RA-18-0101, Unit 1 Cycle 22 and Unit 2 Cycle 23 Reload Safety Analysis Reports

Unit 1 Cycle 22 and Unit 2 Cycle 23 Reload Safety Analysis Reports

ML18211A662
Person / Time
Site:	Brunswick
Issue date:	07/30/2018
From:	Wooten B Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML18221A251	List: RA-18-0101, Unit 1 Cycle 22 and Unit 2 Cycle 23 Reload Safety Analysis Reports
References
RA-18-0101
Download: ML18211A662 (205)

Text

Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461 July 30, 2018 Serial: RA-18-0101 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Unit 1 Cycle 22 and Unit 2 Cycle 23 Reload Safety Analysis Reports

Reference:

Letter from William R. Gideon (Duke Energy) to the U.S. Nuclear Regulatory Commission Document Control Desk, Request for License Amendment Regarding Core Flow Operating Range Expansion, dated September 6, 2016, ADAMS Accession Number ML16257A410 Ladies and Gentlemen:

In the license amendment request (LAR) dated September 6, 2016 (i.e., Reference), Duke Energy Progress, LLC (Duke Energy), stated that the Unit 1 and Unit 2 Brunswick Steam Electric Plant (BSEP) Reload Safety Analysis Reports (RSARs) would be submitted to the NRC, for information, for the initial Maximum Extended Load Line Limit Analysis Plus (MELLLA+)

cycles. BSEP plans to implement MELLLA+ in the Fall of 2018, during Unit 1 Cycle 22 and Unit 2 Cycle 23. Therefore, copies of the Unit 1 Cycle 22 and Unit 2 Cycle 23 RSARs are provided in Enclosures 1 and 4 respectively.

Framatome (i.e., formerly known as AREVA) Reports ANP-3636P, Brunswick Unit 1 Cycle 22 Reload Safety Analysis, Revision 0 (i.e., Enclosure 1), and ANP-3560P, Brunswick Unit 2 Cycle 23 Reload Safety Analysis, Revision 0 (i.e., Enclosure 4), provide information considered proprietary to Framatome. Framatome, as owner of the proprietary information, has executed the affidavits provided in Enclosures 3 and 6, which identify the information as proprietary, is customarily held in confidence, and should be withheld from public disclosure in accordance with 10 CFR 2.390. Non-Proprietary versions of the Framatome reports, ANP-3636NP, Revision 0, and ANP-3560NP, Revision 0, are provided in Enclosures 2 and 5.

No new regulatory commitments are contained in this letter.

Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 832-2487.

U.S. Nuclear Regulatory Commission Page 2 of 2 Bryan B. Wooten Director - Organizational Effectiveness Brunswick Steam Electric Plant SBY/sby

Enclosures:

1. ANP-3636P, Brunswick Unit 1 Cycle 22 Reload Safety Analysis, Revision 0.

[Proprietary Information - Withhold from Public Disclosure in Accordance with 10 CFR 2.390]

2. ANP-3636NP, Brunswick Unit 1 Cycle 22 Reload Safety Analysis, Revision 0.

3. Affidavit Regarding Withholding ANP-3636P, Brunswick Unit 1 Cycle 22 Reload Safety Analysis, Revision 0.

4. ANP-3560P, Brunswick Unit 2 Cycle 23 Reload Safety Analysis, Revision 0.

[Proprietary Information - Withhold from Public Disclosure in Accordance with 10 CFR 2.390]

5. ANP-3560NP, Brunswick Unit 2 Cycle 23 Reload Safety Analysis, Revision 0.

6. Affidavit Regarding Withholding ANP-3560P, Brunswick Unit 2 Cycle 23 Reload Safety Analysis, Revision 0.

cc (with all enclosures):

U.S. Nuclear Regulatory Commission, Region II ATIN: Ms. Catherine Haney, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U.S. Nuclear Regulatory Commission ATIN: Mr. Gale Smith, NRG Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U.S. Nuclear Regulatory Commission ATIN: Mr. Andrew Hon (Mail Stop OWFN 8G9A) 11555 Rockville Pike Rockville, MD 20852-2738 cc (with enclosures 2, 3, 5, and 6):

Chair - North Carolina Utilities Commission (Electronic Copy Only) 4325 Mail Service Center Raleigh, NC 27699-4300 swatson@ncuc.net

RA-18-0101 Enclosure 1 (Proprietary Information - Withhold from Public Disclosure in Accordance With 10 CFR 2.390)

ANP-3636P, Brunswick Unit 1 Cycle 22 Reload Safety Analysis, Revision 0 (Proprietary Information - Withhold from Public Disclosure in Accordance With 10 CFR 2.390)

RA-18-0101 Enclosure 2 ANP-3636NP, Brunswick Unit 1 Cycle 22 Reload Safety Analysis, Revision 0

Controlled Document Brunswick Unit 1 Cycle 22 ANP-3636NP Revision 0 Reload Safety Analysis January 2018 AREVA Inc.

© 2018 AREVA Inc.

Controlled Document Copyright © 2018 AREVA Inc.

All Rights Reserved

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Initial Issue

Controlled Document AREVA Innc. ANP-36 636NP Revission 0 k Unit 1 Cycle Brunswick e 22 Reload Safety S Analysis P

Page ii Contents C

1.0 IN NTRODUCTION ............................................................................................... 1-1 2.0 DISPOSITIO D ON OF EVE ENTS .............................................................................. 2-1 3.0 MECHANIC M AL DESIGN N ANALYSIS .................................................................. 3-1 4.0 THERMAL-H T HYDRAULIC DESIGN N ANALYSIS S .................................................. 4-1 4.1 Thermmal-Hydrau ulic Design and Compa atibility .......................................... 4-1 4.2 Safetty Limit MCPR Analysiis ................................................................... 4-1 4.3 Core Hydrodyna amic Stability................................................................... 4-2 4.3.1 MELLLA BWROG Long Term S Stability So olution Option III ...................................................................................... 4-2 4.3.2 MELLLA+ + Stability DSS-CD D So olution ........................................... 4-3 4.3.3 MELLLA+ + DSS-CD Backup Sta ability Prote ection ........................... 4-3 4.4 Voidinng in the Channel Byp pass Region n ................................................... 4-4 5.0 ANTICIPAT A ED OPERA ATIONAL OCCURRENO NCES ........................................... 5-1 5.1 Syste em Transien nts .................................................................................. 5-1 5.1.1 Load Reje ection No BypassB (LRRNB) .............................................. 5-3 5.1.2 Turbine Trip T No Byp pass (TTNB B) ................................................... 5-4 5.1.3 Feedwate er Controlle er Failure (F FWCF) .......................................... 5-4 5.1.4 Pressure Regulator Failure Dow wnscale (P PRFDS) ........................ 5-5 5.1.5 Loss of Feedwater Heating H ..... ..................................................... 5-5 5.1.6 Control Rod R Withdra awal Error . ..................................................... 5-6 5.2 Slow Flow Runu up Analysis...................................................................... 5-6 5.3 Equippment Out-o of-Service Scenarios S . ..................................................... 5-7 5.3.1 FHOOS ........................................................................................ 5-8 5.3.2 TBVOOS S ...................................................................................... 5-8 5.3.3 Combined d FHOOS and a TBVOO OS ................................................ 5-8 5.3.4 One SRV VOOS .............................................................................. 5-9 5.3.5 One MSIV VOOS............................................................................. 5-9 5.3.6 Single-Lo oop Operation ............ ..................................................... 5-9 5.4 nsing Powerr Shape ........................................................................ 5-10 Licen 6.0 POSTULAT P ED ACCIDENTS ............................................................................. 6-1 6.1 Loss--of-Coolant Accident (LOCA) ...... ..................................................... 6-1 6.2 Contrrol Rod Dro op Accidentt (CRDA) ... ..................................................... 6-1 6.3 Fuel and a Equipm ment Handling Acciden nt .................................................. 6-2 6.4 Fuel Loading L Error (Infrequ uent Event) .................................................... 6-2

Controlled Document AREVA Innc. ANP-36 636NP Revission 0 k Unit 1 Cycle Brunswick e 22 Reload Safety S Analysis Page iii 7.0 SPECIAL S ANNALYSES ....................................................................................... 7-1 7.1 ASME E Overpres ssurization Analysis A ... ..................................................... 7-1 7.2 ATWS S Event Ev valuation.......................................................................... 7-2 7.2.1 ATWS Ov verpressurization Ana lysis ............................................. 7-2 7.2.2 Long-Term Evaluatio on .................................................................. 7-3 7.3 Standdby Liquid Control C Sys stem .......... ..................................................... 7-3 7.4 Fuel Criticality C .................................... ..................................................... 7-4 7.5 Stron ngest Rod Out O Shutdow wn Margin ..................................................... 7-4 8.0 OPERATING O G LIMITS AND A COLR INPUT .......................................................... 8-1 8.1 MCPR Limits ...................................... ..................................................... 8-1 8.2 LHGR R Limits ...................................... ..................................................... 8-1 8.3 MAPL LHGR Limits ............................... ..................................................... 8-2 9.0 REFERENC R CES ............................................. ..................................................... 9-1

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page iv Tables Table 1.1 EOOS Operating Conditions ..................................................................................... 1-2 Table 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses .................. 4-5 Table 4.2 Results Summary for Safety Limit MCPR Analyses ................................................. 4-6 Table 4.3 OPRM Setpoints ....................................................................................................... 4-7 Table 4.4 Option III BSP Endpoints for Brunswick Unit 1 Cycle 22 ........................................... 4-8 Table 4.5 DSS-CD BSP Endpoints For Nominal Feedwater Temperature ................................ 4-9 Table 4.6 DSS-CD BSP Endpoints For Reduced Feedwater Temperature ............................ 4-10 Table 4.7 ABSP Setpoints for the Scram Region ................................................................... 4-11 Table 4.8 Maximum Bypass Voiding at LPRM Level D .......................................................... 4-12 Table 5.1 Exposure Basis for Brunswick Unit 1 Cycle 22 Transient Analysis ......................... 5-11 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 Base 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 EOCLB Base Case FWCF Transient Results ......................................................... 5-18 Table 5.9 Loss of Feedwater Heating Transient Analysis Results .......................................... 5-19 Table 5.10 Control Rod Withdrawal Error CPR Results ....................................................... 5-19 Table 5.11 RBM Operability Requirements ............................................................................ 5-20 Table 5.12 Flow-Dependent MCPR Results ........................................................................... 5-20 Table 5.13 NEOC LHGRFACp EOOS Transient Results ........................................................ 5-21 Table 5.14 EOCLB LHGRFACp EOOS Transient Results ...................................................... 5-22 Table 5.15 Licensing Basis Core Average Axial Power Profile ............................................... 5-23 Table 7.1 ASME Overpressurization Analysis Results ............................................................. 7-5 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 Table 8.4 MCPRp Limits for TSSS Insertion Times BOC to < EOCLB...................................... 8-6 Table 8.5 MCPRp Limits for NSS Insertion Times FFTR/Coastdown ....................................... 8-7

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page v Table 8.6 MCPRp Limits for TSSS Insertion Times FFTR/Coastdown ..................................... 8-8 Table 8.7 Flow-Dependent MCPR Limits ATRIUM 10XM Fuel ................................................ 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 LHGRFACf Multipliers All Cycle 22 Exposures ............................ 8-14 Table 8.14 AREVA Fuel MAPLHGR Limits ............................................................................. 8-14 Figures Figure 1.1 Brunswick Unit 1 MELLLA Power/Flow Map ........................................................... 1-3 Figure 1.2 Brunswick Unit 1 MELLLA+ Power/Flow Map ......................................................... 1-4 Figure 5.1 EOCLB LRNB at 100P/104.5F - TSSS Key Parameters ...................................... 5-24 Figure 5.2 EOCLB LRNB at 100P/104.5F - TSSS Sensed Water Level ................................ 5-25 Figure 5.3 EOCLB LRNB at 100P/104.5F - TSSS Vessel Pressures .................................... 5-26 Figure 5.4 EOCLB TTNB at 100P/104.5F - TSSS Key Parameters ...................................... 5-27 Figure 5.5 EOCLB TTNB at 100P/104.5F - TSSS Sensed Water Level ................................ 5-28 Figure 5.6 EOCLB TTNB at 100P/104.5F - TSSS Vessel Pressures .................................... 5-29 Figure 5.7 EOCLB FWCF at 100P/104.5F - TSSS Key Parameters...................................... 5-30 Figure 5.8 EOCLB FWCF at 100P/104.5F - TSSS Sensed Water Level ............................... 5-31 Figure 5.9 EOCLB FWCF at 100P/104.5F - TSSS Vessel Pressures ................................... 5-32 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 ...... 7-10 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/85.0F - Key Parameters ........... 7-12 Figure 7.6 PRFO ATWS Overpressurization Event at 100P/85.0F - Sensed Water Level .... 7-13 Figure 7.7 PRFO ATWS Overpressurization Event at 100P/85.0F - Vessel Pressures......... 7-14 Figure 7.8 PRFO ATWS Overpressurization Event at 100P/85.0F - Safety/Relief Valve Flow Rates ....................................................................................................... 7-15

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page vi Nomenclature ABSP automated backup stability protection APRM average power range monitor AOO anticipated operational occurrence ARO all control rods out ASME American Society of Mechanical Engineers AST alternative source term ATWS anticipated transient without scram ATWS-RPT anticipated transient without scram recirculation pump trip BOC beginning-of-cycle BPWS banked position withdrawal sequence BSP backup stability protection BWROG Boiling Water Reactor Owners Group CDA confirmation density algorithm CFR Code of Federal Regulations COLR core operating limits report CPR critical power ratio CRDA control rod drop accident CRWE control rod withdrawal error DSS-CD detect and suppress solution - confirmation density EFPD effective full-power days EFPH effective full-power hours EOC end-of-cycle EOCLB end-of-cycle licensing basis EOFP end of full power EOOS equipment out-of-service FFTR final feedwater temperature reduction FHA fuel handling accident FHOOS feedwater heaters out-of-service FWCF feedwater controller failure GE General Electric GSF generic shape function HCOM hot channel oscillation magnitude HFCL high flow control line ICF increased core flow LFWH loss of feedwater heating LHGR linear heat generation rate LHGRFACf flow-dependent linear heat generation rate multipliers LHGRFACp power-dependent linear heat generation rate multipliers LOCA loss-of-coolant accident LPRM local power range monitor LRNB generator load rejection with no bypass

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page vii Nomenclature (Continued)

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

OLMCPR operating limit minimum critical power ratio OOS out-of-service OPRM oscillation power range monitor Pbypass power below which direct scram on TSV/TCV closure is bypassed PCT peak cladding temperature PLU power load unbalance PRFDS pressure regulator failure downscale PRFO pressure regulator failure open PROOS pressure regulator out-of-service RBM (control) rod block monitor RDF recirculation drive flow RHR residual heat removal RPT recirculation pump trip RTP rated thermal power SLC standby liquid control SLMCPR safety limit minimum critical power ratio SLO single-loop operation SRV safety/relief valve SRVOOS safety/relief valve out-of-service SS steady state STP simulated thermal power TBVOOS turbine bypass valves out-of-service TCV turbine control valve TIP traversing incore probe TLO two-loop operation TSSS technical specifications scram speed TSV turbine stop valve TTNB turbine trip with no bypass CPR change in critical power ratio 2PT 2 pump trip

Controlled Document AREVA In nc. ANP-36636NP Revission 0 Brunswick k Unit 1 Cyclee 22 Reload Safety S Analysis Pagge 1-1 1.0 In ntroduction Reload licensing ana alyses resultts generatedd by AREVA Inc. are pre esented in su upport of Brunswic ck Unit 1 Cyc cle 22. The analyses a rep ported in thiss document were perforrmed using methodologies previo ously approv ved for geneeric applicati on to boiling g water reacctors and demonstrated in Refe erence 1 to be applicable to the ME ELLLA+ extended flow op perating dom main, Referenc ce 2. The NR RC technicall limitations associated a w with the application of th he approved methodologies have been satisfied by these analyses.

The Cyclle 22 core co onsists of a total t of 560 fuel f assemb blies, includin ng 234 fresh h ATRIUM' 10XM* as ssemblies and 326 irrad diated ATRIU UM 10XM asssemblies. F Fifteen of thee irradiatedd ATRIUM 10XM assem mblies have le ead Z4B (Zrr-BWR) fuel channels inccluding one assembly y with three ULTRAFLO OW-S spacerrs. The licen nsing analyssis supports the core dessign presente ed in Referen nce 3 and the use of the e MELLLA+ o operating do omain after N NRC approvval is obtained. This analys sis does nott support ME ELLLA+ operration prior tto 4.75 GWd d/MTU. The e analyses s and therma al limits presented within n are applica able for this ccore loading g - including the lead fuel channels an nd spacers.

The Cyclle 22 reload licensing an nalyses were e performed for the pote entially limitin ng events an nd analyses s that were iddentified in the dispositio on of eventss. The resultss of the anallyses are ussed to establishh the Technic cal Specificaations/COLR R limits and e ensure that tthe design a and licensing g criteria are met. The design and safety analy yses are bassed on the d design and o operational assumptions and pla ant paramete ers provided by the utilityy. The resultts of the relo oad licensingg analysis support ope eration for the power/flow w map prese ented in Figuure 1.1 and ffor Figure 1.2 once NR RC approval is obtained. This reload d licensing a lso supportss operation w with the equipment out-of-serrvice (EOOS S) scenarios presented i n Table 1.1..

The resu ults in this report comply with the lice ense conditio on related too the range o of applicabiliity for the channel bow mod del. This lice ense conditioon was adde ed with the innclusion of tthe SAFLIM3 3D methodology to the list of approv ved referencces in Sectio on 5.6.5(b) off the Brunsw wick Techniccal Specifica ations.

• ATRIU UM is a trade emark of AREVA Inc.

The MELLLA+

M deppletion assum mes that operaation is restriccted to the ME ELLLA operattion region un ntil 4.75 GWd/MTU.

G

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 1-2 Table 1.1 EOOS Operating Conditions*

Single-loop operation (SLO)

Turbine bypass valves out-of-service (TBVOOS)

Feedwater heaters out-of-service (FHOOS)

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

One main steam isolation valve out-of-service§ (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.

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.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 1-3 120.0 110.0 100.0 90.0 80.0 MELLLA 70.0

% Power 60.0 R

50.0 I e C g F i 40.0 o n

30.0 20.0 Natural Circulation 10.0 Line Minimum Power 35% Minimum Pump 0.0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr 0 10 20 30 40 50 60 70 80 90 100 110 120 (%)

Core Flow Figure 1.1 Brunswick Unit 1 MELLLA Power/Flow Map

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 1-4 120.0 110.0 100.0 90.0 MELLLA+

80.0 MELLLA 70.0

% Power 60.0 50.0 I C

F 40.0 30.0 20.0 Natural Circulation 10.0 Line Minimum Power Line 35% Minimum Pump 0.0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr 0 10 20 30 40 50 60 70 80 90 100 110 120 (%)

Core Flow Figure 1.2 Brunswick Unit 1 MELLLA+ Power/Flow Map

Controlled Document AREVA In nc. ANP-36636NP Revission 0 Brunswick k Unit 1 Cycle e 22 Reload Safety S Analysis Pagge 2-1 2.0 Disposition D of o Events A disposition of even nts to identify y the limiting g events whi ch need to b be analyzed to support operation n at the Brunnswick Steam m Electric Plant was perrformed for tthe introducttion of ATRIUM 10XM fuel. Events and analyses identified as p potentially lim miting were e either evaluaated generically for the inttroduction off ATRIUM 10 0XM fuel or are perform med on a cyccle-specific b basis.

The resu ults of the dis sposition of events e are presented p in Reference 4 4.

The plant parameter differences between tho ose used in the Brunswiick Unit 1 Cyycle 21 analyses and the planned p anaalyses for thee Brunswick Unit 1 Cycle e 22 reload w were reviewwed to determ mine if the connclusions of the dispositiion of events s remain app plicable. The e review con ncluded that analyses s affected by y the differen nces were included in the e Reference e 5 calculatioon plan.

Starting with w Cycle 22, the follow wing changes s were incorrporated into o the reload licensing analyses s for Brunswick Unit 1.

• The T recirculation pump will w runback to t 34% of ratted pump sp peed at a ratte of 100 rp pm/sec upon n receiving a reactor pro otection syste em (RPS) S SCRAM signa al.

• The T number of o feedwaterr pumps in operation o wil l vary accord ding to the rreactor powe er.

Above A Pbypass (26% of rated power), there t will be two feedwa ater pumps in n operation.

Below B Pbypass, only one fe eedwater pum mp will be in n operation.

• A digital turbine control syystem is being impleme nted. This u upgrade will a add dome pressure control capabilitty.

The Refe erence 4 disposition of events e was reviewed r rela ative to the cchanges outtlined above. TheT review confirmed c the Reference e 4 conclusio ons and iden ntified eventss which are potentially affected byb these changes. Thes se parameterr changes are incorpora ated in the Brunswic ck Unit 1 Cyc cle 22 reloadd licensing analyses.

a

Controlled Document AREVA In nc. ANP-36 636NP Revission 0 Brunswick k Unit 1 Cycle e 22 Reload Safety S Analysis Pagge 3-1 3.0 Mechanical M Design D Anaalysis The mec chanical desiign analyses s for ATRIUMM 10XM are presented iin the appliccable mechanical design re eports (References 6 and 7). The ma aximum exp posure limits for the ATR RIUM 10XM fuel are:

54.0 GWd/MT TU average assembly ex xposure 62.0 GWd/MT TU rod avera age exposurre (full-lengt h fuel rods)

Even tho ough the ATR RIUM 10XM is evaluated d for operati on to the licensed peak rod average e exposure e of 62 GWd d/MTU, they will be limite ed to 60 GW Wd/MTU as p prescribed in n Brunswick Unit 1 license amend dment 124 (R Reference 8).

8 The ATR RIUM 10XM LHGR limits s are presentted in Sectio on 8.2. The ffuel cycle deesign analysses (Reference 3) have verified v that the ATRIUM M 10XM fuel assembliess remain with hin licensed burnup limits.

Controlled Document AREVA In nc. ANP-36636NP Revission 0 Brunswick k Unit 1 Cycle e 22 Reload Safety S Analysis Pagge 4-1 4.0 Thermal-Hyd T draulic Desiign Analysis 4.1 Thermal-Hyd T draulic Desiign and Com mpatibility The resu ults of the the ermal-hydrau ulic characteerization andd compatibiliity analyses are presentted in the thermmal-hydraulic c design report (Referen nce 9). The a analysis resu ults demonsstrate that the e thermal-h hydraulic design and com mpatibility criteria are sa atisfied for B Brunswick Un nit 1.

4.2 Safety S Limit MCPR Ana alysis The safe ety limit MCPPR (SLMCPR R) is defined d as the miniimum value of the critica al power ratio which en nsures that leess than 0.1% of the fue el rods in thee core are exxpected to experience boiling transition n during norm mal operatio on or an anticcipated operrational occu urrence (AOOO). The SLMCPR R for all fuel in the Brunsswick Unit 1 Cycle 22 co ore was dete ermined usinng the methodology describ bed in Refere ence 10. The e analysis wwas performe ed with a power distributtion that consservatively reepresents ex xpected reac ctor operatinng states tha at could both h exist at thee MCPR operating limiit and produce a MCPR equal to the e SLMCPR d during an AOOO.

The Brun nswick Unit 1 Cycle 22 SLMCPR S analysis used the ACE/AT TRIUM 10XM M critical powwer correlatio on additive constants c an nd additive coonstant unce ertainty desccribed in Re eference 11 ffor the ATRIIUM 10XM fu uel.

In the ARREVA metho odology, the effects of ch hannel bow o on the critica al power perrformance a are accounte ed for in the SLMCPR an nalysis. Refference 10 d discusses the e applicationn of a realisttic channel bow model.

The fuel-- and plant-re elated uncerrtainties useed in the SLM MCPR analyysis are pressented in Table 4.1 1. The radiall power unce ertainty usedd in the anal ysis includees the effectss of up to 40 0% of the TIP channels c outt-of-service, up to 50% ofo the LPRM Ms out-of-servvice, and a 2 2500 EFPH LPRM ca alibration inteerval. For TLLO, MELLLA A+ analyses were perforrmed for the minimum and maximum m core flow conditions c as ssociated with rated pow wer (85% an nd 104.5%), as well as the maximum m core powe er at 55% core flow for th he Brunswicck MELLLA+ + power/flow map. For th he maximum m core flow statepoint, s th he TLO core e flow uncerttainty given in Table 4.1 was used. F For the minimmum core flo ow at full powwer, and 55% % core flow statepoints, the SLO co ore flow uncertainnty in Table 4.1 4 was use ed consistentt with the ME ELLLA+ resttrictions liste ed in Section n 2.2.1.1 of o the Refere ence 2 Safety y Evaluationn Report.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 4-2 The analysis results support a two-loop operation (TLO) SLMCPR of 1.07 for both MELLLA and MELLLA+ operation and a single-loop operation (SLO) SLMCPR of 1.08. Consistent with the approved Brunswick Unit 1 Technical Specification SLMCPR values, the Cycle 22 operating limits are based on SLMCPR values of 1.07 for TLO and 1.09 for SLO. Table 4.2 presents a summary of the analysis results including the SLMCPR and the percentage of rods expected to experience boiling transition.

4.3 Core Hydrodynamic Stability As indicated in Section 1, the reload safety analyses presented in this report have been performed to support operation in either the MELLLA or the MELLLA+ operating regions. For hydrodynamic stability, Section 4.3.1 will be applicable until NRC approval is obtained and the reactor begins to operate in the MELLLA+ operating region. Once the reactor begins MELLLA+

operation, Sections 4.3.2 and 4.3.3 will supersede Section 4.3.1.

4.3.1 MELLLA BWROG Long Term Stability Solution Option III For operation in the MELLLA domain, Brunswick has implemented BWROG Long Term Stability Solution Option III (Oscillation Power Range Monitor-OPRM). Reload validation has been performed in accordance with Reference 12. The stability-based Operating Limit MCPR (OLMCPR) is provided for two conditions as a function of OPRM amplitude setpoint in Table 4.3. The two conditions evaluated are for a postulated oscillation at 45% core flow steady-state operation (SS) and following a two recirculation pump trip (2PT) from the limiting full power operation state point. The Cycle 22 power- and flow-dependent limits provide adequate protection against violation of the SLMCPR for postulated reactor instability as long as the operating limit is greater than or equal to the specified value for the selected OPRM setpoint. The results in Table 4.3 are valid for normal and reduced feedwater temperature (including FHOOS and FFTR) operation.

AREVA has performed calculations for the relative change in CPR as a function of the calculated hot channel oscillation magnitude (HCOM). These calculations were performed with the RAMONA5-FA code in accordance with Reference 13. This code is a coupled neutronic-thermal-hydraulic three-dimensional transient model for the purpose of determining the relationship between the relative change in CPR and the HCOM on a plant-specific basis. The stability-based OLMCPRs are calculated using the most limiting of the calculated change in

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 4-3 relative CPR for a given oscillation magnitude or the generic value provided in Reference 12.

The generic value was determined to be limiting for Cycle 22.

In cases where the OPRM system is declared inoperable for Brunswick Unit 1 Cycle 22, Backup Stability Protection (BSP) is provided in accordance with Reference 14. BSP curves have been evaluated using STAIF (Reference 15) to determine endpoints that meet decay ratio criteria for the BSP Base Minimal Region I (scram region) and Base Minimal Region II (controlled entry region). Stability boundaries based on these endpoints are then determined using the generic shape generating function from Reference 14. Analyses have been performed to support operation with both nominal and reduced feedwater temperature conditions (both FFTR and FHOOS). The endpoints for the BSP regions are provided in Table 4.4 and are the same as the regions presented in Reference 4.

4.3.2 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 16. Plant-specific analyses for the DSS-CD Solution are provided in Reference 17. The Detect and Suppress function of the DSS-CD solution based on the OPRM system relies on the Confirmation Density Algorithm (CDA), which constitutes 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 17 will be the stability licensing basis for Brunswick when operation in the MELLLA+ region is approved and implemented. The safety evaluation report for Reference 16 concluded that the DSS-CD solution is acceptable subject to certain cycle-specific limitations and conditions. The reload DSS-CD evaluation is performed by Duke Energy in accordance with the licensing methodology described in Reference 16 to: 1) confirm the DSS-CD Solution is applicable to Brunswick Unit 1 Cycle 22, and 2) confirm the Amplitude Discriminator Setpoint (SAD) of the CDA established in Reference 17 for operation of Brunswick Unit 1 Cycle 22.

4.3.3 MELLLA+ DSS-CD Backup Stability Protection Reference 16 describes two BSP options that are based on selected elements from three distinct constituents: BSP Manual Regions, BSP Boundary, and Automated BSP (ABSP) setpoints.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 4-4 The Manual BSP region boundaries were validated for Brunswick Unit 1 Cycle 22 using STAIF (Reference 15) for nominal and reduced feedwater temperature operation. The endpoints of the regions are defined in Table 4.5 and Table 4.6 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 29.

The ABSP Average Power Range Monitor (APRM) Simulated Thermal Power (STP) setpoints associated with the ABSP Scram Region are listed in Table 4.7. 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 NRCs 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 22 design [

]

[

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Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 4-5 Table 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses Parameter Uncertainty Fuel-Related Uncertainties

[

]

Plant-Related Uncertainties Feedwater flow rate 1.8%

Feedwater temperature 0.8%

Core pressure 0.8%

Total core flow rate TLO 2.5%

SLO 6%

[ ]

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 4-6 Table 4.2 Results Summary for Safety Limit MCPR Analyses Minimum Percentage Power/Flow Supported of Rods in Boiling

(%)

SLMCPR* Transition 100/104.5 TLO - 1.07 0.0706 100/99 TLO - 1.07 0.0785 100/85 TLO - 1.06 0.097 80/55 TLO - 1.07 0.087 71.1/58 SLO - 1.08 0.091

• The OLMCPR shown in Tables 8.1 through 8.7 were developed assuming a TLO SLMCPR of 1.07 and a SLO SLMCPR of 1.09.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 4-7 Table 4.3 OPRM Setpoints OPRM OLMCPR OLMCPR Setpoint (SS) (2PT) 1.05 1.16 1.19 1.06 1.18 1.20 1.07 1.19 1.22 1.08 1.21 1.24 1.09 1.23 1.26 1.10 1.25 1.28 1.11 1.27 1.30 1.12 1.29 1.32 1.13 1.31 1.34 1.14 1.33 1.36 1.15 1.35 1.39 Acceptance Less than or Less than or Criteria equal to the equal to the Off-Rated Rated Power OLMCPR OLMCPR at 45% Flow as described in Section 8.0

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 4-8 Table 4.4 Option III BSP Endpoints for Brunswick Unit 1 Cycle 22 Feedwater Temperature Operation End Point Power Flow Mode Region Designation (% rated) (% rated)

Nominal Scram IA 56.6 40.0 Nominal Scram IB 40.7 31.0 Nominal Controlled entry IIA 64.5 50.0 Nominal Controlled entry IIB 28.5 31.0 FFTR/FHOOS Scram IA 64.9 50.5 FFTR/FHOOS Scram IB 37.3 31.0 FFTR/FHOOS Controlled entry IIA 66.1 52.0 FFTR/FHOOS Controlled entry IIB 28.5 31.0

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 4-9 Table 4.5 DSS-CD BSP Endpoints For Nominal Feedwater Temperature Power Flow Endpoint Definition

(%) (%)

Scram Region A1 57.0 40.6 Boundary, HFCL Scram Region B1 42.0 31.7 Boundary, NCL Controlled Entry A2 64.5 50.0 Region Boundary, HFCL Controlled Entry B2 28.9 31.9 Region Boundary, NCL

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 4-10 Table 4.6 DSS-CD BSP Endpoints For Reduced Feedwater Temperature Power Flow Endpoint Definition

(%) (%)

Scram Region A1 65.9 51.8 Boundary, HFCL Scram Region B1 36.5 31.9 Boundary, NCL Controlled Entry A2 69.8 56.8 Region Boundary, HFCL Controlled Entry B2 28.9 31.9 Region Boundary, NCL

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 4-11 Table 4.7 ABSP Setpoints for the Scram Region Parameter Symbol Value*

Slope of ABSP APRM flow-mTRIP 2.00 %RTP/%RDF biased trip linear segment.

ABSP APRM flow-biased trip setpoint power intercept.

Constant Power Line for Trip PBSP-TRIP 42.0 %RTP from zero Drive Flow to Flow Breakpoint value.

ABSP APRM flow-biased trip setpoint drive flow intercept. WBSP-TRIP 37.5 %RDF Constant Flow Line for Trip.

Flow Breakpoint value WBSP-BREAK 25.0 %RDF

• From Table 6-2 of Reference 29.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 4-12 Table 4.8 Maximum Bypass Voiding at LPRM Level D*

Power (%) Cycle Bypass Flow (%) Exposure Void Condition (GWd/MTU) (%)

[ ]

• The voiding at LPRM level D bounds the voiding at LPRM levels A, B, and C.

Controlled Document AREVA In nc. ANP-36 636NP Revission 0 Brunswick k Unit 1 Cyclee 22 Reload Safety S Analysis Pag ge 5-1 5.0 Anticipated A Operationa O l Occurrenc ces This secttion describe es the analys ses performed to determ mine the pow wer- and flow w-dependent MCPR operating limiits for base case c operatiion at Brunsswick Unit 1 Cycle 22.

COTRAN NSA2 (Referrence 18), XCOBRA-T X (Reference 1 19), XCOBR RA (Referencce 20), and CASMO--4/MICROBU URN-B2 (Re eference 21)) are the majjor codes ussed in the the ermal limits analyses s as describe ed in the ARREVA THERM MEX method dology reporrt (Referencce 20) and neutroniccs methodolo ogy report (R Reference 27). 2 COTRAN NSA2 is a syystem transiient simulation code, wh hich includes s an axial on ne-dimension nal neutroniccs model tha at captures tthe effects o of axial powwer shifts ass sociated with the system m transients.. XCOBRA-T T is a transieent thermal-hhydraulics co ode used in the analysis s of thermal margins for the limiting fuel assemb bly.

XCOBRA A is used in steady-state s e analyses. The T ACE/AT TRIUM 10XM M critical powwer correlation (Reference 11) is use ed to evaluaate the therm mal margin fo or the ATRIU UM 10XM fu uel. Fuel pellet-to-cladding gap cond ductance vallues are bas sed on RODE EX2 (Refere ence 22) calculations forr the Brunswic ck Unit 1 Cyc cle 22 core.

5.1 System S Trannsients The reac ctor plant parrameters forr the system transient an nalyses were e provided bby the utility.

Analyses s have been performed to t determine e power-dep pendent MCP PR limits thaat protect operationn throughoutt the power/fflow domain shown in F igure 1.1 an nd Figure 1.2 2.

At Brunsw wick, direct scram on turbine stop valve (TSV) p position and turbine control valve (T TCV) fast closu ure are bypa assed at pow wer levels les ss than 26%  % of rated (Pbypass). Scram m will occur when the e high pressu ure or high neutron n flux scram setpo oint is reacheed. Reference 23 indica ates that MCP PR limits only need to be e monitored at power levvels greater than or equal to 23% off rated, whhich is the loowest power analyzed fo or this report .

The limiting exposure e for rated power p pressu urization trannsients is typpically at end d of full pow wer (EOFP) when w the control rods arre fully withd drawn. To prrovide additio onal margin to the opera ating limits earrlier in the cy ycle, analysees were also o performed to establish operating limits at a near end-of-cyycle (NEOC)) exposure. Analyses A weere performe ed at cycle eexposures prior to NEOC C to ensure thhat the opera ating limits provide p the necessary n prrotection. Th he end-of-cyycle licensing g basis (EOOCLB) analy ysis was perrformed at EOFP + 15 E EFPD. Analyses were alsso performed to support extended e cyccle operationn with final feedwater f te emperature rreduction (FF FTR) and po ower

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-2 coastdown. The Brunswick Unit 1 Cycle 22 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 and variation of +/- 10 psi in dome pressure 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 event less severe, the TCV initial position is further closed which tends to make the event more severe, especially at higher power levels. LRNB and TTNB events for base case operation were evaluated for both nominal and 10°F reduced feedwater temperatures. The analyses were performed with the limiting feedwater and dome pressure conditions in the allowable ranges.

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 +/- 10 psia and constant rated dome pressure with combined FFTR/Coastdown operation to the maximum licensing exposure (Table 5.1). 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

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-3 nominal scram speed (NSS) insertion times and the Technical Specifications scram speed (TSSS) insertion times used in the analyses are presented in Table 5.2. The NSS MCPRp limits can only be applied if the scram speed test results meet the NSS insertion times. System transient analyses were performed to establish MCPRp limits for both NSS and TSSS insertion times. The Brunswick Unit 1 Technical Specifications (Reference 23) allow for operation with up to 10 slow and 1 stuck control rod. One additional control rod is assumed to fail to scram.

Conservative adjustments to the NSS and TSSS scram speeds were made to the analysis inputs to appropriately account for these effects on scram reactivity. For cases below 26%

power, the results are relatively insensitive to scram speed, and only TSSS analyses are performed.

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.

The excursion of the core power due to the void collapse is terminated primarily by the reactor scram and revoiding of the core.

For power levels less than 50% of rated, the LRNB analyses assume that the power load unbalance (PLU) is inoperable. With the PLU inoperable, the LRNB sequence of events is different than the standard event. Instead of a fast closure, the TCVs close in servo mode and there is no direct scram on TCV closure. The power and pressure excursion continues until the high pressure scram occurs. Given that there is no direct scram when the PLU is inoperable, the above and below Pbypass system responses at 26% power are identical.

LRNB analyses were performed for a range of power/flow conditions to support generation of the thermal limits. Tables 5.3 and 5.4 present the base case limiting LRNB transient analysis results used to generate the NEOC and EOCLB operating limits for both TSSS and NSS insertion times. Figures 5.1 - 5.3 show the responses of various reactor and plant parameters during the LRNB event initiated at 100% of rated power and 104.5% of rated core flow with TSSS insertion times at EOCLB.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-4 5.1.2 Turbine Trip No Bypass (TTNB)

The turbine trip causes a closure of the turbine stop valves. The resulting compression wave travels through the steam lines into the vessel and creates a rapid pressurization. The increase in pressure causes a decrease in core voids, which in turn causes a rapid increase in power.

The closure of the turbine stop valves also causes a reactor scram. Turbine bypass system operation, which also mitigates the consequences of the event, is not credited. The excursion of the core power due to the void collapse is terminated primarily by the reactor scram and revoiding of the core.

TTNB analyses were performed for a range of power/flow conditions for which the TTNB event is potentially limiting to support generation of the thermal limits. Tables 5.5 and 5.6 present the base case TTNB transient analysis results used to generate the NEOC and EOCLB operating limits for both TSSS and NSS insertion times. Figures 5.4 - 5.6 show the responses of various reactor and plant parameters during the TTNB event initiated at 100% of rated power and 104.5% of rated core flow with TSSS insertion times at EOCLB.

5.1.3 Feedwater Controller Failure (FWCF)

The increase in feedwater flow due to a failure of the feedwater control system to maximum demand results in an increase in the water level and a decrease in the coolant temperature at the core inlet. The increase in core inlet subcooling causes an increase in core power. As the feedwater flow continues at maximum demand, the water level continues to rise and eventually reaches the high water level trip setpoint. The initial water level is conservatively assumed to be at the low-level normal operating range to delay the high-level trip and maximize the core inlet subcooling that results from the FWCF. The high water level trip causes the turbine stop valves to close in order to prevent damage to the turbine from excessive liquid inventory in the steam line. The valve closures create a compression wave that travels to the core causing a void collapse and subsequent rapid power excursion. The closure of the turbine stop valves also initiates a reactor scram. Three of the four installed turbine bypass valves are assumed operable and provide pressure relief. The core power excursion is mitigated in part by the pressure relief, but the primary mechanism for termination of the event is reactor scram.

FWCF analyses were performed for a range of power/flow conditions to support generation of the thermal limits. Tables 5.7 and 5.8 present the base case limiting FWCF transient analysis results for NEOC and EOCLB operating limits for both TSSS and NSS insertion times.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-5 Figures 5.7 - 5.9 show the responses of various reactor and plant parameters during the FWCF event initiated at 100% of rated power and 104.5% of rated core flow with TSSS insertion times at EOCLB.

5.1.4 Pressure Regulator Failure Downscale (PRFDS)

The pressure regulator failure downscale event occurs when the pressure regulator system fails and sends a signal to close all four turbine control valves in control mode. Normally, a backup pressure regulator device would take control and maintain the setpoint pressure, resulting in a mild pressure excursion and a benign event. If 5 of the 6 pressure regulator devices were out-of-service, there would be no backup pressure regulator device and the event would be more severe. The core would pressurize resulting in void collapse and a subsequent power increase.

The event would be terminated by scram when either the high-neutron flux or high-pressure setpoint is reached. Operation with only one pressure regulator device is not supported for Brunswick Unit 1 over the entire power/flow map. However, Duke Energy requested that AREVA review the PRFDS event with one pressure device in service to determine if it is bound by the LRNB event at power levels greater than 90% of rated and less than 50% or rated.

Previous analysis results demonstrate that the LRNB is more limiting at power levels greater than 90% of rated. Since LRNB analyses assume the PLU is inoperable below 50% of rated power, the TCVs close in servo or control mode without a direct scram on fast closure.

Therefore, the consequences of the PRFDS event with one pressure regulator out of service are no more severe than the LRNB event at power levels less than 50% of rated.

5.1.5 Loss of Feedwater Heating The loss of feedwater heating (LFWH) event analysis supports an assumed 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 pressurization occurs. For Brunswick Unit 1 Cycle 22, a cycle-specific analysis was performed in accordance with the Reference 24 methodology to

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-6 determine the change in MCPR for the event. The LFWH results, for operation in the MELLLA or MELLLA+ operating regions, are presented in Table 5.9.

5.1.6 Control Rod Withdrawal Error The control rod withdrawal error (CRWE) transient is an inadvertent reactor operator initiated withdrawal of a control rod. This withdrawal increases local power and core thermal power, lowering the core MCPR. The CRWE transient is typically terminated by control rod blocks initiated by the rod block monitor (RBM). The CRWE event was analyzed assuming no xenon and allowing credible instrumentation out-of-service in the rod block monitor (RBM) system. The analysis further assumes that the plant could be operating in either an A or B sequence control rod pattern. The rated power CRWE results are shown in Table 5.10 for selected analytical RBM high power setpoint values from 108% to 117%. An assumed RBM high power setpoint of 111% was used to develop the MCPRp limits. At the corresponding 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 presented in Section 8.2. The recommended operability requirements based on the unblocked CRWE results are shown in Table 5.11 based on the SLMCPR values presented in Section 4.2.

5.2 Slow Flow Runup Analysis Flow-dependent MCPR and LHGR limits are established to support operation at off-rated core flow conditions. The limits are based on the CPR and heat flux changes experienced by the fuel during slow flow excursions. The slow flow excursion event assumes a failure of the recirculation flow control system such that the core flow increases slowly to the maximum flow physically permitted by the equipment (107% of rated core flow). An uncontrolled increase in flow creates the potential for a significant increase in core power and heat flux. Operation with one MSIVOOS causes a larger increase in pressure and power during the flow excursion which results in a steeper flow runup path. A conservatively steep flow runup path was used in the analysis. The slow flow runup analyses were performed to support operation in all the EOOS scenarios.

XCOBRA is used to calculate the change in critical power ratio during a two-loop flow runup to the maximum flow rate. The MCPRf limit is set such that the increase in core power, resulting from the maximum increase in core flow, assures that the TLO safety limit MCPR is not violated.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-7 Calculations were performed for a range of initial flow rates to determine the corresponding MCPR values that put the limiting assembly on the safety limit MCPR at the high flow condition at the end of the flow excursion.

Results of the flow runup analysis are presented in Table 5.12. MCPRf limits that provide the required protection are presented in Table 8.7. The MCPRf limits are applicable for all Cycle 22 exposures.

Flow runup analyses were performed with CASMO-4/MICROBURN-B2 to determine flow-dependent LHGR multipliers (LHGRFACf) 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 assumed to remain constant during the event. The LHGRFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a flow runup. The Cycle 22 LHGRFACf multipliers are presented 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 22 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)

The following EOOS scenarios are supported for Brunswick Unit 1 Cycle 22 MELLLA+

operation:

• TBVOOS

• One SRVOOS

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-8

• One MSIVOOS Table 5.13 and Table 5.14 present the limiting 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. In addition, the steam flow for a given power level decreases since more power is required to increase the enthalpy of the coolant to saturated conditions. The consequences of the FWCF event are potentially more severe as a result of the increase in core inlet subcooling during the overcooling phase of the event. While the decrease in steam flow tends to make the LRNB event less severe, the TCV initial position is further closed which tends to make the event more severe, especially at higher power levels. FWCF events were analyzed to ensure that appropriate FHOOS operating limits are established. 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 to establish the TBVOOS operating limits.

5.3.3 Combined FHOOS and TBVOOS FWCF analyses with both FHOOS and TBVOOS were performed. Operating limits for this combined EOOS scenario were established using these FWCF results. This scenario is not allowed in the MELLLA+ region.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-9 5.3.4 One SRVOOS As noted earlier, all pressurization transient analyses were performed with one of the lowest setpoint SRVs assumed inoperable. Therefore, the base case operating limits support operation with one SRVOOS. The EOOS operating limits also support operation with one SRVOOS.

5.3.5 One MSIVOOS Operation with one MSIVOOS is supported for operation less than 70% of rated power. At these reduced power levels, the flow through any one steam line will not be greater than the flow at rated power when all MSIVs are available. Since all four turbine control valves are available, adequate pressure control can be maintained. The main difference in operation with one MSIVOOS is that the steam line pressure drop between the steam dome and the turbine valves is higher than if all MSIVs are available. Since low steam line pressure drop is limiting for pressurization transients, the results of the pressurization events with all MSIVs in service bound the results with one MSIVOOS. In addition, operation with one MSIVOOS has no impact on the other nonpressurization events evaluated to establish power-dependent operating limits.

Therefore, the power-dependent operating limits applicable to base case operation with all MSIVs in service remain applicable for operation with one MSIVOOS for power levels less than or equal to 70% of rated. As noted earlier, slow flow runup analyses were performed to support operation with one MSIVOOS.

5.3.6 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 limiting CPRs 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 SLMCPR as discussed in Section 4.2, and the application of an SLO MAPLHGR multiplier discussed in Section 8.3. The net result is a 0.02 increase in the base case MCPRp limits and a decrease in the MAPLHGR limit. The same situation is true for the EOOS scenarios. Adding 0.02 to the corresponding two-loop operation EOOS MCPRp limits results in SLO MCPRp limits for the EOOS conditions. The TLO EOOS LHGRFAC multipliers remain applicable in SLO. This scenario is not allowed in the MELLLA+ region.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-10 5.4 Licensing Power Shape The licensing axial power profile used by AREVA for the plant transient analyses bounds the projected end of full power axial power profile. The conservative licensing axial power profile generated at the EOCLB core average exposure of 35,484 MWd/MTU is given in Table 5.15.

Cycle 22 operation is considered to be in compliance when:

• The integrated normalized power generated in the bottom 7 nodes from the projected EOFP solution at the state conditions provided in Table 5.15 is greater than the integrated normalized power generated in the bottom 7 nodes in the licensing basis axial power profile, and the individual normalized power from the projected EOFP solution is greater than the corresponding normalized power from the licensing basis axial power profile for at least 6 of the 7 bottom nodes.

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

If the criteria cannot be fully met, the licensing basis may nevertheless remain valid but further assessment will be required.

The licensing basis power profile in Table 5.15 was calculated using the MICROBURN-B2 code.

Compliance analyses must also be performed using MICROBURN-B2. Note that the power profile comparison should be done without incorporating instrument updates to the axial profile because the updated power is not used in the core monitoring system to accumulate assembly burnups.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-11 Table 5.1 Exposure Basis for Brunswick Unit 1 Cycle 22 Transient Analysis Cycle Core Exposure at Average End of Interval Exposure (MWd/MTU) (MWd/MTU)* Comments 0 16,384 Beginning of cycle 17,900 34,284 Break point for exposure-dependent MCPRp limits (NEOC) 19,100 35,484 Design basis rod patterns to EOFP + 15 EFPD (EOCLB) 20,564 36,948 Maximum licensing core exposure - including FFTR

/Coastdown

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

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-12 Table 5.2 Scram Speed Insertion Times Control Rod TSSS NSS Position Time Time (notch) (sec) (sec) 48 (full-out) 0.000 0.000 48 0.200 0.200 46 0.440 0.318 36 1.080 0.829 26 1.830 1.369 6 3.350 2.510 0 (full-in) 3.806 2.852

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-13 Table 5.3 NEOC Base Case LRNB Transient Results ATRIUM 10XM ATRIUM 10XM Supported Power CPR LHGRFACp TSSS Insertion Times 100 0.32 1.00 90 0.33 1.00 80 0.34 1.00 70 0.35 1.00 60 0.32 1.00 50 0.30 1.00 50 at > 65%F PLU inoperable 0.84 0.89 50 at 65%F PLU inoperable 0.71 0.95 26 at > 65%F PLU inoperable 1.25 0.63 26 at 65%F PLU inoperable 1.14 0.77 26 at > 65%F below Pbypass 1.25 0.63 26 at 65%F below Pbypass 1.14 0.77 23 at > 65%F below Pbypass 1.33 0.60 23 at 65%F below Pbypass 1.24 0.71 NSS Insertion Times 100 0.28 1.00 90 0.29 1.00 80 0.30 1.00 70 0.31 1.00 60 0.30 1.00 50 0.29 1.00 50 at > 65%F PLU inoperable 0.82 0.90 50 at 65%F PLU inoperable 0.69 0.97 26 at > 65%F PLU inoperable 1.23 0.64 26 at 65%F PLU inoperable 1.11 0.78

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-14 Table 5.4 EOCLB Base Case LRNB Transient Results ATRIUM 10XM ATRIUM 10XM Supported Power CPR LHGRFACp TSSS Insertion Times 100 0.32 1.00 90 0.33 1.00 80 0.34 1.00 70 0.36 1.00 60 0.32 1.00 50 0.30 1.00 50 at > 65%F PLU inoperable 0.84 0.89 50 at 65%F PLU inoperable 0.71 0.95 26 at > 65%F PLU inoperable 1.25 0.63 26 at 65%F PLU inoperable 1.14 0.77 26 at > 65%F below Pbypass 1.25 0.63 26 at 65%F below Pbypass 1.14 0.77 23 at > 65%F below Pbypass 1.33 0.60 23 at 65%F below Pbypass 1.24 0.71 NSS Insertion Times 100 0.28 1.00 90 0.29 1.00 80 0.30 1.00 70 0.31 1.00 60 0.30 1.00 50 0.29 1.00 50 at > 65%F PLU inoperable 0.82 0.90 50 at 65%F PLU inoperable 0.69 0.97 26 at > 65%F PLU inoperable 1.23 0.64 26 at 65%F PLU inoperable 1.11 0.78

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-15 Table 5.5 NEOC Base Case TTNB Transient Results ATRIUM 10XM ATRIUM 10XM Supported Power CPR LHGRFACp TSSS Insertion Times 100 0.32 1.00 90 0.33 1.00 80 0.34 1.00 26 at > 65%F below Pbypass 1.26 0.63 26 at 65%F below Pbypass 1.13 0.77 23 at > 65%F below Pbypass 1.33 0.59 23 at 65%F below Pbypass 1.24 0.71 NSS Insertion Times 100 0.28 1.00 90 0.29 1.00 80 0.30 1.00

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-16 Table 5.6 EOCLB Base Case TTNB Transient Results ATRIUM 10XM ATRIUM 10XM Supported Power CPR LHGRFACp TSSS Insertion Times 100 0.32 1.00 90 0.33 1.00 80 0.34 1.00 26 at > 65%F below Pbypass 1.26 0.63 26 at 65%F below Pbypass 1.13 0.77 23 at > 65%F below Pbypass 1.33 0.59 23 at 65%F below Pbypass 1.24 0.71 NSS Insertion Times 100 0.29 1.00 90 0.29 1.00 80 0.30 1.00

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-17 Table 5.7 NEOC Base Case FWCF Transient Results ATRIUM 10XM ATRIUM 10XM Supported Power CPR LHGRFACp TSSS Insertion Times 100 0.29 1.00 90 0.31 1.00 80 0.32 1.00 70 0.34 1.00 60 0.36 1.00 50 0.38 1.00 26 0.71 0.90 26 at > 65%F below Pbypass 1.22 0.51 26 at 65%F below Pbypass 1.22 0.53 23 at > 65%F below Pbypass 1.31 0.49 23 at 65%F below Pbypass 1.31 0.50 NSS Insertion Times 100 0.26 1.00 90 0.27 1.00 80 0.30 1.00 70 0.32 1.00 60 0.34 1.00 50 0.37 1.00 26 0.68 0.92

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-18 Table 5.8 EOCLB Base Case FWCF Transient Results ATRIUM 10XM ATRIUM 10XM Supported Power CPR LHGRFACp TSSS Insertion Times 100 0.29 1.00 90 0.31 1.00 80 0.32 1.00 70 0.34 1.00 60 0.36 1.00 50 0.38 1.00 26 0.71 0.90 26 at > 65%F below Pbypass 1.22 0.51 26 at 65%F below Pbypass 1.22 0.53 23 at > 65%F below Pbypass 1.31 0.49 23 at 65%F below Pbypass 1.31 0.50 NSS Insertion Times 100 0.26 1.00 90 0.27 1.00 80 0.30 1.00 70 0.32 1.00 60 0.34 1.00 50 0.37 1.00 26 0.68 0.92

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-19 Table 5.9 Loss of Feedwater Heating Transient Analysis Results Power ATRIUM 10XM

(% rated) CPR 100 0.11 90 0.12 80 0.12 70 0.13 60 0.15 50 0.17 40 0.20 30 0.25 23 0.31 Table 5.10 Control Rod Withdrawal Error CPR Results Analytical RBM Setpoint (without filter) ATRIUM 10XM

(%) CPR 108 0.22 111 0.26 114 0.29 117 0.32

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-20 Table 5.11 RBM Operability Requirements Applicable Thermal Power ATRIUM 10XM,

(% rated) MCPR 1.89 TLO 29% and < 90%

1.92 SLO 90% 1.49 TLO Table 5.12 Flow-Dependent MCPR Results Core ATRIUM 10XM ATRIUM 10XM Flow Limiting MCPR Limiting MCPR

(% rated) MELLLA MELLLA+

31 1.49 1.61 40 1.45 1.55 50 1.45 1.52 60 1.42 1.47 70 1.32 1.33 80 1.26 1.25 90 1.20 1.19 100 1.13 1.13 107 1.07 1.07

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-21 Table 5.13 NEOC LHGRFACp EOOS Transient Results FHOOS/

FHOOS TBVOOS TBVOOS Power FWCF FWCF FWCF TSSS Insertion Times 100 1.00 1.00 1.00 90 1.00 1.00 1.00 80 1.00 1.00 1.00 70 1.00 1.00 0.99 60 1.00 1.00 0.98 50 0.97 1.00 0.96 26 0.81 0.90 0.81 26 at > 65%F below Pbypass 0.46 0.42 0.39 26 at 65%F below Pbypass 0.48 0.48 0.44 23 at > 65%F below Pbypass 0.44 0.39 0.37 23 at 65%F below Pbypass 0.46 0.43 0.40 NSS Insertion Times 100 1.00 1.00 1.00 90 1.00 1.00 1.00 80 1.00 1.00 1.00 70 1.00 1.00 1.00 60 1.00 1.00 1.00 50 0.98 1.00 0.98 26 0.83 0.92 0.83

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-22 Table 5.14 EOCLB LHGRFACp EOOS Transient Results FHOOS/

FHOOS TBVOOS TBVOOS Power FWCF FWCF FWCF TSSS Insertion Times 100 1.00 1.00 1.00 90 1.00 1.00 1.00 80 1.00 1.00 1.00 70 1.00 1.00 0.99 60 1.00 1.00 0.98 50 0.97 1.00 0.96 26 0.81 0.90 0.81 26 at > 65%F below Pbypass 0.46 0.42 0.39 26 at 65%F below Pbypass 0.48 0.48 0.44 23 at > 65%F below Pbypass 0.44 0.39 0.37 23 at 65%F below Pbypass 0.46 0.43 0.40 NSS Insertion Times 100 1.00 1.00 1.00 90 1.00 1.00 1.00 80 1.00 1.00 1.00 70 1.00 1.00 1.00 60 1.00 1.00 1.00 50 0.98 1.00 0.98 26 0.83 0.92 0.83

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-23 Table 5.15 Licensing Basis Core Average Axial Power Profile State Conditions for Power Shape Evaluation Power, MWt 2923.0 MICROBURN-B2 pressure, psia 1044.8 Inlet subcooling, Btu/lbm 20.3 Flow, Mlb/hr 80.5 Control state ARO Core average exposure (EOCLB), MWd/MTU 35,484 Licensing Axial Power Profile (Normalized)

Node Power Top 25 0.315 24 0.913 23 1.164 22 1.348 21 1.443 20 1.494 19 1.482 18 1.464 17 1.412 16 1.345 15 1.293 14 1.204 13 1.253 12 1.200 11 1.124 10 1.059 9 0.987 8 0.899 7 0.802 6 0.716 5 0.619 4 0.529 3 0.462 2 0.368 Bottom 1 0.103

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-24 Figure 5.1 EOCLB LRNB at 100P/104.5F - TSSS Key Parameters

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-25 Figure 5.2 EOCLB LRNB at 100P/104.5F - TSSS Sensed Water Level

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-26 Figure 5.3 EOCLB LRNB at 100P/104.5F - TSSS Vessel Pressures

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-27 Figure 5.4 EOCLB TTNB at 100P/104.5F - TSSS Key Parameters

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-28 Figure 5.5 EOCLB TTNB at 100P/104.5F - TSSS Sensed Water Level

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-29 Figure 5.6 EOCLB TTNB at 100P/104.5F - TSSS Vessel Pressures

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-30 Figure 5.7 EOCLB FWCF at 100P/104.5F - TSSS Key Parameters

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-31 Figure 5.8 EOCLB FWCF at 100P/104.5F - TSSS Sensed Water Level

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 5-32 Figure 5.9 EOCLB FWCF at 100P/104.5F - TSSS Vessel Pressures

Controlled Document AREVA In nc. ANP-36 636NP Revission 0 Brunswickk Unit 1 Cycle e 22 Reload Safety S Analysis Pagge 6-1 6.0 Postulated P Accidents A

6.1 Loss-of-Coo L olant Accide ent (LOCA)

For MELLLA operatio on, the results of the AT TRIUM 10XM M LOCA ana alysis are preesented in Referencces 25 and 26 2 and provide a PCT off 1885°F. Th he peak loca al metal wateer reaction iss 1.04% an nd the core wide w metal water w reactio on is < 0.47%%. The Refe erence 33 10 0 CFR 50.46 6 report do ocuments an n additional PCT P impact of +2°F thatt must be acccounted for.. The SLO MAPLHG GR multiplierr is 0.80.

For MELLLA+ operation, the results of the ATRIUM A 10X XM LOCA an nalysis are ppresented in Referencces 34 and 35 3 and provide a PCT off 1923°F. Th he peak loca al metal wateer reaction iss 1.23% an nd the core wide w metal water w reactio on is < 0.56%%. The SLO MAPLHGR R multiplier iss 0.80. Thee cycle-spec cific OLMCPRs and off-rrated flow de ependent LH HGR setdow wn bounds those assumed d in the MEL LLLA+ plant-specific ECC CS-LOCA an nalyses.

The Brunnswick LOCA A radiologicaal analysis immplementing g the alterna ative source term methodology was pe erformed in consideration c n of ATRIUM M 10XM fuell in the core inventory so ource terms. Duuke Energy has evaluated the radio ological conssequences o of a LOCA an nd determine ed ATRIUM 10XM fuel does d not significantly inccrease the raadiological cconsequence es relative to o

consideraation of ATR RIUM-10 fuel in the core inventory so ource term.

6.2 Control C Rod d Drop Accid dent (CRDA A)

Brunswicck Unit 1 use es a bank po osition withdrawal seque ence (BPWS S) including rreduced notcch worth rodd pull to limitt high worth control rod movements.

m . A CRDA evvaluation wa as performed d for both A annd B sequen nce startups consistent with w the with hdrawal sequ uence speciffied by Duke e Energy. Subsequent S s have show calculations wn that the mmethodology is applicable to fuel modeled with the CA ASMO4/MICR ROBURN-B2 code syste em. The CR RDA analysiss was performed with the approved a meethodology described d in n Reference 27.

The CRD DA analysis results demo onstrate that the maxim um deposite ed fuel rod e enthalpy is le ess than the NRC threshold of 280 cal/g c and that the estima ted number of fuel rods that exceed d the fuel damage thresho old of 170 cal/g is less than the numb ber of failed rods supported by the Brunswicck alternative e source termm (AST) ana alysis. Duke Energy hass determined d the radioloogical release assumed a in the t current Brunswick B CRDA C AST a analysis boun nds 986 rod failures for core source te erms based on ATRIUM 10XM fuel. The numbe er of fuel rodss estimated to exceed th he

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 6-2 fuel damage threshold is below 986. Therefore, the current Brunswick CRDA AST analysis remains applicable.

Dropped Control Rod Worth, mk 9.47 mk Core average Doppler coefficient, k/k/°F -10.5 x 10-6 Effective delayed neutron fraction 0.0052 4-bundle local peaking factor 1.465 Max Deposited Fuel Rod Enthalpy, cal/g 173.3 Bounding number of rods exceeding 170 cal/g 364 6.3 Fuel and Equipment Handling Accident Duke Energy has determined the radiological release assumed in the current fuel handling accident (FHA) analysis implementing the 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 10XM fuel is 161. These results are consistent with the number of failed rods supported by the current Brunswick AST analysis.

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 50.67 limits.

AREVA has compared the BRK1-22 OLMCPR to generic CPR behavior of ATRIUM 10XM fuel assemblies in D lattice boiling water reactors. The effect on fuel centerline melt and 1% strain limits have also been considered. It has been concluded that the BRK1-22 OLMCPR is sufficiently high and that the increase in nodal power is sufficiently low to ensure that a small fraction of the 10 CFR 50.67 limits will not be exceeded if a fuel loading error were to occur.

Controlled Document AREVA In nc. ANP-36 636NP Revission 0 Brunswick k Unit 1 Cyclee 22 Reload Safety S Analysis Pag ge 7-1 7.0 Special S Anallyses 7.1 ASME A Overppressurizatiion Analysis s This secttion describe es the maxim mum overpre essurization analyses pe erformed to demonstrate e complian nce with the ASME Boile er and Press sure Vessel C Code. The a analysis showws that the safety/relief valves at Brunswick Unit 1 have sufficient ca apacity and performance e to preventt the reactor vessel v pressu ure from reaaching the sa afety limit of 110% of the e design preessure.

An MSIV V closure ana alysis was performed with the AREV VA plant sim mulator code COTRANSA A2 (Reference 18) for 10 02% power and 104.5%  % flow and 10 02% power a and 85% flow w at the high hest Cycle 22 exposure wherew rated power p operaation can be attained. Th he MSIV clossure event iss o the other stteam line va similar to alve closure events in tha at the valve closure resu ults in a rapid pressurizzation of the core. The in ncrease in pressure cau ses a decrease in void w which in turnn causes a rapid increa ase in power. The turbin ne bypass va alves do nott impact the system response e and are no ot modeled in n the analysis. The follow wing assumptions were made in the e analysis:

• The T most critical active co omponent (d direct scramm on valve po osition) was assumed to o fail.

However, H scrram on high neutron flux x and high do ome pressurre is availab ble.

• The T plant con nfiguration analyzed ass sumed that o one of the low west setpoinnt SRVs wass in noperable.

• TSSS T insertio on times werre used.

• The T initial dom me pressure e was set at the maximu um allowed b by the Technnical Specifications S s, 1059.7 ps sia (1045 psiig).

• A fast MSIV closure c time of 2.7 seconnds was use ed.

• Both B MELLLA A and MELLLA+ operatio on are supp ported.

Results of o the limiting g MSIV closure overpres ssurization a analysis are presented in Table 7.1..

Figures 7.1 7 - 7.4 sho ow the respo onse of varioous reactor p plant parame eters during the MSIV closure event.

e The maximum m pre essure of 1331 psig occu urs in the lowwer plenum.. The maxim mum dome pre essure for thhe same eve ent is 1300 psig.

p These p peak pressu ure results haave been adjusted to address NRC concerrns associated with the vvoid-quality correlation, exposure e-dependentt thermal con nductivity, and Doppler e effects. The results dem monstrate tha at the maximum m vessel pre essure limit ofo 1375 psig and dome p pressure limit of 1325 pssig are not exceeded d.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-2 A sensitivity analysis was performed to determine the impact of additional drift on the SRV opening setpoint above the 3% identified in the plant Technical Specifications for the highest setpoint SRV banks. Assuming all of the degraded valves are from the highest setpoint SRV banks 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 limit of 120% of the design pressure (1500 psig). The ATWS 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.

• 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 with a -10 psi uncertainty (1035 psia).

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

• Both MELLLA and MELLLA+ operation are supported.

Results of ATWS overpressurization analyses are presented in Table 7.3. Figures 7.5 - 7.8 show the response of various reactor plant parameters during the limiting PRFO event, the event which results in the maximum vessel pressure. The maximum lower plenum pressure is 1468 psig and the maximum dome pressure is 1451 psig. The peak pressure results have been

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-3 adjusted to address NRC concerns associated with the void-quality correlation, exposure-dependent thermal conductivity, and Doppler effects. The results demonstrate that the ATWS maximum vessel pressure limit of 1500 psig is not exceeded.

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 for the highest setpoint SRV banks. Assuming all of the degraded valves are from the highest setpoint SRV banks provides a conservative scenario, and bounds the situation where the drift occurs in other SRVs. Results for the sensitivity analysis are presented in Table 7.4. The results demonstrate that the ATWS maximum vessel pressure limit of 1500 psig is not exceeded for the scenarios considered.

7.2.2 Long-Term Evaluation Fuel design differences may impact the power and pressure excursion experienced during the ATWS event. This in turn may impact the amount of steam discharged to the suppression pool and containment. For Unit 2 Cycle 20 (Reference 4) an evaluation was performed that concluded the introduction of ATRIUM 10XM fuel will not significantly impact the long-term ATWS response (suppression pool temperature and containment pressure) and the current analysis remains applicable for MELLLA operation. This conclusion remains applicable for Unit 1 Cycle 22.

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

ATWS (long-term and instability) for MELLLA+ has been analyzed in Reference 17.

7.3 Standby Liquid Control System In the event that the control rod scram function becomes incapable of rendering the core in a shutdown state, the standby liquid control (SLC) system is required to be capable of bringing the reactor from full power to a cold shutdown condition at any time in the core life. The Brunswick Unit 1 SLC system is required to be able to inject 720 ppm natural boron equivalent at 70°F into the reactor coolant (including a 25% allowance for imperfect mixing, leakage, and volume of other piping connected to the reactor). An analysis that demonstrates that the SLC system meets the required shutdown capability for Cycle 22 has been performed. The analysis was

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-4 performed to support a coolant temperature of 360.8°F with a boron concentration equivalent to 720 ppm at 70°F. The temperature of 360.8°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 2.00 %k.

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 Reference 31. The ATRIUM 10XM fuel assemblies identified for loading in Cycle 22 meet both the new and spent fuel storage requirements.

7.5 Strongest Rod Out Shutdown Margin The BRK1-22 core has a minimum strongest rod out shutdown margin of 1.06 %k*. This value is produced at the beginning of the cycle at the minimum coolant temperature condition (68 °F).

This value assumes that BRK1-21 ended operation at the lowest allowable exposure.

• Relative to a design goal of 0.92 %k.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-5 Table 7.1 ASME Overpressurization Analysis Results*,

Maximum Peak Peak Vessel Maximum Neutron Heat Pressure Dome Flux Flux Lower-Plenum Pressure Event (% rated) (% rated) (psig) (psig)

MSIV closure 265 131 1331 1300 (102P/104.5F)

Table 7.2 ASME Overpressurization Sensitivity Analysis Results*,

High Bank Maximum Pressure SRV (psig)

Number of Setpoint Lower Steam Event Valves Drift Plenum Dome 3 +4%

MSIV closure 2 +6% 1348 1316 (102P/104.5F) 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.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-6 Table 7.3 ATWS Overpressurization Analysis Results*,

Maximum Vessel Peak Peak Pressure Maximum Neutron Heat Lower- Dome Flux Flux Plenum Pressure Event (% rated) (% rated) (psig) (psig)

MSIV closure (100P/104.5F) 245 138 1429 1411 MSIV closure (100P/85F) 250 134 1449 1432 PRFO (100P/104.5F) 246 147 1447 1429 PRFO (100P/85F) 227 140 1468 1451

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

The maximum Technical Specification allowed SRV degradation of 3% was assumed.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-7 Table 7.4 ATWS Overpressurization Sensitivity Analysis Results*,

High Bank Maximum Pressure SRV (psig)

Number of Setpoint Lower Steam Event Valves Drift Plenum Dome 3 +4%

PRFO 2 +6% 1456 1438 (100P/104.5F) 1 +8%

3 +4%

PRFO 2 +6% 1477 1460 (100P/85F) 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.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-8 Figure 7.1 MSIV Closure Overpressurization Event at 102P/104.5F - Key Parameters

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-9 Figure 7.2 MSIV Closure Overpressurization Event at 102P/104.5F - Sensed Water Level

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-10 Figure 7.3 MSIV Closure Overpressurization Event at 102P/104.5F - Vessel Pressures*

• The pressures presented in this figure do not include the adjustments associated with the NRC concerns discussed in Section 7.1.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-11 Figure 7.4 MSIV Closure Overpressurization Event at 102P/104.5F - Safety/Relief Valve Flow Rates

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-12 Figure 7.5 PRFO ATWS Overpressurization Event at 100P/85.0F - Key Parameters

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-13 Figure 7.6 PRFO ATWS Overpressurization Event at 100P/85.0F - Sensed Water Level

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-14 Figure 7.7 PRFO ATWS Overpressurization Event at 100P/85.0F - Vessel Pressures*

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

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 7-15 Figure 7.8 PRFO ATWS Overpressurization Event at 100P/85.0F - Safety/Relief Valve Flow Rates

Controlled Document AREVA In nc. ANP-36 636NP Revission 0 Brunswick k Unit 1 Cycle e 22 Reload Safety S Analysis Pagge 8-1 8.0 Operating O Liimits and COLR Input 8.1 MCPR M Limitss The dete ermination off the MCPR limits for Bru unswick Uni t 1 Cycle 22 2 is based on n the analyses of the limitin ng anticipateed operation nal occurrenc ces (AOOs).. The MCPR R operating limits are establishhed so that leess than 0.1% of the fue el rods in the e core are exxpected to experience boiling transition n during an AOO A initiated d from rated d or off-rated d conditions and are bassed on the Technica al Specifications two-loop operation SLMCPR off 1.07 and a single-loop operation SLMCPR R of 1.09. Ex xposure-dependent MCP PR limits we ere established to support operation from BOC to near n end-of-c cycle (NEOC C), NEOC to o end-of-cyc le licensing basis (EOCLB), and combined d FFTR/Coa astdown as defined d by th he core averrage exposu ures listed in Table 5.1.

MCPR lim mits are estaablished to support s basee case opera ation over thhe full powerr/flow map includingg MELLLA+ and a the EOO OS scenarios presented d in Table 1.1 1.

Cycle 22 two-loop op peration MCPRp limits fo or ATRIUM 1 10XM fuel arre presented d in Tables 88.1 -

8.6 for baase case operation and the EOOS conditions.

c L Limits are pre esented for nominal scra am speed (N NSS) and Te echnical Specification scram speed ((TSSS) insertion times ffor the expossure ranges considered. An A assumed d RBM high powerp setpo oint of 111% was used to o develop the MCPRp limits. Tables s 8.1 and 8.2 2 present the e MCPRp lim mits for the B BOC to NEO OC exposure e range. Ta ables 8.3 annd 8.4 presen nt the MCPR Rp limits appplicable for thhe BOC to E EOCLB expo osure range. Ta ables 8.5 annd 8.6 presen nt the MCPR Rp limits for F FFTR/Coasttdown opera ation. The FFTR/Co oastdown lim mits (both ba ase case and d TBVOOS) support both nominal and constant rated dom me pressure e operation withw feedwatter temperat ures consisttent with a fe eedwater temperatture reductio on of up to 110.3°F at rated power. M MCPRp limitss for single-lloop operation are 0.02 higher for all cases.

MCPRf limits that pro otect againstt fuel failures s during a poostulated slo ow flow excu ursion are presente ed in Table 8.7.

8 These MCPR M f limits are applicab ble for all Cyycle 22 expo osures and thhe EOOS co onditions ideentified in Taable 1.1.

8.2 LHGR L Limitss The LHG GR limits for ATRIUM 10XM fuel are presented i n Table 8.8 (Reference 7). Power- a and flow-depe endent multipliers (LHGRFACp and LHGRFACf) are applied d directly to tthe LHGR lim mits to protecct against fueel melting an nd overstrain ning of the clladding durin ng an AOO for both UO2 and gadoliniaa bearing rod ds.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis 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 ATRIUM 10XM LHGRFACp multipliers for the BOC to EOCLB exposure range are presented in Table 8.9 and Table 8.10. The FFTR/Coastdown LHGRFACp multipliers are presented in Table 8.11 and Table 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.

LHGRFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a postulated slow flow excursion. For the ATRIUM 10XM fuel the LHGRFACf multipliers are presented in Table 8.13, and are applicable for all Cycle 22 exposures and the EOOS conditions identified in Table 1.1.

8.3 MAPLHGR Limits The ATRIUM 10XM TLO MAPLHGR limits are presented in Table 8.14. For operation in SLO, a multiplier of 0.80 must be applied to the TLO MAPLHGR limits.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 8-3 Table 8.1 MCPRp Limits for NSS Insertion Times BOC to < NEOC*,

EOOS Power ATRIUM 10XM Condition (% rated) MCPRp 100.0 1.35 90.0 1.37 50.0 1.66 Base > 65%F 65%F case 50.0 1.91 1.78 operation 26.0 2.34 2.22 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.38 90.0 1.40 50.0 1.66

> 65%F 65%F TBVOOS 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.35 90.0 1.37 50.0 1.66

> 65%F 65%F FHOOS 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.38 90.0 1.40 50.0 1.66 TBVOOS > 65%F 65%F FHOOS 50.0 1.91 1.78 26.0 2.34 2.22 26.0 3.03 2.96 23.0 3.22 3.20

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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+.

Limits do not support MELLLA+ operation prior to 4.75 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 8-4 Table 8.2 MCPRp Limits for TSSS Insertion Times BOC to < NEOC*,

EOOS Power ATRIUM 10XM Condition (% rated) MCPRp 100.0 1.39 90.0 1.40 50.0 1.66 Base > 65%F 65%F case 50.0 1.93 1.80 operation 26.0 2.36 2.25 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.41 90.0 1.44 50.0 1.66

> 65%F 65%F TBVOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.39 90.0 1.40 50.0 1.66

> 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.41 90.0 1.44 50.0 1.66 TBVOOS > 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 3.03 2.96 23.0 3.22 3.20

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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+.

Limits do not support MELLLA+ operation prior to 4.75 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 8-5 Table 8.3 MCPRp Limits for NSS Insertion Times BOC to < EOCLB*,

EOOS Power ATRIUM 10XM Condition (% rated) MCPRp 100.0 1.36 90.0 1.37 50.0 1.66 Base > 65%F 65%F case 50.0 1.91 1.78 operation 26.0 2.34 2.22 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.38 90.0 1.40 50.0 1.66

> 65%F 65%F TBVOOS 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.36 90.0 1.37 50.0 1.66

> 65%F 65%F FHOOS 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.38 90.0 1.40 50.0 1.66 TBVOOS > 65%F 65%F FHOOS 50.0 1.91 1.78 26.0 2.34 2.22 26.0 3.03 2.96 23.0 3.22 3.20

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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+.

Limits do not support MELLLA+ operation prior to 4.75 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 8-6 Table 8.4 MCPRp Limits for TSSS Insertion Times BOC to < EOCLB*,

EOOS Power ATRIUM 10XM Condition (% rated) MCPRp 100.0 1.39 90.0 1.40 50.0 1.66 Base > 65%F 65%F case 50.0 1.93 1.80 operation 26.0 2.36 2.25 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.41 90.0 1.44 50.0 1.66

> 65%F 65%F TBVOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.39 90.0 1.40 50.0 1.66

> 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.41 90.0 1.44 50.0 1.66 TBVOOS > 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 3.03 2.96 23.0 3.22 3.20

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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+.

Limits do not support MELLLA+ operation prior to 4.75 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 8-7 Table 8.5 MCPRp Limits for NSS Insertion Times FFTR/Coastdown*,,

EOOS Power ATRIUM 10XM Condition (% rated) MCPRp 100.0 1.37 90.0 1.37 50.0 1.66 Base > 65%F 65%F case 50.0 1.91 1.78 operation 26.0 2.34 2.22 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.38 90.0 1.40 50.0 1.66

> 65%F 65%F TBVOOS 50.0 1.91 1.78 26.0 2.34 2.22 26.0 3.03 2.96 23.0 3.22 3.20

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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+.

Limits do not support MELLLA+ operation prior to 4.75 GWd/MTU.

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.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 8-8 Table 8.6 MCPRp Limits for TSSS Insertion Times FFTR/Coastdown*,,

EOOS Power ATRIUM 10XM Condition (% rated) MCPRp 100.0 1.39 90.0 1.40 50.0 1.66 Base > 65%F 65%F case 50.0 1.93 1.80 operation 26.0 2.36 2.25 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.41 90.0 1.44 50.0 1.66

> 65%F 65%F TBVOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 3.03 2.96 23.0 3.22 3.20

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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+.

Limits do not support MELLLA+ operation prior to 4.75 GWd/MTU.

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.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 8-9 Table 8.7 Flow-Dependent MCPR Limits ATRIUM 10XM Fuel Core Flow MCPRf MCPRf

(% of rated) MELLLA MELLLA+

0.0 1.55 1.64 31.0 1.55 1.64 60.0 1.47 1.50 80.0 1.30 1.30 100.0 1.30 1.30 107.0 1.30 1.30 Table 8.8 Steady-State LHGR Limits Peak ATRIUM 10XM Pellet Exposure LHGR (GWd/MTU) (kW/ft) 0.0 15.1 6.0 14.1 18.9 14.1 54.0 10.6 74.4 5.4

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 8-10 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 1.00 Base > 65%F 65%F case 50.0 0.89 0.95 operation 26.0 0.63 0.77 26.0 0.51 0.53 23.0 0.49 0.50 100.0 1.00 90.0 1.00 50.0 1.00

> 65%F 65%F TBVOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.42 0.48 23.0 0.39 0.43 100.0 1.00 90.0 1.00 50.0 0.97

> 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.46 0.48 23.0 0.44 0.46 100.0 1.00 90.0 1.00 50.0 0.96 TBVOOS > 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.39 0.44 23.0 0.37 0.40

• LHGRFACp multipliers do not support MELLLA+ operation prior to 4.75 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 8-11 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 1.00 Base > 65%F 65%F case 50.0 0.89 0.95 operation 26.0 0.63 0.77 26.0 0.51 0.53 23.0 0.49 0.50 100.0 1.00 90.0 1.00 50.0 1.00

> 65%F 65%F TBVOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.42 0.48 23.0 0.39 0.43 100.0 1.00 90.0 1.00 50.0 0.97

> 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.46 0.48 23.0 0.44 0.46 100.0 1.00 90.0 1.00 50.0 0.96 TBVOOS > 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.39 0.44 23.0 0.37 0.40

• LHGRFACp multipliers do not support MELLLA+ operation prior to 4.75 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 8-12 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 0.97 Base

> 65%F 65%F case 50.0 0.89 0.95 operation 26.0 0.63 0.77 26.0 0.46 0.48 23.0 0.44 0.46 100.0 1.00 90.0 1.00 50.0 0.96

> 65%F 65%F TBVOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.39 0.44 23.0 0.37 0.40

• 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.

LHGRFACp multipliers do not support MELLLA+ operation prior to 4.75 GWd/MTU.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis 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 0.97 Base

> 65%F 65%F case 50.0 0.89 0.95 operation 26.0 0.63 0.77 26.0 0.46 0.48 23.0 0.44 0.46 100.0 1.00 90.0 1.00 50.0 0.96

> 65%F 65%F TBVOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.39 0.44 23.0 0.37 0.40

• 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.

LHGRFACp multipliers do not support MELLLA+ operation prior to 4.75 GWd/MTU.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis Page 8-14 Table 8.13 ATRIUM 10XM LHGRFACf Multipliers All Cycle 22 Exposures*

Core Flow

(% of rated) LHGRFACf 0.0 0.58 31.0 0.58 75.0 1.00 107.0 1.00 Table 8.14 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

• LHGRFACf multipliers do not support MELLLA+ operation prior to 4.75 GWd/MTU.

Controlled Document AREVA Innc. ANP-36 636NP Revission 0 k Unit 1 Cycle Brunswick e 22 Reload Safety S Analysis Pagge 9-1 9.0 References R

1. ANP-3108P A Revision R 1, Applicability A of AREVA B BWR Methods to Brunsw wick Extend ded Power P Flow Operating O Doomain, July 2015.

2. NEDO-33006 N 6-A Revisionn 3, General Electric Boililing Water RReactor Maxximum Exten nded Load Line Lim mit Analysis Plus, Generral Electric HHitachi Nucle ear Energy AAmerica, LLC C, Ju une 2009. (a available in ADAMS A Acccession Num mber ML0918 800530)

3. 38-9272009-0 000, Revisioon 0, BNEI-0 0400-0013, B B1C22 Finall Fuel Cycle Design El Transmittal T (N NF17-046), May M 2017.

4. ANP-2956(P)

A ) Revision 0,, Brunswick Unit 2 Cycle e 20 Reload d Safety Anallysis, AREVA A NP, October O 2010.

5. FS1-0032402 2, Revision 1.0, 1 Brunswick Unit 1 Cyycle 22 Calcculation Plan n, June 2017 7.

6. ANP-2948(P)

A ) Revision 2,, Mechanica al Design Re eport for Brunnswick ATR RIUM 10XM F Fuel Assemblies, A AREVA A NP, January 2017.

7. ANP-3606P A Revision R 0, ATRIUM A 10XXM Fuel Rod d Thermal-MMechanical E Evaluation foor Brunswick B Unnit 1 Cycle 22, 2 AREVA, November 2 2017.

8. Letter, Edmond G. Tourig gny (NRC) to o E.E. Utley (CP&L), Isssuance of A Amendment No.

N 124 to Fa acility Opera ating License e No. DPR-7 71 - Brunswick Steam E Electric Plantt, Unit U 1, Regarrding Fuel Cycle No. 7 ReloadR (TACC No. 69200)), February 6, 1989 (3

38-9061815--000).

9. ANP-2989(P)

A ) Revision 0,, Brunswick Unit 1 Therm mal-Hydraullic Design RReport for ATRIUM' A 10 0XM Fuel As ssemblies, AREVA A NP, May 2011.

10. ANP-10307P A PA Revision 0,0 AREVA MCPRM Safetyy Limit Meth hodology for Boiling Watter Reactors, R AR REVA NP, Ju une 2011.

11. ANP-10298P A PA Revision 1, ACE/ATR RIUM 10XM Critical Pow wer Correlatio on, AREVA NP, March M 2014.

12. NEDO-32465 N 5-A, Reactorr Stability Deetect and Suuppress Solu utions Licenssing Basis Methodology M and Reload d Applicationns, GE Nucle ear Energy, AAugust 1996 6.

13. BAW-10255P B PA Revision 2,2 Cycle-Spe ecific DIVOMM Methodolo ogy Using thee RAMONA5 5-FA Code, C AREVA A NP, May 2008.

2

14. OG02-0119-2 O 260, Backup p Stability Protection (BSSP) for Inope erable Option III Solutionn, GE G Nuclear Energy, E July 17, 2002.

15. EMF-CC-074 E 4(P)(A) Volum me 4 Revisio on 0, BWR S Stability Ana alysis - Asse essment of S STAIF with w Input from m MICROBU URN-B2, Sie emens Powe er Corporatio on, August 22000.

16. NEDO-33075 N 5-A Revisionn 8, GE Hitac chi Nuclear E Energy, GE Hitachi Boilling Water Reactor, R Dete ect and Supp press Solutiion - Confirm mation Denssity, Novemb ber 2013.

17. NEDO-33728 N 8, Revision 2, 2 Safety Ana alysis Reporrt for Brunswwick Steam EElectric Plannt Units U 1 and 2 Maximum Extended E Looad Line Limmit Analysis PPlus, Octobeer 2015.

(3 38-9251103--001)

18. ANF-913(P)(A A A) Volume 1 Revision 1 and Volume e 1 Supplem ments 2, 3 annd 4, COTRANSA2 C 2: A Computter Program for Boiling W Water Reacttor Transient Analyses, Advanced A Nuuclear Fuels Corporation n, August 19990.

Controlled Document AREVA Inc. ANP-3636NP Revision 0 Brunswick Unit 1 Cycle 22 Reload Safety Analysis 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. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.

22. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.

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-2941(P) Revision 0, Brunswick Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM' 10XM Fuel, AREVA NP, September 2010.

26. ANP-2943(P) Revision 3, Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel, 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. 002N5935 Revision 0, Backup Stability Protection Region Endpoint Determination, November 2015 (38-9248830-000).

30. ANP-2962(P) Revision 0, Brunswick Nuclear Plant New Fuel Storage Vault Criticality Safety Analysis for ATRIUM' 10XM Fuel, AREVA NP, November 2010.

31. ANP-2955(P) Revision 3, Brunswick Nuclear Plant Spent Fuel Pool Criticality Safety Analysis for ATRIUM' 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.

33. FS1-0028074 Revision 1.0, 10 CFR 50.46 PCT Reporting for the Brunswick Units ATRIUM 10XM Fuel, AREVA, December 2016.

34. ANP-3105P Revision 1, Brunswick Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel for MELLLA+, AREVA, July 2015.

35. ANP-3106P Revision 2, Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+, AREVA, December 2015.

RA-18-0101 Enclosure 3 Affidavit Regarding Withholding ANP-3636P, Brunswick Unit 1 Cycle 22 Reload Safety Analysis, Revision 0

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-3636P, Revision 0, "Brunswick Unit 1 Cycle 22 Reload Safety Analysis," dated December 2017 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 Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA 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 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 -~~-/J\?+

t-day of 1)ecember , 2017.

Hailey M Siekawitch NOTARY PUBLIC, STATE OF WASHINGTON MY COMMISSION EXPIRES: 9/28/2020

RA-18-0101 Enclosure 4 (Proprietary Information - Withhold from Public Disclosure in Accordance With 10 CFR 2.390)

ANP-3560P, Brunswick Unit 2 Cycle 23 Reload Safety Analysis, Revision 0 (Proprietary Information - Withhold from Public Disclosure in Accordance With 10 CFR 2.390)

RA-18-0101 Enclosure 5 ANP-3560NP, Brunswick Unit 2 Cycle 23 Reload Safety Analysis, Revision 0

Controlled Document Brunswick Unit 2 Cycle 23 ANP-3560NP Revision 0 Reload Safety Analysis January 2017 AREVA Inc.

(c) 2017 AREVA Inc.

Controlled Document ANP-3560NP Revision 0 Copyright © 2017 AREVA Inc.

All Rights Reserved

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Initial Issue AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page ii Contents

1.0 INTRODUCTION

.............................................................................................................1 2.0 DISPOSITION OF EVENTS ............................................................................................6 3.0 MECHANICAL DESIGN ANALYSIS ................................................................................7 4.0 THERMAL-HYDRAULIC DESIGN ANALYSIS .................................................................8 4.1 Thermal-Hydraulic Design and Compatibility........................................................8 4.2 Safety Limit MCPR Analysis ................................................................................8 4.3 Core Hydrodynamic Stability ................................................................................9 4.3.1 MELLLA BWROG Long Term Stability Solution Option III ........................9 4.3.2 MELLLA+ Stability DSS-CD Solution ......................................................10 4.3.3 MELLLA+ DSS-CD Backup Stability Protection ...................................... 10 4.4 Voiding in the Channel Bypass Region ..............................................................11 5.0 ANTICIPATED OPERATIONAL OCCURRENCES ........................................................20 5.1 System Transients .............................................................................................20 5.1.1 Load Rejection No Bypass (LRNB).........................................................22 5.1.2 Turbine Trip No Bypass (TTNB) .............................................................23 5.1.3 Feedwater Controller Failure (FWCF).....................................................23 5.1.4 Pressure Regulator Failure Downscale (PRFDS) ................................... 24 5.1.5 Loss of Feedwater Heating .....................................................................24 5.1.6 Control Rod Withdrawal Error .................................................................25 5.2 Slow Flow Runup Analysis .................................................................................25 5.3 Equipment Out-of-Service Scenarios .................................................................26 5.3.1 FHOOS ..................................................................................................27 5.3.2 TBVOOS ................................................................................................28 5.3.3 PROOS ..................................................................................................28 5.3.4 Combined FHOOS and TBVOOS ...........................................................28 5.3.5 Combined FHOOS and PROOS.............................................................28 5.3.6 Combined PROOS and TBVOOS...........................................................28 5.3.7 Combined PROOS, FHOOS, and TBVOOS ........................................... 29 5.3.8 One SRVOOS ........................................................................................29 5.3.9 One MSIVOOS .......................................................................................29 5.3.10 Single-Loop Operation............................................................................30 5.4 Licensing Power Shape .....................................................................................30 6.0 POSTULATED ACCIDENTS .........................................................................................50 6.1 Loss-of-Coolant Accident (LOCA) ......................................................................50 6.2 Control Rod Drop Accident (CRDA) ...................................................................50 6.3 Fuel and Equipment Handling Accident .............................................................51 6.4 Fuel Loading Error (Infrequent Event) ................................................................51 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page iii 7.0 SPECIAL ANALYSES ...................................................................................................53 7.1 ASME Overpressurization Analysis ....................................................................53 7.2 ATWS Event Evaluation .....................................................................................54 7.2.1 ATWS Overpressurization Analysis ........................................................54 7.2.2 Long-Term Evaluation ............................................................................55 7.3 Standby Liquid Control System ..........................................................................55 7.4 Fuel Criticality ....................................................................................................56 7.5 Strongest Rod Out Shutdown Margin .................................................................56 8.0 OPERATING LIMITS AND COLR INPUT ......................................................................68 8.1 MCPR Limits ......................................................................................................68 8.2 LHGR Limits ......................................................................................................68 8.3 MAPLHGR Limits ...............................................................................................69

9.0 REFERENCES

..............................................................................................................85 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page iv Tables Table 1.1 EOOS Operating Conditions ......................................................................................3 Table 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses ................... 12 Table 4.2 Results Summary for Safety Limit MCPR Analyses ..................................................13 Table 4.3 OPRM Setpoints.......................................................................................................14 Table 4.4 BSP Endpoints for Brunswick Unit 2 Cycle 23 ..........................................................15 Table 4.5 DSS-CD BSP Endpoints For Nominal Feedwater Temperature................................. 16 Table 4.6 DSS-CD BSP Endpoints For Reduced Feedwater Temperature .............................. 17 Table 4.7 ABSP Setpoints for the Scram Region .....................................................................18 Table 4.8 Maximum Bypass Voiding at LPRM Level D.............................................................19 Table 5.1 Exposure Basis for Brunswick Unit 2 Cycle 23 Transient Analysis ........................... 31 Table 5.2 Scram Speed Insertion Times ..................................................................................32 Table 5.3 EOCLB Base Case LRNB Transient Results ............................................................33 Table 5.4 EOCLB Base Case TTNB Transient Results ............................................................34 Table 5.5 EOCLB Base Case FWCF Transient Results ...........................................................35 Table 5.6 Loss of Feedwater Heating Transient Analysis Results ............................................ 36 Table 5.7 Control Rod Withdrawal Error CPR Results ...........................................................36 Table 5.8 RBM Operability Requirements ................................................................................37 Table 5.9 Flow-Dependent MCPR Results ...............................................................................37 Table 5.10 EOCLB LHGRFACp EOOS ATRIUM 10XM Transient Results .............................. 38 Table 5.11 EOCLB LHGRFACp EOOS ATRIUM 11 LTA Transient Results ............................. 39 Table 5.12 Licensing Basis Core Average Axial Power Profile .................................................40 Table 7.1 ASME Overpressurization Analysis Results ............................................................57 Table 7.2 ASME Overpressurization Sensitivity Analysis Results* .......................................... 57 Table 7.3 ATWS Overpressurization Analysis Results ............................................................58 Table 7.4 ATWS Overpressurization Sensitivity Analysis Results ........................................... 59 Table 8.1 MCPRp Limits for NSS Insertion Times BOC to < EOCLB ........................................ 70 Table 8.2 MCPRp Limits for TSSS Insertion Times BOC to < EOCLB ...................................... 72 Table 8.3 MCPRp Limits for NSS Insertion Times FFTR/Coastdown ........................................ 74 Table 8.4 MCPRp Limits for TSSS Insertion Times FFTR/Coastdown ...................................... 75 Table 8.5 Flow-Dependent MCPR Limits ATRIUM 10XM Fuel .................................................76 Table 8.6 Flow-Dependent MCPR Limits ATRIUM 11 LTA.......................................................76 Table 8.7 Steady-State LHGR Limits .......................................................................................77 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page v Table 8.8 LHGRFACp Multipliers for NSS Insertion Times BOC to < EOCLB ........................... 78 Table 8.9 LHGRFACp Multipliers for TSSS Insertion Times BOC to < EOCLB ......................... 80 Table 8.10 LHGRFACp Multipliers for NSS Insertion Times FFTR/Coastdown ......................... 82 Table 8.11 LHGRFACp Multipliers for TSSS Insertion Times FFTR/Coastdown ...................... 83 Table 8.12 ATRIUM 10XM and ATRIUM 11 LHGRFACf Multipliers All Cycle 23 Exposures ..........................................................................................................84 Table 8.13 AREVA Fuel MAPLHGR Limits ..............................................................................84 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page vi Figures Figure 1.1 Brunswick Unit 2 MELLLA Power/Flow Map .............................................................4 Figure 1.2 Brunswick Unit 2 MELLLA+ Power/Flow Map...........................................................5 Figure 5.1 EOCLB LRNB at 100P/104.5F - TSSS Key Parameters ......................................... 41 Figure 5.2 EOCLB LRNB at 100P/104.5F - TSSS Sensed Water Level .................................. 42 Figure 5.3 EOCLB LRNB at 100P/104.5F - TSSS Vessel Pressures....................................... 43 Figure 5.4 EOCLB TTNB at 100P/104.5F - TSSS Key Parameters ......................................... 44 Figure 5.5 EOCLB TTNB at 100P/104.5F - TSSS Sensed Water Level................................... 45 Figure 5.6 EOCLB TTNB at 100P/104.5F - TSSS Vessel Pressures ....................................... 46 Figure 5.7 EOCLB FWCF at 100P/104.5F - TSSS Key Parameters ........................................ 47 Figure 5.8 EOCLB FWCF at 100P/104.5F - TSSS Sensed Water Level.................................. 48 Figure 5.9 EOCLB FWCF at 100P/104.5F - TSSS Vessel Pressures ...................................... 49 Figure 7.1 MSIV Closure Overpressurization Event at 102P/104.5F - Key Parameters ........... 60 Figure 7.2 MSIV Closure Overpressurization Event at 102P/104.5F - Sensed Water Level .......................................................................................................................61 Figure 7.3 MSIV Closure Overpressurization Event at 102P/104.5F - Vessel Pressures ......... 62 Figure 7.4 MSIV Closure Overpressurization Event at 102P/104.5F - Safety/Relief Valve Flow Rates ....................................................................................................63 Figure 7.5 PRFO ATWS Overpressurization Event at 100P/85F - Key Parameters ................. 64 Figure 7.6 PRFO ATWS Overpressurization Event at 100P/85F - Sensed Water Level .......... 65 Figure 7.7 PRFO ATWS Overpressurization Event at 100P/85F - Vessel Pressures............... 66 Figure 7.8 PRFO ATWS Overpressurization Event at 100P/85F - Safety/Relief Valve Flow Rates ..............................................................................................................67 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page vii Nomenclature ABSP automated backup stability protection APRM average power range monitor AOO anticipated operational occurrence ARO all control rods out ASME American Society of Mechanical Engineers AST alternative source term ATWS anticipated transient without scram ATWS-RPT anticipated transient without scram recirculation pump trip BOC beginning-of-cycle BPWS banked position withdrawal sequence BSP backup stability protection BWROG Boiling Water Reactor Owners Group CDA confirmation density algorithm CFR Code of Federal Regulations COLR core operating limits report CPR critical power ratio CRDA control rod drop accident CRWE control rod withdrawal error DSS-CD detect and suppress solution - confirmation density EFPD effective full-power days EFPH effective full-power hours EOC end-of-cycle EOCLB end-of-cycle licensing basis EOFP end of full power EOOS equipment out-of-service FFTR final feedwater temperature reduction FHA fuel handling accident FHOOS feedwater heaters out-of-service FWCF feedwater controller failure GE General Electric GSF generic shape function HCOM hot channel oscillation magnitude HFCL high flow control line LFWH loss of feedwater heating LHGR linear heat generation rate LHGRFACf flow-dependent linear heat generation rate multipliers LHGRFACp power-dependent linear heat generation rate multipliers LOCA loss-of-coolant accident LPRM local power range monitor LRNB generator load rejection with no bypass LTA lead test assembly (same as LUA - lead use assembly)

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page viii Nomenclature (Continued)

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

OLMCPR operating limit minimum critical power ratio OOS out-of-service OPRM oscillation power range monitor Pbypass power below which direct scram on TSV/TCV closure is bypassed PCT peak cladding temperature PLU power load unbalance PRFDS pressure regulator failure downscale PRFO pressure regulator failure open PROOS pressure regulator out-of-service RBM (control) rod block monitor RDF rated drive flow RHR residual heat removal RPS reactor protection system RPT recirculation pump trip RTP rated thermal power SAD amplitude discriminator setpoint SLC standby liquid control SLMCPR safety limit minimum critical power ratio SLO single-loop operation SRV safety/relief valve SRVOOS safety/relief valve out-of-service SS steady state STP simulated thermal power TBVOOS turbine bypass valves out-of-service TCV turbine control valve TIP traversing incore probe TLO two-loop operation TSSS technical specifications scram speed TSV turbine stop valve TTNB turbine trip with no bypass CPR change in critical power ratio 2PT 2 pump trip AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 1

1.0 INTRODUCTION

Reload licensing analyses results generated by AREVA Inc. are presented in support of Brunswick Unit 2 Cycle 23. The analyses reported in this document were performed using methodologies previously approved for generic application to boiling water reactors and demonstrated in Reference 1 to be applicable to the MELLLA+ extended flow operating domain, Reference 2. The NRC technical limitations associated with the application of the approved methodologies have been satisfied by these analyses.

The Cycle 23 core consists of a total of 560 fuel assemblies, including 232 fresh ATRIUM' 10XM* assemblies, 320 irradiated ATRIUM 10XM assemblies, and 8 once burned ATRIUM 11 lead test assemblies (LTAs). The licensing analysis supports the core design presented in Reference 3 and the use of the MELLLA+ operating domain after NRC approval is obtained. This analysis does not support MELLLA+ operation prior to 13 GWd/MTU. The analyses and thermal limits presented within are applicable for the core loading including the LTAs and the 2 reconstituted assemblies.

The Cycle 23 reload licensing analyses were performed for the potentially limiting events and analyses that were identified in the disposition of events. The results of the analyses are used to establish the Technical Specifications/COLR limits and ensure that the design and licensing criteria are met. The design and safety analyses are based on the design and operational assumptions and plant parameters provided by the utility. The results of the reload licensing analysis support operation for the power/flow map presented in Figure 1.1 and for Figure 1.2 once NRC approval is obtained. This reload licensing also supports 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.

• ATRIUM is a trademark of AREVA Inc.

Assemblies B20400 and B20431 have been reconstituted utilizing natural uranium replacement rods.

The MELLLA+ depletion assumes that operation is restricted to the MELLLA operation region until 13 GWd/MTU.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 2 For many events, the CPR response of the 8 ATRIUM 11 LTAs has been explicitly calculated.

Since the LTAs will be loaded near the edge of the core in non-limiting locations, these results are different than if the LTAs were loaded in the interior of the core.

Relative to previous cycles, the Cycle 23 MCPR safety limit has been reduced and the LHGR fuel design limit for the ATRIUM 10XM fuel has been modified. These changes have allowed more design flexibility for Cycle 23.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 3 Table 1.1 EOOS Operating Conditions*

Single-loop operation (SLO),

Turbine bypass valves out-of-service (TBVOOS)

Feedwater heaters out-of-service (FHOOS)

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

One main steam isolation valve out-of-service§ (MSIVOOS)

One pressure regulator out-of-service (PROOS)

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.

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.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 4 120.0 110.0 100.0 90.0 80.0 MELLLA 70.0

% Power 60.0 R

50.0 I e C g F i 40.0 o n

30.0 20.0 Natural Circulation 10.0 Line Minimum Power 35% Minimum Pump 0.0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr 0 10 20 30 40 50 60 70 80 90 100 110 120 (%)

Core Flow Figure 1.1 Brunswick Unit 2 MELLLA Power/Flow Map AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 5 120.0 110.0 100.0 90.0 MELLLA+

80.0 MELLLA 70.0

% Power 60.0 50.0 I C

F 40.0 30.0 20.0 Natural Circulation 10.0 Line Minimum Power Line 35% Minimum Pump 0.0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr 0 10 20 30 40 50 60 70 80 90 100 110 120 (%)

Core Flow Figure 1.2 Brunswick Unit 2 MELLLA+ Power/Flow Map AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 6 2.0 DISPOSITION OF EVENTS A disposition of events to identify the limiting events which need to be analyzed to support operation at the Brunswick Steam Electric Plant 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 10XM fuel or are performed on a cycle-specific basis.

The results of the disposition of events are presented in Reference 4. For the ATRIUM 11 LTAs, no additional disposition of events is required. The 8 once-burned ATRIUM 11 LTAs are in non-limiting locations and will have an insignificant effect on the core wide transients. As such, there are no additional events that were originally dispositioned as non-limiting which will become limiting with the introduction of the 8 LTAs.

The plant parameter differences between those used in the Brunswick Unit 2 Cycle 22 analyses and the planned analyses for the Brunswick Unit 2 Cycle 23 reload were reviewed to determine if the conclusions of the disposition of events remain applicable. The review concluded that analyses affected by the differences were included in the Reference 5 calculation plan.

Starting with Brunswick Unit 2 Cycle 23, two changes were incorporated into the transient analyses. The first is that the recirculation pump will runback to 34% of rated pump speed at a rate of 100 rpm/sec upon receiving a RPS SCRAM signal. The second change is the number of feedwater pumps in operation will vary according to the reactor power. Above Pbypass (26% of rated power), there will be two feedwater pumps in operation. Below Pbypass, only one feedwater pump will be in operation.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 7 3.0 MECHANICAL DESIGN ANALYSIS The mechanical design analyses for ATRIUM 10XM and ATRIUM 11 fuel assemblies are presented in the applicable mechanical design reports (References 6, 7, 8 and 38). The maximum exposure limits for the ATRIUM 10XM and ATRIUM 11 LTAs are:

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

Even though the ATRIUM 10XM and ATRIUM 11 LTA fuel designs are evaluated for operation to the licensed peak rod average exposure of 62 GWd/MTU, they will be limited to 60 GWd/MTU as prescribed in Brunswick Unit 2 license amendment 153 (Reference 9).

The ATRIUM 10XM and ATRIUM 11 LTA LHGR limits are presented in Section 8.0. The fuel cycle design analyses (Reference 3) have verified that the ATRIUM 10XM and ATRIUM 11 LTA fuel assemblies remain within licensed burnup limits.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 8 4.0 THERMAL-HYDRAULIC DESIGN ANALYSIS 4.1 Thermal-Hydraulic Design and Compatibility The results of the thermal-hydraulic characterization and compatibility analyses are presented in the thermal-hydraulic design report (Reference 10). The analysis results demonstrate that the thermal-hydraulic design and compatibility criteria are satisfied for the Brunswick Unit 2 transition core consisting of ATRIUM 10XM fuel assemblies and 8 ATRIUM 11 LTAs.

4.2 Safety Limit MCPR Analysis The safety limit MCPR (SLMCPR) is defined as the minimum value of the critical power ratio which ensures that less than 0.1% of the fuel rods in the core are expected to experience boiling transition during normal operation or an anticipated operational occurrence (AOO). The SLMCPR for all fuel in the Brunswick Unit 2 Cycle 23 core was determined using the methodology described in Reference 11. 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 2 Cycle 23 SLMCPR analysis used the ACE/ATRIUM 10XM critical power correlation additive constants and additive constant uncertainty described in Reference 12 for the ATRIUM 10XM fuel. The SPCB critical power correlation, described in Reference 13, was conservatively applied to the ATRIUM 11 LTAs.

In the AREVA methodology, the effects of channel bow on the critical power performance are accounted for in the SLMCPR analysis. Reference 11 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, MELLLA+ 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 MELLLA+ 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 power, and 55% core flow statepoints, the SLO core flow AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 9 uncertainty in Table 4.1 was used consistent with the MELLLA+ restrictions listed in Section 2.2.1.1 of the Reference 2 Safety Evaluation Report.

The analysis results support a two-loop operation (TLO) SLMCPR of 1.05 for MELLLA operation and 1.07 for MELLLA+ operation and a single-loop operation (SLO) SLMCPR of 1.07. Duke has submitted a change to the Brunswick Unit 2 Technical Specification SLMCPR values. The Cycle 23 operating limits are based on the expected SLMCPR values of 1.07 for TLO and 1.09 for SLO. Table 4.2 presents a summary of the analysis results including the SLMCPR and the percentage of rods expected to experience boiling transition.

4.3 Core Hydrodynamic Stability As indicated in Section 1, the reload safety analyses presented in this report have been performed to support operation in either the MELLLA or the MELLLA+ operating regions. For hydrodynamic stability, Section 4.3.1 will be applicable until NRC approval is obtained and the reactor begins to operate in the MELLLA+ operating region. Once the reactor begins MELLLA+

operation, Sections 4.3.2 and 4.3.3 will supersede Section 4.3.1.

4.3.1 MELLLA BWROG Long Term Stability Solution Option III For operation in the MELLLA domain, Brunswick has implemented BWROG Long Term Stability Solution Option III (Oscillation Power Range Monitor-OPRM). Reload validation has been performed in accordance with Reference 14. The stability-based Operating Limit MCPR (OLMCPR) is provided for two conditions as a function of OPRM amplitude setpoint in Table 4.3. The two conditions evaluated are for a postulated oscillation at 45% core flow steady-state operation (SS) and following a two recirculation pump trip (2PT) from the limiting full power operation state point. The Cycle 23 power- and flow-dependent limits provide adequate protection against violation of the SLMCPR for postulated reactor instability as long as the operating limit is greater than or equal to the specified value for the selected OPRM setpoint.

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

AREVA has performed calculations for the relative change in CPR as a function of the calculated hot channel oscillation magnitude (HCOM). These calculations were performed with the RAMONA5-FA code in accordance with Reference 15. This code is a coupled neutronic-thermal-hydraulic three-dimensional transient model for the purpose of determining the relationship between the relative change in CPR and the HCOM on a plant-specific basis. The AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 10 stability-based OLMCPRs are calculated using the most limiting of the calculated change in relative CPR for a given oscillation magnitude or the generic value provided in Reference 14.

The generic value was determined to be limiting for Cycle 23.

In cases where the OPRM system is declared inoperable for Brunswick Unit 2 Cycle 23, Backup Stability Protection (BSP) is provided in accordance with Reference 16. BSP curves have been evaluated using STAIF (Reference 17) to determine endpoints that meet decay ratio criteria for the BSP Base Minimal Region I (scram region) and Base Minimal Region II (controlled entry region). Stability boundaries based on these endpoints are then determined using the generic shape generating function from Reference 16. Analyses have been performed to support operation with both nominal and reduced feedwater temperature conditions (both FFTR and FHOOS). The endpoints for the BSP regions are provided in Table 4.4 and are the same as the regions presented in Reference 4.

4.3.2 MELLLA+ Stability DSS-CD Solution Brunswick Unit 2 will implement the stability DSS-CD solution using the Oscillation Power Range Monitor (OPRM) as described in Reference 18. Plant-specific analyses for the DSS-CD Solution are provided in Reference 19. The Detect and Suppress function of the DSS-CD solution based on the OPRM system relies on the Confirmation Density Algorithm (CDA), which constitutes 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 19 will be the stability licensing basis for Brunswick when operation in the MELLLA+ region is approved. The safety evaluation report for Reference 18 concluded that the DSS-CD solution is acceptable subject to certain cycle-specific limitations and conditions. The reload DSS-CD evaluation is performed by Duke Energy in accordance with the licensing methodology described in Reference 18 to: 1) confirm the DSS-CD Solution is applicable to Brunswick Unit 2 Cycle 23, and 2) confirm the Amplitude Discriminator Setpoint (SAD) of the CDA established in Reference 19 for operation of Brunswick Unit 2 Cycle 23.

4.3.3 MELLLA+ DSS-CD Backup Stability Protection Reference 18 describes two BSP options that are based on selected elements from three distinct constituents: BSP Manual Regions, BSP Boundary, and Automated BSP (ABSP) setpoints.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 11 The Manual BSP region boundaries and the BSP Boundary were calculated for Brunswick Unit 2 Cycle 23 using STAIF (Reference 17) for nominal and reduced feedwater temperature operation. The endpoints of the regions are defined in Table 4.5 and Table 4.6 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 18.

The ABSP Average Power Range Monitor (APRM) Simulated Thermal Power (STP) setpoints associated with the ABSP Scram Region are listed in Table 4.7. 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 NRCs 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 23 design [

]

[

]

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 12 Table 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses Parameter Uncertainty Fuel-Related Uncertainties

[

]

Plant-Related Uncertainties Feedwater flow rate 1.8%

Feedwater temperature 0.8%

Core pressure 0.8%

Total core flow rate TLO 2.5%

SLO 6%

[ ]

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 13 Table 4.2 Results Summary for Safety Limit MCPR Analyses Minimum Percentage Power/Flow Supported of Rods in Boiling

(%)

SLMCPR* Transition 100/104.5 TLO - 1.05 0.094 100/85 TLO - 1.06 0.081 80/55 TLO - 1.07 0.077 71.1/58.4 SLO - 1.07 0.079

• The OLMCPR shown in Tables 8.1 through 8.4 were developed assuming a TLO SLMCPR of 1.07 and a SLO SLMCPR of 1.09.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 14 Table 4.3 OPRM Setpoints OPRM OLMCPR OLMCPR Setpoint (SS) (2PT) 1.05 1.16 1.18 1.06 1.18 1.20 1.07 1.19 1.22 1.08 1.21 1.24 1.09 1.23 1.26 1.10 1.25 1.28 1.11 1.27 1.30 1.12 1.29 1.32 1.13 1.31 1.34 1.14 1.33 1.36 1.15 1.35 1.38 Acceptance Less than or Less than or Criteria equal to the equal to the Off-Rated Rated Power OLMCPR OLMCPR at 45% Flow as described in Section 8.0 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 15 Table 4.4 Option III BSP Endpoints for Brunswick Unit 2 Cycle 23 Feedwater Temperature Operation End Point Power Flow Mode Region Designation (% rated) (% rated)

Nominal Scram IA 56.6 40.0 Nominal Scram IB 40.7 31.0 Nominal Controlled entry IIA 64.5 50.0 Nominal Controlled entry IIB 28.5 31.0 FFTR/FHOOS Scram IA 64.9 50.5 FFTR/FHOOS Scram IB 37.3 31.0 FFTR/FHOOS Controlled entry IIA 66.1 52.0 FFTR/FHOOS Controlled entry IIB 28.5 31.0 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 16 Table 4.5 DSS-CD BSP Endpoints For Nominal Feedwater Temperature Power Flow Endpoint Definition

(%) (%)

Scram Region A1 57.0 40.6 Boundary, HFCL Scram Region B1 42.0 31.7 Boundary, NCL Controlled Entry A2 64.5 50.0 Region Boundary, HFCL Controlled Entry B2 28.9 31.9 Region Boundary, NCL AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 17 Table 4.6 DSS-CD BSP Endpoints For Reduced Feedwater Temperature Power Flow Endpoint Definition

(%) (%)

Scram Region A1 65.9 51.8 Boundary, HFCL Scram Region B1 36.5 31.9 Boundary, NCL Controlled Entry A2 69.8 56.8 Region Boundary, HFCL Controlled Entry B2 28.9 31.9 Region Boundary, NCL AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 18 Table 4.7 ABSP Setpoints for the Scram Region Parameter Symbol Value Slope of ABSP APRM flow-mTRIP 2.00 biased trip linear segment.

ABSP APRM flow-biased trip setpoint power intercept.

Constant Power Line for Trip PBSP-TRIP 42.0 %RTP from zero Drive Flow to Flow Breakpoint value.

ABSP APRM flow-biased trip setpoint drive flow intercept. WBSP-TRIP 37.5 %RDF Constant Flow Line for Trip.

Flow Breakpoint value WBSP-BREAK 25.0 %RDF AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 19 Table 4.8 Maximum Bypass Voiding at LPRM Level D*

Power (%) Cycle Bypass Flow (%) Exposure Void Condition (GWd/MTU) (%)

[ ]

• The voiding at LPRM level D bounds the voiding at LPRM levels A, B, and C.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 20 5.0 ANTICIPATED OPERATIONAL OCCURRENCES This section describes the analyses performed to determine the power- and flow-dependent MCPR operating limits for base case operation for Brunswick Unit 2 Cycle 23.

COTRANSA2 (Reference 20), XCOBRA-T (Reference 21), XCOBRA (Reference 22), and CASMO-4/MICROBURN-B2 (Reference 23) are the major codes used in the thermal limits analyses as described in the AREVA THERMEX methodology report (Reference 22) and neutronics methodology report (Reference 29). 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 12) is used to evaluate the thermal margin for the ATRIUM 10XM fuel. The SPCB critical power correlation (Reference 13) is conservatively used in the thermal margin evaluations for the ATRIUM 11 LTAs. Fuel pellet-to-cladding gap conductance values are based on RODEX2 (Reference 24) calculations for the Brunswick Unit 2 Cycle 23 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 and Figure 1.2.

At Brunswick, direct scram on turbine stop valve (TSV) position and turbine control valve (TCV) fast closure are bypassed at power levels less than 26% of rated (Pbypass). Scram will occur when the high pressure or high neutron flux scram setpoint is reached. Reference 25 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. Analyses were performed at cycle exposures prior to EOC to ensure that the operating limits provide the necessary protection. The end-of-cycle licensing basis (EOCLB) analysis was performed at EOFP + 15 EFPD. Analyses were also performed to support extended cycle operation with final feedwater temperature reduction AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 21 (FFTR) and power coastdown. The Brunswick Unit 2 Cycle 23 licensing basis exposures used to develop the neutronics inputs to the transient analyses are presented in Table 5.1.

All pressurization transients assumed that one of the lowest setpoint safety relief valves (SRV) was inoperable. This basis supports operation with 1 SRV out-of-service.

The Brunswick Unit 2 turbine bypass system includes ten bypass valves. However, for base case analyses in which credit is taken for turbine bypass operation, only eight 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 event less severe, the TCV initial position is further closed which tends to make the event more severe, especially at higher power levels. LRNB and TTNB events for base case operation were evaluated for both nominal and 10°F reduced feedwater temperatures.

FFTR is used to extend rated power operation by decreasing the feedwater temperature. The amount of feedwater temperature reduction is a function of power with the maximum decrease of 110.3°F at rated power. Analyses were performed to support both nominal and constant rated dome pressure with combined FFTR/Coastdown operation to the maximum licensing exposure (Table 5.1). 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.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 22 The results of the system pressurization transients are sensitive to the scram speed used in the calculations. To take advantage of average scram speeds faster than those associated with the Technical Specifications requirements, scram speed-dependent MCPRp limits are provided. The nominal scram speed (NSS) insertion times and the Technical Specifications scram speed (TSSS) insertion times used in the analyses are presented in Table 5.2. The NSS MCPRp limits can only be applied if the scram speed test results meet the NSS insertion times. System transient analyses were performed to establish MCPRp limits for both NSS and TSSS insertion times. The Brunswick Unit 2 Technical Specifications (Reference 25) allow for operation with up to 10 slow and 1 stuck control rod. One additional control rod is assumed to fail to scram.

Conservative adjustments to the NSS and TSSS scram speeds were made to the analysis inputs to appropriately account for these effects on scram reactivity. For cases below 26%

power, the results are relatively insensitive to scram speed, and only TSSS analyses are performed.

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.

The excursion of the core power due to the void collapse is terminated primarily by the reactor scram and revoiding of the core.

For power levels less than 50% of rated, the LRNB analyses assume that the power load unbalance (PLU) is inoperable. With the PLU inoperable, the LRNB sequence of events is different than the standard event. Instead of a fast closure, the TCVs close in servo mode and there is no direct scram on TCV closure. The power and pressure excursion continues until the high pressure scram occurs. Given that there is no direct scram when the PLU is inoperable, the above and below Pbypass system responses at 26% power are identical.

LRNB analyses were performed for a range of power/flow conditions to support generation of the thermal limits. Table 5.3 presents the base case limiting LRNB transient analysis results used to generate the EOCLB operating limits for both TSSS and NSS insertion times. Figures AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 23 5.1 - 5.3 show the responses of various reactor and plant parameters during the LRNB event initiated at 100% of rated power and 104.5% of rated core flow with TSSS insertion times.

5.1.2 Turbine Trip No Bypass (TTNB)

The turbine trip causes a closure of the turbine stop valves. The resulting compression wave travels through the steam lines into the vessel and creates a rapid pressurization. The increase in pressure causes a decrease in core voids, which in turn causes a rapid increase in power.

The closure of the turbine stop valves also causes a reactor scram. Turbine bypass system operation, which also mitigates the consequences of the event, is not credited. The excursion of the core power due to the void collapse is terminated primarily by the reactor scram and revoiding of the core.

TTNB analyses were performed for a range of power/flow conditions for which the TTNB event is potentially limiting to support generation of the thermal limits. Table 5.4 presents the base case TTNB transient analysis results used to generate the EOCLB operating limits for both TSSS and NSS insertion times. Figures 5.4 - 5.6 show the responses of various reactor and plant parameters during the TTNB event initiated at 100% of rated power and 104.5% of rated core flow with TSSS insertion times.

5.1.3 Feedwater Controller Failure (FWCF)

The increase in feedwater flow due to a failure of the feedwater control system to maximum demand results in an increase in the water level and a decrease in the coolant temperature at the core inlet. The increase in core inlet subcooling causes an increase in core power. As the feedwater flow continues at maximum demand, the water level continues to rise and eventually reaches the high water level trip setpoint. The initial water level is conservatively assumed to be at the low-level normal operating range to delay the high-level trip and maximize the core inlet subcooling that results from the FWCF. The high water level trip causes the turbine stop valves to close in order to prevent damage to the turbine from excessive liquid inventory in the steam line. The valve closures create a compression wave that travels to the core causing a void collapse and subsequent rapid power excursion. The closure of the turbine stop valves also initiates a reactor scram. Eight of the ten installed turbine bypass valves are assumed operable 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.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 24 FWCF analyses were performed for a range of power/flow conditions to support generation of the thermal limits. Table 5.5 presents the base case limiting FWCF transient analysis results for EOCLB operating limits for both TSSS and NSS insertion times. Figures 5.7 - 5.9 show the responses of various reactor and plant parameters during the FWCF event initiated at 100% of rated power and 104.5% of rated core flow with TSSS insertion times.

5.1.4 Pressure Regulator Failure Downscale (PRFDS)

The pressure regulator failure downscale event occurs when the pressure regulator fails and sends a signal to close all four turbine control valves in control mode. Normally, the backup pressure regulator would take control and maintain the setpoint pressure, resulting in a mild pressure excursion and a benign event. If one of the pressure regulators were out-of-service, there would be no backup pressure regulator and the event would be more severe. The core would pressurize resulting in void collapse and a subsequent power increase. The event would be terminated by scram when either the high-neutron flux or high-pressure setpoint is reached.

PRFDS analyses were performed with one pressure regulator out-of-service for a range of power/flow conditions to support generation of the thermal limits. 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 pressurization occurs. For Brunswick Unit 2 Cycle 23, a cycle-specific analysis was performed in accordance with the Reference 26 methodology to AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 25 determine the change in MCPR for the event. The LFWH results for operation in MELLLA or MELLLA+ operating regions are presented in Table 5.6.

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. No CRWE results are provided for the ATRIUM 11 LTAs because they are loaded near the edge of the core where a CRWE would not have a significant effect. For the ATRIUM 10XM fuel, the rated power CRWE results are shown in Table 5.7 for selected analytical RBM high power setpoint values from 108% to 117%. An assumed RBM high power setpoint of 111%

was used to develop the MCPRp limits. At the corresponding 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 presented in Section 8.2. The recommended operability requirements based on the unblocked CRWE results are shown in Table 5.8 based on the SLMCPR values presented in Section 4.2.

5.2 Slow Flow Runup Analysis Flow-dependent MCPR and LHGR limits are established to support operation at off-rated core flow conditions. The limits are based on the CPR and heat flux changes experienced by the fuel during slow flow excursions. The slow flow excursion event assumes a failure of the recirculation flow control system such that the core flow increases slowly to the maximum flow physically permitted by the equipment (107% of rated core flow). An uncontrolled increase in flow creates the potential for a significant increase in core power and heat flux. Operation with One MSIVOOS causes a larger increase in pressure and power during the flow excursion which results in a steeper flow runup path. A conservatively steep flow runup path was used in the analysis. The slow flow runup analyses were performed to support operation in all the EOOS scenarios.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 26 XCOBRA is used to calculate the change in critical power ratio during a two-loop flow runup to the maximum flow rate. The MCPRf limit is set such that the increase in core power, resulting from the maximum increase in core flow, assures that the TLO safety limit MCPR is not violated.

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

Results of the flow runup analysis are presented in Table 5.9. MCPRf limits that provide the required protection are presented in Tables 8.5 and 8.6. The MCPRf limits are applicable for all Cycle 23 exposures.

Flow runup analyses were performed with CASMO-4/MICROBURN-B2 to determine flow-dependent LHGR multipliers (LHGRFACf) for ATRIUM 10XM and ATRIUM 11 LTAs. The analysis assumes that the recirculation flow increases slowly along the limiting rod line to the maximum flow physically permitted by the equipment. A series of flow excursion analyses were performed at several exposures throughout the cycle starting from different initial power/flow conditions. Xenon is assumed to remain constant during the event. The LHGRFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a flow runup. The Cycle 23 LHGRFACf multipliers are presented in Table 8.12 for ATRIUM 10XM fuel and the ATRIUM 11 LTAs.

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 2 Cycle 23 MELLLA operation:

• Feedwater heater out-of-service (FHOOS) - up to 110.3°F feedwater temperature reduction

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

• Pressure regulator out-of-service (PROOS)

• Combined FHOOS and TBVOOS

• Combined FHOOS and PROOS

• Combined PROOS and TBVOOS AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 27

• Combined PROOS, 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)

The following EOOS scenarios are supported for Brunswick Unit 2 Cycle 23 MELLLA+

operation:

• TBVOOS

• PROOS

• Combined PROOS and TBVOOS

• One SRVOOS

• One MSIVOOS Table 5.10 and Table 5.11 present the limiting LHGRFACp transient analysis results for each EOOS scenario used to generate the 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. In addition, the steam flow for a given power level decreases since more power is required to increase the enthalpy of the coolant to saturated conditions. The consequences of the FWCF event are potentially more severe as a result of the increase in core inlet subcooling during the overcooling phase of the event. While the decrease in steam flow tends to make the LRNB event less severe, the TCV initial position is further closed which tends to make the event more severe, especially at higher power levels. FWCF events were analyzed to ensure that appropriate FHOOS operating limits are established. Operation with FHOOS or the related FFTR scenario is not allowed in the MELLLA+ region.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 28 5.3.2 TBVOOS For this EOOS scenario, operation with TBVOOS means that the fast opening capability of three or more of the turbine bypass valves cannot be assured, thereby reducing the pressure relief capacity during fast pressurization transients. While the base case LRNB and TTNB events are analyzed assuming the turbine bypass valves out-of-service, operation with TBVOOS has an adverse effect on the FWCF event. Analyses of the FWCF event with TBVOOS were performed to establish the TBVOOS operating limits.

5.3.3 PROOS For this EOOS scenario, operation with PROOS means that there is no backup pressure regulator and the PRFDS event would be more severe. All four turbine control valves close in control mode. The core would pressurize resulting in void collapse and a subsequent power increase. The event would be terminated by scram when either the high-neutron flux or high-pressure setpoint is reached. The PROOS EOOS only impacts the PRFDS event. Analyses of the PRFDS with PROOS were performed to establish the PROOS operating limits.

5.3.4 Combined FHOOS and TBVOOS FWCF analyses with both FHOOS and TBVOOS were performed. Operating limits for this combined EOOS scenario were established using these FWCF results. This scenario is not allowed in the MELLLA+ region.

5.3.5 Combined FHOOS and PROOS PRFDS analyses with both FHOOS and PROOS were performed. Operating limits for this combined EOOS scenario were established using these PRFDS results. This scenario is not allowed in the MELLLA+ region.

5.3.6 Combined PROOS and TBVOOS Limits were established to support operation with both PROOS and TBVOOS. No additional analyses are required to construct MCPRp operating limits for PROOS and TBVOOS since PROOS and TBVOOS are independent EOOS conditions (PROOS only impacts PRFDS events and TBVOOS only impacts FWCF events).

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 29 5.3.7 Combined PROOS, FHOOS, and TBVOOS Limits were established to support operation with PROOS, FHOOS, and TBVOOS. No additional analyses are required to construct MCPRp operating limits for PROOS, FHOOS, and TBVOOS since PROOS and TBVOOS are independent EOOS conditions (PROOS only impacts PRFDS events and TBVOOS only impacts FWCF events). This scenario is not allowed in the MELLLA+ region.

5.3.8 One SRVOOS As noted earlier, all pressurization transient analyses were performed with one of the lowest setpoint SRVs assumed inoperable. Therefore, the base case operating limits support operation with one SRVOOS. The EOOS operating limits also support operation with one SRVOOS.

5.3.9 One MSIVOOS Operation with One MSIVOOS is supported for operation less than 70% of rated power. At these reduced power levels, the flow through any one steam line will not be greater than the flow at rated power when all MSIVs are available. Since all four turbine control valves are available, adequate pressure control can be maintained. The main difference in operation with One MSIVOOS is that the steam line pressure drop between the steam dome and the turbine valves is higher than if all MSIVs are available. Since low steam line pressure drop is limiting for pressurization transients, the results of the pressurization events with all MSIVs in service bound the results with One MSIVOOS. In addition, operation with One MSIVOOS has no impact on the other nonpressurization events evaluated to establish power-dependent operating limits.

Therefore, the power-dependent operating limits applicable to base case operation with all MSIVs in service remain applicable for operation with One MSIVOOS for power levels less than or equal to 70% of rated. As noted earlier, slow flow runup analyses were performed to support operation with One MSIVOOS.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 30 5.3.10 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 limiting CPRs 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 SLMCPR as discussed in Section 4.2, and the application of an SLO MAPLHGR multiplier discussed in Section 8.3. The net result is a 0.02 increase in the base case MCPRp limits and a decrease in the MAPLHGR limit. The same situation is true for the EOOS scenarios. Adding 0.02 to the corresponding two-loop operation EOOS MCPRp limits results in SLO MCPRp limits for the EOOS conditions. The TLO EOOS LHGRFAC multipliers remain applicable in SLO. This scenario is not allowed in the MELLLA+ region.

5.4 Licensing Power Shape The licensing axial power profile used by AREVA for the plant transient analyses bounds the projected end of full power axial power profile. The conservative licensing axial power profile generated at the EOCLB core average exposure of 35,915 MWd/MTU is given in Table 5.12.

Cycle 23 operation is considered to be in compliance when:

• The integrated normalized power generated in the bottom 7 nodes from the projected EOFP solution at the state conditions provided in Table 5.12 is greater than the integrated normalized power generated in the bottom 7 nodes in the licensing basis axial power profile, and the individual normalized power from the projected EOFP solution is greater than the corresponding normalized power from the licensing basis axial power profile for at least 6 of the 7 bottom nodes.

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

If the criteria cannot be fully met, the licensing basis may nevertheless remain valid but further assessment will be required.

The licensing basis power profile in Table 5.12 was calculated using the MICROBURN-B2 code.

Compliance analyses must also be performed using MICROBURN-B2. Note that the power profile comparison should be done without incorporating instrument updates to the axial profile because the updated power is not used in the core monitoring system to accumulate assembly burnups.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 31 Table 5.1 Exposure Basis for Brunswick Unit 2 Cycle 23 Transient Analysis Cycle Core Exposure at Average End of Interval Exposure (MWd/MTU) (MWd/MTU)* Comments 0 17,234 Beginning of cycle 18,681 35,915 Design basis rod patterns to EOFP + 15 EFPD (EOCLB) 20,113 37,347 Maximum licensing core exposure - including FFTR

/Coastdown

• Note that the limits presented in Tables 8.1 - 8.4 and Tables 8.8 - 8.11 are based on core average exposure.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 32 Table 5.2 Scram Speed Insertion Times Control Rod TSSS NSS Position Time Time (notch) (sec) (sec) 48 (full-out) 0.000 0.000 48 0.200 0.200 46 0.440 0.322 36 1.080 0.862 26 1.830 1.422 6 3.350 2.593 0 (full-in) 3.806 2.944 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 33 Table 5.3 EOCLB Base Case LRNB Transient Results ATRIUM ATRIUM 11 ATRIUM ATRIUM 11 10XM LTA 10XM LTA Supported Supported Power CPR CPR LHGRFACp LHGRFACp TSSS Insertion Times 100 0.30 0.35 1.00 1.00 90 0.31 0.37 1.00 1.00 80 0.32 0.37 0.98 0.98 70 0.33 0.37 0.96 0.96 60 0.30 0.36 0.94 0.94 50 0.31 0.33 0.92 0.92 50 at > 65%F PLU inoperable 0.73 0.89 0.86 0.86 50 at 65%F PLU inoperable 0.63 0.79 0.86 0.86 26 at > 65%F PLU inoperable 1.12 1.27 0.64 0.64 26 at 65%F PLU inoperable 1.01 1.19 0.66 0.66 26 at > 65%F below Pbypass 1.12 1.27 0.64 0.64 26 at 65%F below Pbypass 1.01 1.19 0.66 0.66 23 at > 65%F below Pbypass 1.21 1.33 0.60 0.60 23 at 65%F below Pbypass 1.09 1.27 0.64 0.64 NSS Insertion Times 100 0.27 0.33 1.00 1.00 90 0.29 0.34 1.00 1.00 80 0.30 0.35 0.98 0.98 70 0.30 0.35 0.96 0.96 60 0.29 0.35 0.94 0.94 50 0.27 0.33 0.92 0.92 50 at > 65%F PLU inoperable 0.72 0.88 0.86 0.86 50 at 65%F PLU inoperable 0.61 0.78 0.86 0.86 26 at > 65%F PLU inoperable 1.11 1.26 0.64 0.64 26 at 65%F PLU inoperable 0.98 1.17 0.66 0.66 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 34 Table 5.4 EOCLB Base Case TTNB Transient Results ATRIUM ATRIUM 11 ATRIUM ATRIUM 11 10XM LTA 10XM LTA Supported Supported Power CPR CPR LHGRFACp LHGRFACp TSSS Insertion Times 100 0.30 0.36 1.00 1.00 90 0.31 0.36 1.00 1.00 80 0.32 0.37 0.98 0.98 26 at > 65%F below Pbypass 1.12 1.26 0.64 0.64 26 at 65%F below Pbypass 1.01 1.19 0.66 0.66 23 at > 65%F below Pbypass 1.21 1.33 0.60 0.60 23 at 65%F below Pbypass 1.09 1.27 0.64 0.64 NSS Insertion Times 100 0.27 0.33 1.00 1.00 90 0.29 0.34 1.00 1.00 80 0.30 0.35 0.98 0.98 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 35 Table 5.5 EOCLB Base Case FWCF Transient Results ATRIUM ATRIUM 11 ATRIUM ATRIUM 11 10XM LTA 10XM LTA Supported Supported Power CPR CPR LHGRFACp LHGRFACp TSSS Insertion Times 100 0.21 0.23 1.00 1.00 50 0.28 0.31 0.92 0.92 26 0.59 0.71 0.86 0.86 26 at > 65%F below Pbypass 0.70 0.84 0.66 0.66 26 at 65%F below Pbypass 0.33 0.33 0.66 0.66 23 at > 65%F below Pbypass 0.78 0.93 0.64 0.64 23 at 65%F below Pbypass 0.36 0.36 0.64 0.64 NSS Insertion Times 100 0.19 0.21 1.00 1.00 50 0.27 0.30 0.92 0.92 26 0.55 0.68 0.86 0.86 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 36 Table 5.6 Loss of Feedwater Heating Transient Analysis Results Power ATRIUM 10XM ATRIUM 11 LTA

(% rated) CPR CPR*

100 0.10 0.28 90 0.11 0.29 80 0.11 0.30 70 0.12 0.31 60 0.14 0.32 50 0.16 0.34 40 0.19 0.37 30 0.24 0.42 23 0.30 0.48 Table 5.7 Control Rod Withdrawal Error CPR Results Analytical RBM Setpoint (without filter) ATRIUM 10XM

(%) CPR 108 0.21 111 0.26 114 0.28 117 0.32

• The higher CPR responses for the ATRIUM 11 LTA LFWH are due to the non-limiting location chosen for these assemblies. These results would be smaller if the assemblies started the event with less CPR margin (i.e. if they were loaded near the center of the core).

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 37 Table 5.8 RBM Operability Requirements Thermal Power

(% rated) Applicable OLMCPR 1.86 TLO 29% and < 90%

1.89 SLO 90% 1.46 TLO Table 5.9 Flow-Dependent MCPR Results Core Flow ATRIUM 10XM ATRIUM 11 LTA

(% rated) Limiting MCPR Limiting MCPR 31 1.58 1.80 40 1.53 1.67 50 1.51 1.58 60 1.46 1.51 70 1.33 1.42 80 1.27 1.35 90 1.21 1.27 100 1.13 1.19 107 1.07 1.12 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 38 Table 5.10 EOCLB LHGRFACp EOOS ATRIUM 10XM Transient Results FHOOS/ FHOOS/

FHOOS TBVOOS TBVOOS PROOS PROOS Power FWCF FWCF FWCF PRFDS PRFDS TSSS Insertion Times 100 1.00 1.00 1.00 1.00 1.00 90 1.00 1.00 1.00 1.00 1.00 80 0.98 0.98 0.98 0.98* 0.98*

70 0.96 0.96 0.96 0.96 0.96 60 0.94 0.94 0.94 0.92 0.92 50 0.92 0.92 0.92 -- --

26 0.86 0.86 0.84 -- --

26 at > 65%F below Pbypass 0.66 0.43 0.40 -- --

26 at 65%F below Pbypass 0.66 0.50 0.46 -- --

23 at > 65%F below Pbypass 0.62 0.40 0.38 -- --

23 at 65%F below Pbypass 0.64 0.46 0.43 -- --

NSS Insertion Times 100 1.00 1.00 1.00 1.00 1.00 90 1.00 1.00 1.00 1.00 1.00 80 0.98 0.98 0.98 0.98* 0.98*

70 0.96 0.96 0.96 0.96 0.96 60 0.94 0.94 0.94 0.92 0.92 50 0.92 0.92 0.92 -- --

26 0.86 0.86 0.86 -- --

• The PRFDS event with PROOS and FHOOS/PROOS at 80% rated power was analyzed with reactor scram on high-neutron flux and high-dome pressure. The bounding LHGRFACp result is from the transient with reactor scram on high-dome pressure and is provided here.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 39 Table 5.11 EOCLB LHGRFACp EOOS ATRIUM 11 LTA Transient Results FHOOS/ FHOOS/

FHOOS TBVOOS TBVOOS PROOS PROOS Power FWCF FWCF FWCF PRFDS PRFDS TSSS Insertion Times 100 1.00 1.00 1.00 1.00 1.00 90 1.00 1.00 1.00 1.00 1.00 80 0.98 0.98 0.98 0.98* 0.98*

70 0.96 0.96 0.96 0.96 0.96 60 0.94 0.94 0.94 0.92 0.92 50 0.92 0.92 0.92 -- --

26 0.86 0.86 0.84 -- --

26 at > 65%F below Pbypass 0.66 0.43 0.40 -- --

26 at 65%F below Pbypass 0.66 0.50 0.46 -- --

23 at > 65%F below Pbypass 0.62 0.40 0.38 -- --

23 at 65%F below Pbypass 0.64 0.46 0.43 -- --

NSS Insertion Times 100 1.00 1.00 1.00 1.00 1.00 90 1.00 1.00 1.00 1.00 1.00 80 0.98 0.98 0.98 0.98* 0.98*

70 0.96 0.96 0.96 0.96 0.96 60 0.94 0.94 0.94 0.92 0.92 50 0.92 0.92 0.92 -- --

26 0.86 0.86 0.86 -- --

• The PRFDS event with PROOS and FHOOS/PROOS at 80% rated power was analyzed with reactor scram on high-neutron flux and high-dome pressure. The bounding LHGRFACp result is from the transient with reactor scram on high-dome pressure and is provided here.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 40 Table 5.12 Licensing Basis Core Average Axial Power Profile State Conditions for Power Shape Evaluation Power, MWt 2923.0 MICROBURN-B2 pressure, psia 1044.8 Inlet subcooling, Btu/lbm 20.2 Flow, Mlb/hr 80.5 Control state ARO Core average exposure (EOCLB), MWd/MTU 35,915 Licensing Axial Power Profile (Normalized)

Node Power Top 25 0.310 24 0.897 23 1.147 22 1.322 21 1.422 20 1.482 19 1.492 18 1.490 17 1.451 16 1.389 15 1.337 14 1.242 13 1.286 12 1.219 11 1.134 10 1.058 9 0.975 8 0.876 7 0.773 6 0.684 5 0.592 4 0.510 3 0.450 2 0.361 Bottom 1 0.101 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 41 Figure 5.1 EOCLB LRNB at 100P/104.5F - TSSS Key Parameters AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 42 Figure 5.2 EOCLB LRNB at 100P/104.5F - TSSS Sensed Water Level AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 43 Figure 5.3 EOCLB LRNB at 100P/104.5F - TSSS Vessel Pressures AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 44 Figure 5.4 EOCLB TTNB at 100P/104.5F - TSSS Key Parameters AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 45 Figure 5.5 EOCLB TTNB at 100P/104.5F - TSSS Sensed Water Level AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 46 Figure 5.6 EOCLB TTNB at 100P/104.5F - TSSS Vessel Pressures AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 47 Figure 5.7 EOCLB FWCF at 100P/104.5F - TSSS Key Parameters AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 48 Figure 5.8 EOCLB FWCF at 100P/104.5F - TSSS Sensed Water Level AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 49 Figure 5.9 EOCLB FWCF at 100P/104.5F - TSSS Vessel Pressures AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 50 6.0 POSTULATED ACCIDENTS 6.1 Loss-of-Coolant Accident (LOCA)

For MELLLA operation, the results of the ATRIUM 10XM LOCA analysis are presented in References 27 and 28 and provide a PCT of 1885°F. The peak local metal water reaction is 1.04% and the core wide metal water reaction is < 0.47%. The Reference 35 10 CFR 50.46 report documents an additional PCT impact of +2°F that must be accounted for. The SLO MAPLHGR multiplier is 0.80. A LOCA evaluation was performed for the ATRIUM 11 LTAs. The ATRIUM 11 LTA PCT is 1762°F. The peak local metal water reaction is 0.70% and the core wide metal water reaction is < 0.36%. The ATRIUM 11 LTAs SLO MAPLHGR multiplier is 0.80.

For MELLLA+ operation, the results of the ATRIUM 10XM LOCA analysis are presented in References 36 and 37 and provide a PCT of 1923°F. The peak local metal water reaction is 1.23% and the core wide metal water reaction is < 0.56%. The SLO MAPLHGR multiplier is 0.80; however SLO is not allowed when operating in the MELLLA+ domain. No additional LOCA evaluations were performed for the ATRIUM 11 LTAs to support MELLLA+ operation. At the time of NRC approval of MELLLA+ at Brunswick, the ATRIUM 11 LTAs will have accumulated significant exposure such that it is well beyond the limiting portion of the MAPLHGR curve, i.e.,

< 15,000 MWd/MTU, and past the limiting beginning of life conditions.

The Brunswick LOCA radiological analysis implementing the alternative source term methodology was performed in consideration of ATRIUM 10XM 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 2 uses a bank position withdrawal sequence (BPWS) including reduced notch worth rod pull to limit high worth control rod movements. A CRDA evaluation was performed for both A and B sequence startups consistent with the withdrawal sequence specified by Duke Energy. Subsequent calculations have shown that the methodology is applicable to fuel modeled with the CASMO4/MICROBURN-B2 code system. The CRDA analysis was performed with the approved methodology described in Reference 29.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 51 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 alternative source term (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 of fuel rods estimated to exceed the fuel damage threshold is below 986 for all fuel designs. Therefore, the current Brunswick CRDA AST analysis remains applicable. (Due to their non-limiting location near the edge of the core, the integrity of the ATRIUM 11 LTAs will not be challenged during a CRDA event).

Maximum dropped control rod worth, mk 9.75 Core average Doppler coefficient, k/k/oF -10.5 x 106 Effective delayed neutron fraction 0.0052 Four-bundle local peaking factor 1.479 Maximum deposited fuel rod enthalpy, cal/g 171.8 Maximum number of rods exceeding 170 cal/g 182 6.3 Fuel and Equipment Handling Accident Duke Energy has determined the radiological release assumed in the current fuel handling accident (FHA) analysis implementing the 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 10XM fuel is 161. These results are consistent with the number of failed rods supported by the current Brunswick AST analysis.

AREVA has also performed an analysis that shows that the number of failed fuel rods due to a fuel handling accident impacting the ATRIUM 11 fuel does not exceed 194. These results are consistent with the number of failed rods supported by the current Brunswick AST analysis.

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 30, the fuel loading error is characterized as an infrequent event. The acceptance criteria are that the offsite dose consequences due to the event shall not exceed a small fraction of the 10 CFR 50.67 limits.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 52 AREVA has compared the BRK2-23 OLMCPR to generic CPR behavior of ATRIUM 10XM fuel assemblies in D lattice boiling water reactors. The effect of fuel centerline melt and 1% strain limits have also been considered. It has been concluded that the BRK2-23 OLMCPR is sufficiently high and that the increase in nodal power is sufficiently low to ensure that a small fraction of the 10 CFR 50.67 limits will not be exceeded if a fuel loading error were to occur.

The effect of a fuel loading error has also been considered for the ATRIUM 11 lead test assemblies. It has been established that a small fraction of the 10 CFR 50.67 limits will not be exceeded if an LTA is misoriented or misloaded into a different position in the core.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 53 7.0 SPECIAL ANALYSES 7.1 ASME Overpressurization Analysis This section describes the maximum overpressurization analyses performed to demonstrate compliance with the ASME Boiler and Pressure Vessel Code. The analysis shows that the safety/relief valves at Brunswick Unit 2 have sufficient capacity and performance to prevent the reactor vessel pressure from reaching the safety limit of 110% of the design pressure.

An MSIV closure analysis was performed with the AREVA plant simulator code COTRANSA2 (Reference 20) for 102% power and 104.5% flow and 85% flow at the highest Cycle 23 exposure where rated power operation can be attained. The MSIV closure event is similar to the other steam line valve closure events in that the valve closure results in a rapid pressurization of the core. The increase in pressure causes a decrease in void which in turn causes a rapid increase in power. The turbine bypass valves do not impact the system response and are not modeled in the analysis. The following assumptions were made in the analysis:

• The most critical active component (direct scram on valve position) was assumed to fail.

However, scram on high neutron flux and high dome pressure is available.

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

• TSSS insertion times were used.

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

• A fast MSIV closure time of 2.7 seconds was used.

• Both MELLLA and MELLLA+ operation are supported.

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 1355 psig occurs in the lower plenum. The maximum dome pressure for the same event is 1311 psig. These peak pressure results have been adjusted to address NRC concerns associated with the void-quality correlation, exposure-dependent thermal conductivity, and Doppler effects. The results demonstrate that the maximum vessel pressure limit of 1375 psig and dome pressure limit of 1325 psig are not exceeded.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 54 A sensitivity analysis was performed to determine the impact of additional drift on the SRV opening setpoint above the 3% identified in the plant Technical Specifications for the highest setpoint SRV banks. Assuming all of the degraded valves are from the highest setpoint SRV banks 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 limit of 120% of the design pressure (1500 psig). The ATWS 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.

• 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.

• Both MELLLA and MELLLA+ operation are supported.

Results of ATWS overpressurization analyses are presented in Table 7.3. Figures 7.5 - 7.8 show the response of various reactor plant parameters during the limiting PRFO event, the event which results in the maximum vessel pressure. The maximum lower plenum pressure is 1460 psig and the maximum dome pressure is 1442 psig. The peak pressure results have been AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 55 adjusted to address NRC concerns associated with the void-quality correlation, exposure-dependent thermal conductivity, and Doppler effects. The results demonstrate that the ATWS maximum vessel pressure limit of 1500 psig is not exceeded.

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 for the highest setpoint SRV banks. Assuming all of the degraded valves are from the highest setpoint SRV banks provides a conservative scenario, and bounds the situation where the drift occurs in other SRVs. Results for the sensitivity analysis are presented in Table 7.4. The results demonstrate that the ATWS maximum vessel pressure limit of 1500 psig is not exceeded for the scenarios considered.

7.2.2 Long-Term Evaluation Fuel design differences may impact the power and pressure excursion experienced during the ATWS event. This in turn may impact the amount of steam discharged to the suppression pool and containment. For Unit 2 Cycle 20 (Reference 4) an evaluation was performed that concluded the introduction of ATRIUM 10XM fuel will not significantly impact the long-term ATWS response (suppression pool temperature and containment pressure) and the current analysis remains applicable for MELLLA operation. This conclusion remains applicable for Unit 2 Cycle 23. The presence of 8 ATRIUM 11 LTAs will not significantly impact the core wide long term response.

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

ATWS (long-term and instability) for MELLLA+ has been analyzed in Reference 19.

7.3 Standby Liquid Control System In the event that the control rod scram function becomes incapable of rendering the core in a shutdown state, the standby liquid control (SLC) system is required to be capable of bringing the reactor from full power to a cold shutdown condition at any time in the core life. The Brunswick Unit 2 SLC system is required to be able to inject 720 ppm natural boron equivalent at 70°F into the reactor coolant (including a 25% allowance for imperfect mixing, leakage, and volume of other piping connected to the reactor). AREVA has performed an analysis that demonstrates AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 56 that the SLC system meets the required shutdown capability for Cycle 23. The analysis was performed to support a coolant temperature of 360.8°F with a boron concentration equivalent to 720 ppm at 70°F. The temperature of 360.8°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 2.20% k/k.

7.4 Fuel Criticality The new fuel storage vault criticality analysis for ATRIUM 10XM fuel is presented in Reference 32. The spent fuel pool criticality analysis for ATRIUM 10XM fuel is presented in Reference 33. The ATRIUM 10XM fuel assemblies identified for loading in Cycle 23 meet both the new and spent fuel storage requirements.

Comparisons have also been performed that show the ATRIUM 11 lead test assemblies are less reactive than the ATRIUM 10XM reference bounding lattices evaluated in References 32 and 33. It then follows that the ATRIUM 11 LTAs can be safely stored in the new and spent fuel storage areas.

7.5 Strongest Rod Out Shutdown Margin The BRK2-23 MELLLA+ core has a minimum strongest rod out shutdown margin of 1.27

%k/k*. This value is produced at the beginning of the cycle at the minimum coolant temperature condition (55 °F). This value assumes that BRK2-22 ended operation at the lowest allowable exposure.

• Relative to a design goal of 1% k/k.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 57 Table 7.1 ASME Overpressurization Analysis Results*

Maximum Vessel Peak Peak Pressure Maximum Neutron Heat Lower- Dome Flux Flux Plenum Pressure Event (% rated) (% rated) (psig) (psig)

MSIV closure 266 130 1355 1311 (102P/104.5F)

Table 7.2 ASME Overpressurization Sensitivity Analysis Results*

High Bank Maximum Pressure SRV (psig)

Number of Setpoint Lower Steam Event Valves Drift Plenum Dome 3 +4%

MSIV closure 2 +6% 1368 1323 (102P/104.5F) 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.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 58 Table 7.3 ATWS Overpressurization Analysis Results*

Maximum Vessel Peak Peak Pressure Maximum Neutron Heat Lower- Dome Flux Flux Plenum Pressure Event (% rated) (% rated) (psig) (psig)

MSIV closure (100P/104.5F) 253 136 1423 1404 MSIV closure (100P/85F) 254 132 1443 1426 PRFO (100P/104.5F) 246 145 1443 1425 PRFO (100P/85F) 227 138 1460 1442

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

The maximum Technical Specification allowed SRV degradation of 3% was assumed.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 59 Table 7.4 ATWS Overpressurization Sensitivity Analysis Results*

High Bank Maximum Pressure SRV (psig)

Number of Setpoint Lower Steam Event Valves Drift Plenum Dome 3 +4%

PRFO 2 +6% 1455 1436 (100P/104.5F) 1 +8%

3 +4%

PRFO 2 +6% 1471 1454 (100P/85F) 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.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 60 Figure 7.1 MSIV Closure Overpressurization Event at 102P/104.5F - Key Parameters AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 61 Figure 7.2 MSIV Closure Overpressurization Event at 102P/104.5F - Sensed Water Level AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 62 Figure 7.3 MSIV Closure Overpressurization Event at 102P/104.5F - Vessel Pressures*

• The pressures presented in this figure do not include the adjustments associated with the NRC concerns discussed in Section 7.1.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 63 Figure 7.4 MSIV Closure Overpressurization Event at 102P/104.5F - Safety/Relief Valve Flow Rates AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 64 Figure 7.5 PRFO ATWS Overpressurization Event at 100P/85F - Key Parameters AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 65 Figure 7.6 PRFO ATWS Overpressurization Event at 100P/85F - Sensed Water Level AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 66 Figure 7.7 PRFO ATWS Overpressurization Event at 100P/85F - Vessel Pressures*

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

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 67 Figure 7.8 PRFO ATWS Overpressurization Event at 100P/85F - Safety/Relief Valve Flow Rates AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 68 8.0 OPERATING LIMITS AND COLR INPUT 8.1 MCPR Limits The determination of the MCPR limits for Brunswick Unit 2 Cycle 23 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 a two-loop operation SLMCPR of 1.07 and a single-loop operation SLMCPR of 1.09. Exposure-dependent MCPR limits were established to support operation from BOC to end-of-cycle licensing basis (EOCLB), and combined FFTR/Coastdown as defined by the core average exposures listed in Table 5.1. MCPR limits are established to support base case operation and the EOOS scenarios presented in Table 1.1.

Cycle 23 two-loop operation MCPRp limits for ATRIUM 10XM and ATRIUM 11 fuel are presented in Tables 8.1 - 8.4 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 111% was used to develop the MCPRp limits. Tables 8.1 and 8.2 present the MCPRp limits for the BOC to EOCLB exposure range. Tables 8.3 and 8.4 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 0.02 higher for all cases.

MCPRf limits that protect against fuel failures during a postulated slow flow excursion are presented in Tables 8.5 and 8.6. These MCPRf limits are applicable for all Cycle 23 exposures and the EOOS conditions identified in Table 1.1.

8.2 LHGR Limits The LHGR limits for ATRIUM 10XM and ATRIUM 11 LTAs are presented in Table 8.7. Power-and flow-dependent multipliers (LHGRFACp and LHGRFACf) are applied directly to the LHGR limits to protect against fuel melting and overstraining of the cladding during an AOO for both UO2 and gadolinia bearing rods.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 69 The ATRIUM 10XM and ATRIUM 11 LTA LHGRFACp multipliers are determined using the RODEX4 thermal-mechanical methodology (Reference 34). Exposure-dependent LHGRFACp multipliers were established to support operation from BOC to EOCLB and combined FFTR/Coastdown for both NSS and TSSS insertion times and for the EOOS conditions identified in Table 1.1. The ATRIUM 10XM and ATRIUM 11 LTA Cycle 23 LHGRFACp multipliers for the BOC to EOCLB exposure range are presented in Tables 8.8 and 8.9. The FFTR/Coastdown LHGRFACp multipliers are presented in Tables 8.10 and 8.11. 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.

LHGRFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a postulated slow flow excursion. For ATRIUM 10XM and ATRIUM 11 LTAs, the LHGRFACf multipliers are presented in Table 8.12 and are applicable for all Cycle 23 exposures and the EOOS conditions identified in Table 1.1.

8.3 MAPLHGR Limits The ATRIUM 10XM TLO MAPLHGR limits are presented in Table 8.13. For operation in SLO, a multiplier of 0.80 must be applied to the TLO MAPLHGR limits.

The ATRIUM 11 LTA TLO MAPLHGR limits are presented in Table 8.13. For operation in SLO, a multiplier of 0.80 must be applied to the TLO MAPLHGR limits.

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Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 70 Table 8.1 MCPRp Limits for NSS Insertion Times BOC to < EOCLB*,

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPRp MCPRp 100.0 1.34 1.45 80.0 1.41 1.47 50.0 1.62 1.60 Base > 65%F 65%F > 65%F 65%F case 50.0 1.81 1.70 2.02 1.92 operation 26.0 2.22 2.09 2.42 2.33 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.37 1.47 80.0 1.41 1.51 50.0 1.62 1.61

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.75 2.56 3.04 2.94 23.0 2.91 2.76 3.20 3.16 100.0 1.34 1.45 80.0 1.41 1.47 50.0 1.62 1.60

> 65%F 65%F > 65%F 65%F FHOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.34 1.45 80.0 1.42 1.53 80.0 1.55 1.70 50.0 1.81 2.02 PROOS > 65%F 65%F > 65%F 65%F 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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+.

Limits do not support MELLLA+ operation prior to 13 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 71 Table 8.1 MCPRp Limits for NSS Insertion Times BOC to < EOCLB*, (Continued)

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPRp MCPRp 100.0 1.37 1.47 80.0 1.41 1.51 50.0 1.62 1.64 TBVOOS > 65%F 65%F > 65%F 65%F FHOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30 100.0 1.34 1.45 80.0 1.42 1.53 80.0 1.55 1.70 50.0 1.81 2.02 PROOS > 65%F 65%F > 65%F 65%F FHOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.37 1.47 80.0 1.42 1.53 80.0 1.55 1.70 50.0 1.81 2.02 PROOS > 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.75 2.56 3.04 2.94 23.0 2.91 2.76 3.20 3.16 100.0 1.37 1.47 80.0 1.42 1.53 80.0 1.55 1.70 PROOS 50.0 1.81 2.02 FHOOS > 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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+.

Limits do not support MELLLA+ operation prior to 13 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 72 Table 8.2 MCPRp Limits for TSSS Insertion Times BOC to < EOCLB*,

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPRp MCPRp 100.0 1.37 1.48 80.0 1.41 1.49 50.0 1.62 1.60 Base > 65%F 65%F > 65%F 65%F case 50.0 1.82 1.72 2.03 1.93 operation 26.0 2.23 2.12 2.43 2.35 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.39 1.50 80.0 1.43 1.54 50.0 1.62 1.63

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.75 2.56 3.04 2.94 23.0 2.91 2.76 3.20 3.16 100.0 1.37 1.48 80.0 1.41 1.49 50.0 1.62 1.60

> 65%F 65%F > 65%F 65%F FHOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.37 1.48 80.0 1.43 1.55 80.0 1.57 1.71 50.0 1.82 2.03 PROOS > 65%F 65%F > 65%F 65%F 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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+.

Limits do not support MELLLA+ operation prior to 13 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 73 Table 8.2 MCPRp Limits for TSSS Insertion Times BOC to < EOCLB*, (Continued)

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPRp MCPRp 100.0 1.39 1.50 80.0 1.43 1.54 50.0 1.62 1.66 TBVOOS > 65%F 65%F > 65%F 65%F FHOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30 100.0 1.37 1.48 80.0 1.43 1.55 80.0 1.57 1.71 50.0 1.82 2.03 PROOS > 65%F 65%F > 65%F 65%F FHOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.39 1.50 80.0 1.43 1.55 80.0 1.57 1.71 50.0 1.82 2.03 PROOS > 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.75 2.56 3.04 2.94 23.0 2.91 2.76 3.20 3.16 100.0 1.39 1.50 80.0 1.43 1.55 80.0 1.57 1.71 PROOS 50.0 1.82 2.03 FHOOS > 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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+.

Limits do not support MELLLA+ operation prior to 13 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 74 Table 8.3 MCPRp Limits for NSS Insertion Times FFTR/Coastdown*,,

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPRp MCPRp 100.0 1.35 1.45 80.0 1.41 1.47 50.0 1.62 1.60 Base > 65%F 65%F > 65%F 65%F case 50.0 1.81 1.70 2.02 1.92 operation 26.0 2.22 2.09 2.42 2.33 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.37 1.47 80.0 1.41 1.51 50.0 1.62 1.64

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30 100.0 1.35 1.45 80.0 1.42 1.53 80.0 1.55 1.70 50.0 1.81 2.02 PROOS > 65%F 65%F > 65%F 65%F 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.37 1.47 80.0 1.42 1.53 80.0 1.55 1.70 50.0 1.81 2.02 PROOS

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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.

Limits do not support MELLLA+ operation prior to 13 GWd/MTU.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 75 Table 8.4 MCPRp Limits for TSSS Insertion Times FFTR/Coastdown*,,

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPRp MCPRp 100.0 1.37 1.48 80.0 1.41 1.49 50.0 1.62 1.60 Base > 65%F 65%F > 65%F 65%F case 50.0 1.82 1.72 2.03 1.93 operation 26.0 2.23 2.12 2.43 2.35 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.39 1.50 80.0 1.43 1.54 50.0 1.62 1.66

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30 100.0 1.37 1.48 80.0 1.43 1.55 80.0 1.57 1.71 50.0 1.82 2.03 PROOS > 65%F 65%F > 65%F 65%F 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.39 1.50 80.0 1.43 1.55 80.0 1.57 1.71 50.0 1.82 2.03 PROOS

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30

• Limits support operation with any combination of 1 SRVOOS, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher. 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.

Limits do not support MELLLA+ operation prior to 13 GWd/MTU.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 76 Table 8.5 Flow-Dependent MCPR Limits ATRIUM 10XM Fuel Core Flow

(% of rated) MCPRf 0.0 1.70 31.0 1.70 55.0 1.59 100.0 1.20 107.0 1.20 Table 8.6 Flow-Dependent MCPR Limits ATRIUM 11 LTA Core Flow

(% of rated) MCPRf 0.0 1.80 31.0 1.80 100.0 1.20 107.0 1.20 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 77 Table 8.7 Steady-State LHGR Limits Peak ATRIUM ATRIUM 11 Pellet 10XM LTA Exposure LHGR LHGR*

(GWd/MTU) (kW/ft) (kW/ft) 0.0 15.1 12.2 6.0 14.1 --

18.9 14.1 12.2 54.0 10.6 --

74.4 5.4 6.4

• -- indicates that the ATRIUM 11 limit does not include any breakpoint at this exposure.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 78 Table 8.8 LHGRFACp Multipliers for NSS Insertion Times BOC to < EOCLB*

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 Base > 65%F 65%F > 65%F 65%F case 50.0 0.86 0.86 0.86 0.86 operation 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92

> 65%F 65%F > 65%F 65%F TBVOOS 0.86 0.86 0.86 0.86 50.0 26.0 0.64 0.66 0.64 0.66 26.0 0.43 0.50 0.43 0.50 23.0 0.40 0.46 0.40 0.46 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92

> 65%F 65%F > 65%F 65%F FHOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86

> 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64

• LHGRFACp multipliers do not support MELLLA+ operation prior to 13 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 79 Table 8.8 LHGRFACp Multipliers for NSS Insertion Times BOC to < EOCLB* (Continued)

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 0.86 0.86 0.86 0.86 FHOOS 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86 FHOOS > 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86 TBVOOS > 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.43 0.50 0.43 0.50 23.0 0.40 0.46 0.40 0.46 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86 TBVOOS, > 65%F 65%F > 65%F 65%F FHOOS 50.0 0.86 0.86 0.86 0.86 PROOS 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43

• LHGRFACp multipliers do not support MELLLA+ operation prior to 13 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 80 Table 8.9 LHGRFACp Multipliers for TSSS Insertion Times BOC to < EOCLB*

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 Base > 65%F 65%F > 65%F 65%F case 50.0 0.86 0.86 0.86 0.86 operation 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92

> 65%F 65%F > 65%F 65%F TBVOOS 0.86 0.86 0.86 0.86 50.0 26.0 0.64 0.66 0.64 0.66 26.0 0.43 0.50 0.43 0.50 23.0 0.40 0.46 0.40 0.46 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92

> 65%F 65%F > 65%F 65%F FHOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86

> 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64

• LHGRFACp multipliers do not support MELLLA+ operation prior to 13 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 81 Table 8.9 LHGRFACp Multipliers for TSSS Insertion Times BOC to < EOCLB* (Continued)

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 0.86 0.86 0.86 0.86 FHOOS 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86 FHOOS > 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86 TBVOOS > 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.43 0.50 0.43 0.50 23.0 0.40 0.46 0.40 0.46 100.0 1.00 1.00 90.0 1.00 1.00 TBVOOS 50.0 0.86 0.86 FHOOS

> 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43

• LHGRFACp multipliers do not support MELLLA+ operation prior to 13 GWd/MTU.

Note that FHOOS is not allowed in MELLLA+.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 82 Table 8.10 LHGRFACp Multipliers for NSS Insertion Times FFTR/Coastdown*

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 Base case > 65%F 65%F > 65%F 65%F operation 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 TBVOOS > 65%F 65%F > 65%F 65%F 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86

> 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86 TBVOOS > 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43

• 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.

LHGRFACp multipliers do not support MELLLA+ operation prior to 13 GWd/MTU.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 83 Table 8.11 LHGRFACp Multipliers for TSSS Insertion Times FFTR/Coastdown*

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 Base case > 65%F 65%F > 65%F 65%F operation 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 TBVOOS > 65%F 65%F > 65%F 65%F 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86

> 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86 TBVOOS > 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43

• 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.

LHGRFACp multipliers do not support MELLLA+ operation prior to 13 GWd/MTU.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 84 Table 8.12 ATRIUM 10XM and ATRIUM 11 LHGRFACf Multipliers All Cycle 23 Exposures Core Flow

(% of rated) LHGRFACf 0.0 0.58 31.0 0.58 75.0 1.00 107.0 1.00 Table 8.13 AREVA Fuel MAPLHGR Limits Average Planar ATRIUM 10XM ATRIUM 11 Exposure MAPLHGR LTA (GWd/MTU) (kW/ft) MAPLHGR (kW/ft) 0.0 13.1 10.5 15.0 13.1 10.5 67.0 7.7 5.9 AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 85

9.0 REFERENCES

1. ANP-3108P Revision 1, Applicability of AREVA BWR Methods to Brunswick Extended Power Flow Operating Domain, July 2015.

2. NEDO-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 Accession Number ML091800530)

3. ANP-3508P Revision 0, Brunswick Unit 2 Cycle 23 Fuel Cycle Design (MELLLA),

AREVA, July 2016.

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

5. FS1-0026944 Revision 1.0, Brunswick Unit 2 Cycle 23 Calculation Plan, AREVA, July 2016.

6. ANP-2948(P) Revision 2, Mechanical Design Report for Brunswick ATRIUM 10XM Fuel Assemblies, AREVA, January 2017.

7. ANP-3523P Revision 0, ATRIUM 10XM Fuel Rod Thermal-Mechanical Evaluation for Brunswick Unit 2 Cycle 23, AREVA, November 2016.

8. ANP-3363P Revision 0, Mechanical Design Report for Brunswick Unit 2 Cycle 22 ATRIUM 11 Lead Test Assemblies, AREVA, December 2014.

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

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

10. ANP-3337P Revision 0, Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM' 11 Lead Test Assemblies, AREVA, December 2014.

11. ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.

12. ANP-10298PA Revision 1, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, March 2014.

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

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

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

16. OG02-0119-260, Backup Stability Protection (BSP) for Inoperable Option III Solution, GE Nuclear Energy, July 17, 2002.

17. 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.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 86

18. NEDO-33075-A Revision 8, GE Hitachi Nuclear Energy, GE Hitachi Boiling Water Reactor, Detect and Suppress Solution - Confirmation Density, November 2013.

19. NEDO-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)

20. 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.

21. 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.

22. 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.

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

24. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.

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

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

27. ANP-2941(P) Revision 0, Brunswick Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM' 10XM Fuel, AREVA NP, September 2010.

28. ANP-2943(P) Revision 3, Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM' 10XM Fuel, AREVA NP, December 2015.

29. 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.

30. 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.

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

32. ANP-2962(P) Revision 0, Brunswick Nuclear Plant New Fuel Storage Vault Criticality Safety Analysis for ATRIUM' 10XM Fuel, AREVA NP, November 2010.

33. ANP-2955(P) Revision 3, Brunswick Nuclear Plant Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM' 10XM Fuel, AREVA NP, October 2011.

AREVA Inc.

Controlled Document ANP-3560NP Brunswick Unit 2 Cycle 23 Revision 0 Reload Safety Analysis Page 87

34. BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.

35. FS1-0028074 Revision 1.0, 10 CFR 50.46 PCT Reporting for the Brunswick Units ATRIUM 10XM Fuel, AREVA, December 2016.

36. ANP-3105P Revision 1, Brunswick Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel for MELLLA+ Operation, AREVA, July 2015.

37. ANP-3106P Revision 2, Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Fuel for MELLLA+ Operation, AREVA, December 2015.

38. ANP-3360P Revision 0, Fuel Rod Thermal-Mechanical Evaluation for Brunswick Unit 2 Cycle 22, December 2014.

AREVA Inc.

RA-18-0101 Enclosure 6 Affidavit Regarding Withholding ANP-3560P, Brunswick Unit 2 Cycle 23 Reload Safety Analysis, Revision 0

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-3560P Revision 0, "Brunswick Unit 2 Cycle 23 Reload Safety Analysis," dated January 2017 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 Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained,in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

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-3560P Revision 0, "Brunswick Unit 2 Cycle 23 Reload Safety Analysis," dated January 2017 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 Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained,in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA 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 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.

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 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 d '5 0--

day of ..\ "-"--~{' , 2017.

J SUSANK MCCOY NOTARY PUBLIC* WASHINGTON MY COMMISSION EXPIRES 01-14-2020 Susan K. McCoy NOTARY PUBLIC, STATE OF WASHINGTON CS MY COMMISSION EXPIRES: 1/14/2020