BSEP 10-0093, ANP-2936(NP), Revision 0, Brunswick, Unit 2, Thermal-Hydraulic Design Report for Atrium Tm 10XM Fuel Assemblies.
| ML102160375 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/30/2010 |
| From: | AREVA, AREVA NP |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| BSEP 10-0093, TSC-2010-01, TSC-2010-02 ANP-2936(NP), Rev 0 | |
| Download: ML102160375 (28) | |
Text
BSEP 10-0093 Enclosure 3 AREVA Report ANP-2936(NP), Revision 0 Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUMTM 1 OXM Fuel Assemblies dated July 2010
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ANP-2936(NP)
Revision 0 Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUM TM 1OXM Fuel Assemblies July 2010 AREVA NP Inc. AR EVA
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ANP-2936(NP)
Revision 0 Brunswick Unit 2 Thermal-Hydraulic M
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Revision 0 Copyright © 2010 AREVA NP Inc.
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page i Nature of Changes Item Page Description and Justification
- 1. All This is the initial issue.
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Design Report for ATRIUM TM 1OXM Revision 0 Fuel Assemblies Page ii Contents 1.0 Intro d uctio n ....................................................................................................... . . . . 1-1 2.0 Summary and Conclusions .................... ...... . ..................................... 2-1 3.0 Therm al-Hydraulic Design Evaluation ............................................................................ 3-1 3.1 H ydraulic C haracterization ................................................................................. 3-2 3.2 Hydraulic Compatibility ...................................... 3-3 3.3 T herm al Margin Perform ance .............................................................................. 3-4 3 .4 R o d B o w ............................................................................................................. 3 -5 3.5 Bypass Flow ............................................. 3-5 3 .6 S ta b ility ................................................................... .......................................... 3 -5 4 .0 R e fe re n c es ..................................................................................................................... 4 -1 Tables 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the A T R IU M 1OX M Fuel Assem bly ...................................................................................... 3-7 3.2 Comparative Description of Brunswick Unit 2 ATRIUM 10XM, AT R IU M-10, and G E 14 Fuel ........................................................................................ 3-9 3.3 Hydraulic Characterization Comparison Between Brunswick Unit 2 ATRIUM 1OXM, ATRIUM-10, and GE14 Fuel Assemblies ........................................... 3-10 3.4 Brunswick Unit 2 Thermal-Hydraulic Design Conditions .............................................. 3-11 3.5 Brunswick Unit 2 First Transition Core Thermal-Hydraulic Results at R ated C onditions (100% P / 100% F) ............................................................................ 3-12 3.6 Brunswick Unit 2 First Transition Core Thermal-Hydraulic Results at O ff-Rated C onditions (60% P / 45% F) .......................................................................... 3-13 3:7 Brunswick Unit 2 Thermal-Hydraulic Results at Rated Conditions (100%P / 100%F) for Transition to ATRIUM 1OXM Fuel .............................................. 3-14 3.8 Brunswick Unit 2 Thermal-Hydraulic Results at Off-Rated Conditions (60%P / 45%F) for Transition to ATRIUM 1OXM Fuel .................................................. 3-15 Figures 3 .1 A xia l P owe r S ha pe s ..................................................................................................... 3-16 3.2 First Transition Core: Hydraulic Demand Curves 100%P/100%F ................................ 3-17 3.3 First Transition Core: Hydraulic Demand Curves 60%P/45%F .................................... 3-18 AREVA NP Inc.
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page iii Nomenclature AOO anticipated operational occurrence ASME American Society of Mechanical Engineers BRK2-19 Brunswick Unit 2 Cycle 19 BRK2-20 Brunswick Unit 2 Cycle 20 BWR boiling water reactor CHF critical heat flux CPR critical power ratio CRDA control rod drop accident LOCA loss-of-coolant accident LTP lower tie plate MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio NRC Nuclear Regulatory Commission, U.S.
PLFR part-length fuel rod RPF radial peaking factor UTP upper tie plate AREVA NP Inc.
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page 1-1 1.0 Introduction The results of Brunswick Unit 2 thermal-hydraulic analyses are presented to demonstrate that M
T AREVA NP* ATRIUM 1OXMt fuel is hydraulically compatible with the previously loaded ATRIUM-10 and GE14 fuel designs. This report also provides the hydraulic characterization of the ATRIUM 1OXM, and the coresident GE14 and ATRIUM-10 designs for Brunswick Unit 2.
The generic thermal-hydraulic design criteria applicable to the design have been reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC) in the topical report ANF-89-98(P)(A) Revision 1 and Supplement 1 (Reference 1). In addition, thermal-hydraulic criteria applicable to the design have also been reviewed and approved by the NRC in the topical report XN-NF-80-19(P)(A) Volume 4 Revision 1 (Reference 2).
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t ATRIUM is a trademark of AREVA NP.
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Design Report for ATRIUM TM 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 and ATRUM-10 fuel designs in the Brunswick Unit 2 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 and Tables 3.4 to 3.8.
The ATRIUM 1OXM fuel design is geometrically different from the coresident ATRIUM-10 and GE14 designs, but hydraulically the designs are compatible. [
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 1 OXM fuel design. Analyses at rated conditions show core bypass flow varying between [ ] of rated flow for transition core configurations ranging from the BRK2-19 core loading with GE14 and ATRIUM-10 fuel to a full ATRIUM 1OXM core, respectively.
Analyses demonstrate the thermal-hydraulic design and compatibility criteria discussed in Section 3.0 are satisfied for the Brunswick Unit 2 transition cores consisting of ATRIUM 1OXM, ATRIUM-10, and GE14 fuel for the expected core powerdistributions and core power/flow conditions encountered during operation.
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Design Report for ATRIUM TM 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. To the extent possible, these analyses are performed on a generic fuel design basis. 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. 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. 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 reload licensing report.
Fuel centerline temperature. Fuel design and operation shall be such that fuel centerline melting is not projected for normal operation and AOOs. This criterion evaluation is addressed in the mechanical design report.
Rod bow. 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. 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.
Stability. Reactors fueled with new fuel designs must be stable 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 reload licensing report.
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page 3-2 Loss-of-coolant accident (LOCA) analysis. LOCAs are analyzed in accordance with Appendix K modeling requirements using NRC-approved models. The criteria are defined in 10 CFR 50.46. LOCA analysis results are presented in the break spectrum and MAPLHGR reports.
Control rod drop accident (CRDA) analysis. The deposited enthalpy must be less than 280 cal/gm for fuel coolability. This criterion evaluation is addressed in the reload licensing report.
ASME overpressurization analysis. ASME pressure vessel code requirements must be satisfied. This criterion evaluation is addressed in the reload licensing report.
Seismic/LOCA liftoff. Under accident conditions, the assembly must remain engaged in the fuel support. This criterion evaluation is addressed in the mechanical design report.
A summary of the thermal-hydraulic design evaluations is given in Table 3.1.
3.1 Hydraulic Characterization Basic geometric parameters for the ATRIUM 1OXM, ATRIUM-10, and GE14 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, ULTRAFLOWTM* spacer, UTP and LTP losses for ATRIUM 1OXM and ATRIUM-10 are based on tests performed at AREVA's Portable Hydraulic Test Facility. [
] The local component (LTP, spacer, and UTP) loss coefficients for the GE14 fuel are based on flow test results.
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.
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1OXM Revision 0 Fuel Assemblies Page 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, ATRIUM-10, and GE14 fuel designs, has been evaluated. Detailed analyses were performed for the Brunswick Unit 2 Cycle 19 and full core ATRIUM 10XM configurations. Analyses for mixed cores with ATRIUM 1OXM, ATRIUM-10, and GE14 fuel were also performed to demonstrate that 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/100% flow and 60% power/45% flow.
Table 3.4 also defines the core loading for the transition core configurations. Input for other core configurations is similar in that core operating conditions remain the same and the same axial 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 AREVA NP Inc.
Controiled Documnent Brunswick Unit 2 Thermal-Hydraulic ANP-2936(NP)
Design Report for ATRIUM TM 1OXM Revision 0 Fuel Assemblies Page 3-4 ATRIUM 1OXM, ATRIUM-10, 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, ATRIUM-10, 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 the BRK2-19 core loading to a full core of ATRIUM 1OXM, the ATRIUM 1OXM design is considered hydraulically compatible with the coresident fuel designs 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.
CPR values for ATRIUM 1OXM are calculated with the ACE/ATRIUM 10XM critical power correlation (Reference 10) while the CPR values for the ATRIUM-10 and GE14 fuel are calculated with the SPCB critical power correlation (Reference 7). The NRC-approved methodology to demonstrate the acceptability of using the SPCB correlation for computing GE14 fuel CPR is presented in Reference 8.. Assembly design features are incorporated in the CPR calculation through the K-factor term in the ACE correlation and the F-eff term for the SPCB correlation. The K-factors and F-effs are based on the local power peaking for the nuclear design and on additive constants determined in accordance with approved procedures.
The local peaking factors are a function of assembly void fraction and exposure.
For the compatibility evaluation, steady-state analyses evaluated ATRIUM 1OXM, ATRIUM-10, and GE14 assemblies with radial peaking factors (RPFs) between [
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] Table 3.5 and Table 3.6 show CPR results of the ATRIUM 1OXM, ATRIUM-10, 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 designs.
3.4 Rod Bow The bases for rod bow are discussed in the mechanical design report. Rod bow magnitude is determined during the fuel-specific 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 the BRK2-19 core loading to a full ATRIUM 1OXM core (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 9). The study AREVA NP Inc.
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Design Report for ATRIUM 1OXM M
T Revision 0 Fuel Assemblies Page 3-6 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 reload licensing report.
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies 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 ATRIUM-10 and such that there is no significant impact on total core flow or flow [
distribution among assemblies.
3.3 Thermal margin Fuel design shall be ACE/ATRIUM 1OXM is applied to performance within the limits of the ATRIUM 1OXM fuel.
applicability of an approved CHF correlation.
< 0.1% of rods in boiling Verified on cycle-specific basis for transition. Chapter 15 analyses..
Fuel centerline No centerline melting. Plant- and fuel-specific analyses temperature are performed.
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, thermal margins.
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.
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page 3-8 Table 3.1 Design Evaluation of Thermal and Hydraulic Criteria for the ATRIUM-10 Fuel Assembly (Continued)
Report Section Description Criteria Results or Disposition Thermal and Hydraulic Criteria (Continued) 3.6 Stability New fuel designs are ATRIUM 1OXM channel and core stable in the approved decay ratios have been power and flow operating demonstrated to be equivalent to or region, and stability better than other approved AREVA performance will be fuel designs.
equivalent to (or better Core stability behavior is evaluated.
than) existing (approved) on a cycle-specific basis.
AREVA fuel designs.
LOCA analysis LOCA analyzed in Approved Appendix K LOCA accordance with model.
Appendix K modeling Plant- and fuel-specific analysis requirements. Criteria defuineedint1 Criteria 5 with cycle-specific verifications.
defined in 10 CFR 50.46.
CRDA analysis < 280 cal/grn for Cycle-specific analysis is coolability. performed.
ASME over- ASME pressure vessel Cycle-specific analysis is pressurization core requirements shall performed.
analysis be satisfied.
Seismic/LOCA Assembly remains Plant- and fuel-specific analyses liftoff engaged in fuel support. are performed.
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page 3-9 Table 3.2 Comparative Description of Brunswick Unit 2 ATRIUM 1OXM, ATRIUM-110, and GE14 Fuel Fuel Parameter ATRIUM 1OXM ATRIUM-10 GEl4 Number of fuel rods Full-length fuel rods 79 83 78 PLFRs 12 8 14 Fuel clad OD, in 0.4047 0.3957 0.404 Number of spacers 9 8 8 Active fuel length, ft Full-length fuel rods 12.500 12.454 12.500 PLFRs 6.25 7.5 7.0 Hydraulic resistance characteristics Table 3.3 Table 3.3 Table 3.3 Number of water rods 1 1 2 Water rod OD, in 1.378* 1.378* 0.980
- Square water channel outer width.
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T 1OXM Revision 0 Fuel Assemblies Page 3-10 Table 3.3 Hydraulic Characterization Comparison Between Brunswick Unit 2 ATRIUM 1OXM, ATRIUM-10, and GE14 Fuel Assemblies I
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T Revision 0 Fuel Assemblies Page 3-11 Table 3.4 Brunswick Unit 2 Thermal-Hydraulic Design Conditions Reactor Conditions 100%P / 100%F 60%P /45%F Core power level, MWt 2923.0 1753.8 Core exit pressure, psia 1058.3 988.2 Core inlet enthalpy, Btu/lbm 528.3 504.6 Total core coolant flow, Mlbm/hr 77.0 34.65 Axial power shape Middle-peaked Middle-peaked (Figure 3.1) (Figure 3.1)
Number of Assemblies Central Peripheral Region Region Current Core Loading (BRK2-19)
[ ]
First Transition Loading (BRK2-20)
[ ]
[ ]
Second Transition Core Loading
[ ]
[ ]
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page 3-12 Table 3.5 Brunswick Unit 2 First Transition Core Thermal-Hydraulic Results at Rated Conditions (100%P / 100%F)
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page 3-13 Table 3.6 Brunswick Unit 2 First Transition Core Thermal-Hydraulic Results at Off-Rated Conditions (60%P / 45%F)
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page 3-14 Table 3.7 Brunswick Unit 2 Thermal-Hydraulic Results at Rated Conditions (100%P I 100%F) for Transition to ATRIUM 1OXM Fuel I
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page 3-15 Table 3.8 Brunswick Unit 2 Thermal-Hydraulic Results at Off-Rated Conditions (60%P / 45%F) for Transition to ATRIUM 1OXM Fuel I
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page 3-16 I
Figure 3.1 Axial Power Shapes AREVA NP Inc.
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Design Report for ATRIUM 1OXM Revision 0 Fuel Assemblies Page 3-17
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I Figure 3.2 First Transition Core:
Hydraulic Demand Curves 100%P/100%F AREVA NP Inc.
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Figure 3.3 First Transition Core:
Hydraulic Demand Curves 60%P/45%F AREVA NP Inc.
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Design Report for ATRIUM TM 10XM 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 PressureDrop 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. EMF-2209(P)(A) Revision 3, SPCB CriticalPower Correlation,AREVA NP, September 2009.
- 8. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation'sCriticalPower Correlationsto Co-Resident Fuel, Siemens Power Corporation, August 2000.
- 9. EMF-CC-074(P)(A) Volume 1, STAIF - A Computer Programfor BWR Stability Analysis in the Frequency Domain; and Volume 2, STAIF - A Computer Program for BWR Stability Analysis in the FrequencyDomain - Code QualificationReport, Siemens Power Corporation, July 1994.
- 10. ANP-10298PA Revision 0, ACE/ATRIUM IOXM CriticalPower Correlation,AREVA NP, March 2010.
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