Text

ANP-2695(NP), Rev 0, Enclosure 3, Sequoyah Nuclear Plant Unit 1, Realistic Large Break Loss of Coolant Accident Analysis
	ML081080338
Person / Time
Site:	Sequoyah
Issue date:	02/29/2008
From:	; AREVA NP
To:	; Office of Nuclear Reactor Regulation
References
	ANP-2695(NP), Rev 0
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ENCLOSURE3 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNIT 1 SEQUOYAH NUCLEAR PLANT UNIT I REALISTIC LARGE BREAK LOSS OF COOLANT ACCIDENT ANALYSIS ANP-2695(NP)

REVISION 0 FEBRUARY 2008 NON-PROPRIETARY E3-1

AREVA NP Inc.

ANP-2695(NP)

Revision 0 Sequoyah Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis February 2008

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page ii Nature of Changes Item Page Description and Justification

1. All This is a new document.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page iii Contents 1.0 Introduction ..................................................................................................................... 1-1 2.0 Sum mary ........................................................................................................................ 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 Summ ary of Analysis Parameters ................................... 3-6 3.4 SER Compliance ............................................................................................... 3-9 3.5 Realistic Large Break LOCA Results ............................................................... 3-10 4.0 Conclusions .................................................................................................................... 4-1 5.0 References ..................................................................................................................... 5-1 6.0 Addendum -Additional Information Supporting EMF-2103 Revision 0 ......................... 6-1 6.1 Reactor Power ................................................................................................... 6-1 6.2 Rod Quench ....................................................................................................... 6-2 6.3 Rod-to-Rod Thermal Radiation ......................................................................... 6-2 6.4 Film Boiling Heat Transfer Lim it ......................................................................... 6-9 6.5 Do'wncomer Boiling ............. .............................. 6-9 6.6 Break Size ........................................................................................................ 6-25 6.7 ICECON Model ................................................................................................. 6-36 6.8 Cross-References to North Anna ..................................................................... 6-36 6.9 Containment Model .......................................................................................... 6-38 6.10 GDC 35- LOOP and No-LOOP Case Sets ..................................................... 6-43 6.11. Statement ......................................................................................................... 6-44 AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page iv Tables Table 2-1 Summary of Major Parameters for Limiting Transient .............................................. 2-1 Table 3-1 Sampled LBLOCA Parameters ............................................................................... 3-11 Table 3-2 Plant Operating Range Supported by the LOCA Analysis ...................................... 3-12 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 ...................... 3-18 Table 3-7 Heat Transfer Parameters for the Limiting Case .................................................... 3-19 Table 3-8 Containment Initial and Boundary Conditions ......................................................... 3-20 Table 3-9 Passive Heat Sinks in Containment .................................... .............................. 3-21 Table 6-1 Typical Measurement Uncertainties and Local Peaking Factors.............................. 6-5 Table 6-2 FLECHT-SEASET & 17x17 FA Geometry Parameters ..................... 6-6 Table 6-3 FLECHT-SEASET Test Parameters ......................................................................... 6-7 Table 6-4 Minimum Break Area for Large Break LOCA Spectrum ......................................... 6-27 Table 6-5 Minimum PCT Temperature Difference - True Large and Intermediate B re a ks .................................................................................................................... 6-2 9 AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page v Figures Figure 3-1 Prim ary System Noding ........................................................................................ 3-22 Figure 3-2 Secondary System Noding .................................................................................... 3-23 Figure 3-3 Reactor Vessel Noding ......................................................................................... 3-24 Figure 3-4 C ore Noding D etail ................................................................................................ 3-25 Figure 3-5 Upper Plenum Noding Detail ................................................................................. 3-26 Figure 3-6 Containment Noding Diagram ..................................... 3-27 Figure 3-7 Scatter Plot of Operational Parameters ................................................................. 3-28 Figure 3-8 PCT versus PCT Time Scatter Plot from 59 Calculations .................. 3-30 Figure 3-9 PCT versus Break Size Scatter Plot from 59 Calculations .................... 3-31 Figure 3-10 Maximum Oxidation versus PCT Scatter Plot from 59 Calculations ................... 3-32 Figure 3-11 Total Oxidation versus PCT Scatter Plot from 59 Calculations .......................... 3-33 Figure 3-12 Peak Cladding Temperature (Independent of Elevation) for the Lim iting C ase ....................................................................................................... . . 3-34 Figure 3-13 Break Flow for the Limiting Case ......................................................................... 3-35 Figure 3-14 Core Inlet Mass Flux for the Limiting Case .......................................................... 3-36 Figure 3-15 Core ..................................................

Outlet Mass Fluxfor the Limiting Case 3-37 Figure 3-16 Void Fraction at RCS Pumps for the Limiting Case ............................................ 3-38 Figure 3-17 ECCS Flows (Includes Accumulator, Charging, Sl and RHR) for the Lim iting C ase ............................................................................................................. 3-39 Figure 3-18 Upper Plenum Pressure for the Limiting Case ............................................... 3-40 Figure 3-19 Collapsed Liquid Level in the Downcomer for the Limiting Case ........................ 3-41 Figure 3-20 Collapsed Liquid Level in the Lower Plenum for the Limiting Case .................... 3-42 Figure 3-21 Collapsed Liquid Level in the Core for the Limiting Case ................................... 3-43 Figure 3-22 Containment and Loop Pressures for the Limiting Case .................................... 3-44 Figure 3-23 GDC 35 LOOP versus No-LOOP Cases ............................................................ 3-45 Figure 6-1 R2RRAD 5 x 5 Rod Segment .................................................................................. 6-6 Figure 6-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 F A .................................................................................................................................. 6 -8 Figure 6-3 Reactor Vessel Downcomer Boiling Diagram ........................................................ 6-10 Figure 6-4 S-RELAP5 versus Closed Form Solution ........................... 6-13 Figure 6-5 Downcomer Wall Heat Release - Wall Mesh Point Sensitivity ............................... 6-14 Figure 6-6 PCT Independent of Elevation - Wall Mesh Point Sensitivity ................................ 6-15 Figure 6-7 Downcomer Liquid Level - Wall Mesh Point Sensitivity ........................................ 6-16 Figure 6-8 Core Liquid Level -Wall Mesh Point Sensitivity .................................................. 6-17 AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page vi Figure 6-9 Azim uthal Noding ................................................................................................. 6-19 Figure 6-10 Lower Compartment Pressure versus Time ........................................................ 6-20.

Figure 6-11 Downcomer Wall Heat Release - Axial Noding Sensitivity Study .......... 6-21 Figure 6-12 PCT Independent of Elevation - Axial Noding Sensitivity Study ........... 6-22 Figure 6-13 Downcomer Liquid Level - Axial Noding Sensitivity Study .................................. 6-23 Figure 6-14 Core Liquid Level - Axial Noding Sensitivity Study ............................................. 6-24 Figure 6-15 Plant A - Westinghouse 3-Loop Design ........................... 6-30 Figure 6-16 Plant B - Westinghouse 3-Loop Design .............................................................. 6-31 Figure 6-17 Plant C - W estinghouse 3-Loop Design .............................................................. 6-32 Figure 6-18 Plant D - Combustion Engineering 2x4 Design .................................................. 6-33 Figure 6-19 Plant E - Combustion Engineering 2x4 Design ................................................... 6-34 Figure 6-20 Plant F - W estinghouse 4-Loop Design .............................................................. 6-35 Figure 6-21 Energy Addition in Lower Compartment ........................... 6-39 Figure 6-22 Energy Rates in Lower Compartment ................................................................. 6-39 Figure 6-23 Energy Removal Rates in Lower Compartment ........................ .. 6-40 Figure 6-24 Energy Removal Rates in Upper Compartment ................................................. 6-40 Figure 6-25 Heat Removal Rates (log) .................................... 6-41 Figure 6-26 Fraction of Ice Rem aining ................................................................................... 6-4 1 Figure 6-27 Mass Addition to Lower Compartment ................................................................ 6-42 Figure 6-28 Upper Compartment versus Lower Compartment Pressure ................................. 6-42 Figure 6-29 Temperature of Upper and Lower Compartments ............................................... 6-43 This document contains a total of 103 pages.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larqe Break LOCA Analysis Page vii Nomenclature AFD Axial Flux Difference BLEU Blended Low Enriched Uranium CFR Code of Federal Regulations CCTF Cylindrical Core Test Facility CSAU Code Scaling, Applicability, and Uncertainty DEGB Double-Ended Guillotine Break DNB Departure from Nucleate Boiling ECCS Emergency Core Cooling System EFPH Effective Full Power Hours EM Evaluation Model FQ Total Peaking Factor FAH Nuclear Enthalpy Rise Factor HFP Hot Full Power LBLOCA Large Break Loss of Coolant Accident LANL Los Alamos National Laboratory LEFM Leading Edge Flow Meter LOCA Loss of Coolant Accident MSIV Main Steam Isolation Valve MTC Moderator Temperature Coefficient 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 Recirculation Actuation Signal RCP Reactor Coolant Pump RCS Reactor Coolant System RLBLOCA Realistic Large Break LOCA RV Reactor Vessel RHR Residual Heat Removal RWST Refueling Water Storage Tank AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page viii Nomenclature (Continued)

SER Safety Evaluation Reportý SI Safety Injection SIAS Safety Injection Actuation Signal TVA Tennessee Valley Authority UHI Upper Head Injection AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 1-1 1.0 Introduction This report describes and provides results from a RLBLOCA analysis for the Sequoyah Unit 1 Station. The, plant is a Westinghouse 4-loop design with a rated thermal power of 3455 MWt and ice condenser containment. The loops contain four RCPs, four U-tube steam generators and a pressurizer. The ECCS is provided by two independent injection trains and four accumulators.

The analysis supports operation for Cycle 17 and beyond with AREVA NP's Mark-BW fuel design using either BLEU or standard U0 2 fuel and M5 cladding, unless changes in the Technical Specifications, Core Operating Limits Report, core design, fuel design, plant hardware, or plant operation invalidate the results presented herein. The analysis was performed in compliance with the NRC-approved RLBLOCA EM (Reference 1) with exceptions noted below. Analysis results confirm the 10CFR50.46(b) acceptance criteria presented in Section 3.0 are met and serve as the basis for operation of the Sequoyah Unit 1 Station with AREVA NP fuel.

The non-parametric statistical methods inherent in the AREVA NP RLBLOCA methodology provide' for the consideration of a full spectrum of break sizes, break configuration (guillotine or split break), axial shapes, and plant operational parameters. A conservative single-failure assumption is applied in which the loss of one train of the pumped ECCS injection is simulated.

Regardless of the 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 approved RLBLOCA EM (Reference 1) that were requested by the NRC.

The assumed reactor core power for the Sequoyah realistic large break loss-of-coolant accident is 3479 MWt. This value represents the plant rated thermal power of 3455 MWt with a maximum power measurement uncertainty of 0.7 percent (24 MWt) added to the rated thermal power. The power measurement uncertainty assumption discussed in 10CFR50, Appendix K was previously reduced for Sequoyah from 2.0. percent of the plant rated thermal power to 0.7 percent based on the installation of a LEFM system to. measure main feedwater flow. The improved feedwater flow measurement accuracy provided by the LEFM allowed for a power AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 1-2 measurement uncertainty recovery of 1.3 percent. The basis for the current 0.7 percent measurement uncertainty assumption is documented in Topical Report No. WCAP-15669, Revision 0. The power was not sampled in the analysis. This is not expected to have an adverse effect on the PCT results.

The RLBLOCA analysis was performed with a version of S-RELAP5 that requires both the void fraction to be less than 0.95 and the clad temperature to be less than 900 OF before the rod is allowed to quench. This may result in a slight increase in PCT results when compared to an analysis not subject to these constraints.

The RLBLOCA analysis was performed with a version of S-RELAP5 that limits the contribution of the Forslund-Rohsenow model to no more than 15 percent of the total heat transfer at and above a void fraction of 0.9. This may result in a slight increase in PCT results when compared to previous analyses for similar plants.

The split versus double-ended break type is no longer related to break area. In concurrence with Regulatory Guide 1.157, both the split and the double-ended break will range in area between the minimum break area (Amin) and an area of twice the size of the broken pipe. The determination of break configuration, .split versus double-ended, will be made after the break area is selected based on a uniform probability for each occurrence. Arin was calculated to be 33 percent of the DEGB area (see Section 6.6 for further discussion). This is not expected to have an effect on PCT results.

In concurrence with the NRC's interpretation of GDC 35, a set of 59 cases was run with a LOOP assumption and a second set with a No-LOOP assumption. The set of 59 cases that predicted the highest PCT is reported in Section 2 and Section 3, herein. The results from both case sets are shown in Figure 3-23. The effect on PCT results is expected to be minor.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 2-1 2.0 Summary The limiting PCT analysis is based on the parameter specification given in Table 2-1 for the limiting case. The limiting PCT is 1809 OF for an U0 2 rod in a case with LOOP conditions.

Gadolinia-bearing rods of 2, 4, 6 and 8 w/o Gd 20 3 were also analyzed, but were not 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 17x17 thermal hydraulically compatible fuel designs; hence, there is no mixed core consideration.

The analysis assumes full core power operation at 3479 MWt (including uncertainties), a steam generator tube plugging level of 15 percent in all steam generators, a total peaking factor (FQ) up to a value of 2.65 (including uncertainties, but no axial dependency), and a nuclear enthalpy rise factor (FAH) up to a value of 1.706 (including uncertainty). This analysis also addresses typical operational ranges or technical specification limits (whichever is applicable) with regard to pressurizer pressure and level; accumulator pressure, temperature (based on containment temperature), and level; core average temperature; core flow; containment pressure and temperature; and RWST.

The AREVA RLBLOCA methodology explicitly analyzes only fresh fuel assemblies (see Reference 1, Appendix B). Previous analyses have shown that once- and twice-burnt fuel will not be limiting up to peak rod average exposures of .62,000 MWd/MTU. The 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) 4200 Core Power (MWt) 3479 Total Peaking F(F) 2.531 Radial Peak (FAH) 1.706 Axial Offset 0.2883.

Break Type Split Break Size (ft2/side) 3.078 Offsite Power Availability Not available Decay Heat Multiplier 0.96386 AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 ReVision 0 Realistic Large Break LOCA Analysis Page 3-1 3.0 Analysis The purpose of the analysis is to verify typical technical specification peaking factor limits and the adequacy of the ECCS by demonstrating that the following 10CFR 50.46(b) criteria are met:

The calculated maximum fuel element cladding temperature shall not exceed 2200 'F.

The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel excluding the cladding surrounding the plenum volume were to react.

The calculated changes in core geometry shall be such that the core remains amenable to cooling.

Long-term cooling is established and maintained after the LOCA.

The analysis did not evaluate core coolability due to seismic events, nor did it consider the 10CFR 50.46(b) long-term cooling criterion. The RLBLOCA analysis conservatively considers blockage effects due to clad swelling and rupture in the prediction of the hot fuel rod PCT.

Since the analysis purpose is solely to change the LBLOCA licensing basis (from deterministic to realistic) of Unit 1, prior coolable geometry (LOCA-seismic loads) and long-term cooling licensing bases remain unaffected and valid. Therefore, compliance with Criteria 4 and 5 is assured.

Section 3.1 of this report describes the postulated LBLOCA event. Section 3.2 describes the models used in the analysis. Section 3.3 describes the 4-loop PWR plant and summarizes the system parameters used in the analysis. Compliance to the SER is addressed in Section 3.4.

Section 3.5 summarizes the results of the RLBLOCA analysis.

3.1 Description of the LBLOCA Event A LBLOCA is initiated by a postulated rupture of the RCS primary piping. Based on deterministic studies, the worst break location is in the cold leg piping between the reactor coolant pump and the reactor vessel for the RCS loop containing the pressurizer. The break initiates a rapid depressurization of the RCS. A reactor trip signal is initiated when the low AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-2 pressurizer pressure trip setpoint is reached; however, reactor trip is conservatively neglected in the analysis. The reactor is shut down by coolant voiding in the core.

The plant is assumed to be operating normally at full power prior to the accident. The cold leg break is assumed to open instantaneously. For this break, a rapid depressurization occurs, along with a core flow stagnation and reversal. This causes the fuel rods to experience DNB.

Subsequently, the limiting fuel rods are cooled by film boiling and convection to steam. The coolant voiding creates a strong negative reactivity effect and core fission ends. As heat transfer from the fuel rods is reduced, the cladding temperature rises.

Coolant in all regions of the RCS begins to flash. At the break plane, the loss of subcooling in the coolant results in substantially reduced break flow. This reduces the depressurization rate, and leads to a period of positive core flow or reduced downflow as the reactor coolant pumps in the intact loops continue to supply water to the vessel (in No-LOOP conditions). Cladding temperatures may be reduced and some portions of the core may rewet during this period. The positive core flow or reduced downflow period ends as two-phase conditions occur in the reactor coolant pumps, reducing their effectiveness. Once again, the core flow reverses as most of the vessel fluid mass flows out through the broken cold leg.

Mitigation of the LBLOCA begins when the SIAS is tripped. This signal is initiated by either high containment pressure or low pressurizer pressure. Regulations require that a worst single-failure be considered. This single-failure has been determined to be the loss of one ECCS pumped injection train. The AREVA RLBLOCA methodology conservatively assumes an on-time start and normal lineups of the containment spray to conservatively reduce Containment pressure and increase break flow. Hence, the analysis assumes that one charging pump, one Sl pump, one RHR pump and two containment spray pumps are operating.

When the RCS pressure falls below the accumulator pressure, fluid from the accumulators is injected into the cold legs. In the early delivery of accumulator water, high pressure and high break flow will drive some of this fluid to bypass the core. During this bypass period, core heat transfer remains poor and fuel rod cladding temperatures increase. As RCS and containment pressures equilibrate, ECCS water begins to fill the lower plenum and eventually the lower portions of the core; thus, core heat transfer improves and cladding temperatures decrease.

Eventually, the relatively large volume of accumulator water is exhausted and core recovery must rely on pumped ECCS coolant delivery alone. As the accumulators empty, the nitrogen AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-3 gas used to pressurize the accumulators exits through the break. This gas release may result in a short period of improved core heat transfer as the nitrogen gas displaces water in the downcomer. After the nitrogen gas has been expelled, the ECCS temporarily may not be able to sustain full core cooling because of the core decay heat and the higher steam temperatures created by quenching in the lower portions of the Core. Peak fuel rod cladding temperatures may increase for a short period until more energy is removed from the core by the charging, SI and RHR while the decay heat continues to fall. Steam generated from fuel rod rewet will entrain liquid and pass through the core, vessel upper plenum, the hot legs, the steam generator, and the reactor coolant pump before it is vented out the break. Some steam may flow to the upper head and pass through the spray nozzles, which provide a vent path to the break. The resistance of this flow path to the steam flow is balanced by the driving force of water filling the downcomer. This resistance may act to retard the progression of the core reflood and postpone core wide cooling. Eventually (within a few minutes of the accident), the core reflood will progress sufficiently to ensure core wide cooling. Full core quench occurs within a few minutes after core wide cooling. Long-term cooling is then sustained with the RHR 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 methodology (Reference 2). This method outlines an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifies the uncertainties in a LOCA analysis.

The RLBLOCA methodology consists of the following computer codes:

RODEX3A for computation of the initial fuel stored energy, fission gas release, and fuel-cladding gap conductance.

S-RELAP5 for the system calculation (includes ICECON for containment response).

AUTORLBLOCA for generation of ranged parameter values, transient input, transient runs, and general output documentation.

The governing two-fluid (plus non-condensibles) model with conservation equations for mass, energy, and momentum transfer is used. The reactor core is modeled in S-RELAP5 with heat AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-4 generation rates determined from reactor kinetics equations (point kinetics) with reactivity feedback, and with actinide and decay heating.

The two-fluid formulation uses a separate set of 'conservation equations and constitutive relations for each phase. The effects of one phase on the other are accounted for by interfacial friction, and heat and mass transfer interaction terms in the equations. The conservation equations have the same form for each phase; only the constitutive relations and physical properties differ.

The modeling of plant components is performed by following guidelines developed to ensure accurate accounting for physical dimensions and that the dominant phenomena expected during

.the LBLOCA event are captured. The basic building blocks for modeling are hydraulic volumes for fluid paths and heat structures for heat transfer. In addition, special purpose components exist to represent specific components such as the RCPs or the steam generator separators.

All geometries are modeled at the resolution necessary to best resolve the flow field and the phenomena being modeled within practical computational limitations.

System nodalization details are shown in Figures 3-1 through 3-5. A point of clarification: in Figure 3-1, break modeling uses two junctions regardless of break type-split or guillotine; for guillotine breaks, Junction 151 is deleted, it is retained fully open for split breaks. Hence, total break area is the sum of the areas of both break junctions.

A typical calculation using S-RELAP5 begins with the establishment of a steady-state initial condition with all loops intact. The input parameters and initial conditions for this steady-state calculation are chosen to reflect plant technical specifications or to match measured data.

Additionally, the RODEX3A code provides initial conditions for the S-RELAP5 fuel models.

Specific parameters are discussed in Section 3.3.

Following the establishment of an acceptable steady-state condition, the transient calculation is initiated by introducing a break into one of the loops (specifically, the loop with the pressurizer).

The evolution of the transient through blowdown, refill and reflood is computed continuously using S-RELAP5. Containment pressure is also calculated by S-RELAP5 using containment models derived from ICECON (Reference 4), which is based on the CONTEMPT-LT code (Reference 3) and has been updated for modeling ice condenser containments.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-5 The methods used in the application of S-RELAP5 to the LBLOCA are described in Reference 1. A detailed assessment of this computer code was made through comparisons to experimental data, many benchmarks with cladding temperatures ranging from 1,700 OF (or less) to above 2,200 OF. These assessments were used to develop quantitative estimates of the ability of the code to predict key physical phenomena in a PWR LBLOCA. Various models-for example, the core heat transfer, the decay heat model and the fuel cladding oxidation correlation-are defined based on code-to-data -comparisons and are, hence, plant independent.

The RV internals are modeled in detail (Figures 3-3 through 3-5) based on specific inputs supplied by TVA. Nodes and connectivity, flow areas, resistances and heat structures are all accurately modeled. The location of the hot assembly/hot pin(s) is unrestricted; however, the channel is always modeled to restrict appreciable upper plenum liquid fallback.

The final step of the best-estimate methodology is to combine all the uncertainties related to the code and plant parameters, and estimate the PCT at a high probability level. The steps taken to derive the PCT uncertainty estimate are summarized below:

1. Base Plant Input File Development First, base RODEX3A and S-RELAP5 input files for the plant (including the containment input file) are developed. Code input development guidelines are applied to ensure that model nodalization is consistent with the model nodalization used in the code validation.

2. Sampled Case Development The non-parametric statistical approach requires that many "sampled" cases be created and processed. For every set of input created, each "key LOCA parameter" is randomly sampled over a range established through code uncertainty assessment or expected operating limits (provided by plant technical specifications or data). Those parameters considered "key LOCA parameters" are listed in Table 3-1. This list includes both parameters related to LOCA phenomena (based on the PIRT provided in Reference 1) and to plant operating parameters.

3. Determination of Adequacy of ECCS The RLBLOCA methodology uses a non-parametric statistical approach to determine values of PCT at the 95 percent probability level. Total oxidation and total hydrogen are based on the limiting PCT case. The adequacy of the ECCS is demonstrated when these results satisfy the criteria set forth in Section 3.0.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-6 3.3 PlantDescriptionand Summary of Analysis Parameters The plant analysis presented in this report is for a Westinghouse-designed PWR, which has four loops, each with a hot leg, a U-tube steam generator, and a cold leg with a RCP 1. The RCS also includes one pressurizer connected to a hot leg. The core contains (193) 17x17 thermal-hydraulic compatible AREVA Mark-BW fuel assemblies. The ECCS includes one charging and one accumulator/SI/RHR injection path per RCS loop. The SI and RHR feed into common headers which are connected to the accumulator lines. The charging pumps are also cross-connected. The break is modeled in the same loop as the pressurizer, as directed by the RLBLOCA methodology. The RLBLOCA transients are of sufficiently short duration that the switchover to'sump cooling water (i.e., RAS) for ECCS pumped injection need not be considered The S-RELAP5 model explicitly describes the RCS, RV, pressurizer, and accumulator lines.

The charging injection flows are connected to the RCS, and the Sl and RHR injection flows are connected to the accumulator lines, consistent with the plant layout. This model also describes the secondary-side steam generator that is instantaneously isolated (closed MSIV and feedwater trip) at the time of the break. A symmetric steam *generator tube plugging level of 15 percent per steam generator was assumed.

Plant input modeling parameters were provided by TVA specifically for the Sequoyah Unit 1 Station. By procedure, TVA maintains plant documentation current, and directly communicates with AREVA on plant design and operational issues regarding reload cores. TVA and AREVA will continue to interact in that fashion. regarding the use of AREVA fuel in the Sequoyah Unit 1 Station. Both entities have ongoing processes that assure the ranges and values of input parameters for the Sequoyah Unit 1 Station RLBLOCA analysis bound those of the as-operated plant.

As described in the AREVA RLBLOCA methodology, many parameters associated with LBLOCA phenomenological uncertainties and plant operation ranges are sampled. A summary of those parameters is given in. Table 3-1. The LBLOCA phenomenological uncertainties are provided in Reference 1. Values for process or operational parameters, including ranges of sampled process parameters, and fuel design parameters used in the analysis are given in Table 3-2. Plant data are analyzed to develop uncertainties for the process parameters AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-7 sampled in the analysis. Table 3-3 presents a summary of the uncertainties used in the analysis. Two parameters, RWST temperature for ECCS flows and diesel start time, are set at conservative bounding values for all calculations. 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 as nominal or conservatively biased. The PIRT outcome yielded two important (relative to PCT) containment parameters-containment pressure and temperature. In many instances, the conservative guidance of CSB 6-1 (Reference 5) was used in setting the remainder of the containment model input parameters. As noted in Table 3-3, containment temperature is a sampled parameter. Containment pressure response is indirectly ranged by sampling the upper containment volume (Table 3-3). The minimum value is carried over from use in the long-term containment integrity analysis of record for Sequoyah. The maximum value is a simplified value computed as the volume available within the upper dome of the containment and within the crane wall above the control rod drive missile shield with no accounting for internal structures and the volumes of the refueling canal and the annular region separating the ice compartments neglected. This volume is maximized by neglecting the volume of internal structures. The lower compartment volume is biased low in order to promote flow through the ice baskets. In accordance with Reference 1, the condensing heat transfer coefficient is intended to be closer to a best-estimate instead of a bounding high'value. A [ ] Uchida heat transfer coefficient multiplier was specifically validated for use in Sequoyah through application of the process used in the RLBLOCA EM (Reference 1) sample problems. The ice condenser containment noding is shown in Figure 3-6. In the ice compartment, the water formed by melted ice and condensed steam flows to the lower ice compartment sump where it accumulates, if the ice bay drains are not large enough to accommodate the rate of water production. When the water level in the lower ice compartment sump rises above the bottom of the lower doors, water spillage through the lower doors occurs in addition to flow through the drain ports. The water drainage (spillage plus drainage) from the ice compartment falls through the lower compartment vapor. This condenses steam and reduces the containment pressure. The ice compartment drainage flow is treated as a 100 percent efficient spray during the post-blowdown period of the transient.

'The RCPs are Westinghouse 93A type pumps. The homologous pump performance curves for this type of pump were input to the S-RELAP5 plant model.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-8 The initial conditions and boundary conditions are given in Table 3-8. The building spray is modeled at maximum heat removal capacity. While there is an option within the computer code model to deliver spray to the lower compartment, this option is not applicable to Sequoyah Unit 1. All spray flow is delivered to the upper compartment. Because the start time for the recirculation fan is 600 seconds, forced flow from the upper compartment to the lower compartment is not likely to occur during the time period analyzed. The flow of steam or air, from the lower compartment to the upper compartment, backwards through the back draft dampers, is not modeled (no reverse direction flow). This approach is conservative in that no bypass of the ice beds (from lower to upper compartments) is allowed, and all flow from the lower compartment is directed through the ice beds. The passive flow of air and steam, from the upper compartment to the lower compartment, is modeled however. This is a passive flow, which is only a function of the excess pressure of the upper compartment compared to the lower compartment, the flow area of the recirculation fan back draft dampers, and the loss coefficient of the dampers. The back draft dampers are designed such that reverse flow from the lower to the upper compartment is prevented. However, when the upper compartment pressure is at least 0.5 psi greater than the lower compartment, the actual dampers open and allow flow from the upper compartment to the lower compartment. Flow in this manner, from the upper to lower compartment, is modeled without this minimum pressure difference, i.e. any excess pressure is modeled-as resulting in flow.

Passive heat sink parameters are listed in Table 3-9. Surface coatings, where they existed, were incorporated as an equivalent thickness of base material in order to eliminate any insulating effects on the exposed surfaces of the heat structures. Because the original basis for the size of each heat sink was biased low (for a different application), the values listed in Table 3-9 reflect a 10 percent increase in heat transfer surface area as compensation. Passive heat sinks were. added to the lower containment to represent new sump screens being installed in the Sequoyah Unit 1 plant (17 ft3 of steel). Additionally, all heat structure exposed surfaces remain available for condensing steam, even when they may become covered by ice melt or condensate.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-9 3.4 SER Compliance A number of requirements on the methodology are stipulated in the conclusions section of the SER for the RLBLOCA methodology (Reference 1). These requirements have all been fulfilled during the application of the methodology as addressed in Table 3-4.

Six non-limiting PCT cases were potential candidates for blowdown quench (SER Item 7) and, were closely inspected. For this set of calculations, no evidence of blowdown quench was found and compliance to-the SER restriction has been demonstrated.

Several measures have been taken to prevent the top-down quench (SER Item 8). The upper plenum nodalization features include:

" the homogenous 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-flow is allowed between the first axial level Upper Plenum nodes above the hot channel to the average channel;

" the CCFL model is applied on all core exit junctions.

Seven non-limiting cases were closely examined for top-down quench. These cases exhibited a decrease in the integrated mass flux at the hot assembly exit late into the transient, after the PCT time. The heat structure temperature displays a drop to saturation temperature starting at the bottom and sequentially progresses upward through the PCT elevation, the nodes above the PCT node experiencing the quench at times later than at the PCT node. This conclusion is further supported by the evolution of flow into the upper plenum node situated directly above the hot assembly. No evidence of a secondary quench front moving from the top down was noted.

Case 35 (non-limiting) deserves added scrutiny. Hot rod nodes situated above the PCT location quenched earlier than the PCT node, exhibiting an apparent secondary top-down quench front.

Since the power is lower in the top core region, the temperatures in the top region can drop below the CHF temperature before the entire core is quenched. In the S-RELAP5 documentation, this is termed as the top quench front or secondary quench front. It does not mean there is a top-down quench front propagating downwards from the top of the core due to liquid flowing down from the upper plenum. For this particular case, cooling at the top of the AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-10 core occurs due to fluid entrainment from the bottom. The modeling precautions taken to prevent top-down quench are sufficient and no top-down quench has been observed.

Therefore, compliance to the SER restriction has been demonstrated.

3.5 Realistic Large Break LOCA Results Two case sets of 59 transient calculations were performed sampling the parameters listed in Table 3-1. For each case set, PCT was calculated for a U0 2 rod and for Gadolinia-bearing rods with concentrations of 2, 4, 6 and 8 w/o Gd 20 3. The limiting case set, that contained the PCT, was the set with no offsite power available. The limiting PCT (1809 OF) occurred in Case 1 for a U0 2 rod. The major parameters for the limiting transient are presented in Table 2-1. Table 3-5 lists the results of the limiting case. The fraction of total hydrogen generated was not directly calculated; however, it is conservatively bounded by the calculated total percent oxidation, which is well below the 1 percent limit. The best-estimate PCT case is Case 13, which corresponded to the median case out of the 59-case set with no offsite power available. The nominal PCT was 1426 OF. This result can be used to quantify the relative conservatism in the limiting case result. In this analysis, it was 383 OF.

The case results, event times and analysis plots for the limiting PCT case are shown in Table 3-5, Table 3-6, and in Figures 3-12 through 3-22. Figure 3-7 shows linear scatter plots of the key parameters sampled for the 59 calculations. Parameter labels appear to the left of each individual plot. These figures show the parameter ranges used in the analysis. Figures 3-8 and 3-9, show the time of PCT and break size versus PCT scatter plots for the 59 calculations, respectively. Figures 3-10 and 3-11 show the maximum oxidation and total oxidation versus PCT scatter plots for the 59 calculations, respectively. Key parameters for the limiting PCT case are shown in Figures 3-12 through 3-22. Figure 3-12 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 3-23.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-11 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)

Decay 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 Break size Pressurizer pressure Pressurizer liquid level Accumulator pressure Accumulator liquid level Accumulator temperature (based on lower compartment containment temperature)

Containment temperature Containment volume Initial RCS flow rate Initial operating RCS temperature Diesel start (for loss of offsite power only)

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 6.10.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larce Break LOCA Analysis Paqe 3-12 Table 3-2 Plant Operating Range Supported by the LOCA Analysis

__ _ Event 1 Operating Range 1.0 Plant Physical Description 1.1 Fuel a) Cladding outside diameter 0.374 in.

_ b) Cladding inside diameter 0.326 in.

Sc) Cladding thickness I 0.024 in.

d) Pellet outside diameter 0.3195 in.

I e) Pellet density 96 percent of theoretical f) Active fuel length 144 in.

g) Resinter densification h) Gd 2O3 concentrations 2, 4, 6, 8 w/o 1 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 17x17 e) SG tube plugging _ 15 percent 2.0 1 Plant Initial Operating Conditions

_ 2.1 Reactor Power-a) Nominal reactor power 3479 MWt 1 b) FQ _2.652 c) FAH < 1.7063 d) MTC 0 at HFP 2.2 Fluid Conditions a) Loop flow 131.6 Mlbm/hr*_ M *152.8 Mlbm/hr

{b) RCS average temperature 578.2 OF _< T

583 OF 4 c) Upper head temperature -Tcold Temperature 1

Includes uncertainties 2

Ensures that a minimum 7percent peaking margin is maintained to the FQ limits when operating at the positive or negative AFD limit 3

Includes 4 percent measurement uncertainty 4

Upper head temperature will change based on sampling of RCS temperature AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-13 Table 3-2 Plant Operating Range Supported by the LOCA Analysis (Continued)

I d) Pressurizer pressure 1859.7 psia _ P _ 2459.7 psia

_ e) Pressurizer level 57 percent _ L < 95percent f) Accumulator pressure, 614.7 psia < P _<697.7 psia g) Accumulator liquid volume V *1095.4 ft 3 1004.6 ft 3 _<

h) Accumulator temperature 95 OF < T _<130 OF (coupled to

_containment lower volume temperature) i) Accumulator fL/D As-built piping configuration j) Minimum ECCS boron > 2400 ppm 3.0 I Accident Boundary Conditions _

_a) Break location Any RCS piping location

_ b) Break type Double-ended guillotine or split c) Break size (each side, relative to cold 0.33 _ A _ 1.0 full pipe area (split) leg pipe area) 0.33 _ A < 1.0 full pipe area (guillotine) d) Worst single-failure Loss of one train of ECCS e) Offsite power On or Off f) Charging pump flow Bounding minimum of current pump delivery g) SI pump flow Bounding minimum of current pump delivery h) RHR pump flow Bounding minimum of current pump delivery h) ECCS pumped injection temperature 110 OF i) Charging pump delay 37 s (w/ offsite power) 27 s (w/o offsite power) j) SI pump delay 37 s (w/ offsite power) 27 s (w/o offsite power) k) RHR pump delay 37 s (w/ offsite power) 27 s (w/o offsite power)

I) Containment pressure 14.3 psia, nominal value m) Containment upper compartment 80 OF_ T < 110 'F temperature n)Containment lower compartment 95 OF_ T _ 130 OF temperature

0) Containment sprays delay 8 s p) Containment spray water 55 OF temperature AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larqe Break LOCA Analysis Page 3-14 Table 3-3 Statistical Distributions Used for Process Parameters 1 Operational Measurement Standard Parameter Uncertainty Parameter Range Uncertainty2 Deviation Distribution Distribution Pressurizer Pressure (psia) Uniform 1859.7 - 2459.7 N/A N/A Pressurizer Liquid Level (percent) Uniform 57-95 N/A N/A Accumulator Liquid Volume (Wt 3

) Uniform j 1004.6 - 1095.4 N/A N/A Accumulator Pressure (psia) Uniform 614.7 - 697.7 N/A N/A Containment Lower Compartment Uniform 95-130 N/A N/A

/Accumulator Temperature (°F)

Containment Upper Compartment Uniform 80-110 Temperature (CF)

Containment Upper Volume ( It3) Uniform 651,000 - 692,600 N/A N/A Initial RCS Flow Rate (Mlbm/hr) Uniform 131.6 - 152.8 N/A N/A Initial RCS Operating Temperature Uniform 578.2- 583 N/A N/A (Tavg) (°F)

RWST Temperature for ECCS ('F) Point 110 N/A N/A RWST Temperature for Point 55 N/A N/A Containment Sprays (°F)

Offsite Power Availability 3 Binary 0,1 N/A N/A Delay for Containment Cooling (s) Point 8.0I N/A N/A Charging Pump Delay (s) Point 37 (w/ offsite power) N/A N/A 27 (w/o offsite power) N LHSI Pump Delay (s) Point 37 (w/ offsite power) N/A N/A

.1. _________ 27 (w/o offsite power) ________ _______

RHR Pump Delay (s) Point *_______ 37 (w/ offsite power) N/A N/A

_ (w/o

_27offsite power) N/A_ N/A 1

Note that core power is not sampled, see Section 1.0 2

All measurement uncertainties were incorporated into the operational ranges 3

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.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larae Break LOCA Analysis Paqe 3-15 Table 3-4 SER Conditions and Limitations SER Conditions and Limitations I Response

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

2. AREVA NP has agreed that it is not to use nodalization Hot leg nozzle gaps were not modeled.

with hot leg to downcomer nozzle gaps.

3. If AREVA NP applies the RLBLOCA methodology to plants The PLHGR for Sequoyah Unit 1 is lower than that used in using a higher planar linear heat generation rate (PLHGR) the development of the RLBLOCA EM (Reference 1). An than used in the current analysis, or if the methodology is end-of-life calculation was not performed; thus, the need for to be applied to an end-of-life analysis for which the pin a blowdown cladding rupture model was not reevaluated.

pressure is significantly higher, then the need for a blowdown clad rupture model will be reevaluated. The evaluation may be based on relevant engineering experience and should be documented in either the RLBLOCA guideline or plant specific calculation file.

4. Slot breaks on the top of the pipe have not been evaluated. The evaluation of slot breaks is documented in the AREVA These breaks could cause the loop seals to refill during late RLBLOCA analysis guidelines.

reflood and the core to uncover again. These break locations are an oxidation concern as opposed to a PCT concern since the top of the core can remain uncovered for extended periods of time. Should an analysis be performed for a plant with loop seals with bottom elevations that are below the top elevation of the core, AREVA NP will evaluate the effect of the deep loop seal on the slot breaks.

The evaluation may be based on relevant engineering experience and should be documented in either the RLBLOCA guideline or plant-specific calculation file.

5. The model applies to 3 and 4 loop Westinghouse- and Sequoyah Unit 1 is a Westinghouse 4 -loop plant.

CE-designed nuclear steam systems.

6. The model applies to bottom reflood plants only (cold side Sequoyah Unit 1 is a bottom reflood plant.

injection into the cold legs at the reactor coolant discharge piping).

7. The model is valid as long as blowdown quench does not The limiting case did not show any evidence of a blowdown occur. If blowdown quench occurs, additional justification quench. The possibility of blowdown quench was observed for the blowdown heat transfer model and uncertainty are in seven non-limiting cases which were discussed in Section needed or the calculation is corrected. A blowdown 3.4.

quench is characterized by a temperature reduction of the peak cladding temperature (PCT) node to saturation temperature during the blowdown period.

8. The reflood model applies to bottom-up quench behavior. Core quench initiated at the bottom of the core and If a top-down quench occurs, the model is to be justified or proceeded upward.

corrected to remove top quench. A top-down quench is characterized by the quench front moving from the top to the bottom of the hot assembly.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-16 Table 3-4 SER Conditions and Limitations (Continued)

SER Conditions and Limitations I Response

9. The model does not determine whether Long-term cooling was not evaluated in this analysis.

Criterion 5 of 10 CFR 50.46, long term cooling, has been satisfied. This will be determined by each applicant or licensee as part of its application of this methodology.

10. Specific guidelines must be used to develop The nodalization in the plant model is consistent with the Westinghouse the plant-specific nodalization. Deviations 4-loop sample calculation that was submitted to the NRC for review.

from the referenceplant must be addressed. Figure 3-1 shows the loop noding used in this analysis. (Note only Loop 1 is shown in the figure; Loops 2, 3 and 4 are identical to loop 1, except that only Loop 1 contains the pressurizer and the break.) Figure 3-2 shows the steam generator model. Figures 3-3, 3-4, and 3-5 show the reactor vessel noding diagrams. Some minor differences that are included in the plant specific model include:

1) The RV upper internals are of the inverted top-hat type, therefore an additional node was added to the upper head volume in order to model the region situated below the top hat brim and above the upper support plate;

2) The plant was designed to use Upper Head Injection which utilized columns. However it was modified and the upper head safety injection was disconnected and capped. The flow path of the UHI Columns was modeled with an extra set of pipe components connecting the lower most volume of the upper head to the inlet into the corresponding radial region of the upper plenum;

3) The pumped piping branches into the accumulator discharge .piping slightly differently;

4) The hydraulic model of the core employs 22 axial nodes instead of 23;

5) There are no standpipes present in the Sequoyah Unit 1 RV upper plenum;

6) The plant has safety grade charging which is included in the model;

7) The lower support plate that separates the lower plenum from the lower head of the reactor vessel is curved;

8) Sequoyah Unit 1 is a cold upper head type plant.

9) The ICECON noding is representative for an ice condenser plant and represents a change from Reference 1.

10) Component 154 has only one cell instead of the two in Reference 1.

11. A table that contains the plant-specific Simulation of clad temperature response is a function of parameters. and the range of the values phenomenological correlations that have been derived either analytically considered for the selected parameter during or experimentally. The important correlations have been validated for the the topical report approval process must be RLBLOCA methodology and a statement of the range of applicability has provided. When plant-specific parameters been documented. The correlations of interest are the set of heat transfer are outside the range used in demonstrating correlations as described in Reference 1. Table 3-7 presents the acceptable code performance, the licensee or summary of the full range of applicability for the important heat transfer applicant will submit sensitivity studies to correlations, as well as the ranges calculated in the limiting case of this show the effects of that deviation, analysis. Calculated values for other parameters of interest are also provided. .As is evident, the plant-specific parameters fall within the methodology's range of applicability.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-17 Table 3-4 SER Conditions and Limitations (Continued)

SER Conditions and Limitations I Response

12. The licensee or applicant using the approved Analysis results are discussed in Section 3.5.

methodology must submit the results of the plant-specific analyses, including the calculated worst break size, PCT, and local and total oxidation.

13. The licensee or applicant wishing to apply The Sequoyah Unit 1 plant has previously been operating with M5 clad AREVA NP realistic large break loss-of- fuel and thus this restriction has been satisfied.

coolant accident (RLBLOCA) methodology to M5 clad fuel must request an exemption for its use until the planned rulemaking to modify 10 CFR 50.46(a)(i) to include M5 cladding material has been completed.

Table 3-5 Summary of Results for the Limiting PCT Case Case##

PCT Temperature 1809 OF Time 58.4 s Elevation 8.406 ft Metal-Water Reaction _

Percent Oxidation Maximum 1.8600 Percent Total Oxidation 0.0432 AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-18 Table 3-6 Calculated Event Times for the Limiting PCT Case Event L Time (s)

Break Opened 0.0 RCP Trip 0.0 SIAS Issued 0.1 Start of Broken Loop Accumulator Injection 12.7 Start of Intact Loop Accumulator Injection (Loops 2, 3 14.4, 14.5, 14.5 and 4 respectively)

Start of Charging 27.1 SI/RHR Available 27.1 Broken Loop SI Delivery Began 27.1 Intact Loop SI Delivery Began (Loops 2, 3 and 4 27.1, 27.1, 27.1 respectively)

Broken Loop RHR DeliVery Began 27.1 Intact Loop RHR Delivery Began (Loops 2, 3 and 4 respectively) 27.1,27.1,27.1 Beginning of Core Recovery (Beginning of Reflood) 46.2 PCT Occurred 58.4 Broken Loop Accumulator Emptied 83.3 Intact Loop Accumulators Emptied 83.8, 84.1, 83.4 (Loops 2, 3 and 4 respectively) 83.8,_84.1,83.4 Transient Calculation Terminated 580.3 AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-19 Table 3-7 Heat Transfer Parameters for the Limiting Case AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larae Break LOCA Analvsis Paae 3-20 Table 3-8 Containment Initial and Boundary Conditions Containment Net Free Volume Volume (ft3)

Upper Compartment 651,000 - 692,600 Lower Compartment (minimum) 248,500 Ice Condenser 181,400 Dead Ended Compartments 129,900 Initial Mass of Ice 2.448 x 108 Ibm Initial Conditions Containment Pressure (nominal) 14.3 psia Upper Containment Temperature 80 OF - 110 OF Lower Containment Temperature 95 OF - 130 OF Humidity 100 percent Containment Spray Maximum Total Flow 2 x 7700 = 15,400 gpm Minimum Spray Temperature 55 OF Fastest Post-LOCA initiation of 10 s (ramped to full flow spray between 8 and 10 s)

Containment Air Return Fan11 Post-LOCA initiation at 600 s Total Flow = 120,000 cfm 11 Due to the relatively late start of the recirculation fan, it is not modeled in this analysis.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-21 Table 3-9 Passive Heat Sinks in Containment Inside Area Thickness Thickness Height Heat ft2 Sink ft Radius ft ft Material Left Side Right Side fH Reactor Cavity Walls 6438 2.02 concrete Lower Comp. insulated Concrete Floor 4444 2.00 concrete Lower Comp. insulated Interior Concrete 8464 1.00 concrete Lower Comp. insulated Reactor Vessel Biological 11 6,0 19.88 concrete Lower Comp. Lower Comp.

Shield Wall Steel Lined Refueling Canal in 13. 0.02083 21.48 stainless steel Lower Comp.

LC 4.0 21.48 concrete Lower Comp.

Crane Wall between LC & DE 41.5 3.0 33.72 concrete Lower Comp. Dead End Crane Wall in LC 41.5 3.0 29.37 concrete Lower Comp. insulated Crane Wall in UC 41.5 3.0 32.44 concrete Upper Comp. insulated Refueling Canal in Contact with 2551 0.02083 stainless steel Upper Comp.

Upper and Lower Compartment 3.87 concrete Lower Comp.

Refueling Canal in Contact with 1,260 0.02083 stainless steel Upper Comp.

Annular Region 3.0 concrete annulus Concrete Structure between 13081 concrete Upper Comp. Lower Comp.

Upper and Lower Compartment Interior Concrete 2278 3.0 concrete Upper Comp. insulated Containment Shell 24,646 0.05417 carbon steel Upper Comp. annulus LC Steel Heat Sink 24,999 0.03674 carbon steel Lower Comp. insulated UC Steel Heat Sink 11669 0.4229 carbon steel Upper Comp. insulated Dead-End Steel Heat Sink 8610 0.074375 carbon steel DE Comp. insulated Material Properties Thermal Conductivity Volumetric Heat Capacity (BTU/hr-ft-OF) (BTU/ft3-OF)

Concrete 0.84 30.24 Carbon Steel 27.3 59.2 Stainless Steel 9.87 59.22 AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-22 Figure 3-1 Primary System Noding AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-23 Figure 3-2 Secondary System Noding AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-24 Figure 3-3 Reactor Vessel Noding AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larme Break LOCA Analysis Page 3-25 Figure 3-4 Core Noding Detail AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-26 Figure 3-5 Upper Plenum Noding Detail AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larme Break LOCA Analysis Page 3-27 Figure 3-6 Containment Noding Diagram AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larcme Break LOCA Analysis Page 3-28 One-Sided Break Area S m om m Nm m imem 2

(ft /side)

.0 2.0 3.0 4.0 5 .0 Burn Time .. 0 0 0.* 00.0.0 ... 0 O. 0 @e@O -

(hours) 0 .0 5000.0 10000.0 150 00.0 Core Power (MW) 34778.0 3478.5 3479.0 3479.5 3480.0 3480.5 34811.0 Fq Peaking

.5 1.7 19 21 23 25 27 29 AO 0.4 -0.2 0.0 0.2 0 .4 L i I Pressurizer Pressure (psia) 18100.0 2000.0 2200.0 2400.0 26(00.0 Pressurizer Liquid Level wm. me 040 4W 0m-E

(%)

5'0.0 60.0 70.0 80.0 90.0 10)0.0 RCS (Tavg)

Temperature (CF)

, 0 5 0I I 5 57'8.0 579.0 580.0 581.0 5820o 5S3.0 Figure 3-7 Scatter Plot of Operational Parameters AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-29 Total Loop Flow *O *m m O. inc0 (Mlb/hr) 130.0 140.0 150.0 160.0 Accumulator ' m c' Liquid Volume s o W me* o ft (ft3) 1000.0 1020.0 1040.0 1060.0 1080.0 1100.0 Accumulator Pressure " I em IN (psia) 600.0 620.0 640.0 660.0. 680.0 700.0 Upper Compartment { ' ' " , I -

Containment Volume 040 o* o mn m ej (ft3) mmc 6.50e+05 6.60e+05 6.70e+05 6.80e+05 6.90e+05 7.00e+05 Upper Compartment .

Containment TemperatureL-00 o

('F)

mo* m* m om s a H

80.0 90.0 100.0 110.0 Lower Compartment _

(Accumulator) L oo m mumel mo me

Containment Temperature-(°F) 90.0 100.0 110.0 120.0 130.0 Figure 3-7 Scatter Plot of Operational Parameters (Continued)

AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-30 PCT vs Time of PCT 2000 1800 - U 1600 1400 J-- 1200[

a3 1000 800 0 Split Break

[] Guillotine Break 600 I 400 I I I I I 0 100 200 300 400 500 Time of PCT (s)

Figure 3-8 PCT versus PCT Time Scatter Plot from 59 Calculations AREVA NP Inc.

ANP-2695(NP).

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-31 PCT vs One-sided Break Area 2000 1800 [ 0 U

1600 F U 0ED U

U 1400 ONLuIN ET EYEU -

N jý 1200 0 El 1000 0

F ]

800 l-600 I 0 Split Break 0 Guillotine Break 400 L 1.0 2.0 3.0 4.0 5.0 Break Area (ft2/side)

Figure 3-9 PCT versus Break Size Scatter Plot from 59 Calculations AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-32 Maximum Oxidation vs PCT 3.0 2.8 E Split Break LI Guillotine Break 2.6 -

2.4 2.2 LI 2.0 MEN 1.8 c 1.6 X

1.4 LI 1.2 1.0 0.8 I*

0.6 0, E.,

0.4 0ýl IMr 0.2

0.0 400 600 800 1000 1200 1400 1600 1800 2000 PCT (°F)

Figure 3-10 Maximum Oxidation versus PCT Scatter Plot from 59 Calculations AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-33 Total Oxidation vs PCT 0.10 0 Split Break , I LI Guillotine Breaki 0.08 F 0.06 F

._R C

0

-o_

0X 0.04 F M mUm n

"D'm,*[] ,00 0 L.

0.02 0.00 n I I I *]atll I P rrll f [ ! ln* ii i n I f I i 40 0 800 1200 1600 2000 POT (0F)

Figure 3-11 Total Oxidation versus PCT Scatter Plot from 59 Calculations AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larqe Break LOCA Analysis Paqe 3-34 PCT Trace for Case #1 PCT = 1808.5 OF, at Time = 58.39 s, on Hot U02 Rod 2000 1500 E 1000 t-C 0

500 0

0 200 400 600 Time (s)

Figure 3-12 Peak Cladding Temperature (Independent of Elevation) for the Limiting Case AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-35 Break Flow 80 Vessel Side

Pump Side Total 60 40 E

-o 0

U- 20 -I 0

-20 0 200 400 600 Time (s)

Figure 3-13 Break Flow for the Limiting Case AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-36 Core Inlet Mass Flux 1000 500 U) 0

-500 0 200 400 600 Time (s)

Figure 3-14 Core Inlet Mass Flux for the Limiting Case AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larcie Break LOCA Analysis Paqe 3-37 Core Outlet Mass Flux 1000 500.

El

.0 0

-500

-1000 0 200 400 600 Time (s)

Figure 3-15 Core Outlet Mass Flux for the Limiting Case AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-38 Pump Void Fraction 1.0 0.8 0.6 0

U--

0 0.4

____ Broken Loop 1

Intact Loop 2

--- - Intact Loop 3 0.2 Intact Loop 4 0.0 L 0 200 400 600 Time (s)

Figure 3-16 Void Fraction at RCS Pumps for the Limiting Case AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-39 ECCS Flows 1500 1000

-o L0 0

LL 500 0

0 200 400 600 Time (s)

Figure 3-17 ECCS Flows (Includes Accumulator, Charging, Sl and RHR) for the Limiting Case AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larae Break LOCA Analysis Paqe.3-40 Upper Plenum Pressure 2000 1500

~?1000 a.

500 0

0 200 400 600 Time (s)

Figure 3-18 Upper Plenum Pressure for the Limiting Case AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-41 Downcomer Liquid Level 30 20 a,J 2-

.J 10 0

0 200 400 600 Time (s)

Figure 3-19 Collapsed Liquid Level in the Downcomer for the Limiting Case AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larae Break LOCA Analysis Pacle 3-42 Lower Vessel Liquid Level 14 12 10 8

-J 6

4 2

0 0 200 400 600 Time (s)

Figure 3-20 Collapsed Liquid Level in the Lower Plenum for the Limiting Case.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-43 Core Liquid Level 15 10

-J a-5 0 j 0 200 400 600 Time (s)

Figure 3-21 Collapsed Liquid Level in the Core for the Limiting Case AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larqe Break LOCA Analysis Paqe 3-44 Containment and Loop Pressures 100 90 80 70 60 0~

U)

2 50 (I)

U) 0)

0~

40 30 20 10 0

0 200 400 600 Time (s)

Figure 3-22 Containment and Loop Pressures for the Limiting Case AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 3-45 2000 T - -- - 2000

[LOOP 0 No LOOP 1800 -I,--- 1800 0

0 0 i 4 --

-c-----------

1600 - - -aJ 1600

- -c - - - - - ~-

ci

.9 ci 0 0ý 1400 + ---------------------------------- -- - - - - 1400 0

0 ci n r-. D 0 ---- - - -- - - -- - - - - --

1200 - 1200 00 11 1000 + 1000 ci 800 800 0 10 20 30 40 50 60 Case Number Figure 3-23 GDC 35 LOOP versus No-LOOP Cases AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 4-1 4.0 Conclusions The results of the RLBLOCA analysis show that the limiting LOOP case has a PCT of 1809 OF, and a maximum oxidation thickness and hydrogen generation that fall well within regulatory requirements.

The analysis supports operation at a nominal power level of 3479 MWt (including uncertainty), a steam generator tube plugging level of up to 15 percent in all steam generators, a total peaking factor (FQ) of 2.65 (including uncertainty) and a nuclear enthalpy rise factor (FAH) of 1.706 (including uncertainty) with no axial or burnup dependent power peaking limit.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 5-1 5.0 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 2, Standard Review Plan, June 1987.

6. NUREG/CR-1532, EPRI NP-1459, WCAP-96991 "PWR FLECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Task Data Report," June 1980.

7. Letter from Ronald W. Hernan, U.S. NRC, to J. A. Scalice, Tennessee Valley Authority, "Sequoyah Nuclear Plant, Units 1 and 2 Issuance of Amendments RE: 1.3-Percent Power Uprate (TAC NOS. MB3435, AND MB3436) (TSC NO. 01-08)," April 30, 2002 (US NRC ADAMS Accession # ML021220060)

8. NUREG/CR-0994, "A Radiative Heat Transfer Model forthe TRAC Code" November 1979.

9. J.P. Holman, Heat Transfer, 4 th 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.

13. Letter from Pedro Salas, Tennessee Valley Authority to U.S. NRC, TVA-SQN-TS-01-08, Sequoyah Nuclear Plant (SQN), Units 1 & 2, Technical Specification (TS) Change No.

01-08, "Increase Maximum Allowed Reactor Power Level to 3455 Mega-Watt Thermal (MWt)," November 15, 2001 (US NRC ADAMS Accession # ML013470345)

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-1 6.0 Addendum - Additional Information Supporting EMF-2103 Revision 0 The following sections are responses to typical RAI questions posed by the NRC on EMF-2103 Revision 0 plant applications. In some instances, these requests cross-referenced documentation provided on dockets other than those for which the request is made. AREVA discussed these and similar questions from the NRC draft SER for Revision 1 of EMF-2103 in a meeting with the NRC on December 12, 2007. AREVA agreed to provide the following additional information within new submittals of a Realistic Large Break LOCA report.

6.1 Reactor Power Question: Reactor Power - Table 3-2, Item. 2.1, and its associatedFootnote I indicate that the

assumed reactor core power "includes uncertainties." The use of a reactorpower 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, "Requiredand Acceptable Features of The Evaluation Models, 'Sources of Heat During a LOCA." However, Appendix K.L.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.. ."

Response: As indicated in Item 2.1 of Table 3-2 herein, the assumed reactor core power for the Sequoyah realistic large break loss-of-coolant accident is 3479 MWt. This value represents the plant rated thermal power (i.e., total reactor core heat transfer rate to the. reactor coolant system) of 3455 MWt with a maximum power measurement uncertainty of 0.7 percent (24 MWt) added to the rated thermal power.

The power measurement uncertainty assumption discussed in 10CFR50, Appendix K was previously reduced for Sequoyah from 2.0 percent of the plant rated thermal power to 0.7 percent.based on the installation of a leading edge flow meter (LEFM) system to measure main feedwater flow. The improved feedwater flow measurement accuracy provided by the LEFM allowed for a power measurement uncertainty recovery of 1.3 percent. This power level assumption is a change to the approved RLBLOCA EM (Reference 1).

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-2 The basis for the current 0.7 percent measurement uncertainty assumption is documented in Topical Report No. WCAP-15669, Revision 0. This report was submitted to NRC in Reference

13. NRC review and acceptance of the current power measurement uncertainty has been documented in Reference 7.

6.2 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 employed for the Sequoyah Unit 1 LAR requires that both the void fraction is less than 0.95 and the clad temperature is less than the minimum temperature for film boiling heat transfer (Tmin) before the rod is allowed to quench. Tmin is a sampled parameter in the RLBLOCA methodology with a mean value of 626 K and a standard deviation of 33.6 K, making it very unlikely that Tmin would exceed 755 K (900 'F). For the Sequoyah Unit 1 cycle 17 case set Tmin was never sampled above 696 K (793.4 'F). This is a change to the approved RLBLOCA EM (Reference 1).

6.3 Rod-to-Rod Thermal Radiation Question: Providejustification that the S-RELAP5 rod-to-rod thermal radiationmodel applies to the SQN-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: film boiling, convection to steam, rod to liquid radiation and rod to vapor radiation. This heat transfer package was assessed against various experimental data sets involving both moderate (1600 OF - 2000 OF) and high (2000 OF to over 2200 OF) peak cladding temperatures and shown to be conservative when applied nominally. The normal distribution of the experimental data was then determined. During the execution of an RLBLOCA evaluation, the heat transferred from a fuel rod is determined by the application of a multiplier to the nominal heat transfer model. This multiplier is determined by a random sampling of the normal AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-3 distribution of the experimental data benchmarked. Because the data include the effects of rod-to-rod radiation, it is reasonable to conclude that the modeling implicitly includes an allocation for rod-to-rod radiation effects. As will be demonstrated, the approach is reasonable because the conditions within actual limiting fuel assemblies assure that the actual rod-to-rod radiation is larger than the allocation provided through normalization to the experiments.

The FLECHT-SEASET tests evaluated covered a range of PCTs from 1,651 to 2,239 OF and the THTF tests covered a range of PCTs from 1,000 to 2,200 OF. Since the test bundle in either FLECHT-SEASET or THTF is surrounded by a test vessel, which is relatively cool compared to the heater rods, substantial radiation from the periphery rods to the vessel wall can occur. The rods selected for assessing the RLBLOCA reflood heat transfer package were chosen from the interior of the test assemblies to minimize the impact of radiation heat transfer to the test vessel.

The result was that the assessment rods comprise a set which is primarily isolated from cold wall effects by being surrounded by powered rods at reasonably high temperatures.

As a final assessment, three benchmarks independent of THTF and FLECHT-SEASET were performed. These benchmarks were selected from the Cylindrical Core Test Facility (CCTF),

LOFT, and the Semiscale facilities. Because these facilities are more integral tests and together cover a wide range of scale, they also serve to show that scale effects are accommodated within the code calculations.

The results of these calculations are provided in Section 4.3.4, Evaluation of Code Biases, page 4-100, of Reference 1. The CCTF results are shown in Figures 4.180 through 4.192, the LOFT results in Figures 4.193 through 4.201, and the Semiscale results in Figures 4.202 through 4.207. As expected, these figures demonstrate that the comparison between the code calculations and data is improved with the application of the derived biases. The CCTF, LOFT, and Semiscale benchmarks further indicate that, whatever consideration of rod-to-rod radiation is implicit in the S-RELAP5 reflood heat transfer modeling, it does not significantly effect code predictions under conditions where radiation is minimized. The measured PCTs in these assessments ranged from approximately 1,000 to 1,540 OF. At these temperatures, there is little rod-to-rod radiation. Given the good agreement between the biased code calculations and the CCTF, LOFT, and Semiscale data, it can be concluded that there is no significant over prediction of the total heat transfer coefficient.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-4 Notwithstanding any conservatism evidenced by experimental benchmarks, the application of the model to commercial nuclear power plants provides some additional margins due to limitations within the experiments. The benchmarked experiments, FLECHET SEASET and ORNL Thermal Hydraulic Test Facility (THTF), used to assess the S-RELAP5 heat transfer model, Reference 1, incorporated constant rod powers across the experimental assembly.

Temperature differences that occurred were the result of guide tube, shroud or local heat transfer effects. In the operation of a pressurized water reactor (PWR) and in the RLBLOCA evaluation, a radial local peaking factor is present, creating power differences that tend to enhance the temperature differences between rods. In turn, these temperature differences lead to increases in net radiation heat transfer from the hotter rods. The expected rod-to-rod radiation will likely exceed that embodied within the experimental results.

6.3.1 Assessment of Rod-to-Rod Radiation Implicit in the RLBLOCA Methodoloqy As discussed above, the FLECHT-SEASET and THTF tests were selected to assess and determine the S-RELAP5 code heat transfer bias and uncertainty. Uniform radial power distribution was used in these test bundles. Therefore, the rod-to-rod temperature variation in the rods away from the vessel wall is caused primarily by the variation in the sub-channel fluid conditions. In the real operating fuel bundle, on the other hand, there can be 5 to 10 percent rod-to-rod power variation. In addition, the methodology includes a provision to apply the uncertainty measurement to the hot pin. Table 6-1 provides the hot pin measurement uncertainty and a representative local pin peaking factor for several plants. These factors, however, relate the pin to the assembly average. To more properly assess the conditions under which rod-to-rod radiation heat transfer occurs, a more local peaking assessment is required.

Therefore, the plant rod-to-rod radiation assessments herein set the average pin power for those pins surrounding the hot pin at 96 percent of that of the peak pin. For pins further removed the average power is set to 94 percent.

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I I

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-5 Table 6-1 Typical Measurement Uncertainties and Local Peaking Factors FAH Measurement Local Pin Peaking Plant Uncertainty Factor (-)

(percent) 1 4.0 1.068 2 4.0 1.050 3 6.0 1.149 4 4.0 1.113 5 4.25 1.135 6 4.0 1.058 6.3.2 Quantification of the Impact of Thermal Radiation usinq R2RRAD Code The R2RRAD radiative heat transfer model was developed by Los Alamos National Laboratory (LANL) to be incorporated in the BWR version of the TRAC code. The theoretical basis for this code is given in References 8 and 11 and is similar to that developed in the HUXY rod heatup code (Reference 10, Section 2.1.2) used by AREVA for BWR LOCA applications. The version of R2RRAD used herein was obtained from the NRC to examine the rod-to-rod radiation characteristics of a 5x5 rod segment of the 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 limitation affects the view factor calculations for guide tubes, R2RRAD is a reasonable tool to estimate rod-to-rod radiation heat transfer.

The FLECHT-SEASET test series was intended to simulate a 17x17 fuel assembly and there is a close similarity, Table 6-2, between the test bundle and a modern 17x17 assembly.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-6 Table 6-2 FLECHT-SEASET & 17x17 FA Geometry Parameters Design Parameter FLECHT-SEASET 17x17 Fuel Assembly Rod Pitch (in) 0.496 0.496 Fuel Rod Diameter (in) 0.374 0.374 Guide Tube Diameter (in) 0.474 0.482 Five FLECHT-SEASET tests (Reference 6) were selected for evaluation and comparison with expected plant behavior. Table 6-3 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.

00 0 00 0 Guide Tube 00 0 0 Hot Rod 00 *Adjacent Rods 00 0 O0 00 0 00 Figure 6-1 R2RRAD 5 x 5 Rod Segment Two sets of runs were made simulating each of the five experiments and one set of cases was run to simulate the RLBLOCA evaluation of a limiting fuel assembly in an operating plant. For the simulation of Tests 31805, 31504, 31021, and 30817, the thimble tube (guide tube) temperatures were set to the measured values. For Test 34420, the thimble tube temperature was set equal to the measured vapor temperature.. For the first experimental simulation set, the temperature of all 21 rods and the exterior boundary was set to the measured PCT of the simulated test. For the second experimental set, the hot rod temperature was set to the PCT value and the remaining 20 rods and the boundary were set to a temperature 25 'F cooler AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-7 providing a reasonable measure of the variation in surrounding temperatures. To estimate the rod-to-rod radiation in a real fuel assembly at LOCA conditions and compare it. to the experimental results, each of the above cases was rerun with the hot rod PCT set to the experimental result and the remaining rods conservatively set to temperatures expected within the bundle. The guide tubes (thimble tubes) were removed for conservatism and because peak rod powers frequently occur at fuel assembly corners away from either guide tubes or instrument tubes. In line with the discussion in Section 6.3.1, thesurrounding 24 rods were set to a temperature estimated for rods of 4 percent lower power. The boundary temperature was estimated based an average power 6 percent below the hot rod power. For both of these, the temperature estimates were achieved using a ratio of pin power to the difference in temperature between the saturation temperature and the PCT.

T24 rods = 0.96 (PCT - Tsat) + Tsat , and Trurrounding region = 0.94 * (PCT - Tsat) + Tsat.

Tsat was taken as 270 F.

Figure 6-2 shows the hot rod thermal radiation heat transfer for the two FLECHT-SEASET sets and for the plant set. The figure shows that for PCTs greater than about 1700 OF, the hot rod thermal radiation in the plant cases exceeds that of the same component within the experiments.

Table 6-3 FLECHT-SEASET Test Parameters Test Rod 7J PCT PCT htc hcaat Steam Thimble PCTtime Temperature -at Temperature t at 6-ft (F) Time (s) (Btu/hr~ft 2 .OF) 71 (6-ft) (0 F) at 6-ft (OF) 34420 2205 34 10 1850 1850*

31805 2150 110 10 1800 1800 31504 2033 100 10 1750 1750 31021 1684 29 9 1400 1350 30817 1440 70 13 900 750

set to steam temp AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-8 4.5 E-0 i4-I-

9k DI-0 11 1400 1500 1600 1700 1800 1900 2000 2100 2200 2300 2400 PCT (TF)

Figure 6-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 FA 6.3.3 Rod-to-Rod Radiation Summary In summary, the conservatism of the heat transfer modeling established by benchmark can be reasonably extended to plant applications, and the plant local peaking provides a physical reason why rod-to-rod radiation should be more substantial within a plant environment than in the test environment. Therefore, the lack of an explicit rod-to-rod radiation model, in the version of S-RELAP5 applied for realistic LOCA calculations, does not invalidate the conclusion that the cladding temperature and local cladding oxidation have been demonstrated to meet the criteria of 10 CFR 50.46 with a high level of probability.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-9 6.4 Film Boiling Heat Transfer Limit Question: In the SQN-1 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 Sequoyah 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 to 15 percent or less. This is a change to the approved RLBLOCA EM (Reference 1).

6.5 Downcomer Boiling Question: If the PCT is greater than 1800'F or the containment pressure is less than 30 psia, has the Sequoyah 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 Sequoyah Unit 1 has been established generically as adequate for the computation of downcomer phenomena including the prediction of potential local boiling effects. The model was benchmarked against the UPTF tests and the LOFT facility in the RLBLOCA methodology, Revision 0 (Reference 1). Further, AREVA addressed the effects of boiling in the downcomer in a letter, from James Malay to U.S. NRC, April 4, 2003.

The letter cites the lack of direct experimental evidence but contains sensitivity studies on high and low pressure containments, the impact of additional azimuthal noding within the downcomer, and the influence of flow loss coefficients. Of these, the study on azimuthal noding is most germane to this question; indicating that additional azimuthal'nodalization allows higher liquid buildup in portions of the downcomer away from the broken cold leg and increases the liquid driving head. Additionally, AREVA has conducted downcomer axial noding and wall heat release studies. Each of these studies supports the Revision 0 methodology and is documented later in this section.

This question is primarily concerned with the phenomena of downcomer boiling and the extension of the Revision 0 methodology and sensitivity studies to plants with low containment pressures and high cladding temperatures. Boiling, wherever it occurs, is a phenomenon that AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-10 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).

Figure 6-3 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 a result, coolant entering the downcomer may be less subcooled. When the downcomer AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-11 coolant approaches saturation, boiling on the walls initiates, reducing the downcomer hydraulic static level.

With the reduction of the downcomer level, the core inlet flow rate is reduced which, depending on the existing core inventory, may result in a cladding temperature excursion or a slowing of the core cooldown rate.

While downcomei" boiling may impact clad temperatures, it is somewhat of a self-limiting.

process. If cladding temperatures increase, less energy is transferred in the core boiling process and the loop steam flows are reduced. This reduces the required driving head to support continued core reflood and reduces the steam available to heat the ECCS water within the cold legs resulting in greater subcooling of the water entering the downcomer; The impact of downcomer boiling is primarily dependent on the wall heat release rate and on the ability to slip steam up the downcomer and out of the break. The higher the downcomer wall heat release, the more steam is generated within the downcomer and the larger the impact on core reflooding. Similarly, the quicker the passage of steam up the downcomer, the less resident volume within the downcomer is occupied by steam and the lower the impact on the downcomer average density. Therefore, the ability to properly simulate downcomer boiling depends on both the heat release (boiling) model and on the ability to track steam rising through the downcomer. Consideration of both of these is provided in the following text. The heat release modeling in S-RELAP5 is validated by a sensitivity study on wall mesh point spacing and through benchmarking against a closed form solution. Steam tracking is validated through both an axial and an azimuthal fluid control volume sensitivity study done at low pressures. The results indicate that the modeling accuracy within the RLBLOCA methodology is sufficient to resolve the effects of downcomer boiling and that, to the extent that boiling occurs; the methodology properly resolves the impact on the cladding temperature and cladding oxidation rates.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-12 6.5.1 Wall Heat Release Rate The downcomer wall heat release rate during reflood is conduction limited and depends on the vessel wall mesh spacing used in the S-RELAP5 model. The following two approaches are used to evaluate the adequacy of the downcomer vessel wall mesh spacing used in the S-RELAP5 model.

6.5.1.1 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 mesh spacing used in S-RELAP5, it is assumed that the material has constant thermal properties, is initially at temperature Ti, and, at time zero, has one surface, the surface simulating contact with the downcomer fluid, set to a constant temperature, To, representing the fluid temperature. Section 4.3 of Reference 9 gives the exact solution for the temperature profile as a function of time as (T(x,t) - T.) / (Ti - TO) = erf {x / (2.(a t)0 5)}, (1) where, a is the thermal diffusivity of the material given by a = kt(p Cp),

k = thermal conductivity, p = density, Cp = specific heat, and erf{} is the Gauss error function (given in Table A-1 of Reference 9).

The conditions of the benchmark are Ti = 500 °F and To = 300 OF. The mesh spacing in S-RELAP5 is the same as that used for the downcomer vessel wall in the RLBLOCA model.

Figure 6-4 shows the temperature distributions in the metal at 0.0, 100 and 300 seconds as calculated. by using Equation 1 and S-RELAP5, respectively. The solutions are identical confirming the adequacy of the mesh spacing used in the downcomer wall.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-13 550 500 ILL450 o.400 E

F-I-

---g--Closed Form, 0 s 350

-Closed Form, 100 s

-~- Closed Form, 300 s 30 -- O S-RELAP5, 0 s

-- 0O- S-RELAP5, 100 s

--- O-S-RELAP5, 300 s 250 T 1 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 Distance from Inner Wall, feet Figure 6-4 S-RELAP5 versus Closed Form Solution AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-14 6.5.1.2 Plant Model Sensitivity Study As additional verification, a typical 4-loop plant case was used to evaluate the adequacy of the mesh spacing within the downcomer wall heat structure. Each mesh interval in the base case downcomer vessel wall was divided into two equal intervals. Thus, a new input model was created by increasing the number of mesh intervals from 9 to 18. The following four figures show the total downcomer metal heat release rate, PCT independent of elevation, downcomer liquid level, and the core liquid level, respectively, for the base case and the modified case.

These results confirm the conclusion from the exact solution study that the mesh spacing used in the plant model for the downcomer vessel wall is adequate.

!li!!t~~ .......... . ~ ...~~~ .......... .

(Bse VS Wall 1,9-mesh) 8..

-_ ....=1a-Mesh VSL Wall 240CO.[10 .......

r i I ' I %

Ca

-... __ .1._ _ _ _ _.__.

r-r-/ 1 I Time (sec)

Figure 6-5 Downcomer Wall Heat Release - Wall Mesh Point Sensitivity AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-15 2401GC0

--- Base VSjWall (9meshl L.. 18-Mesh .VSL Wall U-0 1200.00 I E

a) 6W.00 0.03 80.0 160.0 240 .0320.0 4we.

Time (sec)

Figure 6-6 PCT Independent of Elevation - Wall Mesh Point Sensitivity AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-16

---Base VSU Wall (9- meshN i

a)

.2,

-j S10.00

0. 0O0 0[J0 160.0 24n.0 320.0 4,00 Time (sec)

Figure 6-7 Downcomer Liquid Level - Wall Mesh Point Sensitivity AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larae Break LOCA Analysis Page 6-17 D

160.0 4000 Time (sec)

Figure 6-8 Core Liquid Level - Wall Mesh Point Sensitivity AREVA NP Inc.

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Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-18 6.5.2 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.

6.5.2.1 Azimuthal Nodalization In a letter to the NRC dated April, 2003 (Reference 1), AREVA documented several studies on downcomer boiling. Of significance here is the study on further azimuthal break up of the downcomer noding. The study, based on a 3-loop plant with a containment pressure of approximately 30 psia during reflood, consisted of several calculations examining the affects on clad temperature and other parameters.

The base model, with 6 axial by 3 azimuthal regions, was expanded to 6 axial by 9 azimuthal regions (Figure 6-9). The base calculation simulated the limiting PCT calculation given in the EMF-2103. three-loop sample problem. This case was then repeated with the revised 6 x 9 downcomer noding.

The change resulted in an alteration of the blowdown evolution of the transient with little evidence of any affect during reflood. To isolate any possible reflood impact that might have an influence on downcomer boiling, the case was repeated with a slightly adjusted vessel-side break flow. Again, little evidence of impact on the- reflood portion of the transient was observed.

The study concluded that blowdown or near blowdown events could be impacted by refining the azimuthal resolution in the downcomer but that reflood would not be impacted. Although the study was performed for a somewhat elevated system pressure, the flow regimes within the downcomer will not differ for pressures as low as atmospheric. Thus, the azimuthal downcomer modeling employed for the RLBLOCA methodology is reasonably converged in its ability to represent downcomer boiling phenomena.

AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Lame Break LOCA Analysis Page 6-19 Base model 9D-ý.

ý-C-F4L-) ý-:

.'CCO1 L~~~PH j F *, dr Revised 9 Region Model Figure 6-9 Azimuthal Noding 6.5.2.2 Axial Nodalization The RLBLOCA methodology divides the downcomer into six nodes axially. In both 3-loop and 4-loop models, the downcomer segment at the active core elevation is represented by two equal length nodes. For most operating plants, the active core length is 12 feet and the downcomer segments at the active core elevation are each 6-feet high. (For a 14 foot core, these nodes would be 7-feet high.) The model for the sensitivity study presented here comprises a 4-loop plant with an ice condenser containment and a 12 foot core. For the study, the two nodes spanning the active core height are divided in half, revising the model to include eight axial nodes. Further, the refined noding is located within the potential boiling region of the AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-20 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 6-10 provides the containment back pressure for the base modeling. Figures 6-11 through 6-14 show the total downcomer metal heat release rate, PCT independent of elevation, downcomer liquid level, and the core liquid level, respectively, for the base case and the modified case.

The results demonstrate that the axial resolution provided in the base case, 6 axial downcomer node divisions with 2 divisions spanning the core active region, are sufficient to accurately resolve void distributions within the downcomer. Thus, this modeling is sufficient for the prediction of downcomer driving head and the resolution of downcomer boiling effects.

40.00 -

i-

Base 6x6 casei Z't200 . . ........ .................................................

cl, C,,

24.0C "S.

"-.9 16,00 --

0.0 R0.0 100.0 240 .02~20.-) 4030.0 Time (sec)

Figure 6-10 Lower Compartment Pressure versus Time AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larme Break LOCA Analysis Page 6-21 U) 24000 .0 180(0 Ob LU 12(*0U0.(

0.0 0.00 s0o 06.0.t 240.0 3200 40U.0 Time (sec)

Figure 6-11 Downcomer Wall Heat Release - Axial Noding

,Sensitivity Study AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-22

-- Base 6x6 .Case!

. 8--

x6 Case

1 I500.00 --

U, U- /

/3 0

JW .... ....

120000 I E ..

(D . . ...... .....

a-..- a.-- s---

0.0I O.0 It0110 150.0 Time (sec) 240.0 320,0 400.0 Figure 6-12 PCT Independent of Elevation - Axial Noding Sensitivity Study AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larme Break LOCA Analvsis Page 6-23 30.00 T

Base M-- as Ix6 e6Case 20.00 ci, ci,

-j

-o lIT' 0~

10.00 0.00.

0,0 82.0 160.0 240.0 320.0 400.0 Time (sec)

Figure 6-13 Downcomer Liquid Level - Axial Noding Sensitivity Study AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-24 W0.OO (D

-J 4c~-o Time (sec)

Figure 6-14 Core Liquid Level - Axial Noding Sensitivity Study 6.5.3 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. B6th 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-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-25 6.6 Break Size Question: 'Were all break sizes assumed greaterthan or equal to 1.0 It?

Response: Yes.

The NRC has requested that the break spectrum for the realistic LOCA evaluations be limited to accidents that evolve through a range of phenomena similar to those encountered for the larger break area accidents. This is a change to the approved RLBLOCA EM (Reference 1). The larger break area LOCAs are typically characterized by the occurrence of dispersed flow film boiling at the hot spot, which sets them apart from smaller break LOCAs. This occurs generally in the vicinity of 0.2 DEGB (double-ended guillotine break) size (i.e., 0.2 times the total flow area of the pipe on both sides of the break). However, this transitional break size varies from plant to plant and is verified only after the break spectrum has been executed. AREVA NP has sought to develop sufficient criteria for defining the minimum large break flow area prior to performing the break spectrum. The purpose for doing so is to assure a valid break spectrum is performed.

6.6.1 Break / Transient Phenomena In determining the AREVA NP criteria, the characteristics of larger break area LOCAs are examined. These LOCA characteristics involve a rapid and chaotic depressurization of the reactor coolantsystem (RCS) during which the three historical approximate states of the system can be identified.

Blowdown The blowdown phase is defined as the time period from initiation of the break until flow from the accumulators begins. This definition is somewhat different from the traditional definition of blowdown which extends the blowdown until the RCS pressure approaches containment pressure. The blowdown phase typically lasts about 12 to 25 seconds, depending on the break size.

Refill is that period that starts with the end of blowdown, whichever definition is used, and ends when water is first forced upward into the core. During this phase the core experiences a near adiabatic heatup.

Reflood is that portion of the transient that starts With the end of refill, follows through the filling of the core with water and ends with the achievement of complete core quench.

AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-26 Implicit in this break-down is that the core liquid inventory has been completely, or nearly so, expelled from the primary system leaving the core in a state of near core-wide dispersed flow film boiling and subsequent adiabatic heatup prior to the reflood phase. Although this break down served as the basis for the original deterministic LOCA evaluation approaches and is valid for most LOCAs that would classically be termed large breaks, as the break area decreases the depressurization rate decreases such that these three phases overlap substantially. During these smaller break events, the core liquid inventory is not reduced as much as that found in larger breaks. Also, the adiabatic core heatup is not as extensive as in the larger breaks which results in much lower cladding temperature excursions.

6.6.2 New Minimum Break Size Determination No determination of the lower limit can be exact. The values of critical phenomena that control the evolution of a LOCA transient will overlap and interplay. This is especially true in a statistical evaluation where parameter values are varied randomly with a strong expectation that the variations will affect results. In selecting the lower area of the RLBLOCA break spectrum, AREVA sought to preserve the generality of a complete or nearly complete core dry out accompanied by a substantially reduced lower plenum liquid inventory. It was reasoned that such conditions would be unlikely if the break flow rate was reduced to less than the reactor coolant pump flow. That is, if the reactor coolant pumps are capable of forcing more coolant toward the reactor vessel than the break can extract from the reactor vessel, the downcomer and core must maintain'some degree of positive flow (positive in the normal operations sense).

The circumstance is, of course, transitory. Break flow is altered as the RCS blows down and the RC pump flow may decrease as the rotor and flywheel slow down if power is lost. However, if the core flow was reduced to zero or became negative immediately after the break initiation, then the event was quite likely to proceed with sufficient inertia to expel most of the reactor vessel liquid to the break. The criteria base, thus established, consists of comparing the break flow to the initial flow through all reactor coolant pumps and setting the minimum break area such that these flows match. This is done as follows:

Wbreak = Abreak

Gbreak = Npump

WRCP.

This gives AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-27 Abreak = (Npump

WRcP)/Gbreak.

The break mass flux is determined from critical flow. Because the RCS pressure in the broken cold leg will decrease rapidly during the first few seconds of the transient, the critical mass flux is averaged between that appropriate for the initial operating conditions and that appropriate for the initial cold leg enthalpy and the saturation pressure of coolant at that enthalpy.

V Gbreak ": (Gbreak(Po, HCLO) + Gbreak(PCLsat, HCLO))/ 2 .

The estimated minimum LBLOCA break area, Amin, is 2.76 ft2 and the break area percentage, based on the full double-ended guillotine break total area, is 33 percent.

Table 6-4 provides a listing of the plant type, initial condition, and the fractional minimum RLBLOCA break area, for all the plant types presented as generic representations in the next section.

Table 6-4 Minimum Break Area for Large Break LOCA Spectrum Saturated Spectrum Spectrum System Pt Pressure Cold Leg Enthalpy Subcooled Gbreak Gbrak No.

of RCP flow Minimum Break Minimum Break Description (psia) (Btu/Ibm) (Ibmrft2-s) (HEM) s(ft')

(s)t (IbMft 2 RCPs (lbmls) Area Area (DEGB)

A 3-Loop Design W 2250 555.0 23190 5700 3 31417 2.18 0.26 B 3-Loop 23880 5450 4 28124 1.92 0.23 Design W 2250 544.5 C 3-Loop Design W .2250 550.0 23540 5580 4 29743 2.04 .0.25 2x4CE Design 2100 538.8 22860 5310 4 21522 1.53. 0.24 E 2x4 CE Design 2055 535.8 22630 5230 3 37049 2.66 0.27 F 4-Loop W 2160 540.9 23290 5370 3 39500 2.76 0.33 Design 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 (Amon) and an area of twice the size of the broken pipe. The determination of break configuration, split versus double-ended, is made after the break area is selected based on a uniform probability for each occurrence.

AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-28 6.6.3 Intermediate Break Size Disposition With the revision of the smaller break area for the RLBLOCA analysis, the break range for small breaks and large breaks are no longer contiguous. Typically the lower end of the large break spectrum occurs at between 0.2 to 0.3 times the total area of a 100 'percent double-ended guillotine break (DEGB) and the upper end of the small break spectrum occurs at approximately 0.05 times the area of a 100 percent DEGB. This leaves a range of breaks that are not specifically analyzed during a LOCA licensing analysis. The premise for allowing this gap is that these breaks do not comprise accidents that develop high cladding temperature and thus do not comprise accidents that critically challenge the emergency core cooling systems (ECCS).

Breaks within this range remain large enough to blowdown to low pressures. Resolution is provided by the large break ECC systems and the pressure-dependent injection limitations that determine critical small break performance are avoided. Further, these accidents develop relatively slowly, assuring maximum effectiveness of those ECC systems.

A variety of plant types for which analysis within the intermediate range have been completed were surveyed. Although statistical determinations are extracted from the consideration of breaks with areas above the intermediate range, the AREVA best-estimate methodology remains suitable to characterize the ECCS performance of breaks within the intermediate range.

Table 6-4 provides a listing of the plant type, initial condition, and the fractional minimum RLBLOCA break area. Figures 6-15 through 6-20 provide the enlarged break spectrum results with the upper end of the small break spectrum and the lower end of the large break spectrum indicated by bars. Table 6-5 provides differences between the true large break region and the intermediate break region (break areas between that of-the largest SBLOCA and the smallest RLBLOCA). The minimum difference is 141 OF; however, this case is not representative of the general trend shown by'the other comparisons. The next minimum difference is 704 OF (see Figure 6-15). Considering this point as an outlier, the table shows the minimum difference between the highest intermediate break spectrum PCT and large break spectrum PCT, for the six plants, as at least 463 OF, and including this point would provide an average difference of 427 OF and a maximum difference of 840 OF.

Thus, by both measures, the peak cladding temperatures within the intermediate break range will be several hundred degrees below those in the true large break range. Therefore, these breaks will not provide a limit or a critical measure of the ECCS performance. Given that the AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-29 large break spectrum bounds the intermediate spectrum, the use of only the large break spectrum meets the requirements of 10CFR50.46 for breaks within the intermediate break LOCA spectrum, and the method demonstrates that the ECCS for a plant meets the criteria of 10CFR50.46 with high probability.

Table 6-5 Minimum PCT Temperature Difference - True Large and Intermediate Breaks Generic Maximum Maximum Plant Plant PCT (OF) PCT (OF) Delta PCT Average Delta Description Label Intermediate Large Size (OF) PCT (OF)

(Table 6-4) Size Break Break A 17461 1887 1411 3-LoopW B 1273 1951 678 4271 Design C 1326 1789 463 2x4 CE D 984 1751 767 Design E 869 1636 767 4-LoopW F 1127 1967 840 840 Design Note: 1. The 2 nd highest PCT was 1183 OF. This changes the Delta PCT to 704 OF and the average delta increases to 615 OF.

AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis FageU- 0 2000 T r- 1 2 Upper En dof Large Break SSBLOCA Spectrum I, Break Siz e Minimum 1800 _ _Spectrum

- - Break Area 41 1600 4-- 4

----- A-4

I 4, 4 4

-- 44 8

- 4L 4

4

I 4

1400- .4-, -. - 4 41 r... 4, 4

4 e'I T4 1200 +

0 1000 + 4 4--

800 +

600 ---

0.0000 0. 1000 0,2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine Figure 6-15 Plant A- Westinghouse 3-Loop Design AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-31 2000 T - "1 - - - -

Upper End of Large Break SBLOCA

Spectrum Break Size Minimum 1800- - _Spectrum

.... Break Area

4; 1600 -- -- * - --- *4; - -

4*I j- 4; 4

'4 # *1

,, -L-1400 C, *j

1 I.

o,',

1200 +---

I

1000 +--

I, 800 I

600 0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine Figure 6-16 Plant B - Westinghouse 3-Loop Design AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Rprali~tic I arnp. Rrp~k I COCA Analysis Paqe 6-32 Realistic Larne Break LOCA Analvsis 2000 .--- -I- - --.... -I- - - - -

U'pper End of Large Break BLOCA Spectrum

-Speak Size Minimum pectrum - Break Area 1800 -

9

-I 9 1600 +--- ..... I--

91

9 14 9 9 1400 +

j ---

I -

91 U,

~~.2..

1200 +

I*

T ---

1000 800 111)1'4 !

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 Eided Guillotine Figure 6-17 Plant C - Westinghouse 3-Loop Design AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Paae 6-33 2000 r Upper End of Large Break SBLOCA Spectrum Break Size Minimum 1800 + Spectrum - -

Break Area Ikoo - --- ----

I i

1

I.

I.

1400 + -i- 4 I I r- . -... -

I I I I 1200 + I I

. . . .I . . * . . .

1000 -

800 -

600 --

0.0000 0.1000 0.2000 0.3000 0.4000 0.5000 0.6000 0.7000 0.8000 0.9000 1.0000 Break Area Normalized to Double Ended Guillotine Figure 6-18 Plant D - Combustion Engineering 2x4 Design AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larae Break LOCA Analysis Paae 6-34 2000 ~1 d

Upper End of Large Brecak SBLOCA Spectrum Break Size Minimum 1800 Spectrum Break Ar I.I 0' 4

1600 -

1

1 I.

.1 I.

1400 -

I II 4 I.

U M.

L 1200

-.. T~1 1000 --

t4

~* .

800 -

1 600 1 wH 0.00 00 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 Figure 6-19 Plant E - Combustion Engineering 2x4 Design AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Larqe Break LOCA Analysis Page 6-35 2200.0000 T --- 4 Large Break Spectrum Upper End of _ Minimum 2000.0000 -

SBLOCA IL" Break Area Break Size --- -- ---.

Spectrum 1800.0000 +

A- - - - - - -Iý I.

1600.0000 - 'I 14' o *I.

1400.0000 4-

- -I--

- ...I - - - - ~L. - -

1200.0000 -

I4 I.

1000.0000 -

II 800.0000 -- -I- - 4 I,

600.0000 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 Figure 6-20 Plant F - Westinghouse 4-Loop Design AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-36 6.7 ICECON Model Question: Verify that the SQN-1 ICECON model is that shown in Figure 5.1 of EMF-CC-39(P)

Revision 2, "ICECON: A Computer Program Used to Calculate Containment Back Pressure.for L OCA Analysis (Including Ice CondenserPlants)."

See Section 3.3.

6.8 Cross-References to North Anna Question: In order to conduct its review of the SQN-1 application of AREVA's realisticLBLOCA 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 SQN-1 except for that information related specifically to North Anna and to sub-atmospheric containments.

AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis. Page 6-37 Response: The responses provided to questions 1, 2, 4, and 6 are for the most part generic and related to the ability of ICECON to calculate containment pressures. Excepting as follows they are applicable to the Sequoyah Unit 1 RLBLOCA submittal.

Question 1 - Completely Applicable Question 2 - Completely Applicable Question 4 - Completely Applicable (the reference to CSB 6-1 should now be to CSB Technical Position 6-2). The NRC altered the identification of this branch technical position in Revision 3 of NUREG-0800.

Question 6 - The direct response is completely applicable excepting that the reference to "North Anna Units 1 and 2" should be deleted. The statement in which the North Anna units are referenced is equally valid without identification of any specific plant.

The supplemental request and response are specific to. North Anna and are not applicable to Sequoyah Unit 1.

The response provided to question 1 contains both generic and plant specific content. The portions that are generic remain applicable to Sequoyah Unit 1. However, the North Anna Units use sub-atmospheric containment designs and Sequoyah Unit 1 is of the ice condenser type.

This leads to several differences in the way the information would be presented.

AREVA NP'Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-38 6.9 Containment Model Question: ANP-2695(P) shows that the containment parameters treated statistically are: (1) upper compartment containment volume, (2) upper compartment containment temperature, and (3) lower compartment containment temperature. ANP-2695(P) states that "in many instances" the guidance of NRC Branch Technical Position CSB 6-1 was used in determining the other containment parameters.

[AREVA NP: Note that the same Containment System Branch Technical Position is now designated 6-2 instead of 6-1.]

(a) How is the mixing of containment steam and ice melt modeled so as to minimize the containment pressure?

See Section 3.3.

(b) Verify that all containment spray and fan coolers are assumed operating at maximum heat removal capacity.

See Section 3.3.

(c) Describe how the limits on the volume of the upper containment were determined.

See Section 3.3.

(d) How are the containment air return fans modeled and what is the effect of this modeling on the containment pressure?

See Section 3.3.

(e) Describe how passive heat sink areas and heat capacities are modeled so as to minimize containment pressure.

See Section 3.3.

The following are a set of containment plots that are produced to supplement the NRC's review of the Sequoyah Unit 1 RLBLOCA analysis.

AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit*1 Revision 0 Realistic Large Break LOCA Analysis Page 6-39 4SOO -i-Energy Addition trom RCS break 4000 -0 steam 4500 3500 [- ~totaI 3000 0 10 20 30 "40 50 60 70 80 90 100 110 120 Time (s)

Figure 6-21 Energy Addition in Lower Compartment o1500t 1000 0

4500

-- Total heat rermoval 4000 D:5540)2J-020%819.V:4*R510-1--0 Energy addition by steam 3500 3000 2500

- 2000 500

-50'0-- --

0 10 20 30 40 50 60 70 80 90 100 110 120 Time (s)

Figure 6-22 Energy Rates in Lower Compartment AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-40 4500-r---'-- ---

Heat removal by slabs 4000 --- 0 Pool condensation i "

Liquid energy dropout 3500 --,*-- Chest sump drainage 7--A Total removal 3000 - totalv 2500

" 2000 1500o 1000 500

-500 ... . ......

0 10 20 30 40 50 60 70 80 90 100 110 120 Time (s)

Figure 6-23 Energy Removal Rates in Lower Compartment 8 00 0 . . . . . , . . . . ,. . . . . . . . .,. . . . . . . . . .,. .

7500 -- Heat removal by slabs

-0 Pool condensation 7000 -- 500 Liquid energy dropout 6500 -K-Spray energy removal 6000 ----- Total removal 5500 5000 D 4500

'a 4000 3500 o 3000 E 2500 2000 1500 1000 500 *

-500 0 10 20 30 40 50 60. 70 80 90 100 110 120 ID:554 2JOOBn1 rre I'eSi WX Time (s)

Figure 6-24 Energy Removal Rates in Upper Compartment AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-41 15.0 10.0 F-5.0 10 15 20 25 30 106545-0 1Q37-4 40%0.X I 1Jw2000 Time (s)

Figure 6-25 Heat Removal Rates (log) 1.0 0.9 E

0 0.8 0.7 0.6 0.5 0 100 200 300 400 500 600 C55400 2s&v2oo i931460001441 Time (s)

Figure 6-26 Fraction of Ice Remaining AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page6-42 11000 I I II Break 10000 ------o Steam

--- Liquid 9000 .OTotal 8000 -

7000 6000 5000 4000 3000 2000 1000 LPL 0'

0 10 20 30 40 50 60 70 80 90 100 110 120 0:5- 214 Ja22 193a:ý MEW: I Time (s)

Figure 6-27 Mass Addition to Lower Compartment 22.0 2. 2 0. . * . . . . . . i. . . . .. . . i . . .. . . . .. . . . . ' ' . .

21.0 - Upper Compartment

--+ Lower Compartment 20.0 "Z 19.0 2 18.0 17.0 16.0 15.0 - ' "

0 10 20 30 40 50 60 70 80 90 100 11c ID554ee2J r*2e0l37:l nýMa:, ,Time(s)

Figure 6-28 Upper Compartment versus Lower Compartment Pressure AREVA NP, Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0.

Realistic Large Break LOCA Analysis Page 6-43 250.0 200.0 U-150.0 100.0 50 .0 .... .. . - .. . : . .

0 10 20 30 40 50 60 70 80 90 100 110 120 ID:55400 ,37:14P04.X:1 1Ja2(OC Time (s)

Figure 6-29 Temperature of Upper and Lower Compartments 6.10 GDC 35- LOOP and No-LOOP Case Sets In concurrence with the NRC's interpretation of GDC 35, a set of 59 cases each was run with a LOOP and No-LOOP assumption. The set of 59 cases that predicted the highest figure of merit, PCT, is reported in Section 2 and Section 3, herein. The results from both case sets are shown in Figure 3-23. This is a change to the approved RLBLOCA EM (Reference 1).

AREVA NP Inc.

ANP-2695(NP)

Sequoyah Nuclear Plant Unit 1 Revision 0 Realistic Large Break LOCA Analysis Page 6-44 6.11 Statement Question: Provide a statement confirming that TVA and its LBLOCA analyses vendor have ongoing processes that assure that the input variablesand ranges of parameters for the SQN-1 LBLOCA analyses conservatively bound. the values and ranges of those parameters for the as operated SQN-1 plant. This statement addresses certain programmatic requirements of 10 CFR 50.46, Section (c).

Response: TVA and the LBLOCA Analysis Vendor have an ongoing process to ensure that all input variables and parameter ranges for the Sequoyah Unit 1 realistic large break loss-of-coolant accident are verified as conservative with respect to plant operating and design conditions. In accordance with TVA Quality Assurance program requirements, this process involves 1) definition of the required input variables and parameter ranges by the Analysis Vendor, 2) compilation of the specific values from existing plant design input and output documents by TVA and Vendor personnel in a formal analysis input summary document issued by the Analysis Vendor and 3) formal review and approval of the input summary document by TVA. Formal TVA approval of the input document serves as the release for the Vendor to perform the analysis.

Continuing review of the input summary document is performed by TV.A as part of the plant design change process and cycle-specific core design process. Changes to the input summary required to support plant modifications or cycle-specific core alternations are formally communicated to the Analysis Vendor by TVA. Revisions and updates to the analysis parameters are documented and approved in accordance with the process described above for the initial analysis.

AREVA NP Inc.

ENCLOSURE4 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNIT 1 PROPRIETARY INFORMATION WITHHOLDING AFFIDAVIT E4-1

AFFIDAVIT COMMONWEALTH OF VIRGINIA

) ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. 1 am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the report ANP-2695(P), Revision 0, 'Sequoyah Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis,"

dated February 2008, and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information, reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this ____

day of -72008.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/10 Reg. # 7079129 SH1ERR L. MCFAVIN Commonwealth of VIigI o uO27079129

'My CommidsslOnt . 0 Expires Oct 31. 2010

v t e Letter Sequence Request
TAC:MD8532, (Approved, Closed)
Initiation Request Acceptance... Administration Withholding Request Acceptance Questions Answers Results Approval Other: ML081080337
MONTHYEAR2008-04-14 [Table View]

ML081080338

Text

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