ML15251A387
| ML15251A387 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 08/31/2015 |
| From: | Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15251A394 | List:
|
| References | |
| RS-15-237 ANP-3328NP, Rev 1 | |
| Download: ML15251A387 (82) | |
Text
Attachment 12 LOCA Break Spectrum Analysis Report (Non-Proprietary Version)
A ARE VA Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM IOXM Fuel August 2015 AN P-3328N P Revision 1
© 2015 AREVA Inc.
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ANP-3328NP Revision 1 Copyright © 2015 AREVA Inc.
All Rights Reserved
Controiied Doc-m,.
Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Pagei Nature of Changes Section(s) or Item Page(s)
Description and Justification I
6-5, 6-6 Modified proprietary brackets AREVA Inc.
Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM Fuel Page ii Contents PaQe 1.0 Introduction.................................................................................. 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 Limiting 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 Water 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 Limiting 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 Limiting 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 Inc.
Controt~ed Document Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision I for ATRIUM IOXM Fuel Page iii List of Tables Table 4.1 Initial Conditions....................................................................... 4-7 Table 4.2 Reactor System Parameters.......................................................... 4-8 Table 4.3 ATRIUM I0XM Fuel Assembly Parameters......................................... 4-9 Table 4.4 High-Pressure Coolant Injection Parameters...................................... 4-10 Table 4.5 Low-Pressure Coolant Injection Parameters....................................... 4-11 Table 4.6 Low-Pressure Core Spray Parameters............................................. 4-12 Table 4.7 Automatic Depressurization System Parameters.................................. 4-13 Table 4.8 Recirculation Discharge Isolation Valve and LPCI Loop Selection Logic Parameters........................................................................... 4-14 Table 5.1 Available ECCS for Recirculation Line Break LOCAs............................... 5-7 Table 6.1 Results for Limiting TLO Recirculation Line Break 0.13 ft2 Split Pump Discharge SF-HPCI Top-Peaked Axis! 102% Power [
3......... 6-3 Table 6.2 Event Times for Limiting TLO Recirculation Line Break 0.13 ft Split Pump Discharge SF-HPCI Top-Peaked Axial 102% Power [
3......... 6-4 Table 6.3 TLO Recirculation Line Break Spectrum Results for 102% Power [
] SF-HPCI................................................................ 6-5 Table 6.4 Summary of TLO Recirculation Line Break Results Highest PCT Cases.......... 6-6 Table 7.1 Results for Limiting SLO Recirculation Line Break 0.1 ft2 Split Pump Discharge SF-HPCI Top-Peaked Axial 102% Power [
].........
7-3 Table 7.2 Event Times for Limiting SLO Recirculation Line Break 0.1 ft Split Pump Discharge SF-HPCI Top-Peaked Axial 102% Power [
]3......... 7-4 Table 7.3 Single-and Two-Loop Operation PCT Summary.................................... 7-5 List of Figures Figure 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model................... 4-15 Figure 4.2
[
]................ 4-16 Figure 4.3 RELAX System Model............................................................... 4-17 Figure 4.4 RELAX Hot Channel Model Top-Peaked Axial.................................... 4-18 Figure 4.5 RELAX Hot Channel Model Mid-Peaked Axial.................................... 4-19 Figure 4.6 ECCS Schematic..................................................................... 4-20 Figure 4.7 Rod Average Power Distributions for 102%P [
] Mid-and Top-Peaked.......................................................................... 4-21 Figure 6.1 Limiting TLO Recirculation Line Break Upper Plenum Pressure................... 6-7 Figure 6.2 Limiting TLO Recirculation Line Break Total Break Flow Rate..................... 6-7 Figure 6.3 Limiting TLO Recirculation Line Break Core Inlet Flow Rate....................... 6-8 Figure 6.4 Limiting TLO Recirculation Line Break Core Outlet Flow Rate..................... 6-8 Figure 6.5 Limiting TLO Recirculation Line Break Intact Loop Jet Pump Exit Flow Rate.................................................................................... 6-9 Figure 6.6 Limiting TLO Recirculation Line Break Broken Loop Jet Pump Exit Flow Rate..................................................................................... 6-9 Figure 6.7 Limiting TLO Recirculation Line Break Relief Valve Flow Rate................... 6-10 Figure 6.8 Limiting TLO Recirculation Line Break ADS Flow Rate............................ 6-10 Figure 6.9 Limiting TLO Recirculation Line Break HPCI Flow Rate........................... 6-11 Figure 6.10 Limiting TLO Recirculation Line Break LPCS Flow Rate.......................... 6-11 AREVA Inc.
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Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page iv Figure 6.11 Figure 6.12 Figure 6.13 Figure 6.14 Figure 6.15 Figure 6.16 Figure 6.17 Figure 6.18 Figure 6.19 Figure 6.20 Figure 6.21 Figure 6.22 Figure 6.23 Figure 6.24 Limiting TLO Recirculation Line Break Intact Loop LPCI Flow Rate.............. 6-12 Limiting TLO Recirculation Line Break Broken Loop LPCI Flow Rate............ 6-12 Limiting TLO Recirculation Line Break Upper Downcomer Mixture Level........ 6-13 Limiting TLO Recirculation Line Break Lower Downcomer Mixture Level........ 6-13 Limiting TLO Recirculation Line Break Intact Loop Discharge Line Liquid Mass.................................................................................. 6-14 Limiting TLO Recirculation Line Break Upper Plenum Liquid Mass............... 6-14 Limiting TLO Recirculation Line Break Lower Plenum Liquid Mass............... 6-15 Limiting TLO Recirculation Line Break Hot Channel Inlet Flow Rate............. 6-15 Limiting TLO Recirculation Line Break Hot Channel Outlet Flow Rate........... 6-16 Limiting TLO Recirculation Line Break Hot Channel Coolant Temperature at the Hot Node at EOB.............................................. 6-16 Limiting TLO Recirculation Line Break Hot Channel Quality at the Hot Node at EOB......................................................................... 6-17 Limiting TLO Recirculation Line Break Hot Channel Heat Transfer Coeff. at the Hot Node at EOB...................................................... 6-17 Limiting TLO Recirculation Line Break Hot Channel Reflood Junction Liquid Mass Flow Rate............................................................... 6-18 Limiting TLO Recirculation Line Break Cladding Temperatures.................. 6-18 Figure 7.1 Limiting SLO Recirculation Line Break Upper Plenum Pressure................... 7-6 Figure 7.2 Limiting SLO Recirculation Line Break Total Break Flow Rate..................... 7-6 Figure 7.3 Limiting SLO Recirculation Line Break Core Inlet Flow Rate....................... 7-7 Figure 7.4 Limiting SLO Recirculation Line Break Core Outlet Flow Rate..................... 7-7 Figure 7.5 Limiting SLO Recirculation Line Break Intact Loop Jet Pump Exit Flow Rate.................................................................................... 7-8 Figure 7.6 Limiting SLO Recirculation Line BreakBroken Loop Jet Pump Exit Flow Rate.................................................................................... 7-8 Figure 7.7 Limiting SLO Recirculation Line Break Relief Valve Flow Rate.................... 7-9 Figure 7.8 Limiting SLO Recirculation Line Break ADS Flow Rate............................. 7-9 Figure 7.9 Limiting SLO Recirculation Line Break HPCI Flow Rate........................... 7-10 Figure 7.10 Limiting SLO Recirculation Line Break LPCS Flow Rate.......................... 7-10 Figure 7.11 Limiting SLO Recirculation Line Break Intact Loop LPCI Flow Rate.............. 7-11 Figure 7.12 Limiting SLO Recirculation Line Break Broken Loop LPCI Flow Rate............ 7-11 Figure 7.13 Limiting SLO Recirculation Line Break Upper Downcomer Mixture Level........ 7-12 Figure 7.14 Limiting SLO Recirculation Line Break Lower Downcomer Mixture Level........ 7-12 Figure 7.15 Limiting SLO Recirculation Line Break Intact Loop Discharge Line Liquid Mass........................................................................... 7-13 Figure 7.16 Limiting SLO Recirculation Line Break Upper Plenum Liquid Mass............... 7-13 Figure 7.17 Limiting SLO Recirculation Line Break Lower Plenum Liquid Mass............... 7-14 Figure 7.18 Limiting SLO Recirculation Line Break Hot Channel Inlet Flow Rate............. 7-14 Figure 7.19 Limiting SLO Recirculation Line Break Hot Channel Outlet Flow Rate........... 7-15 Figure 7.20 Limiting SLO Recirculation Line Break Hot Channel Coolant Temperature at the Hot Node at EOB.............................................. 7-15 Figure 7.21 Limiting SLO Recirculation Line Break Hot Channel Quality at the Hot Node at EOB......................................................................... 7-16 Figure 7.22 Limiting SLO Recirculation Line Break Hot Channel Heat Transfer Coeff. at the Hot Node at EOB...................................................... 7-16 Figure 7.23 Limiting SLO Recirculation Line Break Hot Channel Reflood Junction Liquid Mass Flow Rate............................................................... 7-17 Figure 7.24 Limiting SLO Recirculation Line Break Cladding Temperatures.................. 7-17 ARE VA Inc.
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Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page v Nomenclature ADS ANS APE BL BOL BWR CFR DEG ECCS EDG EOB EPU HPCI ID LOCA LPCI LPCS MAPLHGR MCPR MSIV MWR NRC OD PCT PD PS RDIV RPF SF-ADS SF-DGEN SF-HPCI SF-LPCI SF-LSL SLO TCV TLO TSV UFSAR VDC automatic depressurization system American Nuclear Society axial peaking factor broken loop beginning of life boiling-water reactor Code of Federal Regulations double-ended guillotine emergency core cooling system emergency diesel generator end of blowdown extended power uprate high-pressure coolant injection inside diameter loss-of-coolant accident low-pressure coolant injection low-pressure core spray maximum average planar linear heat generation rate minimum critical power ratio main steam isolation valve metal-water reaction Nuclear Regulatory Commission, U.S.
outside diameter peak cladding temperature pump discharge pump suction recirculation discharge isolation valve radial peaking factor single failure of ADS valve single failure of diesel generator single failure of the HPCI system single failure of an LPCI injection valve single failure of the loop selection logic single-loop operation turbine control valve two-loop operation turbine step valve updated safety analysis report volts direct current AREVA Inc.
Controlled Document Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM 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 Quad Cities 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 (POT) during a postulated LOCA. The LOCA parameters addressed in this report include the following:
Break location Break type (double-ended guillotine (DEG) or split)
Break size Limiting emergency core cooling system (EGCS) single failure Axial power shape (top-or mid-peaked)
The analyses documented in this report were performed with LOCA Evaluation Models developed by ARE VA Inc. (ARE VA) 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 10XM* fuel at beginning-of-life (BOL) conditions. Calculations assumed an initial core power of 102% of 2957 MWt, providing a licensing basis power of 3016.14 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 11.7 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 Inc.
AREVA Inc.
Coritro!!ed Documents Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM Fuel Page 1-2 versus exposure for ATRIUM 10XM fuel contained in Quad Cities. Even though the limiting break will not change with exposure or nuclear fuel design, the value of POT calculated for any given set of break characteristics is dependent on exposure and local power peaking.
Therefore, heatup analyses are performed to determine the POT 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 POT versus exposure from the heatup analyses are documented in the MAPLHGR report.
[
] Limiting reactor power and core flow conditions were selected with consideration for the EPU range of operating conditions. This report also addresses long-term coolability.
The impact on LOCA of operation with equipment out-of-service and their combinations as per the COLR have been considered. The LOCA analyses include the effects of I relief valve out-of-service. Operation with an ADS valve out-of-service is not currently allowed by the Quad Cities Technical Specification. Since the consequences of a LOCA would be more severe during operation with one of the recirculation lines out-of-service, this report presents results for single-loop operation (SLO). Operation with other allowed equipment out-of-service conditions are supported (turbine bypass valves, feedwater heaters, one MSIV, power load unbalance, pressure controller, TCV slow closure, and one stuck closed TCV or TSV).
AREVA Inc.
Con~trolede Document~
Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 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 discharge pipe Type!/ size Split break / 0.13 ft2 Single failure HPCI Axial power shape Top-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 10XM MAPLHGR multiplier of 0.80 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 I0XM core for an extended period after a LOCA.
AREVA Inc.
Contro~ed Dnc~ U ~ (&
Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM 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 IOXM 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 8. The NRC acknowledged the modified approach in Reference 9.
Differences between the break spectrum results for the AREVA fuel and the co-resident legacy fuel will primarily be the result of differences between the fuel vendor LOCA methodologies.
AREVA Inc.
Ccntrofled Document Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM 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 rated LPCS flow is attained.
AREVA Inc.
Controlled Document Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision I for ATRIUM 10XM 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 CER 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 Inc.
Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM Fuel Page 3-3 These criteria are commonly referred to as the peak cladding temperature (POT) 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 POT must not exceed 2200°F. LOCA POT results are provided in Sections 6.0 - 7.0 to determine the limiting LOCA event.
LOCA analysis results demonstrating that the POT, 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 10OXM 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 Inc.
Controlled Document Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM 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 POT 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 Inc.
Controlled Document Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision I for ATRIUM 10XM 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 10OXM 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 Refihll/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 ref lood).
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 Inc.
Controlled Docunment Quad Cities Units I and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM I0XM 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 1OX1 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 Inc.
Controlled Document Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM 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 Exelon.
Limiting reactor power and core flow conditions were selected with consideration for the EPU range of operating conditions. The control blades are modeled as inserting with the scram timing required by the Quad Cities Technical Specification. 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 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.
AREVA Inc.
Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM Fuel Page 4-5 The rod average axial power profiles resulting from the application of this process for 1 02%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 10XM fuel. Some of the key ATRIUM 10XM fuel parameters used in the break spectrum analysis are summarized in Table 4.3.
4.6 ECCS Parameters The ECOS configuration is shown in Figure 4.6. Table 4.4 - Table 4.7 provide the important ECCS characteristics assumed in the analysis. 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 flow is credited until the injection valve is fully open. LPCI and LPCS flow is governed by the valve characteristics provided by Exelon. Also, no credit for ECCS flow is assumed until ECCS pumps reach rated speed.
AREVA Inc.
CotrollVed Doc et Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM l0XM Fuel Page 4-6 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 five ADS valves are assumed operable during the LOCA except when a single failure is assumed to prevent one ADS valve from opening.
In the ARE VA 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. [
]
Recirculation discharge isolation valve (RDIV) and loop selection logic parameters are presented in Table 4.8. For recircutation line breaks sizes > 0.15 if2, the loop selection logic directs all available LPCI flow to the intact loop and closes the RDIV in the intact loop. For break sizes < 0.15 if2, all available LPCl flow is assumed to be injected into the broken loop and the RDIV in the broken loop is closed.
AREVA Inc.
C, u3ntlr.,ed Doun teni Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 4-7 Table 4.1 Initial Conditions*
I Reactor power (% of rated) 102 102 Reactor power (MWt) 3016.14 3016.14 Steam flow rate (MIb/hr) 11.98 11.98 Steam dome pressure (psia) 1020 1020 Core inlet enthalpy (Btu/Ib) 521.6 518.4 ATRIUM 10OXM hot assembly MAPLHGR (kW/ft) 11.7 11.7
]
]
- 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. Initial conditions are based on nominal values, except for initial core power and dome pressure which use conservative values.
I I
AREVA Inc.
C.n r 1k~d D~~cvv~nt Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 4-8 Table 4.2 Reactor System Parameters Parameter Value Vessel ID (in) 251 Number of fuel assemblies 724 Recirculation suction pipe area (ft2) 3.581 Recirculation discharge pipe area (ft2) 3.477 AREVA Inc.
ConltrOl° Docum....nt Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 4-9 Table 4.3 ATRIUM I0XM Fuel Assembly Parameters Parameter Fuel rod array Number of fuel rods per assembly Non-fuel rod type Value Fuel rod OD (in)
Active fuel length (in)
(including blankets)
Water channel outside width (in)
Fuel channel thickness (in) 10xl10 79 (full-length rods) 12 (part-length rods)
Water channel replaces 9 fuel rods 0.4047 145.24 (full-length rods) 75.0 (part-length rods) 1.378
- 0. 075 (minimum wall)
- 0. 100 (corner) 5.278 Fuel channel internal width (in)
AREVA Inc.
Cont rotIed Docutmenst Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Paqie 4-10 Table 4.4 High-Pressure Coolant Injection Parameters Parameter Value Coolant temperature (maximum) (0F) 140 Initiating Signals and Setpoints Water level (in)*
444 High drywell pressure (psig)t 2.5 Time Delays Time for HPCI pump to reach rated speed and injection valve wide open (sec) 55 Coolant Flow Rate Vs. Pressure Vessel to Torus AP Flow Rate (psid)
(gpm) 0 0
150 5000 1120 5000 t
Relative to vessel zero.
The time when the drywell reaches 2.5 psig is not explicitly calculated in the EXEM BWR-2000 LOCA methodology. [
]
AREVA Inc.
Controlled Documnent Quad Cities Units 1 and 2 LOCA Break Spectrum Anaiysis for ATRIUM I0XM Fuel ANP-3328NP Revision 1 Page 4-11 Table 4.5 Low-Pressure Coolant Injection Parameters Parameter Value Reactor pressure permissive for opening injection valves - analytical (psig) 300 Coolant temperature (maXimum) (0F) 160 Initiating Signals and Setpoints Water level (in)*
444 High drywell pressure (psig)t 2.5 Time Delays Initiating signal processing delayI Time from EDG started and initiated signal to LPCI pump at rated speed (sec) 14 Time for power at the injection valve (sec) 26 LPCI injection valve stroke time (sec) 28 Coolant Flow Rates Vs. Pressure Vessel to Flow Rate for Flow Rate for Torus AP 2 Pumps 4 Pumps (psid)
(gpm)
(gpm) 0 9,300 15,700 20 9,000 15,200 150 6,200 10,200 257 0
0 t
Relative to vessel zero.
The time when the drywell reaches 2.5 psig is not explicitly calculated in the EXEM BWR-2000 LOCA methodology. ["
]
[
]
AREVA Inc.
~on~ro~ec~ L)Ocur'~ ~
Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision I Pa~qe 4-12 Table 4.6 Low-Pressure Core Spray Parameters Parameter Value Reactor pressure permissive for opening injection valves - analytical (psig) 300 Coolant temperature (maximum) (0F) 160 Initiating Signals and Setpoints Water level (in)*
444 High drywell pressure (psig)t 2.5 Time Delays Initiating signal processing delay 1
Time from EDG started and initiated signal to LPCS pump at rated speed (sec) 17 Time for power at the injection valve (sec) 17 LPCS injection valve stroke time (sec) 53 Coolant Flow Rates Vs. Pressure Vessel to Flow Rate for Torus AP 1 Pump (psid)
(gpm) 0 5,650 90 4,500 200 3,000 325 0
Relative to vessel zero.
1-The time when the drywell reaches 2.5 psig is not explicitly calculated in the EXEM BWR-2000 LOCA methodology. [
]
r
]
AREVA Inc.
Contro~ed Lx~ur~c~A Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 4-13 Table 4.7 Automatic Depressurization System Parameters Parameter Value Number of valves installed 5
Number of valves available*
5 Minimum flow capacity per valve 558,000 of available valves (4 relief) at (Ibm/hr at psig) 1120 Minimum flow capacity of 598,000 available valves (1 Target Rock) at (Ibm/hr at psig) 1080 Initiating Signals and Setpoints Water level (in)t 444 High drywell pressure (psig)*
2.5 Time Delays ADS timer (delay time from initiating signal to time valves are open) (sec) 120 All 5 valves are assumed operable in the analyses except when analyzing the potential single failure of I ADS valve during the LOCA.
t Relative to vessel zero.
SThe time when the drywell reaches 2.5 psig is not explicitly calculated in the EXEM BWR-2000 LOCA methodology. [
]
AREVA Inc.
Oontro~1ed ~r~r~rnen'~
Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 4-14 Table 4.8 Recirculation Discharge Isolation Valve and LPCI Loop Selection Logic Parameters Parameter Value Minimum break area for loop selection logic to select intact loop*
(ft2) 0.15 LPCI loop selection logic pressure permissive (minimum) (psig) 860 Time to power at the RDIV (sec) 26 RDIV stroke time (sec) 48 For break sizes > 0.15 ft2, loop selection logic opens the LPCI injection valves in the intact loop and closes the RDIV in the intact loop. For break sizes < 0.15 if2, the available LPCI flow is assumed to be injected into the broken loop and the RDIV in the broken loop is closed.
AREVA Inc.
Control led lDcoumen~
Quad Cities Units I and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Paije 4-15 Fuel Stored Energy Boundary Conditions (power, upper & lower plenum conditions)
Hot Assembly Analysis*
(RELAX)
Fuel Stored Energy
- Gap, Gap Coefficient, Fission Gas Boundary Conditions (Pressure, Temperature, Power, Quality, Heat Transfer Coefficient)
Time of Hot Node Reflood Heatup Analysis
- End of Blowdown, Time of Bypass Reflood (HUXY)
- The hot assembly calculation may be combined with the system calculation or executed separately Peak Cladding Temperature, Metal Water Reaction Figure 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model AREVA Inc.
Controlled Document Quad Cities Units I and 2 LOCA Break Spectrum Analysis for ATRIUM 1OXM Fuel ANP-3328NP Revision I Page 4-16 Figure 4.2 [
I AREVA Inc.
Ccrh*L* lld Docu ent Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 4-17 Figure 4.3 RELAX System Model AREVA Inc.
Controlled Document Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM IOXM Fuel ANP-3328NP Revision 1 Page 4-18 Figure 4.4 RELAX Hot Channel Model Top-Peaked Axial AREVA Inc.
C~ontrFo~ed DJOGLIcument Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 4-19 Figure 4.5 RELAX Hot Channel Model Mid-Peaked Axial AREVA Inc.
f~
Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 4-20 feedwater Figure 4.6 ECGS Schematic AREVA Inc.
Contr-olled Docu menit Quad Cities Units 1 and 2 LOCA Break, Spectrum Analysis for ATRIUM I0XM Fuel ANP-3328NP Revision 1 Page 4-21 Figure 4.7 Rod Average Power Distributions for 102%P [
]
Mid-and Top-Peaked AREVA Inc.
Controlle Document Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision I for ATRIUM 10XM 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 ECOS 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 UFSAR (Reference 6) are shown below:
LPCI injection valve (SF-LPCI))
Diesel generator or 125-VDC (SF-DGEN)
High-pressure coolant injection system (SF-HPCI)
Loop Select Logic (SF-LSL)
ADS valve (SF-ADS)
The single failures and the available ECOS 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.
The loop selection logic single failure results in all available LPCl flow directed to the default recirculation loop, and the closure of the RDIV in the same loop. The limiting scenario would be if th~e break is in the default loop, i.e., the loop in which the LPCI flow is injected. Depending on the break size and location relative to the RDIV, some or all of the LPCI flow will flow out the break, minimizing the amount of LPCI flow reaching the core. A comparison of the available EGOS equipment available for the other assumed failures shows that the loop selection logic failure has more or the same amount of EGOS capacity (even if one assumes no benefit from the LPOI system) as the SF-LPCl condition. Therefore, POT results obtained for SF-LPCI are AREVA Inc.
ControlIed Document Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision I for ATRIUM 10XM Fuel Page 5-2 the same as or bound those obtained for loop selection logic failure. As a result, results reported here are those considering SF-LPCl, SF-DGEN, SF-HPCl 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 LPCl 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. HPCI and ADS operation are important emergency systems for small breaks where they assist in depressurizing the reactor system faster, and thereby reduce 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 large break events more severe. For recirculation line breaks with areas >- 0.15 ft, the LPCI loop selection logic directs all available LPCI flow to the intact loop and also closes the RDIV in the intact loop.
For recirculation line break areas < 0.15 ft2, the loop selection logic is not able to determine which loop is broken so the limiting scenario is injecting all available LPCI flow into the broken loop where at least some of the flow will exit out the break. In this limiting scenario, the RDIV in the broken loop will close. 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 DE(G 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 DEG3 break is determined by varying the discharge coefficient. The labeling convention for guillotine breaks is to list the AREVA Inc.
"p L
Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM 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 CER 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 ft2 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 6.0 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-Recirculation Line 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 EGOS 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 Inc.
r~ ~
Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 5-4 address non-fuel related regulatory requirements, these breaks are not limiting relative to fuel acceptance criteria such as POT.
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 AREVA Inc.
Controlled Document Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 5-5 5.3.3 HPCI Line Breaks The HPCl 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.
[
]
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 Breaks 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. [
]
AREVA Inc.
C ontro~ed Document Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 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 Coolingq 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
[
AREVA Inc.
Oontro~ed Document Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM I0XM Fuel ANP-3328NP Revision 1 Page 5-7 Table 5.1 Available ECCS for Recirculation Line Break LOCAs Assumed Systems Failure Remaining*, t LPCI injection valve 2 [PCS + HPC1 + 5 ADS (S F-LPClI)
Diesel generator or 1 LPCS + 2 LPCl + HPCI + 5 ADS 1 25-VDC (SF-DG EN)
HPCl system 2 LPCS + 4 LPCl + 5 ADS (SF-H PCI)
Loop select logic 2 LPCS + 4 LPCl + HPCl + 5 ADS (SF-LSL)
ADS valve 2 LPCS + 4 LPCl + HPCI + 4 ADS (SF-ADS)
Systems remaining, as identified in this table for recirculation line breaks, are applicable to all non-ECOS line breaks. For ECCS line breaks, the systems remaining are those listed for recirculation breaks, less the ECCS in which the break is assumed t
With loop selection logic operational, all available LPCI flow is directed to the intact loop for breaks >
0.15 ft2. All available LPCI flow is directed to the broken loop for breaks < 0.15 ft2. The limiting condition for a loop selection logic failure would result in all available LPCI flow directed to the broken loop for all break sizes.
AREVA Inc.
Controlled Document Quad Cities Units I and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision I for ATRIUM 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-LPOI, SF-DGEN, SF-HPCI, SF-LSL and SF-ADS.
6.1 Limiting Break Analysis Results The analyses demonstrate that the limiting (highest PCT) recirculation line break is the 0.13 ft2 split break in the pump discharge piping with an SF-HPCI single failure and a top-peaked axial power shape when operating at 102% rated core power [
]. The POT is 2127°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 6.23 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 6.24.
Table 6.3 presents SF-HPCI 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 POT recirculation line break calculations for each of the single failures, state points, and axial power shapes. The results of the break spectrum analyses are discussed in the following sections.
6.2 Break Location Analysis Results Table 6.4 shows that the maximum POT calculated for a recirculation line break occurs in the pump discharge piping.
AREVA Inc.
ControlIed Documient Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision I for ATRIUM I0XM Fuel Page 6-2 6.3 Break Geometry and Size Analysis Results Recirculation line break PCT results versus break geometry (DEG or spiit) and size were performed for DEG breaks with discharge coefficients of 1.0, 0.8, 0.6, and 0.4.
Split breaks ranging in size from the full pipe diameter to 0.1 ft2 in increments of 0.1 ft2 and a final size of 0.05 ft2. Based on the results for small breaks, for some conditions the break size increment was 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 0.13 ft2.
6.4 Limiting Single-Failure Analysis Results The results in Table 6.4 show that the limiting single-failure is SF-H PCI.
6.5 Axial Power Shape Analysis Results The results in Table 6.4 show that the top-peaked axial power shape is generally limiting compared to the mid-peaked shape analyses for the limiting break size.
6.6 State Point Analysis Table 6.4 shows that 102% rated core power [
] is generally the limiting state point for the recirculation line breaks.
AREVA Inc.
Cont roled Oocument Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 6-3 Table 6.1 Results for Limiting TLO Recirculation Line Break 0.13 ft2 Split Pump Discharge SF-HPCI Top-Peaked Axial 102% Power [
PCT 2127° F Maximum local cladding oxidation 3.02%
Maximum planar average cladding oxidation 2.40%
AREVA Inc.
S.ro lei, Dou n
Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 6-4 Table 6.2 Event Times for Limiting TLO Recirculation Line Break 0.13 ft2 Split Pump Discharge SF-HPCI Top-Peaked Axial 102% Power [
Event Initiate Break Initiate Scram Diesel Generators Started Low-Low Liquid Level, L2 (444 in)
Power at LPCS Injection Valves ADS with ECCS Pump(s) Running LPCS Pump at Rated Speed LPCS High-Pressure Cutoff LPCS Valve Pressure Permissive LPCS Valve Starts to Open LPCS Flow Starts Top Down Cooling Restriction Begins in the Hot Channel LPCS Valve Fully Open Rated LPCS Flow LPCI Pump at Rated Speed LPCI Valve Pressure Permissive LPCI Valve Starts to Open LPCI Flow Starts LPCI Valve Fully Open Jet Pump Uncovers Recirculation Suction Uncovers ADS Valves Open RDIV Pressure Permissive RDIV Starts to Close RDIV Closed PCT Time of Bypass Reflood Time of Rated Spray Time of Hot Node Reflood Time (sec) 0.0 0.6 17.0 48.5 17.0 56.5 66.5 338.4 348.5 348.5 351.2 351.2 401.5 549.1 63.5 348.5 348.5 366.7 376.5 139.8 414.1 169.5 207.8 207.8 255.8 452.8 513.1 549.1 549.1 AREVA Inc.
Co ntro)!ed Docun ent Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Paqie 6-5 Table 6.3 TLO Recirculation Line Break Spectrum Results for 102% Power [
] SF-HPCI Break Size PT(F and Type
- Pump Suction Pump Discharge Mid-Peaked Top-Peaked Mid-Peaked Top-Peaked 0.13 ft2 split 22
]
AREVA Inc.
Controi1ed Doc~ur ent Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 6-6 Table 6.4 Summary of TLO Recirculation Line Break Results Highest PCT Cases Axial Power Shape Single Mid-Peaked Top-Peaked Failure Break Size POT Break Size POT and Location (0F) and Location
(°F) 102% Power [
]
SF-LPCl
[
]
[
]
SF-HPCI
[
] 0.13 ft2 pump discharge 2127 SF-ADS
[
]
102% Power [
]
SF-LPCI
[
]
SF-DGEN
[
]
SF-HPCI
[
]
SF-ADS
[]
AREVA Inc.
Contro~ed Docurr~nt Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 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 Inc.
C~ontrc;uied Do~cu* me n Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 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 Inc.
C*.
ro e Doc u.men t Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision I 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 Inc.
Contro~ted D Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision I Page 6-10 I-
]I 4J l
I I
(-
i i
i I
i i
i 0
60 120 100 2410 300 360 420 400 540 liME (SEQ Figure 6.7 Limiting TLO Recirculation Line Break Relief Valve Flow Rate L-
]1 iF 021/)
iF I
I I
I I
I I
=
I i
I 0
00 120 100 240 300 200 420 430 040 TIME (SEC)
Figure 6.8 Limiting TLO Recirculation Line Break ADS Flow Rate AREVA Inc.
C.ol [.tiede*
r D o*
ei Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 6-11 E-
]
ho 0.
"1 I
r I
I I
I I
I I
I I
I 0
00 120 180 240
,300 000 420 480 540 TIME (SEC)
Figure 6.9 Limiting TLO Recirculation Line Break HPCl Flow Rate I-J1 L.
U.
I i
i r
=
I Q
f
=
3
=
I i
I f
i I
0 00 120 100 240 300 300 420 400 040 TIME (SEC)
Figure 6.10 Limiting TLO Recirculation Line Break LPCS Flow Rate AREVA Inc.
Gontro~e.~ L.
Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 6-12 I
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Contso7*Ied Document -
Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision I Pane 6-13 El
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Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 6-15 I-
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Controlled Document Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision I Pane 6-17 Figure 6.21 Limiting TLO Recirculation Line Break Hot Channel Quality at the Hot Node at EOB Figure 6.22 Limiting TLO Recirculation Line Break Hot Channel Heat Transfer Coeff. at the Hot Node at EOB AREVA Inc.
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Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 6-18 Figure 6.23 Limiting TLO Recirculation Line Break Hot Channel Reflood Junction Liquid Mass Flow Rate I
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Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM 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. [
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AREVA Inc.
Controlled Document Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision I Paae 7-2
[
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7.2 SL C Analysis Results The SLO analyses are performed with a 0.80 multiplier applied to the two-loop MAPLHGR limit resulting in an SLO MAPLHGR limit of 9.36 kW/ft. The analyses are performed at BOL fuel conditions. The limiting SLO LOCA is the 0.1 ft2 split pump discharge line break with SF-HPCI and a top-peaked axial power shape. The PCT for this case is 20470°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 7.23 show important RELAX system and hot channel results from the SLO limiting LOCA analysis. Figure 7.24 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 0.80 multiplier is applied to the two-loop MAPLHGR limit.
AREVA Inc.
Contro,!ed Docu ient Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel AN P-3328N P Revision 1 Page 7-3 Table 7.1 Results for Limiting SLO Recirculation Line Break 0.1 ft2 Split Pump Discharge SF-HPCl Top-Peaked Axial 102% Power [
I PCT 2047°F Maximum local cladding oxidation 2.21%
Maximum planar average cladding oxidation 1.81 %
AREVA Inc.
Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 7-4 Table 7.2 Event Times for Limiting SLO Recirculation Line Break 0.1 ft2 Split Pump Discharge SF-H PCI Top-Peaked Axial 102% Power [
]
Event Initiate Break Initiate Scram Diesel Generators Started Low-Low Liquid Level, L2 (444 in)
Power at LPCS Injection Valves ADS with ECOS Pump(s) Running LPCS Pump at Rated Speed LPCS High-Pressure Cutoff LPCS Valve Pressure Permissive LPCS Valve Starts to Open LPCS Flow Starts Top Down Cooling Restriction Begins in the Hot Channel LPCS Valve Fully Open Rated LPCS Flow Power at LPCI Injection Valves LPCI Pump at Rated Speed LPCI Valve Pressure Permissive LPCI Valve Starts to Open LPCI Flow Starts LPCI Valve Fully Open Jet Pump Uncovers Recirculation Suction Uncovers ADS Valves Open RDIV Pressure Permissive RDIV Starts to Close RDIV Closed PCT Time of Rated Spray Time of Hot Node Reflood Time of Bypass Reflood Time (sec) 0.0 0.6 17.0 58.8 17.0 66.8 76.8 375.3 387.5 387.5 390.2 390.2 440.5 704.3 26.0 73.8 387.5 387.5 410.9 415.5 175.6 No Uncovery 179.8 219.8 219.8 267.8 493.2 704.3 704.3 538.1 AREVA Inc.
Contr-oled Docurnen Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 7-5 Table 7.3 Single-and Two-Loop Operation PCT Summary Operation Single-loop Two-loop Limiting Case 0.1 ft2 split pump discharge top-peaked SF-HPCI 0.13 ft2 split pump discharge top-peaked SF-HPCI POT (0F) 2047 2127 AREVA Inc.
~Controlled Document Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 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 Inc.
Controlled Document Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 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 Inc.
Contoed Document Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 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 Inc.
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Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 7-12
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Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Paqie 7-16 Figure 7.21 Limiting SLO Recirculation Line Break Hot Channel Quality at the Hot Node at EOB Figure 7.22 Limiting SLO Recirculation Line Break Hot Channel Heat Transfer Coeff. at the Hot Node at EOB AREVA Inc.
Quad Cities Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel ANP-3328NP Revision 1 Page 7-17 Figure 7.23 Limiting SLO Recirculation Line Break Hot Channel Reflood Junction Liquid Mass Flow Rate I-Time (sec)
Fri Oci.3 16:44:06 2014 Figure 7.24 Limiting SLO Recirculation Line Break Cladding Temperatures AREVA Inc.
Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM 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 7).
The first step in the long-term cooling analysis is to demonstrate that sufficient EGGS flow is always available to fill the inside of the core shroud to two-thirds core height. This was done by reviewing a subset of the breaks including the break that has the minimum mass inventory in the lower plenum region and the reactor core anytime during the event and the large break case with the minimum EGGS flow. The review considered the minimum number of EGGS available for any recirculation line break combined with the effects of axial power shape. In each case, the analysis was extended to 10 minutes and all potential leakage paths were conservatively accounted for. For the minimum EGGS flow case, the limiting break is the 1.0 DEG PS SF-LPGI. For this case, two core sprays are available and provide 9854 gpm at the core spray sparger. The results of these analyses show that in all cases the time to fill the core to the jet pump nozzle elevation (two-thirds core height) was less than 10 minutes.
The second step in the analysis is to demonstrate that after 10 minutes, sufficient EGGS flow is available to remove decay heat and maintain the water level at two-thirds core height. The results of the first step showed that at 10 minutes the vessel pressure dropped far enough to ensure rated core spray (i.e. 4500 gpm) is attained at the core spray sparger (even with leakages considered) and distribution of the core spray to all assemblies. EGOS flow of 3300 gpm is needed to remove decay heat and maintain the water level. This flow can be provided from one core spray; meeting the long-term coolability needs demonstrated by the NSSS vendor discussed above.
In summary, maintaining water level at two-thirds core height with one core spray operating or the core flooded to the top of active fuel is required for long-term cooling beyond 10 minutes for AREVA Inc.
Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 10XM Fuel Page 8-2 ATRIUM 10XM fuel. When the core is not flooded to the top of active fuel, the conclusions demonstrate that as long as there is adequate water from LPCI or core spray to maintain the two-thirds core height water level, a single core spray of 4500 gpm to the top of the core is needed to meet the fuel safety limits for long-term core cooling.
AREVA Inc.
Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 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 0.13 ft2 split break in the pump discharge piping with single failure SF-HPCI and a top-peaked axial shape when operating at 102% rated core power [
1.
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 0.80 for ATRIUM 10XM 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 Quad Cities analyses to establish the MAPLHGR limit versus exposure for ATRIUM 10OXM fuel.
AREVA Inc.
Quad Cities Units 1 and 2 ANP-3328NP LOCA Break Spectrum Analysis Revision 1 for ATRIUM 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/ Mu/tirod 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 -
Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
- 5.
EMF-2292(P)(A) Revision 0, A TR/UM TM -IO: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
- 6.
Updated Final Safety Analysis Report, Quad Cities Nuclear Station.
- 7.
NEDO-20566A, General Electric Company Analytical Model for Loss of Coolant Analysis in Accordance with IOCFR5O Appendix K, September 1986.
- 8.
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: 11:096, September 22, 2011.
- 9.
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.
AREVA Inc.