ML110140233

Licensee Slides from 1/13/11 Meeting, Spent Fuel Pool Criticality Analysis

ML110140233
Person / Time
Site:	Fort Calhoun
Issue date:	01/13/2011
From:	Omaha Public Power District
To:	Office of Nuclear Reactor Regulation
Wilkins, L E, NRR/DORL/LPL4, 415-1377
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ML110140223	List: ML110140233
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Download: ML110140233 (44)

Text

Fort Calhoun Station Extended Power Uprate (EPU)

Project Spent Fuel Pool Criticality Analysis January 13, 2011 12/13/2010 1

Introductions

• Bob Dulee EPU NSSS/Licensing Manager

• Carmen Ovici EPU NSSS/Licensing Supervisor

• Steve Callis EPU Senior Licensing Engineer

• Bill Hansher FCS Licensing Supervisor

• Bob Schomaker AREVA Project Manager

• Linda Farrell AREVA Principal Engineer

• Jeremy Dean AREVA Supervisory Engineer 12/13/2010 2

Agenda

• EPU Overview - Bob Dulee

• EPU Impact on Spent Fuel Pool Criticality A l i - Linda Analysis Li d FFarrellll

• Summary and Conclusions - Linda Farrell

• Questions

• Follow-up Actions 12/13/2010 3

Project Overview

• 17% EPU: 1500 MWt to 1755 MWt 80 MWe

• Engineering & Licensing work in progress

- AREVA, AREVA S&L and d WEC

• Submit LAR to NRC: March 31, 2011

• Plant Modifications: 2011/2012 RFOs

• Implement EPU: Post-2012 RFO

Fort Calhoun Spent Fuel Pool Criticality Analysis Linda M. Farrell, AREVA NP

Purpose

• Allow fresh fuel of up to 5 w/o to be stored in the spent fuel racks at the Fort Calhoun Nuclear Station

• Support Extended Power Uprate (EPU) submittal 12/13/2010 6

Analysis Overview

• Methodology is similar to previous AREVA analyses

• All topics listed in DSS-ISG-2010-01 addressed

• Analysis uses

- Soluble boron credit of 500 ppm at all times

- Burnup credit

- Equivalencing of CASMO-3 lumped fission products

- Existing burnup curve from Technical Specification

• CASMO-3 for in-core depletion

• KENO-V.a for all keff and tolerance calculations 12/13/2010 7

Fuel Assembly Selection

• Bounding Fuel Assembly Description

- 14 x 14 fuel assembly

- Nominal planar average enrichment of 5 wt% U-235 (4.95 (4 95 +/

+/- 0.05 0 05 wt%)

• 10CFR50.68(b)(7)

- No axial blankets or cutback regions

- No axial or radial enrichment zoning

- No integral burnable neutron absorber 12/13/2010 8

Fuel Assembly Selection

• Bounds all existing fuel in the Spent Fuel Pool

- Currently approved for 4.5 wt%

- Verification V ifi i off llegacy ffuell 12/13/2010 9

Depletion Analysis Overview

• In-Reactor depletion analysis

- Used NUREG/CR-6801 burnup profiles

- Used CASMO-3 to generate isotopic number densities for burnup credit

- Treatment of CASMO-3 Lumped Fission Product

- Transferred isotopic number densities to KENO-V.a for keff calculations 12/13/2010 10

Depletion Analysis Depletion Uncertainty

• Kopp 5% reactivity decrement to account for uncertainty in number densities generated by CASMO 3 CASMO-3

- 5% of difference between fresh fuel and burnup of interest

- Calculated based on values generated by KENO-V.a 12/13/2010 11

Depletion Analysis Depletion Uncertainty 12/13/2010 12

Depletion Analysis Reactor Parameters

• Values determined for 10 axial nodes and an average assembly

• Used CASMO-3

• Parameters chosen to maximize reactivity of depleted fuel assembly 12/13/2010 13

Depletion Analysis Reactor Parameters

- Fuel Temperature

• Selected to increase Pu production

- Moderator Temperature node specific

• Harden spectrum and increase Pu production

- Soluble boron concentration (cycle average)

- Core Power - (EPU condition)

- Operating History

- No fixed/integral burnable absorbers modeled 12/13/2010 14

Isotopic Number Densities

• Isotope atom densities in both CASMO-3 and KENO-V.a libraries easily transferred

• The short lived isotopes are decayed prior to input in KENO-V.a 12/13/2010 15

Lumped Fission Products

• Isotopes represented by LFP (ORNL-TM-1658) 46 Non-Saturating (401)

Ge-76 Se-78 Se-80 Br-81 Kr-84 Kr-85 Rb-85 Kr-86 Rb-87 Y-89 Sr-90 Zr-91 Zr-92 Zr-93 Zr-94 Zr-96 Mo-97 Mo-98 Mo-100 Ru-101 Ru-102 R 104 Ru-104 R 106 Ru-106 Pd 110 Pd-110 Cd 111 Cd-111 Cd 112 Cd-112 Cd 114 Cd-114 Cd 116 Cd-116 Sn-117 Sn-119 Sn-120 Sn-122 Te-128 Te-130 Xe-132 Xe-134 Xe-136 Cs-137 Ba-138 Ce-140 Nd-144 Nd-148 Nd-150 Sm-154 Gd-156 Gd-158 15 Slowly-Saturating (402)

Se-77 Se-79 Mo-95 Tc-99 Pd-107 Pd-108 In-115 Sb-121 Sb-123 I-127 I-129 La-139 Pr-141 Nd-145 Tb-159 12/13/2010 16

Lumped Fission Products

• Transfer of LFP isotopic concentrations from CASMO-3 to KENO-V.a

• For the LFP, CASMO-3 output lists

- Two-group fluxes

- Cross sections

- Number density

• Determined isotope that has an absorption rate ratio similar to the LFP fast-to-thermal absorption rate ratio

• Determined an equivalent isotopic concentration for input to KENO-V.a 12/13/2010 17

Equivalencing

• Total absorption cross section is defined as (1

• a1 + 2

• a2 )

a - total =

(1 + 2 )

( 1 / 2 )

• a1 + a2

=

(1 + 2 ) / 2 Where:

1 - group 1 flux 2 - group 2 flux a1, a2, - absorption cross sections for groups 1 and 2 for each isotope 12/13/2010 18

Equivalencing

• Derive an equivalent amount for the two pseudo-isotopes by deriving an equivalent macroscopic cross section X eq

• a eq = X 401

• a 401 + X 402

• a 402 or (X 401

• a 401 + X 402

• a 402 )

X eq =

a eq Where:

X - atomic number density for each isotope 12/13/2010 19

Equivalencing

• Equivalent isotopic concentration is calculated by X 401 * ((1/ 2 )

• 1-401 + 2-401 ) + X 402 * ((1/ 2 )

• 1-402 + 2-402 )

X eq =

(1/ 2 )

• 1-orig + 2-orig

• This value is then added to the concentration determined by CASMO-3 and the total used in KENO-V.a calculations

• This approach yields conservative results within CASMO-3 calculations, and is thus appropriate

- Compared LFP kinfinity to Equivalent kinfinity 12/13/2010 20

Depletion Analysis Burnable Absorbers & Rodded Operation

• As previously discussed 12/13/2010 21

Criticality Analysis Axial Burnup Profile

• Used NUREG/CR-6801

• Used axial burnup shapes for all burnups

• Reviewed axial burnup shapes from 7 cycles of operating and planned future cycles

- Included one case at 3.5 wt% and 24.1 GWD/MTU to verify axial burnup shape is conservative and appropriate in 10-30 GWD/MTU range

• Burnup shapes from NUREG/CR-6801 are valid 12/13/2010 22

Criticality Analysis Rack Model 12/13/2010 23

Criticality Analysis Rack Model Three Regions:

- Region 1 of-4 loading U-235 enrichments less than or equal to 5.0 wt% with no burnup credit

- Region 2 of-4 loading U-235 enrichments less than or equal to 5.0 wt% and burnup credit

- Region 2 peripheral cells of-4 loading U-235 enrichments less than or equal to 5.0 wt% and (lower) burnup credit

- Region 2 and Region 2 peripheral cells dimensions are effectively identical, the peripheral cells do not have neutron absorber between the assembly and the concrete wall

- Region 2 peripheral established due to gamma heating analysis 12/13/2010 24

Criticality Analysis Rack Model

• No degradation of installed neutron absorber (Boral)

- Nominal B B-10 0 0151 g/cm2 10 loading = 0.0151

- Modeled B-10 loading = 0.014 g/cm2

• Fuel extends above Boral plate by 1.3 12/13/2010 25

Criticality Analysis Rack Model

• The K95/95 is calculated using:

K95/95 = keff+biasm+ksys+[C2(k2+m2+sys2+tol2)+ktol2]1/2+kBu

- keff= the KENO-V.a KENO V a calculated result

- biasm= the bias associated with the calculation methodology

- ksys= summation of k values associated with the variation of system and base case modeling parameters

- C= confidence multiplier based upon the number of benchmark cases

- k, m, sys= standard deviation of the calculated keff, methodology bias, and system

- ktol. tol= statistical combination and standard deviation of statistically independent k values due to manufacturing tolerances

- kBu= Burnup credit penalty to account for depletion reactivity uncertainty 12/13/2010 26

Criticality Analysis Interfaces

• Interaction effects between Regions 1 and 2 evaluated

- All cases run with 5 wt% U-235 and 4-of-4 loading pattern

- Evaluated at both 0 ppm and 500 ppm soluble boron

- Both 7/8 and 2 separation distances evaluated

• If keff increases above the maximum of the two individual regions, then there is an interaction effect

• No interaction effect was demonstrated 12/13/2010 27

Criticality Analysis Normal Conditions

• System Effects

• Fuel Tolerance Effects

- Moderator Temperature - Theoretical Density

- Rack-to-rack interactions - Pellet OD

- Non Fuel Bearing Components - Clad Inner Dimension

- Cl d Outer Clad O Di Dimension i

• Rack Tolerance Effects - Guide Tube Dimension, ID

- Centered to Off-Centered - Guide Tube Dimension, OD Assembly in Fuel Cell - Rod Pitch

- Cell Inner Dimension - Active Fuel Height

- Absorber Thickness

- Absorber Width

- Region 1 Cell Pitch x & y

- Region 2 Cell Pitch x & y 12/13/2010 28

Criticality Analysis Accident Conditions

• Used 500 ppm soluble boron

• Analyzed conditions

- Drop of a fresh fuel assembly outside the rack

- Misload of a fresh assembly into unapproved location

- Boron dilution

- Straight deep drop accident 12/13/2010 29

Criticality Code Validation Introduction

• NUREG/CR-6698 general methodology utilized

• Benchmark configurations

- 100 International Handbook configurations

- 145 actinide HTC configurations (NUREG/CR-6979)

- 28 fission product configuration from International Handbook

• Complete trend analysis

• Statistical treatment

• Lumped Fission Products

• No code-to-code validation necessary 12/13/2010 30

Criticality Code Validation Area of Applicability

• Experiments fully cover the range of the Fort Calhoun system, thus no extrapolation required

• The most significant physical parameters affecting criticality:

- the fuel enrichment

- the absorber materials

- the lattice spacing

- Other parameters have a smaller effect but have also been included in the analysis

• Sufficient number of experiments to be statistically significant 12/13/2010 31

Criticality Code Validation Area of Applicability 12/13/2010 32

Criticality Code Validation Trend Analysis

• Results analyzed to identify any trends in the bias

• Linear regression analysis used

• Trending Parameters:

12/13/2010 33

Criticality Code Validation Trend Analysis

• Analysis of data

- Weighted and non-weighted linear regression analysis

- Goodness-of-fit tests

• Coefficient of determination

• Students T-distribution

• Test residuals of the regression for trends indicated by Students T-distribution

• Results of the trending analysis

- Very small slopes with no statistical validity, with the exception of the non-weighted trend for fissile isotopic content

- Single-sided lower tolerance band used to establish the bias and uncertainty for fissile isotopic content

- Lower tolerance limit (KL) is conservative 12/13/2010 34

Criticality Code Validation Trend Analysis

• Removal of 11 MOX cases analyzed to determine if they influence bias in non-conservative direction

- Inclusion of cases determined to be insignificant 12/13/2010 35

Criticality Code Validation Statistical Treatment

• Performed two separate trending analysis/bias calculations

- 100 Benchmark Cases

- 173 cases (HTC actinides ti id and d fi fission i products) d t)

• Bias and bias uncertainty established

- Bias uncertainty uses variation of population about the mean

• Used 95/95 single-sided tolerance limits for confidence factor 12/13/2010 36

Criticality Code Confirmation Lumped Fission Products

• Replacement of lumped fission product with equivalent

• Cross-code comparison of CASMO-3/KENO-V.a using equivalencing show method is acceptable 5.00 wt%, 42.3 GWd/mtU 4.20 wt%, 32.161 GWd/mtU CASMO-3: 0.90972 0.92371 KENO-V.a: 0.9109 +/- 0.0004 0.9230 +/- 0.0004

- Note: KENO-V.a is 3D versus CASMO-3 2D

- KENO-V.a incorporates axial leakage and reflection, CASMO-3 does not 12/13/2010 37

Criticality Code Validation Summary

• The bias and its uncertainty

- Used 95/95 weighted single-sided tolerance limit

- Used methodology presented in NUREG/CR-6698

- Took into account the possible trending of keff

• These results support the criticality analysis of the Fort Calhoun spent fuel pool.

- Equivalencing method supported by both equivalent cross-section data and CASMO/KENO-V.a comparisons 12/13/2010 38

Additional Information

• Spacer grid modeling

- Analyzed for including versus not including spacer grids

- Four different soluble boron concentrations (0, 850, 1700, 2500)

- Overall results are statistically insignificant

- Acceptable to not model the spacer grids 12/13/2010 39

Additional Information

• Geometry changes during irradiation

• Results show negative k - no additional uncertainty is warranted 12/13/2010 40

Conclusions

• All acceptance requirements are met

• No change to burnup loading curve in existing Technical Specification

- 500 ppm soluble boron during normal conditions

• Very simple loading pattern of-4 for all storage locations 12/13/2010 41

Conclusions

• Margin to limits

- 0.0069 k margin for the boron dilution events

- 0.0028 k margin for misload conditions at 500 ppm of boron

• Analysis meets requirements for NRC Acceptance for Review 12/13/2010 42

Questions?

Meeting Follow-up Actions