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ANP-3280(NP), Revision 1, Brunswick Unit 1 Cycle 19 Mellla+ Reload Safety Analysis.
	ML16257A404
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Site:	Brunswick
Issue date:	05/31/2016
From:	; AREVA
To:	; Office of Nuclear Reactor Regulation
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ML16257A418	List: BSEP 16-0056, ANP-3108(NP), Revision 1, Applicability of Areva BWR Methods to Brunswick Extended Power Flow Operating Domain.; ML16257A404; ML16257A406; ML16257A407; ML16257A408; ML16257A410; ML16257A411;
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	BSEP 16-0056 ANP-3280(NP), Rev 1
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ANP-3280NP, Revision 1, Brunswick Unit 1Cycle19 MELLLA+ Reload Safety Analysis, May 2016 BSEP 16-0056 Enclosure 16 Controlled Document A AREVA Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis May 2016 (c) 2016 AREVA Inc. ANP-3280NP Revision 1 AREVA Inc. Controlled Document Copyright© 2016 AREVA Inc. All Rights Reserved ANP-3280NP Revision 1 Controlled DocumeQt Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page i Item Page 1. 1-1 2. 1-1 3. 1-2 4. 4-1 5. 4-2 6. 4-2 thru 4-3 and Tables 4.3 thru 4.5 7. Table 4.6 8. 5-1 9. 5-8 10. 5-9 11. 6-1 12. 7-3 13. 8-1 14. 9-2 AREVA Inc. Nature of Changes Description and Justification Updated information about the fuel assembly types that will be in the core when MELLLA+ is approved and implemented.

Added discussion on the channel bow model with respect to the SAFLIM3D methodology at the end of Section 1.0. Updated the second footnote to state that MSIVOOS was not analyzed and is prohibited for Brunswick Unit 1 Cycle 19 operation while in MELLLA+. Added discussion on the channel bow model with respect to the SAFLIM3D methodology in Section 4.2 (added as the third paragraph in Section 4.2). In the last paragraph of Section 4.2 "ACE/ATRIUM 1 OXM methodology" was replaced with "SAFLIM3D methodology

.. " Updated Section 4.3 to reflect the implementation of DSS-CD solution for core hydrodynamic stability for MELLLA+ operation.

Tables 4.3 thru

4.5 were updated and Table 4.6 was removed. Table 4.7 from Revision 0 of this document was re-numbered to Table 4.6. Updated reference in Section 5,0 for neutronics methodology report to Reference

27. Added sentence to the end of Section 5.3 to point to the tables that include plant thermal-mechanical response during the limiting transient for each allowed EOOS scenario.

Updated the last sentence of Section 5.3.5 to state that MSIVOOS was . not analyzed for MELLLA+ and is therefore not allowed while operating in MELLLA+ during Brunswick.Unit 1 Cycle 19. Updated LOCA results in Section 6.1 to reflect current analyses.

Also added text providing confirmation that the LOCA transition statepoint does not have a flow-dependent LHGR setdown. Updated *section 7.2.2 to reflect current MELLLA+ analyses.

' ' Last sentence of the first paragraph of. Section 8.2 was updated to state that LHGR multipliers meet the acceptance criteria for both U0 2 and gad bearing rods. Updated References 1, 11-15, 17, 25, 26, and 29.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page ii Contents 1.0 Introduction

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

1-1 2.0 Disposition of Events for ATRIUM 10XM Fuel lntroduction

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

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

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

4-1 4.1 Thermal-Hydraulic Design and Compatibility

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

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

.4-1 4.3 Core Hydrodynamic Stability

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

.4-2 4.3.1 MELLLA+ Stability DSS-CD Solution ..................................................

.4-2 4.3.2 MELLLA+ DSS-CD Backup Stability Protection

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

.4-2 4.4 Voiding in the Channel Bypass Region .............................................................

.4-3 5.0 Anticipated Operational Occurrences

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

5-1 5.1 System Transients

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

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

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

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

- - '-..................................

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

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

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

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

..........

5-6 5.3 Equipment Out-of-Service Scenarios

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

5-7 5.3.1 FHOOS .................................................................................................

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

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

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

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

5-9 5.3.6 Single-Loop Operation

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

5-9 5.4 Licensing Power Shape .................................................................................. 5-1 O 6.0 Postulated Accidents

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

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

6-1 6.2 . Control Rod Drop Accide_nt (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 ASME Overpressurization Analysis ....................................................................

7-1 7.2 ATWS Event Evaluation

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

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

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

7-2 7.2.2 Long-Term Evaluation

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

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

7-3 7 .4 Fuel Criticality

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

7-3 7.5 Rod Out Shutdown Margin .................................................................

7-4 AREVA Inc.

Controlled Document I Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page iii 8.0 Operating Limits and COLR lnput .................................................................. ...............

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

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

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

8-2 9.0 References

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

9-1 AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document Tables ANP-3280NP Revision 1 Page iv Table 1.1 MELLLA+ EOOS Operating Conditions

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

1-2 Table 4.1 Fuel-and Plant-Related Uncertainties for Safety Limit MCPR Analyses ..................

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

.4-5 Table 4.3 DSS-CD BSP Endpoints For Nominal Feedwater Temperature

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

.4-6 Table 4.4 DSS-CD BSP Endpoints For Reduced Feedwater Temperature

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

.4-7 Table 4.5 ABSP Setpoints for the Scram Region ....................................................................

.4-8 Table 4.6 Maximum Bypass Voiding at LPRM Level D ....... : ....................................................

4-9 Table 5.1 Exposure Basis for Brunswick Unit 1 Cycle 19 Transient Analysis .........................

5-1 i Table 5.2 Scram Speed Insertion Times .................................................................................

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

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

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

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

5-15 Table 5.6 EOCLB Base Case TTNB Transient Results ............................................. , .............

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

5-17 Table 5.8 EOGLB Base Case FWCF Transient Results .........................................................

5-18 Table 5.9 NEOC FHOOS FWCF Transient Results ...............................................................

5-19 Table 5.10 EOCLB FHOOS FWCF Transient Results ............................................................

5-20 Table 5.11 NEOC !BVOOS FWCF Transient Results ...........................................................

5-21 Table 5.12 EOCLB TBVOOS FWCF Transient Results .........................................................

5-22 Table 5.13 NEOC FHOOS/TBVOOS FWCF Transient Results .............................................

5-23 Table 5.14 EOCLB FHOOSITBVOOS FWCF Transient Results ..........................................

.5-24 Table 5.15 Loss of Feedwater Heating Transient Analysis Results ........................................

5-25 Table 5.16 Control Rod Withdrawal Error Results .......................................................

5-25 Table 5.17 RBM Operability Requirements

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

5-26 Table 5.18 Flow-Dependent MCPR Results ...........................................................................

5-26 Table 5.19 Licensing Basis Core Average Axial Power Profile ...............................................

5-27 Table 7.1 ASME Overpressurization Analysis Results* .............................................................

i'-5 Table 7.2 ASME Overpressurization Sensitivity Analysis Results**

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

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

7-6 Table 7.4 ATWS Overpressurization Sensitivity Analysis Results* ............

.-..............................

7-7 Table 8.1 MCPRp Limits for NSS Insertion Times BOC.to< NEOC .........................................

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

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

8-5 AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Pagev Table 8.4 MCPRp Limits for TSSS Insertion Times B,OC to < EOCLB .. *********.********

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

8-6 Table 8.5 MCPRp Limits for NSS Insertion Times FFTR/Coastdown-

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

8-7 Table 8.6 MCPRp Limits for TSSS Insertion Times FFTR/Coastdown-

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

8-8 Table 8. 7 Flow-Dependent MCPR Limits ..................................................................................

8-9

Table 8.8 Steady-State LHGR Limits .........................................................................................

8-9 Table 8.9 LHGRFACp Multipliers for NSS Insertion Times BOC to< EOCLB ........................

8-10 Table 8.10 LHGRFACp Multipliers for TSSS Insertion Times BOC to< EOCLB ....................

8-11 Table 8.11 LHGRFACp Multipliers for NSS Insertion Times FFTR/Coastdown

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

8-12 Table 8.12 LHGRFACp Multipliers for TSSS Insertion Times FFTR/Coastdown

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

8-13 Table 8.13 ATRIUM 10XM LHGRFACt Multipliers All Cycle 19 Exposures

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

8-14 Table 8.14 AREVA Fuel MAPLHGR Limits .............................................................................

8-14 AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document Figures ANP-3280NP Revision 1 Page vi Figure 1.1 Brunswick Unit 1 Power/Flow Map ..........................................................................

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

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

5-28 Figure 5.2 EOCLB LRNB at 100P/104.5F -TSSS Sensed Water Level. ...............................

5-29 Figure 5.3 EOCLB LRNB at 1OOP/104.5F

-TSSS Vessel Pressures

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

5-30 Figure 5.4 EOCLB TTNB at 1 OOP/104.5F

-TSSS Key Parameters

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

5-31 Figure 5.5 EOCLB TTNB at 100P/104.5F -TSSS Sensed Water Level. ...............................

5-32 Figure 5.6 EOCLB TTNB at 100P/104.5F -TSSS Vessel Pressures

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

5-33 Figure 5. 7 EOCLB FWCF at 100P/104.5F

-TSSS Key Parameters

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

5-34 Figure 5.8 EOCLB FWCF at 1OOP/104.5F

-TSSS Sensed Water Level. ..............................

5-35 Figure 5.9 EOCLB FWCF at 100P/104.5F -TSSS Vessel Pressures

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

5-36 Figure 7.1 MSIV Closure Overpressurization Event at 102P/104.5F

-Key Parameters

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

7-8 Figure 7.2 MSIV Closure Overpressurization Event at 102P/104.5F

-Sensed Water Level ...............................................................................................

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

7-9 Figure 7.3 MSIV Closure Overpressurization Event at 102P/104.5F

-Vessel Pressures

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

7-1 O Figure 7.4 MSIV Closure Overpressurization Event at 102P/104.5F

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

7-11 Figure 7.5 PRFO ATWS Overpressurization Event at 100P/85F -Key Parameters

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

7-12 Figure 7.6 PRFO ATWS Overpressurization Event at 100P/85F -Sensed Water Level ........................ , ....................................................................................................

7-13 Figure 7.7 PRFO ATWS Overpressurization Event at 100P/85F -Vessel Pressures

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

7-14 Figure 7.8 PRFO ATWS Overpressurization Event at 100P/85F -Safety/Relief Valve Flow Rates .........................................................................................................

7-15 AREVA Inc.

r Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ABSP APRM AOO ARO ASME AST ATWS ATWS-RPT BOC BPWS BSEP BSP CDA CFR COLR CPR CRDA CRWE DSS-CD EFPD EFPH EOC EOCLB EOFP EOOS FFTR FHOOS FWCF GE GSF HFCL ICF LFWH LHGR LHGRFACt LHGRFACp LOCA LPRM LRNB AREVA Inc. Nomenclature automated backup stability protection average power range monitor 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 confirmation density algorithm Code of Federal .Regulations core operating limits report critical power ratio control rod drop accident control rod withdrawal error detect and suppress solution.-

confirmation density 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 generic shape function high flow control line 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 ANP-3280NP Revision 1 Page vii Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis MAPLHGR MCPR MCPR, MCPRp MELLLA MELLLA+ MSIV MSIVOOS NCL NEOC NSS NRC OLMCPR OPRM Pbypass PCT PLU PRFDS PRFO RBM RDF RHR RPT RTP SLC SLMCPR SLO SRV SRVOOS STP TBVOOS TCV TIP TLO TSSS TSV TTNB UFSAR L\CPR AREVA Inc. Nomenclature (Continued) maximum average planar linear heat generation rate minimum crit.ical power ratio flow-dependent minimum critical power ratio power-dependent minimum critical power ratio maximum extended load line limit analysis maximum extended load line limit analysis plus main steam isolation valve main steam isolation valve out-of-service natural circulation line near end-of-cycle nominal scram speed Nuclear Regulatory Commission, U.S. operating limit minimum critical power ratio oscillation power range monitor power below which direct scram on TSV/TCV closure is bypassed peak cladding temperature power load unbalance pressure regulator failure downscale pressure regulator failure open (control) rod block monitor rated drive flow residual heat removal recirculation pump trip rated thermal power standby liquid control safety limit minimum critical power ratio single-loop operation safety/relief valve

safety/relief valve out-of-service simulated thermal power turbine bypass valves out-of-service turbine control ";alve

traversing incore probe two-loop operation technical specifications scram speed turbine stop valve turbine trip with no bypass updated final safety analysis report change in critical power ratio ANP-3280NP Revision 1 Page viii I' Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis 1.0 Introduction Controlled Document ANP-3280NP Revision 1 Page 1-1 Reload licensing analyses results generated by AREVA Inc. are presented in support of the Brunswick MELLLA+ licensing submittal.

The analyses reported in this document were performed using methodologies previously approved for generic application to t:>oiling water reactors and demonstrated in Reference 1 to be applicable to the MELLLA+ extended flow operating domain, Reference

2. The NRC technical requirements 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 assemb'lies, including 234 fresh ATRIUMŽ 10XM* assemblies and 326 irradiated ATRIUM-10 assemblies.

By the time MELLLA+ is approved and implemented in Brunswick, the ATRIUM-10 fuel will be completely discharged and the core will be 100% ATRIUM 10XM fuel. Therefore, separate ATRIUM-10 limits and analyses are not presented in this report. This licensing analysis for the ATRIUM 1 OXM supports the core design presented in Reference

3. The Cycle 19 reload licensing analyses were performed'for the potentially limiting events and analyses that were identified in the disposition of events (se.e §2.0). 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. The results in this report comply with the license condition related to the range of applicability for the channel bow model. This license condition was added with the inclusion of the SAFLIM3D methodology to the list of approved references in Section 5.6.5 (b) of the Brunswick Technical Specifications.

t< ATRIUM is a trademark of AREVA Inc. AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 1-2

t :j: § Table 1.1 MELLLA+ EOOS Operating Conditions*

Single-loop operation (SLO)t .:t: Turbine bypass valves out-of-service (TBVOOS) Feedwater heaters out-of-service (FHOOS)t One safety relief valve out-of-service (SRVOOS) ' One main steam isolation valve out-of-servicet.

§ (MSIVOOS)

One pressure regulator out-of-service**

Up to 40% of the TIP channels out-of-service (100% available at startup) Up to 50% of the LPRMs out-of-service Each EOOS condition is supported in combination with 1 SRVOOS, up to 40% of the TIP channels . out-of-service, and/or up to 50% of the LPRMs out-of-service.

Note that single-loop' operation and feedwater heaters out-of-service conditions are not allowed when operating in the MELLLA+ domain. For Brunswick Unit 1 Cycle 19, operation with MSIVOOS while in MELLLA+ was not analyzed and is prohibited.

Operation in SLO is only supported up to a maximum power level of 71.1 % of rated. 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 Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis 120.0 110.0 100.0 90.0 80.0 70.0 .. ., 60.0 a. Controlled Document / ,,,,.,, / MELLLA+ ............

v / v I \ MELLLA --h. v \ / \ / 50.0 / \ I 40.0 30.0 20.0 10.0 0.0 o.o 0 AREVA Inc. 7.7 10 Natural __} Circulation Linel/ 15.4 20 23.1 30 I I I I I I "' __.I l,..____ 35% Minimum Pump Minimuj Power Line I I 30.8 40 38.5 50 46.2 60 53.9 70 Core Flow 61.6' 80 Figure 1.1 Brunswick Unit 1 Power/Flow Map 69.3 90 c F v / 77.0 100 84.7 110 ANP-3280NP Revision 1 Page 1-3 92.4 Mlbs/hr 120 (%)

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document 2.0 Disposition of Events for ATRIUM 1 OXM Fuel Introduction ANP-3280NP Revision 1 Page 2-1 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 10XM fuel. Events and analyses identified as potentially limiting were either evaluated generically for the introduction of ATRIUM 1 OXM fuel or are performed on a cycle-specific basis. The results of the disposition of events are presented in Reference

4. AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document 3.0 Mechanical Design Analysis ANP-3280NP Revision 1 Page 3-1 The mechanical design analyses for ATRIUM 10XM are presented in the applicable mechanical design reports (References 5 and 6). The maximum expos'ure limits for the ATRIUM 1 OXM fuel are: 54.0 GWd/MTU average assembly exposure 62.0 GWd/MTU rod average exposure (full-length fuel rods) Even though the ATRIUM 10XM design is 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 LHGR limits are presented in Section 8.0. The fuel cycle design analyses (Reference

3) have v,erified that the ATRIUM 10XM fuel assemblies remain within licensed burnup limits. AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document 4.0 Thermal-Hydraulic Design Analysis 4.1 Thermal-Hydraulic Design af!d Compatibility ANP-3280NP Revision 1 Page 4-1 The results of the thermal-hydraulic characterization and compatibility analyses are presented in the thermal-hydraulic design report (Reference 8). The analyses performed to support the Reference 8 report 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 in the MELLLA+ operating domain. 4.2 Safety Limit MCPR Analysis The safety limit MCPR (SLM CPR) 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 MELLLA+ 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 used the ACE/ATRIUM 10XM critiC?al power correlation additive constants and additive constant uncertainty for ATRIUM 1 OXM fuel described in Reference

10. In the AREVA methodology, the effects of channel bow on the critical power performance are accounted for in the SLM CPR analysis; Reference 9 discusses the application of a realistic channel bow model. 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.

For TLO, analyses were performed for the minimum and maximum core flow conditions associated with rated power (85% and 104.5%), as well as the maximum core power at 55% core flow for the Brunswick power/flow map. For the maximum core flow statepoint, the TLO core flow uncertainty given in Table 4.1 was used. For the minimum core flow at full and 55% core flow statepoints, the SLO core flow uncertainty in Table 4.1 was used consistent with the restrictions listed in Section 2.2.1.1 of the Reference 2 Safety AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 . Page 4-2 Evaluation Report. The calculations performed are consistent with the currer:it Brunswick licensing restrictions for SAFLIM3D.

The results for the minimum supportable SLM CPR of 1.09 two-foop operation (TLO) and 1.11 single-loop operation (SLO) are shown in Table 4.2. Note that single-loop operation is not permitted in the MELLLA+ domain. Although the SAFLIM3D methodology was not implemented during actual Cycle 19 operations, Unit 1 Technical Specifications were subsequently amended to allow a two-loop operation (TLO) SLMCPR of 1.08 and a single-loop operation (SLO) SLM CPR of 1.11. The OLM CPR shown in Tables 8.1 through 8. 7 were developed assuming a TLO SLMCPR of 1.09 and a SLO SLMCPR of 1.11 .. 4.3 Core Hydrodynamic Stability 4.3.1 MELLLA+ Stability DSS-CD Solution Brunswick Unit 1 will implement the stability DSS-CD solution using the Oscillation Power Range Monitor (OPRM) as described in Reference

11. Plant-specific analyses for the QSS-CD Solution are provided in Reference

12. The Detect and Suppress function of the DSS-CD

solution based on the OPRM system relies on the Confirmation Density Algorithm (CDA), which the licensing basis. The Backup Stability Protection (BSP) solution may be used by the plant in the event that the OPRM system is declared inoperable.

The CDA enabled through the OPRM system and the BSP solution described in Reference 12 will be the stability licensing basis for Brunswick., The safety evaluation report for Reference 11 concluded that the DSS-CD solution is acceptable subject to certain limitations and conditions.

The reload DSS-CD evaluation is performed by Duke Energy in accordance with the licensing methodology described in Reference 11 to: 1) confirm the DSS-CD Solution is applicable to , Brunswick Unit 1 Cycle 19, and 2) confirm the Amplitude Discriminator Setpoint (SAo) of the CDA established in Reference 12 for operation in Brunswick Unit 1 Cycle 19. 4.3.2 MELLLA+ DSS-CD Backup Stability Protection

Reference 11 describe: s two -BSP options that are based on selected elements from three distinct constituents:

BSP Manual Regions, BSP Boundary, and Automated BSP (ABSP) setpoints.

! The Manual BSP region boundaries and the BSP Boundary were calculated for Brunswick Unit 1 Cycle 19 using STAIF (Reference

16) for nominal and reduced feedwater temperature AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document*

ANP-3280NP Revision 1 Page 4-3 operation.

The endpoints of the regions are defined in Table 4.3 and Table 4.4 for nominal and reduced feedwater temperature, respectively.

The Manual BSP region boundary endpoints are connected using the Generic Shape Function (GSF). The BSP Boundary for nominal and reduced feedwater temperature is defined by the MELLLA boundary line, per Reference

11. The ABSP Average Power Range Monitor (APRM) Simulated Thermal Power (STP) setpoints associated with the ABSP Scram Region are listed in Table 4.5. These ABSP setpoints are applicable to both TLO and SLO as well as nominal and reduced feedwater temperature operation.

4.4 Voiding

in the Channel Bypass Region To demonstrate compliance with the NRC's requirement that there be less than 5% bypass voiding around the LPRMs (see Section 5.1.1.5.1 of the Reference 2 Safety Evaluation), the bypass void level has been evaluated throughout the cycle. The maximum bypass void value applicable to the Cycle 19 design [ [ [ ] AREVA Inc. ] ]

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis [ [ AREVA Inc. Table 4.1 Fuel-and Plant-Related Uncertainties for Safety Limit MCPR Analyses Parameter Uncertainty Fuel-Related Uncertainties Plant-Related Uncertainties Feedwater flow rate Feedwater temperature Core pressure Total core flow rate TLO SLO 1.8% 0.8% 0.8% 2.5% 6% ] ] ANP-3280NP Revision 1 Page 4-4 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Table 4.2 Results Summary for Safety Limit MCPR Analyses Power/Flow Minimum Percentage

(%) Supported of Rods in Boiling SLM CPR* Transition 100/104.5 TL0-1.07 0.061 100/85 TL0-1.07 0.097 80/55 TL0-1.09 0.085 71.1/58 SL0-1.09 0.083 ANP-3280NP Revision 1 Page 4-5

The OLMCPR shown in Tables 8.1 through 8.7 developed assuming a TLO SLMCPR of 1.09 and a SLO SLMCPR of 1.11. AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document Table 4.3 DSS-CD BSP Endpoints For Nominal Feedwater Temperature Endpoint Power Flow Definition

(%) (%) A1 57.0 40.6 Scram Region Boundary, HFCL B1 42.0 31.7 Scram Region Boundary, NCL Controlled Entry A2 64.5 50.0 Region Boundary, HFCL Controlled Entry B2 28.9 31.9 Region Boundary, NCL AREVA Inc. ANP-3280NP Revision 1 Page 4-6 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document Table 4.4 DSS-CD BSP Endpoints For Reduced Feedwater Temperature Endp.oint Power Flow Definition

(%) (%) A1 65.9 51.8 Scram Region Boundary, HFCL B1 36.5 31.9 Scram Region Boundary, NCL Controlled Entry A2 69.8 56.8 Region Boundary, HFCL Controlled Entry B2 28.9 31.9 Region Boundary, NCL AREVA Inc. ANP-3280NP Revision 1 Page 4-7 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document Table 4.5 ABSP Setpoints for the Scram Region AREVA Inc. Parameter Slope of ABSP APRM biased trip linear segment. ABSP APRM flow-biased trip setpoint power intercept.

Constant Power Line for Trip from zero Drive Flow to Flow Breakpoint value. ABSP APRM flow-biased trip setpoint drive flow intercept.

Constant Flow Line for Trip. Flow Breakpoint value Symbol mTRIP PssP-TRIP WssP-TRIP WssP-BREAK Value 2.00 42.0 %RTP ;::: 37.5 %RDF 25.0 %RDF ANP-3280NP Revision 1 Page 4-8 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Table 4.6 Maximum Bypass Voiding at LPRM Level D* Power(%) Flow(%) Condition

[ Cycle Exposure (GWd/MTU)

Bypass Void (%) ]

The voiding at LPRM level D bounds the voiding at LPRM levels A, B, and C. AREVA Inc. ANP-3280NP Revision 1 Page 4-9 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document 5.0 Anticipated Operational Occurrences

/' ANP-3280NP Revision 1 Page 5-1 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.

I COTRANSA2 (Reference 18),

19), XCOBRA (Reference 20), and CASM0-4/MICROBURN-B2 (Reference

21) are the major codes used in the thermal limits analyses as described in the AREVA THERMEX methodology report (Reference

20) and neutronics methodology report (Reference 27). 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 10XM critical power correlation (Reference

10) is used to evaluate the thermal margin for the ATRIUM 1 OXM fuel. Fuel to-cladding gap conductance values are based on RODEX2 (Reference

22) 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 Referen.ce 23 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 (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 (18,661 MWd/MTU).

Analyses were also performed to support extended cycle operation with AREVA Inc.

Brunswick Unit 1 Cycle 19 *MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 5-2 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 10°F from the nominal feedwater temperature are considered base case operation, not an EOOS condition. This decrease in feedwater temperature causes a small increase in the core inlet subcooling which changes the axial power shape and core void fraction.

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

The consequences of the FWCF event 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 on the FWCF event. While a decrease in steam flow tends to make the LRNB 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°F at rated power. Analyses were performed to support both nominal and constant rated dome pressure with combined FFTR/Coastdown operation to a cycle exposure of 20,414 MWd/MTU. The FWCF analyses were performed with the lowest feedwater temperature associated with the initial power level. Operation with FFTR is not allowed in the MELLLA+ extension of the Brunswick operating domain.* 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 Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 5-3 nominal scram speed (NSS) insertion times and the Technical Specifications scram speed (TSSS) insertion tif'!1es 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

23) allow for operation with up to 1 O "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 Rejection 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.

l The excursion of the core power due to the void collapse is terminated primarily by the reactor scram and revoiding of the core. For power levels less than 50% of rated, the LRNB analyses assume that the power load unbalance (PLU) is inoperable.

With the PLU inoperable, the LRNB sequence of 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 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.

1,.RNB analyses were performed for a range of power/flow conditions covering the full ; ICF/MELLLA+

regi;on 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 ' operatin'g

limits for both TSSS and NSS insertion times. Figures 5.1 -5.3 show the responses of various reactor and plant parameters during the limiting LRNB event initiated at 100% of rated power and 104.5% of rated core flow with TSSS insertion times. AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document 5.1.2 Turbine Trip No Bypass (TTNB) ANP-3280NP Revision 1 Page 5-4 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. 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 limiting 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 in*crease in feedwater flow due to a failure of the feedwater control system to maximum demand results in an increase in the water level and a decrease in the coolant temperature at the core inlet. The increase in core inlet subcooling causes an increase in core power. As the feedwater flow continues at maximum demand, the water level continues to rise and eventually l 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 covering the full ICF/MELLLA+

region to support generation of the thermal limits. Tables 5.7 and 5.8 present the base case limiting FWCF transient analysis results used to generate the NEOC and EOCLB operating limits for both TSSS and NSS insertion times. Figures 5. 7 -5.9 show the responses AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 5-5 of various reactor and plant parameters during the FWCF event initiated at 100% of rated power and 104.5% of rated core flow with TSSS insertion times. 5.1.4 Pressure Regulator Failure Downscale (PRFDS) The pressure regulator failure downscale event occurs when the pressure regulator fails and sends a signal to close all four turbine.control 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 setpoint is reached. Operation with one pressure regulator out-of-service is not supported for Brunswick over the entire power/flow map. However, Duke 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 100°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 the turbine bypass valves, so no occurs. For Brunswick Unit 1 Cycle 19, a AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 5-6 cycle-specific analysis*

was performed in accordance with the Reference 24 methodology to determine the change in MCPR for the event. The LFWH results are presented in Table 5.15. 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 5.16 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 with the LHGR limits and their associated multipliers presented in Section The recommended operability requirements for the unblocked CRWE results are shown in Table 5.17 -based on the two loop 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. For in the MELLLA region, 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.

This evaluat.ion included specific MICROBURN-82 computer runs at 104.5%P/100%F, 100%P/99%F, 100%P/85%F, 71.1 %P/58.4%F, and 77.6%P/55%F.

AREVA Inc.

Brunswick Unit'1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 5-7 XCOBRA is used to calculate the change in critical power ratio during a two-loop flow runup to the maximum flow rate. The MCPRt 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 5.18 . MCPRt limits that provide the required.protection are presented in Table 8.7. The MCPRt limits are applicable for all Cycle 19 exposures.

Flow runup analyses were performed with CASM0-4/MICROBURN-82 to determine dependent LHGR multipliers (LHGRFACt) for the ATRIUM 10XM 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 to remain constant during the event. The LHGRFACt multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a flow runup. The Cycle 19 LHGRFACt multipliers are in Table 8.13. 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 for MELLLA operation:

Feedwater heater out-of-service (FHOOS) -up to 110.3°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 Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 5-8 The following equipment out-of-service (EOOS) scenarios are supported for Brunswick Unit 1 Cycle 19 operation for MELLLA+ operation:

Turbine bypass valves out-of-service (TBVOOS)

One safety/relief valve out-of-service (One SRVOOS) Tables 5.9 through 5.14 present the limiting dCPR and LHGRFACp transient analysis results for each EOOS scenario used to generate the NEOC and EOCLB operating limits for both TSSS and NSS insertion times. 5.3.1 FHOOS The FHOOS scenario assumes a feedwater temperature reduction of 110.3°F at rated power and steam flow. The effect of the reduced feedwater temperature is an increase in the core inlet subcooling

_which can change the axial power shape and core void fraction.

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

Operation with FHOOS or the related FFTR scenario is not allowed in the MELLLA+ region. 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 I 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. This scenario is not allowed in the MELLLA+ region. AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis 5.3.4 One SRVOOS Controlled Document ANP-3280NP Revision 1 Page 5-9 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 than the flow at rated power when all MS IVs are available.

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

The main difference in operation with One MSIVOOS is that.the steam line pressure drop between the steam dome and the turbine valves is higher than if a'll MS IVs are available.

Since low steam line pressure drop is limiting for pressurization transients, the results of the pressurization events with all MS IVs 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 MS IVs *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. This scenario was not analyzed for MELLLA+ conditions and therefore is not allowed while in MELLLA+ during Brunswick Unit 1 Cycle 19. 5.3.6 Single-Loop Operation

Operation in SLO is only supported up to a maximum core flow of 45 Mlbm/hr which corresponds to a maximum power level of 71.1 % of rated at the MELLLA boundary.

In SLO, the two-loop operation and LHGRFAC multipliers remain applicable.

The only impacts on the MCPR, LHGR, and MAPLHGR limits for SLO are an increase of 0.02 in the SLM CPR as discussed in Section 4.2, and the application of an SLO MAPLHGR multiplier discussed in Section 8.3. The net result is a 0.02 increase in the base case MCPRp limits and a decrease in the MAPLHGR limit.. The same situation is true for the EOOS. scenarios.

Adding 0.02 to the corresponding two-loop operation EOOS MCPRp limits results in SLO MCPRp limits for the EOOS conditions.

The TLO EOOS LHGRFAC multipliers remain applicable in SLO. This scenario is not allowed in the MELLLA+ region. AREVA Inc. r /

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document 5.4 Licensing Power Shape ANP-3280NP Revision 1 Page 5-10 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 EOCL8 core average exposure of 33, 159 MWd/MTU is given in Table 5.19. 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 5.19 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 EOCL8. If the criteria cannot be fully met (i.e., not all 7 nodes are at a higher power than the licensing profile), the licensing basis may nevertheless remain valid but further assessment will be required.

The licensing basis power profile in Table 5.19 was calculated using the MICR08URN-82 code. Compliance analyses must also be performed using MICR08URN-82.

Note thatthe 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 Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis

Cycle Exposure at End of Interval (MWd/MTU) 0 16,500 18,661 20,414 Controlled Document Table 5.1 Exposure Basis for Brunswick Unit 1 Cycle 19 Transient Analysis Core Average Exposure (MWd/MTU)*

Comments 14,498 Beginning of cycle 30,998 Break point for exposure-dependent MCPRp limits (NEOC) 33, 159 Design basis rod patterns to EOFP + 15 EFPD (EOCLB) 34,912 Maximum licensing core exposure -including FFTR /Coastdown ANP-3280NP Revision 1 Page 5-11

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

AREVA Inc.

Controlled Docun1e'nt Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Table 5.2 Scram Speed Insertion Times Control Rod TSSS Position Time (notch) (sec) 48 (full-out) 0.000 48 0.200 46 0.440 36 1.080 26 1.830 6 3.350 O (full-in) 3.806 AREVA Inc. NSS Time (sec) 0.000 0.200 0.318 0.829 1.369 2.510 2.852 ANP-3280NP Revision 1 Page 5-12 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis AREVA Inc. Table 5.3 NEOC Base Case LRNB Transient Results ATRIUM 10XM Power TSSS Insertion Times 100 0.32 90 0.33 80 0.34 70 0.34 60 0.33 50 0.31 50 at > 65%F PLU inoperable 0.77 50 at s 65%F PLU inoperable 0.62 26 at > 65%F PLU inoperable 1.17 26 at s 65%F PLU inoperable 1.03 26 at > 65%F below Pbypass 1.17 26 at S 65%F below Pbypass 1.03 23 at > 65%F below Pbypass 1.25 23 at S 65%F below Pbypass 1.12 NSS Insertion Times 100 0.27 90 0.29 80 0.30 70 0.31 60 0.30 50 0.29 50 at > 65%F PLU inoperable 0.76 50 at s 65%F PLU inoperable 0.62 26 at > 65%F PLU inoperable 1.16 26 at s 65%F PLU inoperable 1.02 ATRIUM 10XM Supported LHGRFACµ 1.00 1.00 0.98 0.96 0.94 0.92 0.86 0.86 0.68 0.72 0.66 0.66 0.64 0.64 1.00 1.00 0.98 0.96 0.94 0.92 0.86 0.86 0.70 0.66 ANP-3280NP Revision 1 Page 5-13 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document Table 5.4 EOCLB Base Case LRNB Transient Results ATRIUM 10XM ATRIUM 10XM Supported Power LHGRFACp TSSS lnserlion Times 100 0.33 1.00 90 0.34 1.00 80 0.34 0.98 70 0.34 0.96 60 0.33 0.94 50 0.31 0.92 50 at > 65%F PLU inoperable 0.77 0.86 50 at :5 65%F PLU inoperable 0.62 0.86 26 at > 65%F PLU inoperable 1.17 0.68 26 at :5 65%F PLU inoperable 1.03 0.72 26 at > 65%F below Pbypass 1.17 0.66 26 at :5 65%F below Pbypass 1.03 0.66 23 at > 65%F below Pbypass 1.25 0.64 23 at :5 65%F below Pbypass 1.12 0.64 NSS lnserlion Times 100 0.29 1.00 90 0.31 1.00 80 0.32 0.98 70 0.33 0.96 60 0.32 0.94 50 0.30 0.92 50 at > 65%F PLU inoperable 0.76 0.86 50 at :5 65%F PLU inoperable 0.62 0.86 26 at > 65%F PLU inoperable 1.16 0.70 26 at :5 65%F PLU inoperable 1.02 0.66 AREVA Inc. ANP-3280NP Revision 1 Page 5-14 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis AREVA Inc. Table 5.5 NEOC Base Case TTNB Transient Results ATRIUM 10XM Power ilCPR TSSS Insertion Times 100 0.33 90 0.33 80 0.33 26 at > 65%F below Pbypass 1.18 26 at ::; 65%F below Pbypass 1.01 23 at > 65%F below Pbypass 1.25 23 at ::; 65%F below Pbypass 1.11 NSS Insertion Times 100 0.28 90 0.28 80 0.29 ATRIUM 10XM Supported LHGRFACp 1.00 1.00 0.98 0.66 0.66 0.64 0.64 1.00 1.00 0.98 ANP-3280NP Revision 1 Page 5-15 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Table 5.6 EOCLB Base Case TTNB

Transient Results ATRIUM 10XM Power Tsss Insertion Times 100 0.33 90 0.33 80 0.34 26 at > 65%F below Pbypass 1.18 26 at 65%F below Pbypass 1.01 23 at > 65%F below Pbypass 1.25 23 at 65%F below Pbypass 1.11 NSS Insertion Times 100 0.30 90 0.31 80 0.32 I ' AREVA Inc. ATRIUM 10XM Supported LHGRFACp 1.00 1.00 0.98 0.66 0.66 0.64 0.64 1.00 1.00 0.98 ANP-3280NP Revision 1 , Page 5-16 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis-AREVA Inc. Table 5.7 NEOC Base Case FWCF Transient Results ATRIUM 10XM Power L\CPR TSSS Insertion Times 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 :5 65%F below Pbypass 1.36 23 .at > 65%F below Pbypass 1.49 23 at :5 65%F below Pbypass 1.46 NSS Insertion Times 100 0.25 90 0.27 80 0.30 70 0.33 60 0.36 50 0.39 26 0.59 ATRIUM 10XM Supported LHGRFACp 1.00 1.00 0.98 0.96 0.94 0.92 0.86 0.44 0.46 0.42 0.42 1.00 1.00 0.98 0.96 0.94 0.92 0.86 ANP-3280NP Revision 1 Page 5-17 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document Table 5.8 EOCLB Base Case FWCF Transient Results ATRIUM 10XM ATRIUM 10XM Supported Power LHGRFACµ TSSS Insertion Times 100 0.30 1.00 90 0.32 1.00 80 0.34 0.98 70 0.36 0.96 60 0.38 0.94 50 0.41 0.92 26-0.60 0.86 26 at > 65%F below Pbypass 1.39 0.44 26 at :s; 65%F below Pbypass 1.36 0.46 23 at > 65%F below Pbypass 1.49 0.42 23 at :s; 65%F below Pbypass 1.46 0.42 NSS Insertion Times 100 0.27 1.00 90 0.29 1.00 80 0.31 0.98 70 0.33 0.96 60 0.36 0.94 50 -o.39 0.92 26 0.59 0.86 AREVA Inc. ANP-3280NP Revision 1 Page 5-18 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis AREVA Inc. Table 5.9 NEOC FHOOS FWCF Transient Results ATRIUM 10XM Power TSSS Insertion

"{imes 100 0.30 90 0.33 80 0.35 70 0.38 60 0.42 50 0.46 26 0.76 26 at > 65%F below Pbypass 1.56 26 at ::> 65%F below Pbypass 1.53 23 at > 65%F below Pbypass 1.70 23 at ::> 65%F below Pbypass 1.67 NSS Insertion Times 100 0.27 90 0.29 80 0.32 70 0.36 60 0-40 50 0.44 26 0.75 ATRIUM 10XM Supported LHGRFACp 1.00 1.00 0.98 0.96 0.94 0.92 0.86 0.40 0.42 0.36 0.38 1.00 1.00 0.98 0.96 0.94 0.92 0.86 ANP-3280NP Revision 1 Page 5-19 Controlled Docurnent Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis AREVA Inc. Table 5.10 EOCLB FHOOS FWCF Transient Results ATRIUM 10XM Power TSSS Insertion Times 100 0.30 90 0.33 80 0.35 70 0.38 60 0.42 50 0.46 26 0.76 26 at > 65%F below Pbypass 1.56 26 at :::;; 65%F below Pbypass 1.53 23 at > 65%F below Pbypass 1.70 23 at :::;; 65%F below Pbypass 1.67 NSS Insertion Times 100 0.28 90 0.30 80 0.32 70 0.36 60 0.40 50 0.44 26 0.75 ATRIUM 10XM Supported LHGRFACp 1.00 1.00 0.98 0.96 0.94 0.92 0.86 0.40 0.42 0.36 0.38 1.00 1.00 0.98 0.96 0.94 0.92 0.86 ANP-3280NP Revision 1 Page 5-20 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis AREVA Inc. Table 5.11 NEOC TBVOOS FWCF Transient Results ATRIUM 10XM Power Tsss Insertion Times 100 0.36 90 0.38 80 0.40 70 0.43 60 0.46 50 0.49 26 0.68 26 at > 65%F below Pbypass 1.93 26 at :5 65%F below Pbypass 1.67 23 at > 65%F below Pbypass 2.13 23 at :5 65%F below Pbypass 1.93 NSS Insertion Times 100 0.31 90 0.33 80 0.36 70 0.40 60 0.43 50 0.47 26 0.65 ATRIUM 10XM Supported LHGRFACp 1.00 1.00 0.98 0.96 0.94 0.92 0.86 0.41 0.48 0.37 0.42 1.00 1.00 0.98 0.96 0.94 0.92 0.86 ANP-3280NP Revision 1 Page 5-21 Controlled Document . Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis AREVA Inc. Table 5.12 EOCLB TBVOOS FWCF Transient Results ATRIUM 10XM Power TSSS Insertion Times 100 0.36 90 0.38 80 0.40 70 0.43 60 0.46 50 0.49 26 0.68 26 at > 65%F below Pbypass 1.93 26 at :s; 65%F below Pbypass 1.67 23 at > 65%F below Pbypass 2.13 23 at :s; 65%F below Pbypass 1.93 NSS Insertion Times 100 0.33 90 0.35 80 0.37 70 0.40 60 0.43 50 0.47 26 0.65 ATRIUM 10XM Supported LHGRFACp 1.00 1.00 0.98 0.96 0.94 0.92 0.86 0.41 0.48 0.37 0.42 1.00 1.00 0.98 0.96 0.94 0.92 0.86 ANP-3280NP Revision 1 Page 5-22 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document Table 5.13 NEOC FHOOS/TBVOOS FWCF Transient Results ATRIUM 10XM ATRIUM 10XM Supported Power LHGRFACp TSSS Insertion Times 100 0.35 1.00 90 0.38 1.00 80 0.41 0.98 70 0.45 0.96 60 0.48 0.94 50 0.53 0.92 26 0.77 0.84 26 at > 65%F below Pbypass 2.07 0.36 26 at :5 65%F below Pbypass 1.87 0.43 23 at > 65%F below Pbypass 2.30 0.32 23 at :5 65%F below Pbypass 2.11 0.37 NSS Insertion Times 100 0.31 1.00 90 0.34 1.00 80 0.38 0.98 70 0.42 0.96 60 0.46 0.94 50 0.51 0.92 26 0.75 0.86 AREVA Inc. ANP-3280NP Revision 1 Page 5-23 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document Table 5.14, EOCLB FHOOS/TBVOOS FWCF Transient Results ATRIUM 10XM ATRIUM 10XM Supported Power LHGRFACp Tsss Insertion Times 100 0.35 1.00 90 0.38 1.00 80 0.41 0.98 70 0.45 0.96 60 0.48 0.94 50 0.53 0.92 26 0.77 0.84 26 at> 65%F below Pbypass 2.07 0.36 26 at ::5 65%F below Pbypass 1.87 0.43 23 at > 65%F below Pbypass 2.30 0.32 23 at ::5 65%F below Pbypass 2.11 0.37 NSS Insertion Times 100 0.33 1.00 90 0.35 1.00 80 0.38 0.98 '70 0.42 0.96 60 0.46 0.94 50 0.51 0.92 26 0.75 0.86 AREVA Inc. ANP-3280NP Revision 1 Page 5-24 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis AREVA Inc. Table 5.15 Loss of Feedwater Heating Transient Analysis Results Power ATRIUM 10XM (%rated) LlCPR 100 0.10 90 0.11 80 0.12 70 0.13 60 0.14 50 0.16 40 0.19 30 0.24 23 0.30 Table 5.16 Control Rod Withdrawal Error Results Analytical RBM Setpoint ATRIUM 10XM (without filter) LlCPR (%) 108 0.19 111 0.25 114 0.28 117 0.33 ANP-3280NP Revision 1 Page 5-25 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document Table 5.17 RBM Operability Requirements Thermal Power ATRIUM 10XM (%rated) MCPR 1.71 (TLO) 29% and < 90% 1.74 (SLO) 1.51 (TLO)* Table 5.18 Flow-Dependent MCPR Results Core Flow ATRIUM 10XM (%rated) Limiting MCPR 31 1.61 40 1.54 50 1.51 60 1.46 70 1.34 80 1.29 90 1.23 100 1.16 107 1.09

Greater than 90% rated power is not attainable in SLO AREVA Inc. ANP-3280NP Revision 1 Page 5-26 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis AREVA Inc. Table 5.19 Licensing Basis Core Average Axial Power Profile State Conditions for Power Shape Evaluation Power, MWt 2923.0 MICROBURN-82 pressure, psia 1044.7 Inlet subcooling, Btu/lbm 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 0.254 24 0.752 23 0.990 22 1.148 21 1.252 20 1.317 19 1.352 18 1.370 17 1.355 16 1.388 15 1.372 14 1.320 13 1.338 12 1.299 11 1.242 10 1.179 9 1.113 8 1.018 7 0.903 6 0.799 5 0.680 4 0.570 3 0.490 2 0.390 Bottom 1 0.109 ANP-3280NP Revision 1 Page 5-27 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 5-28 400.0-,--------------------------------, Core Power Heat Flux Core Flow 300.0 Steam Flow -------Feed Flow -0 Q) 200.0 .... 0 Cl'.'. ..... 0 .... c Q) u L 100.0 Q) Q_ .0

.0 AREVA Inc. 1.0 2.0 3.0 4.0 Time, (seconds)

Figure 5.1 EOCLB LRNB at 100P/104.5F-TSSS Key Parameters

5.0 Controlled

Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ..--.. i:::: :.::;, 0 '-QJ N ...., 185.0 i:::: QJ E ::J '-...., en E 0 180.0 ...., ...., () QJ a. en QJ Cl::'. ..r: ...., 175.0 Qi > QJ _J '-QJ ...., 170.0 0 :;;: -0 QJ en i:::: QJ VJ AREVA Inc. .0 1.0 2.0 3.0 4.0 Time, (seconds)

Figure 5.2 EOCLB LRNB at 100P/104.5F-TSSS Sensed Water Level ANP-3280NP Revision 1 Page 5-29 5.0 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 5-30 1300.0.-------------------,,-,,.,,..,_;:--_-_-

.... -_-------------.

/ ____ .,,,.,,,..

1250.0 1200.0 ,..-.... c "(jj 8 ai 1150.0 L ::J "' "' Q) L Cl. 1100.0 1050.0 ___ _,, I I I I I I I I I I I I I I ,. / / ,,,. Dome Pressure I I / / / I I / I / / Lower Plenum Pressure -.... .... .... ....

.0 AREVA Inc. 1.0 2.0 3.0 4.0 Time, (seconds)

Figure 5.3 EOCLB LRNB at 100P/104.5F-TSSS Vessel Pressures

5.0 Brunswick

Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 5-31 300.0 ""O 2 200.0 0 Cl:'. -0 ...., c Q} u L.. Q} 0.. 100.0 .0 Core Power Heat Flux Core Flow Steam Flow -------Feed Flow -100.0-+------.,.---------..-------,--------.--------i

.0 AREVA Inc. 1.0 2.0 3.0 Time, (seconds)

Figure 5.4 EOCLB TTNB at 1OOP/104.SF

-TSSS Key Parameters 4.0 5.0 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis 190.0 ,...... 0 '-Q) N +' 185.0 c Q) E ;:J '-+' "' .E 0 180.0 +' ...., u Q) a. "' Q) ..c ...., 175.0 3: Qj > Q) _J '-Q) ...., 170.0 0 s: -0 Q) "' c Q) (/) 165.0 AREVA Inc. .0 1.0 2.0 3.0 Time, (seconds)

Figure 5.5 EOCLB TTNB at 100P/104.5F-TSSS Sensed Water Level 4.0 ANP-3280NP Revision 1 Page 5-32 5.0 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 5-33 0 "Cii -8 ai L.. ::::J en en Q) L.. 0... AREVA Inc.

1300.0 1250.0 1200.0 1150.0 1100.0 1050.0 I ___ I I I I I I I I I I I I _, I I I I I I ,.,,,,,,.---

.... /----/ ...... ______ _ I .._ // ................

I --I ' I ',, Dome Pressure Lower Plenum Pressure 1000.0-+--------.----------.--------,.-------,--------1

.0 1.0 2.0 3.0 4.0 Time, (seconds)

Figure 5.6 EOCLB TTNB at 100P/104.5F-TSSS Vessel Pressures

5.0 Controlled

Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 5-34 400.0-,------------------------------, 300.0--0 2 200.0 CJ 0:: .._ 0 ..., c Q) () ._ Q) Q 100.0 .o-Core Power Heat Flux Core Flow --------------

Steam Flow Feed Flow

, > ,,-------------------

! \ ' "\ f I Iv* I ) .\ 1 -100.0 -+------.-----.,,----.,...-----.-------,-----r-------l

.0 AREVA Inc. 5.0 10.0 15.0 20.0 Time, (seconds) 25.0 Figure 5.7 EOCLB FWCF at 100P/104.5F-TSSS Key Parameters 30.0 35.0 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis 220.0 ,...... c :.::;.. 0 I.... Q) N ..... c Q) 200.0 E ::::i I.... ..... (/'.) 0 ..... ..... u Q) Q. 180.0 (/'.) Q) 0:: ..c ..... Qi > Q) __J I.... 160.0 Q) ..... c s: -c Q) UJ c Q) UJ 140.0 .o AREVA Inc. 5.0 10.0 15.0 20.0 25.0 Time, (seconds)

Figure 5.8 EOCLB FWCF at 100P/104.5F-TSSS Sensed Water Level 30.0 ANP-3280NP Revision 1 Page 5-35 35.0 Controlled Docun1ent Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 5-36 ,......_ 0 "iii ..:; a) I... :::J en en Q) I... CL AREVA Inc. 1280.0-,---------------------------------., 1240.0 1200.0 1160.0 1120.0 1080.0 / I I I I I I I I I I I *1 I I I I I I I I I I I I I I I I I I I I I I , , , , I , , I I , , ________________________________ , Dome Pressure Lower Plenum Pressure 1040.0-+-------r-----,--------.----....------..-----....-------1

.0 5.0 10.0 15.0 20.0 Time, (seconds) 25.0 Figure 5.9 EOCLB FWCF at 100P/104.5F-TSSS Vessel Pressures 30.0 35.0 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis 6.0 Postulated Accidents Controlled Document 6.1 Loss-of-Coolant Accident (LOCA) ANP-3280NP Revision 1 Page 6-1 The results of the ATRIUM 1 OXM LOCA analysis that bound the ICF/MELLLA+

region are presented in References 25 and 26. For these analyses, an initial MCPR of 1.34 was assumed. The ATRIUM 10XM PCT is 1923°F. The peak local metal water reaction is 1.23% and the core wide metal water reaction is < 0.56%. The limiting LOCA statepoint was determined to be 102% of rated power and [ ] of rated flow. The SLO MAPLHGR multiplier is 0.80. The limits presented in Section 8 bound the rated and off-rated limits assumed in the Reference 25 and 26 analyses.

In addition, it is confirmed that a setdown to the ATRIUM 10XM flow-dependent LHGR is applied at core flows below 70% rated core flow (Table 8.13). This is consistent with the analyses performed in Reference

25. The Brunswick LOCA radiological analyl:?is implementing the alternative source term methodology was performed in consideration of ATRIUM 1 OXM fuel in the core inventory source terms. Duke 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 Duke Energy. Reference 27 describes the approved AREVA generic CRDA methodology.

Subsequent calculations have shown that the methodology is applicable to fuel modeled with the CASM04/MICROBURN-.B2 code system. The CRDA analysis was performed with the approved methodology described in Reference

27. The CRDA analysis results demonstrate that the maximum deposited fuel rod enthalpy is less than the NRC threshold of 280 cal/g and that the estimated number of fuel rods that exceed the fuel damage threshold of 170 cal/g is less than the number of failed rods supported by the Brunswick AST analysis.

Duke 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 10XM fuel. The number offfuel rods estimated to exceed the fuel damage threshold is AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 6-2 below 986 for all fuel designs. Therefore, the current Brunswick CRDA AST analysis remains applicable.

Maximum dropped control rod worth, mk Core average Doppler coefficient, Lik/kf'F Effective delayed neutron fraction Four-bundle local peaking factor Maximum deposited fuel rod enthalpy, cal/g Maximum number of rods exceeding 170 cal/g 6.3 Fuel and Equipment Handling Accident 10.29 -10.0 x 10"" 6 0.0052 1.381 180.9 182 Duke Energy has determined the radiological release assumed in the current fuel handling accident (FHA) analysis implementing the alternative source term (AST) methodology bounds 161 rod failures for core source terms based on ATRIUM 10XM fuel. AREVA has performed an analysis that shows that the number of failed fuel rods due to a fuel handling accident impacting the ATRIUM 1 OXM fuel is 161. These results are consistent with the number of failed rods supported by the current Brunswick AST analysis.

Therefore, the current FHA AST analysis remains applicable.

6.4 Fuel Loading Error (Infrequent Event) There are two types of fuel loading errors possible in a BWR: the mislocation of a fuel assembly in a core position prescribed to be loaded with another fuel assembly, and the misorientation of a fuel assembly with respect to the control blade. As described in Reference 28, the fuel loading error is characterized as an infrequent event. The acceptance criteria are that the offsite dose consequences due to the event shall not exceed a small fraction of the 10 CFR 50.6,7 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. AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document 6.4.2 Misoriented Fuel Bundle ANP-3280NP Revision 1 Page 6-3 AREVA has performed a fuel assembly misorientation analysis for the ATRIUM 1 OXM fuel assemblies in Brunswick Unit 1 Cycle 19. The analysis was performed assuming that the limiting assembly was loaded in the worst orientation (rotated 180°) and depleted through the cycle without operator interaction.

This analysis demonstrated that the small fraction of 1 O 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 Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis 7.0 Special Analyses Controlled Document 7 .1 ASME Overpressurization Analysis ANP-3280NP Revision 1 Page 7-1 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

18) for 102% power and 104.5% flow and 102% power and 85% 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 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.

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

TSSS insertion times were used.

Th'e 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 conductivify, and Doppler The results demonstrate that the maximum vessel pressure limit of 1375 psig and dome pressure limit of 1325 psig are .not exceeded.

AREVA Inc. , '

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 7-2 A sensitivity analysis was performed to deter,mine the impact of additional drift on the SRV opening setpoint above the 3% identified in the plant Technical Specifications.

Assuming all of the degraded valves are from the highest setpoint SRV bank provides a conservative scenario, and bounds the situation where the drift occurs in other 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 ATWS Event Evaluation 7.2.1 ATWS Overpressurization Analysis This section describes the analyses performed to demonstrate that the peak vessel pressure for the limiting ATWS event is less than the ASME Service Level C li111it of 120% of the design pressure (1500 psig). The A TWS overpressurization analyses were performed at 100% power at 85% 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:

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

The 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 the limiting 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 1452 psig and the maximum dome pressure is 1435 psig. The peak pressure results have been adjusted to address NRC concerns associated with the void-quality correlation, AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 7-3 exposure-dependent thermal conductivity, and Doppler effects. The results demonstrate that the ATWS maximum vessel pressure limit of 1500 psig is not exceeded.

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.

The limiting statepoint (100/85) from the base case was rerun with the degradation scheme shown* in Table 7.4. Results for the sensitivity analysis are also 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 Long-Term Evaluation ATWS (long-term and instability) for MELLLA+ has been analyzed in Reference

12. 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 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 70°F. The temperature of 360°F corresponds to the low pressure permissive for the RHR shutdown cooling suction valves, and represents the maximum reactivity condition with soluble boron in the coolant. The analysis shows the core to be subcritical throughout the cycle by at least 2.67%

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

30. The spent fuel pool criticality analysis for ATRIUM 10XM fuel is presented in

t This is a conservative representation of the shutdown margin when greater than 720 ppm natural boron equivalent at 70 °F is present in the core. Relative to a 95/95 uncertainty limit of 0.88%t.k/k.

AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 7-4 Reference

31. The ATRIUM 1 OXM fuel assemblies identified for loading in Cycle 19 meet both the new and spent fuel storage requirements.

7 .5 Strongest Rod Out Shutdown Margin Detailed results for the strongest rod out shutdown margin evaluation are reported in Section 3.3 of Reference

3. In summary, the BRK1-19 MELLLA+ core has a minimum strongest rod out shutdown margin of 1.29 %8k/k*. This value is produced at the beginning of the cycle at the minimum coolant temperature condition (55 °F). This value assumes that BRK1-18 ended operation at the lowest allowable exposure.

Relative to a design goal of 1%

AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 7-5

t :j: Event MSIV closure (102P/104.5F)

MSIV closure (102P/85F)

Event MSIV closure (102P/104.5F)

Table 7.1 ASME Overpressurization Analysis Results**t Maximum Peak Peak Vessel Neutron Heat Pressure Flux Flux Lower-Plenum

(%rated) (%rated) (psig) 287 131 1347 315 131 1332 Table 7.2 ASME Overpressurization Sensitivity Analysis Results***

Maximum Dome Pressure (psig) 1310 1302 Maximum Pressure SRV Bank (psig) Number of* Setpoint Lower . Steam Valves Drift Plenum Dome 1 oos 3 +4% 1q62 1324 2 +6% 1 +8% The peak pressure results include adjustments to address the NRC concerns discussed in Section 7 .1. The maximum Technical Specification allowed SRV degradation of 3% was assumed. The SRV degradation scheme is based on actual plant performance using a 95/95 approach.

AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Event MSIV closure (1OOP/104.5F)

MSIV closure (100P/85F)

PRFO (1OOP/104.5F)

PRFO (100P/85F)

Table 7.3 ATWS Overpressurization Analysis Results**t Maximum Vessel Peak Peak Pressure Neutron Heat Lower-Flux Flux Plenum (%rated) (%rated) (psig) 232 135 1427 236 131 1436 247 144 1446 226 137 1452 ANP-3280NP Revision 1 Page 7-6 Maximum Dome Pressure (psig) 1408 1419 1428 1435

The peak pressure results include adjustments to address the NRG concerns discussed in Section 7.2. t The maximum Technical Specification allowed SRV degradation of 3% was assumed. AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 7-7

t Event

PRFO (100P/85F)

Table 7.4 ATWS Overpressurization Sensitivity Analysis Results**t Maximum Pressure SRV Bank (psig) Number of Setpoint Lower Steam Valves Drift Plenum Dome 1 oos ' 3 +4% 1461 1445 2 +6% 1 +8% The peak pressure results.include adjustments to address the NRC concerns discussed in Section 7.2. The SRV degradation scheme is based on actual plant performance using a 95/95 approach.

AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 7-8 .... 0 ...., c Q) u Q) 0... AREVA Inc.

Core Power Heat Flux Core Flow Steam Flow Feed Flow 200.0 -100.0-r-----.------,-------.------.-----.--------i .o 2.0 4.0 6.0 8.0 10.0 Time, (seconds)

Figure 7.1 MSIV Closure Overpressurization Event at 102P/104.SF

-Key Parameters 12.0 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis 190.0 0 c -.:;.. 0 I.... Q) N ..... 185.0 c Q) E ::l I.... ..... (/l .!; 0 180.0 ..... ..... u Q) a. (/l Q) 0::: ..c ..... 175.0 Q3 > Q) _J I.... Q) ..... 170.0 0 ;: "U Q) (/l c Q) Cf) 165.0 .0 2.0 Controlled Document 4.0 6.0 8.0 10.0 Time, (seconds)

Figure 7.2 MSIV Closure Overpressurization Event at

102P/104.5F

-Sensed Water Level AREVA Inc. ANP-3280NP Revision 1 Page 7-9 12.0 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280N-P Revision 1 Page 7-10

1300.0 I I I I Iv 1' I I I 1"' ..... ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ai 1200.0 I I I I ' L. :J "' "' Q) L. Cl.. 1100.0 J ------I I I I I I I Dome Pressure Lower Plenum Pressure ' ' ' '

.0 2.0 4.0 6.0 8.0 10.0 Time, (seconds)

Figure 7.3 MSIV Closure Overpressurization Event at 102P/104.SF

-Vessel Pressures*

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

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 7-11 ........ CJ) ' E ..0 .;::., :;= 0 [;: > Cl:: (/) AREVA Inc.

SRV Bank 1 SRV Bank 2 ------------

SRV Bank 3 1000.0 800.0 600.0 400.0 200.0 ---------r Iv'"',----, I I I I I I I I I I ----

.0 2.0 4.0 6.0 8.0 10.0 Time, (seconds)

Figure 7.4 MSIV Closure Overpressurization Event at 102P/104.5F

-Safety/Relief Valve Flow Rates 12.0 Controlled.

Document Brunswick Unit 1 Cycle 19* MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 7-12 " Q) ...... 0 0:: ..... 0 ...... c Q) 0 .._ Q) 0... 250.0-,-----------------------------------, Core Power Heat Flux Core Flow 200.0 Steam Flow -------Feed Flow 150.0 100.0 50.0 ----------------------------------------------

.0 -50.0-+---------.-------,-------.--------.---------t

.0 10.0 20.0 30.0 40.0 Time, (seconds)

Figure 7.5 PRFO ATWS Overpressurization Event at 100P/85F -Key Parameters 50.0 AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis 205.0 ,........

c .:.:=-0 L Q.) 200.0 N ...., c Q.) E :::J L ...., rn 195.0 .f: 0 ...., ...., u Q.) 0.. 190.0 rn Q.) 0:: ..r::. ...., Qi 185.0 > Q.) _J L Q.) ...., c ;: 180.0 "C Q.) rn c Q.) (/) 175.0 AREVA Inc. .0 10.0 20.0 30.0 40.0 Time, (seconds)

Figure 7.6 PRFO ATWS Overpressurization Event at 1 OOP/85F -Sensed Water Level ANP-3280NP Revision 1 Page 7-13 50.0 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 7-14

1400.0 a) 1200.0 I... :J Vl Vl Q) I... a.. 1000.0 ----..... ........ Lower Plenum Pressure ........ ........ .... .... .... .... .... .... .... .... .... .... ........ .... .... .... .... .... .... .... .... 800.0 +---------,,------.....,.---------,..--------.---------l

.0 10.0 20.0 30.0 40.0 Time, (seconds)

Figure 7.7 PRFO ATWS Overpressurization Event at 1 OOP/85F -Vessel Pressures*

The pressures presented in this figure do not include the adjustments associated with the NRG concerns discussed in Section 7.2. 50.0 AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 7-15 SRV Bank 1 SRV Bank 2 ------------

SRV Bank 3 1000.0 800.0 ,....... en ..........

E .0 c.. ::: 600.0 0 Li:: > 0::: Cf) 400.0 200.0 " I I I I I I I I I / / / / ,,,,,.---------

--

.0 AREVA Inc. 10.0 20.0 30.0 40.0 Time, (seconds) .Figure 7.8 PRFO ATWS Overpressurization Event at 1 OOP/85F -Safety/Relief Valve Flow Rates 50.0 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document 8.0 Operating Limits and COLR Input 8.1 MCPR Limits ANP-3280NP Revision 1 Page 8-1 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 1.09 and a single-loop operation SLMCPR of 1.1-1. 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 over the full power/flow map including MELLLA+ and the EOOS scenarios presented in Table 1.1. i Cycle 19 two-loop operation MCPRp limits for ATRIUM 1 OXM 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 110.3°F at rated power. MCPRp limits for single-loop operation are 0.01 higher for all cases. MCPRt 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 10XM fuel are presented in Table 8.8. Power-and flow-dependent multipliers (LHGRFACp and LHGRFACt) are applied directly to the LHGR limits to protect against fuel melting and overstraining of the cladding during an AOO for both U02 and gadolinia

' . bearing rods. AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 8-2 The ATRIUM 10XM LHGRFACp multipliers are determined using the RODEX4 thermal-. mechanical methodology (Reference 32). 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 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 110.3°F at rated power. LHGRFACt multipliers are established to provide protection against fuel centerline melt and overstraining ofthe cladding during a postulated slow flow excursion.

The ATRIUM 10XM LHGRFACt multipliers are presented in Table 8.13, and are applicable for all Cycle 19 . . exposures and the EOOS conditions identified in Table 1.1. 8.3 MAPLHGR Limits The ATRIUM 1 OXM TLO MAPLHGR limits are presented in Table 8.14. For operation in SLO, a -( multiplier of 0.8 must be applied to the TLO MAPLHGR limits. AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 8-3

t EOOS Table 8.1 MCPRp Limits for NSS Insertion Times BOC to < NEOC* Power ATRIUM 10XM Condition

(%rated) MCPRp 100.0 1.37 90.0 1.40 80.0 1.41 Base 65.0 1.62 50.0 1.66 case >65%F S65%F operation 50.0 1.90 1.76 26.0 2.31 2.17 26.0 2.55 2.52 23.0 2.65 2.62 100.0 1.40 90.0 1.44 80.0 1.47 65.0 1.62 TBVOOS 50.0 1.66 >65%F S65%F 50.0 1.90 1.76 26.0 2.31 2.17 26.0 3.09 2.83 23.0 3.29 3.09 100.0 1.37 90.0 1.40 80.0 1.43 65.0 1.62 FHOost 50.0 1.66 >65%F S65%F 50.0 1.90 1.76 26.0 2.31 2.17 26.0 2.72 2.69 23.0 2.86 2.83 100.0 1.40 90.0 1.45 80.0 1.49 65.0 1.62 TBVOOS 50.0 1.66 and FHOost >65%F S65%F 50.0 1.90 1.76 26.0 2.31 2.17 26.0 3.23 3.03 23.0 3.46 3.27 Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRp limits will be 0.02 higher. Note that operation in SLO is only supported up to a maximum power level of 71.1 % of rated and is not allowed in MELLLA+. Note that FHOOS is not allowed in MELLLA+. AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 8-4

t EOOS Table 8.2 MCPRp Limits for TSSS Insertion Times BOC to < NEOC* Power ATRIUM 10XM Condition

(%rated) MCPRp 100.0 1.42 90.0 1.44 80.0 1.45 Base 65.0 1.62 50.0 1.66 case > 65%F S65%F operation 50.0 1.91 1.76 I 26.0 2.32 2.18 26.0 2.55 2.52 23.0 2.65 2.62 100.0 1.45 90.0 1.49 80.0 1.51 65.0 1.62 TBVOOS 50.0 1.66 >65%F S65%F 50.0 1.91 1.76 26.0 2.32 2.18 26.0 3.09 2.83 23.0 3.29 3.09 , 100.0 1.42 90.0 1.44 80.0 1.46 65.0 1.62 FHoost 50.0 1.66 > 65%F S65%F 50.0 1.91 1.76 26.0 2.32 2.18 26.0 2.72 2.69 23.0 2.86 2.83 100.0 1.45 90.0 1.49 80.0 1.52 65.0 1.62 TBVOOS 50.0 1.67 and FHOost > 65%F S65%F 50.0 1.91 1.76 26.0 2.32 2.18 26.0 3.23 3.03 23.0 3.46 3.27 Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRp limits will be 0.02 higher. Note that operation in SLO is only supported up to a maximum power level of 71.1 % of rated and is not allowed in MELLLA+. Note that FHOOS is not allowed in MELLLA+. AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 8-5

t EOOS Condition Base case operation TBVOOS FHOost TBVOOS and FHOost Table 8.3 MCPRp Limits for NSS Insertion Times BOC to < EOCLB* Power ATRIUM 10XM (%rated) MCPRp 100.0 1.39 90.0 1.42 80.0 1.43 65.0 1.62 50.0 1.66 > 65%F S65%F 50.0 1.90 1.76 26.0 2.31 2.17 26.0 2.55 2.52 23.0 2.65 2.62 100.0 1.42 90.0 1.46 80.0 1.48 65.0 1.62 50.0 1.66 >65%F S65%F 50.0 1.90 1.76 26.0 2.31 2.17 26.0 3.09 2.83. 23.0 3.29 3.09 100.0 1.39 90.0 1.42 80.0 1.43 65.0 1.62 50.0 1.66 >65%F S65%F 50.0 1.90 1.76 26.0 2.31 2.17 26.0 2.72 2.69 23.0 2.86 2.83 100.0 1.42 90.0 1.46 80.0 1.49 65.0 1.62 50.0 1.66 > 65%F S65%F 50.0 1.90 1.76 26.0 2.31 2.17 26.0 3.23 3.03 23.0 3.46 3.27 Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRp limits will be 0.02 higher. Note that operation in SLO is only supported up to a maximum power level of 71.1 % of rated and is not allowed in MELLLA+. Note that FHOOS is not allowed in MELLLA+. AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 8-6

t EOOS Table 8.4 MCPRp Limits for TSSS 1n*sertion Times BOC to < EOCLB* Power ATRIUM 10XM Condition

(%rated) MCPRp 100.0 1.42 90.0 1.45 80.0 1.45 Base 65.0 1.62 50.0 1.66 case >65%F ::;;65%F operation 50.0 1.91 1.76 26.0 2.32 2.18 26.0 2.55 2.52 23.0 2.65 2.62 100.0 1.45 90.0 1.49 80.0 1.51 65.0 1.62 TBVOOS 50.0 1.66 >65%F ::;;65%F 50.0 1.91 1.76 26.0 2.32 2.18 26.0 3.09 2.83 23.0 3.29 3.09 100.0 1.42 90.0 1.45 80.0 1.46 65.0 1.62 FHOost 50.0 1.66 >65%F ::;;65%F 50.0 1.91 1.76 26.0 2.32 2.18 26.0 2.72 2.69 23.0 2.86 2.83 100.0 1.45 90.0 1.49 80.0 1.52 TBVOOS 65.0 1.62 and 50.0 1.67 FHOost >65%F ::;;65%F 50.0 1.91 1.76 26.0 2.32 2.18 26.0 3.23 3.03 23.0 3.46 3.27 Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRp limits will be 0.02 higher. Note that operation in SLO is only supported up to a maximum power level of 71.1 % of rated and is not allowed in MELLLA+. Note that FHOOS is not allowed in MELLLA+. AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 8-7

t EOOS Condition Base case operation TBVOOS Table 8.5 MCPRp Limits for NSS Insertion Times FFTR/Coastdown**t Power ATRIUM 10XM (%rated) MCPRp 100.0 1.42 90.0 1.44 80.0 1.44 65.0 1.62 50.0 1.66 >65%F 50.0 1.90 1.76 26.0 2.31 2.17 26.0 2.72 2.69 23.0 2.86 2.83 100.0 1.43 90.0 1.47 80.0 1.49 65.0 1.62 50.0 1.66 >65%F 50.0 1.90 1.76 26.0 2.31 2.17 26.0 3.23 3.03 23.0 3.46 3.27 Limits support operation with any combination of 1 SRVOOS, up to 40% of the Tl P channels service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRp limits will be 0.02 higher. Note that operation in SLO is only supported up to a maximum power level of 71.1 % of rated and is not allowed in MELLLA+. Note that reduced feedwater temperatures such as FFTR are not allowed in MELLLA+; 'however, the FFTR/Coastdown limits may be conservatively applied to operation in the MELLLA+ domain at these exposures.

AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 8-8

t EOOS Table 8.6 MCPRp Limits for TSSS Insertion Times FFTR/Coastdown**t Power ATRIUM 10XM Condition

(%rated) MCPRp 100.0 1.48 90.0 1.48 80.0 1.48 Base 65.0 1.62 50.0 1.66 case >65%F S65%F operation 50.0 1.91 1.76 26.0 2.32 2.18 26.0 2.72 2.69 23.0 2.86 2.83 100.0 1.48 90.0 1.49 80.0 1.52 65.0 1.62 TBVOOS 50.0 1.67 > 65%F, S65%F 50.0 1.91 1.76 26.0 2.32 2.18 26.0 3.23 3.03 23.0 3.46 3.27 Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, MCPRp limits will be 0.02 higher. Note that operation in SLO is only supported up to a maximum power level of 71.1 % of rated and is not allowed in MELLLA+. Note that reduced feedwater temperatures such as FFTR are not allowed in MELLLA+; however, the FFTR/Coastdown limits may be conservatively applied to operation in the MELLLA+ domain at these exposures.

AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document Table 8.7 Flow-Dependent MCPR Limits AREVA Inc. Core Flow (%of rated) MCPRt 0.0 1.70 31.0 1.70 55.0 1.60 100.0 1.18 107.0 1.18 Table 8.8 Steady-State LHGR Limits Peak ATRIUM 10XM Pellet Exposure LHGR (GWd/MTU) (kW/ft) 0.0 14.1 18.9 14.1 74.4 7.4 ANP-3280NP Revision 1 Page 8-9 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Table 8.9 LHGRFACp Multipliers for NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM Condition

(%rated) LHGRFACp 100.0 1.00 90.0 1.00 50.0 0.92 Base > 65%F S65%F case 50.0 0.86 0.86 operation 26.0 0.68 0.72 26.0 0.44 0.46 23.0 0.42 0.42 100.0 1.00 90.0 1.00 50.0 0.92 >65%F S65%F TBVOOS 50.0 0.86 0.86 26.0 0.68 0.72 26.0 0.41 0.46 23.0 0.37 0.42 100.0 1.00 90.0 1.00 50.0 0.92 FHOOS* >65%F S65%F 50.0 0.86 0.86 26.0 0.68 0.72 26.0 0.40 0.42 23.0 0.36 0.38 100.0 1.00 90.0 1.00 50.0 0.92 TBVOOS >65%F S65%F and 50.0 0.86 0.86 FHOOS* 26.0 0.68 0.72 26.0 0.36 0.42 23.0 0.32 0.37

Note that FHOOS is not allowed in MELLLA+. AREVA Inc. / ANP-3280NP Revision 1 Page 8-10 Controlled Document Brunswick Unit 1 Cycle 19 , MELLLA+ Reload Safety Analysis Table 8.10 LHGRFACp Multipliers for TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM Condition

(%rated) LHGRFACp 100.0 1.00 90.0 1.00 50.0 0.92 Base >65%F S65%F case 50.0 0.86 0.86 operation 26.0 0.68 0.72 26.0 0.44 0.46 23.0 0.42 0.42 100.0 1.00 90.0 1.00 50.0 0.92 > 65%F S65%F TBVOOS I 50.0 0.86 0.86 26.0 0.68 0.72 26.0 0.41 0.46 23.0 0.37 0.42 100.0 1.00 90.0 1.00 50.0 0.92 FHOOS* >65%F S65%F 50.0 0.86 0.86 26.0 0.68 0.72 26.0 0.40 0.42 23.0 0.36 0.38 100.0 1.00 90.0 1.00 50.0 0.92 TBVOOS > 65%F S65%F and 50.0 0.86 0.86 FHOOS* 26.0 0.68 0.72 26.0 0.36 0.42 23.0 0.32 0.37

that FHOOS is not allowed in MELLLA+. AREVA Inc. ANP-3280NP Revision 1 Page 8-11 Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Table 8.11 LHGRFACp Multipliers for NSS Insertion Times FFTR/Coastdown*

EOOS Power ATRIUM 10XM Condition

(%rated) LHGRFACp 100.0 1.00 90.0 1.00 50.0 0.92 Base >65%F ::;65%F case 50.0 0.86 0.86 operation 26.0 0.68 0.72 26.0 0.40 0.42 23.0 0.36 0.38 100.0 1.00 90.0 1.00 50.0 0.92 TBVOOS >65%F ::;65%F 50.0 0.86 0.86 26.0 0.68 0.72 26.0 0.36 0.42 23.0 0.32 0.37 ANP-3280NP Revision 1 Page 8-12

Note that reduced feedwater temperatures such as FFTR are not allowed in MELLLA+; however, the FFTR/Coastdown limits may be conservatively applied to operation in the MELLLA+ domain at these exposures.

AREVA Inc.

Controlled Document Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis ANP-3280NP Revision 1 Page 8-13

Table 8.12 LHGRFACp Multipliers for TSSS Insertion Times FFTR/Coastdown*

EOOS Power ATRIUM 10XM Condition

(%rated) LHGRFACp 100.0 1.00 90.0 1.00 50.0 0.92 Base >65%F :;;65%F case 50.0 0.86 0.86 operation 26.0 0.68 0.72 26.0 0.40 0.42 23.0 0.36 0.38 100.0 1.00 90.0 1.00 50.0 0.92 TBVOOS >65%F :;;65%F 50.0 0.86 0.86 26.0 0.68 0.72 26.0 0.36 0.42

\ 23.0 0.32 0.37 Note that reduced feedwater temperatures such as FFTR are not allowed in MELLLA+; however, the FFTR/Coastdown limits may be conservatively applied to operation in the MELLLA+ domain at these exposures.

AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis AREVA Inc. Controlled Document Table 8.13 ATRIUM 1 OXM LHGRFACt Multipliers All Cycle 19 Exposures Core Flow (%of rated) LHGRFACt 0.0 0.58 31.0 0.58 75.0 1.00 107.0 1.00 Table 8.14 AREVA Fuel MAPLHGR Limits Average Planar ATRIUM 10XM Exposure MAPLHGR (GWd/MTU) (kW/ft) 0.0 13.1 15.0 13.1 67.0 7.7 ANP-3280NP Revision 1 Page 8-14 Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis 9.0 References Controlled Document ANP-3280NP Revision 1 Page 9-1 1. ANP-3108(P)

Revision 1, Applicability of AREVA NP BWR Methods to Brunswick Extended Power Flow Operating Domain, July 2015. 2. NED0-33006-A Revision 3, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, General Electric Hitachi Nuclear Energy America, LLC, June 2009. (available in ADAMS Folder ML091800530)

3. ANP-3013(P)

Revision 0, Brunswick Unit 1 Cycle 19 Fuel Cycle Design MELLLA+ Operating Domain, AREVA NP, May 2013. 4. ANP-2956(P)

Revision 0, Brunswick Unit 2 Cycle 20 Reload Safety Analysis, AREVA NP, October 2010. 5. ANP-2948(P)

Revision 1, Mechanical Design Report for Brunswick A TR/UM 10XM Fuel Assemblies, AREVA NP, November 2013. 6. ANP-3027(P)

Revision 0, ATRIUM 10XM Fuel Rod Thermal and Mechanical Evaluation for Brunswick Unit 1 Cycle 19 Reload BRK1-19, AREVA NP, November 2011. 7. Letter, Edmond G. Tourigny (NRC) to E.E. Utley (CP&L), "Issuance of Amendment No. 124 to Facility Operating License No. DPR-71 -Brunswick Steam Electric Plant, Unit 1, Regarding Fuel Cycle No. 7 Reload (TAC No. 69200)," February 6, 1989 (38-9061815-000).

8. ANP-2989(P)

Revision 0, Brunswick Unit 1 Thermal-Hydraulic Design Report for . ATRIUMŽ 10XM Fuel Assemblies, AREVA NP, May 2011. 9. ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011. 10. ANP-10298(P)(A).Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, March 2010. 11. NED0-33075-A Revision 8, GE Hitachi Nuclear Energy,, GE Hitachi Boiling Water Reactor, Detect and Suppress Solution -Confirmation Density, November 2013. 12. NED0-33728, Revision 2, Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2 Maximum Extended Load Line Limit Analysis Plus, October 2015. (38-9251103-001)

13. Not used. 14. Not used.* 15. Not used. 16. 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. 17. Not used. 18. ANF-913(P)(A)

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

A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990. AREVA Inc.

Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis Controlled Document ANP-3280NP Revision 1 Page 9-2 19. 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 .. 20. 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. 21. EM F-2158(P)(A)

Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors:

Evaluation and Validation of CASM0-4/M/CROBURN-B2, Siemens Power Corporation, October 1999.

22. XN-NF-81-58(P)(A)

Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Mechanica/

Response Evaluation Model, Exxon Nuclear Company, March 1984. 23. Operating License and Technical Specifications, Brunswick Steam Electric Plant, Unit No 1, Duke Energy, as amended. 24. ANF-1358(P)(A)

Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, September 2005. 25. ANP-3105 (P) Revision 1, Brunswick Units 1and2 LOCA Break Spectrum Analysis for ATRIUMŽ 10XM Fue/forMELLLA+

Operation, AREVA, July 2015. 26. ANP-3106(P)

Revision 2, Brunswick Units 1and2 LOCA-ECCS Analysis MAPLHGR Limit for A TR/UMŽ 1 OXM Fuel for MELLLA + Operation, AREVA, December 2015. 27. XN-NF-80-19(P)(A)

Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors -Neutronic Methods for Design and Analysis, Exxon' Nuclear Company, March 1983.

28. XN-NF-80-19(P)(A)

Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads,.Exxon Nuclear Company, June 1986. 29. Not used. 30. ANP-2962(P)

Revision 0, Brunswick Nuclear Plant New Fuel Storage Vault Criticality Safety Analysis for ATRIUMŽ 10XM Fuel, AREVA NP, November 2010. 31. ANP-2955(P)

Revision 3, Brunswick Nuclear Plant Spent Fuel Pool Criticality Safety Analysis for ATRIUMŽ 10XM Fuel, AREVA NP, October 2011. 32. BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008. AREVA Inc.

BSEP 16-0056 Enclosure 17 AREVA NP Affidavit Regarding Withholding ANP-3280P, Revision 1, Brunswick Unit 1 Cycle 19 MELLLA+ Reload Safety Analysis, May 2016 AFFIDAVIT STATE OF WASHINGTON ) ) SS. COUNTY OF BENTON ) 1. My name is Alan B. Meginnis.

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

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

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

3. I am familiar with the AREVA information contained in the report ANP-3280P, Revision 1, "Brunswick Unit 1Cycle19 MELLLA+ Reload Safety Analysis," dated May 2016 and referred to herein as "Document." Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA for the control and protection of proprietary and confidential information.

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

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

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

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

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief. SUBSCRIBED before me this I 2016, Susan K. McCoy . a NOTARY PUBLIC, STATE OF WASHINGTON MY COMMISSION EXPIRES: 1/14/2020 . SUSAN K MCCOY NOTARY PUBLIC* WASHINGTON MY COMMISSION EXPIRES 01*14-2020

ML16257A404

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4.4 Voiding

5.0 Controlled

5.0 Brunswick

5.0 Controlled

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