enclosure 17 areva report anp-3213(np) monticello fuel ...approved areva parametric crda methodology...

Enclosure 17

AREVA Report ANP-3213(NP)

Monticello Fuel Transition Cycle 28 Reload Licensing Analysis(EPU/MELLLA)

Revision 1

121 pages follow

ANP-3213(NP)Revision 1

MonticelloFuel Transition Cycle 28Reload Licensing Analysis(EPU/MELLLA)

June 2013

AAREVA NP Inc. AR EVA

Uontroned Uocument

AREVA NP Inc.


MonticelloFuel Transition Cycle 28

Reload Licensing Analysis(EPU/MELLLA)

uontroIued Uocument

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Copyright © 2013

AREVA NP Inc.All Rights Reserved

MonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


Page i

Nature of Changes

Item Page Description and Justification

Changes in Revision 1 (as shown below) have been made to sectionswhich affect Neutronics Richland, Thermal-Hydraulics Richland, andMechanics Richland.

(Materials and Thermal-Mechanics Richland sections are unchanged.)

1. p. 2-4 USAR Section 3.6Added "App. A"Added sentence for additional clarity.

2. p. 2-14 USAR Section 14.8Added "GE14" for added clarity.

3. p. 6-2 Section 6.3 Control Rod Drop Accident (CRDA)

.... Approved AREVA parametric CRDA methodology is described inReference 26....

Changed to

.... Approved AREVA parametric CRDA methodology is described inReference 32....

4. p. 9-1 Reference 6 has been updated to reflect correct document name, date, andrevision number.

Changed items are further identifiedby yellow highlighting.

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Contents

1.0 Introduction .................................................................................................................. 1-1

2.0 Disposition of Events .................................................................................................... 2-1

3.0 M echanical Design Analysis ......................................................................................... 3-1

4.0 Therm al-Hydraulic Design Analysis .............................................................................. 4-14.1 Therm al-Hydraulic Design and Com patibility ..................................................... 4-14.2 Safety Lim it M CPR Analysis ............................................................................. 4-14.3 Core Hydrodynam ic Stability ............................................................................. 4-2

5.0 Anticipated O perational O ccurrences ........................................................................... 5-15.1 System Transients ............................................................................................ 5-1

5.1.1 Load Rejection No Bypass (LRNB) ..................................................... 5-25.1.2 Turbine Trip No Bypass (TTNB) .......................................................... 5-35.1.3 Pneumatic System Degradation - Turbine Trip With

Bypass and Degraded Scram (TTW B) ................................................ 5-35.1.4 Feedwater Controller Failure (FW CF) ................................................. 5-45.1.5 Inadvertent HPCI Start-Up (HPCI) ....................................................... 5-45.1.6 Loss of Feedwater Heating ................................................................. 5-55.1.7 Control Rod W ithdrawal Error ............................................................. 5-65.1.8 Fast Flow Runup Analysis ................................................................... 5-6

5.2 Slow Flow Runup Analysis ................................................................................ 5-75.3 Equipm ent O ut-of-Service Scenarios ................................................................ 5-8

5.3.1 Single-Loop O peration ........................................................................ 5-85.3.2 Pressure Regulator Failure Downscale (PRFDS) ................................ 5-9

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

6.0 Postulated Accidents .................................................................................................... 6-16.1 Loss-of-Coolant-Accident (LO CA) ..................................................................... 6-16.2 Pum p Seizure Accident ..................................................................................... 6-16.3 Control Rod Drop Accident (CRDA) .................................................................. 6-26.4 Fuel and Equipm ent Handling Accident ............................................................ 6-36.5 Fuel Loading Error (Infrequent Event) ............................................................... 6-3

6.5.1 M islocated Fuel Bundle ....................................................................... 6-36.5.2 M isoriented Fuel Bundle ..................................................................... 6-3

7.0 Special Analyses .......................................................................................................... 7-17.1 ASM E Overpressurization Analysis ................................................................... 7-17.2 Anticipated Transient W ithout Scram Event Evaluation ..................................... 7-2

7.2.1 O verpressurization Analysis ................................................................ 7-27.2.2 Long-Term Evaluation ......................................................................... 7-3

7.3 Reactor Core Safety Limits - Low Pressure Safety Limit, PressureRegulator Failed O pen Event (PRFO ) ............................................................... 7-4

7.4 Appendix R - Fire Protection Analysis .............................................................. 7-57.5 Standby Liquid Control System ......................................................................... 7-57.6 Fuel Criticality ................................................................................................... 7-6

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8.0 Operating Limits and COLR Input ................................................................................. 8-18 .1 M C P R L im its ..................................................................................................... 8 -18 .2 L H G R L im its ..................................................................................................... 8 -18 .3 M A P LH G R Lim its .............................................................................................. 8-2

9 .0 R e fe re n ce s ................................................................................................................... 9 -1

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

Tables

1.1 EOD and EOOS Operating Conditions ......................................................................... 1-3

2.1 Disposition of Events Summary .................................................................................... 2-32.2 Disposition of Operating Flexibility and EOOS Options on Limiting Events ................. 2-202.3 Methodology and Evaluation Models for Cycle-Specific Reload Analyses .................. 2-21

4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses ....................... 4-34.2 Results Summary for Safety Limit MCPR Analyses ...................................................... 4-44 .3 O P R M S etpo ints ........................................................................................................... 4 -54.4 BSP Endpoints for Monticello Cycle 28 ......................................................................... 4-6

5.1 Exposure Basis for Monticello Cycle 28 Transient Analysis ........................................ 5-105.2 Scram Speed Insertion Times .................................................................................... 5-115.3 Licensing Basis EOFP Base Case LRNB Transient Results ....................................... 5-125.4 Licensing Basis EOFP Base Case TTNB Transient Results ....................................... 5-135.5 Licensing Basis EOFP Base Case TTWB Transient Results ...................................... 5-145.6 Licensing Basis EOFP Base Case FWCF Transient Results ...................................... 5-155.7 Licensing Basis EOFP Base Case HPCI Transient Results ........................................ 5-165.8 Licensing Basis EOFP Base Case CRWE Results ..................................................... 5-175.9 RBM Operability Requirements .................................................................................. 5-185.10 Licensing Basis EOFP PRFDS (PROOS) Transient Results ...................................... 5-19

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

7.1 ASME Overpressurization Analysis Results ................................................................. 7-77.2 ATWS Overpressurization Analysis Results ................................................................. 7-8

8.1 MCPRP Limits for Two-Loop Operation (TLO), NSS Insertion Times BOCto Licensing B asis E O F P .............................................................................................. 8-3

8.2 MCPRP Limits for Two-Loop Operation (TLO), TSSS Insertion Times BOCto Licensing B asis E O FP .............................................................................................. 8-4

8.3 MCPRP Limits for Two-Loop Operation (TLO), NSS Insertion Times BOCto C o a std o w n ............................................................................................................... 8 -5

8.4 MCPRP Limits for Two-Loop Operation (TLO), TSSS Insertion Times BOCto C o a std o w n ............................................................................................................... 8 -6

8.5 MCPRP Limits for Single-Loop Operation (SLO), TSSS Insertion Times BOCto C o a std o w n'. ............................................................................................................. 8 -7

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8.6 Flow-Dependent MCPR Limits ATRIUM 1OXM and GE14 Fuel,NSSITSSS Insertion Times, TLO and SLO, PROOS All Cycle 28E x p o s u re s .................................................................................................................... 8 -8

8.7 ATRIUM 1OXM Steady-State LHG R Lim its ................................................................... 8-98.8 ATRIUM 1OXM LHGRFACp Multipliers for NSS/TSSS Insertion Times,

TLO and SLO , All Cycle 28 Exposures ....................................................................... 8-108.9 GE14 LHGRFACp Multipliers for NSS/TSSS Insertion Times, TLO and

S LO , A ll C ycle 28 Exposures ...................................................................................... 8-118.10 ATRIUM 1OXM LHGRFACf Multipliers, NSS/TSSS Insertion Times, TLO

and SLO , PRO O S, All Cycle 28 Exposures ................................................................ 8-12

8.11 GE14 LHGRFACf Multipliers, NSS/TSSS Insertion Times, TLO and SLO,A ll C ycle 2 8 E xposures ............................................................................................... 8-13

8.12 ATRIUM 1OXM MAPLHGR Lim its, TLO ...................................................................... 8-14

Figures

1.1 Monticello Power/Flow Map - EPU/M ELLLA ................................................................. 1-4

5.1 Licensing Basis EOFP LRNB at 100P/105F -TSSS Key Parameters ........................ 5-215.2 Licensing Basis EOFP LRNB at 1OOP/1 05F - TSSS Vessel Pressures ...................... 5-225.3 Licensing Basis EOFP TTNB at looP/1 05F - TSSS Key Parameters ........................ 5-235.4 Licensing Basis EOFP TTNB at 1 OOP/1 05F - TSSS Vessel Pressures ...................... 5-245.5 Licensing Basis EOFP FWCF at 1 OOP/1 05F - TSSS Key Parameters ....................... 5-255.6 Licensing Basis EOFP FWCF at 1 OOP/1 05F - TSSS Vessel Pressures ..................... 5-265.7 Licensing Basis EOFP HPCI at 1 OOP/1 05F - TSSS Key Parameters ......................... 5-275.8 Licensing Basis EOFP HPCI at 1OOP/1 05F - TSSS Vessel Pressures ....................... 5-28

7.1 MSIV Closure Overpressurization Event at 102P/99F - Key Parameters ..................... 7-97.2 MSIV Closure Overpressurization Event at 102P/99F - Vessel Pressures ................. 7-107.3 MSIV Closure Overpressurization Event at 102P/99F - Safety/Relief

V a lve F lo w R a te s ....................................................................................................... 7 -1 17.4 PRFO ATWS Overpressurization Event at 102P/99F - Key Parameters .................... 7-127.5 PRFO ATWS Overpressurization Event at 102P/99F - Vessel Pressures .................. 7-137.6 PRFO ATWS Overpressurization Event at 102P/99F - Safety/Relief

V a lve F low R ate s ....................................................................................................... 7-14

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Nomenclature

2PT

ADSAOOAPLHGRAROASMEASTATWSATWS-PRFOATWS-RPT

BOCBPWSBSPBWRBWROG

CFRCOLRCPRCRDACRWE

DIVOMDSS

ECCSEFPHEOCEODEOFPEOOSEPU

FWFWCF

two pump trip

automatic depressurization systemanticipated operational occurrenceaverage planar linear heat generation rateall control rods outAmerican Society of Mechanical Engineersalternate source termanticipated transient without scramanticipated transient without scram pressure regulator failure openanticipated transient without scram recirculation pump trip

beginning-of-cyclebanked position withdrawal sequencebackup stability protectionboiling water reactorBoiling Water Reactor Owners Group

Code of Federal Regulationscore operating limits reportcritical power ratiocontrol rod drop accidentcontrol rod withdrawal error

delta-over-initial CPR versus oscillation magnitudedegraded scram speed

emergency core cooling systemeffective full-power hourend-of-cycleextended operating domainend of full powerequipment out-of-serviceextended power uprate

feedwaterfeedwater controller failure

GEGNF

General ElectricGlobal Nuclear Fuels

HCOMHFCLHFRHPCI

hot channel oscillation magnitudehigh flow control lineheat flux ratiohigh pressure coolant injection

ICF increased core flow

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Nomenclature(continued)

LFWHLHGRLHGRFACfLHGRFACPLOCALPRMLRNB

MAPLHGRMCPRMCPRfMCPRPMELLLAMNGPMSIV

NCLNSSNRC

OLMCPROLTP00SOPRM

Pbypass

PCTPRFDSPRFOPROOSPUSAR

RBMRHR

SLCSLCSSLMCPRSLOSLPSSRVSRVOOS

loss of feedwater heatinglinear heat generation rateflow-dependent linear heat generation rate multiplierspower-dependent linear heat generation rate multipliersloss-of-coolant accidentlocal power range monitorgenerator load rejection with no bypass

maximum average planar linear heat generation rateminimum critical power ratioflow-dependent minimum critical power ratiopower-dependent minimum critical power ratiomaximum extended load line limit analysisMonticello Nuclear Generating Plantmain steam isolation valve

nominal control linenominal scram speedNuclear Regulatory Commission, U.S.

operating limit minimum critical power ratiooriginal licensed thermal powerout of serviceoscillation power range monitor

power below which direct scram on TSV/TCV closure is bypassedpeak cladding temperaturepressure regulator failure down-scalepressure regulator failure openpressure regulator out-of-servicePower Uprate Safety Analysis Report

(control) rod block monitorresidual heat removal

standby liquid controlstandby liquid control systemsafety limit minimum critical power ratiosingle-loop operationsingle-loop pump seizuresafety/relief valvesafety/relief valve out-of-service

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Nomenclature(continued)

TBV turbine bypass valvesTCV turbine control valveTIP traversing incore probeTIPOOS traversing incore probe out-of-serviceTLO two-loop operationTSSS technical specifications scram speedTSV turbine stop valveTT turbine tripTTNB turbine trip with no bypassTTWB turbine trip with bypass

USAR Updated Safety Analysis Report

ACPR change in critical power ratio

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1.0 Introduction

The licensing analyses described herein were generated by AREVA NP to support Monticello

Nuclear Generating Plant (MNGP) operation for transitioning to ATRIUM TM 1OXM* fuel starting

in Cycle 28. The analyses were performed using methodologies previously approved for

generic application to boiling water reactors with some exceptions which are explicitly described

in this transition licensing amendment request (LAR). The Nuclear Regulatory Commission

(NRC) technical limitations associated with the application of the approved methodologies have

been satisfied by these analyses.

Licensing analyses support a "representative" core design presented in Reference 1. The

representative core design consists of a total of 484 fuel assemblies, including [ ] fresh

ATRIUM 1OXM assemblies and [ ] irradiated GE14 assemblies. The analyses are prepared

to be the best representation of the proposed MNGP configuration (i.e., extended power uprate

(EPU) at maximum extended load line limit analysis (MELLLA)). However, the Cycle 28 core

design used in this process is only a best-estimate design that is used as a representative

design (because key factors such as Cycle 26 and Cycle 27 fuel depletions can only be

estimated at this time). This process of using a representative core for licensing fuel transitions

has precedent. The precedent recognizes that a representative core design is adequate for the

purposes of the LAR, which are: (1) demonstrate that core design meets the applicability

requirements of the new analysis methods, (2) demonstrate that the results can meet the

proposed safety limits, and (3) demonstrate either existing Technical Specification limits do not

need to be revised for the fuel transition or the needed revisions are identified. The

representative core design for these analyses assures that the actual Cycle 28 core design

meets all these objectives. Ultimately, the reload process will confirm the applicability of all plant

inputs (including plant design changes made in the interim period) for all the appropriate safety

analyses and will also perform the final confirmation that safety limits are satisfied for the actual

core design that will be loaded.

These licensing analyses were performed for potentially limiting events and analyses identified

in Section 2.0. Results of analyses are used to establish the Technical Specifications/COLR

limits and ensure design and licensing criteria are met. Design and safety analyses are based

on both 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

* ATRIUM is a trademark of AREVA NP.

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Page 1-2

also support operation with the equipment out-of-service (EOOS) scenarios presented in

Table 1.1.

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Table 1.1 EOD and EiOSOperating Conditions

Extended Operating Domain(EOD) Conditions

Increased core flow (ICF)

Maximum extended load line limit analysis (MELLLA)

Coastdown

Equipment Out-of-Service(EOOS) Conditions*

Pressure regulator out-of-service (PROOS)

Single-loop operation (SLO)

SLO may be combined with the other EOOS conditions. Base case and each EOOS condition issupported in combination with up to 1 traversing incore probe (TIP) machine out-of-service (TIPOOS)or the equivalent number of TIP channels and/or up to 50% of the LPRMs out-of-service and a1200 effective full-power hour (EFPH) LPRM calibration interval.

AREVA NP Inc.

m

z-U

0

0Core Flow (%)

10 20 30 40 50 60i . . . . . . . . i . . . . i . . . . i . .

70 80 90 100 110 1204 '%N

a<n

N

110-

100

90-

80-

70.

60.

50-

40.

30O

20.

10-

0

i , , i , i , i I i i i I L I I I I

Pov'z R W ._=I I I I I I I iPow1 Flow - l00% EFU =2004 MW t ---------------- -- ------ ---- ---- --- -

A- 51-8% 34.2% 100%CLTP = 1775MWtB: 20.8% 39911 100%OLTP = 1670MWt -A B' n y•r-[ i E i 200x4m,

-- C: 59-r/. 43,3% - 100%oC. reFlow = 57.6O lfr i . - . . .. . . .

D-833 100M9.0%/E- : 100.0%. ,00.0/. --- ---, -- ---I1 -- --. . . . . --.. . . . .- ---E- . . . -I _ _ L . . . . . . .

F: 833% 100.0% / fT1MEIu p, ,, p[m dxyi -i- .d. .-------- - - -0: 37.5% 1000% (22.191 + (0=89714*W)-(00011905*W

2))1.20S

R 20-/. i -./ -whwre:_P=% .OLTan• • W C reFlw ---- ------

l 833% 105.0%-- 3: •75% 111.41/ - -----.-------- - ------- ------ 4. --- -- - I --• -- 4 -------- I -- -I - ----- -- . -......K 100.0% 105.0% i I

-7----------21----2FT- FiA inii----- -----1 -----r - - --- -----T, ---- ', ------------ IT --- I-- --- ---

. .. .I . . . I . . . I . . . . - -t. ..I

I I I I

---"--- -- -- - -__ - .....

, . . . B . . .. . . . --

. . . . . I . . . . . "• . .. . . . . . . . . T . . . . . I - . . . . . . . - - . . . . . . T - - - - - - F - - - - - -

I- - I I I I I - I I IIIi I i i i I i i i I

II I I I L I I I I I

I III. . . .

I-2000

CD o

CLCD

> 0

0)

m

1500

1•00

0

0a-

0'CI-

*500

0

0 5 10 15 20- .. . . . . . . . . . . . . . . . . . . ., . . . ., . . . . I . . . ., . . . . ,

25 30 35 40 45 50 55 60 65

Core Flow (Mlb/hr)

Figure 1.1 Monticello Power/Flow Map -EPU/MELLLA

z

-D C

WA <a

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2.0 Disposition of Events

The objective of this section is to identify limiting events for analysis using AREVA methods,

supporting operation with GE14 and ATRIUM 1OXM fuel. Events and analyses identified as

potentially limiting are either evaluated generically for the introduction of AREVA methods and

fuel or on a cycle-specific basis.

The first step is to identify the licensing basis of the plant. Included in the licensing basis are

descriptions of the postulated events/analyses and the associated criteria. Fuel-related system

design criteria must be met, ensuring regulatory compliance and safe operation. The licensing

basis, related to fuel and applicable for reload analysis, is contained in the Updated Safety

Analysis Report (USAR) (Reference 2), the Technical Specifications (References 3 and 4), Core

Operating Limits Report (COLR), and other reload analysis reports. The licensing basis for EPU

operation is obtained from Reference 5 (and supplements). Reference 6 provides the

applicability of AREVA BWR methods to extended power flow operating domain at Monticello.

AREVA reviewed all fuel-related design criteria, events, and analyses identified in the licensing

basis. When operating limits are established to ensure acceptable consequences of an

anticipated operational occurrence (AOO) or accident, the fuel-related aspects of the system

design criteria are met. All fuel-related events were reviewed and dispositioned into one of the

following categories:

No further analysis required. This classification may result from one of the following:

The consequences of the event have been previously shown to be bounded byconsequences of a different event and the introduction of a new fuel design doesnot change that conclusion.

The consequences of the event are benign, i.e., the event causes no significantchange in margins to the operating limits.

The event is not affected by the introduction of a new fuel design and/or thecurrent analysis of record remains applicable.

Address event each following reload. The consequences of the event are potentiallylimiting and need to be addressed each reload.

Address event for initial licensing analysis. This classification may result from one ofthe following:

The analysis is performed using conservative bounding assumptions and inputssuch that the initial licensing analysis results will remain applicable for followingreloads of the same fuel design (ATRIUM 1OXM).

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Results from the initial licensing analysis will be used to quantitativelydemonstrate that the results remain applicable for following reloads of the samefuel design because the consequences are benign or bounded by those ofanother event.

The impact of operation in the EOOS scenarios presented in Table 1.1 was also considered.

A disposition of events summary is presented in Table 2.1. The disposition summary presents a

list of the events and analyses, the corresponding USAR section, the disposition status, and any

applicable comments.

The disposition for the EOOS scenarios is summarized in Table 2.2. Increased Core Flow (ICF)

and MELLLA operation regions of the power/flow map are included in the disposition results

presented in Table 2.1. Methodology and evaluation models used for the cycle-specific

analyses are provided in Table 2.3.

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Table 2.1 Disposition of Events Summary

USAR Design Disposition

Sect. Criteria Status Comment

3.0 Reactor See below.

3.2 Thermal and Address each time Analyses were performed for the introductionHydraulic changes in hydraulic of ATRIUM 1OXM fuel to demonstrate that thisCharacteristics design occur - fuel design is compatible with the expected

Address for initial coresident fuel (Reference 11 ).licensing analysis. Cycle-specific analyses include SLMCPR,

MCPR, LHGR, and MAPLHGR operating limits(Sections 4.2 and 8.0).

Thermal-hydraulic stability performance isdetermined on a cycle-specific basis(Section 4.3).

3.3 Nuclear Address each reload. Comparison to MAPLHGR, LHGR, and MCPRCharacteristics limits is performed during the cycle-specific

design (Reference 1) and during coremonitoring.

Reactivity coefficients for void, Doppler, andpower are evaluated each reload to ensure thatthey are negative.

Shutdown margin is evaluated on a cycle-specific basis and it is reported in Reference 1.Standby liquid control system shutdowncapability is evaluated on a cycle-specific basis(Section 7.5).

The control rod drop accident (CRDA) analysisis evaluated on a cycle-specific basis(Section 6.3).

The introduction of ATRIUM 1OXM fuel willhave no impact on the propensity for thereactor to undergo xenon instability transients.

3.4 Fuel Mechanical Address for initial The fuel assembly structural analyses areCharacteristics and licensing analysis and performed for the initial reload and remainFuel System for each reload, as applicable for follow-on reloads unlessDesign applicable, changes occur. The fuel assembly analysis,

with the fuel channel, includes an evaluation ofpostulated seismic loads (Reference 7).

The fuel rod thermal-mechanical analyses areperformed on a cycle-specific basis.

3.5 Reactivity Control Address for initial The introduction of ATRIUM 1OXM fuel willMechanical licensing analysis. have no impact on the ability of the control rodsCharacteristics to perform their normal and scram functions

(Reference 7).

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Table 2.1 Disposition of Events Summary (continued)

USAR Design DispositionSect. Criteria Status Comment

3.6 Other reactorApp. A vessel internals

Address for initiallicensing analysis.

Analysis performed for the initial reload todetermine the effect of the mechanical loadsintroduced with ATRIUM 1OXM fuel on otherreactor vessel internals (Reference 38). Theintroduction of the ATRIUM 1OXM fuel intoMonticello will not have any adverse effects onthe reactor pressure vessel seismic analysis ofrecord.

4.0 Reactor CoolantSystem

See below.

4.2 Reactor Vessel

4.3 ReactorRecirculationSystem

4.4 Reactor PressureRelief SystemOverpressuri-zation Protection

4.5 Reactor CoolantSystem Vents

4.6 Hydrogen WaterChemistry

No further analysesrequired.

Address each reload.





The introduction of ATRIUM 1OXM fuel will notimpact the neutron spectrum at the reactorvessel. The vessel fluence is primarilydependent upon the EFPH, power distribution,power level, and fuel management scheme.There are no unique characteristics of theATRIUM 1OXM design that would force asignificant change in the power distribution orcore management scheme.

Analyses performed each reload todemonstrate compliance with the ASMEOverpressurization requirements.Demonstration that the peak steam domepressure remains within allowable limits alsodemonstrates compliance with the recirculationsystem pressure limits (Section 7.1).

This event assures compliance with the ASMEcode (Section 7.1).

Analysis of record shows compliance with thelicensing requirements. The introduction ofATRIUM 1OXM fuel and AREVA methodologydoes not affect the normal operation of thissystem.

The hydrogen water chemistry is independentof the reload fuel. MNGP provides waterchemistry data to AREVA to assess the impactof crud/corrosion on licensing analyses.

The zinc water chemistry is independent of thereload fuel. MNGP provides water chemistrydata to AREVA to assess the impact ofcrud/corrosion on licensing analyses.

4.7 Zinc WaterChemistry

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5.0 Containment See below.System

5.2 Primary No further analyses The primary containment characteristicsContainment required. following a postulated LOCA are independentSystem of fuel design.

5.3 Secondary No further analyses The radiological impact is bounded by the mainContainment required. steam line break accident.System andReactor Building

6.0 Plant See below.EngineeredSafeguards

6.2 ECCS Address for initial Break spectrum analyses performed for thePerformance licensing analysis. initial licensing analysis (Reference 29).

Heatup/MAPLHGR analyses (Reference 30)performed each reload for any new nuclear fueldesign.

6.3 Main Steam Line Address for initial AREVA methodology requires plant-specificFlow Restrictors licensing analysis. evaluation of fuel performance in response to

postulated loss-of-coolant accidents uponintroduction of ATRIUM 1OXM fuel in MNGP.Addressed under the LOCA analysis.

The main steam line break outside the primarycontainment will be considered in theidentification of the spectrum of loss-of-coolantaccident events and is expected to be boundedby the limiting loss-of-coolant accident scenario(Reference 29).

6.4 Control Rod No further analysis The introduction of ATRIUM 1OXM fuel willVelocity Limiters required. have no impact on the ability of the control rods

to perform their normal and scram functions.

6.5 Control Rod No further analysis The introduction of ATRIUM 1OXM fuel willDrive Housing required. have no impact on the ability of the control rodsSupports to perform their normal and scram functions.

6.6 Standby Liquid Address each reload. Standby liquid control system shutdownControl System capability is evaluated on a cycle-specific basis(SLCS) (Section 7.5).

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6.8 Main Control Address for initial As part of the alternative source term (AST)Room, licensing analysis. methodology, the nuclide inventory ofEmergency ATRIUM 1OXM fuel must be evaluated versusFiltration Train the inventories in the AST analysis of record.Building and As shown by radiological source termTechnical evaluations, the ATRIUM 1OXM fuel is notSupport Center significantly different than legacy fuel (GE14).Habitability Further, ATRIUM 1OXM fuel is designed and

operated to comparable standards that wouldensure fuel cladding integrity such that fissionproducts will continue to be contained withinthe cladding.

Therefore, the control room habitability systemdesign basis is unaffected by theATRIUM 1OXM inventories.

7.0 Plant Instru- See below.mentation andControlSystems

7.2 Reactor Control See below.Systems

7.2.1 Reactor Manual Address each reload. Analyses to establish/validate the RBMControl setpoints will be performed each reload. The

CRWE event and RBM setpoint analysis areaddressed below (Section 5.1.7).

7.2.2 Recirculation Address each reload. USAR 14.0 transient analyses verify that theFlow Control fuel related safety design basis of theSystem recirculation flow control system prevent a

transient event sufficient to damage the fuelbarrier or exceed the nuclear system pressurelimits (Sections 5.1.7 and 5.1.8).

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7.3 Nuclear Address each reload. The neutron monitoring system reactor tripInstrumentation setpoints are reviewed and agreed uponSystem between AREVA and Xcel Energy each reload

for the AQOs described in Chapter 14.

AREVA performs cycle-specific OPRM tripsetpoint calculations (Section 4.3).

Analyses to establish/validate the RBMsetpoints are performed each reload. Thesetpoint are determined so that the MCPRPoperating limit based on the CRWE will besimilar to the limit supported by othertransients. The CRWE event and RBMsetpoint analysis are addressed inSection 5.1.7.

7.4 Reactor Vessel No further analyses The safety design basis for the reactor vesselInstrumentation required. instrumentation is independent of the fuel

design.

The reload licensing analyses establish theallowable operating conditions during plannedoperations and abnormal and accidentconditions which can be verified by theoperator using the reactor vesselinstrumentation.

7.5 Plant Radiation No further analysis The introduction of ATRIUM 1OXM fuel willMonitoring required. have no impact on the plant radiationSystems monitoring systems.

7.6 Plant Protection Address each reload. AREVA will perform safety analyses to verifySystem that scrams initiated by the RPS adequately

limit the radiological consequences of grossfailure of the fuel or nuclear system processbarriers (Section 5.0).

7.7 Turbine- Address each reload. AREVA will perform safety analyses whichGenerator include the turbine-generator systemSystem instrumentation and control featuresInstrumentation (Section 5.0).and Control

7.8 Rod Worth Address each reload. AREVA will perform safety analyses toMinimizer evaluate the CRDA to verify that the accidentSystem will not result in fuel pellet deposited enthalpy

greater than the control rod drop accident limitand that the number of failed rods does notexceed the limit (Section 6.3).

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7.9 Other Systems No further analysis All the control and instrumentation featuresControl and required. which may affect the safety analyses wereInstrumentation already discussed above. The remaining

systems are not fuel design dependent and donot need further analysis.

7.10 Seismic and No further analysis The operation of these systems is not affectedTransient required. by the introduction of ATRIUM 1OXM fuel andPerformance AREVA methodology.InstrumentationSystems

7.11 Reactor No further analysis Reactor shutdown capability is not affected byShutdown required. the introduction of ATRIUM 1OXM fuel andCapability AREVA methodology.

7.12 Detailed Control No further analysis Control room design is not affected by theRoom Design required. introduction of ATRIUM 1OXM fuel and AREVAReview methodology.

7.13 Safety Parameter No further analysis Safety parameter display system is notDisplay System required. affected by the introduction of ATRIUM 1OXM

fuel and AREVA methodology.

8.0 Plant Electrical See below.Systems

8.2 Transmission No further analysis Transmission system is not affected by theSystem required. introduction of ATRIUM 1OXM fuel and AREVA

methodology.

8.3 Auxiliary Power No further analysis In case of loss of auxiliary power event theSystem required. reactor scrams and if it is not restored the

diesel generator will carry the vital loads. Seedisposition of Station Blackout event below.

8.4 Plant Standby Address for initial The plant standby diesel generator systemDiesel Generator licensing analysis. features are incorporated into the LOCA breakSystem spectrum analysis which is performed for the

ATRIUM 1OXM fuel with the AREVAmethodology (Reference 29).

8.5 DC Power Address for initial The DC power supply system features areSupply Systems licensing analysis. incorporated into the LOCA break spectrum

analysis which is performed for theATRIUM 1OXM fuel with the AREVAmethodology (Reference 29).

8.6 Reactor No further analysis The power supplies for reactor protectionProtection required. system are not affected by the introduction ofSystem Power ATRIUM 1OXM fuel and AREVA methodology.Supplies

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8.7 Instrumentation No further analysis These systems are not affected by theand Control AC required. introduction of ATRIUM 1OXM fuel and AREVAPower Supply methodology.Systems

8.8 Electrical Design No further analysis Independent of fuel design. Analysis of recordConsiderations required. remains valid.

8.9 Environmental No further analysis Independent of fuel design. Analysis of recordQualification of required. remains valid.Safety-RelatedElectricalEquipment

8.10 Adequacy of No further analysis Independent of fuel design. Analysis of recordStation Electrical required. remains valid.DistributionSystem Voltages

8.11 Power Operated Address each reload. Functionality of safety related valves isValves included in the safety analyses performed for

each cycle (Sections 5.0, 7.1, and 7.2).

8.12 Station Blackout No further analysis Decay heat is the only fuel related input forrequired. station blackout. AREVA dispositioned the

impact of ATRIUM 1OXM fuel by comparing thedecay heat for ATRIUM 1 OXM fuel to the decayheat used in the station blackout analysis ofrecord. Since the ATRIUM 1OXM fuel decayheat is expected to be similar to that of theGE14 fuel the analysis of record results boundthe introduction of ATRIUM 1OXM fuel atMonticello.

9.0 RadioactiveWasteManagement


As shown by radiological source termevaluations, the ATRIUM 1OXM fuel is notsignificantly different than legacy fuel.

ATRIUM 1OXM fuel is designed and operatedto comparable standards that would ensurefuel cladding integrity such that fission productswill continue to be contained within thecladding.

Therefore, plant operations following the fueltransition are not expected to increase the ratethat radiological waste is generated.

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10.0 Plant Auxiliary See below.Systems

10.2 Reactor Auxiliary No further analyses Independent of fuel design (except see below).Systems required (except see Analysis of record remains valid.

below).

10.2.1 Fuel Storage and Address for initial Evaluation of k-eff for normal and abnormalFuel Handling licensing analysis. conditions for spent fuel pool storage racks hasSystems been performed generically for the

ATRIUM 1OXM fuel design (Section 6.4).

10.3 Plant Service No further analyses Independent of fuel design (except see below).Systems required (except see Analysis of record remains valid.

below).

10.3.1 Fire Protection Address for initial The introduction of ATRIUM 1OXM fuel will beSystem licensing analysis. evaluated to demonstrate that no clad damage

occurs for Appendix R (Section 7.4).

10.4 Plant Cooling No further analyses Independent of fuel design (except see below).System required (except see Analysis of record remains valid.

below).

10.4.2 Residual Heat Address for initial AREVA methodology requires plant-specificRemoval System licensing analysis. evaluation of fuel performance in response toService Water postulated LOCA upon introduction of theSystem ATRIUM 1OXM fuel in MNGP (Reference 29).

The decay heat removal design basis of theRHR system is not altered by the introductionof ATRIUM 1OXM fuel in MNGP.

Inadvertent RHR shutdown cooling operation isa benign event which does not needevaluation.

11.0 Plant Power Address each reload. These systems are part of the safety analysisConversion models and their features affect the transientSystems analysis results. These systems are modeled

within the plant transient analyses asappropriate for the introduction ofATRIUM 1OXM fuel at MNGP (Section 5.0).

12.0 Plant Structures No further analyses Independent of fuel design. Analysis of recordand Shielding required. remains valid.

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13.0 Plant Operation Address for initial Organization, Responsibilities, andlicensing analysis. Qualifications of staff personnel are not

affected by transitioning to ATRIUM 1OXM fuel.Training in AREVA methodologies will beprovided for the initial reload. The EmergencyOperational Procedures (EOPs) may needto be updated to include the effects ofATRIUM 1OXM fuel. The overall nuclear siteorganization and plant functional organizationare not affected by the introduction of AREVAfuel.

14.0 Plant Safety See below.Analysis

14.2 MCPR Safety Address each reload. Part of the safety licensing analysis done forLimit each reload with AREVA methodology

(Section 4.2).

14.3 Operating Limits Address each reload. Power- and flow-dependent MCPR and LHGRlimits will be established for each reload usingAREVA methodology. In addition MAPLHGRlimits will be established and verified eachcycle for the ATRIUM 1OXM fuel designs(Section 8.0).

14.4 Transient Events See below.Analyzed forCore Reload

14.4.1 Generator Load Address each reload. This event without bypass operable is aRejection potentially limiting AOO. Load Rejection (LR)Without Bypass with bypass operable is normally bounded by

the LR with no bypass case (Section 5.1.1).

14.4.2 Loss of Address each reload. Application of approved generic analysis wasFeedwater evaluated. Since the generic analysis does notHeating apply, this event will be analyzed for the initial

cycle. Since the results of this event show thisis a potentially limiting event, this event willalso be analyzed each reload (Section 5.1.6).

14.4.3 Rod Withdrawal No further analysis Consequences of a RWE below the low powerError - low required. setpoint are bound by the RWE at power duepower to required strict compliance with BPWS.

14.4.3 Rod Withdrawal Address each reload. Analysis to determine the change in MCPRError - at power and LHGR as a function of RBM setpoint will

be performed for each reload. The analysiswill cover the low, intermediate, and highpower RBM ranges (30% to 100% power)(Section 5.1.7).

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14.4.4 Feedwater Address each reload. This event is a potentially limiting AOO and willController Failure be analyzed each reload (Section 5.1.4).- MaximumDemand

14.4.5 Turbine Trip Address each reload. This event without bypass operable is aWithout Bypass potentially limiting AOO. TT with bypass

operable is bounded by the TT with no bypasscase. TT with bypass operable and degradedscram may be a limiting event for MNGP andhas been analyzed historically for each reload.AREVA will analyze for the initial reload(Section 5.1.2) and will address each reload.

14.5 Special Events See below.

14.5.1 Vessel Pressure Address each reload. This event assures compliance with the ASMEASME Code code. The initial analysis will address MSIV,Compliance TCV, and TSV closures under AREVAModel - MSIV methodology. Since the limiting valve closureClosure is MSIV, only this will be run for future reloads

(Section 7.1).

14.5.2 Standby Liquid Address each reload. Standby liquid control system shutdownControl System capability is evaluated on a cycle-specific basisShutdown Margin (Section 7.5).

14.5.3 Stuck Rod Cold Address each reload. This event is potentially limiting and will beShutdown Margin analyzed each reload (Reference 1).

14.6 Plant Stability Address each reload. Option III will be implemented with theAnalysis transition to AREVA methods. DIVOM and

initial MCPR will be analyzed on acycle-specific basis (Section 4.3).

The Backup Stability Protection (BSP) regionswill be verified on a cycle-specific basis andadjusted if necessary based on the results ofthe analyses (Section 4.3).

14.7 Accident See below.EvaluationMethodology

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14.7.1 Control Rod Address each reload. Safety analyses are performed each reload toDrop Accident evaluate the CRDA to verify that the accidentEvaluation will not result in fuel pellet deposited enthalpy

greater than 280 calories per gram and todetermine the number of rods exceeding the170 calories per gram failure threshold. ForMonticello, the analysis will verify thatdeposited enthalpy remains below 230 cal/gm.

Consequences of the CRDA are evaluated toconfirm that the acceptance criteria aresatisfied (Section 6.3).

14.7.2 Loss-of-Coolant Address for initial LOCA calculations will be performed for EPUAccident licensing analysis. to identify the limiting fluid conditions as a

function of single failure, break size, breaklocation, core flow, and axial power shapeusing the NRC-approved EXEM BWR-2000LOCA methodology. This analysis isperformed for the initial introduction ofATRIUM 1OXM fuel (Reference 29).

MAPLHGR heatup analyses are performedevery time a new neutronic design isintroduced in the core (Reference 30).

14.7.3 Main Steam Line Address for initial The main steam line break will be consideredBreak Accident licensing analysis. in the identification of the spectrum of loss-of-Analysis coolant accident events and is expected to be

bounded by the limiting loss-of-coolantaccident scenario (Reference 29).

14.7.4 Fuel Loading Address each reload. The fuel loading error is analyzed on a cycle-Error Accident specific basis and addresses mislocated or

misoriented fuel assembly (Section 6.5).

14.7.5 One Address each reload. Two-loop pump seizure event is bounded byRecirculation LOCA accident analysis and does not needPump Seizure further analysis.AccidentAnalysis Single-loop pump seizure event has beenhistorically analyzed against the more

restrictive criteria for infrequent events (AOO).Using these criteria, this is the limiting event forsingle-loop operation and it will have to beanalyzed each reload (Section 5.3.1).

14.7.6 Refueling Address for initial The number of fuel rods assumed to fail duringAccident licensing analysis. a fuel handling accident for an ATRIUM 1OXMAnalysis assembly dropping over the core has been

determined and the resulting releasedispositioned against the AST analyses ofrecord (Section 6.4).

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14.7.7 AccidentAtmosphericDispersionCoefficients

14.7.8 Core SourceTerm Inventory

14.8 AnticipatedTransientsWithout Scram(ATWS)

No further analysisrequired.

Address for initiallicensing analysis.


Independent of fuel design. The values ofatmospheric dispersion coefficients in theanalysis of record remain valid.

The source terms for ATRIUM 10XM fuel atEPU conditions have been provided and usedto disposition offsite doses against the ASTanalysis of record. As shown by radiologicalsource term evaluations, the ATRIUM 10XMfuel is not significantly different than legacy fuel(GE14).

The peak vessel pressure is calculated foreach reload. For long-term cooling afterATWS, the decay heat is the only fuel-relatedinput. AREVA dispositioned the impact ofATRIUM 1OXM fuel by comparing the decayheat for ATRIUM 1OXM fuel to the GE14 decayheat used in the ATWS long-term coolinganalysis. Containment heatup wasdispositioned by comparing kineticsparameters for ATRIUM 10XM fuel with thosefor the fuel in the analysis of record(Section 7.2).

14.9 Section deleted NA NA

14.10 Other Analyses

14.10.1 Adequate CoreCooling forTransients witha Single Failure

See below.

No further analysisrequired

USAR 14.10.1 identifies the loss of feedwaterflow event as the worst anticipated transient,and loss of a high pressure inventory makeup(HPCI) or heat removal system as the worstsingle failure.

The analysis of record for loss of feedwaterflow (PUSAR 2.8.5.2.3) already assumed thatthe HPCI system fails to inject. The results ofthis analysis showed that the reactor coreremains covered for the combination of theseworst-case conditions, without operator actionto manually initiate the emergency core coolingsystem or other inventory makeup systems,therefore no further analysis is required.

The events identified in the SupplementalReload Licensing Submittal are addressedbelow as part of the PUSAR (Reference 5).

14A SupplementalReload LicensingSubmittal

See below.

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PUSAR Design DispositionSect. Criteria / Event Status Comment

Decrease inReactor CoolantTemperature

2.8.5.7 Pressure Regulator Address for initial Address for initial reload for EPU/MELLLAFailure - Open licensing analysis. conditions.

Consequences of this event, relative toAOO thermal operating limits, arenonlimiting.

This event results in low steam domepressure and is the most challenging eventfor Technical Specification (TS) 2.1.1.1(Reference 3) low steam dome pressuresafety limit. This section of the TS will beupdated to reduce the 785 psig limit to alower pressure limit. The analysis of thisevent (for initial licensing analysis) willsupport this update to TechnicalSpecifications (Section 7.3).

This event is also used for an ATWSinitiator event.

Decrease in HeatRemoval By theSecondary System/ Increase in ReactorPressure

2.8.5.2.1 Pressure RegulatorFailure - Closed

2.8.5.2.1 MSIV Closures

Address eachreload.

No further analysisrequired.

Consequences of this event, relative to onepressure regulator out-of-service may belimiting; therefore this EOOS event will beevaluated on a cycle-specific basis(Section 5.3.2).

Consequences of this event (with directscram on MSIV closure), relative to thermaloperating limits, are bounded by thegenerator load rejection event. This eventdoes not need further analysis.

Closure of all MSIVs with failure of the valveposition scram function is the design basisoverpressurization event, which isevaluated on a cycle-specific basis(Section 7.1).

The MSIV closure event is a potentiallylimiting ATWS overpressurization event,which is evaluated on a cycle-specific basis(Section 7.2).

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2.8.5.2.1 Loss of Condenser No further Consequences of this event are bounded byVacuum analysis required. either the turbine trip with turbine bypass

valve failure or load rejection with bypassvalve failure.

2.3.5 Loss of AC Power No further This event is analyzed as the Stationanalysis required. Blackout event discussed above under

USAR Section 8.12.

2.8.5.2.3 Loss of Feedwater No further The consequences of this event are onlyFlow analysis required. dependent on the fuel decay heat, since this

event was analyzed as initiated at the lowlevel (L3) scram setpoint in the analysis ofrecord. Since the decay heat ofATRIUM 1OXM fuel is similar to that of GE14fuel the results are expected to be similar tothe current analysis of record.

Decrease in ReactorCoolant SystemFlow Rate

Not Recirculation Pump No further Consequences of this event are benign andevaluated Trip analysis required. bounded by the turbine trip with no bypass

failure event (see dispositions above).

Not Recirculation Flow No further This event is bounded by recirculation pumpevaluated Controller Failure - analysis required. trip events.

Decreasing Flow

2.8.5.3.2 Recirculation Pump No further The consequences of this accident areShaft Break analysis required. bounded by the effects of the recirculation

pump seizure event (see above).

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Reactivity andPower DistributionAnomalies

2.8.5.4.1 Control Rod Mal- No further Consequences of this event are bounded byoperation (system analysis required. the RWE at power.malfunction oroperator error) - lowpower

2.8.5.4.2 Control Rod Mal- Address each Analysis to determine the change in MCPRoperation (system reload, and LHGR as a function of RBM setpoint willmalfunction or be performed for each reload. The analysisoperator error) - at will cover the low, intermediate, and highpower power RBM ranges (30% to 100% power)

(Section 5.1.7).

2.8.5.4.3 Abnormal Startup of No further For all operating modes except refueling,Idle Recirculation analysis required. technical specifications restrictions apply toPump control thermal stresses caused by startup of

an inactive recirculation pump. PUSARidentifies this event as being nonlimiting.The introduction of ATRIUM 1OXM fuel willnot affect this conclusion.

2.8.5.4.3 Recirculation Flow Address each The slow runup event determines the MCPRfControl Failure With reload, limit and LHGRf multiplier and therefore willIncreasing Flow (slow be analyzed each reload (Section 5.2)and fast runup The fast runup event, if not bounded by theevents) slow flow runup event, will be considered in

setting the MCPRP limits (Section 5.1.8).

Increase in ReactorCoolant Inventory

USAR Inadvertent HPCI Address each This is a potentially limiting event which will14A Start-up reload, be evaluated on a cycle-specific basis

(Section 5.1.5).

2.8.5.5 Other BWR transients No further The limiting event for this type of events iswhich increase analysis required. the inadvertent HPCI start-up which will bereactor coolant analyzed each reload.inventory

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Decrease in ReactorCoolant Inventory

2.5.4.1 Inadvertent No further This event results in a mild depressurizationand Safety/Relief Valve analysis required. event which is less severe than the pressure2.8.5.6.1 Opening regulator failure open event (see

Section 7.3). Since the power level settlesout at nearly the initial power level, this eventis considered benign.

2.5.1.1.1 Feedwater Line Break Address for initial The feedwater line break will be considered- Outside licensing analysis in the identification of the spectrum of loss-Containment of-coolant accident events and is expected to

be bounded by the limiting toss-of-coolantaccident scenario (Reference 29).

Radioactive ReleaseFrom Subsystemsand Components

2.9.1 Gaseous Radwaste No further As shown by radiological source termSystem Leak or analysis required. evaluations, the ATRIUM 1OXM fuel is notFailure significantly different than legacy fuel

(GE14). Further, ATRIUM 1OXM fuel isdesigned and operated to comparablestandards that would ensure fuel claddingintegrity such that fission products willcontinue to be contained within the cladding.Therefore, plant operations following the fueltransition are not expected to increase therate that radiological waste is generated.

2.9.2 Liquid Radwaste No further The radionuclide source terms are genericSystem Failure analysis required. and are unaffected by the introduction of

ATRIUM 1OXM fuel.

2.9.2 Postulated No further The radionuclide source terms are genericRadioactive Releases analysis required. and are unaffected by the introduction ofDue to Liquid ATRIUM 1OXM fuel.RadwasteTank Failure

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Other Analyses

2.8.3.3 ATWS with Core No further The discussion presented in Reference 41Instability analysis required. indicates that the "Parameters which might

vary between fuel designs (e.g., reactivitycoefficients) are not expected to significantlychange the consequences of large irregularoscillations." Therefore, the generic ATWSstability results of Reference 41 remainapplicable upon the introduction ofATRIUM 1OXM fuel into MNGP.

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Table 2.2 Disposition of Operating Flexibility andEOOS Options on Limiting Events

Affected Limiting CommentOption Event/Analyses

Single-loop operation LOCA The impact of SLO on LOCA is addressed in(SLO) Reference 29.

SLMCPR The SLO SLMCPR is evaluated on a cycle-specific basis.

Pump Seizure Historically at Monticello the pump seizureaccident during SLO has been evaluated againstthe acceptance criteria for AOO. AREVA willcontinue this practice. Therefore, the MCPRoperating limits for SLO will be modified ifnecessary to assure this accident does not violatethe AOO acceptance critieria.

Safety/relief valves ASME All transient analyses (AOOs) and the ASMEout-of-service all AOO overpressurization event considered operation(SRVOOS) with three SRVs OOS (only the safety function is

credited). Therefore the base case operatinglimits already include this condition.

ATWS Peak ATWS peak pressure analysis considers only onePressure SRVOOS.

Pressure regulator If one of the pressure regulators is OOS theout-of-service backup pressure regulator will operate and(PROOS) therefore not affect the severity of a particular

event.

The pressure regulator down-scale failure eventand the pressure regulator failed open event wereaddressed in Table 2.1.

Traversing in-core probe SLMCPR TIP OOS is included in the SLMCPR analysis.(TIP) out-of-service

ICF/MELLLA All All analyses considered the increased core flowoperation and MELLLA core flow window.

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Table 2.3 Methodology and Evaluation Models forCycle-Specific Reload Analyses

AnalysisEvent Methodology Evaluation Acceptance Criteria

lAnalysis Reference Model and Comment

Thermal and Hydraulic 12 SAFLIM3D SLMCPR Criteria: < 0.1% fuel rodsDesign 24 COTRANSA2 experience boiling transition.

No fuel melting and maximumTransient Analyses 25 XCOBRANofemltnadmxiu

transient induced strain < 1 %.

26 XCOBRA-T Power- and flow-dependent MCPR

9 RODEX4 and LHGR operating limitsestablished to meet the fuel failure28 RODEX2crtia criteria.

Standby Liquid Control 27 CASMO-4 SLCS Criteria: Shutdown margin ofSystem /MICROBURN-B2 at least 0.88% Ak/k.

ASME 24 COTRANSA2 Analyses for ASME and ATWSOverpressurization (as supplemented overpressurization.Analysis by considerationsAnalyssbyoAnsie ASME Overpressurization Criteria:

of ANP-3224(P) Maximum vessel pressure limit ofAnticipated Transient (Reference 6, 1375 psig and maximum domeWithout Scram App. E)) pressure limit of 1332 psig.(pressurization)

A TWS Overpressurization Criteria:Maximum vessel pressure limit of1500 psig.

Emergency Core 34 HUXY LOCA Criteria: 1OCFR50.46.Cooling Systems EXEM BWR-2000 Methodology.

LOCA Analyses Only heatup (HUXY) is analyzed forthe reload specific neutronic design.

Appendix R 34 RELAX 10CFR50 Appendix R.

Neutron Design 18 STAIF Long-Term Stability Solution19 RAMONA5-FA Option Ill Criteria: OPRM setpoints

Neutron Monitoring prevent exceeding OLMCPR limits.System 20 CASMO-4 CRWE Criteria: Power-dependent

21 /MICROBURN-B2 MCPR and LHGR operating limits22 established to meet the fuel failure

criteria.23 Backup Stability Protection

27 Criteria: Stability boundaries that donot exceed acceptable global,regional, and channel decay ratios asdefined by the STAIF methodology.

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3.0 Mechanical Design Analysis

The results of mechanical design analyses for ATRIUM 1OXM fuel are presented in

References 7 and 8. The fuel rod analyses use the NRC-approved RODEX4 methodology

described in Reference 9. The maximum exposure limits for the ATRIUM 1OXM reload fuel are:

54.0 GWd/MTU average assembly exposure62.0 GWd/MTU rod average exposure (full-length fuel rods)

GE14 fuel assemblies have a maximum peak pellet exposure limit of [ ] GWd/MTU

(Reference 10).

The fuel cycle design analyses (Reference 1) verified all fuel assemblies remain within licensed

burnup limits.

The ATRIUM 1OXM LHGR limits are presented in Section 8.0. The GE14 LHGR multipliers

presented in Section 8.0 ensure that the thermal-mechanical design criteria for GE14 fuel are

satisfied.

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4.0 Thermal-Hydraulic Design Analysis

4.1 Thermal-Hydraulic Design and Compatibility

The ATRIUM 1OXM fuel is analyzed and monitored with the ACE critical power correlation

(References 15 through 17). The GE14 fuel is analyzed and monitored with the SPCB critical

power correlation (Reference 13). The SPCB additive constants and additive constant

uncertainty for the GE14 fuel were developed using the indirect approach described in

Reference 14.

Results of thermal-hydraulic characterization and compatibility analyses are presented in

Reference 11. Analysis results demonstrate the thermal-hydraulic design and compatibility

criteria are satisfied for the transition core consisting of ATRIUM 1QXM and GE14 fuel.

4.2 Safety Limit MCPR Analysis

The safety limit MCPR (SLMCPR) is defined as the minimum value of the critical power ratio

ensuring less than 0.1% of the fuel rods are expected to experience boiling transition during

normal operation, or an anticipated operational occurrence (AOO). The SLMCPR for all fuel

was determined using the methodology described in Reference 12. Determination of the

SLMCPR explicitly includes the effects of channel bow. Fuel channels cannot be used for more

than one fuel bundle lifetime.

The analysis was performed with a power distribution conservatively representing expected

reactor operation throughout the cycle. Fuel- and plant-related uncertainties used in the

SLMCPR analysis come from valid references and/or the licensee and are presented in

Table 4.1. The radial power uncertainty used in the analysis includes the effects of up to

1 traversing incore probe (TIP) machine out-of-service (TIPOOS) or the equivalent number of

TIP channels and/or up to 50% of the LPRMs out-of-service and a 1200 effective full-power

hour (EFPH) LPRM calibration interval.

Analyses were performed for the minimum and maximum core flow conditions associated with

rated power for the Monticello power/flow map for EPU/MELLLA operation (statepoints identified

as "K" and "D" in Figure 1.1).

Analysis results support two-loop operation (TLO) SLMCPR of 1.12 and single-loop operation

(SLO) SLMCPR of 1.13. Analysis results including the SLMCPR and the percentage of rods

expected to experience boiling transition are summarized in Table 4.2.

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4.3 Core Hydrodynamic Stability

Monticello has implemented BWROG Long Term Stability Solution Option III (Oscillation Power

Range Monitor-OPRM). Reload validation has been performed in accordance with

Reference 18. 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 28

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.

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 19. This code is a coupled neutronic-

thermal-hydraulic three-dimensional transient model for the purpose of determining the

relationship between the relative change in ACPR and the HCOM on a plant specific basis. The

stability-based OLMCPRs are calculated using the most limiting of the calculated change in

relative ACPR for a given oscillation magnitude or the generic value provided in Reference 18.

The generic value was determined to be limiting for Cycle 28.

In cases where the OPRM system is declared inoperable, Backup Stability Protection (BSP) is

provided in accordance with Reference 22. BSP curves have been evaluated using STAIF

(Reference 23) 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 22.

The STAIF acceptance criteria for the BSP endpoints are global decay ratios < 0.85, and

regional and channel decay ratios < 0.80. Endpoints for the BSP regions provided in Table 4.4

have global decay ratios _< 0.85, and regional and channel decay ratios < 0.80.

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Table 4.1 Fuel- and Plant-RelatedUncertainties for

Safety Limit MCPR Analyses

Parameter Uncertainty

Fuel-RelatedUncertainties

I

IPlant-Related

Uncertainties

Feedwater flow rate 1.8%

Feedwater temperature 0.8%

Core pressure 0.8%

Total core flow rate

TLO 2.5%SLO 6.0%

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Table 4.2 Results Summary forSafety Limit MCPR Analyses

Percentageof Rods in

BoilingSLMCPR Transition

TLO - 1.12 0.0924

SLO - 1.13 0.0812

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Table 4.3 OPRM Setpoints

OPRM OLMCPR OLMCPR

Setpoint (SS) (2PT)

1.05 1.23 1.26

1.06 1.25 1.28

1.07 1.27 1.30

1.08 1.29 1.32

1.09 1.32 1.34

1.10 1.34 1.37

1.11 1.37 1.40

1.12 1.40 1.43

1.13 1.43 1.46

1.14 1.46 1.49

1.15 1.48 1.51

Acceptance Off-Rated Rated PowerCriteria OLMCPR OLMCPR as

at Described in45% Flow Section 8.0

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Table 4.4 BSP Endpoints forMonticello Cycle 28

Power FlowEndpoint (%) (%) Definition

Al 56.6 40.0 Scram region boundary,high flow control line (HFCL)

B1 42.6 33.7 Scram region boundary,nominal control line (NCL)

A2 64.5 50.0 Controlled entry region boundary,HFCL

B2 28.6 31.2 Controlled entry region boundary,NCL

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5.0 Anticipated Operational Occurrences

This section describes the analyses performed to determine the power- and flow-dependent

MCPR operating limits and power- and flow-dependent LHGR multipliers for base case

operation (no equipment out-of-service) for Monticello Cycle 28 representative core.

COTRANSA2 (Reference 24), XCOBRA-T (Reference 25), XCOBRA (Reference 26), and

CASMO-4/MICROBURN-B2 (Reference 27) are the major codes used in the thermal limits

analyses as described in the AREVA THERMEX methodology report (Reference 26) and

neutronics methodology report (Reference 32). 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.

Fuel pellet-to-cladding gap conductance values are based on RODEX2 (Reference 28)

calculations for the Monticello Cycle 28 representative core.

The ACE/ATRIUM 1OXM critical power correlation (References 15 through 17) is used to

evaluate the thermal margin for the ATRIUM 1OXM fuel. The SPCB critical power correlation

(Reference 13) is used in the thermal margin evaluations for the GE14 fuel. The application of

the SPCB correlation to GE14 fuel follows the indirect process described in Reference 14.

5.1 System Transients

The reactor plant parameters for the system transient analyses were validated engineering

inputs as provided by the licensee. Analyses have been performed to determine power- and

flow-dependent MCPR limits and power- and flow-dependent LHGR multipliers that protect

operation throughout the power/flow domain depicted in Figure 1.1.

At Monticello, direct scram on turbine stop valve (TSV) position and turbine control valve (TCV)

fast closure are bypassed at power levels less than 40% of rated (Pbypass). For these powers,

scram will occur when the high pressure or high neutron flux scram setpoint is reached.

Reference 3 indicates that thermal limits only need to be monitored at power levels greater than

or equal to 25% 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 several cycle

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exposures prior to EOFP to ensure that the operating limits provide the necessary protection.

The licensing basis EOFP analysis was performed at EOFP + 400 MWd/MTU (cycle exposure

of 16,175 MWd/MTU). Analyses were performed to support coastdown operation to a cycle

exposure of 21,175 MWd/MTU. The licensing basis exposure range used to develop the

neutronics inputs to the transient analyses are presented in Table 5.1.

Pressurization transient analyses only credit the safety setpoints of the safety/relief valves

(SRV). The base operating limits support operations with 3 SRVs out-of-service.

Variations in feedwater temperature of +5/-1 0°F, from the nominal feedwater temperature and

variation of ±10 psi in dome pressure are considered base case operation, not an EOOS

condition. Analyses were performed to determine the limiting conditions in the allowable

ranges.

System pressurization transient results are sensitive to scram speed assumptions. 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, the Technical Specifications scram speed (TSSS) insertion

times, and degraded scram speed (DSS) 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. Technical Specifications (Reference 3) allow for operation

with up to 8 "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

40% power, the results are relatively insensitive to scram speed, and only TSSS analyses are

performed. At 40% power (Pbypass), analyses were performed, both with and without bypass of

the direct scram function, resulting in an operating limits step change.

5.1.1 Load Rejection No Bypass (LRNB)

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. 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

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excursion of the core power due to the void collapse is terminated primarily by the reactor scram

and revoiding of the core.

LRNB analyses were performed for a range of power/flow conditions to support generation of

the thermal limits. Base case limiting LRNB transient analysis results used to generate the

licensing basis EOFP operating limits, for both TSSS and NSS insertion times, are shown in

Table 5.3. Responses of various reactor and plant parameters during the LRNB event initiated

at 100% of rated power and 105% of rated core flow with TSSS insertion times are shown in

Figure 5.1 and Figure 5.2.

5.1.2 Turbine Trip No Bypass (TTNB)

A turbine trip event can be initiated as a result of several different signals. The initiating signal

causes the TSV to close in order to prevent damage to the turbine. The TSV closure creates a

compression wave traveling through the steam lines into the vessel causing a rapid

pressurization. The increase in pressure results in a decrease in core voids, which in turn

causes a rapid increase in power. Closure of the TSV also causes a reactor scram which helps

mitigate the pressurization effects. 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. Base case

limiting TTNB transient analysis results used to generate the licensing basis EOFP operating

limits, for both TSSS and NSS insertion times, are shown in Table 5.4. Responses of various

reactor and plant parameters during the TTNB event initiated at 100% of rated power and 105%

of rated core flow with TSSS insertion times are shown in Figure 5.3 and Figure 5.4.

5.1.3 Pneumatic System Deqradation - Turbine Trip With Bypass and Degraded Scram(TTWB)

This event is similar to a turbine trip event described previously. The difference is the event is

analyzed with a degraded scram speed (DSS) and the turbine bypass is allowed to open to

mitigate the severity of the event. The MCPRp limits for NSS and TSSS insertion times will

protect this event analyzed with DSS insertion times.

TTWB 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 TTWB transient analysis ACPR

results used to generate the licensing basis EOFP operating limits for both TSSS and NSS

insertion times.

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5.1.4 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 resulting 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.

Valve closure creates a compression wave traveling back to the core, causing void collapse and

subsequent rapid power excursion. The closure of the turbine stop valves also initiates a

reactor scram. The turbine bypass valves are assumed operable and provide some pressure

relief. The core power excursion is mitigated in part by pressure relief, but the primary

mechanisms for termination of the event are reactor scram and revoiding of the core.

FWCF analyses were performed for a range of power/flow conditions to support generation of

the thermal limits. Table 5.6 presents the base case limiting FWCF transient analysis ACPR

results used to generate the licensing basis EOFP operating limits for both TSSS and NSS

insertion times. Figure 5.5 and Figure 5.6 show the responses of various reactor and plant

parameters during the FWCF event initiated at 100% of rated power and 105% of rated core

flow with TSSS insertion times.

5.1.5 Inadvertent HPCI Start-Up (HPCI)

The HPCI flow is injected into the downcomer through the feedwater sparger. Injection of this

subcooled water increases the subcooling at the inlet to the core and results in an increase in

core power. The feedwater control system will attempt to control the water level in the reactor

by reducing the feedwater flow. As long as the mass of steam leaving the reactor through the

steam lines is more than the mass of HPCI water being injected, the water level will be

controlled and a new steady-state condition will be established. In this case the HPCI is fairly

mild as a MCPR transient (similar to a loss of feedwater heating (LFWH) event). If the steam

flow is less than the HPCI flow, the water level will increase until the high level setpoint (L8) is

reached. This type of event is more severe for MCPR calculations (the event is similar to a

feedwater controller failure (FWCF)).

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Historically the HPCI event was analyzed forcing a high level (L8) main turbine trip even in

those cases where the event would develop to a new steady state adding conservatism to the

results. The same approach was used in this analysis forcing the high level turbine trip at all

power levels analyzed. The HPCI flow in Monticello is only injected into one of the two

feedwater lines and thus through the feedwater spargers on only one side of the reactor vessel,

resulting in an asymmetric flow distribution of the injected HPCI flow. The asymmetric injection

of the HPCI flow is assumed to cause an asymmetric core inlet enthalpy distribution with a

larger enthalpy decrease for part of the core. This was accounted for by conservatively

increasing the HPCI flow (decreasing enthalpy on both sides of the core).

HPCI analyses were performed for a range of power/flow conditions to support generation of the

thermal limits. Table 5.7 presents the base case limiting HPCI transient analysis results used to

generate the licensing basis EOFP operating limits for both TSSS and NSS insertion times.

Figure 5.7 and Figure 5.8 show the responses of various reactor and plant parameters during

the HPCI event initiated at 100% of rated power and 105% of rated core flow with TSSS

insertion times.

5.1.6 Loss of Feedwater Heating

The loss of feedwater heating (LFWH) event analysis supports an assumed 95.30 F decrease in

the feedwater temperature. The temperature is assumed to decrease linearly over 31 seconds.

The result is an increase in core inlet subcooling, which reduces voids, thereby increasing core

power and shifting 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.

The limiting full-power ACPRs are 0.17 for ATRIUM 1OXM fuel and 0.18 for GE14 fuel.

Results from LFWH at off-rated conditions are shown in the MCPRP limit and LHGRP multiplier

figures in Appendix A.

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5.1.7 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 CPR. 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 in an

ARTS configuration. 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 5.8 for

the analytical RBM high power setpoint values of 110% to 114%. At the intermediate and low

power setpoints results from the CRWE analysis may set the MCPRP limit. Analysis results

indicate standard filtered RBM setpoint reductions are supported. Analyses demonstrate that

the 1% strain and centerline melt criteria are met for ATRIUM 1OXM fuel. For GE14 fuel see

setdown in Table 8.9. The LHGR limits and their associated multipliers are presented in

Sections 8.2 and 8.3. Recommended operability requirements supporting unblocked CRWE

operation are shown in Table 5.9, based on the SLMCPR values presented in Section 4.2.

5.1.8 Fast Flow Runup Analysis

Several possibilities exist for causing an unplanned increase in core coolant flow resulting from

a recirculation flow control system malfunction. Increasing recirculation flow results in an

increase in core flow which causes an increase in power level and a shift in power towards the

top of the core by reducing the void fraction in that region. If the flow increase is relatively rapid

and of sufficient magnitude, the neutron flux could exceed the scram set point, and a scram

would be initiated.

For BWRs, various failures can occur which can result in a speed increase of both recirculation

pumps or failure of one of the motor generator set speed controllers can result in a speed

increase in one recirculation pump.

The failure of recirculation flow control system, affecting both pumps, is provided with rate limits

and therefore this failure is considered a slow event and is analyzed under the flow-dependent

MCPR limits analysis (MCPRf).

The failure of one of the motor generator speed controllers generally results in the most rapid

rate of recirculation flow increase and this event is referred to as fast flow runup.

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The fast flow runup event was initiated at various power/flow statepoints and cycle exposures.

The most limiting initial conditions are on the left boundary of the power flow map. Results from

fast flow runup analysis are shown in the MCPRP limit and LHGRP multipliers figures in

Appendix A.

5.2 Slow Flow Runup Analysis

Flow-dependent MCPR limits and LHGR multipliers are established to support operation at off-

rated core flow conditions. Limits are based on the CPR and heat flux changes experienced by

the fuel during slow flow excursions. The slow flow excursion event assumes recirculation flow

control system failure such that core flow increases slowly to the maximum flow physically

attainable by the equipment (105% of rated core flow). An uncontrolled increase in flow creates

the potential for a significant increase in core power and heat flux. A conservatively steep flow

runup path was used in the analysis. Analyses were performed to support operation in all the

EOOS scenarios.

MCPRf limits are determined for both ATRIUM 1OXM and GE14 fuel. XCOBRA is used to

calculate the change in critical power ratio during a two-loop flow runup to the maximum flow

rate. The MCPRf limit is set so an increase in core power, resulting from the maximum increase

in core flow, assures the TLO safety limit MCPR is not violated. Calculations were performed

over a range of initial flow rates to determine the corresponding MCPR values causing the

limiting assembly to be at the safety limit MCPR for the high flow condition at the end of the flow

excursion.

MCPRf limits providing the required protection are presented in Table 8.6. MCPRf limits are

applicable for all exposures.

Flow runup analyses were performed with CASMO-4/MICROBURN-B2 to determine flow-

dependent LHGR multipliers (LHGRFACf) for ATRIUM 1 OXM fuel. The analysis assumes

recirculation flow increases slowly along the limiting rod line to the maximum flow physically

attainable by the equipment. A series of flow excursion analyses were performed at several

exposures throughout the cycle, starting from different initial power/flow conditions. Xenon is

assumed to remain constant during the event. LHGRFACf multipliers are established to provide

protection against fuel centerline melt and overstraining of the cladding during a flow runup.

LHGRFACf multipliers for ATRIUM 1OXM fuel are presented in Table 8.10. A process

consistent with the GNF thermal-mechanical methodology was used to determine flow-

dependent LHGR multipliers (LHGRFACf) for GE14 fuel. GE14 LHGRFACf multipliers

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protecting against fuel centerline melt, and clad overstrain during operation at off-rated core flow

conditions, are presented in Table 8.11.

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 equipment out-of-service (EOOS) scenarios supported for Monticello Cycle 28 operation

are shown in Table 1.1. The EOOS scenarios supported are:

* Single-loop operation (SLO) - recirculation loop out-of-service

* Pressure regulator out-of-service (PROOS)

The base case thermal limits support operation with 3 SRVs out-of-service, up to 1 TIPOOS (or

the equivalent number of TIP channels), up to 50% of the LPRMs out-of-service, and a

1200 EFPH LPRM calibration interval.

5.3.1 Single-Loop Operation

AQOs that are limiting during TLO (e.g., HPCI, FWCF, TTNB and become the basis for the

power-dependent MCPR limits and the power-dependent LHGR multipliers, are not more

severe when initiated during SLO. Therefore, the power-dependent LHGR multipliers

established for TLO are applicable during SLO. The power-dependent MCPR operating limits

for SLO are established by adding the limiting power-dependent ACPR for TLO to the SLMCPR

for SLO (see Section 4.2).

LOCA is also more severe when initiated during SLO. Therefore, a reduced MAPLHGR limit is

established for SLO (see Section 6.1).

The pump seizure accident is nonlimiting during TLO but more severe during SLO. Historically

at Monticello the pump seizure accident during SLO has been evaluated against the acceptance

criteria for AOO. AREVA will continue this practice. Therefore, the MCPR operating limits for

SLO will be modified if necessary to assure this accident does not violate the AOO acceptance

criteria (see Section 6.2).

The MCPR, LHGR, and MAPLHGR limits for TLO and SLO are provided in Section 8.0.

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5.3.2 Pressure Re-gulator 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.

The PRFDS ACPR results are presented in Table 5.10. These results are used to create the

operating limits supporting the pressure regulator out-of-service (PROOS) conditions.

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 licensing basis EOFP cycle exposure of 16,175 MWd/MTU (core average

exposure of 33,231.5 MWd/MTU) is given in Table 5.11. Cycle 28 operation is considered to be

in compliance when:

The integrated normalized power generated in the bottom 7 nodes from the projectedEOFP solution at the state conditions provided in Table 5.11 is greater than theintegrated normalized power generated in the bottom 7 nodes in the licensing basis axialpower profile in Table 5.11, and the individual normalized power from the projectedEOFP solution is greater than the corresponding individual normalized power from thelicensing basis axial power profile in Table 5.11 for at least 6 of the 7 bottom nodes.

The projected EOFP condition occurs at a core average exposure less than or equal tolicensing basis EOFP.

If the criteria cannot be fully met the licensing basis may nevertheless remain valid but further

assessment will be required. 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 and nodal burnups.

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Table 5.1 Exposure Basis forMonticello Cycle 28Transient Analysis

CoreCycle Average

Exposure Exposure(MWd/MTU) (MWd/MTU) Comments

0.0 17,057 Beginning of cycle

15,775 32,832 Design basis end of full power(EOFP)

16,175 33,232 Design basis rod patterns toEOFP + 400 MWd/MTU(licensing basis EOFP)

21,175 38,232 Maximum licensing coreexposure - including Coastdown

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Table 5.2 Scram SpeedInsertion Times

TSSS NSS DSSControl Rod Analytical Analytical Analytical

Position Time Time Time(notch) (sec) (sec) (sec)

48 (full-out) 0.000 0.000 0.000

48 0.200 0.200 0.250

46 0.520 0.344 0.365

36 1.160 0.860 1.165

26 1.910 1.395 2.010

6 3.550 2.577 3.729

0 (full-in) 4.006 2.914 4.244

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Table 5.3 Licensing Basis EOFP Base CaseLRNB Transient Results

Power ATRIUM 1OXM GE14

(% rated) ACPR ACPR

TSSS Insertion Times

100 0.36

80 0.39

60 0.39

40 (above Pbypass) 0.38

40 at > 50%F (below Pbypass) 1.25

40 at < 50%F (below Pbypass) 0.95


25 at < 50%F below (Pbypass) 1.22

NSS Insertion Times

0.36

0.37

0.35

0.33

1.15

0.92

1.43

1.20

100

80

60

40

0.29

0.34

0.32

0.30

0.29

0.34

0.31

0.26

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Table 5.4 Licensing Basis EOFP Base CaseTTNB Transient Results


(% rated) ACPR ACPR


100 0.41

80 0.41

60 0.40



40 at 5 50%F (below Pbypass) 0.95



NSS Insertion Times

0.40

0.38

0.36

0.33

1.15

0.92

1.43

1.20

100

80

0.37

0.36

0.32

0.30

0.37

0.36

0.32

0.26

60

40

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Table 5.5 Licensing Basis EOFP Base CaseTTWB Transient Results


(% rated) ACPR ACPR

DSS Insertion Times

100

80

60

40 (above Pbypass)

40 at > 50%F (below Pbypass)

40 at < 50%F (below Pbypass)

25 at > 50%F (below Pbypass)

25 at < 50%F (below Pbypass)

0.38

0.37

0.36

0.32

1.08

0.82

1.08

0.98

0.38

0.36

0.32

0.28

1.03

0.80

1.16

1.02

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Table 5.6 Licensing Basis EOFP Base CaseFWCF Transient Results


(% rated) ACPR ACPR


100 0.43

80 0.45

60 0.49






NSS Insertion Times

0.42

0.45

0.50

0.65

1.55

1.21

2.30

2.06

100

80

60

40

0.39

0.42

0.47

0.57

0.38

0.41

0.47

0.57

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Table 5.7 Licensing Basis EOFP Base CaseHPCI Transient Results


(% rated) ACPR ACPR


100 0.47

80 0.47

60 0.53






NSS Insertion Times

0.46

0.47

0.48

0.53

1.28

1.18

1.67

1.62

100

80

60

40

0.43

0.44

0.46

0.54

0.41

0.43

0.44

0.53

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Table 5.8 Licensing Basis EOFP Base CaseCRWE Results

High Intermediate LowPower Range Power Range Power Range

RBM Trip Core RBM Trip Core RBM Trip CoreSetpoint Power Setpoint Power Setpoint Power

(%) (% rated) MCPR (%) (% rated) MCPR (%) (% rated) MCPR

110 100 1.47 115 85 1.56 120 65 1.7785 1.49 65 1.62 30 2.20

111 100 1.48 116 85 1.58 121 65 1.7985 1.50 65 1.63 30 2.24

112 100 1.50 117 85 1.60 122 65 1.8085 1.52 65 1.65 30 2.24

113 100 1.52 118 85 1.60 123 65 1.8085 1.53 65 1.77 30 2.31

114 100 1.52 119 85 1.60 124 65 1.8085 1.54 65 1.77 30 2.31

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Table 5.9 RBM OperabilityRequirements

Thermal ApplicablePower ATRIUM 1OXM / GE14

(% rated) MCPR

2.46 TLO> 27% and < 90% 2.47 SLO

2.47 SLO

_90% 1.65 TLO

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Table 5.10 Licensing Basis EOFPPRFDS (PROOS)Transient Results


(% rated) ACPR ACPR


100 0.38 0.39

85* 0.41 0.42

851 0.77 0.70

80 0.81 0.74

60 1.00 0.91

40 1.25 1.16

25 1.51 1.43

* Scram on high neutron flux.

t Scram on high dome pressure.

AREVA NP Inc.

2~VLO 7~(~



Table 5.11 Licensing Basis Core AverageAxial Power Profile

State Conditions forPower Shape Evaluation

Power, MWt 2,004.0

Core pressure, psia 1,024.6

Inlet subcooling, Btu/Ibm 22.68

Flow, Mlb/hr 60.48

Control state ARO

Core average exposure 33,231.5(licensing basis EOFP),MWd/MTU

Licensing Axial Power Profile(normalized)

Node PowerTop 24 0.325

23 0.73622 1.19421 1.36820 1.47619 1.50818 1.50217 1.47216 1.40715 1.37214 1.39713 1.37812 1.31711 1.23210 1.1379 1.0348 0.9097 0.7736 0.6505 0.5414 0.4553 0.3962 0.321

Bottom 1 0.099

Sum of Bottom 7 Nodes = 3.235

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I~nn n

400.0 -

300.0 -

"0

C:

a-

Relative Core PowerRelative Heat FluxRelative Core Flow

------- - -R • --- -- --- ----- --- ---

Relative Steam FlowRelative Feed Flow

-- - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

N\i

200.0 -

100.0*

.0-

-100.0

.0 2.0 4.0 6.0 8.0 10.0

Time (seconds)

Figure 5.1 Licensing Basis EOFPLRNB at 10OPI105F - TSSS

Key Parameters

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1300.0

2

:3inU)

4.0 6.0

Time (seconds)

Figure 5.2 Licensing Basis EOFPLRNB at 10OP/105F - TSSS

Vessel Pressures

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600.0

500.0-

400.0 -

Relative Core Power

Relative Heat FluxRelative Core Flow

Relative Steam Flow

Relative Feed Flow

-O0)

(D

1

C,,

0)0@

300.0 -

200.0 -

/100.0,

.0'

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

\ I,, /

--I nnI I

.0 1.0 2.0 3.0 4.01

Time (seconds)5.0 6.0 7.0 8.0

Figure 5.3 Licensing Basis EOFPTTNB at 1OOPI105F - TSSS

Key Parameters

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I 'Al A

1300.0-

1250.0-

2V) 1200.0-

U 1150.0-a3

/ "

,/.

Steam DomeLower Plenum

1100.0-

1050.0-

'AnnnA

.0I1 .0 2.0 3.0 4.0

Time (seconds)5.0 6.0 7.0 8.0

Figure 5.4 Licensing Basis EOFPTTNB at 1OOP/105F - TSSS

Vessel Pressures

AREVA NP Inc.



600 0

500.0

400.0

300.0

200.0-

0

C

a-)

Relative Core PowerRelative Heat FluxRelative Core FlowRelativeSteamFlow----------.-------Relative Steam Flow

Relative Feed Flow

......-100.0

- .0 -

-100.0-

.0 10.0 20.0 30.0Time (seconds)

Figure 5.5 Licensing Basis EOFPFWCF at 1OOP/1 05F - TSSS

Key Parameters

40.0 50.0

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1300.0

1200.0

2En

in

)1)

U)CL

1100.0

1000.0

900.0

20.0 30.0

Time (seconds)

Figure 5.6 Licensing Basis EOFPFWCF at 1OOP/105F - TSSS

Vessel Pressures

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bUU.U .

500.0 -

400.0 -

Relative Core PowerRelative Heat FluxRelative Core FlowRelative Steam FlowRelative Feed Flow

0

a,

300.0 -

200.0 -

I h

100.0-

.0-

_i flnn

----------- - - ----------------------------------------------------- -------- :.----------- --------------

I

.0 10.0 20.0 30.0

Time (seconds)40.0 50.0 60.0

Figure 5.7 Licensing Basis EOFPHPCI at 10OP/105F - TSSS

Key Parameters

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uontroiiedj uocurnenMonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


-9,

U)Q),

Figure 5.8 Licensing Basis EOFPHPCI at 10OP/105F - TSSS

Vessel Pressures

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6.0 Postulated Accidents

6.1 Loss-of-Coolant-Accident (LOCA)

As discussed in Section 2.0 of the LOCA break spectrum report (Reference 29), the LOCA

models, evaluation, and results are for a full core of ATRIUM 1OXM fuel. The basis for

applicability of PCT results from full cores of ATRIUM 1 OXM fuel (based on AREVA methods)

and GE14 fuel (based on GNF methods) for a mixed (transition) core is provided in

Reference 6, Appendix C. Thermal-hydraulic characteristics of the GE14 and ATRIUM 1OXM

fuel designs are similar as presented in Reference 11. Therefore, the core response during a

LOCA will not be significantly different for a full core of GE14 fuel or a mixed core of GE14 and

ATRIUM 1OXM fuel. In addition, since about 95% of the reactor system volume is outside the

core region, slight changes in core volume and fluid energy due to fuel design differences will

produce an insignificant change in total system volume and energy. Therefore, the current

GE14 LOCA analysis and resulting licensing PCT and MAPLHGR limits remain applicable for

GE14 fuel in transition cores.

The results of the ATRIUM 10XM LOCA break spectrum analysis are presented in

Reference 29. The MAPLHGR limits for ATRIUM 1OXM fuel are presented in Reference 30.

The ATRIUM 1OXM PCT is 2088°F. The peak local metal-water reaction and planar average

metal-water reaction were calculated to be 3.50% and 0.73%, respectively. The acceptance

criteria of less than 17% local cladding oxidation thickness and less than 1% core wide metal-

water reaction are met.

Analyses and results support the EOD and EOOS conditions listed in Table 1.1. As discussed

in Section 7.0 of the LOCA break spectrum report (Reference 29), a MAPLHGR multiplier of

0.70 is established for SLO since LOCA is more severe when initiated during SLO.

6.2 Pump Seizure Accident

This accident is assumed to occur as a consequence of an unspecified, instantaneous stoppage

of one recirculation pump shaft while the reactor is operating at full power (in two-loop

operation). The pump seizure event is a very mild accident in relation to other accidents such

as the LOCA. This is easily verified by consideration of the two events. In both accidents, the

recirculation driving loop flow is lost extremely rapidly - in the case of the seizure, stoppage of

the pump occurs; for the LOCA, the severance of the line has a similar, but more rapid and

severe influence. Following a pump seizure event, flow continues, water level is maintained, the

core remains submerged, and this provides a continuous core cooling mechanism.

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However, for the LOCA, complete flow stoppage occurs and the water level decreases due to

loss of coolant resulting in uncovery of the reactor core and subsequent overheating of the fuel

rod cladding. In addition, for the pump seizure accident, reactor pressure does not significantly

decrease, whereas complete depressurization occurs for the LOCA. Clearly, the increased

temperature of the cladding and reduced reactor pressure for the LOCA both combine to yield a

much more severe stress and potential for cladding perforation for the LOCA than for the pump

seizure. Therefore, it can be concluded that the potential effects of the hypothetical pump

seizure accident are very conservatively bounded by the effects of a LOCA and specific

analyses of the pump seizure accident are not required.

Consistent with recent licensing analyses for Monticello, seizure of the recirculation pump in the

active loop during SLO was analyzed and evaluated relative to the acceptance criteria for AOO.

Since single loop pump seizure (SLPS) event is more severe as power and flow increase, the

event is analyzed at the maximum core power and core flow during SLO (66% core power and

52.5% core flow). Thermal limits were determined to protect against this event in single-loop

operation (see Sections 5.3.1 and 8.0).

6.3 Control Rod Drop Accident (CRDA)

Plant startup utilizes a banked position withdrawal sequence (BPWS) including rod worth

minimization strategies. CRDA evaluation was performed for both A and B sequence startups

consistent with that allowed by BPWS. Approved AREVA parametric CRDA methodology is

described in Reference 32, which has been shown to continue to apply to ATRIUM 1OXM and

GEl4 fuel modeled with the CASMO-4/MICROBURN-B2 code system.

Analysis results demonstrate the maximum deposited fuel rod enthalpy is less than the NRC

license limit of 280 cal/g and is also less than 230 cal/g; the estimated number of fuel rods that

exceed the fuel damage threshold of 170 cal/g is less than the number of failed rods assumed in

the USAR (850 8x8 equivalent rods).

Maximum dropped control rod worth, mk 12.14

Core average Doppler coefficient, Ak/k/°F -10.5 x 10-6

Effective delayed neutron fraction 0.00611

Four-bundle local peaking factor 1.475

Maximum deposited fuel rod enthalpy, cal/g 227.7

Maximum number of ATRIUM 1OXM rodsexceeding 170 cal/g 736

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6.4 Fuel and Equipment Handling Accident

As discussed in Reference 40, the fuel handling accident radiological analysis of record for the

alternate source term (AST) was dispositioned with consideration of ATRIUM 10XM core source

terms and number of failed fuel rods. No other aspect of utilizing the ATRIUM 1OXM fuel affects

the current analysis; therefore, the AST analysis remains applicable for fuel transition Cycle 28.

6.5 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. The fuel loading error is characterized as an

infrequent event in the Reference 33 AREVA topical report and in the Monticello USAR

(Reference 2). The acceptance criteria for plants with AST is that the offsite dose

consequences due to the event shall not exceed a small fraction of the 10 CFR 50.67 limits.

6.5.1 Mislocated Fuel Bundle

AREVA has performed a fuel mislocation error analysis that considered the impact of a

mislocated assembly against potential fuel rod failure mechanisms due to increased LHGR and

reduced CPR. The results show that no rod approaches the fuel centerline melt or 1 % strain

limits, and the SLMCPR is not violated (mislocation analysis ACPR result of 0.20 is well below

those reported for AQOs in Section 5.0), i.e. less than 0.1% of the fuel rods are expected to

experience boiling transition. Therefore, no rods would be expected to fail and the offsite dose

criteria (a small fraction of 10 CFR 50.67) is conservatively satisfied. A dose consequence

evaluation is not necessary since no rods are predicted to fail.

6.5.2 Misoriented Fuel Bundle

AREVA has performed a fuel assembly misorientation analysis assuming that the limiting

assembly was loaded in the worst orientation (rotated 1800), while simultaneously producing

sufficient power to be on the MCPR operating limit as if it were oriented correctly. The results

show that no rod approaches the fuel centerline melt or 1% strain limits, and the SLMCPR is not

violated (misorientation analysis ACPR result of 0.27 is well below those reported for AQOs in

Section 5.0), i.e. less than 0.1% of the fuel rods are expected to experience boiling transition.

Therefore, no rods would be expected to fail and the offsite dose criteria (a small fraction of

10 CFR 50.67) is conservatively satisfied. A dose consequence evaluation is not necessary

since no rods are predicted to fail.

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7.0 Special Analyses

7.1 ASME Overpressurization Analysis

This analysis is performed to demonstrate the safety/relief valves have sufficient capacity and

performance to satisfy the requirements established by the ASME Boiler and Pressure Vessel

Code. For Monticello the maximum allowable reactor dome pressure is 1332 psig (1347 psia)

and the maximum allowable vessel pressure is 1375 psig (1390 psia) (Reference 4).

MSIV closure, TSV closure, and TCV closure were performed with the AREVA plant simulator

code COTRANSA2 (Reference 24). The maximum pressure resulting from the closure of

valves in the steam lines tends to increase as the closure time of the valves decreases. The

TCV and TSV close much faster than the MSIV. This suggests that the faster closure of the

TCVs or TSVs would result in higher pressures than closure of the MSIVs. However, the slower

closure of the MSIVs are offset by the fact that the rate of steam flow reduction is concentrated

toward the end of the valve stroke and the resulting reactor pressurization must be absorbed in

a smaller volume (because the MSIVs are closer to the reactor vessel than the TCVs or TSVs).

The analysis of the three valve closures showed that the MSIV valve closure is the most limiting

event. The events were analyzed at 102% power and both 99% and 105% flow at the highest

cycle exposure. 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 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.

Opening of the turbine bypass valves was not credited (this would mitigate the peakpressure resulting from closure of the TSV and the TCV).

* Opening of the SRV at the relief setpoints was not credited (open at safety setpoint).

* Analysis considered 3 SRVOOS.

* TSSS insertion times were used.

0 The initial dome pressure was set at the maximum allowed 1040.0 psia (1025.3 psig).

0 A fast MSIV closure time of 2.2 seconds was used.

0 ATWS-RPT was not credited in this event since this event ends up in a scram(Reference 4).

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Results of the MSIV closure overpressurization event are presented in Table 7.1. Various

reactor plant parameters during the limiting MSIV closure event are presented in

Figure 7.1 through Figure 7.3. The maximum pressure of 1360 psig occurs in the lower vessel.

The maximum steam dome pressure for the same event is 1326 psig. Results demonstrate that

the lower vessel pressure limit of 1375 psig and the steam dome pressure limit of 1332 psig are

protected.

Pressure results include various adders totaling 9 psi to account for void-quality correlations,

Doppler void effects, and thermal conductivity degradation (Reference 6).

7.2 Anticipated Transient Without Scram Event Evaluation

7.2.1 Overpressurization Analysis

This analysis is performed to demonstrate that the peak vessel pressure for the limiting

anticipated transient without scram (ATWS) event is less than the ASME Service Level C limit of

120% of the design pressure (1500 psig). Overpressurization analyses were performed at

102% power at both 99% and 105% flow over the cycle exposure range for both the MSIV

closure event and the pressure regulator failure open (PRFO) events. The PRFO event

assumes a step decrease in pressure demand such that the pressure control system opens the

turbine control and turbine bypass valves. Steam flow demand is assumed to increase to

114.5% of rated steam flow (103% of rated steam flow through the turbine control valves fully

open and 11.5% of rated steam flow through the turbine bypass valves). The system pressure

decreases until the low steam line pressure setpoint is reached resulting in the closure of the

MSIVs. The subsequent pressurization wave collapses core voids, thereby increasing core

power.

The following assumptions were made in the analyses.

0 High-pressure recirculation pump trip (ATWS-RPT) was allowed.

* 1 SRVOOS and the remaining 7 valves open at safety mode setpoints.

0 All scram functions were disabled.

* Nominal values were used for initial dome pressure and feedwater temperature

* A nominal MSIV closure time of 4.0 seconds was used for both events.

Analyses results are presented in Table 7.2. The response of various reactor plant parameters

during the limiting ATWS-PRFO event are shown in Figure 7.4 through Figure 7.6. The

maximum lower vessel pressure is 1445 psig and the maximum steam dome pressure is

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1428 psig. The results demonstrate that the ATWS maximum vessel pressure limit of 1500 psig

is not exceeded.

Pressure results include various adders totaling 20 psi to account for void-quality correlations,

Doppler void effects and thermal conductivity degradation (Reference 6).

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.

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7.3 Reactor Core Safety Limits - Low Pressure Safety Limit, Pressure Regulator

Failed Open Event (PRFO)

Technical Specification for Monticello, Section 2.1.1.1, Reactor Core Safety Limits (SL), requires

that thermal power shall be < 25% rated when the reactor steam dome pressure is < 785 psig

(800 psia) or core flow is < 10% of rated. In Reference 35, General Electric identified that for

plants with the main steam isolation valve (MSIV) low-pressure isolation setpoint < 785 psig,

there is a depressurization transient that will cause this safety limit to be violated. In addition,

plants with an MSIV low-pressure isolation setpoint _Ž 785 psig may also experience an AOO

that violates this safety limit (Monticello MSIV low-pressure setpoint is 809 psig).

The AOO of concern is a depressurization transient, i.e., Pressure Regulator Failure -

Maximum Demand (Open) (PRFO). This event can cause the dome pressure to drop below

785 psig (800 psia) while reactor thermal power is above 25% of rated power.

The PRFO event is initiated through a failure of the pressure controller system open

(instantaneous drop of the pressure demand). This will force the turbine control valves (TCV)

and turbine bypass valve (TBV) to fully open up to the maximum combined steam flow limit.

Opening the turbine valves will create a pressure decrease in the reactor system. At some point

the low-pressure setpoint for main steam isolation valve (MSIV) closure will be reached and the

MSIV will start to close. The initiation of the MSIV closure will trigger the reactor scram on MSIV

position which will reduce further the reactor power. The longest MSIV closure time is

conservative for this event. A closure time of 9.9 seconds was assumed. The system

depressurization also creates a water level swell. If the water level swell reaches the high level

setpoint (L8) the turbine stop valves (TSV) will close.

This event was analyzed to determine the lowest steam dome pressure occurring such that a

future Technical Specification change can be established for the low-pressure value. Since the

core power and heat flux drop throughout this event, followed by a direct scram, this event

poses no threat to thermal limits.

The results of the analyses at various power/flow statepoints and cycle exposures showed that

the lowest steam dome pressure that was reached before thermal power was < 25% thermal

power was 665 psia (650 psig).

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As part of the transition to ATRIUM 1OXM fuel and AREVA methods, AREVA will justify that the

critical power correlations being used for ATRIUM 1OXM fuel and for GE14 fuel are applicable

for pressures above 600 psia.

7.4 Appendix R - Fire Protection Analysis

The Appendix R fire protection case matrix for Monticello safe shutdown is identified in

Reference 36. The most limiting cases were analyzed using the NRC approved AREVA EXEM

BWR-2000 Evaluation Model with a modified Monticello LOCA model. The analyses were

performed for a full core of ATRIUM 1 OXM fuel. The first two fire events were evaluated with

and without a stuck open relief valve, two safety relief valves were used for depressurization

when the reactor water level reached the top of the active fuel in the downcomer, and one

operational core spray train. The final two events were evaluated with the two depressurization

safety relief valves activated at 17 minutes into the fire event instead of the water level being at

the top of the active fuel.

The conclusion of this analysis was that in each event the ATRIUM 1OXM fuel in the core

remains covered during the entire event with no increase in cladding temperature. Results are

therefore independent of fuel type. Containment suppression pool temperatures are not fuel

related and therefore were not considered.

7.5 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 Monticello

SLC system is required to be able to inject 660 ppm natural boron equivalent at 70°F into the

reactor coolant. AREVA has performed an analysis demonstrating the SLC system meets the

required shutdown capability for the cycle. The analysis was performed at a coolant

temperature of 319.20 F, with a boron concentration equivalent to 660 ppm at 68 0 F.* The

temperature of 319.20 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

1.34% Ak/k based on the Cycle 27 EOC short window (which is the most limiting exposure

• Monticello licensing basis documents indicate a minimum of 660 ppm boron at a temperature of 70'F.The AREVA cold analysis basis of 68°F represents a negligible difference and the results areadequate to protect the 70'F licensing basis for the plant.

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Page 7-6

bound by the short and long Cycle 27 exposure window) and based on conservative

assumptions regarding eigenvalue biases and uncertainties in the Cycle 28 transition core.

7.6 Fuel Criticality

The spent fuel pool criticality analysis for ATRIUM 1 OXM fuel is presented in Reference 37 and

submitted to the NRC in Reference 40.

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Page 7-7

Table 7.1 ASME OverpressurizationAnalysis Results*

MaximumPeak Peak Vessel Maximum

Neutron Heat Pressure DomeFlux Flux Lower-Plenum Pressure

Event (% rated) (% rated) (psig) (psig)

MSIV closure(102P/99F) 388 132 1360 1326

Pressurelimit --- --- 1375 1332

* Pressure results include various adders totaling 9 psi to account for void-quality correlations, Doppler

void effects, and thermal conductivity degradation.

AREVA NP Inc.



Page 7-8

Table 7.2 ATWS OverpressurizationAnalysis Results*

MaximumPeak Peak Vessel Maximum

Neutron Heat Pressure DomeFlux Flux Lower-Plenum Pressure

Event (% rated) (% rated) (psig) (psig)

MSIV closure(102P/99F) 308 144 1436 1419

PRFO(102P/99F) 263 151 1445 1428

Pressurelimit --- --- 1500 1500

* Pressure results include various adders totaling 20 psi to account for void-quality correlations,Doppler void effects, and thermal conductivity degradation.

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Page 7-9

-o~1)

00::

0

C0)UL0)

0~

4.0 f

Time (seconds)

Figure 7.1 MSIV Closure Overpressurization Event at102P/99F - Key Parameters

AREVA NP Inc.

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U)()

"1"

a.

2.0 4.0 6.0 8.Time (seconds)

Figure 7.2 MSIV Closure Overpressurization Event at102P/99F - Vessel Pressures*

* The pressure results in this plot do not include the adders due to void-quality correlations, Dopplervoid effects, and thermal conductivity degradation.

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uontrolned uocumentiMonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


600.0

Bank 1Bank 2Bank 3Bank 4Bank 5

500.0-

U)`_"

E.. 400.0-

cI,

Q' 300.0

n 200.0

V)

V.,-(

K100.0

.0

.0 2.0 4.0

Time (seconds)I:I I

8.0 10.0

Figure 7.3 MSIV Closure Overpressurization Event at102P/99F - Safety/Relief Valve Flow Rates*

* In the COTRANSA2 code, the 8 SRVs are grouped in 5 banks. 3 SRVOOS are grouped in bank 1.The remaining 5 operating SRVs are grouped as 1 SRV in banks 2, 3, and 4; and 2 SRVs in bank 5.

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ju.5UU-

Relative Core Power

200.0 -

-o0

CW,pW,a-

Relative Heat FluxRelative Core Flow

Relative Steam FlowRelative Feed Flow

""- -:-.-:..:- - -. .- ------

100.0- -,.-----•-: : -L .........- -

.0 -

Ii I'

- I CIA (Ie 1000 -1*

.0 5.0 10.0 1 5oTime (seconds')

20.0 25.0 30.0

Figure 7.4 PRFO ATWS Overpressurization Event at102P/99F - Key Parameters

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0U)0.

0)L

U)U)U,L

0~

Time (seconds)

Figure 7.5 PRFO ATWS Overpressurization Event at102P/99F - Vessel Pressures*

* The pressure results in this plot do not include the adders due to void-quality correlations, Doppler

void effects, and thermal conductivity degradation.

AREVA NP Inc.



Q1)(I)

E

0

W-

15.0Time (seconds)

Figure 7.6 PRFO ATWS Overpressurization Event at102P/99F - Safety/Relief Valve Flow Rates*

* In the COTRANSA2 code, the 8 SRVs are grouped in 5 banks. 1 SRVOOS is grouped in bank 1. The

remaining 7 operating SRVs are grouped as 3 SRVs in bank 2; 1 SRV in banks 3 and 4; and 2 SRVsin bank 5.

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8.0 Operating Limits and COLR Input

8.1 MCPR Limits

The determination of MCPR limits is based on analyses of the limiting AQOs. The MCPR

operating limits are established so that less than 0.1% of the fuel rods in the core are expected

to experience boiling transition during an AOO initiated from rated or off-rated conditions and

are based on a two-loop operation SLMCPR of 1.12 and a single-loop operation SLMCPR of

1.13. Exposure-dependent MCPR limits were established to support operation from BOC to the

licensing basis EOFP and during Coastdown. MCPR limits are established to support base

case operation and the EOOS scenarios presented in Table 1.1.

Two-loop operation MCPRp limits for ATRIUM 1OXM and GE14 fuel are presented in Table 8.1

through Table 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. Both of these sets (NSS and TSSS) protect the TTWB

with degraded scram speed (DSS) event. MCPRP limits for single-loop operation are provided

in Table 8.5.

MCPRf limits protect against fuel failures during a postulated slow flow excursion.

ATRIUM 1OXM and GE14 fuel limits are presented in Table 8.6 and are applicable for all cycle

exposures and EOOS conditions identified in Table 1.1.

The results from the control rod withdrawal error (CRWE) analysis are not used in establishing

the MCPRP limits. Depending on the choice of RBM setpoints the CRWE analysis operating

MCPR limit may be more limiting than the MCPRp limits. Therefore, Xcel Energy may need to

adjust these limits to account for CRWE results.

8.2 LHGR Limits

The LHGR limits for ATRIUM 1OXM fuel are presented in Table 8.7. The LHGR limits for GE14

fuel are presented in Reference 39. 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.

The LHGRFACp and LHGRFACf multipliers for ATRIUM 1OXM fuel are determined using the

RODEX4 methodology (Reference 9). The LHGRFACP and LHGRFACf multipliers for GE14

fuel are developed in a manner consistent with the GNF thermal-mechanical methodology.

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uontrotoed uocumentMonticello ANP-3213(NP)Fuel Transition Cycle 28 Revision 1Reload Licensing Analysis (EPU/MELLLA) Page 8-2

LHGRFACP multipliers were established to support operation at all cycle exposures for both

NSS and TSSS insertion times and for the EOOS conditions identified in Table 1.1. LHGRFACp

limits are presented in Table 8.8 and Table 8.9 for ATRIUM 1OXM and GE14 fuel, respectively.

LHGRFACf multipliers are established to provide protection against fuel centerline melt and

overstraining of the cladding during a postulated slow flow excursion. LHGRFACf limits are

presented in Table 8.10 and Table 8.11 for ATRIUM 1OXM and GE14 fuel, respectively.

LHGRFACf multipliers are applicable for all cycle exposures and EOOS conditions identified in

Table 1.1.

8.3 MAPLHGR Limits

ATRIUM 1OXM MAPLHGR limits are discussed in Reference 30. The TLO operation limits are

presented in Table 8.12. For operation in SLO, a multiplier of 0.7 must be applied to the TLO

MAPLHGR limits. Power- and flow-dependent MAPLHGR multipliers are not required.

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Page 8-3

Table 8.1 MCPRp Limits forTwo-Loop Operation (TLO), NSS Insertion Times

BOC to Licensing Basis EOFP*

MCPRP

Operating Power ATRIUM 1OXM GE14Condition (% of rated) Fuel Fuel

Base 100.0 1.55 1.53case 40.0 1.71 1.71operation 40.0 at > 50%F 2.77 2.72

25.0 at > 50%F 3.39 3.4740.0 at < 50%F 2.33 2.3825.0 at < 50%F 3.09 3.23

PROOS 100.0 1.59 1.5885.0 1.64 1.6485.0 1.91 1.8440.0 2.39 2.3040.0 at > 50%F 2.77 2.7225.0 at > 50%F 3.39 3.4740.0 at < 50%F 2.48 2.3825.0 at < 50%F 3.24 3.23

* Limits support operation with up to 3 SRVOOS, up to 1 TIPOOS or the equivalent number of TIPchannels and/or up to 50% of the LPRMs out-of-service, and a 1200 EFPH LPRM calibration interval.

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Page 8-4

Table 8.2 MCPRp Limits forTwo-Loop Operation (TLO), TSSS Insertion Times

BOC to Licensing Basis EOFP*

MCPRp

Operating Power ATRIUM 10XM GE14Condition (% of rated) Fuel Fuel


25.0 at > 50%F 3.39 3.4740.0 at < 50%F 2.33 2.3825.0 at < 50%F 3.09 3.23



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Page 8-5

Table 8.3 MCPRp Limits forTwo-Loop Operation (TLO), NSS Insertion Times

BOC to Coastdown*

MCPRP



25.0 at > 50%F 3.39 3.4740.0 at < 50%F 2.33 2.3825.0 at < 50%F 3.09 3.23



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Page 8-6

Two-LoopTable 8.4 MCPRp Limits forOperation (TLO), TSSS Insertion Times

BOC to Coastdown*

MCPRP



25.0 at > 50%F 3.39 3.4740.0 at 5 50%F 2.33 2.3825.0 at < 50%F 3.09 3.23



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Page 8-7

Table 8.5 MCPRP Limits forSingle-Loop Operation (SLO), TSSS Insertion Times

BOC to Coastdown* t

MCPRP


Base 66.0 2.13 2.19case 40.0 2.40 2.31/PROOS 40.0 at > 50%F 2.78 2.73

25.0 at > 50%F 3.40 3.4840.0 at < 50%F 2.49 2.3925.0 at < 50%F 3.25 3.24

* Limits support operation with up to 3 SRVOOS, up to 1 TIPOOS or the equivalent number of TIP

channels and/or up to 50% of the LPRMs out-of-service, and a 1200 EFPH LPRM calibration interval.

t Operation in SLO is not allowed above 66% of rated power.

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Page 8-8

Table 8.6 Flow-Dependent MCPR LimitsATRIUM 1OXM and GE14 Fuel,

NSS/TSSS Insertion Times, TLO and SLO, PROOSAll Cycle 28 Exposures

Core Flow

(% of rated) MCPRf

30.0 1.80

80.0 1.50

105.0 1.50

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Page 8-9

Table 8.7 ATRIUM 1OXM Steady-StateLHGR Limits

PeakPellet Exposure LHGR

(GWd/MTU) (kW/ft)

0.0 14.1

18.9 14.1

74.4 7.4

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Table 8.8 ATRIUM 1OXMLHGRFACp Multipliers for

NSS/TSSS Insertion Times, TLO and SLO,All Cycle 28 Exposures*

LHGRFACp

Operating Power ATRIUM 1OXMCondition (% of rated) Fuel

100.0 1.0040.0 0.80Base 40.0 at > 50%F 0.44

case 25.0 at > 50%F 0.30operation 40.0 at < 50%F 0.56

25.0 at• 50%F 0.36

PROOS 100.0 1.0085.0 0.9585.0 0.9240.0 0.6640.0 at > 50%F 0.4425.0 at > 50%F 0.3040.0 at: <50%F 0.5625.0 at• 50%F 0.36


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Table 8.9 GE14LHGRFACp Multipliers for

NSSITSSS Insertion Times, TLO and SLO,All Cycle 28 Exposures*

LHGRFACP

Operating Power GE14Condition (% of rated) Fuel

Base 100.0 0 .9 9tcase 40.0 0.57operation 40.0 at > 50%F 0.42

25.0 at > 50%F 0.3440.0 at• 50%F 0.5325.0 at < 50%F 0.37

PROOS 100.0 0 .9 9t85.0 0.8985.0 0.7540.0 0.5440.0 at > 50%F 0.4225.0 at > 50%F 0.3440.0 at < 50%F 0.5125.0 at• 50%F 0.37

* Limits support operation with up to 3 SRVOOS, up to 1 TIPOOS or the equivalent number of TIP

channels and/or up to 50% of the LPRMs out-of-service, and a 1200 EFPH LPRM calibration interval.t 0.97 setdown required if analytical setpoint is greater than 114% (based on CRWE calculation).

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Table 8.10 ATRIUM 1OXMLHGRFACf Multipliers,

NSS/TSSS Insertion Times, TLO and SLO, PROOS,All Cycle 28 Exposures

Core Flow ATRIUM 1OXM

(% of rated) LHGRFACf

30.0 0.73

40.0 0.73

75.0 1.00

105.0 1.00

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Table 8.11 GE14LHGRFACf Multipliers,

NSSITSSS Insertion Times, TLO and SLO, PROOS,All Cycle 28 Exposures

Core Flow GE14

(% of rated) LHGRFACf

30.0 0.68

40.0 0.68

75.0 1.00

105.0 1.00

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Table 8.12 ATRIUM 1OXMMAPLHGR Limits, TLO*

Average PlanarExposure MAPLHGR

(GWd/MTU) (kW/ft)

0.0 12.5

20.0 12.5

67.0 7.6

* For operation in SLO, a multiplier of 0.7 must be applied to the TLO MAPLHGR limits.

AREVA NP Inc.

Monticello ANP-3213(NP)Fuel Transition Cycle 28 Revision 1Reload Licensing Analysis (EPU/MELLLA) Page 9-1

9.0 References

1. ANP-3215(P) Revision 0, Monticello Fuel Transition Cycle 28 Fuel Cycle Design (EPU/

MELLLA), AREVA NP, May 2013.

2. Monticello Nuclear Generating Plant, Updated Safety Analysis Report, Revision 28.

3. Technical Specification Requirements for Monticello Nuclear Generating Plant Unit 1,Monticello, Amendment 146.

4. Monticello Nuclear Generating Plant, Technical Specifications (Bases), Revision 16.

5. NEDC-33322(P)* Revision 3, Safety Analysis Report for Monticello Constant PressurePower Uprate, GEH, October 2008.

6. ANP-3224P Revision 2, Applicability of AREVA NP BWR Methods to Monticello,AREVA NP, June 2013.

7. ANP-3119(P) Revision 0, Mechanical Design Report for Monticello A TRIUM TM IOXM

Fuel Assemblies, AREVA NP, October 2012.

8. ANP-3221 P Revision 0, Fuel Rod Thermal-Mechanical Design for MonticelloATRIUM IOXM Fuel Assemblies, Cycle 28, AREVA NP, May 2013.

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

10. GNF Design Basis Document, Fuel-Rod Thermal-Mechanical Performance Limits forGE14C, DB-001 2.03 Revision 2, September 2006 (transmitted by letter, D.J. Mienke(Xcel Energy) to R. Welch (AREVA), "Transmittal of Requested Monticello PlantInformation: GE14 Exposure Limits," July 19, 2012).

11. ANP-3092(P) Revision 0, Monticello Thermal-Hydraulic Design Report forATRIUM TM 1OXM Fuel Assemblies, AREVA NP, July 2012.

12. AN P-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling WaterReactors, AREVA NP, June 2011.

13. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September2009.

14. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical PowerCorrelations to Co-Resident Fuel, Siemens Power Corporation, August 2000.

15. ANP-3138(P) Revision 0, Monticello Improved K-factor Model forACE/ATRIUM IOXMCritical Power Correlation, AREVA NP, August 2012.

16. ANP-10298PA Revision 0, ACE/ATRIUM IOXM Critical Power Correlation, AREVA NP,March 2010.

* This reference should be updated to the NRC-approved revision when possible.

AREVA NP Inc.

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17. ANP-10298(P)(A) Revision 0 Supplement 1P Revision 0, Improved K-factor Model forACE/ATRIUM IOXM Critical Power Correlation, AREVA NP, December 2011.

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

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

20. OG04-0153-260, Plant-Specific Regional Mode DIVOM Procedure Guideline,June 15, 2004.

21. BWROG-03047, Resolution of Reportable Condition for Stability Reload LicensingCalculations Using Generic Regional Mode DIVOM Curve, September 30, 2003.

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

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

24. 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.

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

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

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

28. 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.

29. ANP-3211 (P) Revision 0, Monticello EPU LOCA Break Spectrum Analysis forATRIUM TM IOXM Fuel, AREVA NP, May 2013.

30. ANP-3212(P) Revision 0, Monticello EPU LOCA-ECCS Analysis MAPLHGR Limits forATRIUM TM 1OXM Fuel, AREVA NP, May 2013.

31. 0000-0043-8325-SRLR Revision 1, Cycle 27 Extended Power Uprate SupplementalReload Licensing Report, Global Nuclear Fuel, February 2013.

AREVA NP Inc.

Uontroued uocumentMonticello ANP-3213(NP)Fuel Transition Cycle 28 Revision 1Reload Licensing Analysis (EPU/MELLLA) Page 9-3

32. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodologyfor Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon NuclearCompany, March 1983.

33. XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling WaterReactors: Application of the ENC Methodology to BWR Reloads, Exxon NuclearCompany, June 1986.

34. EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, FramatomeANP, May 2001.

35. General Electric 1OCFR Part 21 Communication, Potential Violation of Low PressureTechnical Specification Safety Limit, SC05-03, March 22, 2005.

36. Letter, D.J. Mienke (Xcel Energy) to Tony Will (AREVA NP), "Transmittal of RequestedMonticello Information - MNGP Appendix R Analysis Information Obtained from GNF,"OC-FAB-ARV-MN-XX-2012-007, February 14, 2012.

37. ANP-3113(P) Revision 0, Monticello Nuclear Plant Spent Fuel Storage Pool CriticalitySafety Analysis for A TRIUMTM IOXM Fuel, AREVA NP, August 2012.

38. 51-9187384-000, "Monticello Plant RPV Seismic Assessment with ATRIUM TM 1OXMFuel," AREVA NP, September 2012 (RJW:12:022).

39. Letter, D.J. Mienke (Xcel Energy) to Tony Will (AREVA NP), "Transmittal of RequestedMonticello Core Follow Data," OC-FAB-ARV-MN-XX-2011-005, November 15, 2011.

40. Letter, M.A. Schimmel (NSPM) to Document Control Desk (NRC), "License AmendmentRequest for Fuel Storage Changes," L-MT-12-076, October 30, 2012 (ADAMSaccession no. ML12307A433).

41. NEDO-32047, A TWS Rule Issues Relative to BWR Core Thermal-Hydraulic Stability,DRF A13-00302, GE Nuclear Energy, February 1992.

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UontrolOed UocumentMonticello ANP-3213(NP)Fuel Transition Cycle 28 Revision 1Reload Licensing Analysis (EPU/MELLLA) Page A-1

Appendix A Operating Limits andResults Comparisons

The figures and tables presented in this appendix show comparisons of the Monticello Cycle 28

operating limits and the transient analysis results. The thermal limits for NSS and TSSS

insertion times protect the TTWB event with DSS insertion times. Comparisons are presented

for the ATRIUM 1OXM and GE14 MCPRP limits and LHGRFACP multipliers.

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ANP-3213(NP)Revision 1Page A-2

MONT CY28 EOFPLBNSS(A10XM

16175.0Fuel)

DSS/NSS/TSSS

4.0

3.5

3.0

-j

Q_

C-)

2.5

I I I I I I I I

* FWCF

o HPCI*• LOFWH

+ LRNB

x RUNUP

* TTNB

v TTWB

V

V

x [

Aax

2.0

1.5

1.0

0 10 20 30 40 50 60 70Power (% Rated)

Power MCPRP

(% of rated) Limit

100.0 1.55

40.0 1.71

40.0 > 50%F 2.77

25.0 > 50%F 3.39

40.0 5 50%F 2.33

25.0 < 50%F 3.09

80 90 100 110

Figure A.1 BOC to Licensing Basis EOFPPower-Dependent MCPR Limits for

ATRIUM 1OXM FuelNSS Insertion Times

Base CaseTwo-Loop Operation (TLO)

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MONT CY28 EOFPLBNSS(GEl4

16175.0Fuel)

DSS/NSS/TSSS

4.0

3.5

3.0

-t

E.

0ja-

o FWCF

o HPCI* LOFWI

+ LRNB

x RUNUF

* TTNB

v TTWB

0

+

+

x ~ ~g -+ *.+

H

2.5

2.0

1.5

1.0

a Ax A

x

0 10 20 30 40 50 60Power (% Rated)

70 80 90 100 110

Power MCPRP

(% of rated) Limit

100.0 1.53

40.0 1.71

40.0 > 50%F 2.72

25.0 > 50%F 3.47

40.05 50%F 2.38

25.0 < 50%F 3.23


GE14 FuelNSS Insertion Times


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MONT CY28 CoostNSS 21175.0(A1OXM Fuel)

DSS/NSS/TSSS

4.0

3.5

3.0

I I

o] FWCF

o HPCIA LOFWH

+ LRNB

x RUNUP

0 TTNB

v TTWB

-tE

_-Q_(D

2.5

2.0

1.5

1.0

+

V

V

0x 0

+A A A

A

+

Ax

0 10 20 30 40 50 60 70Power (% Rated)

80 90 100 110

Power MCPRP

(% of rated) Limit

100.0 1.55

40.0 1.74

40.0 > 50%F 2.77

25.0 > 50%F 3.39

40.0: <50%F 2.33

25.0 5 50%F 3.09

Figure A.3 BOC to CoastdownPower-Dependent MCPR Limits for

ATRIUM 10XM FuelNSS Insertion Times


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uontrolned UocumeniiMonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


MONT CY28 CoastNSS 21175.0(GE14 Fuel)

DSS/NSS/TSSS

4.0

3.5

3.0

-J

o_ 2.5

a-)

2.0

1.5

o FWCF

o HPCI* LOFWH

+ LRNB

x RUNUP

o TTNB

v TTWB

00

+

V

V

x I•I

A A A•x

x

1.0

0 10 20 30 40 50 60

Power (% Rated)70 80 90 100 110

Power MCPRP

(% of rated) Limit

100.0 1.53

40.0 1.71

40.0 > 50%F 2.72

25.0 > 50%F 3.47

40.0 < 50%F 2.38

25.0 5 50%F 3.23


GE14 FuelNSS Insertion Times


AREVA NP Inc.



MONT CY28 EOFPLBTSSS 16175.0(A1OXM Fuel)

DSS/TSSS

4.0

3.5

3.0

-t

C-)

2.5

o FWCF

o HPCI* LOFWH+ LRNB

x RUNUP

0 TTNB

V TTWB

+

x 8

x

IIII I I II

2.0

1.5

1.0

0 10 20 30 40 50 60Power (% Rated)

70 80 90 100 110

Power MCPRP

(% of rated) Limit

100.0 1.59

40.0 1.76

40.0 > 50%F 2.77

25.0 > 50%F 3.39

40.05 50%F 2.33

25.0 5 50%F 3.09


ATRIUM 1OXM FuelTSSS Insertion Times


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Uontrolued Uocument:MonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


MONT CY28 EOFPLBTSSS 16175.0(GE 14 Fuel)

DSS/TSSS

4.0

3.5

3.0

-t

a-)

2.5

o] FWCFo HPCIA LOFWH

+ LRNB

x RUNUPo TTNB

v TTWB

00

+

0

4o-

Axx

2.0

1.5

1.0

0 10 20 30 40 50 60Power (% Roted)

70 80 90 100 110

Power MCPRP

(% of rated) Limit

100.0 1.58

40.0 1.79

40.0 > 50%F 2.72

25.0 > 50%F 3.47

40.05 50%F 2.38

25.05 50%F 3.23


GE14 FuelTSSS Insertion Times


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MONT CY28 CoastTSSS 21175.0(AlOXM Fuel)

DSS/TSSS

4.0

3.5

3.0

E_J

CL 2.5

2.0

1.5

1.0

III I III

o FWCF

o HPCIA LOFWH

+ LRNB

x RUNUP

o TTNB

V TTWB

x

x

0 10 20 30 40 50 60 70Power (% Rated)

80 90 100 110

Power MCPRP

(% of rated) Limit

100.0 1.59

40.0 1.77

40.0 > 50%F 2.77

25.0 > 50%F 3.39

40.05 50%F 2.33

25.0 5 50%F 3.09


ATRIUM 1OXM FuelTSSS Insertion Times


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(Jontroiied uocumenitMonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


MONT CY28 CoastTSSS 211(GE14 Fuel)

75.0 DSS/TSSS

4.0

3.5

3.0

-t

E~

_ja)

2.5

III . IIII

o FWCFo HPCI* LOFWH

+ LRNB

x RUNUP

* TTNB

v TTWB

00

+

V

0 8xx

2.0

1.5

1.0

0 10 20 30 40 50 60

Power (% Rated)70 80 90 100 110

Power MCPRP

(% of rated) Limit

100.0 1.58

40.0 1.79

40.0 > 50%F 2.72

25.0 > 50%F 3.47

40.0! <50%F 2.38

25.0 5 50%F 3.23


GE14 FuelTSSS Insertion Times


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Uontrolled UocumenZMonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


MONT CY28 CoastPROOS 21175.0(A1OXM Fuel)

DSS/TSSS

4.0

3.5

3.0

-t

a_2.5

I I I I I I I

o FWCF

o HPCIA LOFWH

+ LRNB

x PRFDS

0 RUNUPD V TTNB

0 TTWB

+

H +

~+

0

III I I I I I I I

2.0

1.5

1.0

0 10 20 30 40 50 60 70 80 90 100Power (% Rated)

110

Power MCPRP

(% of rated) Limit

100.0 1.59

85.0 1.64

85.0 1.91

40.0 2.39

40.0 > 50%F 2.77

25.0 > 50%F 3.39

40.0 < 50%F 2.48

25.0 5 50%F 3.24


ATRIUM 1OXM FuelNSS/TSSS Insertion Times

PROOSTwo-Loop Operation (TLO)

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ANP-3213(NP)Revision 1Page A-i 1

MONT CY28 CoastPROOS 21175.0(GE14 Fuel)

DSS/TSSS

4.0

3.5

3.0

._J

a-n_£-_

2.5

o FWCF

o HPCIA LOFWH

+ LRNB

x PRFDS

0 RUNUP

v TTNB0 TTWB

00

00 0 g

0

III I I I I I I I

2.0

1.5

1.0

0 10 20 30 40 50

Power (%60Rated)

70 80 90 100 110

Power MCPRP

(% of rated) Limit

100.0 1.58

85.0 1.64

85.0 1.84

40.0 2.30

40.0 > 50%F 2.72

25.0 > 50%F 3.47

40.0: <50%F 2.38

25.0 < 50%F 3.23


GE14 FuelNSSITSSS Insertion Times

PROOSTwo-Loop Operation (TLO)

AREVA NP Inc.

UontroUed UocumentMonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


MONT CY28 CoostSLO 21175.0 DSS/NSS/TSSS(A1OXM Fuel)

4.0

3.5

3.0

-t

E~_j

2.5

2.0

1.5

1.0

III I I I I I I I

0 FWCF

o HPCI* LOFWH

+ LRNB

x PRFDS

* RUNUP0 V TTNB

* TTWB* SLPS

0

H+ X

HX

9 g±

III I I I I I I I

0 10 20 30 40 50 60 70 80 90 100Power (% Rated)

110

Power MCPRP

(% of rated) Limit

66.0 2.13

40.0 2.40

40.0 > 50%F 2.78

25.0 > 50%F 3.40

40.0 5 50%F 2.49

25.0 < 50%F 3.25



Base case + PROOSSingle-Loop Operation (SLO)*

* Operation in SLO is not allowed above 66% of rated power.

AREVA NP Inc.

uontromlod uocumrent~MonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


MONT CY28 CoastSLO 21175.0(GE14 Fuel)

DSS/NSS/TSSS

4.0

3.5

3.0

-t

0E

_j

n-

2.5

II I I I I I I

o FWCF

o HPCI* LOFWH

+ LRNB

x PRFDS

o RUNUPv TTNB

* TTWB0 X SLPS

+ x

0110o 8

0£

I I I Ii i i

2.0

1.5

1.0

0 10 20 30 40 50

Power60

(% Rated)70 80 90 100 110

Power MCPRP

(% of rated) Limit

66.0 2.19

40.0 2.31

40.0 > 50%F 2.73

25.0 > 50%F 3.48

40.0 < 50%F 2.39

25.0 5 50%F 3.24


GE14 FuelNSS/TSSS Insertion Times

Base case + PROOSSingle-Loop Operation (SLO)*

* Operation in SLO is not allowed above 66% of rated power.

AREVA NP Inc.

;ontrotued VocumentMonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


MONT CY28 LHGRFACp Base(AT 1OXM

Case COASTFuel)

ALL SCRAM

0-(-

rY(_-_J

1.2

1.1

1.0

.9

.8

.7

.6

.5

.4

.3

.2

0

0

0

t t I

0

0

A

LOFWH

HPCI

FWCF

I I I I I I I I I

0 10 20 30 40 50 60 70 80 90 100 110

Power (% Rated)

Power LHGRFACP

(% of rated) Multiplier

100.0 1.00

40.0 0.80

40.0 > 50%F 0.44

25.0 > 50%F 0.30

40.0 < 50%F 0.56

25.0 5 50%F 0.36

Figure A.13 All ExposuresPower-Dependent LHGR Multipliers for

ATRIUM 1OXM FuelNSSITSSS Insertion Times

Base CaseTwo-Loop Operation (TLO) andSingle-Loop Operation (SLO)

AREVA NP Inc.

uontroiiecd uocumen"1.MonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


MONT CY28 LHGRFACp Base Case COAST(GE14 Fuel)

ALL SCRAM

1.2

1.1

1.0

.9 0

Q_

C-)Of-

.8

.7

.6

.5

.4

.3

.2

+-I

0 0

0+]

FWCF

HPCI

LOFWHRUNUP

I I I I I I I I I I

0 10 20 30 40 50 60

Power (% Rated)70 80 90 100 110

Power LHGRFACp


100.0 0.99*

40.0 0.57

40.0 > 50%F 0.42

25.0 > 50%F 0.34

40.05 50%F 0.53

25.05 50%F 0.37



Base CaseTwo-Loop Operation (TLO) andSingle-Loop Operation (SLO)

* 0.97 setdown required if analytical setpoint is greater than 114% (based on CRWE calculation).

AREVA NP Inc.

uontronDed uocumentMonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


MONT CY28 LHGRFACp PROOS COAST ALL SCRAM(AT1OXM Fuel)

1.2

1.1

1.0

.9

0~r(_j

I_J

.8

.7

.6

.5

.4

.3

.2

0 10 20 30 40 50 60 70 80 90 100 110

Power (% Rated)

Power LHGRFACp


100.0 1.00

85.0 0.95

85.0 0.92

40.0 0.66

40.0 > 50%F 0.44

25.0 > 50%F 0.30

40.0 <50%F 0.56

25.0 5 50%F 0.36



PROOSTwo-Loop Operation (TLO) andSingle-Loop Operation (SLO)

AREVA NP Inc.

uontroiied uocurnentMonticelloFuel Transition Cycle 28Reload Licensing Analysis (EPU/MELLLA)


MONT CY28 LHGRFACp PROOS(GE 14 Fuel)

COAST ALL SCRAM

1.2

1.1

1.0

.9

II

0-C-)

LL_I,rY

(_j

.8

.7

.6

.5

.4

.3

.2

0

o FWCF

o LOFWH* PRFDS PROOS

I I I I I I I I I I

0 10 20 70 8030 40 50 60

Power (% Roted)90 100 110

Power LHGRFACP(% of rated) Multiplier

100.0 0.99*

85.0 0.89

85.0 0.75

40.0 0.54

40.0 > 50%F 0.42

25.0 > 50%F 0.34

40.0 5 50%F 0.51

25.05 <50%F 0.37



PROOSTwo-Loop Operation (TLO) andSingle-Loop Operation (SLO)

* 0.97 setdown required if analytical setpoint is greater than 114% (based on CRWE calculation).

AREVA NP Inc.

enclosure 17 areva report anp-3213(np) monticello fuel ...approved areva parametric crda methodology...

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