cycle 19.

Relative the 10 CFR 50.46 acceptance criteria (i.e., PCT and cladding oxidation), the consequences of an ATWS event are bound by those of the limiting LOCA event.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 7-3 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 2 SLC system is required to be able to inject 720 ppm natural boron equivalent at 70°F into the reactor coolant (including a 25% allowance for imperfect mixing, leakage, and volume of other piping connected to the reactor). AREVA has performed an analysis that demonstrates that the SLC system meets the required shutdown capability for Cycle 19. The analysis was performed at a coolant temperature of 360°F with a boron concentration equivalent to 720 ppm at 70 0 F. The temperature of 360°F 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 2.07% Ak/k.

7.4 Fuel Criticality The new fuel storage vault criticality analysis for ATRIUM-10 fuel is presented in Reference 29.

The spent fuel pool criticality analysis for ATRIUM-1 0 fuel is presented in Reference 30. The ATRIUM-10 fuel assemblies identified for loading in Cycle 19 meet both the new and spent fuel storage requirements.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 7-4 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) 328 128 1362 1318 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis ,Page 7-5 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/99F) 262 135 1459 1438 PRFO (100P/104.5F) 300 145 1484 1462 PRFO (1OOP/99F) 286 143 1485 1463 AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 7-6

_0 a) 1- 200.0 C3 0o C) 100.0 4.0 6.0 Time, (seconds)

Figure 7.1 MSIV Closure Overpressurization Event at 102PI104.5F - Key Parameters AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 7-7 C

0 6)

N C

6)

E

3 C,)

C 0

0 6) 0~

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6) 6)

-j ci) 0

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C 6)

U) 2.0 4.0 6.0 8.0 Time, (seconds)

Figure 7.2 MSIV Closure Overpressurization Event at 102P/104.5F - Sensed Water Level AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 7-8 En U) ci, a-4.0 6 Time, (seconds)

Figure 7.3 MSIV Closure Overpressurization Event at 102P/1104.5F - Vessel Pressures AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 7-9 600.0 -

E

-o

> 400.0-cfY 4.0 6.0 Time, (seconds)

Figure 7.4 MSIV Closure Overpressurization Event at 102P/104.5F - Safety/Relief Valve Flow Rates AREVA NP Inc.

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Brunswick Unit 2 Cycle 19, Revision 0 Reload Safety Analysis Page 7-10

~0 15, W

20.0 Time, (seconds)

Figure 7.5 PRFO ATWS Overpressurization Event at 100P/99F - Key Parameters AREVANP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 7-11 C:

0 E

0 0~

20.0 Time, (seconds)

Figure 7.6 PRFO ATWS Overpressurization Event at 100P/99F - Sensed Water Level AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 7-12 1600.0 (10 CL; D0 (3

U)

Q) 20.0 Time, (seconds)

Figure 7.7 PRFO ATWS Overpressurization Event at 100P/99F - Vessel Pressures AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 7-13 SRV Bank 1 -

SRV Bank 2 SRV Bank 3 / /

1000.0- 'F/

800.0 -

E

~0 600.0-400.0-200.0 -

.0 5.0 10.0 15.o 20.0 25.0 30.0 35.0 40.0 Time, (seconds)

Figure 7.8 PRFO ATWS Overpressurization Event at 100P/99F - SafetylRelief Valve Flow Rates AREVA NP Inc.

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Brunswick Unit 2 Cycle 19 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 2 Cycle 19 is based on the analyses of the limiting anticipated operational occurrences (AOOs). 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 AOO 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 1.13. Exposure-dependent MCPR limits were established to support operation from BOC to near end-of-cycle (NEOC), NEOC to end-of-cycle licensing basis (EOCLB), and combined FFTR/Coastdown as defined by the core average exposures listed in Table 5.1.

MCPR limits are established to support base case operation and the EOOS scenarios presented in Table 1.1.

Cycle 19 two-loop operation MCPRP limits for ATRIUM-10 and GE14 fuel are presented in Tables 8.1 - 8.6 for base case operation and the EOOS conditions. Limits are presented for nominal scram speed (NSS) and Technical Specification scram speed (TSSS) insertion times for the exposure ranges considered. An assumed RBM high power setpoint of 108% was used to develop the MCPRp limits. Tables 8.1 and 8.2 present the MCPRp limits for the BOC to NEOC exposure range. Tables 8.3 and 8.4 present the MCPRP limits applicable for the BOC to EOCLB exposure range. Tables 8.5 and 8.6 present the MCPRp limits for FFTR/Coastdown operation.

The FFTR/Coastdown limits (both base case and TBVOOS) support both nominal and constant rated dome pressure operation with feedwater temperatures consistent with the Reference 2' feedwater temperature reduction characteristic. MCPRP limits for single-loop operation are 0.02 higher for all cases.

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 19 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 AOO.

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Brunswick Unit 2 Cycle 19 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 FFTR/Coastdown for both NSS and TSSS insertion times and for the EOOS conditions identified in Table 1.1. The ATRIUM-10 Cycle 19 LHGRFACp multipliers for the BOC to EOCLB exposure range are presented in Tables 8.9 and 8.10. The FFTR/Coastdown LHGRFACp multipliers are presented in Tables 8.11 and 8.12. The FFTR/Coastdown limits (both base case and TBVOOS) support both nominal and constant rated dome pressure operation with the Reference 2 feedwater temperature reduction characteristic.

LHGRFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a postulated slow flow excursion. For ATRIUM-10 fuel, the multipliers are presented in Table 8.13 and are applicable for all Cycle 19 exposures and the EOOS conditions identified in Table 1.1.

Note that LHGR limits are not applied to the GE14 fuel so there are no GE14 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-10 TLO MAPLHGR 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 31. As discussed in Section 2.1.2, Progress Energy should evaluate the applicability of the Reference 31 GE14 MAPLHGR limits for operation with the VFDs. 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 AOOs. 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 19 exposures, both NSS and TSSS insertion times and all the EOOS conditions identified in Table 1.1. The GE14 MAPFACp multipliers for all Cycle 19 exposures are presented in Table 8.15.

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Brunswick Unit 2 Cycle 19 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 19 exposures and the EOOS conditions identified in Table 1.1.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-4 Table 8.1 MCPRp Limits for NSS Insertion Times BOC to < NEOC*

EOOS Power ATRIUM-10 GE14 Condition (% rated) MCPRp MCPRP 100.0 1.35 1.37 90.0 1.38 1.40 Base 50.0 1.48 1.48

> 65%F < 65%F > 65%F < 65%F case operation 50.0 1.94 1.73 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.39 1.40 90.0 1.42 1.43 50.0 1.57 1.59

> 65%F < 65%F > 65%F < 65%F TBVOOS 50.0 1.94 1.73 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.82 2.54 2.78 2.53 23.0 2.96 2.74 2.94 2.73 100.0 1.36 1.38 90.0 1.39 1.41 50.0 1.51 1.52 FHOOS > 65%F < 65%F > 65%F < 65%F 50.0 1.94 1.73 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.40 1.41 90.0 1.43 1.45 50.0 1.61 1.63 TBVOOS > 65%F < 65%F > 65%F < 65%F and FHOOS 50.0 1.94 1.73 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.95 2.64 2.91 2.63 23.0 3.09 2.87 3.06 2.86

• Limits support operation with any combination of I 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 0.02 higher.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-5 Table 8.2 MCPRp Limits for TSSS Insertion Times BOC to < NEOC*

EOOS Power ATRIUM-10 GE14 Condition (% rated) MCPRP MCPRP 100.0 1.45 1.47 90.0 11.46 1.48 50.0 1.50 1.52 Base case case > 65%F < 65%F > 65%F < 65%F operation 50.0 1.97 1.75 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.47 1.49 90.0 1.49 1.52 50.0 1.63 1.65

> 65%F < 65%F > 65%F < 65%F 50.0 1.97 1.75 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.82 2.54 2.78 2.53 23.0 2.96 2.74 2.94 2.73 100.0 1.46 1.48 90.0 1.47 1.49 50.0 1.53 1.54 FHOOS > 65%F <65%F > 65%F - 65%F 50.0 1.97 1.75 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.48 1.50 90.0 1.50 1.53 50.0 1.67 1.68 TBVOOS > 65%F < 65%F > 65%F < 65%F and FHOOS 50.0 1.97 1.75 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.95 2.64 2.91 2.63 23.0 3.09 2.87 3.06 2.86

• 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 0.02 higher.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-6 Table 8.3 MCPRp Limits for NSS Insertion Times BOC to < EOCLB*

EOOS Power ATRIUM-10 GE14 Condition (% rated) MCPRp MCPRP 100.0 1.43 1.44 90.0 1.45 1.45 50.0 1.53 1.53 Base > 65%F < 65%F > 65%F < 65%F case 50.0 1.95 1.74 2.00 1.77 operation 26.0 2.27 2.02 2.23 2.03 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.45 1.46 90.0 1.48 1.48 50.0 1.59 1.59

> 65%F < 65%F > 65%F < 65%F TBVOOS 50.0 1.95 1.74 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.82 2.54 2.78 2.53 23.0 2.96 2.74 2.94 2.73 100.0 1.44 1.45 90.0 1.46 1.46 50.0 1.53 1.53 FHOOS > 65%F < 65%F > 65%F < 65%F 50.0 1.95 1.74 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0' 1.46 1.50 90.0 1.49 1.52 50.0 1.62 1.63 TBVOOS > 65%F < 65%F > 65%F < 65%F and FHOOS 50.0 1.95 1.74 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.95 2.64 2.91 2.63 23.0 3.09 2.87 3.06 2.86

• 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 0.02 higher.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-7 Table 8.4 MCPRp Limits for TSSS Insertion Times BOC to < EOCLB*

EOOS Power ATRIUM-10 GE14 Condition (% rated) MCPRp MCPRP 100.0 1.48 1.49 90.0 1.49 1.49 50.0 1.53 1.53 Base > 65%F < 65%F > 65%F < 65%F case operation 50.0 1.99 1.77 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.50 1.51 90.0 1.51 1.53 50.0 1.63 1.65

> 65%F < 65%F > 65%F < 65%F 50.0 1.99 1.77 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.82 2.54 2.78 2.53 23.0 2.96 2.74 2.94 2.73 100.0 1.49 1.50 90.0 1.50 1.50 50.0 1.53 1.54 FHOOS > 65%F < 65%F > 65%F < 65%F 50.0 1.99 1.77 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.51 1.59 90.0 1.52 1.61 50.0 1.67 1.72 TBVOOS > 65%F < 65%F > 65%F < 65%F and EHOOS 50.0 1.99 1.77 2.07 1.83 26.0 2.29 2.03 2.29 2.08 26.0 2.95 2.64 2.95 2.67 23.0 3.09 2.87 3.10 2.90 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 0.02 higher.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-8 Table 8.5 MCPRp Limits for NSS Insertion Times FFTR/Coastdown*

EOOS Power ATRIUM-10 GE14 Condition (% rated) MCPRP MCPRP 100.0 1.47 1.46 90.0 1.48 1.47 50.0 1.56 1.55 Base > 65%F < 65%F > 65%F < 65%F case operation 50.0 1.96 1.75 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.47 1.50 90.0 1.49 1.52 50.0 1.62 1.63

> 65%F < 65%F > 65%F < 65%F TBVOOS 50.0 1.96 1.75 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.95 2.64 2.91 2.63 23.0 3.09 2.87 3.06 2.86 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 0.02 higher.

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Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-9 Table 8.6 MCPRp Limits for TSSS Insertion Times FFTR/Coastdown*

EOOS Power ATRIUM-10 GE14 Condition (% rated) MCPRp MCPRP 100.0 1.50 1.50 90.0 1.50 1.50 50.0 1.56 1.55 Base > 65%F -< 65%F > 65%F < 65%F case operation 50.0 2.02 1.81 2.04 26.0 1.80 2.29 2.03 2.25 2.04 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.51 1.59 90.0 1.52 1.61 50.0 1.67 1.72

> 65%F < 65%F > 65%F < 65%F TBVOOS 50.0 2.02 1.81 2.07 1.83 26.0 2.29 2.03 2.29 2.08 26.0 2.95 2.64 2.95 2.67 23.0 3.09 2.87 3.10 2.90

• 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 0.02 higher.

AREVA NP Inc.

ANP-2771 (NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-10 Table 8.7 Flow-Dependent MCPR Limits ATRIUM-10 and GE14 Fuel Core Flow

(% of rated) MCPRf 0.0 1.65 31.0 1.65 100.0 1.20 107.0 1.20 AREVA NP Inc.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis 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.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload SafetyAnalysis Page 8-12 Table 8.9 ATRIUM-10 LHGRFACp Multipliers for NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM-10 Condition (% rated) LHGRFACp 100.0 1.00 90.0 1.00 50.0 0.90 Base > 65%F < 65%F case operation 50.0 0.69 0.78 26.0 0.60 0.65 26.0 0.59 0.63 23.0 0.58 0.62 100.0 0.95 90.0 0.94 50.0 0.86

> 65%F < 65%F TBVOOS 50.0 0.69 0.78 26.0 0.60 0.65 26.0 0.47 0.54 23.0 0.45 0.50 100.0 1.00 90.0 1.00 50.0 0.90 EHOOS > 65%F < 65%F 50.0 0.69 0.78 26.0 0.60 0.65 26.0 0.59 0.63 23.0 0.58 0.62 100.0 0.95 90.0 0.94 50.0 0.85 TBVOOS > 65%F < 65%F and FHOOS 50.0 0.69 0.78 26.0 0.60 0.65 26.0 0.45 0.51 23.0 0.43 0.47 AREVA NP Inc.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-13 Table 8.10 ATRIUM-10 LHGRFACp Multipliers for TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM-10 Condition (% rated) LHGRFACp 100.0 1.00 90.0 0.95 50.0 0.87 Base > 65%F <65%F case operation 50.0 0.67 0.77 26.0 26.0 0.60 0.59 0.64 0.63 23.0 0.58 0.62 100.0 0.93 90.0 0.91 50.0 0.86

> 65%F < 65%F TBVOOS 50.0 0.67 0.77 26.0 0.60 0.64 26.0 0.47 0.54 23.0 0.45 0.50 100.0 1.00 90.0 0.95 50.0 0.87 EHOOS > 65%F < 65%F 50.0 0.67 0.77 26.0 0.60 0.64 26.0 0.59 0.63 23.0 0.58 0.62 100.0 0.93 90.0 0.91 50.0 0.84 TBVOOS > 65%F < 65%F and FHOOS 50.0 0.67 0.77 26.0 0.60 0.64 26.0 0.45 0.51 23.0 0.43 0.47 AREVA NP Inc.

ANP-2771 (NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-14 Table 8.11 ATRIUM-10 LHGRFACp Multipliers for NSS Insertion Times FFTR/Coastdown EOOS Power ATRIUM-10 Condition (%rated) LHGRFACp 100.0 1.00 90.0 1.00 Base 50.0 0.86

> 65%F < 65%F case operation 50.0 0.69 0.77 26.0 0.60 0.65 26.0 0.59 0.63 23.0 0.58 0.62 100.0 0.93 90.0 0.91 50.0 0.85

> 65%F < 65%F 50.0 0.69 0.77 26.0 0.60 0.65 26.0 0.45 0.51 23.0 .0.43 0.47 AREVA NP Inc.

ANP-2771 (NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-15 Table 8.12 ATRIUM-10 LHGRFACp Multipliers for TSSS Insertion Times FFTR/Coastdown EOOS Power ATRIUM-10 Condition (% rated) LHGRFACP 100.0 1.00 90.0 0.94 50.0 0.86 Base > 65%F < 65%F case operation 50.0 0.67 0.76 26.0 0.60 0.64 26.0 0.59 0.63 23.0 0.58 0.62 100.0 0.91 90.0 0.90 50.0 0.82

> 65%F < 65%F 50.0 0.67 0.76 26.0 0.60 0.64 26.0 0.45 0.51 23.0 0.43 0.47 AREVA NP Inc.

ANP-2771 (NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-16 Table 8.13 ATRIUM-10 LHGRFACf Multipliers All Cycle 19 Exposures Core Flow

(% of rated) LHGRFACf 0.0 0.90 31.0 0.90 50.0 1.00 107.0 1.00 AREVA NP Inc.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis 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.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-18 Table 8.15 GE14 MAPFACp Multipliers for NSS and TSSS Insertion Times All Cycle 19 Exposures EOOS Power GE14 Condition (% rated) MAPFACP 100.0 1.00 Base 50.0 0.73 case and all > 65%F -<65%F supported 50.0 0.64 0.73 Conditions 26.0 0.56 0.61 26.0 0.43 0.49 23.0 0.41 0.45 AREVA NP Inc.

ANP-2771 (NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 8-19 Table 8.16 GEl4 MAPFACf Multipliers All Cycle 19 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-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 9-1 9.0 References

1. ANP-2727(P) Revision 0, Brunswick Unit 2 Cycle 19 Fuel Cycle Design, AREVA NP, June 2008.

2. ANP 2700(P) Revision 0, Brunswick Unit 2 Cycle 19 Plant ParametersDocument, AREVA NP, July 2008.

3. ANP-2674(P) Revision 2, Brunswick Unit I Cycle 17 Reload Safety Analysis, AREVA NP, March 2008.

4. 51-9071336-000, "Brunswick Unit 2 Cycle 19 Calculation Plan," AREVA NP, July 24, 2008.

5. ANP-2743(P) Revision 0, MechanicalDesign Report for Brunswick Unit 2 Reload BRK2-19 A TRIUM-IC FuelAssemblies, AREVA NP, August 2008.

6. Letter, Bart C. Buckley (NRC) to E. E. Utley (CP&L), "Issuance of Amendment No. 153 to Facility Operating License No. DPR Brunswick Steam Electric Plant, Unit 2, Regarding Fuel Cycle No. 8 - Reload Extended Burnup Fuel (TAC No. 66155),"

September 20, 1988 (38-9061815-000).

7. ANP-2729(P) Revision 0, Brunswick Unit 2 Thermal-HydraulicDesign Report for ATRIUM TM- 10 Fuel Assemblies, AREVA NP, July 2008.

8. ANF-524(P)(A) Revision 2 and Supplements 1 and 2, ANF CriticalPower Methodology for Boiling Water Reactors, Advanced Nuclear Fuels Corporation, November 1990.

9. EMF-2209(P)(A) Revision 2, SPCB CriticalPower Correlation, Framatome ANP, September 2003.

10. EMF-2209(P) Revision 2 Addendum 1 Revision 0, SPCB Additive Constants for ATRIUM-IC Fuel, AREVA NP, April 2008.

11. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation'sCriticalPower Correlationsto Co-Resident Fuel, Siemens Power Corporation, August 2000.

12. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications, GE Nuclear Energy, August 1996.

13. BAW-1 0255PA Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008.

14. 0G02-0119-260, Backup Stability Protection (BSP) for Inoperable Option III Solution, GE Nuclear Energy, July 17, 2002.

15. EMF-CC-074(P)(A) Volume 4 Revision 0, BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.

16. ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Programfor Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.

17. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements I and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-HydraulicCore Analysis, Exxon Nuclear Company, February 1987.

AREVA NP Inc.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page 9-2

18. XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon NuclearMethodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.

19. EMF-2158(P)(A) Revision 0, Siemens Power CorporationMethodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.

20. 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.

21. Operating License and Technical Specifications, Brunswick Steam Electric Plant, Unit No 2, Progress Energy, as amended.

22. ANF-1358(P)(A) Revision 3, The Loss of FeedwaterHeating Transientin Boiling Water Reactors, Framatome ANP, September 2005.

23. E-mail, E.J. Geyer (PGN) to R.J. DeSteese (AREVA), "Re: B2C19 RBM Setpoint (NF08-254)," October 20, 2008 (38-9095748-000).

24. ANP-2625(P) Revision 0, Brunswick Units I and 2 LOCA Break Spectrum Analysis for ATRIUM TM-IO Fuel, AREVA NP, June 2007.

25. ANP-2624(P) Revision 1, Brunswick Units I and 2 LOCA-ECCS Analysis MAPLHGR Limit forATRIUMTM -IO Fuel, AREVA NP, December 2008.

26. Updated FSAR Brunswick Steam Electric Plant, Units I and 2,. Revision 21.

27. XN-NF-80-19(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.

28. XN-NF-80-19(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.

29. ANP-2661 (P) Revision 0, Brunswick Nuclear Plant New Fuel Storage Vault Criticality Safety Analysis for A TRIUM TM-IO Fuel, AREVA NP, September 2007.

30. ANP-2642(P) Revision 0, Brunswick Nuclear Plant Spent Fuel Storage Pool Criticality Safety Analysis for A TRIUM TM-IO Fuel, AREVA NP, September 2007.

31. Brunswick Unit 2, Cycle 18 Core OperatingLimits Report Revision 1, Progress Energy, June 2008.

32. Letter, E.J. Geyer (PGN) to C.M. Powers (AREVA), "Brunswick Recirculation Pump .

Coast Down Inertia," NF08-300, December 11, 2008 (38-9100527-000).

33. 2B21-1296 Revision 1, "Generation of the B2C19 Plant Parameters Document,"

Progress Energy Calculation, October 2008 (38-9095541-000).

AREVA NP Inc.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload'Safety Analysis Page A-1 Appendix A MELLLA+ Operation Progress Energy is considering applying for approval to operate in the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) domain which Would provide greater flow flexibility, especially at rated and near-rated power. The combined MELLLA/MELLLA+ power flow map is presented in Figure A.1. The EOOS conditions presented in Table A.1 are supported. While approval is not expected for Brunswick Unit 2 Cycle 19, the cycle-specific reload licensing analyses were performed to support operation in the MELLLA+ domain. Special consideration of MELLLA+ was given during the analyses for the anticipated operational occurrences (AOOs),

ASME and ATWS overpressurization, stability, LOCA, and the other accidents.

Transient analyses results at state points along the MELLLA+ low flow boundary demonstrate that the events are limiting at high flow. Results for the other AOOs show that the operating limits presented in Section 8.0 remain applicable for MELLLA+. LOCA analysis results also show that high flow is limiting. Since no SRVs will be allowed out of service in MELLLA+

operation, the MELLLA+ ASME and ATWS overpressurization results are non-limiting. While most of the limiting analysis results that support the MELLLA operating limits remain limiting for the MELLLA+ domain, the stability analysis is impacted as discussed below.

Stability The Enhanced Option III (EO-III) Long Term Stability Solution (Reference A.1) was used in the stability evaluations to support MELLLA+ operation. The cycle-specific setpoint determination was performed in accordance with the approved methodology presented in References A.1 and A.2, including the 10% penalty on the DIVOM slope discussed in the Reference A.2 NRC safety evaluation. The EO-Ill solution consists of two components: a single channel exclusion region, and a stability-based Operating Limit MCPR (OLMCPR).

The first component is the single channel exclusion region which is protected by automatic scram. The endpoints of the channel exclusion region are given in Table A.2. The exclusion region boundary is defined by a straight line fit through these two endpoints and is only valid for nominal feedwater temperature since FHOOS and FFTR operation are not allowed in the MELLLA+ region.

AREVA NP Inc.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page A-2 The cycle-specific DIVOM curves were generated consistent with the approach described in Reference A.1 using state points located inside the channel exclusion region. A review of the DIVOM analysis results shows that no single channel instabilities occurred, confirming that the channel exclusion region has been adequately set. The limiting calculated DIVOM results for all state points and exposures were used in the evaluation of the stability-based OLMCPRs.

The stability-based OLMCPR is provided for two conditions as a function of OPRM amplitude setpoint in Table A.3. The two conditions evaluated are for 1) a postulated oscillation at 45%

core flow steady-state operation (SS) and 2) following a two recirculation pump trip (2PT) from the limiting full power state point. The Cycle 19 power- and flow-dependent MCPR limits provide adequate protection againstviolation of the SLMCPR for postulated reactor instability as long as the MCPR operating limit is greater than or equal to the specified value for the selected OPRM setpoint. The results in Table A.3 are valid for the full ICF/MELLLA+ operating domain.

For Backup Stability Protection (BSP) during MELLLA+ operation, AREVA has extended the Base Minimal Region I (scram region) using the generic shape generating function (Reference A.3) to the point where it intersects the extension of the MELLLA+ upper flow boundary line (68.8 %P/ 43.2 %F). STAIF analyses based on a pump-trip runback to this intersection have been performed to demonstrate that the decay ratio criteria are met. Since FFTR and FHOOS operation is not allowed in the MELLLA+ domain, these analyses have been performed only for nominal feedwater temperature operation.

There is no need to extend Base Minimal Region II (controlled entry) for MELLLA+ operation.

Potential pump trips from MELLLA+ operating pqints that could result in a post-trip endpoint outside (above) the extended Region I boundary and above the MELLLA line but below the MELLLA+ region, are by definition outside the allowed power/flow map domain. Such a condition would require intervention to bring the core back within the allowable power/flow map operating domain. STAIF analyses at the low-flow high-power corner of the MELLLA+ domain show significant stability margin.

As discussed in Reference A.4, additional analyses are needed to ensure the acceptability of the ATWS with core instability analyses with ATRIUM-10 fuel at MELLLA+ conditions.

References A. 1 ANP-1 0262PA Revision 0, Enhanced Option Ill Long Tenm Stability Solution, AREVA NP, May 2008.

AREVA NP Inc.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page A-3 A.2 BAW-1 0255PA Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008.

A.3 0G02-0119-260, Backup Stability Protection (BSP) for Inoperable Option III Solution, GE Nuclear Energy, July 17, 2002.

A.4 51-9071336-000, "Brunswick Unit 2 Cycle 19 Calculation Plan," AREVA NP, July 24, 2008.

AREVA NP Inc.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page A-4 Table A.1 MELLLA+ EOOS Operating Conditions*, t Turbine bypass valves out-of-service (TBVOOS)

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 up to 40% of the TIP channels out-of-service, and/or up to 50% of the LPRMs out-of-service.

t Note: Feedwater heater out-of-service, SRV out-of-service, main steam isolation Valves out-of-service, and single-loop operation are not allowed in the MELLLA+ region.

AREVA NP Inc.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page A-5 Table A.2 Channel Exclusion Region Endpoints Power Flow

(% rated) (% rated) 50.0 31.0 65.8 39.2 AREVA NP Inc.

ANP-2771(NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page A-6 Table A.3 OPRM Setpoints OPRM OLMCPR OLMCPR Setpoint (SS) (2PT) 1.05 1.21 1.24 1.06 1.24 1.26 1.07 1.26 1.28 1.08 1.28 1.30 1.09 1.30 1.33 1.10 1.33 1.35 1.11 1.35 1.38 1.12 1.37 1.40 1.13 1.40 1.43 1.14 1.43 1.45 1.15 1.45 1.48 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.

ANP-2771 (NP)

Brunswick Unit 2 Cycle 19 Revision 0 Reload Safety Analysis Page A-7 0

0.

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 A.1 Brunswick Unit 2 Power Flow Map with MELLLA+

AREVA NP Inc.