ML13200A195

ANP-3213(NP), Rev. 1, Monticello Fuel Transition Cycle 28 Reload Licensing Analysis (Epu/Mellla).

ML13200A195
Person / Time
Site:	Monticello
Issue date:	06/30/2013
From:	AREVA NP
To:	Office of Nuclear Reactor Regulation
Shared Package
ML13200A185	List: L-MT-13-055, License Amendment Request for Transition to Areva Atrium 10XM Fuel and Areva Safety Analysis Methodology ML13200A189 ML13200A190 ML13200A191 ML13200A192 ML13200A193 ML13200A194 ML13200A195 ML13200A196 ML13200A197 ML13200A199 ML13200A200
References
L-MT-13-055 ANP-3213(NP), Rev 1
Download: ML13200A195 (122)

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

June 2013 A AREVA NP Inc. AR EVA Uontroned Uocument AREVA NP Inc.ANP-3213(NP)

Revision 1 Monticello Fuel Transition Cycle 28 Reload Licensing Analysis (EPU/MELLLA) uontroIued Uocument AREVA NP Inc.ANP-3213(NP)

Revision 1 Copyright

© 2013 AREVA NP Inc.All Rights Reserved Monticello Fuel Transition Cycle 28 Reload Licensing Analysis (EPU/MELLLA)

ANP-3213(NP)

Revision 1 Page i Nature of Changes Item Page Description and Justification Changes in Revision 1 (as shown below) have been made to sections which affect Neutronics Richland, Thermal-Hydraulics Richland, and Mechanics Richland.(Materials and Thermal-Mechanics Richland sections are unchanged.)

1. p. 2-4 USAR Section 3.6 Added "App. A" Added sentence for additional clarity.2. p. 2-14 USAR Section 14.8 Added "GE14" for added clarity.3. p. 6-2 Section 6.3 Control Rod Drop Accident (CRDA).... Approved AREVA parametric CRDA methodology is described in Reference 26....Changed to.... Approved AREVA parametric CRDA methodology is described in Reference 32....4. p. 9-1 Reference 6 has been updated to reflect correct document name, date, and revision number.Changed items are further identified by yellow highlighting.

AREVA NP Inc.

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Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page ii 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 ..............................................................................

4-1 4.1 Therm al-Hydraulic Design and Com patibility

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

4-1 4.2 Safety Lim it M CPR Analysis .............................................................................

4-1 4.3 Core Hydrodynam ic Stability

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

4-2 5.0 Anticipated O perational O ccurrences

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

5-1 5.1 System Transients

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

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

5-2 5.1.2 Turbine Trip No Bypass (TTNB) ..........................................................

5-3 5.1.3 Pneumatic System Degradation

-Turbine Trip With Bypass and Degraded Scram (TTW B) ................................................

5-3 5.1.4 Feedwater Controller Failure (FW CF) .................................................

5-4 5.1.5 Inadvertent HPCI Start-Up (HPCI) .......................................................

5-4 5.1.6 Loss of Feedwater Heating .................................................................

5-5 5.1.7 Control Rod W ithdrawal Error .............................................................

5-6 5.1.8 Fast Flow Runup Analysis ...................................................................

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

5-7 5.3 Equipm ent O ut-of-Service Scenarios

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

5-8 5.3.1 Single-Loop O peration ........................................................................

5-8 5.3.2 Pressure Regulator Failure Downscale (PRFDS) ................................

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

5-9 6.0 Postulated Accidents

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

6-1 6.1 Loss-of-Coolant-Accident (LO CA) .....................................................................

6-1 6.2 Pum p Seizure Accident .....................................................................................

6-1 6.3 Control Rod Drop Accident (CRDA) ..................................................................

6-2 6.4 Fuel and Equipm ent Handling Accident ............................................................

6-3 6.5 Fuel Loading Error (Infrequent Event) ...............................................................

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

6-3 6.5.2 M isoriented Fuel Bundle .....................................................................

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

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

7-1 7.2 Anticipated Transient W ithout Scram Event Evaluation

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

7-2 7.2.1 O verpressurization Analysis ................................................................

7-2 7.2.2 Long-Term Evaluation

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

7-3 7.3 Reactor Core Safety Limits -Low Pressure Safety Limit, Pressure Regulator Failed O pen Event (PRFO ) ...............................................................

7-4 7.4 Appendix R -Fire Protection Analysis ..............................................................

7-5 7.5 Standby Liquid Control System .........................................................................

7-5 7.6 Fuel Criticality

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

7-6 AREVA NP Inc.

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Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page iii 8.0 Operating Limits and COLR Input .................................................................................

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

8 -1 8 .2 L H G R L im its .....................................................................................................

8 -1 8 .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 ....................................................................................

2-3 2.2 Disposition of Operating Flexibility and EOOS Options on Limiting Events .................

2-20 2.3 Methodology and Evaluation Models for Cycle-Specific Reload Analyses ..................

2-21 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses .......................

4-3 4.2 Results Summary for Safety Limit MCPR Analyses ......................................................

4-4 4 .3 O P R M S etpo ints ...........................................................................................................

4 -5 4.4 BSP Endpoints for Monticello Cycle 28 .........................................................................

4-6 5.1 Exposure Basis for Monticello Cycle 28 Transient Analysis ........................................

5-10 5.2 Scram Speed Insertion Times ....................................................................................

5-11 5.3 Licensing Basis EOFP Base Case LRNB Transient Results .......................................

5-12 5.4 Licensing Basis EOFP Base Case TTNB Transient Results .......................................

5-13 5.5 Licensing Basis EOFP Base Case TTWB Transient Results ......................................

5-14 5.6 Licensing Basis EOFP Base Case FWCF Transient Results ......................................

5-15 5.7 Licensing Basis EOFP Base Case HPCI Transient Results ........................................

5-16 5.8 Licensing Basis EOFP Base Case CRWE Results .....................................................

5-17 5.9 RBM Operability Requirements

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

5-18 5.10 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 .................................................................

7-7 7.2 ATWS Overpressurization Analysis Results .................................................................

7-8 8.1 MCPRP Limits for Two-Loop Operation (TLO), NSS Insertion Times BOC to Licensing B asis E O F P ..............................................................................................

8-3 8.2 MCPRP Limits for Two-Loop Operation (TLO), TSSS Insertion Times BOC to Licensing B asis E O FP ..............................................................................................

8-4 8.3 MCPRP Limits for Two-Loop Operation (TLO), NSS Insertion Times BOC to C o a std o w n ...............................................................................................................

8 -5 8.4 MCPRP Limits for Two-Loop Operation (TLO), TSSS Insertion Times BOC to C o a std o w n ...............................................................................................................

8 -6 8.5 MCPRP Limits for Single-Loop Operation (SLO), TSSS Insertion Times BOC to C o a std o w n'. .............................................................................................................

8 -7 AREVA NP Inc.

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Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

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

8 -8 8.7 ATRIUM 1OXM Steady-State LHG R Lim its ...................................................................

8-9 8.8 ATRIUM 1OXM LHGRFACp Multipliers for NSS/TSSS Insertion Times, TLO and SLO , All Cycle 28 Exposures

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

8-10 8.9 GE14 LHGRFACp Multipliers for NSS/TSSS Insertion Times, TLO and S LO , A ll C ycle 28 Exposures

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

8-11 8.10 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

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

5-21 5.2 Licensing Basis EOFP LRNB at 1OOP/1 05F -TSSS Vessel Pressures

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

5-22 5.3 Licensing Basis EOFP TTNB at looP/1 05F -TSSS Key Parameters

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

5-23 5.4 Licensing Basis EOFP TTNB at 1 OOP/1 05F -TSSS Vessel Pressures

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

5-24 5.5 Licensing Basis EOFP FWCF at 1 OOP/1 05F -TSSS Key Parameters

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

5-25 5.6 Licensing Basis EOFP FWCF at 1 OOP/1 05F -TSSS Vessel Pressures

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

5-26 5.7 Licensing Basis EOFP HPCI at 1 OOP/1 05F -TSSS Key Parameters

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

5-27 5.8 Licensing Basis EOFP HPCI at 1OOP/1 05F -TSSS Vessel Pressures

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

5-28 7.1 MSIV Closure Overpressurization Event at 102P/99F -Key Parameters

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

7-9 7.2 MSIV Closure Overpressurization Event at 102P/99F -Vessel Pressures

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

7-10 7.3 MSIV Closure Overpressurization Event at 102P/99F -Safety/Relief V a lve F lo w R a te s .......................................................................................................

7 -1 1 7.4 PRFO ATWS Overpressurization Event at 102P/99F -Key Parameters

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

7-12 7.5 PRFO ATWS Overpressurization Event at 102P/99F -Vessel Pressures

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

7-13 7.6 PRFO ATWS Overpressurization Event at 102P/99F -Safety/Relief V a lve F low R ate s .......................................................................................................

7-14 AREVA NP Inc.

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Revision 1 Page v Nomenclature 2PT ADS AOO APLHGR ARO ASME AST ATWS ATWS-PRFO ATWS-RPT BOC BPWS BSP BWR BWROG CFR COLR CPR CRDA CRWE DIVOM DSS ECCS EFPH EOC EOD EOFP EOOS EPU FW FWCF two pump trip automatic depressurization system anticipated operational occurrence average planar linear heat generation rate all control rods out American Society of Mechanical Engineers alternate source term anticipated transient without scram anticipated transient without scram pressure regulator failure open anticipated transient without scram recirculation pump trip beginning-of-cycle banked position withdrawal sequence backup stability protection boiling water reactor Boiling Water Reactor Owners Group Code of Federal Regulations core operating limits report critical power ratio control rod drop accident control rod withdrawal error delta-over-initial CPR versus oscillation magnitude degraded scram speed emergency core cooling system effective full-power hour end-of-cycle extended operating domain end of full power equipment out-of-service extended power uprate feedwater feedwater controller failure GE GNF General Electric Global Nuclear Fuels HCOM HFCL HFR HPCI hot channel oscillation magnitude high flow control line heat flux ratio high pressure coolant injection ICF increased core flow AREVA NP Inc.

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Revision 1 Page vi Nomenclature (continued)

LFWH LHGR LHGRFACf LHGRFACP LOCA LPRM LRNB MAPLHGR MCPR MCPRf MCPRP MELLLA MNGP MSIV NCL NSS NRC OLMCPR OLTP 00S OPRM Pbypass PCT PRFDS PRFO PROOS PUSAR RBM RHR SLC SLCS SLMCPR SLO SLPS SRV SRVOOS loss of feedwater heating linear heat generation rate flow-dependent linear heat generation rate multipliers power-dependent linear heat generation rate multipliers loss-of-coolant accident local power range monitor generator load rejection with no bypass maximum average planar linear heat generation rate minimum critical power ratio flow-dependent minimum critical power ratio power-dependent minimum critical power ratio maximum extended load line limit analysis Monticello Nuclear Generating Plant main steam isolation valve nominal control line nominal scram speed Nuclear Regulatory Commission, U.S.operating limit minimum critical power ratio original licensed thermal power out of service oscillation power range monitor power below which direct scram on TSV/TCV closure is bypassed peak cladding temperature pressure regulator failure down-scale pressure regulator failure open pressure regulator out-of-service Power Uprate Safety Analysis Report (control) rod block monitor residual heat removal standby liquid control standby liquid control system safety limit minimum critical power ratio single-loop operation single-loop pump seizure safety/relief valve safety/relief valve out-of-service AREVA NP Inc.

uontroneo uocurnent Monticello ANP-3213(NP)

Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page vii Nomenclature (continued)

TBV turbine bypass valves TCV turbine control valve TIP traversing incore probe TIPOOS traversing incore probe out-of-service TLO two-loop operation TSSS technical specifications scram speed TSV turbine stop valve TT turbine trip TTNB turbine trip with no bypass TTWB turbine trip with bypass USAR Updated Safety Analysis Report ACPR change in critical power ratio AREVA NP Inc.

uontroiied uocumenit Monticello ANP-3213(NP)

Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 1-1 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 T M 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.AREVA NP Inc.

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Revision 1 Page 1-2 also support operation with the equipment out-of-service (EOOS) scenarios presented in Table 1.1.AREVA NP Inc.

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Revision 1 Page 1-3 Table 1.1 EOD and EiOS Operating 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 is supported 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 a 1200 effective full-power hour (EFPH) LPRM calibration interval.AREVA NP Inc.

m z-U 0 0 Core Flow (%)10 20 30 40 50 60 i ........i ....i ....i ..70 80 90 100 110 120 4 '%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 i Pow1 Flow -l00% EFU =2004 MW t ----------------

-- ------ ---- ---- --- -A- 51-8% 34.2% 100%CLTP = 1775MWt B: 20.8% 39911 100%OLTP = 1670MWt -A B' n 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=% W C reFlw ---- ------l 833% 105.0%-- 3: 111.41/ ------.--------

---

------4. --- -- -I ---- 4 -------- I -- -I ------ -- .-......K 100.0% 105.0% i I-7----------21----2FT-F iA 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 I IIi 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 CL CD> 0 0)m 1500 0 0 a-0'C I-*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 uontrolled Uocument Monticello ANP-3213(NP)

Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 2-1 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 by consequences of a different event and the introduction of a new fuel design does not change that conclusion.

The consequences of the event are benign, i.e., the event causes no significant change in margins to the operating limits.The event is not affected by the introduction of a new fuel design and/or the current analysis of record remains applicable.

Address event each following reload. The consequences of the event are potentially limiting and need to be addressed each reload.Address event for initial licensing analysis.

This classification may result from one of the following:

The analysis is performed using conservative bounding assumptions and inputs such that the initial licensing analysis results will remain applicable for following reloads of the same fuel design (ATRIUM 1OXM).AREVA NP Inc.

(Jontrolled Uocument Monticello ANP-3213(NP)

Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 2-2 Results from the initial licensing analysis will be used to quantitatively demonstrate that the results remain applicable for following reloads of the same fuel design because the consequences are benign or bounded by those of another 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.AREVA NP Inc.

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Revision 1 Page 2-3 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 introduction Hydraulic changes in hydraulic of ATRIUM 1OXM fuel to demonstrate that this Characteristics 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 is determined on a cycle-specific basis (Section 4.3).3.3 Nuclear Address each reload. Comparison to MAPLHGR, LHGR, and MCPR Characteristics limits is performed during the cycle-specific design (Reference

1) and during core monitoring.

Reactivity coefficients for void, Doppler, and power are evaluated each reload to ensure that they are negative.Shutdown margin is evaluated on a cycle-specific basis and it is reported in Reference 1.Standby liquid control system shutdown capability is evaluated on a cycle-specific basis (Section 7.5).The control rod drop accident (CRDA) analysis is evaluated on a cycle-specific basis (Section 6.3).The introduction of ATRIUM 1OXM fuel will have no impact on the propensity for the reactor to undergo xenon instability transients.

3.4 Fuel Mechanical Address for initial The fuel assembly structural analyses are Characteristics and licensing analysis and performed for the initial reload and remain Fuel System for each reload, as applicable for follow-on reloads unless Design applicable, changes occur. The fuel assembly analysis, with the fuel channel, includes an evaluation of postulated seismic loads (Reference 7).The fuel rod thermal-mechanical analyses are performed on a cycle-specific basis.3.5 Reactivity Control Address for initial The introduction of ATRIUM 1OXM fuel will Mechanical licensing analysis.

have no impact on the ability of the control rods Characteristics to perform their normal and scram functions (Reference 7).AREVA NP Inc.

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Revision 1 Page 2-4 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 3.6 Other reactor App. A vessel internals Address for initial licensing analysis.Analysis performed for the initial reload to determine the effect of the mechanical loads introduced with ATRIUM 1OXM fuel on other reactor vessel internals (Reference 38). The introduction of the ATRIUM 1OXM fuel into Monticello will not have any adverse effects on the reactor pressure vessel seismic analysis of record.4.0 Reactor Coolant System See below.4.2 Reactor Vessel 4.3 Reactor Recirculation System 4.4 Reactor Pressure Relief System Overpressuri-zation Protection 4.5 Reactor Coolant System Vents 4.6 Hydrogen Water Chemistry No further analyses required.Address each reload.Address each reload.No further analyses required.No further analyses required.No further analyses required.The introduction of ATRIUM 1OXM fuel will not impact the neutron spectrum at the reactor vessel. The vessel fluence is primarily dependent upon the EFPH, power distribution, power level, and fuel management scheme.There are no unique characteristics of the ATRIUM 1OXM design that would force a significant change in the power distribution or core management scheme.Analyses performed each reload to demonstrate compliance with the ASME Overpressurization requirements.

Demonstration that the peak steam dome pressure remains within allowable limits also demonstrates compliance with the recirculation system pressure limits (Section 7.1).This event assures compliance with the ASME code (Section 7.1).Analysis of record shows compliance with the licensing requirements.

The introduction of ATRIUM 1OXM fuel and AREVA methodology does not affect the normal operation of this system.The hydrogen water chemistry is independent of the reload fuel. MNGP provides water chemistry data to AREVA to assess the impact of crud/corrosion on licensing analyses.The zinc water chemistry is independent of the reload fuel. MNGP provides water chemistry data to AREVA to assess the impact of crud/corrosion on licensing analyses.4.7 Zinc Water Chemistry AREVA NP Inc.

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Revision 1 Page 2-5 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 5.0 Containment See below.System 5.2 Primary No further analyses The primary containment characteristics Containment required.

following a postulated LOCA are independent System of fuel design.5.3 Secondary No further analyses The radiological impact is bounded by the main Containment required.

steam line break accident.System and Reactor Building 6.0 Plant See below.Engineered Safeguards 6.2 ECCS Address for initial Break spectrum analyses performed for the Performance licensing analysis.

initial licensing analysis (Reference 29).Heatup/MAPLHGR analyses (Reference 30)performed each reload for any new nuclear fuel design.6.3 Main Steam Line Address for initial AREVA methodology requires plant-specific Flow Restrictors licensing analysis.

evaluation of fuel performance in response to postulated loss-of-coolant accidents upon introduction of ATRIUM 1OXM fuel in MNGP.Addressed under the LOCA analysis.The main steam line break outside the primary containment will be considered in the identification of the spectrum of loss-of-coolant accident events and is expected to be bounded by the limiting loss-of-coolant accident scenario (Reference 29).6.4 Control Rod No further analysis The introduction of ATRIUM 1OXM fuel will Velocity 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 will Drive Housing required.

have no impact on the ability of the control rods Supports to perform their normal and scram functions.

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

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Revision 1 Page 2-6 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 6.8 Main Control Address for initial As part of the alternative source term (AST)Room, licensing analysis.

methodology, the nuclide inventory of Emergency ATRIUM 1OXM fuel must be evaluated versus Filtration Train the inventories in the AST analysis of record.Building and As shown by radiological source term Technical evaluations, the ATRIUM 1OXM fuel is not Support Center significantly different than legacy fuel (GE14).Habitability Further, ATRIUM 1OXM fuel is designed and operated to comparable standards that would ensure fuel cladding integrity such that fission products will continue to be contained within the cladding.Therefore, the control room habitability system design basis is unaffected by the ATRIUM 1OXM inventories.

7.0 Plant Instru- See below.mentation and Control Systems 7.2 Reactor Control See below.Systems 7.2.1 Reactor Manual Address each reload. Analyses to establish/validate the RBM Control setpoints will be performed each reload. The CRWE event and RBM setpoint analysis are addressed below (Section 5.1.7).7.2.2 Recirculation Address each reload. USAR 14.0 transient analyses verify that the Flow Control fuel related safety design basis of the System recirculation flow control system prevent a transient event sufficient to damage the fuel barrier or exceed the nuclear system pressure limits (Sections 5.1.7 and 5.1.8).AREVA NP Inc.

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Revision 1 Page 2-7 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 7.3 Nuclear Address each reload. The neutron monitoring system reactor trip Instrumentation setpoints are reviewed and agreed upon System between AREVA and Xcel Energy each reload for the AQOs described in Chapter 14.AREVA performs cycle-specific OPRM trip setpoint calculations (Section 4.3).Analyses to establish/validate the RBM setpoints are performed each reload. The setpoint are determined so that the MCPRP operating limit based on the CRWE will be similar to the limit supported by other transients.

The CRWE event and RBM setpoint analysis are addressed in Section 5.1.7.7.4 Reactor Vessel No further analyses The safety design basis for the reactor vessel Instrumentation required.

instrumentation is independent of the fuel design.The reload licensing analyses establish the allowable operating conditions during planned operations and abnormal and accident conditions which can be verified by the operator using the reactor vessel instrumentation.

7.5 Plant Radiation No further analysis The introduction of ATRIUM 1OXM fuel will Monitoring required.

have no impact on the plant radiation Systems monitoring systems.7.6 Plant Protection Address each reload. AREVA will perform safety analyses to verify System that scrams initiated by the RPS adequately limit the radiological consequences of gross failure of the fuel or nuclear system process barriers (Section 5.0).7.7 Turbine- Address each reload. AREVA will perform safety analyses which Generator include the turbine-generator system System instrumentation and control features Instrumentation (Section 5.0).and Control 7.8 Rod Worth Address each reload. AREVA will perform safety analyses to Minimizer evaluate the CRDA to verify that the accident System will not result in fuel pellet deposited enthalpy greater than the control rod drop accident limit and that the number of failed rods does not exceed the limit (Section 6.3).AREVA NP Inc.

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Revision 1 Page 2-8 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 7.9 Other Systems No further analysis All the control and instrumentation features Control and required.

which may affect the safety analyses were Instrumentation already discussed above. The remaining systems are not fuel design dependent and do not need further analysis.7.10 Seismic and No further analysis The operation of these systems is not affected Transient required.

by the introduction of ATRIUM 1OXM fuel and Performance AREVA methodology.

Instrumentation Systems 7.11 Reactor No further analysis Reactor shutdown capability is not affected by Shutdown required.

the introduction of ATRIUM 1OXM fuel and Capability AREVA methodology.

7.12 Detailed Control No further analysis Control room design is not affected by the Room Design required.

introduction of ATRIUM 1OXM fuel and AREVA Review methodology.

7.13 Safety Parameter No further analysis Safety parameter display system is not Display 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 the System required.

introduction of ATRIUM 1OXM fuel and AREVA methodology.

8.3 Auxiliary Power No further analysis In case of loss of auxiliary power event the System required.

reactor scrams and if it is not restored the diesel generator will carry the vital loads. See disposition of Station Blackout event below.8.4 Plant Standby Address for initial The plant standby diesel generator system Diesel Generator licensing analysis.

features are incorporated into the LOCA break System spectrum analysis which is performed for the ATRIUM 1OXM fuel with the AREVA methodology (Reference 29).8.5 DC Power Address for initial The DC power supply system features are Supply Systems licensing analysis.

incorporated into the LOCA break spectrum analysis which is performed for the ATRIUM 1OXM fuel with the AREVA methodology (Reference 29).8.6 Reactor No further analysis The power supplies for reactor protection Protection required.

system are not affected by the introduction of System Power ATRIUM 1OXM fuel and AREVA methodology.

Supplies AREVA NP Inc.

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Revision 1 Page 2-9 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 8.7 Instrumentation No further analysis These systems are not affected by the and Control AC required.

introduction of ATRIUM 1OXM fuel and AREVA Power Supply methodology.

Systems 8.8 Electrical Design No further analysis Independent of fuel design. Analysis of record Considerations required.

remains valid.8.9 Environmental No further analysis Independent of fuel design. Analysis of record Qualification of required.

remains valid.Safety-Related Electrical Equipment 8.10 Adequacy of No further analysis Independent of fuel design. Analysis of record Station Electrical required.

remains valid.Distribution System Voltages 8.11 Power Operated Address each reload. Functionality of safety related valves is Valves 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 for required.

station blackout.

AREVA dispositioned the impact of ATRIUM 1OXM fuel by comparing the decay heat for ATRIUM 1 OXM fuel to the decay heat used in the station blackout analysis of record. Since the ATRIUM 1OXM fuel decay heat is expected to be similar to that of the GE14 fuel the analysis of record results bound the introduction of ATRIUM 1OXM fuel at Monticello.

9.0 Radioactive

Waste Management No further analyses required.As shown by radiological source term evaluations, the ATRIUM 1OXM fuel is not significantly different than legacy fuel.ATRIUM 1OXM fuel is designed and operated to comparable standards that would ensure fuel cladding integrity such that fission products will continue to be contained within the cladding.Therefore, plant operations following the fuel transition are not expected to increase the rate that radiological waste is generated.

AREVA NP Inc.

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Revision 1 Page 2-10 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 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 abnormal Fuel Handling licensing analysis.

conditions for spent fuel pool storage racks has Systems 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 be System 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-specific Removal System licensing analysis.

evaluation of fuel performance in response to Service Water postulated LOCA upon introduction of the System ATRIUM 1OXM fuel in MNGP (Reference 29).The decay heat removal design basis of the RHR system is not altered by the introduction of ATRIUM 1OXM fuel in MNGP.Inadvertent RHR shutdown cooling operation is a benign event which does not need evaluation.

11.0 Plant Power Address each reload. These systems are part of the safety analysis Conversion models and their features affect the transient Systems analysis results. These systems are modeled within the plant transient analyses as appropriate for the introduction of ATRIUM 1OXM fuel at MNGP (Section 5.0).12.0 Plant Structures No further analyses Independent of fuel design. Analysis of record and Shielding required.

remains valid.AREVA NP Inc.

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Revision 1 Page 2-11 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 13.0 Plant Operation Address for initial Organization, Responsibilities, and licensing analysis.

Qualifications of staff personnel are not affected by transitioning to ATRIUM 1OXM fuel.Training in AREVA methodologies will be provided for the initial reload. The Emergency Operational Procedures (EOPs) may need to be updated to include the effects of ATRIUM 1OXM fuel. The overall nuclear site organization and plant functional organization are not affected by the introduction of AREVA fuel.14.0 Plant Safety See below.Analysis 14.2 MCPR Safety Address each reload. Part of the safety licensing analysis done for Limit each reload with AREVA methodology (Section 4.2).14.3 Operating Limits Address each reload. Power- and flow-dependent MCPR and LHGR limits will be established for each reload using AREVA methodology.

In addition MAPLHGR limits will be established and verified each cycle for the ATRIUM 1OXM fuel designs (Section 8.0).14.4 Transient Events See below.Analyzed for Core Reload 14.4.1 Generator Load Address each reload. This event without bypass operable is a Rejection 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 was Feedwater evaluated.

Since the generic analysis does not Heating apply, this event will be analyzed for the initial cycle. Since the results of this event show this is a potentially limiting event, this event will also be analyzed each reload (Section 5.1.6).14.4.3 Rod Withdrawal No further analysis Consequences of a RWE below the low power Error -low required.

setpoint are bound by the RWE at power due power to required strict compliance with BPWS.14.4.3 Rod Withdrawal Address each reload. Analysis to determine the change in MCPR Error -at power and LHGR as a function of RBM setpoint will be performed for each reload. The analysis will cover the low, intermediate, and high power RBM ranges (30% to 100% power)(Section 5.1.7).AREVA NP Inc.

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Revision 1 Page 2-12 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 14.4.4 Feedwater Address each reload. This event is a potentially limiting AOO and will Controller Failure be analyzed each reload (Section 5.1.4).-Maximum Demand 14.4.5 Turbine Trip Address each reload. This event without bypass operable is a Without Bypass potentially limiting AOO. TT with bypass operable is bounded by the TT with no bypass case. TT with bypass operable and degraded scram may be a limiting event for MNGP and has 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 ASME ASME Code code. The initial analysis will address MSIV, Compliance TCV, and TSV closures under AREVA Model -MSIV methodology.

Since the limiting valve closure Closure 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 shutdown Control System capability is evaluated on a cycle-specific basis Shutdown Margin (Section 7.5).14.5.3 Stuck Rod Cold Address each reload. This event is potentially limiting and will be Shutdown Margin analyzed each reload (Reference 1).14.6 Plant Stability Address each reload. Option III will be implemented with the Analysis transition to AREVA methods. DIVOM and initial MCPR will be analyzed on a cycle-specific basis (Section 4.3).The Backup Stability Protection (BSP) regions will be verified on a cycle-specific basis and adjusted if necessary based on the results of the analyses (Section 4.3).14.7 Accident See below.Evaluation Methodology AREVA NP Inc.

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Revision 1 Page 2-13 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 14.7.1 Control Rod Address each reload. Safety analyses are performed each reload to Drop Accident evaluate the CRDA to verify that the accident Evaluation will not result in fuel pellet deposited enthalpy greater than 280 calories per gram and to determine the number of rods exceeding the 170 calories per gram failure threshold.

For Monticello, the analysis will verify that deposited enthalpy remains below 230 cal/gm.Consequences of the CRDA are evaluated to confirm that the acceptance criteria are satisfied (Section 6.3).14.7.2 Loss-of-Coolant Address for initial LOCA calculations will be performed for EPU Accident licensing analysis.

to identify the limiting fluid conditions as a function of single failure, break size, break location, core flow, and axial power shape using the NRC-approved EXEM BWR-2000 LOCA methodology.

This analysis is performed for the initial introduction of ATRIUM 1OXM fuel (Reference 29).MAPLHGR heatup analyses are performed every time a new neutronic design is introduced in the core (Reference 30).14.7.3 Main Steam Line Address for initial The main steam line break will be considered Break 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-coolant accident 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 by Recirculation LOCA accident analysis and does not need Pump Seizure further analysis.Accident Analysis Single-loop pump seizure event has been historically analyzed against the more restrictive criteria for infrequent events (AOO).Using these criteria, this is the limiting event for single-loop operation and it will have to be analyzed each reload (Section 5.3.1).14.7.6 Refueling Address for initial The number of fuel rods assumed to fail during Accident licensing analysis.

a fuel handling accident for an ATRIUM 1OXM Analysis assembly dropping over the core has been determined and the resulting release dispositioned against the AST analyses of record (Section 6.4).AREVA NP Inc.

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Revision 1 Page 2-14 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 14.7.7 Accident Atmospheric Dispersion Coefficients 14.7.8 Core Source Term Inventory 14.8 Anticipated Transients Without Scram (ATWS)No further analysis required.Address for initial licensing analysis.Address each reload.Independent of fuel design. The values of atmospheric dispersion coefficients in the analysis of record remain valid.The source terms for ATRIUM 10XM fuel at EPU conditions have been provided and used to disposition offsite doses against the AST analysis of record. As shown by radiological source term evaluations, the ATRIUM 10XM fuel is not significantly different than legacy fuel (GE14).The peak vessel pressure is calculated for each reload. For long-term cooling after ATWS, the decay heat is the only fuel-related input. AREVA dispositioned the impact of ATRIUM 1OXM fuel by comparing the decay heat for ATRIUM 1OXM fuel to the GE14 decay heat used in the ATWS long-term cooling analysis.

Containment heatup was dispositioned by comparing kinetics parameters for ATRIUM 10XM fuel with those for the fuel in the analysis of record (Section 7.2).14.9 Section deleted NA NA 14.10 Other Analyses 14.10.1 Adequate Core Cooling for Transients with a Single Failure See below.No further analysis required USAR 14.10.1 identifies the loss of feedwater flow event as the worst anticipated transient, and loss of a high pressure inventory makeup (HPCI) or heat removal system as the worst single failure.The analysis of record for loss of feedwater flow (PUSAR 2.8.5.2.3) already assumed that the HPCI system fails to inject. The results of this analysis showed that the reactor core remains covered for the combination of these worst-case conditions, without operator action to manually initiate the emergency core cooling system or other inventory makeup systems, therefore no further analysis is required.The events identified in the Supplemental Reload Licensing Submittal are addressed below as part of the PUSAR (Reference 5).14A Supplemental Reload Licensing Submittal See below.AREVA NP Inc.

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Revision 1 Page 2-15 Table 2.1 Disposition of Events Summary (continued)

PUSAR Design Disposition Sect. Criteria / Event Status Comment Decrease in Reactor Coolant Temperature 2.8.5.7 Pressure Regulator Address for initial Address for initial reload for EPU/MELLLA Failure -Open licensing analysis.

conditions.

Consequences of this event, relative to AOO thermal operating limits, are nonlimiting.

This event results in low steam dome pressure and is the most challenging event for Technical Specification (TS) 2.1.1.1 (Reference

3) low steam dome pressure safety limit. This section of the TS will be updated to reduce the 785 psig limit to a lower pressure limit. The analysis of this event (for initial licensing analysis) will support this update to Technical Specifications (Section 7.3).This event is also used for an ATWS initiator event.Decrease in Heat Removal By the Secondary System/ Increase in Reactor Pressure 2.8.5.2.1 Pressure Regulator Failure -Closed 2.8.5.2.1 MSIV Closures Address each reload.No further analysis required.Consequences of this event, relative to one pressure regulator out-of-service may be limiting; therefore this EOOS event will be evaluated on a cycle-specific basis (Section 5.3.2).Consequences of this event (with direct scram on MSIV closure), relative to thermal operating limits, are bounded by the generator load rejection event. This event does not need further analysis.Closure of all MSIVs with failure of the valve position scram function is the design basis overpressurization event, which is evaluated on a cycle-specific basis (Section 7.1).The MSIV closure event is a potentially limiting ATWS overpressurization event, which is evaluated on a cycle-specific basis (Section 7.2).AREVA NP Inc.

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Revision 1 Page 2-16 Table 2.1 Disposition of Events Summary (continued)

PUSAR Design Disposition Sect. Criteria / Event Status Comment 2.8.5.2.1 Loss of Condenser No further Consequences of this event are bounded by Vacuum analysis required.

either the turbine trip with turbine bypass valve failure or load rejection with bypass valve failure.2.3.5 Loss of AC Power No further This event is analyzed as the Station analysis 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 only Flow analysis required.

dependent on the fuel decay heat, since this event was analyzed as initiated at the low level (L3) scram setpoint in the analysis of record. Since the decay heat of ATRIUM 1OXM fuel is similar to that of GE14 fuel the results are expected to be similar to the current analysis of record.Decrease in Reactor Coolant System Flow Rate Not Recirculation Pump No further Consequences of this event are benign and evaluated 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 pump evaluated Controller Failure -analysis required.

trip events.Decreasing Flow 2.8.5.3.2 Recirculation Pump No further The consequences of this accident are Shaft Break analysis required.

bounded by the effects of the recirculation pump seizure event (see above).AREVA NP Inc.

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Revision 1 Page 2-17 Table 2.1 Disposition of Events Summary (continued)

PUSAR Design Disposition Sect. Criteria / Event Status Comment Reactivity and Power Distribution Anomalies 2.8.5.4.1 Control Rod Mal- No further Consequences of this event are bounded by operation (system analysis required.

the RWE at power.malfunction or operator error) -low power 2.8.5.4.2 Control Rod Mal- Address each Analysis to determine the change in MCPR operation (system reload, and LHGR as a function of RBM setpoint will malfunction or be performed for each reload. The analysis operator error) -at will cover the low, intermediate, and high power 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 to Pump control thermal stresses caused by startup of an inactive recirculation pump. PUSAR identifies this event as being nonlimiting.

The introduction of ATRIUM 1OXM fuel will not affect this conclusion.

2.8.5.4.3 Recirculation Flow Address each The slow runup event determines the MCPRf Control Failure With reload, limit and LHGRf multiplier and therefore will Increasing Flow (slow be analyzed each reload (Section 5.2)and fast runup The fast runup event, if not bounded by the events) slow flow runup event, will be considered in setting the MCPRP limits (Section 5.1.8).Increase in Reactor Coolant Inventory USAR Inadvertent HPCI Address each This is a potentially limiting event which will 14A 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 is which increase analysis required.

the inadvertent HPCI start-up which will be reactor coolant analyzed each reload.inventory AREVA NP Inc.

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Revision 1 Page 2-18 Table 2.1 Disposition of Events Summary (continued)

PUSAR Design Disposition Sect. Criteria / Event Status Comment Decrease in Reactor Coolant Inventory 2.5.4.1 Inadvertent No further This event results in a mild depressurization and Safety/Relief Valve analysis required.

event which is less severe than the pressure 2.8.5.6.1 Opening regulator failure open event (see Section 7.3). Since the power level settles out at nearly the initial power level, this event is 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-coolant accident scenario (Reference 29).Radioactive Release From Subsystems and Components 2.9.1 Gaseous Radwaste No further As shown by radiological source term System Leak or analysis required.

evaluations, the ATRIUM 1OXM fuel is not Failure significantly different than legacy fuel (GE14). Further, ATRIUM 1OXM fuel is designed and operated to comparable standards that would ensure fuel cladding integrity such that fission products will continue to be contained within the cladding.Therefore, plant operations following the fuel transition are not expected to increase the rate that radiological waste is generated.

2.9.2 Liquid Radwaste No further The radionuclide source terms are generic System Failure analysis required.

and are unaffected by the introduction of ATRIUM 1OXM fuel.2.9.2 Postulated No further The radionuclide source terms are generic Radioactive Releases analysis required.

and are unaffected by the introduction of Due to Liquid ATRIUM 1OXM fuel.Radwaste Tank Failure AREVA NP Inc.

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Revision 1 Page 2-19 Table 2.1 Disposition of Events Summary (continued)

PUSAR Design Disposition Sect. Criteria / Event Status Comment Other Analyses 2.8.3.3 ATWS with Core No further The discussion presented in Reference 41 Instability analysis required.

indicates that the "Parameters which might vary between fuel designs (e.g., reactivity coefficients) are not expected to significantly change the consequences of large irregular oscillations." Therefore, the generic ATWS stability results of Reference 41 remain applicable upon the introduction of ATRIUM 1OXM fuel into MNGP.AREVA NP Inc.

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Revision 1 Page 2-20 Table 2.2 Disposition of Operating Flexibility and EOOS Options on Limiting Events Affected Limiting Comment Option 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 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 critieria.

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

Therefore the base case operating limits already include this condition.

ATWS Peak ATWS peak pressure analysis considers only one Pressure SRVOOS.Pressure regulator If one of the pressure regulators is OOS the out-of-service backup pressure regulator will operate and (PROOS) therefore not affect the severity of a particular event.The pressure regulator down-scale failure event and the pressure regulator failed open event were addressed 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 flow operation and MELLLA core flow window.AREVA NP Inc.

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Revision 1 Page 2-21 Table 2.3 Methodology and Evaluation Models for Cycle-Specific Reload Analyses Analysis Event Methodology Evaluation Acceptance Criteria lAnalysis Reference Model and Comment Thermal and Hydraulic 12 SAFLIM3D SLMCPR Criteria:

< 0.1% fuel rods Design 24 COTRANSA2 experience boiling transition.

No fuel melting and maximum Transient Analyses 25 XCOBRANofemltnadmxiu transient induced strain < 1 %.26 XCOBRA-T Power- and flow-dependent MCPR 9 RODEX4 and LHGR operating limits established to meet the fuel failure 28 RODEX2crtia criteria.Standby Liquid Control 27 CASMO-4 SLCS Criteria:

Shutdown margin of System /MICROBURN-B2 at least 0.88% Ak/k.ASME 24 COTRANSA2 Analyses for ASME and ATWS Overpressurization (as supplemented overpressurization.

Analysis by considerations AnalyssbyoAnsie ASME Overpressurization Criteria: of ANP-3224(P)

Maximum vessel pressure limit of Anticipated Transient (Reference 6, 1375 psig and maximum dome Without Scram App. E)) pressure limit of 1332 psig.(pressurization)

A TWS Overpressurization Criteria: Maximum vessel pressure limit of 1500 psig.Emergency Core 34 HUXY LOCA Criteria:

1OCFR50.46.

Cooling Systems EXEM BWR-2000 Methodology.

LOCA Analyses Only heatup (HUXY) is analyzed for the reload specific neutronic design.Appendix R 34 RELAX 10CFR50 Appendix R.Neutron Design 18 STAIF Long-Term Stability Solution 19 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 limits 22 established to meet the fuel failure criteria.23 Backup Stability Protection 27 Criteria:

Stability boundaries that do not exceed acceptable global, regional, and channel decay ratios as defined by the STAIF methodology.

AREVA NP Inc.

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Page 3-1 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 exposure 62.0 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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Page 4-1 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.AREVA NP Inc.

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Page 4-2 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.AREVA NP Inc.

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Revision 1 Page 4-3 Table 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses Parameter Uncertainty Fuel-Related Uncertainties I I Plant-Related Uncertainties Feedwater flow rate 1.8%Feedwater temperature 0.8%Core pressure 0.8%Total core flow rate TLO 2.5%SLO 6.0%I AREVA NP Inc.I uontroned Uocument Monticello Fuel Transition Cycle 28 Reload Licensing Analysis (EPU/MELLLA)

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Revision 1 Page 4-4 Table 4.2 Results Summary for Safety Limit MCPR Analyses Percentage of Rods in Boiling SLMCPR Transition TLO -1.12 0.0924 SLO -1.13 0.0812 AREVA NP Inc.

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Revision 1 Page 4-5 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 Power Criteria OLMCPR OLMCPR as at Described in 45% Flow Section 8.0 AREVA NP Inc.

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Revision 1 Page 4-6 Table 4.4 BSP Endpoints for Monticello Cycle 28 Power Flow Endpoint (%) (%) 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 AREVA NP Inc.

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Page 5-1 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 AREVA NP Inc.

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Page 5-2 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 AREVA NP Inc.

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Page 5-3 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.AREVA NP Inc.

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Page 5-4 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)).AREVA NP Inc.

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Page 5-5 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 95.3 0 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.AREVA NP Inc.

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Page 5-6 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.AREVA NP Inc.

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Page 5-7 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 AREVA NP Inc.

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Page 5-8 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.AREVA NP Inc.

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Page 5-9 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 projected EOFP solution at the state conditions provided in Table 5.11 is greater than the integrated normalized power generated in the bottom 7 nodes in the licensing basis axial power profile in Table 5.11, and the individual normalized power from the projected EOFP solution is greater than the corresponding individual normalized power from the licensing 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 to licensing 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.AREVA NP Inc.

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Revision 1 Page 5-10 Table 5.1 Exposure Basis for Monticello Cycle 28 Transient Analysis Core Cycle 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 to EOFP + 400 MWd/MTU (licensing basis EOFP)21,175 38,232 Maximum licensing core exposure -including Coastdown AREVA NP Inc.

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Revision 1 Page 5-11 Table 5.2 Scram Speed Insertion Times TSSS NSS DSS Control 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 AREVA NP Inc.

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Revision 1 Page 5-12 Table 5.3 Licensing Basis EOFP Base Case LRNB 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) 1.51 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 AREVA NP Inc.

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Revision 1 Page 5-13 Table 5.4 Licensing Basis EOFP Base Case TTNB Transient Results Power ATRIUM 1OXM GE14 (% rated) ACPR ACPR TSSS Insertion Times 100 0.41 80 0.41 60 0.40 40 (above Pbypass) 0.38 40 at > 50%F (below Pbypass) 1.25 40 at 5 50%F (below Pbypass) 0.95 25 at > 50%F (below Pbypass) 1.51 25 at < 50%F (below Pbypass) 1.22 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 AREVA NP Inc.

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Revision 1 Page 5-14 Table 5.5 Licensing Basis EOFP Base Case TTWB Transient Results Power ATRIUM 1OXM GE14 (% 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 AREVA NP Inc.

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Revision 1 Page 5-15 Table 5.6 Licensing Basis EOFP Base Case FWCF Transient Results Power ATRIUM 1OXM GE14 (% rated) ACPR ACPR TSSS Insertion Times 100 0.43 80 0.45 60 0.49 40 (above Pbypass) 0.62 40 at > 50%F (below Pbypass) 1.60 40 at < 50%F (below Pbypass) 1.16 25 at > 50%F (below Pbypass) 2.22 25 at < 50%F (below Pbypass) 1.92 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 AREVA NP Inc.

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Revision 1 Page 5-16 Table 5.7 Licensing Basis EOFP Base Case HPCI Transient Results Power ATRIUM 1OXM GE14 (% rated) ACPR ACPR TSSS Insertion Times 100 0.47 80 0.47 60 0.53 40 (above Pbypass) 0.59 40 at > 50%F (below Pbypass) 1.31 40 at < 50%F (below Pbypass) 1.10 25 at > 50%F (below Pbypass) 1.56 25 at < 50%F (below Pbypass) 1.48 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 AREVA NP Inc.

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Revision 1 Page 5-17 Table 5.8 Licensing Basis EOFP Base Case CRWE Results High Intermediate Low Power Range Power Range Power Range RBM Trip Core RBM Trip Core RBM Trip Core Setpoint Power Setpoint Power Setpoint Power (%) (% rated) MCPR (%) (% rated) MCPR (%) (% rated) MCPR 110 100 1.47 115 85 1.56 120 65 1.77 85 1.49 65 1.62 30 2.20 111 100 1.48 116 85 1.58 121 65 1.79 85 1.50 65 1.63 30 2.24 112 100 1.50 117 85 1.60 122 65 1.80 85 1.52 65 1.65 30 2.24 113 100 1.52 118 85 1.60 123 65 1.80 85 1.53 65 1.77 30 2.31 114 100 1.52 119 85 1.60 124 65 1.80 85 1.54 65 1.77 30 2.31 AREVA NP Inc.

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

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Revision 1 Page 5-18 Table 5.9 RBM Operability Requirements Thermal Applicable Power ATRIUM 1OXM / GE14 (% rated) MCPR 2.46 TLO> 27% and < 90% 2.47 SLO 2.47 SLO_90% 1.65 TLO AREVA NP Inc.

Luontrwo~e uocumenz Monticello Fuel Transition Cycle 28 Reload Licensing Analysis (EPU/MELLLA)

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Revision 1 Page 5-19 Table 5.10 Licensing Basis EOFP PRFDS (PROOS)Transient Results Power ATRIUM 1OXM GE14 (% rated) ACPR ACPR TSSS Insertion Times 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~(~Monticello Fuel Transition Cycle 28 Reload Licensing Analysis (EPU/MELLLA)

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Revision 1 Page 5-20 Table 5.11 Licensing Basis Core Average Axial Power Profile State Conditions for Power 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 Power Top 24 0.325 23 0.736 22 1.194 21 1.368 20 1.476 19 1.508 18 1.502 17 1.472 16 1.407 15 1.372 14 1.397 13 1.378 12 1.317 11 1.232 10 1.137 9 1.034 8 0.909 7 0.773 6 0.650 5 0.541 4 0.455 3 0.396 2 0.321 Bottom 1 0.099 Sum of Bottom 7 Nodes= 3.235 AREVA NP Inc.

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

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Revision 1 Page 5-21 I~nn n 400.0 -300.0 -"0 C: a-Relative Core Power Relative Heat Flux Relative Core Flow------- --R --- -- --- ----- --- ---Relative Steam Flow Relative 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 EOFP LRNB at 10OPI105F

-TSSS Key Parameters AREVA NP Inc.

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

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Revision 1 Page 5-22 1300.0 2:3 in U)4.0 6.0 Time (seconds)Figure 5.2 Licensing Basis EOFP LRNB at 10OP/105F

-TSSS Vessel Pressures AREVA NP Inc.

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

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Revision 1 Page 5-23 600.0 500.0-400.0 -Relative Core Power Relative Heat Flux Relative Core Flow Relative Steam Flow Relative Feed Flow-O 0)(D 1 C,, 0)0@300.0 -200.0 -/100.0,.0'------ ---- -----------------------------------------------------------------------------.

\ I,, /--I nn I 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 EOFP TTNB at 1OOPI105F

-TSSS Key Parameters AREVA NP Inc.

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

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Revision 1 Page 5-24 I 'Al A 1300.0-1250.0-2 V) 1200.0-U 1150.0-a3/ " ,/.Steam Dome Lower Plenum 1100.0-1050.0-'AnnnA.0 I 1 .0 2.0 3.0 4.0 Time (seconds)5.0 6.0 7.0 8.0 Figure 5.4 Licensing Basis EOFP TTNB at 1OOP/105F

-TSSS Vessel Pressures AREVA NP Inc.

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

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Revision 1 Page 5-25 600 0 500.0 400.0 300.0 200.0-0 C a-)Relative Core Power Relative Heat Flux Relative Core Flow RelativeSteamFlow----------.-------

Relative Steam Flow Relative Feed Flow......-100.0-.0 --100.0-.0 10.0 20.0 30.0 Time (seconds)Figure 5.5 Licensing Basis EOFP FWCF at 1OOP/1 05F -TSSS Key Parameters 40.0 50.0 AREVA NP Inc.

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

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Revision 1 Page 5-26 1300.0 1200.0 2 En in)1)U)CL 1100.0 1000.0 900.0 20.0 30.0 Time (seconds)Figure 5.6 Licensing Basis EOFP FWCF at 1OOP/105F

-TSSS Vessel Pressures AREVA NP Inc.

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

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Revision 1 Page 5-27 bUU.U .500.0 -400.0 -Relative Core Power Relative Heat Flux Relative Core Flow Relative Steam Flow Relative 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 EOFP HPCI at 10OP/105F

-TSSS Key Parameters AREVA NP Inc.

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

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Revision 1 Page 5-28-9, U)Q), Figure 5.8 Licensing Basis EOFP HPCI at 10OP/105F

-TSSS Vessel Pressures AREVA NP Inc.

Uontrolued Uocument Monticello ANP-3213(NP)

Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 6-1 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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Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 6-2 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 rods exceeding 170 cal/g 736 AREVA NP Inc.

uontrolied Uocument Monticello ANP-3213(NP)

Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 6-3 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.AREVA NP Inc.

uontrotflea uocument Monticello ANP-3213(NP)

Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 7-1 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 peak pressure 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).AREVA NP Inc.

uontroIued Uocument Monticello ANP-3213(NP)

Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 7-2 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 AREVA NP Inc.

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Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 7-3 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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Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 7-4 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 _ 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).AREVA NP Inc.

Uontroned uocument Monticello ANP-3213(NP)

Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 7-5 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 319.2 0 F, with a boron concentration equivalent to 660 ppm at 68 0 F.* The temperature of 319.2 0 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 are adequate to protect the 70'F licensing basis for the plant.AREVA NP Inc.

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

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Revision 1 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.AREVA NP Inc.

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

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Revision 1 Page 7-7 Table 7.1 ASME Overpressurization Analysis Results*Maximum Peak Peak Vessel Maximum Neutron Heat Pressure Dome Flux Flux Lower-Plenum Pressure Event (% rated) (% rated) (psig) (psig)MSIV closure (102P/99F) 388 132 1360 1326 Pressure limit --- --- 1375 1332* Pressure results include various adders totaling 9 psi to account for void-quality correlations, Doppler void effects, and thermal conductivity degradation.

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

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Revision 1 Page 7-8 Table 7.2 ATWS Overpressurization Analysis Results*Maximum Peak Peak Vessel Maximum Neutron Heat Pressure Dome Flux Flux Lower-Plenum Pressure Event (% rated) (% rated) (psig) (psig)MSIV closure (102P/99F) 308 144 1436 1419 PRFO (102P/99F) 263 151 1445 1428 Pressure limit --- --- 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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Revision 1 Page 7-9-o~1)0 0:: 0 C 0)U L 0)0~4.0 f Time (seconds)Figure 7.1 MSIV Closure Overpressurization Event at 102P/99F -Key Parameters AREVA NP Inc.

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Revision 1 Page 7-10 U)()"1" a.2.0 4.0 6.0 8.Time (seconds)Figure 7.2 MSIV Closure Overpressurization Event at 102P/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.

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

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Revision 1 Page 7-11 600.0 Bank 1 Bank 2 Bank 3 Bank 4 Bank 5 500.0-U)`_" E.. 400.0-cI, Q' 300.0 n 200.0 V)V.,-(K 100.0.0.0 2.0 4.0 Time (seconds)I: I I 8.0 10.0 Figure 7.3 MSIV Closure Overpressurization Event at 102P/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.AREVA NP Inc.

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

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Revision 1 Page 7-12 ju.5UU-Relative Core Power 200.0 --o 0 C W, p W, a-Relative Heat Flux Relative Core Flow Relative Steam Flow Relative Feed Flow""- -:-.-:..:-

--. .-------100.0-

-L .........-

-.0 -Ii I'-I CIA (I e 1000-1*.0 5.0 10.0 1 5o Time (seconds')

20.0 25.0 30.0 Figure 7.4 PRFO ATWS Overpressurization Event at 102P/99F -Key Parameters AREVA NP Inc.

uon, ro e Uocument Monticello Fuel Transition Cycle 28 Reload Licensing Analysis (EPU/MELLLA)

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Revision 1 Page 7-13 0 U)0.0)L U)U)U, L 0~Time (seconds)Figure 7.5 PRFO ATWS Overpressurization Event at 102P/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.

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

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Revision 1 Page 7-14 Q1)(I)E 0 W-15.0 Time (seconds)Figure 7.6 PRFO ATWS Overpressurization Event at 102P/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 SRVs in bank 5.AREVA NP Inc.

Uon:roiied uocurnen Monticello ANP-3213(NP)

Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 8-1 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 uocument Monticello ANP-3213(NP)

Fuel Transition Cycle 28 Revision 1 Reload 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.AREVA NP Inc.

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Revision 1 Page 8-3 Table 8.1 MCPRp Limits for Two-Loop Operation (TLO), NSS Insertion Times BOC to Licensing Basis EOFP*MCPRP Operating Power ATRIUM 1OXM GE14 Condition

(% of rated) Fuel Fuel Base 100.0 1.55 1.53 case 40.0 1.71 1.71 operation 40.0 at > 50%F 2.77 2.72 25.0 at > 50%F 3.39 3.47 40.0 at < 50%F 2.33 2.38 25.0 at < 50%F 3.09 3.23 PROOS 100.0 1.59 1.58 85.0 1.64 1.64 85.0 1.91 1.84 40.0 2.39 2.30 40.0 at > 50%F 2.77 2.72 25.0 at > 50%F 3.39 3.47 40.0 at < 50%F 2.48 2.38 25.0 at < 50%F 3.24 3.23* Limits support operation with up to 3 SRVOOS, up to 1 TIPOOS or the equivalent number of TIP channels and/or up to 50% of the LPRMs out-of-service, and a 1200 EFPH LPRM calibration interval.AREVA NP Inc.

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Revision 1 Page 8-4 Table 8.2 MCPRp Limits for Two-Loop Operation (TLO), TSSS Insertion Times BOC to Licensing Basis EOFP*MCPRp Operating Power ATRIUM 10XM GE14 Condition

(% of rated) Fuel Fuel Base 100.0 1.59 1.58 case 40.0 1.76 1.79 operation 40.0 at > 50%F 2.77 2.72 25.0 at > 50%F 3.39 3.47 40.0 at < 50%F 2.33 2.38 25.0 at < 50%F 3.09 3.23 PROOS 100.0 1.59 1.58 85.0 1.64 1.64 85.0 1.91 1.84 40.0 2.39 2.30 40.0 at > 50%F 2.77 2.72 25.0 at > 50%F 3.39 3.47 40.0 at < 50%F 2.48 2.38 25.0 at < 50%F 3.24 3.23* Limits support operation with up to 3 SRVOOS, up to 1 TIPOOS or the equivalent number of TIP channels and/or up to 50% of the LPRMs out-of-service, and a 1200 EFPH LPRM calibration interval.AREVA NP Inc.

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Revision 1 Page 8-5 Table 8.3 MCPRp Limits for Two-Loop Operation (TLO), NSS Insertion Times BOC to Coastdown*

MCPRP Operating Power ATRIUM 10XM GE14 Condition

(% of rated) Fuel Fuel Base 100.0 1.55 1.53 case 40.0 1.74 1.71 operation 40.0 at > 50%F 2.77 2.72 25.0 at > 50%F 3.39 3.47 40.0 at < 50%F 2.33 2.38 25.0 at < 50%F 3.09 3.23 PROOS 100.0 1.59 1.58 85.0 1.64 1.64 85.0 1.91 1.84 40.0 2.39 2.30 40.0 at > 50%F 2.77 2.72 25.0 at > 50%F 3.39 3.47 40.0 at < 50%F 2.48 2.38 25.0 at < 50%F 3.24 3.23* Limits support operation with up to 3 SRVOOS, up to 1 TIPOOS or the equivalent number of TIP channels and/or up to 50% of the LPRMs out-of-service, and a 1200 EFPH LPRM calibration interval.AREVA NP Inc.

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Revision 1 Page 8-6 Two-Loop Table 8.4 MCPRp Limits for Operation (TLO), TSSS Insertion Times BOC to Coastdown*

MCPRP Operating Power ATRIUM 10XM GE14 Condition

(% of rated) Fuel Fuel Base 100.0 1.59 1.58 case 40.0 1.77 1.79 operation 40.0 at > 50%F 2.77 2.72 25.0 at > 50%F 3.39 3.47 40.0 at 5 50%F 2.33 2.38 25.0 at < 50%F 3.09 3.23 PROOS 100.0 1.59 1.58 85.0 1.64 1.64 85.0 1.91 1.84 40.0 2.39 2.30 40.0 at > 50%F 2.77 2.72 25.0 at > 50%F 3.39 3.47 40.0 at < 50%F 2.48 2.38 25.0 at < 50%F 3.24 3.23* Limits support operation with up to 3 SRVOOS, up to 1 TIPOOS or the equivalent number of TIP channels and/or up to 50% of the LPRMs out-of-service, and a 1200 EFPH LPRM calibration interval.AREVA NP Inc.

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Revision 1 Page 8-7 Table 8.5 MCPRP Limits for Single-Loop Operation (SLO), TSSS Insertion Times BOC to Coastdown*

t MCPRP Operating Power ATRIUM 10XM GE14 Condition

(% of rated) Fuel Fuel Base 66.0 2.13 2.19 case 40.0 2.40 2.31/PROOS 40.0 at > 50%F 2.78 2.73 25.0 at > 50%F 3.40 3.48 40.0 at < 50%F 2.49 2.39 25.0 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.AREVA NP Inc.

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Revision 1 Page 8-8 Table 8.6 Flow-Dependent MCPR Limits ATRIUM 1OXM and GE14 Fuel, NSS/TSSS Insertion Times, TLO and SLO, PROOS All Cycle 28 Exposures Core Flow (% of rated) MCPRf 30.0 1.80 80.0 1.50 105.0 1.50 AREVA NP Inc.

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Revision 1 Page 8-9 Table 8.7 ATRIUM 1OXM Steady-State LHGR Limits Peak Pellet Exposure LHGR (GWd/MTU) (kW/ft)0.0 14.1 18.9 14.1 74.4 7.4 AREVA NP Inc.

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Revision 1 Page 8-10 Table 8.8 ATRIUM 1OXM LHGRFACp Multipliers for NSS/TSSS Insertion Times, TLO and SLO, All Cycle 28 Exposures*

LHGRFACp Operating Power ATRIUM 1OXM Condition

(% of rated) Fuel 100.0 1.00 40.0 0.80 Base 40.0 at > 50%F 0.44 case 25.0 at > 50%F 0.30 operation 40.0 at < 50%F 0.56 25.0 at 50%F 0.36 PROOS 100.0 1.00 85.0 0.95 85.0 0.92 40.0 0.66 40.0 at > 50%F 0.44 25.0 at > 50%F 0.30 40.0 at: <50%F 0.56 25.0 at 50%F 0.36* 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.AREVA NP Inc.

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Revision 1 Page 8-11 Table 8.9 GE14 LHGRFACp Multipliers for NSSITSSS Insertion Times, TLO and SLO, All Cycle 28 Exposures*

LHGRFACP Operating Power GE14 Condition

(% of rated) Fuel Base 100.0 0.9 9 t case 40.0 0.57 operation 40.0 at > 50%F 0.42 25.0 at > 50%F 0.34 40.0 at 50%F 0.53 25.0 at < 50%F 0.37 PROOS 100.0 0.9 9 t 85.0 0.89 85.0 0.75 40.0 0.54 40.0 at > 50%F 0.42 25.0 at > 50%F 0.34 40.0 at < 50%F 0.51 25.0 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).

AREVA NP Inc.

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Revision 1 Page 8-12 Table 8.10 ATRIUM 1OXM LHGRFACf 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 AREVA NP Inc.

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Revision 1 Page 8-13 Table 8.11 GE14 LHGRFACf 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 AREVA NP Inc.

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Revision 1 Page 8-14 Table 8.12 ATRIUM 1OXM MAPLHGR Limits, TLO*Average Planar Exposure 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.

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Fuel Transition Cycle 28 Revision 1 Reload 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 Pressure Power 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 T M IOXM Fuel Assemblies, AREVA NP, October 2012.8. ANP-3221 P Revision 0, Fuel Rod Thermal-Mechanical Design for Monticello ATRIUM IOXM Fuel Assemblies, Cycle 28, AREVA NP, May 2013.9. BAW-1 0247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.10. GNF Design Basis Document, Fuel-Rod Thermal-Mechanical Performance Limits for GE14C, DB-001 2.03 Revision 2, September 2006 (transmitted by letter, D.J. Mienke (Xcel Energy) to R. Welch (AREVA), "Transmittal of Requested Monticello Plant Information:

GE14 Exposure Limits," July 19, 2012).11. ANP-3092(P)

Revision 0, Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies, AREVA NP, July 2012.12. AN P-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.13. EMF-2209(P)(A)

Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.14. EMF-2245(P)(A)

Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.15. ANP-3138(P)

Revision 0, Monticello Improved K-factor Model forACE/ATRIUM IOXM Critical 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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Page 9-2 17. ANP-10298(P)(A)

Revision 0 Supplement 1P Revision 0, Improved K-factor Model for ACE/ATRIUM IOXM Critical Power Correlation, AREVA NP, December 2011.18. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications, GE Nuclear Energy, August 1996.19. BAW-1 0255PA Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, 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 Licensing Calculations 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 STAIF with 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:

A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.26. XN-NF-80-19(P)(A)

Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.27. EMF-2158(P)(A)

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

Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.28. 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 for ATRIUM T M IOXM Fuel, AREVA NP, May 2013.30. ANP-3212(P)

Revision 0, Monticello EPU LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM T M 1OXM Fuel, AREVA NP, May 2013.31. 0000-0043-8325-SRLR Revision 1, Cycle 27 Extended Power Uprate Supplemental Reload Licensing Report, Global Nuclear Fuel, February 2013.AREVA NP Inc.

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Fuel Transition Cycle 28 Revision 1 Reload Licensing Analysis (EPU/MELLLA)

Page 9-3 32. XN-NF-80-19(P)(A)

Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors -Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.33. XN-NF-80-19(P)(A)

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

Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.34. EMF-2361(P)(A)

Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001.35. General Electric 1OCFR Part 21 Communication, Potential Violation of Low Pressure Technical Specification Safety Limit, SC05-03, March 22, 2005.36. Letter, D.J. Mienke (Xcel Energy) to Tony Will (AREVA NP), "Transmittal of Requested Monticello 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 Criticality Safety Analysis for A TRIUM T M IOXM Fuel, AREVA NP, August 2012.38. 51-9187384-000, "Monticello Plant RPV Seismic Assessment with ATRIUM T M 1OXM Fuel," AREVA NP, September 2012 (RJW:12:022).

39. Letter, D.J. Mienke (Xcel Energy) to Tony Will (AREVA NP), "Transmittal of Requested Monticello 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 Amendment Request for Fuel Storage Changes," L-MT-12-076, October 30, 2012 (ADAMS accession no. ML12307A433).

41. NEDO-32047, A TWS Rule Issues Relative to BWR Core Thermal-Hydraulic Stability, DRF A13-00302, GE Nuclear Energy, February 1992.AREVA NP Inc.

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Page A-1 Appendix A Operating Limits and Results 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.

AREVA NP Inc.

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Revision 1 Page A-2 MONT CY28 EOFPLBNSS (A10XM 16175.0 Fuel)DSS/NSS/TSSS 4.0 3.5 3.0-j Q_C-)2.5 I I I I I I I I* FWCF o HPCILOFWH+ LRNB x RUNUP* TTNB v TTWB V V x [Aa x 2.0 1.5 1.0 0 10 20 30 40 50 60 70 Power (% 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 EOFP Power-Dependent MCPR Limits for ATRIUM 1OXM Fuel NSS Insertion Times Base Case Two-Loop Operation (TLO)AREVA NP Inc.

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Revision 1 Page A-3 MONT CY28 EOFPLBNSS (GEl4 16175.0 Fuel)DSS/NSS/TSSS 4.0 3.5 3.0-t E.0j a-o FWCF o HPCI* LOFWI+ LRNB x RUNUF* TTNB v TTWB 0++x ~ ~g -+ *.+H 2.5 2.0 1.5 1.0 a A x A x 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 40.05 50%F 2.38 25.0 < 50%F 3.23 Figure A.2 BOC to Licensing Basis EOFP Power-Dependent MCPR Limits for GE14 Fuel NSS Insertion Times Base Case Two-Loop Operation (TLO)AREVA NP Inc.

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Revision 1 Page A-4 MONT CY28 CoostNSS 21175.0 (A1OXM Fuel)DSS/NSS/TSSS 4.0 3.5 3.0 I I o] FWCF o HPCI A LOFWH+ LRNB x RUNUP 0 TTNB v TTWB-t E_-Q_(D 2.5 2.0 1.5 1.0+V V 0 x 0+A A A A+A x 0 10 20 30 40 50 60 70 Power (% 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 Coastdown Power-Dependent MCPR Limits for ATRIUM 10XM Fuel NSS Insertion Times Base Case Two-Loop Operation (TLO)AREVA NP Inc.

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Revision 1 Page A-5 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 0 0+V V x 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 Figure A.4 BOC to Coastdown Power-Dependent MCPR Limits for GE14 Fuel NSS Insertion Times Base Case Two-Loop Operation (TLO)AREVA NP Inc.

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Revision 1 Page A-6 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 60 Power (% 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 Figure A.5 BOC to Licensing Basis EOFP Power-Dependent MCPR Limits for ATRIUM 1OXM Fuel TSSS Insertion Times Base Case Two-Loop Operation (TLO)AREVA NP Inc.

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Revision 1 Page A-7 MONT CY28 EOFPLBTSSS 16175.0 (GE 14 Fuel)DSS/TSSS 4.0 3.5 3.0-t a-)2.5 o] FWCF o HPCI A LOFWH+ LRNB x RUNUP o TTNB v TTWB 0 0+0 4o-A xx 2.0 1.5 1.0 0 10 20 30 40 50 60 Power (% 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 Figure A.6 BOC to Licensing Basis EOFP Power-Dependent MCPR Limits for GE14 Fuel TSSS Insertion Times Base Case Two-Loop Operation (TLO)AREVA NP Inc.

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Revision 1 Page A-8 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 HPCI A LOFWH+ LRNB x RUNUP o TTNB V TTWB x x 0 10 20 30 40 50 60 70 Power (% 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 Figure A.7 BOC to Coastdown Power-Dependent MCPR Limits for ATRIUM 1OXM Fuel TSSS Insertion Times Base Case Two-Loop Operation (TLO)AREVA NP Inc.

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Revision 1 Page A-9 MONT CY28 CoastTSSS 211 (GE14 Fuel)75.0 DSS/TSSS 4.0 3.5 3.0-t E~_j a)2.5 III .IIII o FWCF o HPCI* LOFWH+ LRNB x RUNUP* TTNB v TTWB 0 0+V 0 8 xx 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 Figure A.8 BOC to Coastdown Power-Dependent MCPR Limits for GE14 Fuel TSSS Insertion Times Base Case Two-Loop Operation (TLO)AREVA NP Inc.

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Revision 1 Page A-10 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 HPCI A LOFWH+ LRNB x PRFDS 0 RUNUP D 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 90 100 Power (% 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 Figure A.9 BOC to Coastdown Power-Dependent MCPR Limits for ATRIUM 1OXM Fuel NSS/TSSS Insertion Times PROOS Two-Loop Operation (TLO)AREVA NP Inc.

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Revision 1 Page 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 HPCI A LOFWH+ LRNB x PRFDS 0 RUNUP v TTNB 0 TTWB 0 0 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 (%60 Rated)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 Figure A.10 BOC to Coastdown Power-Dependent MCPR Limits for GE14 Fuel NSSITSSS Insertion Times PROOS Two-Loop Operation (TLO)AREVA NP Inc.

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

ANP-3213(NP)

Revision 1 Page A-12 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* RUNUP 0 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 90 100 Power (% 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 Figure A.11 BOC to Coastdown Power-Dependent MCPR Limits for ATRIUM 1OXM Fuel NSS/TSSS Insertion Times Base case + PROOS Single-Loop Operation (SLO)** Operation in SLO is not allowed above 66% of rated power.AREVA NP Inc.

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

ANP-3213(NP)

Revision 1 Page A-13 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 RUNUP v TTNB* TTWB 0 X SLPS+ x 011 0 o 8 0£I I I Ii i i 2.0 1.5 1.0 0 10 20 30 40 50 Power 60 (% 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 Figure A.12 BOC to Coastdown Power-Dependent MCPR Limits for GE14 Fuel NSS/TSSS Insertion Times Base case + PROOS Single-Loop Operation (SLO)** Operation in SLO is not allowed above 66% of rated power.AREVA NP Inc.

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

ANP-3213(NP)

Revision 1 Page A-14 MONT CY28 LHGRFACp Base (AT 1OXM Case COAST Fuel)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 Exposures Power-Dependent LHGR Multipliers for ATRIUM 1OXM Fuel NSSITSSS Insertion Times Base Case Two-Loop Operation (TLO) and Single-Loop Operation (SLO)AREVA NP Inc.

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

ANP-3213(NP)

Revision 1 Page A-15 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 LOFWH RUNUP 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 (% of rated) Multiplier 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 Figure A.14 All Exposures Power-Dependent LHGR Multipliers for GE14 Fuel NSS/TSSS Insertion Times Base Case Two-Loop Operation (TLO) and Single-Loop Operation (SLO)* 0.97 setdown required if analytical setpoint is greater than 114% (based on CRWE calculation).

AREVA NP Inc.

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

ANP-3213(NP)

Revision 1 Page A-16 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 (% of rated) Multiplier 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 Figure A.15 All Exposures Power-Dependent LHGR Multipliers for ATRIUM 1OXM Fuel NSS/TSSS Insertion Times PROOS Two-Loop Operation (TLO) and Single-Loop Operation (SLO)AREVA NP Inc.

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

ANP-3213(NP)

Revision 1 Page A-17 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 80 30 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 Figure A.16 All Exposures Power-Dependent LHGR Multipliers for GE14 Fuel NSS/TSSS Insertion Times PROOS Two-Loop Operation (TLO) and Single-Loop Operation (SLO)* 0.97 setdown required if analytical setpoint is greater than 114% (based on CRWE calculation).

AREVA NP Inc.