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Holtec International Report No. HI-2125245, Revision 5, Licensing Report for Quad Cities Criticality Analysis for Inserts.
	ML14101A228
Person / Time
Site:	Quad Cities
Issue date:	04/07/2014
From:	; Holtec
To:	; Exelon Generation Co, Office of Nuclear Reactor Regulation
Shared Package
ML14101A213	List: ML14101A227; ML14101A228; RS-14-100, Additional Information Regarding License Amendment Request Associated with Use of Neutron Absorbing Inserts in Spent Fuel Pool Storage Racks;
References
	HI-2125245, Rev. 5
	Download: ML14101A228 (128)
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ATTACHMENT 6 Holtec International Report No. HI-2125245, Revision 5,"Licensing Report for Quad Cities Criticality Analysis for Inserts -Non Proprietary Version" HOLT INTERNATI I EC ONAL-~ -Holtec Center, 555 Lincoln Drive West,_Marlton, NJ 08053 Telephone (856) 797- 0900 Fax (856) 797 -0909 Licensing Report for Quad Cities Criticality Analysis for Inserts -Non Proprietary Version FOR Exelon Holtec Report No: HI-2125245 Holtec Project No: 2127 Sponsoring Holtec Division:

HTS Report Class : SAFETY RELATED mý Summary of Revisions:

Revision 0: Original Issue Revision 1: Supplement I was added to cover a new revision of NETCO-SNAP-INO rack insert.Revision 2: All Revision I revision bars were removed. No other changes were made.Revision 3: Sections 2.3.8, 2.7, 7.6, 8.0 and Appendix B were revised. All changes were marked by revision bars.Revision 4: Minor editorial changes to page 4 description of Table 7.1(c) and Table 2.1(c) was move up one line. Neither change marked by revision bar. All Revision 3 revision bars removed.Revision 5: All Revision 4 changes accepted.

Changes have been made in response to NRC RAIs, these changes are marked by revision bars. The results of the analysis have changed because the super lattice has been removed. The following sections have been updated: 2.3.1.5.1, 2.3.8, 2.6, 2.7 (deleted), 5.1, 7.1.10, 7.6 (deleted), 8, 13.1, B.3 and SI.9. Note that Sections 2.8, 2.9 and 7.7 have been renumbered to 2.7, 2.8 and 7.6. The new Section 7.6 has been updated to reflect the changes in table numbers. The following tables have been updated: 5.2(a), 7.11, 7.13(a), 7.14, 7.15 (deleted), 7.16 (deleted), B.1, S1-5 and SI-6. Note that Table 7.17 has been renumbered Table 7.15 and Table 7.18 has been renumbered TFable 7.16. The tables in Supplement I have also all had the filenames removed. The following figures have been update: 7.2 (deleted), 7.3 (deleted).

Figures B.1 through B.8 have been replaced by new Figures B. I through B.3 in revision 5 only, Figures B.4 through B.8 have been deleted.Project No. 2127 Report No. H1-2125245 Page i Table of Contents 1. INTRODUCTION

.............................................................................................................................

9 2. M ETHODOLOGY

..........................................................................................................................

10 2.1 GENERAL APPROACH ..................................................................................................................

10 2.2 COMPUTER COIES AND CROSS SE.CTION LIBRARIES

..............................................................

10 2 .2 .1 M C N P 5 -1 .5 1 .......................................................................................................................

1 0 2 .2 .1.1 M C N 1 5-1.5 1 V alidalio n ........................................................................................................................................

10 2 .2 .1.1.1 U...................

..................................................

i f 2.2.2 CASM O-4 ......................................................................

...........

...............

12 2.3 ANALYSIS M ETIODS ..........................................................

..............................

12 2.3.1 Design Basis Fuel Assembly ................................................

12 2 .3 .1 .1 P eak R eactiv ity ......................................................................................................................................................

13 2.3.1.1.1 Peak Reacivit ad l'ucl Asscmbly Burnup ..........................................................

13 2.3.1.1.2!

...........................................

13 2.3.1.2 14 2.3.1.3 l)etermination of the Design Basis Fuel Assembly Lattice ............................................................................

15 2.3.1.4 Optima2 CASMO-4 Model Simplification Effect ..........................................................................................

15 2.3.1.5 C ore O peraling Param eters ..................................................................................................................................

17 2.3.1.5.1 R eactor P ow er U prate. ......................................................................................................................................

17 2.3.1.5.2 Integral R eactivity C ontrol D evices ..............................................................................................................

18 2.3.1.5.3 AxMjul adncl Planar Enrichment Variations

..............................................................................................

18 2.3.1.5.4 Fuel A sse bly D -C hanneling

...............................................................................................................

18 2 .3 .1.6 19.......................................................................

........ ........ 19 2.3.2 Reactivity Effect of Spent Fuel Pool Water Temperature

.................................................

19 2.3.3 Fuel Depletion Calculation Uncertainty

.........................................................................

20 2.3.4 Fuel and Storage Rack Manufacturing Tolerances

..........................................................

21 2.3.4 .1 Fuel M anufacturing T olerancces

............................................................................................................................

2 1 2.3.4.2 SFP Storage Rack Manufacturing Tolerances

.................................................................................................

22 2 .3.5 R ad ia l P osition ing ...............................................................................................................

2 2 2.3.5.1 Fuel Assembly Orientation in the Core ................................................................................................................

22 2.3.5.2 Fuel R adial Positioning in the R ack ......................................................................................................................

22 2.3.5.3 Inserts R adial Positioning

......................................................................................................................................

24 2.3.5.4 Fuel O rientation in S F1 R ack C ell ........................................................................................................................

24 2.3.6 .etC.o..........a.y.................................................................

25 2.3.6.18 Ma=imum k a.................

.............

........................................................

25 2.3.64 M ARG ..... ...........

.......................................................................................

26 2.3FUE2 MN...... .............

26 2. 3. in6e' t COup ON Me srmn.n etit....y...........................................................

.....................................

26 2.3.8 Maxim m k~f Calclatio for ormalCondiions......

..... ..... ..... ...2 2.4A6.2 TeG peraLuATIOn

....... ................................................................

.........

2 2.5 .7 F e Copoe n As Ne sPECT AintD ..CONStIt U O ....P..........

E N..S....................................................

26 2.6 .8 MaCC iEN AsONDIlTON.

erti an to. aNorm a g Condition

...............................................................

28 2.6.4 Ste pra tre E nL I Water Dens........................................................................................................

29 2.6.2 iDropped Asse .....embly ..or i.z.o tal... ...................................

bl.........................................

29 2.6.2 Drsop ated F e Assembly -.........l..........................................................................................

29 2.6.3 Dropped Assembly -Vertical into a Storage Cell.. ............................................

29 2.6.4 Storage Cell Distortion

.... .......................................................................

30 2.6.5 M isloaded Fuel Assemibly/Missving Insert ...................................................

.... 30 2.6.6 Mislocated Fuel Assembly.........................................................................

31 2.6.6.1 Mislocation of'a Fuel Assembly in the Water Gap between the Racks and Pool Wall .................................

31 2.6.6.2 Mislocation of a Fuel Assembly in the Corner between Two Racks .............................................................

31 2.6.6.3 Mislocalion of a Fuel Assembly between the SFP Rack and the Inspection Platform .................................

31 2.6. 7 Mis-installment of an Insert on Wrong Side of a Cell ..........................................................

32 Project No. 2127 Repoit No. 1-11-2125245 Page 1

2.6.8 Insert

M echanical Wear ..................................................................................................

32 2.6.9 Rack M ovement .................................................................................

.......................

32 2.7 SPENT FUEL RACK INTERFACES

.............................................................................................

33 2.8 RECONS'TI'UTED FUEL ASSEMBI3 ES ........................................................................................

33 3. ACCEPTANCE CRITERIA ...........................................................................................................

34 3.1 APPLICABLE CODES, STANDARDS AND GUIDANCE'S

............................................................

34 4. ASSUM PTIONS ..............................................................................................................................

35 5. INPUT DATA ..................................................................................................................................

36 5.1 FUiEL ASSEMBIY SPECIFICATION

.............................

.................................................

36 5.2 REACTOR PARAMETERS

..............................................................................

........ 36 5.3 SPENT POOl. PARAMEI1 RS ............................................................................................

36 5.4 STORAGE RACK SPElC1FCATION

..........

....................................................

37 5.4.1 M aterial Composi/ions

.................................................................................................

37 6. COM PUTER CODES ....................................................................................................................

38 7. ANALYSIS .......................................................................................................................................

39 7.1 DESIGN BASIS AND LJNCER'TAINTIY EVAIAJAT1ONS

..................................................................

39 7 .1 .1 ...............................................

3 9 7.1.2 Determination of the Design Basis Fuel Assembly Lattice .............................................

39 7.1.2.1 Fuel A ssem bly D e-C hanneling

.............................................................................................................................

39 7.1.3 Optima2 CASM O-4 M odel Simplification Effect .............................................................

39 7.1.4 Core Operating Parameters

...........................................................................................

40 7 .1.4 .1 R eactor P o w er U prate ............................................................................................................................................

4 0 7. 1.5 Water Temperature and Density EI"ffect ............................................................................

40 7.1.6 Depletion Uncertainty

..........

...............................

...........

40 7.1.7 Fuel and Rack Manufacturing Tolerances

.............................

......41 7.1.7.1 Fuel A ssem bly T olerances

...................................................................................................................................

4 1 7.1.7 .2 S F P R ack T olerances

.............................................................................................................................................

41 7.1.8 Radial Positioning

.................................................................................................................

41 7.1.8.1 Fuel Assembly Radial Positioning in SFI- Rack .............................................................................................

41 7.1.8.2 Fuel O rientation in SF R ack ................................................................................................................................

4 1 7.1.9 FuelRod Geometry Change ...........................................................................................

42 7 .1.9 .1 .. .........................................................................................................

..4 2 7.1.9.2....

...........

I..........................................................................................

42 7.1.10 ..............

I.... ... .........

42 7.2 M AXIMUM K,,;, CALCULATIONS FOR NORMAL CONDITIONS

..................................................

42 7.3 M ARGIN EVALUATION

................................................................................................................

42 7.4 ABNORMAL AND ACCIDENT CONDITIONS

...............................................................................

43 7.5 MAXIMUM Kar CALCULATIONS FOR ABNORMAL AND ACCIDFNT CCONDITIONS

....................

43 7.6 SPENT FUEL RACK INTERFACES

.............................................................................................

43 8. CONCLUSION

................................................................................................................................

44 9. REFERENCES

................................................................................................................................

45 Project No. 2127 Report No. [-11-2125245 Page 2 Supplement 1: Additional Calculations to Support the Revised NETCO-SNAP-IN Rack Insert D esign ......................................................................................................

S I-I Project No. 2127 Report No. 111-2125245 Page 3 List of Tables Description Table"able 7.2(a)Table 7.2(b)Table 7.3'Fable 7.4 Table 7.5 Table 7.6(a)Page 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 I 65 I 67 I 68 I 69 70 71 72 73 74 Results of the MCNP5-1.51 Calculations for SVEA-96 Optima2 Q122 Lattices Results of the MCNP5-1.51 Calculations for GEl4 Lattice Type 5 Results of the MCNP5-1.51 Calculations for Design Basis and Simplified Model of SVEA-96 Optima2 Q122 Lattice Type 146 Results of the MCNP5-1.51 Calculations for Core Operating Parameters Results of the MCNP5-1.51 Calculations for the Effect of Water Temperature and Density Results of the MCNP5-1.51 Calculations for the Depletion Uncertainty Table 7.7 Table 7.8 Results of the MCNP5-1.51 Calculations for Fuel Tolerances Results of the MCNP5-1.51 Calculations for Rack Tolerances Project No. 2127 Report No. 1-11-2125245 Page 4 Table Table 7.9(a)Table 7.9(b)Table 7.12(a)Table 7.12(b)Table 7.12(c)Table 7.13(b)T'able 7.15 Description Results of the MCNP5-1.51 Calculations for Fuel Radial Positioning in SFP Racks Results of the MCNP5-1.51 Calculations for Fuel Orientation in SFP Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate the Effect of Nominal Values Instead of Using Minimum B4C Loading and Minimum Insert Thickness on Reactivity Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate the Effect of the Actual Optima2 Q122 Fuel Assembly Margin Evaluation Summary of the Margin Evaluation Results of the MCNP5-1.51 Calculations for the Empty Storage Rack 01-.1 withllit lnQ1.rt Results of the MCNP5-1.51 Calculations for SFR Interface Page 75 76 77 78 79 80 81 82 83 84 I 85 86 U U U U l 1 U l l ProJect No. 2127 Report No. Ht-2125245 Page 5 Table Table SI-I Table S 1-2 Table S1-3 Table S 1-4 Fuel Rack Insert Revised Dimensions Results of the MCNP5 Calculations for Revised Rack Tolerances Results of the MCNP5-1.51 Calculations for Revised Fuel Radial Positioning in SFP Racks Results of the MCNP5- 1.51 Calculations for Revised Fuel Orientation

.-nrýn " r --Page SI-5 SI-6 SI-7 SI-8 S1-9 SI-10 Project No. 2127 Report No. HI1-2125245 Page 6 List of Figures Description Figure Page 87 88 89 90 91 92 93 94 95 96 97 98 [99 I Project No. 2127 Report No. HI-2125245 Page 7 Figure Description Page 100 101 102 103 104 105 106 107 108 109 110 SI-11 Project No. 2127 Report No. 1H1-2125245 Page 8

1. INTRODUCTION This report documents the criticality safety evaluation for the storage of spent BWR fuel in the Unit 1 and Unit 2 spent fuel pools (SFPs) at Quad Cities Station operated by Exelon. The Unit I and Unit 2 SFP racks are identical and are designed to accommodate BWR fuel. Currently, the SFP racks credit BORAFLEX for reactivity control. This new analysis will not credit the BORAFLEX but will instead credit new NETCO-SNAP-IN rack inserts, which are new to Quad Cities but not new relative to their use for spent fuel pool reactivity control. This analysis will demonstrate that with credit for the inserts the effective neutron multiplication factor (krff) in the SFP racks fully loaded with fuel of the highest anticipated reactivity, at a temperature corresponding to the highest reactivity, is less than 0.95 with a 95% probability at a 95%confidence level. Reactivity effects of abnormal and accident conditions are also evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit.Criticality control in the SFP, as credited in this analysis, relies on the following:

Fixed neutron absorbers o NETCO-SNAP-IN rack inserts in SFP rack cells" Integrated neutron absorbers o Gadolinium (Gd) in the fuel (peak reactivity isotopic composition).

Criticality control in the SFP, as credited in this analysis, does not rely on the following:

Burnup credit* BORAFLEX.Prqject No. 2127 Report No. HI-2125245 Page 9

2. METHODOLOGY

2.1 General

Approach The analysis is performed consistent with regulatory requirements and guidance.

The calculations are performed using either the worst case bounding approach or the statistical analysis approach with respect to the various calculation parameters.

The approach considered for each parameter is discussed below.2.2 Computer Codes and Cross Section Libraries 2.2.1 MCNP5-1.51 MCNP5-1.51 is a three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory

[1]. MCNP5-l.51 calculations use continuous energy cross-section data based on ENDF/B-VII.

MCNP is selected because it has history of successful use in fuel storage criticality analyses and has most of the necessary features (except for fuel depletion analysis) for the analysis to be performed for Quad Cities Station SFP.The convergence of a Monte Carlo criticality problem is sensitive to the following parameters:

(1) number of histories per cycle, (2) the number of cycles skipped before averaging, (3) the total number of cycles and (4) the initial source distribution.

All MCNP5 calculations are performed with a minimum of 12,000 histories per cycle, a minimum of 150 skipped cycles before averaging, and a minimum of 150 cycles that aource is specified as iinifnrm nv.nr thp. filIPdi rPoannz (.ernrmhi-p')

2.2.1.1 MCNP5-1.51 Validation ProJect No. 2127 Report No. 1I-1-2125245 Page 10 I Project No. 2127 Report No. 111-21.25245 Page 11 2.2.2 CASMO-4 Fuel depletion analyses during core operation are performed with CASMO-4 Version 2.05.14 (using the 70-group cross-section library), which has been approved by the NRC for reactor analysis (depletion) when providing reactivity data for specific 3D simulator codes. CASMO-4 is a two-dimensional multigroup transport theory code based on the Method of Characteristics and it is developed by Studsvik of Sweden [4]. CASMO-4 is used to perform depletion calculations and to perform various sensitivity studies. The uncertainty on the isotopic composition of the fuel (i.e., the number density) is considered as discussed below (see Section 2.3.3). A validation for CASMO-4 to develop a bias and bias uncertainty is not necessary because the results of the CASM.O-4 sensitivity studies are not used as input into the keff calculations.

However, the code authors have validated CASMO-4 against MCNIP and various critical experiments

[5].The version of the CASMO-4 code used in this application has a built-in limitation in a number of isotopes that may be extracted for specific pins. Therefore, two independent CASMO-4 depletion calculations were performed to separately extract the actinides and fission products.The extracted isotopes were further combined and used in MCNP5-1.51 calculations.

2.3 Analysis

Methods 2.3.1 Design Basis Fuel Assembly There are various fuel designs stored in the Quad Cities SFP. For the purpose of this analysis, the reactivity of each design is evaluated and the most reactive fuel bundle lattice is determined for use as the design basis fuel assembly to determine kn.f at the 95/95 level. This approach follows the guidance in [2] and [6], and is further described below.Project No. 2127 Report No. HI-2125245 Page i12 2.3.1.1 Peak Reactivity The BWR fuel designs used at the Quad Cities Station use Gd as an integral burnable absorber.Initially, the Gd in the fuel assembly holds down the fresh fuel assembly reactivity and then, as core depletion occurs, the Gd begins to burnout until it is essentially fully depleted.

As the Gd depletes the reactivity of the fuel assembly increases until it reaches a peak. This peak reactivity is the fuel assembly's most reactive condition.

Note that most BWR fuel designs are composed of various axial lattices (including blankets) that can have different axial lengths, uranium loadings (also mixed oxide loading, for MOX fuel), fuel pin arrangements including partial or part-length rods, Gd pin locations and loading, etc. These various lattice components can all effect at what burnup the peak reactivity occurs and the magnitude of the peak reactivity.

The axial lattices within a single fuel assembly can therefore all have different peak reactivity 2.3.1.1.1 Peak Reactivity and Fuel Assembly Burnup Typically, a spent fuel assembly is characterized by its assembly average burnup (over all lattices or nodes). In this analysis methodology the fuel assembly average burnup is of no concern and is not credited for reactivity control. Rather, the methodology credits the residual Gd and other depletion isotopic compositions at the fuel assembly peak reactivity (most reactive lattice peak reactivity).

While the peak reactivity occurs at some specific lattice burnup, the peak reactivity lattice burnup varies from lattice to lattice within a fuel design. Therefore, independent calculations with MCNP5-1.51 using pin specific compositions (see Section 2.3.1.1.2) are performed for every lattice of the SVEA-96 Optima2 fuel assembly (as will be seen in Section 7, this is the fuel assembly Mith the design basis lattice) over a burnup range to determine the burnup at peak reactivity for every lattice. Since each lattice is considered at its peak reactivity (and therefore the lattice or nodal burnup at which that occurs), the fuel assembly average burnup or fuel assembly burnup profile is not applicable because the analysis already considers each lattice at its most reactive composition, independent of the fuel assembly average burnup.Project No. 2127 Report No. HI-2125245 Page 13 Project No. 2127 Report No. HI-2125245 Page 14 2.3.1.3 Determination of the Design Basis Fuel Assembly Lattice 2.3.1.4 Optima2 CASMO-4 Model Simplification Effect As previously discussed in Section 2.3.1.2, various fuel designs were provided.

Of these fuel designs, the SVEA-96 Optima2 designs were specified to be bounding.

The Optima2 model in CASMO-4 is described as the SVEA-96 model provided in the CASMO-4 manual [4]. This CASMO-4 internal model is slightly different from the actual fuel assembly geometry.Therefore, it is important to evaluate and if necessary quantify the reactivity effect of the CASMO-4 model simplifications inherent in the code. The CASMO-4 model geometry of the SVEA-96 Optima2 fuel differs from the SVEA-96 Optima2 fuel as follows: Project No. 2127 Report No. HI-2125245 Page 15 With respect to the fuel assembly geometry models, the amount of zirconium (and therefore the amount of water) in the CASMO-4 model of the SVEA-96 Optima2 fuel is reasonably similar to that of the actual SVEA-96 Optima2 fuel and therefore these built-in CASMO-4 simplifications are acceptable.

However, to evaluate the CASMO-4 model geometry simplification effect on reactivity, an applicable set of code-to-code comparisons is performed.

The following cases are evaluated.

For the purpose of showing that the two codes calculate an equivalent reactivity the following comparisons are made: Project No. 2127 Report No. 111-2125245 Page 16

Case 2.3.1.4.1 is compared to Case 2.3.1.4.2 at 0 GWD/MTU to show that the two codes calculate similar results with respect to the fuel assembly and storage rack geometry." Case 2.3.1.4.1 is compared to Case 2.3.1.4.2 at peak reactivity burnup to quantify the reactivity difference due to the effect of the spent fuel. The two codes use different cross section library versions and calculation sequences.

The main calculation sequence difference between the two codes is that CASMO-4 uses a thermal expansion of spent fuel pellet which effects the fuel density [4]. The actual density is conservatively used in MCNP5-1.51.

The results are expected to show that the MCNP5-1.51 code is conservative with respect to the CASMO-4 code. Any non-conservative result would be treated as a bias." Case 2.3.1.4.3 is compared to Case 2.3.1.4.2 to show the reactivity difference between the simplified MCNP5-1.51 model and the design basis model that is slightly modified to be similar to the CASM.O-4 insert orientation.

This case is expected to show that the design basis model with respect to the fuel pin pitch (and subsequent sub-bundle pitch) is conservative.

This is expected to be conservative because the design basis model fuel compositions are taken from the average fuel pin pitch CASMO-4 calculations and used in the MCNP5-1.51 design basis actual fuel pin locations.

Any non-conservative result would be treated as a bias.Case 2.3.1.4.3 is compared to the result of the actual design basis results (similar to Case 2.3.1.4.3 but with the bounding insert orientation) to show that the design basis model is conservative.

2.3.1.5 Core Operating Parameters As previously discussed, CASMO-4 is used to perform depletion calculations to determine the spent fuel isotopic composition.

The operating parameters for spent fuel depletion calculations are discussed in this Section. The operating parameters which may have a significant impact on BWR spent fuel isotopic composition are void fraction, control blade history, moderator temperature, fuel temperature, and power density. Other parameters such as axial enrichment distribution and effect of burnable absorbers are discussed in Section 2.3.1.5.3 and Section 2.3.1.5.2, respectively.

Sensitivity studies are performed to show the effect of each individual parameter, and to confirm that the selected values are in fact appropriate when combined at their worst case.2.3.1.5.1 Reactor Power Uprate T 0 determine the eltect ot the power uprate on the reactivity ot tuel assemblies in racks, the following evaluations are performed.

Project No. 2127 Report No. HI-2125245 Page 17 2.3.1.5.2 Integral Reactivity Control Devices The only type of burnable absorber used for the fuel assemblies covered in this analysis is Gd.The use of Gd does not increase the reactivity of the assembly, compared to an assembly lattice where all rods. contain fuel and no Gd. As discussed in Section 2.3.1.1, the Gd in the fuel assembly holds down the fresh fuel assembly reactivity and then, as core depletion occurs, the Gd begins to burnout until it is essentially fully depleted.

As the Gd depletes the reactivity of the fuel assembly increases until it reaches a peak. This peak reactivity is the fuel assembly's most reactive condition, which is used for design basis condition.

Note that integrated absorbers do not change the amount of water in the assembly, which is a large part of the effect of non-integral absorbers.

2.3.1.5.3 Axial and Planar Enrichment Variations 2.3.1.5.4 Fuel Assembly De-Channeling The SVEA-96 Optima2 fuel assembly (the most reactive fuel assembly, as will be shown in Section 7) cannot be de-channeled for storage in the SFP because of its specific design.However, GEM4 (the most second reactive fuel assembly, as will be shown in Section 7) may be de-channeled.

Studies are performed to evaluate the effect of storage of GE14 without the Zr channel at various radial positioning in the storage cells. The following cases are evaluated." Case 2.3.1.5.4.1:

This is the reference for Case 2.3.1.5.4.2 through Case 2.3.1.5.4.4.

The MCNP5-1.51 model used herein is a 2x2 array with the cell centered fuel assembly that includes the Zr channel, as shown in Figure 2.13(a)." Case 2.3.1.5.4.2:

The MCNP5-1.51 is a 2x2 array of GE14 fuel assembly lattice 5 (the most reactive lattice of GEI4, as will be shown in Section 7). The Zr channel is removed, as shown in Figure 2.13(b). The fuel assemblies are cell centered." Case 2.3.1.5.4.3:

The MCNP5-1.51 is the same as that of Case 2.3.1.5.4.2, except the fuel assemblies are eccentric toward the center, as shown in Figure 2.13(c)." Case 2.3.1.5.4.4:

The MCNP5-1.51 is the same as that of Case 2.3.1.5.4.2, except the fuel assemblies are eccentric away from the corner where the insert wings connect, as shown in Figure 2.13(d).Project No. 2127 Report No. 1-11-2125245 Page 18 2.3.1.6 2.3.2 Reactivity Effect of Spent Fuel Pool Water Temperatui-e The Quad Cities Station SFP has a normal pool water temperature operating range below 150 'F.For the nominal condition, the criticality analyses are to be performed at the most reactive temperature and density [2]. Also, there are temperature-dependent cross section effects in MCNP5-1.51 that need to be considered.

In general, both density and cross section effects may not have the same reactivity effect for all storage rack scenarios, since configurations with strong neutron absorbers typically show a higher reactivity at lower water temperature, while configurations without such neutron absorbers typically show a higher reactivity at a higher water temperature.

For the SFP racks which credit inserts, the most reactive SFP water temperature and density is expected to be at 39.2 TF and I g/cc, respectively.

The standard cross section temperature in MCNP5-1.51 is 293.6 K. Cross sections are also available at other temperatures; however, not usually at the desired temperature for SFP criticality analysis.

MCNP5-1.51 has the ability to automatically adjust the cross sections to the specified temperature when using the TMP card. Furthermore, MCNP5-1.51 has the ability to make a molecular energy adjustment for select materials (such as water) by using the S(a,p) card.The S(a,Oj) card is provided for certain fixed temperatures which are not always applicable to SFP criticality analysis.

Rather, there are limited temperature options, i.e., 293.6 K and 350 K, etc. Additionally, MCNP5-1.51 does not have the ability to adjust the S(a,j3) card for temperatures as it does for the TMP card discussed above. Therefore, additional studies are performed to show the impact of the S(a,3) card at the two available temperatures.

To determine the water temperature and density which result in the maximum reactivity, MCNP5-1.51 calculations are run using the bounding values. Additionally, S(a,p3) calculations are performed for both upper and lower bounding S(a,j3) values, if needed.The studies mentioned above are performed for the following cases for the single cell MCNP5-1.51 SFP model (with periodic boundary conditions through the centerline of the surrounding water 2): Project No. 2127 Report No. HI-2125245 Page 19

Case 2.3.2.1 (reference case): Temperature of 39.2 'F (277.15 K) and a density of 1.0 g/cc are used to determine the reactivity at the low end of the temperature range. The S(at,j3) card corresponds to a temperature of 68.81 'F (293.6 K)." Case 2.3.2.2: Temperature of 150 'F (338.71 K) and a corresponding density of 0.98026 g/cc are used to determine the reactivity at the high end of the temperature range. The S(a,3) card corresponds to a temperature of 68.81 'F (293.6 K).* Case 2.3.2.3: Temperature of 150 'F and a corresponding density of 0.98026 g/cc. The S(0,,3) card corresponds to a temperature of 170.33 'F (350 K).The bounding water temperature and density (the temperature and its corresponding density which result in the maximum reactivity) of the above cases are applied to all further calculations so that the most reactive water temperature and density is considered.

Note that the evaluations 2.3.3 Fuel Depletion Calculation Uncertainty To account for the uncertainty of the number densities in the depletion calculations performed in The depletion uncertainty is applied by multiplying it with the reactivity difference (at 95%/95%) between the MCNP5-1.51 calculation with spent fuel at peak reactivity (includes residual Gd) and a corresponding MCNP5-1.51 calculation with fresh fuel (without Gd 2 O 3).Calculations are performed for the single cell model of design basis fuel assembly.The uncertainty is determined by the following:

Uncertaintylsojo 1 0 cs = [ (kcal.2 -kcalc-) + 2

1/(alc.!2 + calc-2 ) ]

0.05 with= 1 , with spent fuel kcaic-2 kae with fresh fuel ocý,,- = Standard deviation of k,,,,i, Ccalc- = Standard deviation of k,, 1 c-2 The result of the MCNP5-1.51 calculation for the fuel depletion calculation uncertainty is statistically combined with other uncertainties to determine keff, Project No. 2127 Report No. HI-2125245 Page 20 2.3.4 Fuel and Storage Rack Manufacturing Tolerances In order to determine the kerf of the SFP at a 95% probability at a 95% confidence level, consideration is given to the effect of the BWR fuel and SFP storage rack manufacturing tolerances on reactivity.

The reactivity effects of significant independent tolerance variations are combined statistically

[2]. The evaluations use the same MCNP5-1.51 models used in the design basis calculation, 2.3.4.1 Fuel Manufacturing Tolerances The BWR fuel tolerances for Optima2 Q122 fuel (which is the most reactive fuel design evaluated herein) are presented in Table 5.1(a). Fuel tolerance calculations are performed using the design basis fuel assembly lattice, and therefore only the tolerances applicable to that lattice are applicable.

Separate CASMO-4 depletion calculations are performed for each fuel tolerance and the full value of the tolerance is applied for each case in both the depletion and in rack calculations.

Pin specific compositions are used. The MCNP5-1.51 tolerance calculation is compared to the MCNP5-1.51 reference case (nominal parameter values) at the 95% probability at a 95% confidence level using the following equation: delta-ki, nc = (k~aU 2 -kc,1cl) +/- 2 * ý(a,2 + a2)The following fuel tolerances are considered in this analysis:* Fuel enrichment

Gd loading* Fuel pellet density (IJ0 2 and U0 2+Gd 2 0 3 fuel rods)* Fuel pellet outer diameter (OD)* Fuel cladding inner diameter (ID)* Fuel cladding OD* Fuel pin pitch* Fuel sub-bundle pitch 3* Combination of 4 o Water wing canal inner width o Channel outer square width o Channel corner inner radius o Central water canal inner square width* Combination of4 o channel wall thickness 3 For fuel sub-bundle pitch uncertainty calculation, the fuel hardware (channel, central water channel and water wings) is fixed. The fuel lattices are moved only.' Conservatively, the various tolerances are considered together.

The tolerance limits that result in an increase of the amount of water in the core are considered together in one set of uncertainty calculations, and the tolerance limits that result in a decrease of the amount of water in the core are considered together in another set of uncertainty calculations.

Project No. 2127 NReport No. HI-2125245 Page 21!

o Water cross wall thickness The maximum positive reactivity effect of the MCNP5-1.51 calculations for each tolerance is statistically combined with the other tolerance results, and this result is then statistically combined with other uncertainties when determining the keff value.2,3.4.2 SFP Storage Rack Manufacturing Tolerances The SFP- rack tolerances are presented in TFables 5.3(a) and 5.3(b). The single cell MCNP5-1.51 model is used to determine the reactivity effect of the tolerance, and the full value of the tolerance is applied for each case. The MCNP5-1.51 tolerance calculation is compared to the MCNP5-1.51 reference case with a 95% probability at a 95% confidence level using the following equation: delta-k,~ic

= (kcalc2 -kcailc) +/- 2 *Y + 22)The following SFIP rack manufacturing tolerances are considered in this analysis: " Storage cells: o Cell ID and cell pitch o Cell wall thickness" Rack inserts (poison)o Width The maximum positive reactivity effect of the MCNP5-1.51 calculations for each tolerance is statistically combined with the other tolerance results, and this result is then statistically combined with other uncertainties when determining the kg value.The evaluations use the same MCNP5-1.51 models used in the design basis calculation.

The isotopic compositions of the fuel rods are the same as those of the design basis fuel assembly.The poison thickness and loading are used at their minimum values; i.e., they are treated as a bias instead of uncertainty, for conservatism and simplification.

2.3.5 Radial

Positioning 2.3.5.1 Fuel Assembly Orientation in the Core The fuel assembly orientation in the core with respect to its control blade does not change and therefore the design basis calculations consider the only possible configuration.

2.3.5.2 Fuel Radial Positioning in the Rack The BWR fuel that is loaded in the SFP racks may not rest exactly in the center of the storage cell. Evaluations are performed to detennine the most limiting fuel radial location.

The following eccentric fuel positioning cases are analyzed: Project No. 2127 Report No. 111-2125245 Page 22

" Case 2.3.5.2.1:

This is the reference for Case 2.3.5.2.2 through Case 2.3.5.2.5.

The MCNP5-1.51 model used herein is a 2x2 array which is the same as the primary single bundle MCNP5-1.51 model used elsewhere in this analysis.

In both models the fuel is centered in the rack cell. See Figure 2.7(a).* Case 2.3.5.2.2:

Every fuel assembly is positioned toward the center, for the 2x2 array, as shown in Figure 2.7(b).* Case 2.3.5.2.3:

Every fuel assembly is positioned toward the corner where the insert wings connect, for the 2x2 array, as shown in Figure 2.7(c)." Case 2.3.5.2.4:

Every fuel assembly is positioned away from the corner where the insert wings connect, for the 2x2 array, as shown in Figure 2.7(d).* Case 2.3.5.2.5:

Every fuel assembly is centered between insert and cell walls, for the 2x2 array, as shown in Figure 2.7(e).* Case 2.3.5.2.6:

This is the reference for Case 2.3.5.2.7 through Case 2.3.5.2.10.

The MCNP5-1.51 model used herein is an 8x8 array which is the same as the primary single bundle MCNP5-1.51 model used elsewhere in this analysis.

In both models the fuel is centered in the rack cell." Case 2.3.5.2.7:

Every fuel assembly is positioned toward the center, for the 8x8 array, as shown in Figure 2.8.* Case 2.3.5.2.8:

Every fuel assembly is positioned toward the corner where the insert wings connect, for the 8x8 array.* Case 2.3.5.2.9:

Every fuel assembly is positioned away from the corner where the insert wings connect, for the 8x8 array." Case 2.3.5.2.10:

Every fuel assembly is centered between insert and cell walls, for the 8x8 array." Case 2.3.5.2.11:

This is the reference for Case 2.3.5.2.12.

The MCNP5-1.51 model used herein is a single rack cell where the fuel is centered.* Case 2.3.5.2.12:

The fuel assembly is centered between insert and cell walls, for the single rack cell.The maximum positive reactivity effect of the MCNP5-1.51 calculations for the fuel radial positioning is added as the bias and the corresponding 95/95 uncertainty is statistically combined with other uncertainties to determine ke'f.Prqject No. 2127 Report No. HI-2 125245 Page 23 Note that the evaluations use the same MCNP5-1.51 models with periodic boundary conditions used in the design basis calculation, except that the array size is larger. The isotopic compositions of the fuel rods are the same as those of the design basis fuel assembly.2.3.5.3 Inserts Radial Positioning Since the insert width and SFR cell inner diameter are comparable, and each insert is installed into the rack cell such that the insert becomes an integral part of the fuel rack, no uncertainty in the positioning for inserts is evaluated.

The water gap between rack wall and insert is not assumed, since it may provide a small flux trap effect. Nevertheless, the orientation of fuel assembly with respect to position of insert is considered in Section 2.3.5.4.2.3.5.4 Fuel Orientation in SFP Rack Cell As described in Section 5. 1, fuel assemblies have various radial fuel enrichments and gadolinium distribution.

Also, one corner of each fuel assembly is adjacent to the control blade during the depletion in the core. As a result, the fuel depletion is not uniform (more discussion is provided in Section 2.3.1.1.2) and one fuel assembly corner may be more reactive than other corners and therefore the fuel assembly orientation in the SFP storage cell may have an impact on reactivity.

Five cases are analyzed to assess the fuel assembly orientation variations and to determine the most limiting fuel orientation in SFP rack cell with respect to the insert.The MCNP5-1.51 model of the reference case is the design basis fuel in the 2x2 array, as shown in Figure 2.9(a). The MCNP5.1.51 models of the other four cases are the same as that of the reference case, except with different orientation of fuel assemblies with respect to the inserts.Figure 2.9(b) through Figure 2.9(e) show the configurations of the fuel assemblies in the SFP cells for the evaluated cases.Note that the evaluations use the same MCNP5-1.51 models with periodic boundary conditions used in the design basis calculation.

The isotopic compositions of the fuel rods are the same as those of the design basis fuel assembly.2.3.6 I Project No. 2127 Report No. 1-11-2125245 Page 24 2.3.6.1 2.3.6.1.1 I Pro.ject No. 2127 Report No. 1I--2 125245 Page 25 2.3.6.1.2 2.3.6.2 2.3.7 Insert Coupon Measurement Uncertainty There is a measurement uncertainty associated with the B-I 0 content in the poison test coupons. In this analysis, the minimum 13-10 loading and the minimum insert thickness are conservatively used for criticality calculations.

Therefore, the coupon measurement uncertainty is not evaluated further in the analysis.2.3.8 Maximum kef Calculation for Normal Conditions The calculation of the maximum kcff of the SFP storage racks fully loaded with design basis fuel assemblies at their maximum reactivity is determined by adding all uncertainties and biases to the calculated reactivity.

Note that the insert thickness and its B-I 0 loading are taken at their worst case values.Project No. 2127 Report No. 2125245 Page 26 krff is determined by the following equation: keff = k,,,, + uncertainty

+ bias where uncertainty includes: I and the bias includes Note that each uncertainty is statistically combined with other uncertainties, while biases are added together in order to determine krj-f.The approach used in this analysis takes credit for residual Gd.2.4 Margin Evaluation The criticality analysis is performed using several conservative assumptions which introduce quantifiable margin into the analysis.

Four main conservative assumptions are:* Minimum insert B 4 C loading* Minimum insert thickness* Minimum amount of B-10 in boron* Bounding lattice throughout the entire length of fuel assembly.To evaluate this margin, the following cases are evaluated:

Case 2.4.1: This is the design basis fuel assembly.

This is the reference for Case 2.4.2 and Case 2.4.3.* Case 2.4.2: This case is the same as Case 2.4.1, except the nominal insert B 4 C loading, nominal insert thickness and nominal amount of B- 0 in boron are used.Project No. 2127 Report No. 141-2125245 Page 27 Case 2.4.3: This case is the same as Case 2.4.1, except the model includes each Optima2 Q122 fuel lattice in the appropriate axial position.

However, the top and bottom blankets were conservatively replaced by adjacent fuel lattices.

The peak reactivity burnup for each individual Optirna2 Q122 lattice under the design basis core operation parameters was determined separately and used in this case (i.e. each lattice is at its individual peak reactivity).

Therefore, the model represents a conservative maximum but unrealistic reactivity of the actual Optima2 fuel assembly.The differences between the reactivity of Cases 2.4.2 and 2.4.3 and the reactivity of reference Case 2.4.1 provide a quantified margin.Note that the evaluations use the same MCNP5-1.51 models used in the design basis calculation.

The isotopic compositions of the fuel rods of Case 2.4.1 and Case 2.4.2 are the same as those of the design basis fuel assembly.2.5 Fuel Movement, Insp)ection and Reconstitution Operations

2.6 Accident

Condition The accidents considered are:* SFP temperature exceeding the normal range* Dropped assemblies

Storage cell distortion

Missing insert* Misloaded fuel assembly (a fuel assembly in the wrong location within the storage rack)/Missing an insert* Mislocated fuel assembly (a fuel assembly in the wrong location outside the storage rack)* Miss-installment of an insert on wrong sides of a cell* Insert mechanical wear* Rack movement Those are briefly discussed in the following sections.Projiect No. 2127 Report No. 1-11-2125245 Page 28 Note that the double contingency principle as stated in [2] specifies that "two unlikely independent and concurrent incidents or postulated accidents are beyond the scope of the required analysis." This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

The kcff calculations performed for the accident conditions are done with a 95%probability at a 95% confidence level.The accident conditions are considered at the 95/95 level using the total corrections friom the design basis case.2.6.1 Temperature and Water Density Effects The SFP water temperature accident conditions for consideration are the increase in SFP water temperature above the maximum SFP operating temperature of 150 'F. The decrease in temperature was already considered fbr the temperature coefficient determination as discussed in Section 2.3.2.To bound the potential increase in reactivity due to increased SFP temperature, the following case is evaluated:

Case 2.6.1: This case uses a temperature of 255 'F (397.04 K) and a density of 0.84591 g/cc. The S(cc,p) card corresponds to a temperature of 260.33 'F (400 K). In this model, it is assumed that the water modeled includes 10% void. Void is modeled as 10% decrease in density, compared to the density of water at 255 'F.The evaluation use the same MCNP5-1.51 model used in the design basis calculation.

Note that as discussed in Section 2.3.2, SFP storage racks with strong neutron absorbers, such as inserts, show a higher reactivity at a lower water temperature.

The case evaluated above is performed to confirm this statement.

2.6.2 Dropped

Assembly -Horizontal For the case in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a separation distance more than 12 inches.Also, the length of the inserts (as indicated in Table 5.3(b)) covers this separation distance.

Thus, the horizontally dropped assembly is decoupled friom the fuel assemblies in the rack. This accident is also bounded by the mislocated case, where the mislocated assembly is closer to the assembly in the racks. Therefore, the horizontally dropped fuel assembly is not evaluated further in the report.2.6.3 Dropped Assembly -Vertical into a Storage Cell It is also possible to vertically drop an assembly into a location that might be occupied by another assembly or that might be empty. Such a vertical impact would at most cause a small compression of the stored assembly, if present, or result in a small deformation of the baseplate for an empty cell.These deformations could potentially increase reactivity.

However, the reactivity increase would be small compared to the reactivity increase created by the 'misloaded fuel assembly/missing insert'accident (discussed in Section 2.6.5) that does not include the insert in one rack cell. The vertical ProJect No. 2127 Report No. HI-2125245 Page 29 drop is therefore bounded by this misload accident and no separate calculation is performed fbr this drop accident.2.6.4 Storage Cell Distortion A storage cell distortion or altered geometry as a result of fuel handling equipment uplift forces is possible.

However, the reactivity increase would be small compared to the possible reactivity increase created by the 'misloaded fuel assembly/missing insert" accident that does not include the insert in one rack cell, as discussed in Section 2.6.5. The storage cell distortion is therefore bounded by the 'misloaded fuel assembly/missing insert' accident and no separate calculation is performed for the storage cell distortion accident.As a result of significant distortion, the storage cell for whatever reason may not be able to contain the insert and also it will be therefore unacceptable for storage of a fuel assembly.

This condition is bounded by the 'misloaded fuel assembly/missing insert' accident.

However to show that it is acceptable for normal operation and that the empty storage cell decreases the reactivity of the SFR, the model with an empty storage cell, i.e. without a fuel assembly and insert, in the center of a 8x8 array, is evaluated.

Two cases with a cell centered and eccentric position of the fuel assemblies are analyzed.2.6.5 Misloaded Fuel Assembly/Missing Insert The fuel storage racks are qualified for storage of fuel assembly with the highest anticipated reactivity; thus it is not possible to misload a fuel assembly if every cell with a fuel assembly has an insert.Ilowever, there are a fiw cells in the SFP racks which are exempt from fuel storage. Those locations are blocked or have partial interferences.

In a hypothetical scenario, it is assumed that a fuel assembly is misloaded into a cell with a missing insert. To evaluate the effect, the following cases are evaluated: " Case 2.6.5.1: The MCNP5-1.51 model includes an 8x8 array. One cell near the center of the rack does not have the insert. The misloaded fuel assembly is the design basis fuel assembly.This fuel assembly is eccentric toward the walls that are not covered by inserts. Other fuel assemblies are also eccentric toward the misloaded fuel assembly.

The periodic boundary conditions are used through the centerline of the surrounding water (BORAFLEX replacement).

The temperature of the model is set to the minimum (39.2 TF) with its corresponding water density and S(a,p) card. These temperature and density are bounding for the SFP racks. See Figure 2.10(a)." Case 2.6.5.2: The MCNP5-1.51 model is the same as Case 2.6.5.1, except with all fuel assemblies centered in the rack cells. See Figure 2. 1 0(b).Project No. 2127 Report No. H1-2125245 Page 30

2.6.6 Mislocated

Fuel Assembly The Quad Cities SFP layout was reviewed to determine the possible worst case locations for a mislocated fuel assembly.

Three hypothetical locations where a fuel assembly may be mislocated are:* In the water gap between the racks and the pool wall* In the corner between two racks* Between the SFP rack and the inspection platform.The three cited scenarios are evaluated, as follows.2.6.6.1 Mislocation of a Fuel Assembly in the Water Gap between the Racks and Pool Wall A fuel assembly may be mislocated in the water gap between the racks and the pool wall. Due to the neutron leakage to the outside the storage rack area, the effect of this mislocation is bounded by that of'mislocation of a fuel assembly between the SFP rack and the inspection platform' accident, as discussed in Section 2.6.6.3. No separate calculation is performed for this accident.2.6.6.2 Mislocation of a Fuel Assembly in the Corner between Two Racks There are some places in the SFP, but outside of the racks, where the mislocated fuel assembly may be in the corner between two racks (thus the mislocated fuel assembly would be adjacent to the fuel assemblies in racks from two sides). To evaluate the effect of the mislocation of a fuel assembly in the corner between two racks, the following cases are evaluated: " Case 2.6.6.2.1:

The MCNP5-1.51 model is three 8x8 arrays of SIP rack cells. The misplaced fuel assembly is in the corner between two racks. The fuel assemblies in the rack are eccentric toward the mislocated fuel assembly.

The misplaced fuel assembly is placed as close to the racks as possible.

All fuel assemblies in the model are the design basis fuel assembly.

Figures 2.1 1(a) and 2.11 (b) show the MCNP5- 1.51 model used for this analysis.* Case 2.6.6.2.2:

The MCNP5-1.51 model is the same as Case 2.6.6.2.1, except with all fuel assemblies are centered.

See Figures 2.11 (a) and 2.11 (c)." Case 2.6.6.2.3:

The MCNP5-1.51 model is the same as Case 2.6.6.2.1, except the temperature of the model is set to the maximum (150 'F).* Case 2.6.6.2.4:

The MCNP5-1.51 model is the same as Case 2.6.6.2.2, except the temperature of the model is set to the maximum (150 'F).2.6.6.3 Mislocation of a Fuel Assembly between the SFP Rack and the Inspection Platform As discussed in Section 2.5, the fuel handling/inspection/reconstitution platform may have one fuel assembly in it at a time. There is a possibility that a fuel assembly is mislocated between the ProJect No. 2127 Report No. HI-2125245 Page 31 SFP racks and the fuel assembly in the platform.

To evaluate the effect of the mislocation of a fuel assembly between the SFP Rack and the Inspection Platform, the following cases are evaluated:

0 Case 2.6.6.3.1:

The MCNP5-1.51 model is an 8x8 array of SFP rack cells. The misplaced fuel assembly is adjacent to the SFP rack and the inspection platform.

The fuel assembly in the platform is lined up with the mislocated fuel assembly.

The fuel assemblies in the rack are eccentric toward the mislocated fuel assembly.

The misplaced fuel assembly is placed as close to the rack and fuel assembly in the inspection station as possible.

All fuel assemblies in the model are design basis fuel assembly.

The side of the fuel in the platform which does not have any fuel has at least 12 inches of water. Figure 2.12(a) shows the MCNP5-1.51 model used fbr this analysis." Case 2.6.6.3.2:

The MCNP5-1.51 model is the same as Case 2.6.6.3.1, except with all fuel assemblies are centered, See Figure 2.12(b)." Case 2.6.6.3.3:

The MCNP5-1.51 model is the same as Case 2.6.6.3.1, except the temperature of the model is set to the maximum (150 'F)." Case 2.6.6.3.4:

The MCNP5-1.51 model is the same as Case 2.6.6.3.2, except the temperature of the model is set to the maximum (150 'F).2.6.7 Mis-installment of an Insert on Wrong Side of a Cell There is a small possibility that an insert is installed on wrong sides of the cell. In this case, there may not be a poison between a fuel assembly placed in that cell and a fuel assembly in an adjacent cell. However, the effect of this mis-installment is bounded by that of 'misloaded fuel assembly/missing insert' accident that does not include the insert in one rack cell, as discussed in Section 2.6.5. No separate calculation is performed for this accident.2.6.8 Insert Mechanical Wear Handing accidents and other environmental damage may cause scratches and local wear of inserts. The effect of this accident is bounded by that of 'misloaded fuel assembly/missing insert'accident, as discussed in .Section 2.6.5.2.6.9 Rack Movement In the event of seismic activity, there is a hypothetical possibility that the storage rack arrays may move and come closer to each other. Since there is no water gap modeled between cells of a storage rack, the reactivity of the rack movement case is bounded by the reactivity of the design basis calculation.

Project No. 2127 Report No. 1-11-2125245 Page 32

2.7 Spent

Fuel Rack Interfaces The spent fuel pool includes a single type of Region I spent fuel racks, which are loaded with the neutron absorbing inserts in every storage cell as well as a uniform fuiel assembly loading pattern.Therefore, any possible water gaps and interfaces between the racks are bounded by the infinite array used in the design basis calculations.

However, since the neutron absorbing inserts are located in the same comers of rack cells (e.g. south-west), there are two peripheral rows of the cells (correspondingly, north and east periphery of the pool), which are loaded with the fuel assemblies that have one side that is not adjacent to the insert. Furthermore, one fuel assembly in the corner of the spent fuel pool (correspondingly, north-east corner) has two sides that are not adjacent to the insert. Due to the neutron leakage on the periphery of the spent fuel pool the reactivity increase is not expected.

Nevertheless, to evaluate the effect of such conditions, the full spent fuel pool model (74x74 array) loaded with the cell centered design basis fuel assemblies and the model where all fuel assemblies are shifted to the fuel assembly in the corner, which is discussed above, were evaluated.

2.8 Reconstituted

Fuel Assemblies The SFP contains various reconstituted assemblies which were examined and determined to be relatively old and low reactivity designs. The reconstitution of these fuel assemblies removed fuel rods and replaced them by either fuel rods that are of the same or less initial enrichment and equal or greater Gd loading (with burnup similar to the rod they replaced) or solid stainless steel rods.The reactivity effect of this reconstitution is not sufficient to make the reconstituted fuel assembly more reactive than the bounding lattice. Therefore, reconstituted assemblies are covered by the design basis Optima2 Q122 lattice 146. Future reconstituted assemblies will replace fuel rods with stainless steel rods.ProJect No. 2127 Report No. HI-2 125245 Page 33

3. ACCEPTANCE CRITERIA 3.1 Applicable Codes, Standards and Guidance's Codes, standard, and regulations or pertinent sections thereof that are applicable to these analyses include the following:

Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criterion 62,"Prevention of Criticality in Fuel Storage and Handling."* Code of Federal Regulations, Title 10, Part 50.68, "Criticality Accident Requirements."* USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Revision 3 -March 2007.* L. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T.Collins, August 19, 1998.* ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors (withdrawn in 2004).* USNRC, NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, January 2001.* DSS-ISG-2010-01, Revision 0, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools.Project No. 2127 Report No. 111-2125245 Page 34

4. ASSUMPTIONS The analyses apply a number of assumptions, either for conservatism or to simplify the calculation approach.

Important aspects of applying those assumptions are as follows: 1. Bounding or sufficiently conservative inputs and assumptions are used essentially throughout the entire analyses, and as necessary studies are presented to show that the selected inputs and parameters are in fhct conservative or bounding.2. Neutron absorption in minor structural members of the fuel assembly is neglected, e.g., spacer grids are replaced by water. It is conservative to neglect the spacer grids because this spent fuel pool contains no soluble boron, the region around the fuel rods is under-moderated, as confirmed by the fuel tolerances calculations that change the fuel to moderator ratio (Section 7.1.7.1);

therefore, neglecting the spacer grid places more water within the calculation model. In addition, the inconel springs within the spacer are a stronger neutron absorber than water. The active fuel region repeats periodically in the vertical direction.

Therefore, neutron absorption in upper and lower tie plates, fuel plenums, etc. is neglected.

3. The neutron absorber length in the rack is more than the active region of the fuel, but it is modeled to be the same length.4. The fuel density is assumed to be equal to the pellet density, and is conservatively modeled as a solid right cylinder over the entire active length, neglecting dishing and chamfering, This is acceptable since the amount of fuel modeled is more than the actual amount.5. For the inserts, only the worst case bounding material specifications are used (minimum B-1 0 loading and minimum thickness).

6. All models are laterally infinite arrays of the respective configuration, neglecting lateral leakage. The exception is where the model boundaries are water, as specified.

7. All fuel cladding materials are modeled as pure zirconium, while the actual fuel cladding consists of one of several zirconium alloys. This is acceptable since the model neglects the trace elements in the alloy which provide additional neutron absorption.

8.9. The full spent fuel pool model is considered as a 74x74 array of storage cells. I gaps between the spent fuel racks were conservatively neglected.

Project No. 2127 Report No. 1-11-2125245 Page 35

5. INPUT DATA 5.1 Fuel Assembly Specification The SFP racks are designed to accommodate the following fuel assembly types used in the Quad Cities Unit I and Unit 2, which are presented in a chronologic order along with the initial maximum planar average enrichment (IMPAE): The specifications for the most reactive fuel assemblies from the fuel product lines discussed above are presented in Table 5.1. The additional specifications for other fuel design variations are presented in Appendix A.The fuel assembly MCNP model used for the design basis calculations is presented in Fi ure 5.4.The fuel rod, cladding and channel are explicitly modeled.Axially, the design basis MCNP model considers the bounding lattice along the entire length and uses water reflectors at the top and bottom. The MCNP model for the margin evaluation calculations discussed in Section 2.4 differ from the design basis model in that the active length specifically considers each actual lattice in its actual axial configuration (i.e. all the lattices from the Q122 bundle are modeled in the same MCNP model).5.2 Reactor Parameters The reactor core parameters are provided in Table 5.2(a). The reactor control blade data are provided in Table 5.2(b). The reactor control parameters used in CASMO-4 screening and design basis calculations are provided in Table 5.2(c).5.3 Spent Fuel Pool Parainmeers The spent fuel pool parameters are provided in Table 5.2(a).Pro.ject No. 2127 Report No. 1-11-2125245 Page 36

5.4 Storage

Rack Specification The storage rack specifications that are used in the criticality analysis are summarized in Tables 5.3(a) and 5.3(b). The Quad Cities Unit I and Unit 2 SFP are shown in Figures 5.2(a) and 5.2(b), respectively.

eMCNP5-.51 SFP model consists of a single rack cell with periodic boundary conditions through the centerline of the water (BORAFLEX replacement), thus simulating an infinite array of storage cells. The storage rack cell is modeled the same length as the active fuel and all other storage rack materials are neglected.

The neutron absorber is modeled with the worst case bounding values (the minimum B-10 loading and the minimum thickness) provided in Table 5.3(b) and Figure 5.3.The cell wall thickness of the boundary is different from that of inner walls. The cell wall thickness of the boundary is thicker than the inner wall thickness.

The SFP model uses the inner cell wall thickness only, as given in Table 5.3(a), because it decreases the amount of steel in the model, which acts a neutron absorber.The MCNP5-1.51 SFP rack cell model is shown in Figure 5.4.5.4.1 Material Compositions The MCNP5-!.51 material specification is provided in Table 5.4(a) for non-fuel materials, and in Table 5.4(b) for fuel materials.

Project No. 2127 Report No. HI-2125245 Page 37

6. COMPUTER CODES The following computer codes were used in this analysis." MCNP5-1.51

[11 is a three-dimensional continuous energy Monte Carlo code developed at Los Alamos National Laboratory.

This code offers the capability of performing full three dimensional calculations for the loaded storage racks. MCNP5-1.51 was run on the PCs at Holtec." CASMO-4 [4] is a two-dimensional multigroup transport theory code developed by Studsvik.

CASMO-4 is used to perform the depletion calculation for the pin-specific approach, and for various studies. CASMO-4 was run on the PCs at Holtec.Project No. 2127 Report No. HI-2125245 Page 38

7. ANALYSIS 7.1 Design Basis and Uncertainty Evaluations 7.1.1 7.1.2 Determination of the Design Basis Fuel Assembly Lattice As discussed in Section 2.3.1.3, MCNP15-1.5l calculations were performed to determine the design basis lattice. The results for the SVEA-96 Optirna2 Q122 fuel assembly are presented in Table 7.2(a) .The results for the GEI4 lattice type 5 are presented in Table 7.2(b), along with the bounding result of the SVEA-96 Optima2 Q122. As can be seen, the SVEA-96 Optirna2 Q122 lattice type 146 is bounding, and thus it is selected as the design basis lattice. The CASMO-4 model of the SVEA-96 Optima2 bundle Q122 lattice 146 used for depletion calculations is shown in Figure 5.1.7.1.2.1 Fuel Assembly De-Channeling As discussed in Section 2.3.1.5.4, the reactivity of die second most reactive assembly with no Zr channel at various radial positioning was evaluated.

The results are provided in Table 7.2(b) and compared with the reactivity of the design basis lattice (SVEA-96 Optima2 Q122 lattice type 146).As can be seen, the SVEA-96 Optima2 Q122 lattice type 146 is bounding.

Therefore, storage of fuel assemblies without channels is acceptable.

7.1.3 Optima2

CASMO-4 Model Simplification Effect As discussed in Section 2.3.1.4, the effect of CASMO-4 model simplifications on the calculated reactivity of the SVEA-96 Optima2 Q122 lattice 146 was evaluated.

The results are provided in Table 7.3. As can be seen, the reactivity of the simplified model is comparable to that of the complete model of SVEA-96 Optima2 Q122 lattice 146 (essentially within the 95/95 uncertainty between the two calculations).

Therefore, the results show that the CASM.0-4 model simplification Project No. 2127 Report No. 111-2125245 Page 39 does not have a significant impact on the analysis conclusions regarding the determination of the design basis lattice.7.1.4 Core Operating Parameters As discussed in Section 2.3.1.5, the effects of the core operating parameters on the reactivity were evaluated.

The results are provided in Table 7.4. The results show that the two dominant core operating parameters are the control blade insertion and void fraction.

The other core operating parameters have an insignificant impact. Therefore, the design basis (bounding) core operating parameters are: control blades inserted, 0% void fraction, maximum fuel and moderator temperature and maximum specific power.7.1.4.1 Reactor Power Uprate As discussed in Section 2.3.1.5.1, the effect of the MUR on the reactivity was evaluated.

The results are provided in Table 7.4. The most important core operating parameters are rodded operation (control blades) and void fraction.

Other parameters have relatively negligible effects on reactivity.

As can be seen, the calculations with the increased power density show statistically equivalent results, which confirms the negligible effect of the reactor power uprate on reactivity.

7.1.5 Water

Temperature and Density Effect As discussed in Section 2.3.2, the effects of water temperature, and the corresponding water density and temperature adjustments (S(ap)) were evaluated for SFP racks. The results of these calculations are presented in Table 7.5.The results of the SFP temperature and density calculations show that as expected (for poisoned racks) the most reactive water temperature and density for the SFP racks is a temperature of 39.2 TF at a density of I g/cc, and these values are used for all calculations in SFP racks.7.1.6 Depletion Uncertainty As discussed in Section 2.3.3, the uncertainty of the number densities in the depletion calculations was evaluated.

The results of these calculations are presented in Table 7.6(a).Also, as discussed in Section 2.2.1. 1. 1, the uncertainty associated with FPs and LFPs was evaluated.

The results of these calculations are presented in Table 7.6(b).These two uncertainties are statistically combined with other uncertainties to determine keff in Table 7.11 and Table 7.14.ProJect No. 2127 Report No. 1-11-2125245 Page 40 7.1.7 Fuel and Rack Manufacturing Tolerances 7.1.7.1 Fuel Assembly Tolerances As discussed in Section 2.3.4.1, the effect of the BWR fuel tolerances on reactivity was determined.

The results of these calculations are presented in Table 7.7. The maximum positive delta-k value for each tolerance is statistically combined.The maximum statistical combination of fuel assembly tolerances is used to determine ken' in Table 7.11 and Table 7.14.7.1.7.2 SFP Rack Tolerances As discussed in Section 2.3.4.2, the effect of the manufacturing tolerances on reactivity of the SFP racks with inserts was determined.

The results of these calculations are presented in Table 7.8. The maximum positive delta-k value for each tolerance is statistically combined.The maximum statistical combination of the SFP rack tolerances is used to determine keff in Table 7.11 and Table 7.14.7.1.8 Radial Positioning 7.1.8.1 Fuel Assembly Radial lPositioning in SFP Rack As discussed in Section 2.3.5.2, twelve fuel assembly radial positioning cases in racks were evaluated.

The results of these calculations are presented in Table 7.9(a). For each eccentric position case, the result for similar but cell centered case is considered as a reference.

The results show that most cases show a negative reactivity effect, however some delta k,,,e values are positive.

Therefore, a maximum delta k.1 , value is applied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in Table 7.11 and Table 7.14.7.1.8.2 Fuel Orientation in SFP Rack As discussed in Section 2.3.5.4, five fuel assembly orientation cases in racks were evaluated.

The results of these calculations are presented in Table 7.9(b). The result for the reference case is also included.

The results show that all cases are statistically equivalent and the reactivity effect of fuel orientation is negligible.

Nevertheless, a maximum positive delta ke,,e value is applied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in Table 7.11 and 'Fable 7.14.Pro.icct No. 2127 Report No. 1-11-2125245 Page 41 7.1.9 Fuel Rod Geometry Change 7.1.9.1 The results are presented in Table 7.10.The maximum 'k.a1, kcaIc~refcrncc' is added as a bias, and the '2 * *J(YC'uc 2 + Ocajcrcfrcncc), (95/95 uncertainty) is added as an uncertainty to determine kc.f in Table 7.11 and Table 7.14.7.1.9.2 7.1.10 7.2 Maximum kOf Calculations./br Normal Conditions As discussed in Section 2.3.8, the maximum ken" for normal conditions is calculated.

The results are tabulated in Table 7.11. The results show that the maximum kc,,r for the normal conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95% confidence level.7.3 Margin Evaluation As discussed in Section 2.4, the margin analyses were performed using the nominal values for poison thickness and loading, as well as the actual lattice configuration of the Optima2 Q122 fuel assembly.

The results of calculations are provided in Table 7.12(a) and Table 7.12(b). As can be seen and is expected, the reactivity of design basis is larger. The use of a minimum B-10 loading relative to use of a nominal B-10 loading with tolerance uncertainty provide an additional

-1%reactivity margin to the regulatory limit with a 95% probability at a 95% confidence level.Project No. 2127 Report No. 1-11-2125245 Page 42 The summary of the margin evaluation is presented in Table 7.12(c). The result shows that quantified margin remains in the analysis to offset potential effects not already considered in the model.7.4 Abnormal and Accident Conditions As discussed in Section 2.6, the effects of empty storage cell, increased temperature, misloaded fuel assembly/missing insert, and mislocated fuel assembly accidents on reactivity were evaluated.

The results are provided in Table 7.13(a) and Table 7.13(b).As can be seen, the increased water temperature will not result in an increase in reactivity.

Both misloaded fuel assembly/missing insert and mislocated fuel accidents may result in an increase in reactivity.

For the SFP racks, the effect on reactivity of the missing insert is the limiting case.Thus, its calculated MCNP5-t.51 kcac is used for maximum kerr calculations for abnormal and accident conditions, discussed in Section 7.5.The condition with the empty storage cell without insert in the spent fuiel rack shows a lower reactivity than a design basis case, therefore, it is acceptable to have the empty storage cell without insert in the spent fuel pool.7.5 Maximum kef Calculations.for Abnormal and Accident Conditions As discussed in Section 2.6, the maximum kjr for abnormal and accident conditions is calculated.

The results are tabulated in Table 7.14. The results show that the maximum k~ff for abnormal and accident conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95%confidence level.7.6 Spent Fuel Rack .Interaces As discussed in Sections 2.7, the interface between SFRs and pool walls, i.e. effect on reactivity of the peripheral fuel assemblies, that have a side non-adjacent to tie insert, was evaluated.

The results are provided in Table 7.15. As can be seen, this condition will not result in an increase of SFR reactivity.

This result is expected because the infinite array design basis model is an infinite array of storage cells with inserts while the full pool model used for these rack interface calculations includes the rack edge along the pool wall where there is no insert along the water gap edge (i.e. no additional cell with an insert). Therefore, this water gap edge allows for neutron leakage and as the calculations show result in statistically equivalent results.Project No. 2127 Report No. 1-11-2125245 Page 43

8. CONCLUSION The criticality analsis for the storage of BWR assemblies in the Quad Cities SFP racks with NETCO-SNAP-IN inserts has been performed.

The results for the normal condition show that krff is M with the storage racks fully loaded with fuel of the highest anticipated reactivity, which is SVEA-96 Optima2 Q122 lattice type 146, at a temperature corresponding to the highest reactivity.

The results for the accident condition show that krff is M with the storage racks fully loaded with fuel of the highest anticipated reactivity, which is SVEA-96 Optinma2 Q122 lattice type 146, at a temperature corresponding to the highest reactivity.

The maximum calcu-lated reactivity for both normal and accident conditions includes a margin for uncertainty in reac-tivity calculations with a 95% probability at a 95% confidence level. Reactivity effects of abnormal and accident conditions have been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95.ProJect No. 2127 Report No. 111-2125245 Page 44

9. REFERENCES

[1] "MCNP -A General Monte Carlo N-Particle Transport Code, Version 5," Los Alamos National Laboratory, LA-UR-03-1987, April 24, 2003 (Revised 2/1/2008).

[23 L.I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T.Collins, August 19, 1998.[3] "Nuclear Group Computer Code Benchmark Calculations," Holtec Report 111-2104790 Revision 1.[4] M. Edenius, K. Ekberg, B.H. Forss~n, and D. Knott, "CASMO-4 A Fuel Assembly Burnup Program User's Manual," Studsvik/SOA-95/1; and J. Rhodes, K Smith,"CASM.O-4 A Fuel Assembly Burnup Programn User's Manual," SSP-01/400, Revision 5, Studsvik of America, Inc. and Studsvik Core Analysis AB (proprietary).

[5] D. Knott, "CASMO-4 Benchmark Against Critical Experiments," SOA-94/13, Studsvik of America, Inc., (proprietary);

and D. Knott, "CASMO-4 Benchmark Against MCNP," SOA-94/12, Studsvik of America, Inc., (proprietary).

[6'] DSS-ISG-2010-01, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, Revision 0.[7.] Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001.[8] 1-11-2002444, Latest Revision, "Final Safety Analysis Report for the HIl-STORM 100 Cask System", USNRC Docket 72-1014.[91 "Sensitivity Studies to Support Criticality Analysis Methodology," HI-2104598 Rev. 1, October 2010.[10] "Atlas of Neutron Resonances", S.F. Mughabghab, 5th Edition, National Nuclear Data Center, Brookhaven National Laboratory, Upton, USA.[11] "Spent Nuclear Fuel Burnup Credit Analysis Validation", ORNL Presentation to NRC, September 21, 2010.[12] An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (kfrr) Predictions, NUREG/CR-7109, April 2012.[13] OECD / NEA Data Bank, Java-based Nuclear Information Software, Janis version 3.3.Pro.ject No. 2127 Report No. HI-2125245 Page 45

[141 EPRI 1003222, "Poolside Examination Results and Assessment, GEl I BWR Fuel Exposed to 52 to 65 GWd/MTU at the Limerick 1 and 2 Reactors," December 2002.Project No. 2127 Report No. 1-11-2125245 Page 46

-i -m I m m 1 0 UI U-I. -I-m I '1-m i1-----I m m m 1 0-I I-I"-4 F -m 1 0.4,-m I m--4 --4--m I 1. ___________________________ .j t --Project No. 2127 Report No. 1I-1-2125245 Page 47

-U. -I -I m Project No. 2127 Report No. HI-2125245 Page 48 Project No. 2127 Report No. HI-2125245 Page 49 Project No. 2127 Report No. HI-2125245 Page 50 EEl Eli I umui I -I ~I -I~ I I -I -I=I 99 19 11 19 -is go 11 Project No. 2127 Report No. HI-2125245 Page 51 I E 1I I I II II I I II I I I *I I-I-f11111MIM Project No. 2127 Report No. HII-2125245 Page 52 Project No. 2127 Report No. 2125245 Page 53 I LM -11 11 11 11 I I in U go Project No. 2127 Report No. 1]1-2125245 Page 54

-~7j~__1 zIz m[A Ui Project No. 2127 Report No. 1-11-2125245 Page 55 Project No. 2127 Report No. 1-11-2125245 Page 56 Project No. 2127 Report No. HI-2125245 Page 57 Project No. 2127 Report No. HI-2 125245 Page 58

-I I I I~U=Fm U-Project No. 2127 Report No. HI-2 125245 Page 59 Project No. 2127 Report No. HI-2125245 Page 60

[I Projcct No. 2127 Report No. HI-2125245 Page 61!

I.'=Jmvmý--l III IN m III IN 0 C=3-----------

............. ..........IN ........ ... 0 IN IN IN IN ---- -------Project No. 2127 Report No. HI-2125245 Page 62

-ml-W VI I.E a.Project No. 2127 Report No. HI-2125245 Page 63 F q-Will ~-11 IN ll::Im IN 0 11 IN 0 11 IN ml IN m ml 0 m mi IN mi 1=1 m Milm--I- Will Project No. 2127 Report No. 1H1-2125245 Page 64 Results of the MCNP5-1.51 Table 7.2(a)Calculations for SVEA-96 Optima2 Q122 Lattices Description Burnup (GWd/mtU)kcleI sgina Max k-,l delta kCk Uncertainty (95/95)15 16 17 Lattice 146 17 (reference) 19 20 21 15 16 17 Lattice 147 (void) 18 19 20 21 15 16 17 Lattice 147 18 (water) tt 18 19 20 21 15 16 17 Lattice 148 18 19 20 21 Reference Reference-0,0081 0.0016-0.0097 0.0016-0.0121 0.0016 Note 2: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta may occur at an exposure which differs from that shown above.ProJect No. 2127 Report No. 11-2125245 Page 65 Table 7.2(a) Continued Max Uncert.Description IBurnup (GWd/mt1J' I sigma Max kc.41C delta k,,,, Uncert, (95/95)15 16 17 Lattice 149 (void) 19 20 21 15 16 17 Lattice 149 18 (w ater) 18 19 20 21 15 16 17 Lattice 150 18 S 19 20 21 15 16 17 Lattice 151 18 19 20_21-0.0207 0.0016-0.0189 0.0016_-0.01.54 0.0016-0.0111 0.0016 Note 2: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta ka,]may occur at an exposure which differs from that shown above.Project No. 2127 Report No. HI-2125245 Page 66 Table 7.2(b)Results of the MCNP5-1.51 Calculations for GE14 Lattice Type 5 Description Burnups Uncert.(GWd/mtU) sigma delta k (95/95)SVEA-96 Optima2 Q122 15.5 Reference Reference lattice type 146 Single GE14 13 -0.0543 0.0016 Single GE14 13.5 -0.0509 0.0015 Single GE14 14 -0.0491 0.0016 Single GEl4 14.5 -0.0469 0.0015 Single GEl4 15 -0.0473 0.0015 Single GEl4 15.5 -0.0479 0.0015 Single GEl4 16 -0.0485 0.0015 Single GE14 16.5 -0.0482 0.0015 Single GE14 17 -0.0500 0.0015 2x2 GEl4 -with channel (cell centered) (Case 2.3.1.5.4.1) 14.5 Reference Reference 2x2 GEI4 -no channel (Case 2.3.1.5.4.2) 14.5 -0.0044 0.0016 2x2 GE 14 -no channel/eccentric center (Case 2.3.1.5.4.3) 14.5 -0.0173 0.0015 2x2 GE14 -no channel / eccentric out (Case 2.3.1.5.4.4) 14.5 -0.0238 0.0015 Note 2: 1 he result ot the SV tA-Yb Uptlma2 Iz lattice type 140 is proviwea as tne reference.

Note 3: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta k,,] may occur at an exposure which differs from that shown above.Project No. 2127 Report No. HI-2125245 Page 67 Table 7.3 Results of the MCNP5-1.51 Calculations for Design Basis and Simplified Model of SVEA-96 Optima2 Q122 Lattice Type 146 Description Burnup Code kcal sigma (GWd/mtU)Simplified model of SVEA-96 Optima2 Q122 lattice 146 15.5 CASMO-4 Simplified model of SVEA-96 Optima2 Q122 lattice 146 15.5 MCNP5-1.51 (Case 2.3.1.4.2)

Model of SVEA-96 Optima2 Q122 lattice 146, similar to design basist 15.5 MCNP5-1.51 (Case 2.3.1.4.3)

Note 1: These calculations were performed using the design basis core operating parameters as indicated in Table 5,2(c).Project No. 2127 Report No. 1-11-2125245 Page 68 Table 7.4 Results of the MCNP5-1.51 Calculations for Core Operating Parameters Power Fuel Moderator Description Density Control Temp. Te Void Burnup kUnet sigma delta kCsk (595.D cpoD s Blade m Tmp. Fraction (%) (GWd/mtU) c (95/95)___ ___ ___ ___(W/gu)

Q09 ML__Design basis (reference) 23.688 Yes 1176 547 0 15.5 Fuel temperature decreasing 23.688 Yes 588 547 0 16 Moderator temperature decreasing 23.688 Yes 1176 528.8 0 15.5 Void fraction increasing 23.688 Yes 1176 547 94 22 Un-rodded operation 23.688 No 1176 547 0 17 24.1617 Yes 1276 547 0 15.5 ----24.1617 Yes 1376 547 0 15.5 20.1348 Yes 1176 547 0 15.5 Note 1: The burnup calculations for core operating parameters were performed from 14 GWd/mtU to 24 GWd/mtU. For each core operating parameter, only reactivity of the burnup in this range which results in the largest reactivity is reported.Note 2: The bounding case is bolded.Note 3: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta k.!c may occur at an exposure which differs from that shown above.Project No. 2127 Report No. HI-2125245 Page 69 Table 7.5 Results of the MCNP5-1.51 Calculations for the Effect of Water Temperature and Density Water Water Temperature Burnup Tep Dest a4)lne.Description (GWdmU Temp. Density Adjustment, kcaUe sigma delta k (9/5rt.(°F) (g/cc) S(alp) (95/95)

Reference:

lower bound temperature 15.5 39.2 1 j 68.81 Reference Ref.(Case 2.3.2.1) !Upper bound T temperature for normal operation, low 15.5 150 0.98026 68.81 -0.0041 0.0015 S(a4,3)(Case 2.3.2.2)Upper bound temperature for normal operation, 15.5 150 0.98026 170.33 -0.0066 0.0015 high S(a,p)(Case 2.3.2.3)Note 1: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kcai, may occur at an exposure which differs from that shown above.Project No. 2127 Report No. HI-2125245 Page 70 Table 7.6(a)Results of the MCNP5-1.51 Calculations for the Depletion Uncertainty Depletion Description kcaie sigma Uncertainty (5%)Design basis Reference Fresh fuel, no Gd 0.0064 Project No. 2127 Report No. 1I-2125245 Page 71.

U-1 -t- -- i ii-4~4~I~-m m I I I ProJect No. 2127 Report No. 1-11-2125245 Page 72 Table 7.7 Results of the MCNP5-1.51 Calculations for Fuel Tolerances Description Peak Reactivity Burnup (GWd/mtU)sigma delta k,.k Max delta kc,,, (95/95) (95/95)Design basis (reference) 15.5 Max fuel enrichment 16 Min fuel enrichment 15.5 Max Gd loading 16 Min Gd loading 15.5 Max pellet density 16 Min pellet density 15.5 Max pellet OD 15.5 Min pellet OD 16 Max clad ID 16 Min clad ID 16 Max clad OD 15.5 Min clad OD 15.5 Max sub-bundle pitch 15 Min sub-bundle pitch 16.5 Max pin pitch 15.5 Mill pin pitch 15.5 Reference 0.0026-0.0009-0.0013 0.0038 0.0000 0.0012 0.0015 0.0011 0.0010 0.0008-0.0002 0.0027 0.0098-0.0089 0.0122-0.0086 0.003 1 Reference 0.0026 0.0038 0.0012 0.0015 0.0010 0.0027 0.0098 Max combined water wing canal inner width, channel outer square width, channel corner inner radius and central water canal inner square width 15 lI -0.0122 0.0031 Min combined water wing canal inner width, channel outer square width, channel corner inner radius and central water canal inner sCIuare width 15.5-IM-0.0011 Max combination of channel wall thickness and water cross wall 16 0.0008 thickness 0.0019 Min combination of channel wall thickness and water cross wall 15.5 0.0019 thickness Statistical combination of fuel tolerances 0.0171 Note 1: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta k,,1, may occur at an exposure which differs from that shown above.Project No. 2127 Report No. HI-2125245 Page 73 Table 7.8 Results of the MCNP5-1.51 Calculations for Rack Tolerances Burnup dMax delta Description (GWd/mtU) ktalc sigma (95/95) kc (95/95)Design basis 15.5 Reference Reference (reference)

Max cell ID 15.5 -0.0093 N/A Max cell pitch Max wall thickness 15.5 0.0025 0,0025 Min wall thickness 15.5 0.0008 Max insert width 1 5.5 0005 0.0004 Min insert width 15.5 0.0004 Statistical combination of rack tolerances 0.0026 Project No. 2127 Report No. HI-2125245 Page 74 Table 7.9(a)Results of the MCNP5-1.51 Calculations for Fuel Radial Positioning in SFP Racks Burnup Unc.Description (GW d/mtU) ke~le sigma delta k,,,, (95/95)2x2 reference (Casre.e5.21e-15.5 Reference Ref.(Case 2.3.5.2. 1)2x2 eccentric center 15.5 -0.0053 0.0015 (Case 2.3.5.2.2) 2x2 eccentric in 15.5 -0.0081 0.0013 (Case 2.3.5.2.3) 2x2 eccentric out -0.0047 0.0014 (Case 2.3.5,2.4)

.....2x2 insert/cell center 15.5 0.0002 0.0013 (Case 2.3.5.2.5) 8x8 reference 15.5 Reference Ref.(Case 2.3,5.2.6) 8x8 eccentric center 15.5 -0.0023 0.0014 (Case 2.3.5.2.7) 8x8 eccentric in 15.5 -0.0080 0.0016 (Case 2.3.5.2.8) 8x8 eccentric out 15.5 -0.0035 0.0014 (Case 2.3.5.2.9) 8x8 insert/cell center 15.5 0.0016 0.0014 (Case 2.3.5.2.10) lxI reference 15.5 Reference Ref.(Case 2.3.5.2.11) lxl insert/cell center 15.5 0.0000 0.0015 (Case 2.3.5.2.12)

Pro-ject No. 2127 Report No. HI-2125245 Page 75 Table 7.9(b)Results of the MCNP5-1.51 Calculations for Fuel Orientation in SFP Racks Burnup k il sigma delta (9/95Unc.Description (GWd/mtU)

(95/95)Reference (Shown in 15.5 Reference Ref.Figure 2.9(a))Rotated fuel assembly (shown in Figure 2.9(b)) 15.5 -o -0 o0 0.0014 Rotated fuel assembly (shown in Figure 2.9(c)) 15.5 -0.0007 0.0014 Rotated fuel assembly (shown in Figure 2.9(d)) 15.5 -00013 0.0013 Rotated fuel assembly (shown in Figure 2.9(e)) 15 5 -0.0007 0.0013 Pro.ject No. 2127 Report No. 1-I1-2125245 Page 76

-Em s am -Err~W mmmm-m M-I UI --nn H* --Project No. 2127 Report No. HJ-2125245 Page 77 I I I____________--I" ________m m-__m Project No. 2127 Report No. HI-2125245 Page 78 Table 7.12(a)Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate the Effect of Nominal Values Instead of Using Minimum B 4 C Loading and Minimum Insert Thickness on Reactivity B-10 Areal Description Burnup )ensity kcaic sigma delta kC,,C (GWd/mtU) (g/cm 2)Reference (design basis) 15.5 0.0116 Reference (Case 2.4.1)Rack with nominal values for B 4 C loading and insert 15.5 0.0133 -0.0103 thickness (Case 2.4.2)Project No. 2127 Report No. 1-11-2125245 Page 79 Table 7.12(b)Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate the Effect of the Actual Optima2 Q122 Fuel Assembly Decipin Burnup keglic sigma Max k,,,,, delta k,,,I, Description (GWd/mtU)Optima2 Q122 Lattice 146 (Design 15.5 Reference Reference basis)(Case 2.4.1)Optima2 Q122 15.5 0.8873 Lattice 147 16 -Optima2 Q122 08 Lattice 148 16.5 0.8843 17 Optima2 Q122 02 Lattice 149 14.5 0.8825 15 Optima2 Q122 14.5 0.8863-Lattice 150 .......14 Optima2 Q122 14.5 0.8876 Lattice 151 Optima2 Q122 Peak Fuel Assembly t Reactivity 0.8925 -0.0066 (Case 2.4.3) Burnups (bolded)t The top and bottom natural lattice.blankets were conservatively neglected and replaced by adjacent Project No. 2127 Report No. HI-2125245 Page 80 Table 7.12(c)Margin Evaluation Summary of the Margin Evaluation Description Value Insert Composition Margin, fiom Table 7.12(a) -0.0103 Actual Optima2 Fuel Assembly Margin, fiotn -0.0066 Table 7.12(b)Calculated Margin -0.0169 Project No. 2127 Report No. HI-2125245 Page 81 U~ST-m 1m m =.mm V. _Vm m mm m m mm m --mm m m --m m m___________

m_____mm mm________

Project No. 2127 Report No. 1-11-2125245 Page 82 Table 7.13(b)Results of the MCNP5-1.51 Calculations for the Empty Storage Rack Cell without Insert Description Burnup Uncertainty (GWd/mtU) kcaie sigma delta kc (95/95)Design basis 15.5 Reference Reference (8x8 array)Empty storage cell (cell 15.5 -0.0041 0.0016 centered)Empty storage cell (eccentric) 15,5 -0.0081 0.0014 Note 1: The design basis fuel assembly (Optima2 Q122 Lattice Type 146) is used for these calculations.

ProJect No. 2127 Report No. HI-2125245 Page 83 I I I Project No. 2127 Report No. HI-2 125245 Page 84 Table 7.15 Results of the MCNP5-1.51 Calculations for SFR Interface Description Burnup kcR 1 c sigma delta k,, Uncertainty (GWd/mtU)

(95/95)Design basis 15.5 Reference Reference Full SFP (cell 15.5 -0.0008 0.0016 centered) 15.5_l _ -0_00_0.01 Full SFP (eccentric to 15.5 -0.0053 0.0015 SFP corner) I I Project No. 2127 Report No. HI-2125245 Page 85 at Project No. 2127 Report No. 1-11-2125245 Page 86 Figure Proprietary Project No. 2127 Report No. HI-2125245 P~age 87 Figure Proprietary ProJect No. 2127 Report No. HI-2125245 Page 88 Figure Proprietary Project No. 2127 Report No. HI-2125245 Page 89 Figure Proprietary Project No. 2127 Report No. HI-2125245 Page 90 Figure Proprietary Project No. 2127 Report No. I1A-2125245 Page 91 Figure Proprietary 11'ro.ject No. 2127 Report No. HI-2125245 Page 92 Figure Proprietary Project No. 2127 Report No. 1-11-2125245 Page 93 Figure Proprietary Project No. 2127 Report No. 1H1-2125245 Page 94 Figure Proprietary Project No. 2127 Report No. HI-2125245 Page 95 Figure Proprietary Project No. 2127 Report No. HI-2 125245 Page 96 Figure Proprietary Project No. 2127 Report No. 1-11-2125245 Page 97 Figure Proprietary Project No. 2127 Report No. 1,11-2125245 Page 98 Figure Proprietary Project No. 2127 Report No. 2125245 Page 99 Figure Proprietary Pro.ject No. 2127 Report No. HI-2125245 Page 100 Figure Proprietary Project No. 2127 Report No. HI-2125245 Page 101 Figure Proprietary Project No. 2127 Report No. HI-2125245 Page 102 Figure Proprietary ProJect No. 2127 Report No. 1-1-2125245 Page 103 Figure Proprietary ProJect No. 2127 Report No. 1-1-2125245 Page 104 Figure Proprietary ProJect No. 2127 Report No. 1-11-2125245 Page 105 Figure Proprietary Project No. 2127 Report No. 1-11-2125245 Pagec 106 Figure Proprietary Project No. 2127 Report No. 111-2125245 Page 107 Figure Proprietary Project No. 2127 Report No. 1-11-21.25245 Page 108 Figure Proprietary Project No. 2127 Report No. 1H1-2125245 Page 109 Figure Proprietary Project No. 2127 Report No. HI-2125245 Page 110 Figure Proprietary Project No. 2127 Report No. HI-2125245 Page I1I1 Appendix A Proprietary Appendix B Proprietary Project No. 2127 Report No. 141-2125245 Page B-1I Appendix C Proprietary Project No. 2127 Report No. HI-2125245 Page C-1 Supplement 1 Additional Calculations to Support the Revised NETCO-SNAP-IN Rack Insert Design (11 pages including this page)Project No. 2127 Report No. HI-2125245 Page S1 -I SL.I Introduction This Supplement documents the criticality safety evaluation for the storage of spent BWR fuel in the Unit 1 and Unit 2 spent fuel pools (SFPs) at Quad Cities Station operated by Exelon. The purpose of this analysis is to justify that the specified changes in the NETCO-SNAP-IN rack insert design [S1.1] are acceptable and bounded by the current analysis, presented in the main part of the report.SI.2 Methodology See Section 2 of the main report and as otherwise discussed below.SI.3 Acceptance Criteria See Section 3 of the main report.S1.4 Assumptions See Section 4 of the main report and as otherwise discussed below.S1.5 Input Data See Section 5 of the main report. The revised dimensions of the NETCO-SNAP-INO rack insert are presented in Table S I!- and Figure Si-I.S1.6 Computer Codes See Section 6 of the main report.S1.7 Analysis The comparison of the revised insert parameters presented in Table S I-I with the previous insert design in Table 5.3(b) shows that changes are minor and therefore a significant impact on the conclusions made in the main part of the report is not expected.

Nevertheless, to verify the negligible or minor impact of the revised insert design on results presented in the main part of the report additional calculations are presented in this Supplement.

The additional calculations presented in this Supplement are similar to those in report for the following cases:* SFP rack tolerances

Fuel assembly radial positioning in the SFP rack* Fuel orientation in the SFP rack These cases are selected because the NETCO-SNAP-IN rack insert design change may impact the reactivity in the rack. All other calculations from the main report are not affected by the NETCO-SNAP-INO rack insert design change and the results of the unaffected calculations are Project No. 2127 Report No. HI-2125245 Page S 1 -2 used in this Supplement where applicable.

This approach is considered for both normal and accident conditions.

S 1.7.1 SFP Rack Tolerances As discussed in Section SI .7, the effect of the manufacturing tolerances on reactivity of the SFP racks with revised inserts was determined.

The results of these calculations are presented in Table S 1-2. The maximum positive delta-k value for each tolerance is statistically combined.The maximum statistical combination of the SE-1P rack tolerances is used to determine kf-f in Tfable S1-5 and "Fable S1-6.S1.7.2 Fuel Assembly Radial Positioning in the SFP Rack As discussed in Section SI.7, twelve fuel assembly radial positioning cases in the racks were evaluated.

The results of these calculations are presented in Table S1-3. For each eccentric position case, the result for similar but cell centered case is considered as a reference.

The results show that most cases show a negative reactivity effect, however some delta k,,,, values are positive.

Therefore, a maximum delta kal, value is applied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in TFable S 1-5 and Table SI-6.S 1.7.3 Fuel Orientation in the SFP Rack As discussed in Section S1.7, five fuel assembly orientation cases in racks were evaluated.

The results of these calculations are presented in Table S 1-4. The result for the reference case is also included.

The results show that all cases are statistically equivalent and the reactivity effect of fuel orientation is negligible.

Nevertheless, a maximum positive delta kca,, value is applied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in Tfable S1-5 and Table S 1-6.S 1.7.4 Maximum ke 1.Calculations for Normal Conditions The calculations of the maximum k~ff for normal conditions are described in Section 2.3.8 of the main part of the report. The results for the revised NETCO-SNAP-IN@D rack insert design and the results from the main part of the report are tabulated in Table S 1-5. The results show that the maximum k~ff for the normal conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95% confidence level for the revised NETCO-SNAP-INO rack insert design and are bounded by the results from the main part of the report.Project No. 2127 Report No. HI-2125245 Page S 1-3 SI.7.5 Maximum kLIT Calculations for Abnormal and Accident Conditions Tile calculations of the maximum kew- for accident conditions are described in Section 2.6 of the main part of the report. The bounding accident case from the main report is recalculated using the revised NETCO-SNAP-IN rack insert design. The results for the revised NETCO-SNAP-IN' rack insert design and the results firom the main part of the report are tabulated in Table S1-6. The results show that the maximum kerr for abnormal and accident conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95% confidence level for the revised NETCO-SNAP-INO rack insert design and are bounded by the results from the main part of the report.S1.8 References

[SI.1] Transmittal of Design Infformation NF 1100434, Revision 1, "Quad Cities SFP Rack Insert Design Information", dated 09/11/2012.

SI.9 Conclusions The criticality analysis for the storage of BWR assemblies in the Quad Cities SFP racks with revised NETCO-SNAP-IN inserts has been performed.

The results show that k~ff is M with the storage racks fully loaded with fuel of the highest anticipated reactivity, which is SVEA-96 Optima2 Q122 Lattice Type 146, at a temperature corresponding to the highest reactivity.

The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations with a 95% probability at a 95% confidence level. Reactivity effects of abnormal and accident conditions have been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95.The results show that the specified changes in the insert de tliri r.i-ripnt naltucie nvirotri-d

n t, mainn rin,' if thei ,-rtnnrt However, tor tie oounaing acciaent case, tne use oi me new insert teaas to larger biases that impact the overall margin. This impact has been accounted for in Table 7.14 in the main report. Therefore, any insert width dimension between the value used in the main report including the specified manufacturing tolerances and the value evaluated in this Supplement is acceptable.

Project No. 2127 Report No. 1-11-2 125245 Page S 1 -4 Table SI-I Fuel Rack Insert Revised Dimensions

[S1.1]-For the details of the insert dimensions, see t See Table 5.3(b)Figure S 1-1.Project No, 2127 Report No. 1H1-2125245 Page S 1 -5 Table S 1-2 Results of the MCNP5 Calculations for Revised Rack Tolerances Revised Reference Burnup delta k,,k Max delta Max delta Description (GWd/mtU) kc~ii sigma (95/95) kcade kcd (95/95) (95/95)Design basis 15.5 Reference Reference Reference (reference)

Max cell ID 15.5 1 1.00o91 o.oooo o.oooo Max cell pitch Max wall thickness 15.5 1 .0.0017 0.0025 Min wall thickness 15.5 .0.00111 Max insert width 15.5 0.0016 Min insert width 15.5 0.0030 Statistical combination of rack tolerances 0.0035 0.0026 I See Table 7.8 Note 1: The CASMO depletion calculation filenames are op146-dbc(-ac).

Project No. 2127 Report No. 1-11-2125245 Page S 1 -6

'fable S 1-3 Results of the MCNP5-1.51 Calculations for Revised Fuel Radial Positioning in SFP Racks Revised Reference Burnup Revised U. Reference Description (GWd/mtU) k~alc sigma delta kC (95/95) delta U9/5___(95/95)__

(9.5195)2x2 reference 15.5 Ref Ref Ref Ref (Case 2.3.5.2. 1)2x2 eccentric center 15.5 -0.0028 0.0015 -0.0053 0.0015 (Case 2.3.5.2.2) 2x2 eccentric in 15.5 -0.0054 0.0015 -0.0081 0.0013 (Case 2.3.5.2.3) 2x2 eccentric out (Case 2.3.5.2.4) 15.5 .-0.0014 0.0015 -0.0047 0.0014 2x2 insert/cell center 15.5 0.0001 0.0016 0.0002 0.0013 (Case 2.3.5.2.5) 8x8 reference 15.5 Ref Ref Re.. Ref (Case 2.3.5.2.6)5e 8x8 eccentric center 15.5 -0.0032 0.0015 -0.0023 0.0014 (Case 2,3.5.2.7)

..........

8x8 eccentric in 15.5 -0.0071 0,0015 -0,0080 0.0016 (Case 2.3.5.2.8) 8x8 eccentric out 155 -0.0035 0.0016 -0.0035 0.0014 (Case 2.3.5.2.9)_.

8x8 insert/cell center 15.5 0.0009 0.0014 0.0016 0.0014 (Case 2.3.5.2.10) lxi reference 15.5 Ref Ref Ref Ref (Case 2.3.5.2.11) 1 lxl insert/cell center 15.5 0.0004 0.0015 0.0000 0.0015 (Case 2.3.5.2.12) , -1 1" See Table 7.9(a)Note 1: The CASMO depletion calculation filenames are opl46-dbc(-ac).

Project No. 2127 Report No. HI-2125245 Page S 1-7 Table S 1-4 Results of the MCNP5-1.51 Calculations fbr Revised Fuel Orientation in SFP Racks Revised Reference B"Imup Revised Reference R Decipinurnup kRevised IJnIc. t Unc.t Description (GWdhntU) kcalc sigma delta k, (95/95) delta k.,, (95/95)Reference (Shown in 15.5 Ref. Ref. Ref. Ref.Figure 2.9(a)) 1 Rotated fuel assembly 15.5 0.0004 0.0014 -0.0008 0.0014 (shown in Figure 2.9(b))Rotated ffiel assembly 5 00011 0.0015 -0.0007 0.0014 (shown in Figure 2.9(c))Rotated fuel assembly 155 0,0016 0.0014 -0.0013 0.0013 (shown in Figure 2.9(d))Rotated ffiel assembly 15.5 0.0024 0.0016 -0,0007 0.0013 (shown in Figure 2.9(e))I See Table 7.9(b)Note 1: The CASMO depletion calculation filenames are opl46-dbc(-ac).

Project No. 2127 Report No. 1-11-2125245 Page S 1 -8 I I I* I-I mI-I_____________________________I Liii!I.-~ ~ .-. --.i ---L L________________-

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-t ~--I I I- -I ______I-- --1-------I L -L Project No. 2127 Report No. HI-2 125245 Page S 1 -9 I I I Project No. 2127 Report No. 2125245 Page SI-10 Figure Proprietary ProJect No. 2127 Report No. 1-11-2125245 Page SI-11I

ML14101A228: Difference between revisions

Revision as of 17:05, 12 October 2018

Contents

Text

2.6.8 Insert

2.1 General

2.3 Analysis

2.3.5 Radial

2.6 Accident

2.6.2 Dropped

2.6.6 Mislocated

2.7 Spent

2.8 Reconstituted

5.4 Storage

7.1.3 Optima2

7.1.5 Water

Reference:

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ML14101A228: Difference between revisions

Revision as of 17:05, 12 October 2018

Text

2.6.8 Insert

2.1 General

2.3 Analysis

2.3.5 Radial

2.6 Accident

2.6.2 Dropped

2.6.6 Mislocated

2.7 Spent

2.8 Reconstituted

5.4 Storage

7.1.3 Optima2

7.1.5 Water

Reference:

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