ML13200A197
| ML13200A197 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/31/2013 |
| From: | AREVA NP |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML13200A185 | List: |
| References | |
| L-MT-13-055 ANP-3212(NP), Rev 0 | |
| Download: ML13200A197 (55) | |
Text
Enclosure 21 AREVA Report ANP-3212(NP)
Monticello LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 1OXM Fuel (EPU/MELLLA)
Revision 0 54 pages follow
ANP-3212(NP)
Revision 0 Monticello EPU LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM TM1OXM Fuel May 2013 A
AREVA NP Inc. AR EVA
uontroiled uocument AREVA NP Inc.
ANP-3212(NP)
Revision 0 Monticello EPU LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM TM1OXM Fuel
uontrolued uocument AREVA NP Inc.
ANP-3212(NP)
Revision 0 Copyright © 2013 AREVA NP Inc.
All Rights Reserved
Uontroloed Uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Paaei Nature of Changes Item Page Description and Justification
- 1. All This is the initial issue.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page ii Contents 1 .0 In trod u c tio n .................................................................................................................. 1-1 2 .0 S u m m a ry ...................................................................................................................... 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 Param eters .............................................................................................. 4-4 4.6 ECCS Param eters ............................................................................................ 4-4 5.0 MAPLHGR Analysis Results ......................................................................................... 5-1 5.1 Therm al Conductivity Degradation .................................................................... 5-1 6.0 Conclusions .................................................................................................................. 6-1 7.0 References ................................................................................................................... 7-1 Appendix A Supplem ental Information .......................................................................... A-1 Tables 2.1 LOCA Results for PCT Lim iting Conditions ................................................................... 2-3 4.1 Initial Conditions ........................................................................................................... 4-6 4.2 Reactor System Parameters ......................................................................................... 4-7 4.3 ATRIUM 1OXM Fuel Assem bly Param eters .................................................................. 4-8 4.4 High-Pressure Coolant Injection Param eters ................................................................ 4-9 4.5 Low-Pressure Coolant Injection Parameters ............................................................... 4-10 4.6 Low-Pressure Core Spray Parameters ....................................................................... 4-11 4.7 Autom atic Depressurization System Param eters ........................................................ 4-12 4.8 Recirculation Discharge Isolation Valve Param eters ................................................... 4-13 4.9 ECCS Single Failure ................................................................................................... 4-14 5.1 Event Times for Limiting Break 3.3 ft2 Split Pump Suction SF-LPCI Mid-Peaked Axial 102% Power ........................................................................................................ 5-3 5.2 ATRIUM 1OXM MAPLHGR Analysis Results ................................................................ 5-4 A.1 Com puter Codes Used for MAPLHGR Lim it Analysis ............................................. A-1 AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page iii Figures 2.1 MAPLHGR Limit for Two-Loop Operation .................................................................... 2-4 4.1 Flow Diagram for EXEM BWR-2000 ECCS Evaluation Model .................................... 4-15 4 .2 [] .............................. 4-16 4.3 RELAX System Blowdown Model ............................................................................... 4-17 4.4 RELAX Hot Channel Blowdown Model Mid-Peaked Axial ........................................... 4-18 4 .5 EC C S S chem atic ........................................................................................................ 4-19 4.6 Axial Power Distribution for Limiting LOCA Case in RELAX Calculation ..................... 4-20 5.1 Limiting Break Upper Plenum Pressure ........................................................................ 5-5 5.2 Limiting Break Total Break Flow Rate ........................................................................... 5-5 5.3 Limiting Break Core Inlet Flow Rate ............................................................................. 5-6 5.4 Limiting Break Core Outlet Flow Rate ........................................................................... 5-6 5.5 Limiting Break Intact Loop Jet Pump Exit Flow Rate ..................................................... 5-7 5.6 Limiting Break Broken Loop Jet Pump Exit Flow Rate .................................................. 5-7 5.7 Lim iting Break A DS Flow R ate ..................................................................................... 5-8 5.8 Lim iting Break HP C I Flow R ate .................................................................................... 5-8 5.9 Limiting Break LPCS Flow Rate .................................................................................... 5-9 5.10 Limiting Break Intact Loop LPCI Flow Rate .................................................................. 5-9 5.11 Limiting Break Upper Downcomer Mixture Level ........................................................ 5-10 5.12 Limiting Break Lower Downcomer Mixture Level ........................................................ 5-10 5.13 Limiting Break Intact Loop Discharge Line Liquid Mass .............................................. 5-11 5.14 Limiting Break Upper Plenum Liquid Mass ................................................................. 5-11 5.15 Limiting Break Lower Plenum Liquid Mass ................................................................. 5-12 5.16 Limiting Break Hot Channel Inlet Flow Rate ............................................................... 5-12 5.17 Limiting Break Hot Channel Outlet Flow Rate ............................................................. 5-13 5.18 Limiting Break Hot Channel Coolant Temperature at the Limiting Node at EOB ......... 5-13 5.19 Limiting Break Hot Channel Quality at the Limiting Node at EOB ............................... 5-14 5.20 Limiting Break Hot Channel Heat Transfer Coefficient at the Limiting Node at EOB ...5-14 5.21 Limiting TLO Recirculation Line Break Hot Channel Reflood Junction Liquid Ma ss F lo w R a te .......................................................................................................... 5-15 5.22 Limiting Break Cladding Temperatures ....................................................................... 5-15 AREVA NP Inc.
uontroIned uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page iv 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 TCD thermal conductivity degradation TLO two-loop operation USAR Updated Safety Analysis Report AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 1-1 1.0 Introduction The results of the loss-of-coolant accident emergency core cooling system (LOCA-ECCS) analyses for Monticello are documented in this report. The purpose of the LOCA-ECCS analysis is to specify the maximum average planar linear heat generation rate (MAPLHGR) limit versus exposure for ATRIUMTM 1OXM* fuel and to demonstrate that the MAPLHGR limit is adequate to ensure that the LOCA-ECCS criteria in 10 CFR 50.46 are satisfied for operation at or below the limit. The report also documents the licensing basis peak cladding temperature (PCT) and corresponding local cladding oxidation from the metal-water reaction (MWR) for ATRIUM 10XM fuel used at Monticello.
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. The EXEM BWR-2000 Evaluation Model and NRC approval are documented in Reference 2. A summary description of the LOCA analysis methodology is provided in Section 4.0.
The application of the EXEM BWR Evaluation Model for the Monticello LOCA break spectrum analysis is documented in Reference 1. 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 8). The LOCA conditions evaluated in Reference 1 include break size, type, location, axial power shape, and ECCS single failure. The limiting LOCA break characteristics identified in Reference 1 are presented below:
Limiting LOCA Break Characteristics Location Recirculation suction pipe Type / size Split / 3.3 ft2 Single failure LPCI injection valve Axial power shape Mid-peaked Initial state 102% power I [ ]
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 1-2 The LOCA break spectrum analysis documented in Reference 1 was based on a generic ATRIUM 1OXM neutronic design at beginning of life conditions. The PCT and MWR calculated for a fuel rod experiencing the fluid conditions during the limiting LOCA are affected by fuel characteristics that depend on the fuel assembly neutronic design and exposure (e.g., local rod power, stored energy). The fuel assembly heatup analysis results presented in this report are for the limiting (minimum margin to acceptance criteria) ATRIUM 1OXM neutronic design currently designed for use at Monticello. This includes all of the enriched lattices designed for an equilibrium cycle of ATRIUM 1OXM fuel and all of the enriched lattices designed for the initial reload of ATRIUM 1OXM in Cycle 28. The heatup analyses were performed using the fluid conditions from the limiting LOCA identified in Reference 1. Cycle-specific heatup analyses are performed to confirm that the MAPLHGR limits in this report ensure the 10 CFR 50.46 acceptance criteria are satisfied for nuclear designs used in future core designs. The results of cycle-specific heatup analyses will be reported in the Monticello cycle-specific Licensing Analysis report.
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.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 10XM Fuel Page 2-1 2.0 Summary The MAPLHGR limit was determined by applying the EXEM BWR-2000 Evaluation Model for the analysis of the limiting LOCA event. The exposure-dependent MAPLHGR limit for ATRIUM 1OXM fuel is shown in Figure 2.1. The MAPLHGR limit is applicable for the range of operating conditions represented by the EPU licensing amendment request (Reference 8).
Power- and flow-dependent MAPLHGR multipliers are not required. Operation with only one recirculation loop (single-loop operation) requires that a MAPLHGR multiplier of 0.70 be applied to the two-loop operation MAPLHGR limit. The results of these calculations confirm that the LOCA acceptance criteria in the Code of Federal Regulations (10 CFR 50.46) are met for operation at or below these MAPLHGR limits.
The response of the reactor system and hot channel during the limiting LOCA analysis from Reference 1 are presented in Section 5.0. The MAPLHGR analysis results for the limiting lattice design are presented in Section 5.0. The peak cladding temperature (PCT) and metal-water reaction (MWR) results for the ATRIUM 1OXM limiting lattice design are presented in Table 2.1.
Note that the analysis PCT documented in this report is lower than the limiting PCT given in the LOCA break spectrum report (Reference 1). [
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM M T
1OXM Fuel Paqe 2-2 I
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Uontrolned Uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM M T
1OXM Fuel Paqe 2-3 Table 2.1 LOCA Results for PCT Limiting Conditions Parameter ATRIUM 10XM Exposure (GWd/MTU) 0.0 Peak cladding temperature ('F) 2088°F Local cladding oxidation (max %) 3.50%
Planar average oxidation (max %) 0.73%
Total hydrogen generated
(% of total hydrogen possible) < 1.0%
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUMTM 1OXM Fuel Paqe 2-4 16.0 14.0 12.0 10.0
-J n
8.0 6.0 4.0
.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 Planar Average Exposure (GWd/MTU)
Average Planar ATRIUM 1OXM Exposure MAPLHGR (GWd/MTU) (kW/ft) 0.0 12.5 20.0 12.5 67.0 7.6 Figure 2.1 MAPLHGR Limit for Two-Loop Operation*
- A MAPLHGR multiplier of 0.7 is required for single-loop operation.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits 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 (ECOS). 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 required. The results of the Monticello ATRIUM 1OXM break spectrum calculations using EXEM BWR-2000 LOCA methodology are summarized in Reference 1.
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.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits 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 2. The ECCS must be designed such that the plant response to a LOCA meets the following acceptance criteria specified in 10 CFR 50.46:
0 The calculated maximum fuel element cladding temperature shall not exceed 2200'F.
Xcel Energy has informally requested that PCT _ 2150'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 operation of the ECCS, the calculated core temperature shall be maintained for the extended period of time required by the long-lived radioactivity remaining in the core.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits 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 is established for each ATRIUM 1OXM fuel type to ensure that these criteria are met. For jet pump BWRs, the most challenging criterion is that PCT must not exceed 2200'F.
LOCA analysis results demonstrating that the PCT, local oxidation, and hydrogen generation criteria are met are provided in Section 5.0. Compliance with these three criteria ensures that a coolable geometry is maintained. Compliance with the long-term coolability criterion is discussed in Reference 1.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM 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 2. 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 3). 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 2) 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 blowdown analysis is shown in Figure 4.3. This nodalization is consistent with that used in the topical report submitted to the NRC (Reference 2).
The RELAX 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 is used to calculate 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.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 4-2 a mid-peaked 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 4) 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.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM MT 1OXM Fuel Page 4-3 swelling and rupture models from NUREG-0630 have been incorporated into HUXY (Reference 5). The HUXY code complies with the 10 CFR 50 Appendix K criteria for LOCA Evaluation Models.
As reported in Reference 6, an improved method for calculation view factors has been implemented in the HUXY code. HUXY models the radiation heat transfer between the fuel rod of interest and other fuel rods, the internal water channel and the fuel channel. The new method for calculating view factors is more precise and has been used in the exposure-dependent HUXY calculations reported here.
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 2. 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 canister and fuel channel. Experimental data for AREVA 1Oxl 0 fuel which supports the use of the convective heat transfer coefficients listed in Appendix K is documented in Reference 7. 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 percent maximum local metal-water reactor (%MWR). The core average metal-water reaction (CMWR) criterion of less than 1.0% can be satisfied by demonstrating that the maximum planar average MWR calculated by HUXY is less than 1.0%.
4.4 [
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Paae 4-4 I
4.4.1 Calculation Approach I
4.5 PlantParameters The LOCA break spectrum analysis is performed using plant parameters provided by Xcel Energy. Table 4.1 provides a summary of reactor initial conditions that were determined to be limiting in the break spectrum analysis. Table 4.2 lists selected reactor system parameters.
The break spectrum analysis is performed for a full core of ATRIUM 1OXM fuel. Some of the key fuel parameters used in the break spectrum analysis are summarized in Table 4.3. A mid-peaked axial power shape, shown in Figure 4.6, was identified as the most conservative power shape for the limiting break (Reference 1).
4.6 ECCS Parameters The ECCS configuration is shown in Figure 4.5. Table 4.4- Table 4.7provide the important ECCS characteristics assumed in the LOCA break spectrum analysis. The assumed ECCS performance has been reconciled with the revised levels of ECCS performance associated with Containment Accident Pressure (Reference 9). The ECCS is modeled as fill junctions AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 4-5 connected to the appropriate reactor locations: LPCS injects into the upper plenum, HPCI injects into the upper downcomer, and LPCI injects into the recirculation line.
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. For the break spectrum analyses 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 automatic depressurization system (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 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 HPCI, LPCS, or LPCI due to high drywell pressure.
The recirculation discharge isolation valve (RDIV) parameters are shown in Table 4.8.
The single failures and the available ECCS for each failure are summarized in Table 4.9. The potential single failures are discussed in Section 5.1 of Reference 1.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 4-6 Table 4.1 Initial Conditions*
Reactor power (% of rated) 102 Reactor power (MWt) 2044.1 Steam flow rate (Mlb/hr) 8.51 Steam dome pressure (psia) 1038.7 Core inlet enthalpy (Btu/Ib) 523.6 ATRIUM 1OXM hot assembly MAPLHGR (kW/ft) 13.1 Axial power shape Figure 4.6
- 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.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel PaQe 4-7 Table 4.2 Reactor System Parameters Parameter Value Vessel ID (in) 205 Number of fuel assemblies 484 Recirculation suction pipe area (ft2) 3.679 Recirculation discharge pipe area (ft2 ) 3.679 AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 4-8 Table 4.3 ATRIUM 1OXM Fuel Assembly Parameters Parameter Value Fuel rod array 10x10 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-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Paae 4-9 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) Not used 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.
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Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 10XM Fuel Pacie 4-10 Table 4.5 Low-Pressure Coolant Injection Parameters Parameter Value Reactor pressure permissive for opening valves - analytical (psig) 350 Coolant temperature (maximum) (OF) 90 InitiatingSignals and Setpoints Water level* 422.1 High drywell pressure (psig) Not used Time Delays Total system delay from initiating signal until the system is ready to inject (sec) 53.2 LPCI injection valve stroke time (sec) 3 5t Delivered CoolantFlow Rate Vs. Pressure Flow Rate for Flow Rate for 2 Pumps 4 Pumps Injecting Into Injecting Into Vessel to 1 Recirculation 1 Recirculation Torus AP 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 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.
AREVA NP Inc.
uonitrouJec uocumenz Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM 1OXM Fuel M
T Page 4-11 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) Not used 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 3540 130 2700 300 1125 338 0
- Relative to vessel zero.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel PaQe 4-12 Table 4.7 Automatic Depressurization System Parameters Parameter Value Number of valves installed 3 Number of valves available* 3 Minimum flow capacity 791,000 of available valves at (Ibm/hr at psig) 1080 InitiatingSignals and Setpoints Water level (in)t 422.1 High drywell pressure (psig) Not used Time Delays 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.
AREVA NP Inc.
uontroiDedj uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 4-13 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-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM M 1OXM Fuel T
Paqe 4-14 Table 4.9 ECCS Single Failure Assumed Systems Failure Remaining*, t, Battery (SF-BATT) 3 ADS + 1 LPCS + 2 LPCI LPCI injection valve (SF-LPCI) 3 ADS + 2 LPCS + 1 HPCI Diesel generator (SF-DGEN) 3 ADS + 1 LPCS + 1 HPCI + 2 LPCI HPCI system (SF-HPCI) 3 ADS + 2 LPCS + 4 LPCI ADS valve (SF-ADS) 2 ADS + 2 LPCS + 1 HPCI + 4 LPCI
- 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 > 0.4 ft 2. No LPCI flow is credited for breaks < 0.4 ft2.
AREVA NP Inc.
Uontroneci uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Paqe 4-15 System Analysis (RELAX)
Boundary Conditions (power, upper & lower plenum conditions)
Hot Assembly Analysis*
Fuel Stored Energy (RELAX)
Boundary Conditions Gap, (Pressure, Temperature, Power, Gap Coefficient, Quality, Heat Transfer Coefficient)
Fission Gas Time of Hot Node Reflood Heatup Analysis End of Blowdown, Time of Bvyass Refloi (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 NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Paae 4-16 Figure 4.2 [
I AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Paoe 4-17 Figure 4.3 RELAX System Blowdown Model AREVA NP Inc.
Controlled Ujocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Paqe 4-18 Figure 4.4 RELAX Hot Channel Blowdown Model Mid-Peaked Axial AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 4-19 Figure 4.5 ECCS Schematic AREVA NP Inc.
Uontrooved uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Paqe 4-20 Figure 4.6 Axial Power Distribution for Limiting LOCA Case in RELAX Calculation AREVA NP Inc.
uon1troiiec uocumenfL Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 5-1 5.0 MAPLHGR Analysis Results An exposure-dependent MAPLHGR limit for ATRIUM 10XM fuel is obtained by performing HUXY heatup analyses using results from the limiting LOCA analysis case identified in Reference 1. The break characteristics for the limiting analysis are summarized in Section 1.0.
Table 5.1 shows event times for the analysis. The response of the reactor system is shown in Figures 5.1 - 5.21. In the MAPLHGR analysis, the ATRIUM 10XM fuel rod stored energy is set to be bounding at all exposures and the RELAX hot channel peak power node is modeled at the highest MAPLHGR, which is 102% of 13.1 kW/ft for the ATRIUM 1OXM fuel. [
I Table 5.2 shows the MAPLHGR analysis results for the ATRIUM 1OXM fuel. The HUXY model of the ATRIUM 1OXM fuel is applied to obtain these results as described in Section 4.3. The HUXY analysis is performed at exposure intervals between 0 and 67 GWd/MTU. Some of the Monticello neutronic lattice designs experienced higher PCT than the generic ATRIUM 10XM lattice design used in the Monticello break spectrum analyses reported in Reference 1. In order to obtain margin to the PCT criterion the MAPLHGR limit was reduced from 13.1 to 12.5 kW/ft.
The HUXY MAPLHGR input is consistent with the data in Figure 2.1. Exposure-dependent ATRIUM 1OXM fuel rod data is provided from RODEX2 results and includes gap coefficient, hot gap thickness, cold gap thickness, gas moles, fuel rod plenum length, and spring relaxation time. This data is provided as a function of linear heat generation rate at each exposure analyzed.
The ATRIUM 10XM limiting PCT is 2088°F at the 0.0 GWd/MTU exposure. The corresponding maximum planar average cladding oxidation at the PCT limiting exposure is 0.73%. Analysis results show the CMWR is less than 1.0% total hydrogen generated.
Error! Reference source not found. shows temperatures for the limiting ATRIUM 1OXM neutronic lattice as a function of time for the limiting break. The PCT is 2088OF and occurs at 141.1 seconds. These results demonstrate the acceptability of the ATRIUM 10XM MAPLHGR limit shown in Figure 2.1.
5.1 Thermal Conductivity Degradation The RODEX2 code was approved by the NRC in the early 1980s. At that time, thermal conductivity degradation with burnup was not well characterized by irradiation or post-irradiation AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 5-2 testing. As a result, all fuel codes at that time did not account for thermal conductivity degradation (TCD). In the past 20 years, requests to the NRC have been made for commercial fuel operation to increasingly higher burnup levels. This has resulted in renewed interest in the degree and nature of burnup-induced TCD. Interactions between AREVA and the NRC on this topic are summarized in Appendix E of Reference 12, which summarizes how TCD is addressed in the AREVA analyses for Monticello.
The newer RODEX4 code explicitly incorporates the impact of TCD with exposure (Reference 13). RODEX4 calculations were performed with and without the models which account for TCD. The differences in the RODEX4 results were used to increase the stored energy calculated by RODEX2 prior to their input to HUXY.
The impact of TCD is included in the results summarized in Table 5.2. The assessment for Monticello ATRIUM 1 OXM fuel shows that the PCT calculated at 0.0 GWd/MTU, which is not affected by TCD, is still the highest exposure-dependent PCT when TCD is evaluated.
AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM M T
1OXM Fuel PaQe 5-3 Table 5.1 Event Times for Limiting Break 3.3 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 4.8 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 LPCS permissive for ADS 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 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 153.0 AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 10XM Fuel Paae 5-4 Table 5.2 ATRIUM 1OXM MAPLHGR Analysis Results Average Local Planar Cladding Exposure MAPLHGR PCT Oxidation (GWd/MTU) (kW/ft) (OF) (%)
0 12.50 2088 3.50 5 12.50 1990 1.65 10 12.50 1902 1.24 15 12.50 1869 1.09 20 12.50 1970 1.67 30 11.46 1780 0.66 40 10.41 1699 0.49 50 9.37 1622 0.34 60 8.33 1620 0.32 67 7.60 1570 0.26 AREVA NP Inc.
Uonirolled Uociument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 5-5 Figure 5.1 Limiting Break Upper Plenum Pressure Figure 5.2 Limiting Break Total Break Flow Rate AREVA NP Inc.
Uontroioed Uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 5-6 Figure 5.3 Limiting Break Core Inlet Flow Rate Figure 5.4 Limiting Break Core Outlet Flow Rate AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Paae 5-7 Figure 5.5 Limiting Break Intact Loop Jet Pump Exit Flow Rate Figure 5.6 Limiting Break Broken Loop Jet Pump Exit Flow Rate AREVA NP Inc.
uontroIned Uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM 1OXM Fuel M
T Page 5-8 I ]
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0 20 40 60 80 100 120 140 160 TIME (SEC)
Figure 5.7 Limiting Break ADS Flow Rate 5
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Figure 5.8 Limiting Break HPCI Flow Rate AREVA NP Inc.
uontroliei Uocumen:
Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM M T
1OXM Fuel Page 5-9 0
-J LL U)q i I i i I i I I I i I 1 20 40 60 80 100 120 140 160 TIME (SEC)
Figure 5.9 Limiting Break LPCS Flow Rate K ]
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9 II I II I I I I I 0 20 40 60 80 100 120 140 160 TIME (SEC)
Figure 5.10 Limiting Break Intact Loop LPCI Flow Rate AREVA NP Inc.
Uontrouled Uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 5-10 K ]
0 -~lo O
IJto 0
Q-Z a-)
zw ,o I 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 5.11 Limiting Break Upper Downcomer Mixture Level K ]
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0 z
0 0
I i I i I I iI I I I i I i I 0 20 40 60 80 100 120 140 160 TIME (SEC)
Figure 5.12 Limiting Break Lower Downcomer Mixture Level AREVA NP Inc.
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Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM M T
1OXM Fuel Page 5-11 EO a 0q co C-r I i I iI I I i I i I i I I i I 0 20 40 60 80 100 120 140 160 TIME (SEC)
Figure 5.13 Limiting Break Intact Loop Discharge Line Liquid Mass K ]
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Uon'trolled Uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUMTM 1OXM Fuel Page 5-12 ILI iIiI IiI]
CY 3'0 I ,,-I I I I uJ0 z
-j 0
0 I i II I I I i I i I 20 40 60 80 100 120 140 160 TIME (SEC)
Figure 5.15 Limiting Break Lower Plenum Liquid Mass
-J q 0-j Q F-z 0
zo 0 20 40 60 80 100 120 140 160 TIME (SEC)
Figure 5.16 Limiting Break Hot Channel Inlet Flow Rate AREVA NP Inc.
(jontrol~edj Uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 5-13 E- I ' I ' I I I
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Figure 5.17 Limiting Break Hot Channel Outlet Flow Rate Figure 5.18 Limiting Break Hot Channel Coolant Temperature at the Limiting Node at EOB AREVA NP Inc.
Uontroaned Uocument Monticello EPU ANP-3212(NP)
LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Paqe 5-14 Figure 5.19 Limiting Break Hot Channel Quality at the Limiting Node at EOB Figure 5.20 Limiting Break Hot Channel Heat Transfer Coefficient at the Limiting Node at EOB AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 5-15 Figure 5.21 Limiting TLO Recirculation Line Break Hot Channel Reflood Junction Liquid Mass Flow Rate MONT CY28 0 GWd/MTU (XMLCB-4309L-13GV75 Iattice) 2500 1 . .I . II . I .
2500 - PCT Rod SWater Channel A- ,Fuel Channel 2000 1500 0
E
¢ 1000 C
50 500 0
20 40 60 80 100 120 140 160 Time (sec)
Figure 5.22 Limiting Break Cladding Temperatures AREVA NP Inc.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 6-1 6.0 Conclusions The EXEM BWR-2000 Evaluation Model was applied to determine the ATRIUM 1OXM MAPLHGR limit for Monticello. The following conclusions were made from the analyses presented.
The acceptance criteria of the Code of Federal Regulations (10 CFR 50.46) are met for operation at or below the ATRIUM 1OXM MAPLHGR limit given in Figure 2.1.
- Peak PCT < 2200 0F.
- Local cladding oxidation thickness < 0.17.
- Total hydrogen generation < 0.01.
- Coolable geometry satisfied by meeting peak PCT, local cladding oxidation, and total hydrogen generation criteria.
- Core long-term cooling satisfied by concluding core flooded to top of active fuel or core flooded to the jet pump suction elevation with one core spray operating (Reference 1).
The MAPLHGR limit is applicable for ATRIUM 1OXM full cores as well as transition cores containing ATRIUM 1OXM fuel.
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LOCA-ECCS Analysis MAPLHGR Revision 0 Limits for ATRIUM TM 1OXM Fuel Page 7-1 7.0 References
- 1. ANP-3211 (P) Revision 0, Monticello EPU LOCA Break Spectrum Analysis for ATRIUM TM IOXM Fuel, AREVA NP, May 2013.
- 3. 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.
- 4. 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.
- 5. XN-NF-82-07(P)(A) Revision 1, Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, Exxon Nuclear Company, November 1982.
- 6. Letter, P. Salas (AREVA) to Document Control Desk, U.S. Nuclear Regulatory Commission, "2010- Annual Reporting of Changes and Errors in Emergency Core Cooling Systems (ECCS) Evaluation Models," NRC: 1:123, December 20, 2011.
- 7. EMF-2292(P)(A) Revision 0, ATRIUM TM-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
- 8. 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).
- 9. Letter, M.A. Schmimel (NSPM) to Document Control Desk (NRC), "Monticello Extended Power Uprate and Maximum Extended Load Line Limit Analysis Plus License 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).
- 10. Letter, P. Salas (AREVA) to Document Control Desk, U.S. Nuclear Regulatory Commission, "Proprietary Viewgraphs and Meeting Summary for Closed Meeting on Application of the EXEM BWR-2000 ECCS Evaluation Methodology," NRC:I 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.
- 12. ANP-3224P Revision 0, Applicability of AREVA NP BWR Methods to Monticello, AREVA NP, May 2013.
- 13. BAW-1 0247PA Revision 0, Realistic Thermal-MechanicalFuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.
- 14. ANP-3158P Revision 0, Fuel Rod Thermal-MechanicalDesign for Monticello ATRIUM 1OXM Fuel Assemblies, Cycle 28, AREVA NP, February 2013.
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LOCA-ECCS Analysis MAPLHGR Preliminary Revision 0 Limits for ATRIUM TM 1OXM Fuel Page A-1 APPENDIX A Supplemental Information Table A.1 Computer Codes Used for MAPLHGR Limit Analysis 7
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