Text

ANP-3352NP, Revision 1, Transition License Amendment Request, Technical Report.
	ML15356A186
Person / Time
Site:	Saint Lucie
Issue date:	11/30/2015
From:	; AREVA
To:	; Office of Nuclear Reactor Regulation
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ML15356A251	List: L-2015-300, ANP-3352NP, Revision 1, Transition License Amendment Request, Technical Report.; ML15356A184; ML15356A185;
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	L-2015-300, TAC MF5494, TAC MF5495 ANP-3352NP, Rev. 1
	Download: ML15356A186 (71)
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L-2015-300 Attachment 6 AN P-3352N P Revision 1 St. Lucie Unit 2 Fuel Transition License Amendment Request Technical Report Following 70 pages

LI ~JEI~ ~J&L~[ E!~E ii ARE EVA St. Lucie Unit 2 Fuel Transition License ARevisinP1 Amendment Request Technical Report November 2015 AREVA Inc.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision I Page ii Nature of Changes Item Section(s)

Page(s) orDecitoanJuifain DecitoanJutfain 1 Section 1.1 Added last paragraph to describe purpose of this revision.

Sections 2.4.1, Clarified that References 4 and 5 have subsequently been approved by 2 2.4.3.1, and the USNRC and that the relative statements are still applicable.

2.4.4 Table 2-2 Updated results for Items 3.3.1 (Guide Tube) and 3.4 (Structural Deformations) within this table.

Mixed-core results were updated and full-core results were added to the table. Added note regarding the update of the grid strength allowable for Table 2-3 hot, OBE conditions. Loads were updated (generally resulting in an increase in loads) to reflect [

Secion5.2Added statement at end of paragraph that additional SBLOCA results are found in a new reference.

Section Changed licensing results in last paragraph and included "charging" in 65.2.1 text related to ECCS.

Updated references 4 and 28. Reference 4 has been approved by the 7 Section 7.0 USNRC. Reference 28 is the latest SBLOCA Summary Report. Added Reference 32.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page iii Contents 1.0 Introduction and Design Overview........................................................... 1-1 1.1 Introduction ........................................................................... I1-1 1.2 Fuel Design Overview................................................................ 1-2 2.0 Mechanical Design ................................................................... '........2-1 2.1 Introduction ........................................................................... 2-I 2.2 Operational Experience of AREVA HTP TM Fuel Assemblies in CE-16 and CE-14 Plants ................................................................. 2-1 2.3 Mechanical Compatibility ............................................................ 2-2

- 2.3.1 Fuel Assembly.............................................................. 2-4

- 2.3.2 Upper Tie Plate............................................................. 2-5

-2.3.3 Lower Tie Plate............................................................. 2-5

- 2.3.4 Guide Tubes................................................................ 2-5 2.4 Mechanical Design Evaluations..................................................... 2-6

- 2.4.1 Description.................................................................. 2-6

- 2.4.2 Input Parameters and Assumptions ...................................... 2-7

- 2.4.3 Recently Identified Analysis Issues....................................... 2-7 2.4.3.1 Thermal Conductivity Degradation (TCD) .................... 2-7 2.4.3.2 Seismic Evaluations............................................ 2-8

- 2.4.4 Mechanical Analyses Results ............................................. 2-9 2.4.4.1 Additional Seismic Analysis Results......................... 2-15 2.5 Mechanical Design Conclusions ................................................... 2-16 3.0 Nuclear Design................................................................................ 3-1 3.1 Introduction ........ .................................................................. 3-1 3.2 Input Parameters .. .................................................................. 3-1 3.3 Methodology ........................................................................... 3-1 3.4 Description of Design Evaluations .................................................. 3-3 3.5 Results ................................................................................. 3-4 3.6 Conclusion........................................................................... :3-5 4.0 Thermal and Hydraulic Design .............................................................. 4-1 4.1 Description............................................................................ 4-1 4.2 Input Parameters and Assumptions ................................................ 4-1 4.3 Acceptance Criteria .................................................................. 4-2 4.4 Method of Analysis ................................................................... 4-3 4.5 Results ................................................................................ 4-5

- 4.5.1 Thermal-Hydraulic Compatibility .......................................... 4-5 4.5.1.1 Core Pressure Drop............................................ 4-5 4.5.1.2 Total Bypass Flow .............................................. 4-6 4.5.1.3 Crossflow Velocity.: ............................................ 4-7 4.5.1.4 RCS Flow Rate ................................................. 4-7 4.5.1.5 Transition Core DNB Performance............................ 4-7 4.5.1.6 Control Rod Drop Times....................................... 4-8

- 4.5.2 Thermo-Hydrodynamic Instability ......................................... 4-8

- 4.5.3 Rod Bow .................................................................... 4-8

-4.5.4 Guide Tube Heating........................................................ 4-9

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page iv

- 4.5.5 Setpoint Analyses.......................................................... 4-9 4.5.5.1 Thermal Margin Safety Limit Line Verification .............. 4-16 5.0 Accident and Transient Analyses ............................................................ 5-1 5.1 Non-LOCA Analyses ................................................................. 5-1

- 5.1.1 Introduction ................................................................. 5-1

- 5.1.2 Computer Codes ........................................................... 5-1

- 5.1.3 Analysis Methodologies ................................................... 5-2

- 5.1.4 Event Disposition and Analysis ........................................... 5-4

- 5.1.5 Conclusions................................................................. 5-7 5.2 Loss-of-Coolant Accident Analyses ................................................. 5-7

- 5.2.1 Small Break Loss-of-Coolant Accident ................................... 5-8

_ 5.2.2 Large Break Loss-of-Coolant Accident ................................... 5-9 6.0 Summary and Conclusion..................................................................... 6-1 7.0 References .................................................................................... 7-1

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page v Tables Table 2-1: Comparison of Nominal Mechanical Design Features ................................. 2-3 Table 2-2: Fuel Mechanical Design Evaluation Results........................................... 2-10 Table 2-3: Seismic and LOCA Loadings ........................................................... 2-16 Table 3-1: Range of Key Safety Parameters....................................................... 3-2 Table 3-2: Projected Transition Cycle Core Characteristics....................................... 3-4 Table 4-1: Thermal-Hydraulic Design Parameters.................................................. 4-1 Table 4-2: Limiting Parameter Directions ........................................................... 4-2 Table 4-3: System Related Uncertainties ........................................................... 4-2 Table 4-4: Minimum Margin Summary for Setpoint Calculations.................................. 4-9 Table 5-1: Non-LOCA Limiting Results.............................................................. 5-6

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page vi Figures Figure 1-1 : AREVA CE-16 Fuel Assembly for St. Lucie Unit 2 ................................... 1-6 TM Figure. 1-2: St. Lucie Unit 2 FUELGUARD Lower Tie Plate ..................................... 1-7 Figure 1-3: St. Lucie Unit 2 Upper Tie Plate ........................................................ 1-7 Figure 1-4: St. Lucie Unit 2 HTP TM Spacer Grid .................................................... 1-8 Figure 4-1: Pressure Drop Profiles................................................................... 4-6 Figure 4-2: LPD - High Trip Setpoint............................................................... 4-10 Figure 4-3: LPD LSSS Verification Results........................................................ 4-1 1 Figure 4-4: TM/LP Trip Setpoint - QR1 Function .................................................. 4-12 Figure 4-5: TM/LP Trip Setpoint - QA Function.................................................... 4-13 Figure 4-6: ASI Limits for DNB vs. Thermal Power................................................ 4-14 Figure 4-7: DNB LCO CEAD Results .............................................................. 4-14 Figure 4-8: DNB LCO LOCF Results............................................................... 4-15 Figure 4-9: ASI Limits for LHR vs. Maximum Allowable Power Level when Using the Excore Detectors........................................................................ 4-15 Figure 4-10: LPD LCO Verification Results........................................................ 4-16 Figure 4-1 1: Thermal Margin Safety Limit Lines................................................... 4-18 Figure 4-12: Axial Power Distribution for Thermal Margin Limit Lines ........................... 4-19

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page vii Nomenclature AQOO............................... Anticipated Operational Occurrence ARO ............................... All Rods Out ASI ................................. Axial Shape Index ASME.............................. American Society of Mechanical Engineers BOO ............................... Beginning of Cycle BOL ................................ Beginning of Life B&W ............................... Babcock &Wilcox CE.................................. Combustion Engineering CEA................................ Control Element Assembly CEAD.............................. Control Element Assembly Drop CFR................................ Code of Federal Regulations CHF................................ Critical Heat Flux CLT................................. Centerline Temperature COLR.............................. Core Operating Limits Report CUF................................ Cumulative Usage Factor CVCS.............................. Chemical and Volume Control System DNB................................ Departure from Nucleate Boiling DNBR.............................. Departure from Nucleate Boiling Ratio DTC................................ Doppler Temperature Coefficient ECCS.............................. Emergency Core Cooling System EFPD............................... Effective Full Power Days EM ................................. Evaluation Methodology EOC ............................... End of Cycle EOL ................................ End of Life FCM ............................... Fuel Centerline Melt FPL................................. Florida Power and Light F0 .............. . . . . . . . . . . . . . . . . . . . . Total Power Peaking Factor F.................................... Assembly Radial Peaking Factor GDC ............................... General Design Criteria HFP ................................ Hot Full Power HMPTM ............................... High Mechanical Performance HPSI ............................... High Pressure Safety Injection HTPTM ............................... High Thermal Performance HZP ................................ Hot Zero Power ID................................... Inner Diameter IN................................... Information Notice LAR ................................ License Amendment Request LBLOCA........................... Large Break Loss-of-Coolant Accident LCO................................ Limiting Condition for Operation LHGR.............................. Linear Heat Generation Rate

[HR ................................ Linear Heat Rate

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page viii Nomenclature (continued)

LOCA.............................. Loss-of-Coolant Accident LOCF............................... Loss of Coolant Flow LPD ................................ Local Power Density LSSS............................... Limiting Safety System Selling LTP................................. Lower Tie Plate MDNBR............................ Minimum Departure from Nucleate Boiling Ratio MSSV.............................. Main Steam Safety Valve MTC ............................... Moderator Tem peratu re Coefficient NAF ................................ Neutron Absorbing Fuel NRC ............................... Nuclear Regulatory Commission OBE................................ Operating Basis Earthquake OD ................................. Outer Diameter PCT ................................ Peak Cladding Temperature PDIL ............................... Power Dependent Insertion Limit PORV.............................. Power-Operated Relief Valve PWR ............................... Pressurized Water Reactor RCS ................................ Reactor Coolant System RPS ................................ Reactor Protection System RTP ................................ Rated Thermal Power SAFDL ............................. Specified Acceptable Fuel Design Limit SBLOCA........................... Small Break Loss-of-Coolant Accident SER ................................ Safety Evaluation Report SIAS ............................... Safety Injection Actuation Signal SIT.................................. Safety Injection Tank SRP ................................ Standard Review Plan SSE ................................ Safe Shutdown Earthquake TCD ................................ Thermal Conductivity Degradation T-H................................. Thermal Hydraulic TMSLL ............................. Thermal Margin Safety Limit Lines TM/LP.............................. Thermal Margin/Low Pressure TS .................................. Technical Specifications UFSAR ............................ Updated Final Safety Analysis Report USNRC............................ United States Nuclear Regulatory Commission UTP ................................ Upper Tie Plate VHPT............................... Variable High Power Trip W................................... Westinghouse WPR ............................... Wetted Perimeter Ratio

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AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 1-1 1.0 Introduction and Design Overview 1.1 Introduction Florida Power and Light (FPL) is planning to transition St. Lucie Unit 2 to AREVA CE-16 High Thermal Performance (HTPTMl) fuel starting in the spring of 2017. The ARE VA fuel design will be the CE-16 HTP TM fuel consisting of a 16x16 assembly configuration with M51 fuel rods, Zircaloy-4 MONOBLOOTMI corner guide tubes, an Alloy 718 High Mechanical Performance (HMPTMl) spacer at the lowermost axial elevation, Zircaloy-4 HTPTM spacers in all other axial elevations, a FUELGUARDTMI lower tie plate (LTP), and the AREVA reconstitutable upper tie plate (UTP).

The AREVA CE-16 HTP TM fuel design for St. Lucie Unit 2 is similar and has the same design features as the AREVA CE-14 HTP TM fuel design operating in St. Lucie Unit 1. It is also similar to the AREVA CE-16 HTP TM lead fuel assemblies operated in San Onofre Unit 2. The fuel rods are also similar to the AREVA CE-16 HTP TM fuel rods operated in the lead fuel assemblies at Palo Verde. The design features of the AREVA CE-16 HTP TM fuel design planned for St. Lucie Unit 2 have demonstrated excellent fuel performance. The HTPTM / HMPTM spacer grids are very resistant to flow induced grid-to-rod fretting failures, the FUELGUARD TM LTP is effective at

protecting the fuel from debris in the reactor coolant system, and the M5@ cladding has very low oxidation and hydrogen pickup rates.

Section 1.2 of this report provides a more detailed discussion of the design features of the AREVA CE-16 HTP TM fuel assembly. Section 2.0 of the report outlines AREVA's mechanical and structural evaluation methodology for the fuel design including the compatibility assessment and review of operating experience. Section 3.0 discusses the nuclear design bases and the methodologies for transitioning from the Westinghouse fuel design to the AREVA CE-16 HTP TM fuel for St. Lucie Unit 2. Section 4.0 provides the thermal and hydraulic design of the reactor that ensures the core can meet steady state and transient performance requirements without violating the acceptance criteria. Section 5.0 provides information related to the St. Lucie Unit 2 transient and accident analyses for the proposed transition. Also, summary reports of analyses for the non-loss-of-coolant accident (non-LOCA), small break LOCA (SBLOCA), and realistic HTP, HMP, MONOBLOC, and FUELGUARD are trademarks of AREVA. M5 is a registered trademark of AREVA.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 1-2 large break LOCA (RLBLOCA) analysis methodologies have been prepared as documented in References 23, 28, and 29, respectively.

Note that demonstration of the evaluation methodologies has been performed with a submittal core design. The submittal core design was developed to provide key safety parameters to support the transition from Westinghouse fuel to AREVA CE-16 HTP TM fuel prior to the development of cycle-specific designs. This provides assurance that the plant licensing bases are met for the operation of St. Lucie Unit 2 with the AREVA CE-16 HTP TM fuel during the transition and full core cycles.

Revision 1 of this document is being issued to update seismic/LOCA results and SBLOCA results, as described in ANP-3440 (Reference 32). Table 2-2 (Items 3.3.1 and 3.4) and Table 2-3 (all items, including footnote) have been revised with the updated seismic/LOCA information (it was determined that the text within Sections 2.4.3.2 and 2.4.4.1 did not need to be revised; Items 3.2.5 and 3.2.7 within Table 2-2 were confirmed to remain bounding and therefore were not revised). Sections 5.2 and 5.2.1 have been revised with the updated SBLOCA information.

1.2 Fuel Design Overview The AREVA fuel assembly for St. Lucie Unit 2 is of a Combustion Engineering (CE) 16x16 lattice design. This lattice contains 236 fuel rods, four (4) corner guide tubes, and one (1) center guide tube. The corner and center guide tubes each occupy four (4) fuel rod positions.

The fuel rods are positioned within the fuel assembly by ten (10) spacer grids that are attached to the guide tubes.

The St. Lucie 2 AREVA design is very similar to the St. Lucie 1 AREVA fuel design. They both use HTP TM / HMP TM spacer grids, M5 fuel rod cladding, the FUELGUARD TM LTP, and the AREVA reconstitutable CE UTP. These components have been demonstrated to have excellent fuel performance and reliability. Figure 1-1 is a schematic of the AREVA fuel assembly.

The fuel rod design uses M5 cladding. The MS material has very low corrosion and hydrogen pickup rates; providing substantial margin for end of life corrosion and hydrogen content. This material was developed in Europe and has been used extensively both in Europe and the United States for fuel rod cladding. The material has been generically reviewed and accepted

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 1-3 by the United States Nuclear Regulatory Commission (USNRC) for use on CE fuel designs (Reference 1). Reloads with M5 cladding have been provided in the United States since 2000 and on CE-14 designs since 2006. Performance has been demonstrated to rod exposures in excess of 80 MWd/kgU. The fuel rod design includes uranium dioxide fuel rods and Gadolinia bearing uranium dioxide fuel rods, both with axial blankets of lower enriched uranium dioxide.

Also, multiple uranium-235 enrichments are used within an assembly.

The lower tie plate design is the FUELGUARDTM structure. This structure uses curved vanes to provide non-line-of-sight flow paths for the incoming coolant to protect the fuel assembly from debris that may be present. This design is very efficient at preventing debris, including small pieces of wire, from reaching the fuel. The design uses the same vane configuration and spacing that has been used on CE-14, CE-15, CE-16, Westinghouse 14x14, Westinghouse 15x15, Westinghouse 17x17, and Babcock & Wilcox (B&W) 15x15 designs in the united States.

This FUELGUARD TM design has been used on reloads in the United States since 1991 and on CE-14 designs since 2001. A schematic of the CE-16 FUELGUARD TM lower tie plate is provided in Figure 1-2.

The upper tie plate (UTP) design is the standard AREVA reconstitutable design for CE configurations. The basic configuration is the same as that used for CE-14 plants supplied by AREVA, with the heights, diameters, and position of the corner and center posts adjusted for the CE-16 lattice and to be compatible with the core plate separation at the St. Lucie Unit 2 plant.

Figure 1-3 shows the St. Lucie 2 UTP configuration. The reaction plate has also been modified to match the interface conditions with the fuel handling grapples consistent with the co-resident fuel. This reconstitutable design uses the corner locking nuts to engage with the upper sleeves on the corner guide tubes. The design allows the reaction plate to be depressed to a setting well beyond the end of life deflections. At the fully depressed setting, the corner nuts can be rotated to disengage the upper tie plate from the guide tube locking sleeves; the upper tie plate can then be removed. This design does not create any loose or disposable parts during the reconstitution. The design has been used for AREVA CE-14 reloads in the United States since 1982. The reconstitution capabilities of the AREVA CE designs have been successfully demonstrated in CE-i14 and CE-i16 fuel examinations.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 1-4 The cage or skeleton uses four (4) Zircaloy-4 MONOBLOCTM corner guide tubes, one (1)

Zircaloy-4 center guide tube, nine (9) Zircaloy-4 HTP TM spacers, and one (1) Alloy 718 HMP TM spacer at the lowest spacer position. The HTPTM spacers are welded directly to the five guide tubes whereas the HMP TM spacer is attached to the guide tubes by mechanically capturing the spacer between rings that are welded to the guide tubes. Since the HMPTM spacer is made from Alloy 718, it cannot be directly welded to the Zirconium alloy guide tubes. The HTP TM spacer design was developed in the late 1980s and has been used on CE-14, CE-15, CE-16, Westinghouse 14x14, Westinghouse 15x15, Westinghouse 17x17, and B&W 15x15 fuel assemblies in the United States. (The CE-16 application was in two lead assembly programs.)

The initial reloads were in 1991, and the initial CE-14 reloads were in 2001.

The CE-14 and CE-16 units have very challenging flow conditions on the peripheral assemblies and the peripheral fuel has been susceptible to flow induced grid-to-rod fretting failures. The AREVA HTPTM / HMPTM configuration has been successful in preventing these types of fuel failures on the core periphery. St. Lucie Unit 1 has operated with this design for eight (8) cycles without failures. The HTPTM design provides eight (8) line contacts as the interface between the fuel rod and the spacer grid. This line Contact is very resistant to fuel rod failures from flow induced vibration fretting.

The HTPTM design is configured to improve heat transfer. As seen in Figure 1-4, the spring structure provides a flow path at an angle relative to the rod longitudinal direction, causing the water to swirl around the rod without creating a large pressure drop across the spacer. The HMPTM has the same line contact configuration but the channel is not angled. Since this spacer is at the lowermost position, the improved heat transfer is not necessary. As stated previously, the HMPTM material is Alloy 718. This material is very stable in irradiation environments, therefore providing additional assurance that the rod I spacer contact will be maintained throughout the design lifetime.

The assembly uses a MONOBLOCTM guide tube design for the corner guide tubes and a constant outer diameter and wall thickness design for the center guide tube. The MONOBLOC TM design maintains the same inner diameters in the dashpot and non-dashpot regions as the co-resident fuel, but has a constant outer diameter for the full length of the tube.

Therefore, the wall thickness in the dashpot region (about the bottom 14 inches of the guide

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 1-5 tube) is increased. The MONOBLOC TM guide tube design has been used for fuel reload batches in Europe and in the United States since 1998. The first application for CE plants was for a CE-14 design in 2010. St. Lucie Unit 1 has used the MONOBLOC TM guide tube design since 2013.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 1-6 UPPER TIE PLATE>

__UPPER END SPACER GRID (9x)

HTp TM

. FUEL S PELLET MONOBLOC TM GUIDE TUBE (4x)

LOWER END CAP HMPTM SPACER GRID FUELGUARD TM FUEL ROD ASSEMBLY LOWER TIE PLATE*--' (236x)

Figure 1-1: AREVA CE-16 Fuel Assembly for St. Lucie Unit 2

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 1-7 AOE

~Q~flON Figure 1-2: St. Lucie Unit 2 FUELGUARDTM Lower Tie Plate Figure 1-3: St. Lucie Unit 2 Upper Tie Plate

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 1-8 TOP O~ SPAcERC*tAM~6Z RAS CURVED FLOW Figure 1-4: St. Lucie Unit 2 HTPTM Spacer Grid

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-1 2.0 Mechanical Design 2.1 Introduction This section evaluates the mechanical design of the AREVA CE-16 HTP TM fuel design intended for batch implementation at St. Lucie Unit 2 and its compatibility with the co-resident fuel during the transition from mixed-fuel type core populations to cores with only AREVA CE-16 HTP TM fuel. AREVA has performed mechanical compatibility evaluations to assure acceptable fit-up with St. Lucie Unit 2 reactor core internals, fuel handling equipment, fuel storage racks, and co-resident fuel. A summary of the mechanical compatibility evaluations performed by AREVA is provided in Section 2.3.

The AREVA CE-16 HTP TM fuel assembly design for St. Lucie Unit 2 was analyzed in accordance with the USNRC-approved generic mechanical design criteria in EMF-92-1 16(P)(A)

(Reference 2) in conjunction with USNRC-approved topical report BAW-10240(P)(A) (Reference 1). Reference I incorporates the M5 cladding material properties that were previously approved by the USNRC in BAW-10227(P)(A) (Reference 3) into the AREVA mechanical design methodology (Reference 2). All the mechanical design criteria were shown to be met up to the licensed fuel rod burnup limit of 62 MWd/kgU.

Section 2.2 provides an overview of operating experience gained by AREVA with the various CE-16 and CE-14 plants. The operating experience of the various components was also discussed in Section 1.2. Section 2.3 provides a description of the mechanical compatibility assessments. Section 2.4 describes the mechanical evaluations performed to show acceptability with the USNRC approved generic design criteria.

2.2 Operational Experience of ARE VA HTPTM Fuel Assemblies in CE-16 and CE-14 Plants The St. Lucie 2 AREVA fuel design is very similar to the AREVA CE-14 HTP TM fuel design and the AREVA CE-16 HTP TM fuel design used by other plants. AREVA provides the fuel for all of the CE-14 units in the United States (St. Lucie Unit 1, Millstone Unit 2, Calvert Cliffs Units 1 and 2, and Ft. Calhoun). All but Ft. Calhoun are sister units with similar fuel features. The current AREVA design for CE-14 fuel for these sister units uses Zircaloy-4 HTP TM spacer grids at every elevation except the bottom grid. The bottom grid is an Alloy 718 HMP TM grid. The guide tubes are either currently a Zircaloy-4 MONOBLOCTM design or in the process of transitioning to a

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-2 Zircaloy-4 MONOBLOC TM design. The fuel rods either currently use M5 cladding or are in the process of transitioning to M5 cladding. The LTPs are the FUELGUARD TM design, and the UTPs are the AREVA reconstitutable design. The initial HTP TM / HMP TM I FUELGUARD TM transition began at St. Lucie Unit 1 in 2001 and that fuel design has operated for eight (8) cycles without failures. Fuel failures did occur at Millstone Unit 2 but this design did not have the lower Alloy-718 HMP TM grid. Since replacing the bottom grid at Millstone Unit 2 with an HMP TM grid, there have been no failures. Calvert Cliffs began their transition to the AREVA CE-14 HTP TM design in 2010. The AREVA fuel has not failed at the Calvert Cliffs units through the transition.

AREVA has supplied lead assemblies of CE-16 HTP TM fuel to Palo Verde Unit I and San Onofre Unit 2 (SONGS2). The Palo Verde lead assemblies completed their lifetime irradiation and have been discharged and examined. The SONGS2 fuel operated for one cycle (at both in-board and core-periphery locations) before the plant was closed for steam generator issues.

Both programs showed excellent fuel performance. The fuel rod at these units has the same radial dimensions and material as the St. Lucie Unit 2 fuel. However, the active fuel length in these lead assemblies was 150.0 inches instead of the 136.7 inches at St. Lucie. The cage structure is different at these two units, but the component features are similar to the standard CE-14 and St. Lucie Unit 2 AREVA designs. The lead assemblies had M5 cladding, MONOBLOC TM guide tubes (Palo Verde has a double expansion ID), HTP TM I HMP TM spacer grids, a FUELGUARDTM LTP (Palo Verde has the incore detectors entering from the bottom),

and an AREVA reconstitutable UTP (both lead assembly UTPs are much taller than the St.

Lucie 2 UTP). These lead assembly programs confirmed the excellent performance of the ARE VA design.

2.3 Mechanical Compatibility AREVA and Florida Power and Light (FPL) have performed an extensive review of the interfaces between the AREVA fuel assembly design and the plant equipment, the core interfaces, the control element assemblies (CEAs), the handling equipment, and the co-resident fuel. Where possible, the ARE VA design maintained the same interface dimensions as the co-resident fuel. Also, where possible, the AREVA design maintained the same configurations and functionality as the AREVA designed CE-14 fuel in St. Lucie Unit 1. Table 2-1 shows a comparison of the major dimensions of the St. Lucie Unit 2 AREVA design, the St. Lucie Unit 2

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision I Page 2-3 co-resident design, and the St. Lucie Unit 1 AREVA design. Additionally, a prototypic UTP was fabricated and tested successfully for compatibility with plant handling equipment.

Table 2-1: Comparison of Nominal Mechanical Design Features Feature St. Lucie 2 St. Lucie 2 St. Lucie 1 AREVA

_ _ _ BAREVA Design Westinghouse Design Design FeAsebyOeal158.529 158.529 157.115 Length, inch Bundle Pitch, inch 8.18 8.18 8.18 Number of Bundles in212727 Core Core Power, MWth 3020 3020 3020 Fuel Rod Overall Length, 146.60 146.899 145.77 inch Fuel Rod Pitch, inch 0.506 0.506 0.580 Number of Fuel Rods / 3 3 7 Assembly Number of Corner Guide444 Tubes / Assembly Number of Center Guide Tubes (Instrumentation111 Fuel Rod Cladding M5 - ZIRLOTM2 M5 (starting in MaterialCycle 26)

Fuel Rod Cladding Outer038032040 Diameter (OD), inch Fuel Rod Cladding 0.025 0.025 0.028 Thickness, inch Fuel Pellet Diameter, inch 0.3255 0.3255 0.3770 2 ZIRLO is a trademark of the Westinghouse Electric Company.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-4 Table 2-1: Comparison of Nominal Mechanical Design Features (continued)

FaueSt. Lucie 2 St. Lucie 2 St. Lucie 1 AREVA FaueAREVA Designj Westinghouse Design Design Fuel Stack Height (BOL, 136.70 136.70 136.70 cold), inch Axial Blanket Length (top / 6.00(U0 2 Rod) 6.00(UO 2 Rod) 6.00(UO2 Rod) bottom), inch 10.50 (NAF Rod) 10.50 (NAF Rod) 11.40 (NAF Rod)

CornerGidea ub Zircaloy-4 Zircaloy-4 Zircaloy-4 Manter Gia ub CenterGidea ub Zircaloy-4 Zircaloy-4 Zircaloy-4 Number of Grids 10 1 Bottom Grid Alloy 718 HMP TM GADInc 625 3 Nel 78HMT Upper Grids Zircaloy-4 HTP TM Grids) Zircaloy-4 HTP TM

________________________Inconel 625 (Top Grid) _________

2.3.1 Fuel Assembly The fuel assembly overall length was confirmed to be compatible with the dimensions of the core internals (spacing between core support plate and fuel alignment plate) at beginning of life cold and hot conditions. Additionally, positive engagement of the center/locking nuts and fuel alignment plate was demonstrated. An axial growth analysis confirmed adequate assembly to core internals and fuel rod I fuel assembly differential growth margins up to the licensed fuel rod and fuel assembly burnup limits.

The array type, the number of fuel rods and guide tubes, the fuel rod pitch dimensions, and the spacer grid centerline beginning of life elevations are the same as for the co-resident fuel.

These evaluations demonstrated that the AREVA design was compatible with the reactor components and co-resident fuel in the core. Additional evaluations of individual fuel assembly SGUARDIAN is a trademark of the Westinghouse Electric Company.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-5 components were also performed including the Upper Tie Plate, the Lower Tie Plate, and the Center and Corner Guide Tubes.

2.3.2 Upper Tie Plate The mechanical compatibility of the UTP is explicitly evaluated because it:

Interfaces with the holes in the fuel alignment plate in the reactor core

Interfaces with all the fuel assembly grapples when moving the fuel assembly

Interfaces with the control elements The UTP evaluations show that the UTP is mechanically compatible. Additionally, FPL has performed compatibility validation testing with the plant equipment using a prototypic UTP.

2.3.3 Lower Tie Plate The LTP also requires extensive compatibility evaluations because the LTP mates with the features (including alignment pins) of the lower core support plate. The AREVA LTP envelope is slightly smaller [ ] than that of the current St. Lucie Unit 2 fuel design, but is the same as the AREVA LTP used in St. Lucie Unit 1. All of the evaluations show that the LTP is compatible.

2.3.4 Guide Tubes Besides being the structural components of the fuel assembly, the guide tubes interface with the control rods. The radial positions of the guide tubes within the assembly, the inner diameters of the guide tubes, and the weep hole diameters of the AREVA design are the same as the co-resident fuel. The axial locations of the guide tube dashpot and weep holes are also similar to the co-resident design. These critical dimensions assure that control element assembly drop times and guide tube cooling are not significantly affected by the introduction of the AREVA fuel assemblies. The only significant difference is that the AREVA design uses MONOBLOC TM corner guide tubes which have a constant outer diameter as discussed in Section 1.2.

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ARE VA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-6 2.4 Mechanical Design Evaluations 2.4.1 Description The mechanical design evaluations are performed using the USNRC approved design methods and evaluated to the USNRC approved generic design criteria (Reference 2). Additional evaluations are included to address the impact of the thermal conductivity degradation with burnup and to address the impact of burnup on the seismic behavior of the fuel. The methods used for these additional evaluations are consistent with the methods previously reviewed by the USNRC for other applications and the updates to the generic criteria currently under USNRC review (References 4 and 5). (References 4 and 5 have recently been approved by the USNRC subsequent to Revision 0 of this document. The statements relevant to these references are still applicable.) These generic criteria are consistent with the specified acceptable fuel design limits (SAFDLs) identified in Chapter 4.2 of the Standard Review Plan (Reference 6). The USNRC-approved generic design criteria used to assess the performance of the fuel assemblies were developed to satisfy certain objectives (Reference 2). The use of M5 cladding required that the AREVA design methods be modified to incorporate the MS properties and generic design criteria be evaluated to assure continued applicability. This implementation was documented in Reference I and generically reviewed and accepted by the USNRC.

The fuel analyses are broadly separated into fuel rod analyses and structural analyses. The fuel rod analyses include evaluations of the SAFDLs such as internal rod pressure, cladding creep collapse, cladding fatigue, corrosion, etc. These evaluations are very dependent on the rod power. For the transition cycles analyzed for this amendment request, the power histories were created using expected typical cycle core designs projected to the design life of the fuel.

These cycle designs were created using the standard AREVA reload analysis codes and methods. The approved AREVA methodology requires these analyses to be redone for each cycle to assure that the actual cycle design will not result in SAFDL non-compliance. The actual reload cycle core designs will be performed by FPL using their standard, USNRC approved codes and methods. The LAR transition cycles are analyzed to demonstrate that the fuel design is acceptable and provide typical results showing SAFDL compliance. The specific reload results will be slightly different, but will continue to show SAFDL compliance.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-7 2.4.2 Input Parameters and Assumptions The input parameters used to perform the mechanical analyses included fuel design information derived from design documents, fuel assembly and component characteristics established by mechanical / hydraulic testing, plant parameters provided by FPL, fatigue duty cycles created using the fatigue transients provided in the UFSAR, and fuel rod power histories generated for the transition cycles by AREVA.

2.4.3 Recently Identified Analysis Issues As described in Section 2.4.1, the USNRC generically approved methods and criteria were used to evaluate the St. Lucia 2 AREVA fuel design (References 1 and 2). The USNRC has issued two Information Notices (IN), IN-2009-23 and IN-2012-09 (References 7 and 8), which identify issues that are not addressed in the previous reviews of the generic methods. The first IN (Reference 7) identified the non-conservative impact of the thermal conductivity degradation of the fuel pellets with irradiation. The second IN (Reference 8) identified concerns about the change in the fuel assembly seismic response from irradiation. As discussed below, these issues have been addressed in the St. Lucie 2 mechanical evaluations.

2.4.3.1 Thermal Conductivity Degradation (TOO)

As identified in Reference 7, at high burnup conditions, the thermal conductivity of uranium dioxide fuel is reduced. This reduction results in higher pellet temperatures, and results in a reduction in margins to various SAFOLs. To account for TOO effects, ARE VA has developed correction factors to be incorporated into evaluations using the currently approved RODEX2 code. These correction factors conservatively penalize the resulting margins for the affected SAFDLs to account for the thermal conductivity degradation. AREVA has submitted the correction factors generically to the USNRC in References 4 and 5. [

] (References 4 and 5 have recently been approved by the USNRC subsequent to Revision 0 of this document. The statements relevant to these references are still applicable.)

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-8 2.4.3.2 Seismic Evaluations As part of the transition, AREVA performed lateral and vertical seismic evaluations. The fuel assembly lateral seismic and LOCA evaluations included the sensitivity studies to address the impact of AREVA and the co-resident fuel in different core locations for the different row lengths in the core. The USNRC-approved methodology, defined in BAW-10133(P)(A) and Addenda 1 and 2 (Reference 10), was used for the evaluations. As a result of recent USNRC concerns with seismic behavior and feedback from recent AREVA submittals for other units, there were additional evaluations and modifications to the AREVA seismic methods.

The basic methodology for the lateral seismic analysis uses full assembly test data to benchmark the bundle design with the finite element code CASAC. Component tests are performed to determine component characteristics such as stiffness and strength. The time /

motion histories provided by the licensee are then imposed on this benchmarked model to determine the deflections of the fuel assemblies at the different core locations and the impact loads between the assemblies and between the assembly and the core shroud. The evaluations addressed the operating basis earthquake (OBE), the safe shutdown earthquake (SSE), and LOCA events. Each event was evaluated independently with lateral and vertical models.

USNRC Information Notice 2012-09, "Irradiation Effects on Spacer Grid Crush Strength,"

(Reference 8) identified the concern about the impact of the change in behavior of the assembly and assembly components during the operational lifetime. Additional testing and evaluations were included in the analyses to address this information notice. A simulated EOL fuel assembly and simulated EOL spacer grids were tested and used to benchmark EOL-specific CASAC models for both lateral and vertical analyses. These models were then applied in the same manner as the standard BOL models to evaluate impact loads and fuel assembly defiections during seismic and LOCA events.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-9 I

[

I

[

I 2.4.4 Mechanical Analyses Results The generic criteria (SAFDLs) for the fuel rod and fuel assembly are listed in Table 2-2 along with the corresponding section number from the criteria topical report (Reference 2) and with the LAR transition cycle results. As noted in the specific items, some of the criteria specified below are addressed in analyses other than the mechanical design evaluations.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-10

_______Table 2-2: Fuel Mechanical Design Evaluation Results Criteria Section Description Criteria Results 3.2 Fuel Rod Criteria Hydrogen content in components Controlled by manufacturing 3.2.1 Internal controlled to a minimum level seiiain n eiidb 321 Hydriding during manufacture to limit Qualifctyiontro inspvection.b internal hydriding.Qult otlinpco.

Sufficient plenum spring Cladding deflection and cold radial gap to Radial gap maintained throughout 3.2.2 Collapse prevent axial gap formation dniiain 3.2.2 Collapse during densification,.esfiain Table 5-1 demonstrates acceptance criteria are met.

Section 4.5.1.5 demonstrates this 95/9 conidece tat uel ods DNB performance is applicable to 3.2.3 Overheating do not experience DNB during transition mixed core of Cladding steady state or AQOs.cofgrtns Section 4.5.5 demonstrates the TM/LP trip and DNB LCO barn are effectively set.

Table 5-1 demonstrates that Overheating No centerline melting during acceptance criteria are met.

3.2.4 of Fuel normal operation and AQOs. Section 4.5.5 demonstrates the Pellets LPD LSSS and LPD LCO barns are effectively set.

3.2.5 Stress and Strain Limits Transient (AQO) strain:

U0 2 rod = 0.498%

NAF rod =0.468%

Pellet / For M5~cladding, strain < 1% Steady-state strain:

Cladding and no centerline melting. U 2 rd=036 Interaction NAF rod = 0.346%

See overheating of pellets (above) for temperature.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-11

_______Table 2-2: Fuel Mechanical Design Evaluation Results (continued)

Criteria Section Description Criteria Results ASME Section III, Division 1, Article 111-2000, in combination with the specified 0.2% offset yield strength and ultimateCopntmatismrgno Cladding strength of the unirradiated CompoEn mrtra.intinsmmarginto Stress cladding. M5 teslmtbsd 28%.

on bi-axial burst strength of cladding and buckling criteria at limiting overpressure at BOL.

Large break LOCA limiting case PCT results are lower than the Not underestimated during LOCA temperature threshold for clad 326 Cladding and used in determination of 10 rupture. Clad rupture did occur for

.26 Rupture CFR 50.46 criteria, the small break LOCA limiting case. Clad rupture effects are incorporated in the LOCA licensing results.

ASME Section III, Division 1, Article 111-2000, in combination Fuel Rod with the specified 0.2% offset 3..7Mehaicl yield strength and ultimate Criteria met with a minimum

3. FrMcancalin strength of the unirradiated margin of 24%.

Fractring cladding. M5' stress limit based on bi-axial burst strength of cladding.

Models included in USNRC Fuelapprvedfuelperfrmace cdesModels included in USNRC-Fuelapproveddfuellperformanceecodes.

3.2.8 Densification adtknioacutinSee Sections 3.2.2, 3.2.4, 3.2.5, analyses contained in Sections ad337o hstbe and Swlling 3.2.2, 3.2.4, 3.2.5, and 3.3.7 of Criteria met.

this table.

3.3 Fuel System Criteria________________

Stress, strain, and loading limits on assembly components. (See 3.3.9 for handling and 3.3.1 3.4 for accident conditions.)

SRP 4.2 Appendix A and ASME Margins:

Section III, Subsection NG for Normal operation + OBE = 17%

GuieTbe Normal Operation and SSE and Normal + SSE = 8%

_____ ________Appendix F for SSE+'LOCA Normal + SSE + LOCA = 3%

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-12 T~ kI~ ')J). ~. .~I IIw~k~ rij~,I fl~o iii ii ~i~Iu I~+uj~n Da~w iltc f,'ririfln, ie~ri1 Criteria Section Description Criteria Results Normal operation bounded by Spacer Grid Lateral load < load limit, handling criteria. Handling criteria

___________met with a margin of 62%.

Components maintain margin to Upper and Limiting loads occur during ASME criteria and approved Lower Tie handling and postulated topical. Shipping and handling Plates accidents, margins bounded by guide tubes with a margin of 34%.

CUE Results:

332 Cladding Cumulative usage factor (CUF) UO 2 rod = 0.635 332 Fatigue [ ]. NAF rod = 0.643 Criteria met.

3.3. Frttin wer Nofue rodfaiuresdueto fretting Supported by fretting test, 3.3.werFrttin Nofue rodfaiuresdueevaluation, and operational wear. experience.

Acceptable maximum oxide thickness. For M5 cladding, best Maximum best estimate oxide of Oxiatin, Oxdaio, estimate oxide < 100 microns.

Effects of oxidation and crud 24.8 microns.

Approved fuel rod performance 3.3.4 Hydriding, included in thermal and code accounts for oxidation and an Cud mechanical fuel rod analyses. crud buildup. Metal loss accounted Buildup Stress analysis to include metal for in stress analysis.

loss due to oxidation. Criteria met.

Lateral displacement of the fuel rods shall not be of sufficient Section 4.5.3 demonstrates that no 3..5Ro Bw magnitude to impact thermal rod bow penalty is required.

___________ margins.

3.3.6 Axial Irradiation Growth Clearance remains between fuel Fuel Rod rod and UTP/LTP at EOL. Criterion is met through design life.

The fuel assembly length shall not exceed the minimum space Fuelbeteenuppr an loer ore Criterion is met through design life.

Assembly plates in the cold condition at EOL.

~AJ~ DLI ~ ~t~)~L~I I I~I L AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-13

_______Table 2-2: Fuel Mechanical Design Evaluation Results (continued)

Criteria Section Description Criteria Results Acceptable maximum internal rod pressure. [ Maximum gas pressure:

U0 2 rod = 1678.6 psia Rod Internal NAF rod = 1456.7 psia Pressure Maximum values remain below criterion limit. Internal pressure

1. does not exceed system pressure.

3.3. AsembyN litof fro coe lwersuport Criterion is met for operation and 3.3.8__ AsmlifofNlftffrmcrloesupt.4th pump startup at 500 °F.

Components maintain margin to Fuel Assembly withstands 2 1/2 times ASME criteria. Anti-hangup HTP TM 3.3.9 Assembly tewihasattcfoe, spacer margin = 73%. The plenum Handling tewihasasaifocspring meets the handling design criteria.

3.4 Fuel Coolability Verification of spacer and guide tube structural integrity under seismic-LOCAk loading calculated based on AREVA + WV mixed-core Maintain coolable geometry and configurations ability to insert control rods. SRP Structural 4.2 Appendix A and ASME BOL spacer grid design margin 4 =

Deformations Section IIl, Appendix F, with 29% (for OBE = 0.3%)

lower Level A stress allowable for EOL spacer grid design margin =

the guide tubes under SSE. 0.4% (for OBE = 37%)

Guide tube margin = 8%

(for SSE only)

Guide tube margin = 3%

(for SSE+LOCA)

LOCA analysis peak clad temperature and maximum local 3.4.1 Cladding Include in LOCA analysis, cladding oxidation are well within Embrttleentlicensing limits, demonstrating protection from cladding

_______________________embrittlement.

Violent 3.42Epulionof < 230 cal/gm energy deposition Table 5-1 demonstrates

______ uel< 150 cal/gm for HZP conditions. acceptance criteria are met.

I 4BOL, OBE margin becomes 14% With revised allowable. See note in Table 2-3.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-14

_______Table 2-2: Fuel Mechanical Design Evaluation Results (continued)

Criteria Section Description Criteria Results The limiting small break LOCA Fuel onsier ipactof flow blockage transient experienced fuel swelling Fue3BalooConsiderOimpactyss and rupture for the hot rod; results Ballonin in OCA nalyis.are well within licensing limits; fuel

_______________________coolability is thus demonstrated.

4.1 Thermal and H ,draulic Criteria 4.1.1 Hydraulic Hydraulic flow resistance similar Hydraulic compatibility acceptable.

Compatibility to resident fuel assemblies. See Section 4.5.1.

Thermal Section 4.5.5 and Table 5-1 4.1.2 Margin 95/95 no DNB. demonstrates acceptance criteria Performance are met Fuel Section 4.5.5 and Table 5-1 4.1.3 Centerline No centerline melting. demonstrates acceptance criteria Temperature are met.

4.1.4 Rod Bow Protect thermal limits. Criterion is met. See Section 4.5.3 5.0 Neutronics Criteria 5.1 Poe nacrac ihTcncl Criterion is met. See Section 3.0.

Distribution jSpecifications.

5.2 Kinetic Parameters Doppler Reactivity Negative. Criterion is met. See Section 3.0.

_______ Coefficient ______________

Power Coefficient Negative relative to HZP. Criterion is met. See Section 3.0.

Moderator Tempeature In accordance with Technical Coefficenatu Speciiain Criterion is met. See Section 3.0.

5.3 Control Rod Technical Specification's margin Criterion is met. See Section 3.0.

_______ Reactivity maintained.________________

The fuel rod analysis results presented above in Table 2-2 include consideration of the fuel Thermal Conductivity Degradation (TCD) issue. Relevant results have been penalized to include TCD corrections. These corrections are consistent with or more conservative than, the generic penalties developed and submitted for USNRC review in References 4 and 5. [

LI ~ ~ I~ IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-15

] (References 4 and 5 have recently been approved by the USNRC subsequent to Revision 0 of this document. The statements relevant to these references are still applicable.)

2.4.4.1 Additional Seismic Analysis Results The AREVA fuel assembly design for St. Lucie Unit 2 has excellent seismic performance. The large corner guide tubes welded to the nine (9) HTPTM spacer grids creates a cage and bundle structure that has high assembly stiffness. The HTPTM spacer design is very sturdy while remaining flexible resulting in robust seismic performance. Therefore, it can more readily absorb the impacts without plastically deforming.

The St. Lucie 2 Seismic / LOCA evaluations included cases for all the different assembly rows in the core. The mixed core behavior was assessed by performing sensitivity analyses for the different rows with different positions for the AREVA and co-resident designs (including the all-AREVA and all-co-resident design cases). The limiting impact loads and margins for the AREVA assemblies occur in specific mixed core conditions in which the AREVA fuel is on the core periphery and adjacent to the co-resident fuel. These limiting cases are shown in Table 2-3. Based on the evaluations, the AREVA fuel assemblies meet design limits for both mixed core and full core conditions.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 2-16 Table 2-3: Seismic and LOCA Loadings Jlargin Row Layout 86% 9 & 15 assembly row FUll Core AAA.. .AA ARE VA 4 assembly row EOL [ ] [ ] AAAA Mixed BOL [ ] [ 1*

SCore EOL [ ] [ 37% I 17 assembly row AWW...WWVA SSE + LOCA Load Allowable Margin Row Layout

... L 6o 4 assembly row BOL [ ] 11[assembly row 15 assembly row EOL [ ] [ ] 29%AW..WV EO .% 17 assembly row I___EOL ___[ ___________ 0.4%__ AWAW... WAWA

This grid allowable has been updated to [ ] based on the inclusion of additional crush test data for St. Lucie 2 specific grid type. The margin for the BOL, OBE, Mixed Core limiting case increases to 14% and the margin for the BOL, OBE, Full Core case increases to 88%.

2.5 Mechanical Design Conclusions The AREVA CE-16 HTP TM fuel design is mechanically compatible with the co-resident fuel design, the plant structures, and fuel handling / interfacing equipment and structures at St. Lucie Unit 2. The AREVA CE-16 HTP TM fuel design has been analyzed in accordance with USNRC-approved mechanical design criteria using transition cycle inputs. Adaptations to the methodologies have been identified and explained to address USNRC Information Notices and to align with recently approved submittals. All of the design criteria were shown to be met up to the licensing fuel rod burnup of 62 MWd/kgU under normal and faulted operating conditions.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 3-1 3.0 Nuclear Design 3.1 Introduction The licensing basis for the reload core nuclear design is defined in UFSAR Section 4.3. The purpose of the core analysis is to verify that the cycle-specific reload design and the key safety parameters are properly addressed in the reload analysis. The effects of transitioning from Westinghouse CE 16x16 fuel to AREVA CE-16 HTP TM fuel on the nuclear design bases and methodologies for St. Lucie Unit 2 are evaluated in this section.

3.2 Input Parameters The AREVA St. Lucie CE1:-16 HTP TM fuel differs from that of existing Westinghouse CE 16x16 fuel design, with the unique features as described in Sections 1.2 and 2.3. Refer to Section 4.5.1.5, for discussion of the application of a mixed core penalty to the departure from nucleate boiling (DNBR) safety limits. The power distribution effects are discussed in the specific analyses presented in Section 5.1.

3.3 Methodology The nuclear design methodology and codes are updated to include the standard AREVA methodology and code package for the transition cycles and future operation of AREVA St.

Lucie CE-16 HTP TM fuel. References 12, 14, and 15 are the USNRC-approved topical reports outlining the approved AREVA neutronics methodology and codes.

The safety evaluation report (SER) for Reference 12 requires that application of the methodology to a CE-16 fuel assembly design be supported by additional validation and that this validation be maintained by ARE VA and available for USNRC audit. This SER requirement has been met for St. Lucie Unit 2.

Benchmarking of the AREVA neutronics methodology and codes was performed and demonstrated acceptable modeling of previous and current St. Lucie Unit 2 cores. [

]

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 3-2 Key safety parameters are calculated as part of the core design neutronics analysis (see Table 3-1). These parameters are then biased in the safety analysis. Key safety parameters are then calculated for the cycle-specific reload and compared with the values used in the safety analysis. These cycle-specific parameters will be generated based on AREVA methodology using both AREVA codes and the current set of codes used by FPL. If the key parameters are not within the reference safety analysis, then the transient will be re-analyzed or re-evaluated on a cycle-to-cycle basis using the stated methods.

Table 3-1: Range of Key Safety Parameters Technical Transition Analysis Safety Parameter Specification Value Nominal Reactor Core Power TS 1.25 3020 (MWt)

TS 3.2.5 Vessel Average Coolant Inlet 551 COLR Table 3.2-2 Temp HFP (0 F)

Nominal Coolant System 25 Not a TS 25 Pressure (psia)

< +5 (Power -- 70%)

Most Positive Moderator < 0 (Power = 100%)

TS 3.1.1.4 Temperature Coefficient (MTC) Linear ramp from (pcm/°F) +5 at 70% to 0 at 100%

COLR Section 2.1 Most Negative MTC (pcm/°F) -33 Not TSDoppler a Temperature Coefficient (DTC) (pcm/0 F) (See footnote5 ) -. 0t 13 Not a TS Beta-Effective (See footnote5 ) 0.0052 to 0.0065 TS 3.2.3 Normal Operation HFP Unrodded

< 1.65 COLR Section 2.5 FmT (without uncertainties)

COLR --3600 (->200°F)

Shutdown Margin (pcm)

Section 2.8 and 2.9 --3000 (< 200°F) 5 Btaeffctveand DTC do not have analyses or TS limits directly associated with them. These parameters are major contributors to transient analysis behavior and are good early indicators of significant physics characteristics changes in the core. Current design values for these parameters are expected ranges only.

'~AJI iLl IJII~ LJ~JLA~JI I ~l ~L AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision I Page 3-3 Table 3-1 : Range of Key Safety Parameters (continued)

Technical Transition Analysis Safety Parameter Specification Value TS 3.2.1 Linear Heat Rate (kW/ft) <--13.0 COLR Section 2.4 TS 3.2.5 DNB LCO Axial Shape Index > -0.08 COLR Section 2.6 (100% Power) < 0.15 BOC HZP: 4.867 Maximum Ejected Rod, EQ BOC HFP: 2.681 NoaS(See footnote6 ) EOC HZP: 8.781 EOC HFP: 2.320 BOC HZP: 24.9 Total Deposited Enthalpy, BOC HFP: 144.1 NoaS(cal/gm) (See footnote 6) EOC HZP: 26.9 EOC HFP: 136.9 3.4 Description of Design Evaluations Standard nuclear design analytical models and methods (Reference 12) accurately describe the neutronic behavior of the AREVA St. Lucie CE-16 HTP TM fuel. The specific design bases and their relation to the GDCs in 10 CFR 50, Appendix A for the AREVA St. Lucie CE-16 HTP TM design are discussed in Reference 2.

The effect of extended burnup on nuclear design parameters has been previously approved in detail in Reference 13. That discussion is valid for the AREVA St. Lucie CE-16 HTP TM discharge burnup level.

A transition core design and two additional follow-on core designs have been developed for St.

Lucie Unit 2 to model the transition to AREVA St. Lucie CE-16 HTP TM fuel. The loading patterns were developed based on design requirements (e.g. energy, peaking, and assembly placement) for St. Lucie Unit 2. The loading patterns were depleted at a core power of 3020 MWt. These cycles were not developed to be bounding of future cycle designs, but were developed to be representative of future cycle designs to demonstrate acceptable margins. The 6 The control rod ejection analysis values do not have TS limits directly associated with them. The design values listed are expected based on the transition.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 3-4 first transition cycle contains fresh AREVA St. Lucie CE-16 HTP TM fuel with once-burnt and twice-burnt Westinghouse CE 16x16 fuel. The second transition cycle contains fresh and once-burnt AREVA St. Lucie CE-16 HTP TM fuel with twice-burnt Westinghouse CE 16x16 fuel. The third transition cycle contains only AREVA St. Lucie CE-16 HTP TM fuel. These models show that enough margin exists between typical safety parameter values and the corresponding limits to allow flexibility in designing actual reload cores. Table 3-2 contains key information based on the nominal transition cycle designs. Key safety parameters were verified for the core design in Table 3-1.

The standard methods of fresh fuel enrichment loading and integrated burnable poisons will be applied to control the peaking factors and maintain compliance with the Technical Specifications and COLR. Changes in boron concentration and axial offset are typical of normal cycle-to-cycle variations in the core design.

Table 3-2: Projected Transition Cycle Core Characteristics Number of Maximum HFP ARO FrT Maximum HFP ARC FQ Cycle______ __

Transition Feed Cce Energy AEAAREVA Westinghouse AREVA Westinghouse (EFPD) FulFuel Fuel Fuel Assemblies Fe N 518.7 88 1.538 1.220 1.859 1.428 N+1 515.9 84 1.571 1.312 1.894 1.567 N+2 504.9 84 1.556 N/A 1.858 N/A 3.5 Results Margin to key safety parameter limits (Table 3-1) is maintained during the transition from Westinghouse CE 16x16 fuel to AREVA St. Lucie CE-16 HTP TM fuel.

The changes in fuel design and discharge burnup result in only a small impact on the results of the reload transition core analysis relative to the current design. The variations in these parameters are typical of the normal cycle-to-cycle variations that occur as fuel loading patterns are changed each cycle.

IL~ LAEI~~ L'~JYLH I E~I IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 3-5 Changes to the core power distributions and peaking factors are the result of the normal cycle-to-cycle variations in core loading patterns. These will vary cycle-to-cycle based on actual energy requirements. The normal methods of feed enrichment variation and insertion of fresh burnable absorbers will be employed to control peaking factors. Compliance with the peaking factor TS will be assured using these methods.

3.6 Conclusion The nuclear core design analysis of the core design for the transition from Westinghouse CE 16x16 fuel to AREVA St. Lucie CE-16 HTP TM fuel has confirmed peaking factor and key safety parameters can be maintained within their specified limits using only AREVA methodologies and codes. The key safety parameters generated with the core design are used in the applicable analyses and evaluated to meet the acceptance criteria.

The key safety parameters and the peaking factor limits will be verified on a cycle specific basis.

However, the values are planned to be created using FPL and AREVA methods.

C ~

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-1 4.0 Thermal and Hydraulic Design 4.1 Description This section describes the Thermal Hydraulic (T-H) analysis supporting the transition to AREVA CE-16 HTP TM Fuel at St. Lucie Unit 2.

4.2 Input Parameters and Assumptions XCOBRA-IIlC is the core T-H sub-channel analysis code that was used for the AREVA HTP TM fuel analysis. USNRC approval of the XCOBRA-IIIC code was issued in the SER attached to Reference 20.

For the Thermal Hydraulic analysis, fuel-related safety and design parameters of the AREVA CE-16 HTPTM fuel design have been used. These parameters have been used in safety and design analyses discussed in this section and in other relevant sections of this LAR.

Table 4-1 lists T-H parameters used for the fuel transition thermal-hydraulic analysis.

Table 4-1 : Thermal-Hydraulic Design Parameters Parameter Value Reactor core heat output, MWt 3020 Heat generated in fuel, % 97.5 Pressurizer/core pressure, psia 2250 Nominal vessel/core inlet temperature, °F 551 RCS minimum flow rate (including bypass), gpm 370,000 Core bypass flow, % 4.2 Core area, ft2 54.39 Core inlet mass velocity (excluding bypass, based on TS minimum flow 2.45 rate, 106 Ibm/hr-ft2 Pressure drop across core, psi (full-core AREVA CE-16 HTP TM ) [ ]

Core average heat flux, kW/ft 5.2 The limiting directions for biased parameters are shown in Table 4-2. Biases were applied to input parameters according to the approved methodology (Reference 21). For the transient analyses, uncertainties were deterministically applied. Thus, steady-state measurement and instrumentation errors were taken into account in an additive fashion to ensure a conservative

U ~JU ~U ~

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-2 analysis. For statistical departure from nucleate boiling (DNB) calculations, uncertainties were statistically treated according to the approved methodology (Reference 19). The system related uncertainties bounded by the non-loss of coolant accident (non-LOCA) safety analyses are listed in Table 4-3.

Table 4-2: Limiting Parameter Directions Parameter Limiting Direction for DNB Reactor core heat output (MWt) maximum Heat generated in fuel (%) maximum Nominal vessel / core inlet temperature maximum Fr, enthalpy rise hot channel factor maximum Pressurizer/core pressure (psia) minimum RCS flow (See note 1 below) (gpm) minimum Note 1: The limiting (minimum) value of the RCS flow is the TS minimum flow.

Table 4-3: System Related Uncertainties Parameter Uncertainty Reactor Thermal Power +/-0.3% (at 100% RTP)

ROS Flow +12,500 gpm RCS Pressure +/-+45.0 psi Core Inlet Temperature +/-+3.0 °F Control grade equipment was modeled in such a way that it does not mitigate the effects of an event. The reactor trip setpoints and time delays modeled in the transient analyses were conservatively applied to provide bounding simulations of the plant response. To the extent that the reactor protection system and engineered safety features system are credited in the accident analyses, the setpoints have been verified to adequately protect the plant for the fuel transition.

4.3 Acceptance Criteria The reactor core is designed to meet the following limiting T-H criteria:

'~s!JEiLI1JH~U ~JLPYL$Ai~~IIL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision I Page 4-3

There is at least a 95% probability at a 95% confidence level that DNB will not occur on the limiting fuel rods during Modes 1 and 2, operational transients, or any condition of moderate frequency.

No fuel melting during any anticipated normal operating condition, operational transients, or any conditions of moderate frequency.

The ratio of the heat flux causing DNB at a particular core location, as predicted by a DNB correlation, to the actual heat flux at the same core location is the DNBR. Analytical assurance that DNB will not occur is provided by showing the calculated DNBR to be higher than the 95/95 limit DNBR for conditions of normal operation, operational transients and transient conditions of moderate frequency.

4.4 Method of Analysis The T-H analysis of the AREVA CE-I16 HTP TM fuel is based on the approved methodologies for performing DNB calculations (References 25 and 21). The S-RELAP5 code was used for the transient analysis. The XCOBRA-IIIC code was used to calculate minimum DNBR (MDNBR) using the HTP and Biasi critical heat flux (CHF) correlations. RODEX2-2A (References 9 and

22) was developed to perform calculations for a fuel rod under normal operating conditions.

For non-LOCA applications, RODEX2-2A was used to establish the fuel centerline melt linear heat rate (LHR) as a function of exposure. The HTP DNB correlation is based entirely on rod bundle data and takes credit for the significant improvements in DNB performance due to the flow mixing nozzles effects. USNRC acceptance of a 95/95 HTP correlation safety limit DNBR of 1.141 for HTP CHF Correlation is documented in Reference 18. The Biasi CHE correlation (Reference 26) is used to calculate the DNBR for post-scram reactor conditions. The 95/95 Biasi correlation safety limit DNBR used in analysis is [ ]. The ranges of parameters used in TM the AREVA CE-16 HTP design have been verified to fall within the range of applicability for these correlations.

[

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-4 The approved methodology for performing DNB calculations using the XCOBRA-IIIC code in a mixed core is described in Reference 25. The SER for the Reference 20 topical report states that the use of XCOBRA-IIIC is limited to the "snapshot" mode. Thus, MDNBR calculations were performed using a steady-state XCOBRA-IIIC model with core boundary conditions at the time of MDNBR from the S-RELAP5 transient analyses.

The Reference 19 topical report describes the method for performing statistical DNB analyses.

Two conditions were noted in the SER for the Reference 19 methodology:

The methodology is approved only for Combustion Engineering (CE) type reactors which use protection systems as described in the Reference 19 topical report.

The methodology includes a statistical treatment of specific variables in the analysis; therefore, if additional variables are treated statistically, Siemens Power Corporation, now AREVA, should re-evaluate the methodology and document the changes in the treatment of the variables. The documentation will be maintained by AREVA and will be available for USNRC audit.

Protection against the fuel centerline melting (FCM) SAFOL is expressed as a limit on LHR allowed in the core. The FCM limit was explicitly calculated for the AREVA fuel transition. Due to the reduced thermal conductivity of gadolinia fuel rods, the FCM limit may be set by gadolinia fuel. A FCM limit is established for UO2 fuel rods such that, FCM is precluded for all fuel rod types. A penalty to address thermal conductivity degradation (TCD) was applied where applicable.

The impact of rod bowing on the MDNBR and peak LHR was evaluated using the rod bow methodology described in Reference 27.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-5 4.5 Results 4.5.1 Thermal-Hydraulic Compatibility 4.5.1.1 Core Pressure Drop The Westinghouse fuel assemblies have a lower overall resistance to flow than the AREVA HTPTM fuel assemblies; therefore, as the core transitions from a full core of Westinghouse fuel to a full core of AREVA fuel, the core pressure drop increases. An analysis was performed to assess the change in core pressure drop associated with the fuel transition.

The core pressure drop for a full core of ARE VA HTPTM fuel assemblies is [ ]

The total pressure drop associated with the full core of AREVA HTP TM fuel is []

than the total pressure drop of the Westinghouse core. The pressure drop profile between the two assembly types is shown in Figure 4-1.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-6 Figure 4-1: Pressure Drop Profiles 4.5.1.2 Total Bypass Flow The change in total bypass flow was examined to determine if the active heat transfer coolant flow will be adversely impacted by the fuel transition. The bypass flow includes the following flow paths: guide tubes, vessel upper head, inlet-to-exit nozzle, and core barrel/baffle. The change in total bypass flow was determined by examining the change due to non-guide tube paths and guide tube paths. Bypass flow for the non-guide tube paths is affected by changes in

Eli ~JEE~U LJ~J~'Ul ~E IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-7 core pressure drop, while guide tube bypass flow is dependent on both core pressure drop and assembly geometry.

The core pressure drop for a full core of AREVA fuel is higher than the core pressure drop for a Westinghouse core. As a result, the driving force for bypass flow increases and the total bypass flow increases. [

4.5.1.3 Crossflow Velocity The Inter-Assembly Crossflow velocities affecting the AREVA HTPTM fuel assemblies were analyzed to assure satisfactory performance during the transition. Different core configurations were considered in the analysis, ranging between bounding configurations with a single AREVA assembly and a single Westinghouse assembly.

Although other geometries and operating conditions may result in different crossflow velocity profiles, the analyzed scenario provides representative crossflow velocities to cover core configurations associated with the fuel transition. The results are representative of anticipated operating conditions and are used to develop bounding inputs for mechanical analyses.

4.5.1.4 RCS Flow Rate An analysis was performed to assess the change in primary system loop flow attributed to the fuel transition. The change in the Reactor Coolant System (RCS) loop flow will not impact the Technical Specification minimum loop flow rate.

]

4.5.1.5 Transition Core DNB Performance XCOBRA-IIIC was used to analyze the effect of the fuel transition on the DNB performance of the AREVA CE-16 HTP TM fuel assemblies. The power level was selected to achieve MDNBR

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-8 close to the HTP CHE correlation limit. A mixed core penalty was applied to all core configurations, including the full core of AREVA HTP TM fuel.

The AREVA HTPTM fuel assembly is associated with more overall flow resistance than the co-resident Westinghouse fuel. This results in flow transferring from the AREVA HTPTM fuel into the Westinghouse fuel, which is detrimental to DNB performance of the AREVA fuel. [

] The impact will decrease for subsequent transition cycles.

4.5.1.6 Control Rod Drop Times An assessment was performed to validate that the Technical Specification requirement for the control rod drop time is not challenged as a result of the fuel transition. The control rod drop time is primarily dependent on the number, size, and location of the guide tube weep holes, as well as the inner diameter and height of the guide tube dashpot region.

Due to the similarities between the Westinghouse and AREVA guide tube designs, the control rod drop times will not be significantly impacted by the fuel transition and will remain below the required drop time of 3.25 seconds.

4.5.2 Thermo-Hydrodynamic Instability AREVA has evaluated the St. Lucie reactor for its susceptibility to a wide range of potential thermo-hydrodynamic instabilities, It concludes that St. Lucie Unit 2 will not experience thermo-hydrodynamic instabilities during normal operation and AOOs.

4.5.3 Rod Bow The impact of rod bowing on the MDNBR and peak LHR was evaluated using the rod bow methodology described in Reference 27. The objective was to determine the threshold burnup level at which a rod bow penalty must be applied to either the MDNBR or peak LHR results.

The results show that no rod bow penalty is required for DNB or LHR calculations.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-9 4.5.4 Guide Tube Heatinq Boiling of coolant within the guide tubes has the potential to increase corrosion rates and be detrimental for neutron moderation. An analysis was performed to demonstrate that boiling will not occur within the guide tubes of the AREVA fuel assemblies. For conservatism, severe operating conditions were used in the analysis.

Guide tube heating is most severe when a neutron absorbing material is inserted into the guide tube. The analysis considered a high powered assembly with the control rods at PDlL conditions. The analysis demonstrates that control rod linear heat generation rates less than or equal to 9.2 kW/ft will preclude boiling within the guide tube.

4.5.5 Setpoint Analyses The setpoint analyses ensure there is sufficient margin for the Limiting Safety System Settings (LSSS) and Limiting Condition for Operation (LCO) systems that monitor various reactor system variables designed to protect the SAFDLs and other design limits. The results of the setpoint analyses are presented in Table 4-4.

Table 4-4: Minimum Margin Summary for Setpoint Calculations Setpoint Analysis Margin LPD LCO (see note 1 below) 1.2%

LPD LSSS 29%

TM/LP LSSS 4 psid DNB LCO LOCF 5%

DNB LCO CEAD 5%

Note: The setpoints are verified every cycle based on cycle specific core design Note 1: Applicable only when Incore Monitoring System is unavailable.

The TS LSSS are designed to scram the reactor if the monitored parameters reach values that are conservatively set to protect the fuel SAFOLs. The LSSS include reactor trips such as thermal margin/low pressure (TM/LP), local power density (LPD) LSSS, variable high power trip (VHPT), low flow trip, and component pressure and water level trips. The analyses discussed in this section verified the TM/LP and LPD LSSS trip settings.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-10 The TS LOOs provide requirements for parameters associated with the DNB LCO and LPD LCO. The DNB LCO is designed to protect the DNB SAFDL. The LPD LCO is more restrictive and is designed to protect against the LOCA linear heat generation rate (LHGR) limit when the incore detectors are not in service.

The methodology used in the setpoint verification analyses has been approved by the USNRC and is described in Reference 19.

The LPD LSSS barn and results are presented in Figure 4-2 and Figure 4-3, respectively. The TM/LP trip functions analyzed are presented in Figure 4-4 and Figure 4-5. The DNB LCO barn.

and results of the transient simulations are presented in Figure 4-6, Figure 4-7, and Figure 4-8, respectively. The LPD LCO barn and results are presented in Figure 4-9 and Figure 4-10, respectively. The verification of DNB LCO, LPD Leo, TM/LP LSSS and LPD LSSS is redone for each reload to ensure margin to SAFDLs.

The LSSS and LCO functions are unchanged from the current TSICOLR settings.

140 120 100 ci.

a 0O ci) 60 40

-0.6 -0.4 -0.2 "-0 0.200 0.400 0.600 Axial Shape Index (ASI)

Figure 4-2: LPD - High Trip Setpoint

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request AN P-3352N P Technical Report Revision 1 Page 4-11 J~ ~

r V~6.

EG 0

Figure 4-3: LPD LSSS Verification Results

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-12 1.25 1.00 0

t0.75 I--

0.2 0.000.00 '

0.25 0.50 0.75 1.00 1.25 Power (% of Rated)

Figure 4-4: TM/LP Trip Setpoint - QR1 Function

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-13 1.75

,q,1.50 a.

I-J 1.25 1.00-0.75

'~L 0.25 0.50 0.75 1.00

-0.50 -0.25 0.00 Axial Shape Index (AS I)

Figure 4-5: TM/LP Trip Setpoint - QA Function

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-14 1 10 . . . . . . . . .,

100 90 70

60 0

0.

50 4

40 20 .

0-

'_5 0o4 ~03 41.2 4)1 o O.i o0? 413 0.4 415 o05 AKial Shape Ind~ex (aslu)

Fi igure 4-6: ASI Limits for DNB vs. Thermal Power Lo~

£3F3 -E3 110 10oo 90

a. 0 30 10 0-

-.4.7 4.16 -0.5 -0.4 -0.3 -0.2 -.0i 0 0.1 (12 0.3 0.4 0.5 0k6 03*

Axial Shape Index (a~Iu)

Figure 4-7: DNB LCO CEAD Results

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request AN P-3352N P Technical Report Revision 1 Page 4-15 12.0%

110%

00o 000 0 90%

g I'

60%

40%

-0~6 ~0A ~02 0 0.2 04 0.6 ASI (%)

Figure 4-8: DNB LCO LOCF Results 70

~.0 I

10 O ..

02 0.3 0,4 0.5 06 Figure 4-9: ASI Limits for LHR vs. Maximum Allowable Power Level when Using the Excore Detectors

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-16 140 130 4=.1 110 11)0 90., 'i I

80 p

70 5 5 40 40 05,40 10 0

0:7 06 05 04 0.3 02 io, 0 L:*I 2* 1e3 0*4 05 0.6 01 AE*a Sbap. idex {a*u)

Figiure 4-10: LPD LCO Verification Results 4.5.5.1 Thermal Margin Safety Limit Line Verification The Thermal Margin Safety Limit Lines (TMSLLs) at St. Lucie Unit 2 are a series of isobars in power and inlet temperature that establish the operating frontiers in power and temperature at each pressure such that DNB in the core and hot leg saturation are both nominally avoided.

The St. Lucie Unit 2 TMSLLs are nominally based lines and therefore are analyzed using nominal values for all parameters without accounting for uncertainties.

The St. Lucie Unit 2 TMSLLs are verified using the following approach:

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St. Lucie Unit 2 Fuel Transition License Amendment Request AN P-3352N P Technical Report Revision 1 Page 4-17 Each isobar is made up of two regions. The first, flatter region is established by hot leg saturation and the second, steeper portion is established by DNB. The axial shape used is the TMSLL design basis shape for St. Lucie Unit 2 and is shown in Figure 4-12.

))

The TMSLLs presented in Figure 4-11 are the same as in the current TS and were verified to be conservative for use with the HTP correlation for the St. Lucie Unit 2 transition.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-18 I

z FRA(T~Q~O~ AATW ThERMAL FOWt*

Figure 4-11: Thermal Margin Safety Limit Lines

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 4-19 S 1.8 1.6 1.4 0.8 06 cli 0.6 0.2 0 0.1 0.2 0.3: 0.4* 0.5 0.5, 0.7' 0.8 0.9i

... Percent of Active Core Height from Bottom:

Figure 4-12: Axial Power Distribution for Thermal Margin Limit Lines

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision I Page 5-1 5.0 Accident and Transient Analyses 5.1 Non-LOCA Analyses 5.1.1 Introduction This section provides information related to the St. Lucie Unit 2 nuclear power plant transient and accident analyses for the proposed transition to AREVA fuel. It includes a brief description of methodology used to evaluate the St. Lucie Unit 2 UFSAR Chapter 15 events affected by the transition to AREVA fuel. Also, a discussion is included on the basis by which the St. Lucie Unit 2 UFSAR Chapter 15 events not affected by the transition to AREVA fuel have been dispositioned. A summary report that provides a detailed description of analyses for the non-LOCA events using the ARE VA methodology is found in Reference 23.

5.1.2 Computer Codes Descriptions of the principal computer codes used in the non-LOCA transient analyses are provided below.

S-RELAP5 The S-RELAP5 (Reference 21) code is an AREVA modification of the RELAP5/MOD2 code. S-RELAP5 is used for simulation of the transient system response to loss-of-coolant accident (LOCA) as well as non-LOCA events. Control volumes and junctions are defined which describe all major components in the primary and secondary systems that are important for the event being analyzed. The S-RELAP5 hydrodynamic model is a two-dimensional, transient, two-fluid model for flow of a two-phase steam-water mixture. S-RELAP5 uses a six-equation model for the hydraulic solutions. These equations include two-phase continuity equations, two-phase momentum equations, and two-phase energy equations. The six-equation model also allows both non-homogeneous and non-equilibrium situations encountered in reactor problems to be modeled.

I AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision I Page 5-2 RODEX2-2A For non-LOCA applications, RODEX2-2A (References 9 and 22) is used to establish the fuel centerline melt linear heat rate (LHR) as a function of exposure as part of the Thermal Hydraulics portion of the AREVA fuel transition, which is discussed in Section 4.0.

COPERNIC COPERNIC (Reference 24) performs thermal-mechanical calculations for a fuel rod under normal operating conditions. The code incorporates models to describe the thermal-hydraulic condition of the fuel rod in a flow channel; the gas release, swelling, densification and cracking in the pellet; the gap conductance; the radial thermal conduction; the free volume and gas pressure internal to the fuel rod; the fuel and cladding deformations; and the cladding corrosion.

The code has been extensively benchmarked and its predictive capabilities were correlated over a wide range of conditions applicable to light water reactor fuel conditions.

COPERNIC accounts for thermal conductivity degradation (TCD) with increasing rod exposure.

To account for the effects of TCD in the non-LOCA S-RELAP5 simulations, COPERNIC was used to generate the fuel thermal-conductivity, heat capacity and fuel pellet-to-clad gap coefficient inputs for the average core and hot spot models. The properties from COPERNIC were developed for beginning-of-cycle (BOO) and end-of-cycle (EOC) conditions in accordance with Reference 21 and replaces RODEX2 for this purpose in the approved topical report. The COPERNIC fuel properties and gap coefficients were conservatively implemented relative to the RODEX2 inputs as approved in Reference 21.

XCOBRA-IIIC The XCOBRA-IIIC analyses are performed as part of the Thermal Hydraulics portion of the AREVA fuel transition, which is discussed in Section 4.0.

5.1.3 Analysis Methodologqies The approved AREVA methodology for evaluating non-LOCA transients is described in Reference 21. For each non-LOCA transient event analysis, the nodalization, chosen parameters, conservative input and sensitivity studies are reviewed for applicability to the fuel

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St. Lucie Unit Technical 2 Fuel Transition License Amendment Request Report ANP-3352NP Revision 1 Page 5-3 transition in compliance with the SER for Revision 0 of the non-LOCA topical report (Reference 21).

The nodalization used for the calculations supporting the fuel transition is specific to St.

Lucie Unit 2 and is in accordance to the (Reference 21) methodology.

The parameters and equipment states are chosen to provide a conservative estimate of the challenge to the acceptance criteria. The biasing and assumptions for key input parameters are consistent with or more conservative relative to the approved Reference 21 methodology.

The S-RELAP5 code assessments in Reference 21 validated the ability of the code to predict the response of the primary and secondary systems to Chapter 15 non-LOCA transients and accidents. No additional model sensitivity studies are needed for this application.

The method used for the non-LOCA system transient analyses differs from that in the approved Reference 21 topical report as described below:

[

]

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 5-4 Another change allowed by the Reference 21 methodology was to replace RODEX2 with COPERNIC for the purpose of generating the fuel thermal-conductivity, heat capacity and fuel pellet-to-clad gap coefficient inputs for the average core and hot spot models. This change was made to explicitly account for the effects of TCD. The properties from COPERNIC were developed for BOC and EOC conditions in accordance with Reference 21 and COPERNIC replaces RODEX2 for this purpose in the approved topical report. The COPERNIC fuel properties and gap coefficients were conservatively implemented relative to the RODEX2 inputs as approved in Reference 21.

Reference 1 incorporates M5 properties into the S-RELAP5 based non-LOCA methodology.

No restrictions or requirements were identified in the SER for the Reference 1 methodology relative to its application to S-RELAP5 non-LOCA analyses.

The approved methodology for calculating the enthalpy deposition for a CEA ejection accident is given in Reference 14. No restrictions or requirements were identified in the SER for this methodology.

5.1.4 Event Disposition and Analysis Reference 23 summarizes the Chapter 15 non-LOCA safety analyses supporting the transition to AREVA fuel. The analyses provide the required elements to demonstrate applicability of the method to St. Lucie Unit 2 and addresses the SER requirements as discussed in Section 5.1.3.

A review of each UFSAR Chapter 15 event was conducted relative to the transition to AREVA fuel.

Several events (or subevents) are affected by the transition to AREVA fuel, specifically because of changes in thermal hydraulic performance and neutronics inputs to the safety analyses. The events (or subevents) that challenge the non-LOCA fuel related criteria, i.e., DNB and fuel centerline melt, were analyzed using the AREVA safety analysis methodology (Reference 21), as supplemented in Section 5.1.3. In addition, event specific criteria, i.e., time-to-criticality for Boron Dilution and deposited enthalpy for CEA Ejection, were analyzed with the Reference 21 and Reference 14 methodologies, respectively. The following events were analyzed for the fuel transition with respect to the fuel related criteria:

FL! ~)II~~J ~J~LFF FI~F EL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 5-5 o Feedwater System Malfunctions That Result in a Decrease in Feedwater Temperature (UFSAR 15.1.1) o Feedwater System Malfunctions That Result in an Increase in Feedwater Flow (UFSAR 15.1.2) o Excessive Increase in Secondary Steam Flow (UFSAR 15.1.3) o Pre-Trip Steam System Piping Failure (UFSAR 15.1.5) o Post-Trip Steam System Piping Failure (UFSAR 15.1.6) o Loss of Condenser Vacuum (UFSAR 15.2.3) o Loss of Load to One Steam Generator (UFSAR 15.2.9) o Complete Loss of Forced Reactor Coolant Flow (UFSAR 15.3.2) o Reactor Coolant Pump Shaft Seizure (UFSAR 15.3.3) o Uncontrolled CEA Bank Withdrawal from a Subcritical or Low Power Startup Condition (UFSAR 15.4.1) o Uncontrolled CEA Bank Withdrawal at Power (UFSAR 15.4.2) o CEA Misoperation (Dropped CEA) (UFSAR 15.4.3) o Chemical and-Volume Control System (CVCS) Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant (UFSAR 15.4.6) o Spectrum of CEA Ejection Accidents (UFSAR 15.4.8) o Inadvertent Opening of a Pressurizer Safety or Relief Valve (UFSAR 15.6.1)

Other UFSAR Chapter 15 events (or subevents) are not affected by the AREVA fuel transition because the key parameters for these events are plant related system responses (e.g., core power, decay heat, auxiliary feedwater capability, offsite power availability, safety valve setpoints and capacities, safety injection and/or charging

~iU~ IU ~JFI~U LJ~tJLtUI I ~ IL ARE VA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 5-6 capability, etc.) rather than the fuel design parameters. These events (or subevents) challenge criteria other than the SAFDLs, e.g., system overpressure. As such, these events will not be analyzed as part of the transition to ARE VA fuel. These events (or subevents) remain bounded by the current analyses of record.

Reference 23 (Section 2.0) provides the key input parameters assumed for the non-LOCA analyses. A summary of the initial conditions assumed for each Chapter 15 non-LOCA event that was analyzed using S-RELAP5 to support the fuel transition is provided in Reference 23 (Table 2.3). Reference 23 (Table 3.1) provides a summary of the non-LOCA disposition of events. Reference 23 (Section 4.0) discusses each UFSAR Chapter 15 event in detail. The results in Reference 23 demonstrate that acceptance criteria are met for each non-LOCA event that was analyzed for the transition to AREVA fuel. The results are summarized in Table 5-1.

_______ ~~Table 5-1: Non-LOCA Limiting Results _______

UFSAR Analytical Section Event Description Criterion Limit Limiting Result 15.1.1 Decrease in Feedwater MDNBR 1.164 1.257 Temperature Peak LHR, kW/ft [ ] 18.24 15.1.2 Increase in Feedwater MDNBR 1.164 1.220 Flow Peak LHR, kW/ft [ ] 18.50

___________Peak CLT, °F [ ] 3385 (HZP) 15.1.3 *Increase in Steam MDNBR 1.164 1.271 Flow Peak LHR, kW/ft [ ]19.12 Peak CLT, °F j 3491 (HZP) 15.1.5 Pre-scram Main Steam MDNBR (%fuel failure) 1.164 1.203 (0%)

Line Break Peak LHR, kW/ft (% fuel* [ ] 17.67 (0%)

failure) 15.1.6 Post-scram Main MDNBR (% fuel failure) [ 1 1.740 (0%)

Steam Line Break Peak LHR, kW/ft (% fuel I" 1 17.02 (0%)

________failure) 15.2.3 Loss of Condenser MDNBR -. 1.164 1.553 Vacuum Peak LHR, kW/ft [ ]16.04 15.2.9 Transients Resulting MDNBR 1.164 1.713 from the Malfunction of Peak LHR, kW/ft [ 1 15.74 One Steam Generator________

15.3.2 Loss of Forced Reactor MDNBR 1.164 1.227

_______Coolant Flow 15.3.3 Reactor Coolant Pump MDNBR (% fuel failure) 1.164 1.205 (0%)

_______ Rotor Seizure _____________ _______

~AJ~ ELI LJII~U LJtJ~UI ~ IL ARE VA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision I Page 5-7

_______ ~ Table 5-1: Non-LOCA Limiting Results (Continued) _______

UFSAR Analytical Section Event Description Criterion Limit Limiting Result 15.4.1 Uncontrolled CEA MDNBR 1.164 1.994 Wtdaafrma Peak CLT, 0F [ ] 3194 Subcritical or Low Power Startup Condition 15.4.2 Uncontrolled CEA MDNBR 1.164 1.177 Withdrawal at Power Peak LHR, kW/ft [ ] 16.43 15.4.3 CEA Misoperation/CEA MDNBR 1.164 1.554 Drop Peak LHR, kW/ft [ ] 15.71 15.4.6 CVCS Malfunction that Min. time to loss of shutdown 15 15.08 Results in a Decrease margin, mai. 30 30.59 in the Boron Concentration in the Reactor Coolant/Boron Dilution 15.4.8 CEA Ejection MDNBR (% fuel failure) 1.164 1.179 (0%)

Peak CLT, 0F (% fuel failure) [ 1 4876 (0%)

Total deposited enthalpy limit, 230 (HFP) 144.1 (HFP)

____cal/gm 150 (HZP) 26.9 (HZP) 15.6.1 Inadvertent Opening of MDNBR 1.164 1.237 Pressurizer Safety or

_______ Relief Valve__ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _

5.1.5 Conclusions The non-LOCA transient analyses were performed in accordance with the Reference 21 non-LOCA methodology, as supplemented in Section 5.1.3. Reference 23 demonstrates the application of the AREVA non-LOCA safety analysis methodology to St. Lucie Unit 2 for the fuel transition and shows that acceptance criteria are met for each non-LOCA event that was analyzed for the transition to ARE VA fuel.

5.2 Loss-of-Coolant Accident Analyses The loss-of-coolant accident (LOCA) is analyzed to assure that the design bases for the Emergency Core Cooling System (ECCS) satisfy the requirements of 10 CFR 50.46 acceptance criteria for the St. Lucie Unit 2 transition to AREVA fuel. Summary reports that provide a detailed description of supporting small break LOCA and realistic large break LOCA (SBLOCA and RLBLOCA) analyses are found in References 28 and 29, respectively. Additional results

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AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 5-8 supporting the SBLOCA analysis are found in Reference 32 (responses to SNPB RAI-11I through SNPB RAI-20).

5.2.1 Small Break Loss-of-Coolant Accident A SBLOCA is defined as a break in the RCS pressure boundary which has an area of up to approximately 10% of a cold leg pipe area. The most limiting break location is in the cold leg pipe on the discharge side of the reactor coolant pump, which results in the largest amount of inventory loss and the largest fraction of ECCS fluid being lost to the break. This behavior produces the greatest degree of core uncovery and the longest fuel rod heatup time.

The SBLOCA event is characterized by a slow depressurization of the RCS with a reactor trip occurring on a low pressurizer pressure signal. The safety injection actuation signal (SIAS) occurs when the system pressure continues to drop. For some of the break sizes, the rate of Sinventory loss from the primary system is such that the charging system and High Pressure Safety Injection (HPSI) pumps cannot preclude significant core uncovery. The slow RCS depressurization rate extends the time required to reach the safety injection tank (SIT) pressure Sor to recover core liquid level on charging and HPSI flow. Core recovery for the limiting break begins when the charging and HPSI flow to the RCS exceeds the mass flow rate out of the break, followed by injection of SIT flow.

The AREVA SBLOCA evaluation methodology (EM) simulates thermal-hydraulic response of the primary and secondary systems and hot fuel rod and requires the use of two computer codes, S-RELAP5 and RODEX2/2A (Reference 16). The appropriate conservatisms, as prescribed by Appendix K of 10 CFR 50, are incorporated. The EM has been reviewed and approved by the USNRC to perform SBLOCA analyses.

Results from the St. Lucie Unit 2 SBLOCA analysis show that the 10 CER 50.46(b) acceptance criteria for PCT, maximum oxide thickness, and hydrogen generation are met with significant margin. Analysis results show that the limiting PCT occurred for a 2.70-inch diameter cold leg pump discharge break. This case yielded a limiting PCT of 2057 °F as provided in Reference 32 (response to SNPB RAI-15). The transient maximum local oxidation is less than 9%. The total maximum local oxidation is less than 12%, including a pre-transient oxidation of 2.3925%. The maximum core-wide oxidation is less than 0.3%.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 5-9 5.2.2 Largqe Break Loss-of-Coolant Accident A large break loss-of-coolant accident (LBLOCA) is initiated by a postulated large rupture in the RCS cold leg. The RCS depressurizes rapidly and the reactor is shut down by coolant voiding in the core. An SIAS occurs on either high containment pressure or low RCS pressure.

Pumped ECCS and passive SIT fluid injection actuates to mitigate the transient.

The St. Lucie Unit 2 RLBLOCA analysis is performed by applying the S-RELAP5, RODEX3A, and ICECON computer codes. The EM is documented in Reference 30; specific alternative methods to the EM are outlined in the RLBLOCA summary report. These alternative methods are a response to USNRC inquiries related to the methodology updates to the EM. This altered methodology is referred to as the "transition program or transition package." This methodology follows the Code Scaling, Applicability, and Uncertainty evaluation approach (Reference 31),

which outlines an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifies the uncertainties for the RLBLOCA analysis. The approach described in the summary report has been used successfully in multiple applications for support of licensing AREVA fuel transitions.

Results from the St. Lucie Unit 2 RLBLOCA analysis show that the 10 CFR 50.46(b) acceptance criteria for PCT, maximum oxide thickness, and hydrogen generation are met with significant margin. Analysis results show that the limiting PCT occurred for a fresh UO2 rod in a case with no offsite power availability. This case yielded a limiting PCT of 1732 °F.

EU 1jjj~~ L~j~.jL~t I I~I IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 6-1 6.0 Summary and Conclusion This report shows acceptability for the application of the AREVA CE-16 HTP TM fuel design at St.

Lucie Unit 2. The results displayed within the report show compliance of the AREVA CE-16 HTPTM fuel design with USNRC-approved topical reports regarding mechanical and structural analyses, nuclear design analyses, thermal-hydraulics analyses for steady state and transient core performance, and non-LOCA / LOCA safety analyses addressing transient and accident conditions. Alternative methods to the approved topical reports are conservatively applied and clearly described within the document, where appropriate.

Note that demonstration of the evaluation methodologies has been performed with a submittal core design. The submittal core design was developed to provide key safety parameters to support the transition from Westinghouse fuel to AREVA CE-16. HTP TM fuel prior to the development of cycle-specific designs. This provides assurance that the plant licensing bases are met for the anticipated operation of the AREVA CE-16 HTPTM fuel during the transition and full core cycles.

The AREVA fuel design will be the CE-16 HTP TM fuel consisting of a 16x16 assembly configuration with M5 fuel rods, Zircaloy-4 MONOBLOC TM corner guide tubes, an Alloy 718 HMPTM spacer at the lowermost axial elevation, Zircaloy-4 HTPTM spacers in all other axial elevations, a FUELGUARD TM lower tie plate (LTP), and the AREVA reconstitutable upper tie plate (UTP).

The AREVA CE-16 HTP TM fuel design for St. Lucie Unit 2 is similar and has the same design features as the AREVA CE-14 fuel design operating in St. Lucie Unit 1. It is also similar to the AREVA CE-16 HTP TM lead fuel assemblies operated in San Onofre Unit 2 as well as the fuel rods operated in the AREVA CE-16 HTP TM Palo Verde Lead Fuel Assemblies. The design features of the AREVA CE-16 HTP TM fuel design planned for St. Lucie Unit 2 have demonstrated excellent fuel performance. The HTPTM / HMPTM spacer grids are very resistant to flow induced grid-to-rod fretting failures, the FUELGUARD TM LTP is effective at protecting the fuel from debris in the reactor coolant system, and the M5 cladding has very low oxidation and hydrogen pickup rates.

In conclusion, this report supports the use of AREVA CE-i16 HTP TM fuel at St. Lucie Unit 2.

'AJF ILl !JEI~U LJLA~I I ~ IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 7-1 7.0 References

1. BAW-1 0240(P)(A), Revision 0, "Incorporation of M5 Properties in Framatome ANP Approved Methods."

2. EMF-92-1 16(P)(A), Revision 0, "Generic Mechanical Design Criteria for PWR Fuel Designs."

3. BAW-1 0227(P)(A), Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel."

4. EMF-92-116(P)(A), Revision 0, Supplement 1, Revision 0(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs." (Supplement 1 is pending approval).

5. Letter NRC:14:049, P.Salas (AREVA Inc.) to USNRC, "Response to a Request for Additional Information Regarding EMF-92-11I6(P)(A), Revision 0, Supplement 1, Revision 0."

6. NUREG-0800, Revision 2, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition."

7. USNRC Information Notice, IN-2009-23, "Nuclear Fuel Thermal Conductivity Degradation."

8. USNRC Information Notice, IN-2012-09, "Irradiation Effects on Spacer Grid Crush Strength."

9. XN-NF-81-58(P)(A), Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model."

10. BAW-1 01 33(P)(A), Revision 1, and Addenda I and 2, "Mark C Fuel Assembly LOCA-Seismic Analysis."

11. BAW-1 01 72(P)(A), Revision 0, "Mark-BW Mechanical Design Report."

12. EMF-96-029(P)(A), Volumes I and 2, "Reactor Analysis System for PWRs, Volume 1 Methodology Description, Volume 2 Benchmarking Results."

13. ANF-88-1 33(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU."

14. XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors."

15. XN-75-27(A) and Supplements 1 through 5, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors", Exxon Nuclear Company, Report and Supplement 1 dated April 1977, Supplement 2 dated December 1980, Supplement 3 dated September 1981 (P), Supplement 4 dated December 1986 (P), and Supplement 5 dated February 1987 (P).

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 7-2

16. EMF-2328(P)(A) Revision 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based."

17. EMF-2087(P)(A), Revision 0, "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications."

18. EMF-92-1 53(P)(A), Revision 1, "HTP: Departure From Nucleate Boiling Correlation for High Thermal Performance Fuel."

19. EMF-1 961 (P)(A) Revision 0, "Statistical SetpointlTransient Methodology for Combustion Engineering Type Reactors."

20. XN-NF-75-21(P)(A), Revision 2, "XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient Core Operation."

21. EMF-2310(P)(A), Revision 1, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors."

22. ANF-81-58(P)(A), Revision 2 and Supplements 3 and 4, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model."

23. ANP-3347(P), Revision 0, "St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report."

24. BAW-1 0231 (P)(A) Revision 1, "COPERNIC Fuel Rod Design Computer Code."

25. XN-NF-82-21 (P)(A), Revision 1, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations.

26. Energia Nucleare, Volume 14, No. 9, September 1967, "Studies on Burnout, Part 3 - A New Correlation for Round Ducts and Uniform Heating and Its Comparison with World Data" L. Biasi et. al.

27. XN-75-32(P)(A), Supplements 1, 2, 3, and 4, "Computational Procedure for Evaluating Fuel Rod Bowing."

28. ANP-3345(P), Revision 1, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report."

29. ANP-3346(P), Revision 0, "St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report."

30. EMF-2103(P)(A), Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors."

31. NUREG/CR-5249, EGG-2552, Technical Program Group, "Quantifying Reactor Safety Margins," October 1989.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 1 Page 7-3 S32. AN-40PRevision 1,"St. Lucie Unit 2 Fuel Transition: Responses to NRC Questions SRXB-RAI-1 and SNPB RAI-2 thru SI*JPB RAI-20."

v t e Letter Sequence Request
TAC:MF5494, (Open) TAC:MF5495, (Open)
Initiation Request, Request, Request, Request, Request, Request Acceptance... Administration Withholding Request Acceptance Questions 13 July 2015 4 September 2015 Answers Results Approval... Other: L-2015-177, Fuel Transition Small Break LOCA Summary Report, ANP-3345NP, Revision 1, L-2015-206, ANP-3428NP, Revision 0, St. Lucie Unit 2 Fuel Transition: Response to SNPB-RAI-1, Attachment 4 to L-2015-206, L-2015-259, ANP-3440NP, Revision 1, Response to NRC Questions SRXB-RAI-1 and Snpb RAI-2 Thru Snpb RAI-20., L-2015-279, RAI Response Clarification for SRXB-RAI-1 and Snpb RAI-2 Thru RAI-20 for the Technical Specification LAR and Exemption Request Regarding the Transitioning to Areva Fuel, ML15197A339, ML15219A184, ML15279A226, ML15356A184, ML15356A185
MONTHYEAR2015-06-18 [Table View]

L-2015-300, ANP-3352NP, Revision 1, Transition License Amendment Request, Technical Report.

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