ML11172A064

ANP-2970(NP), Rev 0, Sequoyah, Units 1 and 2 Htp Fuel Realistic Large Break LOCA Analysis, Attachment 10

ML11172A064
Person / Time
Site:	Sequoyah
Issue date:	04/30/2011
From:	AREVA NP
To:	Office of Nuclear Reactor Regulation
References
TS-SQN-2011-07 ANP-2970(NP), Rev 0
Download: ML11172A064 (163)

Text

ATTACHMENT 10 ANP-2970(NP), Revision 0 Sequoyah Units 1 and 2 HTP Fuel Realistic Large Break LOCA Analysis April 2011 (NON-PROPRIETARY VERSION)

AREVA NP Inc.

ANP-2970(NP)

Revision 0 Sequoyah Units 1 and 2 HTP Fuel Realistic Large Break LOCA Analysis April 2011 ZA AREVA HP Inc. AREVA

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page i Copyright © 2011 AREVA NP Inc.

All Rights Reserved AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page ii Nature of Changes Item Page Description and Justification

1. All This is a new document.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page iii Contents 1.0 Introduction .................................................................................................................... 1-1 2.0 Sum m ary ........................................................................................................................ 2-1 3.0 Analysis .......................................................................................................................... 3-1 3.1 Description of the LBLOCA Event ...................................................................... 3-2 3.2 Description of Analytical Models ......................................................................... 3-3 3.3 Plant Description and Sum mary of Analysis Parameters ................................... 3-5 3.4 SER Com pliance ................................................................................................ 3-8 3.5 Realistic Large Break LOCA Results ............................................................... 3-10 4.0 Generic Support for Transition Package ........................................................................ 4-1 4.1 Reactor Power .................................................................................................... 4-1 4.2 Rod Quench ....................................................................................................... 4-2 4.3 Rod-to-Rod Thermal Radiation .......................................................................... 4-2 4.4 Film Boiling Heat Transfer Lim it ......................................................................... 4-8 4.5 Downcomer Boiling ............................................................................................ 4-8 4.6 Break Size ........................................................................................................ 4-24 4.7 Detail information for Containment Model (ICECON) ....................................... 4-35 4.8 Cross-References to North Anna ..................................................................... 4-48 4.9 GDC 35 - LOOP and No-LOOP Case Sets ..................................................... 4-49 4.10 Statement ......................................................................................................... 4-49 5.0 Conclusions .................................................................................................................... 5-1 6.0 Recent NRC Request for Additional Information (RAI) and AREVA Responses ..................................................................................................................... 6-1 7.0 References ..................................................................................................................... 7-1 This document contains a total of 152 pages.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page iv Tables Table 2-1 Summary of Major Parameters for Limiting Transient .............................................. 2-2 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 4-1 Typical Measurement Uncertainties and Local Peaking Factors .............................. 4-4 Table 4-2 FLECHT-SEASET & 17x17 FA Geometry Parameters ............................................ 4-5 Table 4-3 FLECHT-SEASET Test Parameters ......................................................................... 4-6 Table 4-4 Minimum Break Area for Large Break LOCA Spectrum ......................................... 4-26 Table 4-5 Minimum PCT Temperature Difference - True Large and Intermediate B re a ks .......................................................................................................................... 4 -2 8 AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page v Figures Figure 3-1 Primary 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 ....................................................................... 3-30 Figure 3-9 PCT versus Break Size Scatter Plot ...................................................................... 3-31 Figure 3-10 Maximum Transient Oxidation versus PCT Scatter Plot ...................................... 3-32 Figure 3-11 Total Oxidation versus PCT Scatter Plot ............................................................. 3-33 Figure 3-12 Peak Cladding Temperature (Independent of Elevation) for the Lim itin g C a se ............................................................................................................... 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 Flux for 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, SI and RHR) for the Lim itin g Ca se ............................................................................................................... 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 Reactor Vessel Liquid Mass (Ibm) versus Time (sec) ......................................... 3-45 Figure 3-24 GDC 35 LOOP versus No-LOOP Cases ............................................................. 3-46 Figure 3-25: PCT Node Cladding Surface Temperature and Saturation T em perature, Case 29 ................................................................................................. 3-47 Figure 3-26: PCT Node Cladding Surface Temperature and Saturation T em perature, C ase 32 ................................................................................................. 3-47 Figure 3-27: PCT Node Cladding Surface Temperature and Saturation T em perature, C ase 35 ................................................................................................. 3-48 Figure 3-28: PCT Node Cladding Surface Temperature and Saturation T em perature, C ase 42 ................................................................................................. 3-48 AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page vi Figure 3-29: PCT Node Cladding Surface Temperature and Saturation T em perature, C ase 68 ................................................................................................. 3-49 Figure 3-30: PCT Node Cladding Surface Temperature and Saturation T em perature, C ase 87 ................................................................................................. 3-49 Figure 4-1 R2RRAD 5x5 Rod Segment .................................................................................... 4-5 Figure 4-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 FA ................................................................................................................................... 4 -7 Figure 4-3 Reactor Vessel Downcomer Boiling Diagram .......................................................... 4-9 Figure 4-4 S-RELAP5 versus Closed Form Solution .............................................................. 4-12 Figure 4-5 Downcomer Wall Heat Release - Wall Mesh Point Sensitivity .............................. 4-13 Figure 4-6 PCT Independent of Elevation - Wall Mesh Point Sensitivity ................................ 4-14 Figure 4-7 Downcomer Liquid Level - Wall Mesh Point Sensitivity ........................................ 4-15 Figure 4-8 Core Liquid Level - Wall Mesh Point Sensitivity ................................................... 4-16 Figure 4-9 Azim uthal Noding .................................................................................................. 4-18 Figure 4-10 Lower Compartment Pressure versus Time ........................................................ 4-19 Figure 4-11 Downcomer Wall Heat Release - Axial Noding Sensitivity Study ....................... 4-20 Figure 4-12 PCT Independent of Elevation - Axial Noding Sensitivity Study ......................... 4-21 Figure 4-13 Downcomer Liquid Level - Axial Noding Sensitivity Study .................................. 4-22 Figure 4-14 Core Liquid Level - Axial Noding Sensitivity Study ............................................. 4-23 Figure 4-15 Plant A - Westinghouse 3-Loop Design .............................................................. 4-29 Figure 4-16 Plant B - Westinghouse 3-Loop Design .............................................................. 4-30 Figure 4-17 Plant C - Westinghouse 3-Loop Design .............................................................. 4-31 Figure 4-18 Plant D - Combustion Engineering 2x4 Design .................................................. 4-32 Figure 4-19 Plant E - Combustion Engineering 2x4 Design ................................................... 4-33 Figure 4-20 Plant H - Westinghouse 4-Loop Design .............................................................. 4-34 Figure 4-21 PCT vs. Containment Volume ............................................................................. 4-36 Figure 4-22 PCT vs. Initial Containment Temperature ........................................................... 4-37 Figure 4-23 Containment Pressure for Limiting Case ............................................................. 4-38 Figure 4-24 Energy Addition in Lower Compartment .............................................................. 4-39 Figure 4-25 Energy Rates in Lower Compartment ................................................................. 4-40 Figure 4-26 Energy Removal Rates in Lower Compartment .................................................. 4-41 Figure 4-27 Energy Removal Rates in Upper Compartment .................................................. 4-42 Figure 4-28 Heat Removal Rates (log) ................................................................................... 4-43 Figure 4-29 Fraction of Ice Remaining .................................................................................... 4-44 Figure 4-30 Mass Addition to Lower Compartment ................................................................ 4-45 AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page vii Figure 4-31 Upper Compartment and Lower Compartment Pressure .................................... 4-46 Figure 4-32 Temperature of Upper and Lower Compartments ............................................... 4-47 Figure 6-1 Fractional Fuel Centerline Temperature Delta Between RODEX3A a n d Da ta ......................................................................................................................... 6 -6 Figure 6-2 Fuel Centerline Temperature Delta of RODEX3A Calculations to Data (Original and Using the New Correlation) ...................................................................... 6-7 Figure 6-3 Correction Factor (as applied for temperatures in Kelvin) ....................................... 6-8 Figure 6-4 K(B urnup) C urve ...................................................................................................... 6-9 Figure 6-5 Radial Temperature Profile for Hot Rod ................................................................ 6-10 Figure 6-6 Temperature versus Time for Fuel Centerline, Clad Surface, and Fuel Av e ra g e ........................................................................................................................ 6 -1 1 Figure 6-7 Fresh and Once-Burned U0 2 Rod PCT Trace ..................................................... 6-12 Figure 6-8 Clad Temperature Response from Single Failure Study ....................................... 6-16 Figure 6-9 Comparison of PCT Independent of Elevation for Max ECCS and Min E C C S ........................................................................................................................... 6 -18 Figure 6-10 Comparison of Containment and System Pressure for Max ECCS a nd Min E C C S ............................................................................................................. 6-19 Figure 6-11 Comparison of ECCS Flows for Max ECCS and Min ECCS ............................... 6-20 Figure 6-12 Downcomer Level ................................................................................................ 6-21 Figure 6-13 Decay Heat Comparisons, Infinite Operation U235, Finite Operation A ll Isotopes (0.1 - 10 sec) ............................................................................................ 6-32 Figure 6-14 Decay Heat Comparisons, Infinite Operation U235, Finite Operation A ll Isotopes (10 - 1000 sec) ......................................................................................... 6-33 Figure 6-15 Decay Heat Ratios, Finite Operation over Infinite Operation U235 All Isotopes (0 - 10 sec) .................................................................................................... 6-34 Figure 6-16 Decay Heat Ratios, Finite Operation over Infinite Operation U235 All Isoto pes (> 10 sec) ........................................................................................................ 6-35 AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae viii Nomenclature AFD Axial Flux Difference ASI Axial Shape Index BLEU Blended Low Enriched Uranium BWR Boiling Water Reactor CCTF Cylindrical Core Test Facility CFR Code of Federal Regulations CPHS Containment Pressure High Signal CSAU Code Scaling, Applicability, and Uncertainty DC Downcomer DEGB Double-Ended Guillotine Break DFSS Design For Six Sigma DNB Departure from Nucleate Boiling ECCS Emergency Core Cooling System EFPH Effective Full Power Hours EM Evaluation Model FA Fuel Assembly Full-Length Emergency Core Heat Transfer Separate Effects and Systems Effects FLECHT-SEASET Tests FQ Total Peaking Factor FAH Nuclear Enthalpy Rise Factor HPSI High Pressure Safety Injection HFP Hot Full Power LANL Los Alamos National Laboratory LEFM Leading Edge Flow Meter LHGR Linear Heat Generation Rate LOCA Loss of Coolant Accident LOFT Loss of Fluid Test LOOP Loss of Offsite Power LPSI Low Pressure Safety Injection MSIV Main Steam Isolation Valve NRC U. S. Nuclear Regulatory Commission NSSS Nuclear Steam Supply System AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page ix Nomenclature (Continued)

PCT Peak Clad Temperature PIRT Phenomena Identification and Ranking Table PLHGR Planar Linear Heat Generation Rate PPLS Pressurizer Pressure Low Signal PWR Pressurized Water Reactor RAI Request for Additional Information RAS Recirculation Actuation Signal RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RLBLOCA Realistic Large Break Loss of Coolant Accident RV Reactor Vessel RWST Refueling Water Storage Tank SGLS Steam Generator Low (pressure) Signal SIAS Safety Injection Activation Signal SI Safety Injection SER Safety Evaluation Report THTF Thermal Hydraulic Test Facility TVA Tennessee Valley Authority UHI Upper Head Injection w/o Weight Percent AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 1-1 1.0 Introduction This report describes and provides results from a RLBLOCA analysis for the Sequoyah Unit 1 and Unit 21 Stations. The plants are 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 19 and beyond with AREVA NP's 17x17 HTP fuel design using either BLEU or standard U0 2 fuel and M5 cladding. The analysis was performed in compliance with the U.S. Nuclear Regulatory Commission (NRC) approved RLBLOCA Evaluation Model (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 Units 1 and 2 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 diesel (loss of one train of the pumped ECCS) 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 for fresh and once-burned fuel.

The following are deviations from the approved RLBLOCA EM (Reference 1) that were requested by the NRC are referred to as the "Transition Package." The "Transition Package" is fully described in Section 4.

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 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.

1 The analysis models the Unit 1 steam generator, which is identical to the Unit 2 replacement steam generator, scheduled to be installed in the fall of 2012. If the Unit 2 steam generator is not replaced in the fall 2012, which is the same outage for the HTP fuel transition, this RLBLOCA analysis will need to be dispositioned for the impact of the old Unit 2 steam generators.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 1-2 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 'F 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 guillotine break (DEGB) type is no longer related to break area.

In concurrence with Regulatory Guide 1.157, both the split and the double-ended break will range in area between the minimum break area (Amin) and an area of twice the size of the broken pipe. The determination of break configuration, split versus double-ended, will be made after the break area is selected based on a uniform probability for each occurrence. Amin was calculated to be 33-percent of the DEGB area (see Section 4.6.2 for further discussion). This is not expected to have an effect on PCT results.

In concurrence with the NRC's interpretation of GDC 35, a set of cases was run with a LOOP assumption and a second set with a No-LOOP assumption. The set of 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 3-24. The effect on PCT results is expected to be minor.

During recent RLBLOCA EM modeling studies, it was noted that cold leg condensation efficiency may be under-predicted. Water entering the downcomer (DC) post-accumulator injection remained sufficiently subcooled to absorb DC wall heat release without significant boiling. However, tests (Reference 14) indicate that the steam and water entering the DC from the cold leg, subsequent to the end of accumulator injection, reach near saturation resulting from the condensation efficiency ranging between 80- to 100-percent. To assure that cold leg condensation would not be under-predicted, a RLBLOCA EM update was made. Noting that saturated fluid entering the DC agrees with industry tests, steam and liquid multipliers were developed so as to approximately saturate the cold leg fluid before it enters the DC. The multipliers were developed through scoping studies using a number of plant configurations-Westinghouse-designed 3- and 4-loop plants, and CE-designed plants. The results of the scoping study indicated that multipliers of 10 and 150 for liquid and steam, respectively, were appropriate to produce saturated fluid entering the DC. This RLBLOCA EM departure was recently discussed with the NRC and the NRC agreed that the approach described immediately above was satisfactory in the interim. The modification is implemented post-accumulator injection, 10 seconds after the vapor void fraction in the bottom of the accumulator becomes greater than 90-percent. Thus, the accumulators have injected all their water into the cold legs, and the nitrogen cover gas has entered the system and been mostly discharged through the break before the condensation efficiency is increased by the factors of 10 and 150, for liquid and vapor respectively. Providing saturated fluid conditions at the DC entrance conservatively reduces both the DC driving head and the core flooding rate. Recall that test results indicate AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 1-3 that fluid conditions entering the DC range from saturated to slightly subcooled. Hence, it is conservative to force an approximation of saturated conditions for fluid entering the DC.

The NRC raised the issue concerning fuel thermal conductivity degradation as a function of burnup in Information Notice 2009-23. In order to manage this issue, AREVA Inc. is modifying the way RODEX3A temperatures are compensated in the RLBLOCA Transition Package methodology. In the current process, the RLBLOCA computes PCTs at many different times during an operating cycle. For each specific time in cycle, the fuel conditions are computed using RODEX3A prior to starting the S-RELAP5 portion of the analysis. A steady-state condition for the given time in cycle using S-RELAP5 is established. A base fuel centerline temperature is established in this process. Then two-transformation adjustments to the base fuel centerline temperature are computed. The first transformation is a linear adjustment for an exposure of 10 Mwd/MtU or higher. The second adjustment is performed in the S-RELAP5 initialization process for the transient case. In the new process, a polynomial transformation is used for the first transformation instead of a linear transformation. The rest of the RLBLOCA process for initializing the S-RELAP5 fuel rod temperature should not be altered and the rest of LOCA transient should also continue in the original fashion. Section 6 will provide additional information on the adjustment and adding once-burned fuel to the analysis. Note that these changes are also deviations required by the NRC that are departures from the approved evaluation model.

Recent NRC concerns raised in the form of RAIs are responded to in Section 6.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 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 1941 OF for a fresh U0 2 hot rod (Case 86) with offsite power available (No-LOOP) conditions. From the same case, the PCT for a once-burned U0 2 rod is 1899 IF. Gadolinia-bearing rods of 2, 4, 6 and 8 w/o Gd 20 3 were analyzed for fresh and once-burned fuel 2 in all cases. The limiting PCT for all once-burned rods in Case 86 was in a 6 w/o Gd 20 3 rod (1917 OF). This RLBLOCA result is based on a case set of 933 individual transient cases for offsite power not available (LOOP) and 93 individual transient cases for offsite power available (No-LOOP) conditions. The core is composed only of AREVA NP 17x1 7 fuel and was modeled in a mixed-core configuration due to thermal-hydraulic differences in the Mark BW and HTP fuel designs.

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) of 1.7056 (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 Transition Package methodology (on a forward fit basis) explicitly analyzes fresh and once-burned fuel assemblies to respond to recent NRC RAIs. The second-burned fuel assemblies have minimal power and are typically located on the periphery; therefore would not be limiting in regards to PCT for a RLBLOCA analysis. The AREVA core management design process ensures that reinsert fuel does not have the limiting FAH. The analysis demonstrates that the 10 CFR 50.46(b) criteria listed in Section 3.0 are satisfied.

2 The once-burned GAD rods were not reduced by the once-burned U0 2 peaking reduction (0.9168) shown in Figure 6-4 at the sampled time-in-life, thus producing higher PCT results.

3 AREVA decided to run 93 cases (reporting the 9 2 nd case) for the Sequoyah analysis for three main reasons: 1) current AOR Sequoyah RLBLOCA results, 2) high peaking values analyzed, and 3) the greater pressure drop in the 17x17 HTP fuel w/ integral flow mixers. The allowance to execute more cases is discussed in EMF-2103(P)(A) Section 5.2.1.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 2-2 Table 2-1 Summary of Major Parameters for Limiting Transient Fresh U0 2 Fuel Once-Burned U0 2 Fuel Core Average Burnup (EFPH) 5400 54004 Core Power (MWt) 3479 3479 Total Peaking (FQ) 2.634 2.634 Radial Peak (FAH) 1.7056 1.56375 Axial Offset 0.3182 0.3182 Break Type Split Split Break Size (ft2/side) 3.8673 3.8673 Offsite Power Availability Available Available Decay Heat Multiplier 1.0 1.0 4 The first cycle ends at 13200 EFPH and this time is from the start of 2nd cycle; the assembly total burnup is 18600 EFPH.

5 Calculated using Fresh Fuel FAH x K(Burnup) multiplier (0.9168 at 18600 EFPH) from Figure 6-4.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-1 3.0 Analysis The purpose of the analysis is to verify the adequacy of the ECCS and to demonstrate compliance to the 10CFR 50.46(b) criteria.

• 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 analyze a new fuel design for Sequoyah, the coolable geometry calculation (LOCA-seismic loads) will be revised and testing will validate the 4 th criterion is met. The long-term cooling licensing bases would remain applicable. 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. Section 4 discusses the additional information provided under the "Transition Package" on EMF-2103. Section 5 provides the conclusions, Section 6 addresses recent NRC RAIs on RLBLOCA submittals and Section 7 contains the reference list.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-2 3.1 Descriptionof 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 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 diesel (one ECCS pumped injection train) with fully functional containment sprays. 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 SI 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 cover gas used to pressurize the accumulators exits through the break. This gas release may AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-3 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 Descriptionof 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 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.

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Realistic Large Break LOCA Analysis Rev. 0 Page 3-4 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 operational characteristics 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.

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 'F (or less) to above 2,200 'F. 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.

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Realistic Large Break LOCA Analysis Rev. 0 Page 3-5 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.

3.3 Plant Descriptionand 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 6 . The RCS also includes one pressurizer connected to a hot leg. The core contains (193) 17x17 AREVA fuel assemblies. The ECCS includes one charging and one accumulator/Sl/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.

6 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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Realistic Large Break LOCA Analysis Rev. 0 Page 3-6 The S-RELAP5 model explicitly describes the RCS, RV, pressurizer, and accumulator lines.

The charging injection flows are connected to the RCS, and the SI 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 Units 1 and 2 Stations. 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 Units 1 and 2 Stations. Both entities have ongoing processes that assure the ranges and values of input parameters for the Sequoyah Units 1 and 2 Stations 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 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 6-2 (Reference 5) was used in setting the remainder of the containment model input parameters. As noted in Table 3-3, containment temperature is a sampled parameter. Containment pressure response is indirectly ranged by sampling the 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 AREVA NP Inc.

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Realistic Large Break LOCA Analysis Rev. 0 Page 3-7 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 containment 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. 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 Units (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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Realistic Large Break LOCA Analysis Rev. 0 Page 3-8 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.

Blowdown Quench (SER Restriction #7)

Item: The model is valid as long as blowdown quench does not occur. If blowdown quench occurs, additionaljustification for the blowdown heat transfermodel and uncertainty are needed or the run corrected.A blowdown quench is characterizedby a temperature reduction of the PCT node to saturationtemperature during the blowdown period.

Treatment: Examine plots of PCT vs. Time at the PCT node for all cases for evidence of blowdown quench using the RLBLOCA definition of blowdown (i.e., accumulatordischarge).

The highest clad surface temperatures independent of elevation and also at the PCT node are calculated as a function of time for all cases. These plots were examined and compared to the end of blowdown (EOB) to verify that no quench of the PCT node occurred prior to EOB. As a result of this inspection, several cases exhibit lingering cladding temperatures near saturation temperature during blowdown and were flagged for further evaluation. Six cases were "screened" for further evaluation regarding possible quench prior to EOB. The hot pin, PCT axial heat structure node, and the corresponding core fluid node for each case are identified.

These parameters form the basis for illustrating the clad temperature, fluid (saturation) temperature, and accumulator injection transient at the PCT location for the screened cases.

Figure 3-25 through Figure 3-30 indicates that, of the six screened cases, evidence of blowdown quench is apparent in three of the cases, Cases 32, 35, and 68 (Note that these cases represent three of the lowest 4 cases sorted in order of decreasing PCT). Clad temperature superheat equivalents achieved for these cases range from about 200 to 450 OF, much lower than the approximately 900 °F blowdown superheat resulting from the limiting case, Case 86. As a result, the hot pins are more easily quenched by a minimal intrusion of liquid into the hot channel.

In comparison to the limiting case, the linear heat generation rates (LHGRs) and break areas are quite small and this explains why the blowdown clad temperature response is minimal.

Given the combined low hot pin power and the small break area, with its comparatively low blowdown rate and slower transient, Case 32, 35, and 68 could not achieve PCTs that are limiting or even high in the order of cases ranked according to PCT.

Effectively, there is no blowdown quench and the cases of interest regarding blowdown quench have no impact. Therefore, compliance to the SER restriction has been demonstrated.

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Realistic Large Break LOCA Analysis Rev. 0 Page 3-9 Bottom-Up Quench Item: The reflood model applies to bottom-up quench behavior.If a top-down quench occurs, the model is to be justified or corrected to remove top down quench. A top-down quench is characterizedby the quench front moving from the top to the bottom of the hot assembly.

Treatment: The reflood heat transfermodel in S-RELAP5 and the prescribed upperplenum nodalization have been developed so that the peak clad temperature location quenches from the consequences of a bottom-up quench only. No additions to the calculation notebook or Design Report are required.

Several provisions have been implemented in the S-RELAP5 model to prevent top-down quench. These measures are:

• The CCFL model is applied on all core exit junctions.

" The reverse form loss at the hot channel and central core exit is increased by a factor of 1000 at the beginning of core reflood.

To confirm that top-down quench does not occur in the analysis, the liquid flow at the exit of the hot channel was plotted. Junction 065 is the hot channel exit junction designation for the exit.

Because the liquid flow rate is not particularly steady, it was integrated to smooth the result.

Cases can exhibit periodic positive integrated flow (indicating flow in an upward direction),

constant integrated flow (no flow), or negative integrated flow (flow moving from the top-down).

Core flows can reverse at break initiation, during blowdown, and reverse flow can continue until the core begins to be recovered by accumulator and ECCS injection. Top-down quench, then, is judged as a potential problem ifthe hot channel exit flow reverses after the beginning of core recovery.

Using the beginning of core recovery (BOCR) for each case analyzed and examining the integrated hot channel exit flows, it is apparent that, for the most part, core exit post-BOCR liquid flows are positive, ruling out the potential for top-down quench. There are some cases, however, that exhibit temporary flow reversals. These cases are screened out and after close examination of all of the cases; it is shown that core exit reverse flow is not significant after BOCR and any top-down quench effects are not sufficient to have an effect on PCT predictions resulting from the analysis. Therefore, compliance to the SER restriction has been demonstrated.

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Realistic Large Break LOCA Analysis Rev. 0 Page 3-10 3.5 Realistic Large Break LOCA Results Two case sets of 93 transient calculations were performed sampling the parameters listed in Table 3-1 and the results of the 9 2 nd case are reported. 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 (1941 OF) occurred in Case 86 for a fresh U0 2 rod. The major parameters for the limiting transient are presented in Table 2-1. The once-burned rod limiting PCT (1917 OF) occurred in Case 86 for a 6 w/o GAD rod 7 . 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 72, which corresponded to the median case out of the 93-case set with no offsite power available. The nominal PCT was 1484 OF. This result can be used to quantify the relative conservatism in the limiting case result. In this analysis, it was 457 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 3-23. Figure 3-7 shows linear scatter plots of the key parameters sampled for the 93 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 93 calculations, respectively. Figures 3-10 and 3-11 show the maximum oxidation and total oxidation versus PCT scatter plots for the 93 calculations, respectively. Key parameters for the limiting PCT case are shown in Figures 3-12 through 3-23. 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 3-24.

7 The once-burned GAD rods were not reduced by the once-burned U0 2 peaking reduction (0.9168) shown in Figure 6-4 at the sampled time-in-life (total = 18600 EFPH), thus producing higher PCT results. The burned U0 2 rod PCT for Case 86 was 1899 OF.

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Realistic Large Break LOCA Analysis Rev. 0 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) 8 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 Plant 9 Offsite power availability 1 ° 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) 8 Not sampled in analysis, multiplier set to 1.0.

9 Uncertainties for plant parameters are based on typical plant-specific data.

10 Not sampled, see Section 4.9.

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Realistic Large Break LOCA Analysis Rev. 0 Paqe 3-12 Table 3-2 Plant Operating Range Supported by the LOCA Analysis Event Operating Range 1.0 Plant Physical Description 1.1 Fuel a) Cladding outside diameter 0.374 in.

b) Cladding inside diameter 0.326 in.

c) Cladding thickness 0.024 in.

d) Pellet outside diameter 0.3195 in.

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

g) Resinter densification I I h) Gd 20 3 concentrations 2, 4, 6, 8 w/o 1.2 RCS a) Flow resistance Analysis b) Pressurizer location Analysis assumes location giving most limiting PCT (broken loop) c) Hot assembly location Anywhere in core d) Hot assembly type 17x17 e) SG tube plugging < 15-percent 2.0 Plant Initial Operating Conditions 2.1 Reactor Power a) Nominal reactor power 3479 MWt11 b) FQ *2.6512 c) FAH *1.705613 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 °F _ T _ 583 OF c) Upper head temperature -Tcold Temperature 14 11 Includes uncertainties 12 Ensures that a minimum 7-percent peaking margin is maintained to the FQ limits when operating at the positive or negative AFD limit 13 Includes 4-percent measurement uncertainty 14 Upper head temperature will change based on sampling of RCS temperature AREVA NP Inc.

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Realistic Large Break LOCA Analysis Rev. 0 Paae 3-13 Table 3-2 Plant Operating Range Supported by the LOCA Analysis (Continued) d) Pressurizer pressure 1859.7 psia _ P *2459.7 psia e) Pressurizer level 57-percent _<L _<95-percent f) Accumulator pressure 614.7 psia _ P _ 697.7 psia g) Accumulator liquid volume 1004.6 ft 3 -<V _ 1095.4 ft3 h) Accumulator temperature 95 °F _< T _ 130 IF (coupled to containment lower volume temperature) i) Accumulator fLJD As-built piping configuration j) Minimum ECCS boron > 2400 ppm 3.0 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 °F i) Charging pump delay 37 s (w/ offsite power) 37 s (w/o offsite power) j) SI pump delay 37 s (w/ offsite power) 37 s (w/o offsite power) k) RHR pump delay 37 s (w/ offsite power) 37 s (w/o offsite power)

I) Containment pressure 14.3 psia, nominal value m) Containment upper compartment 80 OF< T < 110 IF temperature n)Containment lower compartment 95 °F

• T < 130 OF temperature

0) Containment sprays delay 8 s p) Containment spray water 5 OF temperature I AREVA NP Inc.

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Realistic Large Break LOCA Analysis Rev. 0 Page 3-14 Table 3-3 Statistical Distributions Used for Process Parameters 1 5 Operational Measurement Standard Parameter Uncertainty Parameter Range Uncertainty16 Distribution Distribution Deviation Pressurizer Pressure (psia) Uniform 1859.7 - 2459.7 N/A N/A Pressurizer Liquid Level (percent) Uniform 57 - 95 N/A N/A 3

Accumulator Liquid Volume (ft ) Uniform 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 (°F)

Containment Upper Volume ( ft3 ) 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 17 Binary 0,1 N/A N/A Delay for Containment Cooling (s) Point 8.0 N/A N/A Charging Pump Delay (s) Point 37 (w/ offsite power) N/A N/A 37 (w/o offsite power)

N/A N/A LHSI Pump Delay (s) Point 37 37 (w/ offsitepower)

(w/ooffsite power)

RHR Pump Delay (s) Point 37 (w/ offsite power) N/A N/A 37 (w/o offsite power) 15 Note that core power is not sampled, see Section 1.0 16 All measurement uncertainties were incorporated into the operational ranges 17 This is no longer a sampled parameter. One set of 93 cases is run with LOOP and one set of 93 cases is run with No-LOOP.

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Realistic Large Break LOCA Analysis Rev. 0 Page 3-15 Table 3-4 SER Conditions and Limitations SER Conditions and Limitations Response

1. A CCFL violation warning will be added to alert the analyst There was no significant occurrence of CCFL 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 Units 1 and 2 (15.08 kW/ft) are using a higher planar linear heat generation rate (PLHGR) lower than that used in the development of the RLBLOCA than used in the current analysis, or if the methodology is EM (Reference 1). An end-of-life calculation was not to be applied to an end-of-life analysis for which the pin performed; thus, the need for a blowdown cladding rupture pressure is significantly higher, then the need for a model was not reevaluated1 8 .

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 and in response to RAI #25 reflood and the core to uncover again. These break for EMF-2103(P)(A) Rev. 0.

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 Units 1 and 2 are Westinghouse 4-loop plants.

CE-designed nuclear steam systems.

6. The model applies to bottom reflood plants only (cold side Sequoyah Units 1 and 2 are bottom reflood plants.

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 six cases; these cases are discussed in Section 3.4.

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

8. The reflood model applies to bottom-up quench behavior. 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.

18 In response to recent NRC RAI questions, a blowdown rupture check was performed and the results reported in Section 6. There were no cases that exhibited blowdown rupture.

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Realistic Large Break LOCA Analysis Rev. 0 Page 3-16 Table 3-4 SER Conditions and Limitations (Continued)

SER Conditions and Limitations 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 reference plant 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 are slightly differently than the sample problem;

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

5) There are no standpipes present in the Sequoyah Units 1 and 2 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 Units 1 and 2 are 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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Realistic Large Break LOCA Analysis Rev. 0 Paae 3-17 Table 3-4 SER Conditions and Limitations (Continued)

SER Conditions and Limitations 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 Units 1 and 2 plants have previously been operating with AREVA NP realistic large break loss-of- M5 clad 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 #86 Fresh Fuel Once-Burned (offsite power available) U0 2 Rod Fuel 6 w/o GAD' 9 PCT Temperature 1941 OF 1917 OF Time 265.9s 265.9s Elevation 10.043 ft 10.043 ft Metal-Water Reaction Pre-transient Oxidation (%) 0.7624 1.3144 Transient Oxidation Maximum (%) 2.9906 2.5376 Total Oxidation Maximum (%) 3.7530 3.8520 Total Whole-Core Oxidation (%) 0.0982 N/A 19 The once-burned GAD rods were not reduced by the once-burned U0 2 peaking reduction (0.9168) shown in Figure 6-4 at the sampled time-in-life (total = 18600 EFPH), thus producing higher PCT results. The burned U0 2 rod PCT for Case 86 was 1899 OF.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-18 Table 3-6 Calculated Event Times for the Limiting PCT Case Event Time (s)

Break Opened 0.0 RCP Trip N/A SIAS Issued 0.0 Start of Broken Loop Accumulator Injection 9.5 Start of Intact Loop Accumulator Injection (Loops 2, 3 and 4 respectively) 12.1, 12.1, 12.1 Start of Charging 37.0 SI/RHR Available 37.0 Broken Loop SI/RHR Delivery Began 37.0 Intact Loop SI/RHR Delivery Began (Loops 2, 3 and 4 37.0, 37.0, 37.0 respectively)

Beginning of Core Recovery (Beginning of Reflood) 48.2 PCT Occurred 265.9 Broken Loop Accumulator Emptied 82.6 Intact Loop Accumulators Emptied 83.5, 83.8, 83.3 (Loops 2, 3 and 4 respectively)

Transient Calculation Terminated 839.7 AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 3-19 20 Table 3-7 Heat Transfer Parameters for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paqe 3-20 Table 3-8 Containment Initial and Boundary Conditions Containment Net Free Volume Volume (ftz)

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 10 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 4

Air Return 2

Fan Post-LOCA initiation at 600 s Total Flow = 120,000 cfm 24 Due to the relatively late start of the recirculation fan, it is not modeled in this analysis.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-21 Table 3-9 Passive Heat Sinks in Containment Area Thickness Inside Tikes Hih Heat Sink Areaft Thickness Radius Thicknessft Heighft Material Left Side Right Side ft 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 UppeanduLoweStructure between C ptmen 13,081 2.34 concrete Upper Comp.

Upper and Lower Compartment Lower Comp.

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-°F) (BTU/ft3-°F)

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

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-22 Figure 3-1 Primary System Noding AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Pace 3-23 Figure 3-2 Secondary System Noding AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-24 Figure 3-3 Reactor Vessel Noding AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-25 Figure 3-4 Core Noding Detail AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-26 Figure 3-5 Upper Plenum Noding Detail AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-27 Figure 3-6 Containment Noding Diagram AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-28 One-Sided Break Area 2

(ft /side) 1.0 2.0 3.0 4.0 5 .0 Burn '

Time * ** 0 * ** see..... 5 (hours)

F I 0.0 5000.0 10000.0 15000.0 Core L I I 1 Power *

(MW) 3478.0 3478.5 3479.0 3479.5 3480.0 3480.5 3481.0 Fq m -II .

Peaking 1.5 1.7 1.9 2.1 2.3 2.5 2.7 2.9 AO 0 .. . imuim

-0.4 -0.2 0.0 0.2 0.4 Pressurizer '

Pressure m o l*

• m (psia) 1800.0 2000.0 2200.0 2400.0 2600.0 Pressurizer Liquid Level I Ni

(%)

50.0 60.0 70.0 80.0 90.0 100.0 RCS (Tavg) '

Temperature mine iO*i imm

('F) 578.0 579.0 580.0 581.0 582.0 583.0 Figure 3-7 Scatter Plot of Operational Parameters AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-29 Total Loop Flow (Mlb/hr) 130.0 140.0 150.0 160'.0 Accumulator Liquid Volume 3

(Wt )

10'00.0 1020.0 1040.0 1060.0 1080.0 11010.0 Accumulator Pressure (psia) 600.0 620.0 640.0 660.0 680.0 700I.0 Upper Compartment L Containment Volume NN ami......... - I u (if3) F 6.50e+05 6.60e+05 6.70e+05 6.80e+05 6.90e+05 7.00e+05 Upper Compartment L *mm anmm m moe.m Containment Temperature (0 F) -

100.0 80.0 90.0 100.0 110.0 Lower Compartment (Accumulator)

Containment Temperature

('F) --

F 90.0 uM 0Mm meuum 100.0 110.0 ommm 120.0 I

130.0 Figure 3-7 Scatter Plot of Operational Parameters (Continued)

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-30 PCT vs Time of PCT 2200 2000 E]

1800 1600 L-1400 toL C-I-- []

0~

1200 EP 1000 800 ] Split Break El Guillotine Break 600 400 0 100 200 300 400 500 Time of PCT (s)

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

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-31 PCT vs One-sided Break Area 2200 2000 1800

~E 0 EE Elm1

• mm 1600

• E* .N [:] E M m m El I _m0 0 ] m A 1400 E-I- []m 0

10_

1200

• E*

1000 EJ 800 600 MSplit Break El Guillotine Break 400 1.0 2.0 3.0 4.0 5.0 Break Area (ft2/side)

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

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-32 Maximum Oxidation vs PCT 4.0 3.8 0 Split Break 3.6 E[ Guillotine Break 3.4 3.2 3.0 2.8 2.6 2.4 Z.-2.2 F-22.0 1.8 0

1.6 1.4 1.2 [L 1.0 I.]ll oa]

0.8 0.6 0.4 0.2 0.0 400 600 800 1000 1200 1400 1600 1800 2000 2200 PCT (°F)

Figure 3-10 Maximum Transient Oxidation versus PCT Scatter Plot AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 3-33 Total Oxidation vs PCT 0.20 0.18 0.16 0.14 0.12 F

. 0.10 Cu

-0 0.08 0.06 0.04 0.02 0.00 -

.... ' -I 400 600 800 1000 1200 1400 1600 1800 2000 2200 PCT (°F)

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

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-34 PCT Trace for Case #86 PCT = 1940.6 OF, at Time = 265.91 s, on Hot U02 Rod 2000 1500 C)

E 1oo 0

CL 500 0

0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2011 13:44:12 R5DMX Figure 3-12 Peak Cladding Temperature (Independent of Elevation) for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-35 Break Flow 80 60 40 E

a) 0 20 0

-20 0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 3-13 Break Flow for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paqe 3-36 Core Inlet Mass Flux 1000 500 E

9 X 0 C,)

C,

-500

-1000 0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2011 13:44:12 R5DMX Figure 3-14 Core Inlet Mass Flux for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paqe 3-37 Core Outlet Mass Flux 1000 500 E

U-Cu 0

-500 0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2011 13:44:12 R5DMX Figure 3-15 Core Outlet Mass Flux for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-38 Pump Void Fraction 1.0 0.8 0.6 C

0 U-

~0 0.4 Broken Loop 1 Intact Loop 2 Intact Loop 3 0.2 Intact Loop 4 0.0 0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2011 13:44:12 R5DMX Figure 3-16 Void Fraction at RCS Pumps for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-39 ECCS Flows 1500

-- Loop 1 (broken)

---

Loop 2 Loop 3 Loop 4 1000 cn E

.0 Cu 0

500 0

0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2011 13:44:12 R5DMX Figure 3-17 ECCS Flows (Includes Accumulator, Charging, SI and RHR) for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paqe 3-40 Upper Plenum Pressure 3000 2000 clu

3 C',

(n, 1000 0

0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 3-18 Upper Plenum Pressure for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paqe 3-41 Downcomer Liquid Level 30 20 2,--

-j 10 0

0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2011 13:44:12 R5DMX Figure 3-19 Collapsed Liquid Level in the Downcomer for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-42 Lower Vessel Liquid Level 15 10

-I 5

0-0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 3-20 Collapsed Liquid Level in the Lower Plenum for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-43 Core Liquid Level 15 10 2

ci

-J 5

0 0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 3-21 Collapsed Liquid Level in the Core for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-44 Containment and Loop Pressures 100 90 80 70 60 50 Sa-40 30 20 10 0

0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2011 13:44:12 R5DMX Figure 3-22 Containment and Loop Pressures for the Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 3-45 300000 250000 200000 150000 CD 100000 50000 0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 3-23 Reactor Vessel Liquid Mass (Ibm) versus Time (sec)

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 3-46 2200 2200

-LOOP 0 No LOOP 2000 2000 0

0 1800 - 1800 E

0 *

• 0 o0 A 0 1600 = V 1600 o-o 0 03 0 *0 # uu

• 0 0.0 * [] 4 31 13

• 0 n 00 Np~ 0 0

.03 0 (3 1400 t- 1400 4-c0 41 0 VC3 0 0 0 0

[]

1200 T O3

• i1200 0] 0 0

0 0 1000 0 1000 0

800 800 0

600 , - -..... - _____.- - __ r - 600 0 10 20 30 40 50 60 70 80 90 Case Number Figure 3-24 GDC 35 LOOP versus No-LOOP Cases AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 3-47 Case 29 TD Quench Screen - Limiting Rod PCT Node Temperature

- PCT Node Saturation Temp 1000 k

• Accumulator Injection flow-mflowli27 IL (D

CL 0~

a) 500 0) a)

E0 0

0 10 20 30 40 ID:5014215Mar201120:23W02 R5DMX Time (s)

Figure 3-25: PCT Node Cladding Surface Temperature and Saturation Temperature, Case 29 Case 32 TD Quench Screen - Limiting Rod

- PCT Node Temperature

-- PCT Node Saturation Temp 6--

. Accumulator Injection flow-mftowj927 1000 CL a) 0~

a)

'a 500 V---

(I) 0 0.

10 20 30 40 ID-3024515Mar201120-16-47RSDMX Time (s)

Figure 3-26: PCT Node Cladding Surface Temperature and Saturation Temperature, Case 32 AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 3-48 Case 35 TD Quench Screen - Limiting Rod

- PCT Node Temperature

-- PCT Node Saturation Temp 1000 k + Accumulator Injection flow-mflowi927 U-C)

E 0)

U) 500 CO 0) a.

0.

0 10 20 30 40 ID:1537915M.r201120:59:37R5DMX Time (s)

Figure 3-27: PCT Node Cladding Surface Temperature and Saturation Temperature, Case 35 Case 42 TD Quench Screen - Limiting Rod

- PCT Node Temperature a PCT Node Saturation Temp 1000 F

• Accumulator Iniection flow-mftowi927 0.

E 500 cc a) a-0 .1 30 0 10 20 40 ID25750 15Mar201122:47:52R5DMX Time (s)

Figure 3-28: PCT Node Cladding Surface Temperature and Saturation Temperature, Case 42 AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paqe 3-49 Case 68 TD Quench Screen - Limiting Rod PCT Node Temperature

- -- a PCT Node Saturation Temp 1000 F . Accumulator Injection flow-mflowj927 Co E

L) 500 F 0)

C-)

CD a-)

0 .

0 10 20 30 40 ID:0927116Mar201104 03 38 R5DMX Time (s)

Figure 3-29: PCT Node Cladding Surface Temperature and Saturation Temperature, Case 68 Case 87 TD Quench Screen - Limiting Rod

--- PCT Node Temperature

• PCT Node Saturation Temp 9- 1000 F Accumulator Injection flow-mflowj927 to (D

E I

-)

0) 500

'D 0

z to 0) a.

0 10 20 30 40 ID:5087616Mar201106:49M32R50MX Time (s)

Figure 3-30: PCT Node Cladding Surface Temperature and Saturation Temperature, Case 87 AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-1 4.0 Generic Support for Transition Package The following sections are responses to typical RAI questions posed by the NRC on EMF-2103 Revision 0 plant applications, these responses and changes are known as the "Transition Package." In some instances, these requests cross-reference 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.

4.1 Reactor Power Question: Reactor Power - Table 3-2, Item 2.1, and its associatedFootnote I indicate that the assumed reactor core power "includesuncertainties." The use of a reactorpower assumption other than 102-percent, regardlessof BE or Appendix K methodology, is permitted by Title 10 of the Code of FederalRegulations (10 CFR), Part 50, Appendix K.I.A, "Requiredand Acceptable Features of The Evaluation Models, 'Sources of Heat During a LOCA." However, Appendix K.I.A also states: "... An assumed power level lower than the level specified in this paragraph

[1.02 times the licensed power level], (but not less than the licensed power level) may be used provided..."

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. Not sampling the power level is a change to the approved RLBLOCA EM (Reference 1).

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Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-2 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.

4.2 Rod Quench Question: Does the version of S-RELAP5 used to perform the computer runs assure that the void fraction is less than 95-percent and the fuel cladding temperature is less than 900 'F before it allows rod quench?

Response: Yes, the version of S-RELAP5 employed for the Sequoyah 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 (Tmjn) 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 case set Tmin was never sampled above 703.4 K (806.5 OF). This is a change to the approved RLBLOCA EM (Reference 1).

4.3 Rod-to-Rod Thermal Radiation Question: Providejustification that the S-RELAP5 rod-to-rod thermal radiationmodel applies to the SQN 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 *F - 2000 °F) and high (2000 'F to over 2200 'F) peak cladding temperatures and shown to be conservative when applied nominally. The normal distribution of the experimental data was then determined. During the execution of an RLBLOCA evaluation, the heat transferred from a fuel rod is determined by the application of a multiplier to the nominal heat transfer model. This multiplier is determined by a random sampling of the normal distribution of the experimental data benchmarked. Because the data include the effects of rod-to-rod 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 The Full-Length Emergency Core Heat Transfer Separate Effects and Systems Effects Tests (FLECHT-SEASET) evaluated covered a range of PCTs from 1,651 to 2,239 'F and the THTF tests covered a range of PCTs from 1,000 to 2,200 *F. Since the test bundle in either FLECHT-SEASET or Thermal Hydraulic Test Facility (THTF) is surrounded by a test vessel, AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-3 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),

Loss of Fluid Test (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 (Reference 1). 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 'F. 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.

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.

4.3.1 Assessment of Rod-to-Rod Radiation Implicit in the RLBLOCA Methodology As discussed above, the FLECHT-SEASET and THTF tests were selected to assess and determine the S-RELAP5 code heat transfer bias and uncertainty. A 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 AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-4 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 4-1 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.

Table 4-1 Typical Measurement Uncertainties and Local Peaking Factors FAH Measurement Plant Unceait(erent Uncertainty (percent) Local Pin Peaking Factor 1 4.0 1.068 2 4.0 1.050 3 6.0 1.149 4 4.0 1.113 5 4.25 1.135 6 4.0 1.058 4.3.2 Quantification of the Impact of Thermal Radiation using R2RRAD Code The R2RRAD radiative heat transfer model was developed by Los Alamos National Laboratory (LANL) to be incorporated in the BWR version of the TRAC code. The theoretical basis for this code is given in References 8 and 11 and is similar to that developed in the HUXY rod heatup code (Reference 10, Section 2.1.2) used by AREVA for BWR LOCA applications. The version of R2RRAD used herein was obtained from the NRC to examine the rod-to-rod radiation characteristics of a 5x5 rod segment of the 161 rod FLECHT-SEASET bundle. The output provided by the R2RRAD code includes an estimate of the net radiation heat transfer from each rod in the defined array. The code allows the input of different temperatures for each rod as well as for a boundary surrounding the pin array. No geometry differences between pin locations are allowed. Even though this limitation affects the view factor calculations for guide tubes, R2RRAD is a reasonable tool to estimate rod-to-rod radiation heat transfer.

The FLECHT-SEASET test series was intended to simulate a 17x17 fuel assembly and there is a close similarity, Table 4-2, between the test bundle and a modern 17x17 assembly.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 4-5 Table 4-2 FLECHT-SEASET & 17x17 FA Geometry Parameters Design Parameter FLECHT-SEASET 17x17 Fuel Assembly Rod Pitch (in) 0.496 0.496 Fuel Rod Diameter (in) 0.374 0.374 Guide Tube Diameter (in) 0.474 0.482 Five FLECHT-SEASET tests (Reference 6) were selected for evaluation and comparison with expected plant behavior. Table 4-3 characterizes the results of each test. The 5x5 selected rod array comprises the hot rod, 4 guide tubes and 20 near adjacent rods. The simulated hot rod is rod 7J in the tests.

0 0 0 00 Guide Tube 0-0 0 0 Hot Rod 0 0 Adjacent Rods 0 0 00' 0 0 00 0 Figure 4-1 R2RRAD 5x5 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 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 AREVA NP Inc.

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Realistic Large Break LOCA Analysis Rev. 0 Paqe 4-6 experimental result and the remaining rods conservatively set to temperatures expected within the bundle. Because peak rod powers frequently occur at fuel assembly corners away from either guide tubes or instrument tubes and for added conservatism, the guide tubes (thimble tubes) were replaced by fuel rods in the input model described above. In line with the discussion in Section 4.3.1, the surrounding 24 rods were set to a temperature estimated for rods of 4-percent lower power. The boundary temperature was estimated based an average power 6-percent below the hot rod power. For both of these, the temperature estimates were achieved using a ratio of pin power to the difference in temperature between the saturation temperature and the PCT.

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

Tsat was taken as 270 'F.

Figure 4-2 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 'F, the hot rod thermal radiation in the plant cases exceeds that of the same component within the experiments.

Table 4-3 FLECHT-SEASET Test Parameters HTC at PCT Steam Thimble Test Time 2(Btu/hr- Temperature at 71 Temperature at 6-ft (OF) Time (s) ft OF) (6-if) (OF) 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 I I AREVA NP Inc.

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Realistic Large Break LOCA Analysis Rev. 0 Paae 4-7 4.5 C

0 CM I-i-

0 4-(U 0 +i + 1-1400 1500 1600 1700 1800 1900 2000 2100 2200 2300 2400 PCT (TF)

Figure 4-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 FA 4.3.3 Rod-to-Rod Radiation Summary In summary, the conservatism of 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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Realistic Large Break LOCA Analysis Rev. 0 Page 4-8 4.4 Film Boiling Heat TransferLimit Question: In the SQN 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 Units 1 and 2 RLBLOCA analysis limits the contribution of the Forslund-Rohsenow model to no more than 15-percent of the total heat transfer at and above a void fraction of 0.9. Because the limit is applied at a void fraction of 0.9, the contribution of Forslund-Rohsenow within the 0.7 to 0.9 interpolation range is limited to 15 -percent or less. This is a change to the approved RLBLOCA EM (Reference 1).

4.5 Downcomer Boiling Question: If the PCT is greaterthan 1800°F or the containment pressure is less than 30 psia, has the Sequoyah Units 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 the Sequoyah Units 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 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).

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Realistic Large Break LOCA Analysis Rev. 0 Paqe 4-9 M~b ,dc Figure 4-3 Reactor Vessel Downcomer Boiling Diagram Hot downcomer walls penalize PCT by two mechanisms: by reducing subcooling of coolant entering the core and through the reduction in downcomer hydraulic head which is the driving force for core reflood. Although boiling in the downcomer occurs during blowdown, the biggest potential for impact on clad temperatures is during late reflood following the end of accumulator injection. At this time, there is a large step reduction in coolant flow from the ECC systems. As a result, coolant entering the downcomer may be less subcooled. When the downcomer coolant approaches saturation, boiling on the walls initiates, reducing the downcomer hydraulic static level.

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.

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Realistic Large Break LOCA Analysis Rev. 0 Page 4-10 While downcomer boiling may impact clad temperatures, it is somewhat of a self-limiting process. If cladding temperatures increase, less energy is transferred in the core boiling process and the loop steam flows are reduced. This reduces the required driving head to support continued core reflood and reduces the steam available to heat the ECCS water within the cold legs rdsufting 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.

4.5.1 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.

4.5.1.1 Exact Solution In this benchmark, the downcomer wall is considered as a semi-infinite plate. Because the benchmark uses a closed form solution to verify the wall 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 AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-11 (T(x,t) - To) / (T, - To) = erf {x / (2.(a t)° 5)}, (1) where, a is the thermal diffusivity of the material given by a = k/(p Cp),

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

The conditions of the benchmark are 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 4-4 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 Units 1 and 2 HTP Fuel Rev. 0 Realistic Larqe Break LOCA Analysis Page 4-12 550 500 uL 450

.4)

O 400 E

I--

4) 350 300 250 4-0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 Distance from Inner Wall, feet Figure 4-4 S-RELAP5 versus Closed Form Solution AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Large Break LOCA Analysis Page 4-13 4.5.1.2 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.

• VSL Wall (9-msh)

-Bawe 18-Mesh VSL Wall ._S 2400 0W 18N00 -_)__

2 1200000 2)

Cu Cu 6000,00 -4.

wý 0001'0.0 _

80.0

_ ~ . _

160.0

_o 240 0 320.0 400,0 Time (sec)

Figure 4-5 Downcomer Wall Heat Release - Wall Mesh Point Sensitivity AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Larme Break LOCA Analysis Page 4-14 240000

--- oBase VSý Wall (9-mesh) l.-Mesh _SL Wall 1800¸00 2 IL 0 a S

(-

M O0 E

60O~ ___

0.00O00 80,0 ¶60 240.0 3200 400 0 Time (sec)

Figure 4-6 PCT Independent of Elevation - Wall Mesh Point Sensitivity AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Large Break LOCA Analysis Page 4-15 30.00

-. Base VSL Wall (9-meshl I.a18- Mesh VS L Wail I a)

I4-20.00 a)

-. 1 0o-0,0 80.0 160.0 2400 320b0 400.0 Time (sec)

Figure 4-7 Downcomer Liquid Level - Wall Mesh Point Sensitivity AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Larae Break LOCA Analysis Page 4-16 12.00

- fVI~l a Base VSL Wall (9-mesh) 10,00 8.00

_j L7-

_j 4.00 r _I-2.00 0.0 0160,0 240.0 320.0 400.0 Time (sec)

Figure 4-8 Core Liquid Level - Wall Mesh Point Sensitivity AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Large Break LOCA Analysis Page 4-17 4.5.2 Downcomer Fluid Distribution To justify the adequacy of the downcomer nodalization in calculating the fluid distribution in the downcomer, two studies varying separately the axial and the azimuthal resolution with which the downcomer is modeled have been conducted.

4.5.2.1 Azimuthal Nodalization In a letter to the NRC 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 4-9). 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.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Large Break LOCA Analysis Page 4-18 Base model Revised 9 Region Model

- I_ I1I 1IfII I[I CHD0 CHiý Figure 4-9 Azimuthal Noding 4.5.2.2 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 ice condenser containment and a 12 foot core. For the study, the two nodes spanning the active core height are divided in half, revising the model to include eight axial nodes. Further, the refined noding is located within the potential boiling region of the downcomer where, if there is an axial resolution influence, the sensitivity to that impact would be greatest.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Larqe Break LOCA Analysis Page 4-19 r

The results show that the axial noding used in the base methodology is sufficient for plants experiencing the very low system pressure characteristics of ice condenser containments.

Figure 4-10 provides the containment back pressure for the base modeling. Figures 4-11 through 4-14 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 0 0 - . .. . ... .. .. __..

-- I Base 6x6 Cas 3200 24.00 I-U) 9, *--*O-- .-

• 9 9 * -- -- -

CL , ....-

8.00 0.00 t 00 80.0 1600 -. 240.0 320.0 400.0 Time (sec)

Figure 4-10 Lower Compartment Pressure versus Time AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Larqe Break LOCA Analysis Page 4-20

,30&000

- Base 6x6 ase::

U) ..

... 8x6 Caselc C', 2400000 o$

Cl, Cu U) 180O000 A

0:: -U!

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C ii I w 12000.00 8M0000 0.00 80.0 0ý0 1600 -'2400 3200 4000 Time (sec)

Figure 4-11 Downcomer Wall Heat Release - Axial Noding Sensitivity Study AREVA NP Inc.

I

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Larme Break LOCA Analysis Page 4-21 240D.00

[-0Bas 6.6 se

• xCa 8Kr-iax~cE__ 0, U-0 Q) 1200.00 a.

E woIw - _ _

.00 - __ ___

0.0 80.0 180.0 240.0 3200 400.0 Time (sec)

Figure 4-12 PCT Independent of Elevation - Axial Noding Sensitivity Study AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Large Break LOCA Analysis Page 4-22 30.00 20.00 a)

I

'4-

,-J J.

7J 10.00 * - '*

800 160.0 2400 3200 4000 Time (sec)

Figure 4-13 Downcomer Liquid Level -Axial Noding Sensitivity Study AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Large Break LOCA Analysis Page 4-23 12.00 -- _

--- Base 6x6 b~e 10.00 1- ...... ______u .'

8.001

.J 6.00

_1 II 4.00 200 -I-

,~. !~ V 0=0. 0 8C00 2400 320.0 400.0 Time (sec)

Figure 4-14 Core Liquid Level - Axial Noding Sensitivity Study 4.5.3 Downcomer Boilinq Conclusions To further justify the ability of the RLBLOCA methodology to predict the potential for and impact of downcomer boiling, studies were performed on the downcomer wall heat release modeling within the methodology and on the ability of S-RELAP5 to predict the migration of steam through the downcomer. Both azimuthal and axial noding sensitivity studies were performed. The axial noding study was based on an ice condenser plant that is near atmospheric pressure during reflood. These studies demonstrate that S-RELAP5 delivers energy to the downcomer liquid volumes at an appropriate rate and that the downcomer noding detail is sufficient to track the distribution of any steam formed. Thus, the required methodology for the prediction of downcomer boiling at system pressures approximating those achieved in plants with pressures as low as ice condenser containments has been demonstrated.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Large Break LOCA Analysis Page 4-24 4.6 Break Size Question: Were all break sizes assumed greaterthan or equal to 1.0 ft2 ?

Response: Yes.

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

4.6.1 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 coolant system (RCS) during which the three historical approximate states of the system can be identified.

Blowdown The blowdown phase is defined as the time period from initiation of the break until flow from the 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 refilling of the core with water and ends with the achievement of complete core quench.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Large Break LOCA Analysis Page 4-25 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.

4.6.2 New Minimum Break Size Determination No determination of the lower limit can be exact. The values of critical phenomena that control the evolution of a LOCA transient will overlap and interplay. This is 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 Abreak = (Npump

• WRCP)/Gbreak.

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

Gbreak = (Gbreak(Po, HCLO) + Gbreak(PCLsat, HCLO))/ 2 .

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Large Break LOCA Analysis Page 4-26 The estimated minimum LBLOCA break area, Amin, is 2.76 ft 2 and the break area percentage, based on the full double-ended guillotine break total area, is 33-percent.

Table 4-4 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 4-4 Minimum Break Area for Large Break LOCA Spectrum Saturated Spectrum Spectrum System Cold Leg Subcooled Gbreak No. RCP Minimum Minimum Plant Pressure Desciptin (pia) Enthalpy Btulbm)(HEM)Gbreak R of flow Break Break (psia) (Btu/Ibm) (Ibm/ft 2-s) (Ibm/ft 2_s) RCPs (Ibm/s) Area2 Area (ft ) (DEGB)

A 3-Loop W 2250 554.0 22198 6330 3 31558 2.21 0.27 Design B 3-LoopW 2250 544.5 23880 5450 4 28124 1.92 0.23 Design C 3-Loop W 2250 550.0 23540 5580 4 29743 2.04 0.25 Design D 2x4CE 2100 538.8 22860 5310 4 21522 1.53 0.24 Design E 2x4 CE 2060 531.0 22068 5694 4 38277 2.76 0.28 Design F 2x4CE 2250 544 22930 5834 4 41230 2.87 0.29 Design G 2x4 CE 2250 548 22637 6091 4 42847 2.98 0.30 Design H 4-LoopW 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 (Amin) and an area of twice the size of the broken pipe. The determination of break configuration, split versus double-ended, is made after the break area is selected based on a uniform probability for each occurrence.

4.6.3 Intermediate Break Size Disposition With the revision of the smaller break area for the RLBLOCA analysis, the break range for small breaks and large breaks are no longer contiguous. 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 AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Large Break LOCA Analysis Page 4-27 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.

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 4-4 provides a listing of the plant type, initial condition, and the fractional minimum RLBLOCA break area. Figures 4-15 through 4-20 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 4-5 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 222 'F; however, this case is not representative of the general trend shown by the other comparisons. 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 eight plants, as at least 463 'F, and including this point would provide an average difference of 640 'F for the CE 2x4 design plants and a maximum difference of 840 'F for the 4-loop W plant design.

Thus, by both measures, the peak cladding temperatures within the intermediate break range will be several hundred degrees below those in the true large break range. Therefore, these breaks will not provide a limit or a critical measure of the ECCS performance. Given that the large break spectrum bounds the intermediate spectrum, the use of only the large break spectrum meets the requirements of 10CFR50.46 for breaks within the intermediate break LOCA spectrum, and the method demonstrates that the ECCS for a plant meets the criteria of 10CFR50.46 with high probability.

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Sequoyah Units 1 and 2 HTP Fuel Rev. 0 Realistic Larae Break LOCA Analysis Paae 4-28 Table 4-5 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 4-4) Size Break Break A 1206 193025 724 3-LoopW B 1273 1951 678 622 Design C 1326 1789 463 D 984 1751 767 E 1049 1740 691 2x4 CE 640 Design F 791 1670 879 G 1464 1686 22226 4-LoopW H 1127 1967 840 840 Design 25 The analysis for this W 3-Loop plant was performed with the Transition methodology and no break sizes fell into the intermediate break range. The PCT value of 1206 *F is the closest point to the maximum end of the intermediate break spectrum.

26 The analysis for this 2x4 CE plant was performed with the Transition methodology and no break sizes fell into the intermediate break range. The PCT value of 1464 *F is the closest point to the maximum end of the intermediate break spectrum. From the trends of the other 2x4 CE analyses, breaks falling within the intermediate break spectrum would be significantly lower.

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Realistic Large Break LOCA Analysis Rev. 0 Paae 4-29 2000 -

Upper End of Large Break SBLOCA Spectrum Break Size Minimum Spectrum 1800 + Break Area

• t I.

1600 - Fl ~ .

I *t

• I.

I 4 1400 + *

• 0- * *1 C

I.

~1 1200 -

4 1000 +

800 4 600 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 4-15 Plant A - Westinghouse 3-Loop Design AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-30 2000 Upper End of Large Break SBLOCA Spectrum Break Size Minimum Spectrum

• F 1800 Break Area 1600 *

---

1

-- -- - F.1

• 1400 F *F

• F F S Lk -

_1_

1200 -

1000 F F F F 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 4-16 Plant B - Westinghouse 3-Loop Design AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 4-31 2000 -T Upper End of Large Break SBLOCA Spectrum Break Size Minimum 1800 Spectrum Break Area 1600 1400 ----- - .- ---

o

[,.

1200 1000 800 600 4- I 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 4-17 Plant C - Westinghouse 3-Loop Design AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 4-32 2000 Upper End of Large Break SBLOCA Spectrum Break Size i¢" Minimum 1800 Spectrum Break Area 1600 +

• 4*

4 1400 +

oC 1200 +

1000 -k 800 f f6OA E 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 4-18 Plant D - Combustion Engineering 2x4 Design AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-33 2000 Upper End of Large Break SBLOCA Spectrum Break Size e" Minimum 1800 Spectrum Break Area 1600 -

• 4 4 4 1400 C

0 4

U . 4 Ow 1200 1000

• 4 800T 600 i =

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 4-19 Plant E - Combustion Engineering 2x4 Design AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-34 2200.0000 T Large Break Spectrum Upper End of Minimum 2000.0000 SBLOCA k, Break Area Break Size Spectrum 1800.0000 1600.0000 4.

1400.0000 law 1200.0000 1000.0000 800.0000 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 4-20 Plant H - Westinghouse 4-Loop Design AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-35 4.7 Detail informationfor ContainmentModel (ICECON)

Question: Verify that the 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 LOCA Analysis (Including Ice CondenserPlants)."

The AREVA RLBLOCA Report 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-2970(P) states that "in many instances" the guidance of NRC Branch Technical Position CSB 6-2 was used in determining the other containment parameters.

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

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

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

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

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

See Section 3.3 for discussion of questions (a) through (e). Containment initial conditions and cooling system information are provided in Table 3-8 and Heat Sinks are provided in Table 3-9.

For Sequoyah Units 1 and 2, the scatter plots of PCT versus the sampled containment volumes and initial atmospheric temperature are shown in Figure 4-21 and Figure 4-22. Containment pressure as a function of time for limiting case is shown in Figure 4-23. Figures 4-24 through 4-32 are provided to supplement the NRC's review of the Sequoyah Units 1 and 2 RLBLOCA analysis.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Pacqe 4-36 PCT vs Upper Compartment Containment Volume 2200 2000 k 0

1800 F M 1ý El: M E]

1600 I El -

U

-E] 0 E

0 WE E# E]M U 0 1400 k 0 lEl, El Igo U-E No 0

0~ El 1200 F [] El F-1000 k 800 F 600 F U Split Break W]Guillotine Break 400 650()00 660000 670000 680000 690000 700000 Upper Compartment Volume (ft3)

Figure 4-21 PCT vs. Containment Volume AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page m

4-37 PCT vs Upper Compartment Containment Temperature 2200 2000 F 0

MI 1800 F1 mE U El LIE U 1600 U HEf~m LI~ LI M N 0EL IFE El. U 0 El *EM U 1400 LI LI U EL U.

U N I-0 13_

1200 h

[]

1000 F-800 600 0 Split Break LI Guillotine Break 400 80 90 100 110 Upper Compartment Temperature (OF)

Figure 4-22 PCT vs. Initial Containment Temperature AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paqe 4-38 Containment and Loop Pressures 100 90 80 70 60

.CO 50 U) 40 30 20 10 0

0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 4-23 Containment Pressure for Limiting Case AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-39 Energy Addition in Lower Compartment 4500 4000 3500 3000

-c 2500 I-m (0

0 2000 C,

0)

C 1500 w

1000 500 0

-500 1....1.................. .... i .... I.... .....I ...... I ...

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

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 4-24 Energy Addition in Lower Compartment AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-40 Energy Rates in Lower Compartment 4500 4000 3500 3000 2500 CO 2000 0)

C: 1500 LU 1000 500 0

-500 0 10 20 30 40 50 60 70 80 90 100 110 120 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 4-25 Energy Rates in Lower Compartment AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-41 Energy Removal Rates in Lower Compartment 4500 4000 3500 3000 L--

2500 I--

2000 C:

w.

2)

U) 1500 1000 500 0

-500 0 10 20 30 40 50 60 70 80 90 100 110 120 Time (s)

ID:56939 16Mar2011 13:44:12 R5DMX Figure 4-26 Energy Removal Rates in Lower Compartment AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paqe 4-42 Energy Removal Rates in Upper Compartment 4500 4000 3500 3000 2500 I--

(0 0

2000 C)

C-UJ 1500 1000 500 0

-500 0 10 20 30 40 50 60 70 80 90 100 110 120 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 4-27 Energy Removal Rates in Upper Compartment AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 4-43 log of Heat Removal Rates 15.0 10.0 Z-I-

0) 5.0 0.0 0 5 10 15 20 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 4-28 Heat Removal Rates (log)

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-44 Fraction of Ice Remaining 1.0-0.9 Eý

? 0.8 C

0 t-Cu c,, 0.7 0.6 0.5 0 100 200 300 400 500 600 700 800 900 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 4-29 Fraction of Ice Remaining AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-45 Mass Addition to Lower Compartment 11000 10000 9000 8000 7000 aI) 6000 5000 0

U-4000 3000 2000 1000 0

0 10 20 30 40 50 60 7 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 4-30 Mass Addition to Lower Compartment AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-46 Upper and Lower Containment Pressure 22.0 21.0 20.0

.T 19.0 17.0 16.0 15.0 0 10 20 30 40 50 60 70 80 90 100 110 120 Time (s)

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 4-31 Upper Compartment and Lower Compartment Pressure AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 4-47 Temperature of Upper and Lower Compartments 250 .0 .... .......

-+ Lower Compartment

-o Upper Compartment 200.0

-D 2 150.0 a)

E 100.0 50 0 . . . . . . . . . . . . . . . . . . . . . . ..

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

ID:56939 16Mar2O11 13:44:12 R5DMX Figure 4-32 Temperature of Upper and Lower Compartments AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-48 4.8 Cross-References to North Anna Question: In order to conduct its review of the Sequoyah application of AREVA's realistic LBLOCA methods in an efficient manner, the NRC staff would like to make reference to the responses to NRC staff requests for additional information that were developed for the application of the AREVA methods to the North Anna Power Station, Units I 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 I Please verify that the information in these letters is applicable to the AREVA model applied to SQN except for that information related specifically to North Anna and to sub-atmospheric containments.

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 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 Units 1 and 2.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 4-49 The response provided to question 1 contains both generic and plant specific content. The portions that are generic remain applicable to Sequoyah Units 1 and 2. However, the North Anna Units use sub-atmospheric containment designs and Sequoyah Units 1 and 2 are of the ice condenser type. This leads to several differences in the way the information would be presented.

4.9 GDC 35 - LOOP and No-LOOP Case Sets In concurrence with the NRC's interpretation of GDC 35, a set of 93 cases each was run with a LOOP and No-LOOP assumption. The set of 93 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 3-24. This is a change to the approved RLBLOCA EM (Reference 1).

4.10 Statement Question: Provide a statement confirming that TVA and its LBLOCA analyses vendor have ongoing processes that assure that the input variables and ranges of parameters for the Sequoyah Units I and 2 LBLOCA analyses conservatively bound the values and ranges of those parameters for the as operated Sequoyah plants. This statement addresses certain programmaticrequirementsof 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 Units 1 and 2 realistic large break loss-of-coolant accident are verified as applicable 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;

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 TVA 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.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 5-1 5.0 Conclusions The results of the RLBLOCA analysis show that the limiting run (Case 86) had offsite power available and has a PCT of 1941 'F for a fresh U0 2 rod. The maximum oxidation thickness and hydrogen generation fall well within regulatory requirements.

The analysis supports operation at a 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 1.7056 (including uncertainty) with no axial dependent power peaking limit.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-1 6.0 Recent NRC Request for Additional Information (RAI) and AREVA Responses The NRC staff has found that strict adherence to currently referenced, or proposed for referencing AREVA methodologies are inconsistent with the NRC's requirements and review guidance without appropriate justification. This section addresses the NRC staffs concerns for the AREVA RLBLOCA methodology.

Question:

1. Please provide more information about the management of the fuel thermal conductivity degradation issue identified in NRC Information Notice 2009-23, "NuclearFuel Thermal Conductivity Degradation." Specifically:

a. Page 1-3, states, "For each specific time in cycle, the fuel conditions are computed using RODEX3A prior to starting the S-RELAP5 portion of the analysis. A steady-state condition for the given time in cycle using S-RELAP5 is established. A base fuel centerline temperature is established in this process.

Then two-transformation adjustment to the base fuel centerline temperature is computed. The first transformation is a linear adjustment for an exposure of 10 MWd/MTU or higher. In the new process, a polynomial transformation is used in the first transformation instead of a linear transformation." Please clarify the following:

i. Explain how the fuel pellet radialtemperatureprofile is computed.

ii. Explain which code is used to calculate this profile, both for initial conditions and through the postulated accident.

iii. Explain whether the polynomial transformation is applied merely to the centerline temperature,or to the entire pellet temperature.

b. Provide additional information to describe the polynomial transformation.

Summarize data used to develop the polynomial transformation and discuss considerationof applicable uncertainties.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-2

Response

The NRC concern covers a wide range of specific items but can be paraphrased as: "How does the AREVA RLBLOCA analysis for Sequoyah provide a licensing basis for fuel throughout its operational life with particular attention to the phenomena of thermal conductivity degradation with burnup?" In response, the following explanation of the methodology employed for Sequoyah is provided and followed by specific responses to each of the particular questions.

The AREVA transition package has been updated to specifically model both first and second cycle fuel rods. Third cycle fuel does not retain sufficient energy potential to achieve significant cladding temperatures nor cladding oxidation and is not included in the RLBLOCA individual pin calculations. The burnup for the individual first and second cycle rods analyzed is assigned according to the sampled time in cycle. The time in cycle is sampled once and is the same for both the fresh (first cycle) and once-burnt (second cycle) fuel. Burnup for the fresh and once-burnt rods is different in accordance with the cycle management. Likewise, pin pressure and thermal conductivity differ.

In addition to the thermal conductivity and fuel temperature adjustments for burnup, a burnup dependent reduction in allowed peaking is needed for the once-burnt fuel. For first cycle fuel, the RLBLOCA methodology increases the FAH to the Technical Specification maximum (including uncertainty) for the first cycle hot rods in the model. Shortly into the cycle, once-burnt fuel has insufficient energy potential to achieve this peaking. A burnup dependent reduction in allowed peaking is therefore applied through an adjustment in the second cycle FAH. Not modeling the reduction would result in the simulation of an operational state for the once-burnt fuel that would be impossible to achieve. Figure 6-4 provides the bounding once-burnt fuel U0 2 power ratio curve. The curve expresses the relative power to which the once-burnt fuel pins will be controlled as a function of burnup during the cycle.

1.a.i The RODEX3 topical report, ANF-90-145(P)(A), Appendix B (Reference 16) details the calculation of the radial temperature distribution.

1.a.ii A portion of the RODEX3A fuel model was incorporated into the S-RELAP5 code to calculate fuel response for transient analyses. This coding, referred to as the S-RELAP5/RODEX3A model, deals only with transient predictions and does not calculate the burnup response of the fuel. Instead, fuel conditions at the burnup of interest are transferred via a binary data file from RODEX3A to S-RELAP5/RODEX3A, establishing the initial state of the fuel prior to the transient. The data transferred from RODEX3A describes the fuel at zero power. A steady-state S-RELAP5/RODEX3A calculation is required to establish the fuel state at power. The transient fuel pellet radial temperature profile is computed by solving the conduction equation in S-RELAP5.

Material properties are calculated in S-RELAP5/RODEX3A.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-3 1.a.iii The adjustment is applied to the entire fuel pellet. The polynomial transformation provides a bias adjustment to the fuel centerline temperature. A sampled parameter provides a random assessment and adjustment of the centerline temperature uncertainty. These are combined and the total adjustment is achieved by iterating a multiplicative adjustment to the fuel thermal conductivity until the desired fuel centerline temperature is reached.

1.b. Paraphrased concern: Provide information on the treatment of thermal conductivity degradation.

Thermal conductivity degradation impacts the ability to transfer energy from within the pellet to the pellet surface and consequently through the cladding to the coolant. Both the initial pellet temperature and the transient release of energy from the pellet are affected. The impact of thermal conductivity changes with burnup are treated by applying a bias. This bias and a measure of the uncertainty in the data were determined by benchmarking the fuel performance code, RODEX3A, to a set of data that extends past the licensed burnup. The bias adjusts the initial fuel temperature to the mean of the benchmark results. The sampled uncertainty is used to provide for the variance of the benchmarks.

The database for the benchmarks is that used to qualify and approve the RODEX4 code (Reference 15). The data from three experimental rods (cases 432R2, 432R6, and 597R8) were not used in the benchmarks. Test 597R8 was not appropriate for this application. Cases 432R2 and 432R6 are rod studies that are not configured appropriately these types of comparisons.

Essentially, these fuel rods are not representative of commercial PWR fuel. Part of the benchmark activity was to incorporate a fractional representation of difference between the RODEX3A calculated results and the data. The fractional adjustment provides a better adjustment over a range of initial temperatures. Therefore, for each benchmark case the Tfraction was determined.

Zfraction T= d3 Trodex3A data where:

Tfraction = Delta fractional temperature of computed to data (K),

Trodex3A = Temperature computed by RODEX3A (K) and Tdata = Temperature from the RODEX4 database (K)

Figure 6-1 shows the RODEX3A benchmark results along with a polynomial fitted to the results using the least squares method. The negative of this polynomial is the bias which is added to RODEX3A predictions to achieve agreement with the data. Figure 6-2 shows the results of applying this bias in comparison to the results of applying the original RLBLOCA methodology Revision 0 bias. It is evident from Figure 6-2 that the bias makes the adjustment for burnup effects in accordance with the data.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-4 The application of the bias within the methodology proceeds as follows: The burnup for the case hot rods, fresh and once burned, is determined by sampling the time in cycle and a RODEX3A calculation of the initial fuel centerline temperature performed. From the fit in Figure 6-1 an adjusted temperature is determined as per the equation below.

K where:

I Tnew = Adjusted fuel centerline temperature (K),

B = Burnup (Gwd/MtU or Mwd/KgU) and Toriginal = Unadjust RODEX3A fuel centerline temperature (K).

Figure 6-3 provides the bias adjustment T"aIToriginat, as a function t offuburnup, using the h above b

polynomial curve fit.

The uncertainty is determined from a Gaussian distribution characterized by a [ ]

standard deviation and added to Tnew. The fuel temperature calculation is then repeated with a multiplier, fuel K, on the code calculated fuel thermal conductivity. The fuel centerline temperature is compared to 'Tnew + uncertainty' and the calculation is repeated with an adjusted fuel K as necessary. The process is continued until the calculated centerline fuel temperature matches 'Tnew + uncertainty'. Since the process applies an adjustment to the fuel thermal conductivity, the temperature throughout the pellet is adjusted appropriately. The final multiplier is applied to the thermal conductivity throughout the transient.

Because the data fitting covers the complete range of applicable burnup it is applied as such and the zero bias offset used in Revision 0 for the first 10 GWd/mtU burnup is eliminated.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-5 Follow-on Questions to #11:

The issue described in IN 2009-23 invalidatesAREVA's generic disposition for analyzing fresh fuel only, which is based on sensitivity studies indicating that mid-second-cycle fuel had a PCT of 80*F lower than the limiting PCT. This work needs to be repeated accounting for fuel thermal conductivity degradation. Please provide several cases run at various times-in-life for once-burnt fuel, with information similar to the above list provided; burnup for the limiting rod is only necessary for the most limiting second-cycle case analyzed.

For the PCT-limiting RLBLOCA case, please provide:

a. Correctedand uncorrectedradialtemperatureprofile of the hot rod at the time and location of peak cladding temperature.

b. Temperature vs. time for the limiting PCT case at the limiting location, including the fuel centerline, fuel average,and clad surface temperatures. Indicate the end of blowdown, startof refill, and start of reflood on this graph.

c. Burnup for the limiting rod.

Response

Figure 6-5 shows the corrected and uncorrected radial temperature profile for the limiting case hot rod at the initiation of the transient. Because the uncorrected radial profile is never used or recorded in the methodology, it cannot be provided. However, the uncorrected centerline temperature is available and shown on Figure 6-5. As the pellet power is not adjusted the radial temperature profile must follow the corrected profile closely and the two must converge at the surface of the pellet. Figure 6-6 shows the centerline, surface, and average fuel temperatures of the fresh U0 2 rod at the PCT elevation for the limiting PCT case. In this case, all of the fresh rods have higher PCTs than the once-burnt rods. The most limiting once-burnt rod is the U0 2 rod. With a cycle burnup of approximately 5400 EFPH, the fresh fuel has a burnup of 11.6 GWd/MTU while the once-burnt fuel has a burnup of -30 GWd/MTU. A plot comparing the PCT of the fresh and once-burnt U0 2 rods for this case is shown in Figure 6-7.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-6 Figure 6-1 Fractional Fuel Centerline Temperature Delta Between RODEX3A and Data AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 6-7 Figure 6-2 Fuel Centerline Temperature Delta of RODEX3A Calculations to Data (Original and Using the New Correlation)

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 6-8 Figure 6-3 Correction Factor (as applied for temperatures in Kelvin)

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 PaQe 6-9 Figure 6-4 K(Burnup) Curve AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 6-10 Figure 6-5 Radial Temperature Profile for Hot Rod AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paqe 6-11 Figure 6-6 Temperature versus Time for Fuel Centerline, Clad Surface, and Fuel Average AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Pacqe 6-12 PCT Trace for Case #86 2000 1500 4 0

a)

E 1000 C

0~

500 Hot U02 Rod Burned U02 Rod 0 L 0 200 400 600 800 1000 Time (s)

ID:56939 16Mar2011 13:44:12 R5DMX Figure 6-7 Fresh and Once-Burned U0 2 Rod PCT Trace AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-13 Question:

2. The current licensing basis, deterministic loss of coolant accident (LOCA) analysis concluded that the limiting condition did not involve a worst-case single failure, but rather that it depended on injected coolant delivered in such a condition that the resultant containment environment, specifically the lower containmentpressure, contributed to the limiting peak cladding temperature (PCT). Please provide information describing how this potentially limiting scenariowas evaluated using the proposed best-estimate methodology.

Response

2. The current licensing basis for both Sequoyah Units is AREVA's NRC-approved RLBLOCA evaluation model and the worst single failure considered is loss of diesel with fully functional containment sprays. The EM also conservatively prescribes:

(1) The use of full containment sprays without a time delay at the minimum technical specification temperature; (2) Pumped ECCS injection at the maximum technical specification temperature; and (3) Sampling of the containment volume (indirectly sampling containment pressure) from its nominal volume to its empty volume.

Studies, comparing several failure assumptions, including a no-failure assumption (see EMF-2103(P)(A) Revision 0, RAI response Numbers 26 and 111) validate that the ECCS and containment modeling of the AREVA methodology trends to the conservative. The containment pressure response is indirectly ranged by sampling the containment volume. The possible range to be sampled from was 6.510E+5 to 6.926E+5 ft 3 for the Sequoyah Units upper containment volume. Figure 4-21 shows that there is little sensitivity between containment volume (indirectly pressure) and PCT for a statistical application. Thus, the methodology is responsive to the goal of a realistic evaluation, yet slightly conservative.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-14 Question:

3. Please provide additional information summarizing the single-failure evaluation performed to establish compliance with General Design Criterion (GDC) 35 requirements. Identify which single failures were considered, discuss whether each failure was evaluated or explicitly analyzed, and for those failures which were explicitly analyzed, explain whether they were analyzed in a reference case or explicitly as a part of the statisticalmethodology. Also discuss the basis for the single failure evaluation.

For example, were single failures considered as a matter of experience with SEQUOYAH specifically, or with a generic Westinghouse nuclear steam supply system design?

Response

3. Section 4.9 discusses GDC 35. The single failure prescribed by EMF-2103(P)(A)

(AREVA's RLBLOCA EM) is a loss of one train of ECCS (See response to RAI Number 2).

AREVA satisfies the GDC-35 criteria by running one set of 93 cases with offsite power available and one set of 93 cases with no offsite power available. The sampling seeds are held constant between these two case sets, with the only difference being the offsite power assumption. The case set that produces the most limiting PCT is reported, for Sequoyah, this was offsite power available. Figure 3-24 in this document displays the results from the two case sets.

Follow-on Questions to #2 and #3:

a. The staff also needs to understandhow the limiting single failure for the 4-loop W NSSS was determined, since the basis for the RAI response defers to NRC-approved methodology. Poringthrough EMF-2103, the staff only locatedsensitivity results on 3-loop W systems. In some cases, the limiting failure would be a single LPSI and in others it was a diesel. The staff could not locate a clear,generic disposition for the single failure at any place in EMF-2103.

b. What was done under the auspices of EMF-2103 development to ensure that the containment analysis produced a sufficiently conservativeprediction that a no failure, max SI spillage case, for a 4-loop W NSSS, is bounded by the chosen single failure?

The staff will need to see that work.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Pace 6-15

Response

The definition for loss of a diesel scenario by itself would mean that in addition to loss of one LPSI and one HPSI pump, one train of containment spray would not be available. The current method models all containment pressure-reducing systems as fully functional. Containment fans start at time zero and containment sprays have a 10 second delay (Table 3-8).

The response to RAI #111 for EMF-2103 (Reference 26, Attachment 1 page 185 - 189) was based on sensitivities to 3-loop W plants. The Base Case, which produced the most limiting results, is described in the RAI #111 response as the loss of one diesel with full containment spray.

Figure 6-8 (recreated from RAI #111, Figure 111.2) shows that for the sample plant analysis, W 3-loop, the base case, AREVA ECCS failure assumptions, is 35 OF higher in PCT than a fully consistent loss of diesel and over 170 OF greater than the loss of one LPSI case.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-16

-- M Figure 6-8 Clad Temperature Response from Single Failure Study AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-17 A sensitivity study of the Sequoyah limiting case (Case 86) from the Analysis of Record (AOR) was conducted with "maximum" ECCS flow conditions to demonstrate that the minimum ECCS single failure assumption is conservatively bounding.

Sensitivity studies were run for the limiting case (Case 86) in both offsite power configurations with a maximum ECCS delivery. The loss of offsite power (LOOP) case for the max ECCS configuration had a PCT value of 1698 *F compared to AOR LOOP with minimum ECCS, which had a PCT of 1893 'F. The no loss of offsite power (NOLOOP) case for the max ECCS configuration had a PCT value of 1650 *F compared to AOR NOLOOP case with minimum ECCS, which had a PCT of 1941 *F. This demonstrates that the AREVA single failure assumption produces conservative results. Figures 6-9 through 6-12 show the respective PCT trace, containment and system pressure, ECCS injection rates, and downcomer level for both the AOR and the max ECCS sensitivity.

Figure 6-9 demonstrates that the maximum ECCS flow does not have a significant impact on the containment pressure up to about 100 seconds (approximately the time that the accumulator empties); the max ECCS containment pressure overlaps the AOR containment pressure.

Figure 6-12 gives the downcomer level for both the AOR and the max ECCS case. It can be seen that the downcomer level in the max ECCS case is higher than the AOR, consequently providing more driving head for the reflood of the core. The higher driving head in the max ECCS case is enough to compensate for small differences in containment pressure (Figure 6-10) resulting in a faster post peak cooldown.

The AREVA RLBLOCA application, regardless of the loss of diesel assumption, models all containment pressure-reducing systems and conservatively assumes them to be fully functional.

The AOR conservatively assumes an on-time start and normal lineups of the containment spray and fan coolers to conservatively reduce containment pressure and increase break flow. The results of the study demonstrate that the AOR ECCS configuration is PCT-limiting and oxidation-limiting.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-18 Figure 6-9 Comparison of PCT Independent of Elevation for Max ECCS and Min ECCS AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-19 Figure 6-10 Comparison of Containment and System Pressure for Max ECCS and Min ECCS AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-20 Figure 6-11 Comparison of ECCS Flows for Max ECCS and Min ECCS AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-21 Figure 6-12 Downcomer Level AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-22 Question:

4. Page 3-6 states, "the RLBLOCA transients are of sufficiently short duration that the switchover to sump cooling water (i.e., RAS) for ECCS pumped injection need not be considered." For the limiting transient, the collapsed core liquid level from 200-350 seconds appears to trend downward (Figure 3-21). An indication of stable and increasing collapsedliquid level would substantiate the statement quoted above, but this is not the case for Figure 3-21. Is the SRELAP-5 model of the limiting case capable of generating credible results after 350s? If so, please provide results for a period of the transient sufficient to demonstrate that the core collapsed liquid levels are stable or increasing.

Response

4. Not applicable to the Sequoyah analyses; Figure 3-21 clearly shows a steady level increase from 200 seconds through transient termination.

Question:

5. Pleaseprovide information to enable comparison between Technical Specifications (TS) requirements and analytic input parametersfor PressurizerLevel. The TS requirement is given in inches and the input parametersare specified in percent span.

Response

5. Technical Specification LCO 3/4.4.4 states "The pressurizer shall be OPERABLE with a water volume of less than or equal to 1656 cubic feet (equivalent to an indicated level of less than or equal to 92% on narrow range instrumentation), and at least two groups of pressurizer heaters each having a capacity of at least 150 kw."

The sampled range for the liquid level uncertainty in the pressurizer was 57- to 95-percent of span. The Technical Specifications for Sequoyah do not have the requirement in inches, just percent span.

Question:

6. Please provide discussion to confirm that the assumed 80°F upper containment temperature and 95 0 F lower containment temperature are acceptable minimums without a TS requirement.

Response

6. Sequoyah does have an Technical Specification LCO for containment temperature.

Technical Specifications LCO 3.6.1.5 states "between 85°F and 105 0 F in the containment upper compartment, and between 100°F and 125 0 F in the containment AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-23 lower compartment." The inputs TVA provided AREVA for the RLBLOCA analysis bound these LCO values.

Question:

7. The TS minimum for the refueling water storage tank (RWST) temperature is 60°F.

Previous, deterministic analyses demonstrated that minimum safety injection temperaturesresulted in a limiting PCT. In light of this information, please explain why a minimum RWST temperature case was not evaluated, or if a minimum RWST temperature case was evaluated, please summarize the evaluation and discuss its conclusions.

Response

7. As stated in the response to Question 2 in Section 6, the NRC-approved RLBLOCA EM, EMF-2103(P)(A), prescribes use of the maximum temperature for the ECCS pumped injection and use of the minimum temperature for containment sprays. Sequoyah's temperatures were 110 °F for pumped injection and 55 OF for the containment sprays.

While inconsistent, the choice of the two temperatures is conservative. AREVA's RLBLOCA analysis complies with, and does not deviate from, it's approved EM requirements.

Question:

8. As noted in Section 1 of the AREVA RLBLOCA Summary Report, deviations from the approved RLBLOCA evaluation model (EMF-2103(P)(A), Revision 0) are necessary to demonstrate compliance with 10 CFR 50.46 requirements. Please provide a commitment to adhere to the deviations noted in Section 1 of AREVA RLBLOCA Summary Report until such time as:

a. AREVA develops a new revision of EMF-2103,

b. The NRC approves the new revision of EMF-2103, and

c. Sequoyah implements the new, NRC-approved revision of EMF-2103.

The commitment should include language to indicate that meeting Conditions a, b, and c, above, or submitting a license action request to implement a different evaluation method, will obviate the need for this commitment.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-24

Response

8. TVA commits to the following:

Sequoyah Units 1 and 2 will adhere to the deviations noted in Section 1 of ANP-2970(P)(A) until such time as:

"AREVA develops a new revision of EMF-2103, "The NRC approves the new revision of EMF-2103, and "TVA implements the new, NRC-approved revision of EMF-2103.

This commitment will terminate when the above items are met or a license amendment is approved to permit the use of a different evaluation method to replace ANP- 2970(P)(A).

Question:

9. The following questions are based on a July 14, 2009, letter from Gardner,AREVA NP, to the USNRC, re: Informational Transmittal Regarding Requested White Paperson the Treatment of Exposure Dependent Fuel Thermal Conductivity Degradation in Legacy Fuel Performance Codes and Methods.

A) AREVA postulates that clad swelling and rupture produces a benefit to PCT, and because of this, the realistic large break loss of coolant accident (RLBLOCA) model does not include a clad swelling and rupture model. Does this conjecture include considerationof test data, which has shown that following fuel rupture, the ballooned region fills with fuel fragments? What analytic studies support this conclusion? How are they applicable to Sequoyah? Please also address the potential for co-planar blockage with the fuel relocation evaluation.

B) Since blowdown ruptures can occur at end of life conditions, show that blowdown ruptures do not occur at the end of life for the postulated Sequoyah large break LOCA.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-25

Response

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-26 Experience with Appendix K methodologies has shown that the aggregate of these effects acts to decrease the cladding temperatures when no fuel relocation occurs. This was demonstrated in Appendix B, Section B.2 of RLBLOCA EM Topical (Reference 1) and the response to RAI 28 on the topical (page 79 of Amendment 1 to Reference 17) with sensitivity studies on both 3- and 4-loop PWRs with 15x15 and 17x17 fuel designs similar to the 17x17 fuel design used at Sequoyah. The studies included increased heat transfer surface area, increased local coolant velocities, a decrease in gap heat transfer, flow diversion, and interior cladding oxidation. The effects of increased turbulence, droplet shattering, and potential local quenching were not included within the modeling. Decrease in pellet thermal conductivity and a clad heating load increase also were not included since the studies were not meant to address fuel relocation.

Even without half of the cooling mechanisms modeled, the cladding temperatures and local oxidations were reduced. This effect has also been observed experimentally in the FEBA (Reference 18) and FLECHT (Reference 19) test series.

Under a condition of fuel relocation, wherein the fuel above the ballooned region drops into the ballooned region, it has been postulated that increased decay heat generation will lead to an increase in cladding heat flux resulting in higher cladding temperatures. Various presentations (e.g., Reference 20 Articles 1 and 12) purport to show the effect. However, these studies have uniformly incorporated extreme assumptions on the conditions of relocation and the resultant heat transfer processes. Few include provisions for rupture-induced cooling mechanisms. Most assume that the cladding expands circularly without being encumbered by the surrounding pins in the fuel assembly. In fact, a free expansion of the fuel rod is only possible up to pin strains in the mid-30 percents. For higher strains the local gap volume no longer increases faster than the clad surface area. Finally, the packing factor of the rubble filling the ballooned region is over-predicted. If reasonable, yet conservative, assumptions are made, study results would lead to the expectation that fuel relocation, which is real, does not pose a condition by which the ruptured or ballooned regions will exceed the consequence of the non-ballooned regions of the hot pin.

The above conclusion was observed experimentally in the KfK experiments as reported in RAI 131 on the RLBLOCA EM topical report (page 120 of Amendment 1 to Reference 17). In the KfK in Pile Tests, fuel relocation into the ballooned area of the fuel rod occurred but did not adversely affect the subsequent clad temperature behavior. To determine when the fuel relocates two tests were performed with thermocouples located at the top of the pellet stack.

One test comprised low burnup fuel, which maintained its pellet geometry after rupture. The other test was of higher burnup fuel which relocated. Relocation, for the test that relocated, was demonstrated by temperatures from the upper thermocouples showing a significant drop, loss of energy source, at the time of fuel rod rupture. For this test, the heatup rate, at the rupture elevation, following the rupture was reduced relative to the heatup rate prior to rupture. This reduction in heatup rate indicates that the PCT at the time of turnover would be less than what would have be reached if rupture had not occurred, even with the increase in localized decay heat from the pellet rubble residing at the ruptured region. Thus, the KfK experiments AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-27 demonstrate that analyses which ignore the beneficial effects of swelling and rupture provide conservatively high clad temperature estimates for the ruptured region during reflood even when fuel relocation occurs.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-28 Question:

10. Provide information to illustrate the conservative nature of the single-side only oxidation model and its application to the SEQUOYAH RLBLOCA analysis.

Response

10. AREVA's NRC-approved RLBLOCA EM uses the maximum un-ruptured cladding oxidation as representative or bounding of the oxidation that would have been computed at a rupture location. The position is supported by three aspects of the performed oxidation calculation.

• The cladding is initialized with no initial corrosion layer. Because the oxidation rate is inversely proportional to the oxidation layer present, the use of clean cladding at the start of the accident leads to substantially higher reaction rates. For corrosions in the range of the first cycle of M5 cladding, the difference in rate is a minimum of a 50-percent increase and increases during the cycle. The increase applies to both exterior and post-rupture interior oxidation.

" The cladding temperature even in the presence of fuel relocation is reduced for the ruptured region of the cladding. In the KfK experiments (page 210 of NRC:02:062 Attachment 1 to Reference 17 and included in Reference 18) the temperature drop at rupture was between 50 and 75 K. Since the oxidation rate is exponentially proportional to the cladding temperature, a drop of 50 to 75 K for Sequoyah provides an oxidation rate reduction of 50-percent or more.

" For ruptured cladding either the cladding interior oxidation rate is reduced by attached pellet fragments, moderate to highly burned fuel, or the cladding temperature decrease at rupture is much more than the 50 to 75 K explained in Item 2. In either case, an additional mechanism exists to reduce the local oxidation at the rupture location.

In conclusion, insights into the EM oxidation process and those that will evolve after rupture clearly identify differences that will reduce the oxidation at the rupture location to less than that which the EM calculates at un-ruptured locations. Thus, the RLBLOCA Revision 0 EM approach to reporting local oxidation is clearly appropriate to demonstrate compliance with the local oxidation criterion of 10CFR50.46.

Follow-on Question to #10:

NRC Generic letter 98-29 calls of the initial corrosionlayer on the fuel pin to be included in the reportedlocal oxidation value demonstrating compliance with the criteriaof IOCFR50.46.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-29

Response

The initial corrosion layer was calculated to be 0.7624-percent for the Fresh U0 2 rod (at 15 GWd/MTU) and 1.3144-percent for the once-burned 6% GAD rod. The initial corrosion layer was added to the transient calculated value and the total is in Table 3-5.

Question:

11. Provide additional information to justify the use of the selected analytic treatment for decay heat uncertaintyin the RLBLOCA model.

Response

11. The RLBLOCA EM decay heat calculations are based on the 1979 ANSI/ANS standard (Reference 25). The standard is applicable to light water reactors containing low enriched uranium as the initial fissile material; all plants, to which the RLBLOCA EM is applicable, including Sequoyah, are such plants. The selected approach to simulate fission product decay assures a representative yet conservative treatment. The EM fission product decay heat uncertainty and the basis for the conservatism of the approach are outlined in the remainder of the response.

Non-Samplincq Approach to Decay Heat The RLBLOCA methodology proposed herein utilizes the U235 decay curve from the 1979 ANSI/ANS standard for fully saturated decay chains as the decay for all fission products. The fully saturated chains result from an assumption of infinite operation. The total energy per fission is assumed to be 200 MeV (Reference 25). No bias or uncertainty is assigned to the fission product decay heat. Differing from the base EMF-2103 evaluation model approach, the uncertainty for the decay heat parameter is set to zero and no sampling is done on this parameter, resulting in the decay heat being used with a 1.0 multiplier. The decay heat in the analysis is always the 1979 ANS standard for decay heat from U235 with fully saturated decay chains, corresponding to infinite operation, assuming 200 MeV per fission.

Conservatism in the Approach In the approach used, the total energy per fission is assumed to be 200 MeV whereas a more accurate value for U235 would be greater than 202 MeV per fission. This imparts a direct 1-percent conservatism.

During irradiation, plutonium accumulates such that the ratio of plutonium-to-uranium fission-energy production rate is substantial and increasing. Because the decay energy resulting from plutonium fissions is less than that from U235, the decay energy is reduced from U235 fully saturated decay chains as the fuel is burned. Thus, as burnup increases, the RLBLOCA decay AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-30 heat modeling with U235 only, accrues conservatism. This conservatism applies to all regions of the core according to the mix of burnups represented within each region.

The fresh fuel, hot pin and hot assembly, begin operation with no plutonium. Therefore, the reduction in decay heat due to plutonium build-up is not applicable to the low burnup fuel in the initial period of the cycle. However, for fresh fuel, the concentrations of long decay term fission products will not have built up. The lack of long decay term sources comprises a reduction in decay heat rate of several percent over the first year of operation, making the infinite operation assumption conservative while the plutonium concentration is accumulating.

Calculations of these considerations based on the 1979 ANS standard have been performed to demonstrate the conservatism of the selected approach. Figure 6-13 and Figure 6-14 show the decay heat versus time for:

1) Infinite Operation of U235 (the AREVA decay heat model)

2) Finite Operation to 0 GWD/mtU of all fissionable isotopes with uncertainties added

3) Finite Operation to 1 GWD/mtU of all fissionable isotopes with uncertainties added

4) Finite Operation to 1 GWD/mtU of all fissionable isotopes without uncertainties

5) Finite Operation to 20 GWD/mtU of all fissionable isotopes with uncertainties added

6) Finite Operation to 40 GWD/mtU of all fissionable isotopes with uncertainties added

7) Finite Operation to 60 GWD/mtU of all fissionable isotopes with uncertainties added In order to treat the Plutonium buildup effect conservatively, the finite operations curves are based on cycle management and enrichment assumptions that minimize the build up of Plutonium. No uncertainty is included in the infinite operation curve. The uncertainties incorporated in the other curves are 2 sigma values for the individual isotopes as published in the 1979 ANS standard. This provides greater than a 95/95 confidence in each of the decay heat contributions. The contributions are added linearly according to the individual isotopes fractional occurrence of fission.

Because of the range of the decay heat parameter, the early comparison of the relationships is difficult to ascertain. Clearly the U235 infinite operation curve is conservative for all times after a few seconds (-2 seconds). To better demonstrate the relationships, Figure 6-15 and Figure 6-16 provide the ratios of the finite operation curves to the infinite operation curves. The curvature of the plotted ratios during the first 2 to 3 seconds is due to the increased uncertainties during this time phase. The 1979 ANS standard is based on measured data and the difficulty of measuring decay heat within a few seconds of shutdown is reflected in these uncertainties. The highest combined finite operation decay heat curve with uncertainties exceeds the AREVA decay heat curve by only 2.5 percent at shutdown and falls below the AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-31 AREVA curve in less than 2 seconds. Thus, there is only a 5 percent probability that the infinite operation curve of decay heat will be exceeded by up to 2.5 percent and that possibility exists for the firs 2 seconds of the transient. The potential accumulated underprediction is of short duration and of no consequence to the LOCA evaluation. The decay heat curve selected is suitable while somewhat conservative for the realistic evaluation of LOCA.

In conclusion, the choice of infinite operation with pure U235 fission product decay heat provides a base model that is conservative relative to the decay heat for finite operation. For RLBLOCA evaluation, the sampling of a decay heat multiplier has been removed such that the decay heat for all cases is 1.0 times the infinite operation U235 decay chain providing conservative treatment of the 1979 ANS standard with the assumption of 200 Mev/fission.

Follow-on Question to #11:

The NRC needs to understand the sensitivity that PCT has with respect to the decay heat uncertainty, please re-execute the limiting case with a 1.03 decay heat multiplier and report the results.

Response

Not applicable to the Sequoyah analysis. The decay heat was not sampled.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 6-32 Figure 6-13 Decay Heat Comparisons, Infinite Operation U235, Finite Operation All Isotopes (0.1 - 10 sec)

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 6-33 Figure 6-14 Decay Heat Comparisons, Infinite Operation U235, Finite Operation All Isotopes (10 - 1000 sec)

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paqe 6-34 Figure 6-15 Decay Heat Ratios, Finite Operation over Infinite Operation U235 All Isotopes (0 - 10 sec)

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Paae 6-35 Figure 6-16 Decay Heat Ratios, Finite Operation over Infinite Operation U235 All Isotopes (>10 sec)

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 7-1 7.0 References

1. EMF-2103(P)(A) Revision 0, Realistic Large Break LOCA Methodology, Framatome ANP, Inc., April 2003.

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

3. Wheat, Larry L., "CONTEMPT-LT A Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-Of-Coolant-Accident," Aerojet Nuclear Company, TID-4500, ANCR-1219, June 1975.

4. XN-CC-39 (A) Revision 1, "ICECON: A Computer Program to Calculate Containment Back Pressure for LOCA Analysis (Including Ice Condenser Plants)," Exxon Nuclear Company, October 1978.

5. U. S. Nuclear Regulatory Commission, NUREG-0800, Revision 3, Standard Review Plan, March 2007.

6. NUREG/CR-1532, EPRI NP-1459, WCAP-9699, "PWR FLECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Task Data Report," June 1980.

7. 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 for the TRAC Code,"

November 1979.

9. J.P. Holman, Heat Transfer, 4 th Edition, McGraw-Hill Book Company, 1976.

10. EMF-CC-1 30, "HUXY: A Generalized Multirod Heatup Code for BWR Appendix K LOCA Analysis Theory Manual," Framatome ANP, May 2001.

11. D. A. Mandell, "Geometric View Factors for Radiative Heat Transfer within Boiling Water Reactor Fuel Bundles," Nucl. Tech., Vol. 52, March 1981.

12. EMF-2102(P)(A) Revision 0, S-RELAP5: Code Verification and Validation, Framatome ANP, Inc., August 2001.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 7-2

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).

14. G.P. Liley and L.E. Hochreiter, "Mixing of Emergency Core Cooling Water with Steam:

1/3 - Scale Test and Summary," EPRI Report EPRI-2, June 1975.

15. EMF-2994(P) Rev. 4, "RODEX4: Thermal-Mechanical Fuel Rod Performance Code Theory Manual," December 2009.

16. ANF-90-145(P)(A), "RODEX3 Fuel Rod Thermal-Mechanical Response Evaluation Model," April 1996.

17. AREVA Letter NRC:02:062, December 20, 2002, Responses to a Request for Additional Information on EMF-2103(P) Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," (TAC No. MB2865).

18. P. IhIe, Heat Transfer in Rod Bundles with Severe Clad Deformations, KfK 3607 B, April 1984.

19. M.J. Loftus, et al., PWR FLECHT SEASET 163-Rod Bundle Flow Blockage: Task Data Report, No. 13, NUREG/CR-3314, October 1983.

20. NEA/CSNI/R(2004)19, SEGFSM Topical Meeting on LOCA Fuel Issues, Argonne National Laboratory, May 25-26 2004, Published by Organization for Economic Cooperation and Development Nuclear Energy Agency, Isy-les-Moulineaux, France, November 2004.

21. NUREG/CR-0103, ORNL/NUREG/TM-200, Multirod Burst Test Program Progress Report for July - December 1997, US Nuclear Regulatory Commission, Washington, D.C.

22. NUREG/CR-0655, ORNLINUREG/TM-297, Multirod Burst Test Program Progress Report for July - December 1998, US Nuclear Regulatory Commission, Washington, D.C.

23. NUREG/CR-1 023, ORNL/NUREG/TM-351, Multirod Burst Test Program Progress Report for April - June 1999, US Nuclear Regulatory Commission, Washington, D.C.

AREVA NP Inc.

Sequoyah Units 1 and 2 HTP Fuel ANP-2970(NP)

Realistic Large Break LOCA Analysis Rev. 0 Page 7-3

24. K. Wieher and U. Harten, Datenbericht REBEKA-6, KfK 3986, March 1986.

25. ANSI/ANS-5.1-1979, American National Standard for Decay Heat Power in Light Water Reactors, approved August 29, 1979.

26. AREVA Letter NRC:02:062, December 20, 2002, Responses to a Request for Additional Information on EMF-2103(P) Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," (TAC No. MB2865).

AREVA NP Inc.

ATTACHMENT 11 AREVA NP Affidavits Attached are the affidavits supporting the request to withhold proprietary information (included in Attachments 5, 6, and 7) from the public.

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

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

2. I 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 ANP-2986(P),

Revision 002, entitled "Sequoyah HTP Fuel Transition," dated June 2011 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 ,2011.

/

Kathleen Ann Bennett NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 8/31/11 Reg. # 110864

• "KATHLEEN ANN BENNETT JNotary Public

.Commonwealth of Virginia 110864 My Commlsslon Expires Aug 31, 2011

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

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

2. I 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-2971(P), Revision 1, entitled "Sequoyah Units 1 and 2 HTP Fuel S-RELAP5 Small Break LOCA Analysis," dated May 2011 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 secret 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), 6(c) and 6(e) 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 M AAI / 2011.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/14 Reg. # 7079129 SHERRFY L. MCFADEN Notary. Public L 7079129 My Commission Expires Oct 31, 2014

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

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

2. I 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-2970(P), Revision 0, entitled "Sequoyah Units 1 and 2 HTP Fuel Realistic Large Break LOCA Analysis," dated March 2011 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 secret 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), 6(c) and 6(e) 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 f,/UYYT.4'-' 2011.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/14 Reg. # 7079129 dSHERRY L.MCFADEN Notary Public Commonwealth of Virginia I 7079129 My Commission Expires Oct 31, 2014