ML17306A165
| ML17306A165 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 11/02/2017 |
| From: | April Pulvirenti Plant Licensing Branch IV |
| To: | Entergy Operations |
| Pulvirenti A | |
| References | |
| CAC MG0183, EPID L-2017-LRM-0037 | |
| Download: ML17306A165 (65) | |
Text
REACTOR VESSEL NEUTRON FLUENCE CALCULATION METHOD LICENSE AMENDMENT REQUEST PRE-SUBMITTAL MEETING WATERFORD 3 OCTOBER 19, 2017 1
Introductions
Mandy Halter - Director, Nuclear Licensing John Jarrell Manager, Regulatory Assurance Nicholas Petit Supervisor, Design Engineering Paul HymelDesign Engineer Alan Cox Entergy Corp. License Renewal (Consultant)
Ben AmiriWestinghouse Radiation Analysis Engineer Kevin FitzsimmonsProject Manager (Consultant)
Alan HarrisProject Licensing Specialist (Consultant) 2
Agenda
- Waterford 3 Current Licensing Basis (CLB)
- Requested change to CLB
- Justification for change
- Considerations for period of extended operation
- Discussion of NRC staff considerations 3
Waterford 3 Current Licensing Basis DORT method
- Approved for Waterford 3 in Amendment 196 (2004)
- Used for 2nd surveillance capsule, 263°
- Basis for P-T curves valid through 32 EFPY (approximate end of current license period) 4
Agenda
- Waterford 3 Current Licensing Basis (CLB)
- Requested change to CLB
- Justification for change
- Considerations for period of extended operation
- Discussion of NRC staff considerations 5
Requested Change to Current Licensing Basis Approve RAPTOR-M3G for Neutron Fluence Calculation Method for Reactor Vessel Beltline at Waterford 3
- Meets current regulatory guidance (RG 1.190 criteria)
- Considered other licensees submittals and approvals
- Addresses violation of 10 CFR 50.59 issued to Waterford 3 for use in 2015 Appendix H Report without prior NRC approval 6
Agenda
- Waterford 3 Current Licensing Basis (CLB)
- Requested change to CLB
- Justification for change
- Considerations for period of extended operation
- Discussion of NRC staff considerations 7
Justification for Change Current License Complies with applicable Regulatory Positions (RP) in Reg. Guide 1.190
- 1. Fluence Calculational Methods 16 13
- 2. Fluence Measurement Methods 7 0
- 3. Reporting Provisions 8 0 Derived from Table 1 Summary of Regulatory Positions(Reg. Guide 1.190) 8
Reg Guide 1.190 Regulatory Positions (Section 1)
Reg.
Position Topic Description 1.1.1 Modeling Data Modeling based on plant specific data
- 1.1.2 Applicable Nuclear Data RP ComplianceUseSummary latest Evaluated Nuclear Data File (ENDF/B) 1.1.2 X-Section Angular Representation Use minimum P3 angular decomposition 1.1.2 X-Section Group Collapse Demonstrate adequacy of collapsed job library (BUGLE-96) 1.2 Neutron Sources Neutron source model attributes requirements 1.2 End-of-Life Predictions Use best estimate or conservative power distribution 1.3 Fluence Determination Use fluence calculation, not extrapolated measurements 1.3.1 Spatial Representation Discrete ordinate neutron transport calculation minimum requirements 1.3.1 Multiple Transport Calculations N/A - Bootstrapping not performed 1.3.2 Point Estimates N/A - only applies to Monte Carlo analysis, which was not used Statistical Tests Variance Reduction 1.3.3 Capsule Monitoring Capsule Geometry modeling requirements 1.3.3 Spectral Effects on RTNDT Use of spectral lead factor required 1.3.5 Cavity Calculations N/A - No dosimeters in Reactor Cavity 1.4.1 Analytic Uncertainty Analysis Sensitivity analysis of components comprising uncertainty 1.4.2 Comparisons with Benchmark Validation of method against plant model, vessel simulator model, Measurements and Calculations calculational benchmark model 1.4.3 Estimate of Fluence Calculational Requirement to meet 20% (maximum) vessel fluence uncertainty for RTPTS Bias and Uncertainty and RTNDT 9
Reg Guide 1.190 Regulatory Positions (Section 1)
Reg.
Position Topic Description 1.1.1 Modeling Data Modeling based on plant specific data
- 1.1.2 Applicable Nuclear Data RP ComplianceUseSummary latest Evaluated Nuclear Data File (ENDF/B) 1.1.2 X-Section Angular Representation Use minimum P3 angular decomposition 1.1.2 X-Section Group Collapse Demonstrate adequacy of collapsed job library (BUGLE-96) 1.2 Neutron Sources Neutron source model attributes requirements 1.2 End-of-Life Predictions Use best estimate or conservative power distribution 1.3 Fluence Determination Use fluence calculation, not extrapolated measurements 1.3.1 Spatial Representation Discrete ordinate neutron transport calculation minimum requirements 1.3.1 Multiple Transport Calculations N/A - Bootstrapping not performed 1.3.2 Point Estimates N/A - only applies to Monte Carlo analysis, which was not used Statistical Tests Variance Reduction 1.3.3 Capsule Monitoring Capsule Geometry modeling requirements 1.3.3 Spectral Effects on RTNDT Use of spectral lead factor required 1.3.5 Cavity Calculations N/A - No dosimeters in Reactor Cavity 1.4.1 Analytic Uncertainty Analysis Sensitivity analysis of components comprising uncertainty 1.4.2 Comparisons with Benchmark Validation of method against plant model, vessel simulator model, Measurements and Calculations calculational benchmark model 1.4.3 Estimate of Fluence Calculational Requirement to meet 20% (maximum) vessel fluence uncertainty for RTPTS Bias and Uncertainty and RTNDT 10
Focus Discussion Topics Regulatory Position Topic 1.1.2 X-Section Angular Representation 1.3.1 Spatial Representation 1.4.1 Analytic Uncertainty Analysis 1.4.2 Comparisons with Benchmark Measurements and Calculations 1.4.3 Estimate of Fluence Calculational Bias and Uncertainty 11
Reg. Position 1.1.2 Cross-Section Angular Representation In discrete ordinates transport calculations, a P3 angular decomposition of the scattering cross section (at a minimum) must be employed.
- WF3 RAPTOR-M3G calculations performed using P5 angular decomposition 12
Reg. Position 1.3.1 Spatial Representation Discrete ordinates calculations should incorporate mesh of 2 and 3 intervals/in at core periphery and water, respectively. Steel mesh should include at least 1.5 intervals/in in steel. Axial mesh of 0.5 intervals/in is acceptable, except where material and source interfaces cause high flux gradients. Calculations must employ at least an S8 quadrature and at least 40 intervals per octant.
- 4 intervals/in in water and peripheral assemblies
- 1.7 intervals/in in steel areas
- 160 radial x 121 azimuthal x 247 axial intervals
- Mesh size chosen to achieve proper convergence of inner iterations on point-wise basis
- S16 level-symmetric quadrature 13
Reg. Position 1.4.1 Analytic Uncertainty Analysis The calculation methodology must be qualified by an analytic uncertainty analysis
- Westinghouse 4-loop model inputs as base case
- Differences between Westinghouse 4-loop and Combustion Engineering 2-loop accounted for because the plant-specific parameters are model inputs
- Variation in input parameters considered bounding for Westinghouse 4-loop design
- Expected variation in WF3 model inputs essentially the same or bounded by Westinghouse 4-loop design 14
Reg. Position 1.4.1 Analytic Uncertainty Analysis The calculation methodology must be qualified by an analytic uncertainty analysis UNCERTAINTY COMPONENT Peripheral Assembly Source Strength Pin Power Distribution Peripheral Assembly Burnup Axial Power Distribution Internals Dimensions Vessel Inner Radius Vessel Thickness Coolant Temperature Core Periphery Modeling 15
Reg. Position 1.4.1 Analytic Uncertainty Analysis: Geometric Uncertainty Base Case (Catawba)
- WF3 thickness tolerance not bounded by variation in generic uncertainty analysis.
- Thickness tolerance contributes 1.5% to analytic uncertainty in generic model at +12 cm from core midplane; if doubled, this would change the total analytic uncertainty to 10.37%, rather than 10.02%. Net uncertainty would increase from 14.12% to 14.37%.
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Reg. Position 1.4.1 Analytic Uncertainty Analysis: Operational Parameters Summary of Operational Parameters used in the Catawba Unit 1 Uncertainty Analysis and in Waterford 3 Base Case Uncertainty Analysis Waterford 3 Parameter (Catawba) Parameter System Pressure (psia) 2250 2250 Core average 583 ( Cycles 1-5) moderator 585 574-575 (Cycles 6-19) temperture (oF)
Core inlet (downcomer) 553 ( Cycles 1-5) 553 temperture (oF) 543-545 (Cycles 6-19)
Peripheral assembly average burnup range 3,000 TO 50,000 2,548 TO 51,356 (MWD/MTU)
Cycle-average axial peak 1.2 ( Cycle 1) relative power 1.1 1.1 (Cycles 2-19) 17
Reg. Position 1.4.1 Analytic Uncertainty Analysis:
Core Average Moderator and Core Inlet Temperatures The +/-10°F variation in temperatures in the generic model results in a greater density range than the same temperature variation would in the WF3 model.
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Reg. Positon 1.4.1 Analytic Uncertainty Analysis:
Peripheral Assembly Average Burnup Range
- Burnup range of 3,000-50,000 MWD/MTU used for generic uncertainty analysis
- Fuel assembly burnup variation +/-5000 MWD/MTU applied at boundaries of range
- WF3 operating range is 2,548-51,356 MWD/MTU
- Because differences at each end of the range between the generic model and WF3 conditions are small, no significant difference in analytic uncertainty is expected.
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Reg. Position 1.4.2.1 Operating Reactor Measurements Well-documented fluence dosimetry measurements for operating power reactors may be used for methods and data qualification. Comparisons must be performed for the specific reactor being analyzed or for reactors of similar design.
- Operating power reactor database: 18 plants, 156 capsules (in- and ex-vessel), 749 foil sensors
- Includes common PWR designs
- M/C = 1.03, = 5% for in-vessel
- M/C = 0.92, = 6% for ex-vessel
- WF3 surveillance capsule database: 3 in-vessel capsules, 11 foil sensors
- M/C = 1.11, = 7%
- Validates RAPTOR-M3G method uncertainty of 14% for WF3
- Catawba in-vessel dosimetry M/C = 0.998, = 12.8%
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Reg. Position 1.4.2.1 Operating Reactor Measurements Standard Capsule Measured/Calculated Deviation Database Location M/C Operating Power In-Vessel 1.03 5%
Reactor database Operating Power Ex-Vessel 0.92 6%
Reactor database Waterford 3 In-Vessel 1.11 7%
Catawba In-Vessel 0.998 13%
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Reg. Position 1.4.2.2 Pressure Vessel Simulator Measurements Several pressure vessel simulator benchmarks are available and may be used for methods qualification.
- Pool Critical Assembly (PCA) benchmark experiment
- VENUS-1 benchmark experiment
- H.B. Robinson-2 cavity measurement
- Contributions to overall method uncertainty were:
- 5% PCA/VENUS
- 7% H.B. Robinson
- Consistent with Catawba Unit 1 submittal 22
Reg. Position 1.4.2.3 Calculational Benchmark The vessel fluence benchmark problems provided by the NRC in NUREG/CR-6115 should be used for methods qualification
- PWR standard core loading problem
- Mean Percent difference: -2.74%
- Not used as input to method uncertainty
- Consistent with Catawba Unit 1 submittal 23
Reg. Position 1.4.3 Estimate of Calculational Bias and Uncertainty The overall fluence calculational bias and uncertainty must be determined by an appropriate combination of the (1) analytic uncertainty analysis results of Reg. Position 1.4.1 and (2) the results of the uncertainty analysis based on the comparisons to the operating reactor and simulator benchmark measurements of Reg. Position 1.4.2. Vessel fluence uncertainty of 20% is acceptable for RTPTS and RTNDT determination.
Simulator Benchmark Comparisons 5%
H.B Robinson Benchmark Comparisons 7%
Analytic Sensitivity Studies 10%
Peripheral Assembly Source Strength 5%
Pin Power Distribution 1%
Peripheral Assembly Burnup 1%
Axial Power Distribution 1%
Internals Dimensions 1%
Vessel Inner Radius 4%
Vessel Thickness 0%
Coolant Temperature 6%
Core Periphery Modeling 5%
Other Factors 5%
Net Uncertainty 14%
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Requested Change to Current Licensing Basis
- Approve RAPTOR-M3G for Neutron Fluence Calculation Method at Waterford 3
- Intended submittal date 12/1/2017
- Separate attachments
- 10 CFR Part 54 supplemental information
- LAR Questions/Open Discussion 25
Agenda
- Waterford 3 Current Licensing Basis (CLB)
- Requested change to CLB
- Justification for change
- Considerations for period of extended operation
- Discussion of NRC staff considerations 26
Considerations for Period of Extended Operation
- Current LRA submittal based on using RAPTOR-M3G
- MRP-227A remains applicable to Reactor Vessel Internals
- Extended Beltline Region
- Analytic Uncertainty Analysis
- Benchmarking
- Operating power reactor database includes dosimetry at upper extent of core 27
Reg. Position 1.4.3 Estimate of Calculational Bias and Uncertainty (Applied to Extended Beltline)
The overall fluence calculational bias and uncertainty must be determined by an appropriate combination of the (1) analytic uncertainty analysis results of Reg. Position 1.4.1 and (2) the results of the uncertainty analysis based on the comparisons to the operating reactor and simulator benchmark measurements of Reg. Position 1.4.2. Vessel fluence uncertainty of 20% is acceptable for RTPTS and RTNDT determination.
Extended Beltline Beltline Simulator Benchmark Comparisons 5% 5%
H.B Robinson Benchmark Comparisons 7% 7%
Analytic Sensitivity Studies 10% 15%
Peripheral Assembly Source Strength 5% 3%
Pin Power Distribution 1% 0%
Peripheral Assembly Burnup 1% 1%
Axial Power Distribution 1% 10%
Internals Dimensions 1% 2%
Vessel Inner Radius 4% 3%
Vessel Thickness 0% 0%
Coolant Temperature 6% 10%
Core Periphery Modeling 5% 5%
Other Factors 5% 5%
Net Uncertainty 14% 18%
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Surveillance Capsule Removal Schedule
- License Renewal Commitment for submittal of withdrawal schedule within one year following receipt of renewed license
- Anticipated removal of capsule 277° is 48 EFPY 29
Agenda
- Waterford 3 Current Licensing Basis (CLB)
- Requested change to CLB
- Justification for change
- Considerations for period of extended operation
- Discussion of NRC staff considerations 30
Discussion of NRC considerations Item 1: Bias and uncertainty estimates Item 2: Sensitivity calculations in extended beltline Item 3: Measurement benchmarks for extended beltline Item 4: Calculation benchmark for extended beltline 31
Item 1: Bias and Uncertainty Estimates
- RVI aging effects managed by MRP-227A; No request for fluence calculations
- Use of generic Westinghouse 4-loop for analytic uncertainty analysis addressed in LAR submittal as discussed
- Estimate of total method uncertainty will be given for beltline +12 cm from core midplane and for upper circumferential weld (limiting extended beltline material) +52 cm from top of core 32
Item 2: Sensitivity Calculations in Extended Beltline
- RVI aging effects managed by MRP-227A; No request for fluence calculations
- Nozzle forgings and welds not included in extended beltline (fluence for weld is 4.07 x 1016 n/cm2 at 60 EFPY)
- Step change increase in cavity gap in upper portion of core (~1m)
- Meets 20% guidance for traditional beltline
- 5% other factors uncertainty category
- Off-midplane EVND validates calculations in regions of increased axial power distribution sensitivity (M/C = 0.93, = 7 %)
- Investigate M/C comparisons referenced in Concern Item 3 to determine sensitivity to items a)-e) of referenced Enclosure 3 response 33
34 35 36
Item 3: Measurement Benchmarks for Extended Beltline
- RVI aging effects managed by MRP-227A; No request for fluence calculations
- Off-midplane EVND validates calculations in regions of increased axial power distribution sensitivity
- Investigate use of M/C comparisons referenced in Concern Item 3 to provide benchmarking of fluence calculations in extended beltline region.
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Item 4: Calculation Benchmark for Extended Beltline
- RAPTOR-M3G compared with calculational benchmark in NUREG-6115
- No precedent for requiring site-specific calculational benchmarking to other than NUREG-6115
- Demonstrated uncertainty estimate at upper circumferential weld
<20%
- EVND benchmarking in off-midplane regions validates calculations in regions of increased axial power distribution sensitivity
- Nozzles and associated welds have <1017 n/cm2 Suggested action would not significantly contribute to increased confidence of adequate margin in P-T curves 38
Questions/Open Discussion/Next Steps
- Approve RAPTOR-M3G for Neutron Fluence Calculation Method at Waterford 3
- Intended submittal date 12/1/2017
- Separate attachments
- 10 CFR Part 54 supplemental information
REACTOR VESSEL NEUTRON FLUENCE CALCULATION METHOD LICENSE AMENDMENT REQUEST PRE-SUBMITTAL MEETING WATERFORD 3 OCTOBER 19, 2017 Conclusions & Questions 40
REACTOR VESSEL NEUTRON FLUENCE CALCULATION METHOD LICENSE AMENDMENT REQUEST PRE-SUBMITTAL MEETING WATERFORD 3 OCTOBER 19, 2017 Backup Slides Follow 41
42 43 Reg. Position 1.1.1, Modeling Data To the extent possible, plant-specific as-built dimensions and materials should be used as model inputs
- As-built water annulus thickness & surveillance capsule positions
- Design values used for water temperatures, internals dimensions.
- Materials data taken from BUGLE-96 documentation, where available, or internal Westinghouse data sources 44
Reg. Position 1.1.2, Nuclear Data The latest version of the ENDF/B should be used for determining nuclear cross-sections. Cross-section sets based on earlier or equivalent nuclear data sets that have been thoroughly benchmarked are also acceptable.
- RAPTOR-M3G uses BUGLE-96 cross section library, based on ENDF/B-VI.
- BUGLE-97 (ENDF/B-VII) is most current
- Evaluation by Oak Ridge1 analyzed PCA, VENUS-3, and H.B.
Robinson benchmarks using BUGLE-96 and BUGLE-97; determined insignificant differences for high energy neutron applications 1RSICC Data Library Collection DLC-245, VITAMIN-B7/BUGLE-B7, Broad-Group and FineGroup and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data, Radiation Safety Information Computational Center, Oak Ridge National Laboratory (ORNL), October 2011.
45
Reg. Position 1.1.2, Cross-Section Angular Representation In discrete ordinates transport calculations, a P3 angular decomposition of the scattering cross section (at a minimum) must be employed.
- WF3 RAPTOR-M3G calculations performed using P5 angular decomposition 46
Reg. Position 1.1.2, Cross-Section Group Collapsing The adequacy of the collapsed job library must be demonstrated by comparing calculations for a representative configuration performed with both the master and job library.
- WF3 RAPTOR-M3G calculations used BUGLE-96 library
- BUGLE-96 validated by Oak Ridge National Laboratory1; in widespread use 1RSIC Data Library Collection DLC-185, BUGLE-96, Coupled 47 Neutron, 20 Gamma Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications, Radiation Shielding Information Center, Oak Ridge National Laboratory (ORNL), March 1996.
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Reg. Position 1.2, Neutron Source Core neutron source should account for local fuel isotopes and moderator density. Neutron source normalization and energy dependence must account for the fuel exposure dependence of the fission spectra, number of neutrons produced per fission, and energy released per fission.
- Energy distribution of source determined using fuel burnup representative of conditions averaged over applicable irradiation period
- Initial fuel enrichment characteristic of core designs used over applicable period
- Composite values of energy/fission, neutrons/fission, fission spectrum 48
Reg. Position 1.2, End-of-Life Predictions Predictions of end-of-life fluence should be made with best-estimate or conservative generic power distribution.
- Future projections based on current reactor power level of 3716 MWt
- Include 5% positive bias to power generated by peripheral assemblies for margin 49
Reg. Position 1.3, Fluence Determination Absolute fluence calculations, rather than extrapolated fluence measurements, must be used for the fluence determination.
- RAPTOR-M3G was used to calculate the fluence values directly based on the neutron source at the reactor core.
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Reg. Position 1.3.1, Spatial Representation Discrete ordinates calculations should incorporate mesh of 2 and 3 intervals/in at core periphery and water, respectively. Steel mesh should include at least 1.5 intervals/in in steel. Axial mesh of 0.5 intervals/in is acceptable, except where material and source interfaces cause high flux gradients. Calculations must employ at least an S8 quadrature and at least 40 intervals per octant.
- 4 intervals/in in water and peripheral assemblies
- 1.7 intervals/in in steel areas
- 160 radial x 121 azimuthal x 247 axial intervals
- Mesh size chosen to achieve proper convergence of inner iterations on point-wise basis
- S16 level-symmetric quadrature 51
Reg. Position 1.3.1, Multiple Transport Calculations If the calculation is performed using two or more bootstrap calculations, the adequacy of the overlap regions must be demonstrated.
- Bootstrapping was not performed; therefore, this part of Reg. Position 1.3.1 does not apply.
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Reg. Position 1.3.2, Point Estimates If the dimensions of the tally region or the definition of the average-flux region introduce a bias in the tally edit, the Monte Carlo prediction should be adjusted to eliminate the bias.
- RAPTOR-M3G is a deterministic method; therefore, this part of Reg. Position 1.3.2 does not apply 53
Reg. Position 1.3.2, Statistical Tests The Monte Carlo estimated mean and relative error should be tested and satisfy all statistical criteria
- RAPTOR-M3G is a deterministic method; therefore, this part of Reg. Position 1.3.2 does not apply 54
Reg. Position 1.3.2, Variance Reduction All variance reduction methods should be qualified by comparison with calculations performed without variance reduction
- RAPTOR-M3G is a deterministic method; therefore, this part of Reg. Position 1.3.2 does not apply 55
Reg. Position 1.3.3, Capsule Modeling The adequacy of the capsule representation and mesh must be demonstrated.
- Measurement/Calculation (M/C) ratios from operating power reactor database:
- 295 in-vessel (M/C = 1.03)
- 454 ex-vessel (M/C = 0.92)
- WF3 surveillance capsules
- M/C = 1.11 56
Reg. Position 1.3.3, Spectral Effects When fluence is extrapolated from inside surface of vessel to T/4 and 3T/4 locations using E > 1MeV fluence, a spectral lead factor must be applied to the fluence for the calculation of RTNDT
- Westinghouse fluence determination methodology uses Eq. 3 of Reg. Guide 1.991
- Reg. Guide 1.190 states this spectral lead factor has been included in the Eq. 3 attenuation formula of Rev. 2 of Reg. Guide 1.99, and therefore is not required when this formula is used.
1Regulatory Guide 1.99, rev. 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 1988.
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Reg. Position 1.3.5, Cavity Calculations In discrete ordinates transport calculations, the adequacy of the S8 angular quadrature used in cavity transport calculations must be demonstrated.
- WF3 RAPTOR-M3G calculations used S16 order of quadrature
- Cavity dosimetry not analyzed; therefore, this Reg. Position does not apply.
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Analytic Uncertainty Analysis:
Key Computational Parameters 59
Analytic Uncertainty Analysis:
Key Computational Parameters 60
Analytic Uncertainty Analysis: Axial Distribution and Isotopic Content of U-235 and Pu-239 in fuel
- Base axial power distribution representative of mid-cycle conditions
- Fuel burnup representative of conditions averaged over irradiation period
- Initial fuel enrichment characteristic of core designs used
- 10% uncertainty in axial power distribution shape considered in sensitivity study 61
Analytic Uncertainty Analysis:
Biological Shield Concrete Composition
- Standard, generic concrete composition used for reactor cavity model
- Concrete composition not considered in sensitivity study
- Deemed unlikely to significantly affect values of fluence for materials near or above 1017 threshold
- Concrete hydrogen (H) number density 7.8 x 10-3 atoms/b-cm assumed.
- Near low end of range from ORNL studies
- Low H tends to provide conservative fluence results
- Overall method uncertainty contains 5% other factors 62
Analytic Uncertainty Analysis: Cavity Gap
- Plant drawings using as-designed/as-built reactor cavity dimensions used in construction of WF3 model
- Cavity gap dimensions uncertainty not considered in sensitivity study
- Not deemed significant contributor to fluence for materials near or above 1017 threshold
- Overall method uncertainty contains 5% other factors 63
Analytic Uncertainty Analysis: Homogenized Materials Above and Below the Core Active Region
- WF3 model uses approximate mixture of stainless steel and borated water for regions directly above and below core
- Effect of varying balance between water and steel not considered in analytic uncertainty analysis
- Model forming base case of uncertainty analysis used mixtures derived from reactor internals drawings
- Uncertainties not deemed significant contributor to fluence for materials near or above 1017 threshold
- Overall method uncertainty contains 5% other factors 64
Analytic Uncertainty Analysis: Discretization Effects
- WF3 model uses S16 level-symmetric quadrature
- Appendix F of WCAP-18060-NP, Rev. 11 compares S8 and S12 quadrature with no significant difference
- S16 not expected to vary significantly from S8 or S12
- Difference between QR16 and S16 not considered in sensitivity study
- ORNL presentation indicates difference between QR16 and S16 is negligible in regions with 1017 fluence
- Uncertainties not deemed significant contributor to fluence for materials near or above 1017 threshold
- Overall method uncertainty contains 5% other factors 1Westinghouse Report WCAP-18060-NP, Rev. 1, Response to RAIs Concerning the Use of RAPTOR-M3G for the Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate Fluence Evaluations, November 2015. (ADAMS Accession Number ML15324A084.)
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