L-2015-259, ANP-3440NP, Revision 1, Response to NRC Questions SRXB-RAI-1 and Snpb RAI-2 Thru Snpb RAI-20.

ANP-3440NP, Revision 1, Response to NRC Questions SRXB-RAI-1 and Snpb RAI-2 Thru Snpb RAI-20.

ML15279A227
Person / Time
Site:	Saint Lucie
Issue date:	10/02/2015
From:	AREVA
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15279A222	List: L-2015-259, ANP-3440NP, Revision 1, Response to NRC Questions SRXB-RAI-1 and Snpb RAI-2 Thru Snpb RAI-20. ML15279A226
References
L-2015-259, TAC MF5494, TAC MF5495 ANP-3440NP, Rev. 1
Download: ML15279A227 (111)

Text

L-2015-259 Attachment 4 AREVA non-proprietary report AN P-3440N P, Revision 1, Response to NRC Questions SRXB-RAI-1 and SNPB RAI-2 thru SNPB RAI-20 Next 110 Pages

~L[ ~JEF~~J ~d~1 FE~F IL AR EVA AN P-3440N P St. Lucie Unit 2 Fuel Transition: Revision Response to NRC Questions SRXB-RAI-I and SNPB RAI-2 thru SNPB RAI-20 Technical Report September 2015 AREVA Inc.

(c) 2015 AREVA Inc.

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page Copyright © 2015 AREVA Inc.

All Rights Reserved

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ANP-3440NP Revision 1 Page ii TecniclSepotins Nature of Changes Item Section(s) or Page(s) Description and Justification 1 Page 2-30 Removed brackets on the co-resident Lower Tie Plate diagonal dimension in Table 2-7.

2 Page 2-32 Added footnote relative to the Guardian TM grid height dimension in Table 2-7.

3 Page 2-32 Changed statement from "max spacer" to "min spacer" and removed brackets on the co-resident design spacer grid diagonal dimension in Table 2-7.

4 Page 2-33 Changed Center Post OD dimension for co-resident design and corresponding difference evaluation in Table 2-7.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAi-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page iii Contents 1.0 Introduction ................................................................................... 1-1 2.0 Responses to Information Request.......................................................... 2-1 2.1 SRXB-RAI-1........................................................................... 2-1 2.2 SNPB RAI-2........................................................................... 2-9

- 2.2.1 Upper Tie Plate Compatibility............................................ 2-11

- 2.2.2 Lower Tie Plate Compatibility............................................ 2-12

- 2.2.3 Corner and Center Guide Tubes ........................................ 2-13

- 2.2.4 Fuel Assembly ............................................................ 2-13 2.2.4.1 LTP Elevation ................................................. 2-14 2.2.4.2 Fuel Rod Elevations........................................... 2-14 2.2.4.3 Spacer Grid Elevations and Overlap ........................ 2-14 2.2.4.4 Envelope Evaluations ........................................ 2-15 2.3 SNPB RAI-3 ......................................................................... 2-16 2.4 SNPB RAI-4 ......................................................................... 2-17

- 2.4.1 SNPB RAI-4, Sub-item (a):............................................... 2-17 2.4.1.1 Mixed Core Flow Distribution Effects ........................ 2-17 2.4.1.2 Mixed Core DNB Performance............................... 2-19

- 2.4.2 SNPB RAI-4, Sub-item (b):............................................... 2-21 2.4.2.1 Upper Tie Plate (UTP)........................................ 2-21 2.4.2.2 Lower Tie Plate (LTP) ........................................ 2-22 2.4.2.3 Center and Corner Guide Tubes............................. 2-23 2.5 SNPB RAI-5 ......................................................................... 2-25

- 2.5.1 Non-LOCA Analyses Response ......................................... 2-25 2.5.1.1 Thermal Hydraulics Analysis ................................. 2-25 2.5.1.2 Fuel Rod Performance Analysis ............................. 2-25

- 2.5.2 LOCA Analyses Response............................................... 2-26 2.6 SNPB RAI-6 ......................................................................... 2-28 2.7 SNPB RAI-7 ......................................................................... 2-35 2.8 SNPB RAI-8 ......................................................................... 2-36

- 2.8.1 SNPB RAI-8, Sub-item (a):............................................... 2-36

- 2.8.2 SNPB RAI-8, Sub-item (b):............................................... 2-37 2.9 SNPB RAI-9 ......................................................................... 2-38 2.10 SNPB RAI-10........................................................................ 2-39 2.11 SNPB RAI-11........................................................................ 2-40 2.12 SNPB RAI-12........................................................................ 2-47 2.13 SNPB RAI-13........................................................................ 2-52 2.14 SNPB RAI-14........................................................................ 2-54 2.15 SNPB RAI-15........................................................................ 2-79 2.16 SNPB RAI-16........................................................................ 2-83 2.17 SNPB RAI-17........................................................................ 2-87 2.18 SNPB RAI-18........................................................................ 2-91 2.19 SNPB RAI-19........................................................................ 2-92 2.20 SNPB RAI-20........................................................................ 2-94 3.0 References .................................................................................... 3-1

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page iv Tables Table 2-1 : Key In put and Output Parameters for all 59 Cases (1 of 2) ........................... 2-3 Table 2-2: Key In put and Output Parameters for all 59 Cases (2 of 2) ........................... 2-6 Table 2-3: Component Interfaces................................................................... 2-10 Table 2-4: Resident Westinghouse Assembly Flow Fraction Relative to a Core with All Resident Westinghouse Assemblies .............................................. 2-18 Table 2-5: AREVA Assembly Flow Fraction Relative to a Core with All AREVA Assemblies................................................................................... 2-18 Table 2-6: Fuel Rod Performance TCD Factors Comparison .................................... 2-26 Table 2-7: Configuration Comparison for SLU2 (Nominal Design Dimensions) ................ 2-29 Table 2-8: Manufacturer's Pump Performance Data.............................................. 2-41 Table 2-9: 1ST Criteria and Surveillance Data..................................................... 2-41 Table 2-10: IST Pump Performance Data ......................................................... 2-42 Table 2-11: 2A HPSI Minimum Pump delivery..................................................... 2-44 Table 2-12: SBLOCA Analysis HPSI Pump Flow Delivery........................................ 2-45 Table 2-13: Summary of Results - Updated Charging Flow Configuration ..................... 2-82 Table 2-14: Sequence of Events - Updated Charging Flow Configuration ..................... 2-82 Table 2-15: Reactor Vessel Upper Head Pressure - 2.70 inch break case..................... 2-88 Table 2-16: Moderator Density Feedback ......................................................... 2-91 Table 2-17: PCT Time............................................................................... 2-94

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page v Figures Figure 2-1: UTP I FAP Interface.................................................................... 2-12 Figure 2-2: Pressure Drop Profile Comparison.................................................... 2-19 Figure 2-3: Mixed Core DNB Performance ........................................................ 2-20 Figure 2-4: Cross-Section Illustration of LTP Corner Post........................................ 2-23 Figure 2-5: Manufacturers Pump Performance vs 1ST Pump Performance .................... 2-43 Figure 2-6: SBLOCA Analysis Flow vs IST Performance Flow .................................. 2-46 Figure 2-7: Example Only - No Bias............................................................... 2-48 Figure 2-8: Example Only - Bias on Intact Legs .................................................. 2-49 Figure 2-9: Loop Seal Void Fraction - 2.70-in Break .............................................. 2-50 Figure 2-10: Loop Seal Steam Velocity - 2.70-in Break........................................... 2-51 Figure 2-1 1: Steam Generator Level % - 2.70-in Break .......................................... 2-53 Figure 2-12: Reactor Power - 2.60 inch Break.................................................... 2-54 Figure 2-13: Primary and Secondary System Pressures - 2.60 inch Break..................... 2-55 Figure 2-14: Break Void Fraction - 2.60 inch Break .............................................. 2-55 Figure 2-15: Break Mass Flow Rate - 2.60 inch Break ........................................... 2-56 Figure 2-16: Loop Seal Void Fraction - 2.60 inch Break ......................................... 2-56 Figure 2-17: RCS Loop Mass Flow Rate - 2.60 inch Break ...................................... 2-57 Figure 2-18: Main Feedwater Mass Flow Rate - 2.60 inch Break............................... 2-57 Figure 2-19: Auxiliary Feedwater Mass Flow Rate - 2.60 inch Break ........................... 2-58 Figure 2-20: Steam Generator Total Mass - 2.60 inch Break.................................... 2-58 Figure 2-21: High Pressure Safety Injection Mass Flow Rates - 2.60 inch Break.............. 2-59 Figure 2-22: Low Pressure Safety Injection Mass Flow Rates - 2.60 inch Break .............. 2-59 Figure 2-23: Safety Injection Tank Mass Flow Rates - 2.60 inch Break ........................ 2-60 Figure 2-24: Reactor Vessel Mass Inventory -2.60 inch Break ................................. 2-60 Figure 2-25: Hot Assembly Mixture Level - 2.60 inch Break..................................... 2-61 Figure 2-26: Hot Spot Cladding Temperature - 2.60 inch Break ................................ 2-61 Figure 2-27: Charging Flow - 2.60 inch Break .................................................... 2-62 Figure 2-28: Reactor Power - 2.70 inch Break.................................................... 2-62 Figure 2-29: Primary and Secondary System Pressures - 2.70 inch Break..................... 2-63 Figure 2-30: Break Void Fraction - 2.70 inch Break .............................................. 2-63 Figure 2-31: Break Mass Flow Rate - 2.70 inch Break ........................................... 2-64

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page vi Figures (continued)

Figure 2-32: Loop Seal Void Fraction - 2.70 inch Break ......................................... 2-64 Figure 2-33: RCS Loop Mass Flow Rate - 2.70 inch Break ...................................... 2-65 Figure 2-34: Main Feedwater Mass Flow Rate - 2.70 inch Break............................... 2-65 Figure 2-35: Auxiliary Feedwater Mass Flow Rate - 2.70 inch Break ........................... 2-66 Figure 2-36: Steam Generator Total Mass - 2.70 inch Break.................................... 2-66 Figure 2-37: High Pressure Safety Injection Mass Flow Rates - 2.70 inch Break.............. 2-67 Figure 2-38: Low Pressure Safety Injection Mass Flow Rates - 2.70 inch Break .............. 2-67 Figure 2-39: Safety Injection Tank Mass Flow Rates - 2.70 inch Break ........................ 2-68 Figure 2-40: Reactor Vessel Mass Inventory - 2.70 inch Break ................................. 2-68 Figure 2-41: Hot Assembly Mixture Level - 2.70 inch Break..................................... 2-69 Figure 2-42: Hot Spot Cladding Temperature - 2.70 inch Break ................................ 2-69 Figure 2-43: Charging Flow - 2.70 inch Break .................................................... 2-70 Figure 2-44: Reactor Power - 2.80 inch Break.................................................... 2-70 Figure 2-45: Primary and Secondary System Pressures - 2.80 inch Break..................... 2-71 Figure 2-46: Break Void Fraction - 2.80 inch Break .............................................. 2-71 Figure 2-47: Break Mass Flow Rate - 2.80 inch Break ........................................... 2-72 Figure 2-48: Loop Seal Void Fraction - 2.80 inch Break ......................................... 2-72 Figure 2-49: RCS Loop Mass Flow Rate - 2.80 inch Break ...................................... 2-73 Figure 2-50: Main Feedwater Mass Flow Rate - 2.80 inch Break............................... 2-73 Figure 2-51: Auxiliary Feedwater Mass Flow Rate - 2.80 inch Break ........................... 2-74 Figure 2-52: Steam Generator Total Mass - 2.80 inch Break.................................... 2-74 Figure 2-53: High Pressure Safety Injection Mass Flow Rates - 2.80 inch Break.............. 2-75 Figure 2-54: Low Pressure Safety Injection Mass Flow Rates - 2.80 inch Break .............. 2-75 Figure 2-55: Safety Injection Tank Mass Flow Rates - 2.80 inch Break ........................ 2-76 Figure 2-56: Reactor Vessel Mass Inventory - 2.80 inch Break ................................. 2-76 Figure 2-57: Hot Assembly Mixture Level - 2.80 inch Break..................................... 2-77 Figure 2-58: Hot Spot Cladding Temperature - 2.80 inch Break ................................ 2-77 Figure 2-59: Charging Flow - 2.80 inch Break .................................................... 2-78 Figure 2-60:S5-RELAP5 SBLOCA Reactor Coolant System Nodalization....................... 2-80 Figure 2-61: PCT vs. Break Diameter.............................................................. 2-81 Figure 2-62: Heat Transfer Coefficient - 2.60-in Break ........................................... 2-83

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page vii Figures (continued)

Figure 2-63: Hot Spot Vapor Temperature - 2.60-in Break....................................... 2-84 Figure 2-64: Heat Transfer Coefficient - 2.70-in Break ........................................... 2-84 Figure 2-65: Hot Spot Vapor Temperature - 2.70-in Break....................................... 2-85 Figure 2-66: Heat Transfer Coefficient - 2.80-in Break ........................................... 2-85 Figure 2-67: Hot Spot Vapor Temperature - 2.80-in Break....................................... 2-86 Figure 2-68: PCT and Integrated SIT Flow - 2.50-in Break ...................................... 2-88 Figure 2-69: PCT and Integrated SIT Flow - 2.60-in Break ...................................... 2-89 Figure 2-70: PCT and Integrated SIT Flow- 2.70-in Break ...... ............................... 2-90 Figure 2-71: Axial Core Void Distribution at PCT Time - 2.60-in Break.......................... 2-95 Figure 2-72: Axial Core Void Distribution at PCT Time - 2.70-in Break.......................... 2-96 Figure 2-73: Axial Core Void Distribution at PCT Time - 2.80-in Break.......................... 2-97

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page viii Nomenclature AFW ............................... Auxiliary Feedwater AQOO............................... Anticipated Operational Occurrence ASI ................................. Axial Shape Index BOC ............................... Beginning of Cycle BOL ................................ Beginning of Life CE ................................. Combustion Engineering CEA................................ Control Element Assembly CEOG.............................. Combustion Engineering Owners Group CFR................................ Code of Federal Regulations DNB................................ Departure from Nucleate Boiling ECCS.............................. Emergency Core Cooling System EDG ............................... Emergency Diesel Generator EOC ............................... End of Cycle EOL ................................ End of Life EOP................................ Emergency Operating Plan EPU ................................ Extended Power Uprate FAP ................................ Fuel Alignment Plate FCM ............................... Fuel Centerline Melt FPL................................. Florida Power & Light GDC ............................... General Design Criteria gpm ................................ Gallon per Minute GT ................................. Guide Tube GWd/MTU......................... Gigawatt-days per Metric Ton of Uranium HMPTM ............................. High Mechanical Performance HPSI ............................... High Pressure Safety Injection HTC................................ Heat Transfer Coefficient HTPTM ............................. High Thermal Performance ID................................... Inner Diameter IN................................... Information Notice IST ................................. In-service Testing LAR ................................ License Amendment Request LBLOCA........................... Large Break Loss-of-Coolant Accident LCSP............................... Lower Core Support Plate LHGR.............................. Linear Heat Generation Rate LOCA.............................. Loss-of-Coolant Accident LOOP.............................. Loss-of-offsite-power LPSI ............................... Low Pressure Safety Injection LTA................................. Lead Test Assembly LTP................................. Lower Tie Plate

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page ix Nomenclature (continued) mils................................. equivalent to 0.001 inches MW................................. Megawatt MWd/kgU.......................... Megawatt-days per Kilograms of Uranium NRC ............................... Nuclear Regulatory Commission 00.................................. Outer Diameter PCT ................................ Peak Cladding Temperature PLHGR ............................ Peak Linear Heat Generation Rate PWR ..................... .Pressurized Water Reactor RAI................................. Request for Additional Information RCP ................................ Reactor Coolant Pump RCS ................................ Reactor Coolant System RLBLOCA......................... Realistic Large Break LOCA RWST.............................. Refueling Water Storage Tank SBLOCA........................... Small Break Loss-of-Coolant Accident SG.................................. Steam Generator SIT ................................. Safety Injection Tank SL-2................................ St. Lucie Unit 2 SNPB .............................. Nuclear Performance and Code Branch (USNRC)

SRXB .............................. Reactor Systems Branch (USNRC)

TCD................................ Thermal Conductivity Degradation TDH ................................ Total Developed Head TH .................................. Thermal-Hydraulic USNRC............................ United States Nuclear Regulatory Commission UTP ................................ Upper Tie Plate W................................... Westinghouse

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 1-1 1.10 Introduction The US Nuclear Regulatory Commission (NRC) staff requested additional information to support the review of the License Amendment Request ([AR) and exemption for M5 fuel rod cladding regarding St. Lucie Unit 2 transitioning to AREVA fuel as submitted by Florida Power & Light (FPL) (Reference 4). The Reactor Systems Branch (SRXB) and the Nuclear Performance and Code Branch (SNPB) at the NRC have provided a total of twenty (20) Requests for Additional Information (RAls) that will be addressed within this document (SRXB-RAI-1 and SNPB RAI-2 thru SNPB RAI-20). The RAls from the NRC can be found in Reference 1.

Two issues have been identified which affect the responses to SNPB RAI-9 and SNPB RAI-1 5.

1. For SNPB RAl-9, [

This issue has the potential to affect the response to SNPB RAI-9. Therefore, this response is not included in this submittal.

2. For SNPB RAI-15, the Small Break LOCA (SBLOCA) results reported in ANP-3345P (Reference 9) are based on a charging flow modeled with all the flow from one charging pump going into the intact loops rather than using a conservative flow split with 60% of the flow going to the broken loop. The assumed charging flow injection was corrected. The licensing results and associated tables and figures reported in the RAI responses related to SBLOCA have been updated to reflect this more conservative charging flow split.

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AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-1 2.0 Responses to Information Requests 2.1 SRX(B-RAI-1 Section 50.46 of 10 CER requires that ECCS performance be calculated in accordance with an acceptable evaluation model, and uncertainty must be accounted for such that there is a high level of probability that the criteria would not be exceeded. The licensee provided ANP-3346P, Revision 0, "St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA (RLBLOCA) Summary Report," which does not contain the information needed by the staff to confirm that the AREVA RLBLOCA methodology has been implemented as described in TR EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors."

Please provide, preferably in a table, the statistically sampled conditions, input data, etc. for the following parameters for all of the sampled break sizes:

Peak Clad Temperature Case number (PCT) Case end time PCT elevation Hot rod Assembly burnup Core power PLHGR (Peak Linear Heat Axial skew Generation Rate)

One sided break size Axial shape index (ASI) Break type Minimum temperature (T min) Decay heat multiplier Initial stored energy Film boiling heat transfer Dispersed flow film boiling Condensation interphase HTC coefficient (HTC) HTC Initial RCS flow rate Initial T cold Pressurizer pressure Safety Injection Tank (SIT)

Pressurizer level SIT pressure temp Start of broken loop SIT Start of intact loop SIT SIT liquid volume injection injection

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-2 Start of High Pressure Safety Broken loop SIT empty time Intact loop SIT empty time Injection (HPSI) flow Start of Low Pressure Safety End of refill time/start of Beginning of refill time Injection (LPSI) flow ref lood Time of annulus downflow/end Containment pressure at time Beginning of bypass time of bypass of PCT Refueling Water Storage Tank Containment volume (RWST) temp

Response

The requested sampled parameter values are provided in Table 2-1 and Table 2-2 for all 59 cases of the loss-of-offsite-power (LOOP) case set. The limiting case set was determined by running the uncertainty analysis with LOOP and No-LOOP conditions. The case set with highest Peak Cladding Temperature (PCT) is considered limiting and for this analysis the LOOP case was determined to be limiting, It should be noted that the [

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-3

- Table 2-1: Key Input and Output Parameters for all 59 Cases (1 of 2)

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E &w 1 1641 7 500.0 2 1469 6.4 400.0 3 1520 8.9 262.2 4 1444 7.3 298.8 5 1209 36.6 400.0 6 1572 8.3 346.6 7 1494 5.7 400.0 8 1312 14.7 400.0 9 1282 33.2 225.7 10 1471 7.7 315.6 11 1028 94 200.0 ____

12 1542 7.9 400.0 13 1385 9.2 320.8 14 1491 8 400.0 ____

15 1496 31.9 322.8 16 1525 7.7 400.0 17 1418 31.6 263.6 18 1423 5.8 400.0 19 1519 6.8 400.0 20 1609 9.2 399.7 21 1465 30.7 381.5 22 1512 7.9 400.0 23 1422 5.5 400.5 24 1613 7.2 466.6 25 1323 32.1 292.2

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision I Technical Report Page 2-4 Tbe2-1: Key Input and Output Parameters for all 59 Cases (1 of 2) (continued)

A o E z I- E 26 1476 10.1 400.0 27 1399 6.1 400.0 28 1456 4.9 400.0 29 1571 8.8 293.0 30 1496 7.5 400.0 31 1541 7.8 400.0 32 1485 8 400.0 33 1592 6.1 400.0 34 1102 85.8 200.0 35 1232 33.5 240.4 36 1549 6.2 400.0 37 1531 8.6 326.3 38 1645 8.2 415.9 39 1544 9.2 310.0 40 1601 9.7 414.8 41 1596 7.5 400.0 42 1457 9 360.7 43 1332 30.5 271.6 44 1518 8.3 383.5 45 1503 9.9 400.0 46 1540 8.5 341.4 47 1461 6.6 268.7 48 1328 36.5 400.0 49 1497 7.1 400.0 50 1636 8.8 500.0 51 1360 6.5 400.0

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-I ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-5

• Table 2-1: Key Input and Output Parameters for all 59 Cases (1 of 2) (continued)

E o-z *- E mw 52 1572 7.2 449.8 53 1226 32.4 200.0 54 1396 7 403.6 55 1612 8.9 500.0 56 1568 7.8 400.0 57 1732 105.6 551.9 58 1506 7.9 400.0 ____ ____

59 1587 6.3 400.0 _____

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-6 Table 2-2: Ke, )~ut and Output Parameters for all 59 Cases (2 of 2) 0)

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1 11.66 12.66 90.84 90.04 31.00 31.00 12.68 25.90 12.68 25.90 343 2 8.99 10.31 65.13 65.93 30.98 30.98 10.33 26.52 10.33 26.52 34.68 3 6.78 11.93 74.16 77.08 30.96 30.96 11.96 26.86 11.96 26.86 36.91 4 5.99 9.65 73.88 75.65 30.85 30.85 9.69 25.57 9.69 25.57 37.14 5 20.68 20.99 74.08 75.36 31.54 31.54 21.01 37.68 21.01 37.68 0.00 6 6.06 10.96 81.47 83.64 31.03 31.03 11.00 26.82 11.00 26.82 35.75 7 9.40 10.85 87.13 86.72 30.97 30.96 10.85 25.39 10.85 25.39 34.37 8 20.71 20.96 78.75 80.77 31.49 31.49 20.98 34.90 20.98 34.90 0.00 9 15.27 18.15 76.76 80.45 31.29 31.29 18.21 33.64 18.21 33.64 0.00 10 7.77 11.61 68.73 71.08 30.96 30.96 11.63 25.26 11.63 25.26 36.32 11 18.24 21.63 94.33 98.93 31.73 31.73 21.67 43.08 21.67 43.08 0.00 12 16.44 17.05 71.81 74.24 31.14 31.14 17.03 31.42 17.03 31.42 33.47 _____

13 8.42 13.47 78.52 81.87 30.98 30.98 13.51 25.46 13.51 25.46 37.68 14 17.54 17.92 95.67 94.96 31.32 31.32 17.94 32.79 17.94 32.79 32.45 _____

15 8.13 14.46 81.63 86.18 31.37 31.37 14.50 32.21 14.50 32.21 0.00 16 11.81 12.69 73.82 72.47 30.96 30.96 12.72 26.73 12.72 26.73 35.46 17 14.59 18.11 87.83 90.67 31.23 31.22 18.15 32.82 18.15 32.82 0.00 18 9.14 10.76 100.96 98.76 31.00 31.00 10.76 26.04 10.76 26.04 34.47 19 10.99 12.01 71.31 69.33 30.91 30.91 12.02 24.92 12.02 24.92 36.00 20 7.01 12.20 78.28 81.25 31.11 31.11 12.15 26.73 12.15 26.73 36.35 21 10.62 15.35 81.32 85.82 31.36 31.36 15.45 31.83 15.45 31.83 0.00 22 10.85 11.86 72.59 73.10 30.96 30.96 11.87 26.82 11.87 26.82 34.91 _____

23 7.82 9.36 70.38 70.11 31.03 31.03 9.38 25.22 9.38 25.22 33.72 24 6.88 11.54 66.88 68.91 30.97 30.96 11.56 26.64 11.56 26.64 35.67 25 10.33 14.99 78.27 81.18 31.28 31.28 15.02 32.46 15.02 32.46 0.00 26 19.53 19.85 92.03 94.23 31.33 31.33 19.87 32.23 19.87 32.23 0.00 _____

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-7 Table 2-2: Key Input and Output Parameters for all 59 Cases (2 of 2) (continued) ____

z W '* ".- '4---4 ..

,,2 ,- I.,. i,,1,.iE 271038 1174 808 78.3 0.9 ,9 11.7 26.01 117 2601 341 27 10.38 11.74 80.81 78.913 30.96 30.96 11.76 26.01 11.76 26.01 34.13 29 71.317 11.943 92.16 95.53 30.89 30.89 11.97 25.75 11.97 25.75 38.06 30 15.30 16.00 74.79 74.52 31.14 31.134 16.023 30.59 16.02 30.59 33.12 31 11.78 12.649 83.21 81.64 31019 3101.129.61 28.79 12.61 28.79 35.19 32 12.95 13.72 86.77 86.11 31.00 31.00 13.473 27.385 13.473 27.385 35.89 338 10.15 11.36 96.59 795.91 30.90 30.90 11.39 24.75 11.39 24.75 35.06 34 21.17 234.3 88.48972.53 31.85 31.8513.14.6 408.04 234.6 408.04 05.00 35 15.30 17.76 671.11 73.61 31.34 31.34 18.23 34.61 18.23 34.61 06.00 36 8.608 92.49 73.66718.45 30.92 30.92 9.510 23.76 9.510 23.76 34.85 437 7.60 13.43 70.06 72.76 31.11 31.11 13.47 27.858 13.47 27.858 35826 38 6.634 11.215 77.04 79.378 30.88 30.88 11.20 26559 11.20 26559 37.02 _____

39 85.40 14.340 6899.3 72.41 31.13 31.13 14.6.4 28.047 14.6.4 28.047 35.82 40 8.27 13.29 67.920 700.91 31.108 31.108 13.33 26.65 13.33 26.65 36.94 42 7.60 12.25826.77 84.40 31048 310.4 12.30 29.01329.30 29.01 36.642 43 70.19 213.09 70.704 72.20 31.19 31.19 213.19 30.584 213.19 30.584 38.61 46 818.6 13.77 897.320 100.73 31.08 31.08 13.85 28.93 13.85 28.93 35.64 50 15.57 16.15 81.24 83.12 31.13 31.13 16.17 29.85 16.17 29.85 34.66 51 9.14 10.51 71.49 71.88 31.00 31.00 10.50 25.62 10.50 25.62 35.31

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-8 Table 2-2: Kev Inout and Output Parameters for all 59 Cases (2 of 2) (continued) m E o -

0)* I 1 ..Jj O m0 *E ,,i mmmu C

52 6.35 11.40 72.13 74.11 31.02 31.02 11.42 27.02 11.42 27.02 34.86 53 14.47 17.87 82.68 87.02 31.25 31.24 17.92 33.46 17.92 33.46 0.00 54 6.12 9.98 67.74 70.86 30.90 30.90 10.02 26.21 10.02 26.21 36.49 55 12.93 13.65 81.96 82.62 31.13 31.13 13.73 30.15 13.73 30.15 34.41 56 15.12 15.73 87.92 87.42 31.14 31.14 15.75 29.28 15.75 29.28 33.90 57 7.65 13.40 75.80 80.55 31.20 31.20 13.40 30.22 13.40 30.22 0.00 58 19.37 19.64 81.44 83.88 31.39 31.39 19.66 33.27 19.66 33.27 32.26 59 _________ __ ___ _______ 8.72 10.16 68.44 68.07 30.99 30.99 10.18 26.55 10.18 26.55 33.91 _____

m

t.EUIJE UE*U LJ'L-,L4!I F*I IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-9 2.2 SNPB RAI-2 Section 2.1 of ANP-3352P indicates that AREVA CE-I16 HTP TM fuel design is compatible with the co-resident fuel in the SL-2 core. Provide details of the compatibility analysis and evaluations that assure acceptable fit-up with SL-2 reactor core internals, fuel handling equipment, fuel storage racks, and co-resident fuel.

Response

The AREVA design has been evaluated by AREVA and FPL to ensure that the design is compatible with the St. Lucie Unit 2 core structure, the co-resident fuel, the spent fuel pool and new fuel storage, and the handling equipment. In addition to the evaluations described below, a prototypic Upper Tie Plate was tested by FPL to assure compatibility with the new fuel grapple, and a prototypic fuel assembly (containing tungsten carbide pellets to provide the necessary weight) has been received by FPL and is available for testing to assure compatibility with the fuel handling equipment and consistency with the plant equipment, processes, and procedures.

The process used to establish the mechanical compatibility is to evaluate the fuel assembly interfaces with the upper and lower core support plates, the co-resident fuel, the CEAs, the storage areas, and the handling equipment. This evaluation uses plant drawings and co-resident fuel drawings provided by FPL. FPL has been an active participant in these evaluations, both reviewing the AREVA design drawings and reviewing the compatibility evaluations. The specific compatibility assessments are separated into AREVA fuel component reviews. Table 2-3 describes the components and the interfaces that have to be addressed for each component.

~.AJI ILl 'JEI~k~A ULtLH I r~t ~L ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-10 Table 2-3: Component Interfaces Fuel Assembly Compatibility Interfaces Evaluated Conclusions Component Upper core support plate (Fuel UTP posts position, diameter, and Alignment Plate, FAP) height set to maintain engagement UTP engagement for holddown (i.e. Reaction plate in contact with FAP reaction plate engagement throughout throughout the design life at room Upper Tie Plate design life) and operating temperatures (UTP) UTP posts engage with CEAs.

Control Element Assembly (CEA) Center post provides scram support Clearances between post and Handling Equipment reaction plate dimensions on UTP allow grapple engagement Lower core support plate (LCSP) LTP supported by support plate.

Lower Tie Plate LTP hole diameters for alignment (LTP) LCSP alignment pins pins and hole depths sufficient to engage with the alignment pins.

ID and length sufficient to accommodate CEA rods. Internal Corner Guide Control Element Assemblies configuration similar to co-resident Tubes design so no impact on CEA insertion times (e.g. dashpot elevation, length, and ID, etc.)

ID and length sufficient to Control Element AssembliesacomdtCErds Configuration similar to co-resident Center Guide Tubedein ID and length sufficient to In-cre istruentaionaccommodate in-core detectors.

Configuration similar to co-resident design.

Bundle engages with no growth Core plate separation (including and has sufficient clearance at EOL temperature and irradiation changes) with maximum growth at hot/cold conditions withco-esient Grid overlap maintained throughout Fuel Assembly Spacer grid overlap lifeime inludngehenEdLnex fuel (including irradiation changes) toBlifutie,nl.dn hn Lnx AREVA envelope bounded by co-Fuel assembly component envelopes resident fuel envelope. Therefore (basis for storage assessments) compatible with spent fuel and new fuel storage.

t, IULIItJlI*'-.i L t.C Il If I 1IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-11 Table 2-3: Component Interfaces (continued)

Fuel Assembly Compatibility Interfaces Evaluated Conclusions Component Fuelassmbl coponnt iagnal AREVA diagonal dimensions Fuelassebly ompnentdiagnal bounded by co-resident fuel.

dimensions (basis for storage Therefore compatible with spent Fuel Assembly assessments) fuel and new fuel storage.

(continued)

Fuelassmblycomonet elvatons Spacer grids and fuel column Fuelassmblycomonet elvatons elevations similar between AREVA relative to co-resident fuelancorsdtdeis 2.2.1 Upper Tie Plate Compatibility The majority of the compatibility evaluations are for the upper tie plate because this component interfaces with the FAP, CEAs, and handling grapples. Figure 2-1 is illustrative of the interface of the upper tie plate with the FAP. The details of the UTP are different for the St. Lucie Unit 2 design (e.g. there is no box around the grillage at St. Lucie Unit 2), but the components and interfaces are the same. The center and corner posts fit into holes in the FAP. The reaction plate pushes against the FAP to provide holddown. The interface evaluations have to ensure that:

• The posts will insert into the FAP

• The posts remain engaged at beginning of life hot conditions when the core barrel thermal expansion will increase the core plate separation relative to the fuel assembly

• The reaction plate will remain engaged with the FAP at beginning of life, hot conditions when the core barrel thermal expansion will increase the core plate separation relative to the fuel assembly

• The posts will not bottom out in the FAP at end of life, cold conditions when the fuel assembly has irradiation growth and the differential thermal expansion is not beneficial For St. Lucie Unit 2, all of these evaluations showed margin throughout the fuel design lifetime.

~s~JE EU ~ L~k~J~UE I E~I EL ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-12 PPLAT[

Figure 2-1 :UTP / FAP Interface The FAP is open in locations where the CEAs are located. The CEA rods insert into the posts of the UTP. When the CEA scrams, the center hub of the CEA hits and rests on the UTP center post. AREVA verified that the positions of the posts are the same as the position of the CE.A rods, the size of the UTP holes in the posts will accommodate the CEA rods, and the UTP contact area and strength will accommodate the CEA impact and support the CEA at the proper full insertion elevation.

The handling equipment, new fuel and irradiated fuel grapples use a tube with "j-hooks" to attach to the reaction plate between the corner posts. The assessments addressed the reaction plate thickness relative to the j-hook size, the insertion depth of the j-hook, and potential interactions with other UTP components (e.g. springs, spring cups, etc.). All of the interfaces showed clearance margins.

2.2.2 Lower Tie Plate Compatibility The lower tie plate (LTP) rests on the lower core support plate (LCSP). Four alignment pins are attached to the lower core support plate at each fuel assembly location in the core. The pins insert into holes in the LTP to properly position the fuel assembly in the core. The alignment pins on the LCSP and the holes in the fuel alignment plate (FAP) establish the bundle pitch in the core. The inlet flow is through holes in the LCSP in positions between the alignment pins.

The mechanical compatibility assessments verified that the LTP could accommodate the alignment pins, including the diameter, pin height, and position. The evaluations demonstrated margin in the clearances.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-13 2.2.3 Corner and Center Guide Tubes The corner guide tubes must accommodate the CEA outer rods. The nominal inner diameters of the AREVA guide tube is set to the same nominal diameter as the co-resident fuel, in both the dashpot and non-dashpot elevations. The AREVA design has the same number of weep holes at the same diameter as the co-resident fuel and at approximately the same elevations as the co-resident fuel to assure that the insertion, hydraulic breaking and coolant flows are comparable. In the AREVA design, there is a small hole in the screw that attaches to the bottom of the guide tube to allow drainage of the guide tube for dry storage. The co-resident fuel does not have this drain hole. This small screw hole has an insignificant impact on the hydraulic breaking (i.e. there is no significant change (tendency to result in faster insertion) because the hole is very small compared to the total weep hole area in the tube). The insertion length of the CEA rods with respect to the available guide tube length demonstrated there is more margin than with the current design for the CEA rod insertion.

The center guide tube does not have a dashpot; it has a uniform ID the full length of the tube.

The center guide tube must be capable of accommodating the CEA center rod and, in different core locations, the in-core detectors. The nominal inner diameter of the AREVA design is the same as the co-resident fuel. The AREVA design has the same number and size of weep holes as the co-resident design. The length available for insertion is slightly greater than the co-resident fuel, thus providing more margin for the insertion of both the CEA rods and in-core detectors.

2.2.4 Fuel Assembly The AREVA design is compared with the co-resident fuel. The primary comparisons are:

• LTP elevation

• Fuel rod (fuel column) elevations

• Spacer grid elevation and overlap

• UTP bottom surface elevation

• Component envelopes

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ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-14 All elevations are made relative to the height above the LCSP so that there is a common reference for the different designs.

2.2.4.1 LTP Elevation The LTP elevation provides the plafform for the attachment of the guide tubes and a resting surface for the fuel rods after the spacer grid springs relax due to thermal and irradiation effects.

The AREVA designed LTP has the same elevation as the co-resident fuel.

2.2.4.2 Fuel Rod Elevations The nominal fuel column lengths for the AREVA and co-resident designs are the same. The elevation of the bottom of the fuel column at beginning of life (BOL) depends on the height of the fuel rod lower end cap and the gap between the bottom of the end cap and the top of the LTP. Both the AREVA and the co-resident fuel have the same nominal length for the fuel rod lower end cap. Because of the differences in fuel assembly fabrication processes, the AREVA design has a gap between the bottom of the end cap and the top surface of the LTP of about 0.12 inch. Therefore, the fuel column is initially elevated in the AREVA design relative to the co-resident design at BOL by this gap. The impact of this small gap is insignificant. The minimum axial node length in the neutronic calculations is typically at least 10 times greater than this gap.

2.2.4.3 Spacer Grid Elevations and Overlap At BOL, the centerline elevations of the spacer grids (except for the bottommost and topmost spacers) relative to the LCSP align within 0.005 inch. The AREVA grids have different grid heights than the co-resident design. At beginning of life, these differences in grid height assure that there is 100% overlap between the grids in adjacent fuel assemblies, including the top and bottom most spacers. Because fresh fuel with no irradiation growth may be placed next to irradiated fuel during the follow-on reloads, the overlap between the spacer grids for adjacent fresh and end of life (EOL) discharged assemblies is evaluated. The minimum overlap occurs at the top grid and is greater than [ ] between a BOL AREVA assembly and an EOL co-resident assembly. This evaluation is very conservative because fresh fuel will never operate adjacent to an EOL discharged fuel assembly. Fresh fuel can operate adjacent to fuel that has at least one more cycle of irradiation prior to discharge.

'.~jIJ! ELI L.~EI~U EJ~JLUt FE~I IL ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-15 2.2.4.4 Envelope Evaluations The envelopes of the components on the AREVA design, including the diagonal dimension, are compared with the co-resident fuel. This comparison assures that the assemblies can fit beside each other within the bundle pitch set by the core plates. The comparison also indicates if there may be a potential for handling issues. The maximum component envelope for the AREVA design is about [" ] smaller than the maximum nominal co-resident component envelope. Also, the AREVA diagonal dimensions are slightly less, thus reducing the likelihood of corner interactions during handling. Because the envelopes and diagonal dimensions are bounded by the co-resident fuel, compatibility with the spent fuel and new fuel storage is demonstrated.

• .-UE ELI IJSk* P*,.U-I E[I IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-16 2.3 SNPB RAI-3 AREVA has used CE-16 HTP TM lead test assemblies (LTAs) at Palo Verde Nuclear Generating Station, Unit 1 (Palo Verde), and completed a lifetime irradiation before discharge and inspection. Provide the details of the inspection results as to their performance of these LTAs at Palo Verde.

Response

Lead Test Assemblies completed three I18-month cycles at Palo Verde Unit 1. Post Irradiation Examinations were completed at the end of the final cycle to verify the CE16-HTP TM operational performance. The following bullets provide a summary of these 3rd-cycle fuel exams:

• The fuel assembly visuals indicated excellent performance of the CEI6-HTP TM fuel design with no evidence of rod bow, a large shoulder gap present (gap between fuel rods and UTP), no evidence of handling damage, and a very light layer of crud.

• The fuel assembly length data indicated a consistent positive growth after the third cycle of irradiation. The maximum growth observed was [ ] %dL/L at [ ] GWd/MTU. The growth was well within the design correlation.

• The maximum shoulder gap closure observed was [ ], consistent with the fuel rod growth database.

• The grid growth data collected indicated the grids were dimensionally stable with minimal growth or slight shrinkage during the three cycles of operation.

• The grid oxide data showed a maximum oxide thickness of [ ] microns, which was within predictions.

• The maximum rod oxide observed was [ ] microns, well below the licensing limit of 100 microns.

• The rod creep down (change in rod OD) was between [ ], and was consistent with the design models.

• The maximum grid to rod fretting scar observed on the [ ] rods measured was [

]. There is significant grid to rod fretting margin with the CE16-HTP TM design at Palo Verde.

LiI~~ L~'L~L~I i~I IL ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-17 2.4 SNPB RAI-4 Section 2.3.1 of ANP-3352P reports that the AREVA fuel design is compatible with the reactor components and the co-resident fuel in the core.

(a) Provide a detailed summary of the analysis results that confirms the mechanical and thermal compatibility of the AREVA fuel design with the co-resident fuel and the core internals.

Hydraulic compatibility analysis should include the assessment of the impact of AREVA fuel on core flow distribution. Thermal compatibility analysis should evaluate the impact on departure from nuclear boiling ratio calculations due to the introduction of AREVA fuel.

(b) Provide summary results of the additional mechanical compatibility evaluations of the upper tie plate, the lower tie plate, and the center and corner guide tubes of the AREVA fuel assembly while in the SL-2 core.

Response

2.4.1 SNPB RAI-4, Sub-item (a):

Mechanical Compatibility: The detailed summary of the mechanical compatibility is provided in the response to SNPB FRAI-2 found in Section 2.2.

Thermal-Hydraulic Compatibility: Section 4.5.1 of Reference 2 provides a summary of the Thermal-Hydraulic Compatibility Analysis. Additional details regarding the impact of the mixed core on core flow distribution and Departure from Nucleate Boiling (DNB) performance are provided below.

2.4.1 .1 Mixed Core Flow Distribution Effects The AREVA HTP TM assemblies have a slightly higher pressure drop relative to resident Westinghouse assemblies (Figure 2-2 and Reference 2 Section 4.5.1.1). This difference is small and within previous AREVA transition experience. This results in the DNB-limiting AREVA HTPTM assembly receiving less flow in mixed core analyses relative to a full core of AREVA HTP TM fuel. This difference in flow is explicitly included in the thermal-hydraulic analyses.

Pressure drop tests were conducted for both the AREVA St. Lucie Unit 2 fuel assembly and for a co-resident Westinghouse fuel assembly at the same AREVA test facility. These test results provide a common and consistent basis in the mixed core analyses. Table 2-4 and Table 2-5

~AJF ILl ~JiI~U ~F~JL~L~ F~l II.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-18 show the calculated relative flow changes under nominal conditions through resident Westinghouse fuel and AREVA fuel during transition cycles compared to a full core of the respective fuel type.

Table 2-4: Resident Westinghouse Assembly Flow Fraction Relative

- to a Core with All Resident Westinghouse Assemblies Table 2-5: AREVA Assembly Flow Fraction Relative to a Core with All ARE VA Assemblies

ILl JEI~J ~fIJ~L'II I E~I IL AREVA Inc.

St. Lucie Unit 2 Fuei Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-19 Figure 2-2: Pressure Drop Profile Comparison 2.4.1.2 Mixed Core DNB Performance Mixed core DNB performance is addressed in section 4.5.1.5 of Reference 2. As more HTPTM fuel assemblies are inserted into the core in subsequent reloads, less flow will be diverted into the lower pressure drop Westinghouse assemblies; increased flow in the AREVA HTPTM assemblies will result in an increase in DNB performance. The improvement in DNB performance in AREVA HTP TM fuel during the transition is illustrated in Figure 2-3.

~JULr~JE~U ~}~H~HL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-20 Transition DNB analyses performed in Reference 3 were performed using an XCOBRA-IIIC model representative of a single batch of AREVA HTP TM fuel (88 AREVA HTP TM assemblies) which will bound the DNB performance of a full core of AREVA HTP TM fuel. In addition to modeling the mixed core explicitly, the USNRC-approved additional 2% mixed core penalty is applied to the AREVA HTP TM correlation limit as discussed in Section 4.4 of Reference 2.

The Westinghouse co-resident fuel is expected to be at least 5% lower in power peaking and receive increased RCS flow due to slightly lower pressure drop as compared to the AREVA HTP TM fuel. This combination of lower peaking and higher flow will make the co-resident Westinghouse fuel non-limiting for DNB during the transition cycles. This will be verified on a cycle specific basis, as part of the reload process, to confirm that the Westinghouse fuel continues to meet the respective fuel design limits during the transition cycles using the applicable Westinghouse methodologies (Reference 4, Section 7.0).

Figure 2-3: Mixed Core DNB Performance

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-21 2.4.2 SNPB RAI-4. Sub-item (b):

When the fuel assembly is in the core, it mechanically interfaces with the upper fuel alignment plate (FAP), the lower core support plate (LCSP), the control assemblies, the in-core detectors, and the co-resident fuel.

2.4.2.1 Upper Tie Plate (UTP)

The UTP interfaces with the FAP, the CEA, the in-core detectors, and the co-resident fuel when in the core. The UTP corner posts insert into holes in the FAP. The post positions on the UTP are consistent with the FAP hole pattern, and the post diameters will allow the posts to insert into the holes. As discussed in the response to SNPB RAI-2 (Section 2.2), at BOL, hot conditions (the most limiting condition), the UTP posts must be in the holes to maintain the assembly position. At EOL, cold conditions (the most limiting condition), the UTP posts must have clearance between the top of the post and the bottom of the FAP hole. The evaluations of the UTP determined that it will be engaged at all conditions (hot and cold, EOL and BOL), and will have clearance at EOL.

The reaction plate is the part of the UTP that is supported by the holddown springs. This plate must be in contact with the FAP to provide the holddown force from the springs, and should have sufficient travel to accommodate the deflection changes resulting from differential thermal expansion between the core and the fuel assembly and the fuel assembly irradiation growth.

The UTP mechanical compatibility evaluations demonstrate that the reaction plate is in contact at all conditions (hot/cold, BOL/EOL) and maintains holddown capability at all conditions.

The CEA rods enter the fuel assembly through the posts in the UTP. Therefore, the posts have to be in the proper positions and have an inner diameter sufficient to accommodate the CEA rods. These dimensions were verified by comparisons with the FAP and CEA rod dimensions.

When the CEA drops, the center hub of the control assembly contacts and comes to rest on the top of the center guide tube post. The elevation of this center post determines the distance that the CEA rod inserts into the assembly. The mechanical compatibility evaluations demonstrated that the AREVA center post has the same contact area as the co-resident fuel. Therefore the contact area and loads are similar to the co-resident fuel. The center post elevation on the AREVA UTP is the same as the co-resident fuel. Thus, the control rod insertion is the same.

~.jL~E ~LI 'J~~U ~J~jL~1 I ~ IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-22 For fuel assemblies in the instrumented locations of the St. Lucie Unit 2 core, the in-core instrument enters the fuel assembly through the hole in the center post. The center post on the AREVA design has the same inner diameter as the co-resident design. Therefore, the UTP is compatible with the in-core instrumentation.

The envelopes of the AREVA and co-resident UTPs are the same and are much smaller than the envelopes and diagonal dimensions of the spacer grids and LTP. Therefore, there is no interaction of the UTPs with adjacent fuel assemblies.

2.4.2.2 Lower Tie Plate (LTP)

The LTP interfaces with the LCSP, the four alignment pins located in the LCSP, and the co-resident fuel when in the core. In the LTP, the fuel assembly support is provided by posts located underneath the four corner guide tubes. The LTP is attached to the corner guide tubes by cap screws that screw into the plugs on the bottom of the corner guide tubes. These screws insert through the underside of the LTP posts to join the guide tubes to the LTP. The LCSP alignment pins insert into the center of the corner LTP holes and terminate below the bottom of the cap screws. Figure 2-4 is a cross section illustrating the LTP corner post configuration. At the top of the post hole is the counterbore for the cap screw head and the hole for the cap screw shaft. The distance from the bottom of the LTP to the cap screw head must be sufficient to accommodate the alignment pin height. The AREVA evaluations confirmed that the LTP corner post hole was positioned appropriately to engage the pins, that the hole diameter was sufficient for the alignment pins to insert, and that the height of the hole (including the cap screw) was sufficient to accommodate the alignment pin height. The evaluations also showed that the contact area for the LTP post and the LCSP was sufficient.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-23 Figure 2-4: Cross-Section Illustration of LTP Corner Post 2.4.2.3 Center and Corner Guide Tubes When the fuel is in CEA locations in the St. Lucie Unit 2 core, corner and center guide tubes interface with the CEA rods. The mechanical compatibility assessments include the inner diameter of the guide tubes, the elevation of the dashpots of the corner guide tubes, and size and elevation of the weep holes in the corner guide tubes, and the axial length of the guide tube available for control rod insertion. The AREVA corner guide tubes have the same nominal diameters in both the dashpot and non-dashpot regions as the co-resident fuel. The transition length from the dashpot region to the non-dashpot region is longer in the AREVA design, but the difference has an insignificant impact on the volume of the annulus between the control rod and the guide tube. The dashpot elevation is the same in the AREVA and co-resident fuel (about 0.001 inch difference). The number of weep holes and size are the same in both designs. The AREVA design also has a drain hole in the guide tube cap screws, but the increase in flow exit area is small (about a 2% change). The bottom of the AREVA corner guide tube has a slightly lower elevation than the corresponding elevation in the co-resident design. Therefore, the AREVA design is mechanically compatible with the control assemblies: the corner guide tube will allow for CEA rod insertion, the rod insertion times are not affected by the corner guide tube design, and the corner guide tube has slightly more margin to accommodate the control rod length.

ELI ~JII~U ~41JL~Ui I I~I I AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-24 The center guide tube does not have a dashpot. The AREVA center guide tube design has the same nominal inner and outer diameters as the co-resident design. Both designs have the same number and size of weep holes in the tube, and the AREVA design has a drain hole in the LTP. Because the center guide tube does not have a dashpot, it does not provide the hydraulic breaking that the corner guide tubes provide when the CEA is scrammed. The area of the drain hole does not significantly affect the total exit area (<5% change in the non-dashpot tube). The AREVA center guide tube has more axial length for CEA rod insertion than the co-resident fuel design, thus providing additional margin to prevent the control rods from bottoming out.

Therefore, the center guide tube is mechanically compatible with the control assemblies; the inner diameter will allow CEA insertion, the minor configuration differences have negligible effects on control rod insertion times, and the AREVA design has more axial length to accommodate the control rod.

In some locations in the St. Lucie Unit 2 core, an in-core instrument is inserted into the center guide tube. As discussed previously, the AREVA design has the same nominal inner diameter as the co-resident design and has more length to accommodate the insertion than the co-resident design. Therefore, the AREVA design is mechanically compatible with the in-core detectors.

t.JE Il IJ f t. L.J!~l.tl I f I ti AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-25 2.5 SNPB RAI-5 Section 2.4.3.1 of ANP-3352P states that AREVA has developed correction factors to be incorporated into evaluations using the approved legacy code, RODEX2, to account for fuel thermal conductivity degradation with exposure effects. Provide details of the use of RODEX2 with correction factors (NRC-approved Reference 4 of ANP-3352P) for all applicable non-LOCA and LOCA analyses.

Response

2.5.1 Non-LOCA Analyses Response 2.5.1.1 Chapter 15 Non-LOCA and Thermal-Hydraulics Analyses For Chapter 15 Non-LOCA analysis, RODEX2 was replaced with COPERNIC for the purposes of generating the fuel thermal-conductivity, heat capacity, and fuel pellet-to-clad gap coefficient inputs for the average core and hot spot models. This change was made to explicitly account for the effects of thermal conductivity degradation (TCD). The properties from COPERNIC were developed for beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions in accordance with the approved topical report, and COPERNIC replaces RODEX2 for this purpose. The COPERNIC fuel properties and gap coefficients were conservatively implemented relative to the RODEX2 inputs as allowed in the approved topical report (Section 2.7 of Reference 3).

For Thermal-Hydraulic analysis, correction factors have also been applied in the establishment of the core peak linear heat generation rate limit that would preclude Fuel Centerline Melt (FCM) in AREVA HTPTM fuel. These correction factors are melt temperature penalties that were generated through comparison of the RODEX2 code to the COPERNIC code (which accurately models TCD) as described in Section 2.8 of Reference 3 and are consistent with the requirements of recently approved Reference 5. This process ensures that the core peak linear heat generation rate limit set for AREVA HTPTM fuel is based on a conservative prediction of TCD.

2.5.1.2 Fuel Rod Performance Analysis Correction factors for the steady state strain, AOO strain, clad fatigue and rod internal pressure have been incorporated into the evaluations provided in ANP-3352P (Reference 2) to account for the thermal conductivity degradation effects as applied to the legacy code RODEX2. These

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-26 correction factors are determined to be consistent with or conservative as compared to the recently approved Supplement I methodology provided in Reference 5.

Table 2-6 compares the limiting values in ANP-3352P (Reference 2) to values generated using the recently approved Supplement 1 methodology (Reference 5).

Table 2-6: Fuel Rod Performance TCD Factors Comparison Limiting Value Value Using Parameter from ANP- NRC Approved Comment 3352P Reference 5 AQO Strain [ ] [ ]

Steady State Strain [ ] [ ]

Fatigue [ ] [ ]

Rod Internal ]

Pressure []

2.5.2 LOCA Analyses Response For the Small Break LOCA (SBLOCA) analysis in ANP-3345P (Reference 9), RODEX2 is used to determine the initial core and hot pin stored energy for SBLOCA evaluations. Small breaks evolve through a pump coastdown and natural circulation phase to a loop draining phase followed by a boil-down and refill phase. [

~LI UEI~ L/~~I F~I IL ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-27 The peak cladding temperatures, which occur later in the transient, depend on decay heat versus heat transfer and have no relationship to the initial stored energy within the fuel. This was demonstrated in a sensitivity study performed for the U.S. EPR (discussed within RAI-31 b of Reference 10). [

] Thus, with any adjustments (corrections) made to the initial fuel temperature there will be no significant effect on the SBLOCA cladding temperatures or the local oxidation. From this, there are no Thermal Conductivity Degradation (TCD) related adjustments necessary for SBLOCA evaluations.

For the Realistic Large Break LOCA (RLBLOCA) analysis, TCD effect is explicitly accounted for as described in ANP-3346P, Section 6.1 (Reference 11) and is not affected by the RODEX2 correction factors discussed in Section 2.4.3.1 of Reference 2.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-28 2.6 SNPB RAI-6 It is customary that for a mixed core such as SL-2 transition cores, the licensee performs a mixed core analysis if there are geometry differences between the resident and co-resident fuel designs. Provide details of the mixed core analysis that show that the resident fuel is compatible with the new fuel, despite geometrical differences between them.

Response

Table 2-7 provides a review of key nominal dimensions of the AREVA fuel design and the co-resident fuel design on a bundle assembly and component basis. This list of attributes is not all inclusive, but contains key attributes used in the mechanical compatibility evaluations. This comparison shows where differences occur and the impact of these differences on the mechanical compatibility. This comparison demonstrates that the AREVA design configuration is very similar to the co-resident configuration and is mechanically compatible with the core, plant equipment, storage areas, and co-resident fuel design. The mixed core thermal compatibility analysis, discussed in the response to SNBP RAI-4 (Section 2.4.1), has explicitly modeled the fuel assembly geometries of the two fuel types. The bracketed values in Table 2-7 for the co-resident fuel were provided to AREVA as Westinghouse proprietary.

U ~ ~L AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-29 Table 2-7: Configuration Comparison for SLU2 (Nominal Design Dimensions)

Coinponent AREVA Design Co-Resident Design Difference Conclusion Fuel Assembly Height, inches 158.529 158.529 None

[ ] - HMp TM [ ] - GuardianTrM TM

[ ]- HTP [ ] -HID-1L

[ ]- HTP TM [ ]- HID-1L inhe ]

J-HTP TM

[ ] -HID-1L Overlaps assure grids Grid Elevations, inhs [ ] - HTP TM [ ] - HID-1L ajcn oec te (centerline) [ ] - HTP TM [ ] - HID-1L arjaen compatiblter

[ ] -HTP TM [ ] -HID-1L aecmail

[ ]- HTP TM [ ]- HID-1L

[ ]- HTP TM [ ]- HID-IL

[ ] -HTP TM [ ] -Alloy 625]

Bundle pitch, inch 8.18 8.18 None Rod pitch, inch 0.506 0.506 None Difference is small Elevation of start of fuel rpercentage of analysis column, inch [L code node size and not significant Lower Tie Plate (L TP)

Height of Top Surface, 3.112 3.112 None inch Elevation of Corner Guide Tube Interface, inch [ ] [ ]Compatible with CEAs Diameter of Alignment Pin I[ ]Nn Hole,_inch __ _"__ __] _ _ _ _ __ _ _ _ _ _ _ None__

\~IJD IU 'JDO'~U ~UI I A~I HL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAi-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-30 Table 2-7: Configuration Comparison for SLU2 (Nominal Design Dimensions) (continued)

Component AREVA Design Co-Resident Design Difference Conclusion

[ Compatible with Envelope, inch [ ][ ]storage areas, and less likely for handling

] interactions Compatible with DiagnalDimnsin, DiagnalDimesio,

[

)) inh ich11400[ 1.40 [storage areas, and less

] likely for handling

_____________interactions CornerGuide Tube OD, non-dashpot region, [ ][ ]Nn inch [__]__[__ ]__None_

ID, non-dashpot region, [ ][ INn inch [ ][ ]Nn OD, dashpot region, inch [ ] [ ]Compatible with CEA ID,dashpot region, inch [ ] [ ]None

[ ] Cap screw allows Number of weep holes [ [ ] [ ] drainage of GT for dry

_ _ __ _ _ _ _ _ _ _storage Upper [ ] [

[ ] Upper [ ]

Weep hole diameter(s), Lower [' Compatible with CEAs inch [ ] Lower [ ]

Cap screw [ ]

__ffii___

AREVA Inc.

St. ANP-3440NP andLucie SNPBUnit 2 Fuel RAI-2 thruTransition:

SNPB RAI-20Response to NRC Questions SRXB-RAI-1 Revision 1 Technical Report Page 2-31 Table 2-7: Configuration Comparison for SLU2 (Nominal Design Dimensions) (continued)

Component AREVA Design Co-Resident Design Difference Conclusion Elevation of weep hole [ ] upper [ ] upper Cmail ihC~

centerlines, inch [ ] lower [ ] lower Cmail ihC~

Elevation of start of[

dashpot, inch [ ][ ] Compatible with CEAs

[]

Elevation of top of guide tube end plug, inch [ ][ JCompatible with CEAs Center Guide Tube OD, inch [ ][ ]None ID, inch [ ] [ ]None

[ ]

Number of weep holes (( [I

[ ] in tube Wep[oe imee sCompatible with CEAs Wephlimtrs, [ ] in LTP [

inch center boss and in-cores Elevation of weep holes, Compatible with CEAs inch [ ][ ]and in-cores

iLl ~ LJ~.JYUI I D~D IL AREVA Inc.

St. AN P-3440N P andLucie SNPBUnit 2 FuelthruTransition: Response to NRC Questions SRXB-RAI-1 RAI-2 SNPB RAI-20 Revision 1 Technical Report Page 2-32 Table 2-7: Configuration Comparison for SLU2 (Nominal Design Dimensions) (continued)

Component ]~ARE VA Design C-esident Design Difference Conclusion Elevation of bottom of [Compatible with CEAs center guide tube [ ] [ ] and in-cores column, inch Spacer Grids I ________ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

IGuardian T Number of grids and type 9 TT 8 HID-IL Same total number of grids

____ ____ ____ ___1 ____ Alloy 625 Grd ich[ eiht _HTTM [ ]1 _ GuardianTM Gridheigt, I HTP M nchC ]-HI-ILCompatible with co-C ] - HMp TM [ ] -Alloy 625reintfl Compatible with Envelope, inch [ ] ] (nom, max storage areas, and less grid) likely for handling interactions

[ Compatible with Diagonal Dimension, inch [ ]11.368 (for min storage areas, and less spacer) likely for handling interactions Upper Tie Plate _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Corner post CD, inch [___]___[ ___] None __________

Top of corner post rrCompatible with CEAs elevation, inch LLand FAP 1 Value includes the skirt.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-33 Table 2-7: Configuration Comparison for SLU2 (Nominal Design Dimensions) (continued)

Component AREVA Design Co-Resident Design Difference Conclusion Center post OD, inch [ ] [ ] Compatible with FAP Top of center post 15.2 5.2 oeCEA impact elevation elevation, inch 15.2 5.2 oethe same Area of ring at top of Ring OD [ ] Ring OD [ None Same contact area center post, inch [ Ring ID FigI]

Reaction plate arm width, inch [ ] ] None Reaction plate arm [ ] [ ] Compatible with thickness, inch Tab [ ] Tab [ ] grapples None. Much smaller than envelope of other assembly UTPenvlop,~components ich on both designs. Compatible with core UTPenvlop, ich ] ]Because envelope smaller, and co-resident fuel diagonal dimension acceptable

[

Bottom of UTP elevation, F1 1Compatible with core inch [ ] and co-resident fuel

er~ ~ ~ ii .~i o AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-34 Table 2-7: Configuration Comparison for SLU2 (Nominal Design Dimensions) (continued)

Cornponent AREVA Design Co-Resident Design Difference Conclusion Fuel Rod AREVA rod is slightly shorter. Does not affect mechanical compatibility. Compatible with co-Legt, nc 466014.89Acceptable rod growth resident fuel assessed with approved

____________________M5 correlation Cladding OD, inch 0.382 0.382 None Cladding ID, inch 0.332 0.332 None Cladding material M5 ZIRLO TM Mtra eemnsgot correlation Pellet 00, inch 0.3255 0.3255 None Total Fuel Column 167167Nn Length,_inch ________

Lower end cap length, ][ ]Nn inch _______]__None Note: All elevations are from the lower core support plate and are at room temperature.

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St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-35 2.7 SNPB RAI-7 Provide details of the relative axial positions between CE 16x1 6 and AREVA's CE 1 6x1 6 fuel bundles such as locations of spacer grids and straps and bottom/top of active fuel length.

Response

The detailed comparisons of the elevations of the fuel assembly components are provided in Table 2-7 in the SNPB RAI-6 response (Section 2.6). The spacer grid elevations are very similar with the differences in grid heights assuring overlap throughout the design lifetime. The fuel column lengths are the same for both designs. The AREVA design has a nominal 0.12 inch gap between the bottom of the fuel rod and the top of the lower tie plate. Therefore, the elevation of the bottom and top of the AREVA fuel column is nominally 0.12 inch higher than the co-resident fuel. This difference is small and insignificant. It is much smaller than the minimum node length used in the neutronics analyses.

Based on the various comparisons, the AREVA fuel design is compatible with the co-resident fuel.

'.**L ILbS .. }SEC'-f I'J V, Lf L'*U -*

AREVA Inc.

St. Lucie Unit 2 Fuei Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-36 2.8 SNPB RAI-8 Section 4.5.3 of ANP-3352P states that the impact of rod bow on the minimum departure from nucleate boiling ratio and peak linear heat rate was evaluated using the rod bow methodology described in Reference 27, which is XN-75-32(P)(A), Supplements 1 ,2, 3 and 4, "Computational Procedure for Evaluating Fuel Rod Bowing," October 1983.

(a) Considering the updates to rod bow analysis described in Section 3.9 of Topical Report (TR)

BAW-10227P-A (February 2000) and Section 6.16 of TR BAW-l10240(P)-A (May 2004),

determine whether the legacy methodology above is still appropriately applicable to the AREVA's CE 16x16 fuel design with M5 cladding in analyzing rod bow impact on thermal margin analysis and provide an explanation.

(b) Explain how the rod bow analysis is performed for the transition mixed core at SL-2, specifically with the resident CE 16x16 fuel.

Response

2.8.1 SNPB RAI,-8. Sub-item (a):

The rod bow methodology originally submitted and approved in XN-75-32, Supplements, 1, 2, 3, and 4 (Reference 6), has been revisited and reviewed by the USNRC subsequently as the application was extended. For example, when the burnup limits were increased to the currently approved 62 MWd/kgU peak rod limit in ANF-88-1 33(P)(A) (Reference 7), the USNRC concurred that the applicability of the rod bow correlation and resulting DNB and LHGR penalties would be a conservative application, overpredicting the rod bow and thus resulting in a conservative rod bow penalty. When MS implementation was submitted in BAW-10240(P)(A)

(Reference 8), the USNRC again reviewed the applicability of this correlation and concluded that the "growth characteristics of MS will not have a detrimental effect on rod-bow." This BAW-1 0240(P)(A) SER was issued in 2004. Subsequent to that USNRC review, AREVA has completed a lead assembly program for a CE16 fuel design. In this program, the fuel rods were M5 clad, had a 0.382 inch rod OD, and had a rod pitch of 0.506 inch. These attributes are the same as the AREVA St. Lucie Unit 2 fuel. The fuel inspections performed on the fuel after each cycle and after end-of-life discharge showed no noticeable rod distortion in the visual inspections. This performance was well within the correlation predictions. Based on the

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-37 correlation, there would be significant end of life rod bow (approximately 50% gap closure) which would be easily detected by the visual examinations. Therefore, the continued validity and conservatism of the rod bow methodology was confirmed for the CE1 6 design, with M5 cladding.

2.8.2 SNPB RAI-8, Sub-item (b):

The methodology of Reference 6 was applied in evaluation of the impact of Fuel Rod Bow for AREVA HTPTM fuel. The penalty factors prescribed in Reference 6 for an AREVA HTPTM assembly do not depend on the adjacent fuel assembly type. Westinghouse fuel will continue to be evaluated on a cycle-specific basis using the Westinghouse methodology, as part of the reload process, to ensure that the Westinghouse fuel continues to meet the respective fuel design limits during the transition cycles (Reference 4, Section 7.0).

~LF LJIt~U ~1L)L~UI ~ I AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-38 2.9 SNPB RAI-9 The following questions are related to the seismic and seismic/LOCA evaluations of AREVA CE 16x16, HTP TM , co-resident CE 16x16, and mixed core at SL-2 associated with its request for introduction of the new fuel. They are based on the relevant sections of ANP-3352P and ANP-3396P that were submitted to the NRC.

(a) ANP-3396P, Section 3.2 indicates that, "for St. Lucie Unit 2, the events were analyzed for a full core of the current fuel design, a full core of the AREVA CE 16x16 HTP TM fuel, and for a wide range of mixed core configurations, in order to verify that the limiting loads and deflections remain within acceptable fuel design limits." Provide a summary of the results from the above-mentioned analyses for the three different configurations of the SL-2 core.

(b) TR BAW-1 0133(P) (A) originally modeled a Mark-C fuel assembly for seismic and LOCA analyses. The licensee claims that Addenda 1 and 2 of this TR has demonstrated its acceptability for other generic pressurized-water reactor fuel assembly designs, including the CE 16x16 HTP TM fuel design. Table 3.1 of ANP-3396P for Nominal Beginning of Life Mechanical Design Data Comparison indicates significant differences in several listed parameters for CE 16x1 6 HTP TM and Mark-C fuel designs. Therefore, explain in detail how the differences are accounted for in the components testing.

(c) ANP-3396P states that additional testing and evaluations are included in the analyses to address this NRC Information Notice (IN) 2012-09. Provide detailed information on testing performed in response to IN 2012-09.

(d) Provide a detailed description of both the [

] that are mentioned in the ANP-3396P report. Provide results from the analysis that used the [ ] and explain in detail.

Response

], the response to SNPB RAI-9 will be delayed until AREVA can assess the impact of this issue.

ELI ~J~U ~~UI I E~ IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-39 2.10 SNPB RAI-IO Section 2.8 of ANP-3347P, "St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report," states that the RODEX2 code was used to establish the fuel centerline melt linear heat generation rate as a function of exposure and a penalty to address thermal conductivity degradation with burn up was applied where applicable. Provide details of the penalty applied and indicate which fuel performance, mechanical, and thermal models were affected.

Response

Correction factors are applied consistent with the approved methodology provided in Reference 5 in the establishment of the core peak linear heat generation rate limit that would preclude Fuel Centerline Melt (FCM) in AREVA HTPTM fuel. These correction factors are melt temperature penalties that were generated through comparison of the RODEX2 code to the COPERNIC code (which accurately models TCD) consistent with recently approved Reference 5. This process ensures that the core peak linear heat generation rate limit set for AREVA HTPTM fuel is based on a conservative prediction of TCD.

~pr;fIrJIrVtI AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-40 2.11 SNPB RAI-11 This question is from ANP-3345P, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," and is needed to conclude whether or not this LAR provides acceptable results governed by 10 CER 50.46, GDC 36 and 38; and NUREG-800, Chapter 15.

The analysis assumed symmetric injection into all four emergency core cooling (ECC) lines for the evaluation of the spectrum of cold leg breaks. Surveillance records routinely show that the ECC lines are rarely balanced symmetrically. As such, the line containing the maximum high pressure safety injection delivery rate is connected to the broken loop. Demonstrate that a perfect symmetry in ECC delivery rates to the ECC lines is valid for SL-2. Show that the delivery to the ECC lines supports the latest surveillance ECC flow measurement data. Also, verify that the appropriate error is applied to the head and flow conditions describing the ECC flow delivery curves.

Response

The high pressure safety injection (HPSI) pump flow delivery is calculated for each emergency core cooling system (ECCS) line taking into account the asymmetric line configurations. The lowest three-loop flows are used for delivery into the intact loops and the maximum ECCS line flow is assumed to be delivered to the broken loop. The HPSI pump performance used for flow delivery into the reactor coolant system (RCS) assumes some pump degradation as compared to the manufacturer's performance curve. Additionally, the calculations for the ECCS line delivery flows account for the uncertainties associated with the head and flow measurements. A 1% emergency diesel generator (EDG) under frequency is also accounted for so as to obtain conservative minimum flows for use in the small break LOCA (SBLOCA) analysis of the spectrum of cold leg breaks. The recent pump performance shows that the pump performance curves used for the minimum flow delivery are conservative.

Since the 2A HPSI flows provided the minimum flow delivery, the pump performance data and the ECCS lines flow delivery is provided below for the 2A HPSI pump:

Table 2-8 provides the manufacturer's pump performance data for 2A HPSI pump total developed head (TDH).

~.AJF ILl 1JII~U L'L~UI I ~I IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-41 Table 2-9 shows the in-service testing (IST) criteria for 2A HPSI pump along with recent surveillance data.

Table 2-10 provides the IST 2A HPSI pump performance curve which accounts for IST degradation, uncertainties associated with the head and flow measurements and 1% EDG under frequency effects.

Figure 2-5 shows the plots for Table 2-8 and Table 2-10.

Table 2-1 1 shows the 2A HPSI minimum pump delivery for each of the 4 ECC lines, along with the average of the lowest 3-loop delivery flows.

Table 2-12 shows the conservative flow used in the SBLOCA analysis for each of the loops.

The flow used in the analysis for delivery into the 3 intact loops, as shown in Figure 2-6, is thus seen to be conservative as compared to the calculated 3-loop minimum flow.

Table 2-8: Manufacturer's Pump Performance Data HPSI Pump 2A Manufacturer's Curve Flow (gpm) TDH (ft) Flow (gpm) TDH (ft) 0 2866 350 2518 25 2863 400 2348 50 2860 450 2184 100 2860 475 2094 150 2849 500 1997 175 2846 550 1786 200 2836 600 1545 250 2796 625 1415 300 2679 631 1378 Table 2-9: IST Criteria and Surveillance Data 2A HPSI Pump Full Flow Test 2A HPSI Pump Code Run(30gm IST Criterion > 2854 ft IST Criterion > 2486 ft 5/5/2014 Test 2947 ft 3/9/2014 Test 2572 ft 8/6/2014 Test 2901 ft 8/15/2012 Test 2576 ft

BLJE I,*,, IJ~ LJLP-L,-I 1 I IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-42 Table 2-10: IST Pump Performance Data HPSI Pump 2A IST Minimum Performance wl Underfrequency Flow (gpm) TDH (ift 0.0 2783.2 24.8 2780.2 29.7 2780.2 49.5 2777.3 74.3 2777.3 99.0 2777.3 148.5 2735.5 173.3 2716.9 198.0 2691.9 222.8 2667.0 247.5 2638.6 272.3 2591.5 297.0 2528.2 321.8 2452.7 346.5 2376.3 371.3 2297.9 396.0 2215.8 420.8 2136.5 445.5 2061.1 470.3 1976.2 495.0 1884.6 519.8 1789.3 544.5 1685.5 569.3 1575.0 594.0 1458.0 618.8 1335.4 624.7 1300.4 653.4 1149.5 678.2 1019.2

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-43 3000.0 26 0. '.. . . . ". l* -- *. . . . . ... . . .................... .....

28000

"-'*"*"""".... "*" ... '"" ."" 'Per"form"nce...".............

2400.0 1600.0 .

1000.0,,

0.0 100.0 200.0 300.0 400.0 500.0 600.0 700.0 800.0 Flow (gpm)

Figure 2-5: Manufacturer's Pump Performance vs IST Pump Performance

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-44

___________ Table 2-11: 2A HPSI Minimum Pump delivery______

RCS Flow Flow Flow Flow Average Flow (Lowest 3 Pressure (Loop 2B2) (Loop 2B1) (Loop 2A2) (Loop 2A1 ) Legs)

(psia) (gpm) (gpm) (gpm) (gpm) (gpm) 1212 0.5 0.5 0.5 0.5 0.5 1205 16.2 16.0 15.9 15.8 15.9 1198 19.6 19.4 19.2 19.1 19.2 1177 27.9 27.6 27.3 27.1 27.4 1104 49.7 49.2 48.5 48.2 48.6 1035 62.5 61.8 61.0 60.6 61.1 943 74.2 73.4 72.4 72.0 72.6 829 87.1 86.0 84.9 84.4 85.1 699 100.6 99.3 98.0 97.4 98.3 551 114.8 113.4 111.9 111.2 112.2 393 128.1 126.5 124.8 124.0 125.1 217 141.2 139.5 137.6 136.7 137.9 0 155.7 153.8 151.7 150.8 152.1

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-45 Table 2-12: SBOAAnalyi HSPupFlow Delivery RCS Pressure Flow to Each (psia) Loop (gpm) 1063.1 0 1062.6 10.6 1062 21.3 1045.8 31.9 1009.7 42.5 954.3 53.1 883.3 63.8 800.7 74.4 708.4 85 603.7 95.6 476.6 106.3 307.6 116.9 148.1 124.3 125.5 125.2 117.9 125.5 94.3 126.4 59.9 127.7 12.4 129.3 0 129.8

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-46 1400 1200 Minimum IST Performance Delivery/Loop 1000 (average of 3 lowest legs)

S800

0. 600 C..

1 Analysis Flow Curve t (Flow/loop) 400 200 0

0 20 40 60 80 100 120 140 160 Flow (gpm)

Figure 2-6: SBLOCA Analysis Flow vs IST Performance Flow

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-47 2.12 SNPB RAI-12 This question is from ANP-3345P, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," and is needed to conclude whether or not this LAR provides acceptable results governed by 10 CFR 50.46, GDC 36 and 38; and NUREG-800, Chapter 15.

As stated on page 3-5, [

]. Describe how this is accomplished. Is this dynamically calculated by the code or is the broken loop suction leg forced to clear? Explain and show if any residual water is retained in the broken loop suction leg.

Response

I

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-.20 Revision 1 Technical Report Page 2-48 Figure 2-7: Example Only - No Bias

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-49 Figure 2-8: Example Only - Bias on Intact Legs

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-50 1,0 i* iL

• 0.8 r ..... Loop IA

• Loop 18B I

• Loop 2A 0.6 4' Loop 2B - broken LL "5

0.2 0,0 e u *mm I4 *mmaM a a * ............ ' *a

• q, a* a mm= o D si aK 4aa 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-9: Loop Seal Void Fraction - 2.70-in Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-51 60.0 .......... ............ ...... ................. ................... . . . .... . . ... .. ........ . ...................... .............. ...

-a Loop 1A

• Loop 18 50.0 SLoop 2A "qLoop 2B - broken 40.0 30.0 U,  ! 4 20.0 10.0 lb~4~~

-10.0

-20.0 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-10: Loop Seal Steam Velocity - 2.70-in Break

AREVA Inc.

St.

and Lucie SNPBUnit 2 Fuel RAI-2 thruTransition:

SNPB RAI-20Response to NRC Questions SRXB-RAI-1 ANP-3440NP Revision 1 Technical Report Page 2-52 2.13 SNPB RAI-13 This question is from ANP-3345P, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," and is needed to conclude whether or not this LAR provides acceptable results governed by 10 CFR 50.46, GDC 36 and 38; and NUREG-800, Chapter 15.

Show the secondary steam generator two-phase level versus time for the limiting 2.7 inch small break. Indicate the top of the tube bundle elevation in the plots. Does the auxiliary feedwater flow match system boil-off upon actuation? If not, and the bundle is exposed to vapor, does the model account for this potential behavior? Provide an explanation.

Response

Note that the data presented for this response reflects the updated charging flow configuration discussed in SNPB RAI-15 (Section 2.15).

Figure 2-1 1 shows the steam generator's narrow range level. The upper level tap (100%) is at 474.43 inches above the top of the tubesheet; the lower level tap is at 294.052 inches above the top of the tubesheet; the longest U-tube is at 347.259 inches above the top of the tubesheet, which corresponds to 29.5% level. Auxiliary Feedwater (AFW) delivery begins at 934 seconds.

The figure shows the Steam Generator (SG) level increase as a result of AFW delivery, reaching the top of the U-tubes about 480 seconds after AFW begins. [

] AFW delivery stops once the SG level reaches its normal operation value. The SG level is conservatively based on collapsed liquid level signal, which will produce a lower level indication relative to a two-phase level signal.

Note that in this analysis the single failure is the loss of one Emergency Diesel Generator (EDG), which results in a loss of one motor-driven AFW pump. [

] The peak cladding temperature for the 2.70 inch break occurs at 2125 seconds.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-53 80.0 SIntact Loop SG Level

• .Broken Loop SG Level 60.0 a

• ~ 400 0

-J

• .

• T* o U-a 20,0 -

1 0.0 ... n... UU UU iUa ammUml

• S U . U U UUU *DU E UU*

0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-11: Steam Generator Level % - 2.70-in Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-54 2.14 SNPB RAI-14 This question is from ANP-3345P, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," and is needed to conclude whether or not this LAR provides acceptable results governed by 10 CFR 50.46, GDC 36 and 38; and NUREG-800, Chapter 15.

Provide the transient plots for the 2.6 and 2.8 inch breaks.

Response

Note that the data presented for this response reflects the updated charging flow configuration discussed in SNPB RAI-15 (Section 2.15).

The transient plots for the 2.60 inch, 2.70 inch, and 2.80 inch break cases are provided below in Figure 2-12 through Figure 2-59.

4000.0 3000.0

--=Reactor Power

".. 2000.0 0.

1000.0 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-12: Reactor Power- 2.60 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-55 2500.0 2250.0*

2000.0 1750.0 *

........4 12500.0 7500.0 7500.0 250.0 5000a0 Time (s)

Figure 2-13: Primary and Secondary System Pressures - 2.60 inch Break 1.0 U-.

lime (s)

Figure 2-14: Break Void Fraction - 2.60 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-56 1000.0 800.0 600.0 Uo IU 400.0 200.0 lime (s)

Figure 2-15: Break Mass Flow Rate - 2.60 inch Break 1.0 0.8 o0.6 ........

• Loop 2A

-u-- Loop 1BA o Li.

S0.4 0.2 2000.0 4000.0 (s)e Figure 2-16: Loop Seal Void Fraction - 2.60 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-57 10000.0 5000.0 a

.0

,0 U

0.0

-5000.0 0.0~

Figure 2-17: RCS Loop Mass Flow Rate - 2.60 inch Break 3000.0 2000.0 --. Loop 1

- Loop 2 - broken S

.0 U_

1000.0 0.0 100.0 200.0 000.04000.0 50.

Figure 2-18: Main Feedwater Mass Flow Rate - 2.60 inch Break

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-58 40.0

-- a Loop I

-.

• Loop 2 - brokeni 30.0k a

m 20.0 a

10.0 1 0.00 * &*b*ll,-B*

o m

1000.0 ~2o00.0

=====================================

3000.0 4000.0 5000.0 Time (s)

Figure 2-19: Auxiliary Feedwater Mass Flow Rate - 2.60 inch Break 250000.0 200000.0

.0 2

aE 150000.0 100000.0 0.0 L-1000.0 2000.0 3000.0 4000.0 5000.0 Timeo(s)

Figure 2-20: Steam Generator Total Mass - 2.60 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-59 15.0 10.0 /-

,, /

Time (s)

Figure 2-21: High Pressure Safety Injection Mass Flow Rates - 2.60 inch Break 1.0 0.5

-- a Loop 1A

--. Loop 1B

,Loop 2A U -- 4 Loop 2B - broken L.0 0

-0.5

-1.0 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-22: Low Pressure Safety Injection Mass Flow Rates - 2.60 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP

)-3440NP and SNPB RAI-2 thru SNPB RAI-20 R*evision 1 Technical Report P age 2-60 80.0 ,1 40.0

-- u Loop 1A

--

• Loop 1B ALoop 2A

-- 4 Loop 2B - broken L.0 Sl 20.0 UE 0.04-

, I , 0kI

-2"0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-23: Safety Injection Tank Mass Flow Rates - 2.60 inch Break

E 50000.0 0.0I 1000.0 2000.0 3000.0 4000.0 5000.0 Time (a)

Figure 2-24: Reactor Vessel Mass Inventory - 2.60 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-61 12.0~

-I8.0 Thn (s)

Figure 2-25: Hot Assembly Mixture Level - 2.60 inch Break 2200.0 2000.0 1800.0 1600.0 1400.0 U-1200.0 2

1 1000.0 E

U I-800.0 60.

400.0 200.0 lime (s)

Figure 2-26: Hot Spot Cladding Temperature - 2.60 inch Break

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-62 20 .0 ........ .......... .. . . .. ........... ........... .. ........ .. .*...... . ......... ... .. .... .......-....... ..................

15.0

-. Loop 1A

• Loop 2B - broken 0

0

~ 10.0 U-0 0

5.0

• 4*---* . 444 n4-* - ...-- *-

u-*.. . .. 4- 4 0.o0U 0.0... .. ..* 1000.0 2000.0 3000.0 4000.0 5000.0 Time (S)

Figure 2-27: Charging Flow - 2.60 inch Break 4000.0 ,-

3000.0

. Reactor Power 2000.0-1000.0 0.0 1000.0 2000,0 3000.0 4000.0 5000.0 Time (s)

Figure 2-28: Reactor Power - 2.70 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-I ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-63 2 50 0 .0 .. . . . . .. .... ... . . .. . . . . . . .. .... ....... ... ... .... ... . . .. ..

2250.0 2000.0 1750.0 -u R\

/ Upper Head 3-1 3-2 1500.0 I .

72150.0 L, 500.0 ..UI,.

.. . o -U.... U... .

• ul-U d-U U *Ui it U-U*U 250.0 0.0L.. .........

0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (S)

Figure 2-29: Primary and Secondary System Pressures - 2.70 inch Break a w mm m m m n aigs mu a.

1.0 .. . . ..

0.8 U-t U Void Fraction 0,

0.4 0.2 0.0 SI*U U a- U

• U.... . . .

0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-30: Break Void Fraction - 2.70 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-64 1000.0 r-800.0

-nBreak Flow 0

~ 600.0 ii 0

0 0

IL

~(U 400.0 3.

200.0~

.'S.~g. S.4 ~ >Ug~* -u ~ ~ -U--. ~ .e~15~U..* U-.4--I-*~U-U-U 0.0O...

0.0 1000.0 2000.0 3000.0 4*000. 5000.0 Time (S)

Figure 2-31: Break Mass Flow Rate - 2.70 inch Break 1.0

.. .......r ... . ! * ......4.... . ... ....

0.8

• gLoop 1A

• Loopl1B

• Loop 2A 2 0.6 .4 Loop 2B - broken

,LI 0

I.4 0.2 0,0

• m,, t9=,.UkU, U-m* =a *1u,. m** a am .- -. a~a U. n am D U......

• a 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-32: Loop Seal Void Fraction - 2.70 inch Break

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision *1 Technical Report Page 2-65 Ul 9000.0 7000.0 5000.0 - -uLoop 1A

...* Loop 18 0

~.Loop 2A 4 Loop 28 - broken n-3000.0 -~

ii-0=

1000.0 -

- ~ ~*1-V~~ ~.' - -~ -~

-1000.0 -

-3000.0 -

-5000.0 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-33: RCS Loop Mass Flow Rate - 2.70 inch Break 30()00.0 ..... . .. .. . . . . ..

20()00.0 a'op

• Loop 2 - broken U-0 10C 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-34: Main Feedwater Mass Flow Rate - 2.70 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-66 4 0.0 .. . . ... . ... ..... .... . .. . .... .... ... . ........... .. .. ..... ..... . . . .. ...................... ...........

Loop 1 a....

a Loop 2 -broken 30.0 0

0 0

CL:20.0 a

0 0

0 10.0 0.0Sa 0.0 . a glN0im...

1000.0a a l a a. a a a 2000.0,m a. a ai a a. aa 3000.0 ai a a a a ma a a a a a 4000.0 a a a a a am 5000.0 Time (s)

Figure 2-35: Auxiliary Feedwater Mass Flow Rate - 2.70 inch Break 250000-0

.... Loop I

. Loop 2- brokn

• 200000,0 4

'S

'S 150000.0 a a

a Time (s)

Figure 2-36: Steam Generator Total Mass - 2.70 inch Break

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-67 20.0 15.0

• Loop lB p ,* Loop 2A "4Loop 2B - broken

£ tr 10.0 4

£1 4

5.0-0.00.0

• ,Um~

1000.0 2000.0 3000.0 4*0,00 5000.0 Time (s)

Figure 2-37: High Pressure Safety Injection Mass Flow Rates - 2.70 inch Break 1.0. .. . . ... .. . . . . .. .. .. . ... . .

0.5

• Loop IA

• Loop lB

• ~Loop 2,A

-* *4Loop 2B -btroken 0

~ 0.0'-~-

S 0

U.

0

-0.5

-1.00.0 1000.O 2000.0 3000.0 4000.0 5000.0 Time (S)

Figure 2-38: Low Pressure Safety Injection Mass Flow Rates - 2.70 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-68 60.0 40.0 i

• LOOp 18 SLOOP 2A "4Loop 28-b ~ken

• 20.0 0.0 * £ =m *,o~

• t,= A a a  ! Iii

-20.0,0.0 1000,0 2000.0 3000.0 4*0+00 5000.0 Time (s)

Figure 2-39: Safety Injection Tank Mass Flow Rates - 2.70 inch Break 200000.0 . . . .. .. . .. ....... ... .... ....... .... .... . .. . .. . .. . .. .. . .. .... .... ................... ............... ...

UI

-u Mass 150000.0~ -*u.

S N

.5 i U

• .-R a-E

• .al 3" U.

-a/

II,41ll - lt**

'S p I~l~l Il' 100000.0

.. Elg.

500 00 .00.0 1000.0 . ...2000.0

.. . . . . . . . . 3000.0 . . . . . .. 4000.0 .. 5000.0 Time (s)

Figure 2-40: Reactor Vessel Mass Inventory - 2.70 inch Break

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-69 12.0 eP nnU P m mm mmm.

10.0 I

u p

4

-J 8.0 0

I, P u Mixture Level U

6.0 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (e)

Figure 2-41: Hot Assembly Mixture Level - 2.70 inch Break 2000.0 p =

1500.0 -, hot rod node 45 @11.02 ft U

U X\

S1000.0-E F--

Pi

.u uP i --

500.0 0,0 1000.0 2000.0 3000,0 4000.0 5000.0 Time (s)

Figure 2-42: Hot Spot Cladding Temperature - 2.70 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-70 20.0 15.0

• Loop 1A SLoop 2B - broken n

S10.0 5.0

• UU..SSSSSS... ema.....S.S..... .. .a ***S**U* SW 0.0m 5m 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-43: Charging Flow - 2.70 inch Break 4000.0 3000.0'

-uReactor Power

• . 2000.0 01.

1000.0 0.0 ==============================================

0 1000.0 2000.0 3000.0 4000.0 5000.0 Timne (s)

Figure 2-44: Reactor Power - 2.80 inch Break

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-71 2500.0 2250.0 2000.0 1750.0 1500.0 21250.0 U

1000.0 750.0 500.0 250.0 Time (s)

Figure 2-45: Primary and Secondary System Pressures - 2.80 inch Break 1.0 0.8

-- a Void Fraction 0.8 lime (s)

Figure 2-46: Break Void Fraction - 2.80 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI.-20 Revision 1 Technical Report Page 2-72 1000.0,,,

800.0 U

600.0

.0 U

.2 U-U U

400.0 200.0 lime (s)

Figure 2-47: Break Mass Flow Rate - 2.80 inch Break 1.0 06

• 0.4 Figure 2-48: Loop Seal Voidlime (s)

Fraction - 2.80 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-73 10000.0 5000.0 -- Loop 1A

........Loop 1B

... ALOOp2A

--4 Loop 28 - broken

.0 SE 0.0

-5000.00.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-49: RCS Loop Mass Flow Rate - 2.80 inch Break 3000.0

-oLoop1 2000.0 -- Loop 2 - broken a

U.

I UE 1000.0 [

0.0 0.4 0 1000.0 2000.0 3000.0 4000.0 5000.0 Time(s)

Figure 2-50: Main Feedwater Mass Flow Rate - 2.80 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-74 40.0

--- Loop 1

--. Loop 2 - broken 30.0

.-8

'p20.0 U.

U 10.0 0.0:0 1000.0 2000.0 3000.0 4000.0 5000.0 lime (s)

Figure 2-51: Auxiliary Feedwater Mass Flow Rate - 2.80 inch Break

_--b

.0 2l UE Tmie (s)

Figure 2-52: Steam Generator Total Mass - 2.80 inch Break

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-75 20

.0i 15.0

/ -aLoop IA a /........

/ ~--

-- 4 Loop 1B Loop 2A Loop 28- broken U.

10.0 2 /

5.0F

/

0.0 0.0 U j 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-53: High Pressure Safety Injection Mass Flow Rates - 2.80 inch Break 1.0 0.5 I

-- , Loop 1A

.....* Loop 18 a ......*ALoop 2A

--4 Loop 28 - broken

.0 LU 0.0 ~ U ~ ZN... m......~ ~ .~ m 3

U

-0.5

-1.00.0 1000.0 2000.0 3000.0 4000.0 5000.0 Tuiie (s)

Figure 2-54: LOW Pressure Safety Injection Mass Flow Rates - 2.80 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-76 4 00

.O ,

300.0 200.0 ii

.0

.8 100.0 0.0

-0.0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-55: Safety Injection Tank Mass Flow Rates - 2.80 inch Break 200000.0

.0 aE

)1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-56: Reactor Vessel Mass Inventory - 2.80 inch Break

L J EL~ E*JE U L'YL. '*II E [L ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-77 12.0,,,

-I 8.0 6.0 4.0ifi 0".0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-57: Hot Assembly Mixture Level - 2.80 inch Break 2200.0 2000.0 1800.0 1600.0 1400.0 U-C) 1200.0 a)

a. 1000.0 E

C)

I-800.0 60.

400.0 200.0 0.0k0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-58: Hot Spot Cladding Temperature - 2.80 inch Break

~4J5 III LJI1~k=~ ~ [~I IL ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-78 2 0.0 ., . .... . . . . ....... . .. ... .. . .. . .......... .. . ........

15.0

• Loop 1A

.... Loop 2B - broken rr 10.0 1

LL 5.0 6- .444444444**

• 4.. 4. 4 *44i - .- "

0.00,0 n-.B-1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-59: Charging Flow - 2.80 inch Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-79 2.15 SNPB RAI-15 This question is from ANP-3345P, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," and is needed to conclude whether or not this LAR provides acceptable results governed by 10 CFR 50.46, GDC 36 and 38; and NUREG-800, Chapter 15.

What charging flow is assumed and which loops are injecting in the break spectrum analyses?

Response

It was discovered that the charging flow configuration used in the SBLOCA analysis presented in ANP-3345P Revision 1 (Reference 9) modeled all the flow from one charging pump going into the intact loops rather than using a conservative flow split with 60% of the flow going to the broken loop. The assumed charging flow injection was corrected and set to 14 gpm to Loop lA (Intact Leg) and 21 gpm to Loop 2B (Broken Leg). The S-RELAP5 primary system nodalization is shown below in Figure 2-60.

SBLOCA results for the updated charging flow configuration are presented here, as part of this RAI response. Other RAI responses in this document pertaining to SBLOCA results also use these updated results.

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-80 Figure 2-60: S-RELAP5 SBLOCA Reactor Coolant System Nodalization

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI.-20 Revision 1 Technical Report Page 2-81 Figure 2-61 shows the Peak Clad Temperature (POT) versus break size for the updated analysis. ANP-3345P Revision 1 (Reference 9) results are also shown for comparison. Table 2-13 shows the key results with this updated configuration. Table 2-14 shows the time sequence evolution. The transient plots are shown in the SNPB RAI-14 response (Section 2.14). These results show that the 2.70 inch diameter (0.040-ft2 ) break produced the highest POT of 2057 0F. This break size also produced the maximum local oxidation of less than 12%.

The 2.60 inch case produced the highest core wide oxidation of less than 0.3%.

The results of this study show that LOCA licensing limits for clad temperature, local oxidation, and core wide oxidation continue to be met.

Figure 2-61: PCT vs. Break Diameter

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-82 Table 2-13: Summary of Results - Updated Charging Flow Configuration Break diameter (in) 2.50 2.60 2.70 2.80 Break Area (ft2) 0.034 0.037 0.040 0.043 Peak Clad Temperature (0F) 1952 2030 2057 2003 Time of PCT (sec) 2575 2333 2125 1937 Time of Rupture (sec) 2257 2079 1921 1775 Transient Local Maximum Oxidation (%) 7.5666 8.5535 8.9253 7.61 24 Total Local Maximum Oxidation (%)* 9.9591 10.946 11.3178 10.0049 Core Wide Oxidation (%) 0.2440 0.2467 0.2353 0.1955 PCT Elevation (ft) 11.02 11.02 11.02 11.02 Table 2-14: Sequence of Events - ling Flow Configuration (A

1~ (A 0 L. (U C 5-I.- (A (A (A (A (A 3-(A C

u C.) C.)

"6 "6 (U 3- 0 .i~ C C ~ 0 (U 0 ~

~ 0 >

0 oc (U

0 C 0

~

(A ~

~

~

~ .i~

0

~52 0 ~

0 C.)

o 0 0

~

(A (A

~

0 (A

'~

(A 0

E~ ~-

-~

- -~

.~

m LU (U (U 4 C.,

0 C ~

C.'

U 0

o.~Cfl

> > ~ 4 0 0 (U

C 0 -

.. ~ 0. 0 0

(A O~

(U 0 ~ Cfl'-< ~

Ci~

~ LL LL C.) 0 ~ .~ ~ 2 i-C.)

0 (U ~ ~j *2 Z~5 ~ (A

0. e .E E

.~

0 0-

~ .~ 0-

= a..

~I

~u 0.- 0 0 0 0

.C 0

-J -J U)

C 0 E

., ~n A QA A I 7"2 I " q*-/A 7(}A "7*A I I~ I 4 A*

• . .J4O 'j ..s Ot t.J t. I.0 JI't IOt -- O0 -- -- - 000 IOQ LDOL LLUU LLf OI ZO/ 'ff0 --

2.60 2030 0 30 32 40 70 70 372 724 - 932 . . ..- 634 726 2328 2102 2079 2333 384 -

2.70 2057 0 28 29 38 68 68 368 668 -- 934 . .. ..- 604 670 2118 1956 1921 2125 352 -

2.80 2003 0 26 28 35 65 65 366 624 - 942 - - - 572 630 1930 1846 1775 1937 330 --

AREVA Inc.

St. Lucie Unit 2 Fuei Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-83 2.16 SNPB RAI-16 This question is from ANP-3345P, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," and is needed to conclude whether or not this LAR provides acceptable results governed by 10 CFR 50.46, GDC 36 and 38; and NUREG-800, Chapter 15.

Provide the heat transfer coefficient and steam temperature at the hot spot versus time for the 2.7 inch break. Also, show these plots for the 2.6 and 2.8 inch breaks.

Response

Note that the data presented for this response reflects the updated charging flow configuration discussed in SNPB RAI-15 (Section 2.15).

Figure 2-62 through Figure 2-67 show the heat transfer coefficient and vapor temperature at the hot spot for break cases 2.60-inch, 2.70-inch and 2.80-inch.

3.0- - - -

q -- a ~Hot Rod -Node 45 (11.02 ft) 2.0 I--

-I-1.0 p

-E -U -u-U~.~

0.0-0.0 1000.0 2000.0 Time (s) 3000.0 I 4000.0 5000.0 Figure 2-62: Heat Transfer Coefficient - 2.60-in Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-84 2000.0 ......... . ................. . . .* . .............................................. .. ...............................

A --a Hot Assy - Node 45 (11.02 ft)

~i w~

/(

1500.0

/

LL S

E II 0) t.-

U

-. C C--U -U- ..-- Uo-i-.

500.0

• *m N* CU

• U*Ci-u 0.0 ....

0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-63: Hot Spot Vapor Temperature - 2.60-in Break 3.0*

U -in Hot Rod - Node 45 (11.02 ft) 2.0-I-

2Z 1.0 -1,,

0.0 1000.0 20, 00040. 5000.0 Figure 2-64: Heat Transfer Coefficient - 2.70-in Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-85 2000.0-u Hot Assy - Node 45 (11.02 It) 1500.0 p

U 0

~ 1000.0 E

0 U

U 500.0 *l a a-am l. m 0.0 ...

0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-65: Hot Spot Vapor Temperature - 2.70-in Break 3.0 r a Hot Rod - Node 45 (11.02 ft) 2,0 I-i-

1.0

0. 00020. 00.0 4000.0 5000.0 Figure 2-66: Heat Transfer Coefficient - 2.80-in Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-86 2000.0 4 --- Hot Assy - Node 45 (11.02 It) 1500.0 3

Ui 6) 1000.0

/

a U

a.3f.4 ft U #3-U ft 500.0

• ft 0.0 ...

0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-67: Hot Spot Vapor Temperature - 2.80-in Break

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-87 2.17 SNPB RAI-17 This question is from ANP-3345P, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," and is needed to conclude whether or not this LAR provides acceptable results governed by 10 CFR 50.46, GDC 36 and 38; and NUREG-800, Chapter 15.

What is the minimum Reactor Coolant System (RCS) pressure achieved for the 2.7 inch break around the 1,900 to 2,500 second time frame? [

]. Provide an explanation.

Response

Note that the data presented for this response reflects the updated charging flow configuration discussed in SNPB RAI-15 (Section 2.15).

The ROS pressure between the 1900 and 2500 second time frame is provided in Table 2-15.

The ROS depressurizes to the SIT setpoint pressure causing the SIT discharge that begins at 2118 seconds and terminates around 2400 seconds for the 2.70 inch break case. Afterwards, the High Pressure Safety Injection (HPSI) delivers enough Emergency Core Cooling System (ECCS) flow to eventually quench the core around 3700 seconds into the transient. The smaller 2.60 inch break behaves similarly (plots provided within the response to SNPB RAI-14 in Section 2.14).

The 2.50 inch break case is the largest break size that results in the PCT occurring before the cold leg SIT discharge and is turned around by only HPSI and charging flow injection. The break spectrum provides reasonable refinement after the 2.50 inch break case to identify the 2.70 inch break case as the limiting break. The cold leg SIT discharge time compared to the PCT time is shown for the 2.50, 2.60, and 2.70 inch break cases in Figure 2-68 through Figure 2-70.

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-88 Table 2-15: Reactor Vessel Upper Head Pressure - 2.70 inch break case Time (sec) Pressure (psia) 1900 644 2000 578 2100 525 2200 470 2300 450 2400 444 2500 444 2200.0 2000.0 1800.0 1600.0

.400.

"* 1200.0

~.1000.0 a800.0 600.0 400.0 200.0 5000.0 Time (s)

Figure 2-68: PCT and Integrated SIT Flow - 2.50-in Break

LiLr*~

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-89 2200.0,,,

2000.0 1800.0 1600.0 E

1400.0

"* 1200.0

• 1000.0

• . 800.0 600.0 400.0 200.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (s)

Figure 2-69: PCT and Integrated SIT Flow - 2.60-in Break

EJ~~~J LiY'~,L~ i ~ I AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-90 2200.0,,,

2000.0 1800.0 E1600.0 1400.0

• 1200.0 01000.0 a800.0 E

60 0. __

600.0 400.0 Time (s)

Figure 2-70: PCT and Integrated SIT Flow - 2.70-in Break

JHI~J~~U LLL~UE~HL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-91 2.18 SNPB RAI..18 This question is from ANP-3345P, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," and is needed to conclude whether or not this [AR provides acceptable results governed by 10 CFR 50.46, GDC 36 and 38; and NUREG-800, Chapter 15.

Provide the moderator density feedback curve used in the analysis.

Response

The moderator density feedback curve is provided in Table 2-16.

Table 2-16: Moderator Density Feedback

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-92 2.19 SNPB RAI-19 This question is from ANP-3345P, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," and is needed to conclude whether or not this LAR provides acceptable results governed by 10 CFR 50.46, GDC 36 and 38; and NUREG-800, Chapter 15.

Provide the analysis of the effect of reactor coolant pump (RCP) operation on the limiting small break loss of coolant accident (SBLOCA) and identify the Emergency Operating Plan (EOP) timing for tripping RCPs following a SBLOCA.

Response

The effect of RCP trip on SBLOCA for St. Lucie Unit 2 was addressed during the review of Extended Power Uprate (EPU) license amendment request in response to RAI SRX(B-96 (RAI 2.8.5.6.3-1 9) (Reference 12). The response was based on a comparative evaluation with the St. Lucie Unit 1 EPU analyses for the effect of RCP operation (Reference 13). The comparative evaluation determined that the St. Lucie Unit 2 RCP trip criteria of 1736 psia pressurizer pressure for tripping one RCP in each loop and a minimum of 20 0 F subcooling for tripping all four RCPs remain unchanged for EPU. The comparative evaluation and the conclusion, with both the St. Lucie Units operating at a thermal power level of 3020 MWth and favorable minimum HPSI delivery (shown in Table 2-11 of response to SNPB RAI-1 1 in Section 2.11) for St. Lucie Unit 2, continue to remain applicable for the operation of St. Lucie Unit 2 with the fuel transition to AREVA 1 6x1 6 fuel. There are no significant changes to the plant parameters, related to the SBLOCA analysis, due to the implementation of the fuel design change to AREVA 16x16 fuel.

The St. Lucie Unit 2 RCP trip criteria in the LOCA EOP followed the recommendations of Combustion Engineering Owners Group (CEOG) Report CEN-268 (Reference 14) which is reviewed and found to be acceptable for referencing by the NRC in Reference 15. The methodology used in the Reference 14 CEOG study followed the guidelines and criteria provided in Generic Letter 83-10a. The CEOG study, applicable to all CE plants, showed that:

1. Operator has at least 2 minutes to trip the RCPs after RCP trip criterion is reached and remain in compliance with 10 CER 50.46 using Appendix K assumptions.

~.AJE RU ~  ! ~ ~.sL~A I !I AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-93

2. Using most probable best estimate analyses, Operator has at least 10 minutes to trip the RCPs after ROP trip criterion is reached and not exceed the 10 CFR 50.46 limit of 2200 0F.

The study, using the best estimate analysis, also showed that the actual time for tripping the RCPs for SBLOCA would be in excess of the above stated time of 10 minutes without exceeding the peak cladding temperature of 2200 0F. A similar conclusion was reached in the St. Lucie Unit 1 SBLOCA RCP trip analysis (Reference 13) done at 3020 MWth power, using better estimate assumptions, and used in the St. Lucie Unit 2 RCP trip comparative evaluation (Reference 12).

St. Lucie Unit 2 RCP trip strategy based on the pressurizer pressure and RCS subcooling, implemented in the EOP thus remains acceptable for the fuel transition to AREVA 16x1 6 fuel.

IU ~ I E~I ~

AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-94 2.20 SNPB RAI-20 This question is from ANP-3345P, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," and is needed to conclude whether or not this LAR provides acceptable results governed by 10 CFR 50.46, GDC 36 and 38; and NUREG-800, Chapter 15.

Show the axial core void distribution at the time of peak clad temperature (PCT) for the 2.7 inch break. Also, provide this information for the 2.6 and 2.8 inch breaks.

Response

Note that the data presented for this response reflects the updated charging flow configuration discussed in SNPB RAI-15 (Section 2.15).

Figure 2-71 through Figure 2-73 show the axial core void distribution at PCT Time for break cases 2.60-inch, 2.70-inch and 2.80-inch, respectively. [

The PCT time for each break size is as follows (Table 2-17):

able 2-17: POT Time Break Size (in) 2.70 2.80 Time of PCT (sec) I2333~ 2125 1937

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-95 Figure 2-71" Axial Core Void Distribution at PCT Time - 2.60-in Break

JHLI~IIV~U jt~tUflF~HL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 AN P-3440N P and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-96 Figure 2-72: Axial Core Void Distribution at PCT Time - 2.70-in Break

ARE VA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 2-97 Figure 2-73: Axial Core Void Distribution at POT Time - 2.80-in Break

~AJ~ ILl ~JEI~ Li~J~jL.~l FE~I IL AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 3-1 3.0 References

1. ML15233A036 (USNRC ADAMS), Letter to FPL, "St. Lucie Plant, Unit No. 2 - Request for Additional Information regarding License Amendment Request and Exemption Request regarding the Transitioning to AREVA Fuel (TAC NOS. MF5494 And MF5495),"

USNRC, September 2015.

2. ANP-3352P, Revision 0, "St. Lucie Unit 2 Fuel Transition License Amendment Request -

Technical Report," AREVA Inc., December 2014.

3. ANP-3347P, Revision 0, "St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report," ARE VA Inc., December 2014.

4. ML15002A091 (USNRC ADAMS), L-2014-366, Enclosure 1, Attachment 1, Letter to USNRC, "Application for Technical Specification Change and Exemption Request Regarding the Transition to AREVA Fuel," FPL, December 30, 2014.

5. EMF-92-11I6(P)(A), Revision 0 Supplement 1I(P)(A), Revision 0, "Generic Mechanical Design Criteria for PWR Fuel Designs," AREVA Inc., February, 2015.

6. XN-75-32(P)(A), Supplements 1, 2, 3, and 4, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, Inc., October 1983.

7. ANF-88-1 33(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation, December 1991.

8. BAW-10240(P)-A, Revision 0, "Incorporation of M5TM Properties in Framatome ANP Approved Methods," Framatome ANP, Inc., May 2004.

9. ANP-3345P, Revision 1, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," AREVA Inc., June 2015.

10. ML090960483 (USNRC ADAMS), Letter to NRC, "Response to Third Request for Additional Information Regarding ANP-10285P, "U.S. EPR Fuel Assembly Mechanical Design Topical Report,"" USNRC, April 2009

11. ANP-3346P, Revision 0, "St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report," AREVA Inc., December 2014.

12. ML12025A196 (USNRC ADAMS), L-2012-009, Letter to USNRC, "Response to NRC Reactor System Branch and Nuclear Performance Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request," FPL, January 21, 2012.

13. ML11153A048 (USNRC ADAMS), L-2011-206, Letter to USNRC, "Information Regarding Areva LOCA and Non-LOCA Methodologies Provided in Support of the St.

Lucie Unit 1 License Amendment Request for Extended Power Uprate," FPL, May 27, 2011.

'~JEELI~jE:~u Lt~L~UHf~I1L AREVA Inc.

St. Lucie Unit 2 Fuel Transition: Response to NRC Questions SRXB-RAI-1 ANP-3440NP and SNPB RAI-2 thru SNPB RAI-20 Revision 1 Technical Report Page 3-2

14. CEN-268, Rev. 1, "Justification of Trip Two/Leave Two Reactor Coolant Pump Strategy During Transients," CEOG, May 1987.

15. ML031150282 (USNRC ADAMS), USNRC Letter, "Implementation of TMI Action Item II.K.3.5, "Automatic Trip of Reactor Coolant Pumps" (Generic Letter No. 86-06)",

USNRC, May 29, 1986.

16. ANP-3396P, Revision 0, "St. Lucie Unit 2 Fuel Transition Supplemental Information to Support the LAR," AREVA Inc., March 2015.

17. ML15210A252 (USNRC ADAMS), USNRC Letter to AREVA Inc., "Final Safety Evaluation by the Office of Nuclear Reactor Regulation for Topical Report EMF-2328(P)(A), Revision 0, Supplement 1, Revision 0, "PWR [Pressurized Water Reactor]

Small Break LOCA [Loss-of-Coolant Accident] Evaluation Model, S-RELAP5 Based" (TAC No. ME8227)," USNRC, September 1,2015.