Text

ANP-3352NP, Revision 0, St. Luice, Unit 2, Fuel Transition License Amendment Request, Technical Report
	ML15002A092
Person / Time
Site:	Saint Lucie
Issue date:	12/31/2014
From:	; AREVA, AREVA
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML15002A105	List: L-2014-366, ANP-3352NP, Revision 0, St. Luice, Unit 2, Fuel Transition License Amendment Request, Technical Report; L-2014-366, St. Lucie, Unit 2 - Application for Technical Specification Change and Exemption Request Regarding the Transitioning to Areva Fuel; ML15002A091;
References
	L-2014-366 ANP-3352NP, Rev 0
	Download: ML15002A092 (528)
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L-2014-366 Attachment 7 AREVA Reports Non-Proprietary Versions ANP-3352NP ANP-3347NP ANP-3345NP ANP-3346NP Following 527 Pages

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AREVA St. Lucie Unit 2 Fuel Transition License ANP-3352NP Revision 0 Amendment Request Technical Report December 2014 AREVA Inc.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page ii Nature of Changes Section (s) Description and Justification Item or Page(s) 1 All This is a new document.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page iii Contents 1.0 Introduction and Design Overview ................................................................................. 1-1 1 .1 In tro d u ctio n ......................................................................................................... 1 -1 1.2 Fuel Design Overview ........................................................................................ 1-2 2.0 Mechanical Design ......................................................................................................... 2-1 2.1 Introduction ......................................................................................................... 2-1 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 Assem bly ...................................................................................... 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 Assum ptions ..................................................... 2-7

- 2.4.3 Recently Identified Analysis Issues ...................................................... 2-7 2.4.3.1 Therm al Conductivity Degradation (TCD) ............................ 2-7 2.4.3.2 Seism ic Evaluations ............................................................. 2-8

- 2.4.4 Mechanical Analyses Results ............................................................... 2-9 2.4.4.1 Additional Seism ic Analysis Results .................................. 2-15 2.5 Mechanical Design Conclusions ....................................................................... 2-15 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 Therm al and Hydraulic Design ....................................................................................... 4-1 4.1 Description ......................................................................................................... 4-1 4.2 Input Parameters and Assum ptions ................................................................... 4-1 4.3 Acceptance Criteria ............................................................................................ 4-2 4.4 Method of Analysis ............................................................................................. 4-3 4 .5 Re s u lts ............................................................................................................... 4 -5

- 4.5.1 Thermal-Hydraulic Com patibility .......................................................... 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 Therm o-Hydrodynam ic 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 0 Page iv 4.5.5 Setpoint Analyses ................................................................................. 4-9 4.5.5.1 Therm al Margin Safety Lim it 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 Com puter 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 Sm all Break Loss-of-Coolant Accident ................................................. 5-8

- 5.2.2 -Large Break Loss-of-Coolant Accident ................................................. 5-8 6.0 Sum m ary 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 0 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 - Mixed Core ............................................................ 2-15 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: Lim iting Param eter D irections ................................................................................... 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-LO CA Lim iting R esults ...................................................................................... 5-6

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page vi Figures Figure 1-1: AREVA CE-16 Fuel Assembly for St. Lucie Unit 2 .................................................. 1-5 Figure 1-2: St. Lucie Unit 2 FUELGUARD M T

Lower Tie Plate ................................................... 1-6 Figure 1-3: St. Lucie Unit 2 U pper Tie Plate .............................................................................. 1-6 M

T Figure 1-4: St. Lucie Unit 2 HTP Spacer Grid ........................................................................ 1-7 Figure 4-1: P ressure D rop P rofiles ............................................................................................ 4-6 Figure 4-2: LP D - High Trip S etpoint ........................................................................................ 4-10 Figure 4-3: LPD LSSS Verification Results .............................................................................. 4-11 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: D NB LC O C EA D Results ....................................................................................... 4-14 Figure 4-8: D NB LC O LO C F R esults ....................................................................................... 4-15 Figure 4-9: ASI Limits for LHR vs. Maximum Allowable Power Level when Using the E xco re Detectors .................................................................................................... 4-15 Figure 4-10: LPD LCO Verification Results ............................................................................. 4-16 Figure 4-11: 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 0 Page vii Nomenclature AOO ........................................... Anticipated Operational Occurrence ARO ........................................... All Rods Out ASI ............................................. Axial Shape Index ASME ......................................... American Society of Mechanical Engineers BOC ........................................... 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 E O L............................................ E nd of Life FCM ........................................... Fuel Centerline Melt FPL ............................................. Florida Power and Light FQ .......................... . . . .. . . . . .. . . .. . . .. . . Total Power Peaking Factor Fr ................................................ Assembly Radial Peaking Factor GDC ........................................... General Design Criteria HFP ............................................ Hot Full Power HMP TM ...................... . . . . .. . . .. . . .. . . . . High Mechanical Performance HPSI ........................................... High Pressure Safety Injection HTPTM ........................................ High Thermal Performance HZP ............................................ Hot Zero Power ID................................................ Inner Diam eter IN................................................ Information Notice LAR ............................................ License Amendment Request LBLOCA ..................................... Large Break Loss-of-Coolant Accident LCO ............................................ Limiting Condition for Operation LHGR ......................................... Linear Heat Generation Rate LHR ............................................ Linear Heat Rate

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page viii Nomenclature (continued)

LOCA ......................................... Loss-of-Coolant Accident LOCF .......................................... Loss of Coolant Flow LPD ............................................ Local Power Density LSSS .......................................... Limiting Safety System Setting LTP ............................................. Low er Tie Plate MDNBR ...................................... Minimum Departure from Nucleate Boiling Ratio MSSV ......................................... Main Steam Safety Valve MTC ........................................... Moderator Temperature Coefficient NAF .................... Neutron Absorbing Fuel NRC .......................................... Nuclear Regulatory Commission, OBE ............................................ Operating Basis Earthquake O D.............................................. O uter Diam eter 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 ............................................. Therm al 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 W PR ........................................... Wetted Perimeter Ratio

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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 (HTPTM1) fuel starting in the spring of 2017. The AREVA fuel design will be the CE-16 HTP M T

fuel consisting of a 16x16 assembly configuration with M51 fuel rods, Zircaloy-4 MONOBLOCTM1 corner guide tubes, an Alloy 718 High Mechanical Performance (HMPTM1) spacer at the lowermost axial elevation, Zircaloy-4 HTPTM spacers in all other axial elevations, a FUELGUARDTM1 lower tie plate (LTP), and the AREVA reconstitutable upper tie plate (UTP).

The AREVA CE-16 HTP M T

fuel design for St. Lucie Unit 2 is similar and has the same design features as the AREVA CE-14 HTP M T

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 MT 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 0 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.

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 M T

/ HMP M T

spacer grids, M5 fuel rod cladding, the FUELGUARD MT 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 M5 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 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.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 1-3 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 M T

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-14 and CE-16 fuel examinations.

The cage or skeleton uses four (4) Zircaloy-4 MONOBLOC M T corner guide tubes, one (1)

Zircaloy-4 center guide tube, nine (9) Zircaloy-4 HTPTM spacers, and one (1) Alloy 718 HMP M

T spacer at the lowest spacer position. The HTPTM spacers are welded directly to the five guide tubes whereas the HMP M T

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 M

T 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

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 1-4 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 / HMP M T

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 HTP M T 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 / spacer contact will be maintained throughout the design lifetime.

The assembly uses a MONOBLOC TM guide tube design for the corner guide tubes and a constant outer diameter and wall thickness design for the center guide tube. The MONOBLOCTM 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 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 M T guide tube design since 2013.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 1-5 UPPER TIE PLATE

<_UPPER END CAP

, PLENUM SPRING M

T HTP SPACER GRID (9x)'

<-CLADDING

___ FUEL 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-1 6 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 0 Page 1-6 BADE Figure 1-2: St. Lucie Unit 2 FUELGUARD TM 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 0 Page 1-7 TOP OFSPACtHM cuLOWCHANEL 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 0 Page 2-1 2.0 Mechanical Design 2.1 Introduction This section evaluates the mechanical design of the AREVA CE-16 HTP M T

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 HTPTM 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-92-116(P)(A)

(Reference 2) in conjunction with USNRC-approved topical report BAW-10240(P)(A) (Reference 1). Reference 1 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 OperationalExperience of AREVA 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 M T

spacer grids at every elevation except the bottom grid. The bottom grid is an Alloy 718 HMP M T

grid. The guide tubes are either currently a Zircaloy-4 MONOBLOC M T 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 0 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 MT design, and the UTPs are the AREVA reconstitutable design. The initial HTP M T

/ HMP M T

/ 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 M T

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 M

T 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 1 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 / HMP M T

spacer grids, a FUELGUARD TM 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 AREVA design.

2.3 MechanicalCompatibility 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 AREVA 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 0 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 IAREVA Design Westinghouse Design Design Fuel Assembly Overall 158.529 158.529 157.115 Length, inch Bundle Pitch, inch 8.18 8.18 8.18 Number of Bundles in 217 217 217 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 /236 236 176 Assembly Number of Corner Guide 4 4 4 Tubes / Assembly Number of Center Guide Tubes (Instrumentation 1 1 1 Tubes) / Assembly Fuel Rod Cladding M5 ZIRLOTM2 M5 (starting in Material Cycle 26)

Fuel Rod Cladding Outer 0.382 0.382 0.440 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 0 Page 2-4 Table 2-1: Comparison of Nominal Mechanical Design Features (continued)

Feature St. Lucie 2 St. Lucie 2 St. Lucie I AREVA AREVA Design Westinghouse Design Design Fuel Stack Height (BOL, 136.70 136.70 136.70 cold), inch Axial Blanket Length (top / 6.00(UO 2 Rod) 6.00(UO 2 Rod) 6.00(UO 2 Rod) bottom), inch 10.50 (NAF Rod) 10.50 (NAF Rod) 11.40 (NAF Rod)

Corner Guide Tube aloy-4 Zircaloy-4 Zircaloy-4 Material Center Guide Tube Zircaloy-4 Zircaioy-4 Zircaloy-4 Material Number of Grids 10 10 9 Bottom Grid Alloy 718 HMP M T

Inconel62) Alloy 718 HMPTM T

_____________ (GUARDIAN TM3) Alo71HP Zircaloy-4 HID-1L (Mid 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 / 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 3 GUARDIAN is a trademark of the Westinghouse Electric Company.

%..jLJI L U1I;U LJULaUMM IIIL AREVA Inc.

St, Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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 M

T assemblies. The only significant difference is that the AREVA design uses MONOBLOC corner guide tubes which have a constant outer diameter as discussed in Section 1.2.

fU IIZL)J LJUtjUi I Ik:I IL OU.L1 AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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). 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 M5 properties and generic design criteria be evaluated to assure continued applicability. This implementation was documented in Reference 1 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.

11"Pul Ito UIIk::;U uuqýlul I It:;1 It AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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. Lucie 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 (TCD)

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 SAFDLs. To account for TCD effects, AREVA 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. [

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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 deflections during seismic and LOCA events.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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.

VIul sU5I1L.J LJutlJljI I M11I L AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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 specifications and verified by Hydriding during manufacture to limit Quality Control inspection.

internal hydriding.

Sufficient plenum spring Cladding deflection and cold radial gap to Radial gap maintained throughout 3.2.2 Collapse prevent axial gap formation densification.

during densification.

Table 5-1 demonstrates acceptance criteria are met.

Section 4.5.1.5 demonstrates this 95/95eatin onfridence tt fuelros DNB performance is applicable to 3.2.3 Overheating do not experience DNB during transition mixed core of Cladding steady state or AOOs. confitions.

configurations.

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 (AOO) strain:

U0 2 rod = 0.498%

Pellet/ For M5 cladding, strain < 1% NAF rod y 0.468%

Cladding and no centerline melting. Steady-state strain:

Interaction U02 rod 0.346%

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 0 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 ultimate Component maintains margin to Cladding strength of the unirradiated ASME criteria. Minimum margin is Stress cladding. M5 stress limit based 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 3.2.6 Cladding and used in determination of 10 rupture. Clad rupture did occur for 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 3.2.7 Mechanical yield strength and ultimate Criteria met with a minimum Fracturing strength of the unirradiated margin of 24%.

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

Models included in USNRC Fuel approved fuel performance codes Fuelapproveddfuellpeerormanceecodes. aModels included in USNRC-3.2.8 Densification and taken into account in See Sections 3.2.2, 3.2.4, 3.2.5, and Swelling analyses contained in Sections and 3.3.7 of this table.

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 = 18.4%

Normal Operation and SSE and Normal + SSE = 12.9%

Appendix F for SSE+LOCA Normal + SSE + LOCA = 5%

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 2-12 Table 2-2: Fuel Mechanical Design Evaluation Results (continued)

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%.

CUF Results:

33.2 Cladding Cumulative usage factor (CUF) U0 rod = 0.635 Fatigue [ ]. NAF2 rod = 0.643 Criteria met.

rNo fuel rod failures due to fretting Supported by fretting test, 3.3.3 Fretting wear Nuelro evaluation, and operational wear. experience.

Acceptable rfiaximum oxide thickness. For M5 cladding, best Maximum best estimate oxide of Oxidation, estimate oxide < 100 microns. 24.8 microns.

Oxridaion, Effects of oxidation and crud Approved fuel rod performance 3.3.4 Hydriding, included in thermal and code accounts for oxidation and and 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 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 Fuel between upper and lower core Criterion is met through design life.

Assembly plates in the cold condition at EOL.

%.elU OL~I U1.O LJUU.UI I Mi11 It AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 2-13 Table 2-2: Fuel Mechanical Design Evaluation Results (continued)

Criteria SeCction 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

]. does not exceed system pressure.

338 Assembly N from core lower support. Criterion is met for operation and 3.3.8 A mLiftoff No liftoff 4th um startup at 500 'F.

Components maintain margin to Fuel . AM rtra nihnu T T 3.3.9 Assembly Assembly withstands 2 1/2 times ASME criteria. Anti-hangup HTPTM Handling the weight as a static force. spacer margin = 73%. The plenum Handlingspring meets the handling design criteria.

3.4 Fuel Coolability Verification of spacer and guide tube structural integrity under seismic-LOCA loading calculated based on AREVA + W 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 =

Deformations Section III, Appendix F, with 41% (for OBE = 34%)

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

the guide tubes under SSE. 31% (for OBE = 59%)

Guide tube margin = 12.9%

(for SSE only)

Guide tube margin = 5%

(for SSE+LOCA)

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

Violent 3.uioent < 230 cal/gm energy deposition Table 5-1 demonstrates Euel< 150 cal/gm for HZP conditions, acceptance criteria are met.

Fuel

%--,. I II ICUL L-gi~ k.Ul 0IQ:I IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 2-14 Table 2-2: Fuel Mechanical Design Evaluation Results (continued)

Criteria Description Criteria Results Section ______

The limiting small break LOCA FuelConide imact f fow locagetransient experienced fuel swelling 3.4.3 Fuel Consider impact of flow blockage and rupture for the hot rod; results 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 Power In accordance with Technical.

5.1 Distr Distribution Sccans.

In Specifications. Criterion is met. See Section 3.0.

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 In accordance with Technical Temperature Specification. Criterion is met. See Section 3.0.

Coefficient 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. [

-UPI1 LI UHDIU, LJUL, D IMI*II AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 2-15 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 HTP TM 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.

Table 2-3: Seismic and LOCA Loadings - Mixed Core

_ _OBE Loads OBE Allowable OBE Margin Row Layout BOIL 133.8 15 assembly row AWWV... WWA EOL r58.8 17 assembly row I AWWV...WWA SSE+LOCA SSE+LOCA SSE+LOCA Row Layout Loads Allowable Margin Rowaou BOIL 40.7[ 15 assembly row I AWW...WWA EOL 30.9 17 assembly row EOL_ _ [_ _]i ] 30[9 AWAW... WAWA 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 M T

fuel design has been analyzed in accordance with USNRC-

LJIIllt~ I-~L JU6LOUI I IUI IL.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 2-16 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.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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 T M

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 CE-16 HTP M T

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 AREVA 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 0 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 Safety Parameter Transition Analysis 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 (OF)

Not a TS Nominal Coolant System 2250 25 Pressure (psia)

< +5 (Power < 70%)

Most Positive Moderator 5 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 a TS Doppler Temperature Coefficient (DTC) (pcm/°F) (See footnote 4)

Not a TS Beta-Effective (See footnote 4) 0.0052 to 0.0065 TS 3.2.3 Normal Operation HFP Unrodded T < 1.65 COLR Section 2.5 FrT (without uncertainties)

COLR Shutdown Margin (pcm) Ž 3600 (> 200°F)

Section 2.8 and 2.9 > 3000 (< 200°F) 4 Beta-effective and 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.

%'.epulPLO Ullt.;u LJUIýOul I B-11 Ot AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 3-3 Table 3-1: Range of Key Safety Parameters (continued)

Technical Safety Parameter Transition Analysis Specification Value TS 3.2.1 GOLR Section 2.4 Linear Heat Rate (kW/ft) -<13.0 TS 3.2.5 DNB LCO Axial Shape Index > -0.08 COLR Section 2.6 (100% Power) < 0.15 BOC HZP: 4.867 Not a TS Maximum Ejected Rod, FQ BOG HFP: 2.681 (See footnote 5 ) EOC HZP: 8.781 EOC HFP: 2.320 BOC HZP: 24.9 Not a TS Total Deposited Enthalpy, BOC HFP: 144.1 (cal/gm) (See footnote 5 ) 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 T

M their relation to the GDCs in 10 CFR 50, Appendix A for the AREVA St. Lucie CE-16 HTP design are discussed in Reference 2.

The effect of extended burnup on nuclear design parameters has been previously approved in M

T detail in Reference 13. That discussion is valid for the AREVA St. Lucie CE-16 HTP 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 5 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.

lip U11C;U LJULOUI 0 MN It AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 3-4 first transition cycle contains fresh AREVA St. Lucie CE-16 HTPTM 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 Cycle Number of Maximum HFP ARO FrT Maximum HFP ARO FQ Transition Feed Cycle Energy AREVA AREVA Westinghouse AREVA Westinghouse (EFPD) Assemblies Fuel Fe Fuel Fuel Fuel 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 M T

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.

%-,f U1 I UUOl JIIU L-JJLaUI I I1=1I IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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.

.J*UIEILJI Ulqu LJUlfI 0MIIL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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 M T Fuel at St. Lucie Unit 2.

4.2 Input Parametersand Assumptions XCOBRA-IIIC is the core T-H sub-channel analysis code that was used for the AREVA HTP M

T 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, OF 551 RCS minimum flow rate (including bypass), gpm 370,000 Core bypass flow, % 4.2 Core area, ft 2 54.39 Core inlet mass velocity (excluding bypass, based on TS minimum flow rate, 106 Ibm/hr-ft 2 2.45 Pressure drop across core, psi (full-core AREVA CE-16 HTP M T

) [

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

%.A#I ILI VJIIM Ljut.Asl I EIQ;I IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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)

RCS Flow +/-12,500 gpm RCS Pressure +/-45.0 psi Core Inlet Temperature +/-3.0 OF 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:

%AJI ILl I1iI;U L.-utouII AIIL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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-16 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 CHF 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 the AREVA CE-16 HTP TM design have been verified to fall within the range of applicability for these correlations.

I

k-suI OI UIMIIu LULMOUI 0 B51 it AREVA Inc.

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

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) SAFDL 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 U02 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 0 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 AREVA HTPTM fuel assemblies is [

The total pressure drop associated with the full core of AREVA HTP M T

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.

%ýeulILI V11C;U LJULOUI I MN 11 AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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

%..Oru ILI UJl~lc~ L.JULOUI I IMI OL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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 HTPTM fuel assemblies. The power level was selected to achieve MDNBR

MAJ LJIL U1cU LURUsLOI I MNI IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 4-8 close to the HTP CHF correlation limit. A mixed core penalty was applied to all core configurations, including the full core of AREVA HTPTM 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 AQOs.

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 0 Page 4-9 4.5.4 Guide Tube Heating 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 PDIL 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 SAFDLs. 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 0 Page 4-10 The TS LCOs provide requirements for parameters associated with the DNB LCO and LPD

[CO. 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 LCO, 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 TS/COLR settings.

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Figure 4-2: LPD - High Trip Setpoint

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 4-11 1401 120]

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

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

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 4-13 1.75 1.50 U-0 0.

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

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 4-14 110 I I I r I F-- -- Lr -- - --- - 4 4 - T*- 4--

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Figure 4-7: DNB LCO CEAD Results

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 4-15

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 4-16 1(301i 0 I------ r 4.

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Figure 4-10: LPID 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 INB 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:

I.Ol1MVIICU LJu~duIOIc;IIL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 4-18 94,0

.40A 0.60 0.70 0180 0.w 110

iwCONO PAIED *CPAt POWER 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 0 Page 4-19 1.8 1.6 1.4 1.2 w1 S0.8

0.6 0

0.4 0

0.2 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 Percent of Active Core Height from Bottom Figure 4-12: Axial Power Distribution for Thermal Margin Limit Lines

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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 AREVA 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.

%-OJI ILI UNIIZLU LuutouI I IMl It AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 5-2 RODEX2-2A For non-LOCA application-s, 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 (BOC) 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 Methodologies 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

%.#lIMUOIICL Ljukoui I lulIt AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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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%-.jJIli OLU IMU LJU 1LIUI I MINI L AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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:

flU "4....AJIt K;IU L.;ULOLUI I B5I It AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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

WAiVOIC:IIU LJLJ~fuI I IqzI KL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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 AREVA 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 [J ] 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 [ 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 [ 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 [ ] 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 II

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 5-7 Table 5-1: Non-LOCA Limiting Results rContinued)

UFSAR Analytical Section Event Description Criterion Limit Limiting Result 15.4.1 Uncontrolled CEA MDNBR 1.164 1.994 Withdrawal from a Peak CLT, 'F [ ] 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, min. 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, 'F (%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 AREVA fuel.

5.2 Loss-of-CoolantAccident 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.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 5-8 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 inventory loss from the primary system is such that the 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 or to recover core liquid level on HPSI flow. Core recovery for the limiting break begins when the 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 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 2.7-inch cold leg break. This case yielded a limiting PCT of 1926 IF.

5.2.2 Large 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.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 5-9 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 U02 rod in a case with no offsite power availability. This case yielded a limiting PCT of 1732 OF.

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 6-1 6.0 Summary and Conclusion This report shows acceptability for the application of the AREVA CE-16 HTP M T

fuel design at St.

Lucie Unit 2. The results displayed within the report show compliance of the AREVA CE-16 HTP M T

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 M T

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 M T

fuel consisting of a 16x16 assembly configuration with M5 fuel rods, Zircaloy-4 MONOBLOC TM corner guide tubes, an Alloy 718 High Mechanical Performance (HMPTM) spacer at the lowermost axial elevation, Zircaloy-4 HTP M T

spacers in all other axial elevations, a FUELGUARD TM lower tie plate (LTP), and the AREVA reconstitutable upper tie plate (UTP).

M T

The AREVA CE-16 HTP 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 M T

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 M T

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 FUELGUARDTM 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-16 HTP TM fuel at St. Lucie Unit 2.

%-#UIILI LJIIU L-/~tAAAI I It-I OL AREVA Inc.

St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 Page 7-1 7.0 References

1. BAW-10240(P)(A), Revision 0, "Incorporation of M5 Properties in Framatome ANP Approved Methods."

2. EMF-92-116(P)(A), Revision 0, "Generic Mechanical Design Criteria for PWR Fuel Designs."

3. BAW-10227(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, "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-92-116(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-10133(P)(A), Revision 1, and Addenda 1 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 1 and 2, "Reactor Analysis System for PWRs, Volume 1 Methodology Description, Volume 2 Benchmarking Results."

13. ANF-88-133(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).

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St. Lucie Unit 2 Fuel Transition License Amendment Request ANP-3352NP Technical Report Revision 0 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-92-153(P)(A), Revision 1, "HTP: Departure From Nucleate Boiling Correlation for High Thermal Performance Fuel."

19. EMF-1961(P)(A) Revision 0, "Statistical Setpoint/Transient 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-10231(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 0, "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.

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AREVA ANP-3347NP St. Lucie Unit 2 Fuel Transition Revision 0 Chapter 15 Non-LOCA Summary Report December 2014 AREVA Inc.

%,FUIILI VIBIIU LJULJPUU! I It:I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 1 Nature of Changes Section(s)

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

ILEI UIt~iI;U JU%.U LI It:; ItL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 2 Contents Pagqe

1.0 INTRODUCTION

........................................................................................................... 17 2.0 CHAPTER 15 GENERAL DESCRIPTION ................................................................. 18 2.1 Initial Conditions ............................................................................................. 18 2.2 RPS and ESF Functions ................................................................................ 25 2.3 Fuel Mechanical Design ................................................................................ 27 2.4 Peaking Factors ............................................................................................. 28 2.5 Reactivity Coefficients ................................................................................... 29 2.6 CEA Insertion Characteristics ........................................................................ 29 2.7 Analysis Methodologies ................................................................................ 32 2.8 Computer Codes ........................................................................................... 34 3.0 CHAPTER 15 DISPOSITION OF EVENTS .............................................................. 37 4.0 CHAPTER 15 ACCIDENT AND TRANSIENT ANALYSES ...................................... 40 4.1 Feedwater System Malfunctions That Result in a Decrease in Feedwater Temperature (UFSAR 15.1.1) ..................................................... 40 4.1.1 Accident Description .......................................................................... 40 4.1.2 Input Parameters and Assumptions .................................................... 40 4.1.3 Acceptance Criteria ........................................................................... 41 4.1.4 Method of Analysis ............................................................................ 42 4.1.5 Results ............................................................................................... 42 4.2 Feedwater System Malfunctions That Result in an Increase in Feedwater Flow (UFSAR 15.1.2) ................................................................... 52 4.2.1 Accident Description .......................................................................... 52 4.2.2 Input Parameters and Assumptions .................................................... 52 4.2.3 Acceptance Criteria ........................................................................... 54 4.2.4 Method of Analysis ............................................................................ 54 4.2.5 Results ............................................................................................... 54 4.3 Excessive Increase in Secondary Steam Flow (UFSAR 15.1.3) .................... 72 4.3.1 Accident Description .......................................................................... 72 4.3.2 Input Parameters and Assumptions .................................................... 72 4.3.3 Acceptance Criteria ........................................................................... 73 4.3.4 Method of Analysis ............................................................................ 73 4.3.5 Results ............................................................................................... 74 4.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve (UFSAR 15.1.4) ............................................................................................. 94 4.5 Pre-Trip Steam System Piping Failure (UFSAR 15.1.5) ................................ 95 4.5.1 Accident Description .......................................................................... 95

AJI LI UL1J;UI -jU.l*UIl I IM IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 3 4.5.2 Input Parameters and Assumptions .................................................... 96 4.5.3 Acceptance Criteria ........................................................................... 97 4.5.4 Method of Analysis ............................................................................ 97 4.5.5 Results ............................................................................................... 98 4.6 Post-Trip Steam System Piping Failure (UFSAR 15.1.6) .................................. 109 4.6.1 Accident Description ............................................................................. 109 4.6.2 Input Parameters and Assumptions ...................................................... 109 4.6.3 Acceptance Criteria .............................................................................. 111 4.6.4 Method of Analysis ............................................................................... 111 4 .6 .5 R e s u lts ................................................................................................. 112 4.7 Loss of Condenser Vacuum (UFSAR 15.2.3) ................................................... 137 4.7.1 Accident Description ............................................................................. 137 4.7.2 Input Parameters and Assumptions ...................................................... 138 4.7.3 Acceptance Criteria .............................................................................. 139 4.7.4 Method of Analysis ............................................................................... 140 4 .7 .5 R e su lts ................................................................................................. 14 0 4.8 Inadvertent Closure of Main Steam Isolation Valves (BWR)

(UFSAR 15.2.4) ............................................................................................... 152 4.9 Steam Pressure Regulator Failure (UFSAR 15.2.5) ......................................... 152 4.10 Loss of Non-Emergency AC Power to the Station Auxiliaries (U F SA R 15 .2 .6 ) ............................................................................................... 15 2 4.11 Loss of Normal Feedwater Flow (UFSAR 15.2.7) ............................................. 152 4.12 Feedwater System Pipe Break (UFSAR 15.2.8) ............................................... 153 4.13 Transients Resulting from the Malfunction of One Steam Generator (UFSAR 15.2.9) .............................................................................. 154 4.13.1 Accident Description ............................................................................. 154 4.13.2 Input Parameters and Assumptions ...................................................... 155 4.13.3 Acceptance Criteria .............................................................................. 156 4.13.4 Method of Analysis .............................................................................. 156 4 .1 3 .5 R e s u lts ................................................................................................. 15 6 4.14 Partial Loss of Forced Reactor Coolant Flow (UFSAR 15.3.1) ......................... 166 4.15 Complete Loss of Forced Reactor Coolant Flow (UFSAR 15.3.2) .................... 166 4.15.1 Accident Description ............................................................................. 166 4.15.2 Input Parameters and Assumptions ...................................................... 166 4.15.3 Acceptance Criteria .............................................................................. 167 4.15.4 Method of Analysis ............................................................................... 167 4 .1 5 .5 R e s u lts ................................................................................................. 16 8 4.16 Reactor Coolant Pump Shaft Seizure (UFSAR 15.3.3) .................................... 176 4.16.1 Accident Description ............................................................................. 176 4.16.2 Input Parameters and Assumptions ...................................................... 176 4.16.3 Acceptance Criteria .............................................................................. 177

IL1i1 "IU IJI L.JUtUI I K:II L AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 4 4.16.4 Method of Analysis ............................................................................... 177 4 .1 6 .5 R e s u lts ................................................................................................. 17 8 4.17 Reactor Coolant Pump Shaft Break (UFSAR 15.3.4) ....................................... 187 4.18 Uncontrolled CEA Bank Withdrawal from a Subcritical or Low Power Startup Condition (UFSAR 15.4.1) ........................................................ 187 4.18.1 Accident Description ............................................................................. 187 4.18.2 Input Parameters and Assumptions ...................................................... 187 4.18.3 Acceptance Criteria .............................................................................. 188 4.18.4 Method of Analysis ............................................................................... 188 4 .18 .5 R e s u lts ................................................................................................. 18 9 4.19 Uncontrolled CEA Bank W ithdrawal at Power (UFSAR 15.4.2) ........................ 197 4.19.1 Accident Description ............................................................................. 197 4.19.2 Input Parameters and Assumptions ...................................................... 197 4.19.3 Acceptance Criteria .............................................................................. 198 4.19.4 Method of Analysis ............................................................................... 199 4 .19 .5 R e s u lts ................................................................................................. 200 4.20 CEA Misoperation (UFSAR 15.4.3) .................................................................. 209 4.20.1 Accident Description ............................................................................. 209 4.20.2 Input Parameters and Assumptions ...................................................... 209 4.20.3 Acceptance Criteria .............................................................................. 210 4.20.4 Method of Analysis ............................................................................... 210 4 .2 0 .5 R e s u lts ................................................................................................. 2 11 4.21 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (UFSAR 15.4.4) .......................................................................... 220 4.22 A Malfunction or Failure of the Flow Controller in a BWR Recirculation Loop That Results in an Increased Reactor Coolant Flow Rate (UFSAR 15.4.5) .............................................................................. 220 4.23 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant (UFSAR 15.4.6) .................................... 220 4.23.1 Accident Description ............................................................................. 220 4.23.2 Input Parameters and Assumptions ...................................................... 221 4.23.3 Acceptance Criteria ............................................................................. 221 4.23.4 Method of Analysis .............................................................................. 222 4 .2 3 .5 R e s u lts ................................................................................................. 22 2 4.24 Inadvertent Loading of a Fuel Assembly (UFSAR 15.4.7) ................................ 228 4.25 Spectrum of CEA Ejection Accidents (UFSAR 15.4.8) ..................................... 228 4.25.1 Accident Description ............................................................................. 228 4.25.2 Input Parameters and Assumptions ...................................................... 228 4.25.3 Acceptance Criteria .............................................................................. 229 4.25.4 Method of Analysis ......................................... 230 4 .2 5 .5 R e s u lts ................................................................................................. 23 0

%.41ul ILI UIRZIU Ldut-oul I MN it AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 5 4.26 Inadvertent Operation of the ECCS During Power Operation (UF S A R 15 .5 .1) ............................................................................................... 2 52 4.27 CVCS Malfunction that Increases Reactor Coolant Inventory (U F S A R 15 .5 .2 ) ............................................................................................... 2 52 4.28 Inadvertent Opening of Pressurizer Safety or Relief Valve (UFSAR 15 .6 .1 ) ............................................................................................................. 2 52 4 .28.1 A ccident D escription ............................................................................. 252 4.28.2 Input Parameters and Assumptions ...................................................... 253 4.28.3 Acceptance Criteria .............................................................................. 254 4 .28.4 Method of A nalysis ............................................................................... 254 4 .28 .5 R e s u lts ................................................................................................. 2 54 4.29 Break in Instrument Line or Other Lines from the Reactor Coolant Pressure Boundary that Penetrate the Containment (UFSAR 1 5 .6 .2 ) ............................................................................................................. 2 64 4.30 Steam Generator Tube Rupture (UFSAR 15.6.3) ............................................. 264 4.31 Spectrum of Boiling Water Reactor (BWR) Steam System Piping Failures Outside of the Containment (UFSAR 15.6.4) ...................................... 265 4.32 Loss of Coolant Accidents (UFSAR 15.6.5) ...................................................... 265 4.33 Radioactive Releases from a Subsystem or Component (UFSAR 1 5 .7 ) ................................................................................................................ 2 65 4.34 Primary System Pressure Deviation Events (UFSAR 15.8) .............................. 265 4.35 Anticipated Transients without Scram (UFSAR 15.9) ....................................... 265 4.36 Station Blackout (UFSAR 15.10) ...................................................................... 265 5.0

SUMMARY

OF RESULTS ........................................................................................... 267 6 .0 R E F E R E NC E S ............................................................................................................ 2 70

.. AIto~ VoIlcJ LJULLJUI I IuI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary ReDort P*n# R 6 Pnne List of Tables Table 2.1 Key Component Setpoints and Capacities ............................................................ 20 Table 2.2 Plant Operational Modes ...................................................................................... 21 Table 2.3 Summary of Initial Non-LOCA Conditions and Computer Codes Used ................... 22 Table 2.4 RPS Trip Setpoints and Response Times ............................................................. 26 Table 2.5 ESF Actuation Setpoints and Response Times .................................................... 27 Table 2.6 Fuel Design Param eters ...................................................................................... 28 Table 2.7 Core Power Distribution Parameters ..................................................................... 28 Table 2.8 Reactivity Parameters........................................ ..... 29 Table 3.1 Event Disposition Summary of Results ................................................................. 38 Table 4.1 Decrease in Feedwater Temperature: Sequence of Events ................................ 43 Table 4.2 Decrease in Feedwater Temperature: Results ................................................... 43 Table 4.3 Increase in Feedwater Flow: Sequence of Events ............................................... 56 Table 4.4 Increase in Feedwater Flow: Results ................................................................... 56 Table 4.5 Increase in Steam Flow: Sequence of Events ...................................................... 75 Table 4.6 Increase in Steam Flow: Results .......................................................................... 76 Table 4.7 Pre-Scram Main Steam Line Break: Sequence of Events ...................................... 100 Table 4.8 Pre-Scram Main Steam Line Break: Results .......................................................... 100 Table 4.9 Pre-Scram Main Steam Line Break: Sequence of Events For Re-run with LO O P at B reak Initiation ........................................................................................ 10 1 Table 4.10 Post-Scram Main Steam Line Break: Sequence of Events (HZP) ........................ 113 Table 4.11 Post-Scram Main Steam Line Break: Sequence of Events (HFP) ......................... 114 Table 4.12 Post-Scram Main Steam Line Break: Results ...................................................... 115 Table 4.13 Loss of Condenser Vacuum: Sequence of Events (MDNBR Case) ..................... 142 Table 4.14 Loss of Condenser Vacuum: Results ................................................................... 142 Table 4,15 Loss of Load to One Steam Generator: Sequence of Events (0% SGTP) ............ 158 Table 4.16 Loss of Load to One Steam Generator: Limiting Results ..................................... 158 Table 4,17 Loss of Forced Coolant Flow: Sequence of Events .............................................. 169 Table 4,18 Loss of Forced Coolant Flow : Result ................................................................... 169 Table 4.19 Reactor Coolant Pump Rotor Seizure: Sequence of Events ................................ 179 Table 4.20 Reactor Coolant Pump Rotor Seizure: Result ...................................................... 179 Table 4,21 Uncontrolled CEA Withdrawal from Subcritical: Sequence of Events .................... 190 Table 4,22 Uncontrolled CEA Withdrawal from Subcritical: Results ...................................... 190

~AJIILHJtI~U LJU~UIH~IIL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 7 Table 4.23 Uncontrolled CEA Withdrawal at Power: Sequence of Events ............................. 201 Table 4.24 Uncontrolled CEA Withdrawal at Power: Results ................................................. 201 Table 4.25 CEA Drop: Sequence of Events (25 pcm) ............................................................ 212 T able 4 .26 C EA D rop: R esults .............................................................................................. 2 12 Table 4.27 Boron Dilution: Inputs and Boron Requirements (3 Charging Pumps) ................... 224 Table 4.28 Boron Dilution: Inputs and Boron Requirements (2 Charging Pumps) ................... 225 Table 4.29 Boron Dilution: Inputs and Boron Requirements (1 Charging Pump) .................... 226 Table 4.30 Boron D ilution: R esults ........................................................................................ 227 Table 4.31 CEA Ejection: Assumptions ................................................................................. 233 Table 4.32 CEA Ejection: Sequence of Events ...................................................................... 234 Table 4.33 CEA Ejection: Full Power and 65% RTP Results ................................................. 237 Table 4.34 CEA Ejection: 20% RTP and HZP Results .......................................................... 237 Table 4.35 Inadvertent Opening of Pressurizer PORVs: Sequence of Events ....................... 255 Table 4.36 Inadvertent Opening of Pressurizer PORVs: Result ............................................ 255 Table 5.1 Analytical Limits and Limiting Results ...................................................................... 268

V.OLi ILI iJIIUU LjutýLII I it-,itI AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paae 8 Paae 8 St. Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summarv Report List of Figures Figure 2.1 Power Measurement Uncertainty vs. Power ...................................................... 24 Figure 2.2 CEA Insertion Position vs. Tim e ......................................................................... 31 Figure 4.1 Decrease in Feedwater Temperature: Reactor Power ....................................... 44 Figure 4.2 Decrease in Feedwater Temperature: Total Core Heat Flux Power ................... 45 Figure 4.3 Decrease in Feedwater Temperature: Pressurizer Pressure .............................. 46 Figure 4.4 Decrease in Feedwater Temperature: RCS Loop Temperatures ....................... 47 Figure 4.5 Decrease in Feedwater Temperature: RCS Total Loop Flow Rate .................... 48 Figure 4.6 Decrease in Feedwater Temperature: Steam Generator Pressures ................... 49 Figure 4.7 Decrease in Feedwater Temperature: Steam and Feedwater Flow Rates .......... 50 Figure 4.8 Decrease in Feedwater Temperature: Reactivity Feedback ............................... 51 Figure 4.9 Increase in Feedwater Flow: Reactor Power (HFP) ........................................... 57 Figure 4.10 Increase in Feedwater Flow: Total Core Heat Flux Power (HFP) ..................... 58 Figure 4.11 Increase in Feedwater Flow: Pressurizer Pressure (HFP) ................................ 59 Figure 4.12 Increase in Feedwater Flow: RCS Loop Temperatures (HFP) .......................... 60 Figure 4.13 Increase in Feedwater Flow: Steam Generator Pressures (HFP) ..................... 61 Figure 4.14 Increase in Feedwater Flow: Steam and Feedwater Flow Rates (HFP) ........... 62 Figure 4.15 Increase in Feedwater Flow: Reactivity Feedback (HFP) ................................ 63 Figure 4.16 Increase in Feedwater Flow: Reactor Power (HZP) ......................................... 64 Figure 4.17 Increase in Feedwater Flow: Total Core Heat Flux Power (HZP) .................... 65 Figure 4.18 Increase in Feedwater Flow: Pressurizer Pressure (HZP) ................................ 66 Figure 4.19 Increase in Feedwater Flow: RCS Loop Temperatures (HZP) .......................... 67 Figure 4.20 Increase in Feedwater Flow: Steam Generator Pressures (HZP) ..................... 68 Figure 4.21 Increase in Feedwater Flow: Steam and Feedwater Flow Rates (HZP) ........... 69 Figure 4.22 Increase in Feedwater Flow: Reactivity Feedback (HZP) ................................ 70 Figure 4.23 Increase in Feedwater Flow: Peak Fuel Centerline Temperature (HZP) ........... 71 Figure 4.24 Increase in Steam Flow: Reactor Power (HZP) ............................................... 77 Figure 4.25 Increase in Steam Flow: Total Core Heat Flux Power (HZP) ........................... 78 Figure 4.26 Increase in Steam Flow: Pressurizer Pressure (HZP) ....................................... 79 Figure 4.27 Increase in Steam Flow: RCS Loop Temperatures (HZP) ................................ 80 Figure 4.28 Increase in Steam Flow: RCS Total Loop Flow Rate (HZP) ............................. 81 Figure 4.29 Increase in Steam Flow: Steam Generator Pressures (HZP) ............................ 82

%.flILI LJIlC;U LJUtLU I IUI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Pag~e 9 Figure 4.30 Increase in Steam Flow: Steam Line and Feedwater Flow Rates (HZP) ........... 83 Figure 4.31 Increase in Steam Flow: Reactivity Feedback (HZP) ....................................... 84 Figure 4.32 Increase in Steam Flow: Peak Fuel Centerline Temperature (HZP) ................ 85 Figure 4.33 Increase in Steam Flow: Reactor Power (HFP) ............................................... 86 Figure 4.34 Increase in Steam Flow: Total Core Heat Flux Power (HFP) ............................ 87 Figure 4.35 Increase in Steam Flow: Pressurizer Pressure (HFP) ....................................... 88 Figure 4.36 Increase in Steam Flow: RCS Loop Temperatures (HFP) ................................ 89 Figure 4.37 Increase in Steam Flow: RCS Total Loop Flow Rate (HFP) ............................. 90 Figure 4.38 Increase in Steam Flow: Steam Generator Pressures (HFP) ............................ 91 Figure 4.39 Increase in Steam Flow: Steam Line and Feedwater Flow Rates (HFP) ........... 92 Figure 4.40 Increase in Steam Flow: Reactivity Feedback (HFP) ....................................... 93 Figure 4.41 Pre-Scram Main Steam Line Break: Break, SG Steam and Feedwater Flow R a te s .................................................................................................................. 10 2 Figure 4.42 Pre-Scram Main Steam Line Break: Steam Generator Secondary P re s s u re s ........................................................................................................... 10 3 Figure 4.43 Pre-Scram Main Steam Line Break: Pressurizer Pressure .................................. 104 Figure 4.44 Pre-Scram Main Steam Line Break: RCS Loop Temperatures ........................... 105 Figure 4.45 Pre-Scram Main Steam Line Break: Core Inlet Flow Rates ................................. 106 Figure 4.46 Pre-Scram Main Steam Line Break: Reactivity Feedback ................................... 107 Figure 4.47 Pre-Scram Main Steam Line Break: Reactor Power ............................................ 108 Figure 4.48 Post-Scram Main Steam Line Break: Break Flow Rates (HZP Offsite Power A v a ila b le ) ............................................................................................................ 1 16 Figure 4.49 Post-Scram Main Steam Line Break: Steam Generator Pressures (HZP Off site P ow er A vailable) ...................................................................................... 117 Figure 4.50 Post-Scram Main Steam Line Break: MFW Flow Rates (HZP Offsite Power A va ila b le ) ............................................................................................................ 118 Figure 4.51 Post-Scram Main Steam Line Break: AFW Flow Rates (HZP Offsite Power A va ila b le ) ............................................................................................................ 119 Figure 4.52 Post-Scram Main Steam Line Break: Steam Generator Mass Inventories (HZP Offsite Power Available) ............................................................................ 120 Figure 4.53 Post-Scram Main Steam Line Break: Core Inlet Fluid Temperatures (HZP Offsite P ow er A vailable) ...................................................................................... 12 1 Figure 4.54 Post-Scram Main Steam Line Break: Pressurizer Pressure (HZP Offsite P o we r Ava ila b le ) ................................................................................................. 12 2 Figure 4.55 Post-Scram Main Steam Line Break: Pressurizer Liquid Level (HZP Offsite P ow e r Ava ila b le ) ................................................................................................. 12 3

kjLJI DLI L0lIU LJUk.,AI I II L AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report P:3 =in St. Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summary Report Figure 4.56 Post-Scram Main Steam Line Break: Total HPSI Flow Rate (HZP Offsite P ow e r Availab le ) ................................................................................................. 124 Figure 4.57 Post-Scram Main Steam Line Break: Reactivity Feedback (HZP Offsite P o w e r A va ila b le ) ................................................................................................. 12 5 Figure 4.58 Post-Scram Main Steam Line Break: Core Power (HZP Offsite Power A v a ila b le ) ............................................................................................................ 12 6 Figure 4.59 Post-Scram Main Steam Line Break: Break Flow Rates (HFP Offsite Power A v a ila b le ) ............................................................................................................ 12 7 Figure 4.60 Post-Scram Main Steam Line Break: Steam Generator Pressures (HFP Offsite P ow er A vailable) ...................................................................................... 128 Figure 4.61 Post-Scram Main Steam Line Break: MFW Flow Rates (HFP Offsite Power A v a ila b le ) ............................................................................................................ 12 9 Figure 4.62 Post-Scram Main Steam Line Break: AFW Flow Rates (HFP Offsite Power A v a ila b le ) ............................................................................................................ 13 0 Figure 4.63 Post-Scram Main Steam Line Break: Steam Generator Mass Inventories (HFP Offsite Power Available) ............................................................................ 131 Figure 4.64 Post-Scram Main Steam Line Break: Core Inlet Fluid Temperatures (HFP Offsite P ower A vailable) ...................................................................................... 132 Figure 4.65 Post-Scram Main Steam Line Break: Pressurizer Pressure (HFP Offsite P o we r A va ila b le ) ................................................................................................. 13 3 Figure 4.66 Post-Scram Main Steam Line Break: Pressurizer Liquid Level (HFP Offsite P ow e r Ava ila b le ) ................................................................................................. 134 Figure 4.67 Post-Scram Main Steam Line Break: Reactivity Feedback (HFP Offsite P o we r Ava ila b le ) ................................................................................................. 13 5 Figure 4.68 Post-Scram Main Steam Line Break: Core Power (HFP Offsite Power A v a ila b le ) ............................................................................................................ 13 6 Figure 4.69 Loss of Condenser Vacuum: Reactor Power (MDNBR Case) ............................. 143 Figure 4.70 Loss of Condenser Vacuum: Total Core Heat Flux Power (MDNBR Case) ........ 144 Figure 4.71 Loss of Condenser Vacuum: Pressurizer Pressure (MDNBR Case) ................... 145 Figure 4.72 Loss of Condenser Vacuum: Pressurizer Liquid Level (MDNBR Case) .............. 146 Figure 4.73 Loss of Condenser Vacuum: Pressurizer PORVs Flow Rate (MDNBR C a se ) .................................... .............................................................................. 14 7 Figure 4.74 Loss of Condenser Vacuum: RCS Loop Temperatures (MDNBR Case) ............. 148 Figure 4.75 Loss of Condenser Vacuum: RCS Total Loop Flow Rate (MDNBR Case) .......... 149 Figure 4.76 Loss of Condenser Vacuum: Steam Generator Pressures (MDNBR Case) ........ 150 Figure 4.77 Loss of Condenser Vacuum: Reactivity Feedback (MDNBR Case) .................... 151 Figure 4.78 Loss of Load to One Steam Generator: Reactor Power (0% SGTP) ................... 159

%aUl ILI UIIUU LJUtaU1 I It-,[ IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Reoort Paae 11 Figure 4.79 Loss of Load to One Steam Generator: Total Core Heat Flux Power (0%

S G T P) ................................................................................................................ 16 0 Figure 4.80 Loss of Load to One Steam Generator: Pressurizer Pressure (0% SGTP) ......... 161 Figure 4.81 Loss of Load to One Steam Generator: Pressurizer Liquid Level (0%

S G T P ) ................................................................................................................ 16 2 Figure 4.82 Loss of Load to One Steam Generator: RCS Loop Temperatures (0%

S G T P ) ................................................................................................................ 16 3 Figure 4.83 Loss of Load to One Steam Generator: RCS Total Loop Flow Rate (0%

S G T P ) ................................................................................................................ 16 4 Figure 4.84 Loss of Load to One Steam Generator: Steam Generator Pressures (0%

S G T P ) ................................................................................................................ 16 5 Figure 4.85 Loss of Forced Coolant Flow: Reactor Power ..................................................... 170 Figure 4.86 Loss of Forced Coolant Flow: Total Core Heat Flux Power ................................. 171 Figure 4.87 Loss of Forced Coolant Flow: Pressurizer Pressure ........................................... 172 Figure 4.88 Loss of Forced Coolant Flow: RCS Loop Temperatures ..................................... 173 Figure 4.89 Loss of Forced Coolant Flow: RCS Total Loop Flow Rate .................................. 174 Figure 4.90 Loss of Forced Coolant Flow: Reactivity Feedback ............................................ 175 Figure 4.91 Reactor Coolant Pump Rotor Seizure: Reactor Power ....................................... 180 Figure 4.92 Reactor Coolant Pump Rotor Seizure: Total Core Heat Flux Power ................... 181 Figure 4.93 Reactor Coolant Pump Rotor Seizure: Pressurizer Pressure .............................. 182 Figure 4.94 Reactor Coolant Pump Rotor Seizure: RCS Loop Temperatures ....................... 183 Figure 4.95 Reactor Coolant Pump Rotor Seizure: RCS Total Loop Flow Rate ..................... 184 Figure 4.96 Reactor Coolant Pump Rotor Seizure: Reactivity Feedback ............................... 185 Figure 4.97 Reactor Coolant Pump Rotor Seizure: RCS Loop Flow Rates ............................ 186 Figure 4.98 Uncontrolled CEA Withdrawal from Subcritical: Reactivity Feedback ................. 191 Figure 4.99 Uncontrolled CEA Withdrawal from Subcritical: Power and Heat Flux ................ 192 Figure 4.100 Uncontrolled CEA Withdrawal from Subcritical: Hot Spot Temperatures .......... 193 Figure 4.101 Uncontrolled CEA Withdrawal from Subcritical: RCS Temperatures ................. 194 Figure 4.102, Uncontrolled CEA Withdrawal from Subcritical: Pressurizer Pressure .............. 195 Figure 4.103 Uncontrolled CEA Withdrawal from Subcritical: Cold Leg Flow Rates .............. 196 Figure 4.104 Uncontrolled CEA Withdrawal at Power: Reactor Power (BOC Full Power 0 .0 8 p c m /se c) .................................................................................................... 202 Figure 4.105 Uncontrolled CEA Withdrawal at Power: Total Core Heat Flux Power (BOC Full Power 0.08 pcm/sec) ........................................................................ 203 Figure 4.106 Uncontrolled CEA Withdrawal at Power: Pressurizer Pressure (BOC Full P ow er 0.08 pcm /sec) ......................................................................................... 204

%,.AJIILI VOIIU L/;Utf.UI I ICI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Reoort Paae 12 Figure 4.107 Uncontrolled CEA Withdrawal at Power: Pressurizer Liquid Level (BOC Full P ow er 0.08 pcm /sec) .................................................................................. 205 Figure 4.108 Uncontrolled CEA Withdrawal at Power: RCS Loop Temperatures (BOC Full P ow er 0.08 pcm /sec) .................................................................................. 206 Figure 4.109 Uncontrolled CEA Withdrawal at Power: Total RCS Loop Flow Rate (BOC Full P ow er 0.08 pcm /sec) .................................................................................. 207 Figure 4.110 Uncontrolled CEA Withdrawal at Power: Reactivity Feedback (BOC Full P ow er 0 .08 pcm /sec) ......................................................................................... 208 Figure 4.111 CEA Drop: Reactor Power (25 pcm) ................................................................. 213 Figure 4.112 CEA Drop: Reactivity Feedback (25 pcm) ......................................................... 214 Figure 4.113 CEA Drop: RCS Loop Temperatures (25 pcm) ................................................. 215 Figure 4.114 CEA Drop: RCS Total Loop Flow Rate (25 pcm) .............................................. 216 Figure 4.115 CEA Drop: Pressurizer Pressure (25 pcm) ....................................................... 217 Figure 4.116 CEA Drop: Pressurizer Liquid Level (25 pcm) .................................................. 218 Figure 4.117 CEA Drop: Steam Generator Pressures (25 pcm) .......................................... 219 Figure 4.118 CEA Ejection: Reactor Power (BOC HFP) ....... 238 Figure 4.119 CEA Ejection: Total Core Heat Flux Power (BOC HFP) ................................... 239 Figure 4.120 CEA Ejection: RCS Loop Temperatures (BOC HFP) ........................................ 240 Figure 4.121 CEA Ejection: RCS Total Loop Flow Rate (BOC HFP) ..................................... 241 Figure 4.122 CEA Ejection: Reactivity Feedback (BOC HFP) ............................................... 242 Figure 4.123 CEA Ejection: Peak Fuel Centerline Temperatures (BOC HFP) ....................... 243 Figure 4.124 CEA Ejection: Pressurizer Pressure (BOC HFP) .............................................. 244 Figure 4.125 CEA Ejection: Reactor Power (BOC 20% RTP) ................................................ 245 Figure 4.126 CEA Ejection: Total Core Heat Flux Power (BOC 20% RTP) ........................... 246 Figure 4.127 CEA Ejection: RCS Loop Temperatures (BOC 20% RTP) ................................ 247 Figure 4.128 CEA Ejection: RCS Total Loop Flow Rate (BOC 20% RTP) ............................. 248 Figure 4.129 CEA Ejection: Reactivity Feedback (BOC 20% RTP) ....................................... 249 Figure 4.130 CEA Ejection: Peak Fuel Centerline Temperatures (BOC 20% RTP) ............... 250 Figure 4.131 CEA Ejection: Pressurizer Pressure (BOC 20% RTP) ...................................... 251 Figure 4.132 Inadvertent Opening of Pressurizer PORVs: Reactor Power ............................ 256 Figure 4.133 Inadvertent Opening of Pressurizer PORVs: Total Core Heat Flux Power ........ 257 Figure 4.134 Inadvertent Opening of Pressurizer PORVs: Pressurizer Pressure .................. 258 Figure 4.135 Inadvertent Opening of Pressurizer PORVs: Pressurizer PORVs Flow Ra te .................................................................................................................. 259 Figure 4.136 Inadvertent Opening of Pressurizer PORVs: RCS Loop Temperatures ............ 260

%.,PUI IluLJI IBU LdUkwlI I MI I L AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report P~ru* i3 Pane 13 Figure 4.137 Inadvertent Opening of Pressurizer PORVs: RCS Total Loop Flow Rate ......... 261 Figure 4.138 Inadvertent Opening of Pressurizer PORVs: Reactivity Feedback .................... 262 Figure 4.139 Inadvertent Opening of Pressurizer PORVs: Pressurizer Liquid Level .............. 263

EU 010~ IIZ;U LIIJULJI I 1t;1 IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 14 Nomenclature Acronym Definition AC Alternating Current ADVs Atmospheric Dump Valves AFW Auxiliary Feedwater AOO Anticipated Operational Occurrence AOR Analysis of Record ASGPT Asymmetric Steam Generator Pressure Trip ASI Axial Shape Index ATWS Anticipated Transient without Scram BDAS Boron Dilution Alarm System BOC Beginning-of-Cycle BWR Boiling Water Reactor CE Combustion Engineering CEA(s) Control Element Assembly (Assemblies)

CEDM Control Element Drive Mechanism CHF Critical Heat Flux CLT Centerline Temperature COLR Core Operating Limits Report CVCS Chemical and Volume Control System DEGB Double Ended Guillotine Break DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Boiling Ratio ECCS Emergency Core Cooling System EOC End-of-Cycle ESF Engineered Safety Feature ESFAS Engineered Safety Feature Actuation System FCM Fuel Centerline Melt FFBT Failure of Fast Bus Transfer FPL Florida Power and Light Fr Radial Peaking Factor HFP Hot Full Power HMPTM High Mechanical Performance HPP High Pressurizer Pressure HPSI High Pressure Safety Injection HTPTM High Thermal Performance HZP Hot Zero Power

%.AFIIM USII;U 9IN 8 ULJUIH~

AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 15 Nomenclature (Continued)

Acronym Definition LAR License Amendment Request LCO(s) Limiting Condition(s) for Operation LHR Linear Heat Rate LOCA Loss of Coolant Accident LOOP Loss of Offsite Power MDNBR Minimum Departure from Nucleate Boiling Ratio MFW Main Feedwater MFWIV Main Feedwater Isolation Valve MSIS Main Steam Isolation Signal MSIV(s) Main Steam Isolation Valve(s)

MSLB Main Steam Line Break MSSV(s) Main Steam Safety Valve(s)

MTC Moderator Temperature Coefficient NI Nuclear Instrumentation NR Narrow Range NRC Nuclear Regulatory Commission PA Postulated Accident PCI Pellet / Cladding Interaction PCMI Pellet / Cladding Mechanical Interaction PHLA Pressurizer High Level Alarm PORV(s) Power Operated Relief Valve(s)

PSV(s) Pressurizer Safety Valve(s)

PWR Pressurized-Water Reactor RCP(s) Reactor Coolant Pump(s)

RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RPS Reactor Protection System RTD Resistance Temperature Detector RTP Rated Thermal Power RWT Refueling Water Tank SAFDL(s) Specified Acceptable Fuel Design Limit(s)

SBCS Steam Bypass Control System SBO Station Blackout SDCS Shutdown Cooling System SDM Shutdown Margin SER Safety Evaluation Report SG(s) Steam Generator(s)

SGTP Steam Generator Tube Plugging SGTR Steam Generator Tube Rupture

~#II LJIIZ;U LJULoJI I I;I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 16 Nomenclature (Continued)

Acronym Definition SI Safety Injection SIAS Safety Injection Actuation Signal SPC Siemens Power Corporation TCD Thermal Conductivity Degradation TCV Turbine Control Valve TM/LP Thermal Margin / Low Pressure TS Technical Specification(s)

UFSAR Updated Final Safety Analysis Report VCT Volume Control Tank VHP Variable High Power WPR Wetted Perimeter Ratio

%-.AjI Ll LJIIU LJ'UJI.JI IICI L AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 17

1.0 INTRODUCTION

The purpose of this document is to provide non-Loss of Coolant Accident (non-LOCA) input to the fuel transition License Amendment request (LAR) in support of the transition to AREVA fuel in St. Lucie Unit 2. A summary of the Chapter 15 non-LOCA transient analyses and evaluations is contained herein.

Section 2.0 presents the initial conditions, plant operating conditions, trip setpoints, equipment capability, analysis methods and computer codes. Section 3.0 discusses the Chapter 15 Disposition of Events supporting the fuel transition. Section 4.0 presents a summary of the non-LOCA event analyses performed for the fuel transition and Section 5.0 summarizes the results.

ýýlILI UOIRILlu LJLALUI I IC;I it AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 18 2.0 CHAPTER 15 GENERAL DESCRIPTION 2.1 Initial Conditions Key features of St. Lucie Unit 2 that were considered in transient analyses supporting the transition to AREVA fuel were as follows:

Rated core power level of 3020 MWt.

AREVA's 16x16 fuel design for Combustion Engineering (CE) designed plants.

Nominal hot full power (HFP) cold leg temperature range of 535°F to 551 *F.

a Nominal hot zero power (HZP) cold leg temperature of 532 0 F.

Minimum reactor coolant system (RCS) flow rate of 370000 gpm.

Maximum steam generator tube plugging (SGTP) of 20% +/- 4%.

Nominal pressurizer pressure of 2250 psia.

Radial peaking factor (Fr) of 1.65.

Peak linear heat rate (LHR) of 13.0 kW/ft.

Maximum core bypass flow of 4.2%.

Biases were applied to key parameters consistent with or conservative relative to the approved methodology (Reference 1). 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 analysis. For statistical departure from nucleate boiling (DNB) calculations, uncertainties were statistically treated.

The system related uncertainties bounded by the safety analyses are:

Power measurement uncertainty versus reactor power is shown in Figure 2.1.

a RCS pressure measurement uncertainty of +/-45 psi.

Core inlet temperature measurement uncertainty of +/-3 0 F.

0 RCS flow rate measurement uncertainty of 12500 gpm.

R1J VPPiJI~I L/-jLt.UI I MAI HL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 19 Table 2.1 shows the assumed key component setpoints and capacities. Only safety grade equipment was credited to mitigate an event. Control grade equipment was modeled in such a way that it does not mitigate the effects of an event. For example, the pressurizer power operated relief valves (PORVs) and pressurizer spray system were assumed operable while the pressurizer heaters were assumed inoperable for departure from nucleate boiling ratio (DNBR) transient events where suppressing primary side pressure is conservative.

Table 2.2 lists the plant operational modes, consistent with the Technical Specifications (TS)

(Reference 2), which are supported by the AREVA non-LOCA safety analysis. Table 2.3 provides a summary of the initial conditions assumed for each event.

%--UI ILI l ltjiU LJUýJal.I I IUI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summary ReDort Paae 20 20 Paoe Table 2.1 Key Component Setpoints and Capacities Nominal Setpoint (1)

Item Setpoint Tolerance Total Capacity Pressurizer Safety Valves (3 valves) 2500 psia +/-3% 212182 Ibm/hr/valve at 2500 psia + 3% accumulation Pressurizer PORVs (2 valves) 2370 psia +/-4 psi(2) 401000 Ibm/hr/valve +1% at 2400 psia MSSVs (8 total valves per SG)

" Group 1 (4 valves per SG) 1000 psia +/-3% 744210 Ibm/hr/valve at 1000 psia + 3% accumulation

" Group 2 (4 valves per SG) 1040 psia +2%, -3% 774000 Ibm/hr/valve at 1040 psia + 3% accumulation Steam Bypass Control System 7.7 Mlbm/hr (max)

Charging Flow (3 pumps) 49 gpm/pump (max)

Auxiliary Feedwater 712.5 gpm (max. run-outflow (2 motor-driven pumps and per motor-driven pump) 1 steam-driven pump) 1075 gpm (max. run-out flow from steam-driven pump)

1. Analyses support the specified or more conservative values.

2. An additional PORV tolerance of +/-45 psi was assumed based on the Pressurizer Pressure - High RPS trip uncertainty since both use the same signal

%ýoUIWVW=U UULU1 I IM IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summary ReDort Paae 21 Table 2.2 Plant Operational Modes Reactivity % Rated Thermal Average Coolant Operational Mode Condition, Keff Power"1 ) Temperature

1. Power Operation > 0.99 > 5% > 325 0 F

2. Startup >0.99 < 5% > 325°F

3. Hot Standby < 0.99 0% > 325°F

4. Hot Shutdown < 0.99 0% 325°F > Tavq > 200°F

5. Cold Shutdown < 0.99 0% < 200'F

6. Refueling(2) < 0.95 0% < 140°F

1. Does not include decay heat.

2. Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Pnrip 99 Table 2.3 Summary of Initial Non-LOCA Conditions and Computer Codes Used Initial Nominal Initial Core Nominal Principal Statistical Initial Inlet Pressurizer Computer DNB DNB Initial Nominal RCS Flow Temp. Pressure Event Codes Used Correlation Method Core Power Rate (gpm) (OF) (psia)

Decrease in Feedwater S-RELAP5 HTP No 100% 370000 551 2250 Temperature XCOBRA-IIIC Increase in Feedwater Flow S-RELAP5 HTP No 100% 370000 551 2250 XCOBRA-IIIC Increase in Steam Flow S-RELAP5 HTP No 100% 370000 551 2250 XCOBRA-IIIC HZP 532 Pre-Trip Steam System S-RELAP5 HTP Yes 100% 370000 551 2250 Piping Failure XCOBRA-IIIC Post-Trip Steam System S-RELAP5 Biasi No 100% 370000 551 2250 Piping Failure XCOBRA-IIIC HZP 532 PRISM Loss of Condenser Vacuum S-RELAP5 HTP No 100% 370000 551 2250 XCOBRA-IIIC Transients Resulting from S-RELAP5 HTP No 100% 370000 551 2250 the Malfunction of One XCOBRA-IIIC Steam Generator Loss of Forced Reactor S-RELAP5 HTP Yes 100% 370000 551 2250 Coolant Flow XCOBRA-IIIC I I I II

AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Pani* 23 Pane 23 Table 2.3 Summary of Initial Non-LOCA Conditions and Computer Codes Used (Continued)

Initial Nominal Initial Core Nominal Principal Statistical Initial Inlet Pressurizer Computer DNB DNB Initial Nominal RCS Flow Temp. Pressure Event Codes Used Correlation Method Core Power Rate (gpm) (OF) (psia)

Reactor Coolant Pump S-RELAP5 HTP Yes 100% 370000 551 2250 Rotor Seizure XCOBRA-IIIC Uncontrolled CEA S-RELAP5 HTP No HZP 370000 532 2250 Withdrawal from a XCOBRA-IIIC Subcritical or Low Power Startup Condition Uncontrolled CEA S-RELAP5 HTP Yes 100% 370000 551 2250 Withdrawal at Power XCOBRA-IIIC 65% 544.4 20% 535.8 CEA Drop S-RELAP5 HTP No 100% 370000 551 2250 XCOBRA-IIIC CEA Ejection S-RELAP5 HTP Yes 100% 370000 551 2250 XCOBRA-IIIC 65% 544.4 20% 535.8 HZP 532 Inadvertent Opening of a S-RELAP5 HTP Yes 100% 370000 551 2250 Pressurizer Pressure Relief XCOBRA-IIIC Valve

LE JHelt7 OAI LJU.L4U IPl;l RL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paae 24 Figure 2.1 Power Measurement Uncertainty vs. Power

%-API ILI 'Jll;U LJIJOlU I MAI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 25 2.2 RPS and ESF Functions Table 2.4 and Table 2.5 list the Reactor Protection System (RPS) and Engineered Safety Feature (ESF) functions credited in the transient analyses, respectively. Uncertainties and response times associated with each of these functions are also given in Table 2.4 and Table 2.5. The setpoints and response times modeled in the transient analyses were conservatively applied to provide bounding simulations of the plant response. In addition, the transient analyses accounted for the effect of the pressure operating range on the timing of the pressure-related RPS trips. To the extent that the RPS and ESF system are credited in the accident analyses, the setpoints have been verified to adequately protect the plant for operation with AREVA fuel. There are no changes to the current TS specified RPS and ESF setpoints.

AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summary Report Paae 26 Table 2.4 RPS Trip Setpoints and Response Times Harsh Condition Response Time Trip Allowable Trip Setpoint Uncertainty Uncertainty (sec)

Variable Power Level - High(1 ) < 9.61% above thermal power with Note (2) N/A <0.4(3)

Four Reactor Coolant Pumps a minimum setpoint of 15% RTP (excluding RTD Operating and a maximum of < 107.0% RTP delay time)

Thermal Margin/Low Pressure PVAR = f(TIN, Power, ASI) +/-160 psi N/A < 0.9 (TM/LP)(1) Min. floor = 1900 psia (excluding RTD delay time)

Reactor Coolant Flow - Low(1 ) > 94.9% of four pump design +/-3.5% +/-+7.5% < 0.9 reactor coolant flow(4)

Pressurizer Pressure - High < 2374 psia +/-45 psi +/-90 psi 51.15 Steam Generator Pressure - Low > 621 psia +/-40 psi +/-80 psi 5 1.15 Steam Generator Pressure < 132 psi +/-60 psi +/-115 psi 5 1.15 Difference - High(')

Containment Pressure - High < 3.1 psi N/A +/-0.55 5 1.15 (pressure measurement uncertainty)

+/-1.65 psi (trip uncertainty)

1. Trip may be bypassed below 0.5% RTP; bypass is automatically removed when Wide Range Logarithmic Neutron Flux power is > 0.5%

RTP.

2. Minimum uncertainty is 5% at >75% RTP and a maximum of 12% for power levels < 25% RTP, when combined with power measurement uncertainty. The uncertainty is linear between 25% and 75% power.

3. Additional delay of 0.7 sec. added for events initiated at power levels less than or equal to 1% RTP.

4. Minimum analytical reactor coolant flow with 4 pumps operating is 370000 gpm after accounting for measurement uncertainty.

%-AJI Ito LJIIKL L/utrJI I I1ZI It AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paae 27 Table 2.5 ESF Actuation Setpoints and Response Times Allowable Actuation Normal Harsh Condition Response Time Actuation Setpoint Uncertainty Uncertainty (sec)

> 567 psia Not used +/-80 psi < 7.0(1)(2)

Main Steam Isolation 0 SG Pressure - Low Feedwater Isolation > 567 psia Not used +/-80 psi < 5.65(1)(2)

SG Pressure- Low Auxiliary Feedwater Isolation < 275 psi Not used +/-125 psi < 200

High SG Differential Pressure Safety Injection ->1728 psia Not used +/-90 psi < 30(l)

Pressurizer Pressure _<20(2)

- Low

1. Diesel generator starting time and sequence loading delays included.

2. Diesel generator starting time and sequence loading delays not included (offsite power available).

2.3 Fuel MechanicalDesign The AREVA CE-16 fuel design incorporates High Thermal Performance (HTPTM*) grid spacers, a High Mechanical Performance (HMPTMt) lower spacer grid and a FUELGUARDTM lower tie plate. The fuel rod cladding material is M50() and the cladding material for the guide tubes and instrument tube is Zr-4. The key fuel design parameters are summarized in Table 2.6.

HTPTM is a trademark of AREVA.

t HMPTM is a trademark of AREVA.

FUELGUARDTM is a trademark of AREVA.

§ M5 is a registered trademark of AREVA.

fU~LJI IJII:L LdUL.oUl I IUI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report , riv 9v Pane 28 Table 2.6 Fuel Design Parameters Fuel Assembly Array 16 x 16 Number of Fuel Rods per Assembly 236 Guide Tubes per Assembly 4 Instrument Tubes per Assembly 1 Fuel Rod Pitch, inches 0.506 Fuel Pellet Outside Diameter, inches 0.3255 Clad Inside Diameter, inches 0.332 Clad Outside Diameter, inches 0.382 Heated Fuel Length, inches 136.7 Number of Spacers 10 2.4 Peaking Factors The power distribution limits for the AREVA fuel transition are shown in Table 2.7. For non-LOCA events, an Fr of 1.65 and a peak LHR of 13 kW/ft were modeled. Fr is important for transients that are analyzed to assess DNB concerns. For "fast" events that challenge fuel centerline melt (FCM), e.g., Uncontrolled Control Element Assembly (CEA) Withdrawal from HZP and CEA Ejection, event-specific hot spot power peaking factors were used to calculate fuel centerline temperature (CLT). For events that evolve relatively slowly, the radial peaking factor, a conservative axial peaking factor and an event-specific augmentation factor (if required) were combined to determine the challenge to the LHR corresponding to fuel centerline melt.

Table 2.7 Core Power Distribution Parameters Fr limit (without uncertainties) 1.65 Fr measurement uncertainty 6%

Peak LHR limit, including uncertainties (kW/ft) 13.0 Engineering tolerance 3%

IJI U1U%IIU LIU~wU I IMie IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 29 2.5 Reactivity Coefficients Transient response of the reactor core is dependent on reactivity feedback effects, in particular the moderator and Doppler feedback. Depending on the event-specific characteristics, e.g.,

RCS heatup or cooldown, conservatism dictates the use of either maximum or minimum reactivity coefficient values. Justification for the use of the reactivity coefficient values was treated on an event-specific basis. Table 2.8 presents the key core kinetics parameters and reactivity feedback coefficients supported by the transient analyses. The current TS limits on moderator temperature coefficient (MTC) were supported. The Doppler reactivity coefficients were biased according to the approved Reference 1 methodology with additional conservatism to bound expected cycle-to-cycle changes. The continued applicability of the transient analyses is verified for each fuel reload.

The neutronic parameters were validated as representative of core designs expected to be implemented at St. Lucie Unit 2. Key neutronic parameters were evaluated for the first fuel transition cycle (Cycle N). To ensure that these parameters are representative of the third cycle (Cycle N+2), where the core has essentially reached equilibrium, the parameters were re-generated and examined for large deviations. No significant deviations in parameters were observed between Cycle N and Cycle N+2.

Table 2.8 Reactivity Parameters Parameter BOC EOC Moderator Temperature Coefficient (TS limit), pcm/°F +5(1) -33(2)

Doppler Temperature Coefficient, pcm/°F (bounding -1.0 -2.5 analysis value range)

Delayed Neutron Fraction 0.006427 0.005258 U-238 Capture-to-Fission Ratio 0.70 Fraction of heat generated in the fuel 0.975

1. TS limit of +5 pcm/°F for powers < 70% RTP was conservatively applied to all power levels

2. Value used in the analysis is conservative relative to the TS limit of -32 pcm/°F.

2.6 CEA Insertion Characteristics A time delay of 0.74 second was assumed after the trip breakers open until the CEAs start to insert into the core to account for the time required for the magnetic flux of the CEA holding coils

\..gLI Itl U1IICU LJU%.OUI I MN IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 30 to decay sufficiently to release the CEAs. Including this delay, the time to 90% insertion used in the transient analyses was 3.25 seconds, as shown in Figure 2.2.

For events initiated from HFP conditions, a conservative minimum HFP scram worth was used accounting for the most reactive CEA being fully withdrawn. For events initiated from HZP and part-power conditions, the scram worth was set to the TS minimum shutdown margin (SDM) requirement (i.e., 3600 pcm). The shutdown margin requirements are verified for each reload cycle.

ýljul OLD U01t7u uukpul I IM R AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paae Paae 3131 St. Lucie Unit 2 Fuel Transition Char)ter 15 Non-LOCA Surnmarv Reoort 120 110 100__

-. 90 __

2 80__

70 60 _ __

0 S50__ __ ____

0

0. 40 -__

u 30 20 10 0

0.00 0.25 0.50 0.75 1.00 1.25 1.50 1.75 2.00 2.25 2.50 2.75 3.00 3.25 3.50 3.75 Time (seconds)

Figure 2.2 CEA Insertion Position vs. Time

%_.*UI ILI LJIIOU L/UtL,UI 9 I-oI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 32 2.7 Analysis Methodologies The approved methodology for evaluating non-LOCA transients is described in Reference 1.

For each non-LOCA transient event analysis, the nodalization, chosen parameters, conservative input and sensitivity studies were reviewed for applicability to the AREVA fuel transition in compliance with the Safety Evaluation Report (SER) for Revision 0 of the non-LOCA topical report (Reference 1).

The nodalization used for the calculations supporting the AREVA fuel transition was specific to St. Lucie Unit 2 and was consistent with the Reference 1 methodology.

The parameters and equipment states were chosen to provide a conservative estimate of the challenge to the acceptance criteria. The biasing and assumptions for key input parameters were consistent with or conservative to the approved Reference 1 methodology.

The S-RELAP5 code assessments in Reference 1 validated the ability of the code to predict the response of the primary and secondary systems to non-LOCA transients and accidents. No additional model sensitivity studies were needed for this application.

The method used for the non-LOCA system transient analyses differs from that in the approved Reference 1 topical report as described below.

Another change allowed by the Reference 1 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 thermal conductivity degradation (TCD). The properties from COPERNIC were developed for beginning-of-cycle (BOC) and end-of-cycle

ILI UIM:U I-JULOUI I It:-,I OL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 33 (EOC) conditions in accordance with Reference 1, 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 1.

The approved methodology for performing DNB calculations using the XCOBRA-IIIC code is described in Reference 3. The SER for the Reference 3 topical report states that the use of XCOBRA-IIIC is limited to the "snapshot" mode. Thus, minimum departure from nucleate boiling ratio (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 4 topical report describes the method for performing statistical DNB analyses.

Two conditions were noted in the SER for the Reference 4 methodology:

The methodology is approved only for CE type reactors which use protection systems as described in the Reference 4 topical report.

The methodology includes a statistical treatment of specific variables in the analysis; therefore, if additional variables are treated statistically, Siemens Power Corporation (SPC), now AREVA Inc., 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 Nuclear Regulatory Commission (NRC) audit.

Both these conditions are met since St. Lucie Unit 2 is a CE reactor, and no additional variables were used in the statistical DNB analysis.

The DNB calculations were performed utilizing the NRC-approved HTP critical heat flux (CHF)

(or DNB) correlation described in the Reference 5 topical report. The fuel design parameters for AREVA's CE 16x16 HTPTM assembly are within the applicable range for the HTP CHF correlation. The AREVA fuel transition operating conditions are within the applicable range of coolant conditions for the HTP CHF correlation. A mixed core penalty was applied to the CHF correlation limit in accordance with Reference 3.

%_fJIILI LJIIZLA LJUL.UI I It:I II.

AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 34 The post-scram Main Steam Line Break (MSLB) event was analyzed using the Biasi CHF correlation (Reference 12) due to the fuel rod diameter for St. Lucie Unit 2 being outside the range of applicability for the Modified Barnett correlation. The fuel design parameters for AREVA's CE 16x16 HTPTM assembly are within the applicable range for the Biasi CHF correlation. The post-scram MSLB conditions are within the applicable range of coolant conditions for the Biasi CHF correlation, or correlation bounds were conservatively used as allowed in the Reference 1 methodology. A mixed core penalty was applied to the CHF correlation limit in accordance with Reference 3.

CEA ejection analysis was performed to determine the margin to DNB and fuel centerline melt or amount of fuel failures based on DNB and fuel centerline melt using the approved Reference 1 methodology. The approved methodology for calculating the enthalpy deposition for a CEA ejection accident is given in Reference 6.

Reference 13 incorporates M5 properties into the S-RELAP5 based non-LOCA methodology.

No restrictions or requirements were identified in the SER for the Reference 13 methodology relative to its application to S-RELAP5 non-LOCA analyses.

2.8 Computer Codes Descriptions of the computer codes used in the safety analyses are provided below. Table 2.3 lists the principal computer codes used in each of the non-LOCA analyses.

S-RELAP5 - The S-RELAP5 code is an AREVA modification of the RELAP5/MOD2 code. S-RELAP5 was used for simulation of the transient system response to accidents.

Control volumes and junctions are defined which describe the 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.

RODEX2 - RODEX2 (References 7 and 8) 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

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'%AJ lU AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 35 benchmarked and its predictive capabilities were correlated over a wide range of conditions applicable to light water reactor fuel conditions.

RODEX2 was used to establish the fuel centerline melt LHR as a function of exposure.

A penalty to address TCD was applied where applicable.

COPERNIC - Like RODEX2, COPERNIC (Reference 9) 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.

However, COPERNIC accounts for TCD with increasing rod exposure whereas RODEX2 does not. To account for the effects of TCD in the 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 in accordance with Reference 1, 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 1. Average core and hot spot fuel properties and gap coefficients were conservatively calculated for U0 2 and Gd 2-0 3 bearing fuel rods at BOC and EOC.

XCOBRA-IIIC - The XCOBRA-IIIC code (Reference 10) is a steady-state thermal-hydraulics code that calculates the axial and radial flow and enthalpy distributions within assemblies and sub-channels for non-LOCA events. When used in conjunction with core boundary conditions from the S-RELAP transient analysis and the HTP DNB correlation (Reference 5), XCOBRA-IIIC also calculates the corresponding MDNBR.

MDNBR calculations are performed in a two-step process. Calculations are first performed on a core-wide basis to calculate the axially varying flow and enthalpy distributions in the peak powered fuel assembly. Next, these flow and enthalpy boundary conditions are applied to a sub-channel model of the peak powered assembly to determine the local conditions for the calculation of MDNBR.

The Biasi correlation (Reference 12) was used with XCOBRA-IIIC for the post-scram MSLB analysis as approved in Reference 1.

PRISM - AREVA's Pressurized Water Reactor (PWR) neutronics methodology uses the NRC-approved advanced nodal simulator code system SAV95. SAV95 is built around the assembly spectrum/depletion code system MICBURN-3/CASMO-3 developed by Studsvik Scandpower and the three-dimensional reactor code PRISM. PRISM is a three-dimensional, coarse-mesh reactor simulator using two-group diffusion theory. The simulator code models the reactor core in three-dimensional (X-Y-Z) geometry, and the reactor calculations can be performed in quarter- or full-core geometry. The code calculates the reactor core reactivity, nodal power distribution, pin power distribution,

k_,UI ILI* VlIZU LUtoUl 0 IC1;: IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Pacie 36 and in-core detector responses and can be used to simulate fuel shuffling, insertion, and discharge. A summary of the key validation results for the SAV95 code system is presented in Reference 11.

%_LIu ILO UJIICU LJut-fA..J 0 M-1I Et AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 37 3.0 CHAPTER 15 DISPOSITION OF EVENTS The non-LOCA events are presented in St. Lucie Unit 2 Updated Final Safety Analysis Report (UFSAR) Chapter 15. Each event is categorized with respect to its potential consequences.

The events fall into two principal classifications: anticipated operational occurrence (AOO) and postulated accident (PA). Where applicable, the RPS and/or ESF were assumed to fulfill their functions, as needed, to mitigate the consequences of a given event.

A summary of the Chapter 15 non-LOCA disposition of events is provided in Table 3.1. For the fuel transition, detailed analyses were performed for the non-LOCA events that potentially challenge the fuel related acceptance criteria (e.g., minimum DNBR and fuel centerline melt), as well as event specific criteria (time-to-criticality for Boron Dilution and deposited enthalpy for CEA Ejection).

Several St. Lucie Unit 2 UFSAR Chapter 15 events are affected by the transition to AREVA CE-16 HTPTM fuel, specifically because of changes in thermal-hydraulic performance and neutronics inputs to the safety analyses. On the other hand, events and/or sub-events whose key parameters are plant related system responses, e.g., core power, decay heat, auxiliary feedwater (AFW) capability, offsite power availability, safety injection and/or charging capability, safety valve performance, etc., rather than the fuel design parameters, were not analyzed for the fuel transition.

Section 4.0 discusses the Chapter 15 non-LOCA disposition of events and analyses for the AREVA fuel transition.

~,UI ILl 1)11 U LJ~J~LAI I I~I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary ReDort Paae 38 Table 3.1 Event Disposition Summary of Results UFSAR Seo Event Description Disposition Section 15.1.1 Feedwater System Malfunctions That Result in a Reanalyzed for challenge to Decrease in Feedwater Temperature SAFDLs.

15.1.2 Feedwater System Malfunctions That Result in an HZP and HFP cases reanalyzed for Increase in Feedwater Flow challenge to SAFDLs.

15.1.3 Excessive Increase in Secondary Steam Flow (Excess Reanalyzed for challenge to Load) SAFDLs.

or HFP (pre-trip): Bounded by 15.1.3 Inadvertent Opening of a Steam Generator Relief 15.1.4 Safety Valve HZP (post-trip): Bounded by 15.1.6 since 15.1.6 meets AOO criteria.

15.1.5 Pre-Trip Steam System Piping Failure Reanalyzed for challenge to fuel failure limits Reanalyzed for challenge to fuel 15.1.6 Post-Trip Steam System Piping Failure failure limits The radiological analysis is outside the scope of this document.

15.2.1 Loss of External Electrical Load Bounded by 15.2.3 15.2.2 Turbine Trip Bounded by 15.2.3 Reanalyzed for challenge to 15.2.3 Loss of Condenser Vacuum SAFDLs.

Bounded by AOR for overpressure.

15.2.4 Inadvertent Closure of MSIVs (BWR) Not applicable to St. Lucie Unit 2 15.2.5 Steam Pressure Regulator Failure Not part of the licensing basis Loss of Non-Emergency AC Power to the Station Bounded by 15.3.2 for DNBR.

Auxiliaries Bounded by AOR for overpressure.

15.2.7 Loss of Normal Feedwater Flow Bounded by AOR DNB prior to reactor trip is bounded by the AREVA analysis for 15.1.5.

Bounded by AOR for overpressure 15.2.8 Feedwater System Pipe Break ed andlong-term d c heatre moval and decay heat removal.

The radiological analysis is outside the scope of this document..

Transients Resulting from the Malfunction of One Steam Generator Loss of Load to One SG (Single MSIV Closure) Reanalyzed for challenge to 15.2.9 SAFDLs.

Excess Load to One SG Bounded by Loss of Load to One SG Loss of Feedwater to One SG Bounded by Loss of Load to One SG Excess Feedwater to One SG Bounded by Loss of Load to One SG 15.3.1 Partial Loss of Forced Reactor Coolant Flow Bounded by 15.3.2 15.3.2 Complete Loss of Forced Reactor Coolant Flow Reanalyzed for challenge to SAFDLs.

Reanalyzed for challenge to fuel failure limit.

Bounded by AOR for RCS Reactor Coolant Pump Shaft Seizure (Locked Rotor) o u re.

15.3.3 overpressure.

The radiological analysis is outside the scope of this document.

15.3.4 Reactor Coolant Pump Shaft Break Bounded by 15.3.3

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Section Event Description Disposition 15.4.1 Uncontrolled CEA Bank Withdrawal from a Subcritical or Reanalyzed for challenge to Low Power Startup Condition SAFDLs.

Reanalyzed for challenge to 15.4.2 Uncontrolled CEA Bank Withdrawal at Power SAFDLs.

I Bounded by AOR for overpressure.

15.4.3 CEA Misoperation (Dropped CEA) Reanalyzed for challenge to SAFDLs.

15.4.4 Startup of an Inactive RCP at an Incorrect Temperature Not applicable to St. Lucie Unit 2 A Malfunction or Failure of the Flow Controller in a BWR 15.4.5 Recirculation Loop That Results in an Increased Reactor Not applicable to St. Lucie Unit 2 Coolant Flow Rate Modes 2 through 6 reanalyzed for time-to-criticality.

CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant Mode 1 is bounded by 15.4.2.

15.4.7 Inadvertent Loading of a Fuel Assembly Bounded by AOR Reanalyzed for challenge to fuel failure limits and energy deposition.

15.4.8 Spectrum of CEA Ejection Accidents RCS overpressure bounded by AOR.

Radiological analysis is outside the scope of this document.

15.5.1 Inadvertent Operation of the ECCS During Power Bounded by 15.5.2 15.5.1__ Operation 15.5.2 CVCS Malfunction that Increases Reactor Coolant Bounded by AOR 15.5.2___ Inventory Reanalyzed for challenge to Inadvertent Opening of a Pressurizer Safety or Relief SAFDLs.

15.6.1 Valve Pressurizer overfill case is bounded by the AOR.

Break in Instrument Line or Other Lines from the Reactor Bounded by 15.6.1 relative to DNBR.

15.6.2 Coolant Pressure Boundary that Penetrate the Radiological analysis is outside the Containment scope of this document.

Transient analysis is bounded by the AOR.

15.6.3 Steam Generator Tube Rupture (SGTR) Radiological analysis is outside the scope of this document.

15.6.4 BWR Steam System Piping Failures Outside Not applicable to St. Lucie Unit 2 Containment 15.6.5 Loss-of-Coolant Accidents Outside scope of this document 15.7 Radioactive Releases from a Subsystem or Component Outside scope of this document.

15.8 Primary System Pressure Deviation Events Bounded by AOR 15.9 Anticipated Transients Without Scram (ATWS) Bounded by AOR 15.10 Station Blackout Bounded by AOR.

J#I ILI %UHIýU LJýU~.sU 9IZI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 40 4.0 CHAPTER 15 ACCIDENT AND TRANSIENT ANALYSES 4.1 FeedwaterSystem Malfunctions That Result in a Decreasein Feedwater Temperature (UFSAR 15.1.1) 4.1.1 Accident Description The bounding initiator for the feedwater malfunction event resulting in a decrease in feedwater temperature (or the decrease in feedwater temperature event), which causes an increase in primary-to-secondary heat transfer, is the loss of a train of the high-pressure feedwater heaters.

In the event of a loss of high-pressure feedwater heaters, there will be an immediate reduction in feedwater temperature to the steam generators.

At HFP, the decrease in feedwater temperature creates a mismatch between the energy being generated in the reactor core and the energy being removed by the secondary system and results in a cooldown of the primary system. A power increase will occur if the moderator temperature reactivity feedback coefficient is negative. If the power increase is sufficiently large, either overpower or thermal margin limits will be reached with the event being terminated by a reactor trip. If the power increase is less significant, the reactor will stabilize at an elevated power level without reaching a reactor trip with no significant challenge to the SAFDLs.

At HZP, the addition of cold feedwater may cause a decrease in RCS temperature and thus a reactivity insertion due to the effects of the negative moderator temperature coefficient, consistent with the HFP event response. However, the rate of energy change introduced by the event is reduced as the initial load and feedwater flow decrease below their HFP values, so that the transient initiated from a lower power is less severe than the full power case.

4.1.2 Input Parameters and Assumptions Key input parameters and their values used in the analysis of this event were consistent with or conservative relative to the approved Reference 1 methodology, as supplemented in Section 2.7. See Section 2.0 for key input parameter values.

Initial Conditions - The event was assumed to initiate from HFP conditions with a maximum core inlet temperature and TS minimum RCS flow. This set of conditions minimizes the initial margin to DNB.

%-IllulOLD11jI1k=U Ljutoul I ICA It AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 41 Reactivity Feedback - The reactivity feedback coefficients were biased according to the approved methodology. The moderator reactivity feedback was analyzed using the most negative limit. The negative MTC leads to higher power levels during the event as a result of the primary system cooldown. Doppler reactivity was biased to minimize the effects of negative feedback from increasing fuel temperatures.

Reactor Protection System Trips and Delays - The event is primarily protected by the Variable High Power (VHP) trip, which terminates the moderator feedback driven power excursion. The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, rod insertion was delayed to account for CEA holding coil delay time.

The overcooling of the primary system results in decalibration of both the excore nuclear detectors and calculated thermal power signal used as inputs to RPS trips.

This results in a delay for the RPS reactor power to reach the VHP trip setpoint. In addition, the resistance temperature detector (RTD) delay times for the hot leg and cold leg were conservatively biased to minimize the measured thermal power and further delay the VHP trip.

Feedwater Systems - The main feedwater (MFW) temperature was reduced to 326°F and the MFW control valves were assumed to maintain the nominal full power volumetric flow rate to both steam generators during the event.

Main Steam System - The turbine control valve (TCV) is assumed to be operating in automatic control mode to maintain constant steam flow during the event. The TCV is assumed to be capable of passing 120% of the nominal steam flow at its valve wide open position. Maximizing the capacity of the TCV worsens the cooldown of the RCS and the moderator reactivity driven power response.

Gap Conductance - Gap conductance was set to a conservative EOC value to maximize the heat flux through the cladding and minimize the negative reactivity inserted due to Doppler feedback. The gap conductance accounts for TCD.

Steam Generator Tube Plugging - No SGTP was assumed so as to maximize the primary-to-secondary side heat transfer which exacerbates the reactivity insertion due to moderator feedback.

Single Failure - There is no single failure that will adversely affect the consequences since the systems designed to mitigate this event (namely, the RPS) are redundant.

4.1.3 Acceptance Criteria The principally challenged acceptance criterion for this event is:

Fuel cladding integrity should be maintained by ensuring that the Specified Acceptable Fuel Design Limits (SAFDLs) are not exceeded. This criterion is met by assuring that the minimum calculated DNBR is not less than the 95/95 DNB

ILI Ullt-'U LjuqýUl 01r;I it AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 42 correlation limit. Additionally, FCM is demonstrated to be precluded in the most adverse location in the core.

4.1.4 Method of Analysis For the fuel transition, detailed analyses were performed with the approved non-LOCA methodology given in Reference 1, as supplemented in Section 2.7. For this event, the S-RELAP5 code was used to model the key system components and calculate neutron power, fuel thermal response, surface heat transport, fluid conditions (such as coolant flow rates, temperatures, and pressures), and an estimated time of MDNBR. The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code (Reference 3), which was used to calculate the MDNBR using the HTP CHF correlation (Reference 5).

4.1.5 Results The sequence of events is given in Table 4.1 and the results are given in Table 4.2. State-points for the DNB calculations were chosen at and near the time of peak heat flux. The limiting MDNBR was calculated to be above the 95/95 CHF correlation limit. The peak LHR was calculated to be less than the fuel centerline melt limit.

The transient response is shown in Figure 4.1 through Figure 4.8. Figure 4.1 shows the reactor power as a function of time and Figure 4.2 shows the core power based on rod surface heat flux. Figure 4.3 through Figure 4.8 show the pressurizer pressure, the RCS loop temperatures, the total RCS flow rate, the steam generator pressures, the steam and feedwater flow rates, and the reactivity feedback, respectively.

%"IPIIU WJII~U LjUL,1UI 0 II;D It AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Pnni~A441 Pane Table 4.1 Decrease in Feedwater Temperature: Sequence of Events Case Event Time (sec.)

Hot Full Power Loss of feedwater heaters occurs 0.00 VHP trip setpoint reached 43.58 Reactor scram on VHP trip (including trip response 44.00 delay)

CEA insertion begins 44.75 Minimum (prior to scram) core inlet temperature 44.75 reached Peak neutronic power 44.78 MDNBR 44.80 Maximum clad surface heat flux 44.83 Table 4.2 Decrease in Feedwater Temperature: Results Criterion Result Limit MDNBR 1.257 1.164 Peak LHR 18.24 kW/ft [ ] kW/ft

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0 a- 50 25 0

0 10 20 30 40 50 60 Time (s)

Figure 4.1 Decrease in Feedwater Temperature: Reactor Power

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2000 0

it 1000 0

0 10 20 30 40 50 60 Time (s)

Figure 4.2 Decrease in Feedwater Temperature: Total Core Heat Flux Power

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2150 a.

2100 2050 2000 0 10 20 30 40 50 60 Time (s)

Figure 4.3 Decrease in Feedwater Temperature: Pressurizer Pressure

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E 560 540 0 10 20 30 40 50 60 Time (s)

Figure 4.4 Decrease in Feedwater Temperature: RCS Loop Temperatures

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CD

.0 39000 368000 0

-n U

0 38800 366000 U-ci, 364000 a)

U 38600 364000 38400 362000 38200 - 360000 0 10 20 30 40 50 60 Time (s)

Figure 4.5 Decrease in Feedwater Temperature: RCS Total Loop Flow Rate

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0-900 800 L 0 10 20 30 40 50 60 Time (s)

Figure 4.6 Decrease in Feedwater Temperature: Steam Generator Pressures

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-a S

0

.0

-- 0 SG-1 Steam

.......... mSG-2 Steam ca 1000 ---

SG-1 MFW 0 -- A SG-2MFW LL 0

0 S

OFý

-1000 1 0 10 20 30 40 50 60 Time (s)

Figure 4.7 Decrease in Feedwater Temperature: Steam and Feedwater Flow Rates

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0 i;;

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Figure 4.8 Decrease in Feedwater Temperature: Reactivity Feedback

%jLII ILl I.JIIU LJ~t-UI 5 I.I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paqe 52 4.2 FeedwaterSystem Malfunctions That Result in an Increase in FeedwaterFlow (UFSAR 15.1.2) 4.2.1 Accident Description A change in steam generator feedwater conditions that results in an increase in feedwater flow could result in excessive heat removal from the RCS. Such changes in feedwater flow are a result of a failure of a feedwater control valve or feedwater bypass valve, failure in the feedwater control system, operator error or accidental starting of the auxiliary feedwater system. An example of an increase in feedwater flow event resulting in excessive heat removal is the accidental opening of the feedwater regulating valves, which results in an increase in feedwater flow to both steam generators.

At HFP, the excess feedwater flow creates a mismatch between the energy being generated in the reactor core and the energy being removed by the secondary system and results in a cooldown of the primary system. A power increase will occur if the moderator temperature reactivity feedback coefficient is negative. If the power increase is sufficiently large, either overpower or thermal margin limits will be reached with the event being terminated by a reactor trip. If the power increase is less significant, the reactor will stabilize at an elevated power level without reaching a reactor trip with no significant challenge to the SAFDLs.

At HZP, the addition of cold feedwater may cause a decrease in RCS temperature and thus a reactivity insertion due to the effects of the negative moderator temperature coefficient, consistent with the HFP event response.

4.2.2 Input Parameters and Assumptions The key input parameters and their values used in the analysis of this event were consistent with or conservative relative to the approved Reference 1 methodology, as supplemented in Section 2.7. See Section 2.0 for key input parameter values.

Initial Conditions - This event was analyzed from both HFP and HZP conditions with a maximum core inlet temperature and TS minimum RCS flow. This set of conditions minimizes the initial margin to DNB.

Reactivity Feedback - The reactivity feedback coefficients were biased according to the approved methodology. The moderator reactivity feedback was analyzed using the most negative limit. The negative MTC leads to higher power levels during the event

%_'U11UU11t.;U Liul-fuillivilL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 53 as a result of the primary system cooldown. Doppler reactivity was biased to minimize the effects of negative feedback from increasing fuel temperatures.

Reactor Protection System Trips and Delays - The event is primarily protected by the VHP trip, which terminates the moderator feedback driven power excursion. The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, rod insertion was delayed to account for CEA holding coil delay time.

The overcooling of the primary system results in decalibration of both the excore nuclear detectors and calculated thermal power signal used as inputs to RPS trips.

This results in a delay for the RPS reactor power to reach the VHP trip setpoint. In addition, the RTD delay times for the hot leg and cold leg were conservatively biased to minimize the measured thermal power and further delay the VHP trip.

Feedwater Systems - The MFW flow was increased to 130% of the nominal full power volumetric flow rate for the HFP case and delivered to both steam generators during the event. The MFW temperature was reduced to 340"F in conjunction with the increase in MFW flow to account for the lower MFW heater residency time, and thus decreased MFW heating, that may be associated with the increased MFW flow through the heaters. For the HZP case, MFW flow was increased to 120% of the nominal full power volumetric flow rate and delivered to both steam generators during the event at a constant temperature of 235°F.

Main Steam System - The TCV is assumed to be operating in automatic control mode to maintain constant steam flow during the HFP event. The TCV is assumed to be capable of passing 120% of the nominal steam flow at its valve wide open position.

Maximizing the capacity of the TCV worsens the cooldown of the RCS and the moderator reactivity driven power response. At HZP, the TCV is assumed to be in manual mode and does not open in response to the increase In MFW flow.

Gap Conductance - Gap conductance was set to a conservative EOC value to maximize the heat flux through the cladding and minimize the negative reactivity inserted due to Doppler feedback. For the HZP case, The gap conductance accounts for the effect of TCD.

Steam Generator Tube Plugging - No SGTP was assumed so as to maximize the primary-to-secondary side heat transfer, which exacerbates the reactivity insertion due to moderator feedback.

Single-Failure - There is no single-failure that will adversely affect the consequences since the systems designed to mitigate this event (namely, the RPS) are redundant.

%.I#UI ILI UJII;U LJUtoUI I IUZI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 54 4.2.3 Acceptance Criteria The principally challenged acceptance criterion for this event is:

Fuel cladding integrity should be maintained by ensuring that the SAFDLs are not exceeded. This criterion is met by assuring that the minimum calculated DNBR is not less than the 95/95 DNB correlation limit. Additionally, FCM is demonstrated to be precluded in the most adverse location in the core.

4.2.4 Method of Analysis For the fuel transition, detailed analyses were performed with the approved non-LOCA methodology given in Reference 1, as supplemented in Section 2.7. For this event, the S-RELAP5 code was used to model the key system components and calculate neutron power, fuel thermal response, surface heat transport, fluid conditions (such as coolant flow rates, temperatures, and pressures), and an estimated time of MDNBR. The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code (Reference 3), which was used to calculate the MDNBR using the HTP CHF correlation (Reference 5). For the HZP analysis, a hot spot fuel centerline temperature model was employed to evaluate margin to fuel melt.

4.2.5 Results The sequence of events for the HFP analysis is given in Table 4.3 and the results are given in Table 4.4. State points for the DNB calculations were chosen at and near the time of peak heat flux. The limiting MDNBR was calculated to be above the 95/95 CHF correlation limit. The peak LHR was calculated to be less than the fuel centerline melt limit.

The transient response is shown in Figure 4.9 through Figure 4.15. Figure 4.9 shows the reactor power as a function of time and Figure 4.10 shows the core power based on rod surface heat flux. Figure 4.11 through Figure 4.15 show the pressurizer pressure, the RCS loop temperatures, the steam generator pressures, the steam and feedwater flow rates, and the reactivity feedback, respectively.

The sequence of events for the HZP analysis is given in Table 4.3 and the results are given in Table 4.4. The fuel centerline temperature was less than the melt temperature for any fuel composition in the core and the limiting MDNBR was calculated to be above the 95/95 CHF

%.OUIILIUIICU L/UUjLII IUI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paqe 55 correlation limit. There is no post-trip return to power, and therefore, no challenge to the minimum DNBR criterion. The transient response is shown in Figure 4.16 through Figure 4.23.

Figure 4.16 shows the reactor power as a function of time and Figure 4.17 shows the core power based on rod surface heat flux. Figure 4.18 through Figure 4.23 show the pressurizer pressure, the RCS loop temperatures, the steam generator pressures, the steam and feedwater flow rates, the reactivity feedback, and the peak centerline fuel temperature, respectively.

%-#lILI UlIc~u LJUOL.UI I IMI it AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report P*a* 5*

56 Pane Table 4.3 Increase in Feedwater Flow: Sequence of Events Case Event Time (sec)

Hot Full Power Increase in feedwater flow occurs 0.00 VHP trip setpoint reached 32.08 Reactor scram on VHP trip (including trip response 32.48 delay)

CEA insertion begins 33.23 Minimum (prior to scram) core inlet temperature reached 33.23 Peak neutronic power 33.24 Maximum clad surface heat flux 33.29 MDNBR 33.30 Hot Zero Power Increase in feedwater flow occurs 0.00 VHP trip setpoint reached 24.75 Peak neutronic power 24.85 MDNBR 24.95 Maximum clad surface heat flux 24.97 Reactor scram on VHP trip (including trip response 25.15 delay)

CEA insertion begins 26.59 Maximum fuel centerline temperature 29.30 Scram Complete 29.44 SG Level reaches 100% NR Span / MFW Isolated 29.53 Table 4.4 Increase in Feedwater Flow: Results Case Criterion Result Limit Hot Full Power MDNBR 1.220 1.164 Peak LHR 18.50 kW/ft [ ]kW/ft Hot Zero Power MDNBR 5.857 1.164 Peak CLT 3385 0 F [ ]F

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0 10 20 30 40 50 60 Time (s)

Figure 4.9 Increase in Feedwater Flow: Reactor Power (HFP)

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- 2000 0

0 1 1000 0 L 0 10 20 30 40 50 60 Time (s)

Figure 4.10 Increase in Feedwater Flow: Total Core Heat Flux Power (HFP)

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~I1 a.

0 0

£1-2100 2000 0 10 20 30 40 50 60 Time (s)

Figure 4.11 Increase in Feedwater Flow: Pressurizer Pressure (HFP)

I.f[ILI lJIIZLU LJUUUI I RI IEL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Pagje 60 620 600 a) 580 0.

E 56 560 540 L 0 10 20 30 40 50 60 Time (s)

Figure 4.12 Increase in Feedwater Flow: RCS Loop Temperatures (HFP)

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Figure 4.13 Increase in Feedwater Flow: Steam Generator Pressures (HFP)

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-* 2000

.. .... = 4 -: -team

--. SG-1 Steam

...... SG-2 Steam

- - -. SG-1 MFW U. -- A SG-2 MFW 1000 0 0 10 20 30 40 50 60 Time (s)

Figure 4.14 Increase in Feedwater Flow: Steam and Feedwater Flow Rates (HFP)

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0 0

(U S

-5

-10L 0 10 20 30 40 50 60 Time (s)

Figure 4.15 Increase in Feedwater Flow: Reactivity Feedback (HFP)

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--. Core Power 300 I I-200 1-ED 0

0~

100 r 0

0 10 20 30 40 50 60 Time (s)

Figure 4.16 Increase in Feedwater Flow: Reactor Power (HZP)

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x U-400 w

0 "I

0 0* 200 0

-200 0 10 20 30 40 50 60 Time (s)

Figure 4.17 Increase in Feedwater Flow: Total Core Heat Flux Power (HZP)

LIIIVlJIe;U LJU~oUl I It;I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 66 2300 2200 "i 2100 2000 1900 1800 0 10 20 30 40 50 60 Time (s)

Figure 4.18 Increase in Feedwater Flow: Pressurizer Pressure (HZP)

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CL E

510 500 490 0 10 20 30 40 50 60 Time (s)

Figure 4.19 Increase in Feedwater Flow: RCS Loop Temperatures (HZP)

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Figure 4.20 Increase in Feedwater Flow: Steam Generator Pressures (HZP)

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-- 0 SG-1 Steam

.. SG-2 Steam 2000

- -- SG-I MFW

-- A--SG-2 MFW U-a)

CO 1000 0f

-1000 0 10 20 30 40 50 60 Time (s)

Figure 4.21 Increase in Feedwater Flow: Steam and Feedwater Flow Rates (HZP)

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2 0

Z; -2

-4

-6

-8

-10 L 0 10 20 30 40 50 60 Time (s)

Figure 4.22 Increase in Feedwater Flow: Reactivity Feedback (HZP)

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. 2000 0.

E I- 0 1O00 0 L 0 5 10 15 20 25 30 35 40 45 50 Time (s)

Figure 4.23 Increase in Feedwater Flow: Peak Fuel Centerline Temperature (HZP)

\.AJI ILI V~IIC;U LUUUU~I I M51 IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 72 4.3 Excessive Increase in Secondary Steam Flow (UFSAR 15.1.3) 4.3.1 Accident Description This event is defined as a rapid increase in the steam flow that causes a power mismatch between the reactor core power and the steam generator load demand. For this event, the steam flow at HFP could increase by up to 58% of the initial value.

This accident could result from one of the following:

An administrative violation such as excess loading by the operator.

Equipment malfunction in the steam dump control.

Turbine throttle valve control malfunction.

The increased steam flow draws more heat from the primary side. This reduces the temperature of the water in the RCS. In the presence of a negative moderator temperature coefficient, the RCS temperature reduction can result in a nuclear power increase. The reduced coolant temperature also results in a reduction in the pressurizer water volume, due to the increased density of the cooler RCS water, and a reduction in the pressurizer pressure. Given an increase in core power and the reduction in the RCS pressure, the possible consequence of this event (assuming no protective functions) is violation of the DNB and FCM SAFDLs with subsequent fuel damage.

4.3.2 Input Parameters and Assumptions The key input parameters and their values used in the analysis of this event are consistent with or conservative relative to the approved Reference 1 methodology, as supplemented in Section 2.7. See Section 2.0 for key input parameter values.

Initial Conditions - This event was analyzed from both HFP and HZP conditions with a maximum core inlet temperature and TS minimum RCS flow. This set of conditions minimizes the initial margin to DNB.

Reactivity Feedback - The reactivity feedback coefficients were biased according to the approved methodology. A spectrum of EOC moderator temperature feedback was conservatively assumed for the HFP event. The most negative MTC was assumed for the HZP case.

Reactor Protection System Trips and Delays - This event is primarily protected by the VHP and the TM/LP trips. The RPS trip setpoints and response times were conservatively biased to

%AJ-nU11UU1=U L-Ut-UI81tI1L AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 73 delay the actuation of the trip function. In addition, rod insertion was delayed to account for CEA holding coil delay time.

Fuel Rod Gap Conductance - Gap conductance was set to a conservative EOC value, to be consistent with the time-in-cycle for the reactivity coefficients and accounts for the effect of TCD.

Steam Generator Tube Plugging - This analysis conservatively assumed 0% SGTP.

a Auxiliary Feedwater - A maximum AFW flow rate and a minimum AFW temperature were modeled to maximize the cooldown for the HZP case.

Reactor Coolant Pumps - In the HZP case, the Reactor Coolant Pumps (RCPs) were assumed to be tripped at the time of the Safety Injection Actuation Signal (SIAS) in order to delay the time for boron from the High Pressure Safety Injection (HPSI) pumps to reach the core.

4.3.3 Acceptance Criteria The principally challenged acceptance criterion for this event is:

Fuel cladding integrity should be maintained by ensuring that the SAFDLs are not exceeded.

This criterion is met by assuring that

The minimum calculated DNBR is not less than the 95/95 DNB correlation limit, and

The FCM limit is not exceeded.

Note that this event results in RCS depressurization and, thus, does not challenge the reactor coolant pressure boundary limit.

4.3.4 Method of Analysis The analysis was performed with the approved non-LOCA methodology given in Reference 1, as supplemented in Section 2.7. For this event, the S-RELAP5 code was used to model the key system components and calculate neutron power, fuel thermal response, surface heat transport, fluid conditions (such as coolant flow rates, temperatures, and pressures), and an estimated time of MDNBR. The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code (Reference 3), which was used to calculate the MDNBR using the HTP CHF correlation (Reference 5). This event was also addressed as part of the Thermal Margin / Low Pressure (TM/LP) trip statistical setpoint analyses using the Reference 4 methodology.

'jIELI LJEIC:A L/Lju...,U I 11;1IEL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 74 A spectrum of calculations, varying the MTC from -10 pcm/°F to the most-negative limit of -33 pcm/°F, was performed at EOC HFP conditions, maximum TS core inlet temperature, and minimum TS RCS flow rate. The analysis simulated an inadvertent and instantaneous full opening of the Steam Bypass Control System (SBCS) valves, while holding the TCV at the HFP position. This produced the maximum reactor coolant system cooling and, thus, the minimum margins to the DNB and FCM limits.

A single HZP case was analyzed, using the most negative moderator temperature reactivity feedback. The analysis simulated an inadvertent and instantaneous full opening of the TCV, which maximizes the cooling of the reactor coolant system and the challenge to the DNB and FCM SAFDLs.

4.3.5 Results The sequence of events for the HZP case is shown in Table 4.5. The MDNBR and peak CLT results are given in Table 4.6. The MDNBR was calculated to be above the 95/95 limit for the HTP CHF correlation and the maximum fuel centerline temperature was less than the melting temperature for any fuel composition in the core. The system responses are shown in Figure 4.24 through Figure 4.32.

The sequence of events for the limiting HFP case is shown in Table 4.5. The MDNBR and peak LHR results are given in Table 4.6. The MDNBR was calculated to be above the 95/95 limit for the HTP CHF correlation and the peak LHR was less than the melt limit for any fuel composition in the core. The system responses are shown in Figure 4.33 through Figure 4.40.

'%.-OJI ILI LUIZl:U LJýJLUkdI IC;I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report panp 75~

Pane 75 Table 4.5 Increase in Steam Flow: Sequence of Events Case Event Time (sec)

Hot Zero Power Turbine control valves fully opened; AFW initiated at 0.00 maximum delivery VHP trip setpoint reached 11.54 Peak neutron power in transient 11.65 Maximum cladding surface heat flux 11.75 MDNBR 11.75 Reactor scram on VHP trip initiated 11.95 CEAs began to drop 13.40 Maximum fuel centerline temperature 16.1 Scram completed 16.25 MSIS setpoint reached 25.32 SIAS reached and RCPs tripped 28.35 MSIVs fully closed(*) 39.1 HPSI flow began 48.36 Calculation terminated 48.84 Hot Full Power SBCS actuated and instantaneously opened fully 0.00 Peak steam flow through the SBCS valves and TCVs 0.1 Maximum MFW flow rate 6.7 VHP trip setpoint 25.7 Reactor scram 26.1 CEAs started to drop into the core 26.84 Minimum core inlet temperature (pre-scram) reached 26.84 Peak neutronics power 26.85 Maximum cladding surface heat flux 26.9 MDNBR 26.90 CEAs fully inserted 29.69 Calculation terminated 41.86 The time between reaching the MSIS setpoint and full closure of the MSIVs conservatively bounds the TS.

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~-200 100 k 0 I ..I , - , 1 . Is pp,, ,, , , , 1 , , , , I , , ý 0 5 10 15 20 25 30 35 40 45 50 Time (s)

Figure 4.24 Increase in Steam Flow: Reactor Power (HZP)

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300 0~

W 100

-100

-300

-500 0 5 10 15 20 25 30 35 40 45 50 Time (s)

Figure 4.25 Increase in Steam Flow: Total Core Heat Flux Power (HZP)

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1500 0

1000 500 '

0 5 10 15 20 25 30 35 40 45 50 Time (s)

Figure 4.26 Increase in Steam Flow: Pressurizer Pressure (HZP)

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týLl A(n 540 520 U-L.

500 0.

E 480 460 0 5 10 15 20 25 30 35 40 45 50 Time (s)

Figure 4.27 Increase in Steam Flow: RCS Loop Temperatures (HZP)

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Figure 4.28 Increase in Steam Flow: RCS Total Loop Flow Rate (HZP)

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Figure 4.29 Increase in Steam Flow: Steam Generator Pressures (HZP)

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Figure 4.30 Increase in Steam Flow: Steam Line and Feedwater Flow Rates (HZP)

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Figure 4.31 Increase in Steam Flow: Reactivity Feedback (HZP)

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Figure 4.32 Increase in Steam Flow: Peak Fuel Centerline Temperature (HZP)

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Figure 4.33 Increase in Steam Flow: Reactor Power (HFP)

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Figure 4.34 Increase in Steam Flow: Total Core Heat Flux Power (HFP)

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Figure 4.35 Increase in Steam Flow: Pressurizer Pressure (HFP)

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Figure 4.36 Increase in Steam Flow: RCS Loop Temperatures (HFP)

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Figure 4.38 Increase in Steam Flow: Steam Generator Pressures (HFP)

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Figure 4.40 Increase in Steam Flow: Reactivity Feedback (HFP)

V-41ul M U11%.;U L.Jut-Jul I Iul It AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 94 4.4 InadvertentOpening of a Steam GeneratorRelief or Safety Valve (UFSAR 15.1.4)

The inadvertent opening of a main steam relief or safety valve results in a transient similar to a steamline break event. Upon the inadvertent opening of a main steam relief or safety valve, steam flow would increase causing a mismatch between the reactor core power and the steam generator load demand. The increased steam flow draws more heat from the primary side.

This reduces the temperature of the water in the RCS. In the presence of a negative moderator temperature coefficient, the RCS temperature reduction can result in a small increase in nuclear power if analyzed from an at power condition. If analyzed from sub-critical conditions, the assumption of a stuck CEA in conjunction with a negative moderator temperature coefficient could result in a reactivity transient that may overcome the shutdown margin and may result in a subsequent return to power. In both cases, the reduced coolant temperature results in a reduction in the RCS pressure due to the increase in coolant density. The combination of primary coolant system pressure decrease due to the coolant contraction and a potential power increase may introduce a challenge to the DNBR SAFDL.

An inadvertent opening of one Main Steam Safety Valve (MSSV) would result in an increase in steam flow of about 6 to 7% of normal steam flow. This is very small compared to an increase in steam flow of about 58% for the Excess Load event (UFSAR 15.1.3). With a small increase in steam flow for this event, there would be a small increase in power, the plant would reach a new steady state condition, and there would be no reactor trip. In addition, although this is an asymmetric event, the amount of asymmetry is minor such that there would be no significant augmented power peaking. Therefore, the HFP case is bounded by the Excess Load event (UFSAR 15.1.3), which does experience a reactor trip and does challenge the SAFDLs.

A case initiated from HZP will be similar to a HZP Excess Load event (UFSAR 15.1.3); however this event will progress at a slower rate due to lower steam release rates relative to a UFSAR 15.1.3 event (6% to 7% vs. 120%). Compared to a UFSAR 15.1.3 event, lower steam release rates will decrease the rate of RCS cooldown and decrease the positive reactivity addition resulting in a less significant power increase. Prior to a reactor trip, the plant would reach a new steady state condition at approximately 6% to 7% rated thermal power (RTP) with no significant challenge to the SAFDLs. Thus, prior to reactor trip, the HZP UFSAR 15.1.4 case is bounded

~UI ILl L~II~U LJUL~UI I I~I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 95 by the HZP Excess Load event (UFSAR 15.1.3) due to a greater load increase for the UFSAR 15.1.3 event.

A steam generator (SG) low pressure RPS trip would eventually occur for this event but sustained RCS cooldown from the open MSSV may erode shutdown margin and result in a return-to-power due to positive moderator reactivity feedback. If shutdown margin is eroded, the plant would again reach a steady state power condition of approximately 6% to 7% RTP.

For comparison, the Main Steam Line Break (UFSAR 15.1.6) case with the highest return-to-power (i.e., HZP with offsite power available) reached a maximum of approximately 11% RTP.

Both this event and the UFSAR 15.1.6 event assume the most reactive rod is stuck out of the core after scram which results in a localized peaking increase once a return-to-power occurs.

Since this event results in lower steam release rates and a less severe power transient, as well as sharing a common assumption of a stuck out control rod, this event is bounded after scram by the UFSAR 15.1.6 event since the more restrictive DNBR and fuel centerline melt AOO criteria are met (Table 5.1).

Any differences in the plant response for this event relative to the UFSAR 15.1.6 event will not invalidate this disposition. Regardless of the timing of AFW and HPSI, the reactivity transient and resulting core power for this event will be at most 6% to 7% RTP which is less than that for the limiting UFSAR 15.1.6 case. Once the HPSI system introduces boron into the core, the power will reduce for this event and the plant will be brought into a shutdown condition.

4.5 Pre-Trip Steam System Piping Failure(UFSAR 15.1.5) 4.5.1 Accident Description The MSLB event results from a piping break in a main steam line. The rupture of a main steam line will cause the SG pressure and temperature to rapidly decrease. With no check valves in the main steam lines, both SGs are affected.

The pre-scram portion of the MSLB analysis is concerned with the behavior prior to and just after reactor scram where there is a potential for fuel failure to occur. Analysis is terminated shortly after a scram occurs. The pre-scram analysis is similar to an Excess Load event, in that a break in the steam line causes a significant cooldown of the RCS, which when combined with an EOC MTC, results in a significant increase in power. The pre-scram MSLB analysis,

%-AJI ILI VI.JlU LJIUU~UI I ICI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 96 however, considers larger break sizes that may result in secondary side reactor trips. The pre-scram phase of an MSLB event can challenge DNB and FCM limits due to the potential increase in power and the VHP trip power input decalibration for both the ex-core detector and the AT power signals. There is also a challenge caused by harsh containment conditions on reactor trips when the break is located inside containment.

This event is analyzed for a spectrum of break sizes from 0.1 ft2 to 6.8533 ft2 and MTC values from -8 pcm/°F to -33 pcm/°F. Three scenarios for a break location are used: a break of a steam line inside containment (referred to as an asymmetric inside break), a break of a steam line outside containment but upstream of the common turbine header (referred to as an asymmetric outside break), and a break located at the common steam line header at the turbine inlet (referred to as a symmetric break). The asymmetry of the first two scenarios is caused by the difference in length of steam lines connecting each SG to the break location.

The SGs contain integral flow restrictors at the exit of the SG, which will limit the consequences of larger breaks since the steam flow will choke at the flow restrictor.

4.5.2 Input Parameters and Assumptions The key input parameters and their values used in the analysis of this event were consistent with or conservative relative to the approved Reference 1 methodology. See Section 2.0 for key input parameter values.

Initial operating conditions (hot full power plus uncertainty, TS maximum core inlet temperature, nominal RCS pressure, TS minimum RCS flow rate)

Reactivity Feedback - Since this event involves a decrease in the core coolant temperature, the event was assumed to occur at EOC with a range of cases analyzed with MTCs varying up to the most negative MTC at full power.

Reactor Protection System Trips and Delays - The event can actuate the Low SG Pressure, Containment High Pressure, VHP, TM/LP, or Low RCS Flow trips. An Asymmetric SG Pressure trip is possible, but unlikely because asymmetrical response of the SGs is only due to the differences in the lengths of steam lines to the break. The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. For cases with a break inside containment, harsh condition setpoints are used. In addition, control rod insertion is delayed to account for the CEA holding coil delay time.

V.jIJI ILI LJII;U LJULduI I K51I HL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 97 Gap Conductance and Fuel Thermal properties - Gap conductance was set to a conservatively high EOC value and the fuel thermal properties also were set to EOC values, accounting for thermal conductivity degradation.

The ex-core detector signal for the unaffected side only is used for input to the nuclear instrumentation (NI)-Power signal. The signal is decalibrated to account for detector shadowing from the cooler fluid entering the downcomer on the affected side. In addition, the AT power signal from the hot and cold leg RTDs is also decalibrated, using an RTD delay time of 0 seconds for the hot leg temperature and a maximum delay of 12 seconds for the cold leg temperature.

Main Feedwater flow is modeled with a maximum capacity of 130% of nominal.

SG Tube Plugging - 0% plugging in each SG was used since this increases the primary to secondary heat transfer and causes more of a primary cooldown.

A Failure of the Fast Bus Transfer (FFBT) was assumed, causing a RCP in each loop to trip at the time of reactor trip. A Loss of Offsite Power (LOOP) was assumed to occur 3 seconds after the time of reactor trip, causing the other RCPs to trip.

For cases inside containment, a single volume containment was modeled using a physical volume of a nominal value plus uncertainty, with a conservatively low initial pressure and a conservative (high) heat transfer from internal structures.

4.5.3 Acceptance Criteria For the pre-scram MSLB event, the principally challenged acceptance criteria are with respect to radiological consequences. This is a Condition III (Infrequent Fault) or Condition IV (Limiting Fault) event, depending on the size of the break. The amount of fuel failure is determined and compared against the values assumed in the radiological analysis.

4.5.4 Method of Analysis For the fuel transition, detailed analyses were performed with approved Reference 1 methodology. For this event, the S-RELAP5 code was used to model the key system components and calculate neutron power, fuel thermal response, surface heat transport, and fluid conditions (such as coolant flow rates, temperatures, and pressures). The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code (Reference 3), which was used to calculate the MDNBR using the HTP CHF correlation (Reference 5).

%AJI I1l LJIIZL.U L.JPt.U1sL!II II L AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 98 Cases were analyzed at EOC HFP initial conditions, maximum TS core inlet temperature and minimum TS RCS flow rate. The factors affecting scram time were biased using conservative trip signal delay and holding coil delay times to produce the most significant challenge to the DNB limit. These cases assumed an FFBT, causing an RCP in each loop to trip at the time of reactor trip, followed by a LOOP 3 seconds later which tripped the remaining RCPs. Three sets of cases were analyzed for different break locations. Each set considered break sizes from 0.1 ft 2 to the largest steam line pipe diameter at each location. The locations were:

1. Breaks located inside containment

2. Breaks located upstream of the common turbine header and outside containment

3. Breaks located at the common steam line header just upstream of the turbine inlet The presence of a flow restrictor at the SG outlet nozzle precludes the need to analyze breaks even larger than the maximum steam line piping area, i.e., up to a double-ended guillotine break (DEGB). Results showed that the severity of the event was clearly bounded using the break areas that were analyzed.

A further case was analyzed using the break size from the overall limiting case from the three scenarios above. This was the 3.0 ft2 break size from the asymmetric breaks outside containment. This case assumed a break upstream of the common turbine header and inside containment with a coincident LOOP at the time of the break resulting in the tripping of all four RCPs. The most negative MTC limit (-33 pcm/°F) was assumed.

4.5.5 Results State points for the DNB analyses were chosen at or near the time of the maximum power based on heat flux. The limiting results for each break location are given in Table 4.8. The overall limiting case was initiated from HFP conditions by a postulated 3.0 ft 2 break in a main steam line outside of the reactor containment and upstream of the Main Steam Isolation Valve (MSIV) with an MTC of -16 pcm/ 0 F. The MDNBR was calculated to be greater than the 95/95 limit for the HTP CHF correlation resulting in no fuel failure. The peak LHR was calculated to be less than the fuel centerline melt limit; thus, no fuel failure due to FCM was predicted to occur.

The sequence of events for the overall limiting case is given in Table 4.7. Plots of key system parameters for the overall limiting case are shown in Figure 4.41 through Figure 4.47. Figure

'%..,UiI LI UOICU LJUtolA I Mii IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paqe 99 4.41 shows the break, steam and feedwater flow rates as a function of time. Figure 4.42 shows the steam generator pressures. Figure 4.43 shows pressurizer pressure. Figure 4.44 shows RCS loop temperatures. Figure 4.45 shows the core inlet flow rates. Figure 4.46 shows the reactivity feedback. Figure 4.47 shows the reactor power (kinetics and power based on heat flux).

The limiting case was rerun with the same break size (3.0 ft2 from the asymmetric breaks outside containment), the most negative MTC (-33 pcm/°F), and inside containment with coincident LOOP at the time of the break.

The sequence of events for this case (given in Table 4.9) shows that this case tripped on Low RCS Loop Flow (harsh setpoint) at 2.33 seconds. The maximum power based on heat flux and the maximum actual power (from the kinetics model) was 100.3% of RTP. The values of these two parameters never exceeded their initial values before the time of the break. Therefore, the MDNBR was not calculated for this transient.

%-fAJI ILI UIJIIU LJ;UtUI I IVI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chanter 15 Non-LOCA Summarv Renort Pnoe 100 Table 4.7 Pre-Scram Main Steam Line Break: Sequence of Events Event Time (sec)

With reactor at full-power / -16 pcm/°F MTC condition, postulated 3.0 ft2 0.0 break in main steam line outside containment and upstream of a MSIV occurred Indicated power (AT signal) reached VHP trip setpoint 29.01 Reactor trip signal on a VHP trip was received and turbine was tripped 29.41 FFBT caused RCPs in loops 1A and 2A to trip 29.42 Core power (based on heat flux) reached maximum value 29.45 Insertion of scram CEAs began 30.16 MDNBR 30.50 LOOP occurred and RCPs in loops 1 B and 2B were tripped 32.42 Table 4.8 Pre-Scram Main Steam Line Break: Results Case Criterion Result Limit Asymmetric Inside MDNBR (%fuel failure) 1.381 (0%) 1.164 Containment 0

(1 ft 2, -33 pcm/ F) Peak LHR (%fuel failure) Bounded (Note 2) [ ] kW/ft Asymmetric Inside MDNBR (%fuel failure) 1.431 (0%) 1.164 Containment (1.5 ft 2, -33 pcm/0 F) Peak LHR (%fuel failure) Bounded (Note 2) [ ] kW/ft Asymmetric Outside MDNBR (%fuel failure) 1.277 (Note 1) (0%) 1.164 Containment (3.0 ft 2 -16 pcm/ F) Peak LHR (%fuel failure) 17.67 kW/ft (0%) [ ] kW/ft Symmetric Outside MDNBR (%fuel failure) 1.203 (Note 1) (0%) 1.164 Containment (5 ft 2, -12 pcm/ F) Peak LHR (%fuel failure) Bounded (Note 2) [ ] kW/ft Note 1: Although the symmetric outside containment (5 ft2, -12 pcm/*F MTC) case shows the lowest MDNBR in this table, this is only due to conservatisms in the analysis. The analysis shows that the asymmetric outside containment (3.0 ft2, -16 pcm/°F MTC) case is the true limiting case.

Note 2: Bounded by value for asymmetric outside containment 3.0 ft break with -16 pcm/°F MTC.

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With reactor at full-power / -33 pcm/°F MTC condition, postulated 3.0 ft2 break in main steam line inside containment. Simultaneous Loss of Offsite Power. All RCPs tripped. Core power (based on heat flux) at maximum value.

RCS flow reaches RPS Low Flow setpoint (87.4% with harsh 1.43 conditions)

Reactor trip signal on Low Flow was received and turbine was tripped. 2.33 Insertion of scram CEAs began. 3.08 Transient analysis terminated. 13.1

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Figure 4.42 Pre-Scram Main Steam Line Break: Steam Generator Secondary Pressures

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Figure 4.43 Pre-Scram Main Steam Line Break: Pressurizer Pressure

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Figure 4.45 Pre-Scram Main Steam Line Break: Core Inlet Flow Rates

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Figure 4.47 Pre-Scram Main Steam Line Break: Reactor Power

%-.*ULJI VU'.IICLU LJUJ1 LJUI BIt5I OL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 109 4.6 Post-Trip Steam System Piping Failure(UFSAR 15.1.6) 4.6.1 Accident Description For the post-scram analysis, the MSLB event is initiated by a postulated break in a main steam line coincident with reactor scram.

The SG pressures and temperatures will decrease rapidly following the initiating event. The drop in SG pressures will initiate a Main Steam Isolation Signal (MSIS). Following appropriate delays, the MSIVs will close and terminate the blowdown from the SG with the intact main steam line (i.e., the unaffected SG).

The cooldown of the RCS will insert positive reactivity from both moderator and fuel temperature reactivity feedbacks, particularly at EOC conditions with a most-negative MTC. The magnitude of core subcriticality depends on the scram worth and the moderator and fuel temperature reactivity feedbacks.

With the most reactive control rod assumed to be stuck out of the core, the radial neutron flux (and, therefore, power) distribution will be highly peaked in the region of the stuck control rod.

The consequences would be most limiting if the core sector with the stuck control rod is also the sector being cooled primarily with coolant delivered from the cold legs of the affected loop.

The event will be terminated by the injection of boron from high pressure safety injection pumps and/or by the dryout of the affected steam generator which will stop the RCS cooldown.

Maximum AFW flow is used prior to SG isolation to maximize cooldown of the affected SG.

Automatic isolation of AFW on SG differential pressure is assumed. AFW flow is isolated after a conservatively long time delay following the isolation signal.

4.6.2 Input Parameters and Assumptions The parameters and equipment states were chosen to provide conservative calculation of fuel failures. The key input parameters and their values used in the analysis of this event are consistent with or conservative relative to the approved Reference 1 methodology. See Section 2.0 for key input parameter values.

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AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paqe 110 Initial Conditions - Two sets of initial conditions were considered. First, the event was assumed to initiate from rated power conditions with a maximum core inlet temperature.

Rated power conditions (i.e., coolant temperatures) represent the largest potential cooldown and consequential reactivity insertion. A second set of conditions assumed that the event initiated from a hot zero power condition with the minimum allowed TS shutdown margin. For both HFP and HZP initial conditions, cases were run both with and without offsite power.

Break Size and Location - A full double-ended guillotine break of a main steam line upstream of the MSIV was considered. The blowdown of the SG is limited by the flow area of the integral flow restrictor. This break size and location produces the largest cooldown which maximizes the potential return-to-power.

Break Flow - Moody critical flow model was applied at the SG integral flow restrictor. The break was modeled to maximize break flow and rate of cooldown. Steam-only flow out the break was also assumed to maximize the secondary and RCS cooldown rate.

Reactivity Feedback - This event is primarily driven by moderator feedback as a result of the cooldown of the RCS. The MTC analysis value of -33 pcm/°F was modeled which bounds the most negative TS limit of -32 pcm/OF. Minimum scram worth, appropriate for the assumed initial condition, was assumed. For the post-scram analyses, the most reactive rod was assumed to be stuck out of the core.

Gap Conductance - Gap conductance and fuel properties that account for TCD were applied. EOC gap conductance was conservatively chosen to maximize the heat flux through the cladding and minimize the negative reactivity inserted due to Doppler feedback.

Steam Generator Tube Plugging - No SGTP was assumed so as to maximize the primary-to-secondary side heat transfer, which exacerbates the reactivity insertion due to moderator feedback.

RCS Flow - Cases with all RCPs running (offsite power available) and with all RCPs stopped (loss of offsite power) were analyzed to evaluate the effects of RCS flow during the post-scram phase of the event.

Single Failure - A single failure of one of the two HPSI pumps required to be operable during plant normal operation was assumed. This single failure assumption resulted in an additional delay for boron to reach the core.

Main Steam Isolation Valve - The closure of the MSIVs was conservatively modeled to be complete at 7.0 seconds after the low steam generator pressure (487 psia) setpoint is reached.

Main Feedwater - For the post-trip analysis from HFP, MFW flow was allowed to increase up to 130% of the nominal (100%) value in response to the increase in steam flow due to the MSLB. For the post-trip analysis from HZP, MFW flow equal to the nominal (100% power) value was assumed to initiate coincident with the postulated

%ý#lI ILI VIIIZU LJLAuto I I1MI L AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 111 break. Feedwater flow was maintained until a feedwater isolation signal was generated.

The closure of the Main Feedwater Isolation Valve (MFWIV) on both affected and unaffected SGs was conservatively modeled to be complete at 5.65 seconds after the low steam generator pressure (487 psia) setpoint is reached.

4.6.3 Acceptance Criteria The principally challenged acceptance criterion for this event is the radiological consequences.

The analysis documented herein does not address radiological consequences directly; rather, the extent of fuel failure is determined which is an input to the radiological dose analyses.

4.6.4 Method of Analysis This event was analyzed to assess the impact to fuel failure for fuel transition to AREVA fuel.

For the fuel transition, detailed analyses were performed with approved methodology (Reference 1), which consists mainly of the following computer codes. The S-RELAP5 code was used to model the key system components and calculate neutron power, fuel thermal response, surface heat transport, and fluid conditions (such as coolant flow rates, temperatures, and pressures) and produce an estimated time of MDNBR. Core asymmetry was modeled by dividing the core into a sector adjacent to the affected loop and a sector adjacent to the unaffected loop. The stuck-rod region was modeled as a separate region in the affected sector.

The S-RELAP5 core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code, which was used to calculate the MDNBR using the Biasi correlation (Reference 12) for the post-scram cases as stated in Section 2.7. The PRISM code was used to calculate power distribution information and kinetics parameters.

The event was analyzed from both HZP and HFP conditions to assess the potential amount of fuel failure. Offsite power available and loss of offsite power cases were considered. For the loss of offsite power cases, loss of offsite power was assumed at event initiation. The largest break size in combination with the MTC analysis value of -33 pcm/°F that bounds the TS / Core Operating Limits Report (COLR) MTC limit of -32 pcm/°F was analyzed. Breaks located inside or outside containment were bounded because harsh conditions were assumed for all cases.

%..*UlILE VOICU LJUUUI I IUI It AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 112 4.6.5 Results The sequences of events are summarized in Table 4.10 for the cases initiated from HZP, both with and without offsite power. Table 4.11 summarizes the sequences of events for the cases initiated from HFP, both with and without offsite power. Table 4.12 summarizes the results of the analysis for this event.

The greatest challenge to the FCM limit occurred for the case initiated from HZP with offsite power available, and the greatest MDNBR challenge occurred for the case initiated from HFP with offsite power available. Key system parameters illustrating the transient response for the HZP and HFP limiting cases are presented in Figure 4.48 to Figure 4.58 and Figure 4.59 to Figure 4.68, respectively. The calculated MDNBRs were greater than the 95/95 limit for the Biasi CHF correlation; thus, no fuel failure due to DNB was predicted to occur. The peak LHRs were calculated to be less than the FCM limit; thus, no fuel failure due to FCM was predicted to occur.

%_OU11L1U11t.;U LJUUU11R-_11P;,

AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chanter 15 Non-LOCA Summary Renort Pane 113 Table 4.10 Post-Scram Main Steam Line Break: Sequence of Events (HZP)

HZP HZP Offsite Power Loss of Offsite Avail. Time Power Event (sec) Time (sec)

DEGB in main steam line occurs upstream of the MSIV, 0.0 0.0 Reactor trip CEA insertion begins 0.0 0.0 CEAs fully inserted. 0.01 0.01 MSIS on low SG pressure 9.44 8.72 Low pressurizer pressure ESF trip 14.36 18.09 MFWIVs closed, AFW flow begins 15.09 14.37 MSIVs closed 16.44 15.72 Minimum unaffected sector core inlet temperature 21.0 40.6 HPSI begins (unborated water begins to clear from the SI 34.4 48.1 lines)

Shutdown worth has been fully overcome by moderator 76.2 169.6 and Doppler feedback High SG AP ESF Trip 90.3 62.3 AFW to affected SG isolated (on high SG AP ESF trip 290.3 262.3 plus delay)

Peak post-scram reactor power 293.6 340.6 MDNBR 293.6 340.6 Borated HPSI flow begins (unborated water has been 303.7 304.7 cleared from the SI lines)

%.,PUIILI QUIt-,U L,/Ut.,uI I Icl IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Pagie 114 Table 4.11 Post-Scram Main Steam Line Break: Sequence of Events (HFP)

HFP HFP Offsite Power Loss of Offsite Avail. Power Event Time (sec) Time (sec)

DEGB in main steam line occurs upstream of the MSIV 0.0 0.0 Reactor and turbine trip 0.0 0.0 CEA insertion begins 0.74 0.74 CEAs fully inserted. 3.59 3.59 MSIS on low SG pressure 13.85 11.62 Low pressurizer pressure ESF trip 14.06 15.64 MFWIVs closed, AFW flow begins 19.50 17.27 MSIVs closed 20.85 18.62 Minimum unaffected sector core inlet temperature 25.2 46.2 HPSI pumps available (RCS pressures higher than the 34.1 45.6 HPSI pump shutoff head)

High SG AP ESF Trip 89.2 51.1 Shutdown worth has been fully overcome by moderator 194.2 464 and Doppler feedback AFW to affected SG isolated (on high SG AP ESF trip plus 289.2 251.1 delay)

Peak post-scram reactor power 293.2 885.0 MDNBR 293.2 885.0

%jIJUI ILI VIIMLl Ljutsl I ItI-I 1IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summary Rerort Paae 115 Table 4.12 Post-Scram Main Steam Line Break: Results Case Criterion Result Limit HFP, Offsite MDNBR (%fuel failure) 1.740 (0%) [ ]

Power Available Peak LHR (%fuel failure) 16.35 kW/ft (0%) [ ] kW/ft HFP, Loss of MDNBR (%fuel failure) 6.331 (0%) [ ]

Offsite Power Peak LHR (%fuel failure) 9.83 kW/ft (0%) [ ] kW/ft HZP, Offsite MDNBR (%fuel failure) 1.979 (0%) [ ]

Power Available Peak LHR (%fuel failure) 17.02 kW/ft (0%) [ ] kW/ft HZP, Loss of MDNBR (%fuel failure) 11.215 (0%) [ ]

Offsite Power Peak LHR (%fuel failure) 7.43 kW/ft (0%) [ ] kW/ft

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Figure 4.48 Post-Scram Main Steam Line Break: Break Flow Rates (HZP Offsite Power Available)

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0 200 400 600 Time (s)

Figure 4.49 Post-Scram Main Steam Line Break: Steam Generator Pressures (HZP Offsite Power Available)

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Figure 4.50 Post-Scram Main Steam Line Break: MFW Flow Rates (HZP Offsite Power Available)

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F E

-e Affected SG

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Figure 4.51 Post-Scram Main Steam Line Break: AFW Flow Rates (HZP Offsite Power Available)

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Figure 4.52 Post-Scram Main Steam Line Break: Steam Generator Mass Inventories (HZP Offsite Power Available)

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Figure 4.53 Post-Scram Main Steam Line Break: Core Inlet Fluid Temperatures (HZP Offsite Power Available)

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Figure 4.54 Post-Scram Main Steam Line Break: Pressurizer Pressure (HZP Offsite Power Available)

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Figure 4.55 Post-Scram Main Steam Line Break: Pressurizer Liquid Level (HZP Offsite Power Available)

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Figure 4.56 Post-Scram Main Steam Line Break: Total HPSI Flow Rate (HZP Offsite Power Available)

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Figure 4.57 Post-Scram Main Steam Line Break: Reactivity Feedback (HZP Offsite Power Available)

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........ Stuck-Rod Region

- -- *Rest of Affected Sector

- A- Unaffected Sector

.- 30 0

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Figure 4.58 Post-Scram Main Steam Line Break: Core Power (HZP Offsite Power Available)

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Figure 4.59 Post-Scram Main Steam Line Break: Break Flow Rates (HFP Offsite Power Available)

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Figure 4.60 Post-Scram Main Steam Line Break: Steam Generator Pressures (HFP Offsite Power Available)

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-* Affected SG a .......... E Unaffected SG (U

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Figure 4.61 Post-Scram Main Steam Line Break: MFW Flow Rates (HFP Offsite Power Available)

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Figure 4.62 Post-Scram Main Steam Line Break: AFW Flow Rates (HFP Offsite Power Available)

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Figure 4.63 Post-Scram Main Steam Line Break: Steam Generator Mass Inventories (HFP Offsite Power Available)

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Unit 2 FuelT1N Summa Re ort St. Lucie Tson Cha ter 15 Non-LOCA Pe132 600 F-CL 400 300 400 800 Time (s)

Figure 4.64 Steam (HPMainLine Break: Core Inlet Fluid Temperatures (H-FP Offsite Power Available)

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Figure 4.65 Post-Scram Main Steam Line Break: Pressurizer Pressure (HFP Offsite Power Available)

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Figure 4.66 Post-Scram Main Steam Line Break: Pressurizer Liquid Level (HFP Offsite Power Available)

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Figure 4.67 Post-Scram Main Steam Line Break: Reactivity Feedback (HFP Offsite Power Available)

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Figure 4.68 Post-Scram Main Steam Line Break: Core Power (HFP Offsite Power Available)

'jIIl LJIIML LJLJUtIU I lt:I iL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 137 4.7 Loss of CondenserVacuum (UFSAR 15.2.3) 4.7.1 Accident Description The loss of condenser vacuum event is defined as a complete loss of steam load or a turbine trip from full power without a direct reactor trip. This anticipated transient is analyzed as a turbine trip from full power with a simultaneous loss of feedwater to both steam generators due to low suction pressure on the feedwater pumps. The atmospheric dump valves and the steam dump and bypass system valves are assumed to be unavailable, which minimizes the amount of cooling and maximizes the RCS and secondary peak pressures during the event.

Offsite electrical power is available to operate the RCPs and other station auxiliaries. Following the loss of condenser vacuum, turbine stop valves close on a turbine trip, terminating the steam flow, and causing the secondary system temperature and pressure to increase. Primary-to-secondary heat transfer decreases as the secondary system temperature increases. No credit is taken for a direct reactor trip on turbine trip.

If the reactor is not tripped when the turbine is tripped, the primary system temperature and pressure will continue to rise. If this continues, the reactor will trip on high pressurizer pressure, reducing the primary heat source. As the heat load into the primary decreases, the primary system pressurization will begin to diminish. If the setpoint for opening pressurizer safety valves (PSVs) or, if appropriate, the pressurizer PORVs is exceeded during the initial system over-pressurization, these valves will open to relieve pressure and to mitigate the pressure transient.

Energy is removed during the early phase of the transient through the SG safety valves, when the SG pressure exceeds the safety valve opening setpoint.

This event bounds the following events under this category:

Turbine Trip A turbine trip event is similar to the loss of condenser vacuum event, but is not accompanied by a coincident feedwater pump trip due to low suction pressure associated with a loss of condenser vacuum. The assumption of a coincident feedwater pump trip with a turbine trip in the loss of condenser vacuum event bounds the turbine trip event, since continued feedwater

%.."UlILI LjI1C;U LJU%.OUI rl K-11 R AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 138 pump operation would be expected. The turbine trip event is thus bounded by the loss of condenser vacuum event.

Loss of Electric Load A loss of electric load event is similar to the turbine trip event, but is initiated by the closure of the TCV with a slower closure time. The assumption of a faster turbine stop valve closure without concurrent reactor trip in the loss of condenser vacuum bounds the loss of electric load event, since a less severe response from a TCV closure would be expected. The loss of electric load event is thus bounded by the loss of condenser vacuum event, which models a turbine stop valve closure on turbine trip.

Closure of Main Steam Isolation Valve A closure of MSIV event postulates that one or both of the MSIVs close to initiate the event.

The closure of both the MSIVs is not worse than the closure of the turbine stop valves on a turbine trip. The more rapid closure of the turbine stop valves produces a more severe system transient than does the closure of both MSIVs. Closure of both MSIVs is thus bounded by the loss of condenser vacuum event, which models a turbine stop valve closure on turbine trip.

The evaluation of a closure of a single MSIV is provided in Section 4.13.

4.7.2 Input Parameters and Assumptions Key input parameters and their values used in the analysis of this event were consistent with or conservative relative to the approved Reference 1 methodology, as supplemented in Section 2.7. See Section 2.0 for key input parameter values.

Initial Conditions - The event was assumed to initiate from HFP conditions with a maximum core inlet temperature and TS minimum RCS flow. This set of conditions minimizes the initial margin to DNB.

Reactivity Feedback - The reactivity feedback coefficients were biased according to the approved methodology. The most positive moderator reactivity feedback was analyzed using the most positive TS limit for any power level. Positive moderator reactivity feedback leads to higher power levels during the event as a result of the primary system heatup. Doppler reactivity was biased to minimize the effects of negative feedback from increasing fuel temperatures.

%-IJIILE Ullieu LJULIUI I IUI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 139 Reactor Protection System Trips and Delays - This event was analyzed without a direct reactor trip from the turbine trip. This assumption conservatively delayed reactor trip until conditions in the RCS resulted in a High Pressurizer Pressure (HPP) trip. The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, rod insertion was delayed to account for CEA holding coil delay time.

Feedwater Systems -MFW pumps are assumed to trip at the start of the event initiating feedwater isolation to both steam generators.

Pressurizer Pressure Control - Pressurizer pressure control (i.e., pressurizer sprays, heaters and power operated relief valves) parameters and equipment states were selected to reduce the primary system pressure, which provided a conservative calculation of the MDNBR during the transient.

Pressurizer and Main Steam Safety Valves - For this case, which evaluated the MDNBR, the PSVs and MSSVs were conservatively modeled. The opening setpoints of the MSSVs were biased high to the TS upper tolerance limits. The opening setpoints of the PSVs were biased low to the TS lower tolerance limits to provide a conservative calculation of the MDNBR during the transient by ensuring a limited pressure increase.

Gap Conductance - Gap conductance was set to a conservative BOC value to be consistent with the time-in-cycle for the reactivity coefficients. The gap conductance accounts for TCD.

Steam Generator Tube Plugging - Maximum (20%) SGTP was assumed so as to minimize the primary-to-secondary side heat transfer, which exacerbates the reactivity insertion due to moderator feedback.

Single Failure - There is no single failure that will adversely affect the consequences since the systems designed to mitigate this event (namely, the RPS) are redundant.

4.7.3 Acceptance Criteria The main purpose of analyzing this event is to demonstrate that the SAFDLs are not exceeded under the limiting assumptions of no credit for a direct reactor trip on turbine trip.

The principally challenged acceptance criterion for this event is:

Fuel cladding integrity should be maintained by ensuring that the SAFDLs are not exceeded. This criterion is met by assuring that the minimum calculated DNBR is

%fIILE 'JIIVU L-/UULI I IMI NL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 140 not less than the 95/95 DNB correlation limit. Additionally, FCM is demonstrated to be precluded in the most adverse location in the core.

Key parameters for the RCS and secondary side overpressure cases are associated with the plant system responses, rather than fuel design parameters, and include the MSSV and PSV setpoints with tolerances and capacities. Fuel design parameters are not key parameters regarding overpressure, and therefore have no significant impact on the overpressure aspects of this event. The power level, operating parameters, reactor trip setpoints, the safety valves (primary and secondary sides) opening setpoints and tolerance limits and the CEA drop characteristics and timing are unchanged from the current analysis values. There are no changes to the instrumentation uncertainties associated with the operating parameters and trip setpoints. Thus, the current UFSAR analyses of record for the RCS and secondary overpressure cases remain bounding for the fuel transition.

4.7.4 Method of Analysis For the fuel transition, detailed analyses were performed with the approved non-LOCA methodology given in Reference 1, as supplemented in Section 2.7. For this event, the S-RELAP5 code was used to model the key system components and calculate neutron power, fuel thermal response, surface heat transport, fluid conditions (such as coolant flow rates, temperatures, and pressures), and an estimated time of MDNBR. The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code (Reference 3), which was used to calculate the MDNBR using the HTP CHF correlation (Reference 5).

4.7.5 Results The sequence of events is given in Table 4.13 and the results are given in Table 4.14. State-points for the DNB calculations were chosen at and near the time of peak heat flux. The limiting MDNBR was calculated to be above the 95/95 CHF correlation limit. The peak LHR was calculated to be less than the fuel centerline melt limit.

The transient response is shown in Figure 4.69 through Figure 4.77. Figure 4.69 shows the reactor power as a function of time and Figure 4.70 shows the core power based on rod surface heat flux. Figure 4.71 through Figure 4.77 show the pressurizer pressure, pressurizer liquid

%\jLJIILI U!INZU L/LjL.UI I ICI RL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paae 141 level, pressurizer PORVs flow rate, the RCS loop temperatures, the total RCS flow rate, the steam generator pressures, and the reactivity feedback, respectively.

%'.,UI ILKUIIVU LJULJ,UI I IWIL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summarv Report Paae 142 Table 4.13 Loss of Condenser Vacuum: Sequence of Events (MDNBR Case)

Event Time (sec.)

Turbine Trip 0.0 Main Feedwater Isolation Initiated (both loops) 0.0 Steam generator Bank 1 MSSVs opened (both SGs) 4.25 HPP trip setpoint reached 6.11 Pressurizer PORVs Opened 6.11 Reactor scram on HPP trip (including trip response delay) 7.26 CEA insertion begins 8.01 Maximum (prior to scram) core inlet temperature reached 8.01 Peak neutronic power 8.01 Maximum clad surface heat flux 8.18 MDNBR 8.60 Table 4.14 Loss of Condenser Vacuum: Results Criterion Result Limit MDNBR 1.553 1.164 Peak LHR 16.04 kW/ft [ ] kW/ft

%-#U1 ILI LJItCU LJLJUtoI 11M IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 143 120 100 80 n

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Figure 4.69 Loss of Condenser Vacuum: Reactor Power (MDNBR Case)

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Figure 4.70 Loss of Condenser Vacuum: Total Core Heat Flux Power (MDNBR Case)

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Figure 4.71 Loss of Condenser Vacuum: Pressurizer Pressure (MDNBR Case)

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Figure 4.72 Loss of Condenser Vacuum: Pressurizer Liquid Level (MDNBR Case)

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Figure 4.73 Loss of Condenser Vacuum: Pressurizer PORVs Flow Rate (MDNBR Case)

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Figure 4.74 Loss of Condenser Vacuum: RCS Loop Temperatures (MDNBR Case)

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Figure 4.75 Loss of Condenser Vacuum: RCS Total Loop Flow Rate (MDNBR Case)

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Figure 4.76 Loss of Condenser Vacuum: Steam Generator Pressures (MDNBR Case)

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Figure 4.77 Loss of Condenser Vacuum: Reactivity Feedback (MDNBR Case)

Vk-OJI ILI IjKIou LJU~jUl 0 ItzI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 152 4.8 Inadvertent Closure of Main Steam Isolation Valves (BWR) (UFSAR 15.2.4)

This event is not applicable to St. Lucie Unit 2.

4.9 Steam PressureRegulatorFailure (UFSAR 15.2.5)

This event is not in the St. Lucie Unit 2 design basis.

4.10 Loss of Non-Emergency AC Power to the Station Auxiliaries (UFSAR 15.2.6)

This event is identical to the Loss of Normal Feedwater Flow event (UFSAR 15.2.7) except that a loss of power to the RCPs occurs simultaneously with the loss of feedwater flow. Key parameters for this event are associated with plant system responses, rather than fuel design parameters, and include low RCS flow trip setpoint, low SG water level Engineered Safety Feature Actuation System (ESFAS) setpoint, minimum AFW flow rate and delay time, initial core power (decay heat) and MSSV and PSV setpoints and capacities.

With a LOOP, the RCPs coast-down immediately in addition to the loss of feedwater flow. Prior to trip, the LOOP event has similar RCS conditions as the Complete Loss of Forced Reactor Coolant Flow event (UFSAR 15.3.2). For the Complete Loss of Forced Reactor Coolant Flow event, reactor trip occurs very quickly on the low RCS flow trip and main feedwater flow was assumed to be lost at event initiation. Therefore, the MDNBR is bounded by the Complete Loss of Forced Reactor Coolant Flow event.

The disposition of this event regarding overpressure is the same as for the Loss of Normal Feedwater Flow event (UFSAR 15.2.7).

4.11 Loss of Normal FeedwaterFlow (UFSAR 15.2.7)

This event is defined as a complete loss of main feedwater flow while the reactor is operating at the maximum power level. The immediate consequence of a loss of main feedwater flow is a reduction in the steam generator water level, which if left unmitigated will ultimately result in a reactor trip and auxiliary feedwater actuation on a low steam generator water level signal.

Following reactor trip, the rate of heat generation in the RCS (decay heat plus reactor coolant pump heat) may exceed the heat removal capability of the secondary system. In this case, there will be an increase in the steam generator pressure and an increase in RCS pressure,

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%ý.AJIL LJULOUI I IVI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 153 RCS temperature, and pressurizer water level. This trend continues until the RCS heat generation rate falls below the secondary-side heat removal capability.

Key parameters for this event are associated with plant system responses, rather than fuel design parameters, and include low SG water level trip setpoint, low SG water level ESFAS signal setpoint, minimum AFW flow rate and delay time, initial core power (decay heat), MSSV and PSV setpoints and capacities and pump heat.

Since there is only a slight increase in RCS temperature and no significant increase in core power prior to reactor trip, there is no significant challenge to the SAFDLs prior to reactor trip.

In addition, as long as AFW is sufficient to remove decay heat and pump heat following reactor trip, there is no significant challenge to the SAFDLs following reactor trip. This decay heat removal capability is not affected by the fuel design change.

Key parameters for the overpressure aspects of this event are identified above. The reactor power level, the safety valves opening setpoints and capacities, the SG low level reactor trip setpoint, and the auxiliary feedwater system configuration, actuation setpoint and flows, remain unchanged from the current analysis values. The decay heat removal capability is not affected.

There are no changes to the instrumentation uncertainties associated with the operating parameters and trip setpoints. Fuel design parameters are not key parameters regarding overpressure, and therefore, have no significant effect on the overpressure aspects of this event.

Thus, the current UFSAR analysis of record for this event remains bounding for the fuel transition.

4.12 FeedwaterSystem Pipe Break (UFSAR 15.2.8)

This event is defined as a break in a feedwater line at the steam generator inlet nozzle, resulting in an uncontrolled discharge of fluid from the steam generator. Depending on the size of the rupture, the event can cause either a cooldown or a heatup of the RCS. A cooldown of the RCS resulting from a secondary system pipe break is bounded by Main Steam Line Break (UFSAR 15.1.5 and 15.1.6). Therefore, only the RCS heatup aspects are considered for the Feedwater System Pipe Break event.

~%A.JI OLEU1L)ICL LJUIUI1 I IMI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 154 The DNB aspects of this event prior to reactor trip are bounded by the AREVA fuel transition analysis for the Main Steam Line Break (UFSAR 15.1.5 and 15.1.6) event with either a FFBT at the time of reactor trip or a LOOP at event initiation. In addition, as long as AFW is sufficient to remove decay heat and pump heat following reactor trip, there is no significant challenge to the SAFDLs following reactor trip.

Key parameters for the overpressure aspects of this event are associated with plant system responses, rather than fuel design parameters, and include low SG water level ESFAS signal setpoint, minimum AFW flow rate and delay time, initial core power (decay heat), MSSV and PSV setpoints and capacities and pump heat. The reactor power level, the safety valves opening setpoints and capacities, the SG low level reactor trip setpoint, and the auxiliary feedwater system configuration, actuation setpoint and flows, remain unchanged from the current analysis values. The decay heat removal capability is not affected. There are no changes to the instrumentation uncertainties associated with the operating parameters and trip setpoints. Fuel design parameters are not key parameters regarding overpressure, and therefore, have no significant effect on the overpressure aspects of this event. The overpressure aspects of this event are bounded by the current UFSAR analysis of record.

4.13 TransientsResulting from the Malfunction of One Steam Generator(UFSAR 15.2.9) 4.13.1 Accident Description Transients resulting from the malfunction of one steam generator are characterized by increased load in one steam generator and decreased load in the other steam generator. The overpressure response of the decreased-load steam generator is limited by relief valves, but the depressurization response of the increased-load steam generator is not correspondingly limited.

This leads to a downward-trending asymmetric core inlet coolant temperature distribution, which when combined with a negative MTC results in an increase in power. The asymmetric core inlet coolant temperature may also lead to augmented radial peaking, thus potentially challenging the DNB and FCM SAFDLs.

The UFSAR identifies four asymmetric events which are initiated by the malfunction of one steam generator: Loss of Load to One Steam Generator, Excess Load to One Steam Generator, Loss of Feedwater to One Steam Generator and Excess Feedwater to One Steam

ý.wUl ILE U11CU L-JUOýoUl i 1U1 IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 155 Generator. The limiting event of those four, a Loss of Load to One Steam Generator (Single MSIV Closure) event, is initiated by the inadvertent closure of a single MSIV. Upon the loss of load to the affected SG, its pressure and temperature rapidly increase, which causes a heatup of its associated RCS loop. The single MSIV closure isolates the affected SG from the turbine header causing the turbine header pressure to decrease. The decreased turbine header pressure then causes the unaffected SG to experience an increase in steam flow to "pick up" the lost load, with its associated pressure and temperature decrease, which causes a cooldown of the unaffected RCS loop. The result is asymmetry in the coolant temperatures entering the core from the affected and unaffected RCS loops. An increase in core power occurs with the potential for augmented radial peaking. However, a reactor trip will occur soon enough such that there is insufficient time for asymmetric core inlet temperatures to produce any significant increase in augmented radial peaking.

4.13.2 Input Parameters and Assumptions Key input parameters and their values used in the analysis of this event were consistent with or conservative relative to the approved Reference 1 methodology, as supplemented in Section 2.7. See Section 2.0 for key input parameter values.

Initial Conditions - Full power, maximum core inlet temperature and TS minimum RCS flow, in order to ensure that initial conditions most challenging to the SAFDLs were used. Cases that would assess challenge to the RCS overpressure limit were not evaluated in support of the fuel transition, because such challenge is not affected by the fuel transition.

Reactivity Feedback - The reactivity feedback coefficients were biased according to the approved methodology. EOC kinetics were analyzed. The MTC was assumed to be equal to the most negative limit for the EOC cases. Doppler reactivity was biased to bound EOC.

Scram worth was conservatively set to a minimum value appropriate for the initial power level being analyzed.

Reactor Protection System Trips and Delays - The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, control rod insertion is delayed to account for CEA holding coil delay time.

Pressurizer Pressure Control - Pressurizer pressure control (i.e., pressurizer sprays, heaters, and PORVs) parameters and equipment states were selected to reduce the

V-9VIRIU1LIIIU L.JLJdlUirEI" H AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 156 primary system pressure, which provided a conservative calculation of the MDNBR during the transient.

Gap Conductance - Gap conductance was set to conservative values consistent with the time-in-cycle for the reactivity coefficients and account for the effect of TCD.

Steam Generator Tube Plugging - Both 0% and 20% SGTP level were analyzed for this event. Results are presented for the limiting case.

4.13.3 Acceptance Criteria The principally challenged acceptance criteria for this event are:

Fuel cladding integrity should be maintained by ensuring that the SAFDLs are not exceeded.

This criterion is met by assuring that (1) the minimum calculated DNBR is not less than the 95/95 DNB correlation limit and (2) the peak LHR is less than the LHR limit corresponding to the FCM temperature.

The pressure in the reactor coolant system should be less than 110% of the design value. The Loss of Load to One SG event can challenge the primary and secondary overpressure criteria; however, the over-pressurization aspect of this event is bounded by the Loss of Load to Both SGs Event, in which the heat removal from both SGs is simultaneously lost.

4.13.4 Method of Analysis The analyses were performed with the approved non-LOCA methodology given in Reference 1, as supplemented in Section 2.7. For this event, the S-RELAP5 code was used to model the key system components and calculate neutron power, fuel thermal response, surface heat transport, fluid conditions (such as coolant flow rates, temperatures, and pressures), and an estimated time of MDNBR. The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code (Reference 3), which was used to calculate the MDNBR using the HTP CHF correlation (Reference 5).

Cases with 0% and 20% SGTP were analyzed.

4.13.5 Results Calculations were performed to evaluate the challenge to the SAFDLs for this event.

.AII LJII:U LJUt-ol 0IM-,[IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 157 The limiting case was the case that assumed 0% steam generator tube plugging. The limiting case sequence of events is given in Table 4.15 and the results are given in Table 4.16. The MDNBR was calculated to be above the 95/95 limit for the HTP CHF correlation and the peak LHR was calculated to be less than the FCM limit. The transient response is shown in Figure 4.78 through Figure 4.84. Figure 4.78 shows the reactor power as a function of time. Figure 4.79 shows the core power based on rod surface heat flux. Figure 4.80 shows pressurizer pressure. Figure 4.81 shows pressurizer liquid level. Figure 4.82 shows the RCS loop temperatures. Figure 4.83 shows the total RCS flow rate. Figure 4.84 shows the steam generator pressures.

'~AJlIuLJHleu LJU~sUIIl~IIL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summarv Report pn P 1 R Table 4.15 Loss of Load to One Steam Generator: Sequence of Events (0% SGTP)

Event Time (sec)

MSIV-1 begins to close 0.00 MSIV-1 fully closed 0.10 MSSV flow from affected SG began (970 psia) 0.20 Peak average rod surface heat flux reached (100.6% RTP) 3.00 ASGPT setpoint reached (247 psid) 7.66 ASGPT signal received (following 1.15 second delay) 8.81 Maximum kinetics power reached (101.6% RTP) 9.50 CEAs insertion began 9.55 MDNBR 9.70 CEAs were fully inserted (2.85 second insertion time) 12.40 Table 4.16 Loss of Load to One Steam Generator: Limiting Results Criteria Result Limit MDNBR 1.713 1.164 Peak LHR 15.74 kW/ft [ ]kW/ft

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%_-UIILI LJIIC;U LJUtoUI I Iý-I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 166 4.14 PartialLoss of ForcedReactor Coolant Flow (UFSAR 15.3.1)

This event is caused by loss of electrical power to one, two or three of the RCPs. The core and system performance following a partial loss of forced reactor coolant flow is less limiting than that of a Complete Loss of Forced Reactor Coolant Flow event (UFSAR 15.3.2). Therefore, the consequences of this event are bounded by the Complete Loss of Forced Reactor Coolant Flow event.

4.15 Complete Loss of ForcedReactor Coolant Flow (UFSAR 15.3.2) 4.15.1 Accident Description A complete loss of coolant flow event may result from a simultaneous loss of electrical power to all of the four RCPs.

A decrease in reactor coolant flow while a plant is at power results in degraded core heat transfer, reduction in DNB margin and a challenge to the DNB SAFDL. The primary concern with this event is the challenge to the DNB SAFDL. The fuel centerline melt SAFDL is not challenged, since there is no significant increase in core power for this event. The reduction in primary system flow and the associated increase in core coolant temperatures, result in a reduction in DNBR margin. The increasing primary system coolant temperatures also result in expansion of the primary coolant volume, causing an in-surge into the pressurizer and an increase in the pressure of the primary system. However, the pressure increase is small and the pressure limits are not challenged by this event. For St. Lucie Unit 2, this event is analyzed to verify the RPS low flow trip setpoint in combination with the TS / COLR Limiting Conditions for Operation (LCOs), namely the initial core power level, the maximum initial core inlet temperature, the minimum initial RCS flow rate, and the axial shape index limits.

4.15.2 Input Parameters and Assumptions The key input parameters and their values used in the analysis of this event were consistent with or conservative relative to the approved Reference 1 methodology. See Section 2.0 for key input parameter values.

Initial operating conditions (maximum power, TS maximum core inlet temperature, nominal RCS pressure, TS minimum RCS flow rate)

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%-#U1 II IICU L-JU~.sUI I IMI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 167 Reactivity Feedback - Since this event involves an increase in the core coolant temperature, the event was assumed to occur at BOC with a maximum TS MTC.

However, this event occurs quickly and is generally not sensitive to neutronics parameters. A minimum HFP scram worth was used to conservatively maintain relatively high core power during the degradation in flow.

Reactor Protection System Trips and Delays - The event is primarily protected by the low flow RPS trip. The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, control rod insertion is delayed to account for the CEA holding coil delay time.

Pressurizer Pressure Control - Pressurizer pressure control (i.e., pressurizer sprays, heaters, and PORVs) parameters and equipment states were selected to reduce the primary system pressure, which provided a conservative calculation of the MDNBR during the transient.

Gap Conductance - Gap conductance was set to a BOC value accounting for TCD to delay the transfer of heat from the fuel rod to the coolant allowing the primary system flow to decay further thus leading to a conservative prediction of DNBR.

RCS Flow - A conservative pump coastdown model is used for this event.

SGTP - The maximum of 20% plugging in each SG was considered.

4.15.3 Acceptance Criteria For the loss-of-coolant flow event, the principally challenged acceptance criterion is:

Fuel cladding integrity should be maintained by ensuring that the SAFDLs are not exceeded.

This criterion is met by assuring that the minimum calculated DNBR is not less than the 95/95 DNB correlation limit. Since this event does not involve a significant power transient or augmented peaking, the fuel centerline melt limit is not challenged.

4.15.4 Method of Analysis For the fuel transition, detailed analyses were performed with approved non-LOCA methodology (Reference 1). For this event, the S-RELAP5 code was used to model the key system components and calculate neutron power, fuel thermal response, surface heat transport, and fluid conditions (such as coolant flow rates, temperatures, and pressures), and an estimated time of MDNBR. The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code, which was used to calculate the MDNBR using the HTP

%..JUL1 JIIU1l LJU1JWU11i;IIL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 168 CHF correlation. The loss of coolant flow event was also addressed as part of the DNB LCO statistical setpoint analyses using the Reference 4 methodology.

A single case was analyzed at BOC HFP initial conditions, maximum TS core inlet temperature and minimum TS RCS flow rate. The factors affecting scram time were biased using conservative trip signal delay and holding coil delay times to produce the most significant challenge to the DNB limit.

4.15.5 Results The sequence of events is given in Table 4.17. The result is given in Table 4.18. State points for the DNB analyses were chosen at or near the time of the most adverse combination of power and flow. The MDNBR with statistically applied uncertainties was calculated to be greater than the 95/95 DNB correlation limit. Statistical evaluation of this event is also performed as part of the DNB LCO analyses.

Plots of key system parameters are shown in Figure 4.85 through Figure 4.90. Figure 4.85 shows the reactor power as a function of time. Figure 4.86 shows the core power based on rod surface heat flux. Figure 4.87 shows pressurizer pressure. Figure 4.88 shows RCS loop temperatures. Figure 4.89 shows the total RCS flow rate. Figure 4.90 shows the reactivity feedback.

%.fUl ILI UII%=:U L-JU%.oUI I It-51 I L AREVA Inc. ANP-3347NP Revision 0 I Trmneitinn r'hmrtmr 1J;lIUJKI,,n-lL C'(A (uimmnP,= Pr~rfr Do m 1RQ qZt tia "nit ') P11DI sitV ii ILI.ft J ~lIA~. ~ l*CL*II LA Q LIýf b! ILM Table 4.17 Loss of Forced Coolant Flow: Sequence of Events Event Time (sec)

Pump coastdown initiates 0.0 Low RCS flow trip setpoint reached 0.87 Reactor scram on low RCS flow rate (including trip response delay) 1.78 CEA insertion begins 2.52 Peak core power 2.52 Peak pressurizer pressure 3.09 MDNBR 4.05 CEAs fully inserted 5.37 Table 4.18 Loss of Forced Coolant Flow: Result Criterion Result Limit MDNBR 1.227 1.164

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Figure 4.88 Loss of Forced Coolant Flow: RCS Loop Temperatures

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k's'l ILU UIJIIZU LJULý,Ul I1tI It AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 176 4.16 Reactor Coolant Pump Shaft Seizure (UFSAR 15.3.3) 4.16.1 Accident Description The RCP locked rotor event bounds the consequences of a shaft break event, and is postulated as the instantaneous seizure of a (single) RCP rotor. Flow through the faulted RCS loop rapidly decreases, causing a reactor trip on a low total RCS flow signal and a turbine trip on the reactor trip.

FFBT is assumed to occur at the time of reactor trip breaker opening. The FFBT at the time of reactor trip breaker opening results in the coastdown of two of the three non-faulted reactor coolant pumps. The remaining non-faulted RCP is assumed to coast-down at 3.0 seconds following reactor trip breaker opening due to a LOOP.

Following the reactor trip, heat stored in the fuel rods continues to be transferred to the reactor coolant. The combination of the relatively high fuel rod surface heat fluxes, decreasing core flow, and increasing core coolant temperatures challenges the DNBR safety limit and may result in fuel failure.

At the same time, the SG primary-to-secondary heat transfer rate decreases, because (1) the decreasing primary coolant flow degrades the SG tube primary side heat transfer coefficients and (2) the turbine trip causes the secondary-side temperature to increase. Decreasing rate of heat removal in the SGs, combined with decreasing flow of coolant removing heat from the reactor core, cause the reactor coolant to heat up. The resultant reactor coolant expansion causes fluid to surge into the pressurizer and an increase in RCS pressure, with the potential for lifting safety valves. The event may challenge the RCS overpressure criterion.

Since the systems designed to mitigate this event (namely, the RPS) are redundant, there is no single active failure that will adversely affect the consequences of the event.

4.16.2 Input Parameters and Assumptions The key input parameters and their values used in the analysis of this event were consistent with or conservative relative to the approved Reference 1 methodology. See Section 2.0 for key input parameter values.

'.# IIl UIIUU LJUI.Lo~I I It:-I It AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 177 Initial operating conditions (maximum power, TS maximum core inlet temperature, nominal RCS pressure, TS minimum RCS flow rate)

Reactivity Feedback - Since this event involves an increase in the core coolant temperature, the event was assumed to occur at BOC with a maximum TS/COLR MTC at or below 70% RTP. However, this event occurs quickly and is generally not sensitive to neutronics parameters. A minimum HFP scram worth assuming the most reactive rod is stuck out of the core was used to conservatively maintain relatively high core power during the degradation in flow.

Reactor Protection System Trips and Delays - The event is primarily protected by the low flow RPS trip. The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, control rod insertion is delayed to account for the CEA holding coil delay time.

Pressurizer Pressure Control - Pressurizer pressure control (i.e., pressurizer sprays, heaters, and PORVs) parameters and equipment states were selected to reduce the primary system pressure, which provided a conservative calculation of the MDNBR during the transient.

Gap Conductance - Gap conductance was set to a BOC value accounting for TCD to delay the transfer of heat from the fuel rod to the coolant allowing the primary system flow to decay further thus leading to a conservative prediction of DNBR.

Steam Generator Tube Plugging - The maximum of 20% symmetric plugging in each SG was considered.

4.16.3 Acceptance Criteria The principally challenged acceptance criterion for this event is with respect to radiological consequences. The analysis documented in this section does not address radiological consequences directly; rather, the extent of fuel failure is determined which is an input to the radiological analyses.

The locked rotor event does not represent a significant challenge to fuel centerline melting because there is no large power increase and no significant adverse power redistribution within the core. Also, the increase in RCS pressure for this event is much less severe than that resulting from the Loss of Condenser Vacuum event.

4.16.4 Method of Analysis For the fuel transition, detailed analyses were performed with the approved Reference 1 non-LOCA methodology. For this event, the S-RELAP5 code was used to model the key system

1-fUI ILI V01I;U LJUIoUlIII;I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 178 components and calculate neutron power, fuel thermal response, surface heat transport, and fluid conditions (such as coolant flow rates, temperatures, and pressures), and an estimated time of MDNBR. The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code, which was used to calculate the MDNBR using the HTP CHF correlation.

A single case was analyzed for DNBR at BOC HFP initial conditions, maximum TS core inlet temperature and minimum TS RCS flow rate.

The factors affecting scram time were biased using conservative trip signal delay and holding coil times to produce the most significant challenge to the DNB limit.

4.16.5 Results The sequence of events is given in Table 4.19. The result is given in Table 4.20. The MDNBR was calculated to be greater than the 95/95 limit for the HTP DNB correlation.

Plots of key system parameters are shown in Figure 4.91 through Figure 4.97. Figure 4.91 shows the reactor power as a function of time. Figure 4.92 shows the core power based on rod surface heat flux. Figure 4.93 shows pressurizer pressure. Figure 4.94 shows RCS loop temperatures. Figure 4.95 shows the total RCS flow rate. Figure 4.96 shows the reactivity feedback. Figure 4.97 shows the loop flow rates.

'%.OUI LIUIIIZU LJIJ-sUIII M1 L AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Pag 19 Table 4.19 Reactor Coolant Pump Rotor Seizure: Sequence of Events Time Event (sec)

Event initiation (seizure of RCP 1A) 0.0 Low RCS flow trip setpoint reached 0.12 Reactor scram on low RCS flow rate (including trip 1.02 response delay)

CEA insertion begins 1.76 Peak core power 1.76 Peak pressurizer pressure 2.04 MDNBR 3.08 CEAs fully inserted 4.61 Table 4.20 Reactor Coolant Pump Rotor Seizure: Result Criterion Result Limit MDNBR (% fuel failure) 1.205 (0%) 1.164

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Figure 4.91 Reactor Coolant Pump Rotor Seizure: Reactor Power

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Figure 4.92 Reactor Coolant Pump Rotor Seizure: Total Core Heat Flux Power

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Figure 4.93 Reactor Coolant Pump Rotor Seizure: Pressurizer Pressure

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Figure 4.94 Reactor Coolant Pump Rotor Seizure: RCS Loop Temperatures

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Figure 4.95 Reactor Coolant Pump Rotor Seizure: RCS Total Loop Flow Rate

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Figure 4.97 Reactor Coolant Pump Rotor Seizure: RCS Loop Flow Rates Note: Loop 1A is the coolant loop with the seized rotor.

k..AJI ILl IJII;U LJLJoU..,UI l L AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 187 4.17 Reactor Coolant Pump Shaft Break (UFSAR 15.3.4)

The consequences of this event are similar to the Reactor Coolant Pump Shaft Seizure (UFSAR 15.3.3) event. With a broken shaft, the impeller is free to spin as opposed to being fixed in position during the Reactor Coolant Pump Shaft Seizure event. Therefore, the initial rate of reduction in core flow is greater during a Reactor Coolant Pump Shaft Seizure event than in a Reactor Coolant Pump Shaft Break event because the fixed shaft causes greater resistance than a free-spinning impeller early in the transient, when flow through the affected loop is in the positive direction. As the transient continues, the flow direction through the affected loop is reversed. If the impeller is able to spin free, the flow to the core will be less than that available with a fixed-shaft during periods of reverse flow in the affected loop. Because peak pressure and DNB occur very early in the transient, the reduction in core flow during the period of forward flow in the affected loop dominates the severity of the results. Therefore, the consequences of a Reactor Coolant Pump Shaft Break event are bounded by the consequences of the Reactor Coolant Pump Shaft Seizure event.

4.18 Uncontrolled CEA Bank Withdrawal from a Subcriticalor Low Power Startup Condition (UFSAR 15.4.1) 4.18.1 Accident Description This event is initiated by a continuous CEA withdrawal that could result from a malfunction in the reactor regulating system or control element drive system. The event is initiated from a Mode 2 startup (critical) condition at zero power. The event is characterized by a large and rapid positive reactivity insertion that can challenge the DNB and FCM SAFDLs. Reactor trip occurs on a VHP trip signal; however, the power excursion is mitigated by Doppler reactivity feedback prior to reactor trip.

4.18.2 Input Parameters and Assumptions Key input parameters and their values used in the analysis of this event were consistent with or conservative relative to the approved Reference 1 methodology. See Section 2.0 for key input parameter values.

Initial Conditions - HZP initial conditions, maximum HZP core inlet temperature and minimum TS RCS flow rate for four-pump operation were assumed in order to maximize the challenge to DNB.

'.JIIl LJIIQZ:A LJUkoUI I IMI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 188 Reactivity Feedback - Since this event involves a rapid increase in core power, the Doppler reactivity feedback is a key parameter. BOC Doppler reactivity feedback was assumed to minimize the negative reactivity resulting from the heatup of the fuel during the event. Since the event involves a rapid increase in core power, the effect of moderator feedback is minimal. A maximum TS MTC at HZP was assumed. Scram worth was conservatively set to the minimum TS shutdown margin.

Reactor Protection System Trips and Delays - The event is primarily protected by the VHP trip (low setting). The reactor protection system trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, a maximum CEA holding coil delay was assumed. For conservatism, the High Rate-of-Change trip was not credited in the analysis except to the extent that it was used as justification for not initiating this event from subcritical modes of operation.

Pressurizer Pressure Control - Pressurizer pressure control, i.e., pressurizer sprays, heaters and PORVs, parameters and equipment states were selected to reduce the primary system pressure, which provided a conservative calculation of the MDNBR during the transient.

CEA Withdrawal Characteristics - To maximize the reactivity insertion from the CEA withdrawal, a bounding differential worth was assumed together with a maximum CEA withdrawal speed. The reactivity insertion rate used for this event bounds the reactivity insertion rate corresponding to a Boron Dilution event.

Gap Conductance - Gap conductance was set to a conservative BOC value, consistent with the time-in-cycle for the reactivity coefficients, to maximize the heat flux through the cladding and minimize the negative reactivity inserted due to Doppler feedback.

4.18.3 Acceptance Criteria This event is classified as an AOO event and, consistent with the current licensing basis, the principally challenged acceptance criterion is:

Fuel cladding integrity should be maintained by ensuring that the SAFDLs are not exceeded.

This criterion is met by assuring that

The minimum calculated DNBR is not less than the 95/95 DNB correlation limit, and,

Fuel centerline melt is precluded in the most adverse location in the core.

4.18.4 Method of Analysis The analysis was performed with the approved non-LOCA methodology given in Reference 1.

For this event, the S-RELAP5 code was used to model the key system components and

'4..LJI ILI U~I1ZU LJULOUI I It:-I It AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 189 calculate neutron power, fuel thermal response, surface heat transport, fluid conditions (such as coolant flow rates, temperatures, and pressures), maximum fuel centerline temperature, and an estimated time of MDNBR. The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code (Reference 3), which was used to calculate the MDNBR using the HTP CHF correlation (Reference 5).

4.18.5 Results The sequence of events is given in Table 4.21.

The transient response is shown in Figure 4.98 through Figure 4.103. Figure 4.98 shows the total reactivity feedback and its components. Figure 4.99 shows the reactor power and the core power based on rod surface heat flux, as functions of time. Figure 4.100 shows the hot spot centerline fuel temperature. Figure 4.101 shows reactor coolant loop temperatures. Figure 4.102 shows pressurizer pressure. Figure 4.103 shows the cold leg flow rates.

The results with respect to acceptance criteria are shown in Table 4.22. The MDNBR was calculated to be above the 95/95 limit for the HTP CHF correlation. The peak fuel centerline temperature was calculated to be less than the fuel centerline melt temperature.

AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summary Report Pace 190 Table 4.21 Uncontrolled CEA Withdrawal from Subcritical: Sequence of Events Event Time (sec)

Bank withdrawal began 0.00 Indicated core power reached VHP trip setpoint 36.46 Reactor trip signal generated 37.57 CEAs began insertion 38.32 Maximum actual power reached 38.34 Maximum heat flux power occurred 39.64 MDNBR occurred 39.64 Maximum hot spot fuel centerline temperature occurred 41.00 Table 4.22 Uncontrolled CEA Withdrawal from Subcritical: Results Criterion Result Limit MDNBR 1.994 1.164 Peak CLT 3194°F [ ]F

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Figure 4.98 Uncontrolled CEA Withdrawal from Subcritical: Reactivity Feedback

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Figure 4.99 Uncontrolled CEA Withdrawal from Subcritical: Power and Heat Flux

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%--,IILI 'JIIUA IJUkwUl I IUI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 197 4.19 Uncontrolled CEA Bank Withdrawal at Power (UFSAR 15.4.2) 4.19.1 Accident Description An inadvertent CEA bank withdrawal at power could be caused by two potential initiators: (1) operator error, or (2) a malfunction of either the CEAs or of the Control Element Drive Mechanism (CEDM) which results in an uncontrolled, continuous CEA bank withdrawal. The positive reactivity addition from the CEA withdrawal results in a power transient. Due to relatively constant heat extraction from the steam generators during the event, the increase in reactor power produces an increase in reactor coolant temperatures and core heat flux, thereby decreasing the margin to the DNB and FCM SAFDLs, and the RCS overpressure limit.

While a continuous CEA withdrawal is considered unlikely, the reactor protection system is designed to terminate any such transient before fuel thermal design and RCS overpressure limits are reached. Protection against violation of the SAFDLs and RCS overpressure limit is provided primarily by the VHP, TM/LP, local power density, and HPP trips.

4.19.2 Input Parameters and Assumptions Key input parameters and their values used in the analysis of this event were consistent with or conservative relative to the approved Reference 1 methodology, as supplemented in Section 2.7. See Section 2.0 for key input parameter values.

Initial Conditions - Full power, 90% RTP, 65% RTP, and 20% RTP levels were evaluated, with a maximum core inlet temperature and TS minimum RCS flow, in order to ensure that initial conditions most challenging to the SAFDLs were used. Cases that would assess challenge to the RCS overpressure limit were not evaluated in support of the fuel transition, because such challenge is not affected by the fuel transition.

Reactivity Feedback - The reactivity feedback coefficients were biased according to the approved methodology. Both BOC and EOC kinetics were analyzed, to assess the impact of moderator and Doppler feedback. The MTCs were assumed to be either a value that bounds the most positive TS limits for the BOC cases or a value that bounds the most negative TS limit for the EOC cases. Doppler reactivity was biased to bound a range of feedback from BOC to EOC.

Scram worth was conservatively set to a minimum value appropriate for the initial power level being analyzed.

%..A..JI ILI VINIIU LjutoslJ I MNI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 198 Reactor Protection System Trips and Delays - The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, control rod insertion is delayed to account for CEA holding coil delay time.

Pressurizer Pressure Control - Pressurizer pressure control (i.e., pressurizer sprays, heaters, and PORVs) parameters and equipment states were selected to reduce the primary system pressure, which provided a conservative calculation of the MDNBR during the transient.

CEA Withdrawal Characteristics - A bounding range of differential bank worths was analyzed together with a bounding range of CEA withdrawal speeds. As a conservative analytical assumption, withdrawal of the CEAs is not terminated at reactor trip.

Decalibration of the NI power signal due to CEA shadowing as the CEAs are withdrawn was conservatively accounted for.

Gap Conductance - Gap conductance was set to conservative values that are consistent with the time-in-cycle for the reactivity coefficients and that account for the effect of TCD.

4.19.3 Acceptance Criteria The principally challenged acceptance criteria for this event are:

Fuel cladding integrity should be maintained by ensuring that the SAFDLs are not exceeded.

This criterion is met by assuring that (1) the minimum calculated DNBR is not less than the 95/95 DNB correlation limit and (2) the peak LHR is less than the LHR limit corresponding to the FCM temperature.

The pressures in the reactor coolant and main steam systems should be less than 110%

of the design values.

Challenge to this criterion, which is met by assuring that the peak RCS pressure is less than the acceptance criterion of 2750 psia (i.e., 110% of the design pressure) and the peak main steam system pressure is less than the acceptance criterion of 1100 psia (i.e., 110% of the design pressure), was not evaluated in support of the fuel transition.

Key parameters for the RCS and secondary side overpressure cases are associated with plant system responses, rather than fuel design parameters, and include the MSSV and PSV setpoints and capacities. Fuel design parameters are not key parameters regarding overpressure, and therefore, have no significant effect on the overpressure aspects of this event. The power level, operating parameters, reactor trip setpoints, the safety valves (primary and secondary sides) opening setpoints and tolerance limits and the CEA drop time requirements are unchanged from the current analysis values. There are no changes to the instrumentation uncertainties associated with the operating parameters and trip setpoints. Thus, the current UFSAR analysis of record for overpressure remains bounding for the fuel transition.

%_0,UI ILI UIIZ;U LJUkokoll I IUI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 199 4.19.4 Method of Analysis The analyses were performed with the approved non-LOCA methodology given in Reference 1, as supplemented in Section 2.7. For this event, the S-RELAP5 code was used to model the key system components and calculate neutron power, fuel thermal response, surface heat transport, fluid conditions (such as coolant flow rates, temperatures, and pressures), and an estimated time of MDNBR. The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code (Reference 3), which was used to calculate the MDNBR using the HTP CHF correlation (Reference 5).

A spectrum of positive reactivity insertion rates is possible, from very slow to fast, limited only by bank worth and maximum drive speed. Two reactivity feedback matrices of cases are analyzed:

One for most-positive reactivity feedback (most-positive MTC and least-negative Doppler coefficient), and the other for most-negative feedback (most-negative MTC and most-negative Doppler coefficient). A range of initial reactor power levels was analyzed.

For both reactivity feedback matrices of cases, the reactivity insertion rates used bound the respective lowest MDNBR point and the maximum value for CEA bank withdrawal:

Range of Reactivity Insertion Rates (pcm/sec)

Case Minimum Maximum BOC, HFP 0.001 14 EOC, HFP 2.4 20 BOC, 90% RTP 0.001 16 EOC, 90% RTP 2.4 24 BOC, 65% RTP 0.001 26 EOC, 65% RTP 2.3 32 BOC, 20% RTP 0.011 38 EOC, 20% RTP 2.9 50 The lower bound of the reactivity insertion rates also bounds the reactivity insertion rate corresponding to a Mode 1 Boron Dilution event.

%.OUIILI U11t.;:U L-/UkUl I 1UI it AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 200 4.19.5 Results Calculations were performed to evaluate the challenge to the SAFDLs for this event. Part-power levels were analyzed, as well as full power conditions. Both BOC and EOC kinetics were included in the analyses. The DNB and peak LHR results are given in Table 4.24. The DNB SAFDL is most challenged at BOC HFP initial conditions. The peak LHR does not significantly challenge the LHR limit for this event.

BOC (Full Power, 0.08 pcm/sec)

The sequence of events is given in Table 4.23. Results are given in Table 4.24. The MDNBR was calculated to be above the 95/95 limit for the HTP CHF correlation. The peak LHR was calculated to be less than the LHR limit corresponding to fuel centerline melt.

The transient response is shown in Figure 4.104 through Figure 4.110. Figure 4.104 shows the reactor power as a function of time. Figure 4.105 shows the core power based on rod surface heat flux. Figure 4.106 through Figure 4.110 show the pressurizer pressure, pressurizer liquid level, RCS loop temperatures, total RCS flow rate, and reactivity feedback, respectively.

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%..IJI AREVA Inc. ANP-3347NP Revision 0 St Lucie Unit 2 Fuel Transition Chanter 15 Non-LOCA Summarv Renort Paoe 201 Table 4.23 Uncontrolled CEA Withdrawal at Power: Sequence of Events Case Event Time (sec)

BOC full power Bank withdrawal begins 0.00 (0.08 pcmlsec) Pressurizer pressure reaches HPP trip setpoint, 166.62 pressurizer PORVs open, and pressurizer pressure peaks NI power reaches VHP trip setpoint 166.92 Reactor scram on VHP (including signal processing 167.32 delay)

Core heat flux through cladding peaks (117.8% RTP) 167.71 MDNBR 167.85 CEA insertion begins, and core power peaks (118.50 168.07 RTP)

Table 4.24 Uncontrolled CEA Withdrawal at Power: Results Initial BOC EOC Power Level MDNBR Peak LHR MDNBR Peak LHR Full power 1.177 Bounded'1 ) 1.184 Bounded(1 )

90% RTP 1.206 16.43 kW/ft 1.241 Bounded(')

65% RTP 1.484 Bounded(1 ) 1.435 Bounded(')

20% RTP 2.735 Bounded(1 ) 2.744 Bounded(')

Limit 1.164 [ ] kW/ft 1.164 [ ] kW/ft Bounded by value for BOC 90% RTP

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Figure 4.104 Uncontrolled CEA Withdrawal at Power: Reactor Power (BOC Full Power 0.08 pcm/sec)

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Figure 4.105 Uncontrolled CEA Withdrawal at Power: Total Core Heat Flux Power (BOC Full Power 0.08 pcm/sec)

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Figure 4.106 Uncontrolled CEA Withdrawal at Power: Pressurizer Pressure (BOC Full Power 0.08 pcm/sec)

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Figure 4.108 Uncontrolled CEA Withdrawal at Power: RCS Loop Temperatures (BOC Full Power 0.08 pcm/sec)

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Figure 4.109 Uncontrolled CEA Withdrawal at Power: Total RCS Loop Flow Rate (BOC Full Power 0.08 pcm/sec)

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%-flILIUIJICU LI-JLUi-U I IMI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 209 4.20 CEA Misoperation(UFSAR 15.4.3) 4.20.1 Accident Description The CEA drop event is defined as the inadvertent release of a full-length CEA or full-length CEA subgroup, causing it to drop into the core. A dropped CEAICEA subgroup will be detected by either a position limit switch on each CEDM or by a reduction in power as measured by the ex-core detectors.

The negative reactivity insertion when the CEAICEA subgroup drops into the core causes a reduction in the core power and reactor coolant temperature and pressure. The magnitude of the decrease in core power and RCS temperature and pressure depends on the worth of the dropped CEAICEA subgroup. At EOC conditions, a strongly negative MTC will produce a positive reactivity insertion that will return the reactor to the full-power condition (with the assumption of a constant HFP turbine load demand) with augmented radial power peaking corresponding to the new radial power distribution caused by the dropped CEAICEA subgroup. Increased cladding heat fluxes and fuel temperatures in the hot assembly result in a challenge to the DNB and FCM SAFDLs. The event analysis accounts for the changes in power distribution by applying radial peaking augmentation factors.

Protection against exceeding the SAFDLs is provided by the combination of the initial steady-state margin to DNB, defined by maintaining the Axial Shape Index (ASI) and power within the DNB LCO band, the VHP trip and the TM/LP trip. The event may be terminated by a reactor trip, or there may be no reactor trip and the plant returns to the original power level. Other than the RPS, no other automatic functions were credited that would mitigate this event. No single active failure will adversely affect the consequences of this event.

4.20.2 Input Parameters and Assumptions The key input parameters and their values used in the analysis of this event are consistent with or conservative relative to the Reference 1 approved methodology. See Section 2.0 for key input parameter values.

Initial Conditions - This event was assumed to initiate from EOC HFP conditions with a maximum core inlet temperature and TS minimum RCS flow. This set of conditions minimizes the initial margin to DNB.

~AIIl UlINU LJUtoUI I 1Il IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 210 Reactivity Feedback - The reactivity feedback coefficients were biased according to the approved methodology. The MTC was set to the most negative limit to produce the most positive moderator reactivity feedback as the RCS cools down due to a dropped CEA.

Dropped CEA Worth - A bounding range of dropped CEA worth (from 25 pcm to 1000 pcm) was analyzed to allow determination of the most limiting combination of augmented radial peaking and core boundary conditions.

Load demand - A constant turbine load demand at HFP was assumed.

Trip setpoints - Conservative trip setpoints and delay times were used.

4.20.3 Acceptance Criteria The principally challenged acceptance criterion for this event is that the fuel cladding integrity should be maintained by ensuring that the SAFDLs are not exceeded.

This criterion is met by assuring that the minimum calculated DNBR is not less than the 95/95 DNB correlation limit. Additionally, FCM is demonstrated to be precluded in the most adverse location in the core.

4.20.4 Method of Analysis Detailed analyses were performed with approved non-LOCA methodology given in Reference 1.

For this event, the S-RELAP5 code was used to model the key system response and calculate neutron power, fuel thermal response, surface heat transport, and fluid conditions (such as coolant flow rates, temperatures, and pressures). The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code, which was used to calculate the MDNBR using the HTP CHF correlation. Evaluation of the dropped CEA event was also performed as part of the DNB LCO setpoint verification analyses.

Calculations were performed at EOC HFP conditions, maximum TS core inlet temperature, and minimum TS RCS flow rate. This produces the minimum margin to the DNB limit. The event was analyzed with the most negative MTC limit, which results in the most positive moderator reactivity feedback as the RCS cools down due to the dropped CEA.

A range of dropped CEA worths from 25 to 1000 pcm was analyzed to allow determination of the most limiting combination of augmented radial peaking and core boundary conditions.

%..jlJILI U11OU I-JU~JUI I IUI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 211 4.20.5 Results Detailed results are given for the limiting MDNBR case (25 pcm case).

The sequence of events for the 25 pcm case is given in Table 4.25. For the 25 pcm case, the indicated thermal power does not change significantly and the VHP trip setpoint is not reached.

Thus, no reactor trip occurs. Without the occurrence of a reactor trip, the power asymptotically reaches a new steady value near the initial power level. The maximum return-to-power was calculated to be 3029.8 MWt and is reported at the end of the calculation, when there is no significant increase in core power and no significant change in any of the parameters that are input to the DNBR calculation. The limiting MDNBR was calculated to be 1.554, which is above the 95/95 limit of 1.164 for the HTP CHF correlation. The peak LHR forthe 1000 pcm Case was calculated to be 15.71 kW/ft, which bounds the other cases and is less than the FCM limit of

[ ] kW/ft. Statistical evaluation of this event was performed as part of the DNB LCO setpoint verification analysis.

The transient response for the 25 pcm case is shown in Figure 4.111 through Figure 4.117.

The MDNBR and peak LHR results for all cases are given in Table 4.26.

%-oUlILI VIIC;U LJULoUl I IUI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Pmni 91'9 Table 4.25 CEA Drop: Sequence of Events (25 pcm)

Time (sec) Event Value 0.0 Rod drop initiated ---

3.0 Minimum actual power 2940.2 MWt 300.0 Maximum return-to-power 3029.8 MWt Table 4.26 CEA Drop: Results Criterion 25 pcm 200 pcm 500 pcm 1000 pcm Limit MDNBR 1.554 1.578 1.630 1.702 1.164 Peak LHR Bounded by Bounded by Bounded by 15.71 kW/ft [ ] kW/ft 1000 pcm Case 1000 pcm 1000 pcm Case Case

%.-fUIILI U61t.;U LJUk-oUl I 11Z-,I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paanp 213 Pane 213 110 105

£ I-100 -- ----------- .--...........- ............................ ... - - - .--

-e Core Power E Decalibrated NI-Power .......

95

---

Thermal Power 90 0 50 100 150 200 250 300 Time (s)

Figure 4.111 CEA Drop: Reactor Power (25 pcm)

%.AJUI ILI UlJIIU L-JU.,FUI I ICIIL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paae 214 0.10 0.05 0.00 a(U C

-0.05

-0.10 0 50 100 150 200 250 300 Time (s)

Figure 4.112 CEA Drop: Reactivity Feedback (25 pcm)

\jLJIl ILI l lJIýu LJUULJIl I I;I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary ReDort Paae 215 620 610 600 590 2 580 --- Avg. Thot

.......... Avg. Tcold

a. 570 E

560 550 .................................................... .. U .. ... ..... ........................................... ......... .... ....... .............. ..... . ....... .

540 530 0 50 100 150 200 250 300 Time (s)

Figure 4.113 CEA Drop: RCS Loop Temperatures (25 pcm)

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. U U H Ll V ýj 50000 45000 cc (A 40000 p p 40 c-

'A 35000 30000 0

Time (s)

Figure 4.114 CEA Drop: RCS Total Loop Flow Rate (25 pcm)

'%..PUI ILI LJIIUL LJUtLUI I I~ICl It AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summary ReDort Paae 217 2300 2250 2200 Au 2150 a.

2100 2050 2000 0 50 100 150 200 250 300 Time (s)

Figure 4.115 CEA Drop: Pressurizer Pressure (25 pcm)

-,FUI ILI U1lOU L-J~JwjlJ I IUI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chanter 15 Non-LOCA Summarv RePnort Paae 218 70 65 I

CU p

CL

-660 (D

-j 55

. . . . . . . . . . ..... ........I........L........

50 0 50 100 150 200 250 300 Time (s)

Figure 4.116 CEA Drop: Pressurizer Liquid Level (25 pcm)

%..#U ILI VIIIUU L-JUkoUI I I~I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Charter 15 Non-LOCA Summary ReDort P~rp 91Q Pane 219 920 910 I

900 O89 0)

&. 880 ci,

-- 0 SG-1 870 .......... mSG -2 860 850 0 50 100 150 200 250 300 Time (s)

Figure 4.117 CEA Drop: Steam Generator Pressures (25 pcm)

IfU ~IieU LJUkwUI I It-I IL ILA AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 220 4.21 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (UFSAR 15.4.4)

Based on Technical Specification 3.4.1.1, both reactor coolant loops and both reactor coolant pumps in each loop must be in operation in Modes 1 and 2. Therefore, this event is precluded for St. Lucie Unit 2. This disposition is not invalidated by the fuel transition.

4.22 A Malfunction or Failure of the Flow Controllerin a BWR Recirculation Loop That Results in an IncreasedReactorCoolant Flow Rate (UFSAR 15.4.5)

This event is not applicable to St. Lucie Unit 2.

4.23 CVCS Malfunction that Results in a Decreasein the Boron Concentrationin the Reactor Coolant (UFSAR 15.4.6) 4.23.1 Accident Description The Chemical and Volume Control System (CVCS) regulates both the chemistry and the quantity of coolant in the RCS. Changing the boron concentration in the RCS is a part of normal plant operation, compensating for long term reactivity effects, such as fuel burnup, xenon buildup and decay, and plant startup and cooldown.

For refueling operations, borated water is supplied from the refueling water tank (RWT), which assures adequate initial boron concentration above the required shutdown margin. An inadvertent boron dilution in any operational mode adds positive reactivity, produces power and possibly temperature increases, and, in Modes 1 and 2 (startup and power operations) can cause an approach to both the DNBR and FCM limits.

Planned boron dilution evolutions are conducted under strict administrative procedures, which specify permissible limits on the rate and magnitude of any required change in boron concentration. Boron concentration is determined by sampling the RCS.

Inadvertent dilution of the reactor coolant can be terminated by isolation of the makeup water system, by stopping either the makeup water pumps or the charging pumps, or by closing the charging isolation valves. A charging pump must be running for boron dilution to take place.

The CVCS is equipped with the following indications and alarm functions, which will inform the reactor operator when a change in boron concentration in the reactor coolant system may be occurring:

%.jIJIILI UIMIIU LJUgLoUI I ICI IL.

AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 221

Volume control tank (VCT) level indication and high and low alarms

Makeup flow indication and alarms

Volume control tank isolation.

In addition to the above, a boron dilution alarm is provided by the boron dilution alarm system (BDAS).

Because of the procedures involved and the numerous alarms and indications available to the operator, the probability of a sustained or erroneous dilution is very low.

4.23.2 Input Parameters and Assumptions Key input parameters are shown in Table 4.27, Table 4.28 and Table 4.29.

The calculated time-to-criticality from the activation of the BDAS alarm is dependent on the critical-to-alarm boron concentration ratio, the RCS coolant volume/mass, and the flow rate of the boron dilution stream. In addition, for the dilution front model, a range of shutdown cooling system (SDCS) flow rates is evaluated. The initial boron concentration of greater than or equal to the shutdown margin requirements for each Operating Mode is defined such that the total time to criticality allows adequate time for operator action subsequent to the activation of the BDAS alarm, which meets acceptance criteria for the respective modes. For Modes 5 and 6, a conservative RCS volume corresponding to mid-loop conditions is used. The sweepout volume is not used for the safety calculations per the methodology and is bounding.

4.23.3 Acceptance Criteria The acceptance criteria for Modes 2 through 6 are such that the time to criticality ensures the alarm activates to allow an operator action to terminate the event. The minimum required time to criticality after alarm activation for Modes 2 through 5 is 15 minutes. The minimum required time to criticality after alarm activation for Mode 6 is 30 minutes. Furthermore, the initial boron concentrations for each mode are calculated to allow sufficient time for the operator to sample the core in the scenario that the alarm is not active, consistent with the current analysis of record.

%-YLI ILI L)IIUU LU,;UI-~ I IMI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 222 4.23.4 Method of Analysis Cases have been considered for all six operational modes, i.e., power operation, startup, hot standby, hot shutdown, cold shutdown, and refueling. The boron dilution event was analyzed with the approved Reference 1 methodology.

An inadvertent boron dilution adds positive reactivity, produces power and temperature increases, and during operation at power (Mode 1) can cause an approach to both the DNBR and FCM limits. Since the TM/LP trip system monitors the transient behavior of core power level and core inlet temperature at power, the TM/LP trip will intervene, if necessary, to prevent the DNBR limit from being exceeded for power increases within the setting of the VHP trip. A Mode 1 boron dilution event is bounded by the reactivity insertion rates considered in the CEA Withdrawal at Power event.

In the event of an unplanned dilution during Mode 2 power escalation, the plant status is such that minimal impact will result. The plant will slowly escalate in power and activate a power-related trip (TM/LP or VHP). The acceptance criteria for Mode 2 must provide sufficient time to prevent a return to criticality. Prior to trip, challenges to the DNB and FCM SAFDLs are bounded by other events, such as CEA Withdrawal at Startup or Low Power.

Two models were used to evaluate the boron dilution transient,

instantaneous mixing model, and

dilution front model.

The instantaneous mixing model is applicable when at least one RCP is operating and assumes the unborated water is instantaneously mixed with the entire water volume in the RCS (Modes 2, 3 and 4). The dilution front model is used when the core is being cooled by the SDCS (no RCPs operating); for these operating modes the RCS flow is much lower than operating with a RCP and the assumption of instantaneous mixing of the unborated water with the entire RCS volumes is not valid (Modes 4, 5, and 6).

4.23.5 Results Table 4.30 shows the results from the boron dilution event.

%--UIILIUIIZ',U UUt,.oU111tZ',11L AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 223 The results of the instantaneous mixing model provide bounding values that ensure operator action prior to a complete loss in shutdown margin for Modes 2, 3, and 4.

The results of the dilution front model provide bounding values that ensure there is adequate time to operator action prior to a significant loss in shutdown margin provided the SDCS flow rate is maintained at or above the limits shown in Table 4.27, Table 4.28 and Table 4.29 and that the initial boron value is greater than the required shutdown margin.

The boron dilution event requirements are verified each refueling cycle using cycle-specific initial and critical boron concentrations. The initial boron concentrations will be provided to ensure protection of the minimum shutdown margin.

%-LIfu IUl UlI'U L/JU%.,UI I PI11 it AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chanter 15 Non-LOCA Summary Reoort Paae 224 Table 4.27 Boron Dilution: Inputs and Boron Requirements (3 Charging Pumps)

Bounding Boron Concentrations Minimum Required for Limiting Case (ppm) Initial Boron Concentration (Initial boron must Bounding Boron be equal to or Ratio greater than SDM Mode Critical Alarm (CriticallAlarm) requirement) (ppm)

Startup (Mode 2) 938 983 0.954 Case protected by TS SDM requirements Hot Standby 1197 1253 0.954 Case protected by TS (Mode 3) SDM requirements Hot Shutdown 1230 1486 0.961 2171 (Mode 4)

Cold Shutdown 1260 1616 0.780 2178 (Mode 5)

Refueling (Mode 1060 1328 0.799 1705 6)

All Rods In Parameter Value Charging Flow, 49 gpm per pump Number of 3 charging pumps Parameter Value Partial RCS 3 volume, ft

" Plant on 3711 SDCS with no RCPs running

" RCS filled to 3410 nozzle mid-plane with one SDCS train

" RCS filled to 2655 bottom of hot legs Full RCS volume, 8332 ft 3 Sweepout3 242.8 volume, ft

%f ILI U1IceU LJL)Iw.,U I I;I IL AREVA Inc. ANP-3347NP Revision 0

-qt I "iiri I mit 9 P~ Trnncifirnn C-h!n tcr 1 r, klrnnJ r)(- A 'Z~mmnr1 Pci rirf Pm g '"X; Table 4.28 Boron Dilution: Inputs and Boron Requirements (2 Charging Pumps)

Bounding Boron Concentrations Minimum Required for Limiting Case (ppm) Initial Boron Concentration (Initial boron must Bounding Boron be equal to or Ratio greater than SDM Mode Critical Alarm (Critical/Alarm) requirement) (ppm)

Hot Standby 1197 1234 0.969 Case protected by TS (Mode 3) SDM requirements Hot Shutdown 1230 1401 0.878 1817 (Mode 4)

Cold Shutdown 1260 1480 0.851 1825 (Mode 5)

Refueling (Mode 1060 1231 0.862 1599 6)

All Rods In Parameter Value Charging Flow, 49 gpm per pump Number of 2 charging pumps Parameter Value Partial RCS 3 volume, ft

" Plant on SDCS 3711 with no RCPs running

" RCS filled to 3410 nozzle mid-plane with one SDCS train

" RCS filled to 2655 bottom of hot legs

" Full RCS 3 8332 volume, ft Sweepout 3 242.8 volume, ft

ILI LJIIZU LJULoUI I It:-I IL AREVA Inc. ANP-3347NP Revision 0 5Zt I m')

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%i Table 4.29 Boron Dilution: Inputs and Boron Requirements (1 Charging Pump)

Bounding Boron Concentrations Minimum Required for Limiting Case (ppm) Initial Boron Concentration (Initial boron must Bounding Boron be equal to or Ratio greater than SDM Mode Critical Alarm (Critical/Alarm) requirement) (ppm)

Hot Standby 1197 1215 0.985 Case protected by TS (Mode 3) SDM requirements Hot Shutdown 1230 1247 0.935 1609 (Mode 4)

Cold Shutdown 1260 1363 0.925 1651 (Mode 5)

Refueling (Mode 1060 1142 0.862 Case protected by TS

6) SDM requirements All Rods In Parameter Value Charging Flow, 49 gpm per pump Number of 1 charging pumps Parameter Value Partial RCS3 volume, ft

" Plant on SDCS 3711 with no RCPs running

" RCS filled to 3410 nozzle mid-plane with one SDCS train

" RCS filled to 2655 bottom of hot legs

" Full RCS 3 8332 volume, ft Sweepout3 242.8 volume, ft

L)IULLjut.LJI I MN1 IL k..AJI OLElltu AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chanter 15 Non-LOCA Summary Report Paae 227 Table 4.30 Boron Dilution: Results Required Operator Response Time Time to Criticality SDCS Min. Flow Mode (minutes) (minutes) (gpm)

Instantaneous Mixing Model Startup (Mode 2) 15 15.08 N/A Hot Standby (Mode 3) 15 15.09 N/A Hot Shutdown (Mode 4) 15 15.09 N/A Dilution Front Model Hot Shutdown (Mode 4) 15 15.95 780 Cold Shutdown (Mode 5) 15 15.55 780 Refueling (Mode 6) 30 30.59 3000

%_A1I ILE UW=IUL LJUULUI I MI I L AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 228 4.24 InadvertentLoading of a Fuel Assembly (UFSAR 15.4.7)

Adequate processes and surveillance requirements are currently in place to preclude any significant misload from going undetected. This includes flux map symmetry check at low powers and continuous monitoring of key parameters during power ascension, such as the trend of axial shape index in each quadrant. There are no changes to these processes due to the planned fuel transition. Therefore, this event remains unaffected.

4.25 Spectrum of CEA Ejection Accidents (UFSAR 15.4.8) 4.25.1 Accident Description A control rod (or CEA) ejection event is initiated by a postulated rupture of a control rod drive mechanism housing. Such a rupture allows the full system pressure to act on the drive shaft, which ejects its control rod from the core. The consequences of the mechanical failure are a rapid positive reactivity insertion and an increase in radial power peaking, which could possibly lead to localized fuel rod damage.

Doppler reactivity feedback mitigates the power excursion as the fuel begins to heat up.

Although the initial increase in power occurs too rapidly for the scram rods to have any significant effect on the power during that portion of the transient, the scram negative reactivity insertion does affect the fuel temperature and fuel rod cladding surface heat flux.

The ejected rod causes localized peaking such that fuel failure may occur, due to DNB or FCM.

4.25.2 Input Parameters and Assumptions The key input parameters and their assumed values (listed in Table 4.31) used in the analysis of this event are consistent with or conservative relative to the approved Reference 1 methodology. See Section 2.0 for additional key input parameter values.

Initial Conditions - The analysis was performed from full power, 65% RTP, 20%

RTP and HZP initial conditions. Respective maximum core inlet temperatures were assumed for each initial condition. TS minimum RCS flow rate was modeled.

Reactivity Feedback - Reactivity feedbacks were modeled that represented BOC and EOC conditions. Due to the rapidity of the transient, moderator feedback has a second-order impact on the consequences. TS/COLR MTC limits were modeled for the cases initiated at BOC, whereas conservatively biased "least negative" MTCs were modeled for the EOC cases. The event is initially mitigated by the negative

IU IJI 1JiIU LJULosl I 1ItZI It AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 229 Doppler reactivity feedback. As such, the Doppler reactivity assumed in the analysis was conservatively biased to minimize the negative feedback due to increasing fuel temperatures. For the HZP initiated cases, fuel temperature dependent Doppler feedback was modeled.

Reactor Protection System Trips and Delays - The event is primarily protected by the VHP trip. The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, rod insertion is delayed to account for the CEA holding coil delay time.

Pressurizer Pressure Control - Pressurizer pressure control (i.e., pressurizer sprays, heaters, and PORVs) parameters and equipment states were selected to reduce the primary system pressure, which provided a conservative calculation of the MDNBR during the transient.

Ejected CEA Worth - To maximize the core power response to the ejected CEA, a conservatively high ejected CEA worth was assumed for each case, based on St.

Lucie Unit 2 specific rod patterns and power-dependent insertion limits.

Gap Conductance - Depending on the time-in-cycle for the reactivity coefficients, the fuel-to-cladding gap conductance was set to either a conservative BOC value or a conservative EOC value, to either The gap conductances account for the effect of TCD.

Single Failure - Since the systems designed to mitigate this event (namely, the RPS) are redundant, there is no single active failure that will adversely affect the consequences of the event.

4.25.3 Acceptance Criteria The acceptance criteria for this event are the following:

Fuel failures due to DNB and FCM should be limited, so as not to impair the capability to cool the core. Additionally, the fuel failures should be within the limits of fuel failures used in the radiological analysis.

Reactivity excursions should not result in a peak radial average fuel enthalpy greater than the following limits:

230 cal/gm for fuel coolability.

For HZP conditions, 150 cal/gm for fuel failure.

The pellet / cladding mechanical interaction (PCMI) failure criterion is a change in radial average fuel enthalpy greater than the corrosion-dependent limit depicted in Reference 14 (Section 4.2, Appendix B, Figure B-i).

'~LIILO 'JIICU LJLJjLJI II-I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 230 Peak RCS pressure must remain below that which would cause the stresses in the RCS to exceed the faulted condition stress limits.

4.25.4 Method of Analysis The S-RELAP5 code was used to model the key system components and calculate neutron power, fuel thermal response, surface heat transport, and fluid conditions (such as coolant flow rates, temperatures, and pressures). The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code (Reference 3), which was used to calculate the MDNBR using the HTP CHF correlation (Reference 5). Deposited enthalpy was calculated using the Reference 6 methodology.

Eight different initial-condition cases are analyzed for the event: (1) BOC full power, (2) EOC full power, (3) BOC 65% RTP, (4) EOC 65% RTP, (5) BOC 20% RTP, (6) EOC 20% RTP, (7) BOC HZP and (8) EOC HZP. Per the TS, the core is held subcritical by more than 1% for Mode 3 (Hot Standby), Mode 4 (Hot Shutdown) and Mode 5 (Cold Shutdown). Since 1% is more than the worth of the ejected control rod, evaluation of these modes is not required. For this analysis, HZP is assumed to be Mode 2 (Startup).

All four reactor coolant pumps are assumed to be in operation in both Mode 1 (Power Operation) and Mode 2 (Startup).

Conservative assumptions are typically used to bias the RCS pressure transient response. For evaluation of DNB, the RCS pressure is held constant at the initial value and is assumed to neither increase nor decrease.

4.25.5 Results The sequence of events is shown in Table 4.32. Results are given in Table 4.33 and Table 4.34.

The BOC full power case presented the most significant challenge to the DNB acceptance criterion. The transient response is shown in Figure 4.118 through Figure 4.124. Figure 4.118 shows the reactor power as a function of time. Figure 4.119 shows the core power based on rod surface heat flux. Figure 4.120 through Figure 4.124 show the RCS loop temperatures, the total RCS flow rate, the reactivity feedback, the peak fuel centerline temperatures, and the RCS pressure, respectively.

~IIIUlIjIVZ: LJUtL.UI I It:,I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 231 The BOC 20% RTP case presented the most significant challenge to the FCM acceptance criterion. The transient response is shown in Figure 4.125 through Figure 4.131. Figure 4.125 shows the reactor power as a function of time. Figure 4.126 shows the core power based on rod surface heat flux. Figure 4.127 through Figure 4.131 show the RCS loop temperatures, the total RCS flow rate, the reactivity feedback, the peak fuel centerline temperatures, and the RCS pressure, respectively.

Fuel Coolability Per Reference 14 (Section 4.2, Appendix B), the acceptance criterion for coolability is 230 cal/gm. The HZP, HFP, and part power total deposited enthalpy results are provided in Table 4.33 and Table 4.34. For the events analyzed for the fuel transition, the total deposited enthalpy is calculated to be less than 150 cal/gm, which is less than the criterion of 230 cal/gm. This criterion is therefore met.

Cladding Failures For HZP, the restrictive acceptance criterion for cladding failures, per Reference 14 (Section 4.2, Appendix B), is 150 cal/gm peak radial average fuel enthalpy. As shown in Table 4.34, the maximum calculated total deposited enthalpy for the HZP event is much less than 100 cal/gm, which meets the acceptance criterion of 150 cal/gm.

For at power events, the acceptance criterion for fuel cladding failure, per Reference 14 (Section 4.2, Appendix B), is the local heat flux not exceeding the thermal design limit (i.e.,

DNBR). The HFP and part power results are provided in Table 4.33 and Table 4.34. For the fuel transition analyses, the MDNBR is calculated to be greater than the DNBR limit for all analyzed power levels, thus meeting the acceptance criteria for cladding failures.

For pellet / cladding interaction (PCI) and PCMI failures, the fuel transition analyses performed at all power levels show that the enthalpy rise for the peak rods, is below 150 cal/gm, which meets the 150 cal/gm limit depicted in Reference 14 (Section 4.2, Appendix B, Figure B-i) for lower burned fuel.

%--JIILl 'JIIU LJUtdUI I I'I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 232 Fuel Centerline Melt The HZP, part power, and HFP fuel centerline temperature results are provided in Table 4.33 and Table 4.34. The peak fuel centerline temperatures are calculated to be below the corresponding fuel centerline melt temperatures. Thus there are no fuel melt failures.

Radiological Consequences The total number of fuel failures for this event is shown to be zero for both DNB and fuel centerline melt. Since the fuel failures calculated for this event are zero, the radiological consequences analysis remains bounding.

RCS Pressure The RCS overpressure criterion was demonstrated to be satisfied in a generic assessment in the current UFSAR. Fuel design parameters are not key parameters regarding overpressure, and therefore, have no significant effect on the overpressure aspects of this event. The maximum predicted ejected rods worths for the fuel transition are bounded by the values supported by the current UFSAR. Thus, the change in fuel design will not affect the conclusion regarding RCS overpressure in the current UFSAR.

%--eJIILI LJIIC;U LJUt-UI I 1C;I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chaoter 15 Non-LOCA Summary Reoort P~n* 9-:2 233 Pane Table 4.31 CEA Ejection: Assumptions Beginning of Cycle Full Power 65% RTP 20% RTP HZP Initial Power Level, % RTP 100.3 65.0 20.0 0.0 Ejected CEA Worth, % Ak 0.280 0.370 0.540 0.620 Doppler Coefficient, % Ak / OF 0.00100 0.00110 0.00116 BOC table(1 )

Trip Reactivity, % Ak 5.20 3.60 3.60 3.60 FQ After Ejection(2) 2.681 4.550 6.240 4.867 Number of RCPs Operating 4 4 4 4.

End of Cycle Full Power 65% RTP 20% RTP HZP Initial Power Level, % RTP 100.3 65.0 20.0 0.0 Ejected CEA Worth, % Ak 0.061 0.115 0.280 0.525 Doppler Coefficient, % Ak / OF 0.00120 0.00121 0.00126 EOC table(3)

Trip Reactivity, % Ak 5.20 3.60 3.60 3.60 FQ After Ejection(2) 2.320 3.984 7.271 8.781 Number of RCPs Operating 4 4 4 4 1

A table of BOC Doppler reactivity versus fuel temperature, biased less negative than nominal 2

Before measurement, design allowance, and engineering uncertainties are added 3

A table of EOC Doppler reactivity versus fuel temperature, biased less negative than nominal

'~AIILU l ltI;L Ljut-suI I IcI RL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summarv Reoort Paae 234 Table 4.32 CEA Ejection: Sequence of Events Case Event Time (sec)

BOC HFP Beginning of reactivity insertion 0.00 VHP trip setpoint reached 0.03 Ejected CEA fully withdrawn 0.05 Reactor scram on VHP trip (including trip response 0.43 delay)

Maximum neutron power 1.11 CEA insertion begins 1.17 Maximum average-core heat flux through cladding 2.07 MDNBR 2.07 Maximum fuel centerline temperature 3.25 EOC HFP Beginning of reactivity insertion 0.00 Ejected CEA fully withdrawn 0.05 VHP trip setpoint reached 0.07 Maximum neutron power 0.11 Reactor scram on VHP trip (including trip response 0.47 delay)

CEA insertion begins 1.21 Maximum average-core heat flux through cladding 1.34 MDNBR 1.34 Maximum fuel centerline temperature 3.11 BOC 65% RTP Beginning of reactivity insertion 0.00 VHP trip setpoint reached 0.03 Ejected CEA fully withdrawn 0.05 Reactor scram on VHP trip (including trip response 0.44 delay)

Maximum neutron power 1.08 CEA insertion begins 1.18 Maximum average-core heat flux through cladding 2.44 MDNBR 2.44 Maximum fuel centerline temperature 3.66

%-PAIU ILI '.jii;U LJU;LWUI I 1I:7I It AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Charter 15 Non-LOCA Summary Reoort P~ncA 93 235 Pane Table 4.32 CEA Ejection: Sequence of Events (Continued)

Case Event Time (sec)

EOC 65% RTP Beginning of reactivity insertion 0.00 Ejected CEA fully withdrawn 0.05 VHP trip setpoint reached 0.07 Maximum neutron power 0.21 Reactor scram on VHP trip (including trip response 0.47 delay)

CEA insertion begins 1.21 Maximum average-core heat flux through cladding 1.66 MDNBR 1.66 Maximum fuel centerline temperature 3.60 BOC 20% RTP Beginning of reactivity insertion 0.00 VHP trip setpoint reached 0.05 Ejected CEA fully withdrawn 0.05 Reactor scram on VHP trip (including trip response 0.45 delay)

Maximum neutron power 1.07 CEA insertion begins 1.19 MDNBR 2.58 Maximum average-core heat flux through cladding 2.59 Maximum fuel centerline temperature 3.84 EOC 20% RTP Beginning of reactivity insertion 0.00 Ejected CEA fully withdrawn 0.05 VHP trip setpoint reached 0.10 Reactor scram on VHP trip (including trip response 0.50 delay)

Maximum neutron power 1.02 CEA insertion begins 1.24 MDNBR 2.35 Maximum average-core heat flux through cladding 2.36 Maximum fuel centerline temperature 3.88

%.,.#UI OU U I IC',U LIUI-eU I I I C; I I L AREVA Inc. ANP-3347NP Revision 0 IZ"na"it I "? Pica Trnncifi,,n r-hm tor 1r, kMe~n- C'IA QI~mm~n P= r~rf Pm 'm ')'IA t& LJL bý ýJ Table 4.32 CEA Ejection: Sequence of Events (Continued)

Case Event Time (sec)

BOC HZP Beginning of reactivity insertion 0.00 Ejected CEA fully withdrawn 0.05 VHP trip setpoint reached 2.70 Reactor scram on VHP trip (including trip response 3.81 delay)

Maximum neutron power 4.48 CEA insertion begins 4.56 Maximum average-core heat flux through cladding 5.79 MDNBR 5.79 Maximum fuel centerline temperature 7.55 EOC HZP Beginning of reactivity insertion 0.00 Ejected CEA fully withdrawn 0.05 VHP trip setpoint reached 2.12 Maximum neutron power 2.19 Reactor scram on VHP trip (including trip response 3.23 delay)

CEA insertion begins 3.98 Maximum average-core heat flux through cladding 4.93 MDNBR 4.93 Maximum fuel centerline temperature 6.62

%_f~JI ILI UIICEL L./Utojl' I ICI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summarv ReDort Paae 237 Table 4.33 CEA Ejection: Full Power and 65% RTP Results Full Power 65% RTP Criterion BOC EOC BOC EOC Limit MDNBR 1.179 2.031 1.246 2.668 1.164

(% fuel failure) (0%) (0%) (0%) (0%)

(U0 2-Gd 2O3) (U0 2-Gd 2O 3) (fresh all-U0 2) (U0 2-Gd 2O 3) Case-Limiting fuel CLT 4500°F 4103°F 4860°F 3813°F specific FCM limit [ ]OF [ ]OF [ ] OF [ ]oF (see (0%) (0%) (0%) left)

(% fuel failure) (0%)

Total deposited 144.1 cal/gm 136.9 cal/gm 142.3 cal/gm 135.7 cal/gm Case-enthalpy specific Total deposited 230 cal/gm 230 cal/gm N/A N/A (see enthalpy limit left)

Enthalpy rise limit 150 cal/gm 150 cal/gm 150 cal/gm 150 cal/gm Table 4.34 CEA Ejection: 20% RTP and HZP Results 20% RTP HZP Criterion, BOC EOC BOC EOC Limit MDNBR 1.193 3.899 2.104 4.213 1.164

(% fuel failure) (0%) (0%) (0%) (0%)

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Figure 4.118 CEA Ejection: Reactor Power (BOC HFP)

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Figure 4.120 CEA Ejection: RCS Loop Temperatures (BOC HFP)

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Figure 4.121 CEA Ejection: RCS Total Loop Flow Rate (BOC HFP)

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Figure 4.122 CEA Ejection: Reactivity Feedback (BOC HFP)

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Figure 4.123 CEA Ejection: Peak Fuel Centerline Temperatures (BOC HFP)

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Figure 4.124 CEA Ejection: Pressurizer Pressure (BOC HFP)

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Figure 4.126 CEA Ejection: Total Core Heat Flux Power (BOC 20% RTP)

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AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 252 4.26 InadvertentOperation of the ECCS During Power Operation(UFSAR 15.5.1)

This event is caused by a malfunction which results in startup of the safety injection pumps such as an inadvertent SIAS. This event is precluded since the shutoff head of the injection pumps is much less than the RCS pressure in Mode 1. The impact of initiation of charging flow upon SIAS is the same as the CVCS Malfunction that Increases Reactor Coolant Inventory (UFSAR 15.5.2) event. Therefore, this event is bounded by the CVCS Malfunction that Increases Reactor Coolant Inventory event.

4.27 CVCS Malfunction that Increases Reactor Coolant Inventory (UFSAR 15.5.2)

This event produces an unplanned increase in reactor coolant system inventory that may be caused by operator error or a failure in the pressurizer level transmitter which causes an erroneous low-low level signal. This event tends to fill the pressurizer and increase the RCS pressure. However, pressurizer sprays mitigate any significant increase in RCS pressure.

Twenty minutes after either the HPP trip or Pressurizer High Level Alarm (PHLA), it is assumed that the operators mitigate the event by reducing charging flow or restoring letdown flow. The event is analyzed to demonstrate that operators have a sufficient amount of time to preclude the pressurizer from filling following a PHLA. Since operator actions preclude filling of the pressurizer, there is no liquid discharge through the PORVs.

The change in fuel design parameters has no effect on the filling of the pressurizer or RCS pressure during this event. There are no changes to the HPP trip and PHLA setpoints. Thus, this event does not require a reanalysis for the fuel transition and is bounded by the current UFSAR analysis of record.

4.28 InadvertentOpening of PressurizerSafety or Relief Valve (UFSAR 15.6.1) 4.28.1 Accident Description An accidental depressurization of the RCS could occur as a result of (1) an inadvertent opening of both PORVs, (2) an inadvertent opening of a single PSV, or (3) a malfunction of the pressurizer spray system. Since a PSV is sized to relieve approximately half the steam flow rate of a PORV, and the pressurizer spray valves, even if fully open, cannot depressurize the RCS at the rate of two open PORVs, the most severe core conditions are associated with an

%ji.JI Ll LJIIU L-JU..LII I I:I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 253 inadvertent opening of both PORVs. It is assumed that a mechanical failure, spurious actuation signal, or unanticipated operator action will cause the opening of both PORVs.

The opening of both PORVs results in a rapid depressurization of the RCS and a challenge to the DNB SAFDL. The core is protected from reaching the DNB SAFDL by the TM/LP trip.

Since the systems designed to mitigate this event in the short term (namely, the RPS) are redundant, there is no single active failure that will adversely affect the consequences of the event.

4.28.2 Input Parameters and Assumptions The key input parameters and their values used in the analysis of this event are consistent with or conservative relative to the approved Reference 1 methodology, as supplemented in Section 2.7. See Section 2.0 for key input parameter values.

Initial Conditions - This event was assumed to initiate from HFP conditions with a maximum core inlet temperature and TS minimum RCS flow. This set of conditions minimizes the initial margin to DNB.

Reactivity Feedback - The reactivity feedback coefficients were biased according to the approved methodology. BOC moderator density feedback was conservatively assumed for this event, although the reactivity feedback is not a significant parameter.

Reactor Protection System Trips and Delays - This event is primarily protected by the TM/LP RPS trip. The analysis did not credit the variable portion of the TM/LP function; it conservatively credited'only the Low Pressurizer Pressure floor setpoint. The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, rod insertion was delayed to account for CEA holding coil delay time.

Pressurizer PORV Flow Rate - A conservative pressurizer PORV flow rate was assumed for this analysis. The value was based on two valves opening, and the flow rate was biased 1% higher than the maximum flow rate, to maximize the depressurization of the RCS and the potential pressure undershoot of the RPS trip setpoint.

Gap Conductance - Gap conductance was set to a conservative BOC value, to be consistent with the time-in-cycle for the reactivity coefficients. The value accounts for the effect of TCD.

".JII LJII;V LJ6J~LJI I IMI FL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 254 4.28.3 Acceptance Criteria The principally challenged acceptance criterion for this event is:

Fuel cladding integrity should be maintained by ensuring that the SAFDLs are not exceeded.

This criterion is met by assuring that the minimum calculated DNBR is not less than the 95/95 DNB correlation limit. Since this event does not involve a significant power transient or augmented peaking, the fuel centerline melt limit is not challenged. Also, this event results in RCS depressurization and, thus, does not challenge the integrity of the Reactor Coolant Pressure Boundary (RCPB).

4.28.4 Method of Analysis The analysis was performed with the approved non-LOCA methodology given in Reference 1, as supplemented in Section 2.7. For this event, the S-RELAP5 code was used to model the key system components and calculate neutron power, fuel thermal response, surface heat transport, fluid conditions (such as coolant flow rates, temperatures, and pressures), and an estimated time of MDNBR. The core fluid boundary conditions and average rod surface heat flux were then input to the XCOBRA-IIIC code (Reference 3), which was used to calculate the MDNBR using the HTP CHF correlation (Reference 5). This event was also addressed as part of the TM/LP statistical setpoint analyses, using the Reference 4 methodology.

A single calculation was performed at BOC HFP conditions, maximum TS core inlet temperature, and minimum TS RCS flow rate. This produced the minimum margin to the DNB limit. A conservative moderator density reactivity feedback was used, based on the HZP TS MTC. The analysis simulated an inadvertent and instantaneous full opening of both pressurizer PORVs, which maximizes the depressurization of the RCS and challenge to the DNB SAFDL 4.28.5 Results The sequence of events is shown in Table 4.35. The result is given in Table 4.36. The MDNBR was calculated to be above the 95/95 limit for the HTP CHF correlation.

The system responses are shown in Figure 4.132 through Figure 4.139. Figure 4.132 shows the reactor power as a function of time. Figure 4.133 shows the core power based on rod surface heat flux. Figure 4.134 through Figure 4.139 show the pressurizer pressure, PORVs

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For the DNB portion of the event, the pressurizer does not fill with liquid water during this event, as shown in Figure 4.139. The early part of the transient is terminated by the TM/LP reactor trip, as described above. The longer-term portion of this event, which is not affected by the fuel transition, will be terminated by operator action, consistent with the current analysis of record.

Table 4.35 Inadvertent Opening of Pressurizer PORVs: Sequence of Events Event Time (sec)

Inadvertent opening of two PORVs 0.00 Pressurizer pressure reaches Low Pressurizer Pressure 18.85 trip setpoint Reactor scram on Low Pressurizer Pressure (including 20.02 signal processing delay)

MDNBR 20.45 CEA insertion begins 20.77 Table 4.36 Inadvertent Opening of Pressurizer PORVs: Result Criterion Result Limit MDNBR 1.237 1.164

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%.JI lM JIItu LJLOsLAI Iq:; IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 264 4.29 Break in Instrument Line or Other Lines from the ReactorCoolant Pressure Boundary that Penetratethe Containment(UFSAR 15.6.2)

This event may result from a break in a letdown line, instrument line, or sample line. The double-ended break of the letdown line (the largest line) outside containment upstream of the outside containment isolation valve results in the largest release of reactor coolant to the environment. This event is analyzed to ensure that the resulting site boundary doses are within the guidelines established for design basis accidents. This event does not directly challenge peak LHR, DNBR, peak RCS pressure, or peak secondary pressure, and is bounded by the Inadvertent Opening of a Pressurizer Safety or Relief Valve (UFSAR 15.6.1) event with respect to RCS depressurization rate and approach to the DNB criteria.

This event is analyzed mainly for RCS mass release input to the dose analysis. The mass release is mainly a function of letdown line configuration, reactor power level, and the plant operating parameters, which remain unchanged. This event is thus unaffected by the fuel transition and the UFSAR analysis remains bounding.

4.30 Steam GeneratorTube Rupture (UFSAR 15.6.3)

This event is assumed to be caused by the instantaneous rupture of a steam generator tube that relieves to the lower pressure secondary system. This event experiences a decrease in RCS pressure, but no significant change in reactor power. This event is protected by the TM/LP trip and poses no significant challenge to the DNBR SAFDL.

This event is analyzed to demonstrate that radiological criteria are satisfied. The transient analysis provides input to a subsequent radiological analysis. Key parameters for this event include system related parameters such as the break flow rate and mass release from the affected steam generator MSSVs and Atmospheric Dump Valves (ADVs). The change in fuel design parameters has no significant effect on this event. The reactor power, reactor trip setpoints and the safety valves opening setpoints with the tolerances remain unchanged. Flow capacities of MSSVs and ADVs remain the same. There is no change to the TS limit for primary to secondary leakage. Therefore, this event does not require a reanalysis for dose consequences due to the fuel transition and is bounded by the current UFSAR analysis of record.

%.OVIILI 'JIRVU L-/Uk.,UI I IIVI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 265 4.31 Spectrum of Boiling Water Reactor (BWR) Steam System Piping FailuresOutside of the Containment (UFSAR 15.6.4)

This event is not applicable to St. Lucie Unit 2.

4.32 Loss of CoolantAccidents (UFSAR 15.6.5)

The LOCA events are outside the scope of this document.

4.33 Radioactive Releases from a Subsystem or Component (UFSAR 15.7)

The radiological analyses are outside the scope of this document.

4.34 PrimarySystem PressureDeviation Events (UFSAR 15.8)

The fuel design parameters have no significant effect on the pressure increase or decrease of any of the primary system pressure deviation events identified in Section 15.8 of the UFSAR.

Also, none of the primary system pressure deviation events identified in Section 15.8 of the UFSAR pose a significant challenge to the DNBR SAFDL. Therefore, none of the events in Section 15.8 of the UFSAR require a re-analysis for the fuel transition.

4.35 Anticipated Transients without Scram (UFSAR 15.9)

The change in fuel design parameters will have no effect on the ATWS as described in the UFSAR. Therefore, ATWS is not affected by the fuel transition.

4.36 Station Blackout (UFSAR 15.10)

The Station Blackout (SBO) event results from a loss of offsite power followed by failure of both standby diesel generators to start. The SBO analysis is performed to demonstrate that the plant can endure a complete loss of alternating current (AC) power for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The initial part of the SBO event is similar to the Complete Loss of Forced Reactor Coolant Flow event (Section 4.15) with RCP coast-down due to loss of power to the RCPs, followed by a reactor trip on low RCS Flow. This part of the SBO event is bounded by the Complete Loss of Forced Reactor Coolant Flow event (Section 4.15).

The remainder of the SBO event is characterized by the system response as a result of associated operator actions. Heat is transferred to the steam generators via natural circulation.

%-fAI Il DU IMUL LJULoUI I IMI IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 266 1

Following a turbine trip due to the loss of offsite power, the MSSVs open. Later on, operators control the SG pressure below the MSSV setpoint via the ADVs. AFW is supplied to the SGs via the steam-driven AFW pump to maintain SG level and remove decay heat. Eventually, AC power is restored.

There is no significant challenge to RCS or secondary side overpressure for this event.

Other than the short period before reactor trip, which is bounded by the Complete Loss of Forced Reactor Coolant Flow event (Section 4.15), fuel design parameters are not key parameters for this event. The major contributor to the RCS coolant inventory loss is the RCP seal leakage. There is no challenge to the RCS pressure which remains below the opening setpoint of both the PSVs and the PORVs, minimizing any loss of reactor coolant inventory via these valves. The secondary side response is governed by the opening of the MSSVs and the AFW flow from the turbine driven pump. None of these parameters change for the fuel transition. Therefore, this event does not require a re-analysis for the fuel transition.

fu IJIIU ItoLJ LJJukoUI I BI1I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 267 5.0

SUMMARY

OF RESULTS Table 5.1 presents the results for the Chapter 15 non-LOCA transient analyses that were performed to support the St. Lucie Unit 2 fuel transition. The results demonstrate that acceptance criteria are met for all events for operation of St. Lucie Unit 2 with AREVA fuel.

AREVA Inc. ANP-3347NP Revision 0 St Lucie Unit 2 Fuel Transition ChaDter 15 Non-LOCA Summary ReDort Paae 268 Table 5.1 Analytical Limits and Limiting Results UFSAR Limiting Analytical Section Event Description Criterion Analytical Limit Result 15.1.1 Decrease in Feedwater MDNBR 1.164 1.257 Temperature Peak LHR, kW/ft J 18.24 15.1.2 Increase in Feedwater Flow MDNBR 1.164 1.220 Peak LHR, kW/ft ] 18.50 Peak CLT, 'F . ] 3385 (HZP) 15.1.3 Increase in Steam Flow MDNBR 1.164 1.271 Peak LHR, kW/ft [ ] 19.12 Peak CLT, 'F [ ] 3491 (HZP) 15.1.5 Pre-scram Main Steam Line MDNBR (%fuel failure) 1.164 1.203 (0%)

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

15.1.6 Post-scram Main Steam Line MDNBR (%fuel failure) [ ] 1.740(0%)

Break Peak LHR, kW/ft (% fuel failure) [ 1 17.02 (0%)

15.2.3 Loss of Condenser Vacuum MDNBR 1.164 1.553 Peak LHR, kW/ft [ ] 16.04 15.2.9 Transients Resulting from the MDNBR 1.164 1.713 Malfunction of One Steam Peak LHR, kW/ft [ ] 15.74 Generator 15.3.2 Loss of Forced Reactor Coolant MDNBR 1.164 1.227 Flow 15.3.3 Reactor Coolant Pump Rotor MDNBR (% fuel failure) 1.164 1.205 (0%)

, Seizure I

AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Paae 269 Table 5.1 Analytical Limits and Limiting Results (Continued)

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

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

Total deposited enthalpy limit, cal/gm 230 (HFP) 144.1 (HFP) 150 (HZP) 26.9 (HZP) 15.6.1 Inadvertent Opening of MDNBR 1.164 1.237 Pressurizer Safety or Relief Valve

%--VI ILI L.PII;U LJULwUI I I--I IL AREVA Inc. ANP-3347NP Revision 0 St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report Page 270

6.0 REFERENCES

1. EMF-231 0(P)(A), Revision 1, SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, Framatome ANP, Inc., May 2004.

2. Florida Power and Light, St. Lucie Plant Unit No. 2 Technical Specifications,Amendment 166, August 30, 2013.

3. XN-NF-82-21 (P)(A), Revision 1, Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,Exxon Nuclear Company, August 1983.

4. EMF-1 961 (P)(A), Revision 0, Statistical Setpoint/TransientMethodology for Combustion EngineeringType Reactors, Siemens Power Corporation, July 2000.

5. EMF-92-153(P)(A), Revision 1, HTP: DepartureFrom Nucleate Boiling Correlationfor High Thermal PerformanceFuel, Siemens Power Corporation, January 2005.

6. XN-NF-78-44(NP)(A), A Generic Analysis of the Control Rod Ejection Transientfor PressurizedWater Reactors, Exxon Nuclear Company, September 1983.

7. XN-NF-81-58(P)(A), Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-MechanicalResponse Evaluation Model, Exxon Nuclear Company, November 1983.

8. ANF-81-58(P)(A), Revision 2 Supplements 3 and 4, RODEX2 Fuel Rod Thermal MechanicalResponse Evaluation Model, Advanced Nuclear Fuels, April 1990.

9. BAW-1 0231 P-A Revision 1, COPERNIC Fuel Rod Design Computer Code, Framatome ANP, Inc., January 2004.

10. XN-75-21 (P)(A), Revision 2, XCOBRA-IIIC: A Computer Code to Determine the Distributionof CoolantDuring Steady State and Transient Core Operation,Exxon Nuclear Company, March 1985.

11. EMF-96-029(P)(A), Volumes 1 and 2, ReactorAnalysis System for PWR's, Volume 1 -

Methodology Description, Volume 2 - Benchmarking Results, Siemens Power Corporation, October 1996.

12. Energia Nucleare, Volume 14, No. 9, September 1967, "Studies on Burnout, Part3 - A New Correlationfor Round Ducts and Uniform Heating and Its Comparison with World Data," L. Biasi et. al.

13. BAW-1 0240(P)(A), Revision 0, Incorporationof M5TM Propertiesin Framatome ANP Approved Methods, Framatome ANP, Inc., May 2004.

14. U.S. Nuclear Regulatory Commission, NUREG-0800, Revision 3, StandardReview Plan, March 2007.

ILI UIICU L-JLUt-UI 5 'Il IL A

AREVA AN P-3345N P St Lucie Unit 2 Fuel St. Lucie Unit 2 Fuel Transition ANP-3345NP Revision 0 Small Break LOCA Summary Report December 2014 AREVA Inc.

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1/4.LIIl 'JIICýU LJUtý,JI I It-, IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page i Nature of Changes Section(s)

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

%..UIILl ,LII*U LJULdUl I I1-i IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page ii Contents Pa1.e 1.0 IN T RO DUCTIO N. ..... ..... ...................................... ...................................... 1-1 2.0

SUMMARY

OF RESULTS ........................................................... 2-1

3.0 DESCRIPTION

OF ANALYSIS .......................................................................... 3-1 3.1 Description of SBLOCA Event ................................................................ 3-1 3.2 A nalytical Models .................................................................................... 3-4 3.3 Plant Description and Summary of Analysis Input P a ra m ete rs ............................................................................................. 3-6 3.4 S E R C om pliance .................................................................................... 3-8 4.0 ANALYTICAL RESULTS ................................................................................. 4-17 4.1 Results for Break Spectrum .................................................................. 4-17 4.2 Discussion of Transient for Limiting Break ............................................ 4-17 4.3 Attached Piping Sensitivity Study ......................................................... 4-19 5.0 R E F E R E NC E S .................................................................................................. 5-1

AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page iii List of Tables Table 3-1 Small break LOCA Development Phases ................................................... 3-9 Table 3-2 System Parameters and Initial Conditions ................................................ 3-10 Table 3-3 HPSI Flow Rate versus RCS Pressure ..................................................... 3-11 Table 3-4 LPSI Flow Rate versus RCS Pressure ...................................................... 3-12 Table 4-1 Summary of SBLOCA Break Spectrum Results ........................................ 4-20 Table 4-2 Sequence of Events for the SBLOCA Break Spectrum ............................. 4-21

%.,PUI ILI UIIULA L.JUtoUI I I1ZI IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page iv List of Figures Figure 3-1 S-RELAP5 SBLOCA Reactor Coolant System Nodalization .................... 3-13 Figure 3-2 S-RELAP5 SBLOCA Secondary System Nodalization ............................ 3-14 Figure 3-3 S-RELAP5 SBLOCA Reactor Vessel Nodalization .................................. 3-15 Figure 3-4 Axial Power Shape ................................................................................... 3-16 Figure 4-1 Peak Cladding Temperature versus Break Size (SBLOCA Break S p e ctrum ) ............................................................................................. 4-2 2 Figure 4-2 Reactor Power - 2.70 inch Break ............................................................ 4-23 Figure 4-3 Primary and Secondary System Pressures - 2.70 inch Break ................. 4-24 Figure 4-4 Break Void Fraction - 2.70 inch Break ..................................................... 4-25 Figure 4-5 Break Mass Flow Rate - 2.70 inch Break ................................................ 4-26 Figure 4-6 Loop Seal Void Fraction - 2.70 inch Break .............................................. 4-27 Figure 4-7 RCS Loop Mass Flow Rate - 2.70 inch Break ......................................... 4-28 Figure 4-8 Main Feedwater Mass Flow Rate - 2.70 inch Break ................................ 4-29 Figure 4-9 Auxiliary Feedwater Mass Flow Rate - 2.70 inch Break .......................... 4-30 Figure 4-10 Steam Generator Total Mass- 2.70 inch Break .................................... 4-31 Figure 4-11 High Pressure Safety Injection Mass Flow Rates - 2.70 inch Break ...... 4-32 Figure 4-12 Low Pressure Safety Injection Mass Flow Rates- 2.70 inch Break ...... 4-33 Figure 4-13 Safety Injection Tank Mass Flow Rates - 2.70 inch Break .................... 4-34 Figure 4-14 Reactor Vessel Mass Inventory - 2.70 inch Break ................................. 4-35 Figure 4-15 Hot Assembly Mixture Level - 2.70 inch Break ...................................... 4-36 Figure 4-16 Hot Spot Cladding Temperature - 2.70 inch Break ................................ 4-37

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l to IIt.U L-Jut.RUI I It:-I IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page v Nomenclature Acronym Definition ADV Atmospheric dump valve AFAS Auxiliary feedwater actuation signal AFW Auxiliary feedwater AOR Analysis of record AREVA AREVA Inc.

BOC Beginning-of-cycle CE Combustion Engineering CEA Control element assembly CFR Code of Federal Regulations CRGT Control rod guide tube DC-HL Downcomer- Hot Leg DC-UH Downcomer - Upper Head ECCS Emergency Core Cooling System EDG Emergency diesel generator EM Evaluation model EOC End-of-cycle EOP Emergency Operating Procedure HPSI High pressure safety injection LHR Linear heat rate LOCA Loss-of-coolant accident LPSI Low pressure safety injection MFW Main feedwater MSIV Main steam isolation valve MSSV Main steam safety valve NRC Nuclear Regulatory Commission RAI Requested for Additional Information RCP Reactor coolant pump RCS Reactor Coolant System PCT Peak cladding temperature PWR Pressurized water reactor PZR Pressurizer SBLOCA Small break loss-of-coolant accident SER Safety Evaluation Report SG Steam generator SGTP Steam generator tube plugging SI Safety injection SIAS Safety injection actuation signal SIT Safety injection tank

ILI UIIUU LJUtoUl I IUI It AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break Summary Report Page 1-1

1.0 INTRODUCTION

The purpose of this report is to summarize the small break loss-of-coolant accident (SBLOCA) analysis performed with AREVA 16x16 HTP fuel with M51 cladding for the St. Lucie Unit 2 plant. This document provides input to the License Amendment Request (LAR) in support of the transition to AREVA fuel in St. Lucie Unit 2. The SBLOCA analysis was performed in accordance with AREVA's S-RELAP5 SBLOCA methodology (References 1 and 2) and the additional considerations discussed in Section 3.2.

A complete spectrum of cold leg break sizes was considered, ranging from 2.0-inch diameter (0.022 ft2) to 9.49-inch diameter (0.491ft 2) range. In addition, a sensitivity study was performed to consider attached piping break sensitivity.

The analysis supports plant operation at a core power level of 3029.06 MWt (including measurement uncertainty), a peak linear heat rate (LHR) of 13.0 kW/ft, a radial peaking factor of 1.65 (1.81 including uncertainty and an augmentation factor) and a steam generator tube plugging level of 20% with +/-4% asymmetry.

A bounding total SIT line and check valve loss coefficient value is used in the SBLOCA analysis for each loop, including both major and minor loss components.

1 M5 is a registered trademark of AREVA Inc.

%_AJIILI 'JIIt:U LJUtUI I ICI IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break Summary Report Page 2-1 2.0

SUMMARY

OF RESULTS A SBLOCA break spectrum analysis was performed for St. Lucie Unit 2 to demonstrate that the following acceptance criteria for Emergency Core Cooling Systems (ECCS), as stated in 10 Code of Federal Regulations (CFR) 50.46(b)(1-4) (Reference 3), have been met.

1. The calculated maximum fuel element cladding temperature shall not exceed 2200 0 F.

2. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

3. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

4. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

The limiting peak cladding temperature (PCT) is 1926°F for a 2.70-inch diameter (0.040 ft2 ) cold leg pump discharge break. The total maximum local oxidation is less than 9.0%,

including a pre-transient oxidation of 2.3925% and transient maximum oxidation of less than 6.0%. The maximum core-wide oxygen generation is less than 0.2%. The results of the analysis demonstrate the adequacy of the ECCS to support the criteria given in 10 CFR 50.46(b).

In addition to the break spectrum analysis, a sensitivity study was performed to consider a break in an attached pipe. The break in an attached piping sensitivity study was performed with a 10.126-inch diameter (double-ended guillotine) break in the safety injection tank (SIT) line. The PCT calculated for this case was 14511F. The SIT line break results are non-limiting compared to the break spectrum results.

'.LIIl LJIIzU Ljut~lJ I Mei IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break Summary Report Page 3-1

3.0 DESCRIPTION

OF ANALYSIS Section 3.1 of this report provides a brief description of the postulated SBLOCA event.

Section 3.2 describes the analytical models used in the analysis. Section 3.3 presents a description of the St. Lucie Unit 2 plant and outlines the system parameters used in the SBLOCA analysis. Section 3.4 describes the Safety Evaluation Report (SER) compliance.

3.1 Description of SBLOCA Event The postulated SBLOCA is defined as a break in the Reactor Coolant System (RCS) pressure boundary for which the area is 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 (RCP). This break location results in the largest amount of inventory loss and the largest fraction of ECCS fluid ejected out through the break. This produces the greatest degree of core uncovery, the longest fuel rod heatup time, and consequently, the greatest challenge to the 10 CFR 50.46(b)(1-4) criteria (Reference 3).

%,.e'JI Ll LJIIU L/JULoUI I I5I AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break Summary Report Paqe 3-2 The SBLOCA event is characterized by a slow depressurization of the primary system with a reactor trip occurring on a Low Pressurizer Pressure signal. The safety injection actuation signal (SIAS) occurs when the system has further depressurized. The capacity and shutoff head of the HPSI pumps are important parameters in the SBLOCA analysis. For the limiting break size, the rate of inventory loss from the primary system is such that the HPSI pumps cannot preclude significant core uncovery. The primary system depressurization rate is slow, extending the time required to reach the SIT injection pressure or to recover core liquid level on HPSI flow. This tends to maximize the heatup time of the hot rod which produces the maximum PCT and local cladding oxidation. Core recovery for the limiting break begins when the SI flow that is retained in the RCS exceeds the mass flow rate out the break, followed by injection of SIT flow. For very small break sizes, the primary system pressure does not reach the SIT injection pressure.

The SBLOCA event develops in the following distinct phases: (1) subcooled depressurization, (2) loop saturation and loop flow coastdown, (3) loss of loop circulation and reflux mode cooling, (4) loop seal clearing and core refill and (5) long-

V-API ILI UliI~U LJU%.UI I ICA IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break Summary Report Page 3-3 term cooling provided by high and low head safety pump and SIT injections. The SBLOCA development phases are outlined in Table 3-1.

Following the break, the RCS rapidly depressurizes to the saturation pressure of the hot leg fluid. During the initial depressurization phase, reactor trip occurs on low pressurizer pressure, and the turbine is tripped on the reactor trip. The assumption of a loss-of-offsite power concurrent with the reactor SCRAM results in reactor coolant pump trip'.

In the second phase of the transient, the reactor coolant pumps coastdown. In this phase, natural circulation flows are sufficient to provide continuous core heat removal via the steam generators. However, mass continues to be lost to the break during this period.

The third phase in the transient is characterized as a period of loop draining that results in the loss of RCS flows. During this period, the core decay heat removal is provided via reflux boiling. The RCS stabilizes at an equilibrium pressure above the steam generator secondary side pressure. The system reaches a quiescent state in which the core decay heat, break flow, steam generator heat removal, and system hydrostatic head balance combine to control the core inventory.

The fourth phase in the transient is characterized by loop seal clearing and core recovery. The RCS inventory continues to decrease. Prior to loop seal clearing, liquid trapped in the reactor coolant pump suction piping can prevent steam from venting via the break. For a small break, the transient develops slowly, and liquid level in the reactor coolant system may descend to the loop seal level prior to establishing a steam vent. The core can become temporarily uncovered in this loop seal clearing process.

Once the loop seal clears, venting of steam through the break causes a rapid RCS depressurization below the secondary pressure and boiling in the core increases. The Tripping the reactor coolant pumps at the time of SCRAM instead of time zero is:

" A small delay relative to the time of loop seal uncovery for the limiting cases.

" Expected to be slightly conservative, due to the additional loss of primary system inventory through the break.

%-/l ILI. UIIOU LJLJLwUI I 1t;I IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break Summary Report Page 3-4 depressurization also promotes an increase in ECCS flows and the mass loss through the break decreases substantially as a result of phase change. These occurrences combine to cause either the core to uncover and heat up if ECCS flow is still not sufficient to offset the inventory lost out the break or an increase in RCS liquid inventory, preventing core uncovery, when ECCS flow is greater than the inventory lost out the break.

The last phase of the transient is characterized as a long-term cooling period during which the RCS inventory control is provided by the Emergency Core Cooling System.

Pumped injection continues and the passive SIT injection occurs when the RCS pressure decreases below the SIT tank pressure. Long-term RCS inventory and decay heat removal are successfully controlled in this manner.

3.2 Analytical Models The AREVA S-RELAP5 SBLOCA evaluation model (References 1) for event response of the primary and secondary systems and the hot fuel rod used in the analysis is based on the use of two computer codes. The appropriate conservatisms, as prescribed by Appendix K of 10 CFR 50, are incorporated. This methodology has been reviewed and approved by the NRC to perform SBLOCA analyses. The two AREVA computer codes used in the analysis are:

1. The RODEX2-2A code (References 4 and 5) was used to determine the burnup-dependent initial fuel rod conditions for the system calculations.

2. The S-RELAP5 code was used to predict the thermal-hydraulic response of the primary and secondary sides of the reactor system and the hot rod response.

'~iIIl UIIVU LJUI...UI I IC;I It AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break Sujmmqrv Rennrt Pane 3-5 m

%.o1ILI UJlIZ-U L.JUt.~.-UI I B5I IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break Summary Report Page 3-6 System nodalization details are shown in Figure 3-1 (RCS), Figure 3-2 (Secondary System), and Figure 3-3 (Reactor Vessel).

3.3 Plant Description and Summary of Analysis Input Parameters St. Lucie Unit 2 is a Combustion Engineering (CE)-designed Pressurized Water Reactor (PWR) with two hot legs, four cold legs, and two vertical U-tube steam generators (SGs). The reactor has a core power of 3029.06 MWt (including measurement uncertainty). The reactor vessel contains a downcomer, upper and lower plena, and a reactor core containing 217 fuel assemblies. The hot legs connect the reactor vessel with the vertical U-tube steam generators. Main feedwater (MFW) is injected into the downcomer of each SG. There are three AFW pumps, two motor-driven and one turbine-driven. The ECCS contains two HPSI pumps, four SITs, and two low pressure safety injection (LPSI) pumps.

The RCS was nodalized in the S-RELAP5 model with control volumes interconnected by flow paths or "junctions." The model includes four SITs, a pressurizer, and two SGs with both primary and secondary sides modeled. All of the loops were modeled explicitly to provide an accurate representation of the plant. A SGTP level of 20% in each steam generator was modeled, which bounds an average SGTP level of up to 20% with an asymmetry of +/-4%. Important system parameters and initial conditions used in the analysis are given in Table 3-2.

The heat generation rate in the S-RELAP5 reactor core model was determined from reactor kinetics equations with actinide and decay heat as prescribed by Appendix K of 10 CFR 50.

The analysis assumed loss of offsite power concurrent with reactor SCRAM, which is based on the low pressurizer pressure reactor trip and includes delays for Reactor

%,LIILl IJII5U LJUtLsUIIIC'IIL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break Summary Report Page 3-7 Protection System (RPS) circulation and control element assembly (CEA) coil delay.

The single failure criterion required by Appendix K of 10 CFR 50 was satisfied by assuming the loss of one emergency diesel generator (EDG). Thus, a single HPSI and LPSI pump were assumed to be available. The charging system is a safety related system and supported by the remaining EDG, therefore flow from the charging system was credited in the analysis. Initiation of the HPSI and LPSI systems was delayed by 30 seconds beyond the time of SIAS. The 30-second delay represents the time required for diesel generator startup and switching. The charging system is delayed 330 seconds following actuation of SIAS.

The HPSI system was modeled to deliver the SI flow symmetrically to all four loops (Table 3-3). The LPSI system was modeled to deliver the total SI flow asymmetrically to the broken loop (Loop 2B) and one intact loop (Loop 2A) (Table 3-4). The HPSI and LPSI flow are modeled differently, asymmetric verses symmetric, due to the influence the single failure criterion has on the LPSI system. When assuming single failure of an EDG, valves are not opened to allow LPSI flow to Loops 1A and 1 B.

The disabling of a motor-driven AFW pump, due to the single failure criterion, leaves one motor-driven pump and the turbine-driven pump available. The initiation of the motor-driven AFW pump was delayed 330 seconds beyond the time of the auxiliary feedwater actuation signal (AFAS) indicating low SG level (4.0% narrow range). The turbine-driven AFW pump was not credited in the analysis. The AFW flow is directed to the SG attached to the broken loop. Although not significant, a sensitivity study performed with AFW directed into the SG attached to the intact loop produced a lower PCT.

The input model included details of both main steam lines from the SGs to the turbine control valve, including the main steam safety valve (MSSV) inlet piping connected to the main steam lines. The MSSVs were set to open at their nominal setpoints plus 3%

tolerance for the first bank and 2% tolerance for the second bank.

The axial power shape for this analysis is shown in Figure 3-4.

%-PLIPLO WK;IUL LJULOJUI I M-11 IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break Summary Report Page 3-8 3.4 SER Compliance A spectrum of cold leg break sizes from 0.022 ft2 (2.0-inch diameter) to 0.491 ft2 (9.49-inch diameter, 10% of cold leg pipe area) was analyzed. This satisfies the limitation placed on EMF-2328 (Reference 1), that the methodology is acceptable for modeling transients where the break flow area is less than or equal to 10% of the cold leg flow area. In addition, to support the operation of St Lucie Unit 2 with M5 cladding, a sensitivity study was performed to consider attached piping break sensitivity. There is no other SER requirement or restriction on EMF-2328 (Reference 1).

1/4,.#IJI flu 4~Jii~U LJIJLjUI I I~I IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Pagje3-9 Table 3-1 Small break LOCA Development Phases (1) (2) (3) (4) (5)

Subcooled Loop Loss of loop Loop seal Long-term SBLOCA depressurization saturation and circulation clearing and cooling Phase >>>> loop flow and reflux core refill coastdown mode cooling_

RCS RCS boils Primary depressurization, hot leg saturates, RCS down and flow isitrued loop seal clears, tamsvned core covered and codbyEG Description reactor trip, flow coasts is interrupted steam is vented cooled by ECCS turbine trip, loss of down by void to break injection offsite power formation subcoole" / sauaed Break subcooled/ saturated subcooled /

Break subcooled liquid saturated superheated saturated liquid Characteristics discharge discharge liquid vapor discharge discharge discharge RCS Flow forced flow and coastdown coastdown to natural stagnant steam break flow to pool boiling circulation forced rfu forced convection foreddensauio RCS Heat forcsteam convection via condensation boiling and boiling and break Removal via steam in steam break flow flow generators generators generators, break flow pressure rapid rapid rapid plateaus just depressurization slow RCS Pressure rapid rapid above below depressurization depressurization depressurization secondary secondary pressure pressure continuous, continuous, ECCS Injection none none initiates potential SIT potential SIT injection injection covered or covered or2 Core Level covered covered uncovered' uncovered core recovery 1 Depending on the loop seal elevation with respect to the top of the active core.

2 Depending on the break size.

%.fIJIILI UlJII;U LJ.J~UIo I Iq:I it AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page 3-10 Table 3-2 System Parameters and Initial Conditions Reactor Power, MWt 3029.061 Peak LHR, kW/ft 13.0 Radial Peaking Factor (1.65 plus uncertainty & augmentation factor) 1.81 RCS Flow Rate, gpm 370000 Pressurizer Pressure, psia 2250 Core Inlet Coolant Temperature, 'F 554 SIT Pressure, psia 499.7 SIT Fluid Temperature, OF 124.5 Average SG Tube Plugging Level, % 20 SG Secondary Pressure, psia 846-849 MFW Temperature, TF 436 AFW Flow Rate per SG, gpm 255 AFW Temperature, IF 110 AFW Delay Time, sec 330 Low SG Level AFAS Setpoint, % 4 HPSI & LPSI Fluid Temperature, TF 104 Charging System Delay Time, sec 330 Reactor Trip - Low Pressurizer Pressure Setpoint, psia 1810 Reactor Trip Delay Time on Low Pressurizer Pressure, sec 1.15 SCRAM CEA Holding Coil Release Delay Time, sec 0.74 SIAS Activation Setpoint Pressure for Harsh Conditions, psia 1638 HPSI & LPSI Pump Delay Time on SIAS, sec 30 Nominal + 3% (Bank 1 Valves)

MSSV Lift Pressures and Tolerances Nominal + 2% (Bank 2 Valves) 1 Includes 0.3% measurement uncertainty

\.AIIl LUIR;U L;Ut.UI I 1C;1 IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Paine 3-1 1 Table 3-3 HPSI Flow Rate versus RCS Pressure RCS Pressure Loop 1A Loop 1B Loop 2A Loop 2B (psia) (gpm) (gpm) (gpm) (gpm) 1063.1 0.0 0.0 0.0 0.0 1062.6 10.6 10.6 10.6 10.6 1062.0 21.3 21.3 21.3 21.3 1045.8 31.9 31.9 31.9 31.9 1009.7 42.5 42.5 42.5 42.5 954.3 53.1 53.1 53.1 53.1 883.3 63.8 63.8 63.8 63.8 800.7 74.4 74.4 74.4 74.4 708.4 85.0 85.0 85.0 85.0 603.7 95.6 95.6 95.6 95.6 476.6 106.3 106.3 106.3 106.3 307.6 116.9 116.9 116.9 116.9 148.1 124.3 124.3 124.3 124.3 125.5 125.2 125.2 125.2 125.2 117.9 125.5 125.5 125.5 125.5 94.3 126.4 126.4 126.4 126.4 59.9 127.7 127.7 127.7 127.7 12.4 129.3 129.3 129.3 129.3 0.0 129.8 129.8 129.8 129.8

'%.oUI ILI LJII'U LJIJ5.A.l I I-,I IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summarv Renort Pa a 3-12 Table 3-4 LPSI Flow Rate versus RCS Pressure RCS Cold Leg Intact Loop I Intact Loop 2 Intact Loop 3 Broken Loop Pressure 1A 1B 2A 2B (psia) (gpm) (gpm) (gpm) (gpm) 125.5 0.0 0.0 0.0 0.0 117.9 0.0 0.0 240.0 240.0 94.3 0.0 0.0 560.0 560.0 59.9 0.0 0.0 880.0 880.0 12.4 0.0 0.0 1200.0 1200.0 0.0 0.0 0.0 1267.2 1267.2

%-eUI ILI VIIlI;U LJUIJLUI I IC-I IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Paae 3-13 Figure 3-1 S-RELAP5 SBLOCA Reactor Coolant System Nodalization

k-vVI ILU UlJII;U L/UI~UI I ICI IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Paae 3-14 Figure 3-2 S-RELAP5 SBLOCA Secondary System Nodalization

~.~#UI flu ~~ii~u LJU~.,UI I I~I IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Paae 3-15 Figure 3-3 S-RELAP5 SBLOCA Reactor Vessel Nodalization

%-APIILI 'JII.;U LJ/UtU1 I ICI It AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Paqe 3-16 Figure 3-4 Axial Power Shape

%..PUI ILI URIIU LJU4.oJI I IUI IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page 4-17 4.0 ANALYTICAL RESULTS The analysis results demonstrate the adequacy of the ECCS to support the criteria given in 10 CFR 50.46(b)(1-4) (Reference 3) for St. Lucie Unit 2 operating with AREVA supplied 16x16 HTP fuel with M5 cladding.

Section 4.1 describes the SBLOCA break spectrum for the cold leg break. Section 4.2 describes the event for the limiting break. Section 4.3 describes the sensitivity study for a break in the attached piping.

4.1 Results for Break Spectrum The St. Lucie Unit 2 cold leg pump discharge break spectrum analysis for SBLOCA includes breaks of varying diameter up to 10% of the flow area for the cold leg. The spectrum includes a wide enough range of break sizes from 2.0 inch diameter to 9.49 inch diameters to establish a PCT trend. Additional break sizes are performed with a smaller break interval once the potential limiting break size is determined to confirm the limiting break size.

The results for the break spectrum calculations are presented in Table 4-1. Time sequence evolution for each break size analysis is reported in Table 4-2. Figure 4-1 shows the calculated PCTs as a function of break size. The limiting break size was determined to be a 2.70-inch diameter (0.040-ft 2) break.

4.2 Discussion of Transient for Limiting Break The results for the limiting break size (2.70-inch) are shown in Figure 4-2 through Figure 4-16. The following discussion pertains to the limiting case.

The primary pressure decreased immediately after break initiation (Figure 4-3). When the primary pressure reached the low pressurizer pressure trip setpoint at 28 seconds the reactor is tripped (Figure 4-2). Within approximately 1 second after reactor trip, reactor SCRAM occurs, which occurs coincident with the loss of offsite power, causing turbine to trip, RCP trip and MFW pump trip (Figure 4-7 and Figure 4-8). As MFW to the

%oJI ILI VJIIt:U LJUk.jUI I IUI IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page 4-18 SGs is ramped down and steam flow to the main steam isolation valves (MSIVs) are closed, the pressure in the SGs increase for approximately 10 seconds until MSSV inlet reaches the lowest opening pressure setpoint. This provides core heat removal in the early stages of the transient.

The primary system depressurization continues at a relatively fast rate for the first 200 seconds. Significant inventory is lost out the break until the break transitions from expelling liquid to steam at approximately 600 seconds (Figure 4-4 and Figure 4-5).

Around this time, the broken leg loop seal begins to clear and the remaining intact loops remain plugged (Figure 4-6).

Prior to loop seal clearing, the core uncovers below top of active fuel (Figure 4-15).

Since there is no loop flow, a large amount of steam is generated and accumulated in the core by the decay heat power until enough pressure is built to blow the upflow leg of the loop seal in the broken leg at 596 seconds into the transient. This causes an abrupt level drop in the downcomer region and a small core level recovery. As the broken leg clears, the plant then enters a fairly slow boil-off phase where mass is lost out the break, and the primary system continues to empty.

From approximately 1000 to 1400 seconds, the primary side pressure is only slightly higher than the secondary side pressure (Figure 4-3), limiting the primary side heat removal. Broken loop steam generator AFW is initiated at approximately 936 seconds (Figure 4-9), providing an increasing SG level (Figure 4-10). However, the small pressure difference between the primary and secondary side limits the AFW's ability to remove heat from the primary side. At approximately 1500 seconds, the primary pressure reduces below the secondary side pressure, ending heat removal by the secondary side.

Meanwhile, the HPSI system becomes available at 68 seconds into the transient but does not began to inject water into the primary system cold legs until 664 seconds into the transient (Figure 4-9). However, HPSI flow does not provide sufficient inventory at this time to offset the large amount of RCS inventory lost out the break. As effective

%-,U1ILI Ublt=:U LJULoUl 1 R51 IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page 4-19 cooling is lost in the core, and the fuel rods begin to heat up at approximately 1350 seconds (Figure 4-16). Hot rod rupture occurs at 2004 seconds. The fuel continues to heat up until the maximum PCT of 1926°F is reached at 2200 seconds. SIT injection begins just prior to reaching the PCT at 2194 seconds, providing sufficient cooling to turn the PCT over. The PCT occurs at approximately 4 inches below the top of active fuel.

For this break size, although LPSI is available, it did not inject (Figure 4-12) due to the slow primary system depressurization. The continued supply of HPSI supports an increasing reactor vessel inventory (Figure 4-14) through event termination.

In conclusion, the limiting PCT break spectrum case is a 2.70-inch diameter cold leg pump discharge break. The PCT of this case is 1926°F. The maximum local oxidation is less than 6% and the maximum core-wide oxidation is less than 0.2%. The total maximum local oxidation is less than 9%, including a pre-transient oxidation of 2.3925%.

4.3 Attached Piping Sensitivity Study Although breaks in the SIT attached piping are not typically PCT limiting, they do result in reduced ECCS flows available to mitigate the event since one SIT inventory spills to containment, in addition to the reduced SI flow due to single failure assumption.

Therefore, an analysis of the limiting break size and location in an attached piping was performed. For St. Lucie Unit 2, the limiting break location and size for an attached piping is considered a double-ended guillotine break of an SIT line (10.126-inch diameter). The break was located in the SIT line connected to Loop 2B.

The calculated PCT was 1451'F, which is less limiting than the maximum PCT of the break spectrum analysis. The maximum local oxidation thickness is 0.1404% and the maximum core-wide oxidation thickness is 0.0017% which are both less limiting than the maximum obtained in the break spectrum analysis. The minimal HPSI and LPSI flows in the analysis are sufficient to prevent a subsequent heatup after the initial quench from the SIT discharge.

%.,UIILI UIIVU LJULoUI I It:I IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summqrv Renort Paae 4-20 Table 4-1 Summary of SBLOCA Break Spectrum Results Break diameter (in) 2.00 2.50 2.60 2.70 2.80 3.00 Break Area (ft2) 0.022 0.034 0.037 0.040 0.043 0.049 Peak Clad Temperature ('F) 1587 1726 1814 1926 1896 1823 Time of PCT (sec) 3752 2615 2456 2200 1997 1659 Time of Rupture (sec) -- 2443 2219 2004 1848 1654 Transient Local Maximum Oxidation (%) 1.3586 3.6120 4.3352 5.7788 5.0864 3.5589 Total Local Maximum Oxidation (%)' 3.7511 6.0045 6.7277 8.1713 7.4789 5.9514 Core Wide Oxidation (%) 0.0955 0.1273 0.1382 0.1664 0.1388 0.0903 PCT Elevation (ft) 10.77 11.02 11.02 11.02 11.02 11.02 Break diameter (in) 3.50 4.00 4.50 5.00 6.00 7.00 Break Area (ft2) 0.067 0.087 0.110 0.136 0.196 0.267 Peak Clad Temperature (°F) 1694 1578 1283 1077 1017 1080 Time of PCT (sec) 1074 802 649 527 145 131 Time of Rupture (sec) 1057 -.......

Transient Local Maximum Oxidation (%) 1.2501 0.3964 0.0683 0.0126 0.0065 0.0116 Total Local Maximum Oxidation (%)1 3.6426 2.7889 2.4608 2.4051 2.3990 2.4041 Core Wide Oxidation (%) 0.0335 0.0121 0.0011 0.0002 0.0001 0.0002 PCT Elevation (ft) 10.77 10.52 10.27 10.27 10.02 10.02 Break diameter (in) 8.00 9.00 9.49 Break Area (ft 2) 0.349 0.442 0.491 Peak Clad Temperature (°F) 1459 1469 1520 Time of PCT (sec) 169 148 136 Time of Rupture (sec) -. -_

Transient Local Maximum Oxidation (%) 0.1539 0.1569 0.1621 Total Local Maximum Oxidation (%)l 2.5464 2.5494 2.5546 Core Wide Oxidation (%) 0.0017 0.0021 0.0026 PCT Elevation (ft) 10.02 10.02 10.02 1 Includes the Pre-transient Oxidation of 2.3925%.

AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Paoe 4-21 Table 4-2 Sequence of Events for the SBLOCA Break Spectrum 3:- - M U) > U o

C -o C

0) 0 o L'E w w

w0 17 350.52 3.. 63 4 0 M 70 749. . w E 0 2 02 0)0

ý0 C 4( o L 63 0 0 E; 21 .a 0 C_ _) 0 E 2.70 0.

192) 28 o00 2964 28 3 6. 56 08 3 19 21 3.00 183 0 22 24 62 16a I3 7. 0 E 1 U 0 0 (n- LL 01 LL0 a) M 0 C M N1 C0 0 13 4 3 266 . .. 0 C. 0.C. C10 0C 2 0 06 64 - 6 0 00 0 04 0 0 0 00

-j -j

-J~~

- ZJ Mz 2.00 1587 0 50 52 63 93 93 1114 90- - 1090 1272 - 3368 -- 3752 2386 -

2.50 1726 3.0 19I 0 3271 34 43555573 73 782 1-.. -- 930 - - 670 8 782 4 2748 1064 2218 068 2443 0.7 2615 107 1510 208 0.-

2.60 1814 0 30 32 40 70 70 724 - 932 - - 626 724 2452 2100 2219 2456 378 -

2.70 1926 0 28 29 38 68 68 664 - 936 -- - - 596 668 2194 1978 2004 2200 354 -

2.80 1896 0 26 28 35 65 65 604 - 948 - - - 566 634 1992 1854 1848 1997 328 -

3.00 1823 0 22 24 32 62 62 536 -1668 -- - - 504 564 1652 1624 1654 1659 288 -

3.50 1694 0 17 19 25 55 55 412 1 - 6 8 70380 444 1064 1068 1057 1074 208 -

4.00 1578 0 13 15 215151 58 2 322 1 - 296 306 896 902 - 802 162 -

4.50 1283 0 11 1317 47 47266 - - - -260 242 258 64 0 644 - 649 128 -

5.00 1077 0 10 1215 45 45222 - 208 198 216 514 518 - 527 102 -

6.00 1017 0 8 10 1 34343 146 - - - 142 138 160 326 330 - 145 76 -

7.00 1080 0 18 10 12 42 142 1104 - I- - I- 112 1104-1 128 232 236 -- 131 54 -

8.00 1459 0 8 9 11 41 41 68 - - 100 102 - 80 106 164 166 - 169 48 -

9.00 1469 0 7 9 11 41 41 52 198 - 68 86 70 60 84 132 134 - 148 40 -

9.49 1520 0 7 9 10 40 40 46 142 - 158 68 64 56 82 1120 122 - 136 38 -

'U..AI ILl UOIC;U LJU.,UII I ICI IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page 4-22 Figure 4-1 Peak Cladding Temperature versus Break Size (SBLOCA Break Spectrum)

%AJ1OI UHIQ=U LJUtUl 1LA:;IIt~I AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page 4-23 Figure 4-2 Reactor Power - 2.70 inch Break 4000.0 3000.0

-- mReactor Power 2000.0 1000.0 0"0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

U.J I IL Uk)I ýZ,7' LJ U t-, U I I CýI AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page 4-24 Figure 4-3 Primary and Secondary System Pressures - 2.70 inch Break 2500.0 2250.0 2000.0 1750.0 - RV Upper Head

SG-1 1500.0 5SG-2 Z

1250.0

'A L*

CO 1000.0

A, 750.0 500.0 500.0 m ~, *UUmm*-*R * ~

* * -u **

250.0 0.0 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Tone (s)

1,-.* kJ I I I,ý UA ýt.",; LftJ k-e U I I AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Rennod Summa Report Pane 4-25 Figure 4-4 Break Void Fraction - 2.70 inch Break 1.0 IF -4F...

0.8 Void Fraction 0.6 0

U-2 0.4 0.2 U

0.0 0.0

=*a 1000.0 20O0.0 3000.0 4000.0 5000.0 Time (s)

AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary ReDort Paae 4-26 Figure 4-5 Break Mass Flow Rate - 2.70 inch Break 1000.0 800.0

-a Break Flow 600.0 P

'U U.

0 fa 400.0 20(0.0 L w-a= II. = tI- *-' = =.' = -==l *11t1..1* lIl 0.0-00 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

\',uI IL IKA UJUkUUI I MNIL~

AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary ReDort Paae 4-27 Figure 4-6 Loop Seal Void Fraction - 2.70 inch Break 1.0 0.8

- Loop 1A

-- Loop 1B a A Loop 2A 0.6

-- Loop 2B -

'0.4 0.2 0.0.

0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

k-IfkIII U111=1U it L)Utw1! I MRZZI AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Siimm~rv Rennrt Paace 4-28 Summarv Report Pane 4-28 Figure 4-7 RCS Loop Mass Flow Rate - 2.70 inch Break 10000.0 - - r Loop 1A 5000.0 Loop 1B Loop 2A

--ALoop 2B - broken

'U IL 0.0

-5000.0 0.0 --

1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

I I LtU I ý U "L) k-U I IU1= I L AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary ReDort Paae 4-29 Figure 4-8 Main Feedwater Mass Flow Rate - 2.70 inch Break 3000.0

Loop 1 2000.0 . Loop 2 - broken cc U-10I00.0

~0.0 0,0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

k~uýý ýL)I K,:u L-,%U LA'.U Nl I L AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary ReDort Paae 4-30 Figure 4-9 Auxiliary Feedwater Mass Flow Rate - 2.70 inch Break 40.0 F

~1

,-*-* , , . . .... .~. *-** . .4 *4 *-*--4--* 4-4--4-*-* 4- 4*-4 +

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- Loop 2 - broken 30.0 Cc 20.0 11 1 4

10.0 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

.1 "U _,LA I I It- IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summarv Rpnort Paae 4-31 Figure 4-10 Steam Generator Total Mass - 2.70 inch Break 250000.0 Loop 1

Loop 2- broken -

200000.0 -

e-

-i I 7 150000.0 4,

-II*'"I-IH * -***** f *- -*

Il~lu-:*I 1000000 t 000.0 2000.0 3000.0 4000.0 5000.0 0.0 Time (s)

k- fILI -~ZtJU'L IIC AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page 4-32 Figure 4-11 High Pressure Safety Injection Mass Flow Rates - 2.70 inch Break 20.0 15.0 Loop 1A

. Loop 1B

/, ..

Loop 2A Loop 2B - broken CC 10.0 5.0 /

0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

k-,U , I~LI ka , Yý, )ý- . !L ; I rU AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Suimm~rv R~nodt Paaie 4-33 Summa Report Pane 4-33 Figure 4-12 Low Pressure Safety Injection Mass Flow Rates - 2.70 inch Break 1.0 0.5

- Loop 1A

Loop 18

, Loop 2A

-- Loop 2B - broken Cc0 ,

0o

-0.5

-1.0 3000.0 0.0 1000.0 2000.0 4000.0 5000.0 Time (s)

k.,UI~~'LU L) UL I , k::,,

AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summarv ReDort Pace 4-34 Figure 4-13 Safety Injection Tank Mass Flow Rates - 2.70 inch Break 150.0 100.0

. Loop 1A

-- Loop 1B Loop 2A C_ -4 Loop 2B - b 6ken C 50.0 0-11 0.0o -u~41.-Na4m q

-50.0 1- 4000.0 ,5000.0 0.0 1000.0 2000.0 3000.0 Time (s)

I I L: L L 1.,:

AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Suimmarv Rennrt Paae 4-35 Su marv Report Paae 4-35 Figure 4-14 Reactor Vessel Mass Inventory - 2.70 inch Break 200000.0 a Mass 15o000.0 iU

- -a

*"0~ , a *W f- ,

100000.0 50000.0 L 0.0 I1000.0 200.0 3000.0 4000.0 5000.0 Time (s)

AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary ReDort Paae 4-36 Figure 4-15 Hot Assembly Mixture Level - 2.70 inch Break 12.0

~1 a 0-9 a 10.0 fat'~U-Lu..,.

8.0 Mixture Level 6.0 4.0 L___

04.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

k-OUI ILI UIJ t2U LJ IA I 1I7- A AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summarv Rennrt Paaqe 4-37 Figure 4-16 Hot Spot Cladding Temperature - 2.70 inch Break 2

000,0.r---- -

U

// \

II 1500.0 I a

- hot rod node 45 @11.02 ft

/

,/

p 1000.0 U r E

12 i

UU~3 U U S U 500.0

U U-U U U *U-g *~* * *~

0.0 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

%..jfUI ILl L1lIU LJULo.UI I I'I IL AREVA Inc. ANP-3345NP Revision 0 St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report Page 5-1

5.0 REFERENCES

1. AREVA Inc. Document EMF-2328(P)(A) Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2001.

2. AREVA Inc. Document BAW-10240(P)(A) Revision 0, Incorporation of M5 Properties in Framatome ANP Approved Methods, May 2004.

3. Code of Federal Regulations, Title 10, Part 50, Section 46, Acceptance Criteria For Emergency Core Cooling Systems For Light-Water Nuclear Power Reactors, January 2010.

4. AREVA Inc. Topical Report XN-NF-81-58(P)(A) Revision 2, Supplements 1 and 2, RODEX2 FUEL Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.

5. AREVA Inc. Topical Report ANF-81-58(P)(A) Revision 2, Supplements 3 and 4, RODEX2 FUEL Thermal-Mechanical Response Evaluation Model, Advanced Nuclear Fuels Corporation, June 1990.

6. Code of Federal Regulations, Title 10, Part 50, Appendix K, ECCS Evaluation Models, March 2000.

%..#IILI UIMJIU LJULwUI I MII IL A

AREVA ANP-3346NP St. Lucie Unit 2 Fuel Transition Revision 0 Realistic Large Break LOCA Summary Report December 2014 AREVA Inc.

(c) December 2014 AREVA Inc.

'&1..J IU .IIUU LU UI Iu ItII AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page i Nature of Changes Section(s)

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

%_JUV[ LI U11 "Z:U "ULUl IIM ;IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page ii Contents 1A.e

1.0 INTRODUCTION

...................................................................... 1-1 2 .0 S UMMA R Y ........................................................................................................ 2 -1 3 .0 A NA LY S IS ......................................................................................................... 3 -1 3.1 Description of the LBLOCA Event ........................................................... 3-2 3.2 Description of Analytical Models ............................................................. 3-4 3.3 Plant Description and Summary of Analysis Parameters ........................ 3-7 3.4 SER Compliance .................................................................................... 3-8 3.5 Realistic Large Break LOCA Results ...................................................... 3-8

4.0 CONCLUSION

S ................................................................................................ 4-1 5.0 GENERIC SUPPORT FOR TRANSITION PACKAGE ...................................... 5-1 5.1 Reactor Power ........................................................................................ 5-1 5.2 Rod Quench ............................................................................................ 5-2 5.3 Rod-to-Rod Thermal Radiation ............................................................... 5-2 5.3.1 Assessment of Rod-to-Rod Radiation Implicit in the RLBLOCA Methodology ............................................................... 5-4 5.3.2 Quantification of the Impact of Thermal Radiation using R2RRAD Code ................................................................... 5-5 5.3.3 Rod-to-Rod Radiation Summary .................................................. 5-9 5.4 Film Boiling Heat Transfer Limit ............................................................ 5-10 5.5 Downcomer Boiling ............................................................................... 5-10 5.5.1 W all Heat Release Rate ............................................................. 5-13 5.5.2 Exact Solution ............................................................................ 5-14 5.5.3 Plant Model Sensitivity Study ..................................................... 5-16 5.5.4 Downcomer Fluid Distribution .................................................... 5-20 5.5.5 Azimuthal Nodalization ............................................................... 5-20 5.5.6 Axial Nodalization ....................................................................... 5-21 5.5.7 Downcomer Boiling Conclusions ................................................ 5-27 5 .6 Bre a k S ize ............................................................................................ 5 -2 8 5.6.1 Break / Transient Phenomena ................................................... 5-28 5.6.2 New Minimum Break Size Determination ................................... 5-29 5.6.3 Intermediate Break Size Disposition .......................................... 5-32 5.7 Detail information for Containment Model ............................................. 5-39 5.8 Cross-References to North Anna .......................................................... 5-42

%..AJI[ ILI VIJIIU LJ_/UI~ I 1t;I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page iii 5.9 GDC 35 - LOOP and No-LOOP Case Sets .......................................... 5-43 5.10 Input V ariables Statem ent ..................................................................... 5-44 6.0 RECENT NRC REQUEST FOR ADDITIONAL INFORMATION (RA I) AND AREVA RESPO NSES ..................................................................... 6-1 6.1 Thermal Conductivity Degradation - Once-Burned Fuel ......................... 6-1 6.2 Decay Heat Treatm ent .......................................................................... 6-14 6.3 Thermal Conductivity Degradation - Swelling, Rupture, and R e lo ca tio n ............................................................................................. 6 -2 1 6.4 Oxidation - Pre-transient and Single-Sided .......................................... 6-22 6.6 Single Failure A ssum ption .................................................................... 6-24 6 .8 C o re Liquid Leve l .................................................................................. 6-3 1 7 .0 R E F E R E N C E S .................................................................................................. 7-1

'~II IU JICU LJUUULI I IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paae iv List of Tables Table 2-1 Summary of Major Parameters for Limiting Transient ................................. 2-2 Table 3-1 Sampled LBLOCA Parameters ................................................................. 3-10 Table 3-2 Plant Operating Range Supported by the LOCA Analysis ........................ 3-11 Table 3-3 Statistical Distributions Used for Process Parameters .............................. 3-14 Table 3-4 SER Conditions and Limitations ................................................................ 3-15 Table 3-5 Summary of Results for the Limiting PCT Case ........................................ 3-17 Table 3-6 Calculated Event Times for the Limiting PCT Case .................................. 3-17 Table 3-7 Heat Transfer Parameters for the Limiting Case ....................................... 3-18 Table 3-8 Containment Initial and Boundary Conditions ........................................... 3-19 Table 3-9 Passive Heat Sinks in Containment .......................................................... 3-20 Table 5-1 Typical Measurement Uncertainties and Local Peaking Factors ................. 5-5 Table 5-2 FLECHT-SEASET & 17x17 FA Geometry Parameters ............................... 5-6 Table 5-3 FLECHT-SEASET Test Parameters ........................................................... 5-8 Table 5-4 Minimum Break Area for Large Break LOCA Spectrum ............................ 5-31 Table 5-5 Minimum PCT Temperature Difference - True Large and Intermediate B re a ks ................................................................................................... 5 -3 3 Table 6-1: Bounding Once-Burned Fuel Power Ratios by Burnup ............................ 6-13 Table 6-2: Local Maximum Oxidation Results ........................................................... 6-23

ILI UlJIIU LJU~oUI I IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paqe v List of Figures Figure 3-1 P rim ary S ystem N oding ........................................................................... 3-2 1 Figure 3-2 Secondary System Noding ....................................................................... 3-22 Figure 3-3 R eactor V essel Noding ............................................................................ 3-23 Figure 3-4 C o re Nod ing D eta il................................................................................... 3-24 Figure 3-5 Upper Plenum Noding Detail .................................................................... 3-25 Figure 3-6 Scatter Plot of Operational Parameters ................................................... 3-26 Figure 3-7 PCT versus PCT Time Scatter Plot from 59 Calculations ........................ 3-28 Figure 3-8 PCT versus Break Size Scatter Plot from 59 Calculations ....................... 3-29 Figure 3-9 Maximum Oxidation versus PCT Scatter Plot from 59 Calculations ........ 3-30 Figure 3-1i ) Total Oxidation versus PCT Scatter Plot from 59 Calculations .............. 3-31 Figure 3-11 Peak Cladding Temperature (Independent of Elevation) for the Lim iting C a se ........................................................................................ 3-3 2 Figure 3-12 Break Flow for the Limiting Case ........................................................... 3-33 Figure 3-13 3 Core Inlet Mass Flux for the Limiting Case ............................................ 3-34 Figure 3-14 I. Core Outlet Mass Flux for the Limiting Case .......................................... 3-35 Figure 3-1i 5 Void Fraction at RCS Pumps for the Limiting Case ............................... 3-36 Figure 3-1 E5 ECCS Flows (Includes SIT, LPSI and HPSI) for the Limiting Case ........ 3-37 Figure 3-1 71 Upper Plenum Pressure for the Limiting Case ....................................... 3-38 Figure 3-1 E3 Collapsed Liquid Level in the Downcomer for the Limiting Case .......... 3-39 Figure 3-1 E Collapsed Liquid Level in the Lower Plenum for the Limiting Case ....... 3-40 Figure 3-2( ) Collapsed Liquid Level in the Core for the Limiting Case ...................... 3-41 Figure 3-21 Containment and Loop Pressures for the Limiting Case ........................ 3-42 Figure 3-22 GDC 35 LOOP versus No-LOOP Cases ................................................ 3-43 Figure 5-1 R2RRAD 5 x 5 Rod Segment ..................................................................... 5-6 Figure 5-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 F A ........................................................................................................... 5-8 Figure 5-3 Reactor Vessel Downcomer Boiling Diagram .......................................... 5-12 Figure 5-4 S-RELAP5 versus Closed Form Solution ................................................. 5-15 Figure 5-5 Downcomer Wall Heat Release - Wall Mesh Point Sensitivity ................ 5-16 Figure 5-6 PCT Independent of Elevation - Wall Mesh Point Sensitivity .................. 5-17 Figure 5-7 Downcomer Liquid Level - Wall Mesh Point Sensitivity ........................... 5-18

ljI.JI ILI UIJIIU UUJ..IIo~ I IUI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page vi Figure 5-8 Core Liquid Level - Wall Mesh Point Sensitivity ...................................... 5-19 Figure 5-9 Azim uthal Noding ..................................................................................... 5-2 1 Figure 5-10 Lower Compartment Pressure versus Time ........................................... 5-23 Figure 5-11 Downcomer Wall Heat Release - Axial Noding Sensitivity Study .......... 5-24 Figure 5-12 PCT Independent of Elevation - Axial Noding Sensitivity Study ............ 5-25 Figure 5-13 Downcomer Liquid Level - Axial Noding Sensitivity Study .................... 5-26 Figure 5-14 Core Liquid Level - Axial Noding Sensitivity Study ................................ 5-27 Figure 5-15 Plant A - Westinghouse 3-Loop Design ................................................ 5-34 Figure 5-16 Plant B - Westinghouse 3-Loop Design ................................................ 5-35 Figure 5-17 Plant C - Westinghouse 3-Loop Design ................................................ 5-36 Figure 5-18 Plant D - Combustion Engineering 2x4 Design .................................... 5-37 Figure 5-19 Plant E - Combustion Engineering 2x4 Design .................................... 5-38 Figure 5-20 Plant F - Westinghouse 3-loop Design ................................................. 5-39 Figure 5-21 PCT vs. Containment Volume ............................................................... 5-40 Figure 5-22 PCT vs. Initial Containment Temperature ............................................. 5-41 Figure 5-23 Containment Pressure versus Time for the Limiting Case .................... 5-42 Figure 6-1: Fractional Fuel Centerline Temperature Delta between RODEX3A a n d D a ta ................................................................................................. 6-8 Figure 6-2: Fuel Centerline Temperature Delta of RODEX3A Calculations to Data (Original and Using the New Correlation) ....................................... 6-9 Figure 6-3: Correction Factor (as applied for temperatures in Kelvin) ....................... 6-10 Figure 6-4: Radial Temperature Profile for Hot Rod .................................................. 6-11 Figure 6-5: Temperature versus Time for Fuel Centerline, Clad Surface, and F ue l Ave ra g e ........................................................................................ 6 -12 Figure 6-6: Fresh and Once-Burned U0 2 Rod PCT Transient at the Hot Spot ....... 6-13 Figure 6-7 Decay Heat Comparisons, Infinite Operation U235, Finite Operation A ll Isotopes (0.1 - 10 sec) .................................................................... 6-17 Figure 6-8 Decay Heat Comparisons, Infinite Operation U235, Finite Operation A ll Isotopes (10 - 1000 sec) ................................................................. 6-18 Figure 6-9 Decay Heat Ratios, Finite Operation over Infinite Operation for U235, A ll Isotopes (0 - 10 sec) ....................................................................... 6-19 Figure 6-10 Decay Heat Ratios, Finite Operation over Infinite Operation for U235, All Isotopes (0 - 600 sec) ........................................................... 6-20 Figure 6-11: Clad Temperature Response from Single Failure Study ....................... 6-26

%,,UI ILIUIICLJ LJUV ,Ul IIICI IIL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page vii Figure 6-12: PCT Independent of Elevation .............................................................. 6-27 Figure 6-13: Containment and System Pressure ...................................................... 6-28 Figure 6-14: ECCS Mass Flow Rates ........................................................................ 6-29 Figure 6-15: Dow ncom er Level ................................................................................. 6-30 Figure 6-16: Reactor Vessel Fluid Mass, Limiting Case ............................................ 6-32

uIL fuuiiCU L-JULUI I IUI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page viii Nomenclature Acronym Definition AFW Auxiliary feedwater AOR Analysis of record AREVA AREVA Inc.

ASI Axial Shape Index BOC Beginning-of-cycle CCFL Counter Current Flow Limitation CCTF Cylindrical Core Test Facility CE Combustion Engineering CFR Code of Federal Regulations CHF Critical Heat Flux DC-HL Downcomer- Hot Leg DC-UH Downcomer- Upper Head DEGB Double Ended Guillotine Break DNB Departure from Nucleate Boiling ECCS Emergency Core Cooling System EFPH Effective Full Power Hours EM Evaluation model EOC End-of-cycle EOP Emergency Operating Procedure EPU Extended Power Uprate FP&L Florida Power & Light GDC General Design Criteria HEM Homogeneous Equilibrium Model HPSI High pressure safety injection HTC Heat Transfer Coefficient HTP High Thermal Performance LBLOCA Large Break Loss of Coolant Accident LHR Linear heat rate LOCA Loss-of-coolant accident LOFT Loss of Fluid Test LOOP Loss of Offsite Power

%-,UI ILI 'JIICL1 L-/U%,UI I IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page ix Nomenclature Acronym Definition LPSI Low pressure safety injection MFW Main feedwater MSIV Main Steam Isolation Valve MSSV Main steam safety valve No-LOOP No Loss of Offsite Power NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System PCT Peak cladding temperature PIRT Phenomena Identification and Ranking Table PLHGR Peak Linear Heat Generation Rate PWR Pressurized water reactor RAI Request for Additional Information RCP Reactor coolant pump RCS Reactor Coolant System RLBLOCA Realistic Large Break Loss of Coolant Accident RV Reactor Vessel RWST Reactor Water Storage Tank SBLOCA Small break loss-of-coolant accident SER Safety Evaluation Report SG Steam generator SI Safety injection SIAS Safety injection actuation signal SIT Safety injection tank SLB St. Lucie Unit 2 SNGLJUN Single Junction Component SS Steady State THTF Thermal Hydraulic Test Facility TMDPJUN Time Dependent Junction Component TMDPVOL Time Dependent Volume Component TWODEE Two Dimensional Volume Component W Westinghouse

%..JI Ul IIJIIC;U L/UkoUI I M;1 IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 1-1

1.0 INTRODUCTION

This report describes and provides results from a RLBLOCA analysis for the St. Lucie Plant Unit 2. The plant is a CE-designed 3020 MWt plant with a large dry containment.

The plant is a 2X4 loop design - two hot legs and four cold legs. The loops contain four RCPs, two U-tube steam generators and one pressurizer. The ECCS is provided by two independent safety injection trains and four SITs.

The analysis supports operation beginning with Cycle 23 of AREVA's HTP 16X16 fuel design using standard U0 2 fuel with 2%, 4%, 6% and 8% Gd 2 0 3 and M5 cladding and Zirc-4 structural materials, unless changes in the Technical Specifications, Core Operating Limits Report, core design, fuel design, plant hardware, or plant operation invalidate the results presented herein. The analysis was performed in compliance with the NRC-approved RLBLOCA EM (Reference 1) with exceptions noted below. Analysis results confirm the 10CFR50.46 (b) (Reference 19) acceptance criteria presented in Section 3.0 are met and serve as the basis for operation of the St Lucie Nuclear Plant Unit 2 with AREVA fuel.

The non-parametric statistical methods inherent in the AREVA RLBLOCA methodology provide for the consideration of a full spectrum of break sizes, break configuration (guillotine or split break), axial shapes, and plant operational parameters. A conservative loss of a diesel single failure assumption is applied in which one LPSI pump injects into the broken cold leg and intact cold leg of the broken loop, and one HPSI pump injects into all four cold legs. Regardless of the failure assumption, all containment pressure-reducing systems are assumed fully functional. The effects of Gadolinia-bearing fuel rods and peak fuel rod exposures are considered.

The following are differences from the approved RLBLOCA EM (Reference 1) that were requested by the NRC.

%...'LJI ILI VIJIIU LI-Jtl.UI I M-11I L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 1-2 The assumed reactor core power for the St Lucie Unit 2 realistic large break loss-of-coolant accident is 3029.06 MWt. This value represents the current rated thermal power of 3020 MWt plus 0.3% power measurement uncertainty.

The RLBLOCA analysis was performed with a version of S-RELAP5 that requires both the void fraction to be less than 0.95 and the clad temperature to be less than 900 OF before the rod is allowed to quench. This may result in a slight increase in PCT results when compared to an analysis not subject to these constraints.

The RLBLOCA analysis was performed with a version of S-RELAP5 that limits the contribution of the Forslund-Rohsenow model to no more than 15 percent of the total heat transfer at and above a void fraction of 0.9. This may result in a slight increase in PCT results when compared to previous analyses for similar plants.

The split versus double-ended break type is no longer related to break area. In concurrence with Regulatory Guide 1.157, both the split and the double-ended break will range in area between the minimum break area (Amin) and an area of twice the size of the broken pipe. The determination of break configuration, split versus double-ended, will be made after the break area is selected based on a uniform probability for each occurrence. Amin was calculated to be 29.9 percent of the DEGB area (see Section 5.6.2 for further discussion). This is not expected to have an effect on PCT results.

In concurrence with the NRC's interpretation of GDC 35, a set of 59 cases was run with a LOOP assumption and a second set with a No-LOOP assumption. The set of 59 cases that predicted the highest PCT is reported in Section 2 and Section 3, herein.

The results from both case sets are shown in Figure 3-22. The effect on PCT results due to offsite power assumption effects is expected to be minor.

During recent RLBLOCA EM modeling studies, it was noted that cold leg condensation efficiency may be under-predicted. Water entering the DC post-SIT injection remained sufficiently subcooled to absorb DC wall heat release without significant boiling. However, tests (Reference 7) indicate that the steam and water entering the

%oUI ILI LJIIC;U L/L5L.oUI I 1:I It AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 1-3 DC from the cold leg, subsequent to the end of SIT injection, reach near saturation resulting from the condensation efficiency ranging between 80 to 100 percent. To assure that cold leg condensation would not be under-predicted, a RLBLOCA EM update was made. Noting that saturated fluid entering the DC is the most conservative modeling scheme, steam and liquid multipliers were developed so as to approximately saturate the cold leg fluid before it enters the DC. The multipliers were developed through scoping studies using a number of plant configurations-Westinghouse-designed 3- and 4-loop plants, and CE-designed plants. The results of the scoping study indicated that multipliers of 10 and 150 for liquid and steam, respectively, were appropriate to produce saturated fluid entering the DC. This RLBLOCA EM departure was recently discussed with the NRC and the NRC agreed that the approach described immediately above was satisfactory in the interim. The modification is implemented post-SIT injection, 10 seconds after the vapor void fraction in the bottom of the SIT becomes greater than 90 percent. Thus, the SITs have injected all their water into the cold legs, and the nitrogen cover gas has entered the system and been mostly discharged through the break before the condensation efficiency is increased by the factors of 10 and 150, for liquid and vapor respectively. Providing saturated fluid conditions at the DC entrance conservatively reduces both the DC driving head and the core flooding rate. Recall that test results indicate that fluid conditions entering the DC range from saturated to slightly subcooled. Hence, it is conservative to force an approximation of saturated conditions for fluid entering the DC.

AREVA has acknowledged an issue concerning fuel thermal conductivity degradation as a function of burnup as raised by the NRC. In order to manage this issue, AREVA is modifying the way RODEX3A temperatures are compensated in the RLBLOCA Revision 0/Transition package methodology. In the current process, the RLBLOCA computes PCTs at many different times during an operating cycle. For each specific time in cycle, the fuel conditions are computed using RODEX3A prior to starting the S-RELAP5 portion of the analysis. A steady state condition for the given time in cycle using S-RELAP5 is established. A base fuel centerline temperature is established in this process. Then two-transformation adjustments to the base fuel centerline temperature

.-#JVI ILI VIKi-U LJ L,,Ul I ICI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 1-4 are computed. The first transformation is a linear adjustment for an exposure of 10 Mwd/MtU or higher. The second adjustment is performed in the S-RELAP5 initialization process for the transient case. In the new process, a polynomial transformation is used for the first transformation instead of a linear transformation. The rest of the RLBLOCA process for initializing the S-RELAP5 fuel rod temperature is not altered and the rest of LOCA transient continues in the original fashion.

%jIJIILI OIJqI;U L-JUt.sUI I IMI It AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 2-1 2.0

SUMMARY

The limiting PCT analysis is based on the parameter specification given in Table 2-1.

The limiting PCT is 1732 OF for a fresh U0 2 Rod in a case with LOOP conditions. Both fresh and once-burned fuel were analyzed. Gadolinia bearing rods of 2, 4, 6, and 8%

were also analyzed, but were not found to be limiting. This RLBLOCA result is based on a case set of 59 individual transient cases for LOOP and 59 individual transient cases for No-LOOP conditions. The analysis considers a mixed hydraulic core of both AREVA and Westinghouse fuel.

The analysis assumed full core power operation at 3029.06 MWt. The value represents the current rated thermal power (3020 MWt) plus 0.3% power measurement uncertainty.

The analysis assumed a steam generator tube plugging level of 20 percent in all steam generators, a total LHR of 13.0 kW/ft (no axial dependency), a total peaking factor (FQ) up to a value of 2.504, and a nuclear enthalpy rise factor (FAH or FT) up to a value of 1.81 (including 6% uncertainty and 3.5% allowance for control rod insertion effect on an FrT of 1.65). This analysis bounds typical operational ranges or technical specification limits (whichever is applicable) with regard to pressurizer pressure and level; SIT pressure, temperature, and level; core inlet temperature; core flow; containment pressure and temperature; and RWST temperature. The analysis analyzes both fresh and once-burned fuel assemblies, valid for exposures up to 62,000 MWd/MTU. The analysis demonstrates that the 10 CFR 50.46(b) (Reference 19) criteria listed in Section 3.0 are satisfied.

%-.JU ILI U1JII;U LJ'.JL,UI I I1ZI L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summarv Reoort Paae 2-2 Table 2-1 Summary of Major Parameters for Limiting Transient Core Average Burnup (EFPH) 1 12365.52 Core Power (MWt) 2 3029 Hot Rod LHR, kW/ft 12.468 Total Hot Rod Radial Peak (FrT) 3 1.810 ASI -0.1456 Break Type Guillotine Break Size (ft2/side) 3.5510 Offsite Power Availability Not available Decay Heat Multiplier 4 1.0 1 For fresh fuel. The limiting hot rod is a fresh bundle.

2 Core power is not sampled in the analysis. The value includes 0.3% measurement uncertainty, such that the actual analyzed power is 3029.06 MWt.

3 Including 6% uncertainty and 3.5% allowance for control rod insertion effect on an FrT of 1.65. This parameter is not sampled in the analysis.

4 Decay heat is not sampled in the analysis.

ILI UOIC;U LJULoUl I IM DL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-1 3.0 ANALYSIS The purpose of the analysis is to verify typical technical specification peaking factor limits and the adequacy of the ECCS by demonstrating that the following 10CFR 50.46(b) (Reference 19) criteria are met:

(1) The calculated maximum fuel element cladding temperature shall not exceed 2200 OF.

(2) The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

(3) The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel excluding the cladding surrounding the plenum volume were to react.

(4) The calculated changes in core geometry shall be such that the core remains amenable to cooling.

The analysis did not evaluate core coolability due to seismic events, nor did it consider the 10CFR 50.46(b) long-term cooling criterion.

The RLBLOCA analysis conservatively considers blockage effects due to clad swelling and rupture in the prediction of the hot fuel rod PCT. AREVA has previously performed an analysis which demonstrates that for all cases of horizontal seismic and LOCA loads, the resulting loads are below the spacer grid elastic load limit and thus the grids sustain no permanent deformation.

Section 3.1 of this report describes the postulated LBLOCA event. Section 3.2 describes the models used in the analysis. Section 3.3 describes the 2X4-loop PWR plant and summarizes the system parameters used in the analysis. Compliance to the SER is addressed in Section 3.4. Section 3.5 summarizes the results of the RLBLOCA analysis.

k..#Ul ILI UIIUU I-Jut...ul I IM It AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-2 3.1 Description of the LBLOCA Event A LBLOCA is initiated by a postulated large rupture of the RCS primary piping. Based on deterministic studies, the worst break location is in the cold leg piping between the reactor coolant pump and the reactor vessel for the RCS loop containing the pressurizer. The break initiates a rapid depressurization of the RCS. A reactor trip signal is initiated when the low pressurizer pressure trip setpoint is reached; however, reactor trip is conservatively neglected in the analysis. The reactor is shut down by coolant voiding in the core.

The plant is assumed to be operating normally at full power prior to the accident. The cold leg break is assumed to open instantaneously. For this break, a rapid depressurization occurs, along with a core flow stagnation and reversal. This causes the fuel rods to experience DNB. Subsequently, the limiting fuel rods are cooled by film convection to steam. The coolant voiding creates a strong negative reactivity effect and core criticality ends. As heat transfer from the fuel rods is reduced, the cladding temperature increases.

Coolant in all regions of the RCS begins to flash. At the break plane, the loss of subcooling in the coolant results in substantially reduced break flow. This reduces the depressurization rate, and leads to a period of positive core flow or reduced downflow as the RCPs in the intact loops continue to supply water to the RV (in No-LOOP conditions). Cladding temperatures may be reduced and some portions of the core may rewet during this period. The positive core flow or reduced downflow period ends as two-phase conditions occur in the RCPs, reducing their effectiveness. Once again, the core flow reverses as most of the vessel mass flows out through the broken cold leg.

Mitigation of the LBLOCA begins when the SIAS is issued. This signal is initiated by either high containment pressure or low pressurizer pressure. Regulations require that a worst single-failure be considered. This single-failure has been determined to be the loss of one ECCS pumped injection train. The AREVA RLBLOCA methodology conservatively assumes an on-time start and normal lineups of the containment spray to conservatively reduce containment pressure and increase break flow. Hence, the

k.-VI ILO LJIIvL L/i..jUlIIIMH AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-3 analysis assumes that the loss of one emergency diesel generator, which takes one train of ECCS pumped injection out. LPSI injects into the broken loop and one intact loop, HPSI injects into all four loops, and all containment spray pumps are operating.

When the RCS pressure falls below the SIT pressure, fluid from the SITs is injected into the cold legs. In the early delivery of SIT water, high pressure and high break flow will drive some of this fluid to bypass the core. During this bypass period, core heat transfer remains poor and fuel rod cladding temperatures increase. As RCS and containment pressures equilibrate, ECCS water begins to fill the lower plenum and eventually the lower portions of the core; thus, core heat transfer improves and cladding temperatures decrease.

Eventually, the relatively large volume of SIT water is exhausted and core recovery continues relying solely on pumped ECCS injection. As the SITs empty, the nitrogen gas used to pressurize the SITs exits through the break. This gas release may result in a short period of improved core heat transfer as the nitrogen gas displaces water in the downcomer. After the nitrogen gas has been expelled, the ECCS temporarily may not be able to sustain full core cooling because of the core decay heat and the higher steam temperatures created by quenching in the lower portions of the core. Peak fuel rod cladding temperatures may increase for a short period until more energy is removed from the core by the HPSI and LPSI while the decay heat continues to fall. Steam generated from fuel rod rewet will entrain liquid and pass through the core, vessel upper plenum, the hot legs, the steam generators, and the reactor coolant pumps before it is vented out the break. The resistance of this flow path to the steam flow is balanced by the driving force of water filling the downcomer. This resistance may act to retard the progression of the core reflood and postpone core-wide cooling. Eventually (within a few minutes of the accident), the core reflood will progress sufficiently to ensure core-wide cooling. Full core quench occurs within a few minutes after core-wide cooling.

Long-term cooling is then sustained with LPSI pumped injection system.

~iIIl UIII4UL L-J~J'.Ul I Ie' It AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-4 3.2 Description of Analytical Models The RLBLOCA methodology is documented in EMF-2103 Realistic Large Break LOCA Methodology (Reference 1). The methodology follows the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach (Reference 2). This method outlines an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifies the uncertainties in a LOCA analysis.

The RLBLOCA methodology consists of the following computer codes:

RODEX3A for computation of the initial fuel stored energy, fission gas release, and fuel-cladding gap conductance.

S-RELAP5 for the system calculation (includes ICECON for containment response).

AUTORLBLOCATRN for generation of ranged parameter values, transient input, transient runs, and general output documentation.

The governing two-fluid (plus non-condensables) model with conservation equations for mass, energy, and momentum transfer is used. The reactor core is modeled in S-RELAP5 with heat generation rates determined from reactor kinetics equations (point kinetics) with reactivity feedback, and with actinide and decay heating.

The two-fluid formulation uses a separate set of conservation equations and constitutive relations for each phase. The effects of one phase on the other are accounted for by interfacial friction, and heat and mass transfer interaction terms in the equations. The conservation equations have the same form for each phase; only the constitutive relations and physical properties differ.

The modeling of plant components is performed by following guidelines developed to ensure accurate accounting for physical dimensions and that the dominant phenomena expected during the LBLOCA event are captured. The basic building blocks for modeling are hydraulic volumes for fluid paths and heat structures for heat transfer. In addition, special purpose components exist to represent specific components such as the RCPs or the steam generator separators. All geometries are modeled at the

%--'UlILI UIIq7,U LJLA.,UI I IM IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-5 resolution necessary to best resolve the flow field and the phenomena being modeled within practical computational limitations.

System nodalization details are shown in Figures 3-1 through 3-5. Note that in Figure 3-1, break modeling uses two junctions regardless of break type-split or guillotine; for guillotine breaks, Junction 151 is deleted, it is retained fully open for split breaks.

Hence, total break area is the sum of the areas of both break junctions.

A typical calculation using S-RELAP5 begins with the establishment of a steady-state initial condition with all loops intact. The input parameters and initial conditions for this steady-state calculation are chosen to reflect plant technical specifications or to match measured data. Additionally, the RODEX3A code provides initial conditions for the S-RELAP5 fuel models. Specific parameters are discussed in Section 3.3.

Following the establishment of an acceptable steady-state condition, the transient calculation is initiated by introducing a break into one of the loops (specifically, the loop with the pressurizer). The evolution of the transient through blowdown, refill and reflood is computed continuously using S-RELAP5. Containment pressure is also calculated by S-RELAP5 using containment models derived from ICECON (Reference 4), which is based on the CONTEMPT-LT code (Reference 3).

The methods used in the application of S-RELAP5 to the LBLOCA are described in Reference 1. A detailed assessment of this computer code was made through comparisons to experimental data, many benchmarks with cladding temperatures ranging from 1,700 OF (or less) to above 2,200 OF. These assessments were used to develop quantitative estimates of the ability of the code to predict key physical phenomena in a PWR LBLOCA. Various models-for example, the core heat transfer, the decay heat model and the fuel cladding oxidation correlation-are defined based on code-to-data comparisons and are, hence, plant independent.

The RV internals are modeled in detail (Figures 3-3 through 3-5) based on St Lucie Unit 2 specific inputs. Nodes and connectivity, flow areas, resistances and heat structures are all accurately modeled. The location of the hot assembly/hot pin(s) is unrestricted;

%\,UIOIl U1IOU LJUk..~lJ I 11ZI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-6 however, the channel is always modeled to restrict appreciable upper plenum liquid fallback.

The final step of the best-estimate methodology is to combine all the uncertainties related to the code and plant parameters, and estimate the PCT at a high probability level. The steps taken to derive the PCT uncertainty estimate are summarized below:

1. Base Plant Input File Development First, base RODEX3A and S-RELAP5 input files for the plant (including the containment input file) are developed. Code input development guidelines are applied to ensure that model nodalization is consistent with the model nodalization used in the code validation.

2. Sampled Case Development The non-parametric statistical approach requires that many "sampled" cases be created and processed. For every set of input created, each "key LOCA parameter" is randomly sampled over a range established through code uncertainty assessment or expected operating limits (provided by plant technical specifications or data). Those parameters considered "key LOCA parameters" are listed in Table 3-1. This list includes both parameters related to LOCA phenomena (based on the PIRT provided in Reference 1) and to plant operating parameters.

3. Determination of Adequacy of ECCS The RLBLOCA methodology uses a non-parametric statistical approach to determine values of PCT at the 95 percent probability level. Total oxidation and total hydrogen are based on the limiting PCT case. The adequacy of the ECCS is demonstrated when these results satisfy the criteria set forth in Section 3.0.

f-lUl L UII1OJ LJUtol ie11 IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-7 3.3 Plant Description and Summary of Analysis Parameters The plant analysis presented in this report is for a CE-designed PWR, which has 2X4-loop arrangement. There are two hot legs each with a U-tube steam generator and four cold legs each with a RCP 1. The RCS includes one pressurizer connected to a hot leg.

The core contains 217 thermal-hydraulic compatible AREVA HTP 16X16 fuel assemblies with 2%, 4%, 6% and 8% gadolinia pins. The ECCS includes one HPSI, one LPSI and one SIT injection path per RCS loop. The break is modeled in the same loop as the pressurizer, as directed by the RLBLOCA methodology. The RLBLOCA transients are of sufficiently short duration that the switchover to sump cooling water for ECCS pumped injection need not be considered.

The S-RELAP5 model explicitly describes the RCS, RV, pressurizer, and ECCS. The ECCS includes a SIT path and a LPSI/HPSI path per RCS loop. The HPSI and LPSI feed into the SIT line that connects to each cold leg pipe downstream of the RCP discharge. The ECCS pumped injection is modeled as a table of flow versus backpressure. This model also describes the secondary-side steam generator that is instantaneously isolated (closed MSIV and feedwater trip) at the time of the break. A symmetric steam generator tube plugging level of 20 percent per steam generator was assumed.

As described in the AREVA RLBLOCA methodology, many parameters associated with LBLOCA phenomenological uncertainties and plant operation ranges are sampled. A summary of those parameters is given in Table 3-1. The LBLOCA phenomenological uncertainties are provided in Reference 1. Values for process or operational parameters, including ranges of sampled process parameters, and fuel design parameters used in the analysis are given in Table 3-2. Plant data are analyzed to develop uncertainties for the process parameters sampled in the analysis. Table 3-3 presents a summary of the uncertainties used in the analysis. Where applicable, the The homologous pump performance curves were input to the S-RELAP5 plant model; the built-in S-RELAP5 curves were not used.

O~jJ LE VIIt;U LJULAw~I I IUI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-8 sampled parameter ranges are based on technical specification limits or supporting plant calculations that provide more bounding values.

For the AREVA RLBLOCA EM, dominant containment parameters, as well as NSSS parameters, were established via a PIRT process. Other model inputs are generally taken as nominal or conservatively biased. The PIRT outcome yielded two important (relative to PCT) containment parameters, containment pressure and temperature. In many instances, the conservative guidance of CSB 6-2 (Reference 5) was used in setting the remainder of the containment model input parameters. As noted in Table 3-3, containment temperature is a sampled parameter. Containment pressure response is indirectly ranged by sampling the containment volume (Table 3-3). Containment heat sink data is given in Table 3-9. In accordance with Reference 1, the condensing heat transfer coefficient is intended to be closer to a best-estimate instead of a bounding high value. A [ ] Uchida heat transfer coefficient multiplier was also applied to the analysis of St. Lucie Unit 2.

The initial conditions and boundary, conditions are given in Table 3-8. The building spray is modeled at maximum heat removal capacity. All spray flow is delivered to the containment.

3.4 SER Compliance A number of requirements on the methodology are stipulated in the conclusions section of the SER for the RLBLOCA methodology (Reference 1). The requirements have been addressed during the application of the methodology. Table 3-4 shows that compliance to the SER restriction has been demonstrated.

3.5 Realistic Large Break LOCA Results Two case sets of 59 transient calculations were performed sampling the parameters listed in Table 3-1. For each case set, PCT was calculated for a U0 2 rod, and for Gadolinia-bearing rods with concentrations of 2, 4, 6 and 8 w/o Gd 2 0 3 , for both fresh and once-burned exposure. The limiting case set, that contained the PCT, was the set with no offsite power available. The limiting PCT (1732 OF) occurred in Case 57 for the

%-.*IILI VIIIVU LJUkosJI IUI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-9 fresh U0 2 rod. The major parameters for the limiting transient are presented in Table 2-1. Table 3-5 lists the results of the limiting case. The fraction of total hydrogen generated was not directly calculated; however, it is conservatively bounded by the calculated total percent oxidation, which is well below the 1 percent limit. The best-estimate PCT case is Case 30, which corresponded to the median case out of the 59-case set with no offsite power available. The nominal PCT was 1496 OF, for an 8%

Gd 2 0 3 once-burned rod. This result can be used to quantify the relative conservatism in the limiting case result. In this analysis, it was 236 OF.

The case results, event times and analysis plots for the limiting PCT case are shown in Table 3-5, Table 3-6, and in Figure 3-11 through Figure 3-21. Figure 3-6 shows linear scatter plots of the key parameters sampled for the 59 calculations. Parameter labels appear to the left of each individual plot. These figures show the parameter ranges used in the analysis. Figure 3-7 and Figure 3-8 show the time of PCT and break size versus PCT scatter plots for the 59 calculations, respectively. Figure 3-9 and Figure 3-10 show the maximum oxidation and total oxidation versus PCT scatter plots for the 59 calculations, respectively. Key parameters for the limiting PCT case are shown in Figure 3-11 through Figure 3-21. Figure 3-11 is the plot of PCT independent of elevation; this figure shows that the transient exhibits a sustained and stable quench. A comparison of PCT results from both case sets is shown in Figure 3-22.

'%..eV Oto IjIIUU L/ukulsIIMI1III AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-10 Table 3-1 Sampled LBLOCA Parameters Phenomenological Time in cycle (peaking factors, axial shape, rod properties, burnup)

Break type (guillotine versus split)

Critical flow discharge coefficients (break)

Critical flow discharge coefficients (surgeline)

Initial upper head temperature Film boiling heat transfer Dispersed film boiling heat transfer Critical heat flux Tmin (intersection of film and transition boiling)

Initial stored energy Downcomer hot wall effects Condensation interphase heat transfer Metal-water reaction Plant1 2

Offsite power availability Break size Pressurizer pressure Pressurizer liquid level SIT pressure SIT liquid level SIT temperature (based on containment temperature)

Containment temperature Containment volume Initial RCS flow rate Initial operating RCS temperature Uncertainties for plant parameters are based on typical plant-specific data with the exception of "Offsite power availability," which is a binary result that is specified by the analysis methodology.

2 Not sampled, see Section 4.9.

fl VII;U L./UkULJ IJI I IUI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary ReDort Paae 3-11 Table 3-2 Plant Operating Range Supported by the LOCA Analysis Event Operating Range 1.0 Plant Physical Description 1.1 Fuel a) Cladding outside diameter 0.382 in.

b) Cladding inside diameter 0.332 in.

c) Cladding thickness 0.025 in.

d) Pellet outside diameter 0.3255 in.

e) Pellet density 95.35 percent of theoretical f) Active fuel length 136.7 in.

g) Resinter densification I I h) Gd 20 3 concentrations 2, 4, 6, 8 w/o 1.2 RCS a) Flow resistance Analysis b) Pressurizer location Analysis assumes location giving most limiting PCT (broken loop) c) Hot assembly location Anywhere in core d) Hot assembly type 16X16 AREVA HTP with M5 cladding and Zirc-4 structural materials e) SG tube plugging 20 percent' 2.0 Plant Initial Operating Conditions 2.1 Reactor Power a) Nominal reactor power 3029.06 MWt 2 b) LHR 13.0 kW/ft 3 c) Fg 2.504 3 d) Fr 1.810 4 2.2 Fluid Conditions a) Loop flow 138.9 Mlbm/hr < M < 161.8 Mlbm/hr b) RCS Cold Leg temperature 535.0 OF _< T _<554.0 °F c) Pressurizer pressure 2180 psia _<P

2320 psia d) Pressurizer level 60.0 percent < L

71.0 percent In the RLBLOCA analysis, only the maximum 20% tube plugging in each steam generator was analyzed. [

2]

2 Includes 0.3% uncertainty.

3 FQ is based on LHR. F 0 is sampled for each case from a lower bound (a case-specific maximum peaking factor determined from the power history data), to an upper bound value of 2.504, The radial power peaking for the hot rod is including 6% measurement uncertainty and 3.5%

allowance for control rod insertion affect. This value is not sampled in the uncertainty analysis.

Fr tech spec*(l+ uncertFr) * (1+uncert cr insertion) = 1.65*(1.0+0.06)*(1+0.035)=1.810

%-AJI ILI lUJlI;U LJUt.,UI I I-,I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Renort Pane 3-12 Table 3-2 Plant Operating Range Supported by the LOCA Analysis (Continued)

Event Operating Range e) SIT pressure 499.7 psia < P < 679.7 psia 3

f) SIT liquid volume 1388 ft3 < V

1588 ft g) SIT temperature 80.5 OF < T _<124.5 OF (coupled with containment temperature) h) SIT resistance fL/D As-built piping configuration i) Minimum ECCS boron Ž1800 ppm 3.0 Accident Boundary Conditions a) Break location Cold leg pump discharge piping b) Break type Double-ended guillotine or split c) Break size (each side, relative to cold leg 0.299 _<A < 1.0 full pipe area (split) pipe area) 0.299 _<A < 1.0 full pipe area (guillotine) d) Worst single-failure Loss of one emergency diesel generator e) Offsite power On or Off f) ECCS pumped injection temperature 104 OF g) HPSI pump delay 20.0 (w/ offsite power) 30.0 (w/o offsite power) h) LPSI pump delay 20.0 (w/ offsite power) 30.0 (w/o offsite power) i) Containment pressure 13.78 psia j) Containment temperature 80.5 OF _<T < 124.5 OF k) Containment sprays delay Os I) LPSI flow Broken Loop RCS press flow

_sia 0.0 1267.2 12.4 1200.0 59.9 880.0 94.3 560.0 117.9 240.0 125.5 0.0 Intact Loop 1 RCS press flow psia qpr 0.0 1267.2 12.4 1200.0 59.9 880.0 94.3 560.0 117.9 240.0 125.5 0.0 Intact Loop 2 RCS press flow oia gp~m_

0.0 0.0 12.4 0.0 59.9 0.0 94.3 0.0 117.9 0.0 125.5 0.0

%..-UI1L1U1It5U LJUL0UI1IC;I1L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-13 Table 3-2 Plant Operating Range Supported by the LOCA Analysis (Continued)

Event Operating Range Intact Loop 3 RCS press flow Psia gp-0.0 0.0 12.4 0.0 59.9 0.0 94.3 0.0 117.9 0.0 125.5 0.0 m) HPSI flow For each Loop, RCS press flow psia

_pm_

0.0 129.8 12.4 129.3 59.9 127.7 94.3 126.4 117.9 125.5 125.5 125.2 148.1 124.3 307.6 116.9 476.6 106.3 603.7 95.6 708.4 85.0 800.7 74.4 883.3 63.8 954.3 53.1 1009.7 42.5 1045.8 31.9 1062.6 10.6 1063.1 0.0

%-.*II1LE UIlCLJ L.RUL~lI I IC~I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Reoort Paae 3-14 Table 3-3 Statistical Distributions Used for Process Parameters 1 Operational Parameter Uncertainty Parameter Range Distribution Pressurizer Pressure (psia) Uniform 2180-2320 Pressurizer Liquid Level (percent) Uniform 60.0 - 71.0 3

SIT Liquid Volume (ft ) Uniform 1388 - 1588 SIT Pressure (psia) Uniform 499.7 - 679.7 Containment Temperature (*F) Uniform 80.5 - 124.5 Containment Volume (ft3) Uniform 2.493E+6 - 2.631 E+6 Initial RCS Flow Rate (Mlbm/hr) Uniform 138.9 - 161.8

.Initial RCS Operating Temperature Uniform 535.0- 554.0 (Tcolg) (°F)

RWST Temperature for ECCS (°F) Point 104 Offsite Power Availability2 Binary 0,1 Delay for Containment Spray (s) Point 0 20.0 (w/ offsite power)

LPSI Pump Delay (s) Point 30.0 (w/ offsite power) 30.0 (w/o offsite power)

HPSI Pump Delay (s) Point 20.0 (w/ offsite power) 30.0 (w/o offsite power) 1 Note that core power and decay heat are not sampled.

2 This is not a sampled parameter. One set of 59 cases is run with LOOP and one set of 59 cases is run with No-LOOP.

kA*U ILI U~IM5U LJIJI.JUI I IUI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Reoort Paae 3-15 Table 3-4 SER Conditions and Limitations SER Conditions and Limitations Response

1. A CCFL violation warning will be added to alert the analyst There was no significant occurrence of CCFL to CCFL violation in the downcomer should such occur. violation in the downcomer for this analysis.

Violations of CCFL were noted in a statistically insignificant number of time steps.

2. AREVA has agreed that it is not to use nodalization with Hot leg nozzle gaps were not modeled.

hot leg to downcomer nozzle gaps.

3. If AREVA applies the RLBLOCA methodology to plants The PLHGR for St Lucie Unit 2 is lower than using a higher planar linear heat generation rate (PLHGR) that used in the development of the RLBLOCA than used in the current analysis, or if the methodology is EM (Reference 1). An end-of-life calculation to be applied to an end-of-life analysis for which the pin was not performed. However, an assessment pressure is significantly higher, then the need for a for St. Lucie Unit 2 against rupture criteria was blowdown clad rupture model will be reevaluated. The performed, which concluded that for the St.

evaluation may be based on relevant engineering Lucie Unit 2 RLBLOCA analysis, cladding experience and should be documented in either the rupture prior to the initiation of reflood does not RLBLOCA guideline or plant specific calculation file. occur for either the first or second cycle fuel.

4. Slot breaks on the top of the pipe have not been evaluated. The AREVA PWR analysis guidelines provide These breaks could cause the loop seals to refill during late detailed discussion on the generic treatment of reflood and the core to uncover again. These break top slot breaks. For St. Lucie Unit 2 the locations are an oxidation concern as opposed to a PCT elevation of the top of active fuel is below the concern since the top of the core can remain uncovered for elevation of the top of the crossover leg piping extended periods of time. Should an analysis be performed that would be susceptible to a top slot break.

for a plant with spillunder (Top crossover pipe (ID) at the Thus no top slot break LOCA analysis is crossover pipes lowest elevation) that are below the top needed, and no additions to the calculation elevation of the core, AREVA will evaluate the effect of the notebook or Design Report are required.

deep loop seal on the slot breaks. The evaluation may be based on relevant engineering experience and should be documented in either the RLBLOCA guideline or plant-specific calculation file.

5. The model applies to 3 and 4 loop Westinghouse- and St Lucie Unit 2 is a CE-designed 2X4 loop CE-designed nuclear steam systems. plant.

6. The model applies to bottom reflood plants only (cold side St Lucie Unit 2 is a bottom reflood plant.

injection into the cold legs at the reactor coolant discharge piping).

7. The model is valid as long as blowdown quench does not The case set was examined and blowdown occur. If blowdown quench occurs, additional justification quench was not an issue in the St. Lucie Unit 2 for the blowdown heat transfer model and uncertainty are RLBLOCA uncertainty analysis.

needed or the calculation is corrected. A blowdown quench is characterized by a temperature reduction of the peak cladding temperature (PCT) node to saturation temperature during the blowdown period.

8. The reflood model applies to bottom-up quench behavior. The CCFL model is applied on all core exit If a top-down quench occurs, the model is to be justified or junctions as a provision to prevent top-down corrected to remove top quench. A top-down quench is quench. No top-down quench effects are characterized by the quench front moving from the top to observed in the St. Lucie Unit 2 RLBLOCA the bottom of the hot assembly. uncertainty analysis.

%.,oJIOki IUlIqU L-JUJLoJ I I~I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summarv Reoort Paae 3-16 Table 3-4 SER Conditions and Limitations (Continued)

SER Conditions and Limitations Response

9. The model does not determine whether Criterion 5 of 10 Long-term cooling was not evaluated in this CFR 50.46, long term cooling, has been satisfied. This will analysis.

be determined by each applicant or licensee as part of its application of this methodology.

10. Specific guidelines must be used to develop the The nodalization in the plant model is similar with plant-specific nodalization. Deviations from the reference the Westinghouse 3-loop sample calculations plant must be addressed. that were submitted to the NRC for review with a slight deviation in the 2x4 loop, the upper plenum, the upper head and the downcomer nodalizations. This deviation should not impact the acceptability of the nodalization. The nodalization used for the St. Lucie Unit 2 RLBLOCA analysis is consistent with oether licensed plant models for CE 2x4 plants. Figure 3-1 to Figure 3-5 show the nodalization of the plant.

11. A table that contains the plant-specific parameters and the Simulation of clad temperature response is a range of the values considered for the selected parameter function of phenomenological correlations that during the topical report approval process must be have been derived either analytically or provided. When plant-specific parameters are outside the experimentally. The important correlations have range used in demonstrating acceptable code been validated for the RLBLOCA methodology performance, the licensee or applicant will submit and a statement of the range of applicability has sensitivity studies to show the effects of that deviation, been documented. The correlations of interest are the set of heat transfer correlations as described in Reference 1. Table 3-7 presents the summary of the full range of applicability for the important heat transfer correlations, as well as the ranges calculated in the limiting case of this analysis. Calculated values for other parameters of interest are also provided. As is evident, the plant-specific parameters fall within the methodology's range of applicability.

12. The licensee or applicant using the approved methodology Analysis results are discussed in Section 3.5.

must submit the results of the plant-specific analyses, including the calculated worst break size, PCT, and local and total oxidation.

13. The licensee or applicant wishing to apply AREVA realistic FPL will request an exemption for use of M5 clad large break loss-of-coolant accident (RLBLOCA) fuel.

methodology to M5 clad fuel must request an exemption for its use until the planned rulemaking to modify 10 CFR 50.46(a)(i) to include M5 cladding material has been completed.

%.jLJI ILI IJICeU L.JIJLUi I It:I 1L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summarv Renort Paae 3-17 Table 3-5 Summary of Results for the Limiting PCT Case Case # 57 Limiting Hot Rod (offsite power unavailable) (fresh U0 2 rod)

PCT Temperature 1732 OF Time 105.6 s Elevation 9.784 ft Metal-Water Reaction*

Percent Oxidation Maximum 2.2802%

Percent Total Oxidation 0.0139%

Pre-transient oxidation is addressed in Section 6.4.

Table 3-6 Calculated Event Times for the Limiting PCT Case Event Time (s)

Break Opened 0.0 RCP Trip N/A SIAS Issued 1.2 Start of Broken Loop SIT Injection 7.7 Start of Intact Loop SIT Injection 13.4, 13.5, and 13.3 (Loops 2, 3 and 4 respectively)

Broken Loop LPSI Delivery Began 31.2 Intact Loop LPSI Delivery Began (Loops 2, 3 and 4 respectively)

Broken Loop HPSI Delivery Began 31.2 Intact Loop HPSI Delivery Began 31.2, 31.2, and 31.2 (Loops 2, 3 and 4 respectively)

Beginning of Core Recovery (Beginning of 30.2 Reflood) 30.2 Broken Loop SIT Emptied 75.8 Intact Loop SITs Emptied 80.6, 75.4, and 76.2 (Loops 2, 3 and 4 respectively)

PCT Occurred 105.6 Transient Calculation Terminated 551.9

%.,7UI ILI U11CU L.YULoUl I IM IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Reoort Pane 3-18 Table 3-7 Heat Transfer Parameters for the Limiting Case1 7-

%JeJIILO LJiit:U L-JUt..1I 1U1 IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary ReDort Paoe 3-19 Table 3-8 Containment Initial and Boundary Conditions Containment Net Free Volume (ft 3) 2.493E+6 ft3 - 2.631 E+6 ft3 Initial Conditions Containment Pressure (nominal) 13.78 psia Containment Temperature 80.5 OF - 124.5 OF Outside Temperature 26 OF Humidity 1.0 Containment Spray Number of Pumps operating 2 Spray Flow Rate (Total, both pumps) 9,000 gpm Minimum Spray Temperature 46 OF Fastest Post-LOCA initiation of spray 0s Containment Fan Coolers Number of Fan Coolers Operating 4 Minimum Post Accident Initiation Time 0 of Fan Coolers (sec)

Fan Cooler Capacity (1 Fan Cooler)

Containment Temperature (F) Heat Removal Rate (BTU/sec) 60 0 120 3167 200 14722 280 27639

%joJIIII UJIIýU L-JUL9UI IIlI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary ReDort Paae 3-20 Table 3-9 Passive Heat Sinks in Containment1 Heat Sink Area (ft2) Thickness (ft) Material Containment Primary Cylinder 58695 0.16 Carbon Steel Containment Primary Dome 32340 0.08 Carbon Steel Ground Floor 6020.0 2.0 Concrete Shield Walls and Pads 67225.0 2.0 Concrete Containment Sump 25093.0 3.0 Concrete Stainless Steel - Embedded 8715.0 0.01563 Stainless Steel Stainless Steel - piping and other comp. 16399.0 0.04249 Stainless Steel Galvanized Steel - conduits and cable trays 165823.0 0.00924 Galvanized Steel Structure and Misc. Exposed Steel 111802.0 0.06166 Carbon Steel Thermal Conductivity Volumetric Heat Capacity Material (BTU/hr-ft-_F) (BTU/ft 3 OF)

Concrete 1.0 34.2 Carbon Steel 25.9 53.57 Stainless Steel 9.8 54.0 Galvanized Steel 64.0 40.6 Passive heat sinks data listed in the table were used for RLBLOCA analysis. Sensitivity studies were previously performed for the AREVA RLBLOCA Transition Package as applied to EMF-2103 to respond to the NRC's concerns. The results showed for a large dry containment, the PCT is not sensitive to a change in containment back pressure. Hence, the heat sinks changes within +/-5% range will not change the presented RLBLOCA

I I.~LJI l ZIU L.JUk.sUI I ICI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-21 Figure 3-1 Primary System Noding

%U.JI ILE UIRLu LJULLUI I IM IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-22 Figure 3-2 Secondary System Noding

%--rJIILI U11t;U L/UtLUI I IM~ IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Reoort Paae 3-23 Figure 3-3 Reactor Vessel Noding

AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-24 Figure 3-4 Core Noding Detail

%JoiILI LJIIC;J L/J~Uto I -IIL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summarv Renort Paae 3-25 Figure 3-5 Upper Plenum Noding Detail

~.AI LI Ujmieu Ljukol I MI1I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summarv ReDort Paae 3-26 Figure 3-6 Scatter Plot of Operational Parameters One-Sided J

I

.. 'om se me mm m Break Area F (fe/side) 0 1.0 2.0 3.0 4.0 5AC Bum Trne (hours) 0.0()e+00 5.6"e+3 I .6(e+04 1.50e+04 Core Power (MW)

LHGR (KW/ft) 13.0 ASI I

.2

-03.2 osmem L

-0.1

=so m

0 0.0 i

M MOsMm 0I 0.1 s

0.

I

.2 Pressurizer Pressure emnmmeme - ommO (psia) 21550.0 2200.0 2250.0 2300.0 235 0.0 Pressurizer _

Liquid Level mo mmm i some

(%)

60.0 65.0 70.0 75.0 RCS (Tcold) L I0.0 I5.0 I0.

Temperature N o mm m m mm o me soo (F) 535.0 540.0 545.0 550.0 555.0

M VIlk6:;U L-/UtwUI I lUl It AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Reoort Paae 3-27 Figure 3-6 Scatter Plot of Operational Parameters (Continued)

Total Loop Flow Oma m see i N (Mlb/hr) 130.0 140.0 150.0 160.0 170.0 SIT Liquid _ ' '

Volume god.. mmms (if) 1 SIT Pressure I oý

inmm OOeNs m (psia) L 500.0 550.0 600.0 650.0 700.0 Containment L Volume

m meO eme a 2.45e+06 2.50e+06 2.55e-706 2.60e+06 2.65e+06 SIT Temperature a*

(8F) 80.0 90.0 100.0 110.0 120.0 130.0

' 0 ~ 'IJI ILl IJlI~U LJUL~UI I I~I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paae 3-28 Figure 3-7 PCT versus PCT Time Scatter Plot from 59 Calculations 2000I I 1800 F El 1600 El 0

1400 1 F1

- 1200 El El 1000 800 N Split Break El Guillotine Break 600 400 0 100 200 300 400 500 Time of PCT (s)

%'-evIILI LJIIcA Liutlul I 0IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paae 3-29 Figure 3-8 PCT versus Break Size Scatter Plot from 59 Calculations 2000 1800 nI 1600 I ME L F-U 0 r-1 F-1 El El El F-]~ r-Mi 1400 F-LI 0

" 1200 U 0

El 1-1 1000 800 600 0 Split Break El Guillotine Break 400 1.0 2.0 3.0 4.0 5.0 Break Area (ft2/side)

%.#lILI V1iieU LJ-l~UkoI 1IC; It AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paae 3-30 Figure 3-9 Maximum Oxidation versus PCT Scatter Plot from 59 Calculations 3.0 I I I I I I 2.8 0 Split Break El Guillotine Break 2.6 2.4 2.2 2.0 F 1.8 0

,. 1.6

.2 1.4 X

0 1.2 1.0 0 0.8

E E 0.6 0.4 Er- 01 0.2 0.0 400 600 800 1000 1200 1400 1600 1800 2000 PCT (0F)

k~.jI f1l1U11tiie LjtJUI.A.I I ;I U AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary ReDort Paae 3-31 Figure 3-10 Total Oxidation versus PCT Scatter Plot from 59 Calculations 0.10 0.08 0.06 C

-0 0

0 0.04 0.02 0.00 L 400 1200 1400 1600 PCT (0F)

LO UOiZ;U LULUU PAI I IMD fL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-32 Figure 3-11 Peak Cladding Temperature (Independent of Elevation) for the Limiting Case PCT = 1731.4 OF, at Time = 105.59 s, on Fresh U02 Rod 2000 1500 LL a) 4)

C.

E 1000 I-0.

(D) 0 0 200 400 600 Time (s)

ý--UJI II HJiC UUUU1.sLI 1M L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Repnort Paae 3-33 Figure 3-12 Break Flow for the Limiting Case 90 F - Vessel Side

Pump Side

- - - Total 70 50 30 U- 10

-10

-30 F

-50 0 200 400 600 Time (s)

k fUI ILI U1cA LJUOUI I IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-34 Figure 3-13 Core Inlet Mass Flux for the Limiting Case 1000 Hot Assembly Surround Assembly

-- - Average Core Outer Core 500 E

x 00 (I

1 $

-500 0 200 400 600 Time (s)

ILOULJICU LjukdlsU I IMI R AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Rerort Paae 3-35 Figure 3-14 Core Outlet Mass Flux for the Limiting Case 900 - Hot Assembly

Surround Assembl Average Core Outer Core 700 F 500 F E 300 F xe X

0-100 1 I

-100 F

-300 F

-500 0 200 400 600 Time (s)

ý,AUI OLD UJHIZLJ L-JLJUUI I a L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-36 Figure 3-15 Void Fraction at RCS Pumps for the Limiting Case 1.0 II II II 0.8 II 0.6 o

IL I 0.4 Broken Loop 1 Intact Loop 2 0.2 --- Intact Loop 3

-- - Intact Loop 4 0.0 0 200 400 600 Time (s)

%.AUI ILI LUIICU LJLJoUl I IFZI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summarv Renort Paae 3-37 Figure 3-16 ECCS Flows (Includes SIT, LPSI and HPSI) for the Limiting Case 3000 2000 0

w 0

1000 0

0 200 400 600 Time (s)

ýLj'ul OLEU81tIkA L-JujuuI 1W OL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-38 Figure 3-17 Upper Plenum Pressure for the Limiting Case 3 0 00 r I 2000 U)

.0 a-1000 0

0 200 400 600 Time (s)

%-#UI ILI WiIICA LJUkL.Ufl0II IFL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-39 Figure 3-18 Collapsed Liquid Level in the Downcomer for the Limiting Case 30 Sector 1 (broken)

............ Sector 2

-- Sector 3 Sector 4 Average 20 10 II7 0

0 200 400 600 Time (s)

OLSJ fUW=giU LJUoUI E~1t;E AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-40 Figure 3-19 Collapsed Liquid Level in the Lower Plenum for the Limiting Case 10 8

6 4

2 0

0 200 400 600 Time (s)

%fUiL1UOZ UILALJLJOUI BMEL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 3-41 Figure 3-20 Collapsed Liquid Level in the Core for the Limiting Case 15 Hot Assembly Center Core

- - - Average Core Outer Core 10 a)I

13 02 0 200 400 600 Time (s)

%-9Ul OLE UICJIU IJlJI-LII 0 M;1 OL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary ReDort Paae 3-42 Figure 3-21 Containment and Loop Pressures for the Limiting Case 100 90 80 70 60 (A

50 CA 40 30 20 10 0

0 0 200 400 600 Time (s)

E~J LI 1JIIzl.J LJULUUI I IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary ReDort Paae 3-43 Figure 3-22 GDC 35 LOOP versus No-LOOP Cases 200) 2000

LOOP oNo LOOP 1800 1800 U 0 1600 1600

'Q r"}

0 # 0 0 30

s *

0 0 *

000

- 0 0 30000 0 1400 00 u [ 1400 B 0 00 0 0 0 (3 a 000 1200 1200 00 00 1000 1000 00 800 800 600 600 0 10 20 30 40 50 60 Case Number

%,.j!J ILI UHUIU LUltoAI I ICI It AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Pagqe 4-1

4.0 CONCLUSION

S A RLBLOCA analysis was performed for the St Lucie Nuclear Plant Unit 2 using NRC -

approved AREVA RLBLOCA methods (Reference 1). Analysis results show that the limiting LOOP case has a PCT of 1732 'F, and a maximum oxidation thickness and hydrogen generation that fall well within regulatory requirements.

The analysis supports operation at a nominal power level of 3029.06 MWt (including 0.3% uncertainty), a steam generator tube plugging level of up to 20 percent in all steam generators, a total LHR of 13.0 kW/ft, a total peaking factor (FQ) up to a value of 2.504, and a nuclear enthalpy rise factor (FAH or FrT) up to a value of 1.81 (including 6%

uncertainty and 3.5% allowance for control rod insertion effect on an FrT of 1.65) with no axial or burnup dependent power peaking limit. Peak rod average exposures of up to 62,000 MWd/MTU are supported, with the analysis considering both fresh and once-burned fuel assemblies as potentially limiting.

For large break LOCA, the 10CFR50.46 (b) (Reference 19) criteria presented in Section 3.0 are met and operation of St Lucie Unit 2 with AREVA supplied 16x16 M5 clad fuel is justified.

T

' LI UIqJIU LJU*Ul IIIZIIL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-1 5.0 GENERIC SUPPORT FOR TRANSITION PACKAGE The following sections are responses to typical RAI questions posed by the NRC on EMF-2103 Revision 0 plant applications. In some instances, these requests cross-referenced documentation provided on dockets other than those for which the request is made. AREVA discussed these and similar questions from the NRC draft SER for Revision 1 of EMF-2103 in a meeting with the NRC on December 12, 2007. AREVA agreed to provide the following additional information within new submittals of a Realistic Large Break LOCA report.

5.1 Reactor Power Question:

It is indicated in the RLBLOCA analyses that the assumed reactorcore power "includes uncertainties." The use of a reactor power assumption other than 102 percent, regardless of BE or Appendix K methodology, is permitted by Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix K.I.A, "Required and Acceptable Features of The Evaluation Models, 'Sources of Heat During a LOCA." However, Appendix K. I.A also states: "... An assumed power level lower than the level specified in this paragraph [1.02 times the licensed power level], (but not less than the licensed power level) may be used provided..."Please explain.

Response

As indicated in Item 2.1 of Table 3-2 herein, the assumed reactor core power for the St Lucie Unit 2 Realistic Large Break Loss-of-coolant Accident is 3029.06 MWt. The value represents the current rated thermal power (3020 MWt) plus 0.3% power measurement uncertainty.

%_e~JI ILI LJII;U LJUt~stoI I IMI It AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-2 5.2 Rod Quench Question:

Does the version of S-RELAP5 used to perform the computer runs assure that the void fraction is less than 95 percent and the fuel cladding temperature is less than 900 OF before it allows rod quench?

Response

Yes, the version of S-RELAP5 employed for the St Lucie Unit 2 requires that both the void fraction is less than 0.95 and the clad temperature is less than the minimum temperature for film boiling heat transfer (Tmin) before the rod is allowed to quench. T"min is a sampled parameter in the RLBLOCA methodology that typically does not exceed 755 K (900 'F). This is a change to the approved RLBLOCA EM (Reference 1). This feature is carried forward into the UJUN14 version of S-RELAP5.

5.3 Rod-to-Rod Thermal Radiation Question:

Providejustification that the S-RELAP5 rod-to-rod thermal radiationmodel applies to the St Lucie Unit 2 core.

Response

The Realistic LBLOCA methodology, (Reference 1), does not provide modeling of rod-to-rod radiation. The fuel rod surface heat transfer processes included in the solution at high temperatures are: film boiling, convection to steam, rod to liquid radiation and rod to vapor radiation. This heat transfer package was assessed against various experimental data sets involving both moderate (1600 OF - 2000 OF) and high (2000 OF to over 2200 OF) peak cladding temperatures and shown to be conservative when applied nominally. The normal distribution of the experimental data was then determined. During the execution of an RLBLOCA evaluation, the heat transferred from a fuel rod is determined by the application of a multiplier to the nominal heat transfer model. This multiplier is determined by a random sampling of the normal distribution of the experimental data benchmarked. Because the data include the effects of rod-to-rod

%_,,fIIL1ULJIIU LJLJtsUI I I-I 1L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paqe 5-3 radiation, it is reasonable to conclude that the modeling implicitly includes an allocation for rod-to-rod radiation effects. As will be demonstrated, the approach is reasonable because the conditions within actual limiting fuel assemblies assure that the actual rod-to-rod radiation is larger than the allocation provided through normalization to the experiments.

The FLECHT-SEASET tests evaluated covered a range of PCTs from 1,651 to 2,239 OF and the THTF tests covered a range of PCTs from 1,000 to 2,200 OF. Since the test bundle in either FLECHT-SEASET or THTF is surrounded by a test vessel, which is relatively cool compared to the heater rods, substantial radiation from the periphery rods to the vessel wall can occur. The rods selected for assessing the RLBLOCA reflood heat transfer package were chosen from the interior of the test assemblies to minimize the impact of radiation heat transfer to the test vessel. The result was that the assessment rods comprise a set which is primarily isolated from cold wall effects by being surrounded by powered rods at reasonably high temperatures.

As a final assessment, three benchmarks independent of THTF and FLECHT-SEASET were performed. These benchmarks were selected from the Cylindrical Core Test Facility (CCTF), LOFT, and the Semiscale facilities. Because these facilities are more integral tests and together cover a wide range of scale, they also serve to show that scale effects are accommodated within the code calculations.

The results of these calculations are provided in Section 4.3.4, Evaluation of Code Biases, page 4-100, of Reference 1. The CCTF results are shown in Figures 4.180 through 4.192, the LOFT results in Figures 4.193 through 4.201, and the Semiscale results in Figures 4.202 through 4.207. As expected, these figures demonstrate that the comparison between the code calculations and data is improved with the application of the derived biases. The CCTF, LOFT, and Semiscale benchmarks further indicate that, whatever consideration of rod-to-rod radiation is implicit in the S-RELAP5 reflood heat transfer modeling, it does not significantly effect code predictions under conditions where radiation is minimized. The measured PCTs in these assessments ranged from approximately 1,000 to 1,540 OF. At these temperatures, there is little rod-to-rod

%..OUIILI UIIq7;U I-jut-oul I ICI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-4 radiation. Given the good agreement between the biased code calculations and the CCTF, LOFT, and Semiscale data, it can be concluded that there is no significant over prediction of the total heat transfer coefficient.

Notwithstanding any conservatism evidenced by experimental benchmarks, the application of the model to commercial nuclear power plants provides some additional margins due to limitations within the experiments. The benchmarked experiments, FLECHET SEASET and ORNL Thermal Hydraulic Test Facility (THTF), used to assess the S-RELAP5 heat transfer model, Reference 1, incorporated constant rod powers across the experimental assembly. Temperature differences that occurred were the result of guide tube, shroud or local heat transfer effects. In the operation of a pressurized water reactor (PWR) and in the RLBLOCA evaluation, a radial local peaking factor is present, creating power differences that tend to enhance the temperature differences between rods. In turn, these temperature differences lead to increases in net radiation heat transfer from the hotter rods. The expected rod-to-rod radiation will likely exceed that embodied within the experimental results.

5.3.1 Assessment of Rod-to-Rod Radiation Implicit in the RLBLOCA Methodology As discussed above, the FLECHT-SEASET and THTF tests were selected to assess and determine the S-RELAP5 code heat transfer bias and uncertainty. Uniform radial power distribution was used in these test bundles. Therefore, the rod-to-rod temperature variation in the rods away from the vessel wall is caused primarily by the variation in the sub-channel fluid conditions. In the real operating fuel bundle, on the other hand, there can be 5 to 10 percent rod-to-rod power variation. In addition, the methodology includes a provision to apply the uncertainty measurement to the hot pin.

Table 5-1 provides the hot pin measurement uncertainty and a representative local pin peaking factor for several plants. These factors, however, relate the pin to the assembly average. To more properly assess the conditions under which rod-to-rod radiation heat transfer occurs, a more local peaking assessment is required. Therefore, the plant rod-to-rod radiation assessments herein set the average pin power for those

'*-..UI ILI UiI1LI LJUtosLI I MAI It AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-5 pins surrounding the hot pin at 96 percent of that of the peak pin. For pins further removed the average power is set to 94 percent.

Table 5-1 Typical Measurement Uncertainties and Local Peaking Factors Plant FAH Measurement Local Pin Peaking tUncertainty Factor (-)

(percent) 1 4.0 1.068 2 4.0 1.050 3 6.0 1.149 4 4.0 1.113 5 4.25 1.135 6 4.0 1.058 5.3.2 Quantification of the Impact of Thermal Radiation using R2RRAD Code The R2RRAD radiative heat transfer model was developed by Los Alamos National Laboratory (LANL) to be incorporated in the BWR version of the TRAC code. The theoretical basis for this code is given in References 8 and 11 and is similar to that developed in the HUXY rod heatup code (Reference 10, Section 2.1.2) used by AREVA for BWR LOCA applications. The version of R2RRAD used herein was obtained from the NRC to examine the rod-to-rod radiation characteristics of a 5x5 rod segment of the 161 rod FLECHT-SEASET bundle. The output provided by the R2RRAD code includes an estimate of the net radiation heat transfer from each rod in the defined array. The code allows the input of different temperatures for each rod as well as for a boundary surrounding the pin array. No geometry differences between pin locations are allowed. Even though this limitation affects the view factor calculations for guide tubes, R2RRAD is a reasonable tool to estimate rod-to-rod radiation heat transfer.

The FLECHT-SEASET test series was intended to simulate a 17x17 fuel assembly and there is a close similarity, Table 5-2, between the test bundle and a modern 17x17 assembly.

ILI UJIUU LýYUt-oUl I MN ýL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summarv Report p~nni X;-A; Table 5-2 FLECHT-SEASET & 17x17 FA Geometry Parameters Design Parameter FLECHT-SEASET 17x17 Fuel Assembly Rod Pitch (in) 0.496 0.496 Fuel Rod Diameter (in) 0.374 0.374 Guide Tube Diameter (in) 0.474 0.482 Five FLECHT-SEASET tests (Reference 6) were selected for evaluation and comparison with expected plant behavior. Table 5-3 characterizes the results of each test. The 5x5 selected rod array comprises the hot rod, 4 guide tubes and 20 near adjacent rods. The simulated hot rod is rod 7J in the tests.

0 0 0 00 Hot Rod Guide Tube -

0 0 0 Adjacent R nd 0 0' 0 0 Figure 5-1 R2RRAD 5 x 5 Rod Segment Two sets of runs were made simulating each of the five experiments and one set of cases was run to simulate the RLBLOCA evaluation of a limiting fuel assembly in an operating plant. For the simulation of Tests 31805, 31504, 31021, and 30817, the thimble tube (guide tube) temperatures were set to the measured values. For Test 34420, the thimble tube temperature was set equal to the measured vapor temperature.

~AIIl LUIM;U L.JUoUUI I ICI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-7 For the first experimental simulation set, the temperature of all 21 rods and the exterior boundary was set to the measured PCT of the simulated test. For the second experimental set, the hot rod temperature was set to the PCT value and the remaining 20 rods and the boundary were set to a temperature 25 OF cooler providing a reasonable measure of the variation in surrounding temperatures. To estimate the rod-to-rod radiation in a real fuel assembly at LOCA conditions and compare it to the experimental results, each of the above cases was rerun with the hot rod PCT set to the experimental result and the remaining rods conservatively set to temperatures expected within the bundle. The guide tubes (thimble tubes) were removed for conservatism and because peak rod powers frequently occur at fuel assembly corners away from either guide tubes or instrument tubes. In line with the discussion in Section 4.3.1, the surrounding 24 rods were set to a temperature estimated for rods of 4 percent lower power. The boundary temperature was estimated based an average power 6 percent below the hot rod power. For both of these, the temperature estimates were achieved using a ratio of pin power to the difference in temperature between the saturation temperature and the PCT.

T24 rods = 0.96 - (PCT - Tsat) + Tsat and Tsurrounding region = 0.94 - (PCT - Tsat) + Tsat.

Tsat was taken as 270 OF.

Figure 5-2 shows the hot rod thermal radiation heat transfer for the two FLECHT-SEASET sets and for the plant set. The figure shows that for PCTs greater than about 1700 OF, the hot rod thermal radiation in the plant cases exceeds that of the same component within the experiments.

%',fJIILI VfI~t;1 L/lUksUI I I-,I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paqe 5-8 Table 5-3 FLECHT-SEASET Test Parameters htc at PCT ht tSteam Thimble Rod 7J PCT PT PCTtimeStaThml Test at 6-ft (PF) Time (Btu/hr-ft 2_ Temperature -at Temperature (s) 0F) 71 (6-ft) (-F) at 6-ft (OF) 34420 2205 34 10 1850 1850*

31805 2150 110 10 1800 1800 31504 2033 100 10 1750 1750 31021 1684 29 9 1400 1350 30817 1440 70 13 900 750

set to steam temp Figure 5-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 FA 4.5 u.,

=,

Io 0

4-r cc wU 1400 1500 1600 1700 1800 1900 2000 2100 2200 2300 2400 PCT (fF)

'4jIIl URIICL L-JLA.eUI I 1:IUL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paqe 5-9 5.3.3 Rod-to-Rod Radiation Summary In summary, the conservatism of the heat transfer modeling established by benchmark can be reasonably extended to plant applications, and the plant local peaking provides a physical reason why rod-to-rod radiation should be more substantial within a plant environment than in the test environment. Therefore, the lack of an explicit rod-to-rod radiation model, in the version of S-RELAP5 applied for realistic LOCA calculations, does not invalidate the conclusion that the cladding temperature and local cladding oxidation have been demonstrated to meet the criteria of 10 CFR 50.46 (Reference 19) with a high level of probability.

%..#IJI ILl 'II0L LJJLUI I II IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-10 5.4 Film Boiling Heat Transfer Limit Question:

In the St Lucie Unit 2 Cycle 23 Fuel Transition calculations, is the Forslund-Rohsenow model contribution to the heat transfer coefficient limited to less than or equal to 15 percent when the void fraction is greaterthan or equal to 0.9?

Response

Yes, the version of S-RELAP5 employed for the St Lucie Unit 2 RLBLOCA analysis limits the contribution of the Forslund-Rohsenow model to no more than 15 percent of the total heat transfer at and above a void fraction of 0.9. Because the limit is applied at a void fraction of 0.9, the contribution of Forslund-Rohsenow within the 0.7 to 0.9 interpolation range is limited to 15 percent or less. This is a change to the approved RLBLOCA EM (Reference 1). This feature is carried forward into the UJUN14 version of S-RELAP5.

5.5 Downcomer Boiling Question:

If the PCT is greaterthan 1800°F or the containment pressure is less than 30 psia, has the St Lucie Unit 2 downcomer model been rebenchmarked by performing sensitivity studies, assuming adequate downcomer noding in the water volume, vessel wall and other heat structures?

Response

The downcomer model for St. Lucie Unit 2 has been established generically as adequate for the computation of downcomer phenomena including the prediction of potential local boiling effects. The model was benchmarked against the UPTF tests and the LOFT facility in the RLBLOCA methodology, Revision 0 (Reference 1). Further, AREVA addressed the effects of boiling in the downcomer in a letter, from James Malay to U.S.

NRC, April 4, 2003. The letter cites the lack of direct experimental evidence but contains sensitivity studies on high and low pressure containments, the impact of additional azimuthal noding within the downcomer, and the influence of flow loss

%..OLJI ILI UIII;L L-JLJf.UI I MI11 IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-11 coefficients. Of these, the study on azimuthal noding is most germane to this question; indicating that additional azimuthal nodalization allows higher liquid buildup in portions of the downcomer away from the broken cold leg and increases the liquid driving head.

Additionally, AREVA has conducted downcomer axial noding and wall heat release studies. Each of these studies supports the Revision 0 methodology and is documented later in this section.

This question is primarily concerned with the phenomena of downcomer boiling and the extension of the Revision 0 methodology and sensitivity studies to plants with low containment pressures and high cladding temperatures. Boiling, wherever it occurs, is a phenomenon that codes like S-RELAP5 have been developed to predict. Downcomer boiling is the result of the release of energy stored in vessel metal mass. Within S-RELAP5, downcomer boiling is simulated in the nucleate boiling regime with the Chen correlation. This modeling has been validated through the prediction of several assessments on boiling phenomenon provided in the S-RELAP5 Code Verification and Validation document (Reference 12).

~LJI IU LIIML LJULUU I ICI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Renort Paae 5-12 Figure 5-3 Reactor Vessel Downcomer Boiling Diagram Mtb...

Hot downcomer walls penalize PCT by two mechanisms: by reducing subcooling of coolant entering the core and through the reduction in downcomer hydraulic head which is the driving force for core reflood. Although boiling in the downcomer occurs during blowdown, the biggest potential for impact on clad temperatures is during late reflood following the end of SIT injection. At this time, there is a large step reduction in coolant flow from the ECC systems. As a result, coolant entering the downcomer may be less subcooled. When the downcomer coolant approaches saturation, boiling on the walls initiates, reducing the downcomer hydraulic static level.

ILI UII%-,U L-/U%.,UI I IUI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-13 With the reduction of the downcomer level, the core inlet flow rate is reduced which, depending on the existing core inventory, may result in a cladding temperature excursion or a slowing of the core cooldown rate.

While downcomer boiling may impact clad temperatures, it is somewhat of a self-limiting process. If cladding temperatures increase, less energy is transferred in the core boiling process and the loop steam flows are reduced. This reduces the required driving head to support continued core reflood and reduces the steam available to heat the ECCS water within the cold legs resulting in greater subcooling of the water entering the downcomer.

The impact of downcomer boiling is primarily dependent on the wall heat release rate and on the ability to slip steam up the downcomer and out of the break. The higher the downcomer wall heat release, the more steam is generated within the downcomer and the larger the impact on core reflooding. Similarly, the quicker the passage of steam up the downcomer, the less resident volume within the downcomer is occupied by steam and the lower the impact on the downcomer average density. Therefore, the ability to properly simulate downcomer boiling depends on both the heat release (boiling) model and on the ability to track steam rising through the downcomer. Consideration of both of these is provided in the following text. The heat release modeling in S-RELAP5 is validated by a sensitivity study on wall mesh point spacing and through benchmarking against a closed form solution. Steam tracking is validated through both an axial and an azimuthal fluid control volume sensitivity study done at low pressures. The results indicate that the modeling accuracy within the RLBLOCA methodology is sufficient to resolve the effects of downcomer boiling and that, to the extent that boiling occurs; the methodology properly resolves the impact on the cladding temperature and cladding oxidation rates.

5.5.1 Wall Heat Release Rate The downcomer wall heat release rate during reflood is conduction limited and depends on the vessel wall mesh spacing used in the S-RELAP5 model. The following two

%'.*lIILI VOICi;U LJ'jIoUl I ICI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-14 approaches are used to evaluate the adequacy of the downcomer vessel wall mesh spacing used in the S-RELAP5 model.

5.5.2 Exact Solution In this benchmark, the downcomer wall is considered as a semi-infinite plate. Because the benchmark uses a closed form solution to verify the wall mesh spacing used in S-RELAP5, it is assumed that the material has constant thermal properties, is initially at temperature Ti, and, at time zero, has one surface, the surface simulating contact with the downcomer fluid, set to a constant temperature, To, representing the fluid temperature. Section 4.3 of Reference 9 gives the exact solution for the temperature profile as a function of time as (T(x,t) - To) / (Ti - To) = erf {x / (2-(a t)0 5)}, (1) where, a is the thermal diffusivity of the material given by a = k/(p Cp),

k = thermal conductivity, p = density, Cp = specific heat, and erf{} is the Gauss error function (given in Table A-1 of Reference 9).

The conditions of the benchmark are T1 = 500 OF and To = 300 OF. The mesh spacing in S-RELAP5 is the same as that used for the downcomer vessel wall in the RLBLOCA model. Figure 5-4 shows the temperature distributions in the metal at 0.0, 100 and 300 seconds as calculated by using Equation 1 and S-RELAP5, respectively. The solutions are identical confirming the adequacy of the mesh spacing used in the downcomer wall.

%-Apu ILI UltIu LjuJ.Osl I lCI ItI AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Rerort Paae 5-15 Figure 5-4 S-RELAP5 versus Closed Form Solution 550 500 e -------- _ _-e- _

LL40 D 400 E

I--

0j 3___ Closed Form, 0 s 350-i-Closed Form, 100 s

--- #-Closed Form, 300 s 00-O- S-RELAP5, 0 s 300

-- 0 S-RELAP5, 100 s

-,.O S-RELAP5, 300 s 250 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 Distance from Inner Wall, feet

U I;ULJUtU ISDJI C;I IL IL AREVA Inc. ANP-3346NP LOCA Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break Summary Report Page 5-16 5.5.3 Plant Model Sensitivity Study As additional verification, a typical 4-loop plant case was used to evaluate the adequacy of the mesh spacing within the downcomer wall heat structure. Each mesh interval in the base case downcomer vessel wall was divided into two equal intervals. Thus, a new input model was created by increasing the number of mesh intervals from 9 to 18.

The following four figures show the total downcomer metal heat release rate, PCT independent of elevation, downcomer liquid level, and the core liquid level, respectively, for the base case and the modified case. These results confirm the conclusion from the exact solution study that the mesh spacing used in the plant model for the downcomer vessel wall is adequate.

Figure 5-5 Downcomer Wall Heat Release - Wall Mesh Point Sensitivity 30000.00 Base VSL Wall (9-mesh)

...... *18-Mesh VSL Wall 2400.00 24000.00 ____________

C-)

Ul) 18000.00 U)

-) () 12000.00 C:

- 6000.00 ___________

Time (sec)

ý,jUl 2Il U1iieU UJLJUU I IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary ReDort Paae 5-17 Figure 5-6 PCT Independent of Elevation - Wall Mesh Point Sensitivity 2400.00 18(0.00 LL 0

E I-1200.00 600.00

.00.0 80.0 Time (sec)

M U11CU UWLoU1 E IC;l OL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Repnort Paae 5-18 Figure 5-7 Downcomer Liquid Level - Wall Mesh Point Sensitivity a)

-J Time (sec)

k-,LJI ILE IHL LJLJIUI I ICI PL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Renort Page 5-19 Surnmarv Renort Figure 5-8 Core Liquid Level - Wall Mesh Point Sensitivity 12.00 U)

-J 0~

Time (sec)

%..jLJI ILI UJIIC;U LJIJ%.oUI I V5I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-20 5.5.4 Downcomer Fluid Distribution To justify the adequacy of the downcomer nodalization in calculating the fluid distribution in the downcomer, two studies varying separately the axial and the azimuthal resolution with which the downcomer is modeled have been conducted.

5.5.5 Azimuthal Nodalization In a letter to the NRC dated April, 2003 (Reference 1), AREVA documented several studies on downcomer boiling. Of significance here is the study on further azimuthal break up of the downcomer noding. The study, based on a 3-loop plant with a containment pressure of approximately 30 psia during reflood, consisted of several calculations examining the affects on clad temperature and other parameters.

The base model, with 6 axial by 3 azimuthal regions, was expanded to 6 axial by 9 azimuthal regions (Figure 5-9). The base calculation simulated the limiting PCT calculation given in the EMF-2103 three-loop sample problem. This case was then repeated with the revised 6 x 9 downcomer noding.

The change resulted in an alteration of the blowdown evolution of the transient with little evidence of any affect during reflood. To isolate any possible reflood impact that might have an influence on downcomer boiling, the case was repeated with a slightly adjusted vessel-side break flow. Again, little evidence of impact on the reflood portion of the transient was observed.

The study concluded that blowdown or near blowdown events could be impacted by refining the azimuthal resolution in the downcomer but that reflood would not be impacted. Although the study was performed for a somewhat elevated system pressure, the flow regimes within the downcomer will not differ for pressures as low as atmospheric. Thus, the azimuthal downcomer modeling employed for the RLBLOCA methodology is reasonably converged in its ability to represent downcomer boiling phenomena.

%-oUIILI 'JIIVU LJ/UtJlI I IfZI, IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paqe 5-21 Figure 5-9 Azimuthal Noding Base model Revised 9 Region Model 5.5.6 Axial Nodalization The RLBLOCA methodology divides the downcomer into six nodes axially. In both 3-loop and 4-loop models, the downcomer segment at the active core elevation is represented by two equal length nodes. For most operating plants, the active core length is 12 feet and the downcomer segments at the active core elevation are each 6-

%-.*UI ILI LJII;U LJLJ%.UII IUIIL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-22 feet high. (For a 14 foot core, these nodes would be 7-feet high.) The model for the sensitivity study presented here comprises a 3-loop plant with an ice condenser containment and a 12 foot core. For the study, the two nodes spanning the active core height are divided in half, revising the model to include eight axial nodes. Further, the refined noding is located within the potential boiling region of the downcomer where, if there is an axial resolution influence, the sensitivity to that impact would be greatest.

The results show that the axial noding used in the base methodology is sufficient for plants experiencing the very low system pressures characteristic of ice condenser containments. Figure 5-10 provides the containment back pressure for the base modeling. Figure 5-11 through Figure 5-14 show the total downcomer metal heat release rate, PCT independent of elevation, downcomer liquid level, and the core liquid level, respectively, for the base case and the modified case.

The results demonstrate that the axial resolution provided in the base case, 6 axial downcomer node divisions with 2 divisions spanning the core active region, are sufficient to accurately resolve void distributions within the downcomer. Thus, this modeling is sufficient for the prediction of downcomer driving head and the resolution of downcomer boiling effects.

%--UlRG U11colu LJULa9ul 0 lrýl R AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary ReDort Page 5-23 Figure 5-10 Lower Compartment Pressure versus Time U)

CD CL Time (sec)

kLJ ~U0,[lID LULJI UI I* DL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-24 Figure 5-11 Downcomer Wall Heat Release - Axial Noding Sensitivity Study base 6x6as C1)")J 240*100 i00 24000.00...* 8x6 Case

-)

Cla a)

Cu Time (sec)

%AJPulU1L1EIV=L LJýULLI 0 IMD OL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paqe 5-25 Figure 5-12 PCT Independent of Elevation - Axial Noding Sensitivity Study 2400.00 B.ase 6x6 3ase

......................................... *8x6 Case 18B00.00 P 0

Il 120D.001 ci')

E a*

c.vi Time (sec)

\,jUD ILI Uglqvu L~JLUl I Iz;I IL.

AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-26 Figure 5-13 Downcomer Liquid Level - Axial Noding Sensitivity Study 3000 Base 6x6 ase

. 8x6 Case 20.00__ _ _ __ _ _ _

1/ .

10- 0 -"__-______

Time (sec)

'ý.=OLI EIl UIZUI LJ,2LiUI I MNl IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Reoort Paae 5-27 Figure 5-14 Core Liquid Level - Axial Noding Sensitivity Study 12.00

_J 400.0 Time (sec) 5.5.7 Downcomer Boiling Conclusions To further justify the ability of the RLBLOCA methodology to predict the potential for and impact of downcomer boiling, studies were performed on the downcomer wall heat release modeling within the methodology and on the ability of S-RELAP5 to predict the migration of steam through the downcomer. Both azimuthal and axial noding sensitivity studies were performed. The axial noding study was based on an ice condenser plant that is near atmospheric pressure during -reflood. These studies demonstrate that S-RELAP5 delivers energy to the downcomer liquid volumes at an appropriate rate and that the downcomer noding detail is sufficient to track the distribution of any steam formed. Thus, the required methodology for the prediction of downcomer boiling at system pressures approximating those achieved in plants with pressures as low as ice condenser containments has been demonstrated.

~AJI ILl LIlI~U LJU~.sUI I I~I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-28 5.6 Break Size Question:

Were all break sizes assumed greaterthan or equal to 1.0 ft2 ?

Response

Yes. The NRC has requested that the break spectrum for the realistic LOCA evaluations be limited to accidents that evolve through a range of phenomena similar to those encountered for the larger break area accidents. This is a change to the approved RLBLOCA EM (Reference 1). The larger break area LOCAs are typically characterized by the occurrence of dispersed flow film boiling at the hot spot, which sets them apart from smaller break LOCAs. This occurs generally in the vicinity of 0.2 DEGB (double-ended guillotine break) size (i.e., 0.2 times the total flow area of the pipe on both sides of the break). However, this transitional break size varies from plant to plant and is verified only after the break spectrum has been executed. AREVA has sought to develop sufficient criteria for defining the minimum large break flow area prior to performing the break spectrum. The purpose for doing so is to assure a valid break spectrum is performed.

5.6.1 Break / Transient Phenomena In determining the break size criteria, the characteristics of larger break area LOCAs are examined. These LOCA characteristics involve a rapid and chaotic depressurization of the reactor coolant system (RCS) during which the three historical approximate states of the system can be identified.

Blowdown The blowdown phase is defined as the time period from initiation of the break until flow from the SITs begins. This definition is somewhat different from the traditional definition of blowdown which extends the blowdown until the RCS pressure approaches containment pressure. The blowdown phase typically lasts about 12 to 25 seconds, depending on the break size.

~LIIl LIR5U LJUtA.Jl I H:;I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-29 Refill is that period that starts with the end of blowdown, whichever definition is used, and ends when water is first forced upward into the core. During this phase the core experiences a near adiabatic heatup.

Reflood is that portion of the transient that starts with the end of refill, follows through the filling of the core with water and ends with the achievement of complete core quench.

Implicit in this break-down is that the core liquid inventory has been completely, or nearly so, expelled from the primary system leaving the core in a state of near core-wide dispersed flow film boiling and subsequent adiabatic heatup prior to the reflood phase. Although this break down served as the basis for the original deterministic LOCA evaluation approaches and is valid for most LOCAs that would classically be termed large breaks, as the break area decreases the depressurization rate decreases such that these three phases overlap substantially. During these smaller break events, the core liquid inventory is not reduced as much as that found in larger breaks. Also, the adiabatic core heatup is not as extensive as in the larger breaks which results in much lower cladding temperature excursions.

5.6.2 New Minimum Break Size Determination No determination of the lower limit can be exact. The values of critical phenomena that control the evolution of a LOCA transient will overlap and interplay. This is especially true in a statistical evaluation where parameter values are varied randomly with a strong expectation that the variations will affect results. In selecting the lower area of the RLBLOCA break spectrum, AREVA sought to preserve the generality of a complete or nearly complete core dry out accompanied by a substantially reduced lower plenum liquid inventory. It was reasoned that such conditions would be unlikely if the break flow rate was reduced to less than the reactor coolant pump flow. That is, if the reactor coolant pumps are capable of forcing more coolant toward the reactor vessel than the break can extract from the reactor vessel, the downcomer and core must maintain some degree of positive flow (positive in the normal operations sense). The circumstance is,

%-,fU[I LE UIIU L_9UUU1A I M;1 OL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-30 of course, transitory. Break flow is altered as the RCS blows down and the RC pump flow may decrease as the rotor and flywheel slow down if power is lost. However, if the core flow was reduced to zero or became negative immediately after the break initiation, then the event was quite likely to proceed with sufficient inertia to expel most of the reactor vessel liquid to the break. The criteria base, thus established, consists of comparing the break flow to the initial flow through all reactor coolant pumps and setting the minimum break area such that these flows match. This is done as follows:

Wbreak = Abreak

Gbreak = Npump

WRCP.

This gives Abreak = (Npump

WRCP)/Gbreak.

The break mass flux is determined from critical flow. Because the RCS pressure in the broken cold leg will decrease rapidly during the first few seconds of the transient, the critical mass flux is averaged between that appropriate for the initial operating conditions and that appropriate for the initial cold leg enthalpy and the saturation pressure of coolant at that enthalpy.

Gbreak = (Gbreak(PO, HCLO) + Gbreak(PCLsat, HCLO))/2.

The estimated minimum LBLOCA break area, Amin, is 2.94 ft2 and the break area percentage, based on the full double-ended guillotine break total area, is 29.9 percent.

Table 5-4 provides a listing of the plant type, initial condition, and the fractional minimum RLBLOCA break area, for all the plant types presented as generic representations in the next section.

%.j.J'IILI VilIzu I-JULOUI I =1I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary RP.nnrt Pn ga q_'Aj Table 5-4 Minimum Break Area for Large Break LOCA Spectrum Spectrum Spectrum Plant System Cold Leg Subcooled Saturated RCP flow Minimum Minimum Pressure Enthalpy Gbreak Gbreak (HEM) (Ibm/s) Break Area Break Area (psia) (Btu/Ibm) (Ibm/ft 2-s) (Ibm/ft2-s) (ft2) (DEGB)

A 3-LoopW 2250 555.0 23190 5700 31417 2.18 0.26 Design ___________

B 3-LoopW 2250 544.5 23880 5450 28124 1.92 0.23

- Design ______

3-Loop D W 2250 550.0 23540 5580 29743 2.04 0.25

- Design ______

D 2x4DCE 2100 538.8 22860 5310 21522 1.53 0.24

- Design____ __

E 2x4 CE 2055 535.8 22630 5230 37049 2.66 0.27 Design F 4-loopW 2160 540.9 23290 5370 39500 2.76 0.33 1Design I _ _ _ _ I_ _ _ _ I _ _ _ _ _ I _ _ _ _ _ I_ _ _ _ I _ _ __ _ I _ _ _ _ _

The split versus double-ended break type is no longer related to break area. In concurrence with Regulatory Guide 1.157, both the split and the double-ended break will range in area between the minimum break area (Amin) and an area of twice the size of the broken pipe. The determination of break configuration, split versus double-ended, is made after the break area is selected based on a uniform probability for each occurrence.

%..#JIILI URIIULJ UtU~lI MI1IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-32 5.6.3 Intermediate Break Size Disposition With the revision of the smaller break area for the RLBLOCA analysis, the break range for small breaks and large breaks are no longer contiguous. Typically the lower end of the large break spectrum occurs at between 0.2 to 0.3 times the total area of a 100 percent double-ended guillotine break (DEGB) and the upper end of the small break spectrum occurs at approximately 0.05 times the area of a 100 percent DEGB. This leaves a range of breaks that are not specifically analyzed during a LOCA licensing analysis. The premise for allowing this gap is that these breaks do not comprise accidents that develop high cladding temperature and thus do not comprise accidents that critically challenge the emergency core cooling systems (ECCS). Breaks within this range remain large enough to blowdown to low pressures. Resolution is provided by the large break ECC systems and the pressure-dependent injection limitations that determine critical small break performance are avoided. Further, these accidents develop relatively slowly, assuring maximum effectiveness of those ECC systems.

A variety of plant types for which analysis within the intermediate range have been completed were surveyed. Although statistical determinations are extracted from the consideration of breaks with areas above the intermediate range, the AREVA best-estimate methodology remains suitable to characterize the ECCS performance of breaks within the intermediate range. Table 5-4 provides a listing of the plant type, initial condition, and the fractional minimum RLBLOCA break area. Figure 5-15 through Figure 5-20 provide the enlarged break spectrum results with the upper end of the small break spectrum and the lower end of the large break spectrum indicated by bars. Table 5-5 provides differences between the true large break region and the intermediate break region (break areas between that of the largest SBLOCA and the smallest RLBLOCA).

The minimum difference is 141 *F; however, this case is not representative of the general trend shown by the other comparisons. The next minimum difference is 704 'F (see Figure 5-15). Considering this point as an outlier, the table shows the minimum difference between the highest intermediate break spectrum PCT and large break

%-.&U11LI UIIVU LJULoU1UI II L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paae 5-33 spectrum PCT, for the six plants, as at least 463 OF, and including this point would provide an average difference of 427 OF and a maximum difference of 840 OF.

Thus, by both measures, the peak cladding temperatures within the intermediate break range will be several hundred degrees below those in the true large break range.

Therefore, these breaks will not provide a limit or a critical measure of the ECCS performance. Given that the large break spectrum bounds the intermediate spectrum, the use of only the large break spectrum meets the requirements of 10CFR50.46 for breaks within the intermediate break LOCA spectrum, and the method demonstrates that the ECCS for a plant meets the criteria of 10CFR50.46 with high probability.

Table 5-5 Minimum PCT Temperature Difference - True Large and Intermediate Breaks Maximum Generic Maximum lant PPCT Plant Label (F(OFDelta PCT (°F) Large PCT Average Delta Description (Table 5-4) Intermediate Size Break (OF) PCT (°F)

Size Break A 17461 1887 1411 3-Loop W B 1273 1951 678 4271 Design C 1326 1789 463 2x4 CE D 984 1751 767 767 Design E 869 1636 767 3-loopW F 1127 1967 840 840 Design 1 The 2 nd highest PCT was 1183 OF. This changes the Delta PCT to 704 OF and the average delta increases to 615 OF.

%-AFIILI L11lI-U L-/UtsUI I IC;I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Rerort Paae 5-34 Figure 5-15 Plant A - Westinghouse 3-Loop Design 2000 Upper End of Large Break SSBLOCA Spectrum Break Size Minimum 1800 Spectrum Break Area 1600 1400 -

10 1200 1000 800 600 0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine

'.~IIl 'JIIZ:U L-JI4LUl I ICI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-35 Figure 5-16 Plant B - Westinghouse 3-Loop Design 2000 Upper SBLOCA End of" Large Break SBLOCASpectrum Break Size SpeMnrum SpectrumMinimum 1800 Spectrum Break Area ---

4 1600 9*j

4 4 4

4 1400 U

1200 1000 800 600 0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine

%-#UlILI UIIVU L-;U%-oUI I 1t::',1 IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Recoort Paqe 5-36 Figure 5-17 Plant C - Westinghouse 3-Loop Design 2000 Upper End of SBLOCA Large Break Spectrum Break Size Minimum 1800 Spectrum Break Area 1600 1400 1200 1000 800 600 0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine

'%..AJI ILI VOIieU LJ/Ut,UlI ItZI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summarv Rpnnrt P £e.537 Pane 5-37 Figure 5-18 Plant D - Combustion Engineering 2x4 Design 2000 Upper End of Large Break SBLOCALagBrk Break SiSpectrum Break Size k Minimum 1800 -Spectrum Break Area 1600

4*

1400 U,

UJ 1200 1000 A

8 Ann 0.0000 IOU1 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000

_T__

0. 7000 Break Area Normalized to Double Ended Guillotine 0.8000 0.9000 1.0000

%.jý LJIU 'jiiu L/JsuI-u I IqZI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Reoort Paae 5-38 Figure 5-19 Plant E - Combustion Engineering 2x4 Design 2000 Upper End of Large Break SBLOCA Spectrum Break Size Minimum 1800 Spectrum Break Area 1600 01 1400 1-1200 1000 800 600 0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine

~.AIIl IJEIU LJUI.,l IH IMit AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paae 5-39 Figure 5-20 Plant F - Westinghouse 3-loop Design 2200.0000 I

Large Break Spectrum Upper End of Minimum 2000.0000 SBLOCA e" Break Area Break Size Spectrum 1800.0000

V 1600.0000 Aov**** **

1400.0000 1200.0000 1000.0000 800.0000 600.0000 0.00000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine 5.7 Detail information for Containment Model Containment initial conditions and cooling system information are provided in Table 3-8 and Heat Sinks are provided in Table 3-9. For St Lucie Unit 2, the scatter plots of PCT versus the sampled containment volumes and initial atmospheric temperature are shown in Figure 5-21 and Figure 5-22. Containment pressure as a function of time for limiting case is shown in Figure 5-23.

'%.jflI ILI LJ1ICU LJULtIUI I 11,5 IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Renort Pane 5-40 Figure 5-21 PCT vs. Containment Volume 2000 1800 F-1600 M U NoIE Ej: El El 4 El.

EP0 U ME % 1 MElI 1400 El M U

0 0 El Ii 0 1200 a.,

1000 800 600

Split Break

[] Guillotine Break 400 2.4500e+06 2.5500e+06 2.6500e+06 Containment Volume (ft3)

%-.#UI ILI UIJIIU L-JULUI I It-I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary ReDort Paae 5-41 Figure 5-22 PCT vs. Initial Containment Temperature 2000 1800 E]

El U M 1600 U a El El

U lElL1 M m

EJEl El U] 0 MR U I 1400 El

[]

U.

0 ý El

j. 1200 M 0-El 1000 800 600 0 Split Break El Guillotine Break 400 80 90 100 110 120 130 SIT Temperature (OF)

ý_u MUHLIZ-LU L1PUU~UI I IaI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Reoort Paae 5-42 Figure 5-23 Containment Pressure versus Time for the Limiting Case 100 90 80 70 60 (A

50 CL 40 30 20 10 0

0 200 400 600 Time (s) 5.8 Cross-References to North Anna Question:

In order to conduct its review of the St Lucie Unit 2 application of AREVA's realistic LBLOCA methods in an efficient manner, the NRC staff would like to make reference to the responses to NRC staff requests for additionalinformation that were developed for the application of the AREVA methods to the North Anna Power Station, Units I and 2,

%-1FIILI IJII.;1 L/;ULUI I DICI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paqe 5-43 and found acceptable during that review. The NRC Staff safety evaluation was issued on April 1, 2004 (Agency-wide Documentation and Management System (ADAMS) accession number ML040960040). The staff would like to make use of the information that was provided by the North Anna licensee that is not applicable only to North Anna or only to subatmospheric containments. This information is contained in letters to the NRC from the North Anna licensee dated September 26, 2003 (ADAMS accession number ML032790396) and November 10, 2003 (ADAMS accession number ML033240451). The specific responses that the staff would like to reference are:

September 26, 2003 letter. NRC Question 1 NRC Question 2 NRC Question 4 NRC Question 6 November 10, 2003 letter: NRC Question I Please verify that the information in these letters is applicable to the AREVA model applied to St Lucie Unit 2 except for that information related specifically to North Anna and to sub-atmosphericcontainments.

Response

The responses provided to questions 1, 2, 4, and 6 are generic and related to the ability of ICECON to calculate containment pressures. They are applicable to the St Lucie Unit 2 RLBLOCA submittal.

Question 1 - Completely Applicable Question 2 - Completely Applicable Question 4 - Completely Applicable (the reference to CSB 6-1 should now be to CSB Technical Position 6-2). The NRC altered the identification of this branch technical position in Revision 3 of NUREG-0800.

Question 6 - Completely applicable. The supplemental request and response are applicable to St Lucie Unit 2.

5.9 GDC 35 - LOOP and No-LOOP Case Sets Question:

10CFR50, Appendix A, GDC [GeneralDesign Criterion]35 [Emergency core cooling]

states that, "Suitableredundancy in components and features and suitable interconnections,leak detection, isolation, and containment capabilitiesshall be

'ki1JI ILI LDIIML L/juIsUI I IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paqe 5-44 provided to assure that for onsite electric power system operation (assuming offsite electric power is not available)and for offsite electric power operation (assumingonsite power is not available) the system function can be accomplished,assuming a single failure."

The Staff interpretationis that two cases (loss of offsite power with onsite power available, and loss of onsite power with offsite power available)must be run independently to satisfy GDC 35.

Each of these cases is separate from the other in that each case is representedby a different statisticalresponse spectrum. To accomplish the task of identifying the worst case would require more runs. However, for LBLOCA analyses (only), the high likelihood of loss of onsite power being the most limiting is so small that only loss of offsite power cases need be run. (This is unless a particularplant design, e.g., CE

[Combustion Engineering]plant design, is also vulnerable to a loss of onsite power, in which situation the NRC may require that both cases be analyzed separately.This would require more case runs to satisfy the statisticalrequirement than forjust loss of offsite power.)

What is your basis for assuming a 50% probabilityof loss of offsite power? Your statisticalruns need to assume that offsite power is lost (in an independent set of runs).

If, as stated above, it has been determined that Palisades,being of CE design, is also vulnerable to a loss of onsite power, this also should be addressed(with an independent set of runs).

Response

In concurrence with the NRC's interpretation of GDC 35, a set of 59 cases each was run with a LOOP and No-LOOP assumption. The set of 59 cases that predicted the highest figure of merit, PCT, is reported in Section 2 and Section 3, herein. The results from both case sets are shown in Figure 3-22. This is a change to the approved RLBLOCA EM (Reference 1).

5.10 Input Variables Statement Question:

Provide a statement confirming that Florida Power & Light (FP&L) and its LBLOCA analyses vendor have ongoing processes that assure that the input variables and ranges of parameters for the LBLOCA analyses conservatively bound the values and

ILI URI%.5U LdUtaUl 0 IM CL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 5-45 ranges of those parameters for the operated St Lucie Nuclear Plant Unit 2 (SLB). This statement addressescertain programmaticrequirements of 10 CFR 50.46, Section (c).

Response

FP&L and the LBLOCA Analysis Vendor have an ongoing process to ensure that all input variables and parameter ranges for the St. Lucie Unit 2 realistic large break loss-of-coolant accident are verified as conservative with respect to plant operating and design conditions. In accordance with FP&L Quality Assurance program requirements, this process involves

1) Definition of the required input variables and parameter ranges by the Analysis Vendor.

2) Compilation of the specific values from existing plant design input and output documents by FP&L and Vendor personnel in a formal analysis input summary document issued by the Analysis Vendor and

3) Formal review and approval of the input document by FP&L. Formal FP&L approval of the input document serves as the release for the Vendor to perform the analysis.

Continuing review of the input document is performed by FP&L as part of the plant design change process and cycle-specific core design process. Changes to the input summary required to support plant modifications or cycle-specific core alternations are formally communicated to the Analysis Vendor by FP&L. Revisions and updates to the analysis parameters are documented and approved in accordance with the process described above for the initial analysis.

%\.#lI LI UJiieu LjutoslI I It:_I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-1 6.0 RECENT NRC REQUEST FOR ADDITIONAL INFORMATION (RAI) AND AREVA RESPONSES The NRC staff has found that strict adherence to currently referenced, or proposed for referencing AREVA methodologies are inconsistent with the NRC's requirements and review guidance without appropriate justification. This section addresses the NRC staffs concerns for the AREVA RLBLOCA methodology.

6.1 Thermal Conductivity Degradation - Once-Burned Fuel Questions:

1. EMF-2103 considers only fresh fuel. Once-burned fuel is more highly oxidized and has a lower thermal conductivity. The cladding of higher bumup fuel may heat differently than the analyzed fuel, and [once-bumed fuel may have a higher linearheat rate than fresh fuel]. This issue results in a potentialnon-conservatism for predictedpeak cladding temperatureand local oxidation.

a. 10 CFR 50.46 requires the ECCS cooling performance calculation to include a number of postulatedloss of coolant accidents of different sizes, locations, and otherpropertiessufficient to provide assurance that the most severe postulated loss of coolant accidents are calculated.

b. The methodology requiressupplementalinformation, sensitivity studies, or revision to include analysis of the effects of once-burnedfuel, to demonstrate compliance with 10 CFR 50.46.

c. For the PCT-limiting RLBLOCA case, please provide:

i. Correctedand uncorrectedradialtemperatureprofile of the hot rod at the time and location of peak cladding temperature.

ii. Temperature vs. time for the limiting PCT case at the limiting location, including the fuel centerline, fuel average, and clad surface temperatures. Indicate the end of blowdown, start of refill, and start of reflood on this graph.

iii. Burnup for the limiting rod.

The NRC concern covers a wide range of specific items but can be paraphrased as:

"How does the AREVA RLBLOCA analysis for St. Lucie provide a licensing basis for fuel throughout its operational life with particular attention to the phenomena of thermal conductivity degradation with burnup?" In response, the following explanation of the methodology employed for St. Lucie is provided and followed by specific responses to each of the particular questions.

%_#U1I ILI UtIIU L-Jut.A..J I IMI it AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-2 The AREVA transition package has been updated to specifically model both first and second cycle fuel rods. Third cycle fuel does not retain sufficient energy potential to achieve significant cladding temperatures nor cladding oxidation and is not included in the RLBLOCA individual pin calculations. The burnup for the individual first and second cycle rods analyzed is assigned according to the sampled time in cycle. The time in cycle is sampled once and is the same for both the fresh (first cycle) and once-burned (second cycle) fuel. Burnup for the fresh and once-burned rods is different in accordance with the cycle management. Likewise, pin pressure and thermal conductivity differ.

In addition to the thermal conductivity and fuel temperature adjustments for burnup, a burnup dependent reduction in allowed peaking is needed for the once-burned fuel.

For first cycle fuel, the RLBLOCA methodology increases the Fr to the Technical Specification maximum (including uncertainty) for the first cycle hot rods in the model. Shortly into the cycle, once-burned fuel has insufficient energy potential to achieve this peaking. A burnup dependent reduction in allowed peaking is therefore applied through an adjustment in the second cycle Fr. Shortly into the cycle, the once-burned fuel (second cycle) has insufficient energy potential to achieve this peaking.

The Fr for the once-burned peak pin is conservatively set to 100% of the fresh peak pin at the beginning of the irradiation cycle. The modeling of the burnup dependent reduction in peaking is applied through an adjustment to the Fr based on the power ratio shown in Table 6-1.

Responses:

1.a, & 1.b. Paraphrasedconcern: Requirement to treat a wide range of conditions and sensitivity studies necessary to cover once-burned fuel.

The inclusion of once-burned fuel rods in each calculation of the case set provides the required range of parameters and sensitivity studies to satisfy the 10 CFR 50.46 requirements.

1.c Paraphrasedconcern: Provide correctedand uncorrectedradial temperature results, temperatures in the pellet versus time, and

%AJI IMUL1JCIU L/J~U1LJI II L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-3 theburnup for the limiting case.

Figure 6-4 shows the corrected radial temperature profile and uncorrected centerline temperature for the limiting case hot rod at the initiation of the transient. Because the uncorrected radial profile is never used or recorded in the methodology, it cannot be provided. However, the uncorrected centerline temperature is calculated from the equation described below and is shown on Figure 6-4. As the pellet power is not adjusted the radial temperature profile must follow the corrected profile closely and the two must converge at the surface of the pellet. Figure,6-5 shows the transient centerline, surface, and average fuel temperatures of the fresh U0 2 rod at the PCT elevation for the limiting PCT case. In this case, all of the fresh rods have higher PCTs than the once-burned rods. The most limiting once-burned rod is the U0 2 rod. With a cycle burnup of approximately 12365.5 EFPH, the fresh fuel has a burnup of 24.2 MWd/kgU while the once-burned fuel has a burnup of 37.6 MWd/kgU.

Thermal Conductivity Degradation Related Questions:

Paraphrasedconcern: Provide information on the treatment of thermal conductivity degradation.

Thermal conductivity degradation impacts the ability to transfer energy from within the pellet to the pellet surface and consequently through the cladding to the coolant. Both the initial pellet temperature and the transient release of energy from the pellet are affected. The impact of thermal conductivity changes with burnup are treated by applying a bias. This bias and a measure of the uncertainty in the data were determined by benchmarking the fuel performance code, RODEX3A, to a set of data that extends past the licensed burnup. The bias adjusts the initial fuel temperature to the mean of the benchmark results. The sampled uncertainty is used to provide for the variance of the benchmarks.

The database for the benchmarks is that used to qualify and approve the RODEX4 code (Reference 13). The data from three experimental rods (cases 432R2, 432R6, and 597R8) were not used in the benchmarks. Test 597R8 was not appropriate for this application. Cases 432R2 and 432R6 are rod studies that are not configured

~~AIILE L.JIIZ1 L/L)'~AJI 1tIe IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-4 appropriately these types of comparisons. Essentially, these fuel rods are not representative of commercial PWR fuel. Part of the benchmark activity was to incorporate a fractional representation of difference between the RODEX3A calculated results and the data. The fractional adjustment provides a better adjustment over a range of initial temperatures. Therefore, for each benchmark case the Tfraction was determined.

Tjaciioni = Trodex3A - Tdata Trodex3 A where:

Tfraction = Delta fractional temperature of computed to data (K),

Trodex 3A = Temperature computed by RODEX3A (K) and Tdata = Temperature from the RODEX4 database (K)

Figure 6-1 shows the RODEX3A benchmark results along with a polynomial fitted to the results using the least squares method. The negative of this polynomial is the bias which is added to RODEX3A predictions to achieve agreement with the data. Figure 6-2 shows the results of applying this bias in comparison to the results of applying the original RLBLOCA methodology Revision 0 bias. It is evident that the bias makes the adjustment for burnup effects in accordance with the data.

m*sJILIYIIU LJUtuU1I IICII L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary ReDort Page 6-5

%_fjlI ILO UOIC~u LUut-OlI 0 IZ;1 IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-6 Figure 6-3 provides the bias adjustment T11.-- ' , as a function of burnup, using the above polynomial curve fit.

The uncertainty is determined from a Gaussian distribution characterized by a [ ]

standard deviation and added to Tnew. The fuel temperature calculation is then repeated with a multiplier, fuel K, on the code calculated fuel thermal conductivity. The fuel centerline temperature is compared to 'Tnew + uncertainty' and the calculation is repeated with an adjusted fuel K as necessary. The process is continued until the calculated centerline fuel temperature matches 'Tnew + uncertainty'. Since the process applies an adjustment to the fuel thermal conductivity, the temperature throughout the pellet is adjusted appropriately. The final multiplier is applied to the thermal conductivity throughout the transient.

Because the data fitting covers the complete range of applicable burnup it is applied as such and the zero bias offset used in Revision 0 for the first 10 GWd/mtU burnup is eliminated.

Paraphrasedconcern: How is radialtemperatureprofile computed?

The RODEX3 topical report, ANF-90-145(P)(A), Appendix B (Reference 14) details the calculation of the radial temperature distribution.

Paraphrasedconcern: Which codes are used?

A portion of the RODEX3A fuel model was incorporated into the S-RELAP5 code to calculate fuel response for transient analyses. This coding, referred to as the S-RELAP5/RODEX3A model, deals only with transient predictions and does not calculate the burnup response of the fuel. Instead, fuel conditions at the burnup of interest are transferred via a binary data file from RODEX3A to S-RELAP5/RODEX3A, establishing the initial state of the fuel prior to the transient. The data transferred .from RODEX3A describes the fuel at zero power. A steady-state S-RELAP5/RODEX3A calculation is required to establish the fuel state at power. The transient fuel pellet

~jII I LJI 11C UJI.JLAI I IUI HLi AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-7 radial temperature profile is computed by solving the conduction equation in S-RELAP5.

Material properties are calculated in S-RELAP5/RODEX3A.

Paraphrasedconcern: Is the adjustment made to the entire pellet?

The adjustment is applied to the entire fuel pellet. The polynomial transformation provides a bias adjustment to the fuel centerline temperature. A sampled parameter provides a random assessment and adjustment of the centerline temperature uncertainty. These are combined and the total adjustment is achieved by iterating a multiplicative adjustment to the fuel thermal conductivity until the desired fuel centerline temperature is reached.

%,#Uil OLUICIU LJUL~,l I It,* I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-8 Figure 6-1: Fractional Fuel Centerline Temperature Delta between RODEX3A and Data

ILI Ulltvu LJUL.Pul C11M 11%

AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summarv Renort Paae 6-9 Figure 6-2: Fuel Centerline Temperature Delta of RODEX3A Calculations to Data (Original and Using the New Correlation)

.jUI ILMUIICU LJUL)UI I IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-10 Figure 6-3: Correction Factor (as applied for temperatures in Kelvin)

%-UIN IUHU LUULUI I1I I AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-11 Figure 6-4: Radial Temperature Profile for Hot Rod

%-Aug ILE UfIlzu Ljuk~ouI I IC~g Ok AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Reoort Paae 6-12 Figure 6-5: Temperature versus Time for Fuel Centerline, Clad Surface, and Fuel Average

k0u ILE JHIZL. LJUýJLdI 9 IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paqe 6-13 Figure 6-6: Fresh and Once-Burned U0 2 Rod PCT Transient at the Hot Spot Table 6-1: Bounding Once-Burned Fuel Power Ratios by Burnup

%-.1U1MU11C;U L-YUkdUlMý11L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-14 6.2 Decay Heat Treatment Question:

1. Provide additionalinformation to justify the use of the selected analytic treatment for decay heat uncertainty in the RLBLOCA model.

a. The NRC needs to understandthe sensitivity that PCT has with respect to the decay heat uncertainty, please re-execute the limiting case with a 1.03 decay heat multiplier and report the results.

Response

The AREVA RLBLOCA EM decay heat calculations are based on the 1979 ANSI/ANS standard (Reference 15). The standard is applicable to light water reactors containing low enriched uranium as the initial fissile material; all plants, to which the RLBLOCA EM is applicable, are such plants. The selected approach to simulate fission product decay assures a representative yet conservative treatment. The AREVA EM fission product decay heat simulation and the basis for the conservatism of the approach are outlined in the remainder of this response.

Non-Sampling Approach to Decay Heat The RLBLOCA methodology proposed herein utilizes the U 2 35 decay curve from the 1979 ANSI/ANS standard for fully saturated decay chains as the decay for all fission products. The fully saturated chains result from an assumption of infinite operation.

The total energy per fission is assumed to be 200 MeV (Reference 15). No bias or uncertainty is assigned to the fission product decay heat. Differing from the base EMF-2103 evaluation model approach, the uncertainty for the decay heat parameter is set to zero and no sampling is done on this parameter, resulting in the decay heat being used with a 1.0 multiplier. The decay heat in the analysis is always the 1979 ANS standard for decay heat from U 2 35 with fully saturated decay chains, corresponding to infinite operation, assuming 200 MeV per fission.

lV.AJI ILI 'jaidu LI-jut-u I M1I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-15 Conservatism in the Approach In the approach used, the total energy per fission is assumed to be 200 MeV whereas a more accurate value for U 23 5 would be greater than 202 MeV per fission. This imparts a 1 percent conservatism.

During irradiation, plutonium accumulates such that the ratio of plutonium-to-uranium fission-energy production rate is substantial and increasing. Because the decay energy resulting from plutonium fissions is less than that from U 2 35 , the decay energy is reduced from U235 fully saturated decay chains as the fuel is burned. Thus, as burnup increases, the RLBLOCA decay heat modeling with U 23 5 only, accrues conservatism.

This conservatism applies to all regions of the core according to the mix of burnups represented within each region.

The fresh fuel, hot pin and hot assembly, begins operation with no plutonium.

Therefore, the reduction in decay heat due to plutonium build-up is not applicable to the low burnup fuel in the initial period of the cycle. However, for fresh fuel, the concentrations of long decay term fission products will not have built up. The lack of long decay term sources comprises a reduction in decay heat rate of several percent over the first year of operation, making the infinite operation assumption conservative while the plutonium concentration is accumulating.

Calculations of these considerations based on the 1979 ANS standard have been performed to demonstrate the conservatism of the selected approach. Figure 6-7 and Figure 6-8show the decay heat versus time for:

- Infinite Operation of U235 (the AREVA decay heat model)

- Finite Operation to 0.1 GWD/mtU of all fissionable isotopes with uncertainties added

- Finite Operation to 1 GWD/mtU of all fissionable isotopes with uncertainties added

- Finite Operation to 1 GWD/mtU of all fissionable isotopes without uncertainties

- Finite Operation to 20 GWD/mtU of all fissionable isotopes with uncertainties added

- Finite Operation to 40 GWD/mtU of all fissionable isotopes with uncertainties added

- Finite Operation to 60 GWD/mtU of all fissionable isotopes with uncertainties added

%-#U1I ILI VJII;U L/Uk~fUI I M;I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-16 In order to treat the Plutonium buildup effect conservatively, the finite operations curves are based on cycle management and enrichment assumptions that minimize the buildup of Plutonium. No uncertainty is included in the infinite operation curve. The uncertainties incorporated in the other curves are 2 sigma values for the individual isotopes as published in the 1979 ANS standard. This provides greater than a 95/95 confidence in each of the decay heat contributions. The contributions are added linearly according to the individual isotopes fractional occurrence of fission.

Because of the range of the decay heat parameter, the early comparison of the relationships is difficult to ascertain. Clearly the U 23 5 infinite operation curve is conservative for all times after a few seconds (-2 seconds). To better demonstrate the relationships, Figure 6-9 and Figure 6-10 provide the ratios of the finite operation curves to the infinite operation curves. The curvature of the plotted ratios during the first 2 to 3 seconds is due to the increased uncertainties during this time phase. The 1979 ANS standard is based on measured data and the difficulty of measuring decay heat within a few seconds of shutdown is reflected in these uncertainties. The highest combined finite operation decay heat curve with uncertainties exceeds the AREVA decay heat curve by only 2.5 percent at shutdown and falls below the AREVA curve in less than 2 seconds.

Thus, there is only a 5 percent probability that the infinite operation curve of decay heat will be exceeded by up to 2.5 percent and that possibility exists for the first 2 seconds of the transient. The potential accumulated under-prediction is of short duration and of no consequence to the LOCA evaluation. The decay heat curve selected is suitable while somewhat conservative for the realistic evaluation of LOCA.

In conclusion, the choice of infinite operation with pure U 23 5 fission product decay heat provides a base model that is conservative relative to the decay heat for finite operation.

For RLBLOCA evaluation, the sampling of a decay heat multiplier has been removed such that the decay heat for all cases is 1.0 times the infinite operation U235 decay chain providing conservative treatment of the 1979 ANS standard with the assumption of 200 Mev/fission. The response to question 1.a above is that it is not applicable to the St.

Lucie Unit 2 analysis. The decay heat was not sampled.

LUI ILI 'JIIt5U LJUUI I IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-17 Figure 6-7 Decay Heat Comparisons, Infinite Operation U235, Finite Operation All Isotopes (0.1 - 10 sec)

%:.A)IU ILI VJII "., LJVt-oJI I I "I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paae 6-18 Figure 6-8 Decay Heat Comparisons, Infinite Operation U235, Finite Operation All Isotopes (10 - 1000 sec)

'%..oJ MLLJIIZ;U LJULrUI I IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Paqe 6-19 Figure 6-9 Decay Heat Ratios, Finite Operation over Infinite Operation for U235, All Isotopes (0 - 10 sec)

\.AJI IL JUIZU LJULJUI I IUI OL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-20 Figure 6-10 Decay Heat Ratios, Finite Operation over Infinite Operation for U235, All Isotopes (0 - 600 sec)

%'-#lII fU IM;U LJUUUI I IUI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-21 6.3 Thermal Conductivity Degradation - Swelling, Rupture, and Relocation Question:

1. EMF-2103 does not model clad ballooning and rupture or two-sided cladding oxidation. AREVA contends that not modeling these phenomena is conservative because cladding ruptures inherently cool the cladding. Citing experimental evidence to the contrary,the staff does not agree with AREVA.

a. This may result in as much as a 50% underpredictionof cladding oxidation; the PCT may also be underpredicted,especially if baseline PCT is higher than 1800'F

b. 10 CFR 50.46(b)(2) states: "If cladding rupture is calculatedto occur, the inside surfaces of the cladding shall be included in the oxidation, beginning at the calculated time of rupture." RG 1.157, Regulatory Position 3.2.1.1, states: "A model to be used in ECCS evaluations to calculate internalfuel rod heat transfershould recognize the effects of fuel burnup, fuel pellet cracking and relocation, cladding creep, and gas mixture conductivity."

c. The NRC staff may consider a plant-specific disposition for this issue based on the unit-specific results of the ECCS evaluation for the large break LOCA.

Response

%-.*jlI ILI LUfluuL Ljukosl I! IK:I IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-22 6.4 Oxidation - Pre-transient and Single-Sided Question:

EMF-2103 models unoxidized fuel rods, on the basis that more transientoxidation will occur on an unoxidized fuel rod, which will serve to generate additionalcladding heat loads and drive the oxidation rate higher. The NRC staff does not agree that this approachis conservative.

1. This issue may result in inaccuratecladding oxidation estimates and potential cladding embrittlement issues unaccounted for in the evaluation model.

2. Information Notice 98-29, "PredictedIncrease in Fuel Rod Cladding Oxidation,"

discusses the effect high-burnupphenomena have on fuel pellet to cladding heat transfer. The staffs position, that pre- transientoxidation be consideredin ECCS evaluation models, is documented in letters to NEI dated 3/31 and 11/8/99

3. The NRC staff may considera plant-specific disposition for this issue based on the unit-specific results of the ECCS evaluation for the large break LOCA.

Response

1. AREVA's NRC-approved RLBLOCA EM uses the maximum un-ruptured cladding oxidation as representative or bounding of the transient oxidation that would have been computed at a rupture location. The position is supported by three aspects of the performed oxidation calculation.

The cladding is initialized with no initial corrosion layer. Because the oxidation rate is inversely proportional to the oxidation layer present, the use of clean cladding at the start of the accident leads to substantially higher reaction rates. For corrosions in the range of the first cycle, the difference in rate is a minimum of a 50-percent increase and increases during the cycle. The increase applies to both exterior and post-rupture interior oxidation.

The cladding temperature even in the presence of fuel relocation is reduced for the ruptured region of the cladding. In the KfK experiments (page 210 of NRC:02:062 Attachment 1 to Reference 17 and included in Reference 18) the temperature drop at rupture was between 50 and 75 K. Since the oxidation rate is exponentially proportional to the cladding temperature, a drop of 50 to 75 K for St. Lucie Unit 2 provides an oxidation rate reduction of 50-percent or more.

For ruptured cladding either the cladding interior oxidation rate is reduced by attached pellet fragments, moderate to highly burned fuel, or the cladding temperature decrease at rupture is much more than the 50 to 75 K explained above.

In either case, an additional mechanism exists to reduce the local oxidation at the rupture location.

JILI U91MU LJU~oI KII L AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Pa-ge 6-23 In conclusion, insights into the EM oxidation process and those that will evolve after rupture clearly identify differences that will reduce the transient oxidation at the rupture location to less than that which the EM calculates at un-ruptured locations.

Thus, the RLBLOCA Revision 0 EM approach of determining local transient oxidation is clearly appropriate to demonstrate compliance with the local oxidation criterion of 10CFR50.46, when combined with the pre-transient oxidation.

2. The initial corrosion layer was calculated to be 1.057 percent for the Fresh U0 2 rod (at 24.2 GWd/MTU) and 1.530 percent for the once-burned U0 2 rod (at 37.6 GWd/MTU). The initial corrosion layer was added to the transient calculated value and the total is in Table 6-2.

Table 6-2: Local Maximum Oxidation Results Fresh UO 2 Once-Burned U0 2 Pre-transient Oxidation (%) 1.057 1.530 Transient Local Maximum Oxidation (%) 2.2802 1.8520 Total Local Maximum Oxidation (%) 3.337 3.382

%-AJI ILI UOIZ:L8 L.YUtsUI I MNI BL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-24 6.6 Single Failure Assumption Questions:

1. The current licensing basis, deterministicloss of coolant accident (LOCA) analysis concluded that the limiting condition did not involve a worst-case single failure, but ratherthat it depended on injected coolant delivered in such a condition that the resultant containment environment, specifically the lower containment pressure, contributedto the limiting peak cladding temperature (PCT). Pleaseprovide information describing how this potentially limiting scenario was evaluated using the proposed best-estimate methodology.

2. Pleaseprovide additionalinformation summarizing the single-failure evaluation performed to establish compliance with General Design Criterion (GDC) 35 requirements. Identify which single failures were considered,discuss whether each failure was evaluated or explicitly analyzed, and for those failures which were explicitly analyzed, explain whether they were analyzed in a reference case or explicitly as a partof the statisticalmethodology. Also discuss the basis for the single failure evaluation. For example, were single failures considered as a matterof experience with St. Lucie Unit 2 specifically, or with a generic CE nuclearsteam supply system design?

a. The staff also needs to understandhow the limiting single failure for the CE 2x4 NSSS was determined, since the basis for the RAI response defers to NRC-approved methodology. Poringthrough EMF-2103, the staff only located sensitivity results on 3-loop W systems. In some cases, the limiting failure would be a single LPSI and in others it was a diesel. The staff could not locate a clear,generic dispositionfor the single failure at any place in EMF-2103.

b. What was done under the auspices of EMF-2103 development to ensure that the containment analysis produced a sufficiently conservative prediction that a no failure, max SI spillage case, for a CE 2x4 NSSS, is bounded by the chosen single failure? The staff will need to see that work.

Responses:

1. The current licensing basis for St. Lucie Unit 2 is AREVA's NRC-approved RLBLOCA Methodology. The RLBLOCA EM single failure considered is loss of diesel with fully functional containment sprays. The EM also conservatively prescribes:

a. The use of full containment sprays without a time delay at the minimum technical specification temperature;

b. pumped ECCS injection at the maximum technical specification temperature; and

%-OUI REAUIZZIU LJUUU1U 0 IMI IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-25

c. sampling of the containment volume (indirectly sampling containment pressure) from its nominal volume to its empty volume.

Studies, comparing several failure assumptions, including a no-failure assumption (see EMF-2103(P)(A) Revision 0, RAI response Numbers 26 and 111) validate that the ECCS and containment modeling of the AREVA methodology trends to the conservative. The containment pressure response is indirectly ranged by sampling the containment volume. The possible range to be sampled from was 2.49E+6 to 2.63E+6 ft3 for the St. Lucie Unit 2 containment volume. The "PCT vs. Containment Volume" figure (presented in the Summary Report) shows that there is little sensitivity between containment volume (indirectly pressure) and PCT for a statistical application. Thus, the methodology is responsive to the goal of a realistic evaluation, yet slightly conservative.

2. The "GDC 35 - LOOP and No-LOOP Case Sets" section in the Summary Report discusses GDC 35. The single failure prescribed by EMF-2103(P)(A) (AREVA's RLBLOCA EM) is a loss of one train of ECCS.

AREVA satisfies the GDC-35 criteria by running one set of 59 cases with offsite power available and one set of 59 cases with no offsite power available. The sampling seeds are held constant between these two case sets, with the only difference being the offsite power assumption. The case set that produces the most limiting PCT is reported, for St. Lucie Unit 2, this was offsite power available. The "GDC 35 LOOP vs. No-LOOP Cases" figure (presented in the Summary Report) displays the results from the two case sets.

a. The definition of loss of a diesel scenario by itself would mean that in addition to loss of one LPSI and one HPSI pump, one train of containment spray would not be available. The current method models all containment pressure-reducing systems as fully functional. Containment fans and containment sprays start at time zero ("Containment Initial and Boundary Conditions" table (presented in the Summary Report)).

The response to RAI #111 for EMF-2103 (Reference 17, Attachment 1 page 185 - 189) was based on sensitivities to 3-loop W plants. The Base Case, which produced the most limiting results, is described in the RAI #111 response as the loss of one diesel with full containment spray. Figure 6-11 (recreated from RAI #111, Figure 111.2) shows that for the sample plant analysis, W 3-loop, the base case, AREVA ECCS failure assumptions, is 35 0 F higher in PCT than a fully consistent loss of diesel and over 170'F greater than the loss of one LPSI case.

In the AOR a single failure of loss of one diesel was modeled, which results in the loss of one LPSI pump and one HPSI pump. Therefore, there is no LPSI flow to loops 1A and 1B and the HPSI flow is modeled for 1 pump only split to all cold legs. For the sensitivity study HPSI and LPSI flows are increased to normal operation using data for injection with no failures. Figure 6-12 through

AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-26 Figure 6-15 show PCT, containment and system pressure, ECCS injection, and downcomer level comparisons between the AOR and the maximum ECCS sensitivity study.

The loss of offsite power case remained limiting for the maximum ECCS cases with a PCT value of 1717OF compared to the AOR LOOP PCT of 17321F. The offsite power available case PCT of 16571F and the AOR NOLOOP PCT was 16630 F (all results are quoted for the most limiting, fresh U0 2 rod).

The maximum ECCS flow results in a slight reduction the containment pressure, generally on the order of 1 psia. Increased pumped safety injection flow characterized in the maximum ECCS sensitivity run does not have much of an effect on the early clad temperature response as it is masked by SIT injection. As expected, in the maximum ECCS sensitivity studies, cladding temperatures turn over comparatively quickly and hot channel quenching occurs quite a bit earlier than demonstrated in the AOR. The PCT occurs at 105.6 seconds for the AOR and 86.8 seconds in the maximum ECCS run, well beyond the start pumped injection and near the time that the SITs empty. The difference is not large but the PCT prediction is less for the sensitivity study and the single failure chosen for the AOR remains limiting.

Figure 6-11: Clad Temperature Response from Single Failure Study

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AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report P~1r, $6-9R 6-28 Pane Figure 6-13: Containment and System Pressure

101A UHIZIU Ljuq-.ýIul 9 M;1 IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summarv Renort P~ioe Summarv Renort Pane A-29 6-29 Figure 6-14: ECCS Mass Flow Rates

OLO UOM;U Ljukl'ýUl I IM IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-30 Figure 6-15: Downcomer Level

%-.,#lI ILI UJII~U L/JULsL8I I =1I It AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 6-31 6.8 Core Liquid Level Question:

Page 3-6 states, "the RLBLOCA transients are of sufficiently short duration that the switchover to sump cooling water (i.e., RAS) for ECCS pumped injection need not be considered." Forthe limiting transient,the collapsed core liquid level from 200-350 seconds appears to trend downward (Figure 3-21). An indicationof stable and increasingcollapsed liquid level would substantiatethe statement quoted above, but this is not the case for Figure 3-21. Is the SRELAP-5 model of the limiting case capable of generatingcredible results after 350s? If so, please provide results for a period of the transientsufficient to demonstratethat the core collapsed liquid levels are stable or increasing.

Response

A plot of collapsed core liquid level for the limiting case is included in the RLBLOCA summary report results for St. Lucie 2 (Section 3.5). From the plot, it is not apparent that the level is increasing at transient termination even though the core is quenched as demonstrated by the cladding temperature plots. The limiting case was extended and the reactor vessel fluid mass was plotted in Figure 6-16. The figure indicates that the vessel is filling, albeit slowly, at the transient termination.

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AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Siimmqrv 0

R Rpnnrt Picip R-32 Figure 6-16: Reactor Vessel Fluid Mass, Limiting Case 300000 250000 200000 E

-D 150000 CO U,

100000 50000 0

0 200 400 600 800 Time (s)

%.AJI IfJLJI IRVA LJfUUUI 0 ICII IL AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Page 7-1

7.0 REFERENCES

1. EMF-2103(P)(A) Revision 0, Realistic Large Break LOCA Methodology, Framatome ANP, Inc., April 2003.

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

3. Wheat, Larry L., "CONTEMPT-LT A Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-Of-Coolant-Accident,"

Aerojet Nuclear Company, TID-4500, ANCR-1219, June 1975.

4. XN-CC-39 (A) Revision 1, "ICECON: A Computer Program to Calculate Containment Back Pressure for LOCA Analysis (Including Ice Condenser Plants)," Exxon Nuclear Company, October 1978.

5. U. S. Nuclear Regulatory Commission, NUREG-0800, Revision 3, Standard Review Plan, March 2007.

6. NUREG/CR-1532, EPRI NP-1459, WCAP-9699, "PWR FLECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Task Data Report," June 1980.

7. G.P. Liley and L.E. Hochreiter, "Mixing of Emergency Core Cooling Water with Steam: 1/3 - Scale Test and Summary," EPRI Report EPRI-2, June 1975.

8. NUREG/CR-0994, "A Radiative Heat Transfer Model for the TRAC Code" November 1979.

9. J.P. Holman, Heat Transfer, 4 th Edition, McGraw-Hill Book Company, 1976.

10. EMF-CC-1 30, "HUXY: A Generalized Multirod Heatup Code for BWR Appendix K LOCA Analysis Theory Manual," Framatome ANP, May 2001.

11. D. A. Mandell, "Geometric View Factors for Radiative Heat Transfer within Boiling Water Reactor Fuel Bundles," Nucl. Tech., Vol. 52, March 1981.

12. EMF-2102(P)(A) Revision 0, S-RELAP5: Code Verification and Validation, Framatome ANP, Inc., August 2001.

13. EMF-2994(P) Rev. 4, RODEX4: "Thermal-Mechanical Fuel Rod Performance Code Theory Manual, Areva, Inc," December 2009.

14. ANF-90-145(P)(A), "RODEX3 Fuel Rod Thermal-Mechanical Response Evaluation Model," April 1996.

15. ANSI/ANS-5.1-1979, American National Standard for Decay Heat Power in Light Water Reactors, approved August 29, 1979.

%.jI.IILI VIRII;U LJLJI.sUI I I~IZ[ It AREVA Inc. ANP-3346NP Revision 0 St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Pa-ge 7-2.

16. USNRC Letter dated May 30, 2012, "Shearon Harris Nuclear Power Plant, Unit 1

- Issuance of Amendment RE: The Revision to Technical Specification Core Operating Limits Report References for Realistic Large Break Loss-of-Coolant-Accident Analysis (TAC No. ME6999)," ADAMS Accession ML12076A103.

17. AREVA Letter NRC:02:062, December 20, 2002, Responses to a Request for Additional Information on EMF-2103(P) Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," (TAC No. MB2865).

18. P. IhIe, Heat Transfer in Rod Bundles with Sever Clad Deformations, KfK 3607 B, April 1984.

19. Code of Federal Regulations, Title 10, Part 50, Section 46, "Acceptance Criteria For Emergency Core Cooling Systems For Light-Water Nuclear Power Reactors," January 2010.

v t e Letter Sequence Request
TAC:MF5494, (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
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L-2014-366, ANP-3352NP, Revision 0, St. Luice, Unit 2, Fuel Transition License Amendment Request, Technical Report

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1.0 INTRODUCTION

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5.0 REFERENCES

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7.0 REFERENCES

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