ML052340753
| ML052340753 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/04/2005 |
| From: | Ellen Brown NRC/NRR/DLPM/LPD2 |
| To: | |
| Brown Eva, NRR/DLPM, 415-2315 | |
| References | |
| +kBR1SISP20050826, TAC MC6454, TAC MC6455 | |
| Download: ML052340753 (141) | |
Text
K R EVA Meeting Agenda August 4, 2005 Topic Time Presenter
> Introduction 15 Holm
> CASMO4/MB2 Methodology 150 Grummer CASMO4 characterization of BWR lattice MB2 nodal XSEC representation Experience with high voids Axial void distribution uncertainty Recent gamma scan data
> Safety Analysis Methodology Uncertainties Treatment of Uncertainties in Safety Analyses 30 Garrett SLMCPR Overview 60 Garrett SLMCPR Sensitivity to Power Distribution Uncertainty
> Bypass Modeling 30 Grummer
> Summary 15 Holm FRAMATOME ANP, INC. Fuel Performance Meeting-August 4, 2005 2
CASMO-4/MICROBURN-B2 Methodology Ralph Grummer Manager, Core Physics Methods Richland, WA August 4, 2005 FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 1
BWR Methodology Applicability
> Objective Describe the cross section re-construction process used by Framatome-ANP Demonstrate that the Framatome-ANP Methodology is accurate for high void conditions FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 2
CASMO-4
> CASMO-4 performs a multi-group (70) spectrum calculation using a detailed heterogeneous description of the fuel lattice components Explicit modeling of fuel rods, absorber rods, water rods/channels and structural components The library has cross sections for 108 materials including 18 heavy metals Depletion performed with a predictor-corrector approach in each fuel or absorber rod Two-dimensional transport solution is based upon the Method of Characteristics FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 3
CASMO-4 (cont.)
Provides pin-by-pin power and exposure distributions Produces homogeneous multi-group (2) micro-scopic cross sections as well as macro-scopic cross sections Determines discontinuity factors Performs 18-group gamma transport calculation Ability to perform colorset (2X2) calculation with different mesh spacings Reflector calculations are easily performed FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 4
MICROBURN-B2
> Microscopic fuel depletion
> Full two energy group neutron diffusion equation solution
> Modern nodal method solution is used
> Uses a higher order spatial method
> Water gap dependent flux discontinuity factors
> Multilevel iteration technique for efficiency
> MICROBURN-B2 treats a total of 11 heavy metal nuclides to account for the primary reactivity components FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 5
MICROBURN-B2 (cont.)
> A model for nodal burnup gradient
> A model for spectral history gradient
> Full three-dimensional pin power reconstruction method
> TIP (neutron and gamma) and LPRM response models
> Steady state thermal hydraulics model
> Direct moderator heat deposition based upon CASMO-4 calculations
> Calculation of CPR, LHGR and MAPLHGR FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 6
BWR Methodology
> Let us look at the cross section representation used in MICROBURN-B2
> MICROBURN-B2 determines the nodal macroscopic cross sections by summing the contribution of the various nuclides FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 7
MICROBURN-B2 Cross Section Representation I
x ( , , , R) = N i =1 i
i x ( , , , R) + bx ( , , , R)
> where:
x = nodal macroscopic cross section bx = background nodal macroscopic cross section (D, f , a , r )
N i = nodal number density of nuclide " i" xi = microscopic cross section of nuclide " i" I = total number of explicitly modeled nuclides
= nodal instantaneous coolant density
= nodal spectral history E = nodal exposure R = control fraction FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 8
MICROBURN-B2 Cross Section Representation
> Functional representation of x and bx i
comes from 3 void depletion calculations with CASMO-4
> Instantaneous branch calculations at alternate conditions of void and control state are also performed
> The result is a multi dimensional table of microscopic and macroscopic cross sections FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 9
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 10
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 11
MICROBURN-B2 Cross Section Representation
> At BOL the relationship is fairly simple The cross section is only a function of void fraction (water density)
The reason for the variation is the change in the spectrum due to the water density variations
> At any exposure point, a quadratic fit of the three CASMO-4 data points is used to represent the continuous cross section over instantaneous variation of void or water density.
FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 12
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 13
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 14
MICROBURN-B2 Cross Section Representation
> Detailed CASMO-4 calculations confirm that a quadratic fit accurately represents the cross sections FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 15
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 16
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 17
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 18
MICROBURN-B2 Cross Section Representation
> With depletion the isotopic changes cause other spectral changes
> Cross sections change due to the spectrum changes
> Cross sections also change due to self shielding as the concentrations change
> These are accounted for by the void (spectral) history and exposure parameters
> Exposure variations utilize a piecewise linear interpolation over tabulated values at 100 exposure points
> The four dimensional representation can be reduced to three dimensions by looking at a single exposure FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 19
MICROBURN-B2 Cross Section Representation This is a smooth well behaved surface FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 20
MICROBURN-B2 Cross Section Representation
> Quadratic interpolation is performed in each direction independently for the most accurate representation.
FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 21
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 22
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 23
MICROBURN-B2 Cross Section Representation
> The results of this process for all isotopes and all cross sections in MICROBURN-B2 were compared for an independent CASMO-4 calculation with continuous operation at 40% void (40 % void history) and branch calculations at 90% void for multiple exposure.
> The results show very good agreement for the whole exposure range.
FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 24
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 25
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 26
MICROBURN-B2 Cross Section Representation
> At the peak reactivity point multiple comparisons were made to show the results for various instantaneous void fractions FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 27
MICROBURN-B2 Cross Section Representation Quadratic fit using 0-40-80 provides excellent representation of data FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 28
MICROBURN-B2 Cross Section Representation
> Why not use higher void CASMO-4 depletions?
For example 0,45,90
> Introduces more error for intermediate void fractions.
> The following example shows the difference between a 0,40,80 and a 0,45,90 interpolation method FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 29
MICROBURN-B2 Cross Section Representation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 30
MICROBURN-B2 Cross Section Representation
> MICROBURN-B2 uses water density rather than void fraction in order to account for pressure changes as well as sub-cooled density changes
> MICROBURN-B2 uses spectral history rather than void history in order to account for other spectral influences due to actual core conditions (fuel loading, control rod inventory, leakage, etc.)
FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 31
MICROBURN-B2 Cross Section Representation
> The doppler feedback due to the fuel temperature is modeled by accumulating the Doppler broadening of microscopic cross sections of each nuclide I
x = ( Teff Tref ) Tx N i i
f i
where :
Teff = Effective Doppler Fuel Temperature Tref = Reference Doppler Fuel Temperature xi = microscopic cross section (fast and thermal absorption) of nuclide i N i = density of nuclide i FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 32
MICROBURN-B2 Cross Section Representation
> The partial derivatives are determined from branch calculations performed with CASMO-4 at various exposures and void fractions for each void history depletion FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 33
MICROBURN-B2 Cross Section Representation
> The tables of cross sections include data for controlled and uncontrolled states.
> Otherwise the process is the same for controlled states
> Other important feedbacks to nodal cross sections are lattice burnup/spectral history gradient and instantaneous spectral interaction between lattices of different spectra FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 34
CASMO-4 / MICROBURN-B2 Methodology
> Conclusion The methods used in CASMO-4 are state of the art The methods used in MICROBURN-B2 are state of the art The methodology accurately models a wide range of thermal hydraulic conditions FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 35
BWR Methodology Experience BWR Methodology Experience
> Objective Describe the experience base for Framatome-ANP methodologies Demonstrate that the Framatome-ANP Methodology is Applicable to EPU conditions at Browns Ferry FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 37
BWR Methodology Experience
> During the last meeting the range of assembly power and void fraction were presented
> Recent experience shows similar ranges of operation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 38
Topical Report Thermal Hydraulic Conditions Maximum assembly powers approaching 8 MW are in the benchmark database FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 39
Evaluation of Power Uprate for Browns Ferry Max assembly powers are less than those presented in the topical report FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 40
BWR Methodology Experience Current Experience is consistent with the topical report FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 41
Topical Report Thermal Hydraulic Conditions Maximum exit voids of 90% are in the benchmark database FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 42
Evaluation of Power Uprate for Browns Ferry Max exit voids are less than those presented in the topical report FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 43
BWR Methodology Experience Current Experience is consistent with the topical report FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 44
BWR Methodology Experience
> At the point of the highest exit void fraction, additional detail was evaluated Core average void axial profile Axial profile of the peak assembly Histogram of the nodal void fractions in core FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 45
Browns Ferry Current Design FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 46
Browns Ferry with Power Uprate FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 47
BWR Methodology Experience FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 48
Browns Ferry Current Design FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 49
Browns Ferry with Power Uprate Nodal void fractions between 70 and 80 percent are most prevalent FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 50
BWR Methodology Experience Current Experience has Similar Void Population as Expected for BFE Power Uprate FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 51
Experience with High Void Fractions
> Conclusions Reactor conditions for Browns Ferry with power uprate is not significantly different from current experience The range of void fractions in the topical report data exceeds that expected for the power uprate conditions The distribution of voids is nearly the same as current experience Cross section representation is accurate for power uprate conditions FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 52
BWR Methodology Power Distribution Uncertainties
Power Distribution Uncertainties
> Objective Describe the process used by Framatome-ANP to define the power distribution uncertainties Demonstrate that the Framatome-ANP Methodology is Applicable to EPU conditions at Browns Ferry FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 54
Power Distribution Uncertainties
> First we will look at how Framatome-ANP determined the measured power distribution uncertainties
> One of the major components is the comparison of measured and calculated TIPs
> This includes measurement uncertainty as well as calculation uncertainty FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 55
Power Distribution Uncertainties FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 56
Power Distribution Uncertainties FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 57
Power Distribution Uncertainties FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 58
Power Distribution Uncertainties FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 59
Power Distribution Uncertainties
> Axial power distribution uncertainties were determined by the simple relationship Nodal = radial
- axial Nodal2 = radial2 + axial2
> Axial uncertainty was determined to be 1.81 % for C-lattice Plants and 2.91% for D-Lattice Plants
> Another component might be the radial uncertainty at an axial level
> The EMF-2158(P)(A) data was re-evaluated by looking at the deviations between measured and calculated TIP response for each axial level FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 60
Power Distribution Uncertainties There does not appear to be any axial dependency on the standard deviation FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 61
BWR Power Distribution Uncertainties
> There is very limited data on measured power distributions
> The measured power is determined by modifying the calculated power distribution using the measured and calculated LPRM values.
Measured LPRM values are calibrated to the TIP measurements
> Assembly gamma scan measurements at Quad Cities were used to define the uncertainty of the correlation coefficients
> These correlation coefficients indicate the accuracy of the UPDATE methodology FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 62
Al AREVA B m Power Distribution Uncertainties
> The Bundle Correlation Coefficient for QC Cycle 2 was [ I
> The Bundle Correlation Coefficient for QC Cycle 4 was [ I
> The average value of [ ] was used in the determination of the measured power uncertainty
> Using the minimum correlation coefficient increases the measured uncertainty by [ 1%
> Using the maximum correlation coefficient decreases the measured uncertainty by [
FRAP.1ATOME ANP. INC. > CASMO-4MCROBURN-62 Methodology - August 4, 2005 - RGG:05:002 63
AREVA Gamma Scan Data
> Pin-by-Pin Gamma scan data is used for verification of the local peaking uncertainty
> Quad Cities Data indicated that this uncertainty was approximately [ 1%
> KWU measurements of 9x9 and ATRIUM-10 assemblies provided additional validation that this uncertainty was accurate.
> Comparisons to Monte Carlo calculations indicated an uncertainty of approximately [
1 FRAMATOF,lE A N P INC. > CASMO-4/lCROBURN-62 Methodology - August 4, 2005 - RGG:05:002 64
Quad Cities Gamma Scan Benchmark Results EMF-2158(P)(A) pp 8-6,7 This data includes measurement uncertainty.
Local power distribution uncertainty is not axial level dependent FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 65
Local Peaking Uncertainty
> Recent gamma scan measurements including ATRIUM-10 show similar comparisons at various axial levels
> These results do not indicate any trend relative to axial position FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 66
KWU-S Gamma Scan Benchmark Results EMF-2158(P)(A) pp 8-8 Local power distribution uncertainty is not axial level dependent FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 67
KWU-S Gamma Scan Benchmark Results EMF-2158(P)(A)
> Full axial scans were performed on 16 fuel rods
> Comparisons to calculated data show excellent agreement at all axial levels
> The dip in power associated with spacers is not modeled in MICROBURN-B2
> There is no indication of reduced accuracy at higher void fractions FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 68
KWU-S Gamma Scan Benchmark Results EMF-2158(P)(A)
Measurements were performed for moderate void fractions FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 69
KWU-S Gamma Scan Benchmark Results EMF-2158(P)(A)
FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 70
KWU-S Gamma Scan Benchmark Results EMF-2158(P)(A)
Indication that the higher voids are accurately represented FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 71
KWU-S Gamma Scan Benchmark Results EMF-2158(P)(A)
Indication that the higher voids are accurately represented FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 72
KWU-S Gamma Scan Benchmark Results EMF-2158(P)(A)
Indication that the higher voids are accurately represented FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 73
KWU-S Gamma Scan Benchmark Results EMF-2158(P)(A)
Indication that the higher voids are accurately represented FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 74
Axial Void Distribution Uncertainty
> Conclusion Recent gamma scan data has confirmed the local power uncertainty There is no axial dependency in the uncertainty There is no void dependency in the local peaking power uncertainty Current uncertainties are applicable to Browns Ferry with power uprate conditions FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:002 75
Gamma Scan Description Ralph Grummer Manager, Core Physics Methods Richland, WA August 4, 2005 FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:003 1
Gamma Scans FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:003 2
Purpose
> Gamma scans have been used to measure the assembly and individual rod power distribution
> These measurements are used to validate core physics methods and determine the associated uncertainties FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:003 3
Gamma Scan Measurements
> Gamma scans measure the relative gamma flux resulting from isotopic decay
> Certain isotopes can be identified by gamma spectroscopy
> Power measurements target the gamma spectrum associated with La140
> La140 is a decay product of Ba140 which is direct fission product FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:003 4
Gamma Scan Measurements
> The half life of Ba140 is 12.8 days
> The half life of La140 is 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />
> La140 activity is therefore related to the density of Ba140
> The Ba140 density is representative of the integrated fissions over the last 25 days
> Gamma scan measurements need to be taken shortly after shutdown before the Ba140 decays to undetectable levels FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:003 5
Gamma Scan Equipment
> Equipment is tailored to the specific application Assembly scans use a broad window to capture gamma particles from all of the rods Individual rod scans use a narrow window to isolate the rod An axial level measurement uses a broader (axial) window to get a higher count rate Axial scans use a narrow (axial) window to get a finer resolution FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:003 6
Gamma Scan Equipment
> Gamma scan measurements are performed on individual fuel rods removed from assemblies using a high-purity germanium (HPGe) detector and an underwater collimator assembly FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:003 7
Gamma Scan Comparisons
> In order to compare core physics models to the gamma scan results the calculated pin power distribution is converted into a Ba140 density distribution A mathematical process using CASMO-4 pin nuclide inventory and MICROBURN-B2 nuclide inventory is used This is an additional uncertainty in the overall comparison FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:003 8
Power Distribution Uncertainties Gamma scanning provides data on relative local and radial power during last few weeks of operation Uncertainty in gamma scan results has small effect on measured radial power distribution uncertainty
- 50% decrease in correlation coefficient results in 0.4% increase in measured radial power distribution uncertainty
- Additional ATRIUM-10 gamma scan data would not significantly affect measured power distribution uncertainty Local gamma scan data available for various designs
- 11 assemblies in two reactors
- 7x7, 8x8, 9x9, ATRIUM-10
- Exposures include once and twice burned assemblies
- Various gadolinia concentrations
- Various water rod configurations No void dependence observed for local power uncertainties More ATRIUM-10 gamma scanning is not expected to change uncertainties No more ATRIUM-10 gamma scanning is necessary FRAMATOME ANP, INC. > CASMO-4/MICROBURN-B2 Methodology - August 4, 2005 - RGG:05:003 9
Safety Analysis Methodology Uncertainties Michael E. Garrett Manager, BWR Safety Analysis michael.garrett@framatome-anp.com (509) 375-8294 Richland, WA August 4, 2005 FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 1
Presentation Topics
> Treatment of Uncertainties in Safety Analysis Deterministic safety analysis approach
> Safety Limit MCPR (SLMCPR) Methodology Overview
> SLMCPR Sensitivity to Power Distribution Uncertainty Local power peaking Radial power peaking Axial power peaking FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 2
Treatment of Uncertainties in Safety Analysis
Safety Analysis Methodology Treatment of Uncertainties
> The MCPR safety limit methodology explicitly considers important uncertainties in the Mont Carlo calculations performed to determine the number of rods in boiling transition
> Other safety analysis methodologies do not explicitly account for uncertainties; deterministic, bounding approaches are used to ensure that all licensing criteria are satisfied FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 4
Safety Analysis Methodology Deterministic Approach
> Current deterministic methods are not best estimate Individual phenomena are not treated statistically
> Current methods provide conservative, bounding results
> Current methods have adequate conservatism to offset methodology uncertainties
> Conservatism incorporated in two ways Computer code models produce conservative results on an integral basis Important input parameters are conservatively bounding
> All conservatisms are additive and not statistically combined Assuming all parameters are bounding at the same time produces very conservative results FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 5
7 AREVA C j Analysis II' Examples of Analysis Conservadlsrn for Limiting Events
" * )logy Pressurization Events
> COTRANSA2 conservative prediction of Peach Bottom turbine trip tests r
- J
> Steady-state CPR correlation demonstrated to be conservative for transients (predicted dryout time occurs earlier than test data)
FRAMATOME ANP. INC. -
Safety Analysls NRC Presentation - August 4, 2005 6
Safety Analysis Methodology Examples of Analysis Conservatism for Limiting Events Pressurization Events (continued)
> The four steam lines are represented as a single, average steam line Accounting for differences causes the pressurization rate to be reduced
> Bounding scram insertion times (delay and insertion rate)
> All control blades assumed to insert at the same time and rate Control blades actually insert at a distribution of speeds Control blades faster than average provide more negative reactivity than lost by control blades slower than average
> All control rods assumed to be initially fully withdrawn (conservative for off-rated conditions and pre-EOC exposures)
FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 7
Safety Analysis Methodology Examples of Analysis Conservatism for Limiting Events Pressurization Events (continued)
> Conservative licensing basis step-through used for neutronics input More top-peaked axial power shape than design basis Longer cycle exposure than design basis
> Bounding setpoints (analytical limits) and delays used Reactor protection system Turbine protection system
> Bounding equipment performance assumed Turbine control and stop valve closure times RPT delay time Turbine bypass Safety and relief valves FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 8
Methodology Overview
SLMCPR Analysis Methodology
> The purpose of the safety limit MCPR (SLMCPR) is to protect the core from boiling transition (BT) during both normal operation and anticipated operational occurrences (transients)
> At least 99.9% of the rods in the core are expected to avoid BT when the minimum CPR during the transient is greater than the SLMCPR
> The SLMCPR is determined by a statistical convolution of uncertainties associated with the calculation of MCPR
> The SLMCPR analysis is performed each cycle using core and fuel design specific characteristics FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 10
SLMCPR Analysis Methodology Plant Limiting Plant Transient Transient Transient Methodology Analysis Delta-CPR Plant MCPR Initial Operating Conditions Limit Thermal Safety MCPR Hydraulic Limit Safety Critical Power Analysis Analysis Limit Methodology FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 11
Thermal Limits Methodology Average Core Conditions Design Peak Core Condition Allowed Operating Range Design Margin (5%-10%)
Operating Limit OLMCPR (1.38)
Transient Margin (DELTA-CPR)
Statistical Margin Defined Overheating MCPR = 1.00 Cladding Damage MCPR < 1.00 FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 12
SLMCPR Analysis Methodology Major Computer Codes Code Use MICROBURN-B2 Provides radial peaking factor and exposure for each bundle in the core and the core average axial power shape CASMO-4 Provides local peaking factor distribution for each fuel type XCOBRA Provides hydraulic demand curves for each fuel type SLPREP Automation code which obtains neutronic data from MICROBURN-B2 and CASMO-4 and prepares SAFLIM2 input SAFLIM2 Calculates the fraction of rods in boiling transition (BT) for a specified SLMCPR FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 13
SLMCPR Analysis Methodology FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 14
SLMCPR Analysis Methodology FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 15
SLMCPR Analysis Methodology Monte Carlo Technique
> A Monte Carlo analysis is a statistical technique to determine the distribution function of a parameter that is a function of random variables Each random variable is characterized by a mean, standard deviation, and distribution function A random value for each input variable is selected The parameter of interest is calculated using the random values for the input variables The process is repeated a large number of times to create a probability distribution for the parameter of interest FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 16
SLMCPR Analysis Methodology SAFLIM2 Computer Code Description SAFLIM2 is a computer code used to determine the number of fuel rods in the core expected to experience boiling transition for a specified core MCPR Use Evaluate the safety limit MCPR (SLMCPR) which ensures that at least 99.9% of the fuel rods in the core are expected to have a MCPR value greater than 1.0 Documentation EMF-2392(P), SAFLIM2 Theory, Programmers, and Users Manual Acceptability ANF-524(P)(A) Rev 2 and Supplements, ANF Critical Power Methodology for Boiling Water Reactors, November 1990 The safety evaluation by the NRC for the topical report approves the SAFLIM2 methodology for licensing applications FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 17
SAFLIM2 Computer Code Major Features
> Convolution of uncertainties via a Monte Carlo technique
> Consistent with POWERPLEX CMSS calculation of MCPR
> Appropriate critical power correlation used directly to determine if a rod is in boiling transition
> BT rods for all bundles in the core are summed
> Non-parametric tolerance limits used to determine the number of BT rods with 95% confidence
> Explicitly accounts for channel bow
> New fuel designs easily accommodated FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 18
SLMCPR Statistical Parameters FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 19
SAFLIM2 Computer Code Reactor System Uncertainties SFW Feedwater flow rate uncertainty. Obtained from NSSS vendor documentation or customer. A typical value is 1.8%
SFWT Feedwater temperature uncertainty. Obtained from NSSS vendor documentation or customer. A typical value is 0.8%
SP Core pressure uncertainty. Obtained from NSSS vendor documentation or customer. A typical value is 0.7%
SCG Total core flow rate uncertainty. Obtained from NSSS vendor documentation or customer. A typical value is 2.5%
FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 20
SAFLIM2 Computer Code Core Monitoring Uncertainties FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 21
SAFLIM2 Computer Code Fuel Design Uncertainties FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 22
SPCB Critical Power Correlation F-eff
> The SPCB correlation is a function of planar averaged fluid properties
> F-eff accounts for local power peaking as well as local flow and enthalpy effects on critical power
> F-eff consists of 2 components:
F-eff,o determined for each rod location based on local peaking distribution Additive constant (l) determined for each rod location from test measurements
> The F-eff for each rod in the bundle is the sum of the 2 components:
F-eff,i = F-eff,o + A
> The assembly F-eff is the maximum F-eff,i at the plane of interest F-eff = max (F-eff,i)
FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 23
SAFLIM2 Computer Code Calculation Procedure
> Initialization
> Monte Carlo Trials Core Calculations (Outer Loop)
Fuel Assembly Calculations (Inner Loop)
> Rods in BT Calculation FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 24
A AREVA SAFLIMZ Computer Code.
1 Initialization
> Establish initial (nominal) operating conditions at which the core MCPR equals the desired SLMCPR
> Initial conditions are required for the following parameters
+ Core flow
- Core pressure Feedwater temperature
+ Feedwater flow Core inlet enthalpy 1 Core power Assembly power (radial peaking)
, : Core average axial power shape Assembly flow FRAMATOME AMP. INC. Safety Analysis - NRC Presentation - August 4, 2005 25
SAFLL"? Ca uter Code lnitialization (continued)
SAFLIMZ Computer Code Core Calculations = Outer Loop
Assembly Calculatjons Inner Loop Safety Analysis - NRC Presentation - August 4,2005
SAFLlM2 Computer Code Fuel Rod Calculations Inner Loop Safely Analysis NRC Presentation - August 4,2005
AREVA SAFLIMZ Computer Code Number of Rods in BT FRAr.lATOP:IE APJ INC. Safety Analysls - NRC Presentation - August 4,2005 30
SAFLIM2 Computer Code Monte Carlo Trial FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 31
SLMCPR Sensitivity to Power Distribution Uncertainty
SLMCPR Sensitivity Power Distribution Uncertainty Topics
> Sensitivity to radial peaking factor (RPF) and local peaking factor (LPF) uncertainty Conservative range for potential changes in RPF and LPF uncertainties from additional gamma scan data were estimated LPF uncertainty: less than 1.5x current estimate RPF uncertainty: -0.3% to +0.4% change in current estimate
> Basis for not explicitly modeling axial power shape uncertainty FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 33
SLMCPR Sensitivity Local Peaking Factor (LPF) Uncertainty
> Sensitivity analyses performed using Browns Ferry equilibrium ATRIUM'-10 EPU core design
> LPF uncertainty increased by 1.5 multiplier
> SLMCPR lpf Rods in BT 1.08 1.48% 60 1.08 2.22% 62 1.0810 2.22% 60
> SLMCPR insensitive (+0.001) to 1.5x increase in LPF uncertainty Additional gamma scan data not expected to result in significant impact to SLMCPR FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 34
SLMCPR Sensitivity Radial Peaking Factor (RPF) Uncertainty
> Sensitivity analyses performed using Browns Ferry equilibrium ATRIUM'-10 EPU core design
> RPF uncertainty increased 0.4%
> SLMCPR rpf Rods in BT 1.08 4.6% 60 1.08 5.0% 71 1.0855 5.0% 60
> SLMCPR not very sensitive (+0.0055) to 0.4% increase in RPF uncertainty Additional gamma scan data not expected to result in significant impact to SLMCPR FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 35
A AREVA SLMCPR Methodology Axial Power Shape FRAF.1ATOr.lE ANP. INC safety Analysls - N K G rresenrauon - ~ u g u s r4,NU:,
SLMCPR Sensitivity Axial Power Shape Assessment for ATRIUM' -10
> Original methodology assessment performed for fuel designs without part-length fuel rods and with ANFB critical power correlation
> CHF tests indicated ATRIUM'-10 fuel more sensitive to axial power shape
> SLMCPR sensitivity to axial power was reassessed (1998)
> Three types of assessments performed Variations in core average axial power shape Use of assembly specific axial power shape for each assembly Perturbing power shape during Monte Carlo trials FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 37
SLMCPR Sensitivity Axial Power Shape Assessment for ATRIUM' -10 (continued)
> Variation in core average axial power shape Range of core average axial power shapes obtained from a core design analysis SLMCPR analysis performed for each axial shape with all other input parameters held constant Variation observed in BT rods typical of Monte Carlo process; no trend with changes in axial power shape Number of rods in BT is not sensitive to changes in core average axial power shape FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 38
Sensitivity to Axial Power Profile Sensitivity to Axial Power Profile Radials & Locals From to Exposure with AO = -23%
45 44 43 42 number of rods in BT 41 40 39 38 37 36 35 34 33 32 31 30
-0.35 -0.3 -0.25 -0.2 -0.15 -0.1 -0.05 0 0.05 0.1 0.15 Axial Offset of Power FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 39
SLMCPR Sensitivity Axial Power Shape Assessment for ATRIUM-10 (continued)
> Use of assembly-specific axial power shape Special code version developed with capability to model a different axial power shape for each assembly Axial power distribution obtained from cycle design step-through for each assembly in the core Rods in BT calculated for each bundle based on bundle-specific axial power shape Number of rods in BT is not sensitive to the use of core average or bundle-specific power distribution FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 40
Sensitivity of Modeling Assembly-Specific Axial Power Profile Number of Rods in BT (maximum from all exposures)
Core Flow Mlb/hr Core Average Assembly-Specific Axial Power Axial Power 108 35 34 70 43 46
Conclusion:
Results are within normal variation for Monte Carlo results FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 41
SLMCPR Sensitivity Axial Power Shape Assessment for ATRIUM'-10 (continued)
> Perturbing power shape during Monte Carlo trials Special code version developed with capability to perturb the core average axial power shape during each Monte Carlo trial The code used a process to adjust the initial axial power shape to produce a power shape with a different axial offset Axial power uncertainty reported in the MICROBURN-B2 topical report is 1.8% for C-lattice and 2.9% for D-lattice Analyses performed assuming an axial power offset uncertainty of 3%
Results showed little variation in the number of rods in BT Number of rods in BT is not sensitive to perturbing the axial power shape in Monte Carlo trials FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 42
Change Axial During Monte Carlo Trials Bottom Peak, Axial Offset Uncertainty 0.03 1.80 1.60 1.40 Axial Peaking Factor 1.20 1.00 0.80 0.60 0.40 0.20 0.00 0.0 0.2 0.4 0.6 0.8 1.0 Axial Position, normalized FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 43
Change Axial During Monte Carlo Trials Mid Peak, Axial Offset Uncertainty 0.03 1.40 1.20 1.00 Axial Peaking Factor 0.80 0.60 0.40 0.20 0.00 0.0 0.2 0.4 0.6 0.8 1.0 Axial Position, normalized FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 44
Change Axial During Monte Carlo Trials Top Peak, Axial Offset Uncertainty 0.03 1.80 1.60 1.40 Axial Peaking Factor 1.20 1.00 0.80 0.60 0.40 0.20 0.00 0.0 0.2 0.4 0.6 0.8 1.0 Axial Position, normalized FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 45
Sensitivity to Changing Axial Power During Monte Carlo Trials Rods in BT Axial Power Shape Constant Core Perturb Core Average Axial Average Axial Bottom peak 27 29 Middle peak 22 22 Top peak 19 17
Conclusion:
Results are within normal variation for Monte Carlo results FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 46
SLMCPR Sensitivity Axial Power Shape Assessment for ATRIUM' -10 (continued)
Conclusion from 1998 assessment
> SLMCPR methodology remains insensitive to axial power shape and axial power shape uncertainty
> Approved methodology is applicable for ATRIUM' -10 fuel FRAMATOME ANP, INC. Safety Analysis - NRC Presentation - August 4, 2005 47
Bypass Modeling Ralph Grummer Manager, Core Physics Methods Ralph.Grummer@framatome-anp.com (509) 375-8427 Richland, WA August 4, 2005 FRAMATOME ANP, INC. Bypass Modeling - August 4, 2005
Modeling Voiding in the bypass
> Objective Demonstrate that anticipated boiling in the bypass does not impact the safety margins Demonstrate that the Framatome-ANP Methodology is Applicable to EPU conditions at Browns Ferry FRAMATOME ANP, INC. Bypass Modeling - August 4, 2005 2
Modeling Voiding in the bypass
> Calculations for Browns Ferry with Power Uprate do not indicate boiling in the bypass at rated power conditions With single lumped bypass channel With multi-channel bypass and explicit water rod models
> Browns Ferry has 10% more inlet subcooling than similar plants due to lower feedwater temperature FRAMATOME ANP, INC. Bypass Modeling - August 4, 2005 3
Multi-Channel Bypass Model
> In order to evaluate the effect of voiding in the bypass a theoretical case was developed
> Voiding in the bypass was forced to 5% voids by decreasing the inlet sub-cooling from 27.15 to 15 BTU/lbm
> The multi-channel bypass produces conservative results Multi-channel bypass model is an independent flow path for each assembly The boundary condition is equal pressure drop from inlet to exit No cross flow between bypass channels Heat deposition based upon single assembly No Gamma smearing FRAMATOME ANP, INC. Bypass Modeling - August 4, 2005 4
A AREVA Bypass Void Distribution 1 EDIT OF AXIALLY AVERAGED VOID FRACTION IN BYPASS CHANNEL IN UNITS OF %
FRAf.1ATOPJE ANP INC. -
Bypass Modeling August 4.2005 5
Bypass Void Distribution EDIT OF VOID FRACTION IN BYPASS CHANNEL IN UNITS OF %
Multi Channel Single Channel r
Axial Level Core Average Peak Assembly TOP 1 2
3 Bottom Bypass Modeling - August 4. 2005 6
Modeling Voiding in the bypass There is a minimal change in the power distribution of the peak assembly FRAMATOME ANP, INC. Bypass Modeling - August 4, 2005 7
Voiding in the Bypass
> Conclusion Boiling in the bypass is not expected at rated power with power uprate conditions The effects of boiling in the bypass, should it occur are very small with exit void fraction of 5%
Voiding in the bypass has a negligible impact on the LPRM instrumentation as the void fraction is near 1% at the top most LPRM FRAMATOME ANP, INC. Bypass Modeling - August 4, 2005 8