Text

ANP-3061(NP), Revision 0, Brunswick, Unit 1, Cycle 19 Reload Safety Analysis
	ML12100A088
Person / Time
Site:	Brunswick
Issue date:	12/31/2011
From:	; AREVA NP
To:	; Office of Nuclear Reactor Regulation
References
	BSEP 12-0040 ANP-3061(NP), Rev 0
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BSEP 12-0040 ANP-3061(NP), Brunswick Unit 1 Cycle 19 Reload Safety Analysis, Revision 0

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ANP-3061 (NP)

Revision 0 Brunswick Unit 1 Cycle 19 Reload Safety Analysis December 2011 A

AREVA AREVA NP Inc.

e Document AREVA NP Inc.

ANP-3061(NP)

Revision 0 Brunswick Unit I Cycle 19 Reload Safety Analysis

AREVA NP Inc.

ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page i Nature of Changes Item Page Description and Justification

1.

All This is the initial issue AREVA NP Inc.

Contt.,oK IDocument ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page ii Contents 1.0 In tro d u ctio n..................................................................................................................

1-1 2.0 Disposition of Events for ATRIUM 10XM Fuel Introduction...........................................

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

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

4-1 4.1 Thermal-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 Downscale (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 Equipment Out-of-Service Scenarios................................................................

5-7 5.3.1 F H O O S...............................................................................................

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

5-8 5.3.3 Com bined FHOOS and TBVOOS......................................................

5-8 5.3.4 One SRVOOS.....................................................................................

5-8 5.3.5 One MSIVOOS...................................................................................

5-8 5.3.6 Single-Loop Operation........................................................................

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

5-9 6.0 Postulated Accidents....................................................................................................

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

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

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

6-2 6.4 Fuel Loading Error (Infrequent Event)...............................................................

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

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

6-3 7.0 Special Analyses..........................................................................................................

7-1 7.1 ASM E Overpressurization Analysis...................................................................

7-1 7.2 ATW S Event Evaluation....................................................................................

7-2 7.2.1 ATW S Overpressurization Analysis.....................................................

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

7-3 7.3 Standby Liquid Control System.........................................................................

7-3 7.4 F u e l C ritic a lity...................................................................................................

7-4 8.0 Operating Lim its and COLR Input.................................................................................

8-1 8.1 M C P R L im its.....................................................................................................

8-1 8.2 LHGR Lim its.................................................................................................

8-1 8.3 MAPLHGR Lim its..............................................................................................

8-2 9.0 R e fe re n ce s...................................................................................................................

9 -1 AREVA NP Inc.

ANP-3061 (NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page iii 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 e tpo ints...........................................................................................................

4-5 4.4 BSP Endpoints for Brunswick Unit 1 Cycle 19..............................................................

4-6 5.1 Exposure Basis for Brunswick Unit 1 Cycle 19 Transient Analysis..............................

5-11 5.2 S cram S peed Insertion T im 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............................................................

5-19 5.11 R B M O perability R equirem ents..................................................................................

5-20 5.12 Flow -D ependent M C PR R esults.................................................................................

5-20 5.13 Licensing Basis Core Average Axial Power Profile.....................................................

5-21 7.1 ASME Overpressurization Analysis" Results-................................................................ 7-5 7.2 ASME Overpressurization Sensitivity Analysis Results..:..............................................

7-5 7.3 ATWS Overpressurization Analysis Results..............................................................

7-6 7.4 ATWS Overpressurization Sensitivity Analysis Results................................................ 7-7 8.1 MCPRP Limits for NSS Insertion Times BOC to < NEOC..............................................

8-3 8.2 MCPRP Limits for TSSS Insertion Times BOC to < NEOC............................................

8-4 8.3 MCPRp Limits for NSS Insertion Times BOC to < EOCLB............................................

8-5 8.4 MCPRP Limits for TSSS Insertion Times BOC to < EOCLB..........................................

8-6 8.5 MCPRP Limits for NSS Insertion Times FFTR/Coastdown............................................

8-7 8.6 MCPRP Limits for TSSS Insertion Times FFTR/Coastdown..........................................

8-8 8.7 Flow-Dependent MCPR Limits ATRIUM 1OXM and ATRIUM-10 Fuel.......................... 8-9 8.8 S teady-State LH G R Lim its............................................................................................

8-9 8.9 LHGRFACP Multipliers for NSS Insertion Times BOC to < EOCLB............................. 8-10 8.10 LHGRFACp Multipliers for TSSS Insertion Times BOC to < EOCLB........................... 8-11 8.11 LHGRFACP Multipliers for NSS Insertion Times FFTR/Coastdown............................. 8-12 8.12 LHGRFACP Multipliers for TSSS Insertion Times FFTR/Coastdown........................... 8-13 8.13 ATRIUM 1OXM LHGRFACf Multipliers All Cycle 19 Exposures...................................

8-14 8.14 ATRIUM-10 LHGRFACf Multipliers All Cycle 19 Exposures........................................

8-14 8.15 AREVA Fuel MAPLHGR Limits...................................................................................

8-15 AREVA NP Inc.

Doc ment ANP-3061 (NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page iv Figures 1.1 B runsw ick U nit 1 Pow er/Flow M ap...............................................................................

1-3 5.1 EOCLB LRNB at 100P/104.5F -TSSS Key Parameters............................................

5-22 5.2 EOCLB LRNB at 1OOP/1 04.5F - TSSS Sensed Water Level......................................

5-23 5.3 EOCLB LRNB at 100P/104.5F -TSSS Vessel Pressures..........................................

5-24 5.4 EOCLB TTNB at 10OP/104.5F - TSSS Key Parameters............................................

5-25 5.5 EOCLB TTNB at 100P/104.5F -TSSS Sensed Water Level......................................

5-26 5.6 EOCLB TTNB at 10OP/104.5F - TSSS Vessel Pressures..........................................

5-27 5.7 EOCLB FWCF at 1OOP/1 04.5F - TSSS Key Parameters...........................................

5-28 5.8 EOCLB FWCF at IOOP/1 04.5F - TSSS Sensed Water Level.....................................

5-29 5.9 EOCLB FWCF at 100P/104.5F -TSSS Vessel Pressures.........................................

5-30 7.1 MSIV Closure Overpressurization Event at 102P/104.5F - Key Parameters................ 7-8 7.2 MSIV Closure Overpressurization Event at 102P/1 04.5F - Sensed Water Level.......... 7-9 7.3 MSIV Closure Overpressurization Event at 102P/104.5F -Vessel Pressures............ 7-10 7.4 MSIV Closure Overpressurization Event at 102P/104.5F - Safety/Relief Valve F lo w R a te s.................................................................................................................

7 -1 1 7.5 PRFO ATWS Overpressurization Event at 1OOP/1 04.5F - Key Parameters............... 7-12 7.6 PRFO ATWS Overpressurization Event at 1OOP/1 04.5F - Sensed Water Level......... 7-13 7.7 PRFO ATWS Overpressurization Event at 1OOP/1 04.5F - Vessel Pressures............. 7-14 7.8 PRFO ATWS Overpressurization Event at 1OOP/1 04.5F - Safety/Relief Valve Flow Rates..........................

7-15 AREVA NP Inc.

o r Document Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

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

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page vi Nomenclature (Continued)

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

OLMCPR operating limit minimum critical power ratio OPRM oscillation power range monitor Pbypass power below which direct scram on TSVITCV closure is bypassed PCT peak cladding temperature PLU power load unbalance PRFDS pressure regulator failure downscale PRFO pressure regulator failure open RBM (control) rod block monitor 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 UFSAR updated final safety analysis report VFD variable frequency drive ACPR change in critical power ratio AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 1-1 1.0 Introduction Reload licensing analyses results generated by AREVA NP are presented in support of Brunswick Unit 1 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 234 fresh ATRIUM TM 1OXM* assemblies and 326 irradiated ATRIUM-10 assemblies. The licensing analysis supports the core design presented in Reference 1.

The Cycle 19 reload licensing analyses were performed for the potentially limiting events and analyses that were identified in the disposition of events. The results of the analyses are used to establish the Technical Specifications/COLR limits and ensure that the design and licensing criteria are met. The design and safety analyses are based on the design and operational assumptions and plant parameters provided by the utility. The results of the reload licensing analysis support operation for the power/flow map presented in Figure 1.1 and also support operation with the equipment out-of-service (EOOS) scenarios presented in Table 1. 1.

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Controled Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 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 regLulator.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.

AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 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 1 i 4 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr 0

10 20 30 40 50 60 70 80 90 100 110 120 (%)

Core Flow Figure 1.1 Brunswick Unit 1 Power/Flow Map AREVA NP Inc.

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rn,n ANP-3061(NP)

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

The results of the disposition of events are presented in Reference 2.

The plant parameter differences between those used in the Brunswick Unit 1 Cycle 18 analyses and the planned analyses for the Brunswick Unit 1 Cycle 19 reload were reviewed to determine if the conclusions of the disposition of events remain applicable. The review concluded that analyses affected by the differences were included in the Reference 3 calculation plan.

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SDocument ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 3-1 3.0 Mechanical Design Analysis The mechanical design analyses for ATRIUM I0XM and ATRIUM-10 fuel are presented in the applicable mechanical design reports (References 4, 5, and 6). The maximum exposure limits for the ATRIUM 1OXM and ATRIUM-10 fuel are:

54.0 GWd/MTU average assembly exposure 62.0 GWd/MTU rod average exposure (full-length fuel rods)

Even though the ATRIUM 1OXM and ATRIUM-10 fuel designs are licensed for operation to a peak rod average exposure of 62 GWd/MTU, they will be limited to 60 GWd/MTU as prescribed in Brunswick Unit 1 license amendment 124 (Reference 7).

The ATRIUM 10XM and ATRIUM-10 LHGR limits are presented in Section 8.0. The fuel cycle design analyses (Reference 1) have verified that the ATRIUM 10XM and ATRIUM-10 fuel assemblies remain within licensed burnup limits.

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Do ument ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 4-1 4.0 Thermal-Hydraulic Design Analysis 4.1 Thermal-Hydraulic Design and Compatibility The results of the thermal-hydraulic characterization and compatibility analyses are presented in the thermal-hydraulic design report (Reference 8). The analysis results demonstrate that the thermal-hydraulic design and compatibility criteria are satisfied for the Brunswick Unit 1 transition core consisting of ATRIUM 10XM and ATRIUM-10 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 1 Cycle 19 core was determined using the methodology described in Reference 9. The analysis was 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 1 Cycle 19 SLMCPR analysis used the ACE/ATRIUM 1OXM critical power correlation additive constants and additive constant uncertainty for ATRIUM 1OXM fuel.

described in Reference 10. The SPCB critical power correlation additive constants and additive constant uncertainty for ATRIUM-10 fuel are 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 at the end of rated power operation in Cycle 19 is less than 45 GWd/MTU for ATRIUM 1OXM and all ATRIUM-10 fuel that is face adjacent to the ATRIUM 1OXM fuel. The maximum assembly average exposure of the ATRIUM-10 fuel is less than 55 GWd/MTU. 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.

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ontc1ed oimpn ANP-3061 (NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 4-2 The analysis results support a two-loop operation (TLO) SLMCPR of 1.11 and a single-loop operation (SLO) SLMCPR of 1.12, the values currently in the Brunswick Unit 1 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 1 Cycle 19, Backup Stability Protection (BSP) is provided in accordance with Reference 14. 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 both nominal 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 16.

AREVA NP Inc.

Controlled Document Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 4-3 Table 4.1 Fuel-and Plant-Related Uncertainties for Safety Limit MCPR Analyses Parameter U

Fuel-Related Uncertainties ncertainty I

I Plant-Related Uncertainties Feedwater flow rate 1.8%

Feedwater temperature 0.8%

Core pressure 0.8%

Total core flow rate TLO 2.5%

SLO 6%

[

I AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 4-4 Table 4.2 Results Summary for Safety Limit MCPR Analyses Percentage SLMCPR of Rods in Boiling Transition TLO - 1.11 0.098 SLO - 1.12 0.098 AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 4-5 Table 4.3 OPRM Setpoints OPRM OLMCPR OLMCPR Setpoint (SS)

(2PT) 1.05 1.24 1.26 1.06 1.25 1.28 1.07 1.27 1.30 1.08 1.29 1.32 1.09 1.31 1.34 1.10 1.33 1.36 1.11 1.35 1.38 1.12 1.37

'1.40 1.13 1.39 1.42 1.14 1.42 1.44 1.15 1.44 1.46 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.

)curnent Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 4-6 Table 4.4 BSP Endpoints for Brunswick Unit 1 Cycle 19 Feedwater Temperature Operation End Point Power Flow Mode Region Designation

(% rated)

Nominal Scram IA 56.6 40.0 Nominal Scram l1B 40.7 31.0 Nominal Controlled IIA 64.5 50.0 entry Nominal Controlled l1B 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 lIB 28.5 31.0 FHOOS entry AREVA NP Inc.

~liroUI JD mnT ANP-3061(NP)

Brunswick Unit 1 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 1 Cycle 19.

COTRANSA2 (Reference 17), XCOBRA-T (Reference 18), XCOBRA (Reference 19), and CASMO-4/MICROBURN-B2 (Reference 20) are the major codes used in the thermal limits analyses as described in the AREVA THERMEX methodology report (Reference 19) and neutronics methodology report (Reference 20). 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 ACE/ATRIUM 1OXM critical power correlation (Reference 10) is used to evaluate the thermal margin for the ATRIUM 1OXM fuel. The SPCB critical power correlation (Reference 11) is used in the thermal margin evaluations for the ATRIUM-10 fuel. Fuel pellet-to-cladding gap conductance values are based on RODEX2 (Reference 21) calculations for the Brunswick Unit 1 Cycle 19 core.

5.1 System Transients The reactor plant parameters for the system transient analyses were provided by the utility.

Analyses have been performed to determine power-dependent MCPR limits that protect operation throughout the power/flow domain shown in Figure 1.1.

At Brunswick, direct scram on turbine stop valve (TSV) position and turbine control valve (TCV) fast closure are bypassed at power levels less than 26% of rated (Pbypass). Scram will occur when the high pressure or high neutron flux scram setpoint is reached. Reference 22 indicates that MCPR limits only need to be monitored at power levels greater than or equal to 23% of rated, which is the lowest power analyzed for this report.

The limiting exposure for rated power pressurization transients is typically at end of full power (EOFP) when the control rods are fully withdrawn. To provide additional margin to the operating limits earlier in the cycle, analyses were also performed to establish operating limits at a near end-of-cycle (NEOC) exposure of 16,500 MWd/MTU. Analyses were performed at cycle exposures prior to NEOC to ensure that the operating limits provide the necessary protection.

The end-of-cycle licensing basis (EOCLB) analysis was performed at EOFP + 15 EFPD AREVA NP Inc.

Conrc Er~

DoCLm nt ANP-3061 (NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 5-2 (18,661 MWd/MTU). Analyses were also performed to support extended cycle operation with final feedwater temperature reduction (FFTR) and power coastdown. The Brunswick Unit 1 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 1 turbine bypass system includes four bypass valves. However, for base case analyses in which credit is taken for turbine bypass operation, only three of the turbine bypass valves are assumed operable.

Reductions in feedwater temperature of less than or equal to 1 0°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 can be more severe as a result of the increase in core inlet subcooling during the overcooling phase of the event. Analyses were performed to evaluate the impact of reduced feedwater temperature&dn 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 and TTNB 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 0F 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 20,414 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 AREVA NP Inc.

itrolled Documer ANP-3061 (NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 5-3 nominal scram speed (NSS) insertion times and the Technical Specifications scram speed (TSSS) insertion times used in the analyses are presented in Table 5.2. The NSS MCPRP limits can only be applied if the scram speed test results meet the NSS insertion times. System transient analyses were performed to establish MCPRP limits for both NSS and TSSS insertion times. The Brunswick Unit 1 Technical Specifications (Reference 22) 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 collapseis 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 are identical.

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.

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ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 5-4 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.

As indicated in the Reference 2 disposition of events summary, with the implementation of the variable frequency drives (VFDs), the TTNB may become limiting. Since the VFD implementation at Brunswick Unit 1 is complete, TTNB analyses were performed. Tables 5.5 and 5.6 present the base case TTNB 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 TTNB event initiated at 100% of rated power and 104.5% of rated core flow with TSSS insertion times.

5.1.3 Feedwater Controller Failure (FWCF)

The increase in feedwater flow due to a failure ofthe 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. Three of the four installed turbine bypass valves are assumed operable 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 AREVA NP Inc.

Controlled Docmet ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 5-5 results used to generate the NEOC and EOCLB operating limits for both TSSS and NSS insertion times. Figures 5.7 - 5.9 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 Requlator 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 or equal to 90% of rated and less than 50% or rated,. Previous analysis results demonstrate that the LRNB is more limiting at power levels greater than or equal to 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 1 00°F decrease in the feedwater temperature. The result is an increase in core inlet subcooling, which reduces voids, thereby increasing core power and shifting the axial power distribution toward the bottom of the core. As a result of the axial power shift and increased core power, voids begin to build up in the bottom region of the core, acting as negative feedback to the increased subcooling effect.

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

The increase in steam flow is accommodated by the pressure control system via the TCVs or AREVA NP Inc.

Controlled DocuLe.

n j ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 5-6 the turbine bypass valves, so no pressurization occurs. For Brunswick Unit 1 Cycle 19, a cycle-specific analysis was performed in accordance with the Reference 23 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 selected analytical RBM high power setpoint values from 108% to 117%. An assumed RBM high power setpoint of 108% was used to develop the MCPRP limits. At all intermediate and lower power setpoint values, the MCPRp values 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 10XM and ATRIUM-10 with the LHGR limits and their associated multipliers presented in Section 8.2. 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 permitted by the equipment (107% of rated core flow). An uncontrolled increase in 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.

AREVA NP Inc.

Controlled Doc..ent ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 5-7 MCPRf limits are determined for all fuel types in the core. 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 such that the increase in core power, resulting from the maximum increase in core flow, assures 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 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 1 OXM and ATRIUM-1 0 fuel. The analysis assumes that the recirculation flow increases slowly along the limiting rod line to the maximum flow physically permitted 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 c&nterline melt and overstraining of the cladding during a flow runup. The Cycle 19 LHGRFACf multipliers are presented in Table 8.13 for ATRIUM 1OXM fuel and Table 8.14 for ATRIUM-10 fuel.

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 multipliers for two-loop operation are applicable for SLO.

5.3 Equipment Out-of-Service Scenarios The following equipment out-of-service (EOOS) scenarios are supported for Brunswick Unit 1 Cycle 19 operation:

Feedwater heater out-of-service (FHOOS) - up to 110.30 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)

AREVA NP Inc.

Controlled F Locument ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 5-8 5.3.1 FHOOS The FHOOS scenario assumes a feedwater temperature reduction of 110.3 0F 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 required to increase the enthalpy of the coolant to saturated conditions. The consequences of the FWCF event are potentially 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 two or more of the turbine bypass valves cannot be assured, thereby reducing the pressure relief capacity during fast pressurization transients. While the base case LRNB and TTNB events are analyzed assuming the turbine bypass valves out-of-service, operation with TBVOOS has an adverse effect on the FWCF event. Analyses of the-FWCF event with TBVOOS were performed to establish the TBVOOS operating limits.

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 AREVA NP Inc.

<~i~oument ANP-3061 (NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 5-9 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 nonpressurization 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 Singqle-Loop Operation In SLO, the two-loop operation ACPRs and LHGRFAC multipliers remain applicable. The only impacts on the MCPR, LHGR, and MAPLHGR limits for SLO are an increase of 0.01 in the SLMCPR as discussed in Section 4.2, and the application of an SLO MAPLHGR multiplier discussed in Section 8.3. The net result is a 0.01 increase in the base case MCPRP limits and a decrease in the MAPLHGR limit. The same situation is true for the EOOS scenarios. Adding 0.01 to the corresponding two-loop operation EOO MCPRP. limits results in SLO MCPRp limits for the EOOS conditions. The TLO EOOS LHGRFAC multipliers remain applicable in SLO.

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 33,159 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.

AREVA NP Inc.

>i

,,~kd Doctv ANP-3061 (NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 5-10 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.

AREVA NP Inc.

Document Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-11 Table 5.1 Exposure Basis for Brunswick Unit 1 Cycle 19 Transient Analysis Cycle Core Exposure at Average End of Interval Exposure (MWd/MTU)

(MWd/MTU)*

Comments 0

14,498 Beginning of cycle 16,500 30,998 Break point for exposure-dependent MCPRp limits (NEOC) 18,661 33,159 Design basis rod patterns to EOFP +. 15 EFPD (EOCLB) 20,414 34,912 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.

AREVA NP Inc.

-d Document Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-12 Table 5.2 Scram Speed Insertion Times Control Rod TSSS NSS Position Time Time (notch)

(sec)

(sec) 48 (full-out) 0.000 0.000 48 0.200 0.200 46 0.440 0.318 36 1.080 0.829 26 1.830 1.369 6

3.350 2.510 0 (full-in) 3.806 2.852 AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-13 Table 5.3 NEOC Base Case LRNB Transient Results ATRIUM 1OXM ATRIUM 1OXM Supported ATRIUM-10 ATRIUM-10 Power ACPR LHGRFACP ACPR HFR 100 90 80 70 60 50 50 at > 65%F PLU inoperable 50 at < 65%F PLU inoperable 26 at > 65%F PLU inoperable 26 at < 65%F PLU inoperable 26 at > 65%F below Pbypass 26 at < 65%F below Pbypass 23 at > 65%F below Pbypass 23 at < 65%F below Pbypass TSSS Insertion Times 0.32 1.00 0.33 1.00 0.34 0.98 0.34 0.96 0.33 0.94 0.31 0.92 0.77 0.86 0.62 0.86 1.17 0.68 1.03 0.72 1.17 0.66 1.03 0.66 1.25 0.64 1.12 0.64 0.39 0.40 0.41 0.41 0.40 0.37 0.84 0.71 1.21 1.08 1:21 1.08 1.29 1.15 1.35 1.35 1.35 1.42 1.40 1.37 1.72 1.54 1.97 1.75

,1.97 1.75 2.01 1.81 100 90 80 70 NSS Insertion Times 0.27 1.00 0.29 1.00 0.30 0.98 0.31 0.96 0.30 0.94 0.29 0.92 0.76 0.86 0.62 0.86 1.16 0.70 1.02 0.66 60 0.34 0.36 0.37 0.38 0.37 0.35 0.82 0.71 1.20 1.08 1.35 1.35 1.35 1.39 1.38 1.35 1.67 1.52 1.94 1.75 50 50 at > 65%F PLU inoperable 50 at < 65%F PLU inoperable 26 at > 65%F PLU inoperable 26 at < 65%F PLU inoperable AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-14 Table 5.4 EOCLB Base Case LRNB Transient Results ATRIUM 1OXM ATRIUM 1OXM Supported ATRIUM-10 ATRIUM-10 Power ACPR LHGRFACP ACPR HFR 100 90 80 70 60 50 50 at > 65%F PLU inoperable 50 at < 65%F PLU inoperable 26 at > 65%F PLU inoperable 26 at < 65%F PLU inoperable 26 at > 65%F below Pbypass 26 at < 65%F below Pbypass 23 at > 65%F below Pbypass 23 at < 65%F below Pbypass TSSS Insertion Times 0.33 1.00 0.34 1.00 0.34 0.98 0.34 0.96 0.33 0.94 0.31 0.92 0.77 0.86 0.62 0.86

.1.17 0.68

.1.03 0.72 1.17 0.66 1.03 0.66 1.25 0.64 1.12 0.64 0.42 1.35 0.43 1.35 0.43 1.35 0.43 1.42 0.42 1.46 0.38 1.41 0.84 1.72 0.71 1.54 1.21 1.97 1.08 1.75 1.21 1.97 1.08 1.75 1.29 2.01 1.15 1.81 100 90 80 70 60 50 50 at > 65%F PLU inoperable 50 at < 65%F PLU inoperable 26 at > 65%F PLU inoperable 26 at 5 65%F PLU inoperable NSS Insertion Times 0.29 1.00 0.31 1.00 0.32 0.98 0.33 0.96 0.32 0.94 0.30 0.92 0.76 0.86 0.62 0.86 1.16 0.70 1.02 0.66 0.39 0.40 0.41 0.41 0.40 0.37 0.82 0.71 1.20 1.08 1.35 1.35 1.35 1.39 1.44 1.40 1.67 1.52 1.94 1.75 AREVA NP Inc.

.,Document Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-15 Table 5.5 NEOC Base Case TTNB Transient Results ATRIUM 10XM ATRIUM 1OXM Supported ATRIUM-10 ATRIUM-10 Power ACPR LHGRFACP ACPR HFR TSSS Insertion Times 100 0.33 1.00 0.40 1.35 90 0.33 1.00 0.40 1.35 80 0.33 0.98 0.40 1.35 26 at > 65%F below Pbypass 1.18 0.66 1.22 1.97 26 at < 65%F below Pbypass 1.01 0.66 1.05 1.75 23 at > 65%F below Pbypass 1.25 0.64 1.29 2.01 23 at5 <65%F below Pbypa, 1.11 0.64 1.14 1.81 NSS Insertion Times 100 0.28 1.00 0.34 1.35 90 0.28 1.00

.0.35 c1.35 80 0.29 0.98 0.36 1.35 AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-16 Table 5.6 EOCLB Base Case TTNB Transient Results ATRIUM 1OXM ATRIUM 1OXM Supported ATRIUM-10 ATRIUM-10 Power ACPR LHGRFACP ACPR HFR TSSS Insertion Times 100 0.33 1.00 0.42 1.35 90 0.33 1.00 0.43 1.35 80 0.34 0.98 0.43 1.35 26 at > 65%F below Pbypass 1.18 0.66 1.22 1.97 26 at < 65%F below Pbypass 1.01 0.66 1.05 1.75 23 at > 65%F below Pbypass 1.25 0.64 1.29 2.01 23 at < 65%F below Pbypass 1.11 0.64 1.14 1.81 NSS Insertion Times 100 0.30 1.00 0.39 1.35 90 0.31 1.00 0.40 1.35 80 0.32 0.98 0.40 1.35 AREVA NP Inc.

Controlled Documnt Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-17 Table 5.7 NEOC Base Case FWCF Transient Results ATRIUM 1OXM ATRIUM 10XM Supported ATRIUM-10 ATRIUM-10 Power ACPR LHGRFACp ACPR HFR TSSS Insertion Times 100 0.30 1.00 0.35 1.35 90 0.32 1.00 0.37 1.35 80 0.34 0.98 0.39 1.37 70 0.36 0.96 0.41 1.39 60 0.38 0.94 0.43 1.41 50 0.41 0.92 0.45 1.41 26 0.60 0.86 0.65 1.64 26 at > 65%F below Pbypass 1.39 0.44 1.51 2.51 26 at < 65%F below Pbypass 1.36

.0.46 1.51 2.42 23 at > 65%F below Pbypass

.1.49 0.42 1.56 2.64 23 at <65%F below Pbypass 1.46 0.42 1.56 2.55 NSS Insertion Times 100 0.25 1.00 0.30 1.32 90 0.27 1.00 0.32 1.32 80 0.30 0.98 0.35 1.33 70 0.33 0.96 0.38 1.36 60 0.36 0.94 0.41 1.38 50 0.39 0.92 0.43 1.40 26 0.59 0.86 0.63 1.63 AREVA NP Inc.

,ument Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-18 Table 5.8 EOCLB Base Case FWCF Transient Results ATRIUM 1OXM ATRIUM 10XM Supported ATRIUM-10 ATRIUM-10 Power ACPR LHGRFACP ACPR HFR Inse 100 0.30 90 0.32 80 0.34 70 0.36 60 0.38 50 0.41 26 0.60 26 at > 65%F below Pbypass 1.39 26 at < 65%F below PbypasS 1.36 23 at > 6 5 %P below Pbypass 1.49 23 at - 65%F below Pbypass 1.46 TSSS rtion Times 1.00 1.00 0.98 0.96 0.94 0.92 0.86 0.44 0.46 0.42 0.42 0.37 0.39 0.40 0.42 0.43 0.45 0.65 1.51 1.51 1.56 1.56 1.35 1.35 1.37 1.41 1.43 1.44 1.64 2.51 2.42 2.64 2.55 100 90 80 70 60 50 26 NSS Insertion Times 0.27 1.00 0.29 1.00 0.31 0.98 0.33 0.96 0.36 0.94 0.39 0.92 0.59 0.86 0.34 0.36 0.38 0.40 0.41 0.43 0.63 1.35 1.35 1.35 1.40 1.42 1.43 1.63 AREVA NP Inc.

Controlled

,ent Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-19 Table 5.9 Loss of Feedwater Heating Transient Analysis Results Power

(% rated) 100 90 80 70 60 50 40 30 23 ATRIUM 1OXM/

ATRIUM-10 ACPR 0.10 0.11 0.12 0.13 0.14 0.16 0.19 0.24 0.30 Table 5.10 Control Rod Withdrawal Error ACPR Results Analytical RBM Setpoint (without filter)

ACPR*

(%)

108 0.19 111 0.25 114 0.28 117 0.33 Results are for the most limiting of the ATRIUM 1OXM or ATRIUM-10 fuel in the core.

AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-20 Table 5.11 RBM Operability Requirements Applicable Thermal Power ATRIUM 1OXM,

(% rated)

ATRIUM 10 MCPR 1.73 TLO

_>29% and < 90%

17 L

1.75 SLO

>90%

1.53 TLO Table 5.12 Flow-Dependent MCPR Results Core Flow ATRIUM 1OXM ATRIUM-10

(% rated)

Limiting MCPR Limiting MCPR 31 1.66 1.67 40 1.59 1.58 50 1.56 1.51 60 1.50 1.45 70 1.37 1.37 80 1.31 1.31 90 1.25 1.24 100 1.18 1.17 107 1.11 1.11 AREVA NP Inc.

Controlled 1o-, 2rt ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 5-21 Table 5.13 Licensing Basis Core Average Axial Power Profile State Conditions for Power Shape Evaluation Power, MWt 2923.0 MICROBURN-B2 pressure, psia 1044.7 Inlet subcooling, Btu/Ibm 20.3 Flow, Mlb/hr 80.5 Control state ARO Core average exposure (EOCLB), MWd/MTU 33,159 Licensing Axial Power Profile (Normalized)

Node Power Top 25.2

'62 24 0.777 23' ~1.02 22 1.188 211 1.297 20 1.365 19, 1., 398.

18 1.408 16 1.406 15 1.380 14 1.320 13 1.331 12 1.285 1i 1.221~

10 1.154

9.

1.087 8

0.990 7

0.874 6

0.767 57 0.6447 4

0.532 3

0.452~

2 0.357 Bottom", 1 0.100 AREVA NP Inc.

Control Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 5-22 300.0 -

~1) 0 C

0

'1) 0~

200.0 -

Core Power Heat Flux Core Flow Steam Flow --

Feed Flow

-- -/

100.0

/

[I

.0-

-lonn

/7 I

.0 1.0 2.0 3.0 Time, (seconds) 4.0 5.0 Figure 5.1 EOCLB LRNB at 100P/104.5F - TSSS Key Parameters AREVA NP Inc.

Contrle D OCU me Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-23 0

N E

0 U)

U)

Figure 5.2 EOCLB LRNB at 100P/104.5F - TSSS Sensed Water Level AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-24 1300.0 cn 0~

ci) crc ci) 0~

Time, (seconds)

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

con Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 5-25 0

0 3 3.0 Time, (seconds)

Figure 5.4 EOCLB TTNB at 100P/104.5F - TSSS Key Parameters AREVA NP Inc.

o~~Ied Dnum rt Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-26 0

E 0

0~

U)1 Time, (seconds)

Figure 5.5 EOCLB TTNB at 100P/104.5F - TSSS Sensed Water Level AREVA NP Inc.

Cont~

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 5-27 0~

V 0)

(0 V

0~

Time, (seconds)

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

Controlled Document Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-28 400.0 300.0

~0 C:

200.0 100.0

.0

-100.0 Figure 5.7 EOCLB FWCF at 100P/104.5F - TSSS Key Parameters AREVA NP Inc.

Contrc-Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 5-29 0

N E

0, 0

(D Time, (seconds)

Figure 5.8 EOCLB FWCF at 100P/104.5F - TSSS Sensed Water Level AREVA NP Inc.

Conitrol'. '

, iument Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 5-30 0

(0 0-0)

(0 (0

0) 0~

16.0 20.0 Time, (seconds)

Figure 5.9 EOCLB FWCF at 100P/104.5F - TSSS Vessel Pressures AREVA NP Inc.

ANP-3061 (NP)

Brunswick Unit 1 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 1OXM LOCA analysis are presented in References 24 and 25. The ATRIUM 10XM PCT is 1871°F. The peak local metal water reaction is 0.99% and the core wide metal water reaction is < 0.46%. The SLO MAPLHGR multiplier is 0.80.

The LOCA analysis results for the previously loaded ATRIUM-10 fuel are presented in Reference 26 as supplemented by Reference 27.

The Brunswick LOCA radiological analysis implementing the alternative source term methodology was performed in consideration of ATRIUM 1OXM fuel in the core inventory source terms. Progress Energy has evaluated the radiological consequences of a LOCA and determined ATRIUM 10XM fuel does not significantly increase the radiological consequences relative to consideration of ATRIUM-10 fuel in the core inventory source term.

6.2 Control Rod Drop Accident (CRDA)

Brunswick Unit 1 uses a bank position withdrawal sequence (BPWS) including reduced notch worth rod pull to limit high worth control rod movements. A CRDA evaluation was performed for both A and B sequence startups consistent with the withdrawal sequence specified by Progress Energy. Reference 28 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 28.

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 supported by the Brunswick AST analysis. Progress Energy has determined the radiological release assumed in the current Brunswick CRDA AST analysis bounds 986 rod failures for core source terms based on ATRIUM 1OXM fuel, and a slightly larger number of failed rods for core source terms based on ATRIUM-10 fuel. The number of fuel rods estimated to exceed the fuel damage threshold is below 986 for all fuel designs. Therefore, the current Brunswick CRDA AST analysis remains applicable for all the above mentioned fuel types.

AREVA NP Inc.

nt~~.l' Dcument ANP-3061 (NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 6-2 Maximum dropped control rod worth, mk 10.29 Core average Doppler coefficient, Ak/k/°F

-10.0 x 10-6 Effective delayed neutron fraction 0.0052 Four-bundle local peaking factor 1.381 Maximum deposited fuel rod enthalpy, cal/g 180.9 Maximum number of rods exceeding 170 cal/g 182 6.3 Fuel and Equipment Handling Accident Progress Energy has determined the radiological release assumed in the current fuel handling accident (FHA) analysis implementing the alternative source term (AST) methodology bounds 163 and 161 rod failures for core source terms based on ATRIUM-10 and ATRIUM 1OXM fuel, respectively. 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. The number of failed rods for ATRIUM 1OXM fuel is 161. These results are consistent with the number of failed rods supported by the current Brunswick AST analysis. The analysis also shows that the slightly higher mass of the ATRIUM 1OXM fuel does not result in an increase in rod failures above 163 when dropped onto ATRIUM-10 fuel. Therefore, the current FHA AST analysis remains applicable for all the above mentioned fuel types.

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 29, the fuel loading error is characterized as an infrequent event. The acceptance criteria are 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 fuel mislocation error analysis for Brunswick Unit 1 Cycle 19. This analysis evaluated the impact of a mislocated assembly against potential fuel rod failure mechanisms due to increased LHGR and reduced CPR. Based on this analysis, the offsite dose criteria (a small fraction of 10 CFR 50.67) is conservatively satisfied. A dose consequence evaluation is not necessary since no rod approached the fuel centerline melt or 1% strain limits, and less than 0.1% of the fuel rods are expected to experience boiling transition which could result in a dryout induced failure.

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ANP-3061 (NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 6-3 6.4.2 Misoriented Fuel Bundle AREVA has performed a fuel assembly misorientation analysis for the ATRIUM 1OXM fuel assemblies in Brunswick Unit 1 Cycle 19. The analysis was performed assuming that the limiting assembly was loaded in the worst orientation (rotated 1800) and depleted through the cycle without operator interaction. AREVA has also performed a bounding fuel assembly misorientation analysis for 1OX10 fuel monitored with the SPCB critical power correlation. These analyses demonstrate that the small fraction of 10 CFR 50.67 offsite dose criteria is conservatively satisfied. A dose consequence evaluation is not necessary since no rod approached the fuel centerline melt or 1% strain limits and less than 0.1% of the fuel rods are expected to experience boiling transition.

AREVA NP Inc.

ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 7-1 7.0 Special Analyses 7.1 ASME Overpressurization Analysis This section describes the maximum overpressurization analyses performed to demonstrate compliance with the ASME Boiler and Pressure Vessel Code. The analysis shows that the safety/relief valves at Brunswick Unit 1 have sufficient capacity and performance to prevent the reactor vessel pressure from reaching the safety limit of 110% of the design pressure.

An MSIV closure analysis was performed with the AREVA plant simulator code COTRANSA2 (Reference 17) 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.

a The plant configuration analyzed assumed that one of the lowest setpoint SRVs was inoperable.

TSSS insertion times were used.

The initial dome pressure was set at the maximum allowed by the Technical Specifications, 1059.7 psia (1045 psig).

A fast MSIV closure time of 2.7 seconds was used.

Results of the limiting MSIV closure overpressurization analysis 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 1347 psig occurs in the lower plenum. The maximum dome pressure for the same event is 1310 psig. These peak pressure results have been adjusted to address NRC concerns associated with the void-quality correlation, exposure-dependent thermal conductivity, and Doppler effects. The results demonstrate that the maximum vessel pressure limit of 1375 psig and dome pressure limit of 1325 psig are not exceeded.

A sensitivity analysis was performed to determine the impact of additional drift on the SRV opening setpoint above the 3% identified in the plant Technical Specifications. Assuming all of AREVA NP Inc.

Cont olled Dc ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 7-2 the degraded valves are from the lowest setpoint SRV bank provides a conservative scenario, and bounds the situation where the drift occurs in the higher bank of SRVs. Results for the sensitivity analysis are presented in Table 7.2. 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 pressure (1500 psig). The ATWS overpressurization analyses were performed at 100% power at 99% and 104.5% flow. 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 steam flow increases until the 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 and an increase in core pressure thereby increasing the core power.

The following assumptions were made in the analyses:

a The analytical limit ATWS-RPT setpoint and function were assumed.

The recirculation pump inertia associated with the variable frequency drives is used.

To support operation with one SRVOOS, the plant configuration analyzed assumed that one of the lowest setpoint SRVs was inoperable.

All scram functions were disabled.

The initial dome pressure was set to the nominal pressure of 1045 psia.

The MSIV closure is based on a nominal closure time of 4.0 seconds for both events.

Results of ATWS overpressurization analyses are presented in Table 7.3. 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 1446 psig and the maximum dome pressure is 1428 psig. The peak pressure results have been adjusted to address NRC concerns associated with the void-quality correlation, exposure-dependent thermal conductivity, and Doppler effects. The results demonstrate that the ATWS maximum vessel pressure limit of 1500 psig is not exceeded.

AREVA NP Inc.

Co troe nent ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 7-3 A sensitivity analysis was performed to determine the impact of operation with additional SRV setpoint drift above the 3% assumed in the plant Technical Specifications. Assuming all of the degraded valves are from the lowest setpoint SRV bank provides a conservative scenario, and bounds the situation where the drift occurs in the higher bank of SRVs. Results for the sensitivity analysis are presented in Table 7.4. The results demonstrate that the ATWS maximum vessel pressure limit of 1500 psig is not exceeded for any of the scenarios considered.

7.2.2 Lonq-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 2 Cycle 20 (Reference 2) an evaluation was performed that concluded the introduction of ATRIUM 10XM 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 for Unit 1 Cycle 19. A previous analysis for the ATRIUM-10 fuel reached a similar conclusion (Reference 30).

Relative to 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.

7.3 Standby Liquid Control System In the event that the control rod scram function becomes incapable of rendering the core in a shutdown state, the standby liquid control (SLC) system is required to be capable of bringing the reactor from full power to a cold shutdown condition at any time in the core life. The Brunswick Unit 1 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 to support a coolant temperature of 360°F with a boron concentration equivalent to 720 ppm at 700F. 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 2.67% Ak/k.

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ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 7-4 7.4 Fuel Criticality The new fuel storage vault criticality analysis for ATRIUM 1OXM fuel is presented in Reference 31. The spent fuel pool criticality analysis for ATRIUM 1OXM fuel is presented in Reference 32. The ATRIUM 1OXM fuel assemblies identified for loading in Cycle 19 meet both the new and spent fuel storage requirements.

AREVA NP Inc.

Contro ed D ocumen Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 7-5 Table 7.1 ASME Overpressurization Analysis Results*

Maximum Peak Peak Vessel Maximum Neutron Heat Pressure Dome Flux Flux Lower-Plenum Pressure Event

(% rated)

(psig)

MSIV closure 131 1347 1310 (102P/104.5F)

Table 7.2 ASME Overpressurization Sensitivity Analysis Results*

Low Bank Maximum Pressure SRV (psig)

Number of Setpoint Lower Steam Event Valves Drift Plenum Dome 1

OOS MSIV closure (102P/104.51) 2

+6%

1359 1322 1

+10%

The peak pressure results include adjustments to address the NRC concerns discussed in Section 7.1.

AREVA NP Inc.

-1.:)curnent Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 7-6 Table 7.3 ATWS Overpressurization Analysis Results*

Maximum Vessel Peak Peak Pressure Maximum Neutron Heat Lower-Dome Flux Flux Plenum Pressure Event

(% rated)

(psig)

MSIV closure (100P/104.5F) 232 135 1427 1408 MSIV closure (100P/99F) 233 134 1427 1408 PRFO (100P/104.5F) 247 144 1446 1428 PRFO (100P/99F) 241 142 1445 1427 The peak pressure results include adjustments to address the NRC concerns discussed in Section 7.2.

AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 7-7 Table 7.4 ATWS Overpressurization Sensitivity Analysis Results*

Low Bank Maximum Pressure SRV (psig)

Number of Setpoint Lower Steam Event Valves Drift Plenum Dome 1

OOS PRFO PR 0F 2

+6%

1456 1437 (1

P/104.51F) 1

+10%

The peak pressure results include adjustments to address the NRC concerns discussed in Section 7.2.

AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 7-8 Qa) 0-Figure 7.1 MSIV Closure Overpressurization Event at 102PI104.5F - Key Parameters AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 7-9 0

Q)

Q) 0 Q) a)I 6.0 Time, (seconds)

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

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 7-10 U)

U)

Time, (seconds)

Figure 7.3 MSIV Closure Overpressurization Event at 102PI104.5F -Vessel Pressures*

The pressures presented in this figure do not include the adjustments associated with the NRC concerns discussed in Section 7.2.

AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 7-11 E

0 M'

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

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 7-12 250.0 200.0 150.0 100.0

")

0 Q)

Q)

CL

-50.0 Figure 7.5 PRFO ATWS Overpressurization Event at 100P/104.5F - Key Parameters AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 7-13 0

()

N 4)E2 0

0 Cr" 0

1) a)

a)

Time, (seconds)

Figure 7.6 PRFO ATWS Overpressurization Event at 100PI104.5F - Sensed Water Level AREVA NP Inc.

Conro~; iDocurnn Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 7-14 Sbuu.OI 1400.0-1300.0-U) 1200.0-1100.0-0-

1000,0-900.0 -

Dome Pressure P-Lower Plenum Pressure

v.

1

.0 10.0 20.0 30.0 Time, (seconds) 40.0 50.0 Figure 7.7 PRFO ATWS Overpressurization Event at 100P/104.5F - Vessel Pressures*

The pressures presented in this figure do not include the adjustments associated with the NRC concerns discussed in Section 7.2.

AREVA NP Inc.

Controlled Document Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 7-15 1000 800 E

-o 600 SRV Bank 1 SRV Bank 2 SRV Bank k

/-37

.0O i.0-

'.0

.0 400 200

.0 10.0 20.0 30.0 40.0 Time, (seconds)

Figure 7.8 PRFO ATWS Overpressurization Event at 100PI104.5F - Safety/Relief Valve Flow Rates 50.0 AREVA NP Inc.

ANP-3061(NP)

Brunswick Unit 1 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 1 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 1.12. 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 1OXM and ATRIUM-10 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 a feedwater temperature reduction of up to 1 10.3 0 F at rated power. MCPRP limits for single-loop operation are 0.01 higher for all cases.

MCPRf limits that protect against fuel failures during a postulated slow flow excursion 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 1OXM and 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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ontrolied Docu rmet ANP-3061(NP)

Brunswick Unit 1 Cycle 19 Revision 0 Reload Safety Analysis Page 8-2 The ATRIUM 1OXM LHGRFACp multipliers are determined using the RODEX4 thermal-mechanical methodology (Reference 33). The ATRIUM-10 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 1OXM and 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 feedwater temperatures consistent with a feedwater temperature reduction of up to 11 0.3 0F at rated power.

LHGRFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a postulated slow flow excursion. For ATRIUM 1OXM and ATRIUM-10 fuel, the LHGRFACf multipliers are presented in Tables 8.13 and 8.14 respectively, and are applicable for all Cycle 19 exposures and the EOOS conditions identified in Table 1.1.

8.3 MAPLHGR Limits The ATRIUM, 1OXM TLO MAPLHGR limits are presentel:din Table 8.15. For operation in SLO, a multiplier of 0.8 must be applied to the TLO MAPLHGR limits.

The ATRIUM-10 TLO MAPLHGR limits are presented in Table 8.15. For operation in SLO, a multiplier of 0.85 must be applied to the TLO MAPLHGR limits.

AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 8-3 Table 8.1 MCPRp Limits for NSS Insertion Times BOC to < NEOC*

EOOS Power ATRIUM 1OXM ATRIUM-10 Condition

(% rated)

MCPRp MCPRp 100.0 1.39 1.48 90.0 1.42 1.50 80.0 1.43 1.51 65.0 1.64 1.67 Bae50.0 1.68 1.71 case

> 65%F

-< 65%F

> 65%F

-< 65%F operation 50.0 1.92 1.78 1.98 1.87 26.0 2.33 2.19 2.38 2.26 26.0 2.57 2.54 2.70 2.70 23.0 2.67 2.64 2.75 2.75 100.0 1.42 1.51 90.0 1.46 1.53 80.0 1.49 1.57 65.0 1.64 1.67 TBVOOS 50.0 1.68 1.71

>65%F

-< 65%F

> 65%F

-< 65%F 50.0 1.92 1.78 1.98 1.87 26.0 2.33 2.19 2.38 2.26 26.0 3.11 2.85 3.22 2.98 23.0 3.31 3.11 3.39 3.19 100.0 1.39 1.48 90.0 1.42 1.50 80.0 1.45 1.51 65.0 1.64 1.67 FOS50.0 1.68 1.71

> 65%F

-< 65%F

> 65%F

-7 65%F 50.0 1.92 1.78 1.98 1.87 26.0 2.33 2.19 2.38 2.26 26.0 2.74 2.71 2.91 2.91 23.0 2.88 2.85 3.04 3.04 100.0 1.42 1.51 90.0 1.47 1.53 80.0 1.51 1.57 65.0 1.64 1.67 TBVOOS 50.0 1.68 1.73 and FHOOS

> 65%F

< 65%F

> 65%F

< 65%F 50.0 1.92 1.78 1.98 1.87 26.0 2.33 2.19 2.38 2.26 26.0 3.25 3.05 3.36 3.20 23.0 3.48 3.29 3.54 3.42 Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher.

AREVA NP Inc.

Controtied Document Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 8-4 Table 8.2 MCPRp Limits for TSSS Insertion Times BOC to < NEOC*

EOOS Power ATRIUM 1OXM ATRIUM-10 Condition

(% rated)

MCPRP MCPRP 100.0 1.44 1.54 90.0 1.46 1.54 80.0 1.47 1.55 65.0 1.64 1.67 Bae50.0 1.68 1.71 case

> 65%F

-6 65%F

> 65%F

< 65%F operation 50.0 1.93 1.78 2.00 1.87 26.0 2.34 2.20 2.39 2.26 26.0 2.57 2.54 2.70 2.70 23.0 2.67 2.64 2.75 2.75 100.0 1.47 1.56 90.0 1.51 1.58 80.0 1.53 1.61 65.0 1.64 1.67 TBOS 50.0 1.68 1.72

> 65%F

< 65%F

> 65%F

< 65%F 50.0 1.93 1.78 2.00 1.87 26.0 2.34 2.20 2.39 2.26 26.0 3.11 2.85 3.22 2.98 23.0 3.31 3.11 3.39 3.19 100.0 1.44 1.54 90.0 1.46 1.54 80.0 1.48 1.55 65.0 1.64 1.67 FHOOS 50.0 1.68 1.71

> 65%F

< 65%F

> 65%F

< 65%F 50.0 1.93 1.78 2.00 1.87 26.0 2.34 2.20 2.39 2.26 26.0 2.74 2.71 2.91 2.91 23.0 2.88 2.85 3.04 3.04 100.0 1.47 1.56 90.0 1.51 1.58 80.0 1.54 1.61 65.0 1.64 1.67 TBVOOS 50.0 1.69 1.76 and FHOOS

> 65%F

< 65%F

> 65%F

< 65%F 50.0 1.93 1.78 2.00 1.87 26.0 2.34 2.20 2.39 2.26 26.0 3.25 3.05 3.36 3.20 23.0 3.48 3.29 3.54 3.42 Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.01 higher.

AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 8-5 Table 8.3 MCPRp Limits for NSS Insertion Times BOC to < EOCLB*

EOOS Power ATRIUM 10XM ATRIUM-10 Condition

(% rated)

MCPRP MCPRP 100.0 1.41 1.53 90.0 1.44 1.54 80.0 1.45 1.55 65.0 1.64 1.67 Base 50.0 1.68 1.71 case

> 65%F

-< 65%F

> 65%F

< 65%F operation 50.0 1.92 1.78 1.98 1.87 26.0 2.33 2.19 2.38 2.26 26.0 2.57 2.54 2.70 2.70 23.0 2.67 2.64 2.75 2.75 100.0 1.44 1.55 90.0 1.48 1.57 80.0 1.50 1.60 65.0 1.64 1.67 TBVOOS 50.0 1.68 1.71

> 65%F

<- 65%F

> 65%F

-< 65%F 50.0 1.92 1.78 1.98 1.87 26.0 2.33 2.19 2.38 2.26 26.0 3.11 2.85 3.22 2.98 23.0 3.31 3.11 3.39 3.19 100.0 1.41 1.53 90.0 1.44 1.54 80.0 1.45 1.55 65.0 1.64 1.67 FOS50.0 1.68 1.71

> 65%F

< 65%F

> 65%F

< 65%F 50.0 1.92 1.78 1.98 1.87 26.0 2.33 2.19 2.38 2.26 26.0 2.74 2.71 2.91 2.91 23.0 2.88 2.85 3.04 3.04 100.0 1.44 1.55 90.0 1.48 1.57 80.0 1.51 1.60 65.0 1.64 1.67 TBVOOS 50.0 1.68 1.73 and FHOOS

> 65%F

< 65%F

> 65%F

< 65%F 50.0 1.92 1.78 1.98 1.87 26.0 2.33 2.19 2.38 2.26 26.0 3.25 3.05 3.36 3.20 23.0 3.48 3.29 3.54 3.42 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 0.01 higher.

AREVA NP Inc.

d Documeni Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 8-6 Table 8.4 MCPRp Limits for TSSS Insertion Times BOC to < EOCLB*

EOOS Power ATRIUM 10XM ATRIUM-10 Condition

(% rated)

MCPRP MCPRP 100.0 1.44 1.56 90.0 1.47 1.57 80.0 1.47 1.57 65.0 1.64 1.67 50.0 1.68 1.71 case

> 65%F

< 65%F

> 65%F

< 65%F operation 50.0 1.93 1.78 2.00 1.87 26.0 2.34 2.20 2.39 2.26 26.0 2.57 2.54 2.70 2.70 23.0 2.67 2.64 2.75 2.75 100.0 1.47 1.60 90.0 1.51 1.61 80.0 1.53 1.63 65.0 1.64 1.68 TBVOOS 50.0 1.68 1.73

>65%F

< 65%F

> 65%F

-< 65%F 50.0 1.93 1.78 2.01 1.88 26.0 2.34 2.20 2.40 2.27 26.0 3.11 2.85 3:23 2.99 23.0

.. 3.31 3.11 3.40 3.20 100.0 1.44 1.56 90.0 1.47 1.57 80.0 1.48 1.57 65.0 1.64 1.67 HOS50.0 1.68 1.71

> 65%F

-8 65%F

> 65%F

< 65%F 50.0 1.93 1.78 2.00 1.87 26.0 2.34 2.20 2.39 2.26 26.0 2.74 2.71 2.91 2.91 23.0 2.88 2.85 3.04 3.04 100.0 1.47 1.60 90.0 1.51 1.61 80.0 1.54 1.63 65.0 1.64 1.68 TBVOOS 50.0 1.69 1.77 and FHOOS

> 65%F

< 65%F

> 65%F

< 65%F 50.0 1.93 1.78 2.01 1.88 26.0 2.34 2.20 2.40 2.27 26.0 3.25 3.05 3.37 3.21 23.0 3.48 3.29 3.55 3.43 Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.01 higher.

AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 8-7 Table 8.5 MCPRp Limits for NSS Insertion Times FFTR/Coastdown*

EOOS Power ATRIUM 1OXM ATRIUM-10 Condition

(% rated)

MCPRP MCPRP 100.0 1.44 1.57 90.0 1.46 1.57 80.0 1.46 1.58 65.0 1.64 1.67 Base 50.0 1.68 1.71 case

> 65%F

< 65%F

> 65%F

< 65%F operation 50.0 1.92 1.78 1.98 1.87 26.0 2.33 2.19 2.38 2.26 26.0 2.74 2.71 2.91 2.91 23.0 2.88 2.85 3.04 3.04 100.0 1.45 1.57 90.0 1.49 1.58 80.0 1.51 1.61 65.0 1.64 1.67 TBVOOS 50.0 1.68 1.73

> 65%F

-< 65%F

> 65%F

< 65%F 50.0 1.92 1,.78 1.98 1.87 26.0 2.33 2.19 2.38 2.26 26.0 3.25 3.05 3.36 3.20 23.0 3.48 3.29 3.54 3.42 Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.01 higher.

AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 8-8 Table 8.6 MCPRp Limits for TSSS Insertion Times FFTR/Coastdown*

EOOS Power ATRIUM 10XM ATRIUM-10 Condition

(% rated)

MCPRP MCPRP 100.0 1.50 1.72 90.0 1.50 1.72 80.0 1.50 1.72 65.0 1.64 1.74 Base 50.0 1.68 1.78 case ope

> 65%F

-- 65%F

> 65%F

-- 65%F 50.0 1.93 1.78 2.07 1.94 26.0 2.34 2.20 2.46 2.33 26.0 2.74 2.71 2.98 2.98 23.0 2.88 2.85 3.11 3.11 100.0 1.50 1.72 90.0 1.51 1.72 80.0 1.54 1.72 65.0 1.64 1.74 TBVOOS 50.0 6%

169 1.83 65%F 65%F

> 65%F

< 65%F:

50.0 1.93 1.78 2.07 1.94 26.0 2.34 2.20 2.46 2.33 26.0 3.25 3.05 3.43 3.27 23.0 3.48 3.29 3.61 3.49 Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.01 higher.

AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 8-9 Table 8.7 Flow-Dependent MCPR Limits ATRIUM 1OXM and ATRIUM-10 Fuel Core Flow

(% of rated)

MCPRf 0.0 1.72 31.0 1.72 55.0 1.62 100.0 1.20 107.0 1.20 Table 8.8 Steady-State LHGR Limits Peak ATRIUM 1OXM ATRIUM-10 Pellet Exposure LHGR LHGR (GWd/MTU)

(kW/ft)

(kW/ft) 0.0 14.1 13.4 18.9 14.1 13.4 74.4 7.4 7.1 AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 8-10 Table 8.9 LHGRFACp Multipliers for NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM ATRIUM-10 Condition

(% rated)

LHGRFACp LHGRFACP 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.90 Base

> 65%F

-2 65%F

> 65%F

-< 65%F case 50.0 0.86 0.86 0.78 0.86 operation 26.0 0.68 0.72 0.65 0.73 26.0 0.44 0.46 0.48 0.50 23.0 0.42 0.42 0.46 0.47 100.0 1.00 0.91 90.0 1.00 0.91 50.0 0.92 0.85

> 65%F

< 65%F

> 65%F

< 65%F TBVOOS 50.0 0.86 0.86 0.78 0.85 26.0 0.68 0.72 0.65 0.73 26.0 0.41 0.46 0.41 0.45 23.0 0.37 0.42-0.3_9 -0.41 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.89

> 65%F

< 65%F

> 65%F

< 65%F 50.0 0.86 0.86 0.78 0.86 26.0 0.68 0.72 0.65 0.70 26.0 0.40 0.42 0.43 0.46 23.0 0.36 0.38 0.39 0.43 100.0 1.00 0.91 90.0 1.00 0.91 50.0 0.92 0.84 TBVOOS

> 65%F

< 65%F

> 65%F

< 65%F and FHOOS 50.0 0.86 0.86 0.78 0.84 26.0 0.68 0.72 0.65 0.70 26.0 0.36 0.42 0.38 0.41 23.0 0.32 0.37 0.35 0.37 AREVA NP Inc.

Cont o~

)

m Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 8-11 Table 8.10 LHGRFACp Multipliers for TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 1OXM ATRIUM-10 Condition

(% rated)

LHGRFACp LHGRFACP 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.89 Base

> 65%F

-< 65%F

> 65%F

-< 65%F case 50.0 0.86 0.86 0.76 0.85 operation 26.0 0.68 0.72 0.64 0.73 26.0 0.44 0.46 0.48 0.50 23.0 0.42 0.42 0.46 0.47 100.0 1.00 0.87 90.0 1.00 0.87 50.0 0.92 0.85

> 65%F 5 65%F

> 65%F

-< 65%F TBVOOS 50.0 0.86 0.86 0.76 0.85 26.0 0.68 0.72 0.64 0.73 26.0 0.41 0.46 0.41 0.45 23.0 0.37 0.42 0.39 0.41 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.88

> 65%F

< 65%F

> 65%F

< 65%F 50.0 0.86 0.86 0.76 0.85 26.0 0.68 0.72 0.64 0.70 26.0 0.40 0.42 0.43 0.46 23.0 0.36 0.38 0.39 0.43 100.0 1.00 0.87 90.0 1.00 0.87 50.0 0.92 0.83 TBVOOS

> 65%F

< 65%F

> 65%F

< 65%F and FHOOS 50.0 0.86 0.86 0.76 0.83 26.0 0.68 0.72 0.64 0.70 26.0 0.36 0.42 0.38 0.41 23.0 0.32 0.37 0.35 0.37 AREVA NP Inc.

Cointrolled DOCr Fmi Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 8-12 Table 8.11 LHGRFACp Multipliers for NSS Insertion Times FFTR/Coastdown EOOS Power ATRIUM 1OXM ATRIUM-10 Condition

(% rated)

LHGRFACP LHGRFACP 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.86 Base

> 65%F

-< 65%F

> 65%F

-6 65%F case 50.0 0.86 0.86 0.78 0.86 operation 26.0 0.68 0.72 0.65 0.70 26.0 0.40 0.42 0.43 0.46 23.0 0.36 0.38 0.39 0.43 100.0 1.00 0.91 90.0 1.00 0.91 50.0 0.92 0.84

> 65%F 5 65%F

> 65%F

-< 65%F 50.0 0.86 0.86 0.78 0.84 26.0 0.68 0.72 T

0.65 0.70 26.0

.:0.36 0.42 0.38 0.41 23.0 0.32 0.37 0.3t, 0.37 AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 8-13 Table 8.12 LHGRFACP Multipliers for TSSS Insertion Times FFTR/Coastdown EOOS Power ATRIUM 1OXM ATRIUM-10 Condition

(% rated)

LHGRFACp LHGRFACp 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.85 Base

> 65%F

-< 65%F

> 65%F s 65%F case 50.0 0.86 0.86 0.76 0.85 operation 26.0 0.68 0.72 0.64 0.70 26.0 0.40 0.42 0.43 0.46 23.0 0.36 0.38 0.39 0.43 100.0 1.00 0.86 90.0 1.00 0.86 50.0 0.92 0.83

> 65%F

< 65%F

> 65%F

- 65%F 50.0 0.86 0.86 0.76 0.83 26.0 0.68 0.72 0.64 0.70 26.0 0.36 0.42 0.38 0.41 23.0 0.32 0.37 0.35 0.37 AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061 (NP)

Revision 0 Page 8-14 Table 8.13 ATRIUM 1OXM LHGRFACf Multipliers All Cycle 19 Exposures Core Flow

(% of rated)

LHGRFACf 0.0 0.58 31.0 0.58 75.0 1.00 107.0 1.00 Table 8.14 ATRIUM-10 LHGRFACf Multipliers All Cycle 19 Exposures Core Flow

(% of rated)

LHGRFACf 0.0 0.85 31.0 0.85 65.0 1.00 107.0 1.00 AREVA NP Inc.

Brunswick Unit 1 Cycle 19 Reload Safety Analysis ANP-3061(NP)

Revision 0 Page 8-15 Table 8.15 AREVA Fuel MAPLHGR Limits Average Planar ATRIUM 1OXM ATRIUM-10 Exposure MAPLHGR MAPLHGR (GWd/MTU)

(kW/ft)

(kW/ft) 0.0 13.1 12.5 15.0 13.1 12.5 67.0 7.7 7.3 AREVA NP Inc.

Cont.olied Do umel t Brunsw Reload 9.0 1.

2.

3.

4.

5.

6.

7.

ick Unit 1 Cycle 19 Safety Analysis ANP-3061(NP)

Revision 0 Page 9-1 8.

9.

10.

11.

12.

13.

14.

15.

16.

17.

References ANP-3005(P) Revision 0, Brunswick Unit I Cycle 19 Fuel Cycle Design, AREVA NP, June 2011.

ANP-2956(P) Revision 0, Brunswick Unit 2 Cycle 20 Reload Safety Analysis, AREVA NP, October 2010.

51-9159796-000, "Brunswick Unit 1 Cycle 19 Calculation Plan," AREVA NP, July 2011.

ANP-2948(P) Revision 0, Mechanical Design Report for Brunswick ATRIUM IOXM Fuel Assemblies, AREVA NP, October 2010.

ANP-3027(P) Revision 0, ATRIUM IOXM Fuel Rod Thermal and Mechanical Evaluation for Brunswick Unit 1 Cycle 19 Reload BRKI-19, AREVA NP, November 2011.

ANP-2817(P) Revision 0, Mechanical Design Report for Brunswick Unit 1 Reload BRK1-18 ATRIUM-10 Fuel Assemblies, AREVA NP, August 2009.

Letter, Edmond G. Touringy (NRC) to E.E. Utley (CP&L), "Issuance of Amendment No. 124 to Facility Operating License No. DPR Brunswick Steam Electric Plant, Unit 1, Regarding Fuel Cycle No. 7 Reload (TAC No. 69200)," February 6, 1989 (38-9061815-000).

ANP-2989(P) Revision 0, Brunswick Unit I Thermal-Hydraulic Design Report for ATRIUM TM IOXM Fuel Assemblies, AREVA NP, May 2011.

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

ANP-10298(P)(A) Revision 0, ACE/ATRIUM IOXM Critical Power Correlation, AREVA NP, March 2010.

EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.

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

BAW-10255PA Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008.

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

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.

ANP-2875(P) Revision 0, Brunswick Unit 1 Cycle 18 Reload Safety Analysis, AREVA NP, December 2009.

ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, CO TRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.

AREVA NP Inc.

ANP-3061 (NP)

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

18.

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

19.

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

20.

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

21.

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.

22.

Operating License and Technical Specifications, Brunswick Steam Electric Plant, Unit No 1, Progress Energy, as amended.

23.

ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, September 2005.

24.

ANP-2941 (P) Revision 0, Brunswick Units I and 2 LOCA Break Spectrum Analysis for ATRIUM TM 1OXM Fuel, AREVA NP, September 2010.

25.

ANP-2943(P) Revision 0, Brunswick Units I and 2 LOCA-ECCS Analysis MAPLHGR Limit for A TRIUM TM 10XM Fuel, AREVA NP, September 2010.

26.

ANP-2624(P) Revision 2,. Brunswick Units I and 2 LOCA-ECCS Analysis MAPLHGR Limit forATRIUMTM-IO Fuel, AREVA NP, October 2009.

27.

51-9143783-000, "10 CFR 50.46 PCT Model Change Reporting for the Brunswick Units,"

AREVA NP, August 2010.

28.

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.

29.

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.

30.

ANP-2674(P) Revision 3, Brunswick Unit 1 Cycle 17 Reload Safety Analysis, AREVA NP, February 2009.

31.

ANP-2962(P) Revision 0, Brunswick Nuclear Plant New Fuel Storage Vault Criticality Safety Analysis for A TRIUM TM IOXM Fuel, AREVA NP, October 2011.

32.

ANP-2955(P) Revision 3, Brunswick Nuclear Plant Spent Fuel Pool Criticality Safety Analysis for A TRIUMTM IOXM Fuel, AREVA NP, October 2011.

33.

BAW-1 0247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.

AREVA NP Inc.

BSEP 12-0040, ANP-3061(NP), Revision 0, Brunswick, Unit 1, Cycle 19 Reload Safety Analysis

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