ML17290A655

Public Meeting, Licensee Presentation Slides Regarding Waterford, Unit 3, Fluence Methodology

ML17290A655
Person / Time
Site:	Waterford
Issue date:	10/19/2017
From:	Entergy Operations
To:	April Pulvirenti Plant Licensing Branch IV
Pulvirenti A
References
Download: ML17290A655 (36)

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 concerns 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 concerns 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 concerns 7

Justification for Change Current License Complies with applicable Regulatory Positions (RP) in Reg. Guide 1.190

1. of RPs that Total # of compliance may RG 1.190 Section RPs change for Waterford 3

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

• 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%.

16

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 Catawba Unit 1 Waterford 3 Parameter Uncertainty Analysis 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.

18

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.

19

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%

20

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%

21

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%

24

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 50 LAR

• 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 concerns 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

• Quantify uncertainty at location of upper circumferential weld on WF3 vessel

- 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%

28

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 concerns 30

Discussion of NRC concerns Item 1: Bias and uncertainty estimates Item 2: Sensitivity calculations Item 3: Measurement benchmarks for extended beltline Item 4: Calculation benchmark for extended beltline 31

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 50 LAR

• 10 CFR Part 54 supplemental information

• All Open issues resolved for LRA to support ACRS 32

REACTOR VESSEL NEUTRON FLUENCE CALCULATION METHOD LICENSE AMENDMENT REQUEST PRE-SUBMITTAL MEETING WATERFORD 3 OCTOBER 19, 2017 Conclusions & Questions 33

34 35 36