ML13200A192
| ML13200A192 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/31/2013 |
| From: | AREVA NP |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML13200A185 | List: |
| References | |
| L-MT-13-055 ANP-3092(NP), Rev 0 | |
| Download: ML13200A192 (29) | |
Text
{{#Wiki_filter:Enclosure 11 AREVA Report ANP-3092(NP) Monticello Thermal-Hydraulic Design Report for ATRIUM 1 OXM Fuel Assemblies Revision 0 28 pages follow uontroliici Luocumrnet ANP-3092(NP) Revision 0 Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies July 2012 A AREVA AREVA NP Inc. (Jontrolled uocument AREVA NP Inc.ANP-3092(NP) Revision 0 Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies Uontrolled uocument AREVA NP Inc.ANP-3092(NP) Revision 0 Copyright © 2012 AREVA NP Inc.All Rights Reserved kIk uontroiled Uocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Paqe i Nature of Changes Item Page Description and Justification
- 1. All This is the initial release.AREVA NP Inc.
uontroled LUocument Monticello Thermal-Hydraulic ANP-3092(NP) Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page ii Contents 1 .0 In tro d u c tio n .................................................................................................................. 1-1 2.0 Summary and Conclusions ........................................................................................... 2-1 3.0 Thermal-Hydraulic Design Evaluation ........................................................................... 3-1 3.1 Hydraulic Characterization ................................................................................ 3-2 3.2 Hydraulic Compatibility ...................................................................................... 3-3 3.3 Thermal Margin Performance ............................................................................ 3-4 3 .4 R o d B o w ........................................................................................................... 3 -5 3 .5 B y p a ss F lo w ..................................................................................................... 3-6 3 .6 S ta b ility ............................................................................................................. 3 -6 4 .0 R e fe re n c e s ................................................................................................................... 4 -1 Tables 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM 1OXM Fuel Assembly ..................................................................................... 3-7 3.2 Comparative Description for Monticello ATRIUM 1OXM and GE14 Fuel T y p e s ......................................................................................................................... 3 -1 0 3.3 Hydraulic Characterization Comparison for Monticello ATRIUM 1OXM and G E 1 4 F u e l T y p e s ....................................................................................................... 3 -1 1 3.4 Monticello Thermal-Hydraulic Design Conditions ........................................................ 3-12 3.5 Monticello First Transition Core Thermal-Hydraulic Results at Rated Conditions (100%P / 100%F) ..................................................................................... 3-13 3.6 Monticello First Transition Core Thermal-Hydraulic Results at Off-Rated Conditions (59.2%P / 43.3%F) ................................................................................... 3-14 3.7 Monticello Thermal-Hydraulic Results at Rated Conditions (100%P /100%F) for Transition to ATRIUM 1OXM Fuel ............................................................. 3-15 3.8 Monticello Thermal-Hydraulic Results at Off-Rated Conditions (59.2%P /43.3%F) for Transition to ATRIUM 1OXM Fuel ............................................................ 3-16 Figures 3.1 Axial Power Shapes ................................................................................................... 3-17 3.2 First Transition Core: Hydraulic Demand Curves 100%P / 100%F ............................. 3-18 3.3 First Transition Core: Hydraulic Demand Curves 59.2%P / 43.3%F ........................... 3-19 AREVA NP Inc. Lontrolled Uocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Page iii Nomenclature AOO ASME BWR CFR CHF CPR CRDA LOCA LTP MAPLHGR MCPR NRC OLMCPR PLFR RPF SLMCPR UTP ACPR anticipated operational occurrence American Society of Mechanical Engineers boiling water reactor code of federal regulations critical heat flux critical power ratio control rod drop accident loss-of-coolant accident lower tie plate maximum average planar linear heat generation rate minimum critical power ratio Nuclear Regulatory Commission, U.S.operating limit minimum critical power ratio part-length fuel rod radial peaking factor safety limit minimum critical power ratio upper tie plate change in critical power ratio AREVA NP Inc. (ontrolled uocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Paqe 1-1 1.0 Introduction This report provides the thermal-hydraulic characterization of the AREVA NP ATRIUM T M 1OXM*and the coresident GE14 fuel designs for Monticello. To ensure the ATRIUM 1OXM fuel will be hydraulically compatible with the coresident GE14 fuel, the results from Monticello thermal-hydraulic analyses will be compared to the acceptance criteria established in U.S. Nuclear Regulatory Commission (NRC) approved topical reports ANF-89-98(P)(A) Revision 1 and Supplement 1 (Reference
- 1) and XN-NF-80-19(P)(A)
Volume 4 Revision 1 (Reference 2).* ATRIUM is a trademark of AREVA NP.AREVA NP Inc. uontroiled Uocument Monticello Thermal-Hydraulic ANP-3092(NP) Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 2-1 2.0 Summary and Conclusions ATRIUM 1OXM fuel assemblies have been determined to be hydraulically compatible with the coresident GE14 fuel design in the Monticello reactor for the entire range of the licensed power-to-flow operating map. Detailed calculation results supporting this conclusion are provided in Section 3.2. Results for coresident GE14 and ATRIUM 10XM fuel in a representative first transition core are provided in Table 3.5 and Table 3.6. Results for a full core of GE14, for coresident GE14 and ATRIUM 1OXM fuel in representative first and second transition cores and for a full core of ATRIUM 1OXM fuel are summarized in Table 3.7 and Table 3.8.The ATRIUM 1OXM fuel design is geometrically different from the coresident GE14 fuel design, but the designs are hydraulically compatible. [I Core bypass flow (defined as leakage flow through the lower tie plate (LTP) flow holes, channel seal, core support plate, and LTP-fuel support interface) is not adversely affected by the introduction of the ATRIUM 1OXM fuel design. Analyses at rated conditions with a middle-peaked power shape show core bypass flow varying between [ ] of rated flow for core configurations ranging from a full core of GE14 fuel to a full core of ATRIUM 1OXM, respectively. Analyses demonstrate the thermal-hydraulic design and compatibility criteria discussed in Section 3.0 are satisfied for the Monticello transition cores consisting of GE14 and ATRIUM 1OXM fuel for the expected core power distributions and core power/flow conditions encountered during operation. AREVA NP Inc. uontrolled uocument Monticello Thermal-Hydraulic ANP-3092(NP) Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-1 3.0 Thermal-Hydraulic Design Evaluation Thermal-hydraulic analyses are performed to verify that design criteria are satisfied and to help establish thermal operating limits with acceptable margins of safety during normal reactor operation and anticipated operational occurrences (AOOs). The design criteria that are applicable to the ATRIUM 1OXM fuel design are described in Reference 1 (Section 4.0). To the extent possible, these analyses are performed on a generic fuel design basis. These analyses remain applicable to the ATRIUM 1OXM fuel transition cycle and to subsequent cycles as long as there are no changes to the mechanical fuel design. However, due to reactor and cycle operating differences, many of the analyses supporting these thermal-hydraulic operating limits are performed on a plant- and cycle-specific basis and are documented in plant- and cycle-specific reports.The thermal-hydraulic design criteria are summarized below: Hydraulic compatibility (Reference 1, Section 4.1.1). The hydraulic flow resistance of the reload fuel assemblies shall be sufficiently similar to the existing fuel in the reactor such that there is no significant impact on total core flow or the flow distribution among assemblies in the core. This criterion evaluation is addressed in Sections 3.1 and 3.2.Thermal margin performance (Reference 1, Section 4.1.2). Fuel assembly geometry, including spacer design and rod-to-rod local power peaking, should minimize the likelihood of boiling transition during normal reactor operation as well as during AQOs.The fuel design should fall within the bounds of the applicable empirically based boiling transition correlation approved for AREVA reload fuel. Within other applicable mechanical, nuclear, and fuel performance constraints, the fuel design should achieve good thermal margin performance. The thermal-hydraulic design impact on steady-state thermal margin performance is addressed in Section 3.3. Additional thermal margin performance evaluations dependent on the cycle-specific design are addressed in the Monticello ATRIUM 1OXM Fuel Transition Reload Licensing Analysis report and in subsequent cycle-specific Monticello Reload Licensing Analysis reports.Fuel centerline temperature (Reference 1, Section 4.1.3). Fuel design and operation shall be such that fuel centerline melting is not projected for normal operation and AQOs. This criterion evaluation is addressed in the Fuel Rod Thermal-Mechanical Design for Monticello ATRIUM 1OXM Fuel report.Rod bow (Reference 1, Section 4.1.4). The anticipated magnitude of fuel rod bowing under irradiation shall be accounted for in establishing thermal margin requirements. This criterion evaluation is addressed in Section 3.4.Bypass flow (Reference 1, Section 4.1.5). The bypass flow characteristics of the reload fuel assemblies shall not differ significantly from the existing fuel in order to provide adequate flow in the bypass region. This criterion evaluation is addressed in Section 3.5.AREVA NP Inc. uontronled Vocument Monticello Thermal-Hydraulic ANP-3092(NP) Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-2 Stability (Reference 1, Section 4.1.6). Reactors fueled with new fuel designs must be stable (decay ratio < 1.0) in the approved power and flow operating region. The stability performance of new fuel designs will be equivalent to, or better than, existing (approved) AREVA fuel designs. This criterion evaluation is addressed in Section 3.6. Additional core stability evaluations dependent on the cycle-specific design are addressed in the Monticello ATRIUM 1OXM Fuel Transition Reload Licensing Analysis report and in subsequent cycle-specific Monticello Reload Licensing Analysis reports.Loss-of-coolant accident (LOCA) analysis (Reference 1, Section 4.2). LOCAs are analyzed in accordance with Appendix K modeling requirements using NRC-approved models. The criteria are defined in title 10 of the code of federal regulations (CFR)50.46. LOCA analysis results are presented in the Monticello LOCA Break Spectrum Analysis for ATRIUM 1OXM Fuel and Monticello LOCA MAPLHGR Analysis for ATRIUM 1OXM Fuel reports.Control rod drop accident (CRDA) analysis (Reference 1, Section 4.3). The deposited enthalpy must be less than 280 cal/gm for fuel coolability. This criterion evaluation is addressed in the Monticello ATRIUM 1OXM Fuel Transition Reload Licensing Analysis report and in subsequent cycle-specific Monticello Reload Licensing Analysis reports.ASME overpressurization analysis (Reference 1, Section 4.4). ASME pressure vessel code requirements must be satisfied. This criterion evaluation is addressed in the Monticello ATRIUM 1OXM Fuel Transition Reload Licensing Analysis report and in subsequent cycle-specific Monticello Reload Licensing Analysis reports.Seismic/LOCA liftoff (Reference 1, Section 4.5). Under accident conditions, the assembly must remain engaged in the fuel support. This criterion evaluation is addressed in the Mechanical Design Report for Monticello ATRIUM 1OXM Fuel Assemblies. A summary of the thermal-hydraulic design evaluations is given in Table 3.1.3.1 Hydraulic Characterization Basic geometric parameters for the GE14 and ATRIUM 1OXM fuel designs are summarized in Table 3.2. Component loss coefficients for the ATRIUM 1OXM are based on tests and are presented in Table 3.3. These loss coefficients include modifications to the test data reduction process [] The bare rod friction, ULTRAFLOW T M* spacer, UTP and LTP losses for ATRIUM 1 OXM are based on tests performed at AREVA's Portable Hydraulic Test Facility.I] The local component (LTP, spacer, and UTP) loss coefficients for the GE14 fuel are based on flow test results.* ULTRAFLOW is a trademark of AREVA NP.AREVA NP Inc. Uontrolled Uocument Monticello Thermal-Hydraulic ANP-3092(NP) Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-3 The primary resistance for the leakage flow through the LTP flow holes is [] The resistances for the leakage paths are shown in Table 3.3.3.2 Hydraulic Compatibility The thermal-hydraulic analyses were performed in accordance with the AREVA thermal-hydraulic methodology for BWRs. The methodology and constitutive relationships used by AREVA for the calculation of pressure drop in BWR fuel assemblies are presented in Reference 3 and are implemented in the XCOBRA code. The XCOBRA code predicts steady-state thermal-hydraulic performance of the fuel assemblies of BWR cores at various operating conditions and power distributions. XCOBRA received NRC approval in Reference 4.The NRC reviewed the information provided in Reference 5 regarding inclusion of water rod models in XCOBRA and accepted the inclusion in Reference 6.Hydraulic compatibility, as it relates to the relative performance of the ATRIUM 1OXM and GE14 fuel designs, has been evaluated. Detailed analyses were performed for full core GE14 and full core ATRIUM 1OXM configurations. Analyses for mixed cores with GE14 and ATRIUM 1OXM fuel were also performed to demonstrate the thermal-hydraulic design criteria are satisfied for transition core configurations. The hydraulic compatibility analysis is based on [Table 3.4 summarizes the input conditions for the analyses. These conditions reflect two of the state points considered in the analyses: 100% power/1 00% flow and 59.2% power/43.3% flow, which is the core flow at the minimum pump speed on the MELLLA line. Table 3.4 also defines the core loading for the representative transition core configurations. Input for other core configurations is similar in that core operating conditions remain the same and the same axial AREVA NP Inc. uontrolled uocument Monticello Thermal-Hydraulic ANP-3092(NP) Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-4 power distribution is used. Evaluations were made with the bottom-, middle-, and top-peaked axial power distributions presented in Figure 3.1. Results presented in this report are for the middle-peaked power distribution. Results for bottom- and top-peaked axial power distributions show similar trends.Table 3.5 and Table 3.6 provide a summary of calculated thermal-hydraulic results using the first transition core configuration. Table 3.7 and Table 3.8 provide a summary of results for all core configurations evaluated. Core average results and the differences between the ATRIUM 1OXM and GE14 results at rated power are within the range considered compatible, as expected. Similar agreement occurs at lower power levels. As shown in Table 3.5, [] Table 3.6 shows that [] Differences in assembly flow between the ATRIUM 1OXM and GE14 fuel designs as a function of assembly power level are shown in Figure 3.2 and Figure 3.3.Core pressure drop and core bypass flow fraction are also provided for the configurations evaluated. Based on the reported changes in pressure drop and assembly flow caused by the transition from a full core of GE14 to a full core of ATRIUM 1OXM, the ATRIUM 1OXM design is considered hydraulically compatible with the coresident fuel design since the thermal-hydraulic design criteria are satisfied. 3.3 Thermal Margin Performance Relative thermal margin analyses were performed in accordance with the thermal-hydraulic methodology for AREVA's XCOBRA code. The calculation of the fuel assembly critical power ratio (CPR) (thermal margin performance) is established by means of an empirical correlation based on results of boiling transition test programs. The CPR methodology is the approach used by AREVA to determine the margin to thermal limits for BWRs.On a cycle-specific basis analyses will be performed to ensure the fuel design minimizes the likelihood of boiling transition during normal operation as well as during AQOs. This protection is accomplished by determining the operating limit minimum critical power ratio (OLMCPR) for each fuel bundle in the reactor core. The OLMCPR is comprised of the core limiting safety limit minimum critical power ratio (SLMCPR) and the limiting transient ACPR. The limiting transient AREVA NP Inc. uontrolled uocument Monticello Thermal-Hydraulic ANP-3092(NP) Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-5 ACPR is determined during the evaluation of AOOs and bounding accidents. Therefore, on a cycle-specific basis good thermal margin performance is achieved by establishing an SLMCPR that predicts < 99.9% of rods to be in boiling transition. CPR values for ATRIUM 1OXM are calculated with the ACE/ATRIUM 1OXM critical power correlation (Reference
- 7) while the CPR values for the GE14 fuel are calculated with the SPCB critical power correlation (Reference 8). The NRC-approved methodology to demonstrate the acceptability of using the SPCB correlation for computing GE14 fuel CPR is presented in Reference
- 9. Assembly design features are incorporated in the CPR calculation through the K-factor term for the ACE correlation and the F-eff term for the SPCB correlation.
The K-factors and F-effs are made up of two parts which are added together. The first part depends on the local power peaking in the fuel assembly, which depends on the nuclear design and is a function of void fraction and exposure. The second part is called an additive constant, which is determined for each rod position based on critical power testing and calculated using the methods approved in References 7 and 8.For the compatibility evaluation, steady-state analyses evaluated ATRIUM 1OXM and GE14 assemblies with radial peaking factors (RPFs) between [I Table 3.5 and Table 3.6 show CPR results of the ATRIUM 1OXM and GE14 fuels. Table 3.7 and Table 3.8 show similar comparisons of CPR and assembly flow for the various core configurations evaluated. Analysis results indicate ATRIUM 1OXM fuel will not cause thermal margin problems for the coresident fuel design.3.4 Rod Bow The bases for rod bow are discussed in the Mechanical Design Report for Monticello ATRIUM 1OXM Fuel Assemblies. Rod bow magnitude is determined during the fuel-specific AREVA NP Inc. Uontrolled Uocument Monticello Thermal-Hydraulic ANP-3092(NP) Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-6 mechanical design analyses. Rod bow has been measured during post-irradiation examinations of BWR fuel fabricated by AREVA.3.5 Bypass Flow Total core bypass flow is defined as leakage flow through the LTP flow holes, channel seal, core support plate, and LTP-fuel support interface. Table 3.7 shows that total core bypass flow (excluding water rod flow) fraction at rated conditions changes from [ ] of rated core flow during the transition from a full core of GE14 to a full core of ATRIUM 1OXM (middle-peaked power shape). In summary, adequate bypass flow will be available with the introduction of the ATRIUM 1OXM fuel design and applicable design criteria are met.3.6 Stability Each new fuel design is analyzed to demonstrate that the stability performance is equivalent to or better than an existing (NRC-approved) AREVA fuel design. The stability performance is a function of the core power, core flow, core power distribution, and to a lesser extent, the fuel design. [] A comparative stability analysis was performed with the NRC-approved STAIF code (Reference 11). The study shows that the ATRIUM 1OXM fuel design has decay ratios equivalent to or better than other approved AREVA fuel designs.As stated above, the stability performance of a core is strongly dependent on the core power, core flow, and power distribution in the core. Therefore, core stability is evaluated on a cycle-specific basis and addressed in the Monticello ATRIUM 1OXM Fuel Transition Reload Licensing Analysis report and in subsequent cycle-specific Monticello Reload Licensing Analysis reports.AREVA NP Inc. UontrouIed Document Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Page 3-7 Table 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM 1OXM Fuel Assembly Report Section Description Criteria Results or Disposition Thermal and Hydraulic Criteria 3.1 / 3.2 Hydraulic Hydraulic flow resistance Verified on a plant-specific basis.compatibility shall be sufficiently ATRIUM 1OXM demonstrated to be similar to existing fuel compatible with GE14 fuel.such that there is no significant impact on total core flow or flow distribution among assemblies.(Reference 1, Section 4.1.1)If there is more than an[additional stability evaluations will be performed with the approved STAIF code.(Reference 1 Supplement 1, page 18)3.3 Thermal margin Fuel design shall be ACE/ATRIUM 1OXM critical power performance within the limits of correlation is applied to the applicability of an ATRIUM 1OXM fuel.approved CHF SPCB critical power correlation is correlation.(Reference 1, applied to the GEl4 fuel.Section 4.1.2)< 99.9% rods in boiling Verified on cycle-specific basis.transition. (Reference 1, Table 4.1)Fuel centerline No centerline melting. Plant- and fuel-specific analyses temperature (Reference 1, are performed. Section 4.1.3)AREVA NP Inc. Uontrolled Uocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Page 3-8 Table 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM 1OXM Fuel Assembly (Continued) Report Section Description Criteria Results or Disposition Thermal and Hydraulic Criteria (Continued) 3.4 Rod bow Rod bow must be The lateral displacement of the fuel accounted for in rods due to fuel rod bowing is not of establishing thermal sufficient magnitude to impact margins. (Reference 1, thermal margins.Section 4.1.4)3.5 Bypass flow Bypass flow Verified on a plant-specific basis.characteristics shall be Analysis results demonstrate that similar among adequate bypass flow is provided.assemblies to provide adequate bypass flow.(Reference 1, Section 4.1.5)3.6 Stability New fuel designs are stable (decay ratio < 1.0)in the approved power and flow operating region, and stability performance will be equivalent to (or better than) existing (approved) AREVA fuel designs.(Reference 1, Section 4.1.6)ATRIUM 1OXM channel and core decay ratios have been demonstrated to be equivalent to or better than other approved AREVA fuel designs.Core stability behavior is evaluated on a cycle-specific basis.LOCA analysis LOCA analyzed in Approved Appendix K LOCA accordance with model.Appendix K modeling Plant- and fuel-specific analysis requirements. Criteria with cycle-specific verifications. defined in 10 CFR 50.46.(Reference 1, Section 4.2)CRDA analysis < 280 cal/gm for Cycle-specific analysis is coolability. performed.(Reference 1, Section 4.3)AREVA NP Inc. Uontrolecd uocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Page 3-9 Table 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM 1OXM Fuel Assembly (Continued) Report Section Description Criteria Results or Disposition Thermal and Hydraulic Criteria (Continued) ASME over- ASME pressure vessel Cycle-specific analysis is pressurization code requirements shall performed. analysis be satisfied.(Reference 1, Section 4.4)Seismic/LOCA Assembly remains Criterion addressed in the liftoff engaged in fuel support. Mechanical Design Report for (Reference 1, Monticello ATRIUM 1OXM Fuel Section 4.5) Assemblies. AREVA NP Inc. uontrolled Uocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Paae 3-10 Table 3.2 Comparative Description for Monticello ATRIUM 1OXM and GE14 Fuel Types Fuel Parameter ATRIUM 1OXM GE14 Number of fuel rods Full-length fuel rods 79 78 PLFRs 12 14 Fuel clad OD, in 0.4047 0.404 Number of spacers 9 8 Active fuel length, in Full-length fuel rods 145.24 145.24 PLFRs 75.0 84.0 Hydraulic resistance characteristics Table 3.3 Table 3.3 Number of water rods 1 2 Water rod OD, in 1.378* 0.980* Square water channel outer width.AREVA NP Inc. Uontrolled uocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Page 3-11 Table 3.3 Hydraulic Characterization Comparison for Monticello ATRIUM 1OXM and GE14 Fuel Types I AREVA NP Inc. uontrolled Uocument Monticello Thermal-Hydraulic ANP-3092(NP) Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 3-12 Table 3.4 Monticello Thermal-Hydraulic Design Conditions Reactor Conditions 100%P / 100%F 59.2%P / 43.3%F Core power level, MWt 2004.0 1186.4 Core exit pressure, psia 1040.0 966.1 Core inlet enthalpy, Btu/Ibm 523.0 498.5 Total core coolant flow, Mlbm/hr 57.6 24.9 Axial power shape Middle-peaked Middle-peaked (Figure 3.1) (Figure 3.1)Number of Assemblies Central Peripheral Region Region Core Loading Full Core GE14[]F T First Transition Core Loading[ C[]Second Transition Core Loading[ ]Full Core ATRIUM 10XM[[AREVA NP Inc. uontrolled Uocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Page 3-13 Table 3.5 Monticello First Transition Core Thermal-Hydraulic Results at Rated Conditions (100%P / 100%F)I I I AREVA NP Inc. uontrolled Uocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Paqe 3-14 Table 3.6 Monticello First Transition Core Thermal-Hydraulic Results at Off-Rated Conditions (59.2%P / 43.3%F)I I AREVA NP Inc. Lontrollec Uocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Page 3-15 Table 3.7 Monticello Thermal-Hydraulic Results at Rated Conditions (100%P / 100%F) for Transition to ATRIUM 1OXM Fuel[I AREVA NP Inc. uontrolled Vocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Page 3-16 Table 3.8 Monticello Thermal-Hydraulic Results at Off-Rated Conditions (59.2%P / 43.3%F) for Transition to ATRIUM 1OXM Fuel I I AREVA NP Inc. uontroned uocurnentc Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Page 3-17[I Figure 3.1 Axial Power Shapes AREVA NP Inc. uontroouea uocumenit Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Page 3-18[Figure 3.2 First Transition Core: Hydraulic Demand Curves 100%P / 100%F AREVA NP Inc. uontroloed Uocument Monticello Thermal-Hydraulic Design Report for ATRIUM T M 1OXM Fuel Assemblies ANP-3092(NP) Revision 0 Page 3-19[I Figure 3.3 First Transition Core: Hydraulic Demand Curves 59.2%P / 43.3%F AREVA NP Inc. (Jontrollecd Uocument Monticello Thermal-Hydraulic ANP-3092(NP) Design Report for ATRIUM T M 1OXM Revision 0 Fuel Assemblies Page 4-1 4.0 References
- 1. ANF-89-98(P)(A)
Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.2. XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.3. XN-NF-79-59(P)(A), Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies, Exxon Nuclear Company, November 1983.4. XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.5. Letter, R.A. Copeland (ANF) to R.C. Jones (USNRC), "Explicit Modeling of BWR Water Rod in XCOBRA," RAC:002:90, January 9,1990.6. Letter, R.C. Jones (USNRC) to R.A. Copeland (ANF), no subject (regarding XCOBRA water rod model), February 1, 1990.7. ANP-10298PA Revision 0, ACE/ATRIUM IOXM Critical Power Correlation, AREVA NP, March 2010.8. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.9. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.10. ANP-1 0298PA Revision 0 Supplement 1 P Revision 0, Improved K-factor Model for ACE/ATRIUM IOXM Critical Power Correlation, AREVA NP, December 2011.11. EMF-CC-074(P)(A) Volume 1, STAIF -A Computer Program for BWR Stability Analysis in the Frequency Domain; and Volume 2, STAIF -A Computer Program for BWR Stability Analysis in the Frequency Domain -Code Qualification Report, Siemens Power Corporation, July 1994.AREVA NP Inc.}}