ML13200A196

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ANP-3211(NP), Rev. 1, Monticello EPU LOCA Break Spectrum Analysis for Atrium 10XM Fuel.
ML13200A196
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Site: Monticello Xcel Energy icon.png
Issue date: 07/31/2013
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AREVA NP
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Office of Nuclear Reactor Regulation
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L-MT-13-055 ANP-3211(NP), Rev 1
Download: ML13200A196 (80)


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Enclosure 19 AREVA Report ANP-321 1 (NP)

Monticello LOCA Break Spectrum Analysis for ATRIUM 1OXM Fuel (EPU/MELLLA)

Revision 1 79 pages follow

Controlled Document ANP-3211(NP)

Revision 1 Monticello EPU LOCA Break Spectrum Analysis for ATRIUM TM 1OXM Fuel July 2013 AREVA NP Inc.

A AR EVA

Controlled Document AREVA NP Inc.

ANP-3211(NP)

Revision 1 Monticello EPU LOCA Break Spectrum Analysis for ATRIUM TM 1OXM Fuel

Controlled Document AREVA NP Inc.

ANP-321 1(NP)

Revision 1 Copyright © 2013 AREVA NP Inc.

All Rights Reserved

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUMTM 1OXM Fuel Page i Nature of Changes Item Page Description and Justification

1. Page 4-5 Section 4-5 Added proprietary marks around core flow.
2. Pages 4-10 Tables 4.4 - 4.7 through The high drywell pressure was changed from "not used" to the value with 4-13 a footnote.
3. Page 4-13 Table 4.7 A time delay was added for an interlock.
4. Page 5-5 Section 5.3.3 Added a paragraph to address SF-ADS.
5. Page 6-2 Section 6.3 Added a sentence at the end of the second bullet.
6. Pages 6-4 Tables 6.2 and 7.2 and 7-4 Changed the event description from "LPCS permissive for ADS" to "RH R/LPCS permissive for ADS" and changed the event time.

Changed the event time for "ADS valves open."

7. Pages 6-5 Tables 6.3 and 6.4 and 6-6 Corrected the PCTs which were impacted by the additional delay before opening the ADS valves.

AREVA NP Inc.

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page ii Contents 1 .0 In tro d u c tio n .................................................................................................................... 1 -1 2.0 Summary of Results ....................................................................................................... 2-1 3.0 LOCA Description .......................................................................................................... 3-1 3.1 Accident Description ........................................................................................... 3-1 3.2 Acceptance Criteria ............................................................................................ 3-2 4.0 LOCA Analysis Description ............................................................................................ 4-1 4.1 Blowdown Analysis ............................................................................................. 4-1 4.2 Refill/Reflood Analysis ........................................................................................ 4-2 4.3 Heatup Analysis ................................................................................................. 4-2 4 .4 [] ................................................... 4 -3 4.4.1 Calculation Approach ........................................................................... 4-4 4.5 Plant Parameters and Initial Conditions ............................................................. 4-4 4.6 ECCS Parameters .............................................................................................. 4-5 5.0 Break Spectrum Analysis Description ............................................................................ 5-1 5.1 Lim iting Single Failure ........................................................................................ 5-1 5.2 Recirculation Line Breaks ................................................................................... 5-2 5.3 Non-Recirculation Line Breaks .......................................................................... 5-3 5.3.1 Main Steam Line Breaks ...................................................................... 5-4 5.3.2 Feedwater Line Breaks ........................................................................ 5-4 5.3.3 HPCI Line Breaks ................................................................................. 5-5 5.3.4 LPCS Line Breaks ................................................................................ 5-5 5.3.5 LPCI Line Breaks ................................................................................. 5-5 5.3.6 Reactor W ater Cleanup Line Breaks .................................................... 5-6 5.3.7 Shutdown Cooling Line Breaks ............................................................ 5-6 5.3.8 Instrument Line Breaks ........................................................................ 5-6 6.0 Recirculation Line Break LOCA Analyses ...................................................................... 6-1 6.1 Lim iting Break Analysis Results ......................................................................... 6-1 6.2 Break Location Analysis Results ........................................................................ 6-1 6.3 Break Geometry and Size Analysis Results ....................................................... 6-2 6.4 Lim iting Single-Failure Analysis Results ............................................................. 6-2 6.5 Axial Power Shape Analysis Results .................................................................. 6-2 6.6 State Point Analysis ........................................................................................... 6-2 7.0 Single-Loop Operation LOCA Analysis .......................................................................... 7-1 7.1 SLO Analysis Modeling Methodology ................................................................. 7-1 7.2 SLO Analysis Results ......................................................................................... 7-2 8.0 Long-Term Coolability ................................................................................................... 8-1 9.0 Conclusions .................................................................................................................... 9-1 10.0 References ................................................................................................................... 10-1 AREVA NP Inc.

Controlled Document Monticello EPU ANP-3211(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page iii Tables 4 .1 In itia l C o n d itio n s ............................................................................................................. 4 -7 4.2 R eactor System Param eters .......................................................................................... 4-8 4.3 ATRIUM 1OXM Fuel Assembly Parameters ................................................................... 4-9 4.4 High-Pressure Coolant Injection Parameters ............................................................... 4-10 4.5 Low-Pressure Coolant Injection Parameters ................................................................ 4-11 4.6 Low-Pressure Core Spray Parameters ........................................................................ 4-12 4.7 Automatic Depressurization System Parameters ......................................................... 4-13 4.8 Recirculation Discharge Isolation Valve Parameters ................................................... 4-14 5.1 Available ECCS for Recirculation Line Break LOCAs .................................................... 5-7 6.1 Results for Limiting TLO Recirculation Line Break 3.3 ft2 Split Pump Suction SF-LPCI Mid-Peaked Axial 102% Power [ ] ............................. 6-3 6.2 Event Times for Limiting TLO Recirculation Line Break 3.3 ft 2 Split Pump Suction SF-LPCI Mid-Peaked Axial 102% Power [ ] ............................. 6-4 6.3 TLO Recirculation Line Break Spectrum Results for 102% Power [

] S F -LP C I ................................................................................................... 6 -5 6.4 Summary of TLO Recirculation Line Break Results Highest PCT Cases ...................... 6-6 7.1 Results for Limiting SLO Recirculation Line Break 3.4 ft 2 Split Pump Suction SF-LPCI Mid-Peaked Axial 102% Power [ ] ............................. 7-3 7.2 Event Times for Limiting SLO Recirculation Line Break 3.4 ft 2 Split Pump Suction SF-LPCI Mid-Peaked Axial 102% Power [ ] ............................ 7-4 7.3 Single- and Two-Loop Operation PCT Summary ........................................................... 7-5 Figures 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model ..................................... 4-15 4 .2 [] ................................ 4 -16 4.3 R E LA X System Model .................................................................................................. 4-17 4.4 RELAX Hot Channel Model Top-Peaked Axial ............................................................ 4-18 4.5 RELAX Hot Channel Model Mid-Peaked Axial ............................................................. 4-19 4 .6 E C C S S che m atic ......................................................................................................... 4-20 4.7 Rod Average Power Distributions for 102%P [ I Mid- and T o p-P eaked ................................................................................................... 4-2 1 6.1 Limiting TLO Recirculation Line Break Upper Plenum Pressure .................................... 6-7 6.2 Limiting TLO Recirculation Line Break Total Break Flow Rate ...................................... 6-7 6.3 Limiting TLO Recirculation Line Break Core Inlet Flow Rate ......................................... 6-8 6.4 Limiting TLO Recirculation Line Break Core Outlet Flow Rate ...................................... 6-8 6.5 Limiting TLO Recirculation Line Break Intact Loop Jet Pump Exit Flow Rate ................ 6-9 6.6 Limiting TLO Recirculation Line Break Broken Loop Jet Pump Exit Flow Rate ............. 6-9 6.7 Limiting TLO Recirculation Line Break ADS Flow Rate ............................................... 6-10 6.8 Limiting TLO Recirculation Line Break HPCI Flow Rate .............................................. 6-10 6.9 Limiting TLO Recirculation Line Break LPCS Flow Rate ............................................. 6-11 6.10 Limiting TLO Recirculation Line Break Intact Loop LPCI Flow Rate ............................ 6-11 AREVA NP Inc.

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page iv 6.11 Limiting TLO Recirculation Line Break Upper Downcomer Mixture Level .................... 6-12 6.12 Limiting TLO Recirculation Line Break Lower Downcomer Mixture Level .................... 6-12 6.13 Limiting TLO Recirculation Line Break Intact Loop Discharge Line Liquid Mass ......... 6-13 6.14 Limiting TLO Recirculation Line Break Upper Plenum Liquid Mass ............................. 6-13 6.15 Limiting TLO Recirculation Line Break Lower Plenum Liquid Mass ............................. 6-14 6.16 Limiting TLO Recirculation Line Break Hot Channel Inlet Flow Rate ........................... 6-14 6.17 Limiting TLO Recirculation Line Break Hot Channel Outlet Flow Rate ........................ 6-15 6.18 Limiting TLO Recirculation Line Break Hot Channel Coolant Temperature at the H ot No d e at E O B ......................................................................................................... 6 -15 6.19 Limiting TLO Recirculation Line Break Hot Channel Quality at the Hot Node at E O B .............................................................................................................................. 6 -1 6 6.20 Limiting TLO Recirculation Line Break Hot Channel Heat Transfer Coeff. at the Hot Nod e at E O B ......................................................................................................... 6 -16 6.21 Limiting TLO Recirculation Line Break Hot Channel Reflood Junction Liquid Mass F lo w R ate ..................................................................................................................... 6 -1 7 6.22 Limiting TLO Recirculation Line Break Cladding Temperatures .................................. 6-17.

7.1 Limiting SLO Recirculation Line Break Upper Plenum Pressure ................................... 7-6 7.2 Limiting SLO Recirculation Line Break Total Break Flow Rate ...................................... 7-6 7.3 Limiting SLO Recirculation Line Break Core Inlet Flow Rate ......................................... 7-7 7.4 Limiting SLO Recirculation Line Break Core Outlet Flow Rate ...................................... 7-7 7.5 Limiting SLO Recirculation Line Break Intact Loop Jet Pump Exit Flow Rate ................ 7-8 7.6 Limiting SLO Recirculation Line Break Broken Loop Jet Pump Exit Flow Rate ............. 7-8 7.7 Limiting SLO Recirculation Line Break ADS Flow Rate ................................................. 7-9 7.8 Limiting SLO Recirculation Line Break HPCI Flow Rate ................................................ 7-9 7.9 Limiting SLO Recirculation Line Break LPCS Flow Rate ............................................. 7-10 7.10 Limiting SLO Recirculation Line Break Intact Loop LPCI Flow Rate ............................ 7-10 7.11 Limiting SLO Recirculation Line Break Upper Downcomer Mixture Level ................... 7-11 7.12 Limiting SLO Recirculation Line Break Lower Downcomer Mixture Level ................... 7-11 7.13 Limiting SLO Recirculation Line Break Intact Loop Discharge Line Liquid Mass ......... 7-12 7.14 Limiting SLO Recirculation Line Break Upper Plenum Liquid Mass ............................. 7-12 7.15 Limiting SLO Recirculation Line Break Lower Plenum Liquid Mass ............................. 7-13 7.16 Limiting SLO Recirculation Line Break Hot Channel Inlet Flow Rate ........................... 7-13 7.17 Limiting SLO Recirculation Line Break Hot Channel Outlet Flow Rate ........................ 7-14 7.18 Limiting SLO Recirculation Line Break Hot Channel Coolant Temperature at the Hot N od e a t E O B ......................................................................................................... 7 -14 7.19 Limiting SLO Recirculation Line Break Hot Channel Quality at the Hot Node at E O B .............................................................................................................................. 7 -1 5 7.20 Limiting SLO Recirculation Line Break Hot Channel Heat Transfer Coeff. at the Hot No d e at E O B ......................................................................................................... 7 -15 7.21 Limiting SLO Recirculation Line Break Hot Channel Reflood Junction Liquid Ma ss F low R ate ........................................................................................................... 7-16 7.22 Limiting SLO Recirculation Line Break Cladding Temperatures .................................. 7-16 AREVA NP Inc.

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page v Nomenclature ADS automatic depressurization system ANS American Nuclear Society BOL beginning of life BWR boiling-water reactor CFR Code of Federal Regulations CHF critical heat flux CMWR core average metal-water reaction DEG double-ended guillotine DG diesel generator ECCS emergency core cooling system EOB end of blowdown EPU extended power uprate HPCI high-pressure coolant injection LOCA loss-of-coolant accident LPCI low-pressure coolant injection LPCS low-pressure core spray MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio MWR metal-water reaction NRC Nuclear Regulatory Commission, U.S.

PCT peak cladding temperature PD pump discharge PS pump suction RDIV recirculation discharge isolation valve SF-BATT single failure of battery (DC) power SF-DGEN single failure of diesel generator SF-HPCI single failure of the HPCI system SF-LPCI single failure of an LPCI injection valve SLO single-loop operation TLO two-loop operation USAR updated safety analysis report AREVA NP Inc.

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 1-1 1.0 Introduction The results of a loss-of-coolant accident (LOCA) break spectrum analysis at extended power uprate (EPU) conditions for Monticello are documented in this report. The purpose of the break spectrum analysis is to identify the parameters that result in the highest calculated peak cladding temperature (PCT) during a postulated LOCA. The LOCA parameters addressed in this report include the following:

0 Break location a Break type (double-ended guillotine (DEG) or split)

  • Break size

The analyses documented in this report were performed with LOCA Evaluation Models developed by AREVA NP and approved for reactor licensing analyses by the U.S. Nuclear Regulatory Commission (NRC). The models and computer codes used by AREVA for LOCA analyses are collectively referred to as the EXEM BWR-2000 Evaluation Model (References 1 - 4). The EXEM BWR-2000 Evaluation Model and NRC approval are documented in Reference 1. A summary description of the LOCA analysis methodology is provided in Section 4.0. The calculations described in this report were performed in conformance with 10 CFR 50 Appendix K requirements and satisfy the event acceptance criteria identified in 10 CFR 50.46. [

The break spectrum analyses documented in this report were performed for a core composed entirely of ATRIUMTM 1OXM* fuel at beginning-of-life (BOL) conditions. Calculations assumed an initial core power of 102% of 2004 MWt, providing a licensing basis power of 2044.08 MWt.

The 2.0% increase reflects the maximum uncertainty in monitoring reactor power, as per NRC requirements. The limiting assembly in the core was assumed to be at a maximum average planar linear heat generation rate (MAPLHGR) limit of 13.1 kW/ft. Other initial conditions used in the analyses are described in Section 4.0.

This report identifies the limiting LOCA break characteristics (location, type, size, single failure and axial power shape) that will be used in future analyses to determine the MAPLHGR limit

  • ATRIUM is a trademark of AREVA NP.

AREVA NP Inc.

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 1-2 versus exposure for ATRIUM 1OXM fuel contained in Monticello. Even though the limiting break will not change with exposure or nuclear fuel design, the value of PCT calculated for any given set of break characteristics is dependent on exposure and local power peaking. Therefore, heatup analyses are performed to determine the PCT versus exposure for each nuclear design in the core. The heatup analyses are performed each cycle using the limiting boundary conditions determined in the break spectrum analysis. The maximum PCT versus exposure from the heatup analyses are documented in the MAPLHGR report.

Analyses do not support operation with one automatic depressurization system (ADS) valve out of service. [

] Limiting reactor power and core flow conditions were selected with consideration for the range of operating conditions represented by the EPU/MELLLA licensing amendment request (Reference 6). This report also presents results for single-loop operation (SLO) and long-term coolability.

AREVA NP Inc.

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 2-1 2.0 Summary of Results Based on analyses presented in this report, the limiting break characteristics are identified below.

Limiting LOCA Break Characteristics Location Recirculation suction pipe 2

Type / size Split break / 3.3 ft Single failure LPCI injection valve Axial power shape Mid-peaked Initial state 102% power/[ ]

A more detailed discussion of results is provided in Sections 6.0 - 7.0.

] The break characteristics identified in this report can be used in subsequent fuel type specific LOCA heatup analyses to determine the MAPLHGR limit appropriate for the fuel type.

The SLO LOCA analyses support operation with an ATRIUM 1OXM MAPLHGR multiplier of 0.70 applied to the normal two-loop operation MAPLHGR limit.

The long-term coolability evaluation confirms that the ECCS capacity is sufficient to maintain adequate cooling in an ATRIUM 1OXM core for an extended period after a LOCA.

AREVA NP Inc.

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUMTM 1OXM Fuel Page 2-2 While the fuel rod temperatures in the limiting plane of the hot channel during a LOCA are dependent on exposure, the factors that determine the limiting break characteristics are primarily associated with the reactor system and are not dependent on fuel-exposure characteristics. Fuel parameters that are dependent on exposure (e.g., stored energy, local peaking) have an insignificant effect on the reactor system response during a LOCA. The limiting break characteristics are determined using BOL fuel conditions for a representative ATRIUM 1OXM lattice design and conservative stored energy. These limiting break conditions are applicable for exposed fuel. Fuel exposure effects are addressed in heatup analyses performed to determine or verify MAPLHGR limits versus exposure for each fuel design.

The break spectrum analysis was performed using the NRC approved AREVA EXEM BWR 2000 LOCA methodology. A modified application approach to [ ] is presented in Section 4.4. This modification is conservative relative to the application approach for the approved methodology utilized in Reference 1. The modified application approach was communicated to the NRC in Reference 10. The NRC acknowledged the modified approach in Reference 11.

AREVA NP Inc.

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 3-1 3.0 LOCA Description 3.1 Accident Description The LOCA is described in the Code of Federal Regulations 10 CFR 50.46 as a hypothetical accident that results in a loss of reactor coolant from breaks in reactor coolant pressure boundary piping up to and including a break equivalent in size to a double-ended rupture of the largest pipe in the reactor coolant system. There is not a specifically identified cause that results in the pipe break. However, for the purpose of identifying a design basis accident, the pipe break is postulated to occur inside the primary containment before the first isolation valve.

For a boiling water reactor (BWR), a LOCA may occur over a wide spectrum of break locations and sizes. Responses to the break vary significantly over the break spectrum. The largest possible break is a double-ended rupture of a recirculation pipe; however, this is not necessarily the most severe challenge to the emergency core cooling system (ECCS). A double-ended rupture of a main steam line causes the most rapid primary system depressurization, but because of other phenomena, steam line breaks are seldom limiting with respect to the event acceptance criteria (10 CFR 50.46). In addition to break location dependence, different break sizes in the same pipe produce quite different event responses, and the largest break area is not necessarily the most severe challenge to the event acceptance criteria. Because of these complexities, an analysis covering the full range of break sizes and locations is performed to identify the limiting break characteristics.

Regardless of the initiating break characteristics, the event response is conveniently separated into three phases: the blowdown phase, the refill phase, and the reflood phase. The relative duration of each phase is strongly dependent upon the break size and location. The last two phases are often combined and will be discussed together in this report.

During the blowdown phase of a LOCA, there is a net loss of coolant inventory, an increase in fuel cladding temperature due to core flow degradation, and for the larger breaks, the core becomes fully or partially uncovered. There is a rapid decrease in pressure during the blowdown phase. During the early phase of the depressurization, the exiting coolant provides core cooling. Later in the blowdown, core cooling is provided by lower plenum flashing as the system continues to depressurize and the injection of ECCS flows. The blowdown phase is defined to end when the system reaches the pressure corresponding to the rated LPCS flow.

AREVA NP Inc.

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 3-2 In the refill phase of a LOCA, the ECCS is functioning and there is a net increase of coolant inventory. During this phase the core sprays provide core cooling and, along with low-pressure and high-pressure coolant injection (LPCI and HPCI), supply liquid to refill the lower portion of the reactor vessel. In general, the core heat transfer to the coolant is less than the fuel decay heat rate and the fuel cladding temperature continues to increase during the refill phase.

In the reflood phase, the coolant inventory has increased to the point where the mixture level reenters the core region. During the core reflood phase, cooling is provided above the mixture level by entrained reflood liquid and below the mixture level by pool boiling. Sufficient coolant eventually reaches the core hot node and the fuel cladding temperature decreases.

3.2 Acceptance Criteria A LOCA is a potentially limiting event that may place constraints on fuel design, local power peaking, and in some cases, acceptable core power level. During a LOCA, the normal transfer of heat from the fuel to the coolant is disrupted. As the liquid inventory in the reactor decreases, the decay heat and stored energy of the fuel cause a heatup of the undercooled fuel assembly.

In order to limit the amount of heat that can contribute to the heatup of the fuel assembly during a LOCA, an operating limit on the MAPLHGR is applied to each fuel assembly in the core.

The Code of Federal Regulations prescribes specific acceptance criteria (10 CFR 50.46) for a LOCA event as well as specific requirements and acceptable features for Evaluation Models (10 CFR 50 Appendix K). The conformance of the EXEM BWR-2000 LOCA Evaluation Models to Appendix K is described in Reference 1. The ECCS must be designed such that the plant response to a LOCA meets the following acceptance criteria specified in 10 CFR 50.46:

  • The calculated maximum fuel element cladding temperature shall not exceed 2200'F.
  • The calculated local oxidation of the cladding shall nowhere exceed 0.17 times the local cladding thickness.

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

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

After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

AREVA NP Inc.

Controlled Document Monticello EPU ANP-3211(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 3-3 These criteria are commonly referred to as the peak cladding temperature (PCT) criterion, the local oxidation criterion, the hydrogen generation criterion, the coolable geometry criterion, and the long-term cooling criterion. A MAPLHGR limit versus fuel exposure is established to ensure that these criteria are met. For jet pump BWRs, the most challenging criterion is that PCT must not exceed 2200 0 F. LOCA PCT results are provided in Sections 6.0 - 7.0 to determine the limiting LOCA event.

LOCA analysis results demonstrating that the PCT, local oxidation, and hydrogen generation criteria are met are provided in follow-on MAPLHGR report and cycle specific heatup analyses performed to determine MAPLHGR limits versus exposure for each fuel design. Cycle-specific heatup analyses are performed to demonstrate that the MAPLHGR limit versus exposure for the ATRIUM 1OXM fuel remains applicable for cycle-specific nuclear designs. Compliance with these three criteria ensures that a coolable geometry is maintained. Long-term coolability criterion is discussed in Section 8.0.

AREVA NP Inc.

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 forATRIUM TM 1OXM Fuel Page 4-1 4.0 LOCA Analysis Description The Evaluation Model used for the break spectrum analysis is the EXEM BWR-2000 LOCA analysis methodology described in Reference 1. The EXEM BWR-2000 methodology employs three major computer codes to evaluate the system and fuel response during all phases of a LOCA. These are the RELAX, HUXY, and RODEX2 computer codes. RELAX is used to calculate the system and hot channel response during the blowdown, refill and reflood phases of the LOCA. The HUXY code is used to perform heatup calculations for the entire LOCA, and calculates the PCT and local clad oxidation at the axial plane of interest. RODEX2 is used to determine fuel parameters (such as stored energy) for input to the other LOCA codes. The code interfaces for the LOCA methodology are illustrated in Figure 4.1.

A complete analysis for a given break size starts with the specification of fuel parameters using RODEX2 (Reference 4). RODEX2 is used to determine the initial stored energy for both the blowdown analysis (RELAX hot channel) and the heatup analysis (HUXY). This is accomplished by ensuring that the initial stored energy in RELAX and HUXY is the same or higher than that calculated by RODEX2 for the power, exposure, and fuel design being considered.

4.1 Blowdown Analysis The RELAX code (Reference 1) is used to calculate the system thermal-hydraulic response during the blowdown phase of the LOCA. For the system blowdown analysis, the core is represented by an average core channel. The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and with decay heating as required by Appendix K of 10 CFR 50. The reactor vessel nodalization for the system analysis is shown in Figure 4.3. This nodalization is consistent with that used in the topical report submitted to the NRC (Reference 1).

The RELAX blowdown analysis is performed from the time of the break initiation through the end of blowdown (EOB). The system blowdown calculation provides the upper and lower plenum transient boundary conditions for the hot channel analysis.

Following the system blowdown calculation, another RELAX analysis is performed to analyze the maximum power assembly (hot channel) of the core. The RELAX hot channel blowdown calculation determines hot channel fuel, cladding, and coolant temperatures during the blowdown phase of the LOCA. The RELAX hot channel nodalization is shown in Figure 4.4 for AREVA NP Inc.

Controlled Document Monticello EPU ANP-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 forATRIUM TM 1OXM Fuel Page 4-2 a top-peaked power shape, and in Figure 4.5 for a mid-peaked axial power shape. The hot channel analysis is performed using the system blowdown results to supply the core power and the system boundary conditions at the core inlet and exit. The initial average fuel rod temperature at the limiting plane of the hot channel is conservative relative to the average fuel rod temperature calculated by RODEX2 for operation of the ATRIUM 1OXM assembly at the MAPLHGR limit. The heat transfer coefficients and fluid conditions at the limiting plane of the RELAX hot channel calculation are used as input to the HUXY heatup analysis.

4.2 Refill/Reflood Analysis The RELAX code is also used to compute the system and hot channel hydraulic response during the refill/reflood phase of the LOCA. The RELAX system and RELAX hot channel analyses continue beyond the end of blowdown to analyze system and hot channel responses during the refill and reflood phases. The refill phase is the period when the lower plenum is filling due to ECCS injection. The reflood phase is the period when some portions of the core and hot assembly are being cooled with ECCS water entering from the lower plenum. The purpose of the RELAX calculations beyond blowdown is to determine the time when the liquid flow via upward entrainment from the bottom of the core becomes high enough at the hot node in the hot assembly to end the temperature increase of the fuel rod cladding. This event time is called the time of hot node reflood. [

] The time when the core bypass mixture level rises to the elevation of the hot node in the hot assembly is also determined.

RELAX provides a prediction of fluid inventory during the ECCS injection period. Allowing for countercurrent flow through the core and bypass, RELAX determines the refill rate of the lower plenum due to ECCS water and the subsequent reflood times for the core, hot assembly, and the core bypass. The RELAX calculations provide HUXY with the time of hot node reflood and the time when the liquid has risen in the bypass to the height of the axial plane of interest (time of bypass reflood).

4.3 Heatup Analysis The HUXY code (Reference 2) is used to perform heatup calculations for the entire LOCA transient and provides PCT and local clad oxidation at the axial plane of interest. The heat generated by metal-water reaction (MWR) is included in the HUXY analysis. HUXY is used to calculate the thermal response of each fuel rod in one axial plane of the hot channel assembly.

These calculations consider thermal-mechanical interactions within the fuel rod. The clad AREVA NP Inc.

Controlled Document Monticello EPU ANP-3211(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-3 swelling and rupture models from NUREG-0630 have been incorporated into HUXY (Reference 3). The HUXY code complies with the 10 CFR 50 Appendix K criteria for LOCA Evaluation Models.

HUXY uses the EOB time and the times of core bypass reflood and core reflood at the axial plane of interest from the RELAX analysis. Until the EOB, HUXY uses RELAX hot channel heat transfer coefficients, fluid temperatures, fluid qualities, and power. Throughout the calculations, decay power is determined based on the ANS 1971 decay heat curve plus 20% as described in Reference 1. After the EOB and prior to the time of hot node reflood, HUXY uses Appendix K spray heat transfer coefficients for the fuel rods, water channel and fuel channel. Experimental data for AREVA 1 OXI 0 fuel which supports the use of the convective heat transfer coefficients listed in Appendix K is documented in Reference 5. After the time of hot node reflood, Appendix K reflood heat transfer coefficients are used in the HUXY analysis. The principal results of a HUXY heatup analysis are the PCT and the percent local oxidation of the fuel cladding, often called the %MWR.

4.4 [

AREVA NP Inc.

Controlled Document Monticello EPU AN P-321 1(NP)

LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-4 4.4.1 Calculation Approach 4.5 Plant Parameters and Initial Conditions The LOCA break spectrum analysis is performed using plant parameters provided by Xcel Energy. Limiting reactor power and core flow conditions were selected with consideration for the range of operating conditions represented by the EPU license amendment request (Reference 6). At the EPU rated thermal power the allowable range of core flow is 99% - 105%

of rated core flow. The NRC is currently reviewing an LAR to reduce the minimum allowable core flow at rated power from 99% to 80% of rated core flow. [

Table 4.1 provides a summary of reactor initial conditions used in the break spectrum analysis.

Table 4.2 lists selected reactor system parameters.

AREVA uses a process for determining initial power distributions that produce conservative LOCA results compared to the power distributions that could exist during actual operation. The initial power distributions are based on a conservatively low MCPR operating limit and the MAPLHGR limit to be supported. The goal is to establish a MAPLHGR limit that is less restrictive than the LHGR limit and satisfies 10 CFR 50.46 acceptance criteria. When this goal is achieved some rods in the HUXY heatup analysis will have LHGRs higher than the LHGR limit. Whether a MAPLHGR limit can be supported that is less restrictive than the LHGR limit depends primarily on the plant-specific ECCS parameters used in the analysis.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 10XM Fuel Page 4-5 The radial peaking factor for the hot bundle is established through calculations that determine the maximum radial that could be achieved without violating a conservatively low MCPR operating limit when the highest planar power is at the MAPLHGR limit. The use of a low MCPR operating limit results in a high bundle power. Since MCPR depends on core flow and axial power shape, a different radial peaking factor is used for each combination of core flow and axial power shape. After the radial peaking is calculated, the axial peaking factor at the peak power plane is calculated to put the nodal power at the MAPLHGR limit. Table 4.1 summarizes the MAPLHGR limit and the MCPR operating limit that were used to establish the radial and axial peaking factors for each of the power/flow conditions that have been analyzed.

The rod average axial power profiles resulting from the application of this process for 102%P

[ ] are shown in Figure 4.7. The axial power profiles for the other power and flow conditions are similar.

The break spectrum analysis is performed for a full core of ATRIUM 1OXM fuel. Some of the key ATRIUM 1OXM fuel parameters used in the break spectrum analysis are summarized in Table 4.3.

4.6 ECCS Parameters The ECCS configuration is shown in Figure 4.6. Table 4.4 - Table 4.7 provide the important ECCS characteristics assumed in the analysis. The assumed ECCS performance has been reconciled with the revised levels of ECCS performance associated with Containment Accident Pressure (Reference 7). The ECCS is modeled as fill junctions connected to the appropriate reactor locations: LPCS injects into the upper plenum, HPCI injects into the upper downcomer, and LPCI injects into the recirculation lines.

The flow through each ECCS valve is determined based on system pressure and valve position.

Flow versus pressure for a fully open valve is obtained by linearly interpolating the pump capacity data provided in Table 4.4 - Table 4.6. No HPCI or LPCS flow is credited until the ECCS injection valves are fully open. Also, no credit for ECCS flow is assumed until ECCS pumps reach rated speed.

The ADS valves are modeled as a junction connecting the reactor steam line to the suppression pool. The flow through the ADS valves is calculated based on pressure and valve flow characteristics. The valve flow characteristics are determined such that the calculated flow is equal to the rated capacity at the reference pressure shown in Table 4.7. All three ADS valves AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision I for ATRIUM TM 1OXM Fuel Page 4-6 are assumed operable during the LOCA except when a single failure is assumed to prevent one ADS valve from opening.

In the AREVA LOCA analysis model, ECCS initiation is assumed to occur when the water level drops to the applicable level setpoint. No credit is assumed for the start of LPCS or LPCI due to high drywell pressure. [

The recirculation discharge isolation valve (RDIV) parameters are shown in Table 4.8.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-7 Table 4.1 Initial Conditions*

Reactor power (% of rated) 102 102

[

Reactor power (MWt) 2044.1 2044.1 ]

]

C Steam flow rate (MIb/hr) 8.51 8.51 Steam dome pressure (psia) 1038.7 1038.7 Core inlet enthalpy (Btu/Ib) 523.6 515.8 ATRIUM 1OXM hot assembly MAPLHGR (kW/ft) 13.1 13.1 C

The AREVA calculated heat balance is adjusted to match the heat balance at 100% power and 100%

core flow. AREVA heat balance calculations establish these initial conditions at the stated power and flow.

C ]

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-8 Table 4.2 Reactor System Parameters Parameter Value Vessel ID (in) 205 Number of fuel assemblies 484 Recirculation suction pipe area (ft 2) 3.679 Recirculation discharge pipe area (ft2) 3.679 AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-9 Table 4.3 ATRIUM 1OXM Fuel Assembly Parameters Parameter Value Fuel rod array 1Ox10 Number of fuel rods per 79 (full-length rods) assembly 12 (part-length rods)

Non-fuel rod type Water channel replaces 9 fuel rods Fuel rod OD (in) 0.4047 Active fuel length (in) 145.24 (full-length rods)

(including blankets) 75.0 (part-length rods)

Water channel outside width (in) 1.378 Fuel channel thickness (in) 0. 075 (minimum wall)

0. 100 (corner)

Fuel channel internal width (in) 5.278 AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-10 Table 4.4 High-Pressure Coolant Injection Parameters Parameter Value Coolant temperature (maximum) (°F) 127 InitiatingSignals and Setpoints Water level (in)* 422.1 High drywell pressure (psig)t 3 Time Delays Time for HPCI pump to reach rated speed and injection valve wide open (sec) 45 Delivered Coolant Flow Rate Vs. Pressure Vessel to Torus AP Flow Rate (psid) (gpm) 0 0 150 2700 1120 2700

  • Relative to vessel zero.

t The time when the drywell reaches 3 psig is not explicitly calculated in the EXEM BWR-2000 LOCA methodology. It is conservatively assumed that the high drywell pressure is reached by the time the reactor coolant inventory has decreased to the low-low water level setpoint.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-11 Table 4.5 Low-Pressure Coolant Injection Parameters Parameter Value Reactor pressure permissive for opening valves - analytical (psig) 350 Coolant temperature (maximum) (°F) 90 InitiatingSignals and Setpoints Water level (in)* 422.1 High drywell pressure (psig)t 3 Time Delays Total system delay from initiating signal until the system is ready to inject (sec) 53.2 LPCI injection valve stroke time (sec) 35*

Delivered Coolant Flow Rate Vs. Pressure Flow Rate for Flow Rate for 2 Pumps 4 Pumps Injecting Into Injecting Into Vessel to 1 Recirculation 1 Recirculation TorusA P Loop Loop (psid) (gpm) (gpm) 0 8,000 12,400 20 7,740 12,000 260 3,000 4,000 300 0 0

  • Relative to vessel zero.

t The time when the drywell reaches 3 psig is not explicitly calculated in the EXEM BWR-2000 LOCA methodology. It is conservatively assumed that the high drywell pressure is reached by the time the reactor coolant inventory has decreased to the low-low water level setpoint.

Rated LPCI flow was assumed to occur when the LPCI injection valve is greater than 50% open. In the analysis, rated LPCI flow was assumed to occur at 35 seconds.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM 1OXM Fuel M

T Page 4-12 Table 4.6 Low-Pressure Core Spray Parameters Parameter Value Reactor pressure permissive for opening valves - analytical (psig) 350 Coolant temperature (maximum) (°F) 90 InitiatingSignals and Setpoints Water level (in)* 422.1 High drywell pressure (psig)t 3 Time Delays Total system delay from initiating signal until the system is ready to inject (sec) 38 LPCS injection valve stroke time (sec) 15 Delivered Coolant Flow Rate Vs. Pressure Vessel to Flow Rate for Torus AP 1 Pump (psid) (gpm) 0 3,540 130 2,700 300 1,125 338 0 Relative to vessel zero.

t The time when the drywell reaches 3 psig is not explicitly calculated in the EXEM BWR-2000 LOCA methodology. It is conservatively assumed that the high drywell pressure is reached by the time the reactor coolant inventory has decreased to the low-low water level setpoint.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM 1OXM Fuel M

T Page 4-13 Table 4.7 Automatic Depressurization System Parameters Parameter Value Number of valves installed 3 Number of valves available* 3 Minimum flow capacity of 791,000 available valves at (Ibm/hr at psig) 1080 InitiatingSignals and Setpoints Water level (in)t 422.1 High drywell pressure (psig)* 3 Time Delays Interlock (delay time from the start of one of the six RHR or core spray pumps to delivering output pressure) (sec)5 18 ADS timer (delay time from initiating signal to time valves are open) (sec) 138

  • All 3 valves are assumed operable in the analyses except when analyzing the potential single failure of 1 ADS valve during the LOCA.

Relative to vessel zero.

The time when the drywell reaches 3 psig is not explicitly calculated in the EXEM BWR-2000 LOCA methodology. It is conservatively assumed that the high drywell pressure is reached by the time the reactor coolant inventory has decreased to the low-low water level setpoint.

5 The analysis conservatively assumes the output pressure is not achieved until one of the RHR pumps reaches rated speed.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-14 Table 4.8 Recirculation Discharge Isolation Valve Parameters Parameter Value Reactor pressure permissive for closing valves - analytical (psia) None RDIV stroke time (sec) 35 AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 forATRIUM TM 1OXM Fuel Page 4-15 Neutronic Data (CASMO, MICROBURN)

  • The hot assembly calculation may be combined with the system calculation or executed Peak Cladding Temperature, separately Metal Water Reaction Figure 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-16 Figure 4.2 [

I AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-17 Figure 4.3 RELAX System Model AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-18 Figure 4.4 RELAX Hot Channel Model Top-Peaked Axial AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM 1OXM Fuel M

T Page 4-19 Figure 4.5 RELAX Hot Channel Model Mid-Peaked Axial AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-20 Figure 4.6 ECCS Schematic AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 4-21 Figure 4.7 Rod Average Power Distributions for 102%P [ ]

Mid- and Top-Peaked AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 forATRIUM TM 1OXM Fuel Page 5-1 5.0 Break Spectrum Analysis Description The objective of these LOCA analyses is to ensure that the limiting break location, break type, break size, and ECCS single failure are identified. The LOCA response scenario varies considerably over the spectrum of break locations. Potential break locations have been separated into two groups: recirculation line breaks and non-recirculation line breaks. The basis for the break locations and potentially limiting single failures analyzed in this report is described in the following sections.

5.1 Limiting Single Failure Regulatory requirements specify that the LOCA analysis be performed assuming that all offsite power supplies are lost instantaneously and that only safety grade systems and components are available. In addition, regulatory requirements also specify that the most limiting single failure of ECCS equipment must be assumed in the LOCA analysis. The term "most limiting" refers to the ECCS equipment failure that produces the greatest challenge to event acceptance criteria. The limiting single failure can be a common power supply, an injection valve, a system pump, or system initiation logic. The most limiting single failure may vary with break size and location. The potential limiting single failures identified in the USAR (Reference 8) are shown below:

  • DC power (SF-BATT)
  • Diesel generator (SF-DGEN)
  • LPCI injection valve (SF-LPCI) 0 High-pressure coolant injection system (SF-HPCI) a ADS valve (SF-ADS)

Reference 8 has not yet been revised based on EPU. However, the review Xcel Energy performed in support of EPU did not prompt the establishment of any new limiting failure to consider (Reference 7). All three ADS valves are assumed operable in the analyses that do not consider SF-ADS. The single failures and the available ECCS for each failure assumed in these analyses are summarized in Table 5.1. Other potential failures are not specifically considered because they result in as much or more ECCS capacity.

A comparison of the systems remaining for each of the assumed failures shows that the diesel generator failure (SF-DGEN) always has more available ECCS capacity than the battery failure (SF-BATT). Therefore, PCT results obtained for SF-BATT bound those obtained for SF-DGEN.

Similarly, a comparison of the systems remaining for HPCI failure (SF-HPCI) shows that it also always provides more available ECCS capacity than the SF-BATT. Therefore, PCT results AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM M 1OXM Fuel T

Page 5-2 obtained for SF-BATT bound those obtained for SF-HPCI. As a result, results reported here are those considering SF-BATT, SF-LPCI, and SF-ADS.

5.2 Recirculation Line Breaks The response during a recirculation line LOCA is dependent on break size. The rate of reactor vessel depressurization decreases as the break size decreases. The high pressure ECCS and ADS will assist in reducing the reactor vessel pressure to the pressure where the LPCI and LPCS flows start. For large breaks, rated LPCS and LPCI flow is generally reached before or shortly after the time when the ADS valves open so the ADS system is not required to mitigate the LOCA. ADS operation is an important emergency system for small breaks where it assists in depressurizing the reactor system faster, and thereby reduces the time required to reach rated LPCS and LPCI flow.

The two largest flow resistances in the recirculation piping are the recirculation pump and the jet pump nozzle. For breaks in the discharge piping (PD), there is a major flow resistance in both flow paths from the reactor vessel to the break. For breaks in the suction piping (PS), the major flow resistances are in the same flow path from the vessel to the break. As a result, pump suction side breaks experience a more rapid blowdown, which tends to make the event more severe. For suction side breaks, the recirculation discharge isolation valve on the broken loop closes which allows the LPCI flow to fill the discharge piping and supply flow to the lower plenum and core. For discharge side breaks with break areas > 0.4 ft 2, the LPCI Selection Logic directs all available LPCI flow to the intact loop. No LPCI flow is credited for breaks

< 0.4 ft2. Both suction and discharge recirculation pipe breaks are considered in the break spectrum analysis.

Two break types (geometries) are considered for the recirculation line break. The two types are the double-ended guillotine (DEG) break and the split break.

For a DEG break, the piping is assumed to be completely severed resulting in two independent flow paths to the containment. The DEG break is modeled by setting the break area (at both ends of the pipe) equal to the full pipe cross-sectional area and varying the discharge coefficient between 1.0 and 0.4. The range of discharge coefficients is used to cover uncertainty in the actual geometry at the break. Discharge coefficients below 0.4 are unrealistic and not considered in the EXEM BWR-2000 methodology. The most limiting DEG break is determined by varying the discharge coefficient. The labeling convention for guillotine breaks is to list the AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 5-3 discharge coefficient before DEG. For example, a guillotine break with a discharge coefficient of 0.8 on the suction side of the recirculation pump would be labeled as 0.8DEGPS.

A split type break is assumed to be a longitudinal opening or hole in the piping that results in a single break flow path to the containment. Appendix K of 10 CFR 50 defines the cross-sectional area of the piping as the maximum split break area required for analysis. The labeling convention for split breaks is to list the flow area using the letter "P" instead of a period. For example, a split break with a flow area of 3.5 ft 2 on the suction side of the recirculation pump would be labeled as 3P5FT2PS. These labeling conventions for double-ended guillotine and split breaks are typically used in figures such as those in Section 7.0 of this report.

Break types, break sizes and single failures are analyzed for both suction and discharge recirculation line breaks.

Section 6.0 provides a description and results summary for breaks in the recirculation line.

5.3 Non-RecirculationLine Breaks In addition to breaks in the recirculation line, breaks in other reactor coolant system piping must be considered in the LOCA break spectrum analysis. Although the recirculation line large breaks result in the largest coolant inventory loss, they do not necessarily result in the most severe challenge to event acceptance criteria. The double-ended rupture of a main steam line is expected to result in the fastest depressurization of the reactor vessel. Special consideration is required when the postulated break occurs in ECCS piping. Although ECCS piping breaks are small relative to a recirculation pipe DEG break, the potential to disable an ECCS system increases their severity.

The following sections address potential LOCAs due to breaks in non-recirculation line piping.

Non-recirculation line breaks outside of the containment are inherently less challenging to fuel limits than breaks inside the containment. For breaks outside containment, isolation or check valve closure will terminate break flow prior to the loss of significant liquid inventory and the core will remain covered. If high-pressure coolant inventory makeup cannot be reestablished, ADS actuation may become necessary. [

] Although analyses of breaks outside containment may be required to AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 5-4 address non-fuel related regulatory requirements, these breaks are not limiting relative to fuel acceptance criteria such as PCT.

5.3.1 Main Steam Line Breaks A steam line break inside containment is assumed to occur between the reactor vessel and the inboard main steam line isolation valve (MSIV) upstream of the flow limiters. The break results in high steam flow out of the broken line and into the containment. Prior to MSIV closure, a steam line break also results in high steam flow in the intact steam lines as they feed the break via the steam line manifold. A steam line break inside containment results in a rapid depressurization of the reactor vessel. Initially the break flow will be high quality steam; however, the rapid depressurization produces a water level swell that results in liquid discharge at the break. For steam line breaks, the largest break size is most limiting because it results in the most level swell and liquid loss out of the break.

]

5.3.2 Feedwater Line Breaks I

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 5-5 5.3.3 HPCI Line Breaks The HPCI injection line is connected to the feedwater line outside of the containment.

]

The HPCI steam supply line is connected to the main steam line inside of containment.

C]

HPCI and ADS are important for small breaks. With an HPCI line break, failure of an ADS valve is a potentially limiting single failure. [

5.3.4 LPCS Line BreaksO A break in the LPCS line is expected to have many characteristics similar to [

] . However, some characteristics of the LPCS line break are unique and are not addressed in other LOCA analyses. Two important differences from other LOCA analyses are that the break flow will exit from the region inside the core shroud and the break will disable one LPCS system. The LPCS line break is assumed to occur just outside the reactor vessel. [

5.3.5 LPCI Line Breaks The LPCI injection lines are connected to the larger recirculation discharge lines. [

]

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 5-6 5.3.6 Reactor Water Cleanup Line Breaks The extraction line is connected to a recirculation suction line with an additional connection to the vessel bottom head. [

The return line is connected to the feedwater line; [

I.

5.3.7 Shutdown Coolinq Line Breaks The shutdown cooling suction piping is connected to a recirculation suction line and the shutdown cooling return line is connected to a recirculation discharge line. [

5.3.8 Instrument Line Breaks

[

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 5-7 Table 5.1 Available ECCS for Recirculation Line Break LOCAs Assumed Systems Failure Remaining*, t, Battery 3 ADS + 1 LPCS + 2 LPCI (SF-BATT)

LPCI injection valve 3 ADS + 2 LPCS + 1 HPCI (SF-LPCI)

Diesel generator 3 ADS + 1 LPCS + 1 HPCI + 2 LPCI (SF-DGEN)

HPCI system 3 ADS + 2 LPCS + 4 LPCI (SF-HPCI)

ADS valve 2 ADS + 2 LPCS + 1 HPCI + 4 LPCI (SF-ADS)

Systems remaining, as identified in this table for recirculation line breaks, are applicable to all non-ECCS line breaks. For a LOCA from an ECCS line break, the systems remaining are those listed for recirculation breaks, less the ECCS in which the break is assumed t 2 LPCI means two RHR pumps injecting to the intact loop, 4 LPCI means four RHR pumps injecting to the intact loop.

Loop selection logic directs all available LPCI flow to the intact loop for breaks z 0.4 ft2. No LPCI flow is credited for breaks < 0.4 ft2 .

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LOCA Break Spectrum Analysis Revision 1 for ATRIUMTM 10XM Fuel Page 6-1 6.0 Recirculation Line Break LOCA Analyses The largest diameter recirculation system pipes are the suction line between the reactor vessel and the recirculation pump and the discharge line between the recirculation pump and the riser manifold ring. LOCA analyses are performed for breaks in both of these locations with consideration for both DEG and split break geometries. The break sizes considered included DEG breaks with discharge coefficients from 1.0 to 0.4 and split breaks with areas ranging between the full pipe area and 0.05 ft2 . As discussed in Section 5.0, the single failures considered in the recirculation line break analyses are SF-BATT, SF-LPCI, and SF-ADS.

6.1 Limiting Break Analysis Results 2

The analyses demonstrate that the limiting (highest PCT) recirculation line break is the 3.3 ft split break in the pump suction piping with an SF-LPCI single failure and a mid-peaked axial power shape when operating at 102% rated core power [ ] . The PCT is 2130*F. The key results and event times for this limiting break are provided in Table 6.1 and Table 6.2, respectively. Figure 6.1 - Figure 6.21 provide plots of key parameters from the RELAX system and hot channel analyses. A plot of cladding temperature versus time in the hot assembly from the HUXY heatup analysis is provided in Figure 6.22.

Table 6.3 details SF-LPCI results obtained from calculations performed for the range of break sizes, break locations, and axial power shapes that were considered when operating at 102%

rated core power [ ]. Calculations were performed for this same range for each combination of initial operating state point and single failure. Table 6.4 provides a summary of the highest PCT recirculation line break calculations for each of the single failures, state points, and axial power shapes. The results of the break analyses are discussed in the following sections.

6.2 Break Location Analysis Results Table 6.4 shows that the maximum PCT calculated for a recirculation line break occurs in the pump suction piping.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 10XM Fuel Page 6-2 6.3 Break Geometry and Size Analysis Results Recirculation line break PCT results versus break geometry (DEG or split) and size were performed for

  • DEG breaks with discharge coefficients of 1.0, 0.8, 0.6, and 0.4.

2 2

  • Split breaks ranging in size from the full pipe diameter to 0.1 ft in increments of 0.1 ft and a final size of 0.05 ft 2. Based on the results for small breaks, the break size interval may be reduced to 0.01 ft2 in order to assure the limiting break size has been identified.

Table 6.4 shows that the maximum PCT calculated for a recirculation line break occurs for a split break of 3.3 ft 2.

6.4 Limiting Single-FailureAnalysis Results The results in Table 6.4 show that the limiting single-failure is SF-LPCI.

6.5 Axial Power Shape Analysis Results The results in Table 6.4 show that the mid-peaked axial power shape is limiting compared to the top-peaked shape analyses for the limiting break size.

6.6 State PointAnalysis Table 6.4 shows that [ ] was the limiting state point for the recirculation line breaks.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 6-3 Table 6.1 Results for Limiting TLO Recirculation Line Break 3.3 ft 2 Split Pump Suction SF-LPCI Mid-Peaked Axial 102% Power [ I PCT 2130°F Maximum local MWR 3.96%

Maximum planar average MWR 1.12%

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 10XM Fuel Page 6-4 Table 6.2 Event Times for Limiting TLO Recirculation Line Break 3.3 ft 2 Split Pump Suction SF-LPCl Mid-Peaked Axial 102% Power [

Event Time (sec)

Initiate break 0.0 Initiate scram 1.5 RDIV pressure permissive 4.8 Low-low liquid level, L2 (422.1 in) 6.0 Jet pump uncovers 6.3 Lower plenum flashes 7.2 Recirculation suction uncovers 8.7 MSIV fully closed 9.9 Diesel generators started 15.0 Power at LPCS injection valves 18.2 RDIV starts to close 21.4 LPCS valve pressure permissive 25.5 LPCS valve starts to open 25.5 LPCS high-pressure cutoff 26.0 RHR/LPCS permissive for ADS 33.0 LPCS pump at rated speed 38.0 Blowdown ends 39.5 LPCS valve open 40.5 LPCS flow starts 40.5 RDIV closed 56.4 Bypass reflood 130.1 Core reflood 141.1 PCT 141.1 ADS valves open 171.0 AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 6-5 Table 6.3 TLO Recirculation Line Break Spectrum Results for 102% Power [ ] SF-LPCI I I AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 6-6 Table 6.4 Summary of TLO Recirculation Line Break Results Highest PCT Cases AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 6-7 Figure 6.1 Limiting TLO Recirculation Line Break Upper Plenum Pressure Figure 6.2 Limiting TLO Recirculation Line Break Total Break Flow Rate AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 6-8 Figure 6.3 Limiting TLO Recirculation Line Break Core Inlet Flow Rate Figure 6.4 Limiting TLO Recirculation Line Break Core Outlet Flow Rate AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUMTM 1OXM Fuel Page 6-9 Figure 6.5 Limiting TLO Recirculation Line Break Intact Loop Jet Pump Exit Flow Rate Figure 6.6 Limiting TLO Recirculation Line Break Broken Loop Jet Pump Exit Flow Rate AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 forATRIUMTM 1OXM Fuel Page 6-10 ca 0

_j U-U)

I I i I i I i I I i I i I i I O 20 40 60 80 100 120 140 160 TIME (SEC)

Figure 6.7 Limiting TLO Recirculation Line Break ADS Flow Rate 0

_j U -

I I I I I t I I  ! I 0 20 40 60 80 100 120 140 160 TIME (SEC)

Figure 6.8 Limiting TLO Recirculation Line Break HPCI Flow Rate AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM 1OXM Fuel M

T Page 6-11 I- 'I' I

-I 6

6 0:

0 uco CLe 0 20 40 60 80 100 120 140 160 TIME (SEC)

Figure 6.9 Limiting TLO Recirculation Line Break LPCS Flow Rate

-i TLI 0

U-Lq 9

I i I i I I i I i I i I I i I 0 20 40 60 80 100 120 140 160 TIME (SEC)

Figure 6.10 Limiting TLO Recirculation Line Break Intact Loop LPCI Flow Rate AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 6-12 l- -]

EL w

0 0

ix M

0L 0L w1 20 40 60 80 100 120 140 160 TIME (SEC)

Figure 6.11 Limiting TLO Recirculation Line Break Upper Downcomer Mixture Level I- -1 7

-ii Ix W9 7

0 C-)

z 0:

0 0

-j I I I I I I I I I 0 20 40 60 80 100 120 140 160 TIME (SEC)

Figure 6.12 Limiting TLO Recirculation Line Break Lower Downcomer Mixture Level AREVA NP Inc

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 6-13 I- -1

-j~

w TIME (SEC)

Figure 6.13 Limiting TLO Recirculation Line Break Intact Loop Discharge Line Liquid Mass

'I ]1 Ui, zw Q-I I . I . I . I . I . I I i I 0 20 40 60 80 100 120 140 160 TIME (SEC)

Figure 6.14 Limiting TLO Recirculation Line Break Upper Plenum Liquid Mass AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM 1OXM Fuel M

T Page 6-14 I -1 z

0.,

I I I I I I I 0 20 40 60 80 100 120 140 160 TIME (SEC)

Figure 6.15 Limiting TLO Recirculation Line Break w Lower Plenum Liquid Mass Li ]

7 0

C-j LL T

o I I I I I I I I -

0 20 40 60 80 100 120 140 160 TIME (SEC)

Figure 6.16 Limiting TLO Recirculation Line Break Hot Channel Inlet Flow Rate AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 6-15 I- I . I . I I I . I . I . I . I

-]

M wo 01 LL c 0

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,-,o

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Figure 6.17 Limiting TLO Recirculation Line Break Hot Channel Outlet Flow Rate Figure 6.18 Limiting TLO Recirculation Line Break Hot Channel Coolant Temperature at the Hot Node at EOB AREVA NP Inc.

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T Page 6-16 Figure 6.19 Limiting TLO Recirculation Line Break Hot Channel Quality at the Hot Node at EOB Figure 6.20 Limiting TLO Recirculation Line Break Hot Channel Heat Transfer Coeff. at the Hot Node at EOB AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 6-17 Figure 6.21 Limiting TLO Recirculation Line Break Hot Channel Reflood Junction Liquid Mass Flow Rate El 3000

]1 F-IOO Time (sec)

Figure 6.22 Limiting TLO Recirculation Line Break Cladding Temperatures AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 7-1 7.0 Single-Loop Operation LOCA Analysis During SLO, the pump in one recirculation loop is not operating. A break may occur in either loop, but results from a break in the inactive loop would be similar to those from a two-loop operation break. If a break occurs in the inactive loop during SLO, the intact active loop flow to the reactor vessel would continue during the recirculation pump coastdown period and would provide core cooling similar to that which would occur in breaks during TLO. The system response would be similar to that resulting from an equal-sized break during two-loop operation.

A break in the active loop during SLO results in a more rapid loss of core flow and earlier degraded core conditions relative to those from a break in the inactive loop. Therefore, only breaks in the active recirculation loop are analyzed.

A break in the active recirculation loop during SLO will result in an earlier loss of core heat transfer relative to a similar break occurring during two-loop operation. This occurs because there will be an immediate loss of jet pump drive flow. Therefore, fuel rod surface temperatures will increase faster in an SLO LOCA relative to a TLO LOCA. Also, the early loss of core heat transfer will result in higher stored energy in the fuel rods at the start of the heatup. The increased severity of an SLO LOCA can be reduced by applying an SLO multiplier to the two-loop MAPLHGR limits. [

]

7.1 SLO Analysis Modeling Methodology AREVA NP Inc.

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T Page 7-2 I

I 7.2 SLO Analysis Results

[

The SLO analyses are performed with a 0.70 multiplier applied to the two-loop MAPLHGR limit resulting in an SLO MAPLHGR limit of 9.17 kW/ft. The analyses are performed at BOL fuel conditions. The limiting SLO LOCA is the 3.4 ft2 split pump suction line break with SF-LPCI and a mid-peaked axial power shape. The PCT for this case is 1932°F. Other key results and event times for the limiting SLO LOCA are provided in Table 7.1 and Table 7.2 respectively.

Figure 7.1 - Figure 7.21 show important RELAX system and hot channel results from the SLO limiting LOCA analysis. Figure 7.22 shows the cladding surface temperature for the limiting rod as calculated by HUXY.

A comparison of the limiting SLO and the limiting two-loop results is provided in Table 7.3. The results in Table 7.3 show that the limiting two-loop LOCA PCT bounds the limiting SLO PCT when a 0.70 multiplier is applied to the two-loop MAPLHGR limit.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUMTM 1OXM Fuel Page 7-3 Table 7.1 Results for Limiting SLO Recirculation Line Break 3.4 ft 2 Split Pump Suction SF-LPCI Mid-Peaked Axial 102% Power [ I PCT 1932 0 F Maximum local MWR 1.99%

Maximum planar average MWR 0.68%

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LOCA Break Spectrum Analysis Revision 1 forATRIUM TM 10XM Fuel Page 7-4 Table 7.2 Event Times for Limiting SLO Recirculation Line Break 3.4 ft2 Split Pump Suction SF-LPCI Mid-Peaked Axial 102% Power [

Event Time (sec)

Initiate break 0.0 Initiate scram 1.5 RDIV pressure permissive 4.7 Low-low liquid level, L2 (422.1 in) 6.0 Jet pump uncovers 6.3 Lower plenum flashes 7.0 Recirculation suction uncovers 8.6 MSIV fully closed 9.9 Diesel generators started 15.0 Power at LPCS injection valves 18.2 RDIV starts to close 21.4 LPCS valve pressure permissive 24.5 LPCS valve starts to open 24.5 LPCS high-pressure cutoff 25.0 RHR/LPCS permissive for ADS 33.0 Blowdown ends 37.7 LPCS pump at rated speed 38.0 LPCS valve open 39.5 LPCS flow starts 39.5 RDIV closed 56.4 Bypass reflood 137.1 Core reflood 137.3 PCT 137.3 ADS valves open 171.0 AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 7-5 Table 7.3 Single- and Two-Loop Operation PCT Summary POT PCT Operation Limiting Case (OF)

Single-loop 3.4 ft 2 split pump suction mid-peaked SF-LPCI 1932 Two-loop 3.3 ft 2 split pump suction mid-peaked SF-LPCI 2130 AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM 1OXM Fuel M

T Page 7-6 Figure 7.1 Limiting SLO Recirculation Line Break Upper Plenum Pressure Figure 7.2 Limiting SLO Recirculation Line Break Total Break Flow Rate AREVA NP Inc.

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T Page 7-7 Figure 7.3 Limiting SLO Recirculation Line Break Core Inlet Flow Rate Figure 7.4 Limiting SLO Recirculation Line Break Core Outlet Flow Rate AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 forATRIUM TM 1OXM Fuel Page 7-8 Figure 7.5 Limiting SLO Recirculation Line Break Intact Loop Jet Pump Exit Flow Rate Figure 7.6 Limiting SLO Recirculation Line Break Broken Loop Jet Pump Exit Flow Rate AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 7-9 I i I i I i I I I I i I i 0

-J

  • -o U-01 0 20 40 60 80 100 120 140 TIME (SEC)

Figure 7.7 Limiting SLO Recirculation Line Break ADS Flow Rate LI ]

0 U-5 uý I i I I I I I I I I 0 20 40 60 80 100 120 140 TIME (SEC)

Figure 7.8 Limiting SLO Recirculation Line Break HPCI Flow Rate AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 7-10 i I I i I I I I I i I i 0

0 1L co Ii I i I i II I I I I i I 0 20 40 60 80 100 120 140 TIME (SEC)

Figure 7.9 Limiting SLO Recirculation Line Break LPCS Flow Rate E ]

5-0.

-j W U-I i I i I i I i I i I i I i I 0 20 40 60 80 100 120 140 TIME (SEC)

Figure 7.10 Limiting SLO Recirculation Line Break Intact Loop LPCI Flow Rate AREVA NP Inc.

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T Page 7-11 C]

-1 w

x_

2 0o z

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w1 0

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z 0

a 0 20 40 60 80 100 120 140 TIME (SEC)

Figure 7.12 Limiting SLO Recirculation Line Break Lower Downcomer Mixture Level AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 7-12 El -1 0

DJ I . . . I . I . I . I . I . I 0 20 40 60 80 100 120 140 TIME (SEC)

Figure 7.13 Limiting SLO Recirculation Line Break Intact Loop Discharge Line Liquid Mass El ]

Zd) z w

0LI CL TIME (SEC)

Figure 7.14 Limiting SLO Recirculation Line Break Upper Plenum Liquid Mass AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 7-13 I- -J mCd JCL Wz W

0-.

j I . I . I . . . I . I . I . I 0 20 40 60 80 100 120 140 TIME (SEC)

Figure 7.15 Limiting SLO Recirculation Line Break Lower Plenum Liquid Mass LI -1 0

J0 k-I-J LI- F 0

.i r i i 0 20 40 60 80 100 120 140 TIME (SEC)

Figure 7.16 Limiting SLO Recirculation Line Break Hot Channel Inlet Flow Rate AREVA NP Inc.

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T Page 7-14

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0 z

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Figure 7.17 Limiting SLO Recirculation Line Break Hot Channel Outlet Flow Rate Figure 7.18 Limiting SLO Recirculation Line Break Hot Channel Coolant Temperature at the Hot Node at EOB AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 7-15 Figure 7.19 Limiting SLO Recirculation Line Break Hot Channel Quality at the Hot Node at EOB Figure 7.20 Limiting SLO Recirculation Line Break Hot Channel Heat Transfer Coeff. at the Hot Node at EOB AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 7-16 Figure 7.21 Limiting SLO Recirculation Line Break Hot Channel Reflood Junction Liquid Mass Flow Rate E- 2000

-1 1500 1000 500 100 200 Time (see)

Figure 7.22 Limiting SLO Recirculation Line Break Cladding Temperatures AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 8-1 8.0 Long-Term Coolability Long-term coolability addresses the issue of reflooding the core and maintaining a water level adequate to cool the core and remove decay heat for an extended time period following a LOCA. For non-recirculation line breaks, the core can be reflooded to the top of the active fuel and be adequately cooled indefinitely. For recirculation line breaks, the core will initially remain covered following reflood due to the static head provided by the water filling the jet pumps to a level of approximately two-thirds core height. Eventually, the heat flux in the core will not be adequate to maintain a two-phase water level over the entire length of the core. Beyond this time, the upper third of the core will remain wetted and adequately cooled by core spray.

Maintaining water level at two-thirds core height with one core spray system operating is sufficient to maintain long-term coolability as demonstrated by the NSSS vendor (Reference 9).

Since fuel temperatures during long-term cooling are low relative to the PCT and are not significantly affected by fuel design, this conclusion is applicable to ATRIUM 1OXM fuel.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 1OXM Fuel Page 9-1 9.0 Conclusions The major conclusions of this LOCA break spectrum analysis are:

The limiting recirculation line break is a 3.3 ft 2 split break in the pump suction piping with single failure SF-LPCI and a mid-peaked axial shape when operating at 102% rated core power [ I.

The limiting break analysis identified above satisfies all the acceptance criteria specified in 10 CFR 50.46. The analysis is performed in accordance with 10 CFR 50.46 Appendix K requirements.

The MAPLHGR limit multiplier for SLO is 0.70 for ATRIUM 1OXM fuel. This multiplier ensures that a LOCA from SLO is less limiting than a LOCA from two-loop operation.

The limiting break characteristics determined in this report can be referenced and used in future Monticello analyses to establish the MAPLHGR limit versus exposure for ATRIUM 1OXM fuel.

AREVA NP Inc.

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LOCA Break Spectrum Analysis Revision 1 for ATRIUM TM 10XM Fuel Page 10-1 10.0 References

1. EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001.
2. ' XN-CC-33(A) Revision 1, HUXY. A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option Users Manual, Exxon Nuclear Company, November 1975.
3. XN-NF-82-07(P)(A) Revision 1, Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, Exxon Nuclear Company, November 1982.
4. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal -

MechanicalResponse Evaluation Model, Exxon Nuclear Company, March 1984.

5. EMF-2292(P)(A) Revision 0, ATRIUM TM-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
6. Letter, T.J. O'Connor (NSPM) to Document Control Desk (NRC), "License Amendment Request: Extended Power Uprate (TAC MD9990)," L-MT-08-052, November 5, 2008 (ADAMS Accession No. ML083230111).
7. Letter, M.A. Schmimel (NSPM) to Document Control Desk (NRC), "Monticello Extended Power Uprate and Maximum Extended Load Line Limit Analysis Plus Licensing Amendment Requests: Supplement to Address SECY 11-0014 Use of Containment Accident Pressure (TAC Nos. MD9990 and ME3145)," L-MT-12-082, September 28, 2012 (ADAMS Accession No. ML12276A057).
8. Updated Safety Analysis Report Monticello Nuclear Plant, Revision 28.
9. NEDO-20566A, General Electric Company Analytical Model for Loss of CoolantAnalysis in Accordance with 10CFR50 Appendix K, September 1986.
10. Letter, P. Salas (AREVA) to Document Control Desk (NRC), "Proprietary Viewgraphs and Meeting Summary for Closed Meeting on Application of the EXEM BWR-2000 ECCS Evaluation Methodology," NRC: 1:096, September 22, 2011.
11. Letter, T.J. McGinty (NRC) to P. Salas (AREVA), "Response to AREVA NP, Inc.

(AREVA) Proposed Analysis Approach for Its EXEM Boiling Water Reactor (BWR)-2000 Emergency Core Cooling System (ECCS) Evaluation Model," July 5, 2012.

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