ML072750045
| ML072750045 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/24/2007 |
| From: | Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| ANP-2631, Rev 0 | |
| Download: ML072750045 (3) | |
Text
Enclosure Browns Ferry Nuclear Plant (BFN)
ANP-2631 Revision 0 Browns Ferry Unit 3 Cycle 14 Reload Analysis Report
ANP-2631 Revision 0 Browns Ferry Unit 3 Cycle 14 Reload Analysis May 2007
AREVA NP Inc.
ANP-2631 Revision 0 Browns Ferry Unit 3 Cycle 14 Reload Analysis sjp/paj
- 0 Customer Disclaimer Important Notice Regarding the Contents and Use of This Document Please Read Carefully AREVA NP Inc.'s warranties and representations concerning the subject matter of this document are those set forth in the agreement between AREVA NP Inc. and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither AREVA NP Inc. nor any person acting on its behalf:
- a. makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or
- b. assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this document.
The information contained herein is for the sole use of the Customer.
In order to avoid impairment of rights of AREVA NP Inc. in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until so authorized in writing by AREVA NP Inc. or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this document.
0 AREVA NP 0 ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page i Nature of Changes Item Page Description and Justification
- 1. All This is a new document.
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page ii Contents 1.0 Introduction .................................................................................................................... 1-1 2.0 Fuel Mechanical Design Analysis .................................................................................. 2-1 3.0 Therm al-Hydraulic Design Analysis ............................................................................... 3-1 3.2 Hydraulic Characterization ................................................................................. 3-1 3.2.1 Hydraulic Com patibility ......................................................................... 3-1 3.2.3 Fuel Centerline Tem perature ............................................................... 3-1 3.2.5 Bypass Flow ......................................................................................... 3-1 3.3 MCPR Fuel Cladding Integrity Safety Lim it (SLM CPR) ...................................... 3-1 3.3.1 Coolant Therm odynam ic Condition ...................................................... 3-1 3.4 Licensing Power and Exposure Shape .............................................................. 3-2 4.0 Nuclear Design Analysis ................................................................................................ 4-1 4.1 Fuel Bundle Nuclear Design Analysis ................................................................ 4-1 4.2 Core Nuclear Design Analysis ............................................................................ 4-3 4.2.1 Core Configuration ............................................................................... 4-3 4.2.2 Core Reactivity Characteristics ........................................................... 4-3 4.2.4 Core Hydrodynam ic Stability ................................................................ 4-5 5.0 Abnorm al Operational Transients .................................................................................. 5-1 5.1 Analysis of Plant Transients at Rated Power Conditions ................................... 5-1 5.1.1 NEOC Licensing Exposure ................................................................... 5-1 5.1.2 EO C Licensing Exposure ..................................................................... 5-2 5.1.3 FFTR/Coastdown Licensing Exposure ................................................. 5-2 5.2 Analysis for Reduced Flow Operation ................................................................ 5-3 5.3 Analysis for Reduced Power Operation .............................................................. 5-3 5.4 ASM E Overpressurization Analysis .................................................................... 5-3 5.5 Control Rod W ithdrawal Error ............................................................................. 5-3 5.6 Fuel Loading Error (Infrequent Event) ................................................................ 5-4 5.6.1 Mislocated Fuel Assem bly .................................................................... 5-4 5.6.2 Misoriented Fuel Bundle ....................................................................... 5-4 5.7 Determ ination of Therm al Margins ..................................................................... 5-5 6.0 Postulated Accidents ...................................................................................................... 6-1 6.1 Loss-of-Coolant Accident ................................................................................... 6-1 6.1.1 Break Location Spectrum ..................................................................... 6-1 6.1.2 Break Size Spectrum ............................................................................ 6-1 6.1.3 MAPLHG R Analyses ............................................................................ 6-1 6.2 Control Rod Drop Accident ................................................................................. 6-2 6.4 Fuel and Equipm ent Handling Accident ............................................................. 6-2 AREVA NP Inc.
0 AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page iii Contents (Continued) 7.0 Technical Specifications ................................................................................................. 7-1 7.1 Limiting Safety System Settings ......................................................................... 7-1 7.1.1 MCPR Fuel Cladding Integrity Safety Limit .......................................... 7-1 7.1.2 Steam Dome Pressure Safety Limit ..................................................... 7-1 7.2 Limiting Conditions for Operation ....................................................................... 7-1 7.2.1 Average Planar Linear Heat Generation Rate ...................................... 7-1 7.2.2 Minimum Critical Power Ratio .............................................................. 7-1 7.2.3 Linear Heat Generation Rate ............................................................... 7-2 8.0 Methodology References ............................................................................................... 8-1 9.0 Additional References .................................................................................................... 9-1 Tables 1.1 EOD and EOOS Operating Conditions ........................................................................... 1-2 3.1 Licensing Basis Core Average Axial Power Profile and Licensing Axial Exposure Ratio ............................................................................................................... 3-3 4.1 Core Composition .......................................................................................................... 4-7 4.2 OPRM Setpoint Versus Stability-Based MCPR Operating Limits .................................. 4-8 4.3 Neutronic Design Values ................................................................................................ 4-9 5.1 Flow-Dependent MCPR Limits for Maximum Flow of 102.5% of Rated Flow ................ 5-6 5.2 Flow-Dependent MCPR Limits for Maximum Flow of 107% of Rated Flow ................... 5-6 5.3 Flow-Dependent LHGRFACf Multipliers for Maximum Flow of 102.5% and 107% of Rated Flow ....................................................................................................... 5-6 5.4 MCPRn Limits for NSS Insertion Times .......................................................................... 5-7 5.5 MCPRp Limits for TSSS Insertion Times ...................................................................... 5-11 5.6 LHGRFACp Multipliers NSS/TSSS Insertion Times All Exposures ............................... 5-15 5.7 Control Rod Withdrawal Error MCPR versus RBM Setpoint Results (for Rated Power and 1.09 SLMCPR) ................................................................................ 5-16 5.8 RBM Setpoint Applicability ........................................................................................... 5-16 Figures 4.1 Lower Right Quarter Core Layout By Fuel Type ......................................................... 4-10 AREVA NP Inc.
0 AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page iv Nomenclature AOT abnormal operational transient ARO all rods out ASME American Society of Mechanical Engineers AST alternative source term BOC beginning of cycle BPWS banked position withdrawal sequence CGU commercial grade uranium CPR critical power ratio CRDA control rod drop accident CRWE control rod withdrawal error EFPD effective full-power days EOC end of cycle EOC-RPT-OOS end of cycle recirculation pump trip out-of-service EOD extended operating domain EOFPL end of full power life (100%P/100%F normal FW temperature)
EOOS equipment out-of-service FFTR final feedwater temperature reduction FHOOS feedwater heaters out-of-service FWCF feedwater controller failure ICF increased core flow LFWH loss of feedwater heating LHGR linear heat generation rate LHGRFACf flow-dependent linear heat generation rate factors LHGRFACp power-dependent linear heat generation rate factors LOCA loss-of-coolant accident LPRM local power range monitor LRNB load rejection no bypass MAPFACf flow-dependent maximum average planar heat generation rate factors MAPFACp power-dependent maximum average planar heat generation rate factors MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio MCPRf flow-dependent minimum critical power ratio MCPRp power-dependent minimum critical power ratio MELLLA maximum extended load line limit analysis MSIV main steam isolation valve MSRVOOS main steam relief valves out-of-service AREVA NP Inc.
AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page v Nomenclature (Continued)
NEOC near end of cycle NRC Nuclear Regulatory Commission, U.S.
NSS nominal scram speed OLMCPR operating limit minimum critical power ratio 00S out of service OPRM oscillation power range monitor PAPT protection against power transient PCT peak clad temperature PLUOOS power load unbalance out-of-service RBM rod block monitor RNW reduced notch worth RPT recirculation pump trip SER safety evaluation report SLC standby liquid control (boron)
SLCSDM standby liquid control shutdown margin (boron)
SLMCPR safety limit minimum critical power ratio SLO single-loop operation TBVOOS turbine bypass valves out-of-service TIP traversing in-core probe TIPOOS traversing in-core probe out-of-service TLO two-loop operation TSSS technical specification scram speed UFSAR updated final safety analysis report ACPR change in critical power ratio AREVA NP Inc.
AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 1-1 1.0 Introduction This report provides results of analyses performed by AREVA NP* as part of the reload analysis. This report is intended to be used in conjunction with the AREVA topical Report 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, which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a generic reference list. Section numbers in this report are the same as corresponding section numbers in XN-NF-80-19(P)(A) Volume 4 Revision 1. Methodology used in this report which supersedes XN-NF-80-19(P)(A) Volume 4 Revision 1 is referenced in Section 8.0. The application of the methodology used in the computer codes that were utilized in performing the analyses presented in this report were applied in accordance with the NRC technical limitations (safety evaluation report (SER) restrictions) as stated in the methodology.
The core consists of a total of 764 fuel assemblies, including 374 unirradiated ATRIUMM-10t assemblies and 390 irradiated ATRIUM-10 assemblies. The reference core configuration is described in Section 4.2.
The effects of channel bow are explicitly accounted for in the safety limit analysis. The Extended Operating Domain (EOD) and Equipment Out-Of-Service (EOOS) conditions presented in Table 1.1 are supported.
- AREVA NP Inc. is an AREVA and Siemens company.
t ATRIUM is a trademark of AREVA NP.
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AREVA NP 1 ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 1-2 Table 1.1 EOD and EOOS Operating Conditions Extended Operating Domain (EOD) Conditions Increased core flow (ICF)
Maximum extended load line limit analysis (MELLLA)
Combined FFTR/coastdown Equipment Out-of-Service (EOOS) Conditions*
Turbine bypass valves out-of-service (TBVOOS)
EOC recirculation pump trip out-of-service (EOC-RPT-OOS)
Feedwater heaters out-of-service (FHOOS)
Power load unbalance out-of-service (PLUOOS)
Combined EOC-RPT-OOS and TBVOOS Combined EOC-RPT-OOS and FHOOS Combined EOC-RPT-OOS and PLUOOS Combined TBVOOS and FHOOS Combined TBVOOS and PLUOOS Combined FHOOS and PLUOOS Combined EOC-RPT-OOS, TBVOOS, and FHOOS Combined EOC-RPT-OOS, TBVOOS, and PLUOOS Combined EOC-RPT-OOS, FHOOS, and PLUOOS Combined TBVOOS, FHOOS, and PLUOOS Combined EOC-RPT-OOS, TBVOOS, FHOOS, and PLUOOS Single-loop operation (SLO)
SLO may be combined with all of the other EOOS conditions. Base case and each EOOS condition is supported in combination with 1 MSRVOOS, up to 2 TIPOOS or the equivalent number of channels (per operating requirements defined in Reference 9.6 Section 3.2), and/or up to 50% of the LPRMs out-of-service.
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S AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 2-1 2.0 Fuel Mechanical Design Analysis Applicable AREVA Fuel Design Reports References 9.11, 9.12, and 9.16 To assure the power history for the ATRIUM-10 fuel is bounded by the assumed power history in the fuel mechanical design analyses, LHGR operating limits have been specified in Section 7.2.3. In addition, ATRIUM-10 LHGR limits for Abnormal Operational Transients (AOTs) have been specified in References 9.11, 9.12, and 9.16 (AOT is equivalent to anticipated operational occurrences used in References 9.11, 9.12, and 9.16). The exposure limits for the ATRIUM-10 bundles are specified in References 9.11, 9.12, and 9.16.
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 3-1 3.0 Thermal-Hydraulic Design Analysis 3.2 Hydraulic Characterization 3.2.1 Hydraulic Compatibility Hydraulic demand curves for the ATRIUM-1 0 fuel are provided in Reference 9.1, Figures 3.2, 3.3, and 3.4. All thermal-hydraulic compatibility criteria have been met.
3.2.3 Fuel Centerline Temperature Applicable Reports ATRIUM-10 References 9.11, 9.12, and 9.16, Figure 3.2 3.2.5 Byiass Flow Calculated Bypass Flow Fraction 14.4%
at 100%P/100%F (includes water channel flow) 3.3 MCPR Fuel CladdingIntegrity Safety Limit (SLMCPR)
Two-Loop Operation* 1.09 Reference 9.6 Single-Loop Operation* 1.11 3.3.1 Coolant Thermodynamic Condition Thermal Power (at SLMCPR)t 5655.11 MWt Feedwater Flow Rate (at SLMCPR) 23.49 Mlbm/hr Steam Dome Pressure (at rated conditions) 1050 psia Feedwater Temperature 394.80 F
- Includes the effects of channel bow, 2 TIPOOS or the equivalent number of TIP channels (per operating requirements defined in Reference 9.6 Section 3.2), a 2500 EFPH LPRM calibration interval, and up to 50% of the LPRMs out of service t Thermal power at SLMCPR is specific to SLMCPR methodology (Reference 8.2). The methodology increases ("pushes") the core power to reach the SLMCPR.
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,qW AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 3-2 3.4 Licensing Power and Exposure Shape The licensing axial power profile used by AREVA for the plant transient analyses bounds the projected end of full power (EOFP) axial power profile. The conservative licensing axial power profile as well as the corresponding axial exposure ratio are given in Table 3.1. Future projected cycle power profiles are considered to be in compliance when the EOFP normalized power generated in the core is greater than the licensing axial power profile at the given state conditions when the comparison is made over the bottom third of the core height.
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0 AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 3-3 Table 3.1 Licensing Basis Core Average Axial Power Profile and Licensing Axial Exposure Ratio State Conditions for Power Shape Evaluation Power, MWt 3952.0 Core pressure, psia 1064.7 Inlet subcooling, Btu/Ibm -27.75 Flow, MIb/hr 107.62 Control state ARO Core average exposure (EOFPL + 15 EFPD), MWd/MTU 31,606 Licensing Axial Power Profile (Normalized)
Node Power Top 25 0.222 24 0.684 23 0.893 22 1.021 21 1.116 20 1.183 19 1.232 18 1.270 17 1.297 16 1.373 15 1.386 14 1.393 13 1.387 12 1.370 11 1.340 10 1.293 9 1.218 8 1.109 7 0.973 6 0.830 5 0.705 4 0.614 3 0.543 2 0.432 Bottom 1 0.116 Licensing Axial Exposure Ratio (EOFP + 15 EFPD, ARO)
Average Bottom 8 ft / 12 ft = 1.0573 AREVA NP Inc.
AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 4-1 4.0 Nuclear Design Analysis 4.1 Fuel Bundle Nuclear Design Analysis The fuel cycle design used as the basis for the reload analysis is described in Reference 9.3.
The core composition is presented in Table 4.1 and Figure 4.1. The detailed fuel bundle design information for the fresh ATRIUM-10 fuel is provided in Reference 9.2. The following summary provides the appropriate cross-reference.
Assembly Average Enrichment (ATRIUM-10 fuel)
A10-4218B-15GV80-FCC (FT4)* 4.22 wt%
A10-4218B-13GV80-FCC (FT5) 4.22 wt%
A10-3757B-10GV80-FCC (FT6) 3.76 wt%
Radial Enrichment Distribution (enriched lattices only)
Al OB-4545L-1 5G80-FCC Reference 9.2, Figure D.2 AlOB-4557L-13G80-FCC Reference 9.2, Figure D.3 AlOT-4386L-13G80-FCC Reference 9.2, Figure D.4 AlOT-4386L-12G50-FCC Reference 9.2, Figure D.5 AlOB-4543L-13G80-FCC Reference 9.2, Figure D.8 Al OT-4399L-l 1G80-FCC Reference 9.2, Figure D.9 A1OT-4399L-11G50-FCC Reference 9.2, Figure D.10 AlOB-3997L-l0G80-FCC Reference 9.2, Figure D.13 Al0T-3997L-8G80-FCC Reference 9.2, Figure D.14 AlOT-3997L-8G50-FCC Reference 9.2, Figure D.15 Axial Enrichment Distribution Reference 9.2, Figures 2.1-2.3 Burnable Absorber Distribution Reference 9.2, Figures 2.4-2.6 Non-Fueled Rods Reference 9.2, Figures 2.4-2.6 Neutronic Design Parameters Table 4.3
- See Figure 4.1 for fuel type definitions.
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 4-2 Fuel Storage Spent Fuel Storage Pool Reference 9.4 The BFE3-14 reload batch fuel design meets the criticality safety limitations defined in Table 2.1 of Reference 9.4 and therefore can be safely stored in the pool.
New Fuel Storage Vault Reference 9.14 The BFE3-14 reload batch can be safely stored in the new fuel storage vault per the criticality safety limits defined in Table 2.1 of Reference 9.14.
Shipping Container References 9.19 and 9.20 The BFE3-14 reload assemblies conform to the nuclear criticality requirements established for the RAJ-11 shipping container in Reference 9.19. Satisfying the Reference 9.19 requirements ensures that the BFE3-14 fuel design may be stacked according to the constraints of the RAJ-11 shipping container stacking analysis provided in Reference 9.20.
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AREVA NP W ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 4-3 4.2 Core Nuclear Design Analysis 4.2.1 Core Configuration Figure 4.1 Core Exposure at EOC 13, MWd/MTU* 34,100 (nominal value)
Core Exposure at EOC 13, MWd/MTU 33,565 (short window)
Core Exposure at EOC 13, MWd/MTU 34,634 (long window)
Core Exposure at BOC 14, MWd/MTU 13,034 (from nominal EOC 13)
Core Exposure at NEOC,t MWd/MTU (from nominal EOC 13) 27,834 Core Exposure at EOC (EOFPL + 15 EFPD),ý MWd/MTU (from nominal EOC 13) 31,606 Maximum Core Exposure,§ MWd/MTU 32,637 4.2.2 Core Reactivity Characteristics** tt BOC 14 cold k-eff, all rods out 1.1279 BOC 14 cold k-eff, all rods in 0.9654 BOC 14 cold k-eff, strongest rod out 0.9896 BOC 14 cold shutdown margin 1.04% Ak/k Reactivity defect/R-value 0.01% Ak/k (minimum CSDM at 18,833 MWd/MTU cycle exposure)
- The thermal limits provided in this report are applicable for an EOC 13 exposure between the long and short windows.
- NEOC analyses and limits are applicable up to this core exposure.
EOC analyses and limits are applicable up to this core exposure.
FFTR/coastdown analyses and limits are applicable up to this core exposure.
- k-eff data are bias corrected. Bias corrected k=1+[k(MCB2)-k(target)].
t Evaluated based on short window.
AREVA NP Inc.
0 AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 4-4 Standby liquid control (SLC)* system reactivity, 816 ppm at 3660 F (equivalent to 720 ppm at 680F)t
- Maximum k-eff 0.9798
- Minimum SLCSDM 2.02% Ak/k
- A minimum SLCSDM of 0.88% Ak/k is required to protect manufacturing and calculational uncertainties when analyzed at temperature of RHR initialization.
t TVA Browns Ferry SLC licensing basis documents indicate a minimum of 720 ppm boron at a temperature of 70 0 F. The AREVA cold analysis basis of 68°F represents a negligible difference and the results are adequate to protect the 70°F licensing basis for the plant.
AREVA NP Inc.
AREVA NP 0 ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 4-5 4.2.4 Core Hydrodynamic Stability Browns Ferry has implemented BWROG Long Term Stability Solution Option III (Oscillation Power Range Monitor-OPRM). Reload validation has been performed in accordance with Reference 9.9. The stability based Operating Limit MCPR (OLMCPR) is provided for two conditions as a function of OPRM amplitude setpoint in Table 4.2. The two conditions evaluated are for a postulated oscillation at 45% core flow steady state operation (SS) and following a two recirculation pump trip (2PT) from the limiting full power operation state point. Current power and flow dependent limits provide adequate protection against violation of the Safety Limit MCPR for postulated reactor instability as long as the operating limit is greater than or equal to the specified value for the selected OPRM setpoint.
Evaluations by General Electric have shown that the generic DIVOM curves specified in NEDO-32465-A, may not be conservative for current plant operating conditions for plants which have implemented Stability Option Ill. Specifically, a non-conservative deficiency has been identified for high peak bundle power-to-flow ratios in the generic regional mode DIVOM curve. The deficiency results in a non-conservative slope of the associated DIVOM curve so that the Option Ill trip setpoint is too high. GE issued a Part 21 Notification in GE 10 CFR Part 21 Notification, Stability Reload Licensing Calculations Using Generic DIVOM Curve, MFN 01-046, August 31, 2001.
To address this issue related to the generic DIVOM slope, AREVA has performed calculations for the relative change in AMCPR as a function of the calculated hot channel oscillation magnitude (HCOM). These calculations have been performed with the RAMONA5-FA code.
This code is a coupled neutronic-thermal hydraulic three-dimensional transient model for the purpose of determining relationship between the relative change in AMCPR and the HCOM on a plant specific basis. This model has been developed consistent with the recommendations of the BWROG in OG04-0153-260, Plant-Specific Regional Mode DIVOM Procedure Guideline, June 15, 2004. The generation of the plant-specific DIVOM data with this model is consistent with the BWROG resolution of the above Part 21 notification as provided in BWROG-03047, Resolution of Reportable Condition for Stability Reload Licensing Calculations Using Generic Regional Mode DIVOM Curve, September 30, 2003.
AREVA NP Inc.
AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 4-6 The stability-based OLMCPRs were calculated using the most limiting calculated change in relative AMCPR for a given oscillation magnitude. The reload validation calculation demonstrated that reactor stability does not produce the limiting OLMCPR as long as the selected OPRM setpoint produces values for OLMCPR(SS) and OLMCPR(2PT) that are less than the corresponding acceptance criteria. The setpoints provided in Table 4.2 support the EOOS operating conditions provided in Table 1.1.
AREVA NP Inc.
0 AREVA NP 0 ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 4-7 Table 4.1 Core Composition Cycle Number of Fuel Description Loaded Assemblies ATRIUM-10 A10-3813B-13GV80 12 22 ATRIUM-10 A10-4077B-15GV80 12 30 ATRIUM-10 A10-4088B-13GV80 12 43 ATRIUM-10 A10-4171B-14GV80-FCB 13 63 ATRIUM-10 Al 0-4163B-1 6GV80-FCB 13 168 ATRIUM-10 Al 0-4181 B-1 3GV80-FCB 13 64 ATRIUM-10 A10-4218B-15GV80-FCC 14 218 ATRIUM-10 A10-4218B-13GV80-FCC 14 92 ATRIUM-10 A10-3757B-10GV80-FCC 14 64 AREVA NP Inc.
AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 4-8 Table 4.2 OPRM Setpoint Versus Stability-Based MCPR Operating Limits BOC to FFTR / Coastdown OPRM Setpoint OLMCPR OLMCPR A(SP) (SS) (2PT) 1.05 1.23 1.15 1.06 1.24 1.16 1.07 1.25 1.17 1.08 1.26 1.18 1.09 1.26 1.18 1.10 1.28 1.20 1.11 1.30 1.22 1.12 1.32 1.24 1.13 1.34 1.26 1.14 1.36 1.28 1.15 1.39 1.30 Rated Power Acceptance Off-rated OLMCPR as OLMCPR Criteria described in
@45% Flow Section 5 AREVA NP Inc.
AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 4-9 Table 4.3 Neutronic Design Values Number of fuel assemblies 764 Rated thermal power,* MWt 3952 Rated core flow,* Mlbm/hr 102.5 Fuel channel dimensions Corner thickness, inch 0.100 Reduced thickness, inch 0.075 Fuel assembly pitch, inch 6.0 Wide water gap thickness, inch 0.630 Narrow water gap thickness, inch 0.414 Control Bladest Total span, inch 9.810 Total support span, inch 1.580 Total thickness, inch 0.312 Total face-to-face internal dimension, inch 0.200 B4C rod absorber Number of rods 21 Rod diameter ID/OD, inch 0.138 / 0.188 Theoretical density of B4C, % 70
- Statepoint parameters for individual solutions are based on consistent heat balance calculations for the core power and flow prescribed for the condition being modeled.
t The control rod data represent the Duralife-100D/BWR-4 blade type.
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 4-10 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 30 4 1 5 2 4 2 4 2 4 2 4 1 5 5 32 0.0 22.2 0.0 23.2 0.0 23.2 0.0 22.7 0.0 22.7 0.0 23.4 0.0 0.0 40.1 28 1 5 2 4 1 4 2 5 1 4 2 6 4 3 32 23.3 0.0 24.0 0.0 22.3 0.0 23.0 0.0 22.3 0.0 21.9 0.0 0.0 18.4 34.9 26 5 2 4 2 4 1 4 2 4 2 4 3 5 6 32 0.0 22.2 0.0 23.2 0.0 23.5 0.0 21.2 0.0 22.3 0.0 17.7 0.0 0.0 40.1 24 2 4 2 5 2 4 2 4 2 4 2 6 4 5 33 23.4 0.0 22.3 0.0 22.2 0.0 23.1 0.0 22.9 0.0 23.6 0.0 0.0 0.0 39.7 22 4 1 4 2 4 1 4 2 4 2 5 3 5 1 33 0.0 20.9 0.0 23.1 0.0 23.2 0.0 23.6 0.0 22.9 0.0 17.8 0.0 21.1 37.5 20 2 4 1 4 1 4 1 4 2 4 2 6 4 6 33 23.3 0.0 20.8 0.0 23.3 0.0 21.1 0.0 23.2 0.0 24.5 0.0 0.0 0.0 38.3 18 4 2 4 2 4 1 4 2 4 2 4 3 5 3 33 0.0 22.4 0.0 23.1 0.0 20.9 0.0 20.2 0.0 23.2 0.0 17.9 0.0 15.8 40.2 16 2 5 2 4 2 4 2 4 2 6 4 6 3 31 23.1 0.0 22.5 0.0 23.5 0.0 22.9 0.0 24.2 0.0 0.0 0.0 17.8 33.5 14 4 1 4 2 4 2 4 2 4 3 5 6 33 0.0 21.1 0.0 20.4 0.0 22.1 0.0 23.5 0.0 18.0 0.0 0.0 38.7 12 2 4 2 4 2 4 2 6 3 5 3 32 32 22.9 0.0 20.4 0.0 23.3 0.0 22.8 0.0 18.6 0.0 17.8 38.2 34.0 10 4 2 4 2 5 2 4 4 5 3 32 0.0 22.1 0.0 22.9 0.0 23.6 0.0 0.0 0.0 17.6 39.7 8 1 6 3 6 3 6 3 6 6 32 23.0 0.0 15.9 0.0 19.4 0.0 13.1 0.0 0.0 39.1 6 5 4 5 4 5 4 5 3 33 33 0.0 0.0 0.0 0.0 0.0 0.0 0.0 15.9 39.3 38.9 4 5 3 6 5 1 6 3 33 Nuclear Fuel Type 0.0 15.8 0.0 0.0 22.5 0.0 16.8 38.1 BOC Exposure (GWd/MTU) 2 31 31 32 33 33 32 32 37.8 40.4 40.2 40.1 39.7 40.3 34.8 No. Per Fuel Type Description Cycle Loaded Quarter core 31 A10-3813B-13GV80 12 3 32 A10-4077B-15GV80 12 10 33 A10-4088B-13GVS0 12 10 1 A1O-417IB-14GV80-FCB 13 16 2 AIO-4163B-I6GV8O-FCB 13 42 3 A10-4181B-13GV80-FCB 13 16 4 A1O-4218B-15GV80-FCC 14 55 5 A10-4218B-I3GV80-FCC 14 23 C AI0-3757B-10GV80-FCC 14 16 Figure 4.1 Lower Right Quarter Core Layout By Fuel Type AREVA NP Inc.
AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-1 5.0 Abnormal Operational Transients Applicable Disposition of Events Reference 9.5 5.1 Analysis of Plant Transients at Rated Power Conditions Reference 9.6 Limiting Transients: Load Rejection No Bypass (LRNB)
Turbine Trip No Bypass (TTNB)
Feedwater Controller Failure (FWCF)
Loss of Feedwater Heating (LFWH) t Control Rod Withdrawal Error (CRWE), see Section 5.5 5.1.1 NEOC Licensing Exoosure Peak Peak Neutron Heat Scram Flux Flux Transient Speed (% Rated) (% Rated) ACPR LRNB* TSSS 295 116 .30 TTNB* TSSS 298 116 .30 FWCF* TSSS 291 121 .33 LRNB* NSS 273 114 .28 TTNB* NSS 275 114 .28 FWCF* NSS 280 119 .31 LFWHt -- - .09
- The results presented are based on base case operation at 100%P/105%F and are the most limiting considering earlier exposures.
The inadvertent HPCI pump startup event (including asymmetric injection effects) has been analyzed generically for Browns Ferry and has been determined to be nonlimiting (Reference 9.5). The EPU inadvertent HPCI pump startup analysis demonstrated that the event did not reach the level 8 trip setpoint (with sufficient margin); therefore, the event does not result in a turbine trip and the resulting pressurization transient.
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-2 5.1.2 EOC Licensing Exposure Peak Peak Neutron Heat Scram Flux Flux Transient Speed (% Rated) (% Rated) ACPR LRNB* TSSS 339 121 .31 TTNB* TSSS 343 122 .31 FWCF* TSSS 343 126 .33 LRNB* NSS 324 120 .30 TTNB* NSS 328 121 .30 FWCF* NSS 334 125 .32 LFWHt -- -- .09 5.1.3 FFTR/Coastdown Licensing Exposure Peak Peak Neutron Heat Scram Flux Flux Transient Speed (% Rated) (% Rated) ACPR LRNB* TSSS 343 122 .31 TTNB* TSSS 348 122 .31 FWCF* TSSS 343 128 .34 LRNB* NSS 327 121 .30 TTNB* NSS 331 122 .30 FWCF* NSS 338 128 .33 LFWHt -- -- .09
- The results presented are based on base case operation at 100%P/1 05%F and are the most limiting considering earlier exposures.
t The inadvertent HPCI pump startup event (including asymmetric injection effects) has been analyzed generically for Browns Ferry and has been determined to be nonlimiting (Reference 9.5). The EPU inadvertent HPCI pump startup analysis demonstrated that the event did not reach the level 8 trip setpoint (with sufficient margin); therefore, the event does not result in a turbine trip and the resulting pressurization transient.
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S AREVA NP G ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-3 5.2 Analysis for Reduced Flow Operation Reference 9.6 Limiting Transient: Slow Flow Excursion MCPRf Tables 5.1 and 5.2 Reference 9.6 Figures 2.1 and 2.2 LHGRFACf Table 5.3 Reference 9.6 Figure 2.3 MCPRf and LHGRFACf results are applicable at all cycle exposures and in all EOD and EOOS scenarios presented in Table 1.1. Since the Cycle 14 core is composed of only ATRIUM-10 fuel, MAPFACf multipliers are not required.
5.3 Analysis for Reduced Power Operation Reference 9.6 Limiting Transients: Load Rejection No Bypass (LRNB)
Turbine Trip No Bypass (TTNB)
Feedwater Controller Failure (FWCF)
MCPRp Base Case and EOOS Operation Tables 5.4 and 5.5 Reference 9.6 Sections 3.0 and 4.0 LHGRFACp All Conditions Table 5.6 Reference 9.6 Sections 3.0 and 4.0 Since the Cycle 14 core is composed of only ATRIUM-10 fuel, MAPFACp multipliers are not required.
5.4 ASME OverpressurizationAnalysis Reference 9.6 Limiting Event MSIV Closure Worst Single Failure Valve Position Scram Maximum Vessel Pressure (Lower Plenum) 1346 psig Maximum Steam Dome Pressure 1318 psig 5.5 Control Rod Withdrawal Error The CRWE event was analyzed assuming no xenon and credible instrumentation out-of-service in the rod block monitor (RBM) system. The analysis further assumes that the plant could be operating in either an A or B sequence control rod pattern.
The rated power CRWE results are shown in Table 5.7. For the analytical RBM high power setpoint values of 107% to 117% and all intermediate and lower power setpoint values, the MCPRp values bound the CRWE MCPR values. The MCPR values are based on an SLMCPR of 1.09. For other values of SLMCPR the CRWE MCPR can be adjusted by the difference in AREVA NP Inc.
AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-4 the SLMCPR and 1.09. AREVA analyses show that the filtered RBM setpoint reductions given in Reference 9.15 are supported.
The ATRIUM-10 fuel design meets the 1% plastic strain and centerline melt criteria by not exceeding the protection against power transient (PAPT) LHGR limit during the event (References 9.11, 9.12, and 9.16).
The recommended operability requirements based on the generic unblocked CRWE results are shown in Table 5.8 based upon the SLMCPR values of Section 3.3. For other values of SLMCPR, the MCPR in Table 5.8 can be adjusted by the ratio of the SLMCPR values. For Cycle 14, the CRWE results at all power levels are bounded by the MCPRp values given in Tables 5.4 - 5.5.
5.6 Fuel Loading Error(Infrequent Event)
As described in the AREVA topical report XN-NF-80-19(P)(A) Volume 4 Revision 1, the Fuel Loading error is characterized as an Infrequent Event and the acceptance criteria is that the offsite dose consequences due to the event shall not exceed a small fraction of the 10 CFR 50.67 limits.
5.6.1 Mislocated Fuel Assembly AREVA has performed a bounding fuel mislocation error analysis and has demonstrated continued applicability of the bounding results. This analysis evaluated the impact of a mislocated assembly against potential fuel rod failure mechanisms due to increased LHGR and reduced CPR. Based on these analyses, the offsite dose criteria (a small fraction of 10 CFR 50.67) is conservatively satisfied. Since no rod LHGR would exceed the transient LHGR limit, and since less than 0.1% of the fuel rods are expected to experience boiling transition which could result in a dryout induced failure, a dose consequence evaluation is not necessary.
5.6.2 Misoriented Fuel Bundle AREVA has performed a bounding fuel assembly misorientation analysis. The analysis was performed assuming that the limiting assembly was loaded in the worst orientation (rotated 1800) while producing sufficient power to be on the MCPR limit if it had been oriented correctly.
The analyses demonstrate that the small fraction of 10 CFR 50.67 offsite dose criteria is conservatively satisfied. A dose consequence evaluation is not necessary since less than 0.1%
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AREVA NP s ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-5 of the fuel rods are expected to experience boiling transition and the change in LHGR for the misoriented assembly remains below the transient LHGR limit.
5.7 Determinationof Thermal Margins The results of the analyses presented in Sections 5.1-5.3 and 5.5 are used for the determination of the MCPR and LHGR operating limits. Section 5.1 provides the results of analyses at rated conditions. Section 5.2 provides for the determination of the MCPRf and LHGRf limits at reduced flow (MCPRf, Tables 5.1-5.2, LHGRFACf, Table 5.3). Section 5.3 provides for the determination of the MCPRp limits and LHGRFACp at conditions of reduced power (Tables 5.4-5.6). Exposure dependent limits are presented for base case operation and the EOOS conditions presented in Table 1.1. MCPRp limits for single-loop operation (SLO) will be 0.02 higher than those for two-loop because the SLO SLMCPR is 0.02 higher.
TLO MCPRf limits and LHGRFACf multipliers are applicable for SLO without any adjustment.
The flow-dependent limits are based on a slow flow excursion of two recirculation loops for TLO, which is conservative relative to a single recirculation loop excursion that could occur in SLO.
For SLO operation, the MAPLHGR multiplier listed in Section 7.2.1 is applied.
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-6 Table 5.1 Flow-Dependent MCPR Limits for Maximum Flow of 102.5% of Rated Flow Core Flow MCPRf
(% of rated) ATRIUM-10 30 1.46 72 1.28 102.5 1.28 Table 5.2 Flow-Dependent MCPR Limits for Maximum Flow of 107% of Rated Flow Core Flow MCPRf
(% of rated) ATRIUM-10 30 1.49 78 1.28 107 1.28 Table 5.3 Flow-Dependent LHGRFACf Multipliers for Maximum Flow of 102.5% and 107% of Rated Flow Maximum Core Flow of Maximum Core Flow of 102.5% Rated 107% Rated Core Flow Core Flow
(% of rated) LHGRFACf (% of rated) LHGRFACf 30 0.94 30 0.92 42.8 1.00 47.1 1.00 102.5 1.00 107 1.00 AREVA NP Inc.
W AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-7 Table 5.4 MCPRp Limits for NSS Insertion Times*
BOC BOC BOC Operating Power to to to End of Condition (% of rated) NEOC EOC COAST 100.0 1.40 1.41 1.42 60.0 1.60 1.60 1.64 55.0 1.62 1.62 1.68 50.0 1.67 1.67 1.73 50.0 1.82 1.82 1.82 Base case 40.0 1.90 1.90 1.90 26.0 2.21 2.21 2.31 26.0 at > 50%F 2.63 2.63 2.72 23.0 at > 50%F 2.78 2.78 2.90 26.0 at - 50%F 2.49 2.49 2.58 23.0 at < 50%F 2.62 2.62 2.72 100.0 1.43 1.44 1.45 60.0 1.63 1.63 1.65 55.0 1.65 1.65 1.70 50.0 1.69 1.69 1.75 50.0 1.82 1.82 1.82 TBVOOS 40.0 1.90 1.90 1.90 26.0 2.23 2.23 2.33 26.0 at > 50%F 3.12 3.12 3.20 23.0 at > 50%F 3.39 3.39 3.48 26.0 at 5 50%F 2.65 2.65 2.72 23.0 at < 50%F 2.88 2.88 2.96 100.0 1.40 1.41 1.42 60.0 1.60 1.60 1.64 55.0 1.62 1.62 1.68 50.0 1.67 1.67 1.73 50.0 1.82 1.82 1.82 EOC-RPT-OOS 40.0 1.90 1.90 1.90 26.0 2.21 2.21 2.31 26.0 at > 50%F 2.63 2.63 2.72 23.0 at > 50%F 2.78 2.78 2.90 26.0 at - 50%F 2.49 2.49 2.58 23.0 at - 50%F 2.62 2.62 2.72 100.0 1.42 1.42 --
60.0 1.64 1.64 ---
55.0 1.68 1.68 ---
50.0 1.73 1.73 ---
50.0 1.82 1.82 FHOOS 40.0 1.90 1.90 26.0 2.31 2.31 ---
26.0 at > 50%F 2.72 2.72 --
23.0 at > 50%F 2.90 2.90 26.0 at s 50%F 2.58 2.58 23.0 at < 50%F 2.72 2.72 ---
Limits support operation with any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher.
FHOOS limits are not provided for BOC to End of COAST since the feewater temperature for FHOOS was assumed to be the same as FFTR. The thermal limit at BOC to End of COAST was developed to bound the corresponding earlier exposure FHOOS limit.
A step change in PLUOOS limits at 50% power is not supported since at 50% and below the LRNB with or without PLUOOS is the same event.
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W S AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-8 Table 5.4 MCPRp Limits for NSS Insertion Times (Continued)
BOC BOC BOC Operating Power to to to End of Condition (% of rated) NEOC EOC COAST 100.0 1.40 1.41 1.42 60.0 1.60 1.60 1.64 55.0 1.79 1.79 1.79 50.0 --- --- --
50.0 1.82 1.82 1.82 PLUOOS 40.0 1.90 1.90 1.90 26.0 2.21 2.21 2.31 26.0 at > 50%F 2.63 2.63 2.72 23.0 at > 50%F 2.78 2.78 2.90 26.0 at < 50%F 2.49 2.49 2.58 23.0 at - 50%F 2.62 2.62 2.72 100.0 1.43 1.44 1.45 60.0 1.63 1.63 1.65 55.0 1.65 1.65 1.70 50.0 1.69 1.69 1.75 50.0 1.82 1.82 1.82 ECPOO S 40.0 1.90 1.90 1.90 26.0 2.23 2.23 2.33 26.0 at > 50%F 3.12 3.12 3.20 23.0 at > 50%F 3.39 3.39 3.48 26.0 at < 50%F 2.65 2.65 2.72 23.0 at < 50%F 2.88 2.88 2.96 100.0 1.42 1.42 ---
60.0 1.64 1.64 ---
55.0 1.68 1.68 ---
50.0 1.73 1.73 ---
50.0 1.82 1.82 ---
FHOOS 40.0 1.90 1.90 --
26.0 2.31 2.31 --
26.0 at > 50%F 2.72 2.72 ---
23.0 at > 50%F 2.90 2.90 --
26.0 at < 50%F 2.58 2.58 ---
23.0 at < 50%F 2.72 2.72 ---
100.0 1.40 1.41 1.42 60.0 1.60 1.60 1.64 55.0 1.79 1.79 1.79 50.0 --- --- --
50.0 1.82 1.82 1.82 ECPOO S 40.0 1.90 1.90 1.90 26.0 2.21 2.21 2.31 26.0 at > 50%F 2.63 2.63 2.72 23.0 at > 50%F 2.78 2.78 2.90 26.0 at < 50%F 2.49 2.49 2.58 23.0 at - 50%F 2.62 2.62 2.72 AREVA NP Inc.
AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-9 Table 5.4 MCPRp Limits for NSS Insertion Times (Continued)
BOC BOC BOC Operating Power to to to End of Condition (% of rated) NEOC EOC COAST 100.0 1.45 1.45 60.0 1.65 1.65 55.0 1.70 1.70 50.0 1.75 1.75 TBVOOS 50.0 1.82 1.82 FHOOS 40.0 1.90 1.90 26.0 2.33 2.33 26.0 at > 50%F 3.20 3.20 23.0 at > 50%F 3.48 3.48 26.0 at < 50%F 2.72 2.72 23.0 at < 50%F 2.96 2.96 100.0 1.43 1.44 1.45 60.0 1.63 1.63 1.65 55.0 1.79 1.79 1.79 50 .0 ---......
50.0 1.82 1.82 1.82 TBVOOS 40.0 1.90 1.90 1.90 26.0 2.23 2.23 2.33 26.0 at > 50%F 3.12 3.12 3.20 23.0 at > 50%F 3.39 3.39 3.48 26.0 at 5 50%F 2.65 2.65 2.72 23.0 at 5 50%F 2.88 2.88 2.96 100.0 1.42 1.42 ---
60.0 1.64 1.64 ---
55.0 1.79 1.79 ---
5 0 .0 ---......
50.0 1.82 1.82 ---
EHOOS 40.0 1.90 1.90 26.0 2.31 2.31 ---
26.0 at > 50%F 2.72 2.72 ---
23.0 at > 50%F 2.90 2.90 ---
26.0 at < 50%F 2.58 2.58 ---
23.0 at < 50%F 2.72 2.72 ---
100.0 1.45 1.45 60.0 1.65 1.65 55.0 1.70 1.70 50.0 1.75 1.75 EOC-RPT-OOS 50.0 1.82 1.82 TBVOOS 40.0 1.90 1.90 FHOOS 26.0 2.33 2.33 26.0 at > 50%F 3.20 3.20 23.0 at > 50%F 3.48 3.48 26.0 at - 50%F 2.72 2.72 23.0 at < 50%F 2.96 2.96 AREVA NP Inc.
W W AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-10 Table 5.4 MCPRp Limits for NSS Insertion Times (Continued)
BOC BOC BOC Operating Power to to to End of Condition (% of rated) NEOC EOC COAST 100.0 1.43 1.44 1.45 60.0 1.63 1.63 1.65 55.0 1.79 1.79 1.79 5 0 .0 . ... ... ..
EOC-RPT-OOS 50.0 1.82 1.82 1.82 TBVOOS 40.0 1.90 1.90 1.90 PLUOOS 26.0 2.23 2.23 2.33 26.0 at > 50%F 3.12 3.12 3.20 23.0 at > 50%F 3.39 3.39 3.48 26.0 at < 50%F 2.65 2.65 2.72 23.0 at < 50%F 2.88 2.88 2.96 100.0 1.42 1.42 ---
60.0 1.64 1.64 ---
55.0 1.79 1.79 ---
50 .0 ---......
EOC-RPT-OOS 50.0 1.82 1.82 --
FHOOS 40.0 1.90 1.90 --
PLUOOS 26.0 2.31 2.31 --
26.0 at > 50%F 2.72 2.72 ---
23.0 at > 50%F 2.90 2.90 ---
26.0 at < 50%F 2.58 2.58 ---
23.0 at < 50%F 2.72 2.72 ---
100.0 1.45 1.45 ---
60.0 1.65 1.65 ---
55.0 1.79 1.79 ---
50 .0 ---......
TBVOOS 50.0 1.82 1.82 --
FHOOS 40.0 1.90 1.90 ---
PLUOOS 26.0 2.33 2.33 --
26.0 at > 50%F 3.20 3.20 ---
23.0 at > 50%F 3.48 3.48 --
26.0 at < 50%F 2.72 2.72 --
23.0 at 5 50%F 2.96 2.96 ---
100.0 1.45 1.45 60.0 1.65 1.65 ---
55.0 1.79 1.79 ---
50.0 -- ---
EOC-RPT-OOS 50.0 1.82 1.82 --
TBOOS 40.0 1.90 1.90 --
FHOOS 26.0 2.33 2.33 ---
PLUOOS 26.0 at > 50%F 3.20 3.20 ---
23.0 at > 50%F 3.48 3.48 ---
26.0 at < 50%F 2.72 2.72 ---
23.0 at < 50%F 2.96 2.96 --
AREVA NP Inc.
W W AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-11 Table 5.5 MCPRp Limits for TSSS Insertion Times*
BOC BOC BOC Operating Power to to to End of Condition (% of rated) NEOC EOC COAST 100.0 1.42 1.42 1.44 60.0 1.62 1.62 1.66 55.0 1.65 1.65 1.70 50.0 1.69 1.69 1.75 50.0 1.83 1.83 1.83 Base case 40.0 1.91 1.91 1.91 26.0 2.23 2.23 2.33 26.0 at > 50%F 2.63 2.63 2.72 23.0 at > 50%F 2.78 2.78 2.90 26.0 at < 50%F 2.49 2.49 2.58 23.0 at < 50%F 2.62 2.62 2.72 100.0 1.45 1.46 1.47 60.0 1.65 1.65 1.68 55.0 1.67 1.67 1.72 50.0 1.71 1.71 1.77 50.0 1.83 1.83 1.83 TBVOOS 40.0 1.91 1.91 1.91 26.0 2.25 2.25 2.35 26.0 at > 50%F 3.12 3.12 3.20 23.0 at > 50%F 3.39 3.39 3.48 26.0 at < 50%F 2.65 2.65 2.72 23.0 at - 50%F 2.88 2.88 2.96 100.0 1.42 1.42 1.44 60.0 1.62 1.62 1.66 55.0 1.65 1.65 1.70 50.0 1.69 1.69 1.75 50.0 1.83 1.83 1.83 EOC-RPT-OOS 40.0 1.91 1.91 1.91 26.0 2.23 2.23 2.33 26.0 at > 50%F 2.63 2.63 2.72 23.0 at > 50%F 2.78 2.78 2.90 26.0 at < 50%F 2.49 2.49 2.58 23.0 at < 50%F 2.62 2.62 2.72 100.0 1.44 1.44 ---
60.0 1.66 1.66 ---
55.0 1.70 1.70 ---
50.0 1.75 1.75 ---
50.0 1.83 1.83 ---
FHOOS 40.0 1.91 1.91 ---
26.0 2.33 2.33 ---
26.0 at > 50%F 2.72 2.72 --
23.0 at > 50%F 2.90 2.90 26.0 at - 50%F 2.58 2.58 ---
23.0 at < 50%F 2.72 2.72 Limits support operation with any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher.
FHOOS limits are not provided for BOC to End of COAST since the feewater temperature for FHOOS was assumed to be the same as FFTR. The thermal limit at BOC to End of COAST was developed to bound the corresponding earlier exposure FHOOS limit.
A step change in PLUOOS limits at 50% power is not supported since at 50% and below the LRNB with or without PLUOOS is the same event.
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0 AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-12 Table 5.5 MCPRp Limits for TSSS Insertion Times (Continued)
BOC BOC BOC Operating Power to to to End of Condition (% of rated) NEOC EOC Coast 100.0 1.42 1.42 1.44 60.0 1.62 1.62 1.66 55.0 1.80 1.80 1.80 5 0 .0 ---......
50.0 1.83 1.83 1.83 PLUOOS 40.0 1.91 1.91 1.91 26.0 2.23 2.23 2.33 26.0 at > 50%F 2.63 2.63 2.72 23.0 at > 50%F 2.78 2.78 2.90 26.0 at < 50%F 2.49 2.49 2.58 23.0 at < 50%F 2.62 2.62 2.72 100.0 1.45 1.47 1.48 60.0 1.65 1.65 1.68 55.0 1.67 1.67 1.72 50.0 1.71 1.71 1.77 50.0 1.83 1.83 1.83 EC TOO S 40.0 1.91 1.91 1.91 26.0 2.25 2.25 2.35 26.0 at > 50%F 3.12 3.12 3.20 23.0 at > 50%F 3.39 3.39 3.48 26.0 at - 50%F 2.65 2.65 2.72 23.0 at < 50%F 2.88 2.88 2.96 100.0 1.44 1.44 --
60.0 1.66 1.66 ---
55.0 1.70 1.70 50.0 1.75 1.75 --
E ROOS 50.0 40.0 1.83 1.91 1.83 1.91 --
26.0 2.33 2.33 26.0 at > 50%F 2.72 2.72 ---
23.0 at > 50%F 2.90 2.90 ---
26.0 at < 50%F 2.58 2.58 -
23.0 at < 50%F 2.72 2.72 --
100.0 1.42 1.42 1.44 60.0 1.62 1.62 1.66 55.0 1.80 1.80 1.80 50 .0 ---......
50.0 1.83 1.83 1.83 PLUOOS 40.0 1.91 1.91 1.91 26.0 2.23 2.23 2.33 26.0 at > 50%F 2.63 2.63 2.72 23.0 at > 50%F 2.78 2.78 2.90 26.0 at - 50%F 2.49 2.49 2.58 23.0 at < 50%F 2.62 2.62 2.72 AREVA NP Inc.
W W AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-13 Table 5.5 MCPRp Limits for TSSS Insertion Times (Continued)
BOC BOC BOC Operating Power to to to End of Condition (% of rated) NEOC EOC Coast 100.0 1.47 1.47 ---
60.0 1.68 1.68 ---
55.0 1.72 1.72 ---
50.0 1.77 1.77 ---
TBVOOS FHOOS 50.0 40.0 1.83 1.91 1.83 1.91 ---
26.0 2.35 2.35 ---
26.0 at > 50%F 3.20 3.20 ---
23.0 at > 50%F 3.48 3.48 ---
26.0 at - 50%F 2.72 2.72 ---
23.0 at < 50%F 2.96 2.96 --
100.0 1.45 1.46 1.47 60.0 1.65 1.65 1.68 55.0 1.80 1.80 1.80 50.0 --- --- --
TBVOOS 50.0 1.83 1.83 1.83 40.0 26.0 1.91 1.91 1.91 2.25 2.25 2.35 26.0 at > 50%F 3.12 3.12 3.20 23.0 at > 50%F 3.39 3.39 3.48 26.0 at 5 50%F 2.65 2.65 2.72 23.0 at _ 50%F 2.88 2.88 2.96 100.0 1.44 1.44 ---
60.0 1.66 1.66 --
55.0 1.80 1.80 ---
5 0 .0 ---......
50.0 1.83 1.83 ---
FHOOS ---
40.0 1.91 1.91 PLUOOS ---
26.0 2.33 2.33 26.0 at > 50%F 2.72 2.72 ---
23.0 at > 50%F 2.90 2.90 ---
26.0 at - 50%F 2.58 2.58 ---
23.0 at < 50%F 2.72 2.72 ---
100.0 1.47 1.47 ---
60.0 1.68 1.68 --
55.0 1.72 1.72 ---
50.0 1.77 1.77 ---
EOC-RPT-OOS 50.0 1.83 1.83 ---
TBVOOS 40.0 1.91 1.91 ---
FHOOS 26.0 2.35 2.35 ---
26.0 at > 50%F 3.20 3.20 ---
23.0 at > 50%F 3.48 3.48 ---
26.0 at - 50%F 2.72 2.72 ---
23.0 at < 50%F 2.96 2.96 ---
AREVA NP Inc.
V AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-14 Table 5.5 MCPRp Limits for TSSS Insertion Times (Continued)
BOC BOC BOC Operating Power to to to End of Condition (% of rated) NEOC EOC Coast 100.0 1.45 1.47 1.48 60.0 1.65 1.65 1.68 55.0 1.80 1.80 1.80 5 0 .0 ---......
EOC-RPT-OOS 50.0 1.83 1.83 1.83 TBVOOS 40.0 1.91 1.91 1.91 PLUOOS 26.0 2.25 2.25 2.35 26.0 at > 50%F 3.12 3.12 3.20 23.0 at > 50%F 3.39 3.39 3.48 26.0 at - 50%F 2.65 2.65 2.72 23.0 at 5 50%F 2.88 2.88 2.96 100.0 1.44 1.44 ---
60.0 1.66 1.66 ---
55.0 1.80 1.80 ---
50.0 --- --
EOC-RPT-OOS 50.0 1.83 1.83 ---
FHOOS 40.0 1.91 1.91 ---
PLUOOS 26.0 2.33 2.33 ---
26.0 at > 50%F 2.72 2.72 ---
23.0 at > 50%F 2.90 2.90 ---
26.0 at _ 50%F 2.58 2.58 --
23.0 at < 50%F 2.72 2.72 ---
100.0 1.47 1.47 ---
60.0 1.68 1.68 ---
55.0 1.80 1.80 ---
50.0 --- -
TBVOOS 50.0 1.83 1.83 ---
FHOOS 40.0 1.91 1.91 ---
PLUOOS 26.0 2.35 2.35 --
26.0 at > 50%F 3.20 3.20 ---
23.0 at > 50%F 3.48 3.48 ---
26.0 at 5 50%F 2.72 2.72 ---
23.0 at < 50%F 2.96 2.96 ---
100.0 1.47 1.47 ---
60.0 1.68 1.68 ---
55.0 1.80 1.80 ---
E O C -R P T -O O S 5 0 .0 .........
50.0 1.83 1.83 ---
TBVOOS 40.0 1.91 1.91 ---
FHOOS 2.35 2.35 ---
PLUOOS 26.0 26.0 at > 50%F 3.20 3.20 ---
23.0 at > 50%F 3.48 3.48 ---
26.0 at - 50%F 2.72 2.72 ---
23.0 at < 50%F 2.96 2.96 ---
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-15 Table 5.6 LHGRFACp Multipliers NSSITSSS Insertion Times All Exposures*
Operating Power ATRIUM-10 Condition (% of rated) LHGRFACp 100 1.00 26 0.64 Base case 26 at > 50%F 0.53 operation t 23 at > 50%F 0.51 26 atS 50%F 0.53 23 at < 50%F 0.53 100 1.00 26 0.64 EOOS with 26 at > 50%F 0.53 TBV in-servicet 23 at > 50%F 0.51 26 at - 50%F 0.53 23 atS 50%F 0.53 100 1.00 26 0.63 EOOS with 26 at > 50%F 0.46 TBVOOS' 23 at > 50%F 0.42 26 atS 50%F 0.53 23 atS 50%F 0.52 Limits support operation with any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. Base case supports single-loop operation..
t Limits are applicable for all the EOOS scenarios presented in Table 1.1 except those that include TBVOOS.
Limits are applicable for all the EOOS scenarios presented in Table 1.1 including those with TBVOOS.
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0 AREVA NP S
ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 5-16 Table 5.7 Control Rod Withdrawal Error MCPR versus RBM Setpoint Results (for Rated Power and 1.09 SLMCPR)
Analytical CRWE RBM MCPR Setpoint (w/o filter)
(%)
107 1.27 111 1.32 114 1.35 117 1.35 Table 5.8 RBM Setpoint Applicability Thermal Power Applicable
(% of rated) MCPR*
< 1.74 TLO
> 27% and < 90%
< 1.77 SLO
> 90% < 1.43 TLO1
t Greater than 90% rated power is not attainable in SLO.
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 6-1 6.0 Postulated Accidents 6.1 Loss-of-CoolantAccident 6.1.1 Break Location Spectrum Reference 9.7 6.1.2 Break Size Spectrum Reference 9.7 6.1.3 MAPLHGR Analyses The MAPLHGR limits presented in Reference 9.8 remain valid for ATRIUM-10 fuel.
Limiting Break: 0.5 ft 2 split Recirculation Pump Discharge Line Battery (DC) power Based on the PCT results in Reference 9.8 and subsequent evaluations to provide 10 CFR 50.46 reporting estimates (Reference 9.10), the current licensing PCT is provided below. The MCPR value used in the LOCA analyses is less than the rated power MCPR limits presented in Section 5.0.
Initial PCT (OF) 2007 (Reference 9.8) 10 CFR 50.46 Estimates -5 net cumulative value (OF)
(Reference 9.10)
CurrentLicensing PCT (OF) 2002 The peak local metal-water reaction for the limiting PCT lattice design is 1.71%. The maximum core wide metal-water reaction (for hydrogen generation) for a full ATRIUM-10 core is <1.0%.
The PCT for Cycle 14 ATRIUM-10 reload fuel was calculated to be 19720 F; therefore, in terms of PCT, the limiting neutronic design used in Reference 9.8 remains bounding. The peak local metal-water reaction and total core wide metal-water reaction were calculated to be 1.63% and
<1%, respectively. When compared to the acceptance criteria of less than 17% local cladding oxidation thickness, the local metal-water reaction result remains acceptable.
The plant parameters for the LOCA analysis (Reference 9.7) bound the cycle-specific plant parameters documented in Reference 9.13. The LOCA analysis and results support the EOD and EOOS conditions listed in Table 1.1. Note that the following EOOS conditions have no AREVA NP Inc.
0 0 AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 6-2 direct influence on the LOCA events: TBVOOS, EOC-RPT-OOS, PLUOOS, and TIPOOS/LPRM out-of-service.
6.2 ControlRod Drop Accident Browns Ferry Unit 3 uses a banked position withdrawal sequence (BPWS) including reduced notch worth (RNW) rod pulls to limit high worth control rod movements. A CRDA evaluation was performed for both A and B sequence startups consistent with the withdrawal sequence specified by TVA.
The CRDA analysis demonstrates that the maximum deposited fuel rod enthalpy is less than the NRC limit of 280 cal/g (fuel dispersal) and that the estimated number of fuel rods that exceed the fuel damage threshold of 170 cal/g is less than the number of failed rods (850 rods) assumed in the Browns Ferry UFSAR radiological assessment. The inputs to the deposited enthalpy calculation are determined on a cycle specific basis using the methods described in Reference 8.5. Key results from the CRDA analysis are summarized below:
Maximum dropped control rod worth, mk 11.3 Core average Doppler coefficient, Ak/k/=F -10.0 x 10-6 Effective delayed neutron fraction 0.0052 Four-bundle local peaking factor 1.39 Maximum deposited fuel rod enthalpy, cal/g 210.2 Maximum number of rods exceeding 170 cal/g 182 6.4 Fuel and Equipment Handling Accident The fuel handling accident radiological analysis implementing the alternative source term (AST) as approved in Reference 9.17 was performed with consideration of ATRIUM-10 core source terms. The number of failed fuel rods for the ATRIUM-10 fuel as previously provided to TVA in Reference 9.18 for use in the AST analysis is unchanged. No other aspect of utilizing the ATRIUM-10 fuel affects the current analysis; therefore, the AST analysis remains bounding for the AREVA ATRIUM-10 fuel.
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 7-1 7.0 Technical Specifications.
7.1 Limiting Safety System Settings 7.1.1 MCPR Fuel Cladding Integrity Safety Limit MCPR Safety Limit (all fuel) - two-loop operation 1.09*
MCPR Safety Limit (all fuel) - single-loop operation 1.11*
7.1.2 Steam Dome Pressure Safety Limit Pressure Safety Limit 1325 psig 7.2 Limiting Conditions for Operation 7.2.1 Average Planar Linear Heat Generation Ratet Reference 9.8 MAPLHGR Limits Average Planar Exposure MAPLHGR (GWd/MTU) (kW/ft) 0.0 12.5 15.0 12.5 67.0 7.3 Single-Loop Operation MAPLHGR Multiplier Reference 9.8 for ATRIUM-10 Fuel is 0.85.
7.2.2 Minimum Critical Power Ratio Flow-Dependent MCPR Limits: Tables 5.1 and 5.2 Exposure-Dependent MCPRp Limits Tables 5.4 and 5.5
- Includes the effects of channel bow, 2 TIPOOS or the equivalent number of TIP channels (per operating requirements defined in Reference 9.6 Section 3.2), a 2500 EFPH LPRM calibration interval, and up to 50% of the LPRMs out-of-service.
1 Limits are applicable for all of the EOOS scenarios presented in Table 1.1. For SLO operation, the MAPLHGR multiplier listed in Section 7.2.1 is applied.
Refer to References 9.11, 9.12, and 9.16 for the maximum licensing exposures.
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S AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 7-2 7.2.3 Linear Heat Generation Rate References 9.11, 9.12 and 9.16 Steady-State LHGR Limits Pellet Exposure LHGR (GWd/MTU) (kW/ft) 0.0 13.4 18.9 13.4 74.4* 7.1 The PAPT LHGR curves are identified in References 9.11, 9.12 and 9.16. The LHGRFACf and LHGRFACp multipliers are applied directly to the steady-state LHGR limits at reduced power and reduced flow to ensure the PAPT LHGR limits are not violated during an AOT.
LHGRFAC Multipliers for Off-Rated Conditions:
LHGRFACf Table 5.3 LHGRFACp Table 5.6
- Refer to References 9.11, 9.12, and 9.16 for the maximum licensing exposures.
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 8-1 8.0 Methodology References See XN-NF-80-19(P)(A) Volume 4 Revision 1 for a complete bibliography.
8.1 ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis, Advanced Nuclear Fuels Corporation, August 1990.
8.2 ANF-524(P)(A) Revision 2 and Supplements 1 and 2, ANF CriticalPower Methodology for Boiling Water Reactors,Advanced Nuclear Fuels Corporation, November 1990.
8.3 EMF-2209(P)(A) Revision 2, SPCB CriticalPower Correlation,Framatome ANP, September 2003.
8.4 EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation'sCriticalPower Correlationsto Co-Resident Fuel, Siemens Power Corporation, August 2000.
8.5 EMF-2158(P)(A) Revision 0, Siemens Power CorporationMethodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.
8.6 EMF-CC-074(P)(A) Volume 4 Revision 0, BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-BZ Siemens Power Corporation, August 2000.
8.7 ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteriafor BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 9-1 9.0 Additional References 9.1 EMF-3218(P) Revision 0, Browns Ferry Unit 3 Thermal-HydraulicDesign Report for ATRIUM TM- 10 Fuel Assemblies, Framatome ANP, September 2005.
9.2 AN P-2617(P) Revision 1, Nuclear Fuel Design Report Browns Ferry Unit 3 Fabrication Batch BFE3-14 A TRIUM TM-1O Fuel, AREVA NP, May 2007.
9.3 ANP-2626(P) Revision 0, Browns Ferry Unit 3 Cycle 14 Fuel Cycle Design, AREVA NP, May 2007.
9.4 EMF-2939(P) Revision 0, Browns Ferry Nuclear PlantSpent Fuel Storage Pool Criticality Safety Analysis for A TRIUMTM-10 Fuel, Framatome ANP, August 2003.
9.5 Letter, A.W. Will (AREVA) to G. C. Storey (TVA), "Disposition of Events for Extended Power Uprate at Browns Ferry Units 2 and 3," AWW:06:065, May 1, 2006.
9.6 ANP-2630(P) Revision 0, Browns Ferry Unit 3 Cycle 14 Plant TransientAnalysis, AREVA NP, May 2007.
9.7 EMF-2950(P) Revision 1, Browns Ferry Units 1, 2, and 3 Extended Power Uprate LOCA Break Spectrum Analysis, Framatome ANP, April 2004.
9.8 EMF-3145(P) Revision 0, Browns Ferry Units 1, 2, and 3 Extended Power Uprate LOCA-ECCS Analysis MAPLHGR Limit forATRIUMTM-O Fuel, Framatome ANP, December 2004.
9.9 NEDO-32465-A, Licensing Topical Report, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, GE Nuclear Energy, August 1996.
9.10 Letter, T.A. Galioto (AREVA) to G.C. Storey (TVA), "10 CFR 50.46 PCT Reporting for BFN Units 2 and 3," TAG:05:056, June 30, 2005.
9.11 EMF-2971 (P) Revision 1, Mechanicaland Thermal-HydraulicDesign Report for Browns Ferry Unit 3 Batches BFC-landBFC-1A A TRIUM-I1 Fuel Assemblies, Framatome ANP, January 2004.
9.12 EMF-3213(P) Revision 0, MechanicalDesign Report for Browns Ferry Unit 3 Reload BFE3-13 A TRIUM-iO Fuel Assemblies, Framatome ANP, September 2005.
9.13 ANP-2589(P) Revision 0, Browns Ferry Unit 3 Cycle 14 Plant ParametersDocument, AREVA NP, January 2007.
9.14 EMF-2978(P) Revision 0, Browns Ferry Nuclear Plant New Fuel Storage Vault Criticality Safety Analysis for ATRIUM-10 Fuel, Framatome ANP, July 2005.
9.15 NEDC-32433P, Maximum Extended Load Line Limit andArts Improvement Program Analyses for Browns Ferry Nuclear Plant Unit 1, 2, and 3, GE Nuclear Energy, April 1995.
9.16 AN P-2628(P) Revision 0, MechanicalDesign Report for Browns Ferry Unit 3 Reload BFE3-14 ATRIUM-I FuelAssemblies, AREVA NP, May 2007.
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AREVA NP ANP-2631 Browns Ferry Unit 3 Cycle 14 Revision 0 Reload Analysis Page 9-2 9.17 Letter, E.A. Brown (NRC) to K.W. Singer (TVA), "Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Issuance of Amendments Regarding Full-Scope Implementation of Alternative Source Term (TAC Nos. MB5733, MB5734, MB5735, MC0156, MC0157, and MC0158)
(TS-405)," September 27, 2004.
9.18 Letter, T.A. Galioto (AREVA) to J.F. Lemons (TVA), "Fuel Handling Accident Assumptions for Browns Ferry," TAG:02:012, January 23, 2002.
9.19 USNRC Certificate of Compliance for Radioactive Material Packages, Model No.: RAJ-II, USA/9309/B(U)F-96 Revision 6.
9.20 Letter, N.J. Carr (AREVA) to G.C. Storey (TVA), "Update on RAJ-11 Shipping Container, Inner Container Stacking Criticality," NJC:04:052 FAB04-761, October 29, 2004.
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AREVA NP Browns Ferry Unit 3 Cycdle 14 AN P-2631 Reload Analysis Revision 0 0
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