Text

Enclosure 2 Areva Report ANP-3295NP, Rev. 3, Monticello Licensing Analysis for EFW (Epu/Mella+).
	ML16063A034
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Site:	Monticello
Issue date:	02/29/2016
From:	; AREVA
To:	; Office of Nuclear Reactor Regulation
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ML16063A032	List: L-MT-16-010, Enclosure 2 Areva Report ANP-3295NP, Rev. 3, Monticello Licensing Analysis for EFW (Epu/Mella+).; ML16063A033;
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L-MT- 16-010 Enclosure 2 ARE VA Report ANP-3295NP Non-Proprietary Monticello Licensing Analysis for EFW (EPU/MELLLA+)

Revision 3 February 2016 134 pages follow

Controlned Document A

ARE VA AN P-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+)

ControD~ed Document AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For PaNe EFW (EPU/MELLLA+)

Nature of Changes Item Page Description and Justification

1. 4-2 APRM reduction updated in response to CR 2015-7013 and CR 201 5-7948.

2. 5-19 Low power range (30% power) MCPR values revised in response to CR 2015-7013, CR 2015-7455, and CR 2015-7948.

Changes are also revision identified bar in by a vertical the right-hand margin.line ( I)

AREVA Inc.

ControU~ed Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page ii Contents 1.0 Introduction................................................................................... 1-1 2.0 Disposition of Events ........................................................................ 2-1 3.0 Mechanical Design Analysis ................................................................ 3-I 4.0 Thermal-Hydraulic Design Analysis ........................................................ 4-1 4.1 Thermal-Hydraulic Design and Compatibility ...................................... 4-1 4.2 Safety Limit MCPR Analysis....................................................... .4-1 4.3 Core Hydrodynamic Stability........................................................ 4-2 4.4 Voiding in the Channel Bypass Region ............................................ 4-3 5.0 Anticipated Operational Occurrences ...................................................... 5-1 5.1 System Transients................................................................... 5-1 5.1.1 Load Rejection No Bypass (LRNB)....................................... 5-3 5.1.2 Turbine Trip No Bypass (TTNB) .......................................... 5-3 5.1.3 Pneumatic System Degradation - Turbine Trip With Bypass and Degraded Scram (TTWB)................................... 5-3 5.1.4 Feedwater Controller Failure (FWCF).................................... 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 Withdrawal Error ............................................ 5-6 5.1.8 Fast Flow Runup Analysis ................................................ 5-6 5.2 Slow Flow Runup Analysis.......................................................... 5-7 5.3 Equipment Out-of-Service Scenarios .............................................. 5-8 5.3.1 Single-Loop Operation .................................................... 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 (LOCA).................................................. 6-1 6.2 Pump Seizure Accident ............................................................. 6-2 6.3 Control Rod Drop Accident (CRDA)................................................ 6-2 6.4 Fuel and Equipment Handling Accident............................................ 6-3 6.5 Fuel Loading Error (Infrequent Event).............................................. 6-3 6.5.1 Mislocated Fuel Bundle ................................................... 6-3 6.5.2 Misoriented Fuel Bundle .................................................. 6-4 7.0 Special Analyses............................................................................. 7-1 7.1 ASME Overpressurization Analysis ................................................ 7-1 7.2 Anticipated Transient Without Scram Event Evaluation........................... 7-2 7.2.1 Overpressurization Analysis .............................................. 7-2 7.2.2 Long-Term Evaluation..................................................... 7-3 7.3 Reactor Core Safety Limits - Low Pressure Safety Limit, Pressure Regulator Failed Open 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 Inc.

Controiied Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page iii 8.0 Operating Limits and COLR Input........................................................... 8-I 8.1 MCPR Limits......................................................................... 8-1 8,2 LHGR Limits ......................................................................... 8-1 8,3 MAPLHGR Limits.................................................................... 8-2 9.0 References ................................................................................... 9-1 Appendix A Operating Limits and Results Comparisons ..................................... A-I Tables Table 1.1 EOD and EQOS Operating Conditions ................................................. 1-3 Table 2.1 Disposition of Events Summary ......................................................... 2-3 Table 2.2 Disposition of Operating Flexibility and EOOS Options on Limiting Events.................................................................................. 2-22 Table 2.3 Methodology and Evaluation Models for Cycle-Specific Reload Analyses............................................................................... 2-23 Table 4.1 Thermal-Hydraulic Results at Rated Conditions (100%P / 80%F).................... 4-5 Table 4.2 Thermal-Hydraulic Results at Off-Rated Conditions (82.5%P I 57.4%F) ............. 4-6 Table 4.3 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses ............. 4-7.

Table 4.4 Results Summary for Safety Limit MCPR Analyses.................................... 4-8 Table 4.5 Channel Instability Exclusion Region Endpoints ....................................... 4-9 Table 4.6 OPRM Setpoints ........................................................................ 4-10 Table 4.7 BSP Endpoints for Monticello Cycle 28 ............................................... 4-11 Table 4.8 Maximum Bypass Voiding at LPRM Level D.......................................... 4-12 Table 5.1 Effect of EFW on Transient Analyses - Comparison of Transient Results for Technical Specifications Scram Speed (TSSS) ........................ 5-1 1 Table 5.2 Exposure Basis for Monticello Cycle 28 Transient Analysis ......................... 5-12 Table 5.3 Scram Speed Insertion Times ......................................................... 5-13 Table 5.4 Licensing Basis EOFP Base Case LRNB Transient Results......................... 5-14 Table 5.5 Licensing Basis EOFP Base Case TTNB Transient Results......................... 5-15 Table 5.6 Licensing Basis EOFP Base Case TTWB Transient Results ........................ 5-16 Table 5.7 Licensing Basis EOFP Base Case FWCF Transient Results ........................ 5-17 Table 5.8 Licensing Basis EOFP Base Case HPCl Transient Results ......................... 5-18 Table 5.9 Licensing Basis EOFP Base Case CRWE Results for TLO ......................... 5-19 Table 5.10 RBM Operability Requirements ...................................................... 5-20 Table 5.11 Licensing Basis EOFP PRFDS (PROOS) Transient Results....................... 5-21 AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page iv Table 5.12 Licensing Basis Core Average Axial Power Profile ................................. 5-22 Table 6.1 Initial Conditions.......................................................................... 6-5 Table 6.2 Summary of TLO Recirculation Line Break Results Highest PCT Cases.................................................................................... 6-6 Table 7.1 ASME Overpressurization Analysis Results............................................ 7-7 Table 7.2 ATWS Overpressurization Analysis Results............................................ 7-8 Table 8.1 MCPRp Limits for Two-Loop Operation (TLO), NSS Insertion Times BOC to Licensing Basis EOFP ........................................................ 8-3 Table 8.2 MCPRp Limits for Two-Loop Operation (TLO), TSSS Insertion Times BOC to Licensing Basis EOFP ........................................................ 8-4 Table 8.3 MCPRp Limits for Two-Loop Operation (TLO), NSS Insertion Times BOC to Coastdown..................................................................... 8-5 Table 8.4 MCPRp Limits for Two-Loop Operation (TLO), TSSS Insertion Times BOC to Coastdown .......................................... i.......................... 8-6 Table 8.5 MCPRp Limits for Single-Loop Operation (SLO), TSSS insertion Times BOC to Coastdown'. ................................................................... 8-7 Table 8.6 Flow-Dependent MCPR Limits ATRIUM 10XM and GEl4 Fuel, NSS/TSSS Insertion Times, TLO and SLO, PROOS All Cycle 28 Exposures............................................................................... 8-8 Table 8.7 ATRIUM 10XM Steady-State LHGR Limits............................................. 8-9 Table 8.8 ATRIUM 10XM LHGRFACp Multipliers for NSS/TSSS Insertion Times, TLO and SLO, All Cycle 28 Exposures .............................................. 8-10 Table 8.9 GEI4 LHGRFACp Multipliers for NSS/TSSS Insertion Times, TLO and SLO, All Cycle 28 Exposures ........................................................ 8-11 Table 8.10 ATRIUM 10XM LHGRFACf Multipliers, NSS/TSSS Insertion Times, TLO and SLO, PROOS, All Cycle 28 Exposures.................................... 8-12 Table 8.11 GEl4 LHGRFACf Multipliers, NSSITSSS Insertion Times, TLO and SLO, PROOS, All Cycle 28 Exposures .............................................. 8-13 Table 8.12 ATRIUM 10XM MAPLHGR Limits, TLO.............................................. 8-14 Figures Figure 1.1 Monticello Power/Flow Map - EPU/EFW .............................................. 1-4 Figure 5.1 Licensing Basis EOFP LRNB at I00P/80F -TSSS Key Parameters............... 5-23 Figure 5.2 Licensing Basis EOFP LRNB at 100OP/80F - TSSS Vessel Pressures............. 5-24 Figure 5.3 Licensing Basis EOFP TTNB at 100P/80F - TSSS Key Parameters............... 5-25 Figure 5.4 Licensing Basis EOFP TTNB at 100P/80F - TSSS Vessel Pressures ............. 5-26 Figure 5.5 Licensing Basis EOFP FWCF at I00P/80F - TSSS Key Parameters.............. 5-27 AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page v Figure 5.6 Licensing Basis EOFP FWCF at 100P/80F - TSSS Vessel Pressures ............ 5-28 Figure 5.7 Licensing Basis EOFP HPCI at 100P/80F - TSSS Key Parameters ............... 5-29 Figure 5.8 Licensing Basis EOFP HPCI at 100OPI80F- TSSS Vessel Pressures ............. 5-30 Figure 7.1 MSIV Closure Overpressurization Event at 102P/80F - Key Parameters ............................................................................. 7-9 Figure 7.2 MSIV Closure Overpressurization Event at 102P/80F -Vessel Pressures.............................................................................. 7-10 Figure 7.3 MSIV Closure Overpressurization Event at 102P/80F - Safety/Relief Valve Flow Rates ..................................................................... 7-11 Figure 7.4 PRFO ATWVS Overpressurization Event at 102P/80F - Key Parameters............................................................................ 7-12 Figure 7.5 PRFO ATWS Overpressurization Event at 102P/80F -Vessel Pressures.............................................................................. 7-13 Figure 7.6 PRFO ATVVS Overpressurization Event at 102P/80F - Safety/Relief Valve Flow Rates ..................................................................... 7-14 AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page vi EFW (EPU/MELLLA+)

Nomenclature 2PT two pump trip ADS automatic depressurization system AOO anticipated operational occurrence APLHGR average planar linear heat generation rate APRM average power range monitor ARO all control rods out ASME American Society of Mechanical Engineers AST alternative source term ATWS anticipated transient without scram ATWS-I anticipated transient without scram instability ATWS-PRFO anticipated transient without scram pressure regulator failure open ATWS-RPT anticipated transient without scram recirculation pump trip BOO beg inning-of-cycle BPWS banked position withdrawal sequence BSP backup stability protection BWR boiling water reactor BWROG Boiling Water Reactor Owners Group CFR Code of Federal Regulations COLR core operating limits report CPR critical power ratio CRDA control rod drop accident CRWE control rod withdrawal error DIVOM delta-over-initial CPR versus oscillation magnitude DSS degraded scram speed ECCS emergency core cooling system EFPH effective full-power hour EFW extended flow window EOC end-of-cycle EOD extended operating domain EOFP end of full power E00S equipment out-of-service EPU extended power uprate FW feedwater FWCF feedwater controller failure GE General Electric GNF Global Nuclear Fuels HCOM hot channel oscillation magnitude HFCL high flow control line HFR heat flux ratio HPCI high pressure coolant injection ICF increased core flow AREVA Inc.

Contronled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page vii EFW (EPU/MELLLA+)

Nomenclature (continued)

LAR license amendment request LFWH loss of feedwater heating LHGR linear heat generation rate LHGRFACf flow-dependent linear heat generation rate multipliers LHGRFACp power-dependent linear heat generation rate multipliers LOCA loss-of-coolant accident LPRM local power range monitor LRNB generator load rejection with no bypass MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio MCPRf flow-dependent minimum critical power ratio MCPRp power-dependent minimum critical power ratio MELLLA maximum extended load line limit analysis MELLLA+ maximum extended load line limit analysis plus MNGP Monticello Nuclear Generating Plant MSIV main steam isolation valve NCL natural circulation line NRC Nuclear Regulatory Commission, U.S.

NSS nominal scram speed OLMCPR operating limit minimum critical power ratio O LTP original licensed thermal power 00S out of service OPRM oscillation power range monitor Pbypass power below which direct scram on TSV/TCV closure is bypassed PCT peak cladding temperature PRFDS pressure regulator failure down-scale PRFO pressure regulator failure open PROOS pressure regulator out-of-service PUSAR Power Uprate Safety Analysis Report RBM (control) rod block monitor RHR residual heat removal SLC standby liquid control SLCS standby liquid control system SLMCPR safety limit minimum critical power ratio SLO single-loop operation SLPS single-loop pump seizure SRV safety/relief valve SRVOOS safety/relief valve out-of-service SS steady-state AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+i) Page viii Nomenclature (continued)

TBV turbine bypass valves TCV turbine control valve TIP traversing incore probe TI P0OS 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 Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 1-1 1.0 Introduction The licensing analyses described herein were generated by AREVA Inc. to support Monticello Nuclear Generating Plant (MNGP) operation with ATRIUMTM 10XM* fuel at Extended Power Uprate (EPU) and Extended Flow Window (EFW) conditions. EFW is the AREVA term used to denote the increased core flow window also known as Maximum Extended Load Line Limit Analysis Plus (MELLLA+). AREVA has performed licensing calculations previously to support MNGP to transition to AREVA ATRIUM 10XM fuel under EPU conditions (Reference 1, 2, 3, 4, 5, 6, 7 and 8).

The applicability of currently approved AREVA codes and methods to EFW is addressed in Reference 9. In support of the first application of ARE VA methods for EFW, a methodology for analyzing the fuel specific impact of ATRIUM 10OXM fuel on Anticipated Transients Without Scram (ATWS) and Instability (ATWS-I) is presented in Reference 10 and the results from the application of this methodology for Monticello are presented in Reference 11. These analyses together with the current report form the License Amendment Request (LAR) package addressing the EFW for ATRIUM 10OXM fuel using ARE VA's methodologies. The analyses presented herein were performed using methodologies previously approved for generic application to boiling water reactors with some exceptions which are explicitly described in this 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 12. Although the first reload of ARE VA fuel has subsequently been delayed until Cycle 29, Cycle 28 remains the representative first transition cycle for Monticello fuel transition and EFW LAR. The representative core design consists of a total of 484 fuel assemblies, including [ ] fresh ATRIUM 10XM assemblies and [ ] irradiated GEl4 assemblies. The analyses are prepared to be the best representation of the proposed MNGP configuration (i.e. EPU at EFW). The Cycle 28 core design was used in this process as a representative design. 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 the 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

ATRIUM is a trademark of AREVA Inc.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 1-2 transition or the needed revisions are identified. The representative core design for these analyses assures the actual 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 licensing analysis support operation for the power/flow map presented in Figure 1.1 and also support operation with the equipment out-of-service (EOOS) scenarios presented in Table 1.1.

AREVA Inc.

ControR~ed Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 1-3 Conditions (Opeatn Extended Operating Domain (EOD) Conditions Increased core flow (ICF)

Extended Flow Window (EFW)

Coastdown Equipment Out-of-Service (EO00S) Conditions*

Pressure regulator out-of-service (PROOS)

Single-loop operation (SLO)t

SLO may be combined with the other EOOS conditions. Base case and each EOOS condition is supported in combination with up to I 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 LPRM calibration interval of 1000 MWd/ST average core exposure.

t SLO is not allowed in EFW operating conditions AREVA Inc.

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Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (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 GEI4 and ATRIUM I0XM 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 13), the Technical Specifications (References 14 and 15),

Core Operating Limits Report (COLR), and other reload analysis reports. The licensing basis for EPU and MELLLA+ operation is obtained from References 16 (and supplements) and 17. In References 18 and 19, the NRC issued operating licenses for EPU and MELLLA+ respectively.

References 2 and 9 provide the applicability of ARE VA BWR methods to extended power and extended flow window operating domain at Monticello.

ARE VA 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 (AQO) 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 and transition to EFW conditions 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, transition to EFW conditions 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:

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 2-2

- The analysis is performed using conservative bounding assumptions and inputs such that the initial licensing analysis results for EFW will remain applicable for following reloads of the same fuel design (ATRIUM 10XM).

- 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 of each event for transitioning to EFW conditions under AREVA methodologies, and any applicable comments.

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

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) 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 10XM fuel to demonstrate that this Characteristics design occur - fuel design is compatible with the expected Address for initial coresident fuel (Reference 5). Analyses were licensing analysis. performed at EPU and EFW conditions.

Results for EFW conditions are provided in Section 4.1.

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 12) 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 12.

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 operation of ATRIUM 10XM fuel at EFW conditions 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 3).

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

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 2-4 EFW (EPU/MELLLA+)

Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 3.5 Reactivity Address for initial The operation of ATRIUM 10OXM fuel at EFW Control licensing analysis. conditions will have no impact on the ability of Mechanical the control rods to perform their normal and Characteristics scram functions (Reference 3). No adverse effects are anticipated in regard to channel bow for ATRIUM 10OXM fuel operating at EFW conditions.

3.6 Other reactor Address for initial Analysis was performed for the initial reload to App. A vessel internals licensing analysis. determine the effect of the mechanical loads introduced with ATRIUM 10XM fuel at EFW conditions on other reactor vessel internals (Reference 20).

4.0 Reactor Coolant See below.

System 4.2 Reactor Vessel No further analyses The operation of ATRIUM 10OXM fuel at EFW required, conditions 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 10OXM design that would force a significant change in the power distribution or core management scheme.

4.3 Reactor Address each reload. Analyses performed each reload to Recirculation demonstrate compliance with the ASME System 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).

4.4 Reactor Pressure Address each reload. This event assures compliance with the ASME Relief System code (Section 7.1).

Overpressuri-zation Protection 4.5 Reactor Coolant No further analyses Analysis of record shows compliance with the System Vents required. licensing requirements. The operation of ATRIUM I0XM fuel at EFW conditions does not affect the normal operation of this system.

4.6 Hydrogen Water No further analyses The hydrogen water chemistry is independent Chemistry required, of the operation of ATRIUM 10XM at EFW conditions. MNGP provides water chemistry data to AREVA to assess the impact of crud/corrosion on licensing analyses.

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 2-5 EFW (EPUIMELLLA+)

Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 4.7 Zinc Water Address for initial The zinc water chemistry is independent of the Chemistry and reload and assess the operation of ATRIUM 10OXM at EFW conditions.

OLNC (On-Line impact of subsequent MNGP will provide water chemistry data to NobleChem) changes in water AREVA to assess the impact of crud/corrosion chemistry for on licensing analyses.

follow-on reloads.

5.0 Containment See below.

System 5.2 Primary No further analyses For events with scram (i.e. LOCA) the only fuel Containment required, dependent characteristic important to the System containment response is decay heat. Since the ATRIUM 10XM fuel decay heat is similar to that of the GEl4 fuel, the analysis of record results for MELLLA+ remain applicable for the operation of ATRIUM 10OXM fuel at EFW conditions.

For events without scram (i.e. ATWS) the fuel dependent characteristics important to the containment response are void coefficient and boron worth. Based on comparison of these fuel characteristics (refer to Section 7.2.2), the analysis of records results for MELLLA+

remain applicable for the operation of ATRIUM 1OXM fuel at EFW conditions.

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 results for EFW Performance licensing analysis. conditions are presented in Section 6.1. The results show that low core flow conditions specific to EFW are not limiting for LOCA.

The limiting break spectrum analyses were performed for the initial licensing analysis (Reference 6).

Heatup/MAPLHGR analyses (Reference 7) performed each reload for any new nuclear fuel design.

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 2-6 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 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 10XM fuel in MNGP.

Addressed under the LOCA analysis.

The main steam line break outside the primary containment was considered in the identification of the spectrum of loss-of-coolant accident events and it is bounded by the limiting loss-of-coolant accident scenario (Reference 6). The operation of ATRIUM 10XM at EFW conditions does not cause a break in the main steam line to become more limiting than a break in the recirculation pipe.

6.4 Control Rod No further analysis The operation of ATRIUM 10OXM fuel at EFW Velocity Limiters required. conditions will 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 operation of ATRIUM 10OXM fuel at EFW Drive Housing required. conditions will have no impact on the ability of Supports the control rods 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).

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 10XM 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 10OXM fuel is not Support Center significantly different than legacy fuel (GEl4).

Habitability Further, ATRIUM 10XM 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 10XM inventories.

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 2-7 EFW (EPU/MELLLA+)

Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 7.0 Plant lnstru- 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 5.1.8 and 5.2).

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 AOOs 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 and EFW conditions.

Bypass voiding impact on LPRM readings have Address each reload been evaluated in the analysis of the OPRM for bypass boiling and APRM systems (Sections 4.3 and 4.4).

impact. 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.

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 2-8 EFW (EPU/MELLLA+)

Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 7.5 Plant Radiation No further analysis The operation of ATRIUM 10XM fuel at EFW Monitoring required. conditions will have no impact on the plant Systems radiation 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).

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 operation of ATRIUM 10XM fuel at EFW Performance conditions.

Instrumentation Systems 7.11 Reactor No further analysis Reactor shutdown capability is not affected by Shutdown required, the operation of ATRIUM 10XM fuel at EFW Capability conditions.

7.12 Detailed Control No further analysis Control room design is not affected by the Room Design required. operation of ATRIUM 10XM fuel at EFW Review conditions.

7.13 Safety Parameter No further analysis Safety parameter display system is not Display System required. affected by the operation of ATRIUM 10XM fuel at EFW conditions.

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 2-9 EFW (EPU/MELLLA+)

Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 8.0 Plant Electrical See below.

Systems 8.2 Transmission No further analysis Transmission system is not affected by the System required. operation of ATRIUM 10XM fuel at EFW conditions.

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 10XM fuel with the AREVA methodology (Reference 6). Results for EFW conditions LOCA break spectrum analyses are presented in Section 6.1.

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 10XM fuel with the AREVA methodology (Reference 6). Results for EFW conditions LOCA break spectrum analyses are presented in Section 6.1.

8.6 Reactor No further analysis The power supplies for reactor protection Protection required. system are not affected by the operation of System Power ATRIUM 10XM fuel at EFW conditions.

Supplies 8.7 Instrumentation No further analysis These systems are not affected by the and Control AC required. operation of ATRIUM 10OXM fuel at EFW Power Supply conditions.

Systems 8.8 Electrical Design No further analysis Independent of fuel design. Analysis of record Considerations required. remains valid for operation of ATRIUM 10XM fuel at EFW conditions.

8.9 Environmental No further analysis Independent of fuel design. Analysis of record Qualification of required. remains valid for operation of ATRIUM 10XM Safety-Related fuel at EFW conditions.

Electrical Equipment AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 2-10 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 8.10 Adequacy of No further analysis Independent of fuel design. Analysis of record Station Electrical required. remains valid for operation of ATRIUM 10OXM Distribution fuel at EFW conditions.

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 10XM fuel by comparing the decay heat for ATRIUM 10XM fuel to the decay heat used in the station blackout analysis of record. Since the ATRIUM 10XM fuel decay heat is expected to be similar to that of the GEl4 fuel, the analysis of record results for MELLLA+ remain applicable for the operation of ATRIUM 10OXM fuel at EFW conditions.

9.0 Radioactive No further analyses As shown by radiological source term Waste required, evaluations, the ATRIUM 10OXM fuel is not Management significantly different than legacy fuel.

ATRIUM 10OXM 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.

10.0 Plant Auxiliary See below.

Systems 10.2 Reactor Auxiliary No further analyses Independent of operation of ATRIUM 10OXM Systems required (except see fuel at EFW conditions. Analysis of record below), remains valid.

10.2.1 Fuel Storage and Address for initial Evaluation of k-eft for normal and abnormal Fuel Handling licensing analysis. conditions for spent fuel pool storage racks has Systems been performed generically for the ATRIUM 10XM fuel design (Sections 6.4 and 7.6). The assumptions made in the Reference 8 criticality evaluation have been reviewed relative to EFW operation. It is concluded that Monticello criticality evaluation (Reference 8) is also valid when operating under EFW conditions (see Section 7.6).

AREVA Inc_

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 2-11 EFW (EPU/MELLLA+)

Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 10.3 Plant Service No further analyses Independent of operation of ATRIUM 10XM Systems required (except see fuel at EFW conditions.

below). Analysis of record remains valid.

10.3.1 Fire Protection Address for initial The operation of ATRIUM 10XM fuel at EFW System licensing analysis. conditions will be evaluated to demonstrate that no clad damage occurs for Appendix R (Section 7.4).

10.4 Plant Cooling No further analyses Independent of operation of ATRIUM 10XM System required (except see fuel at EFW conditions.

below). Analysis of record remains valid.

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 10XM fuel in MNGP (Reference 6).

The decay heat removal design basis of the RHR system is not altered by the operation of ATRIUM 10XM fuel in at EFWV conditions.

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 operation of ATRIUM 10OXM fuel at EFW conditions (Section 5.0).

12.0 Plant Structures No further analyses Independent of operation of ATRIUM 10XM and Shielding required. fuel at EFW conditions. Analysis of record remains valid.

13.0 Plant Operation Address for initial Organization, Responsibilities, and licensing analysis. Qualifications of staff personnel are not affected by operation of ATRIUM 10OXM fuel at EFW conditions. 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 10OXM fuel. The overall nuclear site organization and plant functional organization are not affected by the operation of AREVA fuel at EFW conditions.

14.0 Plant Safety See below.

Analysis AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 2-12 Table 2.1 Disposition of Events Summary (continued)

USAIR Design Disposition Sect. Criteria Status Comment 14.2 MCPR Safety Address each reload. Part of the safety licensing analysis evaluated Limit for 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 for the whole power/flow map (including EFW conditions). In addition MAPLHGR limits will be established and verified each cycle for the ATRIUM 10OXM 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 was analyzed in support of the fuel transition. Since the results show this is a potentially limiting event, this event will be analyzed each reload (Section 5.1.6).

14.4.3 Control Rod No further analysis Consequences of a CRWE below the low Withdrawal Error required. power setpoint are bound by the CRWE at

- low power power due to required strict compliance with BPWS.

14.4.3 Control Rod Address each reload. Analysis to determine the change in MCPR Withdrawal Error and LHGR as a function of RBM setpoint will

- at power 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).

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 AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 2-13 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 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.

This event will be analyzed each reload (Section 5.1.2).

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 fuel transition analysis addressed Compliance MSIV, TCV, and TSV closures under AREVA Model - MSIV methodology. Since the limiting valve closure Closure is MSIV, only this will be analyzed 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 12).

14.6 Plant Stability Address each reload. Enhanced Option Ill will be implemented with Analysis the transition to AREVA methods. DIVOM and initial MCPR will be analyzed on a cycle-specific basis (Section 4.3). The Channel Instability Exclusion Regions will be determined on a cycle specific basis.

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 Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 2-14 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 were performed for Accident licensing analysis. EPU/EFW 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 was performed for the transition to ATRIUM 10XM fuel (Reference 6).

Additional analyses were performed for EFW for the point M in the power/flow map (see Figure 1.1). A summary of all LOCA break spectrum results for EFW conditions is presented in Section 6.1 MAPLHGR heatup analyses are performed every time a new neutronic design is introduced in the core (Reference 7).

14.7.3 Main Steam Line Address for initial The main steam line break was considered in Break Accident licensing analysis, the identification of the spectrum of loss-of-Analysis coolant accident events and is bounded by the limiting recirculation line break scenario (Reference 6).

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).

AREVA Inc.

C ontrolled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 2-15 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 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. This is independent of Accident operation of ATRIUM 10OXM fuel at EFW Analysis conditions.

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 be analyzed each reload (Section 5.3.1). SLO is not allowed in EFW conditions.

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 10XM Analysis assembly dropping over the core has been analyzed in support of the fuel transition. This is independent of operation of ATRIUM 10OXM

at EFW conditions (Section 6.4).

14.7.7 Accident No further analysis Independent of operation of ATRIUM 10XM Atmospheric required, fuel at EFW conditions. The values of Dispersion atmospheric dispersion coefficients in the Coefficients analysis of record remain valid.

14.7.8 Core Source Address for initial The source terms for ATRIUM 10XM fuel at Term Inventory licensing analysis. 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 10OXM fuel is not significantly different than legacy fuel (GEI4). This is independent of operation of ATRIUM 1OXM at EFW conditions.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 2-16 Table 2.1 Disposition of Events Summary (continued)

USAR Design Disposition Sect. Criteria Status Comment 14.8 Anticipated Address each reload. The peak vessel pressure is calculated for Transients each reload. In support of operating Without Scram ATRIUM 10XM at EFW conditions analyses (ATWS) were performed at maximum and minimum core flows allowed at EFW (Section 7.2).

For long-term cooling after ATWS, the decay heat is the only fuel-related input. AREVA dispositioned the impact of ATRIUM 10XM fuel by comparing the decay heat for ATRIUM 10XM fuel to the decay heat used in the ATWS long-term cooling analysis. Since the ATRIUM 10XM fuel decay heat is expected to be similar to that of the GEI4 fuel, the analysis of record results for MELLLA+ remain applicable for the operation of ATRIUM 10OXM fuel at EFW conditions. Containment heatup was dispositioned by comparing kinetics parameters for ATRIUM 10XM fuel with those for the fuel in the analysis of record.

14.9 Section deleted NA NA 14.10 Other Analyses See below.

14.10.1 Adequate Core No further analysis USAR 14.10.1 identifies the loss of feedwater Cooling for required flow event as the worst anticipated transient, Transients with and loss of a high pressure inventory makeup a Single Failure (HPCI) or heat removal system as the worst single failure.

The analysis of record for loss of feedwater flow (Section 2.8.5.2.3 of Reference 16) 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. These conclusions remain valid for the operation of ATRIUM I0XM fuel under EFW conditions.

14A Supplemental See below. The events identified in the Supplemental Reload Licensing Reload Licensing Submittal are addressed Submittal below as part of the PUSAR (Reference 16).

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 2-17 Table 2.1 Disposition of Events Summary (continued)

PUA einDisposition et. Criteria / Event Status Comment Decrease in Reactor Coolant. ..

Temperature 2.8.5.7 Pressure Regulator Address for initial Address for initial reload for EPU/EFW 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 14) 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.

The analysis of this event has shown that this event is more severe at off-rated conditions and outside of the EFW region of the power/flow map (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 Address each Consequences of this event, relative to one Failure - Closed reload. pressure regulator out-of-service may be limiting; therefore this EOOS event will be evaluated on a cycle-specific basis (Section 5.3.2).

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 2-18 Table 2.1 Disposition of Events Summary (continued)

PIJ$A

... Design Disposition Secti.i Criteria/ Event Status Comment 2.8.5.2.1 MSIV Closures No further analysis Consequences of this event (with direct required. 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 Address each position scram function is the design basis reload. overpressurization event, which is evaluated on a cycle-specific basis (Section 7.1).

The MSIV closure event is a potentially Address each limiting ATWS overpressurization event, reload, which is evaluated on a cycle-specific basis.

Analyses have shown that the PRFO used as an ATIWS initiator event is a more limiting event for ATWS overpressure limits.

(Section 7.2).

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 Analysis not impacted by transition to EFWV Flow analysis required. conditions. The consequences of this event are only dependent on the fuel decay heat.

Since the decay heat of ATRIUM 10XM fuel is similar to that of GEl4 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 Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 2-19 EFW (EPU/MELLLA+)

Table 2.1 Disposition of Events Summary (continued)

PSRDesign 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 CRWE 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 operation of ATRIUM 10XM fuel at EFW conditions 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 HPCl start-up which will be reactor coolant analyzed each reload.

inventory AREVA Inc.

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Table 2.1 Disposition of Events Summary (continued)

PImA ein ipsto St. 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 was considered in

- Outside licensing analysis the identification of the spectrum of loss-of-Containment coolant accident events and is expected to be bounded by the limiting loss-of-coolant accident scenario (Reference 6).

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 10XM fuel is not Failure significantly different than legacy fuel (GEl4). Further, ATRIUM 10XM 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.

Transition to EFW conditions will not impact this conclusion.

2.9.2 Liquid Radwaste No further The radionuclide source terms are generic System Failure analysis required. and are unaffected by the operation of ATRIUM 10XM fuel at EFW conditions.

2.9.2 Postulated No further The radionuclide source terms are generic Radioactive Releases analysis required. and are unaffected by the operation of Due to Liquid ATRIUM 10XM fuel at EFW conditions.

Radwaste Tank Failure AREVA Inc.

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Table 2.1 Disposition of Events Summary (continued)

J4$RDesign Disposition S i*ct. Criteria/ Event Status Comment

~Other Analyses 2.8.3.3 ATWS with Core Address for initial The discussion presented in Reference 21 Instability licensing analysis indicates that "Ifa new GE fuel product line or another vendor's fuel is loaded at the plant, the applicability of any generic sensitivity analyses supporting the MELLLA+

application shall be justified in the plant-specific application. If the generic sensitivity analyses cannot be demonstrated to be applicable, the analyses will be performed including the new fuel. For example, the ATWS instability analyses supporting the MELLLA+ condition are based on the GEI4 fuel response. New analyses that demonstrate the ATWS instability performance of the new GE fuel or another vendor's fuel for MELLLA+ operation shall be provided to support the plant-specific application." The results of the ATRIUM 10XM ATWS instability evaluation for EFW conditions are provided in References 10 and 11.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPUIMELLLA+) Page 2-22 Table 2.2 Disposition of Operating Flexibility and EQOS Options on Limiting Events Option Option ~Affected LimitingComn

~Event/AnalysesComn Single-loop operation SLO is not allowed in EFW operating conditions.

(SLO)

Safety/relief valves ASME All transient analyses (AOOs) and the ASME out-of-service all AO0 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/EFW All All analyses considered the increased core flow operation and extended core flow window.

AREVA Inc.

Contro~led Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 2-23 Table 2.3 Methodology and Evaluation Models for Cycle-Specific Reload Analyses Analysis Event Methodology Evaluation Acceptance Criteria

/Analysis Reference Model and Comment Thermal and Hydraulic 22 SAFLIM3D SLMCPR Criteria: < 0.1% fuel rods Design 23 COTRANSA2 experience boiling transition.

No fuel melting and maximum Transient Analyses 24 XCOBRA tasetidcdsri %

25XCBRA-TPower- and flow-dependent MCPR 26 RODEX4 and LHGR operating limits established to meet the fuel failure 27 RODEX2 criteria.

Standby Liquid Control 28 CASMO-4 SLCS Criteria: Shutdown margin of System /MICROBURN-B2 at least 0.88 %Ak/k.

ASME 23 COTRANSA2 Analyses for ASME and ATWS Overpressurization (as supplemented overpressurization.

Analysis by considerations of AP-324(P) ASME Overpressurization Criteria:

of AP-324(P) Maximum vessel pressure limit of Anticipated Transient (Reference 2,an p Without Scram App. E)) 1375 psig admaximum dome (pressurization) pressure limit of 1332 psig.

A TWS Overpressurization Criteria:

Maximum vessel pressure limit of 1500 psig.

Emergency Core 29 HUXY LOCA Criteria: 10CFR50.46.

Cooling Systems EXEM BWR-2000 Methodology.

LOCAAnalses 7 ROEX2Only heatup (HUXY) is analyzed for LOCAAnalsesthe reload specific neutronic design.

Appendix R 29 RELAX 10OCFR50 Appendix R.

Neutron Design 30 STAIF Long-Term Stability Solution Enhanced Option Ill Criteria:

Neutron Monitoring 3 AOA-A OPRM setpoints prevent exceeding System 32 CASMO-4 SLMCPR limits.

3 M3CCBUN-2 CRWE Criteria: Power-dependent MCPR and LHGR operating limits 34 established to meet the fuel failure 35 criteria.

28 Backup Stability Protection Criteria: Stability boundaries that do not exceed acceptable global, regional, and channel decay ratios as defined by the STAlF methodology.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 3-1 3.0 Mechanical Design Analysis The results of mechanical design analyses for ATRIUM 10XM fuel are presented in References 3 and 36. The fuel rod analyses use the NRC-approved RODEX4 methodology described in Reference 26. The maximum exposure limits for the ATRIUM 10XM reload fuel are:

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

GEl4 fuel assemblies have a maximum peak pellet exposure limit of [ ] GWd/MTU (Reference 37).

The fuel cycle design analyses (Reference 12) verified all fuel assemblies remain within licensed burnup limits.

The ATRIUM 10XM LHGR limits are presented in Section 8.0. The GEl4 LHGR multipliers presented in Section 8.0 ensure that the thermal-mechanical design criteria for GEl4 fuel are satisfied.

Reference 36 presents representative fuel rod thermal-mechanical analyses using the RODEX4 methodology for Cycle 28 transition cycle. The cycle design is further described in Section 1.0.

The updated analyses performed for the transition cycle demonstrate the fuel rod thermal-mechanical criteria are satisfied.

AREVA Inc.

Controlled Document AN P-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 4-1 4.0 Thermal-Hydraulic Design Analysis 4.1 Thermal-Hydraulic Design and Compatibility The ATRIUM 10XM fuel is analyzed and monitored with the ACE critical power correlation (References 38, 39, and 40). The GEl4 fuel is analyzed and monitored with the SPCB critical power correlation (Reference 41). The SPCB additive constants and additive constant uncertainty for the GEI4 fuel were developed using the indirect approach described in Reference 42.

Results of thermal-hydraulic characterization and compatibility analyses are presented in Reference 5. Analyses were performed for various state points across the power/flow map (Figure 1.1). Results from the analyses at EFW conditions are presented in Tables 4.1 and 4.2.

Analysis results demonstrate the thermal-hydraulic design and compatibility criteria are satisfied for the transition core consisting of ATRIUM 10XM and GEI4 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 22. 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 4.3. The radial power uncertainty used in the analysis includes the effects of up to 1 traversing incore probe (TIP) machine out-of-service (TI POOS) or the equivalent number of TIP channels and/or up to 50% of the LPRMs out-of-service and a LPRM calibration interval of 1000 MWd/ST average core exposure. The requirements associated with LPRM surveillance permit the frequency to be extended up to 25% of the specified frequency. This is included in the calculations through increased uncertainties for assembly radial peaking and nodal power (see Table 4.3).

Analyses were performed for the minimum and maximum core flow conditions associated with rated power for the Monticello power/flow map for EPU/EFW operation (statepoints identified as AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 4-2 "K" and "L" in Figure 1.1) and also minimum core flow at EFW conditions (statepoint identified as "M"in Figure 1.1).

In the SLMCPR analyses for Monticello MELLLA+ performed by GNF, the core flow uncertainty applied for single-loop operation was also applied to the SLMCPR analyses for the minimum core flow at rated power (point "L"in Figure 1.1) and for the minimum core flow along the MELLLA+ boundary (point "M"in Figure 1.1). This precedent was followed for the AREVA EFW SLMCPR calculations at the same core flow conditions.

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 4.4.

4.3 Core Hydrodynamic Stability Monticello has implemented Long Term Stability Solution Enhanced Option Ill (EO-llI) to support MELLLA+ operation. Reload validation has been performed in accordance with Reference 43. The EO-IlI solution consists of two components; a Channel Instability Exclusion Region (CIER), and a stability-based Operating Limit MCPR (OLMCPR).

The first component is the ClER which is protected by automatic scram. The CIER is defined to prevent operation where the channel decay ratio can approach a value of 1.0. For application with the STAIF frequency domain code, Reference 35, a channel decay ratio less than 0.80 is used to account for the code uncertainty. This constant decay ratio line is then lowered by 5%

of rated power in order to bound any normal operational variations in bundle conditions. The effect of bypass boiling on the APRM signal has been evaluated. The relative reduction in the APRM signal was calculated to be [ ] at the intersection of MELLLA and NCL for the worst case scenario (the maximum allowable number of LPRMs are out-of-service). This represents a

[ ] reduction in terms of rated power, which is less than 5% conservative bias that is already built into EO-lII Long Term Stability Solution. The endpoints of the channel exclusion region are given in Table 4.5.

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 31. This code is a coupled neutronic-thermal-hydraulic three-dimensional transient model for the purpose of determining the relationship between the relative change in CPR and the HCOM on a plant specific basis. The AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 4-3 stability-based OLMCPRs are calculated using the most limiting of the calculated change in relative CPR for a given oscillation magnitude or the generic value provided in Reference 30.

The stability-based OLMCPR is provided for two conditions as a function of OPRM amplitude setpoint in Table 4.6. The two conditions evaluated are for a postulated oscillation at 45% core flow steady-state (SS) operation and following a two recirculation pump trip (2PT) from the limiting full power operation statepoint. These conditions are part of the NRC approved conditions for stability analysis per Reference 30. 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 Oscillation Power Range Monitor (OPRM) setpoint. The results in Table 4.6 are valid for the full ICF/EFW operating domain. It was verified the EQ-Ill solution, which relies on normalized OPRM values, is not affected by LPRM miscalibration and does not require additional uncertainty factors to account for bypass voiding effects.

In cases where the OPRM system is declared inoperable, Backup Stability Protection (BSP) is provided in accordance with Reference 34. BSP curves have been evaluated using STAlF (Reference 35) to determine endpoints that meet decay ratio criteria for the BSP Region I (scram region) and Region II (controlled entry region). Stability boundaries based on these endpoints are then determined using the generic shape generating function from Reference 34.

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

4.4 Voiding in the Channel Bypass Region Bypass voiding is not significant during full power, steady-state operation, so there is no impact on the lattice local peaking or the LPRM response. However, bypass voiding is of great concern for stability analysis due to its direct impact on the fuel channel flow rates and the axial power distributions. The reduced density head in the core bypass due to boiling results in a higher bypass flow rate and consequently a lower hot channel flow rate, which when coupled with a more bottom-peaked power distribution destabilize the core through higher channel decay ratios.

AREVA accounts for the core bypass voiding by modelling in the AREVA steady-state core simulator, transient simulator, LOCA and stability codes (Reference 2). The bypass void level AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 4-4 has been evaluated throughout the cycle and the maximum bypass void value applicable to the Cycle 28 design at statepoint "M" (82.5% power and 57.4% flow) in Figure 1.1 is reported in Table 4.8.

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Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 4-5 Table 4.1 Thermal-Hydraulic Results at Rated Conditions (100%P / 80%F) *

State point corresponding to the "L"point in the power/flow map, Figure 1.1.

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ControH~ed Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 4-6 Table 4.2 Thermal-Hydraulic Results at Off-Rated Conditions (82.5%P / 57.4%F)*

State point corresponding to the "M"point in the power/flow map, Figure 1.1.

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Controlele Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 4-7 Table 4.3 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses Parameter Uncertainty Fuel-Related Uncertainties Plant-Related Uncertainties Feedwater flow rate 1.8%

Feedwater temperature 0.8%

Core pressure 0.8%

Total core flow rate TLO 2.5%

SLO 6.0%

[ ]

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Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 4-8 EFW (EPU/MELLLA+)

Table 4.4 Results Summary for Safety Limit MCPR Analyses Minimum Percentage Powr/lo Supported of Rods in Boiling

() SLMCPR Transition 100/105 TLO -1.12 [ ]

100/80 TLO -1.12 [ ]

82.5/57.4 TLO -1.12 [ ]

66/52.5 SLO -1.13 [ ]

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Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 4-9 EFW (EPU/MELLLA+)

Table 4.5 Channel Instability Exclusion Region Endpoints Location Power (%) Flow (%)

Natural Circulation 72.0 34.2 Line Extended MLL+ 100.0 45.0 Boundary Line AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 4-10 EFW (EPU/MELLLA+)

Table 4.60OPRM Setpoints OPRM OLMCPR OLMCPR Setpoint (SS) (2PT) 1.05 1.24 1.30 1.06 1.26 1.32 1.07 1.29 1.35 1.08 1.31 1.37 1.09 1.34 1.40 1.10 1.37 1.43 1.11 1.39 1.46 1.12 1.42 1.49 1.13 1.46 1.53 1.14 1.48 1.55 1.15 1.51 1.58 Acceptance Off-Rated Rated Power Criteria OLMCPR OLMCPR as at 45% Described in Core Flow Section 8.0 AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+i) Page 4-11 Table 4.7 BSP Endpoints for Monticello Cycle 28 Power Flow Endpoint (%) (%) Definition Al 72.5 44.5 Scram region boundary, high flow control line (HFCL)

B1 42.6 33.7 Scram region boundary, natural circulation line (NCL)

A2 89.3 66.1 Controlled entry region boundary, HFCL B2 28.6 31.2 Controlled entry region boundary, NCL AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 4-12 EFW (EPU/MELLLA+I)

Table 4.8 Maximum Bypass Voiding at LPRM Level D Power (%) Flow Bypass

(%) Condition Void (%)

[ ]

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Controlled Document ANP.-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) 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. Analyses supporting operation at EPU/MELLLA conditions were provided in Reference 1. Analyses results from operating at EPU/EFW are provided in this report. Most of the results did not change since the limiting results come from high core flows state points in the power/flow map.

A comparison of results for various transients operating at high core flow (105% core flow) and low core flow (80% core flow) corresponding to EFW at EPU conditions is provided in Table 5.1.

COTRANSA2 (Reference 23), XCOBRA (Reference 24), XCOBRA-T (Reference 25) and CASMO-4/MICROBURN-B2 (Reference 28) are the major codes used in the thermal limits analyses as described in the AREVA THERMEX methodology report (Reference 24) and neutronics methodology report (Reference 44). 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 27) calculations for the Monticello Cycle 28 representative core.

The ACE/ATRIUM 10XM critical power correlation (References 38, 39, and 40) is used to evaluate the thermal margin for the ATRIUM 10XM fuel. The SPCB critical power correlation (Reference 41) is used in the thermal margin evaluations for the GE14 fuel. The application of the SPCB correlation to GEl4 fuel follows the indirect process described in Reference 42.

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.

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Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-2 Reference 14 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 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 5.2.

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 5.3. The NSS MCPRp limits can only be applied ifthe 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 14) 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.

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Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-3 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 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 5.4. Responses of various reactor and plant parameters during the LRNB event initiated at 100% of rated power and 80% of rated core flow (EFW) 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 5.5. Responses of various reactor and plant parameters during the TTNB event initiated at 100% of rated power and 80%

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

5.1.3 Pneumatic System Deciradation - Turbine Trip With Bypass and Degqraded 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 AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-4 mitigate the severity of the event. The MCPRn limits for NSS and TSSS insertion times will protect this event analyzed with 088 insertion times.

TTWB 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 TTWB transient analysis ACPR results used to generate the licensing basis EOFP operating limits for both TSSS and NSS insertion times.

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 5.7 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 80% of rated core flow (EFW) 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 AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-5 steam lines is more than the mass of HPCl water being injected, the water level will be controlled and a new steady-state condition will be established. In this case the HPCl is fairly mild as a MCPR transient (similar to a loss of feedwater heating (LFWH) event). Ifthe steam flow is less than the HPCl 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)).

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

HPCl analyses were performed for a range of power/flow conditions to support generation of the thermal limits. Table 5.8 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 HPCl event initiated at 100% of rated power and 80% of rated core flow (EFW) with TSSS insertion times.

5.1.6 Loss of Feedwater Heatingq The loss of feedwater heating (LFWH) event analysis supports an assumed 95.30 F decrease in the feedwater temperature. The temperature is assumed to decrease linearly over 39 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 AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-6 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 10XM fuel and 0.19 for GEI4 fuel.

Results from LFWH at off-rated conditions are shown in the MCPRp limit and LHGRp multiplier figures in Appendix A.

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 5.9 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 10XM fuel (Section 3.0 of Reference 36). For GEl4 fuel see setdown in Table 8.9. The LHGR limits and their associated multipliers are presented in Section 8.2. Recommended operability requirements supporting unblocked CRWE operation are shown in Table 5.10, 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.

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Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-7 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.

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 10XM and GEl4 fuel. XCOBRA code 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 10XM fuel. The analysis assumes recirculation flow increases slowly along the limiting rod line to the maximum flow physically AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-8 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 10XM 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 GEl4 fuel. GEl4 LHGRFACf multipliers 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 LPRM calibration interval of 1000 MWd/ST average core exposure. The requirements associated with LPRM surveillance permit the frequency to be extended up to 25% of the specified frequency.

5.3.1 Singqle-Loop Operation AOOs 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).

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Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-9 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).

Operation in single loop is not allowed in EFW. Therefore, the SLO pump seizure results determined for EPU/MELLLA do not change for EPU/EFW.

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

5.3.2 Pressure Regqulator 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 PREDS ACPR results are presented in Table 5.11. 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 ARE VA 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 33,232 MWd/MTU) is given in Table 5.12. 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 5.12 is greater than the integrated normalized power generated in the bottom 7 nodes in the licensing basis axial power profile in Table 5.12, 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 5.12 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.

AREVA Inc.

Contro~led Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-10 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 Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-11 Table 5.1 Effect of EFW on Transient Analyses - Comparison of Transient Results for Technical Specifications Scram Speed (TSSS)

EPU EPU 80% Rated Core Event Parameter 105% Rated Core Flow Flow Peak Neutron Flux (% Rated) 421 298 Peak Heat Flux (% Rated) 124 117 LRNB Peak Vessel Pressure (psig) 1267 1257 ATRIUM 10XM ACPR 0.36 0.30 Peak Neutron Flux (% Rated) 541 388 Peak Heat Flux (% Rated) 129 124 TTNB Peak Vessel Pressure (psig) 1284 1275 ATRIUM 10XM ACPR 0.41 0.33 Peak Neutron Flux (% Rated) 470 339 Peak Heat Flux (% Rated) 127 121 TTWB*

Peak Vessel Pressure (psig) 1282 1271 ATRIUM I0XM ACPR 0.38 0.32 Peak Neutron Flux (% Rated) 536 394 Peak Heat Flux (% Rated) 133 127 FWCF Peak Vessel Pressure (psig) 1274 1264 ATRIUM 10XM ACPR 0.43 0.35 Peak Neutron Flux (% Rated) 537 396 Peak Heat Flux (% Rated) 139 133 HPCI Peak Vessel Pressure (psig) 1275 1267 ATRIUM 10XM ACPR 0.47 0.38 Peak Neutron Flux (% Rated) 115 115 LOWt Peak Heat Flux (% Rated) 115 115 Peak Vessel Pressure (psig) 1046 1036 ATRIUMI10XM ACPR 0.17 0.16

This event assumed a degraded scram speed (DSS) curve, see Table 5.3.

t A scram does not occur during this event.

AREVA Inc.

Controfiled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 5-12 Table 5.2 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 Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 5-13 Table 5.3 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 Inc.

Contronled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 5-14 Table 5.4 Licensing Basis EOFP Base Case LRNB Transient Results Power ATRIUM 10XM GE14

(% rated) ACPR ACPR TSSS Insertion Times 100 0.36 0.36 80 0.39 0.37 60 0.39 0.35 40 (above Pbypass) 0.38 0.33 40 at > 50%F (below Pbypass) 1.25 1.15 40 at < 50%F (below Pbypss) 0.95 0.92 25 at > 50%F (below Pbypass) 1.51 1.43 25 at < 50%F below (Pbypass) 1.22 1.20 NSS Insertion Times 100 0.29 0.29 80 0.34 0.34 60 0.32 0.31 40 0.30 0.26 AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-15 Table 5.5 Licensing Basis EOFP Base Case TTNB Transient Results Power ATRIUM 10XM GEl4

(% rated) ACPR ACPR TSSS Insertion Times 100 0.41 0.40 80 0.41 0.38 60 0.40 0.36 40 (above Pbypass) 0.38 0.33 40 at > 50%F (below Pbyp2 s,) 1.25 1.15 40 at < 50%F (below Pbypass) 0.95 0.92 25 at > 50%F (below Pbypass) 1.51 1.43 25 at < 50%F (below Pbypass) 1 .22 1.20 NSS Insertion Times 100 0.38 0.37 80 0.36 0.36 60 0.32 0.32 40 0.30 0.26 AREVA Inc.

Contro~led Document ANP.-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-16 Table 5.6 Licensing Basis EOFP Base Case TTWB Transient Results Power ATRIUM 10XM GEl4

(% rated) ACPR ACPR DSS Insertion Times 100 0.38 0.38 80 0.37 0.36 60 0.36 0.32 40 (above Pbypass) 0.32 0.28 40 at > 50%F (below Pbypass) 1.08 1.03 40 at < 50%F (below Pbypass) 0.82 0.80 25 at > 50%F (below Pbypass) 1.08 1.16 25 at < 50%F (below Pbypass) 0.98 1.02 AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-17 Table 5.7 Licensing Basis EOFP Base Case FWCF Transient Results Power ATRIUM 10XM GEl4

(% rated) ACPR ACPR TSSS Insertion Times 100 0.43 0.42 80 0.45 0.45 60 0.49 0.50 40 (above Pbypass) 0.62 0.65 40 at > 50%F (below Pbypass) 1.60 1.55 40 at < 50%F (below Pbypass) 1.16 1.21 25 at > 50%F (below Pbypass) 2.22 2.30 25 at < 50%F (below Pbypass) 1.92 2.07 NSS Insertion Times 100 0.40 0.38 80 0.43 0.41 60 0.48 0.47 40 0.57 0.57 AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-18 Table 5.8 Licensing Basis EOFP Base Case HPCl Transient Results Power ATRIUM 10XM GEl4

(% rated) ACPR ACPR TSSS Insertion Times 100 0.47 0.46 80 0.48 0.47 60 0.53 0.48 40 (above Pbypass) 0.59 0.54 40 at > 50%F (below Pbypass) 1 .31 1.28 40 at < 50%F (below Pbypass) 1.10 1.19 25 at > 50%F (below Pbypass) 1.56 1.67 25 at < 50%F (below Pbypass) 1.48 1.62 NSS Insertion Times 100 0.43 0.41 80 0.45 0.43 60 0.46 0.44 40 0.54 0.53 AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-19 Table 5.9 Licensing Basis EOFP Base Case CRWE Results for TLO 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 1.48 65 1.62 30 2.19 I

111 100 85 1.48 1.50 116 85 65 1.58 1.63 121 65 30 1.79 2.19 I

112 100 1.50 117 85 1.60 122 65 1.80 85 1.52 65 1.65 30 2.19 113 100 1.52 118 85 1.69 123 65 1.80 85 1.53 65 1.77 30 2.24 114 100 1.52 119 85 1.69 124 65 1.86 85 1.54 65 1.77 30 2.24 I

AREVA Inc.

Controlned Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 5-20 Table 5.10 RBM Operability Requirements Thermal Applicable Power ATRIUM 10XM /GEl4

(% rated) MCPR 2.54 TLO

>_27% and < 90% 25 L

>90% 1.77 TLO AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-21 Table 5.11 Licensing Basis EOFP PRFDS (PROOS)

Transient Results Power ATRIUM 10XM GEl4

(% rated) ACPR ACPR TSSS Insertion Times 100 0.38 0.39 85* 0.41 0.42 0.77 0.71 80 0.81 0.75 60 1.00 0.91 40 1.25 1.16 25 1.51 1.43 t Scram Scram onon high high neutron flux.

dome pressure.

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 5-22 EFW (EPU/MELLLA+)

Table 5.12 Licensing Basis Core Average Axial Power Profile State Conditions for Power Shape Evaluation Power, MWt 2,004.0 Core pressure, psia 1,020.5 Inlet subcooling, Btu/Ibm 24.62 Flow, Mlb/hr 60.48 Control state ARO Core average exposure 33,232 (licensing basis EOFP),

MWd/MTU Licensing Axial Power Profile (normalized)

Node Power Top 24 0.308 23 0.700 22 1.148 21 1.327 20 1.451 19 1.506 18 1.521 17 1.510 16 1.455 15 1.430 14 1.463 13 1.451 12 1.396 11 1.314 10 1.213 9 1.094 Sum of Bottom 7 Nodes = 2.773 AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For Page 5-23 EFW (EPU/MELLLA+)

300.0 Relative Core Power Relative Heat Flux Relative Core Flow RelativeSteamFlow__

Relative Feed Flow 200.0-

-o-100.0 -

i)_ -

.0 -

- 100.0 .4 2.0 4.0 6.0 8.0 10.0

.0 Time (seconds)

Figure 5.1 Licensing Basis EOFP LRNB at 100PI80F - TSSS Key Parameters AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-24 1300.0

/

1250.0-I-

II N N

I N N

N Ii 1200.0- N

/

N N N..

0. N L. 1150.0- I L.

10-I 1100.0 -

I 1050.0-Steam Dome Lower Plenum I* AA

.0 2.0 4J.0 6.0I 8.0 10.0 Time (seconds)

Figure 5.2 Licensing Basis EOFP LRNB at I00P/80F - TSSS Vessel Pressures AREVA Inc.

Control~ed Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) __ Page 5-25 4UUAI, Relative Core Power Relative Heat Flux Relative Core Flow RelativeSteam__Flow__

300.0 -

Relative Feed Flow "0

n (1)

S100.0 - *! - - .....

.0-

-1 00.0.

.0 2.0 4.0 6.0 8.0 10.0 Time (seconds)

Figure 5.3 Licensing Basis EOFP TTNB at 100 PI8OF - TSSS Key Parameters AREVA Inc.

ControH~ed Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-26 UU.U -r

/

I,

/

/~

1250.0-iI 1200.0-i/

I 0.)

Co 1150.0 -

,/

a)

/

1100o.o-IiI 1050.0-Steam Dome Lower Plenum I AAA A

.0 2.0 4.0 6.0 8.0 10.0 Time (seconds)

Figure 5.4 Licensing Basis EQFP TTNB at 1OOP/8OF - TSSS Vessel Pressures AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-27

-o

'I)

'1, Figure 5.5 Licensing Basis EOFP FWCF at 100PI80F - TSSS Key Parameters AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-28 1300.0 60 o) ci)

L.

0_

Figure 5.6 Licensing Basis EOFP FWCF at IOOPI80F - TSSS Vessel Pressures AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-29 4U(iO -,-

Relative Core Power Relative Heat Flux Relative Core Flow

300.0 - Relative Fteed Flow 200.0 -

"0 Relatve Fed Flo r--,

0 100.0 -

03_

m4AA* A

.0 10.0 20.0 IX*.30.0 40.0 I 50.0 60.0 70.0 Time (seconds)

Figure 5.7 Licensing Basis EOFP HPCI at 100OP/80F - TSSS Key Parameters AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 5-30 I \

1250.0-1200.0-

-S" 1150.0-1I-1100.0 -

1050.0-Steam Dome Lower Plenum ________________________

A Ann A

.0 10.0 2.0.0 30.0 40.0 50.0 60.0 70.0 Time (seconds)

Figure 5.8 Licensing Basis EOFP HPCI at IOOP/8OF - TSSS Vessel Pressures AREVA lnc~

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 6-1 6.0 Postulated Accidents 6.1 Loss-of-Coolant-Accident (LO CA)

As discussed in Section 2.0 of the LOCA break spectrum report (Reference 6), the LOCA models, evaluation, and results are for a full core of ATRIUM 10XM fuel. The basis for applicability of POT results from full cores of ATRIUM 10OXM fuel (based on AREVA methods)

and GEI4 fuel (based on GNF methods) for a mixed (transition) core is provided in Reference 2, Appendix C. Thermal-hydraulic characteristics of the GEl4 and ATRIUM IOXM fuel designs are similar as presented in Reference 5. Therefore, the core response during a LOCA will not be significantly different for a full core of GEl4 fuel or a mixed core of GE14 and ATRIUM 10XM 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 GEI4 LOCA analysis and resulting licensing POT and MAPLHGR limits remain applicable for GEI4 fuel in transition cores.

The results of the ATRIUM 10XM LOCA break spectrum analysis performed in support of the fuel transition LAR are presented in Reference 6. [

] The analysis considered the same full range of break sizes, break locations, break types, and ECCS single failures that were evaluated in Reference 6.

Table 6.1 summarizes the MAPLHGR limit and MCPR operating limit as well as other initial conditions that have been analyzed. Table 6.2 presents the limiting results from the break spectrum calculations [

]

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 6), a MAPLHGR multiplier of 0.70 is established for SLO since LOCA is more severe when initiated during SLO.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 6-2 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.

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).

Operation in single loop is not allowed in EFW. Therefore, the SLO pump seizure results determined for EPU/MELLLA do not change with the addition of EFW to the power/flow map.

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 AREVA Inc.

Co ntrolled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 6-3 described in Reference 44, which has been shown to continue to apply to ATRIUM I0XM 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-e 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 10XM rods exceeding 170 cal/g 736 6.4 Fuel and Equipment Handling Accident As discussed in Reference 45, the fuel handling accident radiological analysis of record for the alternative source term (AST) was dispositioned with consideration of ATRIUM 1OXM core source terms and number of failed fuel rods. No other aspect of utilizing the ATRIUM 10OXM 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 46 AREVA topical report and in the Monticello USAR (Reference 13). 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 CER 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 0.13 is well below AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 6-4 those reported for AOOs in Section 5.0). 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 ifit 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 0.25 is well below those reported for AOOs in Section 5.0). 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 Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 6-5 Table 6.1 Initial Conditions*

Reactor power (% of rated) 102 102 84.5 Reactor power (MWt) 2044.1 2044.1 1693.4 Steam flow rate (Mlb/hr) 8.51 8.51 6.94 Steam dome pressure (psia) 1038.7 1038.7 1007.5 Core inlet enthalpy (Btu/Ib) 523.6 515.8 505.3 ATRIUM 10XM hot assembly MAPLHGR (kW/ft) 13.1 13.1 13.1

The AREVA calculated heat balance is adjusted to match the heat balance at 100% power and 100%

core flow. AREVA heat balance calculations establish these initial conditions at the stated power and flow.

]

AREVA Inc.

Control~ed Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 6-6 Table 6.2 Summary of TLO Recirculation Line Break Results Highest PCT Cases AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 7-1 7.0 Special Analyses 7.1 ASME Overpressurizatien 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 15).

MSIV closure, TSV closure, and TCV closure were performed for the transition to ATRIUM 10XM fuel (Reference 1) with the ARE VA plant simulator code COTRANSA2 (Reference 23). The analysis of the three valve closures showed that the MSIV valve closure is the most limiting event.

The ASME event was analyzed at 102% core power and both 80% and 105% core flow at the highest cycle exposure. The MSIV closure event 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 approximately 5% drift over the Technical Specifications SRVs opening setpoint

Analysis considered 3 SRVOOS.

TSSS insertion times were used.

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

Monticello sensitivity calculations confirmed that using the maximum allowed dome pressure for the initial pressure is conservative for calculating peak dome and vessel pressures.

A fast MSIV closure time of 2.2 seconds was used.

ATWS-RPT was not credited in this event since this event ends up in a scram (Reference 15).

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 1361 psig occurs in the lower vessel.

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Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 7-2 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 2).

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 (ATWVS) event is less than the ASME Service Level C3 limit of 120% of the design pressure (1500 psig). Overpressurization analyses were performed at 102% core power at both 80% and 105% core 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. 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 PRFO event used as initiator for ATWS analyses was determined to be limiting.

The following assumptions were made in the analyses.

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

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

Analysis considered approximately 5% drift over the Technical Specifications SRVs opening setpoint

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 1452 psig and the maximum steam dome pressure is 1437 psig. The results demonstrate that the ATWS maximum vessel pressure limit of 1500 psig is not exceeded.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 7-3 Pressure results include various adders totaling 20 psi to account for void-quality correlations, Doppler void effects and thermal conductivity degradation (Reference 2).

7.2.2 Longq-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.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (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 47, 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. Ifthe 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 (including state points within the EFW region) and cycle exposures showed that the limiting state point was 60% of rated core power and 44% of rated core flow when the event is initiated from early exposures in the cycle.

This statepoint is outside of the EFW region of the power/flow map.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 7-5 The lowest steam dome pressure that was reached before thermal power was < 25% thermal power was 665 psia (650 psig).

As part of the transition to ATRIUM 10OXM fuel and AREVA methods, AREVA justified that the critical power correlations being used for ATRIUM 10XM fuel and for GEl4 fuel are applicable for pressures above 600 psia (see Reference 2, Appendix G).

7.4 Appendix R - Fire Protection Analysis The Appendix R fire protection case matrix for Monticello safe shutdown is identified in Reference 48. 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 10XM 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 10XM 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.

This event is not sensitive to the initial core flow. Therefore, the conclusions are applicable for operation with ATRIUM 10XM fuel in EFW region.

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 AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 7-6 temperature of 319.2°F, with a boron concentration equivalent to 660 ppm at 68°F.* The temperature of 31 9.2°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 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 10XM fuel is presented in Reference 8 and submitted to the NRC in Reference 45. The assumptions made in the Reference 8 criticality evaluation have been reviewed relative to EFW operation. Since the criticality analysis is a peak reactivity analysis, the potential impact of EPU/EFW operation would be on the lattice depletion to peak reactivity. The criticality analysis includes sensitivity analyses that cover a range of power densities and void histories used in the lattice depletions. It also provides a comparison of rodded versus unrodded depletions. These sensitivities were used to demonstrate the small impact on reactivity to relatively large changes in these parameters. For example, power density (PD) sensitivities were included in Table 6.5 of Reference 8 for changes in PD of _+50% which clearly bounds the impact of EPU conditions. Void history sensitivities illustrated in Figure 6.4 of Reference 8 demonstrate that the limiting condition is found at 0%

void history; consequently the analysis is not adversely impacted by potentially higher voids with EFW operation. Operation at lower flow rates can reduce rod density. However, the criticality analysis assumes uncontrolled depletions which are shown to be bounding for ATRIUM 10XM fuel in Table 6.6 of Reference 8. It is concluded that the Monticello criticality evaluation is also valid when operating under EFW conditions.

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 Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) 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/80F) 381 130 1352 1326 MSIV closure (102P/105F) 350 135 1361 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.

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) 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)

PRFO (102P/80F) 240 143 1452 1437 PRFO (102P/105F) 266 153 1445 1427 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.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 7-9 a)

-.-' 200.0-4,-

0 S100.0 -

0).

6.0 Time (seconds)

Figure 7.1 MSIV Closure Overpressurization Event at 102P/80F - Key Parameters AREVA Inc.

ControDled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 7-10 1 4000

/ N

/

N N

N 1300.0 -

H I,

S1200.0-Steam Dome Lower Plenum

.0 2.0 4.0 6.0 8.0 10.0 12.0 Time (seconds)

Figure 7.2 MSIV Closure Overpressurization Event at 102PI80F - Vessel Pressures*

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

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 7-11 600.0 Bank 1 Bank 2 500.0 - Bank 43' Bank,3 -, -.... _

Bank 5 C,

E

__ 400.0-

~)300.0 -

S200.0-6)3 100.0 -

.v 2.0 4.

.0 6.0 8.0 10.0 12.0 Time (seconds)

Figure 7.3 MSIV Closure Overpressurization Event at 102P/80F - SafetylRelief 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 Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 7-12 4-"

C.)

0_

Figure 7.4 PRFO ATWS Overpressurization Event at 102P/80F - Key Parameters AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 7-13 "0"

PI ZI LO P) 03 Figure 7.5 PRFO ATWS Overpressurization Event at 102PI80F - Vessel Pressures *

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

AREVA Inc.

Controt~ed Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 7-14 E

-o

'I, U)

Time (seconds)

Figure 7.6 PRFO ATWS Overpressurization Event at 1 02P/80F - SafetylRelief Valve Flow Rates*

In the COTRANSA2 code, the 8 SRVs are grouped in 5 banks. I 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 Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (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 AOOs. The MCPR operating limits are established so that less than 0.1% of the fuel rods in the core are expected to experience boiling transition during an AOO initiated from rated or off-rated conditions and are based on a two-loop operation SLMCPR of 1.12 and a single-loop operation SLMCPR of 1.13. MCPR limits were established to support operation from BOO 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 10XM and GEl4 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 10XM and GEl4 fuel limits are presented in Table 8.6 and are applicable for all cycle exposures and EGOS 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 10XM fuel are presented in Table 8.7. The LHGR limits for GEI4 fuel are presented in Reference 49. 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 AGO.

The LHGRFACp and LHGRFACf multipliers for ATRIUM 10OXM fuel are determined using the RODEX4 methodology (Reference 26). The LHGRFACp and LHGRFACf multipliers for GEl4 fuel are developed in a manner consistent with the GNF thermal-mechanical methodology.

AREVA Inc.

Contro~led Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (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 10XM and GEl4 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 10XM and GEl4 fuel, respectively.

LHGRFACf multipliers are applicable for all cycle exposures and EOOS conditions identified in Table 1.1.

8.3 MAPLHGR Limits ATRIUM 10XM MAPLHGR limits are discussed in Reference 7. 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 Inc.

ControDled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) 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 10XM GEI4 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 3.24 PROOS 100.0 1.59 1.58 85.0 1.64 1.64 85.0 1.91 1.85 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 3.24

Limits support channels and/oroperation withof up up to 50% thetoLPRMs 3 SRVOOS, up to I TIPOOS out-of-service, or thecalibration and a LPRM equivalent interval numberofof1000 TIP MWd/ST core average exposure.

AREVA Inc.

ControN~ed Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) 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 GEl4 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 3.24 PROOS 100.0 1.59 1.58 85.0 1.64 1.64 85.0 1.91 1.85 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 3.24

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 LPRM calibration interval of 1000 MWd/ST core average exposure.

AREVA Inc~

ControUIed Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 8-5 Table 8.3 MCPRp Limits for Two-Loop Operation (TLO), NSS Insertion Times BOC to Coastdown

MCPRp Operating Power ATRIUM 10XM GEI4 Condition (% of rated) Fuel Fuel Base 100.0 1.56 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 3.24 PROOS 100.0 1.59 1.58 85.0 1.64 1.64 85.0 1.91 1.85 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 3.24

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 LPRM calibration interval of 1000 MWd/ST core average exposure.

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 8-6 Table 8.4 MCPRp Limits for Two-Loop Operation (TLO), TSSS Insertion Times BOC to Coastdown

MCPRp Operating Power ATRIUM 10XM GEl4 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 < 50%F 2.33 2.38 25.0 at < 50%F 3.09 3.24 PROOS 100.0 1.59 1.58 85.0 1.64 1.64 85.0 1.91 1.85 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 3.24

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 LPRM calibration interval of 1000 MWd/ST core average exposure.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 8-7 Table 8.5 MCPRp Limits for Single-Loop Operation (SLO), TSSS Insertion Times BOC to Coastdown*0 t MCPRp Operating Power ATRIUM 10XM GEl4 Condition (% of rated) Fuel Fuel Base 66.0 2.13 2.19 case 40.0 2.40 2.31 IPROOS 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 3.25

Limits support channels and/oroperation withof up up to 50% thetoLPRMs 3 SRVOOS, up to 1 TIPOOS out-of-service, or the equivalent number of TIP and a LPRM calibration interval of 1000 MWd/ST core average exposure.

t Operation in SLO is not allowed above 66% of rated power or in the EFW region.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 8-8 Table 8.6 Flow-Dependent MCPR Limits ATRIUM 10XM and GEl4 Fuel, NSSITSSS 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 Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 8-9 Table 8.7 ATRIUM 10OXM 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 Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 8-10 Table 8.8 ATRIUM 10XM LHGRFACp Multipliers for NSSITSSS Insertion Times, TLO and SLO, All Cycle 28 Exposures*

LHGRFACp Operating Power ATRIUM 10XM Condition (% of rated) Fuel Base 100.0 1.00 case 40.0 0.80 operation 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 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 I TIPOOS or the equivalent number of TIP channels and/or up to 50% of the LPRMs out-of-service, and a LPRM calibration interval of 1000 MWd/ST core average exposure.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 8-11 Table 8.9 GEl4 LHGRFACp Multipliers for NSSITSSS Insertion Times, TLO and SLO, All Cycle 28 Exposures

LHGRFACp Operating Power GE14 Condition (% of rated) Fuel Base 100.0 0.98t case 40.0 0.57 operation 40.0 at > 50%F 0.41 25.0 at > 50%F 0.34 40.0 at < 50%F 0.53 25.0 at -<50%F 0.37 PROOS 100.00.8 85.0 0.89 85.0 0.75 40.0 0.54 40.0 at > 50%F 0.41 25.0 at > 50%F 0.34 40.0 at < 50%F 0.51 25.0 at < 50%F 0.37

Limits support channels and/oroperation withof up up to 50% thetoLPRMs 3 SRVOOS, up to 1 TIPOOS out-of-service, or thecalibration and a LPRM equivalentinterval numberofof1000 TIP MWd/ST core average exposure.

S0.96 setdown required if analytical setpoint is greater than 114% (based on CRWE calculation).

AREVA Inc

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 8-12 Table 8.10 ATRIUM 10XM LHGRFACf Multipliers, NSSITSSS 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 Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page 8-13 Table 8.11 GEl4 LHGRFACf Multipliers, NSSITSSS Insertion Times, TLO and SLO, PROOS, All Cycle 28 Exposures Core Flow GEl4

(% of rated) LHGRFACf 30.0 0.68 40.0 0.68 75.0 1.00 105.0 1.00 AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 8-14 Table 8.12 ATRIUM 10XM 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 Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 9-1 9.0 References

1. ANP-321 3(P) Revision 1, Monticello Fuel Transition Cycle 28 Reload Licensing Analysis (EPU/MELLLA), AREVA NP, June 2013.

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

3. ANP-311I9P Revision 0, Mechanical Design Report for Monticello A TRIUM TM IOXM Fuel Assemblies, AREVA NP, October 2012.

4. ANP-3221 P Revision 0, Fuel Rod Thermal-MechanicalDesign for Monticello ATRIUM IOXM Fuel Assemblies, Cycle 28, ARE VA NP, May 2013.

5. ANP-3092(P) Revision 0, Monticello T-hermal-Hydraulic Design Report for ATRIUMTM I0XM Fuel Assemblles, AREVA NP, July 2012.

6. ANP-321 1(P) Revision 1, Monticello EPU LOCA Break Spectrum Analysis for ATRIUM TM IOXM Fuel, AREVA NP, July 2013.

7. ANP-3212(P) Revision 0, Monticello EPU LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM TM IOXM Fuel, AREVA NP, May 2013.

8. ANP-31 13(P) Revision 0, Monticello Nuclear Plant Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM TM IOXM Fuel, ARE VA NP, August 2012.

9. ANP-3135P Revision 0, Applicability of ARE VA BWR Methods to Extended Flow Window for Monticello, AREVA, April 2014.

10. ANP-3274P Revision 0, Analytical Methods for Monticello A TWS-I, AREVA, April 2014.

11. ANP-3284P Revision 0, Results of Analysis and Benchmarking of Methods for Monticello ATWS-I, ARE VA, April 2014.

12. ANP-3124(P) Revision 0, Monticello Cycle 28 Fuel Cycle Design, AREVA NP, November 2012.

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

14. Technical Specification Requirements for Monticello Nuclear GeneratingPlant Unit 1, Monticello, Amendment 146.

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

16. NEDO-33322P Rev i s i on 3, Safety Analysis Report for Monticello Constant Pressure Power Uprate, GEH, 0October 2008.

17. NEDC-33435P, Revision 1, Safety Analysis Report for Monticello Maximum Extended Load Line Limit Analysis Plus, GEH, December 2009.

18. Letter from T. A. Beltz (NRC) to K. D. Fili (Xcel Energy), "Monticello Nuclear Generating Plant - Issuance of Amendment No. 176 To Renewed Facility Operating License Regarding Extended Power Uprate (TAO No. MD9990)," December 9, 2013, ML13316B298.

AREVA Inc.

ControDned Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 9-2

19. Letter from T. A. Beltz (NRC) to K. D. Fili (Xcel Energy), "Monticello Nuclear Generating Plant - Issuance of Amendment No. 180 To Renewed Facility Operating License Regarding Maximum Extended Load Line Limit Analysis Plus (TAO No. ME3145),"

March 28, 2014, ML14035A248.

20. 51-9187384-000, "Monticello Plant RPV Seismic Assessment with ATRIUMTM I0XM Fuel," AREVA NP, September 2012 (RJW: 12:022).

21. NEDO-33006-A Revision 3, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, General Electric Hitachi Nuclear Energy America, LLC, June 2009. (available in ADAMS Folder ML091800530).

22. AN P-I10307PA Revision 0, ARE VA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.

23. ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor TransientAnalyses, Advanced Nuclear Fuels Corporation, August 1990.

24. XN-NF-80-1 9(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.

25. XN-NF-84-1 05(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. BAW-1 0247PA Revision 0, Realistic Thermal-MechanicalFuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.

27. XN-NF-81-58(P)(A) Revision 2 and Supplements I and 2, RODEX2 Fuel Rod Thermal-MechanicalResponse Evaluation Model, Exxon Nuclear Company, March 1984.

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

29. EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001.

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

31. BAW-1 0255PA Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008.

32. OG04-0153-260, Plant-Specific Regional Mode DIVOM Procedure Guideline, June 15, 2004.

33. BWROG-03047, Resolution of Reportable Condition for Stability Reload Licensing Calculations Using Generic Regional Mode DIVOM Curve, September 30, 2003.

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

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page 9-3

35. 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.

36. ANP-3158P Revision 2, Fuel Rod Thermal-MechanicalDesign for Monticello ATRIUM IOXM Fuel Assemblies, Cycle 28, AREVA NP, August 2013.

37. GNF Design Basis Document, Fuel-Rod Thermal-MechanicalPerformance Limits for GEI4C, DB-0012.03 Revision 2, September 2006 (transmitted by letter, D.J. Mienke (Xcel Energy) to R. Welch (AREVA), "Transmittal of Requested Monticello Plant Information: GEl4 Exposure Limits," July 19, 2012).

38. ANP-1 0298PA Revision 0, ACE/ATRIUM IOXM CriticalPower Correlation, AREVA NP, March 2010.

39. ANP-1 0298PA Revision 0 Supplement 1P Revision 0, Improved K-factor Model for ACE/ATRIUM IOXM Critical Power Correlation,AREVA NP, December 2011.

40. Letter, Sher Bahadur (U.S. Nuclear Regulatory Commission) to P. Salas (AREVA),

"FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR TOPICAL REPORT ANP-10298PA, REVISION 0, SUPPLEMENT 1iP, REVISION 0, "IMPROVED K-FACTOR MODEL FOR ACE/ATRIUM 10XM CRITICAL POWER CORRELATION" (TAC NO. ME7963)", March 31, 2014.

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

42. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation'sCritical Power Correlationsto Co-Resident Fuel, Siemens Power Corporation, August 2000.

43. ANP-1 0262(P)(A) Revision 0, Enhanced Option Ill Long Term Stability Solution, May 2008.

44. 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.

45. 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).

46. XN-NF-80-1 9(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.

47. General Electric 10CFR Part 21 Communication, Potential Violation of Low Pressure Technical Specification Safety Limit, SC05-03, March 22, 2005.

48. 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-20 12-007, February 14, 2012.

49. 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.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) 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 EFW operating limits and the transient analysis results. The thermal limits for NSS and TSSS insertion times protect the TTWB event with OSS insertion times. Comparisons are presented for the ATRIUM 10XM and GEI4 MCPRp limits and LHGRFACp multipliers.

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-2 MONT CY28 EOFPLBNWSS 1 6175.0 DSS/NSS (AlIOXMv Fuel) 4.0 1 I II o] FWCF o HPCl A LOFWH 3.5

+ LRNB x LRWB o RUNUP 3.0 S TTNB

[] TTWB om F~

o-- 2.5

+ 1 2.0 1.5 Ag A AX+/-

A* o 0 A

,V 0 1.0 I r I I I tI I I 0 10 20 30 40 50 60 70 80 90 100 110 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 < 50%F 2.33 25.0 < 50%F 3.09 Figure A.1 BOC to Licensing Basis EOFP Power-Dependent MCPR Limits for ATRIUM 10XM Fuel NSS Insertion Times Base Case Two-Loop Operation (TLO)

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-3 MONT CY28 EOFPLBNSS 1 6175.0 DSS/NSS (GEl4 Fuel) 4.0 SIII I 1 o] FWCF o HPCI 3.5 A* LOFWH

+ LRNB x LRWB

<> RUNUP 3.0 v TTNB

[] TTWB 0

0 C-)_ +

2.0 1'.5 O0 0 A X <

1.0 0 io 20 30 40 50 60 70 80 90 100 110 Power (% Rated)

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 < 50%F 3.24 Figure A.2 BOC to Licensing Basis EOFP Power-Dependent MCPR Limits for GEl4 Fuel NSS Insertion Times Base Case Two-Loop Operation (TLO)

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-4 MONT CY28 CoastNSS 21175.0 DSS/NSS (AlIOXIV Fuel) 4.0 o] FWCF o HPCI 3.5 A LOFWH

+ LRNB 3.0

% co x LRWB RUNUP v TTNB TTWB

\

°--

2.5 4-

[]

[

[]

2.0 1.5 A o 0 A X 0 1.0 I I I I1I I 1 I 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)

Power MCPRp

(% of rated) Limit 100.0 1.56 40.0 1.74 40.0 > 50%F 2.77 25.0 > 50%F 3.39 40.0*<50%F 2.33 25.0 < 50%F 3.09 Figure A.3 BOC to Coastdown Power-Dependent MICPR Limits for ATRIUMIOXM~ Fuel NSS Insertion Times Base Case Two-Loop Operation (TLO)

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-5 MONT CY28 CoastNSS 21175.0 DSS/NSS (GEl4 Fuel) 4.0 o] FWCF o HPCI A LOFWH 3.5

+ LRNB x LRWB o RUNUP 3.0 v TTNB

[] TTWB 0

0

+

o_ 2.5 rY 2.0 1.5 X A i i i v 1.0 I I I I I [ I I I I 0 10 20 30 40 50 60 70 80 90 10O0 110 Power (% Rated)

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 < 50%F 3.24 Figure A.4 BOC to Coastdown Power-Dependent MCPR Limits for GEI4 Fuel NSS Insertion Times Base Case Two-Loop Operation (TLO)

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-6 MONT CY28 EOFPLBTSSS 1 6175.0 DSS/TSSS (AllOXM Fuel) 4.0 I I I I I I I I o] FWCF o HPCI 3.5 A LOFWH

+ LRNB x LRWB

%* v TTNB 3.0

[] TTWB c* 2.5 rY +

2.0 1.5 A A 0 0 0 ax X

1.0 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)

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 40.0 < 50%F 2.33 25.0 < 50%F 3.09 Figure A.5 BOC to Licensing Basis EOFP Power-Dependent MCPR Limits for ATRIUM 10XM Fuel TSSS Insertion Times Base Case Two-Loop Operation (TLO)

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-7 MONT CY28 EOFPLBTSSS 1 6175.0 DSS/TSSS (GEl4 Fuel) 4.0 I III II I I o] FWCF o HPCI 3.5 z, LOFWH

+ LRNB x LRWB E*< RUNUP 3.0 v TTNB

[] TTWB I._

r* 2.5 C) a 2.0 1.5 0 c. 0 0A x

1.0 I I I I r i i i I I 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)

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 < 50%F 3.24 Figure A.6 BOC to Licensing Basis EOFP Power-Dependent MCPR Limits for GEI4 Fuel TSSS Insertion Times Base Case Two-Loop Operation (TLO)

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-8 MONT CY28 CoastTSSS 21175.0 DSS/TSSS (A1OXM Fuel) 4.0 I I I U* FWCF o HP~l A LOFWH 3.5

+ LRNB x LRWB o RUNUP S TTNB 3.0 F [] TTWB E

._]

0D_ 2.5 I rv +

C3_

2.0 F

+

1.5 I A A x x o 1.0 I I I I I I I 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)

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 40.0 < 50%F 2.33 25.0 < 50%F 3.09 Figure A.7 BOC to Coastdown Power-Dependent MCPR Limits for ATRIUM 10XM Fuel TSSS Insertion Times Base Case Two-Loop Operation (TLO)

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-9 MONT CY28 CoestTSSS 21175.0 DSS/TSSS (GEl4 Fuel) 4,0 3.5 3.0

,-I-

¢* 2.5 2.0 1.5 1.0 0 10 20 30 40 Power50 (% Rated) 60 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 -<50%F 3.24 Figure A.8 BOC to Coastdown Power-Dependent MCPR Limits for GEI4 Fuel TSSS Insertion Times Base Case Two-Loop Operation (TLO)

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For *Revision 3 EFW (EPU/MELLLA+) Page A-10 MONT CY28 CoastPROOS 21175.0 DSS/TSSS (AlIOXM Fuel) 4.0 - I I Ii I o] FWCF o HPCl A LOFWH 3.5

+ LRNB o> PRFDS 3.0 o] v RUNUP

[] TTNB

~* TTWB n 2.5 n-) +

[]

N .4-2.0 V +

1.5 v V x x

1.0 0 10 20 ,30 40 50 60 70 80 90 100 110 Power (% Rated)

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 < 50%F 3.24 Figure A.9 BOC to Coastdown Power-Dependent MCPR Limits for ATRIUM 10XM Fuel NSSITSSS Insertion Times PROOS Two-Loop Operation (TLO)

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-il MONT CY28 CoastPROOS 21175.0 DSS/TSSS (GE14- Fuel) 4.0 o] FWCF o HPCI

LOFWH 3.5

+ LRNB x LRWB o PRFDS V RUNUP 3.0 I S TTNB o *( TTWB E

°--

2.5 C-)

0 2.0 1.5 V z x

1.0 I I I I I I I I I I 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)

Power MCPRp

(% of rated) Limit 100.0 1.58 85.0 1.64 85.0 1.85 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 3.24 Figure A.10 BOC to Coastdown Power-Dependent MCPR Limits for GEI4 Fuel NSS/TSSS Insertion Times PROOS Two-Loop Operation (TLO)

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-12 MONT CY28 CoastSLO 211 75.0 DSS/NSS/TSSS (AIOXM Fuel) 4.0 II iII o] FWCF o HPCI 3.5 A LOFWH x PRFDS o] o RUNUP 3.0 S TTNB

[] TTWB

)4 SLPS

a. 2.5 0Y +

C_)

[]+ x 2.0 g

0 o+

1.5 A A* O 1.0 T I I I I I I I 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)

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 < 50%F 2.49 25.0*<50%F 3.25 Figure A.tl BOC to Coastdown Power-Dependent MCPR Limits for ATRIUM 10XM Fuel NSSITSSS Insertion Times Base case + PROOS Single-Loop Operation (SLO)*

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

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-13 MvONT CY28 CoastSLO 21175.0 DSS/NSS/TSSS (CEl4 Fuel) 4-.0 3.5 3.0

£1_ 2.5 a

2.0 1.5 1.0 0 10 20 30 40 50 60 70 80 90 100 110 Power (% Rated)

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 -<50%F 3.25 Figure A.12 BOC to Coastdown Power-Dependent MCPR Limits for GEl4 Fuel NSSITSSS Insertion Times Base case + PROOS Single-Loop Operation (SLO)*

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

AREVA Inc.

Controlied Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-14 MONT(AllIOXM CY28 LHGR~FACp Fuel) 1.2 1.1 1.0

.9 0

.8 0D_

.7

__J

.6

.5

.4 A LOFWH A

HPCI

.3 A FWCF

.2 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*<50%F 0.36 Figure A.13 All Exposures Power-Dependent LHGR Multipliers for ATRIUM I0XM Fuel NSSITSSS Insertion Times Base Case Two-Loop Operation (TLO) and Single-Loop Operation (SLO)

AREVA Inc.

Doument ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+)

Controlled Page A-15 (GE14 Fuel) 1.2 I I I 1.1 1.0

.9 n

.8 0i 0D LL

+

0_ .7 0l

(_9

.6

.5

.4

[] 0] FWCF 0 HPCl A LOFWH

.3 + RUNUP

.2 I I I I f I I I I I 0 1 20 30 40 50 60 70 80 90 100 110 Power (% Rated)

Power LHGRFACp

(% of rated) Multiplier 100.0 0.98*

40.0 0.57 40.0 > 50%F 0.41 25.0 > 50%F 0.34 40.0 -<50%F 0.53 25.0 -<50%F 0.37 Figure A.14 All Exposures Power-Dependent LHGR Multipliers for GEI4 Fuel NSSITSSS Insertion Times Base Case Two-Loop Operation (TLO) and Single-Loop Operation (SLO)

0.96 setdown required ifanalytical setpoint is greater than 114% (based on CRWE calculation).

AREVA Inc.

Controlled Document ANP-3295NP Revision 3 Monticello Licensing Analysis For EFW (EPU/MELLLA+) Page A-16 MONT CY28 LHGRFACp PROOS (ATI OXM Fuel) 1.2 1.1 1.0

.9

.8

.7 CD 7-

.6

.5

.4 A

[] LOFWH o PRFDS PROOS

.3 A FWCF

.2 I I I 0 10 20 .30 40 50 60 70 60 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*<50%F 0.36 Figure A.15 All Exposures Power-Dependent LHGR Multipliers for ATRIUM 10XM Fuel NSSITSSS Insertion Times PROOS Two-Loop Operation (TLO) and Single-Loop Operation (SLO)

AREVA Inc.

Controlled Document ANP-3295NP Monticello Licensing Analysis For Revision 3 EFW (EPU/MELLLA+) Page A-17 MONT CY28 LHGRFACp PROOS SCA (CE14 Fuel) COASTI AL 1.2 I I I I I I I I 1.1

+

1.0

.9 h

.8 CD

.<1 .7

+

.6 I

.5 0]

.4 0] FWCF 0 LOFWH A HPCI

+ PRFDS PROOS

.2 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 0.98*

85.0 0.89 85.0 0.75 40.0 0.54 40.0 > 50%F 0.41 25.0 > 50%F 0.34 40.0 < 50%F 0.51 25.0 < 50%F 0.37 Figure A.16 All Exposures Power-Dependent LHGR Multipliers for GEI4 Fuel NSSITSSS Insertion Times PROOS Two-Loop Operation (TLO) and Single-Loop Operation (SLO)

0.98 setdown required ifanalytical setpoint is greater than 1i14% (based on CRWE calculation).