ML103560434

Extended Power Uprate Licensing Report, Attachment 5, Appendix C, ANP-2903(NP), Revision 0, EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding

ML103560434
Person / Time
Site:	Saint Lucie
Issue date:	02/28/2010
From:	AREVA NP
To:	Office of Nuclear Reactor Regulation
References
L-2010-259 ANP-2903(NP), Rev 000
Download: ML103560434 (99)

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{{#Wiki_filter:St. Lucie Unit 1 L-2010-259 Docket No. 50-335 St. Lucie Unit 1 App C-1 Realistic Large Break LOCA Summary Report St. Lucie Unit 1 Extended Power Uprate Licensing Report Appendix C St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report With Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Areva NP Inc.

AREVA NP Inc. ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding February 2010

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page i AREVA NP Inc. Copyright © 2010 AREVA NP Inc. All Rights Reserved

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page ii AREVA NP Inc. Nature of Changes Item Page Description and Justification

1.

All This is a new document.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page iii AREVA NP Inc. Contents 1.0 Introduction....................................................................................................................1-1 2.0 Summary........................................................................................................................2-1 3.0 Analysis..........................................................................................................................3-1 3.1 Description of the LBLOCA Event......................................................................3-1 3.2 Description of Analytical Models.........................................................................3-3 3.3 Plant Description and Summary of Analysis Parameters...................................3-6 3.4 SER Compliance................................................................................................3-7 3.5 Realistic Large Break LOCA Results.................................................................3-8 4.0 Generic Support for Transition Package........................................................................4-1 4.1 Reactor Power....................................................................................................4-1 4.2 Rod Quench.......................................................................................................4-1 4.3 Rod-to-Rod Thermal Radiation..........................................................................4-2 4.4 Film Boiling Heat Transfer Limit.........................................................................4-8 4.5 Downcomer Boiling............................................................................................4-8 4.6 Break Size........................................................................................................4-24 4.7 Detail information for Containment Model........................................................4-35 4.8 Cross-References to North Anna.....................................................................4-39 4.9 GDC 35 - LOOP and No-LOOP Case Sets.....................................................4-40 4.10 Input Variables Statement................................................................................4-41 5.0 Conclusions....................................................................................................................5-1 6.0 References.....................................................................................................................6-1

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page iv AREVA NP Inc. Tables Table 2-1 Summary of Major Parameters for Limiting Transient..............................................2-1 Table 3-1 Sampled LBLOCA Parameters.................................................................................3-9 Table 3-2 Plant Operating Range Supported by the LOCA Analysis......................................3-10 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 4-1 Typical Measurement Uncertainties and Local Peaking Factors..............................4-4 Table 4-2 FLECHT-SEASET & 17x17 FA Geometry Parameters............................................4-5 Table 4-3 FLECHT-SEASET Test Parameters.........................................................................4-6 Table 4-4 Minimum Break Area for Large Break LOCA Spectrum.........................................4-26 Table 4-5 Minimum PCT Temperature Difference - True Large and Intermediate Breaks..........................................................................................................................4-28

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page v AREVA NP Inc. Figures Figure 3-1 Primary System Noding.........................................................................................3-21 Figure 3-2 Secondary System Noding....................................................................................3-22 Figure 3-3 Reactor Vessel Noding..........................................................................................3-23 Figure 3-4 Core Noding Detail................................................................................................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 3-10 Total Oxidation versus PCT Scatter Plot from 59 Calculations...........................3-31 Figure 3-11 Peak Cladding Temperature (Independent of Elevation) for the Limiting Case...............................................................................................................3-32 Figure 3-12 Break Flow for the Limiting Case.........................................................................3-33 Figure 3-13 Core Inlet Mass Flux for the Limiting Case..........................................................3-34 Figure 3-14 Core Outlet Mass Flux for the Limiting Case.......................................................3-35 Figure 3-15 Void Fraction at RCS Pumps for the Limiting Case.............................................3-36 Figure 3-16 ECCS Flows (Includes SIT, LPSI and HPSI) for the Limiting Case.....................3-37 Figure 3-17 Upper Plenum Pressure for the Limiting Case....................................................3-38 Figure 3-18 Collapsed Liquid Level in the Downcomer for the Limiting Case........................3-39 Figure 3-19 Collapsed Liquid Level in the Lower Plenum for the Limiting Case....................3-40 Figure 3-20 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 4-1 R2RRAD 5 x 5 Rod Segment..................................................................................4-5 Figure 4-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 FA...................................................................................................................................4-7 Figure 4-3 Reactor Vessel Downcomer Boiling Diagram..........................................................4-9 Figure 4-4 S-RELAP5 versus Closed Form Solution..............................................................4-12 Figure 4-5 Downcomer Wall Heat Release - Wall Mesh Point Sensitivity..............................4-13 Figure 4-6 PCT Independent of Elevation - Wall Mesh Point Sensitivity................................4-14 Figure 4-7 Downcomer Liquid Level - Wall Mesh Point Sensitivity........................................4-15 Figure 4-8 Core Liquid Level - Wall Mesh Point Sensitivity...................................................4-16 Figure 4-9 Azimuthal Noding..................................................................................................4-18

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page vi AREVA NP Inc. Figure 4-10 Lower Compartment Pressure versus Time........................................................4-19 Figure 4-11 Downcomer Wall Heat Release - Axial Noding Sensitivity Study.......................4-20 Figure 4-12 PCT Independent of Elevation - Axial Noding Sensitivity Study.........................4-21 Figure 4-13 Downcomer Liquid Level - Axial Noding Sensitivity Study..................................4-22 Figure 4-14 Core Liquid Level - Axial Noding Sensitivity Study.............................................4-23 Figure 4-15 Plant A - Westinghouse 3-Loop Design..............................................................4-29 Figure 4-16 Plant B - Westinghouse 3-Loop Design..............................................................4-30 Figure 4-17 Plant C - Westinghouse 3-Loop Design..............................................................4-31 Figure 4-18 Plant D - Combustion Engineering 2x4 Design..................................................4-32 Figure 4-19 Plant E - Combustion Engineering 2x4 Design...................................................4-33 Figure 4-20 Plant F - Westinghouse 3-loop Design...............................................................4-34 Figure 4-21 PCT vs. Containment Volume..............................................................................4-36 Figure 4-22 PCT vs. Initial Containment Temperature............................................................4-37 Figure 4-23 Containment Pressure as function of time for limiting case..................................4-38 This document contains a total of 98 pages.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page vii AREVA NP Inc. Nomenclature CCTF Cylindrical Core Test Facility CE Combustion Engineering Inc. CFR Code of Federal Regulations CSAU Code Scaling, Applicability, and Uncertainty DC Downcomer DEGB Double-Ended Guillotine Break DNB Departure from Nucleate Boiling ECCS Emergency Core Cooling System EFPH Effective Full Power Hours EPU Extend Power Uprate EM Evaluation Model F Q Total Peaking Factor FP&L Florida Power and Light Company FH Nuclear Enthalpy Rise Factor HPSI High Pressure Safety Injection HFP Hot Full Power LANL Los Alamos National Laboratory LHR Linear Heat Rate RLBLOCA Realistic Large Break Loss of Coolant Accident LOCA Loss of Coolant Accident LPSI Low Pressure Safety Injection MSIV Main Steam Isolation Valve NRC U. S. Nuclear Regulatory Commission NSSS Nuclear Steam Supply System PCT Peak Clad Temperature PIRT Phenomena Identification and Ranking Table PLHGR Planar Linear Heat Generation Rate PWR Pressurized Water Reactor RAS Recirculatio n Actuation Signal RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RV Reactor Vessel RWST Refueling Water Storage Tank SI Safety Injection SIAS Safety Injection Activation Signal SIT Safety Injection Tank SER Safety Evaluation Report

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 1-1 AREVA NP Inc. 1.0 Introduction This report describes and provides results fro m a RLBL OCA analysis for the St Lucie Nuclea r Plant Unit 1 Extended Power Upra te. The plant is a CE-designed 3020 MWt plant with a large dry containment. AREVA NP is the current fuel supplier. The plant is a 2 X4 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 o peration for EPU Cycle and beyond with AREVA NPs HTP 14X 14 fuel design using standard UO 2 fuel with 2%, 4%, 6% and 8% Gd 2O3 and ZIRCALOY-4 cladding, unless changes in the Technical Specifications, Core Operating Limits Report, core design, fuel design, plant hardware, or plant operation invalidate the results present ed herein. The analysis was perfor med in compliance with the NRC-approved RLBLOCA EM (Reference 1) with exceptions noted below. Analysis results confirm the 1 0CFR50.46 (b) acceptance criteria presented in Section 3.0 are met and serve as the basis for operation of the St L ucie Nuclear Plant Unit 1 with AREVA NP fuel. Pe r RLBLOCA EM (Reference 1), fuel assemblies residing in the core for more than one cycle will not be limiti ng. Therefore, this RLBLOCA analysis covers the transition cycle with both fresh fuel and burned fuel. The non-parametric statistical methods inherent in the AREVA NP RLBLOCA methodology provide for the consider ation of a full spectrum of break sizes, break configuration (guillotine or split break), axial shapes, and plant operational paramete rs. A conservative loss of a diesel assumption is applied in which LPSI inject into the broken loop and on e intact loop and HPSI inject into all four lo ops. Regardless of th e single-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 deviations from the appro ved RLBL OCA EM (Reference 1) that were requested by the NRC. The assumed reactor core power for the St L ucie Unit 1 realistic larg e break loss-of-coolant accident is 3029 MWt. This value r epresents the 10% power uprate and 1.7% measureme nt uncertainty recapture (MUR) relative to the c urrent rated thermal p ower of 2700 MWt plus 0.3% power measurement uncertainty. (2700 MWt X (1+10%) X (1+1.7%) X (1+0.3%) = 3029 MWt)

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 1-2 AREVA NP Inc. The RLBLOCA analysis was performed with a version of S-RELAP5 that requires both the voi d fraction to be less than 0.95 and the clad temperature to b e less than 900 °F before the rod is allowed to q uench. This may result in a sligh t 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 th at limits the contribution of the Forslund-Rohsenow model to no more t han 15 percent of the t otal heat transfer at and above a void fraction of 0.9. This may result in a slight incr ease 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 (A min) a nd an area of twice t he size of the broke n pipe. Th e determination of break configuration, split versus double-ended, will be made afte r the break area is selected based on a uniform probability for each occurrence. Amin was calculated to b e 26.7 percent of the DEGB area (se e Section 4.6 for further discussion). This is not expected to have an effect on PCT results. In concurrence with the NRCs 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 is expected to be minor. During rece nt RLBLOCA EM modeling stud ies, it was n oted that cold leg co ndensation efficiency may be under-predicted. Water entering the DC post-accumulator 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 DC from t he cold leg, subsequent to the end of accumulator injectio n, reach ne ar saturatio n resulting from the condensation efficiency ranging between 80 to 10 0 percent. To assure that cold leg condensatio n would not be under-predicted, a RLBLOCA EM update was made. Noting that sa turated 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 mult ipliers were developed t hrough scoping studies using a n umber of pl ant configurationsWestinghouse - designed 3-and 4-loop plants, an d CE-designed plants. The result s of the scoping study indicated that multipliers of 10 and 150 for liquid and steam, respectively, were a ppropriate to

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 1-3 AREVA NP Inc. produce saturated fluid entering the DC. This RLBLOCA EM departure was recently discussed with the NRC and the NRC agr eed that the approach described immediately above wa s satisfactory in the interim. The modification is implement ed post-accumulation injection, 10 seconds after the vapor void fraction in the bottom of the accumulator becomes greater than 90 percent. Thus, the accumulators have injected all their water into the cold legs, and the nitrogen cover gas has entered the system and been mostly discha rged through the break before the condensation efficien cy is in creased by the factors of 10 and 15 0, for liquid and vapor respectively. Providing saturated fluid conditio ns at the DC entrance conservatively reduces both the DC driving head and the core flooding rate. Recall that test results indicat e 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 Inc. has ackno wledged an issue con cerning fuel thermal conductivity degradation as a function of burnup as raised by the NRC. In order to manage this issue, A REVA Inc. is modifying the way RODEX3A temperatures are compensated in the RLBLOCA Re vision 0/Transition package methodology. In the current process, the RLBLOCA comput es PCTs at many different times during an operating cycle. For ea ch specif ic time in cycle, the fuel conditions are computed using RODEX3A prior to starting the S-RELAP5 port ion of the analysis. A steady state condition for the given t ime in cycle using S-RELAP5 is established. A base fuel centerline temperature is e stablished in this process. Then two-transformatio n adjustments to the ba se fuel centerline temperature are computed. The f irst transformation is a linear adjustment for an exposure of 10 Mwd/MtU or higher. The second adjustment is performed in the S-RELAP5 initialization proce ss for the transient case. In the new process, a polynomial transformation is used fo r the first tra nsformation instead of a linear transformation. The rest of the RLBLOCA process for initializing the S-RELAP5 fuel rod temperature should not be altered and the rest of LOCA transient should also continue in the original fashion.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 2-1 AREVA NP Inc. 2.0 Summary The limiting PCT analysis is base d on the p arameter specificat ion g iven in Table 2-1. The limiting PCT is 1672 F for a 6% Gd 2O3 Rod in a case with LOOP conditions (LOOP is loss of offsite power. No-LOOP is with offsite power a vailable). UO2 rods and Gadolinia bearing rods of 2, 4 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 core is composed only of AREVA NP HTP 14x14 thermal hydraulically compatible fuel designs; hence, there is no mixed core consideration. The analysis assumed full core po wer operation at 3029 MWt. The value represents the 10 % power uprate and 1.7 % measurement uncertainty recapture (MUR) rel ative to the current rated thermal power (2700 MWt) plus 0.3% power measurement uncertainty. The analysis assumed a steam generator tube plugging level of 10 percent in all steam generators, a total of LHR of 15.0 kW/ft (no axial depende ncy), a total peaking fa ctor (F Q) up to a value of 2.161, an d a nuclear enthalpy rise factor (FH) up to a value of 1.749 (including 6% uncertainty). This analysis bounds typical operational rang es or technical specifica tion limits ( whichever is applicab le) with regard to Pressurizer pressur e and level; SIT pressure, temperature, and level; core average temperature; core flow; containment pressure and temperature; and RWST. The AREVA RLBLOCA methodology explicitly analyzes only fresh fuel asse mblies (see Reference 1, Appendix B). Pre vious analyses have shown that once-and twice-burnt fuel will not be limiting up to peak rod average exposures of 62,000 MWd/MTU. T he analysis demonstrates that the 10 CFR 50.46(b) criteria listed in Section 3.0 are satisfied. Table 2-1 Summary of Major Parameters for Limiting Transient Core Average Burnup (EFPH) 6874.59 Core Power (MWt) 3029.06 Hot Rod LHR, kW/ft 14.6990 Total Hot Rod Radial Peak (Fr T) 1.810 ASI 0.0393 Break Type Guillotine Break Size (ft2/side) 4.1705 Offsite Power Availability Not available Decay Heat Multiplier 0.98429

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-1 AREVA NP Inc. 3.0 Analysis The purpose of the analysis is to v erify typical technical specification peaking factor limits and the adequacy of the ECCS by demonstrating that the following 10CFR 50.46(b) criteria are met: (1) The calculated maximum fuel element cladding temperature shall not exceed 2200 °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 hyd rogen generated from t he chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be g enerated if all of the metal in the cladding cylinders surro unding 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 re mains amenable to cooling. (5) Long-term cooling is demonstrated outside of this report for operation at EPU conditions. The analysis did not e valuate core coolability due to seismic events, nor did it consider th e 10CFR 50.46(b) long-term cooling criterion. The RLBLOCA analysis conservatively considers blockage effects due to clad swelling and rupture in t he predictio n of the ho t fuel rod PCT. AREVA NP has previously performed an analysis which demonstrates that f or all cases of horizont al seismic and LOCA loads, the resulting lo ads are below the spacer grid ela stic load limit and thus the grids sustain n o permanent deformation. Section 3.1 of this repor t describes the postulated LBLOCA event. Se ction 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 a ddressed in Section 3.4. Section 3.5 summarizes the results of the RLBLOCA analysis. 3.1 Description of the LBLOCA Event A LBLOCA is initiated by a postu lated large rupture of t he RCS primary piping. Based on deterministic studies, t he worst break location is in the cold leg piping between the reactor coolant pump and the reactor vessel for the RCS loop containing th e pressurizer. The bre ak initiates a r apid 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.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-2 AREVA NP Inc. The plant is assumed to be operating normally a t full power prior to the accident. The cold leg break is a ssumed to open instant aneously. For this bre ak, a rapid depressurization occur s, along with a core flow s tagnation and reversal. This causes the fuel rods to expe rience DNB. Subsequently, the limiting fuel rods are cooled by film convection to steam. The coolant voiding creates a st rong 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 bre ak plane, the loss of subcooling in the coolant results in su bstantially reduced break flow. Thi s reduces the depressurization rate, and leads to a period of positive core flow or re duced downflow as the RCPS in the intact loops continue to supply wate r to the RV (in No-LOOP conditions). Cladding temperatures may be reduced and some portions of the core may re wet during this period. The positive core flow o r reduced downflow period ends as two-phase conditions occur in th e RCPs, re ducing 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 consid ered. This single-failur e 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 analysis assumes that the loss of one emergency diesel gene rator, which takes one train of ECCS pu mped injection o ut. LPSI inject into the broken loop and one intact loop, HPSI inject 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 pre ssure 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. A s 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 pu mped ECCS injection. A s the SITs empty, the nitrogen gas used to

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-3 AREVA NP Inc. pressurize the SITs exits through th e break. T his gas rele ase may re sult in a sh ort period of improved core heat transfer as the nitrogen gas displace s water in the downcomer. After the nitrogen ga s has been expelled, t he ECCS t emporarily may not be able to su stain full co re cooling because of the core decay heat an d the higher steam t emperatures created by quenching in the lower portions of t he core. Peak fuel rod cladding temperatures may increase for a short period until more energ y is remo ved from the c ore by the HPSI and L PSI while th e decay heat continues to fall. Stea m generated from fuel rod rewet will entrain liquid and pass through the core, vessel upper plenum, the h ot legs, the steam gen erators, and the reactor coolant pumps before it is vented out the break. Some steam flow to th e upper head and pass through the spray nozzles, which provide a vent path to the break. The resistance o f this flow path to the steam flow is balan ced 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 p rogress 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. 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 a pproach for defining and qualifying a best-estimate thermal-hydraulic cod e and quant ifies the uncertainties in a LOCA analysis. The RLBLOCA methodology consists of the following computer codes:  RODEX3A for computation of the initial fuel st ored energy, fission ga s release, and fuel-cladding gap conductance.  S-RELAP5 for the system calculation (includes ICECON for containment response).  AUTORLBLOCA for ge neration of ranged parameter values, transient input, transient runs, and general output documentation. The governing two-fluid (plus non-condensible s) model with conservation equation s for mass, energy, and momentu m transfer is used. The r eactor 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.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-4 AREVA NP Inc. 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 interactio n terms in the equations. The conservation equations h ave the same form for each phase; only the constitutive relations an d physica l properties differ. The modeling of plan t components is performed by following guideline s developed to ensure accurate accounting for physical dimensions and that the dominant phenomena expected during the LBLOCA event are captured. T he basic building blocks for modeling are hydraulic volumes for fluid paths and heat structures f or heat transfer. In addition, special purpose components exist to represent specif ic components such as the RCPs or the steam generator separators. All geometries are mod eled at the resolution n ecessary to best resolve the flow field and the phenomena being modeled within practical computational limitations. System nodalization det ails are sho wn in Figures 3-1 through 3-5. A point of clarification: in Figure 3-1, break modeling uses tw o junctions regardless of break typesplit or guillotine; for guillotine breaks, Junction 151 is deleted, it is retained fully open for s plit breaks. Hence, total break area is the sum of the areas of both break junctions. A typical ca lculation u sing S-RELAP5 begins with the est ablishment of a steady-state initial condition with all loops intact. The input parameters and initial condit ions for this steady-state calculation are chosen to reflect plant technical specifications or to match me asured data. Additionally, the RODEX 3A code provides init ial cond itions for the S -RELAP5 fu el 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 tr ansient thro ugh blowdo wn, refill an d reflood is computed continuously using S-RELAP5. Con tainment pressure is also calculate d by S-REL AP5 using containment models derived from ICECON (Reference 4), which is b ased 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 assessmen t of this computer code was made t hrough comparisons to experimental data, man y benchmarks with clad ding temperatures ranging from 1,7 00 °F (o r less) to above 2,200 °F. These assessments were used to develop quantitative estimates of the

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-5 AREVA NP Inc. ability of the code to pre dict key physical pheno mena in a PWR LBLOCA. Various modelsfor example, the core hea t transfer, t he decay heat model and the fuel cladd ing oxidation correlationare defined based on code-to-data comparisons and are, hence, plant independent. The RV internals are modeled in detail (Figur es 3-3 thro ugh 3-5) based on St Lucie Unit 1 specific inp uts. Node s and connectivity, flow areas, resistances and heat structures are all accurately modeled. The location of the hot a ssembly/hot pin(s) is un restricted; 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 (includin g the containment input file) are developed. Code inpu t 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 ca ses be created and processed. For every set of input created, each key LOCA parameter is randomly sampled over a range establishe d through cod e uncertaint y 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 b oth parameters related to LOCA pheno mena (based on the PI RT provided in Reference 1) and to plant operating parameters.

3.

Determination of Adequacy of ECCS The RLBLOCA method ology uses a non-para metric statistical approach to d etermine values of PCT at the 95 percent pr obability level. Total oxidation and t otal hydrogen are based on th e limiting P CT case. The adequacy of the ECCS is demonstrated when these results satisfy the criteria set forth in Section 3.0.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-6 AREVA NP Inc. 3.3 Plant Description and Summary of Analysis Parameters The plant a nalysis presented in this report is f or a CE-designed PWR, which has 2X 4-loop arrangement. There are two hot leg s each with a U-tube st eam generator and four cold legs each with a RCP1. The RCS includes one Pressurizer connected to a hot leg. The core contains 217 thermal-hydraulic compatible AREVA HTP 14X14 fuel assemblies with 2%, 4%, 6% and 8% gadolinia pins. The ECCS includes one HPSI, one LPSI a nd one SIT injection path per RCS loop. The b reak is modeled in the same loop as the pressurizer, as dire cted by the RLBLOCA methodology. The RLBLOCA transients are of sufficien tly short duration that the switchover to sump cooling water (i.e., RAS) for ECCS pumped injection need not be considered The S-REL AP5 model explicitly describes the RCS, RV, Pressurizer, and ECCS. The ECCS includes a SIT path and a LPSI/ HPSI path per RCS lo op. The HPSI and LPSI feed into a common header that co nnects to each cold leg pipe downstream of the RCP disch arge. The ECCS pumped injectio n is modeled as a table of flow ve rsus backpressure. This model also describes t he secondar y-side steam generator that is in stantaneously isolated (closed MSI V and feedwater trip) at the time of the break. A symmetric steam generator tube plugging level of 10 percent per steam generator was assumed. As described in the AREVA RL BLOCA me thodology, many para meters associated wit h 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 f or process or operation al parameters, includ ing ranges of sampled process para meters, and fuel design parameters used in th e analysis are given in Table 3-2. Plant data are analyzed to develop uncertainties for th e process parameters sampled in the analysis. Table 3 -3 presents a summary of the un certainties used in the analysis. Where applicable, the sampled parameter ranges are based on technical specification limits or supporting plant calculations that provide more bounding values. For the AREVA NP RLBLOCA EM, dominant containment parameters, as well as NSSS parameters, were established via a PIRT process. Other model inputs are generally taken a s nominal or conservatively biased. The PIRT outcome yielded two important (relative to PCT) 1 The RCPs are Byron-Jackson Type DFSS pumps are specified by FP&L. The homologous pump performance curves were input to the S-RELAP5 plant model; the built-in S-RELAP5 curves were not used.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-7 AREVA NP Inc. Containment parameterscontainment pressur e and temperature. In many instances, the conservative guidance of CSB 6-2 (Reference 5) was u sed in setting the remainder of the containment model input parameters. As note d in Table 3-3, contain ment tempe rature is a sampled parameter. Containment pressure response is indirectly ranged by sa mpling the containment volume (Table 3-3). Containment heat sink data is given in Table 3-9. I n accordance with Reference 1, the condensing h eat transfer coefficient is intended to be closer to a best-e stimate inst ead of a b ounding hig h value. A [ ] Uchida heat transfe r coefficient multiplier was specifically validated for use in St Lucie through application of the process used in the RLBLOCA EM (Reference 1) sample problems. The initia l conditions a nd boundary condition s are given in Table 3-8. The build ing spray is modeled at maximum heat removal capacity. All spray flow is delivered to the containment. 3.4 SER Compliance A number o f requirements on the methodology are stipulated in the c onclusions section of th e SER for the RLBLOCA methodology (Reference 1). The se requirements have all b een fulfilled during the application of the methodology as addressed in Table 3-4. 3.4.1 Item 7: Blowdown Quench One case was potential candidate for blowdown quench and was closely inspected. For this calculation, no evidence of blowdown quench was observed. Therefore, compliance to the SER restriction has been demonstrated. 3.4.2 Item 8: Top-down Quench Several pro visions have been impl emented in the S-REL AP5 model t o prevent th e top-down quench. The upper plenum nodalization features include:  the homoge nous option is selected for the junction that connects the first axial level node above the hot channel to the second axial level node above the hot channel;  no cross-flo w is allowed between t he first axial level Upp er Plenum nodes above the ho t channel to the average channel;  the CCFL model is applied on all core exit junctions. Six cases were closely examined for top-down quench. No evidence of top-down quench was observed. Therefore, compliance to the SER restriction has been demonstrated.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-8 AREVA NP Inc. 3.5 Realistic Large Break LOCA Results Two case sets of 59 transient calculations were performed sampling the paramet ers listed in Table 3-1. For each case set, PCT was calculated for a UO2 rod and for Gadolinia-bearing rods with concentrations of 2, 4, 6 and 8 w/o Gd 2O3. The limiting case set, that contained the PCT, was the set with no offsite power available. The limiting PCT (1672 F) occurred in Case 3 for a 6% Gd 2O3 rod. The major parameters for the limiting tran sient are pr esented in Table 2-1. Table 3-5 lists the results of the limiting case. The fraction of total hydrogen generated was no t directly calculated; however, it is conservatively bounded by the c alculated to tal percent oxidation, which is well below the 1 percent limit. The best-estimate PCT case is Case 17, which corresponded to the me dian case o ut of the 5 9-case set with no offsite power available. The nominal PCT was 1509 F for a 6% Gd2O3 rod. This result can be used to quantify the relative conservatism in the limiting case result. In this analysis, it was 163 F. The case re sults, event times and analysis plots for the limiting PCT case are shown in Table 3-5, Table 3-6, and Figure 3-11 through Figure 3-21. Figure 3-6 shows linear scatter plots of the key parameters sample d for the 59 calculations. Parameter labels ap pear to the left of ea ch individual plot. These figures show the paramet er ranges used in the analysis. Figure 3-7 an d Figure 3-8 show the time of PCT and break size versus PCT scatter plots f or the 59 calculations, respectively. Figure 3-9 and Figur e 3-10 sho w the maxi mum o xidation and total oxidation versus PCT scatter plots for the 59 calculation s, respectively. Key para meters for the limiting PCT case are sh own in Figure 3-11 thr ough Figure 3-21. Figure 3 -11 is the plot of PCT independent of elevation; this figure clearly indicates that the transient exhibits a sustained and stable quench. A comparison of PCT results from both case sets is shown in Figure 3-22.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-9 AREVA NP Inc. 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) De cay heat 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 Steam generator interfacial drag Condensation interphase heat transfer Metal-water reaction Plant1 Offsite power availability2 Brea k size Pressu rizer pressure Pressurizer liquid level SIT pressure SIT liquid level SIT temperature (based on containment temperature) Contai nment temperature Contai nment volume Initial RCS flow rate Initial operating RCS temperature Diesel start (for loss of offsite power only) 1 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.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-10 AREVA NP Inc. 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 0.440 in. b) Cladding inside diameter 0.384 in. c) Cladding thickness 0.028 in. d) Pellet outside diameter 0.377 in. e) Pellet density 95.35 percent of theoretical f) Active fuel length 136.7 in. g) Resinter densification [ ] h) Gd2O3 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 14X14 AREVA NP HTP fuel e) SG tube plugging 10 percent (2% asymmetry)1 2.0 Plant Initial Operating Conditions 2.1 Reactor Power a) Nominal reactor power 3029 MWt2 b) LHR 15.0 kW/ft c) FQ 2.161 d) Fr 1.8103 2.2 Fluid Conditions a) Loop flow 140.8 Mlbm/hr  M  164.6 Mlbm/hr b) RCS Cold Leg temperature 548.0 F  T  554.0 F c) Pressurizer pressure 2210 psia  P  2290 psia d) Pressurizer level 62.6 percent  L  68.6 percent e) SIT pressure 214.7 psia  P  294.7 psia f) SIT liquid volume 1090 ft3  V  1170 ft3 1 In the RLBLOCA analysis, only the maximum 10% tube plugging in each steam generator was analyzed. By independently sampling the break loss discharge coefficients, any flow differences attributed to asymmetry in the SG tube plugging is covered by use of the RLBLOCA methodology. 2 Includes 0.3% uncertainties 3 The radial power peaking for the hot rod is including 6% measurement uncertainty and 3.5% allowance for control rod insertion affect. Fr tech spec *(1+ uncert_Fr) * (1+uncert_cr_insertion) = 1.65*(1.0+0.06)*(1+0.035)=1.810

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-11 AREVA NP Inc. Table 3-2 Plant Operating Range Supported by the LOCA Analysis (Continued) Event Operating Range g) SIT temperature 115.5 F  T  124.5 F (Its coupled with containment temperature) h) SIT resistance fL/D As-built piping configuration i) Minimum ECCS boron 1900 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 pipe area) 0.2997  A  1.0 full pipe area (split) 0.2997  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 120 °F g) HPSI pump delay 19.5 (w/ offsite power) 30.0 (w/o offsite power) h) LPSI pump delay 19.5 (w/ offsite power) 30.0 (w/o offsite power) i) Containment pressure 14.7 psia, nominal value 1 j) Containment temperature 115.5 F  T  124.5 F k) Containment sprays delay 0 s l) LPSI flow BROKEN_LOOP

• LOOP-1A1
• RCS pressure LPSI flow psia gpm 18.32 1287.

23.48 1261. 33.47 1210. 43.02 1158. 47.64 1132. 52.14 1107. 69.04 1005. 87.73 877. 103.73 748. 117.05 620. 127.72 492. 135.41 364. 140.64 236. 143.98 82. 144.37 31. 144.44 0. INTACT_LOOP1

• LOOP-1B1 1 Nominal containment pressure range is -0.7 to 0.5 psig. For RLBOCA, a reasonable value between this range is acceptable.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-12 AREVA NP Inc.

• RCS pressure LPSI flow psia gpm 18.32 0.0 23.48 0.0 33.47 0.0 43.02 0.0 47.64 0.0 52.14 0.0 69.04 0.0 87.73 0.0 103.73 0.0 117.05 0.0 127.72 0.0 135.41 0.0 140.64 0.0 143.98 0.0 144.37 0.0 144.44 0.0 INTACT_LOOP2
• LOOP-1A2
• RCS pressure LPSI flow psia gpm 18.32 0.0 23.48 0.0 33.47 0.0 43.02 0.0 47.64 0.0 52.14 0.0 69.04 0.0 87.73 0.0 103.73 0.0 117.05 0.0 127.72 0.0 135.41 0.0 140.64 0.0 143.98 0.0 144.37 0.0 144.44 0.0 INTACT_LOOP3
• LOOP-1B2
• RCS pressure LPSI flow psia gpm 18.32 926.

23.48 902. 33.47 853. 43.02 804. 47.64 780. 52.14 755. 69.04 657. 87.73 535. 103.73 413. 117.05 291. 127.72 169.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-13 AREVA NP Inc. 135.41 47. 140.64 0. 143.98 0. 144.37 0. 144.44 0. m) HPSI flow BROKEN_LOOP

• RCS pressure HPSI flow psia gpm

15. 160.0 315. 137.0 615. 109.0 815. 85.0 1015. 51.0 1115. 16.0 1125. 8.0 1129. 0.0 INTACT_LOOP1

• RCS pressure HPSI flow psia gpm

15. 151.7 315. 130.0 615. 103.7 815. 81.3 1015. 48.7 1115. 15.3 1125. 5.7 1129. 0.0 INTACT_LOOP2

• RCS pressure HPSI flow psia gpm

15. 151.7 315. 130.0 615. 103.7 815. 81.3 1015. 48.7 1115. 15.3 1125. 5.7 1129. 0.0 INTACT_LOOP3

• RCS pressure HPSI flow psia gpm

15. 0.0 315. 0.0 615. 0.0 815. 0.0 1015. 0.0 1115. 0.0 1125. 0.0 1129. 0.0

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-14 AREVA NP Inc. Table 3-3 Statistical Distributions Used for Process Parameters1 Parameter Operational Uncertainty Distribution Parameter Range Pressurizer Pressure (psia) Uniform 2210 - 2290 Pressurizer Liquid Level (percent) Uniform 62.6 - 68.6 SIT Liquid Volume (ft3) Uniform 1090.0 - 1170.0 SIT Pressure (psia) Uniform 214.7 - 294.7 Containment Temperature (°F) Uniform 115.5 - 124.5 Containment Volume ( ft3) Uniform 2.461E+6 - 2.637E+6 Initial RCS Flow Rate (Mlbm/hr) Uniform 140.8 - 164.6 Initial RCS Operating Temperature (Tcold) (°F) Uniform 548.0 - 554.0 RWST Temperature for ECCS (°F) Point 104 Offsite Power Availability2 Binar y 0,1 Delay for Containment Spray (s) Point 0 LPSI Pump Delay (s) Point 19.5 (w/ offsite power) 30.0 (w/o offsite power) HPSI Pump Delay (s) Point 19.5 (w/ offsite power) 30.0 (w/o offsite power) 1 Note that core power is not sampled, see Section 1.0. 2 This is no longer a sampled parameter. One set of 59 cases is run with LOOP and one set of 59 cases is run with No-LOOP.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-15 AREVA NP Inc. Table 3-4 SER Conditions and Limitations SER Conditions and Limitations

Response

1.

A CCFL violation warning will be added to alert the analyst to CCFL violation in the downcomer should such occur. There was no significant occurrence of CCFL violation in the downcomer for this analysis. Violations of CCFL were noted in a statistically insignificant number of time steps.

2.

AREVA NP h as agr eed th at it is not to u se no dalization with hot leg to downcomer nozzle gaps. Hot leg nozzle gaps were not modeled.

3.

If AREVA NP applies the RLBLOCA methodology to plants using a higher planar linear heat generation rate (PL HGR) than used in the curre nt analysis, or if the method ology is to be applied to an end-of-life an alysis for which the pin pressure is s ignificantly higher, then th e need for a blowdown clad rup ture mo del will be re evaluated. T he evaluation m ay be based on r elevant eng ineering experience a nd sho uld be docum ented in eit her th e RLBLOCA guideline or plant specific calculation file. The PLHGR for St Lucie Unit 1 is lower than that used in the development of the RLBLOC A EM (Reference 1). An end-of-life calc ulation was not pe rformed; thus, the need for a blowdown cladding rupture model was not reevaluated.

4.

Slot breaks on the top of the pipe have not been evaluated. These breaks could cause the loop seals to refill during late reflood a nd the c ore to u ncover ag ain. T hese br eak locations ar e an o xidation c oncern as o pposed to a P CT concern since the top of the core can remain uncovered for extended per iods of time. Sh ould an a nalysis b e performed for a plant with s pillunder (T op crossover pipe (ID) at the crossover pipes lowest elevation) that are below the top e levation of the c ore, AREVA NP will evaluate the effect of the deep l oop seal on the sl ot breaks. T he evaluation m ay be based on r elevant eng ineering experience a nd sho uld be docum ented in eit her th e RLBLOCA guideline or plant-specific calculation file. For St Lucie unit 1, the elevation of the cross-over piping top (ID) relative to the cold leg center line is -57 inches, and the elevation of the top of the act ive core relative to the cold leg center l ine is -66.235 i nches. T herefore, n o ev aluation i s required.

5.

The model ap plies to 3 a nd 4 loop Westi nghouse-an d CE-designed nuclear steam systems. St Lucie Unit 1 is a CE-designed 2X4 loop plant.

6.

The model a pplies to bottom reflood plants only (cold side injection into the cold legs at the reactor coolant discharge piping). St Lucie Unit 1 is a bottom reflood plant.

7.

The model is v alid as long as blowdown quench does not occur. If blo wdown quench occurs, additional j ustification for the blo wdown heat transfer model a nd uncertainty are needed or th e ca lculation is correct ed. A b lowdown quench is characterized by a temperature reduction of t he peak cla dding temper ature (PCT ) node to sat uration temperature during the blowdown period. The limiting case did not show any evidence of a blowdown quench.

8.

The reflood m odel applies to bottom-up quench behavior. If a top-down quench occurs, the model is to be justifi ed or corrected to remove top q uench. A top-do wn q uench is characterized by the quench front moving from the top to the bottom of the hot assembly. Core quench initi ated at the b ottom of the core an d proceeded upward.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-16 AREVA NP Inc. Table 3-4 SER Conditions and Limitations (Continued) SER Conditions and Limitations

Response

9.

The model d oes not d etermine whether Criterion 5 of 10 CFR 50.46, lo ng ter m cooling, has bee n satisfied. This w ill b e determined by each applicant or lic ensee as part of its application of this methodology. Long-term cooling was not evaluated in this analysis.

10. Specific guidelines must be u sed to dev elop the pla nt-specific nod alization. Deviatio ns from the reference plant must be addressed.

The nodaliz ation in the p lant model is con sistent with the CE-desi gned 2X4 lo op sam ple ca lculation that was s ubmitted to the NRC for revi ew. Figure 3-1 shows the loop noding used in this analysis. (Note only Loop 1 is shown in the figure; Loops 2 and 3 are identical to loop 1, except that only Loop 1 contains the pressurizer and the break.) Fig ure 3-2 shows the steam ge nerator mod el. Figures 3-3, 3-4, and 3-5 sh ow t he react or vessel noding diagrams.

11. A tabl e that contains t he plant-specif ic parameters a nd the ra nge of the valu es considered for the se lected parameter during the topic al rep ort approv al pr ocess must be provided. When p lant-specific paramete rs are outside the range used in demonstrating acceptable code performance, the licensee or applicant will submit sensi tivity studies to show the effects of that deviation.

Simulation of clad temp erature r esponse is a function of phenomenological correlations that have be en derived either analytically or experimentally. The important correlations have been validated for the RLBLOCA methodology and a statement of the range of applicability has been documented. The correlations of interest are the set of heat transfer correlations a s descri bed i n R eference 1. T able 3-7 pres ents th e summary of th e full range of applicability for the important heat transfer correlations, as well as th e ranges calculated in th e limiting case of this analysis. Cal culated val ues for other par ameters of i nterest are also provided. As is evident, the plant-spec ific parameters fall within the methodologys range of applicability.

12. The licensee or applicant using the approved methodology must submit the results of the plant-specific ana lyses, incl uding the calculated worst break s ize, PCT, and local and total oxidation.

Analysis results are discussed in Section 3.5.

13. The licens ee or a pplicant wishing to a pply AREVA NP realistic large break loss-of-coolant accident (RLBLOCA) methodology to M5 clad fuel must request an ex emption for its use until the planned rulemaking to modify 10 CFR 50.46(a)(i) to inc lude M5 cl adding material has been completed.

Not applicable.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-17 AREVA NP Inc. Table 3-5 Summary of Results for the Limiting PCT Case Case # 3 1 PCT Temper ature 1672 F Time 26.6 s Elevation 3.406 ft Metal-Water Reaction Percent Oxidation Maximum 0.6517 Percent Total Oxidation 0.0381 Table 3-6 Calculated Event Times for the Limiting PCT Case Event Time (s) Break Opened 0.0 RCP Trip N/A SIAS Issued 1.0 Start of Broken Loop SIT Injection 14.6 Start of Intact Loop SIT Injection (Loops 2, 3 and 4 respectively) 17.3, 17.3 and 17.3 Broken Loop LPSI Delivery Began 31.0 Intact Loop LPSI Delivery Began (Loops 2, 3 and 4 respectively) N/A, N/A and 31.0 Broken Loop HPSI Delivery Began 31.0 Intact Loop HPSI Delivery Began (Loops 2, 3 and 4 respectively) 31.0, 31.0 and N/A Beginning of Core Recovery (Beginning of Reflood) 26.9 Broken Loop SIT Emptied 58.1 Intact Loop SITs Emptied (Loops 2, 3 and 4 respectively) 56.1, 58.6 and 60.9 PCT Occurred 26.6 Transient Calculation Terminated 553.5

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-18 AREVA NP Inc. Table 3-7 Heat Transfer Parameters for the Limiting Case

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-19 AREVA NP Inc. Table 3-8 Containment Initial and Boundary Conditions Containment Net Free Volume (ft3) 2,460,780 - 2,636,550 Initial Conditions Containment Pressure (nominal) 14.7 psia Containment Temperature 115.5 ºF - 124.5 ºF Outside Temperature 38 ºF Humidity 1.0 Containment Spray Number of Pumps operating 2 Spray Flow Rate (Total, both pumps) 9,000 gpm Minimum Spray Temperature 36 ºF Fastest Post-LOCA initiation of spray 0 s Containment Fan Coolers Number of Fan Coolers Operating 4 Minimum Post Accident I nitiation Time of Fan Coolers (sec) 0 Fan Cooler Capacity (1 Fan Cooler) Containment Temperature (F) 60 120 180 220 264 Heat Removal Rate (BTU/sec) 0 3472 8865 13,933 25,000

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-20 AREVA NP Inc. Table 3-9 Passive Heat Sinks in Containment1 Heat Sink Area (ft2) Thickness (ft) Material Containment Shell 86700 0.1171 C Steel Floor Slab 12682 20.0 Concrete Misc Concrete 87751 1.5 Concrete Galvanized Steel 130000 130000 0.0005833 0.01417 Zinc C Steel Carbon Steel 25000 0.03125 C Steel Stainless Steel 22300 0.0375 S Steel Misc Steel 40000 0.0625 C Steel Misc Steel 41700 0.02083 C Steel Misc Steel 7000 0.17708 C Steel Imbedded Steel 18000 18000 0.0708 7.07 C Steel Concrete Sump (GSI-191) 7414 0.02895 C Steel Material Properties Thermal Conductivity (BTU/hr-ft-oF) Volumetric Heat Capacity (BTU/ft3-oF) Concrete 1.0 34.2 Carbon Steel 25.9 53.57 Stainless Steel 9.8 54.0 Galvanizing 64.0 40.6 1 Passive heat sinks data listed in the table were used for RLBOCA analysis. Sensitivity studies were previously performed for the AREVA RLBLOCA Transition Package as applied to EMF-2103 to respond to the NRCs concerns. The results showed for a large dry containment, the PCT is not sensitive to change in containment back pressure. Hence, the heat sinks changes within 5% range will not change the presented RLBLOCA results.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-21 AREVA NP Inc. Figure 3-1 Primary System Noding

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-22 AREVA NP Inc. Figure 3-2 Secondary System Noding

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-23 AREVA NP Inc. Figure 3-3 Reactor Vessel Noding

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-24 AREVA NP Inc. Figure 3-4 Core Noding Detail

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 3-25 AREVA NP Inc. Figure 3-5 Upper Plenum Noding Detail

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 3-26 One-Sided Break Area (ft'/side) 1.0 2.0 3.0 4.0 5.0 Burn Time (hours) I ** *** ************** 0.0 5000.0 10000.0 15000.0 Core Power (MW) LHGR (KW/ft) ASI Pressurizer Pressure (psia) Pressurizer liquid Level (Ok) RCS (Tcold) Temperature ('F) t. :. :'.,. : : :. _ 3028.0 3028.5 3029.0 3029.5 3030.0 3030.5 3031.0 ~

..:...~..:

,..... J 14.4 14.6 14.8 15.0 15.2 15.4 ~ ~_..... ~ -0.1 0.0 0.0 0.1 0.1 t :.... ~:,..~~-__ 1 2200.0 2220.0 2240.0 2260.0 2280.0 2300.0 ~_.:-.-:..~..... 62.0 63.0 64.0 65.0 66.0 ~....-

-_.-.~...~ ::

548.0 550.0 552.0 554.0 Figure 3-6 Scatter Plot of Operational Parameters AREVA NP Inc.

SIT Pressure (psia) St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding L~~:l~ ~_._~._L...:-_. 1 1400 150.0 160.0 170.0 S~;~~~:;d ~ -_. -~--~--. 1 1080.0 1100.0 1120.0 1140.0 1160.0 1180.0 ~ ,_.:~._--- ---~ 200.0 220.0 240.0 260.0 2800 300.0 C~\\~~:'"t ~ - ~ -_. -, *** _~ 1 2.45e+06 2.50e+06 2.55e+06 2.60e+06 2.65e+06 T,m~"~~,"" L:,::~~ ~.~, :, 1 110.0112.0114.0 116.0118.0120.0122.0124.0 1260128.0 130.0 Figure 3-6 Scatter Plot of Operational Parameters (Continued) AREVA NP Inc. ANP-2903(NP) Revision 000 Page 3-27

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding PCT vs Time of PCT 2000 1800 D 1600 D D 1400 D CL ~ 0,-", 1200 I-0 D... 0 1000 0 800 I

• Split Break cl o Guillotine Break 600 ANP-2903(NP)

Revision 000 Page 3-28 400 a 100 200 300 Time of peT (5) 400 500 Figure 3-7 PCT versus PCT Time Scatter Plot from 59 Calculations AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding peT vs One-sided Break Area 2000 ,----~----,-----~-----,------.----------,-----.--------, 1800 0 ~

• • 0 1600 o.0 0

~** I ~

• 0*.80 nO 0

0 1400 o 0 0 E cPO 0 I-1200 0 D.- o 1000 0 800 ANP-2903(NP) Revision 000 Page 3-29 600 I

• Split Break I

o Guillotine Breakl 5.0 400 L--_-----'--__---'I__-----'---__---'-I__~__...LI______'_____ 1.0 2.0 3.0 4.0 Break Area (fe/side) Figure 3-8 PCT versus Break Size Scatter Plot from 59 Calculations AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Maximum Oxidation vs peT 2.0 ANP-2903(NP) Revision 000 Page 3-30 1.8 1.6 1.4 1.2 cg 1.0 ro "0'xo 0.8 0.6 0.4 0.2 I

• Split Break I

D Guillotine Breakl D 0.0 400 600 800 1000 1200 peT (oF) 1400 1600 1800 2000 Figure 3-9 Maximum Oxidation versus PCT Scatter Plot from 59 Calculations AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Total Oxidation vs peT ANP-2903(NP) Revision 000 Page 3-31 0.08 I

• Split Break I

D Guillotine Breakl 0.06 '{i ~ c: 0 co -0 D 'x 0 ~cw. D 0.04 ~ D* 0.02 l~ I I D 0.00 ~ ~ Irnrn LJ L-' 1200 400 600 800 1000 1400 1600 1800 2000 peT CF) Figure 3-10 Total Oxidation versus PCT Scatter Plot from 59 Calculations AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding PCT Trace for Case #3 peT =1671.8 of, at Time =26.63 s, on 6% Gad Rod ANP-2903(NP) Revision 000 Page 3-32 2000 1500 o~ Q)'-

l

+-' co'- Q) 0.. E 1000 Q) l- +-'c*0 Q.. ..c (f) Q) 2: 500 oo 200 Time (s) 400 600 Figure 3-11 Peak Cladding Temperature (Independent of Elevation) for the Limiting Case AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Break Flow 80 -- Vessel Side

---

Pump Side

-- - Total 60 ANP-2903(NP) Revision 000 Page 3-33 "'0 ~ ""(j) 40 --E ..0 ~ Q).... co 0:: 5 0 LL 20 o -20 o 200 Time (5) 400 600 10:22950 160ec2009 15:31:36 R50MX Figure 3-12 Break Flow for the Limiting Case AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Core Inlet Mass Flux 1000 -- Hot Assembly - - - - Surround Assembly - - - Average Core Outer Core 500 if) I N~ E ..0 X

l L!-

V) V) co ~ 0 d ANP-2903(NP) Revision 000 Page 3-34 -500 o 200 Time (s) 400 600 10:22950 160ec2009 15:31 :36 R50MX Figure 3-13 Core Inlet Mass Flux for the Limiting Case AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 3-35 Core Outlet Mass Flux 900 -- Hot Assembly - - - - Surround Assembly - - - Average Core Outer Core 700 500 UJ I N~ E 300 .0

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St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Pump Void Fraction ANP-2903(NP) Revision 000 Page 3-36 1.0 0.8 0.6 c .Q "0 co l.L "'0

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0.4 0.2 0.0 o 200 Time (5) Broken Loop 1 Intact Loop 2 Intact Loop 3 1ntact Loop 4 400 600 10:22950 160ec2009 15:31:36 R50MX Figure 3-15 Void Fraction at RCS Pumps for the Limiting Case AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 3-37 ECCS Flows 3000 Loop 1 (broken) Loop2 Loop3 Loop 4 600 400 ---r-- 200 I' 1\\ I,

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St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Upper Plenum Pressure 3000 2000 ro .iii 8 Q)

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~ 0.. 1000 ANP-2903(NP) Revision 000 Page 3-38 oo \\ 200 Time (5) 400 600 10:22950 160ec2009 15:31:36 R50MX Figure 3-17 Upper Plenum Pressure for the Limiting Case AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Downcomer Liquid Level 30 -- Sector 1 (broken) ........... Sector 2

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--- Sector 4 -- Average 1\\ 20 I! g III I Q) I II I Q) ...J II III "'0

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St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 3-40 Lower Vessel Liquid Level 14 12 10 ( tv §: 8 (I)> (I) --.J 6 4 2 600 400 200 OL----~-----'-------~------'-------~L---------' o Time (5) 10:22950 160ec2009 15:31:36 R50MX Figure 3-19 Collapsed Liquid Level in the Lower Plenum for the Limiting Case AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 3-41 Core Liquid Level 15,----------,----,------------r------.---------,,--------, -- Hot Assembly - - - - Center Core - - - Average Core Outer Core 600 400 200

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St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 3-42 Containment and Loop Pressures 100 ,-----,--~-------,,------~--------,---~--------, 90 80 -- Containment - - - - SG Outlet (primary side) - - - Upper Plenum Oowncomer Inlet 70 60 ro

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St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 3-43 2200 ---

---

- ------ 2200 I.Zr4LOOP

[JZr4No-LOOP I 2000 2000 1800 ~-- 1800 1600 1400 C

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ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-1 AREVA NP Inc. 4.0 Generic Support for Transition Package The following sections are responses to typical RAI questions posed by the NRC on EMF-2103 Revision 0 plant application

s. In some i nstances, t hese requests cross-referenced documentation provided on dockets other than those for w hich the request is made. AREVA discussed these and similar questions from the NRC draft SER for Revisi on 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.

4.1 Reactor Power Question: It is indicated in the RLBLOCA analyses that the assumed reactor core 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 indicate d in Item 2. 1 of Table 3 -2 herein, th e assumed reactor core power for the St Lucie Unit 1 Re alistic Large Break Loss-of-coolant Accident is 3029 MWt. The value represents the 10% power uprate and 1.7 % measurement uncertainty recapture (MUR) relative to the current rated thermal power (2700 MWt) plus 0.3% power measurement uncertainty. 4.2 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 °F before it allows rod quench? Response: Yes, the version of S-RELAP5 e mployed for the St Lucie Unit 1 requires that both the void fraction is le ss than 0. 95 and the clad temp erature is less than t he minimum temperature for film boiling heat tra nsfer (T min) before the rod is allowe d to quench. T min is a sampled parameter in the RLBLOCA methodology that typically does not exceed 755 K (900

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-2 AREVA NP Inc. oF). This is a change to the approved RL BLOCA EM (Re ference 1). This feature is carried forward into the UNOV07 version of S-RELAP5. 4.3 Rod-to-Rod Thermal Radiation Question: Provide justification that the S-RELAP5 rod-to-rod thermal radiation model applies to the St Lucie Unit 1 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: f ilm boiling, convection to steam, rod to liquid radiation and r od to vapor radiation. T his heat tra nsfer packa ge was assessed against various experimenta l data sets involving both moderate (1600 °F - 2000 °F) and high (2000 °F to over 2200 °F) p eak cladding temperatures and shown to be conservative when applied nominally. The normal d istribution of the experimental data was then determined. Du ring the exe cution of an RLBL OCA evaluation, the heat tra nsferred fro m a fuel rod is determined by the applicat ion of a multiplier to the nominal heat transfer model. This m ultiplier 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 radiation, it is rea sonable to conclude th at the mode ling implicitly includes an allocatio n for rod-to-rod radiation effects. As will be demonstrated, t he approach is reason able 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 °F and the THTF tests covered a range of PCTs from 1,000 to 2,200

°F. Since the test bun dle 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 refloo d 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 assessme nt 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 ben chmarks were selecte d from the Cylindrical Core Test Fa cility (CCTF), LOFT, and the Semis cale facilitie s. Because these facilitie s are more integral tests an d

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-3 AREVA NP Inc. together cover a wide range of scale, they also serve to show that scale effects ar e 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 figu res demonstrate that the comparison betwe en 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-REL AP5 reflood heat transf er 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 °F. At these temperatures, there is little rod-to-rod radiation. Given the good agreement between the biased co de calculations and the CCTF, LOF T, and Semiscale dat a, it can be conclu ded that ther e is no significant ove r prediction of the total heat transfer coefficient. Notwithstanding any co nservatism evidenced b y experime ntal 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 an d ORNL Thermal Hydraulic Test Facility (THTF), used to assess the S-RELAP5 he at transfer model, Ref erence 1, incorporated constant rod powers across the experimental assembly. Temperature differences that occurred were t he result of guide tube, shroud or local heat transfer effects. In the operation of a pressurized water re actor (PWR) and in the RLBLOCA evaluation, a radial local peaking factor is present, creat ing power differences that tend to enhance the temperature differences between rods. In turn, these temperature differences lead to increase s in net ra diation heat transfer from the hott er rods. T he expected rod-to-ro d radiation will likely exceed that embodied within the experimental results. 4.3.1 Assessment of Rod-to-Rod Radiation Implicit in the RLBLOCA Methodology As discussed above, the FLECHT -SEASET a nd THTF tests were selected t o assess and determine the S-RELAP5 code heat transfer bias and uncertainty. Uniform radial power distribution was used in these te st bundles. T herefore, the rod-to-rod temperature variation in the rods away fro m the vessel wall is caused primarily b y the variation in the sub-channel fluid conditions. In the real operating fue l bundle, on the other hand, there can be 5 to 10 percent

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-4 AREVA NP Inc. rod-to-rod power variation. In add ition, the methodology includes a provision to apply the uncertainty measureme nt to the h ot pin. Ta ble 4-1 pro vides the hot pin measurement uncertainty and a representative lo cal pin pea king factor for several plants. Th ese 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 loca l peaking assessment is required. Therefore, the plant rod-to-rod radiation asse ssments herein set the average pin power for those pins surrounding the hot pin at 96 percent of that of the pea k pin. For pins further removed the average power is set to 94 percent. Table 4-1 Typical Measurement Uncertainties and Local Peaking Factors Plant F H Measurement Uncertainty (percent) Local Pin Peaking Factor (-) 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 4.3.2 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 1 1 and is sim ilar to that d eveloped in the HUX Y rod heatup code (Reference 10, Section 2.1.2) used by AREVA for BW R LOCA applications. The version of R2RRAD used herein was obtained from t he NRC to exa mine the rod-to-rod radiation characteristics of a 5x5 rod segme nt of the 16 1 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 limitatio n affects t he view factor calcula tions for g uide tubes, R2RRAD is a reasonable tool to estimate rod-to-rod radiation heat transfer.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-5 AREVA NP Inc. The FLECHT-SEASET test series was intended to simulate a 17x17 fuel assembly and there is a close similarity, Table 4-2, between the test bundle and a modern 17x17 assembly. Table 4-2 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 t ests (Reference 6) were selected for evaluation and comparison with expected plant behavior. Table 4-3 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. Figure 4-1 R2RRAD 5 x 5 Rod Segment Two sets of runs were made simul ating each of the five experiments and one set of cases was run to simulate the RLBLOCA e valuation of a limiting fuel a ssembly in an operatin g plant. For the simulation of Test s 31805, 3 1504, 3102 1, and 308 17, the thimble tube (guide tube ) temperatures were set to the measured values. For Te st 34420, the thimble tube temperature was set equal to the measured vapor temperature. For the first experime ntal simulation set, the temperature of all 21 rods and the exterior boundary wa s set to the measured PCT of th e simulated test. For the second experimental set, the hot ro d temperature was set t o the PCT Guide Tube Hot Rod Adjacent Rods

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-6 AREVA NP Inc. value and t he remaining 20 rods and the bou ndary were set to a temperature 25 °F cooler providing a reasonable measure of the variation in surround ing temperatures. To e stimate the rod-to-rod radiation in a real fu el assembly at LOCA conditions and compare it to the experimental results, ea ch of the a bove cases was rerun with the h ot rod PCT set to the experimental result and the remaining rods con servatively set to temperatures expected within the bundle. The guide tubes (thimble tubes) were removed for conservatism and because peak rod powers frequently occur at f uel assemb ly corners away from either guide tubes o r 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 temp erature 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 F. Figure 4-2 shows the hot rod therma l radiation heat transfer for the two FLECHT-SEASET sets and for the plant set. T he figure shows that for PCTs great er than abo ut 1700 °F, the hot rod thermal radiation in th e plant cases exceed s that of t he same component within the experiments. Table 4-3 FLECHT-SEASET Test Parameters Test Rod 7J PCT at 6-ft (°F) PCT Time (s) htc at PCTtime (Btu/hr-ft2-°F) Steam Temperature -at 7I (6-ft) (°F) Thimble Temperature at 6-ft (°F) 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

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-7 AREVA NP Inc. Figure 4-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 FA 4.3.3 Rod-to-Rod Radiation Summary In summary, the conservatism of th e heat transfer modeling established by bench mark can be reasonably extended to plant applications, an d the plant local peaking provides a physical reason why rod-to-rod radiation sho uld be more substantial within a plant environment than i n 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 be en demonstrated to me et the criteria of 10 CFR 50.46 with a high level of probability. 0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 1400 1500 1600 1700 1800 1900 2000 2100 2200 2300 2400 PCT (°F) Radiation HTC (BTU-hr/ft^2-°F) FLECHT_SEASET set-1 FLECHT_SEASET Set-2 Fuel Assembly

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-8 AREVA NP Inc. 4.4 Film Boiling Heat Transfer Limit Question: In the St Lucie Unit 1 Cycle 24 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 greater than or equal to 0.9? Response: Yes, the version of S-RELAP5 employed for the St Lucie Unit 1 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 t o 15 percent or less. This is a change to the approved RLBLOCA EM (Reference 1). This feature is carried forward into the UNOV07 version of S-RELAP5. 4.5 Downcomer Boiling Question: If the PCT is greater than 1800°F or the containment pressure is less than 30 psia, has the St Lucie Unit 1 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 1 ha s been esta blished gen erically as adequate fo r the computation of d owncomer phenomena including the predictio n of potential local boiling effects. The model was benchmarked against the UPTF tests and the LOFT facility in the RLBLOCA meth odology, Re vision 0 (Reference 1). Further, AREVA add ressed 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 sen sitivity studies on high and low pressure co ntainments, the impact of additional azimuthal noding within the downcomer, and the influence of flow loss coefficients. Of these, the study on azimuthal noding is most germane to this question; indicating that additional azimuthal nodalization allows higher liquid buildup in portion s of the do wncomer away from the broken co ld leg and increases th e liquid driving head. Additionally, AREVA has c onducted downcomer axial noding and wall heat release stu dies. Each of these studies supports the Revision 0 methodology and is documented later in this section. This que stion is primarily concern ed with the phenomena of downcomer boilin g and the extension of the Revision 0 methodology and sensitivity stu dies to plants with low containment

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 4-9 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). ~ ~ECC ( 1\\. 0 0 n1b~core \\ /'tr--CHY{ JD~(){.:p--{;

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~b,dC Figure 4-3 Reactor Vessel Downcomer Boiling Diagram 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 accumulator injection. At this time, there is a large step reduction in coolant flow from the ECC systems. As AREVA NP Inc.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-10 AREVA NP Inc. a result, co olant entering the downcomer ma y be less subcooled. When the downcomer coolant approaches saturation, boiling on the wal ls initiates, reducing the downcomer hydraulic static level. With the reduction of the downcomer level, the core inlet f low 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 down comer boiling may impact clad t emperatures, it is so mewhat of a self-limitin g process. If cladding t emperatures increa se, less energ y is transfe rred in the core boilin g process an d the loop steam flows are reduce d. This re duces the r equired driving head t o 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 o n the wall h eat 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 th e larger the impact on core reflood ing. Similarly, the quicker the pa ssage of st eam up the downcome r, the less resident volume within t he downcomer is occupied by steam and the lower the impact on the downcomer average d ensity. Therefore, the ability to properly simulate downcomer boilin g depends on both the heat release (boiling) model and on the ability to track steam rising through the downco mer. Consideration of both of the se is provided in the fo llowing text. The hea t 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 a ccuracy within the RLBLOCA meth odology is sufficien t 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 cladd ing oxidation rates. 4.5.1 Wall Heat Release Rate The downcomer wall he at release rate during reflood is co nduction limited and depends on the vessel wall mesh spacing used in the S-RELAP5 model. The following two approaches are used

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-11 AREVA NP Inc. to evaluate the adequacy of the downcomer ve ssel wall m esh spacing used in the S-RELAP5 model. 4.5.1.1 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 me sh spacing used in S-RELAP5, it is assumed that the material has con stant thermal properties, is initially at temperature T i, and, at time zero, h as one surface, the surface simulating contact with the do wncomer 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*( t)0.5)}, (1) where,  is the thermal diffusivity of the material given by  = k/( Cp), k = thermal conductivity,  = 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 T i = 500 oF and T o = 300 oF. The mesh spacing in S-RELAP5 is the same a s that used for the downcomer vessel wall in the RLBL OCA model. Figure 4-4 shows the temperature distributions in the met al at 0.0, 100 and 300 seconds as calculated by using Equation 1 and S-REL AP5, respectively. The solutions are identica l confirming the adequacy of the mesh spacing used in the downcomer wall.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-12 AREVA NP Inc. Figure 4-4 S-RELAP5 versus Closed Form Solution 250 300 350 400 450 500 550 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 Distance from Inner Wall, feet Metal Temperature, F Closed Form, 0 s Closed Form, 100 s Closed Form, 300 s S-RELAP5, 0 s S-RELAP5, 100 s S-RELAP5, 300 s

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding 4.5.1.2 Plant Model Sensitivity Study ANP-2903(NP) Revision 000 Page 4-13 As additional verification, a typical 4-100p 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, peT 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. 30000.00,----,---_-----,--;,"-' 24000.00 f---+-------j'--- 0' Q) C/) 18000.00

l co Q)

C/) co 12000.00 Q)a; 0::: 01 Q)c W 6000.00 co S 0'000*~.0L.--~---::80'-::-.0--~--1~60:-::-.0--~----""2-L40"'-.0--~--32,.L0.-0--~------.J400.0 Time (sec) Figure 4-5 Downcomer Wall Heat Release - Wall Mesh Point Sensitivity AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 4-14 400.0 320,0 240.0 160.0 80,0 I=:::: ~::~~:hL Wall (9-meSh1 AI..v." VSLWall .rJ ~ -V- ~ '~~ ~ ~". ~ 0.000.0 600.00 2400.00 1800,00 u.. 0 ()) I-

l CO I-1200.00

()) 0.. E ()) f-Time (sec) Figure 4-6 peT Independent of Elevation - Wall Mesh Point Sensitivity AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 4-15 30.00 r---~---,.....--~----,----~----,------~---.---~-----.., 20.00 rt-------+Ii:--IIIfll-+*------fM7nh-----,-------,--+-h...,.......,.,- 10.00 r----~--m-____+--+-------+_------+-------_I_------_____J 0.00 ';;---~--____=_~-------'--------:-:::_:_--~--__=--::-:--~~--__L---"------,J 0.0 80.0 160.0 240.0 320.0 400.0 Time (sec) Figure 4-7 Downcomer Liquid Level - Wall Mesh Point Sensitivity AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 4-16 12.00 '---~--...,----~----,-- ~__--, 10.00 t-------+------+-------l------j1===~:!~:!5'~~'~~!::::j--l

- )1---------+------+-------1---------+-------1 Q)

~ 4.00 m------+--j-----+-------t-------+-------I 2.00 1tt-I:-----,F,--+------+-------l--------+--------1 0.00 ';;----=..:..,.,""-----::-:;----~-_____:-=':::----'---~l,_:__--~--~_,____--~--...J 0.0 80.0 160.0 240.0 320.0 400.0 Time (sec) Figure 4-8 Core Liquid Level - Wall Mesh Point Sensitivity AREVA NP Inc.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-17 AREVA NP Inc. 4.5.2 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. 4.5.2.1 Azimuthal Nodalization In a letter to the NRC d ated April, 2003 (Refere nce 1), AREVA docume nted several studies on downcomer boiling. Of significa nce here is the study on further azimuthal break up of the downcomer noding. T he study, based on a 3-loop plant with a co ntainment 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 azimutha l regions (Figure 4-9). The base calculation simulated the limiting PCT calculation given in th e EMF-2103 t hree-loop sample problem. This c ase was then repeated with the re vised 6 x 9 downcomer noding. The change resulted in an alteration of the b lowdown evolution of t he transient with litt le evidence of any affect during reflood. To isolate any possible reflood impact that might have an influence o n downcomer boiling, t he case wa s repeated with a sligh tly 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 t he downcomer but that reflood wo uld not be i mpacted. Although the study was performed f or 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 re asonably co nverged in its ability to represent downcomer boiling phenomena.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Base model (lH9 @ (Hy @(HY Revised 9 Region Model ~ (IHY ~Q (Hy (Cl~ (HL Figure 4-9 Azimuthal Noding 4.5.2.2 Axial Nodalization ANP-2903(NP) Revision 000 Page 4-18 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-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 AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 4-19 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 4-10 provides the containment back pressure for the base modeling. Figure 4-11 through Figure 4-14 show the total downcomer metal heat release rate, peT 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. 400.0 320.0 240.0 160.0 80.0 1-- Base6x6 ~asel o. 0.00 40.00 8.00 16.00 24.00 32.00 Time (sec) Figure 4-10 Lower Compartment Pressure versus Time AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 4-20 18000.00 r-i-------j---f-+' 12000.00 H'-------+----'------+-----~ 30000.00,-----,----,----r.r; Q) (/) co Q) Q) 0::: 01 Q)c W co S 0- ~ 24000.00 I---j-----+-----.: 6000.00 H-----+------t------j-------+---------1 0.000~.O'-----~-----;;;:80;-;;-.O--~----;';:;60--;::.0--~--7240J.:-.0::---~---3:-::-20L..,.o,---~---4...Joo.O Time (sec) Figure 4-11 Downcomer Wall Heat Release - Axial Noding Sensitivity Study AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 4-21 2400.00 r---~--,---~---,-----~-----,---~-----,--~-----, 1800.00 l.L ~ ~ Q)

Jco 00 Q)

Cl.. E Q)I-600.00 'i~. ~1I~**'\\,/'vt_':.IL"-",'. ':.~ \\. ~ ""~' h -.,~ ~. '~ ~'.. 0.000.'n.0--~---;;8~0.0;;---~---:1~60;-;;.0--~--::;-:24~0.0::---~----=3;:l,20-=-.0 --~--4,.J00.0 Time (sec) Figure 4-12 peT Independent of Elevation - Axial Noding Sensitivity Study AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 4-22 30.00 '---~---,-----~---r----~---,-----~---,-----~----, 10.00 r----l---j!~r.;:_--+-------t_-----_+------+_-----___I 0.000';;-.0--~------:::80:-:.0--~-----:-:160':-:.0~-~---2""4':-:0.0'-----~------=-=32LO.0,-----~-----c::'4oo.0 Time (sec) Figure 4-13 Downcomer Liquid Level - Axial Noding Sensitivity Study AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 4-23 12.00 r--------,--~---r---,---___,__-----,___--~-_____, 10.00 1-----t-----I-----I-t=~~~i==:J----___l 8.00 Q3 ~ Q5> Q) .....J 6.00

Q

l 0-

.:::i 4.00 2.00 1tt-t------;f.It--+------+------+------+----------1 0.000'::-.0-----=::.e '.,."---------;;:80~.0--~-----:1=60c::-.0--~---::-24:'.:-0.0=----~--3=-='20~.0--~------:-:'400.0 Time (sec)

Figure 4-14 Core Liquid Level - Axial Noding Sensitivity Study 4.5.3 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. AREVA NP Inc.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-24 AREVA NP Inc. 4.6 Break Size Question: Were all break sizes assumed greater than 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. T his is a cha nge to the approved RLBLOCA EM (Referen ce 1). Th e larger break area LOCAs are typica lly 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 brea k). However, this transitional break size varies from plant to plant and is verified only after the break spectru m has been executed. AREVA NP has sought to develop sufficient criteria for defining the minimum larg e break flow area prior t o performing the break spectrum. The purpose for doing so is to assure a valid break spectrum is performed. 4.6.1 Break / Transient Phenomena In determining the AREVA NP cri teria, the characteristics of larger break area LOCAs are examined. These LOCA characteristics involve a rapid a nd 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 perio d from initiation of the break until flow from the accu mulators begins. This definition is somewhat different from the traditional definition of blowdown which extends the blowdown until the RCS pres sure approaches containment pressure. The blowdown phase typically la sts about 12 to 25 seconds, depending on the break size. Refill is that period that starts with the end of blowdown, whichever definition is u sed, and ends when water i s first force d upward in to 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.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-25 AREVA NP Inc. Implicit in this break-do wn is that the core liqu id inventory has been completely, o r nearly so, expelled from the prima ry system le aving the core in a stat e of near co re-wide dispersed flow film boiling and subseq uent 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 d ecreases su ch that these three pha ses overlap substant ially. During these smaller break events, the co re liquid inv entory is not reduced a s much as t hat 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. 4.6.2 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 especia lly true in a statistical evaluation where parameter values are varied randomly with a strong expectation that the variations will affe ct 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 sub stantially reduced lower plenum liquid inventory. It was rea soned that such conditions would be unlikely if the break flow rate was reduced to less than the reactor coolant pump flow. Tha t is, if the reactor coolant pumps ar e capable of forcing more coolant toward the reactor vessel than the break can e xtract from t he reactor vessel, the d owncomer and core must maintain some degree of positive flow (positive in the no rmal operations sense). The circumstance is, of course, transitory. Break flow is alt ered 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 red uced to zero or became negative i mmediately after the bre ak initiation, then the event was quite likely to proceed wit h sufficient inertia to expel most of the reactor vessel liquid to the brea k. The criteria base, thus established, consists of comparing the brea k flow to the initial f low through all re actor coolant pumps an d setting the minimum break area such that these flows match. This is done as follows: Wbreak = Abreak

• Gbreak = Npump
• WRCP.

This gives

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-26 AREVA NP Inc. 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 a nd that appropriate for the initial cold leg enthalpy and the saturation pressure of coolant at that enthalpy. Gbreak = (Gbreak(P0, HCL0) + Gbreak(PCLsat, HCL0))/2. The estimated minimum LBLOCA break area, A min, is 2.94 f t2 and the b reak area p ercentage, based on the full double-ended guillotine break total area, is 29.97 percent. Table 4-4 provides a listing of the plant type, initial con dition, and t he fractiona l minimu m RLBLOCA break area, for all the plant types presented as generic representations in the next section. Table 4-4 Minimum Break Area for Large Break LOCA Spectrum Plant Description System Pressure (psia) Cold Leg Enthalpy (Btu/lbm) Subcooled Gbreak (lbm/ft2-s) Saturated Gbreak (HEM) (lbm/ft2-s) RCP flow (lbm/s) Spectrum Minimum Break Area (ft2) Spectrum Minimum Break Area (DEGB) A 3-Loop W Design (sha) 2250 555.0 231 90 5700 31417 2.18 0.26 B 3-Loop W Design (rob) 2250 544.5 238 80 5450 28124 1.92 0.23 C 3-Loop W Design (nab) 2250 550.0 235 40 5580 29743 2.04 0.25 D 2x4 CE Design (ftc) 2100 538.8 228 60 5310 21522 1.53 0.24 E 2x4 CE Design (pal) 2055 535.8 226 30 5230 37049 2.66 0.27 F 4-loop W Design (seq2) 2160 540.9 23290 5370 39500 2.76 0.33 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 minimu m break area (A min) and an area of twice t he size of the broken pipe. Th e determination of break configuration, split versus double-ended, is made after the break area is selected based on a uniform probability for each occurrence.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-27 AREVA NP Inc. 4.6.3 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. Typica lly the lower end of the large break spectrum occurs at between 0.2 to 0.3 times t he total area of a 100 percent double-ended guillotine break (DEGB) and the upper end of the small break spectrum occurs at a pproximately 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 a ccidents th at critically challenge t he emergency core cooling systems (ECCS). Breaks within this rang e remain large enough to blowdown to low pressures. Resolution is provided by the large br eak ECC systems and t he pressure-dependent injection limitations that determine critical small break perf ormance 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 surve yed. Although statist ical determinations are extracted from the consideration o f breaks with areas above the intermediate range, the A REVA best-estimate methodology remains suitable to characterize the ECCS performance of breaks within the intermediate range. Table 4-4 provides a listing of the plant type, initial con dition, and t he fractiona l minimu m RLBLOCA break area. Figure 4-15 through Figure 4-20 provide the enlarged break spectru m results with the upper end of the small break spectrum and the lower end of the large break spectrum indicated by bars. Table 4-5 provides differences between the true large break region and the inte rmediate break region ( break areas between that of the lar gest SBLOCA and th e smallest R LBLOCA). The minimum difference is 14 1 °F; however, this case is no t representative of the g eneral tren d shown b y the other compariso ns. The n ext minimu m difference is 704 °F (see Figure 4-15). Considering this point as an outlier, the table shows the minimum difference bet ween the highest inter mediate break spectrum PCT and large break spectrum PCT, for the six plants, is at lea st 463 °F, and including this point would provide a n average difference of 427 °F and a maximum difference of 840 °F. Thus, by bo th measures, the pea k cladding temperatures within the in termediate break range will be several hundred degrees below those in the true lar ge break ra nge. Therefore, these breaks will not provide a limit or a critical measure of the ECCS performance. Gi ven that the

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-28 AREVA NP Inc. large break spectrum bounds the intermediate spectrum, the use of only the large break spectrum meets the re quirements of 10CFR50.46 for br eaks within the intermediate brea k LOCA spectrum, and the method demonstrates that the ECCS for a plant meets the criteria of 10CFR50.46 with high probability. Table 4-5 Minimum PCT Temperature Difference - True Large and Intermediate Breaks Plant Description Generic Plant Label (Table 4-4) Maximum PCT (°F) Intermediate Size Break Maximum PCT (°F) Large Size Break Delta PCT (°F) Average Delta PCT (°F) A 1746 1 1887 141 1 B 1273 1951 678 3-Loop W Design C 1326 1789 463 4271 D 984 1751 767 2x4 CE Design E 869 1636 767 767 3-loop W Design F 1127 1967 840 840 Note: 1. The 2nd highest PCT was 1183 °F. This changes the Delta PCT to 704 °F and the average delta increases to 615 °F.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-29 AREVA NP Inc. Figure 4-15 Plant A - Westinghouse 3-Loop Design 600 800 1000 1200 1400 1600 1800 2000 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 PCT (°F) Upper End of SBLOCA Break Size Spectrum Large Break Spectrum Minimum Break Area

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-30 AREVA NP Inc. Figure 4-16 Plant B - Westinghouse 3-Loop Design 600 800 1000 1200 1400 1600 1800 2000 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 PCT (°F) Upper End of SBLOCA Break Size Spectrum Large Break Spectrum Minimum Break Area

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-31 AREVA NP Inc. 600 800 1000 1200 1400 1600 1800 2000 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 PCT (°F) Upper End of SBLOCA Break Size Spectrum Large Break Spectrum Minimum Break Area Figure 4-17 Plant C - Westinghouse 3-Loop Design

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-32 AREVA NP Inc. 600 800 1000 1200 1400 1600 1800 2000 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 PCT (°F) Upper End of SBLOCA Break Size Spectrum Large Break Spectrum Minimum Break Area Figure 4-18 Plant D - Combustion Engineering 2x4 Design

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-33 AREVA NP Inc. 600 800 1000 1200 1400 1600 1800 2000 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 PCT (°F) Upper End of SBLOCA Break Size Spectrum Large Break Spectrum Minimum Break Area Figure 4-19 Plant E - Combustion Engineering 2x4 Design

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-34 AREVA NP Inc. 600.0000 800.0000 1000.0000 1200.0000 1400.0000 1600.0000 1800.0000 2000.0000 2200.0000 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 PCT (°F) Upper End of SBLOCA Break Size Spectrum Large Break Spectrum Minimum Break Area Figure 4-20 Plant F - Westinghouse 3-loop Design

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-35 AREVA NP Inc. 4.7 Detail information for Containment Model Containment initia l conditions and cooling syst em information are pro vided in Ta ble 3-8 and Heat Sinks are provided in Table 3 -9. For St Lucie Unit 1, the scatter plots of PCT versus the sampled containment volumes and initial atmo spheric temperature are shown in Figure 4-21 and Figure 4-22. Containment pressure as a function of time for limiting case is shown in Figure 4-23.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding PCT vs Containment Volume 1800 D ~ 1600 D D ~I *

• D...

D ~ **w ibJ] D

• dI D

1400 cJ D D E D eY D I-1200 0 0... D 1000 D 800 ANP-2903(NP) Revision 000 Page 4-36 600 I

• Split Break 1-D Guillotine Breakl 2.5500e+06 Containment Volume (ft3) 2.6500e+06 AREVA NP Inc.

Figure 4-21 PCT vs. Containment Volume

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Page 4-37 PCT vs Containment Temperature 2000 ,----------~-----,-------~---- 1800 0-0- 1600 cJIo III -Iio ~ ~- II. -.. 'd-o _ 00 0 -.- 1400 o~* 0 E 0 0 D f-1200 0 D... D 1000 D 800 600

• Split Break o Guillotine Break 130 120 Containment Temperature CF) 400 L--

~ 110 Figure 4-22 PCT V5. Initial Containment Temperature AREVA NP Inc.

St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Containment Pressures 1-CorUhn~1 10 nn.(B) ANP-2903(NP) Revision 000 Page 4-38 Figure 4-23 Containment Pressure as function of time for limiting case AREVA NP Inc.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-39 AREVA NP Inc. 4.8 Cross-References to North Anna Question: In order to conduct its review of the St Lucie Unit 1 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 additional information that were developed for the application of the AREVA methods to the North Anna Power Station, Units 1 and 2, 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 1 Please verify that the information in these letters is applicable to the AREVA model applied to St Lucie Unit 1 except for that information related specifically to North Anna and to sub-atmospheric containments. Response: The respon ses provided to questions 1, 2, 4, a nd 6 are generic and re lated to the ability of ICECON to ca lculate containment pressures. They are appli cable to the St Lucie Unit 1 RLBLOCA submittal. Question 1 - Completely Applicable Question 2 - Completely Applicable

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-40 AREVA NP Inc. Question 4 - Completely Applicable (the reference to CSB 6-1 should now be to CSB Technical Position 6-2). The NRC altered the identificat ion of this branch technica l position in Revision 3 of NUREG-0800. Question 6 - Completely applicable. The supplemental request and response are applicable to St Lucie Unit 1. 4.9 GDC 35 - LOOP and No-LOOP Case Sets Question: 10CFR50, Appendix A, GDC [General Design Criterion] 35 [Emergency core cooling] states that, Suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite electric power is not available) and for offsite electric power operation (assuming onsite power is not available) the system function can be accomplished, assuming a single failure. The Staff interpretation is 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 represented by a different statistical response 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 particular plant 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 statistical requirement than for just loss of offsite power.) What is your basis for assuming a 50% probability of loss of offsite power? Your statistical runs 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 NRCs 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

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 4-41 AREVA NP Inc. 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). 4.10 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 ranges of those parameters for the operated St Lucie Nuclear Plant Unit 1(SLA). This statement addresses certain programmatic requirements 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 SLA realistic larg e break loss-of-coolant accident are verified as conservative with respe ct 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 existin g plant design input and output documents by FP&L and Vendor personnel in a formal analysis input summary d ocument issued by the Analysis Vendor and

3) Formal review and approval of t he 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 th e 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 alt ernations a re formall y communicated to the Analysis Vendor by FP&L. Revisions and updates to the analysis parameters are documented and ap proved in accordance with the proce ss described above for the initial analysis.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 5-1 AREVA NP Inc. 5.0 Conclusions A RLBL OCA analysis was performed for th e St Lucie Nuclear Plant Unit 1 using NRC approved AREVA NP RLBLOCA methods (Reference 1). Analysis results show that the limiting LOOP case has a PCT of 1672oF, and a maximum oxidation thickness and hydrogen generation that fall well within regulatory requirements. The analysis supports operation a t a nominal power level of 3029 MWt (including 0.3 uncertainty), a steam generator tube plugging le vel of up to 10 percent in all steam generators, a total LHR of 15.0 kW/ft, a total peaking facto r (F Q) up to a value of 2.161, and a nuclear enthalpy rise factor (F H) up to a value of 1.749 (includin g 6% uncertainty) with no axial or burnup dep endent power peaking limit and p eak rod average exposures of up to 62,000 MWd/MTU. For large break LOCA, the three 10CFR50.46 (b) criteria pr esented in Section 3.0 are met and operation of St Lucie Unit 1 with AREVA NP-s upplied 14x14 Zircaloy-4 clad fuel is justified.

ANP-2903(NP) Revision 000 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding Page 6-1 AREVA NP Inc. 6.0 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, 4th Edition, McGraw-Hill Book Company, 1976.

10.

EMF-CC-130, 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.}}