ML11172A072

ANP-2971(NP), Rev 1, Sequoyah Units 1 and 2 Htp Fuel S-RELAP5 Small Break LOCA Analysis, Attachment 9

ML11172A072
Person / Time
Site:	Sequoyah
Issue date:	05/31/2011
From:	AREVA NP
To:	Office of Nuclear Reactor Regulation
References
TS-SQN-2011-07 ANP-2971(NP), Rev 1
Download: ML11172A072 (56)

Text

ATTACHMENT 9 ANP-2971(NP), Revision 1 Sequoyah Units 1 and 2 HTP Fuel S-RELAP5 Small Break LOCA Analysis May 2011 (NON-PROPRIETARY VERSION)

ANP-2971 (NP)

Revision 1 Sequoyah Units 1 and 2 HTP Fuel S-RELAP5 Small Break LOCA Analysis May 2011 A

AREVA MP Inc. AREVA

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 2 AREVA NP Inc.

Copyright © 2011 AREVA NP Inc.

All Rights Reserved

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paae 3 Nature of Changes Item Page Description and Justification 1 All Revision 1. Changed the PCT result for the accumulator line break.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paae 4 Table of Contents Table of Contents .......................................................................................................................... 4 List of Tables ................................................................................................................................. 5 List of Figures ............................................................................................................................... 6 Nom enclature ................................................................................................................................ 7 1.0 Introduction ....................................................................................................................... 8 2.0 Sum m ary of Results ...................................................................................................... 9 3.0 Description of Analysis ................................................................................................ 10 3.1 Description of SBLOCA Event ......................................................................... 10 3.2 Analytical Models ............................................................................................. 10 3.3 Plant Description and Summary of Analysis Parameters ................................ 15 4.0 Analytical Results ............................................................................................................ 22 4.1 Results ................................................................................................................ 22 4.2 Discussion of Transient for 9.76-inch Break ................................................... 23 4.3 RCP Trip Sensitivity Study ............................................................................. 24 4.3.1 Cold Leg Breaks RCP Trip Sensitivity Study ..................................... 25 4.3.2 Hot Leg Breaks RCP Trip Sensitivity Study ..................................... 25 4.4 Attached Piping LOCA Cases ......................................................................... 25 4.5 Results Com parison against AOR ................................................................... 25 5.0 References ...................................................................................................................... 55 This document contains a total of 55 pages.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 5 List of Tables Table 3-1 System Parameters and Initial Conditions ............................................................................ 17 Table 3-2 CCI, HHSI, & RHR Flows as Function of the Primary System Pressure ............................... 18 Table 3-3 Moderator Density Reactivity ................................................................................................ 19 T able 3-4 D oppler R eactivity ...................................................................................................................... 19 Table 4-1 Summary of SBLOCA Break Spectrum Results ................................................................... 27 Table 4-2 Sequence of Events for Break Spectrum Calculations, sec ................................................ 28 Table 4-3 Summary of Results for the Cold Leg Break RCP Trip Sensitivity ........................................ 29 Table 4-4 Summary of Results for the Hot Leg Break RCP Trip Sensitivity Study ............................... 29 AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 6 List of Figures Figure 3-1 S-RELAP5 SBLOCA Reactor Coolant System Nodalization .............................................. 12 Figure 3-2 S-RELAP5 SBLOCA Secondary System Nodalization ....................................................... 13 Figure 3-3 S-RELAP5 SBLOCA Reactor Vessel Nodalization .............................................................. 14 Figure 3-4 A xial Pow er S hape ................................................................................................................... 20 Figure 3-5 Axial Power Distribution Comparison .................................................................................. 21 Figure 4-1 Reactor Power - 9.76-inch Break ........................................................................................ 30 Figure 4-2 Primary and Secondary System Pressures - 9.76-inch Break ............................................ 31 Figure 4-3 Liquid and Vapor Break Flow - 9.76-inch Break ................................................................. 32 Figure 4-4 Vapor Void Fraction at the Break - 9.76 in Break .............................................................. 33 Figure 4-5 Loop Seal Void Fraction - 9.76-inch Break .......................................................................... 34 Figure 4-6 Total Core Inlet Mass Flow Rate - 9.76-inch Break ............................................................ 35 Figure 4-7 RCS Loop Flow Rate - 9.76-inch Break .............................................................................. 36 Figure 4-8 Steam Generator Main Feedwater Flow Rates - 9.76-inch Break ...................................... 37 Figure 4-9 Auxiliary Feedwater Flow - 9.76 in Break ............................................................................ 38 Figure 4-10 Steam Generator Total Mass - 9.76-inch Break ............................................................... 39 Figure 4-11 Steam Generator Level - 9.76 in Break ............................................................................ 40 Figure 4-12 Total HHSI Mass Flow - 9.76-inch Break ......................................................................... 41 Figure 4-13 Total Accumulator Flow - 9.76-inch Break ........................................................................ 42 Figure 4-14 Integrated Break Flow and ECC Flow - 9.76 in Break ...................................................... 43 Figure 4-15 Total Primary Mass - 9.76-inch Break .............................................................................. 44 Figure 4-16 Reactor Vessel Mass - 9.76-inch Break ............................................................................ 45 Figure 4-17 Hot Assembly Collapsed Liquid Level - 9.76-inch Break ................................................. 46 Figure 4-18 Hot Rod Mixture Level - 9.76 in Break ............................................................................... 47 Figure 4-19 Downcomer Collapsed Liquid Level - 9.76 in Break ........................................................ 48 Figure 4-20 Volume Liquid Temperature at PCT Location - 9.76 in Break .......................................... 49 Figure 4-21 Liquid Void Fraction Distribution in the Hot Channel 0.75 - 7.75 ft Elevation

- 9 .7 6 in B re a k ............................................................................................................................... 50 Figure 4-22 Liquid Void Fraction Distribution in the Hot Channel 8.25 - 11.75 ft Elevation

- 9 .7 6 in Bre a k ............................................................................................................................... 51 Figure 4-23 Heat Transfer Coefficient at PCT Location - 9.76-inch Break .......................................... 52 Figure 4-24 Peak Cladding Temperature at PCT Location - 9.76-inch Break ..................................... 53 Figure 4-25 Comparison of PCTs - AOR vs. S-RELAP5 Analysis ........................................................ 54 AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 7 Nomenclature ADV Atmospheric dump valve AFAS Auxiliary feedwater actuation signal AFW Auxiliary feedwater AOR Analysis of Record BOC Beginning-of-cycle CCI Centrifugal Charging Injection CEA Control Element Assembly CFR Code of Federal Regulation CRGT Control Rod Guide Tube DC-HL Downcomer - Upper Head ECCS Emergency core cooling system EDG Emergency diesel generator EOC End-of-cycle EOP Emergency Operating Procedure HHSI High head safety injection HTP High Thermal Performance LHR Linear heat rate LOCA Loss of coolant accident LOOP Loss of offsite power LHSI Low head safety injection MFW Main feedwater MSIV Main steam isolation valve MSSV Main steam safety valve MWt Mega-Watt thermal NRC Nuclear Regulatory Commission PCT Peak cladding temperature PWR Pressurized water reactor RAI Request for Additional Information RCP Reactor coolant pump RCS Reactor coolant system RHR Residual Heat Removal SBLOCA Small Break Loss-of-Coolant-Accident SEQ Sequoyah SG Steam generator SGTP Steam generator tube plugging SI Safety injection SIAS Safety injection actuation signal AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 8 1.0 Introduction This report documents a SBLOCA analysis for the Sequoyah Units 1 & 2 (SEQ1 and SEQ2) plants for the transition from AREVA Mark-BW fuel to AREVA HTP fuel. This analysis was performed with the S-RELAP5 methodology (References [1] and [2]).

A spectrum of break sizes from 0.0055 (1-inch diameter) to 0.5217 ft2 (9.78-inch diameter) was analyzed, including the reactor coolant pump trip sensitivity study and the attached piping break study to support the operation of SEQ1 and SEQ2.

This analysis supports plant operation at a power level of 3479 MWt (including measurement uncertainty), a FQ of 2.65, a radial peaking factor of 1.7056 (1. 64 plus a 4% measurement uncertainty), and 15% steam generator tube plugging.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 9 2.0 Summary of Results A SBLOCA break spectrum analysis was performed for SEQW and SEQ2. The limiting PCT is 1470 OF for a break of 9.76-inch diameter pump discharge cold leg break.

This analysis was performed to demonstrate that the following acceptance criteria for Emergency Core Cooling Systems, as stated in 10 CFR 50.46(b)(1-4), have been met.

(1) The calculated maximum fuel element cladding temperature shall not exceed 2200 0 F.

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

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

(4) Calculated changes in core geometry shall be such that the core remains amenable to cooling.

The break spectrum calculations conservatively assumed RCP trip at reactor trip and loss of offsite power. An evaluation of delayed RCP trip was performed since a delayed RCP trip can potentially produce more limiting results. Section 4.3 discusses the RCP trip sensitivity study.

The results with respect to RCP trip would be "a trip within 2 minutes of loss of subcooling or SIAS actuation" does not result in a violation of 50.46(b).

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 10 3.0 Description of Analysis Section 3.1 below provides a brief description of the postulated SBLOCA event. Section 3.2 describes the analytical models used in the analysis. Section 3.3 presents a description of SEQ1 and SEQ2 plant parameters and outlines the system parameters used in the SBLOCA analysis.

3.1 Description of SBLOCA Event The postulated SBLOCA is defined as a break in the RCS pressure boundary which has an area of up to approximately 10% of a cold leg pipe area. The most limiting break location for the base break spectrum cases assuming LOOP is in the cold leg pipe on the discharge side of the RCP.

This break location results in the largest amount of inventory loss and the largest fraction of ECCS fluid being ejected out through the break. This produces the greatest degree of core uncovery, the longest fuel rod heatup time, and consequently, the greatest challenge to the 10 CFR 50.46(b)(1 -4) criteria.

The SBLOCA event is characterized by a slow depressurization of the primary system with a reactor trip occurring on a Low Pressurizer Pressure signal. The SIAS occurs when the system has further depressurized. The capacity and shutoff head of the HHSI pumps are important parameters in the SBLOCA analysis. For the limiting break size, the rate of inventory loss from the primary system is large enough that the HHSI pumps cannot preclude significant core uncovery. The primary system depressurization rate is slow, extending the time required to reach the accumulator injection pressure (600 psig) or to recover core liquid level on HHSI flow. This tends to maximize the heatup time of the hot rod which produces the limiting PCT and local cladding oxidation. Core recovery for the limiting break begins when the SI flow that is retained in the RCS exceeds the mass flow rate out the break, followed by accumulator injection. For very small break sizes, the primary system pressure does not reach the accumulator injection pressure.

3.2 Analytical Models The AREVA S-RELAP5 SBLOCA evaluation model for event response of the primary and secondary systems and hot fuel rod used in this analysis (Reference [1]) consists of two computer codes. The appropriate conservatisms, as prescribed by Appendix K of 10 CFR 50, are incorporated. This methodology has been reviewed and approved by the NRC to perform SBLOCA analyses. The two AREVA computer codes used in this analysis are:

1. The RODEX2-2A code was used to determine the burnup-dependent initial fuel rod conditions for the system calculations.

2. The S-RELAP5 code was used to predict the thermal-hydraulic response of the primary and secondary sides of the reactor system and the hot rod response.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paae 11 The gap conditions used to initialize S-RELAP5 are taken at EOC, consistent with an EOC top-peaked axial power distribution. The use of EOC fuel rod conditions along with an EOC power shape is bounding of BOC because (1) the gap conductance is higher at EOC, and (2) the power shape is more top-skewed at EOC. The initial stored energy, although higher at BOC, has a negligible impact on SBLOCA results since the stored energy is dissipated long before core uncovery.

Figures 3-1 through 3-3 show the RCS, Secondary, and Reactor Vessel nodalization.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 12 Figure 3-1 S-RELAP5 SBLOCA Reactor Coolant System Nodalization AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 13 Figure 3-2 S-RELAP5 SBLOCA Secondary System Nodalization AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paqe 14 Figure 3-3 S-RELAP5 SBLOCA Reactor Vessel Nodalization AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 15 3.3 Plant Description and Summary of Analysis Parameters The plant analysis presented in this report is for a Westinghouse-designed PWR, which has four loops, each with a hot leg, a U-tube steam generator, and a cold leg with a RCP 1 . The RCS Also includes one pressurizer connected to a hot leg. The core contains (193) 17x17 AREVA fuel assemblies. The ECCS includes one charging and one accumulator/SI/RHR injection path per RCS loop. The SI and RHR feed into common headers which are connected to the accumulator lines. The charging pumps are also cross-connected. The limiting break location will be in a cold leg pump discharge location of piping. The RLBLOCA transients are of sufficiently short duration that the switchover to sump cooling water (i.e., RAS) for ECCS pumped injection need not be considered.

The S-RELAP5 model explicitly describes the RCS, RV, pressurizer, and accumulator lines.

The charging injection flows are connected to the RCS, and the SI and RHR injection flows are connected to the accumulator lines, consistent with the plant layout. A symmetric steam generator tube plugging level of 15-percent per steam generator was assumed.

MFW is injected into the downcomer of each SG. There are three AFW pumps, that is, two motor-driven and one turbine-driven pumps. The ECCS for the SBLOCA contains two HHSI pumps, four accumulators.

The RCS was nodalized in the S-RELAP5 model into control volumes interconnected by flow paths or "junctions." The model includes four accumulators, a pressurizer, and four SGs with both primary and secondary sides modeled. All of the loops were modeled explicitly to provide an accurate representation of the plant. A SGTP level of 15% in each steam generator was assumed. The HHSI system was modeled to deliver the total SI flow asymmetrically to the broken loop and three intact loops in the S-RELAP5 model. The RHR system is included in the model.

The heat generation rate in the S-RELAP5 reactor core model was determined from reactor kinetics equations with actinide and decay heating as prescribed by Appendix K.

The analysis assumed loss of offsite power concurrent with reactor scram on low pressurizer pressure. The single-failure criterion required by Appendix K was satisfied by assuming the loss of one EDG, which resulted in the disabling of one HHSI pump and a motor-driven AFW pump.

This yields only one HHSI pump providing flow to the RCS. It also results in only one motor driven AFW pump being available. Minimum capability, degraded flow was assumed from the single operable HHSI pump. For Sequoyah, where charging pumps are considered safety grade, charging pump flow is included with the HHSI flow. Initiation of the HHSI system was delayed by 37-seconds beyond the time of SIAS. The 37-second delay represents the time required for diesel generator startup and switching, and the time it takes for the centrifugal charging pump cold leg injection valves to fully open.

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

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 16 The initiation of the motor-driven pump was delayed by 60-seconds. No credit is taken for the turbine-driven AFW pump.

The input model included details of both main steam lines from the SGs to the turbine control valve, including the MSSV inlet piping connected to the main steam lines. For this small area break, the primary and secondary pressure remained coupled at the MSSV actuation pressure setpoint.

Important system parameters and initial conditions used in the analysis are given in Table 3-1.

The HHSI flow used in the analysis is shown in Table 3-2. The reactivity feedback tables (Table 3-3 and Table 3-4) are applicable for the HTP transition core designs.

The axial power shapes for this analysis is shown in Figure 3-4. Figure 3-5 compares the axial power shape at mid-node elevation for hot rod, hot assembly, inner and outer core used in the analysis.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paae 17 Table 3-1 System Parameters and Initial Conditions Parameter Value Reactor Power, MWt 34791 Axial Power Shape Figure 3-4 Radial Peaking Factor (includes uncertainty) 1.7056 Total Power Peaking Factor (includes uncertainty) 2.65 RCS Flow Rate, gpm 378400 Pressurizer Pressure, psia 2249.7 Core Inlet Coolant Temperature, OF 583 Accumulator Pressure, psia 614.7 Accumulator Fluid Temperature, OF 105 Accumulator Water Volume, ft3 1050 SG Tube Plugging Level, % 15 SG Secondary Pressure, psia 877.4 MFW Temperature, 'F 435.8 AFW Temperature, OF 120 Low SG Level AFAS Setpoint for harsh conditions, %NR Span 0 CCI, HHSI, and RHR Fluid Temperature, OF 110 Pressurizer Pressure - Low Reactor Trip Setpoint (RPS), psia 1859.70 Reactor Scram Delay Time on Low Pressurizer Pressure, sec 2 Scram CEA Holding Coil Release Delay Time, sec 0.5 SIAS Activation Setpoint Pressure, psia 1744.7 HHSI Pump Delay Time on SIAS, sec 37 MSSV lift pressures 1064.0 1077.0 1090.0 1103.0 1117.0 1 Includes uncertainty.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 18 Table 3-2 CCI, HHSI, & RHR Flows as Function of the Primary System Pressure CCI Flow HHSI Flow RHR Flow Primary Mass Flow Primary Mass Flow Primary Mass Flow Pressure Rate Pressure Rate Pressure Rate (psia) (ILb s) (psia) (Ibm/s) (psia) (Ib s) 2237 0.00 -

1902 2.20 -

1676 12.50 -

1443 20.80 -

1392 22.30 1392 0.00 1220 27.30 1220 23.70 1038 32.10 1038 37.80 861 36.80 861 48.00 675 41.20 675 56.10 482 45.10 482 62.80 291 49.30 291 69.30 242 50.30 242 70.90 194 51.40 194 72.50 179 51.80 179 73.00 179 0.00 149 52.40 149 73.90 149 246.40 125 53.00 125 74.60 125 331.00 100 53.50 100 75.40 100 393.40 75 54.00 75 76.10 75 446.50 50 54.60 50 76.90 50 494.40 15 55.30 15 78.00 15 550.00 AREVA NP Inc.

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Sequoyah Units I and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 19 Table 3-3 Moderator Density Reactivity Table 3-4 Doppler Reactivity AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paae 20 Figure 3-4 Axial Power Shape AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 21 Figure 3-5 Axial Power Distribution Comparison AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Pace 22 4.0 Analytical Results 4.1 Results The break spectrum should include breaks of varying diameter through the break that represents 10% of the cold leg flow area. Once the preliminary limiting break size (DLIM) has been identified, additional calculations should be run with diameters of DLIM+/-0.25 in, DLIM+0.50 in, and DLIM+/-0.75 in. The expectation is to

1. Cover the entire 0-10% small break area range; and

2. Demonstrate that the actual limiting break size has been identified within 0.25-inch diameter.

The current analysis takes into consideration breaks in the attached piping (accumulator line, charging line).

For the RCP trip sensitivity study, a hot leg break was analyzed in addition to the cold leg breaks. The break spectrum calculations were executed for breaks of 1.00, 2.00, 2.75, 3:00, 3.50, 4.00, 4.50, 4.75, 4.90, 4.95, 5.00, 5.05, 5.10, 5.12, 5.13, 5.14, 5.15, 5.20, 5.25, 5.50, 5.75, 6.00, 6.50, 7.00, 8.00, 8.50, 9.00, 9.75, 9.76, 9.77 and 9.78-inch diameter (the 9.76-inch diameter break corresponds to an area equal to 10% of the cold leg area). The results for the break spectrum calculations are presented in Table 4-1. Predicted event times are summarized in Table 4-21 for the break spectrum calculations.

The break spectrum calculations conservatively assumed RCP trip at reactor trip due to an assumed loss of offsite power at reactor trip. However, a delayed RCP trip can potentially produce more limiting results than tripping the RCPs at reactor trip. This delayed trip sensitivity is discussed in Section 4.3.

The ."'

in Table 4-2 indicates that the loop seals cleared after the accumulator injection occurred.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 23 The key parameters for the limiting case (9.76-inch break) are shown in Figure 4-1 through Figure 4-24. The results of the analysis demonstrate the adequacy of the Emergency Core Cooling System (ECCS) to support the criteria given in 10 CFR 50.46(b) for SEQ1 and SEQ2 operating with AREVA supplied 17x17 HTP M5 clad fuel; as follows:

(1) Peak cladding temperature: The calculated limiting fuel element cladding temperature is 1470°F, which is less than the 22000 F limit criterion.

(2) Maximum local cladding oxidation: The calculated maximum local oxidation of the cladding is 0.17% which is less than the 17% limit of the criterion.

(3) Maximum core-wide oxidation: The calculated core-wide total oxidation is 0.0013%,

which is less than the 1% limit of the criterion.

(4) Coolable geometry: The cladding remains amenable to cooling. None of the cases analyzed predicted hot rod rupture, hence no blockage is predicted to occur, which would degrade core cooling. Further, an M5 mechanical deformation analysis presented in Appendix F of Reference [4] is not changed by the fuel design changes. Both thermal and mechanical deformations of the fuel assemblies in the core have been assessed and the resultant deformations have been shown to maintain coolable core configurations. Therefore, the coolable geometry requirements of the criterion are met.

4.2 Discussion of Transient for 9.76-inch Break The break opens at the initiation of the transient t=0 seconds and initiates a subcooled depressurization of the primary system. The Low Pressurizer Pressure Trip setpoint is reached within 0.4 seconds and within 3 seconds the reactor is tripped, the offsite power is lost, concurrent with the turbine trip, RCP trip, and Main Feedwater Pumps trip. After 7 seconds the Safety Injection Actuation Signal is issued on Low Pressurizer Pressure and at the same time the motor-driven Auxiliary Feedwater Pump actuation signal is issued.

The primary depressurization continues at a relatively high rate for the first approximately 20 seconds, after which it decreases significantly for approximately 60 seconds, while the pressure is relatively flat, and then remains relatively constant through the rest of the transient.

The secondary side pressure increases in all four Steam Generators but not enough to reach the MSSV lift setpoint(s) and it stabilizes and remains almost constant through the rest of the transient (Figure 4-2). The single Auxiliary Feedwater Pump left running by the single-failure assumption starts delivering flow to the Steam Generators 2 and 4 around 66 seconds (Figure 4-9), but it is unable to increase the SG level above the lower boundary of the narrow range level span (Figure 4-11).

The continued depressurization causes the flashing of the liquid and the generation of steam bubbles that tend to accumulate to the high elevation points in the primary, e.g. the bend of the Steam Generator U-tubes. The liquid in the pressurizer drains rapidly and liquid in the core AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 24 starts to boil as the saturation pressure is reached and progressively the nodes in the hot assembly start voiding. As a result, the collapsed liquid level decreases rapidly at the beginning, but as the volume heat source decreases rapidly following the reactor trip, some of the steam bubbles formed initially collapse and the level briefly recovers around 10 seconds into the transient, after which it continues to decrease until the entire hot channel is completely void of liquid around approximately 50 seconds (Figure 4-17). At the same time, the hot rod mixture level remains to the top of the core (Figure 4-18) and remains there until around 45 seconds when it abruptly drops to the bottom. As downcomer level decreases (Figure 4-19), it reaches the point where the cold leg nozzles are uncovered and steam can start venting out to the cold legs (Figure 4-3 and Figure 4-4).

As the core is voided of liquid, the heat transfer conditions worsen (Figure 4-23) and reach the point where the hot rod is no longer properly cooled and starts the adiabatic heatup (Figure 4-24). Meanwhile, both Centrifugal Charging Injection and High Head Safety Injection start delivering fluid to the cold legs at 46 seconds (Figure 4-12), but the liquid injected cannot make up for the large amount of fluid lost to the break up to this point (Figure 4-14) and the cladding temperature keeps rising. As break uncovers around 86 seconds, the amount of liquid lost out the break decreases sharply and steam is vented out the break (Figure 4-3) releasing energy from the primary. The primary system pressure starts decreasing again and the loop seals begin to clear between 90 and 102 seconds (Figure 4-5). Seconds later, the accumulator injection begins (Figure 4-13) and the large amount of liquid added to the primary turns the transient around, liquid levels start rising, first in the downcomer (Figure 4-19), and later into the core (Figure 4-17). Cladding temperature reaches its peak of 1469.3 OF at 143 seconds, and then, as cooling conditions in the core starts improving due to the increased presence of liquid, it drops as core quenches.

4.3 RCP Trip Sensitivity Study A Reactor Coolant Pump trip sensitivity study has been performed as required in Reference [3].

For plants that do not have an automatic RCP trip, a delayed RCP trip can potentially result in a more limiting condition than tripping the RCPs at reactor trip. Continued operation of the RCPs can result in earlier loop seal clearing with associated two-phase flow out the break, which would results in less inventory loss out the break early in the transient. It is possible that tripping the pumps at minimum RCS inventory could cause a collapse of core voids, thus depressing the core level and provoking a deeper core uncovery, and a potentially higher PCT.

The TVA Emergency Operating Procedure No. E-1, "Loss of Reactor or Secondary Coolant" requires the RCPs to be tripped when:

(1) Primary system pressure is less than 1250 psig and at least one charging or one safety injection pump is running, or, when (2) A Phase B isolation signal has been generated on high-high containment pressure (i.e.

containment pressure _> 2.9 psig).

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 25 The nominal setpoint of 1250 psig is based on an instrument uncertainty of +/-111 psi, thus the pump trip could occur at actual primary system pressures as high as 1361 psig or as low as 1139 psig.

The RCP trip sensitivity study will include those breaks for which the break uncovery time from the break spectrum case is less than the RCP trip time determined based on the EOPs. From the break spectrum transient study, this means breaks of 6.5-inch diameter and higher, i.e. 6.50, 7.00, 8.00, 9.00, 9.75, 9.76, 9.77 and 9.78-inch diameter.

4.3.1 Cold Leg Breaks RCP Trip Sensitivity Study For the cold leg breaks RCP trip sensitivity study the selected breaks of 6.50, 7.00, 8.00, 9.00, 9.75, 9.76, 9.77 and 9.78-inch diameter are being re-analyzed with a delayed RCP trip. For these cases the RCP trip time based on EOPs is the same, 166.0 seconds.

A summary of results of the cold leg break RCP trip sensitivity runs are presented in Table 4-3.

The results of the cold leg breaks RCP trip sensitivity study do not impact the limiting PCT of 1470°F from the SBLOCA break spectrum study.

4.3.2 Hot Leg Breaks RCP Trip Sensitivity Study For the hot leg breaks RCP trip sensitivity study the same selected breaks of 6.50, 7.00, 8.00, 9.00, 9.75, 9.76, 9.77 and 9.78 inch diameter are being re-analyzed with a delayed RCP trip and a hot leg break. For these cases the RCP trip time based on EOPs is the same, 166.0 seconds.

A summary of results of the cold leg break RCP trip sensitivity runs are presented in Table 4-4.

The results of the hot leg breaks RCP trip sensitivity study do not impact the limiting PCT of 1470°F from the SBLOCA break spectrum study.

4.4 Attached Piping LOCA Cases Sensitivity calculations for breaks in the attached piping have also been performed. These consist of a double ended guillotine (DEG) break of the accumulator line and a DEG break of the charging line.

For the DEG break of the accumulator, the calculated PCT is 1386.1 OF. The results of the charging line break analysis showed that there is no heat up occurred, which is consistent with the conclusion of the Sequoyah FSAR analysis of this break. The results of these two breaks do not challenge the limiting PCT of 1470°F from the SBLOCA break spectrum study.

4.5 Results Comparison against AOR The results from the current analysis break spectrum were compared to the analysis of record (AOR) as provided in Figure 4-25. In the AOR, the limiting break was the 2.75-inch diameter break, which resulted in a PCT of 1403 'F. In the current analysis, the 2.75-inch diameter break AREVA NP Inc.

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Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 26 resulted in a PCT of 1027 'F. The difference in results can be attributed to several factors of which the most important are enumerated below:

The most important difference comes from the axial power shape used by each of the analyses.

The AOR uses a more conservative axial shape which puts more power in the top half of the core. As a result, the fuel nodes at the top of the core experience a faster liquid dry-out, less cooling and a prolonged adiabatic heatup which drives the PCT higher. Also, the FAH values used are different - the AOR uses a very conservative value of 1.8928 (due to an axial peak fixed at 1.4 by the BAW-10168 (P)(A) methodology), while the current analysis complies with the EM requirements of using the Technical Specifications limit of 1.7056.

There are several important modeling differences that account for the different thermal-hydraulic behavior. The AOR model uses a lumped triple loop for modeling the intact loops, while the current analysis uses distinct loop models for each loop. This accounts for major differences in the system response with respect to steam venting and also plays a role in the loop seal clearing. Also, the AOR model for the break is very different using a "dummy" node connected to the bottom of the pipe, and a valve connected to this "dummy" node, while the S-RELAP5 model uses a trip valve directly connected to the centerline cold leg piping. This modeling difference is responsible for the differences in the vapor void fraction at the break, thus affecting the break uncovery time, the fluid mass lost out the break and the steam and energy venting out the break. This modeling difference plays a role in the significant difference in the primary pressure, which is lower in the current analysis from early into the transient and almost to the end, and, as a result, determines a higher pumped SI flow being delivered in the current analysis vs. the AOR. The differences in the SI flow being delivered and the fluid lost out the break account for the difference in reactor vessel mass and consequently the lower collapsed core liquid level in the AOR, the prolonged and deeper core uncovery and thus the higher PCT result in the AOR.

One other important difference is that the AOR model has been built with the original Steam Generators, while the current analysis uses the replacement Steam Generators. While the replacement Steam Generators are functionally identical to the original ones, there are differences in design that come into play and affect the transient behavior, with consequences on the loop seal clearing and steam venting.

AREVA NP Inc.

A AREVA ANP-2971 (NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 27 Table 4-1 Summary of SBLOCA Break Spectrum Results Break diameter (in) 1.00 2.00 2.20 2.40 2.60 2.75 2.90 3.00 Peak Clad Temperature (°F) 690.2 690.2 976.6 986.9 960.7 1027.0 884.5 1069.1 Time of PCT (sec) 1 1 1369 2856 2487 2004 2011 1832 PCT Elevation (ft) 10.125 10.125 10.625 10.625 11.125 10.625 11.125 11.125 Time of Rupture (sec) N/A N/A N/A N/A N/A N/A N/A N/A Core Wide Oxidation (%) 0.0002 0.0002 0.0002 0.0004 0.0003 0.0004 0.0001 0.0006 Local Maximum Oxidation (%) 0.0004 0.0002 0.0036 0.0163 0.0158 0.0238 0.0055 0.0334 Break diameter (in) 3.25 3.50 3.75 4.00 4.25 4.50 4.75 4.80 Peak Clad Temperature (OF) 945.6 947.2 957.2 758.2 816.2 844.4 1246.9 852.0 Time of PCT (sec) 1580 1320 1035 387 340 294 547 255 PCTElevation (ft) 11.125 11.125 11.125 10.625 10.625 10.625 11.125 10.625 Time of Rupture (sec) N/A N/A N/A N/A N/A N/A N/A N/A Core Wide Oxidation (%) 0.0001 0.0001 0.0001 0.0001 0.0001 0.0001 0.0009 0.0001 Local Maximum Oxidation (%) 0.0068 0.0048 0.0061 0.0002 0.0005 0.0007 0.0686 0.0008 Break diameter (in) 4.85 4.90 4.95 5.00 5.50 6.00 6.50 7.00 Peak Clad Temperature (OF) 868.8 889.3 894.5 904.1 1040.5 1056.1 1095.8 1214.4 Time of PCT (sec) 252 249 244 240 212 180 164 182 PCT Elevation (ft) 10.625 10.625 10.625 10.625 10.625 10.625 10.625 10.625 Time of Rupture (sec) N/A N/A N/A N/A N/A N/A N/A N/A Core Wide Oxidation (%) 0.0001 0.0001 0.0001 0.0001 0.0002 0.0002 0.0002 0.0005 Local Maximum Oxidation (%) 0.0010 0.0013 0.0013 0.0015 0.0082 0.0107 0.0150 0.0474 Break diameter (in) 8.00 9.00 9.75 9.76 9.77 9.78 Peak Clad Temperature (OF) 1367.6 1341.5 1385.7 1469.3 1325.6 1379.4 Time of PCT (sec) 175 149 132 143 139 136 PCT Elevation (ft) 10.625 10.625 10.125 10.625 10.125 10.625 Time of Rupture (sec) N/A N/A N/A N/A N/A N/A Core Wide Oxidation (%) 0.0009 0.0004 0.0007 0.0013 0.0003 0.0006 Local Maximum Oxidation (%) 0.1082 0.0812 0.1145 0.1659 0.0848 0.1184 AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 28 Table 4-2 Sequence of Events for Break Spectrum Calculations, sec CLC L w M iu M w

0 -00 w EL 4) - E CD 0 4 >

E2 0 0 C

.. I 3 46 a) 5.-

> - - U 40 1 Z 130 w 39 1 20 0 N7- 2 3 132 33 0M M 36 E a598

.Xa.

I ) U) E .20

=. w

~E -

c 0

'W 0

O 2.70-018 200 - 76 00 04 269 0- 20F 1.00 0 225 227 240 684 231 - - - - - - 4040 - 1 680 2.00 0 43 46 57 176 50 - 1370 - - 1390 - 3993 - 1 2.20 0 34 36 47 152 40 - 1300 - - 1138 - 3506 - 1369 1210 2.40 0 27 29 39 132 33 - 1038 - - 936 - 2648 - 2856 980 2.60 0 21 23 33 116 27 - 872 - - 880 13450 2268 - 2487 704 3

2.75 0 18 20 30 106 24 - 746 - - 764 2690 1982 - 2004 738 -

2.90 0 15 17 27 64 21 - 694 - - 670 2312 1938 - 2011 680 3.00 0 13 16 25 64 19 - 638 - - 648 1878 1664 - 1832 624 3.25 0 10 13 22 60 16 1 566 - - 550 1574 1578 - 1580 544 3.50 0 8 10 19 56 14 486 - - 498 1262 1252 - 1320 276 3.75 0 6 9 17 54 12

• 412 - - 422 1020 1026 - 1035 382 4.00 0 5 7 15 54 11 386 - 384 - 394 914 922 - 387 336 -

4.25 0 4 7 14 52 10 336 336 334 - 346 762 814 - 340 176 -

4.50 0 0.8 3 12 50 7 288 290 290 - 300 652 700 - 294 168 4.75 0 0.7 3 12 50 7 2 - 250 264 514 518 - 547 130 4.80 0 0.7 3 11 50 7 250 250 250 - 260 550 602 - 255 128 4.85 0 0.7 3 11 50 7 1 246 246 - 256 532 540 - 252 126 4.90 0 0.7 3 11 50 7 1 244 1244 252 520 534 - 249 126 4.95 0 0.7 3 11 50 7 240 20 240 4 248 508 554 - 244 124 5.00 0 0.7 3 11 48 7 236 234 234 - 248 492 538 - 240 122 3 5.50 0 0.6 3 10 48 7 198 202 200

• 212 392 428 - 212 106 -

6.00 0 0.6 3 9 48 7 1698166 168 108 328 338 - 180 92 4 6.50 0 0.6 3 9 46 7 146 146 146 152 160 276 298 - 164 82 7.00 0 0.5 3 8 46 7 130 128 128 142 138 234 260 - 182 78 -

8.00 0 0.5 3 8 46 7 126 126 126 136 132 166 118 - 175 56 308 9.00 0 0.5 3 7 46 7 98 98 98 108 90 130 142 - 149 46 244 9.75 0 0.4 3 7 46 7 92 92 92 104 86 110 120 - 132 40 214 9.76 0 0.4 3 7 46 7 90 92 92 102 86 110 130 - 143 40 220 9,77 0 0.4 3 7 46 7 90 98 98 106 92 112 114 - 139 40 208 9.78 0 0.4 3 7 46 7 98 98 98 106 92 112 114 - 136 40 210 AREVA NP Inc.

A ARE VA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 29 Table 4-3 Summary of Results for the Cold Leg Break RCP Trip Sensitivity Break diameter (in) 6.50 7.00 8.00 9.00 9.75 9.76 9.77 9.78 Peak Clad Temperature (°F) 984.2 895.2 872.7 690.2 690.2 690.2 690.2 690.2 Time of PCT (sec) 244 247 209 1 1 1 1 1 PCT Elevation (ft) 10.625 10.625 10.625 10.125 10.125 10.125 10.125 10.125 Time of Rupture (sec) - - - - - - - -

Core Wide Oxidation (%) 0.0001 0.0001 0.0001 0.0000 0.0000 0.0000 0.0000 0.0000 Local Maximum Oxidation (%) 0.0048 0.0018 0.0011 0.0000 0.0000 0.0000 0.0000 0.0000 Table 4-4 Summary of Results for the Hot Leg Break RCP Trip Sensitivity Study Break diameter (in) 6.50 7.00 8.00 9.00 9.75 9.76 9.77 9.78 Peak Clad Temperature (°F) 807.1 777.7 878.7 860.3 690.2 690.2 720.6 781.1 Time of PCT (sec) 349 210 216 197 1 1 167 163 PCT Elevation (ft) 10.625 10.625 10.625 10.125 10.125 10.125 10.125 10.125 Time of Rupture (sec) - - - - - - - -

Core Wide Oxidation (%) 0.0000 0.0000 0.0001 0.0000 0.0000 0.0000 0.0000 0.0000 Local Maximum Oxidation (%) 0.0004 0.0003 0.0011 0.0008 0.0001 0.0001 0.0001 0.0003 AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 30 Reactor Power 4000.0 3000.0 2000.0 1000.0 0.0 0.0 Time (s)

Figure 4-1 Reactor Power - 9.76-inch Break AREVA NP Inc.

A AREVA ANP-2971 (NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 31 System Pressures 3000.0 2000.0 E

1000.0 0.0 1 1 1 1 0.0 100.0 200.0 300.0 400,0 500.0 Time (s)

Figure 4-2 Primary and Secondary System Pressures - 9.76-inch Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis PaQe 32 Liquid&Vapor Break Flow 10000.0 8000.0 6000.0 2

4000.0 2000.0 Time (s)

Figure 4-3 Liquid and Vapor Break Flow - 9.76-inch Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 33 Break Vapor Void Fraction 1.0 a

Time (s)

Figure 4-4 Vapor Void Fraction at the Break - 9.76 in Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 34 Loop Seal Void Fraction 1.0 0.8

-Uvoidg-1 30070000 4voidg-230070000 A voidg-330070000

-A--- voidg-430070000 0.6 0.4 0.2 -

/^

0.

100.0 200.0 300.0 400.0 500.0 Time (s)

ID- - 1 D5,35 51-UX Figure 4-5 Loop Seal Void Fraction - 9.76-inch Break AREVA NP Inc.

A AREVA ANP-2971 (NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 35 Total Core Inlet Mass Flow Rate 40000.0 30000.0 20000.0 E.

10000.0 0.0

-10000.0 500.0 Time (s)

Figure 4-6 Total Core Inlet Mass Flow Rate - 9.76-inch Break AREVA NP Inc.

A AR EVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paae 36 RCS Loop Flow Rates 15000.0 0* 5000.0 50K Time (s)

Figure 4-7 RCS Loop Flow Rate - 9.76-inch Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 37 SG MFW Flow Rates 1500.0 Loop 1 1000.0 . Loop 2

• Loop 3 A- Loop 4 500.0 0.0 1 -.- t - 30A 300.0

• 4, ' * .4 0.0 100.0 200.0 400.0 500.0 Time (s)

Figure 4-8 Steam Generator Main Feedwater Flow Rates - 9.76-inch Break AREVA NP Inc.

A AREVA ANP-2971 (NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paqe 38 Auxiliary Feedwater Flow 6.0

-w4

• 4 -*- -- 4

-Umflowj-503000000 4.0 P -. mflov4-603000000 A mflowj-703000000

-4 mflov4-803000000 2.0 0.- 100A 1, A .

A 40.

0.0 100.0 200.0 300.0 500.0 Time (s)

Figure 4-9 Auxiliary Feedwater Flow - 9.76 in Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Pa~qe 39 SG Total Mass 110000.0 n%

Time (s)

Figure 4-10 Steam Generator Total Mass - 9.76-inch Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analvsis Paae 40 SG Level 50.0 40.0 I -- a cntrlvar-5101

• • crtrvar-6 101 A cntrlvar-7101 30.0 .4 cntrlvar-81 01 20.0 10.0 0.0 IU * '4 - .0) 0'.0 100.0 200.0 300.0 400.0 500.0 Time (s)

Figure 4-11 Steam Generator Level - 9.76 in Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 41 Total SI Flow 80.0 60.0 0E 40.0 u8.

20.0 500.0 ID -m-h , 105 51DEMUX Time (s)

Figure 4-12 Total HHSI Mass Flow - 9.76-inch Break AREVA NP Inc.

A AR EVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 42 Total Accumulator Flow 4000.0 3000.0 2000.0 S

1000.0 0.0O

-1000.0 I 0.0 100.0 200.0 300.0 400.0 500.0 Time (s)

Figure 4-13 Total Accumulator Flow - 9.76-inch Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 43 Integrated Break Flow and Total ECC Flow 800000.0 600000.0 400000.0 2 0 200000.0

-200000.0I I ,

0.0 100.0 200.0 300.0 400.0 500.0 Time (s)

Figure 4-14 Integrated Break Flow and ECC Flow - 9.76 in Break AREVA NP Inc.

A AR EVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 44 Total Primary Mass 500000.0 400000.0 300000.0

.2 200000.0 100000.0 Time (s)

Figure 4-15 Total Primary Mass - 9.76-inch Break AREVA NP Inc.

A AR EVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 45 RV Mass 300000.0 250000.0 200000.0 150000.0 100000.0 50000.0 Time (s)

Figure 4-16 Reactor Vessel Mass - 9.76-inch Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 46 Hot Assembly Collapsed Level 15.0

.0 Time (s)

Figure 4-17 Hot Assembly Collapsed Liquid Level - 9.76-inch Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paae 47 Hot Rod Mixture Level 15.0 2=

Time (s)

Figure 4-18 Hot Rod Mixture Level - 9.76 in Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paae 48 Downcomer Level 30.0 20.0 0L)

Time (s)

Figure 4-19 Downcomer Collapsed Liquid Level - 9.76 in Break AREVA NP Inc.

A ARE VA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 49 Volume Liquid Temperature at PCT Location 700.0 2 500.0 E

400.0 0

Time (s)

Figure 4-20 Volume Liquid Temperature at PCT Location - 9.76 in Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paqe 50 Liquid Void Fraction Distribution in the Hot Channel 0.8 ii

'ft elev.

5ft plev.

-- 3 .7 5 ft eley.

2 0.6 5 ft elev.

P

• 4.7!5 ft elev.

5.7:5 ft elev.

ii 6.7;5 ft elev.

e*7.7!5 ft eAv.

0 0

0.2 0.0 0.0 Time (s)

Figure 4-21 Liquid Void Fraction Distribution in the Hot Channel 0.75 - 7.75 ft Elevation - 9.76 in Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 51 Liquid Void Fraction Distribution in the Hot Channel 1.0:

0.8 2 0.6

.2 I--

O0.4 0.2 500.0 Time (s)

Figure 4-22 Liquid Void Fraction Distribution in the Hot Channel 8.25 - 11.75 ft Elevation - 9.76 in Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 52 Location at PCT Coefficient Transfer Heat Heat Transfer Coefficient at PCT Location 4.0 3.0

-Uhthtc-20102901 4I .

C-,

I-1.0 fl-0.

0.0 100.0 200.0 300.0 400.0 500.0 Time (s)

Figure 4-23 Heat Transfer Coefficient at PCT Location - 9.76-inch Break AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Paae 53 Peak Cladding Temperature 1500.0 1000.0 E

500.0 I 2-.,1o

-55.) 0535)51DEMUX Time (s)

Figure 4-24 Peak Cladding Temperature at PCT Location - 9.76-inch Break AREVA NP Inc.

A AR EVA ANP-2971 (NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 54 Peak Cladding Temperature Comparison 1400.0 1200.0 1000.0 E

800.0 600.0 400.0 0.0 1000.0 2000.0 3000.0 4000.0 Time (s)

Figure 4-25 Comparison of PCTs - AOR vs. S-RELAP5 Analysis AREVA NP Inc.

A AREVA ANP-2971(NP)

Sequoyah Units 1 and 2 HTP Fuel Revision 1 S-RELAP5 Small Break LOCA Analysis Page 55 5.0 References

1. AREVA NP Document EMF-2328(P)(A) Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2001.

2. AREVA NP Document BAW-10240(P)(A) Revision 0, Incorporationof M5 Propertiesin FramatomeANP Approved Methods, May 2004.

3. AREVA NP Document 51-9012330-001, Guidelines for PWR Safety Analysis: Small Break LOCA Analysis Using S-RELAP5, July 2007.

4. Topical Report BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and StructuralMaterial (M5) in PWR ReactorFuel,"AREVA NP (June 2003).

AREVA NP Inc.