ML14024A562: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
 
(Created page by program invented by StriderTol)
Line 16: Line 16:
| page count = 251
| page count = 251
}}
}}
=Text=
{{#Wiki_filter:ATTACHMENT 4Holtec International Report No. HI-2104790, Revision 1,"Nuclear Group Computer Code Benchmark Calculations" U.HOINTERHEWLTECNATIO0NAL Holtec Center. 555 Lincoln Drive West, Marlton, NJ 08053Telephone (856) 797- 0900Fax (856) 797 -0909Nuclear Group Computer Code Benchmark Calculations FORGENERIC:
NON-PROPRIETARY VERSIONHoltec Report No: HI-2104790 Holtec Project No: GENERICSponsoring Holtec Division:
HTSReport Class: SAFETY RELATED Summary of Revisions Revision 0Original IssueRevision 1Additional criticality experiments were added to Appendix B. The index numbers ofcriticality experiments in Appendix C were updated to be consistent with a new revisionof Appendix B. The benchmark of MCNP5-1.51 with ENDF/B-VII was added inAppendix D.REPORT HI-2104790 i
Table of Contents1.0 Introdu ction .........................................................................................................................
32.0 M ethodology
.......................................................................................................................
32.1 Determination of Bias and Bias Uncertainty
................................................................
32.2 Statistical Methods ......................................................................................................
42.2.1 Single Sided Tolerance Limit Method ..................................................................
42.2.2 Confidence Band with Administrative Margin Method ........................................
52.2.3 Non-parametric Statistical Treatment Method ......................................................
62.3 Area of Applicability
..............................................
82.3.1 Key Parameters Identification
................................................................................
82.3.2 Screening Area of Applicability
..........................................................................
93 .0 A ssu m ption s ........................................................................................................................
94.0 Computer Files ............................................................................................................
95 .0 S u m m ary .............................................................................................................................
96 .0 R eferences
.........................................................................................................................
10Appendix A: Holtec Approved Computer Program List .............................................
A-1Appendix B: Description of the Critical Experiments
.............................................
B-1Appendix C: Benchmark of MCNP5-1.51 with ENDF/B-V
.....................................
C-1Appendix D: Benchmark of MCNP5-1.51 with ENDF/B-VII
......................................
D-1REPORT HI-2 104790 iiREPORT HI-2104790 ii
==1.0 Introduction==
This report documents the criticality experiment benchmark validation calculations for thefollowing computer codes and libraries combinations and establishes the criticality code bias andbias uncertainty for these codes:MCNP5-1.51 with ENDF/B-V (Appendix C)MCNP5-1.51 with ENDF/B-VII (Appendix D)For that purpose, results from the codes are compared to the critical experiments referred to asthe Haut Taux de Combustion (HTC) experiments and to the selected
: critical, presented inAppendix B, with geometric and material characteristics similar to that of spent fuel storage andtransport casks. The simulated fuel rods used in these experiments contained uranium or mixtureof uranium and plutonium oxides. In the HTC experiments the plutonium-to-uranium ratio andthe isotopic compositions of both the uranium and plutonium were designed to be similar to whatwould be found in a typical pressurized-water reactor (PWR) fuel assembly that initially had anenrichment of 4.5 wt % 235U and was burned to 37,500 MWd/MTU.The purpose of the calculation is to determine the code bias and bias uncertainty consistent withstandards such as ANSI/ANS-8.1
[1] and ANSI/ANS-8.17
[2]. Criticality safety standards ANSI/ANS-8.1 and ANSI/ANS-8.17 apply to criticality methods validation and to criticality evaluations, respectively.
ANSI/ANS-8.1 requires that a validation be performed on the methodused to calculate criticality safety margins and that the validation must be documented in awritten report describing the method, computer program and cross section libraries used, theexperimental data, the areas of applicability and the bias and margins of safety. ANSI/ANS-8.17 prescribes the criteria to establish sub-criticality safety margins.2.0 Methodology Validation of the computer code and continuous energy data library to perform criticality safetycalculation has been performed following reference
[5] methodology.
The validation allows theunderstanding of the accuracy of the calculational methodology to predict subcriticality.
Validation includes identification of the difference between calculated and experimental neutroneffective multiplication factor (keff), called the bias. A set of appropriate critical experiments areselected so bias trends can be drawn through statistical analyses.
The range of the benchmark parameters used to validate the calculational methodology primarily defines the area ofapplicability (AOA), which establishes the limits of the systems that can be analyzed using thevalidated criticality safety methodology.
Determination of Bias and Bias Uncertainty Following reference
[5] guide, the statistical analysis to determine the mean multiplication factor(keff) and the bias uncertainty (Sp) approach involves determining the weighted mean thatincorporates the uncertainty from both, measurements and calculation method as follows:REPORT HI-2104790 3
= a + Uep (2-)where ai is the uncertainty for the ith keff, exup is the measurement uncertainty and oc,,Ic-i is thecalculation uncertainty.
Then, the weighted mean multiplication factor keffand the biasuncertainty (Sp) are given by:keff --So1ar1(2-2)S,= N[s2 + j 2 (2-3)where s2 is the variance about the mean and j2 is the average total uncertainty, given by:2 21 1(2-4)-2_ n 1(2-5)where n is the number of critical experiments used in the validation and keg.i is the ith value of themultiplication factor.Bias is determined by the relation:
Bias = keff -1 if keff is less than 1, otherwise Bias = 0 (2-6)Because a positive bias may be nonconservative, a bias is set to zero if the calculated average keffis greater than one.Statistical MethodsSingle Sided Tolerance Limit MethodIf the benchmark calculated neutron multiplication factor does not exhibit trends with theparameters, the lower tolerance limit or single sided tolerance limit method can be used. Aweighted lower limit tolerance (KL) is a single lower limit above which a defined fraction of thepopulation of keg is expected to lie, with a prescribed confidence and within the area of theapplicability.
The term "weighted" refers to a specific statistical technique where theREPORT HI-2104790 4
uncertainties in the data are used to weight the data point. Data with high uncertainties will haveless "weight" than data with small uncertainties.
A lower tolerance limit can be used when there are no trends apparent in the critical experiment results and the critical experiment results have a normal distribution.
The method is applicable only within the limits of the validation data without extrapolating the AOA. The single sidedlower tolerance limit is defined by the equation:
KL = keff -U X Sp (2-7)Ifkeff >- 1, then KL = 1 -U X Sp (2-8)where Sp is the square root of the pooled variance used as the mean bias uncertainty whenapplying the single sided tolerance limit for a normally distributed data and U is the single sidedlower tolerance factor, determined from the following equations
[6]. Note that for groups withlarger than 50 samples, the single sided lower tolerance factor for 50 samples was conservatively used.z1Pz+ z -abU =a(2-9)2a = z1-Y2(N-1)(2-10)22 ____b =1- N pz-N(2-11)where zj-p is the critical value from the normal distribution that is exceeded with probability i-pand z1.y is the critical value from the normal distribution that is exceeded with probability 1-y.Confidence Band with Administrative Margin MethodIf the benchmarks calculated neutron multiplication factor exhibit a trend with a given parameter, the method based on a confidence band with administrative margin can be used. This methodapplies a statistical calculation of the bias and its uncertainty plus an administrative margin to alinear fit of the critical experiment benchmark data.The confidence band W is defined for a confidence level of (1-y) using the relationship:
W = max {w(xi,),w(x,,,)}
(2-12)whereREPORT HI-2104790 5
w(x) = tj~ x sp 1 + -n+ --)(2-13)andn is the number of critical experiments used in establishing kcaI(x),tl-Y is the Student-t distribution statistic for 1-y and n-2 degrees of freedom,2 is the mean value of the parameter x in the set of calculations, Xmin, x, are the minimum and maximum values of the independent parameter x,Sp is the pooled standard deviation for the set of criticality calculations given by:= S2(x) + S~w (2-14)where S2k(x) is the variance of the regression fit and is given by:Sk(x) = (n -2) 1 (ke f-1 -k --g) }2-(2-15)k is the mean value of the calculated keff and sw2 is the within-variance of the data:S'2 1 1 ini=l,n(2-16)where q1 -traic-i + rex is the uncertainty for the ith keff, ep is the measurement uncertainty and Ucak.i is the calculated uncertainty.
Non-parametric Statistical Treatment MethodData that do not follow a normal distribution can be analyzed by non-parametric techniques.
Theanalysis results in a determination of the degree of confidence that a fraction of the truepopulation of data lies above the smallest observed value. The more data is available in thesample, the higher the degree of confidence.
The following equation determines the percent confidence that a fraction of the population isabove the lowest observed value:REPORT HI-2 104790 6REPORT HI-2104790 6
m-Ifl = 1 -j! (n --j)! (1 -q)Iqn-j(2-17)whereq is the desired population fraction (normally 0.95),n is the number of data in one data sample,m is the rank order indexing from the smallest sample to the largest (m=l for the smallestsample; m=2 for the second smallest sample, etc.). Non-parametric techniques do notrequire reliance upon distributions, but are rather an analysis of ranks. Therefore, thesamples are ranked from the smallest to the largest.For a desired population fraction of 95% and a rank of order of 1 (the smallest data sample),
theequation reduces to:= l-q" = 1-0.95" (2-18)This information is then used to determine the Non-parametric Margin from Table 2.2 inReference
[5].For non-parametric data analysis, KL is determined by:KL = Smallest keff value -Uncertainty for Smallest keff- Non-parametric Margin (NPM) (2-19)Single-Sided Tolerance Band MethodWhen a relationship between a calculated keff and an independent variable can be determined, asingle-sided lower tolerance band may be used. This is a conservative method that provides afitted curve above which the true population of keff is expected to lie. The tolerance bandequation is actually a calibration curve relation.
The equation for the single-sided lower tolerance band isKL = Kfit(x) -Spfit I2F(a+n +-Z) (n-2a(x)x yn2(2-20)where:KIt(x) is the function derived from the trend analysis, p is the desired confidence (0.95),gat'"'2) is the F distribution percentile with degree of fit, n-2 degrees of freedom.
Thedegree of fit is 2 for a linear fit,REPORT HI-2104790 7
n is the number of critical experiment keff values,x is the independent fit variable, xi is the independent parameter in the data set corresponding to the ,ith,' keff value,;? is the weighted mean of the independent variables, Z2P.l is the symmetric percentile of the Gaussian or normal distribution that contains the Pfraction, y = (I -p)/2, (2-21)2X 1-y,n-2 is the upper Chi-square percentile,
_- S 2 5 2Spfit f= t + (2-22)= n- -2 To.2 [keff. -fitt(Xi)]
21Slit = 1 1(2-23)Area of Applicability The area(s) of applicability refers to the key physical parameter(s) that define a particular fissileconfiguration.
This configuration can either be an actual system or a process.
The determination of the AOA of the validation is determined following NUREG/CR-6698 steps [5]. The approachused in developing the AOA consists of the following steps:i. Identification of the key parameters associated with the system to be evaluated.
ii. Establishment a "screening" AOA for critical experiments.
iii. Identification of criticality experiments that are within the "screening" AOA.iv. Determination of the detailed AOA based on the selected criticality benchmark experiments.
: v. Demonstration that the system to be evaluated in within the AOA provided by the criticalexperiments.
Steps i. and ii. are presented in subsections 2.3.1 and 2.3.2, respectively.
Step iii. is presented inAppendix B. Steps iv. and v. are presented in Appendix C and D.Key Parameters Identification REPORT HI-2 104790 8REPORT HI-2104790 8
This validation will cover a number of designs but all the designs will consider the same keyparameters in defining the applicability area. These parameters fall into three categories:
materials, geometry and neutron energy spectra.Regarding
: material, the fuel is a uranium or mixture of uranium and plutonium oxides pelletsclad in a zirconium alloy. The moderator and reflector is water which in some cases hasdissolved boron. or gadolinium solutions.
Absorber plates made of borated steel, Boral, ZircaloyBoroflex or cadmium and absorber rods made of steel, aluminum, Gd203, Pyrex, Vicor orborated aluminum will be included in this validation.
Some experiments were performed withsteel or lead reflector screens.Regarding
: geometry, the fuel in the HTC experiments is in square lattices with pin diameter
-9.5 mm and pitch in the range found on Table B-1 through Table B-6. The geometry parameters of other selected critical experiments are varied in a wide range and they can be found inreferences
[B.6] through [B.12]. The fuel assemblies may be separated by water, water and anabsorber plate or water and absorber rods. The system may be water reflected or steel/lead reflected.
Regarding the neutron energy spectra, they are thermal with EALF values in the range of 0.07and 1.55 eV.Table 2-1 presents the key physical parameters for AOA selected.
Screening Area of Applicability For the key parameters selected in section 2.3.1, Table 2-1 summarizes the range of parameters for which the validation applies.
These data are the base for the selection of the criticalexperiments, which span the range of parameters.
==3.0 Assumptions==
No substantial simplifying assumptions were made in the modeling of the critical experiments used for benchmarking:
all experiments were modeled as full three-dimensional geometries, fuelrod arrays were modeled as lattices, all fuel rod details were modeled, and the water between therods was modeled as specified in the experiment description.
: However, structures further awayfrom the experiment, such as building walls and foundations, were not included in the models.4.0 Computer FilesAll computer files to support this analysis are provided on the Holtec server in\Projects\0\Reports\HI-2104790 and its subdirectories.
5.0 SummaryThe criticality experiment benchmark validation calculations for the computer codes andlibraries shown in Section 1.0 were performed for the validation of the Holtec International REPORT HI-2104790 9
criticality safety methodology.
The results of calculations and the criticality code bias and biasuncertainty for these codes are presented in appropriate appendices.
The similarity between thechosen experiments and the actual systems has been based on a set of screening criteria as isstated in the NUREG/CR-6698
[5].The summary of biases and bias uncertainties for the validated computer codes is shown in Table5.1.6.0 References
[1] ANSI/ANS 8.1-1983, American National Standard For Nuclear Criticality Safety InOperations With Fissionable Materials Outside Reactors, American Nuclear Society, LaGrange Park, Illinois.
[2] ANSI/ANS-8.17, "American National Standard for Criticality Safety Criteria for theHandling,
: Storage, and Transportation of LWR Fuel Outside Reactors,"
American NuclearSociety, La Grange Park, Illinois.
[3] Criticality Benchmark Guide for Light Water Reactor Fuel in Transportation and StoragePackages, NUREG/CR-6361 (ORNL/TM-1321 1), U.S. Nuclear Regulatory Commission, March 1997.[4] J.R. Taylor, An Introduction to Error Analysis (University Science Books, Mill Valley,California, 1982).[5] Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, U.S. Nuclear Regulatory Commission, January 2001.[6] M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91,August 1963.REPORT I-11-2104790 10 Table 2-1 Key Criticality System Parameters and Range of those Parameters in Expected DesignsParameter Critical Experiment Requirement Range of Key Parameters Fissionable Material 235U, 239Pu, 241Pu 235U, 239Pu, 241PuIsotopic Composition 235U/Ut < 5.Owt% 0. 16wt% to 5.74wt%Pu/(U+Pu)
< 20wt% 1. 104wt% to 20wt%/oPhysical Form UO2,MOX UO2,MOXModerator Material (coolant)
H HPhysical Form H20 H20Density Normal pressure
& temperature around 1.0 g/cm3condition Reflector Material H HPhysical Form H20 H20Density Normal pressure
& temperature around 1.0 g/cm3condition Interstitial Reflector MaterialPlate Steel or Lead Steel or LeadAbsorber MaterialNone, Boron (0 to 2550 ppm) orSoluble None, Boron or Gadolinium Gadolinium (0 to 197 ppm)Rods Boron Pyrex, Vicor, Steel or B-Al ISeparating MaterialWater, B-SS, Boral, Boroflex, Plate Water, B-SS, Boral or Cadmium Zircaloy or CadmiumGeometryFuel Square/Triangle lattice of fuel Square/Triangle lattice of fuelpins pinsNeutron Energy Thermal spectrum Thermal spectrumREPORT HI-2104790 11REPORT HI-2104790 I1I Table 5-1 Summary of Biases and Bias Uncertainties for the Validated Computer CodesComputer Code Total Bias Bias Uncertainty MCNP5-1.51 with ENDF/B-V (Appendix C)MCNP5-1.51 with ENDF/B-VII (Appendix D)REPORT HI-2 104790 12REPORT HI-2104790 12 Appendix AHoltec Approved Computer Program List(total number of pages: 5 including this page)Appendix Proprietary REPORT HI-2 104790 A-IREPORT HI-2104790 A-1 Appendix BDescription of the Critical Experiments (total number of pages: 16 including this page)REPORT HI-2104790 B-iREPORT HI-2104790 B-I B.1. Introduction and PurposeThe purpose of this Appendix is to document the description of the full set of criticalexperiments selected for the benchmark validation of computer codes.B.2. Physical Description of HTC Critical Experiments In the 1980s, a series of critical experiments referred to as the Haut Taux de Combustion (HTC)experiments was conducted by the Institut de Radioprotection et de Sfiretd Nucl~aire (IRSN) atthe experimental criticality facility in Valduc, France, between 1988 and 1990. The fuel rodswere fabricated specifically for this set of experiments.
The fuel consisted of 1-cm-long pelletscontained within Zircaloy-4 cladding.
The plutonium-to-uranium ratio and the isotopiccompositions of both the uranium and plutonium used in the simulated fuel rods were designedto be similar to what would be found in a typical pressurized-water reactor fuel assembly thatinitially had an enrichment of 4.5 wt % 235U and was burned to 37,500 MWd/MTU.
The fuelmaterial also includes 241Am, which is present due to the decay of 241Pu. The fuel rods were heldin place by an upper and a lower grid and were contained in one or four assemblies placed into arectangular tank. The critical approach was accomplished by varying the water or solution levelin the tank containing the fuel pin arrays. The critical condition was extrapolated from asubcritical configuration with a multiplication factor within 0.1% of 1.000.This section provides a summary description of the materials and physical layouts of the 156critical configurations.
Detailed descriptions of the critical experiments are presented inreferences
[B.1] through [B.4]. The HTC experiments include configurations designed tosimulate fuel handling activities, pool storage, and transport in casks constructed of thick lead orsteel and were categorized into four phases.B.2.1. Phase 1: Water-Moderated and Reflected ArraysThe first phase included 18 configurations, each involving a single square-pitched array of rodswith rod pitch varying from 1.3 to 2.3 cm.The tank was incrementally filled with water at room temperature, water being injected at thebottom of the tank. A measurement needle provided water height. Therefore, the water was usedas core moderator and as reflector beneath the fuel and around the array on four sides. Thecritical approach parameter was the water level.Eighteen experiments have been performed with various arrays and all are considered acceptable for use as benchmark experiments:
0 5 square or almost square array -square pitch 1.3, 1.5, 1.7, 1.9, 2.3 cm -15 experiments,
* 1 rectangular centered array -square pitch 1.7 cm -2 experiments,
* 1 rectangular no-centered array -square pitch 1.7 cm -1 experiment.
The experiments key physical parameters are summarized in Table B-1.REPORT HI-2 104790 B-2REPORT HI-2104790 B-2 B.2.2. Phase 2: Reflected Simple Arrays Moderated by Poisoned Water with Gadolinium or BoronThe second phase included 41 configurations that were similar to the first phase except that thewater used as moderator and reflector included either boron or gadolinium in solution at variousconcentrations.
The tank was incrementally filled with poisoned solution at room temperature, this solutionbeing pumped in the bottom of the tank. A measurement needle provided solution height. Thecritical approach parameter was the water level.Forty one experiments are evaluated and all are considered acceptable for use as benchmark experiments.
Twenty of them are performed with gadolinium solutions, and the others withboron solutions.
The experiments key physical parameters are summarized in Table B-2 through Table B-3.B.2.3. Phase 3: Pool StorageThe third phase simulated fuel assembly storage rack conditions and included 26 configurations with 1.6 cm square rods pitch arranged into four assemblies in a 2 x 2 array. These assemblies with, in some cases, canisters, were placed on a pedestal centered inside a parallelepiped tankwhich was itself located on the floor in the middle (approximately) of a large room. The spacingbetween assemblies was varied, and some of the assemblies had B-SS, Boral, or cadmiumplates attached to the sides of the four assemblies.
The tank was incrementally filled with water at room temperature, water being pumped in at thebottom of the tank. A measurement needle provided water height. Therefore, the water was usedas core moderator and as reflector beneath the fuel and around the array on four sides. Thecritical approach parameter was the water level.Twenty six experiments are evaluated and all are considered acceptable for use as benchmark experiments.
Eleven of them were performed with neutron absorbing canisters around the fourarrays, and the others without any.The experiments key physical parameters are summarized in Table B-4.B.2.4. Phase 4: Shipping CaskThe fourth phase simulated cask conditions and included 71 configurations similar to the Phase 3configurations except thick steel or lead shields were placed around the outside of the 2 x 2 arrayof fuel assemblies.
These assemblies with, in some cases, canisters, were placed on a pedestalcentered inside a parallelepiped tank which was itself located on the floor in the middle(approximately) of a large room. Space between assemblies and between assemblies and screenvaried from one case to another.REPORT HI-2104790 B-3REPORT HI-2104790 B-3 The tank was incrementally filled with water at room temperature, water being pumped in at thebottom of the tank. A measurement needle provided water height. Therefore, the water was usedas core moderator and as reflector beneath the fuel and around the array on four sides behind thereflector screens.
The critical approach parameter was the water level.Seventy one experiments are evaluated and all are considered acceptable for use as benchmark experiments.
Thirty eight experiments were performed with lead reflector screens and thirtythree with steel reflector screens.
Twenty six among the former and twenty one among the latterused absorbing canisters around the four arrays, and the others without any.The experiments key physical parameters are summarized in Table B-5 through Table B-6.B.3. Physical Description of the Selected Benchmark Critical Experiments The benchmark experiments are selected to cover a wide range of code applications for fresh andspent fuel storage analysis.
This section provides a summary description of the materials andphysical layouts of the 135 critical configurations with fresh and selected actinides for spent fuel.For the fresh fuel assumption, the code is compared to the critical experiments of un-irradiated U02 systems with geometric and material characteristics similar to that of fuel storage systems.For the spent fuel assumption with bumup credit, additional comparisons are made to un-irradiated mixed-oxide (MOX) fuel of similar characteristics to spent fuel. The U02 experiments 234 235 238 28 239 24 24address U, U and U. The MOX critical experiments address 238Pu, Pu, 24&deg;pu, 241pu,242pu and 241Am. Detailed descriptions of the critical experiments are presented in references
[B.6] through [B. 12].Description of the selected critical experiments is summarized in Table B-7.B.4. References
[B.1] F. Fernex, "Programme HTC -Phase 1 : R~seaux de crayons dans l'eau pure (Water-moderated and reflected simple arrays) Rddvaluation des experiences,"
DSU/SEC/T/2005-33/D.R.,
Institut de Radioprotection et de Sfiret& Nuclraire, 2008.[B.2] F. Fernex, Programme HTC -Phase 2 : Rrseaux simples en eau empoisonnre (bore etgadolinium)
(Reflected simple arrays moderated by poisoned water with gadolinium orboron) Rddvaluation des experiences,"
DSU/SEC/T/2005-38/D.R.,
Institut deRadioprotection et de Sfiretd Nuclraire, 2008.[B.3] F. Fernex, "Programme HTC -Phase 3 : Configurations "stockage en piscine" (Poolstorage)
R66valuation des expdriences,"
DSU/SEC/T/2005-37/D.R.,
Institut deRadioprotection et de SfiretW Nuclaire, 2008.[B.4] F. Fernex, "Programme HTC -Phase 4 : Configurations "chateaux de transport" (Shipping cask) -R66valuation des exp6riences,"
DSU/SEC/T/2005-36/D.R.,
Institut deRadioprotection et de Sfiret6 Nuclraire, 2008.REPORT HI-2104790 B-4
[B.5] C. Portella, C. Woillard "Programme "HTC" -Experiences de criticit6 avec des crayonscombustibles HTC (type REP A haut taux de combustion)
-Rdsultats de l'6tudeparamrtrique avec de l'eau gadolinire."
[Translation:
.... Hbu" program -Criticity Experiments with Hbu fuel rods (LWR type at high bum up) -Results of parametric study with poisoned water with gadolinium."]
Note technique IPSN/SRSC n' 90.01.[B.6] International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03, NEA Nuclear Science Committee, September 2008 Edition[B.7] G.S. Hoovier et al., Critical Experiments Supporting Underwater Storage of TightlyPacked Configurations of Spent Fuel Pins, BAW-1645-4, Babcock & Wilcox Company,November 1991.[B.8] L.W. Newman et al., Urania Gadolinia:
Nuclear Model Development and CriticalExperiment Benchmark, BAW- 1810, Babcock and Wilcox Company, April 1984.[B.9] J.C. Manaranche et al., "Dissolution and Storage Experimental Program with 4.75%Enriched Uranium-Oxide Rods," Trans. Am. Nucl. Soc. 33: 362-364 (1979).[B. 10] S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990.[B.1 1] S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in OrganicModerator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1986.[B. 12] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton PartialPlutonium core, WCAP-3385-54, Westinghouse Electric Corp., Atomic Power Division, December 1965.[B. 13] Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data,NUREG/CR-6979 (ORNL/TM-2007/083),
U.S. Nuclear Regulatory Commission, September 2008.REPORT HI-2 104790 B-5REPORT HI-2104790 B-5 Table B-I Key Physical Parameters of the HTC Phase I Critical Experiments
[B. 1]Case Reference Experiment Pitch Number of Rods Date of Temperature Critical waternumber (cm) Along edge Total experiment
(&deg;C) height (cm) (a)1 MIX-COMP-THERM-HTC-001 2327 50 x 50 2500 05/05/88 22.5 61.41 + 0.062 MIX-COMP-THERM-HTC-002 2335 2.3 38 x 37 1406 06/06/88 21.1 87.68 + 0.063 MIX-COMP-THERM-HTC-003 2336 37 x 37 1369 06/07/88 21.0 90.38 + 0.064 MIX-COMP-THERM-HTC-004 2337 27 x 27 729 06/09/88 20.7 63.77 + 0.065 MIX-COMP-THERM-HTC-005 2339 1.9 25 x 25 625 06/13/88 20.5 81.95 + 0.086 MIX-COMP-THERM-HTC-006 2340 25 x 24 600 06/14/88 20.7 90.22 + 0.067 MIX-COMP-THERM-HTC-007 2341 26 x 26 676 06/15/88 20.2 65.11 +/- 0.078 MIX-COMP-THERM-HTC-008 2342 1.7 25 x 25 625 06/16/88 21.0 74.86 +/- 0.069 MIX-COMP-THERM-HTC-009 2343 25 x 24 600 06/16/88 20.8 82.25 + 0.0610 MIX-COMP-THERM-HTC-010 2345 29 x 29 841 06/26/88 21.1 59.92 + 0.0611 MIX-COMP-THERM-HTC-01 1 2347 1.5 27 x 27 729 06/23/88 21.3 76.72 +/- 0.0612 MIX-COMP-THERM-HTC-012 2348 27 x 26 702 06/23/88 21.1 84.57 +/- 0.0613 MIX-COMP-THERM-HTC-013 2349 39 x 39 1521 06/29/88 21.3 53.77 +/- 0.0614 MIX-COMP-THERM-HTC-014 2352 1.3 34 x 34 1156 07/05/88 21.3 80.16 +/- 0.0615 MIX-COMP-THERM-HTC-015 2353 34 x 33 1122 07/06/88 21.3 86.35 +/- 0.0616 MIX-COMP-THERM-HTC-016 2355 50 x 18 900 07/19/88 21.0 69.07 +/- 0.0617 MIX-COMP-THERM-HTC-017 2357 1.7 50 x 17 850 07/21/88 21.4 83.15 + 0.0818 MIX-COMP-THERM-HTC-018 2361 50 x 18'b) 900 07/28/88 22.4 80.16 +/- 0.07(a) given at a level of confidence of 95%(b) no-centered arrayREPORT HI-2104790 B-6 Table B-2 Key Physical Parameters of the HTC Phase 2 Critical Experiments with Gadolinium Solutions
[B.2]Number of Rods Gadolinium Experiment Pitch Date of Temperature Critical waterCase Reference number (cm) Along (CM) (a) conc. (g/l)number ____ edge Total experiment (C) height (ba)19 MIX-COMP-THERM-HTC-019 2405 38 x 38 1444 01/20/89 20.3 81.86 +/- 0.04 0.05220 MIX-COMP-THERM-HTC-020 2406 38 x 37 1406 01/23/89 19.7 87.16 +/- 0.04 0.05221 MIX-COMP-THERM-HTC-021 2407 42 x 42 1764 01/23/89 20.1 80.13 +/- 0.04 0.10022 MIX-COMP-THERM-HTC-022 2408 42 x 41 1722 01/25/89 19.7 84.38 +/- 0.04 0.09923 MIX-COMP-THERM-HTC-023 2409 1.3 41 x 41 1681 01/25/89 19.6 89.54 + 0.04 0.09924 MIX-COMP-THERM-HTC-024 2410 46 x 46 2116 01/26/89 20.1 81.33 +/- 0.04 0.15125 MIX-COMP-THERM-HTC-025 2411 45 x 45 2025 01/27/89 20.0 89.49 +/- 0.04 0.14826 MIX-COMP-THERM-HTC-026 2412 50 x 50 2500 01/30/89 20.7 85.83 +/- 0.04 0.20027 MIX-COMP-THERM-HTC-027 2415 50 x 49 2450 02/01/89 19.6 90.03 + 0.05 0.19728 MIX-COMP-THERM-HTC-028 2417 50 x 50 2500 02/09/89 19.6 89.67 + 0.04 0.19629 MIX-COMP-THERM-HTC-029 2419 42 x 42 1764 02/14/89 21.4 85.88 +/- 0.05 0.14730 MIX-COMP-THERM-HTC-030 2420 42 x 41 1722 02/15/89 21.0 90.51 +/-0.05 0.14731 MIX-COMP-THERM-HTC-031 2422 1.5 36 x 36 1296 02/21/89 22.1 83.86 +/--0.05 0.09832 MIX-COMP-THERM-HTC-032 2423 36 x 35 1260 02/21/89 22.6 89.85 +/- 0.04 0.09833 MIX-COMP-THERM-HTC-033 2425 32 x 32 1024 02/24/89 20.9 73.60 +/- 0.05 0.04834 MIX-COMP-THERM-HTC-034 2427 31 x 31 961 02/27/89 20.6 84.14 +/- 0.04 0.04835 MIX-COMP-THERM-HTC-035 2430 1.7 31 x 30 930 03/01/89 21.1 85.87 +/- 0.05 0.04836 MIX-COMP-THERM-HTC-036 2434 1.9 35 x 35 1225 03/08/89 21.7 89.61 +/- 0.04 0.04837 MIX-COMP-THERM-HTC-037 2436 39 x 39 1521 03/13/89 22.5 85.86 + 0.05 0.0971.738 MIX-COMP-THERM-HTC-038 2433 50 x 23 1150 03/07/89 21.7 84.35 +/- 0.04 0.048(a) given at a level of confidence of 95%(b) nominal values given in the report [B.5], notretainedREPORT HI-2104790 B-7 Table B-3 Key Physical Parameters of the HTC Phase 2 Critical Experiments with Boron Solutions
[B.2]Experiment Pitch Date of Temperature Boron conc.Case Reference NumbeAlong Rods water heightnedge Total experiment
(&deg;C) (cm) (a) (g/l)39 MIX-COMP-THERM-HTC-039 2437 37 x 37 1369 04/17/89 23.0 78.80 + 0.04 0.100 +/- 0.00140 MIX-COMP-THERM-HTC-040 2438 37 x 36 1332 04/18/89 22.8 83.84 +/- 0.04 0.106 +/- 0.00141 MIX-COMP-THERM-HTC-041 2441 39 x 39 1521 04/20/89 23.5 84.04 + 0.04 0.205 +/- 0.00242 MIX-COMP-THERM-HTC-042 2444 42 x 41 1722 04/26/89 23.0 85.40 + 0.05 0.299 +/- 0.0031.343 MIX-COMP-THERM-HTC-043 2446 45 x 44 1980 05/09/89 24.2 84.14 + 0.04 0.400 +/- 0.00444 MIX-COMP-THERM-HTC-044 2447 44 x 44 1936 05/10/89 24.7 88.63 +/- 0.05 0.399 +/- 0.00445 MIX-COMP-THERM-HTC-045 2448 47 x 47 2009 05/11/89 26.3 88.44 +/- 0.04 0.486 +/- 0.00546 MIX-COMP-THERM-HTC-046 2449 50 x 50 2500 05/17/89 25.1 90.64 +/- 0.04 0.587 +/- 0.00647 MIX-COMP-THERM-HTC-047 2459 49 x 49 2401 06/05/89 24.7 88.88 + 0.04 0.595 +/- 0.00648 MIX-COMP-THERM-HTC-048 2468 43 x 43 1849 06/15/89 22.7 89.46 +/- 0.04 0.499 +/- 0.00549 MIX-COMP-THERM-HTC-049 2470 39 x 39 1521 06/19/89 23.6 85.37 +/- 0.05 0.393 + 0.0041.550 MIX-COMP-THERM-HTC-050 2471 35 x 35 1225 06/21/89 23.6 88.90 +/- 0.04 0.295 +/- 0.00351 M1X-COMP-THERM-HTC-051 2473 32 x 32 1024 06/27/89 23.5 87.02 +/- 0.04 0.200 + 0.00252 MIX-COMP-THERM-HTC-052 2475 30 x 29 870 07/03/89 23.6 82.48 +/- 0.04 0.089 + 0.00153 MIX-COMP-THERM-HTC-053 2478 28 x 28 784 07/06/89 23.8 85.10 +/- 0.04 0.090 + 0.00154 MIX-COMP-THERM-HTC-054 2483 32 x 32 1024 07/19/89 24.2 87.06 +/- 0.04 0.194 +/- 0.00255 MIX-COMP-THERM-HTC-055 2485 1.7 37 x 37 1369 07/21/89 24.5 89.65 + 0.04 0.286 + 0.00356 MIX-COMP-THERM-HTC-056 2487 45 x 44 1980 08/09/89 23.8 88.72 +/- 0.04 0.415 +/- 0.00457 MIX-COMP-THERM-HTC-057 2482 50 x 21 1050 07/17/89 24.0 77.74 +/- 0.04 0.100 +/- 0.00158 MIX-COMP-THERM-HTC-058 2490 39 x 38 1482 09/08/89 22.9 88.41 +/- 0.04 0.220 +/- 0.0021.959 MIX-COMP-THERM-HTC-059 2492 31 x 30 930 09/14/89 22.0 86.95 +/- 0.04 0.110 +/- 0.001(a) given at a level of confidence of 95%/REPORT HI-2104790 B-8 Table B-4 Key Physical Parameters of the HTC Phase 3 Critical Experiments (pin pitch 1.6 cm) [B.3]Experiment Canister Number of Rods Date of Temperature Critical water WaterCase Reference number Type Along edge Total experiment (OC) height (cm) (a) Gap(cm)60 MIX-COMP-THERM-HTC-060 2518 25 x 25 625 01/04/90 18.3 88.83 + 0.34 3.561 MIX-COMP-THERM-HTC-061 2520 25 x 24 600 01/09/90 18.7 49.55 +/- 0.34 0.062 MIX-COMP-THERM-HTC-062 2521 Borated 25 x 24 600 01/10/90 18.8 71.45 +/- 0.34 2.0Steel63 MIX-COMP-THERM-HTC-063 2522 25 x 24 600 01/10/90 19.0 89.96 +/- 0.34 3.064 MIX-COMP-THERM-HTC-064 2523 25 x 24 600 01/12/90 18.9 58.23 +/- 0.34 1.065 MIX-COMP-THERM-HTC-065 2514 Boral 25 x 25 625 12/28/89 20.6 90.03 +/- 0.34 0.066 MIX-COMP-THERM-HTC-066 2511 25 x 25 625 12/21/89 21.1 82.16 +/- 0.34 2.067 MIX-COMP-THERM-HTC-067 2524 25 x 24 600 01/15/90 18.7 55.33 +/- 0.34 0.068 MIX-COMP-THERM-HTC-068 2525 25 x 24 600 01/16/90 19.0 67.95 +/- 0.34 1.0m69 MIX-COMP-THERM-HTC-069 2526 25 x 24 600 01/17/90 19.1 79.83 + 0.34 1.570 MIX-COMP-THERM-HTC-070 2527 25 x 24 600 01/18/90 19.1 58.66 +/- 0.34 0.571 MIX-COMP-THERM-HTC-071 2509 25 x 25 625 12/19/89 20.9 84.75 +/- 0.34 18.072 MIX-COMP-THERM-HTC-072 2531 25 x 24 600 01/23/90 19.0 88.2 +/- 0.34 14.573 MIX-COMP-THERM-HTC-073 2532 24 x 24 576 01/24/90 19.1 81.18 -0.34 11.074 MIX-COMP-THERM-HTC-074 2533 24 x 23 552 01/25/90 19.3 82.12 +/-0.34 10.075 MIX-COMP-THERM-HTC-075 2534 23 x 23 529 01/26/90 19.4 81.2 +/- 0.34 9.076 MIX-COMP-THERM-HTC-076 2535 22 x 22 484 01/30/90 19.7 86.17 +/- 0.34 8.077 MIX-COMP-THERM-HTC-077 2536 20 x 20 400 01/31/90 19.7 82.08 +/- 0.34 6.078 MIX-COMP-THERM-HTC-078 2537 17 x 17 289 02/01/90 19.9 77.92 +/- 0.34 4.079 MIX-COMP-THERM-HTC-079 2538 17 x 16 272 02/02/90 20.0 90.28 +/- 0.34 4.080 MIX-COMP-THERM-HTC-080 2539 14 x 14 196 02/05/90 20.2 75.99 +/- 0.34 2.081 MIX-COMP-THERM-HTC-081 2541 13 x 13 169 02/06/90 20.0 83.17 +/- 0.34 1.082 MIX-COMP-THERM-HTC-082 2544 13 x 13 169 02/07/90 20.4 79.46 +/- 0.34 0.083 MIX-COMP-THERM-HTC-083 2547 25 x 25 625 02/19/90 20.9 29.46 +/- 0.34 0.084 MIX-COMP-THERM-HTC-084 2548 25 x 25 625 02/20/90 20.9 37.96 +/- 0.34 4.085 MIX-COMP-THERM-HTC-085 2549 25 x 25 625 02/20/90 21.0 64.43 +/- 0.34 10.0(a) given at a level of confidence of 95%REPORT HI-2104790 B-9 Table B-5 Key Physical Parameters of the HTC Phase 4 Critical Experiments with the Lead Screen (four 25 x 25 arrays with 1.6 cmpitch) [B.4]Case Reference Experiment Canister Date of Temperature Water Gap Screen array Critical waternumber Type experiment
(&deg;C) (cm) (a) distance (cm) height (cm)(b) (c)86 MIX-COMP-THERM-HTC-086 2562 03/16/90 22.8 0.0 0.0 42.53 + 0.3487 MIX-COMP-THERM-HTC-087 2563 03/19/90 23.1 0.5 0.0 44.79 +/- 0.3488 MIX-COMP-THERM-HTC-088 2564 03/20/90 23.3 1.0 0.0 47.86 + 0.3489 MIX-COMP-THERM-HTC-089 2565 03/21/90 23.1 1.5 0.0 51.3 + 0.3490 MIX-COMP-THERM-HTC-090 2566 03/22/90 23.3 2.0 0.0 54.65 + 0.3491 MIX-COMP-THERM-HTC-091 2567 Borated 03/22/90 23.4 3.0 0.0 62.04 -0.34Steel92 MIX-COMP-THERM-HTC-092 2568 03/23/90 23.6 3.5 0.0 66.10 -0.3493 MIX-COMP-THERM-HTC-093 2569 03/26/90 23.5 2.0 0.5 55.87 -0.3494 MIX-COMP-THERM-HTC-094 2570 03/27/90 23.1 2.0 1.0 57.33 + 0.3495 MIX-COMP-THERM-HTC-095 2571 03/27/90 23.0 2.0 1.5 58.68 + 0.3496 MIX-COMP-THERM-HTC-096 2572 03/28/90 22.9 2.0 2.0 59.78 +/- 0.3497 MIX-COMP-THERM-HTC-097 2586 04/23/90 21.9 0.0 0.0 72.47 +/- 0.3498 MIX-COMP-THERM-HTC-098 2587 04/24/90 22.0 0.0 0.0 72.49 +/- 0.3499 MIX-COMP-THERM-HTC-099 2588 Boral 04/24/90 22.2 0.0 0.5 74.70 +/- 0.34100 MIX-COMP-THERM-HTC-100 2624 07/13/90 21.6 1.0 0.0 86.06 +/- 0.34101 MIX-COMP-THERM-HTC-101 2625 07/18/90 22.4 0.5 0.0 76.69 + 0.34102 MIX-COMP-THERM-HTC-102 2577 04/05/90 22.7 0.0 0.0 46.13 +/- 0.34103 MIX-COMP-THERM-HTC-103 2578 04/05/90 22.6 1.0 0.0 52.89 +/- 0.34104 MIX-COMP-THERM-HTC-104 2579 04/06/90 22.6 2.0 0.0 63.52 +/- 0.34105 MIX-COMP-THERM-HTC-105 2580 Cadmium 04/09/90 22.4 2.5 0.0 69.83 +/- 0.34106 MIX-COMP-THERM-HTC-106 2581 04/11/90 22.5 2.0 0.5 65.84 +/- 0.34107 MIX-COMP-THERM-HTC-107 2582 04/11/90 22.5 2.0 1.0 68.63 + 0.34108 MIX-COMP-THERM-HTC-108 2583 04/12/90 22.4 2.0 1.5 71.21 +/- 0.34REPORT HI-2104790 B-10 Screen array Critical waterCase Reference Experiment Canister Date of Temperature Water Gap Sce (cm) heit(cm) wnumber Type experiment (OC) (cm) (a) (b) (c)109 M1X-COMP-THERM-HTC-109 2584 04/12/90 22.4 2.0 2.0 73.36+/- 0.34110 MIX-COMP-THERM-HTC-1 10 2621 07/03/90 22.3 3.0 0.0 76.25 +/- 0.34111 MIX-COMP-THERM-HTC-1 11 2622 07/04/90 22.3 3.5 0.0 83.38 +/- 0.34112 MIX-COMP-THERM-HTC-1 12 2550 02/23/90 21.4 0.0 0.0 27.45 + 0.34113 MIX-COMP-THERM-HTC-1 13 2551 02/26/90 22.1 1.0 0.0 28.00 +/- 0.34114 MIX-COMP-THERM-HTC-114 2552 02/28/90 21.8 2.0 0.0 29.37 +/- 0.34115 MIX-COMP-THERM-HTC-115 2553 03/01/90 21.8 4.0 0.0 34.65 +/- 0.34116 MIX-COMP-THERM-HTC-1 16 2554 03/02/90 21.3 6.0 0.0 41.60 +/- 0.34117 MIX-COMP-THERM-HTC-117 2555 03/05/90 20.7 8.0 0.0 48.65 +/- 0.34No118 MIX-COMP-THERM-HTC-118 2556 03/06/90 20.7 10.0 0.0 54.74 +/- 0.34119 MIX-COMP-THERM-HTC-119 2557 03/07/90 20.9 12.0 0.0 59.57 +/-0.34120 MIX-COMP-THERM-HTC-120 2558 03/09/90 21.3 2.0 0.5 29.43 +/- 0.34121 MIX-COMP-THERM-HTC-121 2559 03/12/90 21.7 2.0 1.0 29.46 +/- 0.34122 MIX-COMP-THERM-HTC-122 2560 03/13/90 21.9 2.0 1.5 29.55 +/-0.34123 MIX-COMP-THERM-HTC-123 2561 03/14/90 22.3 2.0 2.0 29.62 +/- 0.34(a) Water gap between arrays.(b) Water gap between screen and array.(c) Given at a level of confidence of 95%REPORT HI-2104790 B-11 Table B-6 Key Physical Parameters of the HTC Phase 4 Critical Experiments with the Steel Screen (four 25 x 25 arrays with 1.6 cmpitch) [B.4]Experiment Canister Date of Temperature Water Gap Screen array Critical waterCase Reference (CM) (a) distance (cm) height (cm)number Type experiment (C) ((b) Mc)124 MIX-COMP-THERM-HTC-124 2602 05/21/90 23.6 0.0 0.0 42.11 + 0.34125 MIX-COMP-THERM-HTC-125 2603 05/21/90 23.4 0.5 0.0 44.14 +/- 0.34126 MIX-COMP-THERM-HTC-126 2604 05/22/90 22.9 1.0 0.0 46.96 +/- 0.34127 MIX-COMP-THERM-HTC-127 2605 05/29/90 20.4 1.5 0.0 50.16 + 0.34128 MIX-COMP-THERM-HTC-128 2606 05/30/90 20.1 2.0 0.0 53.43 +/- 0.34129 MIX-COMP-THERM-HTC-129 2607 Borated 05/31/90 20.0 2.0 0.5 54.71 +/--0.34Steel130 MIX-COMP-THERM-HTC-130 2608 06/05/90 20.2 2.0 1.0 56.32 +/- 0.34131 MIX-COMP-THERM-HTC-131 2609 06/05/90 20.1 2.0 1.5 57.96 +/- 0.34132 MIX-COMP-THERM-HTC-132 2610 06/06/90 19.7 2.0 2.0 59.16 +-0.34133 MIX-COMP-THERM-HTC-133 2611 06/08/90 19.5 3.0 0.0 60.38 +/- 0.34134 MIX-COMP-THERM-HTC-134 2612 06/12/90 20.1 3.5 0.0 64.19 + 0.34135 MIX-COMP-THERM-HTC-135 2589 04/26/90 22.4 0.0 0.0 69.82 + 0.34Boral136 MIX-COMP-THERM-HTC-136 2626 07/19/90 22.6 0.5 0.0 73.44 + 0.34137 MIX-COMP-THERM-HTC-137 2613 06/13/90 20.5 0.0 0.0 44.70 +/- 0.34138 MIX-COMP-THERM-HTC-138 2614 06/13/90 20.6 1.0 0.0 51.00 + 0.34139 MIX-COMP-THERM-HTC-139 2615 06/14/90 20.6 2.0 0.0 60.26 +/- 0.34140 MIX-COMP-THERM-HTC-140 2616 06/15/90 20.7 2.0 0.5 62.54 + 0.34141 MIX-COMP-THERM-HTC-141 2617 Cadmium 06/18/90 21.0 2.0 1.0 65.85 + 0.34142 MIX-COMP-THERM-HTC-142 2618 06/19/90 21.3 2.0 1.5 68.70 +/- 0.34143 MIX-COMP-THERM-HTC-143 2619 06/20/90 21.5 2.0 2.0 71.00 +/- 0.34144 MIX-COMP-THERM-HTC-144 2620 06/21/90 21.7 2.5 0.0 65.76 +/- 0.34145 MIX-COMP-THERM-HTC-145 2590 04/27/90 22.4 0.0 0.0 27.77 + 0.34No146 MIX-COMP-THERM-HTC-146 2591 05/09/90 24.4 1.0 0.0 28.34 + 0.34REPORT HI-2104790 B-12 Experiment Canister Date of Temperature Water G Screen array Critical waterCase Reference ECMxperiment C roa Wae) (ap distance (cm) height (cm)number Type experiment (C) ((b) (c)147 MIX-COMP-THERM-HTC-147 2592 05/10/90 24.4 2.0 0.0 29.74 +/- 0.34148 MIX-COMP-THERM-HTC-148 2593 05/10/90 24.3 2.0 0.5 29.68 +/- 0.34149 MIX-COMP-THERM-HTC-149 2594 05/11/90 24.5 2.0 1.0 29.66 +/- 0.34150 MIX-COMP-THERM-HTC-150 2595 05/11/90 24.4 2.0 1.5 29.68 +/--0.34151 MIX-COMP-THERM-HTC-151 2596 05/14/90 24.7 2.0 2.0 29.76 +/- 0.34152 MIX-COMP-THERM-HTC-152 2597 05/15/90 24.6 4.0 0.0 35.33 +/- 0.34153 MIX-COMP-THERM-HTC-153 2598 05/15/90 24.6 6.0 0.0 43.24 +/- 0.34154 MIX-COMP-THERM-HTC-154 2599 05/16/90 24.7 8.0 0.0 51.30 +/- 0.34155 MIX-COMP-THERM-HTC-155 2600 05/17/90 24.7 10.0 0.0 58.73 +/- 0.34156 MIX-COMP-THERM-HTC-156 2601 05/18/90 24.6 12.0 0.0 64.84 +/- 0.34(a) Water gap between arrays.(b) Water gap between screen and array.(c) Given at a level of confidence of 95%REPORT HI-2104790 B- 13 Table B-7 Description of the Selected Benchmark Critical Experiments
[LB.6]Case Reference Identification U, wt% 'wt157 LEU-COMP-THERM-011-001 Core 1 2.46 -158 LEU-COMP-THERM-01 1-002 Core II 2.46 -159 LEU-COMP-THERM-01 1-004 Core IIIB 2.46 -160 LEU-COMP-THERM-0 11-015 Core IX 2.46 -161 LEU-COMP-THERM-051-001 Core X 2.46 -162 LEU-COMP-THERM-051-003 Core XIB 2.46 -163 LEU-COMP-THERM-051-009 Core XII 2.46 -164 LEU-COMP-THERM-051-010 Core XIII 2.46 -165 LEU-COMP-THERM-051-012 Core XIV 2.46 -166 LEU-COMP-THERM-051-013 Core XV 2.46 -167 LEU-COMP-THERM-051-014 Core XVI 2.46 -168 LEU-COMP-THERM-051-015 Core XVII 2.46 -169 LEU-COMP-THERM-051-016 Core XVIII 2.46 -170 LEU-COMP-THERM-051-017 Core XIX 2.46 -171 LEU-COMP-THERM-051-018 Core XX 2.46 -172 LEU-COMP-THERM-051-019 Core XXI 2.46 -173 BAW-1645-4
[B.7] S-type Fuel, w/886 ppm B 2.46 -174 BAW-1645-4
[B.7] S-type Fuel, w/746 ppm B 2.46 -175 BAW-1645-4
[B.7] SO-type Fuel, w/l 156 ppm B 2.46 -176 BAW-1810
[B.8] Case 1 1337 ppm B 2.46 -177 BAW-1810
[B.8] Case 12 1899 ppm B 2.75 -178 French [B.9] Water Moderator 0 gap 4.75 -179 French [B.9] Water Moderator 2.5 cm gap 4.75 -180 French [B.9] Water Moderator 5 cm gap 4.75 -181 French [B.9] Water Moderator 10 cm gap 4.75 -182 LEU-COMP-THERM-0 17-012 Steel Reflector, 1.321 cm separation 2.35 -183 LEU-COMP-THERM-017-013 Steel Reflector, 2.616 cm separation 2.35 -184 LEU-COMP-THERM-017-014 Steel Reflector, 3.912 cm separation 2.35 -185 LEU-COMP-THERM-001-008 Steel Reflector, Infinite separation 2.35 -186 LEU-COMP-THERM-010-016 Steel Reflector, 1.321 cm separation 4.306 -187 LEU-COMP-THERM-010-018 Steel Reflector, 2.616 cm separation 4.306 -188 LEU-COMP-THERM-010-019 Steel Reflector, 5.405 cm separation 4.306 -189 LEU-COMP-THERM-004-010 Steel Reflector, Infinite separation 4.306 -190 LEU-COMP-THERM-013-003 Steel Reflector, with Boral Sheets 4.306 -191 LEU-COMP-THERM-010-021 Lead Reflector, 0.55 cm sepn. 4.306 -192 LEU-COMP-THERM-010-022 Lead Reflector, 1.956 cm sepn. 4.306 -193 LEU-COMP-THERM-010-023 Lead Reflector, 5.405 cm sepn. 4.306 -194 LEU-COMP-THERM-002-004 Experiment 004/032 -no absorber 4.306 -195 LEU-COMP-THERM-009-005 Exp. 009 1.05% Boron Steel plates 4.306 -REPORT HI-2104790 B-14 Case Reference Identification U, wt% Pu,196 LEU-COMP-THERM-009-007 Exp. 009 1.62% Boron Steel plates 4.306 -197 LEU-COMP-THERM-009-009 Exp. 031 -Boral plates 4.306 -198 PNL-7167
[B. 10] Experiment 214R -with flux traps 4.306 -199 PNL-7167
[B.10] Experiment 214V3 -with flux trap 4.306 -200 LEU-COMP-THERM-014-001 Case 173 -0 ppm B 4.306201 LEU-COMP-THERM-014-005 Case 177 -2550 ppm B 4.306 -202 PNL-5803
[B.I 1] MOX Fuel -Type 3.2 Exp. 21 0.71 20203 PNL-5803
[B.1 1] MOX Fuel -Type 3.2 Exp. 43 0.71 20204 PNL-5803
[B.1 1] MOX Fuel -Type 3.2 Exp. 13 0.71 20205 PNL-5803
[B.11] MOX Fuel -Type 3.2 Exp. 32 0.71 20206 MIX-COMP-THERM-003-001 Saxton Case 52 PuO2 0.52" pitch 0.72 6.6207 WCAP-3385
[B.12] Saxton Case 52 U 0.52" pitch 5.74 -208 MIX-COMP-THERM-003-002 Saxton Case 56 PuO2 0.56" pitch 0.72 6.6209 MIX-COMP-THERM-003-003 Saxton Case 56 borated PuO2 0.72 6.6210 WCAP-3385
[B.12] Saxton Case 56 U 0.56" pitch 5.74 -211 MIX-COMP-THERM-003-005 Saxton Case 79 PuO2 0.79" pitch 0.72 6.6212 WCAP-3385
[B.12] Saxton Case 79 U 0.79" pitch 5.74 -213 MIX-COMP-THERM-002-030 0.700-in.
pitch 0 ppm B 0.72 2.0214 MIX-COMP-THERM-002-031 0.700-in.
pitch 688 ppm B 0.72 2.0215 MIX-COMP-THERM-002-032 0.870-in.
pitch 0 ppm B 0.72 2.0216 MIX-COMP-THERM-002-033 0.870-in.
pitch 1090 ppm B 0.72 2.0217 MIX-COMP-THERM-002-034 0.990-in.
pitch 0 ppm B 0.72 2.0218 MIX-COMP-THERM-002-035 0.990-in.
pitch 767 ppm B 0.72 2.0219 MIX-COMP-THERM-003-004 Saxton Case PuO2 0.735" pitch 0.72 6.6220 MIX-COMP-THERM-003-006 Saxton Case PuO2 1.04" pitch 0.72 6.6221 MIX-COMP-THERM-006-001 8 wt% 240Pu 0.80" pitch 0.71 2.0222 MIX-COMP-THERM-006-002 8 wt% 240Pu 0.93" pitch 0.71 2.0223 MIX-COMP-THERM-006-003 8 wt% 240Pu 1.05" pitch 0.71 2.0224 MIX-COMP-THERM-006-004 8 wt% 240Pu 1.143" pitch 0.71 2.0225 MIX-COMP-THERM-006-005 8 wt% 240Pu 1.32" pitch 0.71 2.0226 MIX-COMP-THERM-006-006 8 wt% 240Pu 1.386" pitch 0.71 2.0227 MIX-COMP-THERM-007-001 16 wt% 240Pu 0.93" pitch 0.72 2.0228 MIX-COMP-THERM-007-002 16 wt% 240Pu 1.05" pitch 0.72 2.0229 MIX-COMP-THERM-007-003 16 wt% 240Pu 1.143" pitch 0.72 2.0230 MIX-COMP-THERM-007-004 16 wt%/o 240Pu 1.32" pitch 0.72 2.0231 MIX-COMP-THERM-008-001 24 wt0/o 240Pu 0.80" pitch 0.72 2.0232 MIX-COMP-THERM-008-002 24 wt%/o 240Pu 0.93" pitch 0.72 2.0233 MIX-COMP-THERM-008-003 24 wt%/o 240Pu 1.05" pitch 0.72 2.0234 MIX-COMP-THERM-008-004 24 wt%/o 240Pu 1.143" pitch 0.72 2.0235 MIX-COMP-THERM-008-005 24 wt0o 240Pu 1.32" pitch 0.72 2.0REPORT HI-2104790 B-15 Case Reference Identification U, wt% Pu,236 MIX-COMP-THERM-008-006 24 wt% 240Pu 1.386" pitch 0.72 2.0237 MIX-COMP-THERM-005-001 18 wt% 240Pu 0.85" pitch 0.72 4.0238 MIX-COMP-THERM-005-002 18 wt% 240Pu 0.93" pitch 0.72 4.0239 MIX-COMP-THERM-005-003 18 wt% 240Pu 1.05" pitch 0.72 4.0240 MIX-COMP-THERM-005-004 18 wt% 240Pu 1.143" pitch 0.72 4.0241 MIX-COMP-THERM-005-005 18 wt%/o 240Pu 1.386" pitch 0.72 4.0242 MIX-COMP-THERM-005-006 18 wt% 240Pu 1.60" pitch 0.72 4.0243 MIX-COMP-THERM-005-007 18 wt% 240Pu 1.70" pitch 0.72 4.0244 LEU-COMP-THERM-001
-001 1 Cluster 2.35 -245 LEU-COMP-THERM-00 1-002 3 Clusters, Separation 11.92 cm 2.35 -246 LEU-COMP-THERM-001
-003 3 Clusters, Separation 8.41 cm 2.35 -247 LEU-COMP-THERM-00 1-004 3 Clusters, Separation 10.05 cm 2.35 -248 LEU-COMP-THERM-001-005 3 Clusters, Separation 6.39 cm 2.35 -249 LEU-COMP-THERM-00 1-006 3 Clusters, Separation 9.01 cm 2.35 -250 LEU-COMP-THERM-00 1-007 3 Clusters, Separation 4.46 2.35 -251 LEU-COMP-THERM-002-001 I Cluster, 1Ox 11.51 4.306 -252 LEU-COMP-THERM-002-002 1 Cluster, 9x 13.35 4.306 -253 LEU-COMP-THERM-002-003 I Cluster, 8x16.37 4.306 -254 LEU-COMP-THERM-002-005 3 Clusters, Separation 7.11 cm 4.306 -255 LEU-COMP-THERM-003-001 I Cluster, 614.4 Rods, Gd water impurity 2.35 -256 LEU-COMP-THERM-003-002 1 Cluster, 529.3 Rods 2.35 -257 LEU-COMP-THERM-003-003 I Cluster, 523.9 Rods 2.35 -258 LEU-COMP-THERM-003-004 1 Cluster, 525.3 Rods 2.35 -259 LEU-COMP-THERM-003-005 I Cluster, 595.4 Rods 2.35 -260 LEU-COMP-THERM-003-006 1 Cluster, 485.8 Rods 2.35 -261 LEU-COMP-THERM-003-007 I Cluster, 523.8 Rods 2.35 -262 LEU-COMP-THERM-003-008 1 Cluster, 505.4 Rods 2.35 -263 LEU-COMP-THERM-003-009 4 Clusters, Separation 2.59 cm 2.35 -264 LEU-COMP-THERM-003-010 2 Clusters, Separation 1.68 cm 2.35 -265 LEU-COMP-THERM-003-011 4 Clusters, Separation 4.27 cm 2.35 -266 LEU-COMP-THERM-003-012 4 Clusters, Separation 5.95 cm 2.35 -267 LEU-COMP-THERM-003-013 4 Clusters, Separation 5.11 cm 2.35 -268 LEU-COMP-THERM-003-014 4 Clusters, Separation 6.66 cm 2.35 -269 LEU-COMP-THERM-003-015 4 Clusters, Separation 7.53 cm 2.35 -270 LEU-COMP-THERM-003-016 4 Clusters, Separation 9.00 cm 2.35 -271 LEU-COMP-THERM-003-017 4 Clusters, Separation 9.97 cm 2.35 -272 LEU-COMP-THERM-003-018 4 Clusters, Separation 11.45 cm 2.35 -273 LEU-COMP-THERM-003-019 4 Clusters, Separation 13.87 cm 2.35 -274 LEU-COMP-THERM-003-020 3 Clusters, Separation 9.88 cm 2.35 --275 LEU-COMP-THERM-003-021 3 Clusters, Separation 6.78 cm 2.35 -REPORT HI-2104790 B-16 Case Reference Identification U, wt% Put276 LEU-COMP-THERM-003-023 3 Clusters, Separation 6.176 cm 2.35 -277 LEU-COMP-THERM-004-001 1 Cluster, 225.8 Rods, Gd water impurity 4.306 -278 LEU-COMP-THERM-004-002 1 Cluster, 216.2 Rods 4.306 -279 LEU-COMP-THERM-004-003 1 Cluster, 216.6 Rods 4.306 -280 LEU-COMP-THERM-004-004 1 Cluster, 218.6 Rods 4.306 -281 LEU-COMP-THERM-004-005 I Cluster, 167.85 Rods 4.306 -282 LEU-COMP-THERM-004-006 1 Cluster, 203 Rods 4.306 -283 LEU-COMP-THERM-004-007 1 Cluster, 173.5 Rods 4.306 -284 LEU-COMP-THERM-004-008 2 Clusters, Separation 2.83 cm 4.306 -285 LEU-COMP-THERM-004-009 3 Clusters, Separation 12.27 cm 4.306 -286 LEU-COMP-THERM-004-011 3 Clusters, Separation 12.493 cm 4.306 -287 LEU-COMP-THERM-004-012 4 Clusters, Separation 4.72 cm 4.306 -288 LEU-COMP-THERM-004-013 4 Clusters, Separation 8.38 cm 4.306 -289 LEU-COMP-THERM-004-014 4 Clusters, Separation 10.86 cm 4.306 -290 LEU-COMP-THERM-004-015 4 Clusters, Separation 11.29 cm 4.306 -291 LEU-COMP-THERM-004-016 4 Clusters, Separation 12.02 cm 4.306 -292 LEU-COMP-THERM-004-017 4 Clusters, Separation 13.64 cm 4.306 -293 LEU-COMP-THERM-004-018 4 Clusters, Separation 14.98 cm 4.306 -294 LEU-COMP-THERM-004-019 4 Clusters, Separation 19.81 cm 4.306 -295 LEU-COMP-THERM-004-020 4 Clusters, Separation 8.50 cm 4.306 -296 LEU-COMP-THERM-006-001 19x19, Rod Pitch -1.849 cm 2.596 -297 LEU-COMP-THERM-006-002 20x20, Rod Pitch -1.849 cm 2.596 -298 LEU-COMP-THERM-006-003 21x21, Rod Pitch -1.849 cm 2.596 -299 LEU-COMP-THERM-006-004 17x17, Rod Pitch -1.956 cm 2.596 -300 LEU-COMP-THERM-006-005 18x18, Rod Pitch -1.956 cm 2.596 -301 LEU-COMP-THERM-006-006 19x 19, Rod Pitch -1.956 cm 2.596 -302 LEU-COMP-THERM-006-007 20x20, Rod Pitch -1.956 cm 2.596 -303 LEU-COMP-THERM-006-008 21x21, Rod Pitch -1.956 cm 2.596 -304 LEU-COMP-THERM-006-009 16x16, Rod Pitch -2.15 cm 2.596 -305 LEU-COMP-THERM-006-010 17x17, Rod Pitch -2.15 cm 2.596 -306 LEU-COMP-THERM-006-011 18x18, Rod Pitch -2.15 cm 2.596 -307 LEU-COMP-THERM-006-012 19x19, Rod Pitch -2.15 cm 2.596 -308 LEU-COMP-THERM-006-013 20x20, Rod Pitch -2.15 cm 2.596 -309 LEU-COMP-THERM-006-014 15x 15, Rod Pitch -2.293 cm 2.596 -310 LEU-COMP-THERM-006-015 16x16, Rod Pitch -2.293 cm 2.596 -311 LEU-COMP-THERM-006-016 17x17, Rod Pitch -2.293 cm 2.596 -312 LEU-COMP-THERM-006-017 18x18, Rod Pitch -2.293 cm 2.596 -313 LEU-COMP-THERM-006-018 19x19, Rod Pitch -2.293 cm 2.596 -314 LEU-COMP-THERM-008-001 Core XI, 1511 ppm 2.459 -315 LEU-COMP-THERM-008-002 Core XI, 1335.5 ppm 2.459 -REPORT HI-2104790 B-17 Case Reference Identification U, wt% Pu,316 LEU-COMP-THERM-008-003 Core XI, 1335.5 ppm 2.459317 LEU-COMP-THERM-008-004 Core XI, 1182 ppm, 36 Pyrex Rods 2.459318 LEU-COMP-THERM-008-005 Core XI, 1182 ppm, 36 Pyrex Rods 2.459319 LEU-COMP-THERM-008-006 Core XI, 1032.5 ppm, 72 Pyrex Rods 2.459320 LEU-COMP-THERM-008-007 Core XI, 1032.5 ppm, 72 Pyrex Rods 2.459321 LEU-COMP-THERM-008-008 Core XI, 794 ppm, 144 Pyrex Rods 2.459322 LEU-COMP-THERM-008-009 Core XI, 779 ppm, 144 Pyrex Rods 2.459323 LEU-COMP-THERM-008-010 Core XI, 1245 ppm, 72 Vicor Rods 2.459324 LEU-COMP-THERM-008-011 Core XI, 1384 ppm, 144 A1203 Rods 2.459325 LEU-COMP-THERM-008-012 Core XI, 1348 ppm, 36 A1203 Rods 2.459326 LEU-COMP-THERM-008-013 Core XI, 1348 ppm, 36 A1203 Rods 2.459327 LEU-COMP-THERM-008-014 Core XI, 1363 ppm, 72 A1203 Rods 2.459328 LEU-COMP-THERM-008-015 Core XI, 1362 ppm, 72 A1203 Rods 2.459329 LEU-COMP-THERM-008-016 Core XI, 1158 ppm 2.459330 LEU-COMP-THERM-008-017 Core XI, 921 ppm 2.459331 LEU-COMP-THERM-009-001 0% Boron Steel plates, dist. 0.245 cm 4.306332 LEU-COMP-THERM-009-002 0% Boron Steel plates, dist. 3.277 cm 4.306333 LEU-COMP-THERM-009-003 0% Boron Steel plates, dist. 0.428 cm 4.306334 LEU-COMP-THERM-009-004 0% Boron Steel plates, dist. 3.277 cm 4.306335 LEU-COMP-THERM-009-006 1.05% Boron Steel plates, dist. 3.277 cm 4.306 -336 LEU-COMP-THERM-009-008 1.62% Boron Steel plates, dist. 3.277 cm 4.306 -337 LEU-COMP-THERM-009-024 Al plates, dist. 0.105 cm 4.306 -338 LEU-COMP-THERM-009-025 Al plates, dist. 3.277 cm 4.306 -339 LEU-COMP-THERM-009-026 Zircaloy-4 plates, dist. 0.078 cm 4.306 -340 LEU-COMP-THERM-009-027 Zircaloy-4 plates, dist. 3.277 cm 4.306 -341 LEU-COMP-THERM-010-001 Lead Reflector, 0 cm separation 4.306 -342 LEU-COMP-THERM-010-002 Lead Reflector, 0.660 cm separation 4.306 -343 LEU-COMP-THERM-010-003 Lead Reflector, 1.321 cm separation 4.306 -344 LEU-COMP-THERM-010-004 Lead Reflector, 5.405 cm separation 4.306 -345 LEU-COMP-THERM-010-009 Steel Reflector, 0 cm separation 4.306 -346 LEU-COMP-THERM-010-010 Steel Reflector, 0.660 cm separation 4.306 -347 LEU-COMP-THERM-010-01 1 Steel Reflector, 1.321 cm separation 4.306 -348 LEU-COMP-THERM-010-012 Steel Reflector, 2.616 cm separation 4.306 -349 LEU-COMP-THERM-010-013 Steel Reflector, 5.405 cm separation 4.306 -350 LEU-COMP-THERM-010-014 Steel Reflector, 0 cm separation 4.306 -351 LEU-COMP-THERM-010-015 Steel Reflector, 0.660 cm separation 4.306 -352 LEU-COMP-THERM-010-017 Steel Reflector, 1.956 cm separation 4.306 -353 LEU-COMP-THERM-010-020 Lead Reflector, 0 cm separation 4.306 -354 LEU-COMP-THERM-01 1-003 Core lilA 2.46 -355 LEU-COMP-THERM-01 1-005 Core IIIC 2.46 -REPORT HI-2104790 B-18 Case Reference Identification U, wt% Pu,356 LEU-COMP-THERM-01 1-006 Core 1111 2.46 -357 LEU-COMP-THERM-01 1-007 Core IIIE 2.46 -358 LEU-COMP-THERM-01 1-008 Core IIIF 2.46 -359 LEU-COMP-THERM-01 1-009 Core IIIG 2.46 -360 LEU-COMP-THERM-01 1-010 Core IV 2.46 -361 LEU-COMP-THERM-01 1-011 Core V 2.46 -362 LEU-COMP-THERM-01 1-012 Core VI 2.46 -363 LEU-COMP-THERM-01 1-013 Core VII 2.46 -364 LEU-COMP-THERM-0 11-0 14 Core VIII 2.46 -365 LEU.-COMP-THERM-012-001 0% Boron Steel plate, Gd water impurity 2.35 -366 LEU-COMP-THERM-012-002 1.1% Boron Steel plate 2.35 -367 LEU-COMP-THERM-012-003 1.6% Boron Steel plate 2.35 -368 LEU-COMP-THERM-012-004 Boral B plate 2.35369 LEU-COMP-THERM-012-005 Boral C plate 2.35370 LEU-COMP-THERM-012-006
: Boroflex, 1.84 cm separation 2.35371 LEU-COMP-THERM-012-007
: Boroflex, 1.73 cm separation 2.35372 LEU-COMP-THERM-013-001 Steel Reflector, 0% Boron Steel plate 4.306 -373 LEU-COMP-THERM-013-002 Steel Reflector, 1.1% Boron Steel plate 4.306 -374 LEU-COMP-THERM-01 3-004 Steel Reflector,
: Boroflex, 8.37 cm separation 4.306 -375 LEU-COMP-THERM-014-002 Borated Water, 490 ppm 4.306 -376 LEU-COMP-THERM-014-006 Unborated Water 4.306 -377 LEU-COMP-THERM-014-007 Borated Water, 1030 ppm 4.306 -378 LEU-COMP-THERM-016-001 0% Boron Steel plates, dist. 0.645 cm 2.35 -379 LEU-COMP-THERM-016-002 0% Boron Steel plates, dist. 2.732 cm 2.35 -380 LEU-COMP-THERM-016-003 0% Boron Steel plates, dist. 4.042 cm 2.35 -381 LEU-COMP-THERM-016-004 0% Boron Steel plates, dist. 0.645 cm 2.35 -382 LEU-COMP-THERM-016-005 0% Boron Steel plates, dist. 4.042 cm 2.35 -383 LEU-COMP-THERM-016-006 0% Boron Steel plates, dist. 0.645 cm 2.35 -384 LEU-COMP-THERM-016-007 0% Boron Steel plates, dist. 4.042 cm 2.35 -385 LEU-COMP-THERM-016-008 1.05% Boron Steel plates, dist. 0.645 cm 2.35 -386 LEU-COMP-THERM-016-009 1.05% Boron Steel plates, dist. 4.042 cm 2.35 -387 LEU-COMP-THERM-016-010 1.62% Boron Steel plates, dist. 0.645 cm 2.35 -388 LEU-COMP-THERM-016-011 1.62% Boron Steel plates, dist. 4.042 cm 2.35 -389 LEU-COMP-THERM-016-012 Boral plates, dist. 0.645 cm 2.35 -390 LEU-COMP-THERM-016-013 Boral plates, dist. 4.442 cm 2.35 -391 LEU-COMP-THERM-016-014 Boral plates, dist. 0.645 cm 2.35 -392 LEU-COMP-THERM-016-028 Al plates, dist. 0.645 cm 2.35 -393 LEU-COMP-THERM-016-029 Al plates, dist. 4.042 cm 2.35 -394 LEU-COMP-THERM-016-030 Al plates, dist. 4.442 cm 2.35 -395 LEU-COMP-THERM-016-031 Zircaloy-4 plates, dist. 0.645 cm 2.35 -REPORT HI-2104790 B-19 Case Reference Identification U, wt% Pu,396 LEU-COMP-THERM-016-032 Zircaloy-4 plates, dist. 4.042 cm 2.35 -397 LEU-COMP-THERM-026-001 Hex, 621 Rods, Temperature
: 20. IC 4.92 -398 LEU-COMP-THERM-026-002 Hex, 889 Rods, Temperature 231.4C 4.92 -399 LEU-COMP-THERM-026-003 Hex, 1951 Rods, Temperature 19.3C 4.92 -400 LEU-COMP-THERM-026-004 Hex, 2791 Rods, Temperature 206.0C 4.92 -401 LEU-COMP-THERM-026-005 Hex, 325/680 Rods, Temperature 20.8C 4.92 -402 LEU-COMP-THERM-026-006 Hex, 325/912 Rods, Temperature 212.1C 4.92 -403 LEU-COMP-THERM-051-002 Core XIA 2.46 -404 LEU-COMP-THERM-051-004 Core XIC 2.46 -405 LEU-COMP-THERM-051-005 Core XID 2.46 -406 LEU-COMP-THERM-051-006 Core XIE 2.46 -407 LEU-COMP-THERM-051-007 Core XIF 2.46 -408 LEU-COMP-THERM-051-008 Core XIG 2.46 -409 LEU-COMP-THERM-05 1-011 Core XIIIA 2.46 -410 LEU-COMP-THERM-062-001 No Boron Steel plates 2.6 -411 LEU-COMP-THERM-062-002 0% Boron Steel plates, 3 mm, dist. 0 2.6 -412 LEU-COMP-THERM-062-003 0% Boron Steel plates, 6 mm, dist. 0 2.6 -413 LEU-COMP-THERM-062-004 0% Boron Steel plates, 6 mm, dist. 0.5 2.6 -414 LEU-COMP-THERM-062-005 0% Boron Steel plates, 6 mm, dist. 1 2.6 -415 LEU-COMP-THERM-062-006 0.67% Boron Steel plates, 3 mm, dist. 0 2.6 -416 LEU-COMP-THERM-062-007 0.67% Boron Steel plates, 6 mm, dist. 0 2.6 -417 LEU-COMP-THERM-062-008 0.67% Boron Steel plates, 3 mm, dist. 0.5 2.6 -418 LEU-COMP-THERM-062-009 0.67% Boron Steel plates, 6 mm, dist. 0.5 2.6 -419 LEU-COMP-THERM-062-010 0.67% Boron Steel plates, 3 mm, dist. 1 2.6 -420 LEU-COMP-THERM-062-011 0.67% Boron Steel plates, 6 mm, dist. 1 2.6 -421 LEU-COMP-THERM-062-012 0.98% Boron Steel plates, 3 mm, dist. 0 2.6 -422 LEU-COMP-THERM-062-013 0.98% Boron Steel plates, 6 mm, dist. 0 2.6 -423 LEU-COMP-THERM-062-014 0.98% Boron Steel plates, 6 mm, dist. 0.5 2.6 -424 LEU-COMP-THERM-062-015 0.98% Boron Steel plates, 6 mm, dist. 1 2.6 -425 LEU-COMP-THERM-065-001 No Boron Steel plates 2.6 -426 LEU-COMP-THERM-065-002 0% Boron Steel plates, dist. 0 2.6 -427 LEU-COMP-THERM-065-003 0.67% Boron Steel plates, dist. 0 2.6 -428 LEU-COMP-THERM-065-004 0.98% Boron Steel plates, dist. 0 2.6 -429 LEU-COMP-THERM-065-005 No Boron Steel plates 2.6 -430 LEU-COMP-THERM-065-006 0% Boron Steel plates, dist. 0 2.6 -431 LEU-COMP-THERM-065-007 0% Boron Steel plates, dist. 0.5 2.6 -432 LEU-COMP-THERM-065-008 0% Boron Steel plates, dist. 0 2.6 -433 LEU-COMP-THERM-065-009 0% Boron Steel plates, dist. 0.5 2.6 -434 LEU-COMP-THERM-065-010 0.67% Boron Steel plates, dist. 0 2.6 -435 LEU-COMP-THERM-065-011 0.67% Boron Steel plates, dist. 0.5 2.6 -__jREPORT HI-2104790 B-20 Case Reference Identification U, wt% Pu,436 LEU-COMP-THERM-065-012 0.67% Boron Steel plates, dist. 0 2.6 -437 LEU-COMP-THERM-065-013 0.67% Boron Steel plates, dist. 0.5 2.6 -438 LEU-COMP-THERM-065-014 0.98% Boron Steel plates, dist. 0 2.6 -439 LEU-COMP-THERM-065-015 0.98% Boron Steel plates, dist. 0.5 2.6 -440 LEU-COMP-THERM-065-016 0.98% Boron Steel plates, dist. 0 2.6 -441 LEU-COMP-THERM-065-017 0.98% Boron Steel plates, dist. 0.5 2.6 -442 LEU-COMP-THERM-081-001 Otto Hahn, ZrB2 and B4C rods 5.423 -443 LEU-COMP-THERM-082-001 IPEN/MB-01 (580 pins) 4.3486 -444 LEU-COMP-THERM-082-002 IPEN/MB-01 (560 pins) 4.3486 -445 LEU-COMP-THERM-082-003 670 pins, A1203-B4C rods 4.3486 -446 LEU-COMP-THERM-082-004 672 pins, A1203-B4C rods 4.3486 -447 LEU-COMP-THERM-082-005 668 pins, A1203-B4C rods 4.3486 -448 LEU-COMP-THERM-082-006 668 pins, A1203-B4C rods 4.3486 -449 LEU-COMP-THERM-090-001 664 pins, 16 steel rods 4.3486 -450 LEU-COMP-THERM-090-002 662 pins, 18 steel rods 4.3486 -451 LEU-COMP-THERM-090-003 658 pins, 14 steel rods 4.3486 -452 LEU-COMP-THERM-090-004 660 pins, 12 steel rods 4.3486 -453 LEU-COMP-THERM-090-005 660 pins, 12 steel rods 4.3486 -454 LEU-COMP-THERM-090-006 661 pins, 17 steel rods 4.3486 -455 LEU-COMP-THERM-090-007 662 pins, 16 steel rods 4.3486 -456 LEU-COMP-THERM-090-008 634 pins, 12 steel rods 4.3486 -457 LEU-COMP-THERM-090-009 620 pins, 26 steel rods 4.3486 -458 LEU-COMP-THERM-091-001 668 pins, 0 steel rods, 4 Gd203 rods 4.3486 -459 LEU-COMP-THERM-091-002 648 pins, 0 steel rods, 8 Gd203 rods 4.3486 -460 LEU-COMP-THERM-091-003 672 pins, 0 steel rods, 4 Gd203 rods 4.3486 -461 LEU-COMP-THERM-091-004 646 pins, 4 steel rods, 4 Gd203 rods 4.3486 -462 LEU-COMP-THERM-091-005 656 pins, 4 steel rods, 4 Gd2O3 rods 4.3486 -463 LEU-COMP-THERM-091-006 664 pins, 4 steel rods, 2 Gd203 rods 4.3486 -464 LEU-COMP-THERM-091-007 670 pins, 2 steel rods, 2 Gd203 rods 4.3486 -465 LEU-COMP-THERM-091-008 664 pins, 2 steel rods, 2 Gd203 rods 4.3486 -466 LEU-COMP-THERM-091-009 656 pins, 0 steel rods, 2 Gd203 rods 4.3486 -467 MIX-COMP-THERM-004-001 23x23, 1.825 cm pitch 0.72 3.01468 MIX-COMP-THERM-004-002 23x23, 1.825 cm pitch 0.72 3.011469 MIX-COMP-THERM-004-003 23x23, 1.825 cm pitch 0.72 3.011470 MIX-COMP-THERM-004-004 21x21, 1.956 cm pitch 0.72 3.011471 MIX-COMP-THERM-004-005 21x21, 1.956 cm pitch 0.72 3.011472 MIX-COMP-THERM-004-006 21x21, 1.956 cm pitch 0.72 3.011473 MIX-COMP-THERM-004-007 20x20, 2.225 cm pitch 0.72 3.011REPORT HI-2104790 B-21 Case Reference Identification U, wt% Pu,474 MIX-COMP-THERM-004-008 20x20, 2.225 cm pitch 0.72 3.01475 MIX-COMP-THERM-004-009 20x20, 2.225 cm pitch 0.72 3.011476 MIX-COMP-THERM-004-010 21x21, 2.474 cm pitch 0.72 3.011477 MIX-COMP-THERM-004-011 21x21, 2.474 cm pitch 0.72 3.011478 MIX-COMP-THERM-006-007 8 wt% 240Pu 1.05" pitch, Al Rods 0.72 2.0479 MIX-COMP-THERM-006-013 8 wt% 240Pu 1.05" pitch, B4 Rods 0.72 2.0480 MIX-COMP-THERM-006-014 8 wt% 240Pu 1.05" pitch, B3 Rods 0.72 2.0481 MIX-COMP-THERM-006-015 8 wt% 240Pu 1.05" pitch, B2 Rods 0.72 2.0482 MIX-COMP-THERM-006-016 8 wt% 240Pu 1.05" pitch, BI Rods 0.72 2.0483 MIX-COMP-THERM-006-017 8 wt% 240Pu 1.05" pitch, AI+Cd Rods 0.72 2.0484 MIX-COMP-THERM-006-023 8 wt% 240Pu 1.05" pitch, B4+Cd Rods 0.72 2.0485 MIX-COMP-THERM-006-024 8 wt% 240Pu 1.05" pitch, B3+Cd Rods 0.72 2.0486 MIX-COMP-THERM-006-025 8 wt% 240Pu 1.05" pitch, B2+Cd Rods 0.72 2.0487 MIX-COMP-THERM-006-026 8 wt% 240Pu 1.05" pitch, B 1+Cd Rods 0.72 2.0488 MIX-COMP-THERM-006-027 8 wt% 240Pu 1.05" pitch, Air+Cd Rods 0.72 2.0489 MIX-COMP-THERM-006-028 8 wt% 240Pu 1.05" pitch, H20+Cd Rods 0.72 2.0490 MIX-COMP-THERM-006-029 8 wt% 240Pu 1.32" pitch, Al Rods 0.72 2.0491 MIX-COMP-THERM-006-035 8 wt% 240Pu 1.32" pitch, B4 Rods 0.72 2.0492 MIX-COMP-THERM-006-036 8 wt% 240Pu 1.32" pitch, B3 Rods 0.72 2.0493 MIX-COMP-THERM-006-037 8 wt% 240Pu 1.32" pitch, B2 Rods 0.72 2.0494 MIX-COMP-THERM-006-038 8 wt% 240Pu 1.32" pitch, B I Rods 0.72 2.0495 MIX-COMP-THERM-006-039 8 wt% 240Pu 1.32" pitch, AI+Cd Rods 0.72 2.0496 MIX-COMP-THERM-006-045 8 wt% 240Pu 1.32" pitch, B4+Cd Rods 0.72 2.0497 MIX-COMP-THERM-006-046 8 wt% 240Pu 1.32" pitch, B3+Cd Rods 0.72 2.0498 MIX-COMP-THERM-006-047 8 wt% 240Pu 1.32" pitch, B2+Cd Rods 0.72 2.0499 MIX-COMP-THERM-006-048 8 wt% 240Pu 1.32" pitch, B 1 +Cd Rods 0.72 2.0500 MIX-COMP-THERM-006-049 8 wt% 240Pu 1.32" pitch, Air+Cd Rods 0.72 2.0501 MIX-COMP-THERM-006-050 8 wt%/o 240Pu 1.32" pitch, H20+Cd Rods 0.72 2.0502 MIX-COMP-THERM-007-005 16 wt%/o 240Pu 1.386" pitch 0.72 2.0503 MIX-COMP-THERM-007-006 16 wt% 240Pu 1.05" pitch, Al Rods 0.72 2.0504 MIX-COMP-THERM-007-012 16 wt% 240Pu 1.05" pitch, B4 Rods 0.72 2.0505 MIX-COMP-THERM-007-013 16 wt% 240Pu 1.05" pitch, B3 Rods 0.72 2.0506 MIX-COMP-THERM-007-014 16 wt% 240Pu 1.05" pitch, B2 Rods 0.72 2.0507 MIX-COMP-THERM-007-015 16 wt% 240Pu 1.05" pitch, B I Rods 0.72 2.0508 MIX-COMP-THERM-007-016 16 wt% 240Pu 1.05" pitch, AI+Cd Rods 0.72 2.0509 MIX-COMP-THERM-007-022 16 wt% 240Pu 1.05" pitch, B4+Cd Rods 0.72 2.0510 MIX-COMP-THERM-007-023 16 wt% 240Pu 1.05" pitch, B3+Cd Rods 0.72 2.0511 MIX-COMP-THERM-007-024 16 wt% 240Pu 1.05" pitch, B2+Cd Rods 0.72 2.0512 MIX-COMP-THERM-007-025 16 wt% 240Pu 1.05" pitch, BI+Cd Rods 0.72 2.0513 MIX-COMP-THERM-007-026 16 wt% 240Pu 1.05" pitch, Air+Cd Rods 0.72 2.0 IREPORT HI-2104790 B-22 Case Reference Identification U, wt% Pu,514 MIX-COMP-THERM-007-027 16 wt% 240Pu 1.05" pitch, H20+Cd Rods 0.72 2.0515 MIX-COMP-THERM-008-007 24 wt% 240Pu 1.05" pitch, Al Rods 0.72 2.0516 MIX-COMP-THERM-008-013 24 wt% 240Pu 1.05" pitch, B4 Rods 0.72 2.0517 MIX-COMP-THERM-008-014 24 wt% 240Pu 1.05" pitch, B3 Rods 0.72 2.0518 MIX-COMP-THERM-008-015 24 wt% 240Pu 1.05" pitch, B2 Rods 0.72 2.0519 MIX-COMP-THERM-008-016 24 wt% 240Pu 1.05" pitch, B I Rods 0.72 2.0520 MIX-COMP-THERM-008-017 24 wt% 240Pu 1.05" pitch, AL+Cd Rods 0.72 2.0521 MIX-COMP-THERM-008-023 24 wt% 240Pu 1.05" pitch, B4+Cd Rods 0.72 2.0522 MIX-COMP-THERM-008-024 24 wt% 240Pu 1.05" pitch, B3+Cd Rods 0.72 2.0523 MIX-COMP-THERM-008-025 24 wt% 240Pu 1.05" pitch, B2+Cd Rods 0.72 2.0524 MIX-COMP-THERM-008-026 24 wt% 240Pu 1.05" pitch, B 1 +Cd Rods 0.72 2.0525 MIX-COMP-THERM-008-027 24 wt% 240Pu 1.05" pitch, Air+Cd Rods 0.72 2.0526 MIX-COMP-THERM-008-028 24 wt% 240Pu 1.05" pitch, H20+Cd Rods 0.72 2.0527 MIX-COMP-THERM-009-001 8 wt% 240Pu 0.55" pitch 0.16 1.5528 MIX-COMP-THERM-009-002 8 wt% 240Pu 0.60" pitch 0.16 1.5529 MIX-COMP-THERM-009-003 8 wt% 240Pu 0.71" pitch 0.16 1.5530 MIX-COMP-THERM-009-004 8 wt% 240Pu 0.80" pitch 0.16 1.5531 MIX-COMP-THERM-009-005 8 wt% 240Pu 0.90" pitch 0.16 1.5532 MIX-COMP-THERM-009-006 8 wt% 240Pu 0.93" pitch 0.16 1.5REPORT HI-2 104790 B-23REPORT HI-2104790 B-23 Appendix CBenchmark of MCNP5-1.51 with ENDF[B-V(total number of pages: 27 including this page)REPORT HI-2104790 C-1 C.1 Introduction This Appendix presents the analysis of the validation results for MCNP5-1.51 code and includesthe results of the calculations, normality test, the detailed statistical trending
: analysis, calculation bias and bias uncertainty for each distinct area of applicability of the parameters of interest.
C.2 Computer Code Parameter DataThe computer code MCNP5-1.51
[C. 1] is the continuous energy Monte Carlo codes and treats anarbitrary three-dimensional configuration of materials in geometric cells bounded by first- andsecond-degree surfaces and fourth-degree elliptical tori. Thermal neutrons are described by boththe free gas and S(a,3) models. All calculations were performed using the default data libraries provided with the code: the default continuous energy neutron transport data predominantly based on ENDF/B-V.
The list of ZAIDs that were used in the analysis is presented in Table C.2-1. The criticality source card was set to accumulate a total of 1.8 million neutron histories forevery individual run. The neutrons start from an arbitrary distribution, causing a generally verylarge variance of results from the first cycles in comparison with the following cycles. Therefore, the results from the first 50 cycles were skipped when calculating the average keff. The calculated keff values have associated uncertainties due to the statistical nature of the Monte Carlo codes.C.3 Analysis of MCNP5-1.51 Validation ResultsC.3.1. Calculational ResultsThe calculation results for the 156 HTC critical experiments and for the 135 selected criticalexperiments described in Appendix B are presented and discussed in this section.
The calculation results are summarized by grouping the experiments in terms of the categories as set forth inAppendix B. Calculation
: results, including keff, Ucaic-, and EALF, measurement uncertainties (trp) and the calculation and measurement combined uncertainty (ar) are shown in Table C.3-1through Table C.3-5.Figure C.3-1 and C.3-2 are histograms showing the frequency of calculated k, and EALF for all291 benchmarks.
The nominal calculated k, values range from A. The EALFresults values show a range betweenDescriptive statistics for the different group of experiments is summarized in Table C.3-6.C.3.2. Normality TestIn order to assess the normality assumption, Shapiro and Wilk [5] test has been used for groupswith fewer than 50 samples while the Pearson's chi-square (X2) test [4] has been used for sampleslarger than 20 samples.
The tests are applied to the group of experiments in terms of thecategories as set forth in Appendix B.For the Shapiro and Wilk test, Table C.3-7 shows the computed Wtest value, and W value thatcan be obtained for the number of experiments from [5] to accept the normality hypothesis.
If WREPORT HI-2104790 C-2 is less than the test statistic, Wtest, then the data is considered normally distributed.
For the X2test, it is concluded normal for xf < n, where n is a number of bins for the group of experiments.
The probability Pd(,2 > ,0 2) of obtaining a value of 2 _> in an experiment with d degreesof freedom to confirm quantitatively that the agreement is satisfactory was taken or interpolated, if necessary, from Appendix D in Reference
[4]. Thus, if Pd(k2 >_ 1o2) is large, the obtained andexpected distributions are consistent; if it is small, they probably disagree.
In particular, ifPd(,k2 > i02) is less than 5%, we say that the disagreement is significant and reject the assumeddistributions at the 5% level. If it is less than 1%, the disagreement is called highly significant, and we reject the assumed distributions at the 1% level.As it is shown in Table C.3-7, all cases except Phase 1 test normal. Nevertheless, the group withall 291 experiments shows an agreement with the assumed normal distribution with theprobability Pd = 7.36%.C.3.3. Trending AnalysisTrends are determined through the use of regression fits to the calculated results.
The equations used to identify trends are given below:Y(x) = a + bx (7-1)1L 1X fVX 1'1 'X y1\(7-2)b= ( Zxi- X- YZ T)(7-3)(7-4)The squared term of the linear correlation factor r defined below (from Reference
[5]) is used toquantitatively measure the degree to which a linear relationship exist between two variables.
1-U2 (xi -:0) (yi -Y)r =(7-5)The closer r2 approaches the value of 1, the better the fit of the data to the linear equation.
Amore quantitative measure of the fit can be found by using Appendix C in Reference
[4]. Theinterpolation was applied, if necessary.
For any given observed value ro, PN(Irl -- Irol) is theprobability that N measurements of two uncorrelated variables would give a coefficient r as largeas ro. Thus, if we obtain a coefficient ro for which PN(IrI > Irol) is small, it is correspondingly unlikely that our variables are uncorrelated; that is, a correlation indicated.
In particular, ifREPORT HI-2104790 C-3 PN(IrI > Irol) < 5%, the correlation is called significant; if it is less than 1%, the correlation iscalled highly significant.
The validation results are analyzed by grouping the experiments in terms of the categories as setforth in Appendix B. Independent variables used in the trending analysis by group, correlation coefficients and trending analysis results are summarized in Table C.3-8. The linear regression equations for the independent parameter with the significant correlation of keff were presented inTable C.3-8.C.3.4. Bias and Bias Uncertainty In this section, benchmark results are analyzed using the statistical method described in section2.2.The first step is to evaluate whether the four HTC phases and selected ex eriments, should bereduced to a single set. Th ofe Phase 1 data set is , thof the Phase 2 data set is , the mean ke of the Phase 3 data set is, the mean keff of the Phase 4 data set is and the mean keff of theselected ex eriments data is .The maximum difference between the meansis just &#xfd; which is less than the uncertainty.
These sets are water moderated uranium ormixed plutonium-uranium dioxide lattices.
The addition of a absorber rods, separator plates orreflector plates is not introducing a significant increase in the ability to calculate keff. The Phase 1through Phase 4 sets and the selected experiments are considered one large set of 291experiments from now on.The analysis of the correlation coefficient in Table C.3-8 (combined set) and the plot of datatrend (Figure C.3-3) show that there is a significant trend a function of the rod pitch. This isdiscussed in the Section C.3.5.2.The total bias (systematic error or mean of the deviation from a keff of exactly 1.000) of theMCNP5-1.51 code is shown in the table belowCalculational Bias of the MCNP5-1.51 codeDescription Total Bias Bias Uncertainty HTC and Selected Experiments C.3.5. Applicability of MCNP5-1.51 Validation ResultsThis subsection contains a more detailed evaluation of the set of critical experiments.
Regarding the selected experiments, the following subjects are discussed:
" Neutron absorber and neutron reflector materials
" Fuel rod pitch trend" Neutron absorber geometryREPORT HI-2104790 C-4
* Fuel bumup" Unborated and borated water.The general focus is to justify that using the full set of critical experiments is appropriate.
Insome cases, subsets of full set of experiments are established.
For those subsets, statistical evaluations are performed to determine bias, bias uncertainty, normality and trends. Trends areevaluated for fuel rod outer diameter, fuel rod pitch, fuel density, and EALF.C.3.5. 1. Neutron Absorber and Neutron Reflector Materials The HTC and Selected Experiments consider the following neutron absorbers and reflectors:
* Absorbers o Boron, in the form of soluble boron in the water, boron in solid form (B4C), andboron in borated steelo Soluble gadolinium in watero Cadmium" Reflectors o Steelo Leado WaterSome typical configurations do not contain gadolinium or cadmium neutron absorbers or leadreflectors.
To verify that including those materials does not have a significant effect on theresults of the benchmarking
: analyses, a subset without those experiments containing thosematerials was analyzed.
The comparison with the full set is presented in Table C.3-9 and showsno significant differences when those materials are excluded.
: However, in both cases, asignificant trend is observed, as a function of the rod pitch in the experiment.
This is discussed inthe next section.C.3.5.2.
Fuel Rod Pitch TrendTo better understand the observed rod pitch trend, the results for all 291 experiments are shownin Figure C.3-4 as a function of rod pitch. It appears that the trend is due to the experiments athigher rod pitch value (> 2 cm), which consistently show keff values well above 1.0. To evaluatethe impact of those experiments at larger rod pitches, the Table C.3-10 shows a comparison ofresults with and without those experiments.
When results above 2 cm rod pitch are excluded, aslightly higher absolute bias is observed, in this case with a lower uncertainty, and no significant rod pitch trend. Based on those results it could be concluded that the trend is only caused by theexperiments at higher rod pitch values. To ensure that a potential trend would not be ignored, allfollowing evaluations are performed for the two conditions used above, i.e. for all rod pitchvalues, and for experiments with rod pitch values limited to no more than 2 cm.C.3.5.3.
Absorber GeometryREPORT HI-2 104790 C-5REPORT HI-2104790 C-5 The criticality experiments analyzed in this report include experiments with Boron in the form ofplates, absorber rods and soluble boron in water. No trend relating to these experiments isobserved.
C.3.5.4.
Fuel BurnupThe full set of critical experiments contains experiments with fresh U02 fuel, with simulated spent fuel (37.5 GWd/MTU),
and MOX fuel with Pu content between 2 and 20%, which is evenhigher than typically found in spent fuel. The experiments are therefore reasonably representative of burned fuel at different burnup levels. To verify that the experiments cover theburnup range sufficiently, the experiments are subdivided into fresh U02 fuel, HTC experiments and MOX experiments, and compared to the results of the entire set. The comparison is shown inTable C.3-1 1. The comparison shows no significant differences between the entire set and theU02 and HTC subsets, but for MOX the bias is now positive (i.e. truncated bias of 0.0), with alarger uncertainty, and some trends. However, this is based on relatively small sets ofexperiments.
Bias values are comparable between sets with and without rod pitch values above 2cm, with a maximum absolute value of .C.3.5.5.
Unborated and Borated WaterThe full set of critical experiments contains both experiments with and without soluble boron.The entire set of analyses shows no significant trend when analyzed as a function of the solubleboron level. Nevertheless, sets with and without soluble boron are analyzed and compared to thefull set that contains all experiments.
The results are shown in Table C.3-12. Similar to theprevious subsection, the comparison shows no significant differences between those subsets.Bias values are comparable between sets with and without rod pitch values above 2 cm, with amaximum absolute value of .C.4 SummaryA set of 291 critical experiments has been selected and has been used for the validation of theHoltec International criticality safety methodology.
The similarity between the chosenexperiments and the actual systems has been based on a set of screening criteria as is stated in theNUREG/CR-6698
[5]. Experiments have been categorized by common features as Phase 1through Phase 4 and selected experiments and parameterized by key variables such as latticepitch / assembly pitch, absorber solution concentration, number of fuel rods, rod outer diameter, fuel density, screen array distance, fuel enrichment and EALF. Benchmark calculations havebeen performed using the Monte Carlo code MCNP5-1.51.
It was determined that Phase 1through Phase 4 and selected experiments are in sufficient agreement that this sets are lumpedtogether as a single set of 291 experiments.
The bias and bias uncertainty are presented in sectionC.3.4. The applicability of validation results is considered in section C.3.5.The range of key parameters for the design application, benchmarks and validated AOA aresummarized in Table C.3-13. A point by point comparison between design application andbenchmarks shows that the experimental range covers all the parameters.
The soluble boronREPORT HI-2104790 C-6 concentration is extrapolated generously since 1&deg;B is a 1/v absorber (as permitted on Table 2.3 of[5]).As for the fuel density, Table 2.3 of Reference
[5] states there is "no requirement" and that"experiments should be as close to the desired concentration as possible".
Since the experiment fuel density is 9.2 -10.4 g/cm3 and the design application one is around 10.0 -10.7 g/cm3, it isconsidered that the values are very close so the validated AOA covers the design application range.The fuel enrichment can be up to 5%. The experiments used go up to 5.74 wt% 235U. Therefore, it is considered that the validated AOA covers the design application range.C.5 References
[C. 1] "MCNP -A General Monte Carlo N-Particle Transport Code, Version 5"; Los AlamosNational Laboratory, LA-UR-03-1987 (Revised 2/1/2008).
REPORT HI-2104790 C-7 Table C.2-1 MCNP5-1.51 ZAIDs Used for EachNuclideNuclide ZAIID1H 1001.50c1&deg;B 5010.50c11B 5011.55cC 6000.50c1IN 7014.50c160 8016.50c23Na 11023.51c Mg 12000.50c 27A1 13027.50c Si 14000.51c 31p 15031.50c 32s 16032.51c Ca 20000.51c Ti 22000.50c Cr 24000.50c 55Mn 25055.51c Fe 26000.55c 59Co 27059.50c Ni 28000.50c Cu 29000.50c Zn 30000.40c Zr 40000.56c Mo 42000.50c Cd 48000.50c Sn 50000.40c Gd 64000.35c Pb 82000.50c 234U 92234.50c 235u 92235.50c 236u 92236.50c 238 U 92238.50c 238Pu 94238.50c 239pu 94239.50c 240pu 94240.50c 241pu 94241.50c 242Pu 94242.50c 241, mln 95241.50c REPORT HI-2104790 C-8 Table C.3-1 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase1 Critical Experiments:
Water-Moderated and Reflected ArraysCase Evaluation Identification 1 MIX-COMP-THERM-HTC-001 2 MIX-COMP-THERM-HTC-002 3 MIX-COMP-THERM-HTC-003 4 MIX-COMP-THERM-HTC-004 5 MIX-COMP-THERM-HTC-005 6 MIX-COMP-THERM-HTC-006 7 MIX-COMP-THERM-HTC-007 8 MIX-COMP-THERM-HTC-008 9 MIX-COMP-THERM-HTC-009 10 MIX-COMP-THERM-HTC-010 11 MIX-COMP-THERM-HTC-0 1112 MIX-COMP-THERM-HTC-012 13 MIX-COMP-THERM-HTC-013 14 MIX-COMP-THERM-HTC-014 15 MIX-COMP-THERM-HTC-015 16 MIX-COMP-THERM-HTC-016 17 MIX-COMP-THERM-HTC-017 18 MIX-COMP-THERM-HTC-018 File- l~ri +oaci +/-Ox i !EALFname '~ cclt ~ --UU U --U UU U-REPORT HI-2 104790 C-9REPORT HI-2104790 C-9 Table C.3-2 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 2 CriticalEx riments:
Reflected Simple Arrays Moderated by Poisoned Water with Gadolinium or BoronFile-Case Evaluation Identification name kefi -0cai -Gexp -Gi EALF (eV)19 MIX-COMP-THERM-HTC-019 U UIU UM T20 MIX-COMP-THERM-HTC-020 UU21 MIX-COMP-THERM-HTC-021 UEUV22 MIX-COMP-THERM-HTC-022 U111U[U23 MIX-COMP-THERM-HTC-023 U = U U0 i24 MIX-COMP-THERM-HTC-024 U U U I25 MIX-COMP-THERM-HTC-025 U U U[U i26 MIX-COMP-THERM-HTC-026 U27 MIX-COMP-THERM-HTC-027 U U.U.U..28 MIX-COMP-THERM-HTC-028 U29 MIX-COMP-THERM-HTC-029 U30 MIX-COMP-THERM-HTC-030 U UU31 MIX-COMP-THERM-HTC-031 U32 MIX-COMP-THERM-HTC-032 U33 MIX-COMP-THERM-HTC-033 U UUU34 MIX-COMP-THERM-HTC-034 U U ..U.U.. .35 MIX-COMP-THERM-HTC-035 U36 MIX-COMP-THERM-HTC-036 U lA U U U37 MIX-COMP-THERM-HTC-037 U38 MIX-COMP-THERM-HTC-038 U39 MIX-COMP-THERM-HTC-039 U U.U.U..40 MIX-COMP-THERM-HTC-040 U U U.U.W..41 MIX-COMP-THERM-HTC-041 U42 MIX-COMP-THERM-HTC-042 U UUU-43 MIX-COMP-THERM-HTC-043 U UUU44 MIX-COMP-THERM-HTC-044 U[45 MIX-COMP-THERM-HTC-045 U UUU46 MIX-COMP-THERM-HTC-046 U11UU47 MIX-COMP-THERM-HTC-047 U U[U U i48 MIX-COMP-THERM-HTC-048 U U49 MIX-COMP-THERM-HTC-049 U U.U.U.U..
50 MIX-COMP-THERM-HTC-050 U U[ U W51 MIX-COMP-THERM-HTC-051 U U UUU52 MIX-COMP-THERM-HTC-052 U[U UU U53 MIX-COMP-THERM-HTC-053 U U UUW54 MIX-COMP-THERM-HTC-054 U U UUUW55 MIX-COMP-THERM-HTC-055 U U U UREPORT HI-2104790 C-10 File-Case Evaluation Identification name keff-i + gcalc-i + gexp + ai EALF (eV)56 MIX-COMP-THERM-HTC-056 U UUU57 MIX-COMP-THERM-HTC-057 U[U58 MIX-COMP-THERM-HTC-058 U U WUU59 MIX-COMP-THERM-HTC-059 U"J _"___"REPORT HI-2 104790 c-ilREPORT HI-2104790 C-I11 Table C.3-3 The MCNP5-1.51 Calculational Results and Measurements Uncertainties forPhase 3 Critical Experiments:
Pool StorageCase Evaluation Identification 60 MIX-COMP-THERM-HTC-060 61 MIX-COMP-THERM-HTC-061 62 MIX-COMP-THERM-HTC-062 63 MIX-COMP-THERM-HTC-063 64 MIX-COMP-THERM-HTC-064 65 MIX-COMP-THERM-HTC-065 66 MIX-COMP-THERM-HTC-066 67 MIX-COMP-THERM-HTC-067 68 MIX-COMP-THERM-HTC-068 69 MIX-COMP-THERM-HTC-069 70 MIX-COMP-THERM-HTC-070 71 MIX-COMP-THERM-HTC-071 72 MIX-COMP-THERM-HTC-072 73 MIX-COMP-THERM-HTC-073 74 MIX-COMP-THERM-HTC-074 75 MIX-COMP-THERM-HTC-075 76 MIX-COMP-THERM-HTC-076 77 MIX-COMP-THERM-HTC-077 78 MIX-COMP-THERM-HTC-078 79 MIX-COMP-THERM-HTC-079 80 MIX-COMP-THERM-HTC-080 81 MIX-COMP-THERM-HTC-081 82 MIX-COMP-THERM-HTC-082 83 MIX-COMP-THERM-HTC-083 84 MIX-COMP-THERM-HTC-084 85 MIX-COMP-THERM-HTC-085 File- ~, + ~ ~ aexp + a, EALF (eV)name-U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U --~-U U ---U U ---U U --= ----- ~----- ~----- ~------ ~----- ~REPORT HI-2 104790 c-i 2REPORT HI-2104790 C-12 Table C.3-4 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 4 CriticalExperiments:
Shipping CaskCase Evaluation Identification 86 MIX-COMP-THERM-HTC-086 87 MIX-COMP-THERM-HTC-087 88 MIX-COMP-THERM-HTC-088 89 MIX-COMP-THERM-HTC-089 90 MIX-COMP-THERM-HTC-090 91 MIX-COMP-THERM-HTC-091 92 MIX-COMP-THERM-HTC-092 93 MIX-COMP-THERM-HTC-093 94 MIX-COMP-THERM-HTC-094 95 MIX-COMP-THERM-HTC-095 96 MIX-COMP-THERM-HTC-096 97 MIX-COMP-THERM-HTC-097 98 MIX-COMP-THERM-HTC-098 99 MIX-COMP-THERM-HTC-099 100 MIX-COMP-THERM-HTC-100101 MIX-COMP-THERM-HTC-101102 MIX-COMP-THERM-HTC-102103 MIX-COMP-THERM-HTC-103File- k~ff, +/- aca!c~ +/- ~ + a, (eV)name-m -m m m--m m m--m m m---m m m-U -U U -~-U m m m--m m m-U -U U -~-m m m m-m -m m m-~ m m m-m -m m m--U U U -~--m m m--U U U -~--U U U -~---U U -~--m m -104MIX-COMP-THERM-HTC-104 I I I I II105MIX-COMP-THERM-HTC-105106 MIX-COMP-THERM-HTC-106107 MIX-COMP-THERM-HTC-107108 MIX-COMP-THERM-HTC-108 109 MIX-COMP-THERM-HTC-109110 MIX-COMP-THERM-HTC-I 10111 MIX-COMP-THERM-HTC-11112 MIX-COMP-THERM-HTC-1 12113 MIX-COMP-THERM-HTC-113114 MIX-COMP-THERM-HTC-1 14115 MIX-COMP-THERM-HTC-1 15116 MIX-COMP-THERM-HTC-116117 MIX-COMP-THERM-HTC-117---m m m--U U U ~--U U U ~--m m m-m m m-m m m-U U U U -~-m m m-U U U U -~-U U U U ~-U U U U -~U U U U -~U U U U118MIX-COMP-THERM-HTC-1 118I IN I II119 MIX-COMP-THERM-HTC-119120 MIX-COMP-THERM-HTC-120120 MIX-COMP-THERM-HTC-121121 MIX-COMP-THERM-HTC-121 U U U U U122 MIX-COMP-THERM-HTC-122 123 MIX-COMP-THERM-HTC-123 REPORT HI-2104790 C-13 CaseEvaluation Identification 124MIX-COMP-THERM-HTC-124125 MIX-COMP-THERM-HTC-125126 MIX-COMP-THERM-HTC-126127 MIX-COMP-THERM-HTC-127128 MIX-COMP-THERM-HTC-128129 MIX-COMP-THERM-HTC-129130 MIX-COMP-THERM-HTC-130131 MIX-COMP-THERM-HTC-131File-name kef-132u calc- G~x (y +1EALF JIMD(-COMP-ThERM-HTC-132 I I_32 ... .... .. .... ..... ... .2 --U133 MIX-COMP-THERM-HTC-133 134 MIX-COMP-THERM-HTC-134 135 MIX-COMP-THERM-HTC-135 136 MIX-COMP-THERM-HTC-136 137MIX-COMP-THERM-HTC-137IM 138MIX-COMP-THERM-HTC-138139 MIX-COMP-THERM-HTC-139 140 MIX-COMP-THERM-HTC-140141 MIX-COMP-THERM-HTC-141142 MIX-COMP-THERM-HTC-142143 MIX-COMP-THERM-HTC-143144 MIX-COMP-THERM-HTC-144 145 MIX-COMP-THERM-HTC-145146 MIX-COMP-THERM-HTC-146147 MIX-COMP-THERM-HTC-147148 MIX-COMP-THERM-HTC-148149 MIX-COMP-THERM-HTC-149150 MIX-COMP-THERM-HTC-150151 MIX-COMP-THERM-HTC-151152 MIX-COMP-THERM-HTC-152153 MIX-COMP-THERM-HTC-153 F-F-F-F--F--1F--F--F--1F--1F-1F--1F-1U U U ~ IU U -~ IU U U ~ IU U U ~ IU-UU U U -~ Im m m m I-m m m IU U U ~ IU U U ~ IU U U ~ IU U U -~ IU U U -~ IU U -~ IU U -~ IU U --~ Im -m I--- m IU -- m I-U -m Im U -m I--- m I--m IU --m I-m m I--m m I-m m I-m m I-m m I-m m I-m m I-m mU ~ U ~ I154MIX-COMP-THERM-HTC-154 155MIX-COMP-THERM-HTC-155156 MIX-COMP-THERM-HTC-156 REPORT HI-2 104790 C-14REPORT HI-2104790 C-14 Table C.3-5 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for SelectedCritical Experiments CaseEvaluation Identification File- klff. + ralc +/- 0expname+/- (yiEALF(eV)157 Core I__158 Core II159 Core III I160 Core IX161 Core X162 Core XI163 Core XII164 Core XIII I165 Core XIV I166 Core XV167 Core XVI I168 Core XVII169 Core XVIH I170 Core XIX I171 Core XX172 Core XXI I173 S-type Fuel, w/886 ppm B174 S-type Fuel, w/746 ppm B175 SO-type Fuel, w/1 156 ppm B176 Case 1 1337 ppm B177 Case 12 1899 ppmB B178 Water Moderator 0 gap179 Water Moderator 2.5 cm gap180 Water Moderator 5 cm gap181 Water Moderator 10 cm gap182 Steel Reflector, 1.321 cm separation 183 Steel Reflector, 2.616 cm separation 184 Steel Reflector, 3.912 cm separation 185 Steel Reflector, Infinite separation 186 Steel Reflector, 1.321 cm separation 187 Steel Reflector, 2.616 cm separation 188 Steel Reflector, 5.405 cm separation 189 Steel Reflector, Infinite separation 190 Steel Reflector, with Boral Sheets191 Lead Reflector, 0.55 cm sepn.192 Lead Reflector, 1.956 cm sepn.193 Lead Reflector, 5.405 cm sepn.194 Experiment 004/032 -no absorberREPORT 111-2104790 c-ISREPORT HI-2104790 C-15 CaseEvaluation Identification File-"CIM'nPke~ff-i-acalc-+ (Yexp-aiEALF195 Exp. 009 1.05% Boron Steel plates196 Exp. 009 1.62% Boron Steel plates197 Exp. 031 -Boral plates198 Experiment 214R -with flux traps199 Experiment 214V3 -with flux trap200 Case 173 -0 ppm B201 Case 177 -2550 ppm B202 MOX Fuel -Type 3.2 Exp. 21203 MOX Fuel -Type 3.2 Exp. 43204 MOX Fuel -Type 3.2 Exp. 13205 MOX Fuel -Type 3.2 Exp. 32206 Saxton Case 52 PuO2 0.52" pitch207 Saxton Case 52 U 0.52" pitch208 Saxton Case 56 PuO2 0.56" pitch209 Saxton Case 56 borated PuO2210 Saxton Case 56 U 0.56" pitch211 Saxton Case 79 PuO2 0.79" pitch212 Saxton Case 79 U 0.79" pitch213 0.700-in.
pitch 0 ppm B214 0.700-in.
pitch 688 ppm B215 0.870-in.
pitch 0 ppm B216 0.870-in.
pitch 1090 ppm B217 0.990-in.
pitch 0 ppm B218 0.990-in.
pitch 767 ppm B219 Saxton Case PuO2 0.735" pitch220 Saxton Case PuO2 1.04" pitch221 8 wt% 240Pu 0.80" pitch222 8 wt% 240Pu 0.93" pitch223 8 wt% 240Pu 1.05" pitch224 8 wt% 240Pu 1.143" pitch225 8 wt% 240Pu 1.32" pitch226 8 wt% 240Pu 1.386" pitch227 16 wt% 240Pu 0.93" pitch228 16 wt% 240Pu 1.05" pitch229 16 wt% 240Pu 1.143" pitch230 16 wt% 240Pu 1.32" pitch231 24 wt% 240Pu 0.80" pitch232 24 wt% 240Pu 0.93" pitch233 24 wt% 240Pu 1.05" pitch234 24 wt% 240Pu 1.143" pitchREPORT HI-2104790 C-16 I 1 7 1 T rCaseEvaluation Identification File-keff-i+ Gcalc.I Oexp-oiEALF(eV)L235 24 wt% 240Pu 1.32" pitch236 24 wt% 240Pu 1.386" pitch237 18 wt% 240Pu 0.85" pitch238 18 wt% 240Pu 0.93" pitch239 18 wt% 240Pu 1.05" pitch240 18 wt% 24OPu 1. 143" pitch241 18 wt% 240Pu 1.386" pitch242 18 wt% 240Pu 1.60" pitch243 18 wt% 240Pu 1.70" pitch317 Core XI, 1182 ppm, 36 Pyrex Rods318 Core XI, 1182 ppm, 36 Pyrex Rods319 Core XI, 1032.5 ppm, 72 Pyrex Rods320 Core XI, 1032.5 ppm, 72 Pyrex Rods321 Core XI, 794 ppm, 144 Pyrex Rods322 Core XI, 779 ppm, 144 Pyrex Rods323 Core XI, 1245 ppm, 72 Vicor Rods360 Core IV361 Core V362 Core VI363 Core VII364 Core VIII445 670 pins, A1203-B4C rods446 672 pins, A1203-B4C rods447 668 pins, A1203-B4C rods448 668 pins, A1203-B4C rods479 8 wt% 240Pu 1.05" pitch, B4 Rods480 8 wt% 240Pu 1.05" pitch, B3 Rods481 8 wt% 240Pu 1.05" pitch, B2 Rods482 8 wt% 240Pu 1.05" pitch, BI RodsUUUUUUU m Um -m miU ~ -MiUUUMIm -Urn'm -rn mlm -U MiUUUMIUUUMI~UUMiUUUMIUUUMIUUUMiUUUMIUUUMiUUUMI4848 wt% 240Pu 1.05" pitch, B4+CdP -i485 8 wt% 240Pu Rods105" pitch, B3+Cd U U U ..U -U486 ~RodsII486 I8 wt% 240Pu 1.05" pitch, B2+Cd~Rods4878 wt% 240Pu 1.05" pitch, B 1 +CdRods-i491 8 wt% 240Pu 1.32" pitch, B4 Rods492 8 wt% 240Pu 1.32" pitch, B3 Rods493 8 wt% 240Pu 1.32" pitch, B2 Rods494 , 8 wt% 240Pu 1.32" pitch, B 1 RodsU U ~ U UUUUUMIUUUUMIUUUUMIUUUUMIREPORT HI-2104790 C-17 Case Evaluation Identification File- ke i caic-- (exp 4-j EALFname &#xfd;fj(V496 8 wt% 240Pu 1.32" pitch, B4+Cd496 Rods497 8 wt% 240Pu 1.32" pitch, B3+CdRods498 8 wt% 240Pu 1.32" pitch, B2+Cd498 Rods5_49 16 wt% 240Pu 1.32" pitch, B4 I+CdRods504 16 wt% 240Pu 1.05" pitch, B4 Rods ____ m U ___505 16 wt% 240Pu 1.05" pitch, B3 Rods _____ m506 16 wt% 240Pu 1.05" pitch, B2 Rods ______ m507 16 wt% 240Pu 1.05" pitch, Bl Rods _____ M509 16 wt% 240Pu 1.05" pitch, B4+Cd U509 Rods510 16 wt% 240Pu 1.05" pitch, B3+Cd U510 Rods511 16 wt% 240Pu 1.05" pitch, B2+Cd U~Rods512 16wt%240Pu 1.05" pitch, Bl+CdoU5162 Rods516 24 wt% 240Pu 1.05" pitch, B4 Rods A 517 24 wt% 240Pu 1.05" pitch, B3 Rods_______
518 24 wt% 240Pu 1.05" pitch, B2 Rods ___ _ M519 24 wt% 240Pu 1.05" pitch, BI Rods _ _ _ _l521 24 wt% 240Pu 1.05" pitch, B4+Cd U521 Rods522 24 wt% 240Pu 1.05" pitch, B3+Cd U U U52 Rods -_ M_00 -o523 24 wt% 240Pu 1.05" pitch, B2+Cd U U _524 Rods524] 24 wt% 240Pu 1.05" pitch, Bl+C&#xfd;d _ _ __ R odsREPORT HI-2104790 C-i 8REPORT HI-2104790 C-18 Table C.3-6 Descriptive Statistics of the MCNP5-1.51 Calculational ResultsExperiment Description No. ofexp.keff rangeEALF (eV) rangePhase 1 18Phase 2 41Phase 3 26Phase 4 71Selected Experiments 135All experiments 291Table C.3-7 Normality Test Results for the MCNP5-1.51 calculations Experiment Description No. ofexp.Shapiro-Wilk Pearson's chi-square (x2)WtPePtwX2nPd(x2;d)NormalPhase 1 18Phase 2 41Phase 3 26Phase 4 71HTC Experiments 156Selected Experiments 135All experiments 291-_M I I=REPORT HI-2104790 C-19REPORT HI-2104790 C-19 Table C.3-8 Trending Analysis Results for the MCNP5-1.51 calculations Experiment Description No. ofexp.Correlated Parameter, xCorrelation Coefficient, 2rProbability, Pd(N;r)Correlation Regression
: Equation, k(x)EALFPhase 1 18 PitchNumber of RodsEALFPitchPhase 2 41 Number of RodsGadolinium Conc.Boron Conc.EALFPhase 3 26 Water GapNumber of RodsEALFPhase 4 71 Water GapScreen Array DistanceEALFPitchSelected Rod ODExperiments Fuel DensityU Enrichment Pu Enrichment EALFPitchAll experiments 291 RodchRod ODFuel DensityII* Im II.IUIREPORT HI-2104790 C-20REPORT HI-2104790 C-20 Table C.3-9 Analysis of Neutron Absorbers and Reflector Materials for the MCNP5-1.
51calculations Experiment No. of Bias Normality Desriptint ex. o Bias Uncerta X2 Significant TrendsDescription exp. inty (Pd(X2;d))Allexperiments 291All exceptthose withGadolinium, 201 -Cadmium andLeadTable C.3-10 Analysis of Fuel Rod Pitch Trend for the MCNP5-1.
51 calculations Bias Normality Experiment Rod Bias Uncerta X2 Significant TrendsDescription Pitch inty (Pd(X2;d))All (291All totaln Lexperiments
<2 cm -total)All except All (201 m -those with total) -Gadolinium,
_Cadmium and <=2 cm -Lead (144 -total)REPORT HI-2 104790 C-2 1REPORT HI-2104790 C-21 Table C.3-11 Analysis of Fuel Burnup for the MCNP5-1.
51 calculations Experiment Description RodPitchBiasBiasUncertaintyNormality X2(Pd(X2;d))Significant Trends-I- 4 4All exceptthose withGadolinium, Cadmium andLeadtAll (201total)--<=2 cm(144total)-mAll (61Fresh U02  total)Fuel <=2 cm(52 total) ____All (85HTC total)Experiments
<=2 cm(82 total) _ __ ]All (55total)-LL-4 + 4MOXExperiments
<=2 cm(10 total)M MUtNote: Critical experiments with Gadolinium, Cadmium and Lead were excluded from all subsequent subsets.REPORT HI-2104790 C-22 Table C.3-12 Analysis of the Unborated and Borated Water for the MCNP5-1.51 calculations Experiment Description RodPitchBiasBiasUncertaintyNormality X2(Pd(X ;d))Significant TrendsAll exceptthose withGadolinium, Cadmium andLeadtAll (201total)-m<=2 cm(144total)&#xfd;ImAll (149total)-I-All with FreshWateri i<=2 cm(94 total)E--71All (52All with total) _Borated Water <=2 cm(50 total) _tNote: Critical experiments with Gadolinium, Cadmium and Lead were excluded from all subsequent subsets.REPORT HI-2104790 C-23 Table C.3-13 Comparison of Key Parameters and Definition of Validated AOAParameter Design Benchmarks Validated Application Fissionable Material 235U, 239Pu, 241Pu 235U, 239Pu, 241Pu 235U, 239Pu, 241PuIsotopic Composition 235u/ut < 5.Owt% 1.57-5.74%
< 5wt%Pu/(U+Pu)
< 20wt% 1.104-20%
< 20wt%Physical Form U02 MOX U02 MOX U02 MOXFuel Density (g/cm3) 10.0- 10.7 9.2 -10.4 9.2 -10.7Moderator Material (coolant)
H H HPhysical Form H20 H20 H20Density (g/cm3) around 1.0 g/cm3  around 1.0 g/cm3  around 1.0 g/cm3Reflector Material H H HPhysical Form H20 H20 H20Density (g/cm3) around 1.0 g/cm3  around 1.0 g/cm3  around 1.0 g/cm3Interstitial Reflector MaterialPlate Steel or Lead Steel or Lead Steel or LeadAbsorber MaterialNone, Boron (89 -None, Boron (0 -Soluble None, Boron or 595 ppm) or 1000 ppm) orGadolinium Gadolinium (49.2 -Gadolinium (0 to199.7 ppm) 1000 ppm)Rods Boron Pyrex , Vicor' or BoronB-AlSeparating MaterialWater, B-SS, Water, B-SS, Boral Water, B-SS, BoralPlate Boral orCadmim oor Cadmium or CadmiumCadmiumGeometryLattice type Square Square, Triangle Square, Triangle1.26-1.47 Lattice Pitch (cm) (PWR) 0.968 to 4.318 0.968 to 4.318Lattce Ptch cm) 1.24 -1.88(BWR)ThermalNeutron Energy Thermal Thermal spectrum Thermal spectrumspectrum IIREPORT HI-2104790 C-24REPORT HI-2104790 C-24 Figure Proprietary Figure C.3-1 Frequency Chart for Calculated keff of the Selected 243 Benchmarks for theMCNP5-1.51 codeFigure Proprietary Figure C.3-2 Frequency Chart for Calculated EALF (eV) of the Selected 243 Benchmarks for theMCNP5-1.51 codeREPORT HI-2104790 C-25 Figure Proprietary Figure C.3-3 MCNP5-1.51 Calculated kff Values for Various Values of the Spectral Index (AllExperiments)
REPORT HI-2 104790 C-26REPORT HI-2104790 C-26 Figure Proprietary Figure C.3-4 MCNP5-1.51 Calculated k.f Values as a Function of Rod Pitch (All Experiments)
REPORT HI-2104790 C-27 Appendix DBenchmark of MCNP5-1.51 with ENDF/B-VII (total number of pages: 51 including this page)REPORT HI-2 104790 D- IREPORT HI-2104790 D-1 D.1 Introduction This Appendix presents the analysis of the validation results for MCNP5-1.51 code and includesthe results of the calculations, normality test, the detailed statistical trending
: analysis, calculation bias and bias uncertainty for each distinct area of applicability of the parameters of interest.
D.2 Computer Code Parameter DataThe computer code MCNP5-1.51
[D. 1] is the continuous energy Monte Carlo codes and treats anarbitrary three-dimensional configuration of materials in geometric cells bounded by first- andsecond-degree surfaces and fourth-degree elliptical tori. Thermal neutrons are described by boththe free gas and S(a,3) models. All calculations were performed using the default data libraries provided with the code: the default continuous energy neutron transport data based on ENDF/B-VII. The list of ZAIDs that were used in the analysis is presented in Table D.2-1. The neutronsstart from an arbitrary distribution, causing a generally very large variance of results from thefirst cycles in comparison with the following cycles. Therefore, all MCNP5-1.51 calculations areperformed with 12,000 histories per cycle, 50 skipped cycles before averaging, and 100 cyclesthat are accumulated.
The calculated kff values have associated uncertainties due to the statistical nature of the Monte Carlo codes.D.3 Analysis of MCNP5-1.51 Validation ResultsD.3.1. Calculational ResultsThe calculation results for the 156 HTC critical experiments and for the 376 selected criticalexperiments described in Appendix B are presented and discussed in this section.
The calculation results are summarized by grouping the experiments in terms of the categories as set forth inAppendix B. Calculation
: results, including keg-fc, orcatc-i and EALF, measurement uncertainties (u,,p) and the calculation and measurement combined uncertainty (au) are shown in Table D.3-1through Table D.3-5.Figure D.3-1 and D.3-2 are histograms showing the frequency of calculated ke and EALF for all532 benchmarks.
The nominal calculated k, values range from O i .The EALFresults values show a range betweenDescriptive statistics for the different group of experiments is summarized in Table D.3-6.D.3.2. Normality TestIn order to assess the normality assumption, Shapiro and Wilk [5] test has been used for groupswith fewer than 50 samples while the Pearson's chi-square (X2) test [4] has been used for sampleslarger than 20 samples.
The tests are applied to the group of experiments in terms of thecategories as set forth in Appendix B.For the Shapiro and Wilk test, Table D.3-7 shows the computed Wtest value, and W value thatcan be obtained for the number of experiments from [5] to accept the normality hypothesis.
If WREPORT HI-2104790 D-2 is less than the test statistic, Wtest, then the data is considered normally distributed.
For the X,test, it is concluded normal for x(2 < n, where n is a number of bins for the group of experiments.
The probability Pd(.k' > ,o02) of obtaining a value of,;2 > 102 in an experiment with d degreesof freedom to confirm quantitatively that the agreement is satisfactory was taken or interpolated, if necessary, from Appendix D in Reference
[4]. Thus, if Pd(&2 > 102) is large, the obtained andexpected distributions are consistent; if it is small, they probably disagree.
In particular, ifPd(X2 > ,k02) is less than 5%, we say that the disagreement is significant and reject the assumeddistributions at the 5% level. If it is less than 1%, the disagreement is called highly significant, and we reject the assumed distributions at the 1% level.D.3.3. Trending AnalysisTrends are determined through the use of regression fits to the calculated results.
The equations used to identify trends are given below:Y(x) = a + bx (7-1)(7-2)(7-3)A= zz #- t .-(7-4)The squared term of the linear correlation factor r defined below (from Reference
[5]) is used toquantitatively measure the degree to which a linear relationship exist between two variables.
1 (X, -2)(y,- y)2(7-5)The closer r2 approaches the value of 1, the better the fit of the data to the linear equation.
Amore quantitative measure of the fit can be found by using Appendix C in Reference
[4]. Theinterpolation was applied, if necessary.
For any given observed value ro, PN(IrI >- Irol) is theprobability that N measurements of two uncorrelated variables would give a coefficient r as largeas ro. Thus, if we obtain a coefficient ro for which PN(IrJ > Irol) is small, it is correspondingly unlikely that our variables are uncorrelated; that is, a correlation indicated.
In particular, ifPN(IrI -- Irol) 5 5%, the correlation is called significant; if it is less than 1%, the correlation iscalled highly significant.
REPORT HI-2 104790 D-3REPORT HI-2104790 D-3 The validation results are analyzed for the group of all experiments.
Independent variables usedin the trending
: analysis, correlation coefficients and trending analysis results are summarized inTable D.3-8.D.3.4. Bias and Bias Uncertainty In this section, benchmark results are analyzed using the statistical method described in section2.2.The first step is to evaluate whether the HTC experiments and selected ex eriments, should bereduced to a single set. The mean, of the HTC data set is and the mean kffof the selected experiments data is .The difference between the means is just0.0010 which is less than the uncertainty.
These sets are water moderated uranium or mixedplutonium-uranium dioxide lattices.
The HTC sets of experiments and the selected experiments are considered one large set of 532 experiments from now on.The normality test in Table D.3-7 and in Figure D.3-1 shows that the data is not normallydistributed.
Therefore, the distribution free approach
[6] is used for all subsets with the rejectednormality distribution.
The lower tolerance limit with 95% probability and 95% confidence levelis determined for order data [6] and the difference between weighted average keff and this lowertolerance limit is used to determine the bias uncertainty.
This is conservative since the data isclose to the normal distribution.
The distribution free bias uncertainty is also provided in allsubsequent tables for the subsets with the rejected normality assumption.
The analysis of the correlation coefficient in Table D.3-8 and the plot of data trend (Figure D.3-3) show that there is not a clear trend in the data.The total bias (systematic error or mean of the deviation from a keff of exactly 1.000) of theMCNP5-1.51 code is shown in the table belowCalculational Bias of the MCNP5-1.51 codeDescription Total Bias Bias Uncertainty HTC and Selected Experiments D.3.5. Applicability of MCNP5-1.51 Validation ResultsThis subsection contains a more detailed evaluation of the set of critical experiments.
Regarding the selected experiments, the following subjects are discussed:
* Neutron absorber and neutron reflector materials
" Neutron absorber geometry" Fuel burnup" Unborated and borated waterREPORT HI-2104790 D-4
* Various Combinations of Fuel Bumup and Unborated/Borated Water.The general focus is to justify that using the full set of critical experiments is appropriate.
Insome cases, subsets of full set of experiments are established.
For those subsets, statistical evaluations are performed to determine bias, bias uncertainty, normality and trends. Trends areevaluated for fuel rod outer diameter, fuel rod pitch, fuel density, boron content, U or Puenrichment and EALF. To estimate a significance of observed trend, the residuals from the trendequation were tested for a normal distribution
[D.2]. If residuals are normally distributed thenthere is a significant trend, otherwise there is no linear trend as this violets the basic assumptions of linear regression.
For each significant linear correlation, the bias and bias uncertainty werecalculated as a function of the independent parameter.
D.3.5. 1. Neutron Absorber and Neutron Reflector Materials The HTC and Selected Experiments consider the following neutron absorbers and reflectors:
* Absorbers o Boron, in the form of soluble boron in the water, boron in solid form (B4C), andboron in borated steel, Pyrex, Boroflex and borated aluminumo Soluble gadolinium in water and Gd203 rodso Cadmium" Reflectors o Steelo Leado WaterSome typical configurations do not contain gadolinium or cadmium neutron absorbers or leadreflectors.
To verify that including those materials does not have a significant effect on theresults of the benchmarking
: analyses, a subset without those experiments containing thosematerials was analyzed.
In addition, according to recommendations of NUREG-6979
[B. 13], thefollowing HTC experiments were also excluded:
61, 65, 67, 86, 97, 98, 99, 102, 124, 135, and137. The comparison with the full set is presented in Table D.3-9 and shows no significant differences when those materials are excluded.
: However, a significant correlation as a functionof EALF was determined by the residuals normality test. This correlation is presented in theFigure D.3-4. The bias and bias uncertainty as a function of the EALF were calculated for thistrend and shown in Table D.3-10, with a maximum absolute value of .D.3.5.2.
Absorber GeometryThe criticality experiments analyzed in this report include experiments with Boron in the form ofplates, absorber rods and soluble boron in water. No trend relating to these experiments isobserved.
D.3.5.3.
Fuel BurnupThe full set of critical experiments contains experiments with fresh U02 fuel, with simulated spent fuel (37.5 GWd/MTU),
and MOX fuel with Pu content between 1.5 and 20%, which isREPORT HI-2104790 D-5 even higher than typically found in spent fuel. The experiments are therefore reasonably representative of burned fuel at different burnup levels. To verify that the experiments cover theburnup range sufficiently, the experiments are subdivided into fresh U02 fuel and spent fuel withHTC and MOX experiments, and compared to the results of the entire set. The comparison isshown in Table D.3-1 1. The comparison shows no significant differences between the entire setand the fresh and spent fuel subsets.
: However, in some cases, the correlations are observed.
Thesignificant trends as a function of EALF and Pu enrichment were determined in the spent fuelsubset by the residuals normality test. These correlations are presented in the Figure D.3-5 andFigure D.3-6. The bias and bias uncertainty as a function of the EALF and Pu enrichment werecalculated for these trends and shown in Table D.3-12, with a maximum absolute value of -D.3.5.4.
Unborated and Borated WaterThe full set of critical experiments contains both experiments with and without soluble boron.The entire set of analyses shows no significant trend when analyzed as a function of the solubleboron level. Nevertheless, sets with and without soluble boron are analyzed and compared to thefull set that contains all experiments.
The results are shown in Table D.3-13. Similar to theprevious subsection, the comparison shows no significant differences between those subsets.However, there are significant trends in the fresh water subset as a function of EALF and Uenrichment and in the borated water subset as a function of fuel density that were determined bythe residuals normality test. These correlations are presented in the Figure D.3-7 through FigureD.3-9. The bias and bias uncertainty as a function of the EALF, U enrichment and fuel densitywere calculated for these trends and shown in Table D.3-14, with a maximum absolute value of-D.3.5.5.
Various Combinations of Fuel Burnup and Unborated/Borated WaterTo perform more detailed evaluation of the set of critical experiments, the additional four subsetswith different combinations of fuel bumup and unborated/borated water were analyzed.
Theresults are shown in Table D.3-15. There are significant EALF trend in the subset of fresh U02fuel with fresh water, significant trends in the subset of spent fuel with fresh water as a functionof EALF and Pu enrichment and significant trends in the subset of spent fuel with borated wateras a function of rod OD and fuel density, that were determined by the residuals normality test.These correlations are presented in the Figure D.3-10 through Figure D.3-14.The bias and biasuncertainty as a function of the EALF, Pu enrichment, rod OD and fuel density were calculated for these trends and shown in Table D.3-16, with a maximum absolute value ofD.4 SummaryA set of 532 critical experiments has been selected and has been used for the validation of theHoltec International criticality safety methodology.
The similarity between the chosenexperiments and the actual systems has been based on a set of screening criteria as is stated in theNUREG/CR-6698
[5]. Experiments have been categorized by fuel burnup as fresh U02 fuel andspent fuel with HTC and MOX experiments or by unborated and borated water condition andREPORT HI-2104790 D-6 parameterized by key variables such as lattice pitch / assembly pitch, absorber solutionconcentration, number of fuel rods, rod outer diameter, fuel density, screen array distance, fuelenrichment and EALF. Benchmark calculations have been performed using the Monte Carlocode MCNP5-1.51.
It was determined that HTC experiments and selected experiments are insufficient agreement that this sets are lumped together as a single set of 532 experiments.
Thebias and bias uncertainty are presented in section D.3.4. The applicability of validation results isconsidered in section D.3.5.The range of key parameters for the design application, benchmarks and validated AOA aresummarized in Table D.3-17. A point by point comparison between design application andbenchmarks shows that the experimental range covers all the parameters.
The soluble boronconcentration is extrapolated generously since &deg;B is a 1/v absorber (as permitted on Table 2.3 of[5]).As for the fuel density, Table 2.3 of Reference
[5] states there is "no requirement" and that"experiments should be as close to the desired concentration as possible".
Since the experiment fuel density is 9.2 -10.4 g/cm3 and the design application one is around 10.0 -10.7 g/cm3, it isconsidered that the values are very close so the validated AOA covers the design application range.The fuel enrichment can be up to 5%. The experiments used go up to 5.74 wt% 235U. Therefore, it is considered that the validated AOA covers the design application range.D.5 References
[D. 1] "MCNP -A General Monte Carlo N-Particle Transport Code, Version 5"; Los AlamosNational Laboratory, LA-IJR-03-1987 (Revised 2/1/2008).
[D.2] J. W. Barnes, "Statistical Analysis for Engineers and Scientists",
McGraw-Hill Inc., 1988REPORT HI-2104790 D-7 Table D.2-1 ZAIDs Used for Each NuclideMCNP5.1.51 MCNP5.1.51 MCNP5.1.51 Nuclide ZAD Nuclide ZAD Nuclide ZIZAID ZAID ZAID'H 1001.70c 48Ti 22048.70c
'l&deg;Mo 42100.70c 2 H 1002.70c 49Ti 22049.70c 107Ag 47107.70c 4He 2004.70c 5&deg;Ti 22050.70c 109Ag 47109.70c 10B 5010.70c 50Cr 24050.70c 106Cd 48106.70c "1B 5011.70c 52Cr 24052.70c 108Cd 48108.70c C 6000.70c 53Cr 24053.70c l"0Cd 48110.70c 14N 7014.70c 54Cr 24054.70c 111Cd 48111.70c 160 8016.70c 55Mn 25055.70c 112Cd 48112.70c 2&deg;Ne 10020.42c 54Fe 26054.70c 113Cd 481 13.70c23Na 11023.70c 56Fe 26056.70c 114Cd 48114.70c 24Mg 12024.70c 57Fe 26057.70c 116Cd 48116.70c 25Mg 12025.70c 58Fe 26058.70c 1131n 49113.70c 26Mg 12026.70c 59Co 27059.70c 1151n 49115.70c 27A1 13027.70c 58Ni 28058.70c 12 Sn 50112.70c 28Si 14028.70c 6&deg;Ni 28060.70c 114Sn 50114.70c 29Si 14029.70c 61Ni 28061.70c "15Sn 50115.70c 3&deg;Si 14030.70c 62Ni 28062.70c 116Sn 50116.70c 31p 15031.70c 64Ni 28064.70c 17TSn 50117.70c 32s 16032.70c 63Cu 29063.70c 118Sn 50118.70c 36Ar 18036.70c 65Cu 29065.70c 119Sn 501 19.70c38Ar 18038.70c Zn 30000.70c 120Sn 50120.70c 4&deg;Ar 18040.70c 90Zr 40090.70c 122Sn 50122.70c 39K 19039.70c 91Zr 40091.70c 124Sn 50124.70c 40K 19040.70c 92Zr 40092.70c 144Sm 62144.70c 41K 19041.70c 94Zr 40094.70c 147Sm 62147.70c 4&deg;Ca 20040.70c 96Zr 40096.70c 148Sm 62148.70c 42Ca 20042.70c 93Nb 41093.70c 149sm 62149.70c 43Ca 20043.70c 92Mo 42092.70c
&deg;50Sm 62150.70c 44Ca 20044.70c 94Mo 42094.70c I52Sm 62152.70c 46Ca 20046.70c 95Mo 42095.70c
'54Sm 62154.70c 48Ca 20048.70c 96Mo 42096.70c 152Gd 64152.70c 46Ti 22046.70c 97Mo 42097.70c 154Gd 64154.70c 47Ti 22047.70c 98Mo 42098.70c 155Gd 64155.70c REPORT HI-2104790 D-8 Nuclide MCNP5.1.51 MCNP5.1.51 Nlid MCNP5.1.51 ZAID ZAID ZAID156Gd 64156.70c 179Hf 72179.70c 236U 92236.70c 157Gd 64157.70c 180Hf 72180.70c 238u 92238.70c 158Gd 64158.70c 204Pb 82204.70c 238Pu 94238.70c 16&deg;Gd 64160.70c 206Pb 82206.70c 239pu 94239.70c 174Hf 72174.70c 207Pb 82207.70c 240Pu 94240.70c 176Hf 72176.70c 208Pb 82208.70c 241Pu 94241.70c 177Hf 72177.70c 234u 92234.70c 242Pu 94242.70c 178Hf 72178.70c 235u 92235.70c 241AnM 95241.70c REPORT HI-2 104790 D-9REPORT HI-2104790 D-9 Table D.3-1 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase1 Critical Experiments:
Water-Moderated and Reflected ArraysFile-CaseEvaluation Identification kDif--I- Ccalc-i+/- Gexp+/- (yiEALF(All)I MIX-COMP-THERM-HTC-001 2 MIX-COMP-THERM-HTC-002 3 MIX-COMP-THERM-HTC-003 4 MIX-COMP-THERM-HTC-004 5 MIX-COMP-THERM-HTC-005 6 MIX-COMP-THERM-HTC-006 7 MIX-COMP-THERM-HTC-007 8 MIX-COMP-THERM-HTC-008 9 MIX-COMP-THERM-HTC-009 10 MIX-COMP-THERM-HTC-010 11 MIX-COMP-THERM-HTC-0 1112 MIX-COMP-THERM-HTC-012 13 MIX-COMP-THERM-HTC-013 14 MIX-COMP-THERM-HTC-014 15 MIX-COMP-THERM-HTC-015 16 MIX-COMP-THERM-HTC-016 17 MIX-COMP-THERM-HTC-017 18 MIX-COMP-THERM-HTC-018 REPORT HI-2104790 D-10 Table D.3-2 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 2Experiments:
Reflected Simple Arrays Moderated by Poisoned Water with Gadolinium or BoronCriticalCaseEvaluation Identification File-namekeff-i1 acalc-i-(Yexp+/- "3iEALF (eV)19 MIX-COMP-THERM-HTC-019 20 MIX-COMP-THERM-HTC-020 21 MIX-COMP-THERM-HTC-021 22 MIX-COMP-THERM-HTC-022 23 MIX-COMP-THERM-HTC-023 24 MIX-COMP-THERM-HTC-024 25 MIX-COMP-THERM-HTC-025 26 MIX-COMP-THERM-HTC-026 27 MIX-COMP-THERM-HTC-027 28 MIX-COMP-THERM-HTC-028 29 MIX-COMP-THERM-HTC-029 30 MIX-COMP-THERM-HTC-030 31 MIX-COMP-THERM-HTC-031 32 MIX-COMP-THERM-HTC-032 33 MIX-COMP-THERM-HTC-033 34 MIX-COMP-THERM-HTC-034 35 MIX-COMP-THERM-HTC-035 36 MIX-COMP-THERM-HTC-036 37 MIX-COMP-THERM-HTC-037 38 MIX-COMP-THERM-HTC-038 39 MIX-COMP-THERM-HTC-039 40 MIX-COMP-THERM-HTC-040 41 MIX-COMP-THERM-HTC-041 42 MIX-COMP-THERM-HTC-042 43 MIX-COMP-THERM-HTC-043 44 MIX-COMP-THERM-HTC-044 45 MIX-COMP-THERM-HTC-045 46 MIX-COMP-THERM-HTC-046 47 MIX-COMP-THERM-HTC-047 48 MIX-COMP-THERM-HTC-048 49 MIX-COMP-THERM-HTC-049 50 MIX-COMP-THERM-HTC-050 51 MIX-COMP-THERM-HTC-051 52 MIX-COMP-THERM-HTC-052 53 MIX-COMP-THERM-HTC-053 54 MIX-COMP-THERM-HTC-054 55 MIX-COMP-THERM-HTC-055 REPORT HI-2 104790 D- 11REPORT HI-2104790 D-11 File-nameCaseEvaluation Identification
+- Gcalc-i+/- crexp-(yiEALF (eV)56 MIX-COMP-THERM-HTC-056 57 MIX-COMP-THERM-HTC-057 58 MIX-COMP-THERM-HTC-058 59 MIX-COMP-THERM-HTC-059 IREPORT HI-2 104790 D-12REPORT HI-2104790 D-12 Table D.3-3 The MCNP5-1.51 Calculational Results and Measurements Phase 3 Critical Experiments:
Pool StorageUncertainties forCaseEvaluation Identification File- kff-inn~mp kI fI Gcalc-iI eFXP I+/- c7iEALF (eV)60 MIX-COMP-THERM-HTC-060 61 MIX-COMP-THERM-HTC-061 62 MIX-COMP-THERM-HTC-062 63 MIX-COMP-THERM-HTC-063 64 MIX-COMP-THERM-HTC-064 65 MIX-COMP-THERM-HTC-065 66 MIX-COMP-THERM-HTC-066 67 MIX-COMP-THERM-HTC-067 68 MIX-COMP-THERM-HTC-068 69 MIX-COMP-THERM-HTC-069 70 MIX-COMP-THERM-HTC-070 71 MIX-COMP-THERM-HTC-071 72 MIX-COMP-THERM-HTC-072 73 MIX-COMP-THERM-HTC-073 74 MIX-COMP-THERM-HTC-074 75 MIX-COMP-THERM-HTC-075 76 MIX-COMP-THERM-HTC-076 77 MIX-COMP-THERM-HTC-077 78 MIX-COMP-THERM-HTC-078 79 MIX-COMP-THERM-HTC-079 80 MIX-COMP-THERM-HTC-080 81 MIX-COMP-THERM-HTC-081 82 MIX-COMP-THERM-HTC-082 83 MIX-COMP-THERM-HTC-083 84 MIX-COMP-THERM-HTC-084 85 MIX-COMP-THERM-HTC-085 II-REPORT HI-2 104790 D- 13REPORT HI-2104790 D-13 Table D.3-4 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 4 CriticalExperiments:
Shipping CaskFile-CaseEvaluation Identification keff.i :+/- GcaxpI+/- (7iEALF(eV)86 MDI-COMP-THERM-HTC-086 87 MIX-COMP-THERM-HTC-087 88 MIX-COMP-THERM-HTC-088 89 MIX-COMP-THERM-HTC-089 90 MIX-COMP-THERM-HTC-090 91 MIX-COMP-THERM-HTC-091 92 MIX-COMP-THERM-HTC-092 93 MIX-COMP-THERM-HTC-093 94 MIX-COMP-THERM-HTC-094 95 MIX-COMP-THERM-HTC-095 96 MIX-COMP-THERM-HTC-096 97 MIX-COMP-THERM-HTC-097 98 MIX-COMP-THERM-HTC-098 99 MIX-COMP-THERM-HTC-099 100 MIX-COMP-THERM-HTC-100 101 MDI-COMP-THERM-HTC-101102 MIX-COMP-THERM-HTC-102 103 MIX-COMP-THERM-HTC-103 104 MIX-COMP-THERM-HTC-104 105 MIX-COMP-THERM-HTC-105 106 MIX-COMP-THERM-HTC-106 107 MIX-COMP-THERM-HTC-107 108 MIX-COMP-THERM-HTC-108 109 MIX-COMP-THERM-HTC-109 110 MIX-COMP-THERM-HTC-1 10111 MIX-COMP-THERM-HTC-1 111112 MIX-COMP-THERM-HTC-1 12113 MIX-COMP-THERM-HTC-1 13114 MIX-COMP-THERM-HTC-1 14115 MIX-COMP-THERM-HTC-1 15116 MIX-COMP-THERM-HTC-1 16117 MIX-COMP-THERM-HTC-117118 MIX-COMP-THERM-HTC-1 18119 MIX-COMP-THERM-HTC-119120 MIX-COMP-THERM-HTC-120 121 MIX-COMP-THERM-HTC-121 122 MIX-COMP-THERM-HTC-122 123 MIX-COMP-THERM-HTC-123 REPORT HI-2 104790 D- 14REPORT HI-2104790 D-14 File--I- Ucalc-CaseEvaluation Identification keff-i4- (FexpEALF(,=X'l124 MIX-COMP-THERM-HTC-124125 MIX-COMP-THERM-HTC-125126 MIX-COMP-THERM-HTC-126127 MIX-COMP-THERM-HTC-127128 MIX-COMP-THERM-HTC-128129 MIX-COMP-THERM-HTC-129130 MIX-COMP-THERM-HTC-130131 MIX-COMP-THERM-HTC-131132 MIX-COMP-THERM-HTC-132133 MIX-COMP-THERM-HTC-133 134 MIX-COMP-THERM-HTC-134 135 MIX-COMP-THERM-HTC-135136 MIX-COMP-THERM-HTC-136 137 MIX-COMP-THERM-HTC-137138 MIX-COMP-THERM-HTC-138 139 MIX-COMP-THERM-HTC-139 140 MIX-COMP-THERM-HTC-140 141 MIX-COMP-THERM-HTC-141142 MIX-COMP-THERM-HTC-142 143 MIX-COMP-THERM-HTC-143 144 MIX-COMP-THERM-HTC-144 145 MIX-COMP-THERM-HTC-145 146 MIX-COMP-THERM-HTC-146 147 MIX-COMP-THERM-HTC-147148 MIX-COMP-THERM-HTC-148 149 MIX-COMP-THERM-HTC-149 150 MIX-COMP-THERM-HTC-150 151 MIX-COMP-THERM-HTC-151 152 MIX-COMP-THERM-HTC-152153 MIX-COMP-THERM-HTC-153 154 MIX-COMP-THERM-HTC-154 155 MIX-COMP-THERM-HTC-155 156 MIX-COMP-THERM-HTC-156 REPORT HI-2 104790 D- 15REPORT HI-2104790 D-15 Table D.3-5 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for SelectedCritical Experiments r r 1CaseEvaluation Identification File-keff-iI (cale-EALFI Gexp-+/-' (i157 Core l158 Core II159 Core HII160 Core IX161 Core X162 Core XI163 Core XII164 Core XIII165 Core XIV166 Core XV167 Core XVI168 Core XVII169 Core XVIII170 Core XIX171 Core XX172 Core XXI173 S-type Fuel, w/886 ppm B174 S-type Fuel, w/746 ppm B175 SO-type Fuel, w/1 156 ppm B176 Case 1 1337 ppm B177 Case 12 1899 ppmB B178 Water Moderator 0 gap179 Water Moderator 2.5 cm gap180 Water Moderator 5 cm gap181 Water Moderator 10 cm gap182 Steel Reflector, 1.321 cm separation 183 Steel Reflector, 2.616 cm separation 184 Steel Reflector, 3.912 cm separation 185 Steel Reflector, Infinite separation 186 Steel Reflector, 1.321 cm separation 187 Steel Reflector, 2.616 cm separation 188 Steel Reflector, 5.405 cm separation 189 Steel Reflector, Infinite separation 190 Steel Reflector, with Boral Sheets191 Lead Reflector, 0.55 cm sepn.192 Lead Reflector, 1.956 cm sepn.193 Lead Reflector, 5.405 cm sepn.194 Experiment 004/032 -no absorberREPORT HI-2104790 D- 16 File-CaseEvaluation Identification keff-i-Gcalc._+/-- (yexp+/- GiEALF(P17)195 Exp. 009 1.05% Boron Steel plates196 Exp. 009 1.62% Boron Steel plates197 Exp. 031 -Boral plates198 Experiment 214R- with flux traps199 Experiment 214V3 -with flux trap200 Case 173 -0 ppm B201 Case 177 -2550 ppm B202 MOX Fuel -Type 3.2 Exp. 21203 MOX Fuel -Type 3.2 Exp. 43204 MOX Fuel -Type 3.2 Exp. 13205 MOX Fuel -Type 3.2 Exp. 32206 Saxton Case 52 PuO2 0.52" pitch207 Saxton Case 52 U 0.52" pitch208 Saxton Case 56 PuO2 0.56" pitch209 Saxton Case 56 borated PuO2210 Saxton Case 56 U 0.56" pitch211 Saxton Case 79 PuO2 0.79" pitch212 Saxton Case 79 U 0.79" pitch213 0.700-in.
pitch 0 ppm B214 0.700-in.
pitch 688 ppm B215 0.870-in.
pitch 0 ppm B216 0.870-in.
pitch 1090 ppm B217 0.990-in.
pitch 0 ppm B218 0.990-in.
pitch 767 ppm B219 Saxton Case PuO2 0.735" pitch220 Saxton Case PuO2 1.04" pitch221 8 wt% 240Pu 0.80" pitch222 8 wt% 240Pu 0.93" pitch223 8 wt% 240Pu 1.05" pitch224 8 wt% 240Pu 1.143" pitch225 8 wt% 240Pu 1.32" pitch226 8 wt% 240Pu 1.386" pitch227 16 wt% 240Pu 0.93" pitch228 16 wt% 240Pu 1.05" pitch229 16 wt% 240Pu 1.143" pitch230 16 wt% 240Pu 1.32" pitch231 24 wt% 240Pu 0.80" pitch232 24 wt% 240Pu 0.93" pitch233 24 wt% 240Pu 1.05" pitch234 24 wt% 240Pu 1.143" pitchREPORT HI-2 104790 D- 17REPORT HI-2104790 D-17 CaseEvaluation Identification File-namekeiffi 1 1 acalc- 1 1 exp 1+- CjEALF235 24 wt% 240Pu 1.32" pitch236 24 wt% 240Pu 1.386" pitch237 18 wt% 240Pu 0.85" pitch238 18 wt% 240Pu 0.93" pitch239 18 wt% 240Pu 1.05" pitch240 18 wt%240Pu 1.143" pitch241 18 wt% 240Pu 1.386" pitch242 18 wt% 240Pu 1.60" pitch243 18 wt% 240Pu 1.70" pitch244 1 Cluster245 3 Clusters, Separation 11.92 cm246 3 Clusters, Separation 8.41 cm247 3 Clusters, Separation 10.05 cm248 3 Clusters, Separation 6.39 cm249 3 Clusters, Separation 9.01 cm250 3 Clusters, Separation 4.46251 1 Cluster, l0xll.51252 1 Cluster, 9x13.35253 1 Cluster, 8x16.37254 3 Clusters, Separation 7.11 cmI1 Cluster, 614.4 Rods, Gd water255 impurity256 1 Cluster, 529.3 Rods257 1 Cluster, 523.9 Rods258 1 Cluster, 525.3 Rods259 1 Cluster, 595.4 Rods260 1 Cluster, 485.8 Rods261 1 Cluster, 523.8 Rods262 1 Cluster, 505.4 Rods263 4 Clusters, Separation 2.59 cm264 2 Clusters, Separation 1.68 cm265 4 Clusters, Separation 4.27 cm266 4 Clusters, Separation 5.95 cm267 4 Clusters, Separation 5.11 cm268 4 Clusters, Separation 6.66 cm269 4 Clusters, Separation 7.53 cm270 4 Clusters, Separation 9.00 cm271 4 Clusters, Separation 9.97 cm272 4 Clusters, Separation 11.45 cm273 4 Clusters, Separation 13.87 cm--- --- -REPORT HI-2104790 D-18 CaseEvaluation Identification File-* 1 -'pkeff. + calc-274 -3 Clusters, Separation 9.88 cm275 -3 Clusters, Separation 6.78 cm276 3 Clusters, Separation 6.176 cm277 1 Cluster, 225.8 Rods, Gd waterimpurity278 1 Cluster, 216.2 Rods279 1 Cluster, 216.6 Rods280 1 Cluster, 218.6 Rods281 1 Cluster, 167.85 Rods282 1 Cluster, 203 Rods283 1 Cluster, 173.5 Rods284 2 Clusters, Separation 2.83 cm285 3 Clusters, Separation 12.27 cm286 3 Clusters, Separation 12.493 cm287 4 Clusters, Separation 4.72 cm288 4 Clusters, Separation 8.38 cm289 4 Clusters, Separation 10.86 cm290 4 Clusters, Separation 11.29 cm291 4 Clusters, Separation 12.02 cm292 4 Clusters, Separation 13.64 cm293 4 Clusters, Separation 14.98 cm294 4 Clusters, Separation 19.81 cm295 4 Clusters, Separation 8.50 cm296 19xl9, RodPitchi-1.849 cm297 20x20, Rod Pitch -1.849 cm298 2 1 x2C1, RodPitchi-1.849 cm299 17x 17, RodPitchi-1.956 cm300 18x 18, RodPitchi-1.956 cm301 19x19, Rod Pitch -1.956 cm302 20x20, Rod Pitch -1.956 cm303 21x21, Rod Pitch -1.956 cm304 16x 16, Rod Pitch -2.15 cm305 17x17, RodPitch-2.156cm306 18x18, RodPitch-2.156cm307 19x19, Rod Pitch -2.56 cm308 20x20, Rod Pitch -2.15 cm309 15x215, Rod Pitch -2.293 cm310 16x 16, Rod Pitch -2.293 cm311 17xl7, Rod Pitch -2.293 cm312 18x18, Rod Pitch -2.293 cm-7.+ Gx----T7IREPORT 111-2104790 D- 19REPORT HI-2104790 D- 19 CaseEvaluation Identification File-keff-i I- (ycalc- -+/- Gexp+/- (YiEALF313 19x19, Rod Pitch -2.293 cm314 Core XI, 1511 ppm315 Core XI, 1335.5 ppm316 Core XI, 1335.5 ppm317 Core XI, 1182 ppm, 36 Pyrex Rods318 Core XI, 1182 ppm, 36 Pyrex Rods319 Core XI, 1032.5 ppm, 72 Pyrex Rods320 Core XI, 1032.5 ppm, 72 Pyrex Rods321 Core XI, 794 ppm, 144 Pyrex Rods322 Core XI, 779 ppm, 144 Pyrex Rods323 Core XI, 1245 ppm, 72 Vicor Rods324 Core XI, 1384 ppm, 144 A1203 Rods325 Core XI, 1348 ppm, 36 A1203 Rods326 Core XI, 1348 ppm, 36 A1203 Rods327 Core XI, 1363 ppm, 72 A1203 Rods328 Core XI, 1362 ppm, 72 A1203 Rods329 Core XI, 1158 ppm330 Core XI, 921 ppm331 0% Boron Steel plates, dist. 0.245 cm332 0% Boron Steel plates, dist. 3.277 cm333 0% Boron Steel plates, dist. 0.428 cm334 0% Boron Steel plates, dist. 3.277 cm3351.05% Boron Steel plates, dist. 3.277cm--mmmm-3361.62% Boron Steel plates, dist. 3.277cm--- mlm -337 Al plates, dist. 0.105 cm338 Al plates, dist. 3.277 cm339 Zircaloy-4 plates, dist. 0.078 cm340 Zircaloy-4 plates, dist. 3.277 cm341 Lead Reflector, 0 cm separation 342 Lead Reflector, 0.660 cm separation 343 Lead Reflector, 1.321 cm separation 344 Lead Reflector, 5.405 cm separation 345 Steel Reflector, 0 cm separation 346 Steel Reflector, 0.660 cm separation 347 Steel Reflector, 1.321 cm separation 348 Steel Reflector, 2.616 cm separation 349 Steel Reflector, 5.405 cm separation 350 Steel Reflector, 0 cm separation REPORT HI-2 104790 D-20REPORT HI-2104790 D-20 File-CaseEvaluation IdentificationF-i 'calcE.4 Gexp-(;iEALF351 Steel Reflector, 0.660 cm separation 352 Steel Reflector, 1.956 cm separation 353 Lead Reflector, 0 cm separation 354 Core lilA355 Core IIC356 Core HID357 Core IIIE358 Core IIIF359 Core IIG360 Core IV361 Core V362 Core VI363 Core VII364 Core VIII0% Boron Steel plate, Gd waterimpurity366 1.1% Boron Steel plate367 1.6% Boron Steel plate368 Boral B plate369 Boral C plate370 Boroflex, 1.84 cm separation 371 Boroflex, 1.73 cm separation 372 Steel Reflector, 0% Boron Steel plate.- --- -373Steel Reflector, 1.1% Boron Steelplate--- -I374Steel Reflector,
: Boroflex, 8.37 cmseparation
-=-I375 Borated Water, 490 ppm376 Unborated Water377 Borated Water, 1030 ppm378 0% Boron Steel plates, dist. 0.645 cm379 0% Boron Steel plates, dist. 2.732 cm380 0% Boron Steel plates, dist. 4.042 cm381 0% Boron Steel plates, dist. 0.645 cm382 0% Boron Steel plates, dist. 4.042 cm383 0% Boron Steel plates, dist. 0.645 cm384 0% Boron Steel plates, dist. 4.042 cm3851.05% Boron Steel plates, dist. 0.645cm-m-m--- m386 1.05% Boron.Steel plates, dist. 4.0421___ cm J m_____ J ________=_____________________
REPORT HI-2 104790 D-2 1REPORT HI-2104790 D-21 Case Evaluation Identification File- ( L EALFname ___ (V387 1.62% Boron Steel plates, dist. 0.645cm388 1.62% Boron Steel plates, dist. 4.042cm389 Boral plates, dist. 0.645 cm390 Boral plates, dist. 4.442 cm391 Boral plates, dist. 0.645 cm392 Al plates, dist. 0.645 cm393 Al plates, dist. 4.042 cm394 Al plates, dist. 4.442 cm395 Zircaloy-4 plates, dist. 0.645 cm396 Zircaloy-4 plates, dist. 4.042 cm397 Hex, 621 Rods, Temperature 20.1C398 Hex, 889 Rods, Temperature 231A.4C399 Hex, 1951 Rods, Temperature 19.3C400 Hex, 2791 Rods, Temperature 206.OC401 Hex, 325/680 Rods, Temperature 20.8C402 Hex, 325/912 Rods, Temperature 212.1C403 Core XIA ---404 Core XIC ---405 Core XID ---406 Core XIE407 Core XIF408 Core XIG409 Core XIIIA410 No Boron Steel plates411 0% Boron Steel plates, 3 mm, dist. 0412 0% Boron Steel plates, 6 mm, dist. 0413 0% Boron Steel plates, 6 mm, dist. 0.5414 0% Boron Steel plates, 6 mm, dist. 1415 0.67% Boron Steel plates, 3 mm, dist.0416 0.67% Boron Steel plates, 6 mm, dist.0417 0.67% Boron Steel plates, 3 mm, dist.0.5418 0.67% Boron Steel plates, 6 mm, dist.0.5419 0.67% Boron Steel plates, 3 mm, dist.1420 0.67% Boron Steel plates, 6 mm, dist. _____ ______REPORT HI-2104790 D-22 Case Evaluation Identification 14210.98% Boron Steel plates, 3 mm, dist.0-m- -nm m422 0.98% Boron Steel plates, 6 mm, dist.0423 0.98% Boron Steel plates, 6 mm, dist._____ ~~~~0.5 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _4240.98% Boron Steel plates, 6 mm, dist.1---mm-425 No Boron Steel plates426 0% Boron Steel plates, dist. 0427 0.67% Boron Steel plates, dist. 0428 0.98% Boron Steel plates, dist. 0429 No Boron Steel plates430 0% Boron Steel plates, dist. 0431 0% Boron Steel plates, dist. 0.5432 0% Boron Steel plates, dist. 0433 0% Boron Steel plates, dist. 0.5434 0.67% Boron Steel plates, dist. 0435 0.67% Boron Steel plates, dist. 0.5436 0.67% Boron Steel plates, dist. 0437 0.67% Boron Steel plates, dist. 0.5438 0.98% Boron Steel plates, dist. 0439 0.98% Boron Steel plates, dist. 0.5.440 0.98% Boron Steel plates, dist. 0441 0.98% Boron Steel plates, dist. 0.5442 Otto Hahn, ZrB2 and B4C rods443 f'EN/MB-0l1 (580 pins)444 IPEN/MB-01 (560 pins)445 670 pins, A1203-B4C rods446 672 pins, A1203-B4C rods447 668 pins, A1203-B4C rods448 668 pins, A1203-B4C rods449 664 pins, 16 steel rods450 662 pins, 18 steel rods451 658 pins, 14 steel rods452 660 pins, 12 steel rods453 660 pins, 12 steel rods454 661 pins, 17 steel rods455 662 pins, 16 steel rods456 634 pins, 12 steel rodsREPORT HI-2 104790 D-23REPORT HI-2104790 D-23 File-CaseEvaluation Identification keff-i+ Gcalc- +/- exp+/- cTiEALF(ANA457 620 pins, 26 steel rods458 668 pins, 0 steel rods, 4 Gd203 rods459 648 pins, 0 steel rods, 8 Gd203 rods460 672 pins, 0 steel rods, 4 Gd203 rods461 646 pins, 4 steel rods, 4 Gd203 rods462 656 pins, 4 steel rods, 4 Gd203 rods463 664 pins, 4 steel rods, 2 Gd203 rods464 670 pins, 2 steel rods, 2 Gd203 rods465 664 pins, 2 steel rods, 2 Gd203 rods466 656 pins, 0 steel rods, 2 Gd203 rods467 23x23, 1.825 cm pitch468 23x23, 1.825 cm pitch469 23x23, 1.825 cm pitch470 21x21, 1.956 cm pitch471 21x21, 1.956 cm pitch472 21x21, 1.956 cm pitch473 20x20, 2.225 cm pitch474 20x20, 2.225 cm pitch475 20x20, 2.225 cm pitch476 21 x21,2.474 cm pitch477 21x21, 2.474 cm pitch478 8 wt% 240Pu 1.05" pitch, Al Rods479 8 wt% 240Pu 1.05" pitch, B4 Rods480 8 wt% 240Pu 1.05" pitch, B3 Rods481 8 wt% 240Pu 1.05" pitch, B2 Rods482 8 wt% 240Pu 1.05" pitch, B I Rods4838 wt% 240Pu 1.05" pitch, AI+CdRods---r -- -484 8 wt% 240Pu 1.05" pitch, B4+CdRods485 8 wt% 240Pu 1.05" pitch, B3+CdRods486 8 wt% 240Pu 1.05" pitch, B2+CdRods487 8 wt% 240Pu 1.05" pitch, B I +CdRods488 8 wt%/o 240Pu 1.05" pitch, Air+CdRods489 8 wt% 240Pu 1.05" pitch, H20+CdRods490 8 wt% 240Pu 1.32" pitch, Al Rods _ __ _491 8 wt% 240Pu 1.32" pitch, B4 Rods _ _ _ _REPORT HI-2 104790 D-24REPORT HI-2104790 D-24 File- Irai EALFCase Evaluation Identification name kf alc : GeV +/- (eV)492 8 wt% 240Pu 1.32" pitch, B3 Rods493 8 wt% 240Pu 1.32" pitch, B2 Rods494 8 wt% 240Pu 1.32" pitch, BI Rods495 8 wt% 240Pu 1.32" pitch, AI+CdRods _____496 8 wt% 240Pu 1.32" pitch, B4+CdRods497 8 wt% 240Pu 1.32" pitch, B3+CdRods498 8 wt% 240Pu 1.32" pitch, B2+CdRods499 8 wt% 240Pu 1.32" pitch, B 1 +CdRods _______500 8 wt% 240Pu 1.32" pitch, Air+CdRods501 8 wt% 240Pu 1.32" pitch, H20+CdRods502 16 wt% 240Pu 1.386" pitch503 16 wt% 240Pu 1.05" pitch, Al Rods504 16 wt% 240Pu 1.05" pitch, B4 Rods505 16 wt% 240Pu 1.05" pitch, B3 Rods506 16 wt% 240Pu 1.05" pitch, B2 Rods507 16wt%240Pu1.05"pitch, B1Rods508 16 wt% 240Pu 1.05" pitch, AI+CdRods _______509 16 wt% 240Pu 1.05" pitch, B4+CdRods510 16 wt% 240Pu 1.05" pitch, B3+CdRods511 16 wt% 240Pu 1.05" pitch, B2+CdRods512 16 wt% 240Pu 1.05" pitch, BI+CdRods _____513 16 wt% 240Pu 1.05" pitch, Air+CdRods514 16 wt% 240Pu 1.05" pitch, H20+CdRods515 24 wt% 240Pu 1.05" pitch, Al Rods516 24 wt% 240Pu 1.05" pitch, B4 Rods517 24 wt% 240Pu 1.05" pitch, B3 Rods518 24 wt% 240Pu 1.05" pitch, B2 Rods519 24 wt% 240Pu 1.05" pitch, B1 Rods520 24 wt% 240Pu 1.05" pitch, AI+CdRods _ _ __521 24 wt% 240Pu 1.05" pitch, B4+Cd _ ___REPORT HI-2104790 D-25 CaseEvaluation Identification Rods52224 wt% 240Pu 1.05" pitch, B3+Cd---n--mRods523 24 wt% 240Pu 1.05" pitch, B2+CdRods524 24 wt% 240Pu 1.05" pitch, Bl+CdRods525 24 wt% 240Pu 1.05" pitch, Air+CdRods52624 wt% 240Pu 1.05" pitch, H20+CdRods----m-527 8 wt% 240Pu 0.55" pitch528 8 wt% 240Pu 0.60" pitch529 8 wt% 240Pu 0.71" pitch530 8 wt% 240Pu 0.80" pitch531 8 wt% 240Pu 0.90" pitch532 8 wt% 240Pu 0.93" pitchREPORT HI-2104790 D-26 Table D.3-6 Descriptive Statistics of the MCNP5-1.51 Calculational ResultsExperiment Description No. ofexp.keff rangeHTC Experiments 156Selected Experiments 376All experiments 532EALF (eV) rangeTable D.3-7 Normality Test Results for the MCNP5-1.51 calculations Experiment Description No. ofexp.Shapiro-Wilk Pearson's chi-square (X2)WtestWHTC Experiments 156 N/A N/ASelected Experiments 376 N/A N/AAll experiments 532 N/A N/AnmmmPd(X2;d)NormalmmmTable D.3-8 Trending Analysis Results for the MCNP5-1.51 calculations Experiment Description No. ofexp.Correlated Parameter, xCorrelation Coefficient, 14Probability, Pd(N;r)Correlation Regression
: Equation, k(x)+ 4-Allexperiments 532EALFPitchRod ODFuel Density* m* ImREPORT HI-2104790 D-27REPORT HI-2104790 D-27 Table D.3-9 Analysis of Neutron Absorbers and Reflector Materials for the MCNP5-1.51 calculations Experiment Description No.ofexp.BiasBiasUncertainty Normality X2(Pd(X2;d))
Linear Correlation Residuals Normality, (Pd(X2;d))All 532 -experiments__53 II -IIAll exceptthose withGadolinium, Cadmium andLead365REPORT HI-2104790 D-28 Table D.3-10 Bias and Bias Uncertainty as a Function of Independent REPORT HI-2104790 D-29REPORT HI-2104790 D-29 Table D.3-11 Analysis of Fuel Burnup for the MCNP5-1.51 calculations Normality Residuals xperimen of Bias Bias N i Linear Correlation Normality, Description Uncertainty X2  2exp. (Pd(xE;d))
________________
(Pd(X2;d))
All exceptthose withGadolinium, 365 mmCadmium andLeadtFresh U02  207 mIN mm m m_______m
-Fuel--HT m+MOX 158 mExperiments REPORT HI-2 104790 D-30REPORT HI-2104790 D-30 Table D.3-12 Bias and Bias Uncertainty as a Function of Independent Parameter Indepen Bias Independ BiasExperiment dent Calculated Bias Uncerta ent Calculated Bias UncertaiDescription Paramet kF inty aramete kf ntyer, x r, xEALFPu Enrichment HTC + MOXExperiments IIIIREPORT HI-2 104790 D-3 IREPORT HI-2104790 D-31 Table D.3-13 Analysis of the Unborated and Borated Water for the MCNP5-1.51 calculations No. Normality Residuals Eerimen of Bias Bias Linear Correlation Normality, Description Uncertainty22 exp. (Pd(x';d))
(Pd(x ;d))All except thosewithGadolinium, 365 M Cadmium andLeadtAll with Fresh 287 _ _ _ _ _ MWater 287 mAll with 78Borated Water 7EMO "tNote: Critical experiments with Gadolinium, Cadmium and Lead were excluded from all subsequent subsets.REPORT HI-2104790 D-32 Table D.3-14 Bias and Bias Utas a Function ofParameter All with FreshWaterEALFmElmEimEimEimEimEimEimEimEimEimElmElmEimElmEimElmElmElmElmEimE-DensitymElmEimEiU Enrichment M IAll withBorated WaterN/AREPORT HI-2 104790 D-33REPORT HI-2104790 D-33 Experiment Description IndependentParameta1- "'CalculatedBiasBiasUncertaintyIndepend Biasent Calculated Bias UncertaiParamete kyr, xREPORT HI-2 104790 D-34REPORT HI-2104790 D-34 Table D.3-15 Analysis of the Combinations of Fuel Burnup and Unborated/Borated Water for the MCNP5-1.51 calculations No. Bias Normality Residuals Experiment of Bias Uncertain x2 Linear Correlation Normality, exp. ty (Pd(x;d))
(Pd(X ;d))All except thosewithGadolinium, 365 1 1___ 1ma nCadmium andLeadtFresh U02 Fuelwith Fresh 154 -__ mWaterFresh U02 Fuelwith Borated 53 aIWaterHTC + MOXAFuel with Fresh 133 AWater MIEH-TC + MOXU.Fuel with 25 1__ __Borated WatertNote: Criticalsubsets.experiments with Gadolinium, Cadmium and Lead were excluded from all subsequent REPORT HI-2104790 D-35REPORT HI-2104790 D-35 Table D.3-16 Bias and Bias Uncertainty as a Function of Independent Parameter Indepen Bias Independ BiasExperiment dent Calculated ent Calculated Bias UncertaiDescription Paramet kBfia inct Paramete kefter, x r, xntyEALFmIImEimiImIImElmElmIImEimIIFresh U02Fuel withFresh WaterN/AmmmmmmmmmEALFmiImPu Enrichment HTC + MOXFuel withFresh WaterREPORT HI-2104790 D-36 Experiment Description IndependentParametCalculated keffBiasBiasUncertaintyIndependentParameter. xCalculated keffBiasBiasUncertaintymEmmEmmEmmEmmEmRod ODmElmElE lm UImIIHTC + MOXFuel withBorated WaterDensitymElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElEE==mM==MREPORT HI-2104790 D-37 Table D.3-17 Comparison of Key Parameters and Definition of Validated AOAParameter Design Benchmarks Validated Application Fissionable Material 235U, 239Pu, 241Pu 235U, 239Pu, 241Pu 235U, 239Pu, 241PuIsotopic Composition 235u/ut < 5.Owt% 1.57 -5.74% < 5wt%Pu/(U+Pu)
< 20wt% 1.104-20%
< 20wt%Physical Form UO2,MOX U02,MOX UO2, MOXFuel Density (g/cm3) 10.0 -10.7 9.2 -10.4 9.2 -10.7Moderator Material (coolant)
H H HPhysical Form H20 H20 H20Density (g/cm3) around 1.0 g/cm3  around 1.0 g/cm3  around 1.0 g/cm3Reflector Material H H HPhysical Form H20 H20 H20Density (g/cm3) around 1.0 g/cm3  around 1.0 g/cm3  around 1.0 g/cm3Interstitial Reflector MaterialPlate Steel or Lead Steel or Lead Steel or LeadAbsorber MaterialNone, Boron (15 -None, Boron (0 -Soluble None, Boron or 2550 ppm) or 2550 ppm) orGadolinium Gadolinium (48 -Gadolinium (48 to197 ppm) 197 ppm)Rods Boron Pyrex', Vicor BoronSteel or B-AlSeparating MaterialWater, B-SS, Water, B-SS, Boral, Water, B-SS, Boral,Plate Boral or Boroflex, Zircaloy or Boroflex, Zircaloy orCadmium Cadmium CadmiumGeometryLattice type Square Square, Triangle Square, Triangle1.26-1.47 Lattice Pitch (cm) (PWR) 0.968 to 4.318 0.968 to 4.318LaticePith (m) 1.24 -1.88(BWR)ThermalNeutron Energy spectrum Thermal spectrum Thermal spectrumREPORT HI-2104790 D-38 Figure Proprietary Figure D.3-1 Frequency Chart for Calculated keff of the Selected 532 Benchmarks for theMCNP5-1.51 codeFigure Proprietary Figure D.3-2 Frequency Chart for Calculated EALF (eV) of the Selected 532 Benchmarks for theMCNP5-1.51 codeREPORT HI-2104790 D-39 Figure Proprietary Figure D.3-3 MCNP5-1.51 Calculated keg Values for Various Values of the Spectral Index (AllExperiments)
REPORT HI-2104790 D-40 Figure Proprietary Figure D.3-4 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral IndexREPORT HI-2 104790 D-4 1REPORT HI-2104790 D-41 Figure Proprietary Figure D.3-5 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral IndexREPORT HI-2104790 D-42 Figure Proprietary Figure D.3-6 MCNP5-1.51 Calculated keff Values for Various Values of the Pu Enrichment REPORT HI-2 104790 D-43REPORT HI-2104790 D-43 Figure Proprietary Figure D.3-7 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral IndexREPORT H-Ii-2104790 D-44 Figure Proprietary Figure D.3-8 MCNP5-1.51 Calculated keff Values for Various Values of the U Enrichment REPORT HI-2104790 D-45 Figure Proprietary Figure D.3-9 MCNP5-1.51 Calculated keff Values for Various Values of the Fuel DensityREPORT HI-2104790 D-46REPORT HI-2104790 D-46 Figure Proprietary Figure D.3-10 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral IndexREPORT HI-2 104790 D-47REPORT HI-2104790 D-47 Figure Proprietary Figure D.3-11 MCNP5-1.51 Calculated ke, Values for Various Values of the Spectral IndexREPORT HI-2104790 D-48 Figure Proprietary Figure D.3-12 MCNP5-1.51 Calculated keff Values for Various Values of the Pu Enrichment REPORT HI-2 104790 D-49REPORT HI-2104790 D -49 Figure Proprietary Figure D.3-13 MCNP5-1.51 Calculated keff Values for Various Values of the Rod ODREPORT HI-2 104790 D-50REPORT HI-2104790 D-50 Figure Proprietary Figure D.3-14 MCNP5-1.51 Calculated keff Values for Various Values of the Fuel DensityREPORT HI-2 104790 D-5 1REPORT HI-2104790 D-51 ATTACHMENT 5Holtec International Report No. HI-2125245, Revision 4,"Licensing Report for Quad Cities Criticality Analysis for Inserts -Non Proprietary Version" Eu.'HOLTINTERN AT IINECONALHoltec Center, 555 Lincoln Drive West, Marlton, NJ 08053Telephone (856) 797- 0900Fax (856) 797 -0909Licensing Report for Quad Cities Criticality Analysis for Inserts -Non Proprietary VersionFORExelonHoltec Report No: HI-2125245 Holtec Project No: 2127Sponsoring Holtec Division:
HTSReport Class : SAFETY RELATED Summary of Revisions:
Revision 0: Original IssueRevision 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 markedby revision bars.Revision 4: Minor editorial changes to page 4 description of Table 7.1(c) and Table 2.1(c) wasmove up one line. Neither change marked by revision bar. All Revision 3 revision barsremoved.Project No. 2127Report No. H1-2125245 Page i Table of Contents1. INTRODUCTION
.................................................................................................
102. METHODOLOGY
.............................................................................................
112.1 GENERAL, APPROACH
.......................................................................................
...........
II2.2 COMPUTER CODES AND CROSS SECTION LIBRARIES
.................................................................
112 .2 .1 M C N P 5 -1.5 1 .......................................................
............................................................
1 12 .2.1 .1 M C N P 5-1.5 1 V alidation
........................................................................................................................................
2.2.1.1.1 122 .2 .2 C A S M O -4 .............................
...............................................................................
..132.3 ANALYSIS ME-THODS
..................................................................................................................
132.3.1 Design Basis Fuel Assembly
.............................................................................................
132 .3 ,1.1 P eak R eactiv ity ......................................................................................................................................................
142.3,1.1.1 P ak R eactivity and Fuel A ssem b B urnup .............................................................
....................................
142.3.1.1.2.......
......................................................................................
..142.3.1.2 J- ii i i ; .......................................
152.3.1.3 Determination of the Design Basis Fuel Assembly Lattice ...........................................................................
162.3.1.4 Optima2 CASMO-4 Model Simplification Effect ..........................................................................................
162 .3 .1.5 C ore O perating Param eters ...................................................................................................................................
182.3.1.5.1 R eacto r P ow eer U prate ......................................................................................................................................
182.3.1.5.2 Integral R eactivity C ontrol D evices ...........................................................................................
...................
192.3.1.5.3 A xial and Planar Enrichm ent V ariations
....................................................................................................
192.3.1.5.4 Fuel A ssem bly D c-Channeling
..............................................................................................................
192 .3 .1 .6 .....................
.........................................................................
2 02.3.2 Reactivity Effect of Spent Fuel Pool Water Temperature
................................................
202.3.3 Fuel Depletion Calculation Uncertainy
.........................................................................
212.3.4 Fuel and Storage Rack Manufacturing Tolerances
.........................................................
222.3.4.1 Fuel M anufacturing T olerances
............................................................................................................................
222.3.4.2 SEP Storage Rack Manufacturing Tolerances
...............................................................................................
232.3.5 Radial Positioning
...........................................................................................................
232.3.5.1 Fuel Assembly Orientation in the Core ...........................................................................................................
232.3.5.2 Fuel R adial Positioning in the R ack ......................................................................................................................
232.3.5 .3 Inserts R adial Positioning
......................................................................................................................................
252.3.5.4 Fuel Orientation in SFP Rack Cell ..............................................................................................................
252.3.6 Insert. Co po. Measrement...
rtain.. ............
...... I ................
........
262.3.6.18
.... .. .........................
.........................................
.....................
262.3.64 MA 2G .. E ... ...........................................................................................
272.3.6 .2.......
.. ...........
..... ................................................
272. .1 Te peatr ad atrD siyEf cs.............................................
......2 ..6.2 D o p d ss b y -H o i o t l........................I.........................................
........................................
272.3.7 Insert Coupon Measurement Uncertainty
.............................................
......7.. 22.3.8 Maximum kff Calculation for Normal Conditions..............................................
272.4 MARGIN EVALUATIONstortion.........................................................................................................
282.5 FUEL MOVEMENT, INSPECTION AND RECONSTITUTION OPERATIONS
..............................
............
2.6 ACCID NT CONDITION
..........
...................................................................................................
292.6.1 Temperature and WaF er Density Effects .......................................
...............
302.6.2 Droppeda As Fem sbly -Horizontal
........................................................................................
302.6.3 Dropped A ssembly -Vertical into a Storage Cell..............................................
302.6.4 Storage Cell Distortion...........................................................................
312.6.5 Misloaded Fuel Assembly/Missing Insert.......................................................
312.6.6 Mislocated Fuel Assembly
.......................................................................
322.6.6.1 Mislocation of a Fuel Assembly in the Water (jai) between the Racks and Pool Wall ...........................
322.6.6.2 Mislocation of a Fuel Assembly in the Comner between Two Racks................................................
322.6.6.3 Mislocation of a Fuel Assembly between the SFP Rack and the Inspection Platform
................................
322.6.7 Mis-installment of an Insert on Wrong Side of a Cell .....................................................
33Project No. 2127 Report No. HI-2125245 Page 1 2.6.8 Insert M echanical Wear .....................................................................................................
332.6.9 Rack M ovement ...................................................................................................................
332.7 .. .....................................................................
...............................................
332.8 SPENT FUEL, RACK INTERFACES
..............................................................................................
342.9 RECONSTITtr TED FuEl, ASSEMBLIES
.......................................................................................
353. ACCEPTANCE CRITERIA
...........................................................................................................
363.1 AIPPICABLE CODES, STANDARDS AND GUIDANCE'S
............................................................
364. ASSUM PTIONS ..............................................................................................................................
375. INPUT DATA ..................................................................................................................................
385.1 FUEL.. ASSEMBLY SPECIFICATION
.............................................................................................
385.2 REACTOR PARAMEIERS
..... ........................
..............................
385.3 SPEN'T FUEL POOL PARAMETERS
............................................................................................
385.4 STORAGE RACK SPECIFICATION
..............................................................................................
395.4.1 M aterial Compositions
...................................................................................................
396. CO M PUTER CO DES .........................................................................................................
.. 407. ANALYSIS
.......................................................................................................................................
417.1 DESIGN BASIS AND UNCERTAINTY EVALUATIONS
.................................................................
417.1.1 .........
.. 417.1.2 Determination of the Design Basis Fuel Assembly Lattice ..............................................
417.1.2.1 Fuel Assembly De-Channeling
.............................................................................................................................
417.1.3 Optima2 CASM O-4 M odel Simplification Efl&ct ............................................................
417.1.4 Core Operating Parameters
............................................................................................
427.1.4.1 R eactor Pow er U prate ............................................................................................................................................
427.1.5 Water Temperature and Density Effect ............................................................................
427.1.6 Depletion Uncertainty
......................................................................................................
427.1.7 Fuel and Rack M anzjfacturing Tolerances
.....................................................................
437.1.7.1 Fuel Assembly Tolerances
....................................................................................................................................
437.1.7.2 SFP Rack Tolerances
............................................................................................................................................
437.1.8 Radial Positioning
...........................................................................................................
437,1.8.1 Fuel Assembly Radial Positioning in SFP Rack ............................................................................................
437.1.8.2 Fuel Orientation in SFP Rack ................................................................................................................................
437.1.9 Fu.. .............
...............
................................................
447.1.9 .1 .........................................................................................................
...447.1.10 ; 447.2 MAXIMUM KFF CALCULATIONS FOR NORMAL CONDITIONS
..................................................
447.3 M ARGIN EVALUATION
................................................................................................................
447.4 ABNORMAL AND ACCIDENT CONDITIONS
..............................................................................
457.5 MAXIMUM K3:F,: CALCUi ATIONS FOR ABNORMAL AND ACCIDENT CONDITIONS
......................
457.6 ........................................................................................................................
457.7 SPENT FUEL RACK INTERFACES
.............................................................................................
458. CONCLUSION
................................................................................................................................
479. REFERENCES
................................................................................................................................
48Project No. 2127 Report No. HI-2125245 Page 2 Supplement 1: Additional Calculations to Support the Revised NETCO-SNAP-IN Rack InsertD esign ..............................................................................................
.... Si-IProJect No. 2127Report No. F11-2125245 Page 3 List of TablesTableDescription Table 7.2(a)Table 7.2(b)Table 7.3Table 7.4Table 7.5Table 7.6(a)Table 7.7Table 7.8Page505152535455565758596061626364656667687071727374Results of the MCNP5-1.51 Calculations for SVEA-96 Optima2 Q122LatticesResults of the MCNP5-1.51 Calculations for GEI 4 Lattice Type 5Results of the MCNP5-1.51 Calculations for Design Basis andSimplified Model of SVEA-96 Optima2 Q122 Lattice Type 146Results of the MCNP5-1.51 Calculations for Core Operating Parameters Results of the MCNP5-1.51 Calculations for the Effect of WaterTemperature and DensityResults of the MCNP5-1.51 Calculations for the Depletion Uncertainty Results of the MCNP5-1.51 Calculations for Fuel Tolerances Results of the MCNP5-1.51 Calculations for Rack Tolerances 757677ProJect No. 2127Report No. 1-11-2125245 Page 4 TableTable 7.9(a)Table 7.9(b)Table 7.12(a)Table 7.12(b)Table 7.12(c)Table 7.13(b)Description Results of the MCNP5-1.51 Calculations for Fuel Radial Positioning inSFP RacksResults of the MCNP5-1.51 Calculations for Fuel Orientation in SFPRAAk--Page78798081Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluatethe Effect of Nominal Values Instead of Using Minimum B4C Loadingand Minimum Insert Thickness on Reactivity Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluatethe Effect of the Actual Optima2 Q122 Fuel AssemblyMargin Evaluation Summary of the Margin Evaluation Results of the MCNP5-1.51 Calculations for the Empty Storage RackCell without InsertTable 7.16Table 7.17Results of the MCNP5-i.51 Calculations for Axially Infinite Optima2Q122 LatticesResults of the MCNP5-1.51 Calculations for SFR Interface 828384858687888990lmmmmmmmProject No. 2127Report No. HI-2125245 Page 5 TableDescrintion Table SI-ITable S 1-2Table S 1-3Table S 1-4Fuel Rack Insert Revised Dimensions Results of the MCNP5 Calculations for Revised Rack Tolerances Results of the MCNP5-1.51 Calculations for Revised Fuel RadialPositioning in SFP RacksResults of the MCNP5-1.51 Calculations for Revised Fuel Orientation in SFP RacksPageUS1-5SI-6SI-7S1-8S1-9SIl-10Project No. 2127Report No. 1-11-2125245 Page 6 List of FiguresDescrintion FigurePage9293949596979899100101102103104Project No. 2127Report No. 1H1-2125245 Page 7 FigureDescription Project No. 2127Page105106107108109110III112113114115116117118.... -24, Report No. 1-11-2125245 Page 8 FigureDescription PageSI-I1Project No. 2127Report No. HI-2125245 Page 9
: 1. INTRODUCTION This report documents the criticality safety evaluation for the storage of spent BWR fuel in theUnit I and Unit 2 spent fuel pools (SFPs) at Quad Cities Station operated by Exelon. The Unit Iand Unit 2 SFP racks are identical and are designed to accommodate BWR fuel. Currently, theSFPI racks credit BORAFLEX for reactivity control.
This new analysis will not credit theBORAFLEX but will instead credit new NETCO-SNAP-IN rack inserts, which are new toQuad Cities but not new relative to their use for spent fuel pool reactivity control.
This analysiswill demonstrate that with credit for the inserts the effective neutron multiplication factor (kerf) inthe 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 toassure that under all credible abnormal and accident conditions, the reactivity will not exceed theregulatory limit.Criticality control in the SFP, as credited in this analysis, relies on the following:
* Fixed neutron absorbers o NETCO-SNAP-1N0 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.
Project No. 2127Report No. 1.11-2125245 Page 10
: 2. METHODOLOGY 2.1 General ApproachThe analysis is performed consistent with regulatory requirements and guidance.
Thecalculations 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 NationalLaboratory
[1]. MCNP5-1.51 calculations use continuous energy cross-section data based onENDF/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 theanalysis 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 totalnumber 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 beforeaveraging, and a minimum of 150 cycles that are accumulated, The initial source is specified asnniform cnver the fineled re.oinnq (t-enmhIie0 I2.2.1.1 MCNP5-1.51 Validation ProJect No. 2127Report No. 1-I1-2125245 Page I11 Project No. 2127Report No. HI-2125245 Page 12 U2.2.2 CASMO-4Fuel 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 reactoranalysis (depletion) when providing reactivity data for specific 3D simulator codes. CASMO-4 isa two-dimensional multigroup transport theory code based on the Method of Characteristics andit 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 thefuel (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 theCASMO-4 sensitivity studies are not used as input into the kcff calculations.
: However, the codeauthors have validated CASMO-4 against MCNP and various critical experiments
[5].The version of the CASMO-4 code used in this application has a built-in limitation in a numberof isotopes that may be extracted for specific pins. Therefore, two independent CASMO-4depletion calculations were performed to separately extract the actinides and fission products.
The extracted isotopes were fuirther combined and used in MCNP5-1.51 calculations.
2.3 Analysis Methods2.3.1 Design Basis Fuel AssemblyThere are various fuel designs stored in the Quad Cities SFP. For the purpose of this analysis, thereactivity of each design is evaluated and the most reactive fuel bundle lattice is determined foruse as the design basis fuel assembly to determine keff at the 95/95 level. This approach followsthe guidance in [2] and [6], and is further described below.Project No. 2127Report No. HI-2125245 Page 1 3 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, ascore depletion occurs, the Gd begins to burnout until it is essentially fully depleted.
As the Gddepletes 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 composedof various axial lattices (including blankets) that can have different axial lengths, uraniumloadings (also mixed oxide loading, for MOX fuel), fuel pin arrangements including partial orpart-length rods, Gd pin locations and loading, etc. These various lattice components can alleffect at what burnup the peak reactivity occurs and the magnitude of the peak reactivity.
TheMax ial lattices w ith in a sin g le fu el asse bl ca th r f e a l h ve d f r nt p k r a ti ty2.3.1.1.1 Peak Reactivity and Fuel Assembly BurnupTypically, a spent fuel assembly is characterized by its assembly average burnup (over all latticesor nodes). In this analysis methodology the fuel assembly average burnup is of no concern and isnot credited for reactivity control.
Rather, the methodology credits the residual Gd and otherdepletion isotopic compositions at the fuel assembly peak reactivity (most reactive lattice peakreactivity).
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) areperformed for every lattice of the SVEA-96 Optima2 fuel assembly (as will be seen in Section 7,this is the fuel assembly with the design basis lattice) over a burnup range to determine theburnup 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 burnupor fuel assembly burnup profile is not applicable because the analysis already considers eachlattice at its most reactive composition, independent of the fuel assembly average burnup.2.3.1.1.2 Project No. 2127Report No. HI-2125245 Page 14 2.3.1.2ProJect, No. 2127Report No. I-11-2125245 Page 15 2.3.1.3 Determination of the Design Basis Fuel Assembly Lattice2.3.1.4 Optima2 CASMO-4 Model Simplification EffectAs previously discussed in Section 2.3.1.2, various fuel designs were provided.
Of these fueldesigns, the SVEA-96 Optima2 designs were specified to be bounding.
The Optima2 model inCASMO-4 is described as the SVEA-96 model provided in the CASMO-4 manual [4]. ThisCASMO-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 theCASMO-4 model simplifications inherent in the code. The CASMO-4 model geometry of theSVEA-96 Optimna2 fuel differs from the SVEA-96 Optirna2 fuel as follows:Project No. 2127Report No. HI-2125245 Page 16 With respect to the fuel assembly geometry models, the amount of zirconium (and therefore theamount of water) in the CASMO-4 model of the SVEA-96 Optima2 fuel is reasonably similar tothat 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 onreactivity, an applicable set of code-to-code comparisons is performed.
The following cases areevaluated.
For the purpose of showing that the two codes calculate an equivalent reactivity the following comparisons are made:Project No. 2127Report No. 1-11-2125245 Page 17
" Case 2.3.1.4.1 is compared to Case 2.3.1.4.2 at 0 GWD/MTU to show that the two codescalculate 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 thereactivity difference due to the effect of the spent fuel. The two codes use different crosssection library versions and calculation sequences.
The main calculation sequencedifference between the two codes is that CASMO-4 uses a thermal expansion of spentfuel pellet which effects the fuel density [4]. The actual density is conservatively used inMCNP5-1.51.
The results are expected to show that the MCNP5-1.51 code isconservative with respect to the CASMO-4 code. Any non-conservative result would betreated as a bias." Case 2.3.1.4.3 is compared to Case 2.3.1.4.2 to show the reactivity difference betweenthe simplified MCNP5-1.51 model and the design basis model that is slightly modified tobe similar to the CASMO-4 insert orientation.
This case is expected to show that thedesign basis model with respect to the fuel pin pitch (and subsequent sub-bundle pitch) isconservative.
This is expected to be conservative because the design basis model fuelcompositions are taken from the average fuel pin pitch CASMO-4 calculations and usedin the MCNP5-1.51 design basis actual fuel pin locations.
Any non-conservative resultwould be treated as a bias.Case 2.3.1.4.3 is compared to the result of the actual design basis results (similar to Case2.3.1.4.3 but with the bounding insert orientation) to show that the design basis model isconservative.
2.3.1.5 Core Operating Parameters As previously discussed, CASMO-4 is used to perform depletion calculations to determine thespent 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 onlBWR 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 Section2.3.1.5.2, 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 theirworst case.2.3.1.5.1 Reactor Power UprateTo determine the effect of the power uprate on the reactivity ofassemblies in the SFP racks, the following evaluations are performed.
ProJect No. 2127Report No. HI-2125245 Page !18 2.3.1.5.2 Integral Reactivity Control DevicesThe 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 latticewhere all rods contain fuel and no Gd. As discussed in Section 2.3.1.1, the Gd in the fuelassembly holds down the fresh fuel assembly reactivity and then, as core depletion occurs, theGd begins to burnout until it is essentially fully depleted.
As the Gd depletes the reactivity of thefuel assembly increases until it reaches a peak. This peak reactivity is the fuel assembly's mostreactive condition, which is used for design basis condition.
Note that integrated absorbers do notchange 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 inSection 7) cannot be de-channeled for storage in the SFP because of its specific design.However, GE14 (the most second reactive fuel assembly, as will be shown in Section 7) may bede-channeled.
Studies are performed to evaluate the effect of storage of GE14 without the Zrchannel 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.
TheMCNP5-1.51 model used herein is a 2x2 array with the cell centered fuel assembly thatincludes 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 GEl4 fuel assembly lattice 5 (themost 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 assembliesare 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 fuelassemblies 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 fuelassemblies are eccentric away from the corner where the insert wings connect, as shownin Figure 2.13(d).Project No. 2127 Report No. 1-11-2125245 Page 19 2.3.1.62.3.2 Reactivity Effect of Spent Fuel Pool Water Temperature The Quad Cities Station SFP has a normal pool water temperature operating range below 1 50 'F.For the nominal condition, the criticality analyses are to be performed at the most reactivetemperature and density [2]. Also, there are temperature-dependent cross section effects inMCNP5-1.51 that need to be considered.
In general, both density and cross section effects maynot have the same reactivity effect for all storage rack scenarios, since configurations with strongneutron absorbers typically show a higher reactivity at lower water temperature, whileconfigurations without such neutron absorbers typically show a higher reactivity at a higherwater temperature.
For the SFP racks which credit inserts, the most reactive SFP watertemperature 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 alsoavailable at other temperatures;
: however, not usually at the desired temperature for SFPcriticality analysis.
MCNP5-1.51 has the ability to automatically adjust the cross sections to thespecified temperature when using the TMP card. Furthermore, MCNP5-1.51 has the ability tomake a molecular energy adjustment for select materials (such as water) by using the S(a,p) card.The S(a,13) card is provided for certain fixed temperatures which are not always applicable toSFP 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(C,3) card fortemperatures as it does for the TMP card discussed above. Therefore, additional studies areperformed 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(U,[) calculations are performed for both upper and lower bounding S(a,f3) values, if needed.The studies mentioned above are performed for the following eases for the single cellMCNP5-1.51 SFP model (with periodic boundary conditions through the centerline of thesurrounding water 2):Project No. 2127Report No. 111-2125245 Page 20
" Case 2.3.2.1 (reference case): Temperature of 39.2 IF (277.15 K) and a density of 1.0g/cc are used to determine the reactivity at the low end of the temperature range. TheS(a,o3) card corresponds to a temperature of 68.81 IF (293.6 K).* Case 2.3.2.2:
Temperature of 150 IF (338.71 K) and a corresponding density of 0.98026g/cc are used to determine the reactivity at the high end of the temperature range. TheS(a,j3) card corresponds to a temperature of 68.81 &deg;F (293.6 K).* Case 2.3.2.3:
Temperature of 150 IF and a corresponding density of 0.98026 g/cc. TheS(@,p) card corresponds to a temperature of 170.33 IF (350 K).The bounding water temperature and density (the temperature and its corresponding densitywhich 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 use the same MCNP5-l.51 models used in the design basis calculation.M 2.3.3 Fuel Depletion Calculation Uncertainty To account for the uncertainty of the number densities in the depletion calculations erformed inThe depletion uncertainty is applied by multiplying it with the reactivity difference (at95%/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 Gd2O3).Calculations are performed for the single cell model of design basis fuel assembly.
The uncertainty is determined by the following:
Uncertaintytopic
= [ (kcale-2 -kcale-1) + 2 * ((J .12 + Gcalc.22)
]
* 0.05withkeac-1 = kcalc with spent fuel= kal. with fresh fuelGcal-l =Standard deviation of 0~ceal-2=
Standard deviation of kcale-2'The result of the MCNP5-1.51 calculation for the fuel depletion calculation uncertainty isstatistically combined with other uncertainties to determine keff.Project No. 2127Report No. 2125245 Page 21 2.3.4 Fuel and Storage Rack Manufacturing Tolerances In order to determine the kenr 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 arecombined statistically
[2]. The evaluations use the same MCNP5-1.51 models used in the designbasis calculation.
2.3.4.1 Fuel Manufacturing Tolerances The BWR fuel tolerances for Optima2 Q122 fuel (which is the most reactive fuel designevaluated herein) are presented in Table 5.1(a). Fuel tolerance calculations are petformed usingthe design basis fuel assembly
: lattice, and therefore only the tolerances applicable to that latticeare 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 rackcalculations.
Pin specific compositions are used. The MCNP5-1.51 tolerance calculation iscompared to the MCNP5-1.51 reference case (nominal parameter values) at the 95% probability at a 95% confidence level using the following equation:
delta-kale
= (ka&2 -kclcl) +/- 2
* q (a12 + a22)The following fuel tolerances are considered in this analysis:
" Fuel enrichment
" Gd loading* Fuel pellet density (U02 and U02+Gd2O3 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 4o Water wing canal inner widtho Channel outer square widtho Channel comer inner radiuso Central water canal inner square width* Combination of 4o channel wall thickness 3 For fuel sub-bundle pitch uncertainty calculation, the fuel hardware (channel, central water channel andwater wings) is fixed. The fuel lattices are moved only.4 Conservatively, the various tolerances are considered together.
The tolerance limits that result in anincrease 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 togetherin another set of uncertainty calculations.
Pr(&#xfd;ject No. 2127Report No. H--2125245 Page 22 o Water cross wall thickness The maximum positive reactivity effect of the MCNP5-1.51 calculations for each tolerance isstatistically combined with the other tolerance
: results, and this result is then statistically combined with other uncertainties when determining the kff value.2.3.4.2 SFP Storage Rack Manufacturing Tolerances The SFP rack tolerances are presented in Tables 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 thetolerance is applied for each case. The MCNP5-1.51 tolerance calculation is compared to theMCNIP5-1.51 reference case with a 95% probability at a 95% confidence level using thefollowing equation:
delta-kcac
= (kc.Ic2 -kcalcl) +/- 2 * (G12 + C22)The following SFP rack manufacturing tolerances are considered in this analysis:
" Storage cells:o Cell ID and cell pitcho Cell wall thickness
* Rack inserts (poison)o WidthThe maximum positive reactivity effect of the MCNP5-1.51 calculations for each tolerance isstatistically combined with the other tolerance
: results, and this result is then statistically combined with other uncertainties when determining the keff value.The evaluations use the same MCNP5-1.51 models used in the design basis calculation.
Theisotopic 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 biasinstead of uncertainty, for conservatism and simplification.
2.3.5 Radial Positioning 2.3.5.1 Fuel Assembly Orientation in the CoreThe fuel assembly orientation in the core with respect to its control blade does not change andtherefore the design basis calculations consider the only possible configuration.
2.3.5.2 Fuel Radial Positioning in the RackThe BWR fuel that is loaded in the SFP racks may not rest exactly in the center of the storagecell. Evaluations are performed to determine the most limiting fuel radial location.
The following eccentric fuel positioning cases are analyzed:
Project No. 2127Report No. HI-2125245 Page 23
" Case 2.3.5.2.1:
This is the reference for Case 2.3.5.2.2 through Case 2.3.5.2.5.
TheMCNP5-1.51 model used herein is a 2x2 array which is the same as the primary singlebundle MCNP5-1.51 model used elsewhere in this analysis.
In both models the fuel iscentered 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, asshown in Figure 2.7(b)." Case 2.3.5.2.3:
Every fuel assembly is positioned toward the corner where the insertwings 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 insertwings 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 2x2array, 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.
TheMCNP5-1.51 model used herein is an 8x8 array which is the same as the primary singlebundle MCNP5-1.51 model used elsewhere in this analysis.
In both models the fuel iscentered in the rack cell.* Case 2.3.5.2.7:
Every fuel assembly is positioned toward the center, for the 8x8 array, asshown in Figure 2.8.* Case 2.3.5.2.8:
Every fuel assembly is positioned toward the corner where the insertwings connect, for the 8x8 array." Case 2.3.5.2.9:
Every fuel assembly is positioned away from the corner where the insertwings connect, for the 8x8 array." Case 2.3.5.2.10:
Every fuel assembly is centered between insert and cell walls, for the8x8 array.* Case 2.3.5.2.11:
This is the reference for Case 2.3.5.2.12.
The MCNP5-1.51 model usedherein 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 thesingle rack cell.The maximum positive reactivity effect of the MCNP5-1.51 calculations for the fuel radialpositioning is added as the bias and the corresponding 95/95 uncertainty is statistically combinedwith other uncertainties to determine kr.fProject No. 2127Report No. HI-2 125245Page 24 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 isotopiccompositions 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 inthe positioning for inserts is evaluated.
The water gap between rack wall and insert is notassumed, since it may provide a small flux trap effect. Nevertheless, the orientation of fuelassembly with respect to position of insert is considered in Section 2.3.5.4.2.3.5.4 Fuel Orientation in SFP Rack CellAs 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 thedepletion in the core. As a result, the fuel depletion is not uniform (more discussion is providedin Section 2.3.1.1.2) and one fuel assembly corner may be more reactive than other corners andtherefore 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 themost 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 shownin Figure 2.9(a). The MCNP5.1.51 models of the other four cases are the same as that of thereference 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 SFPcells 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 asthose of the design basis fuel assembly.
2.3.6Project No. 2127Report No. HI-2125245 Page 25 2.3.6.12.3.6.1.1 Prqject No. 2127Report No. H1-2125245 Page 26 2.3.6.1.2
-2.3.6.22.3.7 Insert Coupon Measurement Uncertainty There is a measurement uncertainty associated with the B-! 0 content in the poison test coupons.
Inthis analysis, the minimum B-10 loading and the minimum insert thickness are conservatively usedfor criticality calculations.
Therefore, the coupon measurement uncertainty is not evaluated furtherin the analysis.
2.3.8 Maximum klf Calculation for Normal Conditions The calculation of the maximum kff of the SFP storage racks fully loaded with design basis fuelassemblies at their maximum reactivity is determined by adding all uncertainties and biases to thecalculated reactivity.
Note that the insert thickness and its B-I 0 loading are taken at their worst casevalues.Project No. 2127Report No. 1-11-2125245 Page 27 kefr is determined by the following equation:
keff = k&#xa2;1e + uncertainty
+ biaswhere uncertainty includes:
and the bias includesNote that each uncertainty is statistically combined with other uncertainties, while biases areadded together in order to determine keff.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 B4C loading* Minimum insert thickness
* Minimum amount ofB-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 andCase 2.4.3.Project No. 2127Report No. 1H1-2125245 Page 28
* Case 2.4.2: This case is the same as Case 2.4.1, except the nominal insert B4C loading,nominal insert thickness and nominal amount of B-I10 in boron are used." Case 2.4.3: This case is the same as Case 2.4.1, except the model includes each Optima2Q122 fuel lattice in the appropriate axial position.
: However, the top and bottom blanketswere conservatively replaced by adjacent fuel lattices.
The peak reactivity burnup foreach individual Optima2 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 peakreactivity).
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 ofthe design basis fuel assembly.
2.5 Fuel Movement, Inspection 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 movementProject No. 2127Report No. 1-1-2125245 Page 29 Those are briefly discussed in the following sections.
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."
Thisprinciple precludes the necessity of considering the simultaneous occurrence of multiple accidentconditions.
The kfl" calculations perfomaed 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 from the designbasis cas.2.6.1 Temperature and Water Density EffectsThe SFP water temperature accident conditions for consideration are the increase in SFP watertemperature above the maximum SFP operating temperature of 150 'F. The decrease in temperature was already considered for 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 isevaluated:
Case 2.6.1: This case uses a temperature of 255 'F (397.04 K) and a density of 0.84591g/cc. The S(a,3) card corresponds to a temperature of 260.33 'F (400 K). In this model, itis assumed that the water modeled includes 10% void. Void is modeled as 10% decreasein 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 asinserts, show a higher reactivity at a lower water temperature.
The case evaluated above isperformed 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 assemblywill 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 from the fuel assemblies in the rack. This accidentis also bounded by the mislocated case, where the mislocated assembly is closer to the assemblyin the racks. Therefore, the horizontally dropped fuel assembly is not evaluated further in thereport.2.6.3 Dropped Assembly
-Vertical into a Storage CellIt is also possible to vertically drop an assembly into a location that might be occupied by anotherassembly 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.Project No. 2127Report No. 1-11-2125245 Page 30 These deformations could potentially increase reactivity.
: However, the reactivity increase would besmall 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 verticaldrop is therefore bounded by this misload accident and no separate calculation is performed for thisdrop accident.
2.6.4 Storage Cell Distortion A storage cell distortion or altered geometry as a result of fuel handling equipment uplift forces ispossible.
: 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 theinsert 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 isperformed for the storage cell distortion accident.
As a result of significant distortion, the storage cell for whatever reason may not be able to containthe insert and also it will be therefore unacceptable for storage of a fuel assembly.
This condition isbounded by the 'misloaded fuel assembly/missing insert' accident.
However to show that it isacceptable 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 8x8array, is evaluated.
Two cases with a cell centered and eccentric position of the fuel assemblies areanalyzed.
2.6.5 Misloaded Fuel Assembly/Missing InsertThe 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 aninsert.However, there are a few 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 fuelassembly is misloaded into a cell with a missing insert. To evaluate the effect, the following casesare evaluated:
" Case 2.6.5.1:
The MCNP5-1.51 model includes an 8x8 array. One cell near the center of therack 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 fuelassemblies are also eccentric toward the misloaded fuel assembly.
The periodic boundaryconditions 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 itscorresponding water density and S(a,3) card. These temperature and density are boundingfor 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 fuelassemblies centered in the rack cells. See Figure 2.10(b).Project No. 2127Report No. I-1-2125245 Page 31 2.6.6 Mislocated Fuel AssemblyThe Quad Cities SFP layout was reviewed to determine the possible worst case locations for amislocated 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 WallA fuel assembly may be mislocated in the water gap between the racks and the pool wall. Due to theneutron leakage to the outside the storage rack area, the effect of this mislocation is bounded by thatof'mislocation of a fuel assembly between the SFP rack and the inspection platform'
: accident, asdiscussed 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 RacksThere are some places in the SFP, but outside of the racks, where the mislocated fuel assembly maybe in the corner between two racks (thus the mislocated fuel assembly would be adjacent to the fuelassemblies in racks from two sides). To evaluate the effect of the mislocation of a fuel assembly inthe 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 SFP rack cells. Themisplaced fuel assembly is in the corner between two racks. The fuel assemblies in the rackare eccentric toward the mislocated fuel assembly.
The misplaced fuel assembly is placed asclose to the racks as possible.
All fuel assemblies in the model are the design basis fuelassembly.
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 fuelassemblies are centered.
See Figures 2.1 1(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 thetemperature 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 thetemperature 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 PlatformAs discussed in Section 2.5, the fuel handling/inspection/reconstitution platform may have onefuel assembly in it at a time. There is a possibility that a fuel assembly is mislocated between theProject No. 2127Report No. HI-2125245 Page 32 SFP racks and the fuel assembly in the platform.
To evaluate the effect of the mislocation of a fuelassembly between the SFP Rack and the Inspection
: Platform, the following cases are evaluated:
" 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 inthe platform is lined up with the mislocated fuel assembly.
The fuel assemblies in the rackare eccentric toward the mislocated fuel assembly.
The misplaced fuel assembly is placed asclose 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 doesnot have any fuel has at least 12 inches of water. Figure 2.12(a) shows the MCNP5-1.51 model used for 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 fuelassemblies 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 thetemperature of the model is set to the maximum (150 TF)." Case 2.6.6.3.4:
The MCNP5-1.51 model is the same as Case 2.6.6.3.2, except thetemperature of the model is set to the maximum (150 'F).2.6.7 Mis-installment of an Insert on Wrong Side of a CellThere is a small possibility that an insert is installed on wrong sides of the cell. In this case, theremay not be a poison between a fuel assembly placed in that cell and a fuel assembly in anadjacent cell. However, the effect of this mis-installment is bounded by that of 'misloaded fuelassembly/missing insert' accident that does not include the insert in one rack cell, as discussed inSection 2.6.5. No separate calculation is performed for this accident.
2.6.8 Insert Mechanical WearHanding accidents and other environmental damage may cause scratches and local wear ofinserts.
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 MovementIn the event of seismic activity, there is a hypothetical possibility that the storage rack arraysmay move and come closer to each other. Since there is no water gap modeled between cells of astorage rack, the reactivity of the rack movement case is bounded by the reactivity of the designbasis calculation.
2.7Project No. 2127Report No. HI-2125245 Page 33 2.8 Spent Fuel Rack Interfaces The spent fuel pool includes a single type of Region I spent fuel racks, which are loaded with theneutron absorbing inserts in every storage cell as well as a uniform fuel assembly loading pattern.Therefore, any possible water gaps and interfaces between the racks are bounded by the infinitearray used in the design basis calculations.
: However, since the neutron absorbing inserts are locatedin the same corners 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 ofthe spent fuel pool (correspondingly, north-east corner) has two sides that are not adjacent to theinsert. Due to the neutron leakage on the periphery of the spent fuel pool the reactivity increase isnot 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 allfuel assemblies are shifted to the fuel assembly in the corner, which is discussed above, wereevaluated.
Project No. 2127 Report No. HI-2125245 Page 34 2.9 Reconstituted Fuel Assemblies The SFP contains various reconstituted assemblies which were examined and determined to berelatively old and low reactivity designs.
The reconstitution of these fuel assemblies removed fuelrods and replaced them by either fuel rods that are of the same or less initial enrichment and equalor 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 assemblymore reactive than the bounding lattice.
Therefore, reconstituted assemblies are covered by thedesign basis Optima2 Q122 lattice 146. Future reconstituted assemblies will replace fuel rods withstainless steel rods.Project No. 2127Report No. 111-2125245 Page 35
: 3. ACCEPTANCE CRITERIA3.1 Applicable Codes, Standards and Guidance's Codes, standard, and regulations or pertinent sections thereof that are applicable to theseanalyses 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 Freshand Spent Fuel Storage and Handling, Revision 3 -March 2007.* L. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of FuelStorage 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 andTransportation of LWR Fuel Outside Reactors (withdrawn in 2004).* USNRC, NUREG/CR-6698, Guide for Validation of Nuclear Criticality SafetyCalculational Methodology, January 2001." DSS-ISG-2010-01, Revision 0, Staff Guidance Regarding the Nuclear Criticality SafetyAnalysis for Spent Fuel Pools.Project No. 2127Report No. HI-2125245 Page 36
: 4. ASSUMPTIONS The analyses apply a number of assumptions, either for conservatism or to simplify thecalculation 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 theselected inputs and parameters are in fact 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 becausethis 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 tomoderator ratio (Section 7.1.7.1);
therefore, neglecting the spacer grid places more waterwithin the calculation model. In addition, the inconel springs within the spacer are astronger neutron absorber than water. The active fuel region repeats periodically in thevertical direction.
Therefore, neutron absorption in upper and lower tie plates, fuelplenums, etc. is neglected.
: 3. The neutron absorber length in the rack is more than the active region of the fuel, but it ismodeled 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 andchamfering.
This is acceptable since the amount of fuel modeled is more than the actualamount.5. For the inserts, only the worst case bounding material specifications are used (minimumB-I 0 loading and minimum thickness).
: 6. All models are laterally infinite arrays of the respective configuration, neglecting lateralleakage.
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 claddingconsists of one of several zirconium alloys. This is acceptable since the model neglectsthe 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. The watergaps between the spent fuel racks were conservatively neglected.
Project No. 2127Report No. HI-2125245 Page 37
: 5. INPUT DATA5.1 Fuel Assembly Specification The SFP racks are designed to accommodate the following fuel assembly types used in the QuadCities Unit I and Unit 2, which are presented in a chronologic order along with the initialmaximum 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 Figure 5.4.The fuel rod, cladding and channel are explicitly modeled.,
: Axially, the design basis MCNP modelconsiders the bounding lattice along the entire length and uses water reflectors at the top andbottom. The MCNP model for the margin evaluation calculations discussed in Section 2.4 differfrom the design basis model in that the active length specifically considers each actual lattice inits actual axial configuration (i.e. all the lattices from the Q122 bundle are modeled in the sameMCNP mrNAA1\5.2 Reactor Parameters The reactor core parameters are provided in Table 5.2(a). The reactor control blade data areprovided in Table 5.2(b). The reactor control parameters used in CASMO-4 screening and designbasis calculations are provided in Table 5.2(c).5.3 Spent Fuel Pool Parameters The spent fuel pool parameters are provided in Table 5.2(a).
5.4 Storage Rack Specification The storage rack specifications that are used in the criticality analysis are summarized in Tables5.3(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.
model consists of a single rack cell with periodic boundary conditions through the centerline ofthe water (BORAFLEX replacement),
thus simulating an infinite array of storage cells. Thestorage 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 (theminimum 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 wallthickness of the boundary is thicker than the inner wall thickness.
The SF1P model uses the innercell wall thickness only, as given in Table 5.3(a), because it decreases the amount of steel in themodel, which acts a neutron absorber.
The MCNI15-1.51 SFP rack cell model is shown in Figure 5.4.5.4.1 Material Compositions The MCNP5-1.51 material specification is provided in Table 5.4(a) for non-fuel materials, and inTable 5.4(b) for fuel materials.
Project No. 2127Report No. 1-I1-2125245 Page 39
: 6. COMPUTER CODESThe following computer codes were used in this analysis.
* MCNP5-1.51
[1] is a three-dimensional continuous energy Monte Carlo code developed at Los Alamos National Laboratory.
This code offers the capability of performing fullthree dimensional calculations for the loaded storage racks. MCNP5-1.51 was run on thePCs at Holtec." CASMO-4 [4] is a two-dimensional multigroup transport theory code developed byStudsvik.
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. 2127Report No. f11-2125245 Page 40
: 7. ANALYSIS7.1 Design Basis and Uncertainty Evaluations 7.1.17.1.2 Determination of the Design Basis Fuel Assembly LatticeAs discussed in Section 2.3.1.3, MCNP5-1.51 calculations were performed to determine thedesign basis lattice.
The results for the SVEA-96 Optima2 Q122 lbel assembly are presented inTable 7.2(a) .The results for the GE 14 lattice type 5 are presented in Table 7.2(b), along with thebounding result of the SVEA-96 Optima2 Q122. As can be seen, the SVEA-96 Optima2 Q122lattice type 146 is bounding, and thus it is selected as the design basis lattice.
The CASMO-4model of the SVEA-96 Optima2 bundle Q122 lattice 146 used for depletion calculations is shownin Figure 5.1.7.1.2.1 Fuel Assembly De-Channeling As discussed in Section 2.3.1.5.4, the reactivity of the second most reactive assembly with no Zrchannel at various radial positioning was evaluated.
The results are provided in Table 7.2(b) andcompared 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 offuel assemblies without channels is acceptable.
7.1.3 Optima2 CASMO-4 Model Simplification EffectAs 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 inTable 7.3. As can be seen, the reactivity of the simplified model is comparable to that of thecomplete model of SVEA-96 Optima2 Q122 lattice 146 (essentially within the 95/95 uncertainty between the two calculations).
Therefore, the results show that the CASMO-4 model simplification Project No. 2127Report No. HI-2125245 Page 41 does not have a significant impact on the analysis conclusions regarding the determination of thedesign 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 wereevaluated.
The results are provided in Table 7.4. The results show that the two dominant coreoperating 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 UprateAs discussed in Section 2.3.1.5.1, the effect of the MUR on the reactivity was evaluated.
The resultsare 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 EffectAs discussed in Section 2.3.2, the effects of water temperature, and the corresponding waterdensity and temperature adjustments (S(a,3))
were evaluated for SFP racks. The results of thesecalculations are presented in Table 7.5.The results of the SFP temperature and density calculations show that as expected (for poisonedracks) the most reactive water temperature and density for the SFP racks is a temperature of39.2 'F 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 inTable 7.11 and Table 7.14.Project No. 2127Report No. 111-2125245 Page 42 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 wasdetermined.
The results of these calculations are presented in Table 7.7. The maximum positivedelta-k value for each tolerance is statistically combined.
The maximum statistical combination of fuel assembly tolerances is used to determine keff inTable 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 theSFP racks with inserts was determined.
The results of these calculations are presented in Table7.8. 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 kerr inTable 7.11 and Table 7.14.7.1.8 Radial Positioning 7.1.8.1 Fuel Assembly Radial Positioning in SFP RackAs discussed in Section 2.3.5.2, twelve fuel assembly radial positioning cases in racks wereevaluated.
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 resultsshow that most cases show a negative reactivity effect, however some delta k, 1 values arepositive.
Therefore, a maximum delta k.,, value is applied as a bias and the correspondent 95/95uncertainty is statistically combined with other uncertainties in Table 7.11 and Table 7.14.7.1.8.2 Fuel Orientation in SFP RackAs discussed in Section 2.3.5.4, five fuel assembly orientation cases in racks were evaluated.
Theresults of these calculations are presented in Table 7.9(b). The result for the reference case is alsoincluded.
The results show that all cases are statistically equivalent and the reactivity effect offuel orientation is negligible.
Nevertheless, a maximum positive delta value is applied as abias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties inTable 7.11 and Table 7.14.Project No. 21.27Report No. HI-2125245 Page 43 7.1.9 Fuel Rod Geometry Change7.1.9.1i ne results are presenteo in i aoe /. 1 v.The maximum 'ke,,ac -cIcrercnce is added as a bias, and the '2 *
+ G3cf"2cn, (95/95uncertainty) is added as an uncertainty to determine kerr in Table 7.11 and Table 7.14.7.1.9.27.1.107.2 Maximum kff Calculations/for Normal Conditions As discussed in Section 2.3.8, the maximum kfrf for normal conditions is calculated.
The results aretabulated in Table 7.11. The results show that the maximum keff for the normal conditions in theSFP 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 forpoison thickness and loading, as well as the actual lattice configuration of the Optima2 Q122 fuelassembly.
The results of calculations are provided in Table 7.12(a) and Table 7.12(b).
As can beseen and is expected, the reactivity of design basis is larger. The use of a minimum B-10 loadingrelative 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. 2127Report No. HI-2125245 Page 44 The summary of the margin evaluation is presented in Table 7.12(c).
The result shows thatquantified margin remains in the analysis to offset potential effects not already considered in themodel.7.4 Abnormal and Accident Conditions As discussed in Section 2.6, the effects of empty storage cell, increased temperature, misloaded fuelassembly/missing insert, and mislocated fuel assembly accidents on reactivity were evaluated.
Theresults 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 increasein reactivity.
For the SFP racks, the effect onl reactivity of the missing insert is the limiting case.Thus, its calculated MCNP5-1.51 k.lc is used for maximum keff calculations for abnormal andaccident conditions, discussed in Section 7.5.The condition with the empty storage cell without insert in the spent fuel rack shows a lowerreactivity than a design basis case, therefore, it is acceptable to have the empty storage cell withoutinsert in the spent fuel pool.7.5 Maximum keff Calculations for Abnormal and Accident Conditions As discussed in Section 2.6, the maximum keff 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 andaccident conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95%confidence level.7.67.7 Spent Fuel Rack Interfaces As discussed in Sections 2.8, the interface between SFRs and pool walls, i.e. effect on reactivity ofthe peripheral fuel assemblies, that have a side non-adjacent to the insert, was evaluated.
The resultsare provided in Table 7.17. As can be seen, this condition will not result in an increase of SFRreactivity.
This result is expected because the infinite array design basis model is an infinite array ofProject No. 2127Report No. HI-2125245 Page 45 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. noadditional cell with an insert).
Therefore, this water gap edge allows for neutron leakage and as thecalculations show result in statistically equivalent results.Project No. 2127Report No. HI-2125245 Page 46
: 8. CONCLUSION The criticality analysis for the storage of BWR assemblies in the Quad Cities SFP racks withNETCO-SNAP-IN 6 inserts has been performed.
The results for the normal condition show thatl~ff is = 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 highestreactivity.
The results for the accident condition show that k',T is M with the storage racksfully loaded with fuel of the highest anticipated reactivity, which is SVEA-96 Optima2, at a temperature corresponding to the highest reactivity.
The maximum calculated reactivity for both normal and accident conditions includes a margin for uncertainty in reactivity calculations with a 95% probability at a 95% confidence level. Reactivity effects of abnormaland accident conditions have been evaluated to assure that under all credible abnormal andaccident conditions, the reactivity will not exceed the regulatory limit of 0.95.Pro.ject No. 2127Report No. HI-2125245 Page 47
: 9. REFERENCES
[1] "MCNP -A General Monte Carlo N-Particle Transport Code, Version 5," Los AlamosNational Laboratory, LA-UR-03-1987, April 24, 2003 (Revised 2/1/2008).
[2] L.I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of FuelStorage 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 HI-2104790 Revision 1.[4] M. Edenius, K. Ekberg, B.H. Forss6n, and D. Knott, "CASMO-4 A Fuel AssemblyBurnup Program User's Manual,"
Studsvik/SOA-95/1; and J. Rhodes, K Smith,"CASMO-4 A Fuel Assembly Burnup Program User's Manual,"
SSP-01/400, Revision 5,Studsvik of America, Inc. and Studsvik Core Analysis AB3 (proprietary).
[5] D. Knott, "CASMO-4 Benchmark Against Critical Experiments,"
SOA-94/13, Studsvikof 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 forSpent Fuel Pools, Revision 0.[7] Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001.[8] HI-2002444, Latest Revision, "Final Safety Analysis Report for the HI-STORM 100Cask System",
USNRC Docket 72-1014.[9] "Sensitivity Studies to Support Criticality Analysis Methodology,"
H1-2104598 Rev. 1,October 2010.[10] "Atlas of Neutron Resonances",
S.F. Mughabghab, 5th Edition, National Nuclear DataCenter, 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 (kif) Predictions, NUREG/CR-7109, April 2012.[13] OECD / NEA Data Bank, Java-based Nuclear Information
: Software, Janis version 3.3.Project No. 2127Report No. HI-2125245 Page 48
[14] EPRI 1003222, "Poolside Examination Results and Assessment, GEl I BWR FuelExposed to 52 to 65 GWd/MTU at the Limerick 1 and 2 Reactors,"
December 2002.Project No. 2127Report No. HI-2125245 Page 49
-] -Iim I m-mI j --* I.m Immm I m-f f-+ 4-n
,,m Ii+/- 4-mmUmt -t I --_________
: a. _______________________________
+/- 4 L ..Project No. 2127Report No. HI-2125245 Page 50 mJ6I NmIUmProject No. 2127Report No. HI-2125245 Page 51 Project No. 2127 Report No. HI-2125245 Page 52
-i -Project No. 2127Report No. .HI-2125245 Page 53 El9......-.-
IIIi-I-I-lIl---li-
=I I I I I I -I I I I is -Project No. 2127Report No. 111-2125245 Page 54 ALL----IrIElIIIII II' I I I Iii1I I 4 II -I IM I -IMIMMlProject No. 2127Report No. 1H1-2125245 Page 55 Project No. 2127Report No. HI-2 125245Page 56 IProject No. 2127Report No. HI-2125245 Page 57 Project No. 2127Report No. HI-2125245 Page 58 Project No. 2127Report No. HI-2125245 Page 59 Project No. 2127Report No. HI-2125245 Page 60 Project No. 2127 Report No. H1-2125245 Page 61 I1 -1Ur357LIZLIZ~IIIIPr(&#xfd;Iect No. 2127Report No. HI-2125245 Page 62 Project No. 2127 Report No. 1.tl-2125245 Page 63 Project No. 2127Report No. H1-2125245 Page 64 I J I3I I IU;JIProject No. 2127Report No. Hf-2125245 Page 65
-~ I. I-mIlV-I-I-I-I.a'IILI::::Project No. 2127Report No. HI-2125245 Page 66
! MillLUIN 011 IN m11 IN 0mi IN m1111, IN 0M",ImmI M--7m- --MMM-m-----Mil 101JM mProject No. 2127Report No. HI-2125245 Page 67 Table 7.2(a)Results of the MCNP5-1.51 Calculations for SVEA-96 Optima2 Q122 LatticesNote 2: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta k,,,, may occur atan exposure which differs from that shown above.Project No. 2127Report No. HI1-2125245 Page 68 Table 7.2(a) Continued BurnupDescription (GWd/mtU) 151617Lattice 149 18(void) 1920211516Lattice 149 17(water) ,t 18192021151617Lattice 150 18192021151617Lattice 151 18192021itaC sigma Maxdelta k,,,,Uncert.(95/95)-0.0207 0.0016-0.0189 0.0016-0.0154 0.0016--0.0111 0.0016Note 2: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta k,01Imay occur at an exposure which differs from that shown above.Project No. 2127Report No. 2125245 Page 69 Table 7.2(b)Results of the MCNP5-1.51 Calculations for GE14 Lattice Type 5Description Burnup k Uncert.(GWd/mtU) kdV sigma delta k (95/95)SVEA-96 Optima2 Q122 15.5 Reference Reference lattice type 146Single GEl4 13 -0.0543 0.0016Single GE14 13.5 -0.0509 0.0015Single GEl4 14 -0.0491 0.0016Single GEI4 14.5 1 -0.0469 0.0015Single GEl4 15 -0.0473 0.0015Single GE14 15.5 -0.0479 0.0015Single GE14 16 -0.0485 0.0015Single GE14 16.5 1___l_-0.0482 0.0015Single GEl4 17 -0.0500 0.00152x2 GE14 -with channel (cell 14.5 Reference Reference centered)
(Case 2.3.1.5.4.1) 14.5 __Rfec Rern2x2 GE 14 -no channel 14.5 -0.0044 0.0016_(Case 2.3.1.5.4.2) 2x2 GE14 -no channel /eccentric 14.5 -0.0173 0.0015center (Case 2.3.1.5.4.3) 2x2 GE1 4 -no channel /eccentric 14.5 -0.0238 0.0015out (Case 2.3.1.5.4.4)
Note 2: The result of the SVEA-96 Optima2 Q122 lattice type 146 is provided as the reference.
Note 3: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kcalmay occur at an exposure which differs from that shown above.Project No. 2127Report No. HI-2125245 Page 70 Table 7.3Results of the MCNP5-1.51 Calculations for Design Basis and Simplified Model of SVEA-96 Optima2 Q122 Lattice Type 146Burnup Code kalc sigmaDescription (GWd/mtU)
Simplified model of SVEA-96Optima2 Q122 lattice 146 15.5 CASMO-4 1(Case 2.3.1.4.1)
Simplified model of SVEA-96Optima2 Q122 lattice 146 15.5 MCNP5-1.51 (Case 2.3.1.4.2)
Model of SVEA-96 Optima2 Q122lattice 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. 2127Report No. HI-2125245 Page 71 Table 7.4Results of the MCNP5-1.51 Calculations for Core Operating Parameters Power Fuel Moderat Void BurnupDescription Density Temp. or Temp. Fraction (GWd/ k,,,, sigma delta Uncert.(W/gU) Blade .K) (&deg;FL ) mtU) kcAc (95/95)Design basis 23.688 Yes 1176 547 0 15.5 l l(reference)
Fuel temperature 23.688 Yes 588 547 0 16decreasing Moderator temperature 23.688 Yes 1176 528.8 0 15.5decreasing Void fraction 23.688 Yes 1176 547 94 22increasing Un-rodded operation 23.688No1176547017-il I/-24.1617 Yes 1276 547 0 15.5 1I 2IIL______ Yes_376_47_0
_5. ____ ____20.1348Yes1176547015.5-lIIl-lNote 1: The burnup calculations for core operating parameters were perfonrmed from 14GWd/mtU to 24 GWd/mtU.
For each core operating parameter, only reactivity of the burnup inthis 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,,,may occur at an exposure which differs from that shown above.Project No. 2127Report No. 1-11-2125245 Page 72 Table 7.5Results of the MCNP5-1.51 Calculations for the Effect of Water Temperature and DensityWater Water Temperature Description Burnup Temp. Density Adjustment, Unrtsigma delta k,, (95/5.(GWd/mtU)
(OF) (g/cc) S(aF) (95/95)(OF)
==Reference:==
lowerbound temperature 15.5 39.2 1 68.81 Reference Ref.(Case 2.3.2.1)Upper boundtemperature fornormal operation, low 15.5 150 0.98026 68.81 -0.0041 0.0015S(a4)(Case 2.3.2.2)Upper boundtemperature fornormal operation, 15.5 150 0.98026 170.33 -0.0066 0.0015high S(a,3)(Case 2.3.2.3)Note 1: The maximum calculation uncertainty (sigma) used tomay occur at an exposure which differs from that shown above.determine the 95/95 delta k.,,Project No. 2127Report No. HI-2125245 Page 73 Table 7.6(a)Results of the MCNP5-1.51 Calculations for the Depletion Uncertainty Depletion Description kcilc sigma Uncertainty (5%)Design basis Reference Fresh fuel, no Gd 0.0064Project No. 2127Report No. 141-2125245 Page 74 m-- I~1 ~ 1~-4 -4 4-= I =Project No. 2127Report No. .11-2125245 Page 75 Table 7.7Results of the MCNP5-1.51 Calculations for Fuel Tolerances Description PeakReactivity Burnup(GWd/mtU) k.1,sigmadelta kcnie(95/95)Max delta kca,,(95/95)Design basis (reference) 15.5Max fuel enrichment 16Min fuel enrichment 15.5Max Gd loading 16Min Gd loading 15.5Max pellet density 16Min pellet density 15.5Max pellet OD 15.5Min pellet OD 16Max clad ID 16Min clad ID 16Max clad OD 15.5Min clad OD 15.5Max sub-bundle pitch 15Min sub-bundle pitch 16.5Max pin pitch 15.5Mill pin pitch 15.5Reference Reference 0.0026* 0.0026-0.0009-0.00130.00380.0038 _____0.00000.00 120.00o 12 0 .0010.00150.00150.00110.00 100 0.00100 .0008 0 .0098-0.00020.00270.00270.00980.0098-0.0089 _____0.01220.0122-0.0086 _____Max combined water wing canalinner width, channel outer squarewidth, channel corner inner radiusand central water canal innersquare width15FI -0.00310.0031Min combined water wing canalinner width, channel outer squarewidth, channel corner inner radiusand central water canal innersouare width15.5-I-0.0011Max combination of channel wallthickness and water cross wall 16 l 0.0008thickness 0.0019Min combination of channel wallthickness and water cross wall 15.5 0.0019thickness Statistical combination of fuel tolerances 0.0171Note 1: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kclcmay occur at an exposure which differs from that shown above.Project No. 2127Report No. HI-2125245 Page 76 Table 7.8Results of the MCNP5-1.51 Calculations for Rack Tolerances Burnup delta kai, Max deltaDescription (GWd/mtU) k,,& sigma (95/95) kcle(95/95) (95/95)Design basis 15.5 Reference Reference (reference)
Max cell ID 15.5 -0.0093 N/AMax cell pitchMax wall thickness 15.5 0.0025Min wall thickness 15.5 _ 0.0008Max insert width 15.5 -0.0005 0.0004Min insert width 15.5 0.0004Statistical combination of rack tolerances 0.0026Project No. 2127Report No. 1-11-2125245 Page 77 Table 7.9(a)Results of the MCNP5-1.51 Calculations for Fuel Radial Positioning in SFP RacksBurnupUnc.
Description BurnU) kualc sigma delta kc.I, (95/95)(GWd/mtU) 95/95)_2x2 reference 15.5 Reference Ref.(Case 2.3.5.2.1) 15.5 _lRenc e2x2 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 15.5 1 --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)
: 15. ___efeenc Ref.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)
........1 .__0 0 5 .18x8 insert/cell center 15.5 0.0016 0.0014(Case 2.3.5.2.10)
.......I x I reference 15.5 Reference Ref(Case 2.3.5.2.11)
IxI insert/cell center 15.5 0.0000 0.0015(Case 2.3.5.2.12)
Project No. 2127Report No. 1-11-2125245 Page 78 Table 7.9(b)Results of the MCNP5-1.51 Calculations for Fuel Orientation in SFP RacksBurnup IUnc.Description Burnup k,,,, sigma delta Iellc (95/95)(GWd/mtU) 9595Reference (Shown in 15.5 Reference Ref.Figure 2.9(a)) ........Rotated fuel assembly(shown in Figure 2.9(b)) 15.5 -0.0008 0.0014Rotated fuel assembly(shown in Figure 2.9(c)) 15.5 -0.0007 0.0014Rotated fuel assembly(shown in Figure 2.9(d)) 15.5 -0.0013 0.0013Rotated fuel assembly(shown in Figure 2.9(e)) , 5 -0.0007 0.0013Project No. 2127Report No. 1-11-2125245 Page 79 S-,. --mWm nm M&#xfd;m m mnmmm mmn UUm -- nProject No. 2127Report No. 1-11-2125245 Page 80 I miI. _________________
I ..."mI- ___________________
Project No. 2127Report No. HI-2125245 Page 81 Table 7.12(a)Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate the Effect of Nominal Values Instead ofUsing Minimum B4C Loading and Minimum Insert Thickness on Reactivity B-10 ArealDescription Burnup Density k,,, sigma delta kIc(GWd/mtU)
(g/cm2)Reference (designbasis) 15.5 0.0116 Reference (Case 2.4.1)Rack with nominalvalues for 134Cloading and insert 15.5 0.0133 1 -0.0103thickness (Case 2.4.2)Project No. 2127Report No. 111-2125245 Page 82 Table 7.12(b)Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate the Effect of theActual Optima2 Q122 Fuel AssemblyDescription lurnu) kenic sigma Max kcille delta k,,,,Desripion (GWd/mtU)
Optima2 Q122Lattice 146 (Designbasis)(Case 2.4.1)15.5= I = I Reference IReference 15Optima2 Q122 155Lattice 147'1616Optima2 Q122 16.5Lattice 1481714Optirna2 Q122 14Lattice 149 14.51514_Optima2 Q122 14.5Lattice 1501514Optima2 Q122 14.5Lattice 151150.88730.8843U, 0 J0.88630.88760.8925Optima2 Q122Fuel Assemblyt(Case 2.4.3)PeakReactivity Burnups(bolded)~1--0.0066t The toplattice.and bottom naturalblankets were conservatively neglected and replaced by adjacentProject No. 2127Report No. HI-2125245 Page 83 Table 7.12(c)Margin Evaluation Summary of the Margin Evaluation Description ValueInsert Composition Margin, from Table 7.12(a) -0.0103Actual Optima2 Fuel Assembly Margin, from -0.0066Table 7.12(b)Calculated Margin -0.0169Prqjject No. 2127Report No. 2125245 Page 84 V'mImU I m I Im" I mm--mmm m --F~ m mm-nU m --P~ m m --m- m --mmmmm0mm. m ImmProject No. 2127Report No. 1-11-2125245 Page 85 Table 7.13(b)Results of the MCNP5-1.51 Calculations for the Empty Storage Rack Cell without InsertDescription Burnup keli sigma delta kcal Uncertainty
_________(GWd/mtU)
(95195)Design basis 15.5 Reference Reference (8x8 array)Empty storagecell (cell 15.5 -0.0041 0.0016centered)
Empty storage 15.5 -0,0081 0.0014cell (eccentric) 15.5,_ I -0.0081 _0.0014Note 1: The design basis fuelcalculations.
assembly (Optima2 Q122 Lattice Type146) is used for theseProject No. 2127Report No. HI-2125245 Page 86
-m~IhhEI~im m11EE1~Ihh~n~~m
~ -r .rnuim m_ __ _mEl-_m_I~.m-m-m_Project No. 2127Report No. HI-2125245 Page 87 Project No. 2127Report No. 1.11-2125245 Page 88 Table 7.16Results of the MCNP5-1.51 Calculations for Axially Infinite Optima2 Q122 LatticesBurnup kcai kc.f Delta-K Uncertainty Description (GWd/mtU)
(reference)
(infinite)
Optima2 Q122Lattice 146 15.5 0.0013 0.0015(Design basis) .....Optima2QI22 15.5 0.0018 0.0015Lattice 147Optima2 QI22 16.5 0.0008 0.0014Lattice 148Optima2QI22 14.5 0.0011 0.0015Lattice 149Optima2 Q122 14.5 0.0027 0.0014Lattice 150Optima2 Q122 14.5 0.0010 0.0015Lattice 151 -F 14.5 0.0010 0.0015Note: The difference between the MCNP models under the "reference" column and the MCNP models under the "infinite" column isdescribed in Section 5.1.Project No. 2127Report No. HI-2125245 Page 89 Table 7.17Results of the MCNP5-1.51 Calculations for SFR Interface Description Burnup k,,, sigma delta k,.,, Uncertainty (GWd/mtU) a(95/95)Design basis 15.5 Reference Reference Full SFP (cell 15.5 -0.0008 0.0016centered) 15.5_1 _0.08 .01Full SFP(eccentric to 15.5 -0.0053 0.0015SFP corner)Prqject No. 2127Report No. 1H11-2125245 Page 90 at*1*Project No. 2127Report No. 2125245 Page 91 Figure Proprietary Project No. 2127Report No. HI-2125245 Page 92 Figure Proprietary Project No. 2127Report No. HI-2 125245Page 93 Figure Proprietary ProJect No. 2127Report No. 1-11-2125245 Page 94 Figure Proprietary Project No. 2127Report No. 1-11-2125245 Page 95 Figure Proprietary Project No. 2127Report No. -1.1-2125245 Page 96 Figure Proprietary Project No. 2127Report No. 1-11-2125245 Page 97 Figure Proprietary Project No. 2127Report No. 111-2125245 Page 98 Figure Proprietary Project No. 2127Report No. 1-1-2125245 Page 99 Figure Proprietary Project No. 2127Report No. 1-11-2125245 Page 100 Figure Proprietary Project No. 2127Report No. HI-2 125245Page 1. 0 1 Figure Proprietary Project No. 2127 Report No. 2125245 Page 102 Figure Proprietary Project No. 2127Report No. 1-1-2125245 Page 103 Figure Proprietary Project No. 2127 Report No. HI-2125245 Page 104 Figure Proprietary Project No. 2127Report No. HI-2125245 Page 105 Figure Proprietary ProJect No. 2127Report No. 1-11-2125245 Page 106 Figure Proprietary Project No. 2127Report No. 2125245 Page 107 Figure Proprietary Project No. 2127Report No. HI-2125245 Page 108 Figure Proprietary Project No. 2127Report No. HI-2125245 Page 109 Figure Proprietary Project No. 2127Report No. 2125245 Page 110 Figure Proprietary Project No. 2127Report No. HI-2125245 Page I I I Figure Proprietary Project No. 2127Report No. 1-11-2125245 Page 112 Figure Proprietary Project No. 2127Report No. 111-2125245 Page 113 Figure Proprietary Project No. 2127Report No. H-ti-2125245 Page 114 Figure Proprietary ProJect No. 2127Report No. HIl -2125245Page 1 15 Figure Proprietary Project No. 2127Report No. HI-2 125245Page 116 Figure Proprietary Project No. 2127Report No. 1-11-2125245 Page 117 Figure Proprietary Project No. 2127Report No. HI-2125245 Page H18 Appendix AProprietary Appendix BProprietary Project No. 2127Report No. HI-2125245 Page B-1 Appendix CProprietary Project No. 2127Report No. 1-H-2125245 Page C- I Supplement 1Additional Calculations to Support the Revised NETCO-SNAP-IN Rack Insert Design(I 1 pages including this page)Project No. 2127Report No. HI-2125245 Page S I -I1 SI.1 Introduction This Supplement documents the criticality safety evaluation for the storage of spent BWR fuel inthe Unit I and Unit 2 spent fuel pools (SFPs) at Quad Cities Station operated by Exelon. Thepurpose of this analysis is to justify that the specified changes in the NETCO-SNAP-IN rackinsert design [Sl.1 j are acceptable and bounded by the current analysis, presented in the mainpart of the report.S1.2 Methodology See Section 2 of the main report and as otherwise discussed below.S1.3 Acceptance CriteriaSee Section 3 of the main report.S1.4 Assumptions See Section 4 of the main report and as otherwise discussed below.S1.5 Input DataSee Section 5 of the main report. The revised dimensions of the NETCO-SNAP-INO rack insertare presented in Table SI -I and Figure SI -I.S1.6 Computer CodesSee Section 6 of the main report.S1.7 AnalysisThe comparison of the revised insert parameters presented in Table SI-1 with the previous insertdesign in Table 5.3(b) shows that changes are minor and therefore a significant impact on theconclusions made in the main part of the report is not expected.
Nevertheless, to verify thenegligible or minor impact of the revised insert design on results presented in the main part ofthe 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 rackThese cases are selected because the NETCO-SNAP-IN rack insert design change may impactthe reactivity in the rack. All other calculations from the main report are not affected by theNETCO-SNAP-INO rack insert design change and the results of the unaffected calculations areProject No. 2127Report No. 1-11-2125245 Page S 1 -2 used in this Supplement where applicable.
This approach is considered for both normal andaccident conditions.
S 1.7.1 SFP Rack Tolerances As discussed in Section S1.7, the effect of the manufacturing tolerances on reactivity of the SFPracks with revised inserts was determined.
The results of these calculations are presented inTable S 1-2. 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 kcfr inTFable S 1-5 and Table S1-6.S 1.7.2 Fuel Assembly Radial Positioning in the SFP RackAs discussed in Section S1.7, twelve fuel assembly radial positioning cases in the racks wereevaluated.
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 resultsshow that most cases show a negative reactivity effect, however some delta k,,,, values arepositive.
Therefore, a maximum delta kcIc value is applied as a bias and the correspondent 95/95uncertainty is statistically combined with other uncertainties in 'able S 1-5 and Table S1-6.S 1.7.3 Fuel Orientation in the SFP RackAs discussed in Section S1.7, five filel assembly orientation cases in racks were evaluated.
Theresults of these calculations are presented in Table S1-4. The result for the reference case is alsoincluded.
The results show that all cases are statistically equivalent and the reactivity effect offuel orientation is negligible.
Nevertheless, a maximum positive delta k,1, value is applied as abias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties inTFable S1-5 and Table S1-6.S 1.7.4 Maximum krff Calculations for Normal Conditions The calculations of the maximum kef. for normal conditions are described in Section 2.3.8 of themain part of the report. The results for the revised NETCO-SNAP-IN rack insert design and theresults from the main part of the report are tabulated in Table S1-5. The results show that themaximum ktfr 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-IN rack insert design and arebounded by the results from the main part of the report.Project No. 2127Report No. HI-2125245 Page S 1 -3 S 1.7.5 Maximum kdr Calculations for Abnormal and Accident Conditions The calculations of the maximum k.fr for accident conditions are described in Section 2.6 of themain part of the report. The bounding accident case from the main report is recalculated using therevised NETCO-SNAP-IN rack insert design. The results for the revised NETCO-SNAP-INO rack insert design and the results from the main part of the report are tabulated in Table S1 -6. Theresults show that the maximum k~ff for abnormal and accident conditions in the SFP racks is lessthan 0.95 at a 95% probability and at a 95% confidence level for the revised NETCO-SNAP-IN rack insert design and are bounded by the results from the main part of the report.S1.8 References
[S .1] Transmittal of Design Information NF1 100434, Revision 1, "Quad Cities SFP Rack InsertDesign Information",
dated 09/i 1/2012.S1.9 Conclusions The criticality analysis for the storage of BWR assemblies in the Quad Cities SFP racks withrevised NETCO-SNAP-INO inserts has been performed.
The results show that kefr is M withthe stora racks full loaded with fuel of the highest anticipated reactivity, which is SVEA-96Optima2 , at a temperature corresponding to the highest reactivity.
The maximumcalculated 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 abnomial and accident conditions, thereactivity will not exceed the regulatory limit of 0.95.The results show that the specified changes in the insert desthe current analvsis.
oresented in the main Dart of the report. ITherefore, any insert width dimension between the value used in themain report including the specified manufacturing tolerances and the value evaluated in thisSupplement is acceptable.
Project No. 2127Report No. 1I1-2125245 Page S 1 -4 Table SI-IFuel Rack Insert Revised Dimensions
[S 1.1]-1For the details of the insert dimensions, see Figure S I-i.t See 'Table 5.3(b)Project No. 2127Report No. HI-21 25245Page S 1-5 Table S 1-2Results of the MCNP5 Calculations for Revised Rack Tolerances Revised Reference Burnup delta kac Max delta Max deltaDescription (GWd/mtU)
Filename kclc sigma (95/95) kclc kteit(95/95) (95/95)Design basis 15.5 op146-rt201155r Reference Reference Reference (reference)
Max cell ID 15.5 op146-rt202155r
-0.0091 0.0000 0.0000Max cell pitchMax wall thickness 15.5 op146-rt203155r 0.0017 0.0017 0.0025Min wall thickness 15.5 op146-rt204155r
!_0.0011_
Max insert width 15.5 op146-rt206155r 1_ 1 0.0016 0.0030 0.0004Min insert width 15.5 op]46-rt207155r 0.0030Statistical combination of rack tolerances 0.0035 0.00261 See Table 7.8Note 1: The CASMO depletion calculation filenames are op 146-dbc(-ac).
Project No. 2127Report No. 111-2125245 Page S 1 -6 Table S 1-3Results of the MCNP5-1.51 Calculations for Revised Fuel Radial Positioning in SFP RacksRevised Reference Burnup Revised Refernc.e ncDescription (GWd/mtU)
Filename kCHIC sigma delta kcaic Un(9 delta kcal, 9 5nc.(~95/95)
(95/95)2x2 reference 15.5 2x2dbrot0l55r Ref Ref. Ref. Ref.(Case 2.3.5.2.1) 2W2 eccentric center 15.5 2x2ecnt 155r1 -0.0028 0.0015 -0.0053 0.0015(Case 2.3.5.2.2) 2x2 eccentric in 15.5 2x2ein155r
-0.0054 0.0015 -0.0081 0.0013(Case 2.3.5.2.3)
.0.0.802x2 eccentric out 15.5 2x2eoutl55r
-0.0014 0.0015 -0.0047 0.0014(Case 2.3.5.2.4)
...2x2 insert/cell center 15.5 2x2icnt]
55r 0.0001 0.0016 0.0002 0.0013(Case 2.3.5.2.5) 8x8 reference 15.5 8x8dbc155r 1 Ref Ref Ref. RefCase 2.3.5.2.61R 8x8 eccentric center 15.5 8x8ecntl55r
-0.0032 0.0015 -0.0023 0.0014(Case 2.3.5.2.7)
....8x8 eccentric in 15.5 8einI55r
-0.0071 0.0015 -0.0080 0,0016(Case 2.3.5.2.8) 8x8 eccentric out 15.5 8x8eoutl55r
-0.0035 0.0016 -0.0035 0.0014(Case 2.3.5.2.9) 8x8 insert/cell center 15.5 8xSicnt 155r 0.0009 0.0014 0.0016 0.0014(Case 2.3.5.2.10)
Ix reference Ref Ref. Ref. Ref.(Case 2.3.5.2.11) 5bc 1555rIx] insert/cell center 15.5 lxlicntl55r1 0.0004 0.0015 0.0000 0.0015(Case 2.3.5.2.12) 1 ___ I i I It See Table 7.9(a)Note 1: The CASMO depletion calculation filenames are opl46-dbc(-ac).
Project No. 2127Report No. HI-2125245 Page S 1 -7 Table S1-4Results of the MCNP5-1.51 Calculations for Revised Fuel Orientation in SFP RacksRevised Reference Burnup Revised Reference unc.Description (GWd/mtU)
Filename k,,,&#xfd; sigma delta (9/5 delta (9/5___________(95/95) dlak. (95/95)Reference (Shown in 15.5 2x2dbrotOl55r Ref Ref. Ref RefFigure 2.9(a)) 5 d t 51fe.Rotated fuel assembly 155 2x2dbrotl 155r 00004 0.0014 -0.0008 0.0014(shown in Figure 2.9(b))Rotated fuel assembly 155 2x2dbrot2lS5r 0.0011 0.0015 -0.0007 0.0014(shown in Figure 2.9(c))Rotated fuel assembly 15.5 2x2dbrot3l55r 0.0016 0,0014 -0.0013 0.0013(shown in Figure 2.9(d)) 5.Rotated fuel assembly 155 2x2dbrot4l55r 0.0024 0.0016 -0.0007 0.0013(shown in Figure 2.9(e))t See Table 7.9(b)Note 1: The CASMO depletion calculation filenames are opl46-dbc(-ac).
Project No. 2127Report No. HI-2125245 Page S 1-8 ZJI!-4 -__ __ -_Im_I-If-________________________________________________
I _______________________
I.I--IProject No. 2127Report No. 11I-2125245 Page S 1 -9 4 -4_____1 =1I_ _ _ _ _ _ 11 .. ....m-II =m -m--_ _ _ _ _ _ -_Project No. 2127Report No. HI-2125245 Page S I -10 Figure Proprietary Project No. 2127Report No. H1-2125245 Page SI -I I}}

Revision as of 01:16, 3 July 2018

Quad Cities, Units 1 and 2, Attachment 4 and 5 to RS-14-039, HI-2104790, Rev. 1, Nuclear Group Computer Code Benchmark Calculations, and HI-2125245, Rev. 4, Licensing Report for Quad Cities Criticality Analysis for Inserts-Non Proprietary V
ML14024A562
Person / Time
Site: Quad Cities  Constellation icon.png
Issue date: 01/22/2014
From:
Holtec
To:
Office of Nuclear Reactor Regulation
Shared Package
ML14024A496 List:
References
RS-14-039 HI-2104790, Rev. 1, HI-2125245, Rev. 4
Download: ML14024A562 (251)
v • d • e


Text

ATTACHMENT 4Holtec International Report No. HI-2104790, Revision 1,"Nuclear Group Computer Code Benchmark Calculations" U.HOINTERHEWLTECNATIO0NAL Holtec Center. 555 Lincoln Drive West, Marlton, NJ 08053Telephone (856) 797- 0900Fax (856) 797 -0909Nuclear Group Computer Code Benchmark Calculations FORGENERIC:

NON-PROPRIETARY VERSIONHoltec Report No: HI-2104790 Holtec Project No: GENERICSponsoring Holtec Division:

HTSReport Class: SAFETY RELATED Summary of Revisions Revision 0Original IssueRevision 1Additional criticality experiments were added to Appendix B. The index numbers ofcriticality experiments in Appendix C were updated to be consistent with a new revisionof Appendix B. The benchmark of MCNP5-1.51 with ENDF/B-VII was added inAppendix D.REPORT HI-2104790 i

Table of Contents1.0 Introdu ction .........................................................................................................................

32.0 M ethodology

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

32.1 Determination of Bias and Bias Uncertainty

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

32.2 Statistical Methods ......................................................................................................

42.2.1 Single Sided Tolerance Limit Method ..................................................................

42.2.2 Confidence Band with Administrative Margin Method ........................................

52.2.3 Non-parametric Statistical Treatment Method ......................................................

62.3 Area of Applicability

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

82.3.1 Key Parameters Identification

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

82.3.2 Screening Area of Applicability

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

93 .0 A ssu m ption s ........................................................................................................................

94.0 Computer Files ............................................................................................................

95 .0 S u m m ary .............................................................................................................................

96 .0 R eferences

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

10Appendix A: Holtec Approved Computer Program List .............................................

A-1Appendix B: Description of the Critical Experiments

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

B-1Appendix C: Benchmark of MCNP5-1.51 with ENDF/B-V

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

C-1Appendix D: Benchmark of MCNP5-1.51 with ENDF/B-VII

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

D-1REPORT HI-2 104790 iiREPORT HI-2104790 ii

1.0 Introduction

This report documents the criticality experiment benchmark validation calculations for thefollowing computer codes and libraries combinations and establishes the criticality code bias andbias uncertainty for these codes:MCNP5-1.51 with ENDF/B-V (Appendix C)MCNP5-1.51 with ENDF/B-VII (Appendix D)For that purpose, results from the codes are compared to the critical experiments referred to asthe Haut Taux de Combustion (HTC) experiments and to the selected

critical, presented inAppendix B, with geometric and material characteristics similar to that of spent fuel storage andtransport casks. The simulated fuel rods used in these experiments contained uranium or mixtureof uranium and plutonium oxides. In the HTC experiments the plutonium-to-uranium ratio andthe isotopic compositions of both the uranium and plutonium were designed to be similar to whatwould be found in a typical pressurized-water reactor (PWR) fuel assembly that initially had anenrichment of 4.5 wt % 235U and was burned to 37,500 MWd/MTU.The purpose of the calculation is to determine the code bias and bias uncertainty consistent withstandards such as ANSI/ANS-8.1

[1] and ANSI/ANS-8.17

[2]. Criticality safety standards ANSI/ANS-8.1 and ANSI/ANS-8.17 apply to criticality methods validation and to criticality evaluations, respectively.

ANSI/ANS-8.1 requires that a validation be performed on the methodused to calculate criticality safety margins and that the validation must be documented in awritten report describing the method, computer program and cross section libraries used, theexperimental data, the areas of applicability and the bias and margins of safety. ANSI/ANS-8.17 prescribes the criteria to establish sub-criticality safety margins.2.0 Methodology Validation of the computer code and continuous energy data library to perform criticality safetycalculation has been performed following reference

[5] methodology.

The validation allows theunderstanding of the accuracy of the calculational methodology to predict subcriticality.

Validation includes identification of the difference between calculated and experimental neutroneffective multiplication factor (keff), called the bias. A set of appropriate critical experiments areselected so bias trends can be drawn through statistical analyses.

The range of the benchmark parameters used to validate the calculational methodology primarily defines the area ofapplicability (AOA), which establishes the limits of the systems that can be analyzed using thevalidated criticality safety methodology.

Determination of Bias and Bias Uncertainty Following reference

[5] guide, the statistical analysis to determine the mean multiplication factor(keff) and the bias uncertainty (Sp) approach involves determining the weighted mean thatincorporates the uncertainty from both, measurements and calculation method as follows:REPORT HI-2104790 3

= a + Uep (2-)where ai is the uncertainty for the ith keff, exup is the measurement uncertainty and oc,,Ic-i is thecalculation uncertainty.

Then, the weighted mean multiplication factor keffand the biasuncertainty (Sp) are given by:keff --So1ar1(2-2)S,= N[s2 + j 2 (2-3)where s2 is the variance about the mean and j2 is the average total uncertainty, given by:2 21 1(2-4)-2_ n 1(2-5)where n is the number of critical experiments used in the validation and keg.i is the ith value of themultiplication factor.Bias is determined by the relation:

Bias = keff -1 if keff is less than 1, otherwise Bias = 0 (2-6)Because a positive bias may be nonconservative, a bias is set to zero if the calculated average keffis greater than one.Statistical MethodsSingle Sided Tolerance Limit MethodIf the benchmark calculated neutron multiplication factor does not exhibit trends with theparameters, the lower tolerance limit or single sided tolerance limit method can be used. Aweighted lower limit tolerance (KL) is a single lower limit above which a defined fraction of thepopulation of keg is expected to lie, with a prescribed confidence and within the area of theapplicability.

The term "weighted" refers to a specific statistical technique where theREPORT HI-2104790 4

uncertainties in the data are used to weight the data point. Data with high uncertainties will haveless "weight" than data with small uncertainties.

A lower tolerance limit can be used when there are no trends apparent in the critical experiment results and the critical experiment results have a normal distribution.

The method is applicable only within the limits of the validation data without extrapolating the AOA. The single sidedlower tolerance limit is defined by the equation:

KL = keff -U X Sp (2-7)Ifkeff >- 1, then KL = 1 -U X Sp (2-8)where Sp is the square root of the pooled variance used as the mean bias uncertainty whenapplying the single sided tolerance limit for a normally distributed data and U is the single sidedlower tolerance factor, determined from the following equations

[6]. Note that for groups withlarger than 50 samples, the single sided lower tolerance factor for 50 samples was conservatively used.z1Pz+ z -abU =a(2-9)2a = z1-Y2(N-1)(2-10)22 ____b =1- N pz-N(2-11)where zj-p is the critical value from the normal distribution that is exceeded with probability i-pand z1.y is the critical value from the normal distribution that is exceeded with probability 1-y.Confidence Band with Administrative Margin MethodIf the benchmarks calculated neutron multiplication factor exhibit a trend with a given parameter, the method based on a confidence band with administrative margin can be used. This methodapplies a statistical calculation of the bias and its uncertainty plus an administrative margin to alinear fit of the critical experiment benchmark data.The confidence band W is defined for a confidence level of (1-y) using the relationship:

W = max {w(xi,),w(x,,,)}

(2-12)whereREPORT HI-2104790 5

w(x) = tj~ x sp 1 + -n+ --)(2-13)andn is the number of critical experiments used in establishing kcaI(x),tl-Y is the Student-t distribution statistic for 1-y and n-2 degrees of freedom,2 is the mean value of the parameter x in the set of calculations, Xmin, x, are the minimum and maximum values of the independent parameter x,Sp is the pooled standard deviation for the set of criticality calculations given by:= S2(x) + S~w (2-14)where S2k(x) is the variance of the regression fit and is given by:Sk(x) = (n -2) 1 (ke f-1 -k --g) }2-(2-15)k is the mean value of the calculated keff and sw2 is the within-variance of the data:S'2 1 1 ini=l,n(2-16)where q1 -traic-i + rex is the uncertainty for the ith keff, ep is the measurement uncertainty and Ucak.i is the calculated uncertainty.

Non-parametric Statistical Treatment MethodData that do not follow a normal distribution can be analyzed by non-parametric techniques.

Theanalysis results in a determination of the degree of confidence that a fraction of the truepopulation of data lies above the smallest observed value. The more data is available in thesample, the higher the degree of confidence.

The following equation determines the percent confidence that a fraction of the population isabove the lowest observed value:REPORT HI-2 104790 6REPORT HI-2104790 6

m-Ifl = 1 -j! (n --j)! (1 -q)Iqn-j(2-17)whereq is the desired population fraction (normally 0.95),n is the number of data in one data sample,m is the rank order indexing from the smallest sample to the largest (m=l for the smallestsample; m=2 for the second smallest sample, etc.). Non-parametric techniques do notrequire reliance upon distributions, but are rather an analysis of ranks. Therefore, thesamples are ranked from the smallest to the largest.For a desired population fraction of 95% and a rank of order of 1 (the smallest data sample),

theequation reduces to:= l-q" = 1-0.95" (2-18)This information is then used to determine the Non-parametric Margin from Table 2.2 inReference

[5].For non-parametric data analysis, KL is determined by:KL = Smallest keff value -Uncertainty for Smallest keff- Non-parametric Margin (NPM) (2-19)Single-Sided Tolerance Band MethodWhen a relationship between a calculated keff and an independent variable can be determined, asingle-sided lower tolerance band may be used. This is a conservative method that provides afitted curve above which the true population of keff is expected to lie. The tolerance bandequation is actually a calibration curve relation.

The equation for the single-sided lower tolerance band isKL = Kfit(x) -Spfit I2F(a+n +-Z) (n-2a(x)x yn2(2-20)where:KIt(x) is the function derived from the trend analysis, p is the desired confidence (0.95),gat'"'2) is the F distribution percentile with degree of fit, n-2 degrees of freedom.

Thedegree of fit is 2 for a linear fit,REPORT HI-2104790 7

n is the number of critical experiment keff values,x is the independent fit variable, xi is the independent parameter in the data set corresponding to the ,ith,' keff value,;? is the weighted mean of the independent variables, Z2P.l is the symmetric percentile of the Gaussian or normal distribution that contains the Pfraction, y = (I -p)/2, (2-21)2X 1-y,n-2 is the upper Chi-square percentile,

_- S 2 5 2Spfit f= t + (2-22)= n- -2 To.2 [keff. -fitt(Xi)]

21Slit = 1 1(2-23)Area of Applicability The area(s) of applicability refers to the key physical parameter(s) that define a particular fissileconfiguration.

This configuration can either be an actual system or a process.

The determination of the AOA of the validation is determined following NUREG/CR-6698 steps [5]. The approachused in developing the AOA consists of the following steps:i. Identification of the key parameters associated with the system to be evaluated.

ii. Establishment a "screening" AOA for critical experiments.

iii. Identification of criticality experiments that are within the "screening" AOA.iv. Determination of the detailed AOA based on the selected criticality benchmark experiments.

v. Demonstration that the system to be evaluated in within the AOA provided by the criticalexperiments.

Steps i. and ii. are presented in subsections 2.3.1 and 2.3.2, respectively.

Step iii. is presented inAppendix B. Steps iv. and v. are presented in Appendix C and D.Key Parameters Identification REPORT HI-2 104790 8REPORT HI-2104790 8

This validation will cover a number of designs but all the designs will consider the same keyparameters in defining the applicability area. These parameters fall into three categories:

materials, geometry and neutron energy spectra.Regarding

material, the fuel is a uranium or mixture of uranium and plutonium oxides pelletsclad in a zirconium alloy. The moderator and reflector is water which in some cases hasdissolved boron. or gadolinium solutions.

Absorber plates made of borated steel, Boral, ZircaloyBoroflex or cadmium and absorber rods made of steel, aluminum, Gd203, Pyrex, Vicor orborated aluminum will be included in this validation.

Some experiments were performed withsteel or lead reflector screens.Regarding

geometry, the fuel in the HTC experiments is in square lattices with pin diameter

-9.5 mm and pitch in the range found on Table B-1 through Table B-6. The geometry parameters of other selected critical experiments are varied in a wide range and they can be found inreferences

[B.6] through [B.12]. The fuel assemblies may be separated by water, water and anabsorber plate or water and absorber rods. The system may be water reflected or steel/lead reflected.

Regarding the neutron energy spectra, they are thermal with EALF values in the range of 0.07and 1.55 eV.Table 2-1 presents the key physical parameters for AOA selected.

Screening Area of Applicability For the key parameters selected in section 2.3.1, Table 2-1 summarizes the range of parameters for which the validation applies.

These data are the base for the selection of the criticalexperiments, which span the range of parameters.

3.0 Assumptions

No substantial simplifying assumptions were made in the modeling of the critical experiments used for benchmarking:

all experiments were modeled as full three-dimensional geometries, fuelrod arrays were modeled as lattices, all fuel rod details were modeled, and the water between therods was modeled as specified in the experiment description.

However, structures further awayfrom the experiment, such as building walls and foundations, were not included in the models.4.0 Computer FilesAll computer files to support this analysis are provided on the Holtec server in\Projects\0\Reports\HI-2104790 and its subdirectories.

5.0 SummaryThe criticality experiment benchmark validation calculations for the computer codes andlibraries shown in Section 1.0 were performed for the validation of the Holtec International REPORT HI-2104790 9

criticality safety methodology.

The results of calculations and the criticality code bias and biasuncertainty for these codes are presented in appropriate appendices.

The similarity between thechosen experiments and the actual systems has been based on a set of screening criteria as isstated in the NUREG/CR-6698

[5].The summary of biases and bias uncertainties for the validated computer codes is shown in Table5.1.6.0 References

[1] ANSI/ANS 8.1-1983, American National Standard For Nuclear Criticality Safety InOperations With Fissionable Materials Outside Reactors, American Nuclear Society, LaGrange Park, Illinois.

[2] ANSI/ANS-8.17, "American National Standard for Criticality Safety Criteria for theHandling,

Storage, and Transportation of LWR Fuel Outside Reactors,"

American NuclearSociety, La Grange Park, Illinois.

[3] Criticality Benchmark Guide for Light Water Reactor Fuel in Transportation and StoragePackages, NUREG/CR-6361 (ORNL/TM-1321 1), U.S. Nuclear Regulatory Commission, March 1997.[4] J.R. Taylor, An Introduction to Error Analysis (University Science Books, Mill Valley,California, 1982).[5] Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, U.S. Nuclear Regulatory Commission, January 2001.[6] M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91,August 1963.REPORT I-11-2104790 10 Table 2-1 Key Criticality System Parameters and Range of those Parameters in Expected DesignsParameter Critical Experiment Requirement Range of Key Parameters Fissionable Material 235U, 239Pu, 241Pu 235U, 239Pu, 241PuIsotopic Composition 235U/Ut < 5.Owt% 0. 16wt% to 5.74wt%Pu/(U+Pu)

< 20wt% 1. 104wt% to 20wt%/oPhysical Form UO2,MOX UO2,MOXModerator Material (coolant)

H HPhysical Form H20 H20Density Normal pressure

& temperature around 1.0 g/cm3condition Reflector Material H HPhysical Form H20 H20Density Normal pressure

& temperature around 1.0 g/cm3condition Interstitial Reflector MaterialPlate Steel or Lead Steel or LeadAbsorber MaterialNone, Boron (0 to 2550 ppm) orSoluble None, Boron or Gadolinium Gadolinium (0 to 197 ppm)Rods Boron Pyrex, Vicor, Steel or B-Al ISeparating MaterialWater, B-SS, Boral, Boroflex, Plate Water, B-SS, Boral or Cadmium Zircaloy or CadmiumGeometryFuel Square/Triangle lattice of fuel Square/Triangle lattice of fuelpins pinsNeutron Energy Thermal spectrum Thermal spectrumREPORT HI-2104790 11REPORT HI-2104790 I1I Table 5-1 Summary of Biases and Bias Uncertainties for the Validated Computer CodesComputer Code Total Bias Bias Uncertainty MCNP5-1.51 with ENDF/B-V (Appendix C)MCNP5-1.51 with ENDF/B-VII (Appendix D)REPORT HI-2 104790 12REPORT HI-2104790 12 Appendix AHoltec Approved Computer Program List(total number of pages: 5 including this page)Appendix Proprietary REPORT HI-2 104790 A-IREPORT HI-2104790 A-1 Appendix BDescription of the Critical Experiments (total number of pages: 16 including this page)REPORT HI-2104790 B-iREPORT HI-2104790 B-I B.1. Introduction and PurposeThe purpose of this Appendix is to document the description of the full set of criticalexperiments selected for the benchmark validation of computer codes.B.2. Physical Description of HTC Critical Experiments In the 1980s, a series of critical experiments referred to as the Haut Taux de Combustion (HTC)experiments was conducted by the Institut de Radioprotection et de Sfiretd Nucl~aire (IRSN) atthe experimental criticality facility in Valduc, France, between 1988 and 1990. The fuel rodswere fabricated specifically for this set of experiments.

The fuel consisted of 1-cm-long pelletscontained within Zircaloy-4 cladding.

The plutonium-to-uranium ratio and the isotopiccompositions of both the uranium and plutonium used in the simulated fuel rods were designedto be similar to what would be found in a typical pressurized-water reactor fuel assembly thatinitially had an enrichment of 4.5 wt % 235U and was burned to 37,500 MWd/MTU.

The fuelmaterial also includes 241Am, which is present due to the decay of 241Pu. The fuel rods were heldin place by an upper and a lower grid and were contained in one or four assemblies placed into arectangular tank. The critical approach was accomplished by varying the water or solution levelin the tank containing the fuel pin arrays. The critical condition was extrapolated from asubcritical configuration with a multiplication factor within 0.1% of 1.000.This section provides a summary description of the materials and physical layouts of the 156critical configurations.

Detailed descriptions of the critical experiments are presented inreferences

[B.1] through [B.4]. The HTC experiments include configurations designed tosimulate fuel handling activities, pool storage, and transport in casks constructed of thick lead orsteel and were categorized into four phases.B.2.1. Phase 1: Water-Moderated and Reflected ArraysThe first phase included 18 configurations, each involving a single square-pitched array of rodswith rod pitch varying from 1.3 to 2.3 cm.The tank was incrementally filled with water at room temperature, water being injected at thebottom of the tank. A measurement needle provided water height. Therefore, the water was usedas core moderator and as reflector beneath the fuel and around the array on four sides. Thecritical approach parameter was the water level.Eighteen experiments have been performed with various arrays and all are considered acceptable for use as benchmark experiments:

0 5 square or almost square array -square pitch 1.3, 1.5, 1.7, 1.9, 2.3 cm -15 experiments,

  • 1 rectangular centered array -square pitch 1.7 cm -2 experiments,
  • 1 rectangular no-centered array -square pitch 1.7 cm -1 experiment.

The experiments key physical parameters are summarized in Table B-1.REPORT HI-2 104790 B-2REPORT HI-2104790 B-2 B.2.2. Phase 2: Reflected Simple Arrays Moderated by Poisoned Water with Gadolinium or BoronThe second phase included 41 configurations that were similar to the first phase except that thewater used as moderator and reflector included either boron or gadolinium in solution at variousconcentrations.

The tank was incrementally filled with poisoned solution at room temperature, this solutionbeing pumped in the bottom of the tank. A measurement needle provided solution height. Thecritical approach parameter was the water level.Forty one experiments are evaluated and all are considered acceptable for use as benchmark experiments.

Twenty of them are performed with gadolinium solutions, and the others withboron solutions.

The experiments key physical parameters are summarized in Table B-2 through Table B-3.B.2.3. Phase 3: Pool StorageThe third phase simulated fuel assembly storage rack conditions and included 26 configurations with 1.6 cm square rods pitch arranged into four assemblies in a 2 x 2 array. These assemblies with, in some cases, canisters, were placed on a pedestal centered inside a parallelepiped tankwhich was itself located on the floor in the middle (approximately) of a large room. The spacingbetween assemblies was varied, and some of the assemblies had B-SS, Boral, or cadmiumplates attached to the sides of the four assemblies.

The tank was incrementally filled with water at room temperature, water being pumped in at thebottom of the tank. A measurement needle provided water height. Therefore, the water was usedas core moderator and as reflector beneath the fuel and around the array on four sides. Thecritical approach parameter was the water level.Twenty six experiments are evaluated and all are considered acceptable for use as benchmark experiments.

Eleven of them were performed with neutron absorbing canisters around the fourarrays, and the others without any.The experiments key physical parameters are summarized in Table B-4.B.2.4. Phase 4: Shipping CaskThe fourth phase simulated cask conditions and included 71 configurations similar to the Phase 3configurations except thick steel or lead shields were placed around the outside of the 2 x 2 arrayof fuel assemblies.

These assemblies with, in some cases, canisters, were placed on a pedestalcentered inside a parallelepiped tank which was itself located on the floor in the middle(approximately) of a large room. Space between assemblies and between assemblies and screenvaried from one case to another.REPORT HI-2104790 B-3REPORT HI-2104790 B-3 The tank was incrementally filled with water at room temperature, water being pumped in at thebottom of the tank. A measurement needle provided water height. Therefore, the water was usedas core moderator and as reflector beneath the fuel and around the array on four sides behind thereflector screens.

The critical approach parameter was the water level.Seventy one experiments are evaluated and all are considered acceptable for use as benchmark experiments.

Thirty eight experiments were performed with lead reflector screens and thirtythree with steel reflector screens.

Twenty six among the former and twenty one among the latterused absorbing canisters around the four arrays, and the others without any.The experiments key physical parameters are summarized in Table B-5 through Table B-6.B.3. Physical Description of the Selected Benchmark Critical Experiments The benchmark experiments are selected to cover a wide range of code applications for fresh andspent fuel storage analysis.

This section provides a summary description of the materials andphysical layouts of the 135 critical configurations with fresh and selected actinides for spent fuel.For the fresh fuel assumption, the code is compared to the critical experiments of un-irradiated U02 systems with geometric and material characteristics similar to that of fuel storage systems.For the spent fuel assumption with bumup credit, additional comparisons are made to un-irradiated mixed-oxide (MOX) fuel of similar characteristics to spent fuel. The U02 experiments 234 235 238 28 239 24 24address U, U and U. The MOX critical experiments address 238Pu, Pu, 24°pu, 241pu,242pu and 241Am. Detailed descriptions of the critical experiments are presented in references

[B.6] through [B. 12].Description of the selected critical experiments is summarized in Table B-7.B.4. References

[B.1] F. Fernex, "Programme HTC -Phase 1 : R~seaux de crayons dans l'eau pure (Water-moderated and reflected simple arrays) Rddvaluation des experiences,"

DSU/SEC/T/2005-33/D.R.,

Institut de Radioprotection et de Sfiret& Nuclraire, 2008.[B.2] F. Fernex, Programme HTC -Phase 2 : Rrseaux simples en eau empoisonnre (bore etgadolinium)

(Reflected simple arrays moderated by poisoned water with gadolinium orboron) Rddvaluation des experiences,"

DSU/SEC/T/2005-38/D.R.,

Institut deRadioprotection et de Sfiretd Nuclraire, 2008.[B.3] F. Fernex, "Programme HTC -Phase 3 : Configurations "stockage en piscine" (Poolstorage)

R66valuation des expdriences,"

DSU/SEC/T/2005-37/D.R.,

Institut deRadioprotection et de SfiretW Nuclaire, 2008.[B.4] F. Fernex, "Programme HTC -Phase 4 : Configurations "chateaux de transport" (Shipping cask) -R66valuation des exp6riences,"

DSU/SEC/T/2005-36/D.R.,

Institut deRadioprotection et de Sfiret6 Nuclraire, 2008.REPORT HI-2104790 B-4

[B.5] C. Portella, C. Woillard "Programme "HTC" -Experiences de criticit6 avec des crayonscombustibles HTC (type REP A haut taux de combustion)

-Rdsultats de l'6tudeparamrtrique avec de l'eau gadolinire."

[Translation:

.... Hbu" program -Criticity Experiments with Hbu fuel rods (LWR type at high bum up) -Results of parametric study with poisoned water with gadolinium."]

Note technique IPSN/SRSC n' 90.01.[B.6] International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03, NEA Nuclear Science Committee, September 2008 Edition[B.7] G.S. Hoovier et al., Critical Experiments Supporting Underwater Storage of TightlyPacked Configurations of Spent Fuel Pins, BAW-1645-4, Babcock & Wilcox Company,November 1991.[B.8] L.W. Newman et al., Urania Gadolinia:

Nuclear Model Development and CriticalExperiment Benchmark, BAW- 1810, Babcock and Wilcox Company, April 1984.[B.9] J.C. Manaranche et al., "Dissolution and Storage Experimental Program with 4.75%Enriched Uranium-Oxide Rods," Trans. Am. Nucl. Soc. 33: 362-364 (1979).[B. 10] S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990.[B.1 1] S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in OrganicModerator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1986.[B. 12] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton PartialPlutonium core, WCAP-3385-54, Westinghouse Electric Corp., Atomic Power Division, December 1965.[B. 13] Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data,NUREG/CR-6979 (ORNL/TM-2007/083),

U.S. Nuclear Regulatory Commission, September 2008.REPORT HI-2 104790 B-5REPORT HI-2104790 B-5 Table B-I Key Physical Parameters of the HTC Phase I Critical Experiments

[B. 1]Case Reference Experiment Pitch Number of Rods Date of Temperature Critical waternumber (cm) Along edge Total experiment

(°C) height (cm) (a)1 MIX-COMP-THERM-HTC-001 2327 50 x 50 2500 05/05/88 22.5 61.41 + 0.062 MIX-COMP-THERM-HTC-002 2335 2.3 38 x 37 1406 06/06/88 21.1 87.68 + 0.063 MIX-COMP-THERM-HTC-003 2336 37 x 37 1369 06/07/88 21.0 90.38 + 0.064 MIX-COMP-THERM-HTC-004 2337 27 x 27 729 06/09/88 20.7 63.77 + 0.065 MIX-COMP-THERM-HTC-005 2339 1.9 25 x 25 625 06/13/88 20.5 81.95 + 0.086 MIX-COMP-THERM-HTC-006 2340 25 x 24 600 06/14/88 20.7 90.22 + 0.067 MIX-COMP-THERM-HTC-007 2341 26 x 26 676 06/15/88 20.2 65.11 +/- 0.078 MIX-COMP-THERM-HTC-008 2342 1.7 25 x 25 625 06/16/88 21.0 74.86 +/- 0.069 MIX-COMP-THERM-HTC-009 2343 25 x 24 600 06/16/88 20.8 82.25 + 0.0610 MIX-COMP-THERM-HTC-010 2345 29 x 29 841 06/26/88 21.1 59.92 + 0.0611 MIX-COMP-THERM-HTC-01 1 2347 1.5 27 x 27 729 06/23/88 21.3 76.72 +/- 0.0612 MIX-COMP-THERM-HTC-012 2348 27 x 26 702 06/23/88 21.1 84.57 +/- 0.0613 MIX-COMP-THERM-HTC-013 2349 39 x 39 1521 06/29/88 21.3 53.77 +/- 0.0614 MIX-COMP-THERM-HTC-014 2352 1.3 34 x 34 1156 07/05/88 21.3 80.16 +/- 0.0615 MIX-COMP-THERM-HTC-015 2353 34 x 33 1122 07/06/88 21.3 86.35 +/- 0.0616 MIX-COMP-THERM-HTC-016 2355 50 x 18 900 07/19/88 21.0 69.07 +/- 0.0617 MIX-COMP-THERM-HTC-017 2357 1.7 50 x 17 850 07/21/88 21.4 83.15 + 0.0818 MIX-COMP-THERM-HTC-018 2361 50 x 18'b) 900 07/28/88 22.4 80.16 +/- 0.07(a) given at a level of confidence of 95%(b) no-centered arrayREPORT HI-2104790 B-6 Table B-2 Key Physical Parameters of the HTC Phase 2 Critical Experiments with Gadolinium Solutions

[B.2]Number of Rods Gadolinium Experiment Pitch Date of Temperature Critical waterCase Reference number (cm) Along (CM) (a) conc. (g/l)number ____ edge Total experiment (C) height (ba)19 MIX-COMP-THERM-HTC-019 2405 38 x 38 1444 01/20/89 20.3 81.86 +/- 0.04 0.05220 MIX-COMP-THERM-HTC-020 2406 38 x 37 1406 01/23/89 19.7 87.16 +/- 0.04 0.05221 MIX-COMP-THERM-HTC-021 2407 42 x 42 1764 01/23/89 20.1 80.13 +/- 0.04 0.10022 MIX-COMP-THERM-HTC-022 2408 42 x 41 1722 01/25/89 19.7 84.38 +/- 0.04 0.09923 MIX-COMP-THERM-HTC-023 2409 1.3 41 x 41 1681 01/25/89 19.6 89.54 + 0.04 0.09924 MIX-COMP-THERM-HTC-024 2410 46 x 46 2116 01/26/89 20.1 81.33 +/- 0.04 0.15125 MIX-COMP-THERM-HTC-025 2411 45 x 45 2025 01/27/89 20.0 89.49 +/- 0.04 0.14826 MIX-COMP-THERM-HTC-026 2412 50 x 50 2500 01/30/89 20.7 85.83 +/- 0.04 0.20027 MIX-COMP-THERM-HTC-027 2415 50 x 49 2450 02/01/89 19.6 90.03 + 0.05 0.19728 MIX-COMP-THERM-HTC-028 2417 50 x 50 2500 02/09/89 19.6 89.67 + 0.04 0.19629 MIX-COMP-THERM-HTC-029 2419 42 x 42 1764 02/14/89 21.4 85.88 +/- 0.05 0.14730 MIX-COMP-THERM-HTC-030 2420 42 x 41 1722 02/15/89 21.0 90.51 +/-0.05 0.14731 MIX-COMP-THERM-HTC-031 2422 1.5 36 x 36 1296 02/21/89 22.1 83.86 +/--0.05 0.09832 MIX-COMP-THERM-HTC-032 2423 36 x 35 1260 02/21/89 22.6 89.85 +/- 0.04 0.09833 MIX-COMP-THERM-HTC-033 2425 32 x 32 1024 02/24/89 20.9 73.60 +/- 0.05 0.04834 MIX-COMP-THERM-HTC-034 2427 31 x 31 961 02/27/89 20.6 84.14 +/- 0.04 0.04835 MIX-COMP-THERM-HTC-035 2430 1.7 31 x 30 930 03/01/89 21.1 85.87 +/- 0.05 0.04836 MIX-COMP-THERM-HTC-036 2434 1.9 35 x 35 1225 03/08/89 21.7 89.61 +/- 0.04 0.04837 MIX-COMP-THERM-HTC-037 2436 39 x 39 1521 03/13/89 22.5 85.86 + 0.05 0.0971.738 MIX-COMP-THERM-HTC-038 2433 50 x 23 1150 03/07/89 21.7 84.35 +/- 0.04 0.048(a) given at a level of confidence of 95%(b) nominal values given in the report [B.5], notretainedREPORT HI-2104790 B-7 Table B-3 Key Physical Parameters of the HTC Phase 2 Critical Experiments with Boron Solutions

[B.2]Experiment Pitch Date of Temperature Boron conc.Case Reference NumbeAlong Rods water heightnedge Total experiment

(°C) (cm) (a) (g/l)39 MIX-COMP-THERM-HTC-039 2437 37 x 37 1369 04/17/89 23.0 78.80 + 0.04 0.100 +/- 0.00140 MIX-COMP-THERM-HTC-040 2438 37 x 36 1332 04/18/89 22.8 83.84 +/- 0.04 0.106 +/- 0.00141 MIX-COMP-THERM-HTC-041 2441 39 x 39 1521 04/20/89 23.5 84.04 + 0.04 0.205 +/- 0.00242 MIX-COMP-THERM-HTC-042 2444 42 x 41 1722 04/26/89 23.0 85.40 + 0.05 0.299 +/- 0.0031.343 MIX-COMP-THERM-HTC-043 2446 45 x 44 1980 05/09/89 24.2 84.14 + 0.04 0.400 +/- 0.00444 MIX-COMP-THERM-HTC-044 2447 44 x 44 1936 05/10/89 24.7 88.63 +/- 0.05 0.399 +/- 0.00445 MIX-COMP-THERM-HTC-045 2448 47 x 47 2009 05/11/89 26.3 88.44 +/- 0.04 0.486 +/- 0.00546 MIX-COMP-THERM-HTC-046 2449 50 x 50 2500 05/17/89 25.1 90.64 +/- 0.04 0.587 +/- 0.00647 MIX-COMP-THERM-HTC-047 2459 49 x 49 2401 06/05/89 24.7 88.88 + 0.04 0.595 +/- 0.00648 MIX-COMP-THERM-HTC-048 2468 43 x 43 1849 06/15/89 22.7 89.46 +/- 0.04 0.499 +/- 0.00549 MIX-COMP-THERM-HTC-049 2470 39 x 39 1521 06/19/89 23.6 85.37 +/- 0.05 0.393 + 0.0041.550 MIX-COMP-THERM-HTC-050 2471 35 x 35 1225 06/21/89 23.6 88.90 +/- 0.04 0.295 +/- 0.00351 M1X-COMP-THERM-HTC-051 2473 32 x 32 1024 06/27/89 23.5 87.02 +/- 0.04 0.200 + 0.00252 MIX-COMP-THERM-HTC-052 2475 30 x 29 870 07/03/89 23.6 82.48 +/- 0.04 0.089 + 0.00153 MIX-COMP-THERM-HTC-053 2478 28 x 28 784 07/06/89 23.8 85.10 +/- 0.04 0.090 + 0.00154 MIX-COMP-THERM-HTC-054 2483 32 x 32 1024 07/19/89 24.2 87.06 +/- 0.04 0.194 +/- 0.00255 MIX-COMP-THERM-HTC-055 2485 1.7 37 x 37 1369 07/21/89 24.5 89.65 + 0.04 0.286 + 0.00356 MIX-COMP-THERM-HTC-056 2487 45 x 44 1980 08/09/89 23.8 88.72 +/- 0.04 0.415 +/- 0.00457 MIX-COMP-THERM-HTC-057 2482 50 x 21 1050 07/17/89 24.0 77.74 +/- 0.04 0.100 +/- 0.00158 MIX-COMP-THERM-HTC-058 2490 39 x 38 1482 09/08/89 22.9 88.41 +/- 0.04 0.220 +/- 0.0021.959 MIX-COMP-THERM-HTC-059 2492 31 x 30 930 09/14/89 22.0 86.95 +/- 0.04 0.110 +/- 0.001(a) given at a level of confidence of 95%/REPORT HI-2104790 B-8 Table B-4 Key Physical Parameters of the HTC Phase 3 Critical Experiments (pin pitch 1.6 cm) [B.3]Experiment Canister Number of Rods Date of Temperature Critical water WaterCase Reference number Type Along edge Total experiment (OC) height (cm) (a) Gap(cm)60 MIX-COMP-THERM-HTC-060 2518 25 x 25 625 01/04/90 18.3 88.83 + 0.34 3.561 MIX-COMP-THERM-HTC-061 2520 25 x 24 600 01/09/90 18.7 49.55 +/- 0.34 0.062 MIX-COMP-THERM-HTC-062 2521 Borated 25 x 24 600 01/10/90 18.8 71.45 +/- 0.34 2.0Steel63 MIX-COMP-THERM-HTC-063 2522 25 x 24 600 01/10/90 19.0 89.96 +/- 0.34 3.064 MIX-COMP-THERM-HTC-064 2523 25 x 24 600 01/12/90 18.9 58.23 +/- 0.34 1.065 MIX-COMP-THERM-HTC-065 2514 Boral 25 x 25 625 12/28/89 20.6 90.03 +/- 0.34 0.066 MIX-COMP-THERM-HTC-066 2511 25 x 25 625 12/21/89 21.1 82.16 +/- 0.34 2.067 MIX-COMP-THERM-HTC-067 2524 25 x 24 600 01/15/90 18.7 55.33 +/- 0.34 0.068 MIX-COMP-THERM-HTC-068 2525 25 x 24 600 01/16/90 19.0 67.95 +/- 0.34 1.0m69 MIX-COMP-THERM-HTC-069 2526 25 x 24 600 01/17/90 19.1 79.83 + 0.34 1.570 MIX-COMP-THERM-HTC-070 2527 25 x 24 600 01/18/90 19.1 58.66 +/- 0.34 0.571 MIX-COMP-THERM-HTC-071 2509 25 x 25 625 12/19/89 20.9 84.75 +/- 0.34 18.072 MIX-COMP-THERM-HTC-072 2531 25 x 24 600 01/23/90 19.0 88.2 +/- 0.34 14.573 MIX-COMP-THERM-HTC-073 2532 24 x 24 576 01/24/90 19.1 81.18 -0.34 11.074 MIX-COMP-THERM-HTC-074 2533 24 x 23 552 01/25/90 19.3 82.12 +/-0.34 10.075 MIX-COMP-THERM-HTC-075 2534 23 x 23 529 01/26/90 19.4 81.2 +/- 0.34 9.076 MIX-COMP-THERM-HTC-076 2535 22 x 22 484 01/30/90 19.7 86.17 +/- 0.34 8.077 MIX-COMP-THERM-HTC-077 2536 20 x 20 400 01/31/90 19.7 82.08 +/- 0.34 6.078 MIX-COMP-THERM-HTC-078 2537 17 x 17 289 02/01/90 19.9 77.92 +/- 0.34 4.079 MIX-COMP-THERM-HTC-079 2538 17 x 16 272 02/02/90 20.0 90.28 +/- 0.34 4.080 MIX-COMP-THERM-HTC-080 2539 14 x 14 196 02/05/90 20.2 75.99 +/- 0.34 2.081 MIX-COMP-THERM-HTC-081 2541 13 x 13 169 02/06/90 20.0 83.17 +/- 0.34 1.082 MIX-COMP-THERM-HTC-082 2544 13 x 13 169 02/07/90 20.4 79.46 +/- 0.34 0.083 MIX-COMP-THERM-HTC-083 2547 25 x 25 625 02/19/90 20.9 29.46 +/- 0.34 0.084 MIX-COMP-THERM-HTC-084 2548 25 x 25 625 02/20/90 20.9 37.96 +/- 0.34 4.085 MIX-COMP-THERM-HTC-085 2549 25 x 25 625 02/20/90 21.0 64.43 +/- 0.34 10.0(a) given at a level of confidence of 95%REPORT HI-2104790 B-9 Table B-5 Key Physical Parameters of the HTC Phase 4 Critical Experiments with the Lead Screen (four 25 x 25 arrays with 1.6 cmpitch) [B.4]Case Reference Experiment Canister Date of Temperature Water Gap Screen array Critical waternumber Type experiment

(°C) (cm) (a) distance (cm) height (cm)(b) (c)86 MIX-COMP-THERM-HTC-086 2562 03/16/90 22.8 0.0 0.0 42.53 + 0.3487 MIX-COMP-THERM-HTC-087 2563 03/19/90 23.1 0.5 0.0 44.79 +/- 0.3488 MIX-COMP-THERM-HTC-088 2564 03/20/90 23.3 1.0 0.0 47.86 + 0.3489 MIX-COMP-THERM-HTC-089 2565 03/21/90 23.1 1.5 0.0 51.3 + 0.3490 MIX-COMP-THERM-HTC-090 2566 03/22/90 23.3 2.0 0.0 54.65 + 0.3491 MIX-COMP-THERM-HTC-091 2567 Borated 03/22/90 23.4 3.0 0.0 62.04 -0.34Steel92 MIX-COMP-THERM-HTC-092 2568 03/23/90 23.6 3.5 0.0 66.10 -0.3493 MIX-COMP-THERM-HTC-093 2569 03/26/90 23.5 2.0 0.5 55.87 -0.3494 MIX-COMP-THERM-HTC-094 2570 03/27/90 23.1 2.0 1.0 57.33 + 0.3495 MIX-COMP-THERM-HTC-095 2571 03/27/90 23.0 2.0 1.5 58.68 + 0.3496 MIX-COMP-THERM-HTC-096 2572 03/28/90 22.9 2.0 2.0 59.78 +/- 0.3497 MIX-COMP-THERM-HTC-097 2586 04/23/90 21.9 0.0 0.0 72.47 +/- 0.3498 MIX-COMP-THERM-HTC-098 2587 04/24/90 22.0 0.0 0.0 72.49 +/- 0.3499 MIX-COMP-THERM-HTC-099 2588 Boral 04/24/90 22.2 0.0 0.5 74.70 +/- 0.34100 MIX-COMP-THERM-HTC-100 2624 07/13/90 21.6 1.0 0.0 86.06 +/- 0.34101 MIX-COMP-THERM-HTC-101 2625 07/18/90 22.4 0.5 0.0 76.69 + 0.34102 MIX-COMP-THERM-HTC-102 2577 04/05/90 22.7 0.0 0.0 46.13 +/- 0.34103 MIX-COMP-THERM-HTC-103 2578 04/05/90 22.6 1.0 0.0 52.89 +/- 0.34104 MIX-COMP-THERM-HTC-104 2579 04/06/90 22.6 2.0 0.0 63.52 +/- 0.34105 MIX-COMP-THERM-HTC-105 2580 Cadmium 04/09/90 22.4 2.5 0.0 69.83 +/- 0.34106 MIX-COMP-THERM-HTC-106 2581 04/11/90 22.5 2.0 0.5 65.84 +/- 0.34107 MIX-COMP-THERM-HTC-107 2582 04/11/90 22.5 2.0 1.0 68.63 + 0.34108 MIX-COMP-THERM-HTC-108 2583 04/12/90 22.4 2.0 1.5 71.21 +/- 0.34REPORT HI-2104790 B-10 Screen array Critical waterCase Reference Experiment Canister Date of Temperature Water Gap Sce (cm) heit(cm) wnumber Type experiment (OC) (cm) (a) (b) (c)109 M1X-COMP-THERM-HTC-109 2584 04/12/90 22.4 2.0 2.0 73.36+/- 0.34110 MIX-COMP-THERM-HTC-1 10 2621 07/03/90 22.3 3.0 0.0 76.25 +/- 0.34111 MIX-COMP-THERM-HTC-1 11 2622 07/04/90 22.3 3.5 0.0 83.38 +/- 0.34112 MIX-COMP-THERM-HTC-1 12 2550 02/23/90 21.4 0.0 0.0 27.45 + 0.34113 MIX-COMP-THERM-HTC-1 13 2551 02/26/90 22.1 1.0 0.0 28.00 +/- 0.34114 MIX-COMP-THERM-HTC-114 2552 02/28/90 21.8 2.0 0.0 29.37 +/- 0.34115 MIX-COMP-THERM-HTC-115 2553 03/01/90 21.8 4.0 0.0 34.65 +/- 0.34116 MIX-COMP-THERM-HTC-1 16 2554 03/02/90 21.3 6.0 0.0 41.60 +/- 0.34117 MIX-COMP-THERM-HTC-117 2555 03/05/90 20.7 8.0 0.0 48.65 +/- 0.34No118 MIX-COMP-THERM-HTC-118 2556 03/06/90 20.7 10.0 0.0 54.74 +/- 0.34119 MIX-COMP-THERM-HTC-119 2557 03/07/90 20.9 12.0 0.0 59.57 +/-0.34120 MIX-COMP-THERM-HTC-120 2558 03/09/90 21.3 2.0 0.5 29.43 +/- 0.34121 MIX-COMP-THERM-HTC-121 2559 03/12/90 21.7 2.0 1.0 29.46 +/- 0.34122 MIX-COMP-THERM-HTC-122 2560 03/13/90 21.9 2.0 1.5 29.55 +/-0.34123 MIX-COMP-THERM-HTC-123 2561 03/14/90 22.3 2.0 2.0 29.62 +/- 0.34(a) Water gap between arrays.(b) Water gap between screen and array.(c) Given at a level of confidence of 95%REPORT HI-2104790 B-11 Table B-6 Key Physical Parameters of the HTC Phase 4 Critical Experiments with the Steel Screen (four 25 x 25 arrays with 1.6 cmpitch) [B.4]Experiment Canister Date of Temperature Water Gap Screen array Critical waterCase Reference (CM) (a) distance (cm) height (cm)number Type experiment (C) ((b) Mc)124 MIX-COMP-THERM-HTC-124 2602 05/21/90 23.6 0.0 0.0 42.11 + 0.34125 MIX-COMP-THERM-HTC-125 2603 05/21/90 23.4 0.5 0.0 44.14 +/- 0.34126 MIX-COMP-THERM-HTC-126 2604 05/22/90 22.9 1.0 0.0 46.96 +/- 0.34127 MIX-COMP-THERM-HTC-127 2605 05/29/90 20.4 1.5 0.0 50.16 + 0.34128 MIX-COMP-THERM-HTC-128 2606 05/30/90 20.1 2.0 0.0 53.43 +/- 0.34129 MIX-COMP-THERM-HTC-129 2607 Borated 05/31/90 20.0 2.0 0.5 54.71 +/--0.34Steel130 MIX-COMP-THERM-HTC-130 2608 06/05/90 20.2 2.0 1.0 56.32 +/- 0.34131 MIX-COMP-THERM-HTC-131 2609 06/05/90 20.1 2.0 1.5 57.96 +/- 0.34132 MIX-COMP-THERM-HTC-132 2610 06/06/90 19.7 2.0 2.0 59.16 +-0.34133 MIX-COMP-THERM-HTC-133 2611 06/08/90 19.5 3.0 0.0 60.38 +/- 0.34134 MIX-COMP-THERM-HTC-134 2612 06/12/90 20.1 3.5 0.0 64.19 + 0.34135 MIX-COMP-THERM-HTC-135 2589 04/26/90 22.4 0.0 0.0 69.82 + 0.34Boral136 MIX-COMP-THERM-HTC-136 2626 07/19/90 22.6 0.5 0.0 73.44 + 0.34137 MIX-COMP-THERM-HTC-137 2613 06/13/90 20.5 0.0 0.0 44.70 +/- 0.34138 MIX-COMP-THERM-HTC-138 2614 06/13/90 20.6 1.0 0.0 51.00 + 0.34139 MIX-COMP-THERM-HTC-139 2615 06/14/90 20.6 2.0 0.0 60.26 +/- 0.34140 MIX-COMP-THERM-HTC-140 2616 06/15/90 20.7 2.0 0.5 62.54 + 0.34141 MIX-COMP-THERM-HTC-141 2617 Cadmium 06/18/90 21.0 2.0 1.0 65.85 + 0.34142 MIX-COMP-THERM-HTC-142 2618 06/19/90 21.3 2.0 1.5 68.70 +/- 0.34143 MIX-COMP-THERM-HTC-143 2619 06/20/90 21.5 2.0 2.0 71.00 +/- 0.34144 MIX-COMP-THERM-HTC-144 2620 06/21/90 21.7 2.5 0.0 65.76 +/- 0.34145 MIX-COMP-THERM-HTC-145 2590 04/27/90 22.4 0.0 0.0 27.77 + 0.34No146 MIX-COMP-THERM-HTC-146 2591 05/09/90 24.4 1.0 0.0 28.34 + 0.34REPORT HI-2104790 B-12 Experiment Canister Date of Temperature Water G Screen array Critical waterCase Reference ECMxperiment C roa Wae) (ap distance (cm) height (cm)number Type experiment (C) ((b) (c)147 MIX-COMP-THERM-HTC-147 2592 05/10/90 24.4 2.0 0.0 29.74 +/- 0.34148 MIX-COMP-THERM-HTC-148 2593 05/10/90 24.3 2.0 0.5 29.68 +/- 0.34149 MIX-COMP-THERM-HTC-149 2594 05/11/90 24.5 2.0 1.0 29.66 +/- 0.34150 MIX-COMP-THERM-HTC-150 2595 05/11/90 24.4 2.0 1.5 29.68 +/--0.34151 MIX-COMP-THERM-HTC-151 2596 05/14/90 24.7 2.0 2.0 29.76 +/- 0.34152 MIX-COMP-THERM-HTC-152 2597 05/15/90 24.6 4.0 0.0 35.33 +/- 0.34153 MIX-COMP-THERM-HTC-153 2598 05/15/90 24.6 6.0 0.0 43.24 +/- 0.34154 MIX-COMP-THERM-HTC-154 2599 05/16/90 24.7 8.0 0.0 51.30 +/- 0.34155 MIX-COMP-THERM-HTC-155 2600 05/17/90 24.7 10.0 0.0 58.73 +/- 0.34156 MIX-COMP-THERM-HTC-156 2601 05/18/90 24.6 12.0 0.0 64.84 +/- 0.34(a) Water gap between arrays.(b) Water gap between screen and array.(c) Given at a level of confidence of 95%REPORT HI-2104790 B- 13 Table B-7 Description of the Selected Benchmark Critical Experiments

[LB.6]Case Reference Identification U, wt% 'wt157 LEU-COMP-THERM-011-001 Core 1 2.46 -158 LEU-COMP-THERM-01 1-002 Core II 2.46 -159 LEU-COMP-THERM-01 1-004 Core IIIB 2.46 -160 LEU-COMP-THERM-0 11-015 Core IX 2.46 -161 LEU-COMP-THERM-051-001 Core X 2.46 -162 LEU-COMP-THERM-051-003 Core XIB 2.46 -163 LEU-COMP-THERM-051-009 Core XII 2.46 -164 LEU-COMP-THERM-051-010 Core XIII 2.46 -165 LEU-COMP-THERM-051-012 Core XIV 2.46 -166 LEU-COMP-THERM-051-013 Core XV 2.46 -167 LEU-COMP-THERM-051-014 Core XVI 2.46 -168 LEU-COMP-THERM-051-015 Core XVII 2.46 -169 LEU-COMP-THERM-051-016 Core XVIII 2.46 -170 LEU-COMP-THERM-051-017 Core XIX 2.46 -171 LEU-COMP-THERM-051-018 Core XX 2.46 -172 LEU-COMP-THERM-051-019 Core XXI 2.46 -173 BAW-1645-4

[B.7] S-type Fuel, w/886 ppm B 2.46 -174 BAW-1645-4

[B.7] S-type Fuel, w/746 ppm B 2.46 -175 BAW-1645-4

[B.7] SO-type Fuel, w/l 156 ppm B 2.46 -176 BAW-1810

[B.8] Case 1 1337 ppm B 2.46 -177 BAW-1810

[B.8] Case 12 1899 ppm B 2.75 -178 French [B.9] Water Moderator 0 gap 4.75 -179 French [B.9] Water Moderator 2.5 cm gap 4.75 -180 French [B.9] Water Moderator 5 cm gap 4.75 -181 French [B.9] Water Moderator 10 cm gap 4.75 -182 LEU-COMP-THERM-0 17-012 Steel Reflector, 1.321 cm separation 2.35 -183 LEU-COMP-THERM-017-013 Steel Reflector, 2.616 cm separation 2.35 -184 LEU-COMP-THERM-017-014 Steel Reflector, 3.912 cm separation 2.35 -185 LEU-COMP-THERM-001-008 Steel Reflector, Infinite separation 2.35 -186 LEU-COMP-THERM-010-016 Steel Reflector, 1.321 cm separation 4.306 -187 LEU-COMP-THERM-010-018 Steel Reflector, 2.616 cm separation 4.306 -188 LEU-COMP-THERM-010-019 Steel Reflector, 5.405 cm separation 4.306 -189 LEU-COMP-THERM-004-010 Steel Reflector, Infinite separation 4.306 -190 LEU-COMP-THERM-013-003 Steel Reflector, with Boral Sheets 4.306 -191 LEU-COMP-THERM-010-021 Lead Reflector, 0.55 cm sepn. 4.306 -192 LEU-COMP-THERM-010-022 Lead Reflector, 1.956 cm sepn. 4.306 -193 LEU-COMP-THERM-010-023 Lead Reflector, 5.405 cm sepn. 4.306 -194 LEU-COMP-THERM-002-004 Experiment 004/032 -no absorber 4.306 -195 LEU-COMP-THERM-009-005 Exp. 009 1.05% Boron Steel plates 4.306 -REPORT HI-2104790 B-14 Case Reference Identification U, wt% Pu,196 LEU-COMP-THERM-009-007 Exp. 009 1.62% Boron Steel plates 4.306 -197 LEU-COMP-THERM-009-009 Exp. 031 -Boral plates 4.306 -198 PNL-7167

[B. 10] Experiment 214R -with flux traps 4.306 -199 PNL-7167

[B.10] Experiment 214V3 -with flux trap 4.306 -200 LEU-COMP-THERM-014-001 Case 173 -0 ppm B 4.306201 LEU-COMP-THERM-014-005 Case 177 -2550 ppm B 4.306 -202 PNL-5803

[B.I 1] MOX Fuel -Type 3.2 Exp. 21 0.71 20203 PNL-5803

[B.1 1] MOX Fuel -Type 3.2 Exp. 43 0.71 20204 PNL-5803

[B.1 1] MOX Fuel -Type 3.2 Exp. 13 0.71 20205 PNL-5803

[B.11] MOX Fuel -Type 3.2 Exp. 32 0.71 20206 MIX-COMP-THERM-003-001 Saxton Case 52 PuO2 0.52" pitch 0.72 6.6207 WCAP-3385

[B.12] Saxton Case 52 U 0.52" pitch 5.74 -208 MIX-COMP-THERM-003-002 Saxton Case 56 PuO2 0.56" pitch 0.72 6.6209 MIX-COMP-THERM-003-003 Saxton Case 56 borated PuO2 0.72 6.6210 WCAP-3385

[B.12] Saxton Case 56 U 0.56" pitch 5.74 -211 MIX-COMP-THERM-003-005 Saxton Case 79 PuO2 0.79" pitch 0.72 6.6212 WCAP-3385

[B.12] Saxton Case 79 U 0.79" pitch 5.74 -213 MIX-COMP-THERM-002-030 0.700-in.

pitch 0 ppm B 0.72 2.0214 MIX-COMP-THERM-002-031 0.700-in.

pitch 688 ppm B 0.72 2.0215 MIX-COMP-THERM-002-032 0.870-in.

pitch 0 ppm B 0.72 2.0216 MIX-COMP-THERM-002-033 0.870-in.

pitch 1090 ppm B 0.72 2.0217 MIX-COMP-THERM-002-034 0.990-in.

pitch 0 ppm B 0.72 2.0218 MIX-COMP-THERM-002-035 0.990-in.

pitch 767 ppm B 0.72 2.0219 MIX-COMP-THERM-003-004 Saxton Case PuO2 0.735" pitch 0.72 6.6220 MIX-COMP-THERM-003-006 Saxton Case PuO2 1.04" pitch 0.72 6.6221 MIX-COMP-THERM-006-001 8 wt% 240Pu 0.80" pitch 0.71 2.0222 MIX-COMP-THERM-006-002 8 wt% 240Pu 0.93" pitch 0.71 2.0223 MIX-COMP-THERM-006-003 8 wt% 240Pu 1.05" pitch 0.71 2.0224 MIX-COMP-THERM-006-004 8 wt% 240Pu 1.143" pitch 0.71 2.0225 MIX-COMP-THERM-006-005 8 wt% 240Pu 1.32" pitch 0.71 2.0226 MIX-COMP-THERM-006-006 8 wt% 240Pu 1.386" pitch 0.71 2.0227 MIX-COMP-THERM-007-001 16 wt% 240Pu 0.93" pitch 0.72 2.0228 MIX-COMP-THERM-007-002 16 wt% 240Pu 1.05" pitch 0.72 2.0229 MIX-COMP-THERM-007-003 16 wt% 240Pu 1.143" pitch 0.72 2.0230 MIX-COMP-THERM-007-004 16 wt%/o 240Pu 1.32" pitch 0.72 2.0231 MIX-COMP-THERM-008-001 24 wt0/o 240Pu 0.80" pitch 0.72 2.0232 MIX-COMP-THERM-008-002 24 wt%/o 240Pu 0.93" pitch 0.72 2.0233 MIX-COMP-THERM-008-003 24 wt%/o 240Pu 1.05" pitch 0.72 2.0234 MIX-COMP-THERM-008-004 24 wt%/o 240Pu 1.143" pitch 0.72 2.0235 MIX-COMP-THERM-008-005 24 wt0o 240Pu 1.32" pitch 0.72 2.0REPORT HI-2104790 B-15 Case Reference Identification U, wt% Pu,236 MIX-COMP-THERM-008-006 24 wt% 240Pu 1.386" pitch 0.72 2.0237 MIX-COMP-THERM-005-001 18 wt% 240Pu 0.85" pitch 0.72 4.0238 MIX-COMP-THERM-005-002 18 wt% 240Pu 0.93" pitch 0.72 4.0239 MIX-COMP-THERM-005-003 18 wt% 240Pu 1.05" pitch 0.72 4.0240 MIX-COMP-THERM-005-004 18 wt% 240Pu 1.143" pitch 0.72 4.0241 MIX-COMP-THERM-005-005 18 wt%/o 240Pu 1.386" pitch 0.72 4.0242 MIX-COMP-THERM-005-006 18 wt% 240Pu 1.60" pitch 0.72 4.0243 MIX-COMP-THERM-005-007 18 wt% 240Pu 1.70" pitch 0.72 4.0244 LEU-COMP-THERM-001

-001 1 Cluster 2.35 -245 LEU-COMP-THERM-00 1-002 3 Clusters, Separation 11.92 cm 2.35 -246 LEU-COMP-THERM-001

-003 3 Clusters, Separation 8.41 cm 2.35 -247 LEU-COMP-THERM-00 1-004 3 Clusters, Separation 10.05 cm 2.35 -248 LEU-COMP-THERM-001-005 3 Clusters, Separation 6.39 cm 2.35 -249 LEU-COMP-THERM-00 1-006 3 Clusters, Separation 9.01 cm 2.35 -250 LEU-COMP-THERM-00 1-007 3 Clusters, Separation 4.46 2.35 -251 LEU-COMP-THERM-002-001 I Cluster, 1Ox 11.51 4.306 -252 LEU-COMP-THERM-002-002 1 Cluster, 9x 13.35 4.306 -253 LEU-COMP-THERM-002-003 I Cluster, 8x16.37 4.306 -254 LEU-COMP-THERM-002-005 3 Clusters, Separation 7.11 cm 4.306 -255 LEU-COMP-THERM-003-001 I Cluster, 614.4 Rods, Gd water impurity 2.35 -256 LEU-COMP-THERM-003-002 1 Cluster, 529.3 Rods 2.35 -257 LEU-COMP-THERM-003-003 I Cluster, 523.9 Rods 2.35 -258 LEU-COMP-THERM-003-004 1 Cluster, 525.3 Rods 2.35 -259 LEU-COMP-THERM-003-005 I Cluster, 595.4 Rods 2.35 -260 LEU-COMP-THERM-003-006 1 Cluster, 485.8 Rods 2.35 -261 LEU-COMP-THERM-003-007 I Cluster, 523.8 Rods 2.35 -262 LEU-COMP-THERM-003-008 1 Cluster, 505.4 Rods 2.35 -263 LEU-COMP-THERM-003-009 4 Clusters, Separation 2.59 cm 2.35 -264 LEU-COMP-THERM-003-010 2 Clusters, Separation 1.68 cm 2.35 -265 LEU-COMP-THERM-003-011 4 Clusters, Separation 4.27 cm 2.35 -266 LEU-COMP-THERM-003-012 4 Clusters, Separation 5.95 cm 2.35 -267 LEU-COMP-THERM-003-013 4 Clusters, Separation 5.11 cm 2.35 -268 LEU-COMP-THERM-003-014 4 Clusters, Separation 6.66 cm 2.35 -269 LEU-COMP-THERM-003-015 4 Clusters, Separation 7.53 cm 2.35 -270 LEU-COMP-THERM-003-016 4 Clusters, Separation 9.00 cm 2.35 -271 LEU-COMP-THERM-003-017 4 Clusters, Separation 9.97 cm 2.35 -272 LEU-COMP-THERM-003-018 4 Clusters, Separation 11.45 cm 2.35 -273 LEU-COMP-THERM-003-019 4 Clusters, Separation 13.87 cm 2.35 -274 LEU-COMP-THERM-003-020 3 Clusters, Separation 9.88 cm 2.35 --275 LEU-COMP-THERM-003-021 3 Clusters, Separation 6.78 cm 2.35 -REPORT HI-2104790 B-16 Case Reference Identification U, wt% Put276 LEU-COMP-THERM-003-023 3 Clusters, Separation 6.176 cm 2.35 -277 LEU-COMP-THERM-004-001 1 Cluster, 225.8 Rods, Gd water impurity 4.306 -278 LEU-COMP-THERM-004-002 1 Cluster, 216.2 Rods 4.306 -279 LEU-COMP-THERM-004-003 1 Cluster, 216.6 Rods 4.306 -280 LEU-COMP-THERM-004-004 1 Cluster, 218.6 Rods 4.306 -281 LEU-COMP-THERM-004-005 I Cluster, 167.85 Rods 4.306 -282 LEU-COMP-THERM-004-006 1 Cluster, 203 Rods 4.306 -283 LEU-COMP-THERM-004-007 1 Cluster, 173.5 Rods 4.306 -284 LEU-COMP-THERM-004-008 2 Clusters, Separation 2.83 cm 4.306 -285 LEU-COMP-THERM-004-009 3 Clusters, Separation 12.27 cm 4.306 -286 LEU-COMP-THERM-004-011 3 Clusters, Separation 12.493 cm 4.306 -287 LEU-COMP-THERM-004-012 4 Clusters, Separation 4.72 cm 4.306 -288 LEU-COMP-THERM-004-013 4 Clusters, Separation 8.38 cm 4.306 -289 LEU-COMP-THERM-004-014 4 Clusters, Separation 10.86 cm 4.306 -290 LEU-COMP-THERM-004-015 4 Clusters, Separation 11.29 cm 4.306 -291 LEU-COMP-THERM-004-016 4 Clusters, Separation 12.02 cm 4.306 -292 LEU-COMP-THERM-004-017 4 Clusters, Separation 13.64 cm 4.306 -293 LEU-COMP-THERM-004-018 4 Clusters, Separation 14.98 cm 4.306 -294 LEU-COMP-THERM-004-019 4 Clusters, Separation 19.81 cm 4.306 -295 LEU-COMP-THERM-004-020 4 Clusters, Separation 8.50 cm 4.306 -296 LEU-COMP-THERM-006-001 19x19, Rod Pitch -1.849 cm 2.596 -297 LEU-COMP-THERM-006-002 20x20, Rod Pitch -1.849 cm 2.596 -298 LEU-COMP-THERM-006-003 21x21, Rod Pitch -1.849 cm 2.596 -299 LEU-COMP-THERM-006-004 17x17, Rod Pitch -1.956 cm 2.596 -300 LEU-COMP-THERM-006-005 18x18, Rod Pitch -1.956 cm 2.596 -301 LEU-COMP-THERM-006-006 19x 19, Rod Pitch -1.956 cm 2.596 -302 LEU-COMP-THERM-006-007 20x20, Rod Pitch -1.956 cm 2.596 -303 LEU-COMP-THERM-006-008 21x21, Rod Pitch -1.956 cm 2.596 -304 LEU-COMP-THERM-006-009 16x16, Rod Pitch -2.15 cm 2.596 -305 LEU-COMP-THERM-006-010 17x17, Rod Pitch -2.15 cm 2.596 -306 LEU-COMP-THERM-006-011 18x18, Rod Pitch -2.15 cm 2.596 -307 LEU-COMP-THERM-006-012 19x19, Rod Pitch -2.15 cm 2.596 -308 LEU-COMP-THERM-006-013 20x20, Rod Pitch -2.15 cm 2.596 -309 LEU-COMP-THERM-006-014 15x 15, Rod Pitch -2.293 cm 2.596 -310 LEU-COMP-THERM-006-015 16x16, Rod Pitch -2.293 cm 2.596 -311 LEU-COMP-THERM-006-016 17x17, Rod Pitch -2.293 cm 2.596 -312 LEU-COMP-THERM-006-017 18x18, Rod Pitch -2.293 cm 2.596 -313 LEU-COMP-THERM-006-018 19x19, Rod Pitch -2.293 cm 2.596 -314 LEU-COMP-THERM-008-001 Core XI, 1511 ppm 2.459 -315 LEU-COMP-THERM-008-002 Core XI, 1335.5 ppm 2.459 -REPORT HI-2104790 B-17 Case Reference Identification U, wt% Pu,316 LEU-COMP-THERM-008-003 Core XI, 1335.5 ppm 2.459317 LEU-COMP-THERM-008-004 Core XI, 1182 ppm, 36 Pyrex Rods 2.459318 LEU-COMP-THERM-008-005 Core XI, 1182 ppm, 36 Pyrex Rods 2.459319 LEU-COMP-THERM-008-006 Core XI, 1032.5 ppm, 72 Pyrex Rods 2.459320 LEU-COMP-THERM-008-007 Core XI, 1032.5 ppm, 72 Pyrex Rods 2.459321 LEU-COMP-THERM-008-008 Core XI, 794 ppm, 144 Pyrex Rods 2.459322 LEU-COMP-THERM-008-009 Core XI, 779 ppm, 144 Pyrex Rods 2.459323 LEU-COMP-THERM-008-010 Core XI, 1245 ppm, 72 Vicor Rods 2.459324 LEU-COMP-THERM-008-011 Core XI, 1384 ppm, 144 A1203 Rods 2.459325 LEU-COMP-THERM-008-012 Core XI, 1348 ppm, 36 A1203 Rods 2.459326 LEU-COMP-THERM-008-013 Core XI, 1348 ppm, 36 A1203 Rods 2.459327 LEU-COMP-THERM-008-014 Core XI, 1363 ppm, 72 A1203 Rods 2.459328 LEU-COMP-THERM-008-015 Core XI, 1362 ppm, 72 A1203 Rods 2.459329 LEU-COMP-THERM-008-016 Core XI, 1158 ppm 2.459330 LEU-COMP-THERM-008-017 Core XI, 921 ppm 2.459331 LEU-COMP-THERM-009-001 0% Boron Steel plates, dist. 0.245 cm 4.306332 LEU-COMP-THERM-009-002 0% Boron Steel plates, dist. 3.277 cm 4.306333 LEU-COMP-THERM-009-003 0% Boron Steel plates, dist. 0.428 cm 4.306334 LEU-COMP-THERM-009-004 0% Boron Steel plates, dist. 3.277 cm 4.306335 LEU-COMP-THERM-009-006 1.05% Boron Steel plates, dist. 3.277 cm 4.306 -336 LEU-COMP-THERM-009-008 1.62% Boron Steel plates, dist. 3.277 cm 4.306 -337 LEU-COMP-THERM-009-024 Al plates, dist. 0.105 cm 4.306 -338 LEU-COMP-THERM-009-025 Al plates, dist. 3.277 cm 4.306 -339 LEU-COMP-THERM-009-026 Zircaloy-4 plates, dist. 0.078 cm 4.306 -340 LEU-COMP-THERM-009-027 Zircaloy-4 plates, dist. 3.277 cm 4.306 -341 LEU-COMP-THERM-010-001 Lead Reflector, 0 cm separation 4.306 -342 LEU-COMP-THERM-010-002 Lead Reflector, 0.660 cm separation 4.306 -343 LEU-COMP-THERM-010-003 Lead Reflector, 1.321 cm separation 4.306 -344 LEU-COMP-THERM-010-004 Lead Reflector, 5.405 cm separation 4.306 -345 LEU-COMP-THERM-010-009 Steel Reflector, 0 cm separation 4.306 -346 LEU-COMP-THERM-010-010 Steel Reflector, 0.660 cm separation 4.306 -347 LEU-COMP-THERM-010-01 1 Steel Reflector, 1.321 cm separation 4.306 -348 LEU-COMP-THERM-010-012 Steel Reflector, 2.616 cm separation 4.306 -349 LEU-COMP-THERM-010-013 Steel Reflector, 5.405 cm separation 4.306 -350 LEU-COMP-THERM-010-014 Steel Reflector, 0 cm separation 4.306 -351 LEU-COMP-THERM-010-015 Steel Reflector, 0.660 cm separation 4.306 -352 LEU-COMP-THERM-010-017 Steel Reflector, 1.956 cm separation 4.306 -353 LEU-COMP-THERM-010-020 Lead Reflector, 0 cm separation 4.306 -354 LEU-COMP-THERM-01 1-003 Core lilA 2.46 -355 LEU-COMP-THERM-01 1-005 Core IIIC 2.46 -REPORT HI-2104790 B-18 Case Reference Identification U, wt% Pu,356 LEU-COMP-THERM-01 1-006 Core 1111 2.46 -357 LEU-COMP-THERM-01 1-007 Core IIIE 2.46 -358 LEU-COMP-THERM-01 1-008 Core IIIF 2.46 -359 LEU-COMP-THERM-01 1-009 Core IIIG 2.46 -360 LEU-COMP-THERM-01 1-010 Core IV 2.46 -361 LEU-COMP-THERM-01 1-011 Core V 2.46 -362 LEU-COMP-THERM-01 1-012 Core VI 2.46 -363 LEU-COMP-THERM-01 1-013 Core VII 2.46 -364 LEU-COMP-THERM-0 11-0 14 Core VIII 2.46 -365 LEU.-COMP-THERM-012-001 0% Boron Steel plate, Gd water impurity 2.35 -366 LEU-COMP-THERM-012-002 1.1% Boron Steel plate 2.35 -367 LEU-COMP-THERM-012-003 1.6% Boron Steel plate 2.35 -368 LEU-COMP-THERM-012-004 Boral B plate 2.35369 LEU-COMP-THERM-012-005 Boral C plate 2.35370 LEU-COMP-THERM-012-006

Boroflex, 1.84 cm separation 2.35371 LEU-COMP-THERM-012-007
Boroflex, 1.73 cm separation 2.35372 LEU-COMP-THERM-013-001 Steel Reflector, 0% Boron Steel plate 4.306 -373 LEU-COMP-THERM-013-002 Steel Reflector, 1.1% Boron Steel plate 4.306 -374 LEU-COMP-THERM-01 3-004 Steel Reflector,
Boroflex, 8.37 cm separation 4.306 -375 LEU-COMP-THERM-014-002 Borated Water, 490 ppm 4.306 -376 LEU-COMP-THERM-014-006 Unborated Water 4.306 -377 LEU-COMP-THERM-014-007 Borated Water, 1030 ppm 4.306 -378 LEU-COMP-THERM-016-001 0% Boron Steel plates, dist. 0.645 cm 2.35 -379 LEU-COMP-THERM-016-002 0% Boron Steel plates, dist. 2.732 cm 2.35 -380 LEU-COMP-THERM-016-003 0% Boron Steel plates, dist. 4.042 cm 2.35 -381 LEU-COMP-THERM-016-004 0% Boron Steel plates, dist. 0.645 cm 2.35 -382 LEU-COMP-THERM-016-005 0% Boron Steel plates, dist. 4.042 cm 2.35 -383 LEU-COMP-THERM-016-006 0% Boron Steel plates, dist. 0.645 cm 2.35 -384 LEU-COMP-THERM-016-007 0% Boron Steel plates, dist. 4.042 cm 2.35 -385 LEU-COMP-THERM-016-008 1.05% Boron Steel plates, dist. 0.645 cm 2.35 -386 LEU-COMP-THERM-016-009 1.05% Boron Steel plates, dist. 4.042 cm 2.35 -387 LEU-COMP-THERM-016-010 1.62% Boron Steel plates, dist. 0.645 cm 2.35 -388 LEU-COMP-THERM-016-011 1.62% Boron Steel plates, dist. 4.042 cm 2.35 -389 LEU-COMP-THERM-016-012 Boral plates, dist. 0.645 cm 2.35 -390 LEU-COMP-THERM-016-013 Boral plates, dist. 4.442 cm 2.35 -391 LEU-COMP-THERM-016-014 Boral plates, dist. 0.645 cm 2.35 -392 LEU-COMP-THERM-016-028 Al plates, dist. 0.645 cm 2.35 -393 LEU-COMP-THERM-016-029 Al plates, dist. 4.042 cm 2.35 -394 LEU-COMP-THERM-016-030 Al plates, dist. 4.442 cm 2.35 -395 LEU-COMP-THERM-016-031 Zircaloy-4 plates, dist. 0.645 cm 2.35 -REPORT HI-2104790 B-19 Case Reference Identification U, wt% Pu,396 LEU-COMP-THERM-016-032 Zircaloy-4 plates, dist. 4.042 cm 2.35 -397 LEU-COMP-THERM-026-001 Hex, 621 Rods, Temperature
20. IC 4.92 -398 LEU-COMP-THERM-026-002 Hex, 889 Rods, Temperature 231.4C 4.92 -399 LEU-COMP-THERM-026-003 Hex, 1951 Rods, Temperature 19.3C 4.92 -400 LEU-COMP-THERM-026-004 Hex, 2791 Rods, Temperature 206.0C 4.92 -401 LEU-COMP-THERM-026-005 Hex, 325/680 Rods, Temperature 20.8C 4.92 -402 LEU-COMP-THERM-026-006 Hex, 325/912 Rods, Temperature 212.1C 4.92 -403 LEU-COMP-THERM-051-002 Core XIA 2.46 -404 LEU-COMP-THERM-051-004 Core XIC 2.46 -405 LEU-COMP-THERM-051-005 Core XID 2.46 -406 LEU-COMP-THERM-051-006 Core XIE 2.46 -407 LEU-COMP-THERM-051-007 Core XIF 2.46 -408 LEU-COMP-THERM-051-008 Core XIG 2.46 -409 LEU-COMP-THERM-05 1-011 Core XIIIA 2.46 -410 LEU-COMP-THERM-062-001 No Boron Steel plates 2.6 -411 LEU-COMP-THERM-062-002 0% Boron Steel plates, 3 mm, dist. 0 2.6 -412 LEU-COMP-THERM-062-003 0% Boron Steel plates, 6 mm, dist. 0 2.6 -413 LEU-COMP-THERM-062-004 0% Boron Steel plates, 6 mm, dist. 0.5 2.6 -414 LEU-COMP-THERM-062-005 0% Boron Steel plates, 6 mm, dist. 1 2.6 -415 LEU-COMP-THERM-062-006 0.67% Boron Steel plates, 3 mm, dist. 0 2.6 -416 LEU-COMP-THERM-062-007 0.67% Boron Steel plates, 6 mm, dist. 0 2.6 -417 LEU-COMP-THERM-062-008 0.67% Boron Steel plates, 3 mm, dist. 0.5 2.6 -418 LEU-COMP-THERM-062-009 0.67% Boron Steel plates, 6 mm, dist. 0.5 2.6 -419 LEU-COMP-THERM-062-010 0.67% Boron Steel plates, 3 mm, dist. 1 2.6 -420 LEU-COMP-THERM-062-011 0.67% Boron Steel plates, 6 mm, dist. 1 2.6 -421 LEU-COMP-THERM-062-012 0.98% Boron Steel plates, 3 mm, dist. 0 2.6 -422 LEU-COMP-THERM-062-013 0.98% Boron Steel plates, 6 mm, dist. 0 2.6 -423 LEU-COMP-THERM-062-014 0.98% Boron Steel plates, 6 mm, dist. 0.5 2.6 -424 LEU-COMP-THERM-062-015 0.98% Boron Steel plates, 6 mm, dist. 1 2.6 -425 LEU-COMP-THERM-065-001 No Boron Steel plates 2.6 -426 LEU-COMP-THERM-065-002 0% Boron Steel plates, dist. 0 2.6 -427 LEU-COMP-THERM-065-003 0.67% Boron Steel plates, dist. 0 2.6 -428 LEU-COMP-THERM-065-004 0.98% Boron Steel plates, dist. 0 2.6 -429 LEU-COMP-THERM-065-005 No Boron Steel plates 2.6 -430 LEU-COMP-THERM-065-006 0% Boron Steel plates, dist. 0 2.6 -431 LEU-COMP-THERM-065-007 0% Boron Steel plates, dist. 0.5 2.6 -432 LEU-COMP-THERM-065-008 0% Boron Steel plates, dist. 0 2.6 -433 LEU-COMP-THERM-065-009 0% Boron Steel plates, dist. 0.5 2.6 -434 LEU-COMP-THERM-065-010 0.67% Boron Steel plates, dist. 0 2.6 -435 LEU-COMP-THERM-065-011 0.67% Boron Steel plates, dist. 0.5 2.6 -__jREPORT HI-2104790 B-20 Case Reference Identification U, wt% Pu,436 LEU-COMP-THERM-065-012 0.67% Boron Steel plates, dist. 0 2.6 -437 LEU-COMP-THERM-065-013 0.67% Boron Steel plates, dist. 0.5 2.6 -438 LEU-COMP-THERM-065-014 0.98% Boron Steel plates, dist. 0 2.6 -439 LEU-COMP-THERM-065-015 0.98% Boron Steel plates, dist. 0.5 2.6 -440 LEU-COMP-THERM-065-016 0.98% Boron Steel plates, dist. 0 2.6 -441 LEU-COMP-THERM-065-017 0.98% Boron Steel plates, dist. 0.5 2.6 -442 LEU-COMP-THERM-081-001 Otto Hahn, ZrB2 and B4C rods 5.423 -443 LEU-COMP-THERM-082-001 IPEN/MB-01 (580 pins) 4.3486 -444 LEU-COMP-THERM-082-002 IPEN/MB-01 (560 pins) 4.3486 -445 LEU-COMP-THERM-082-003 670 pins, A1203-B4C rods 4.3486 -446 LEU-COMP-THERM-082-004 672 pins, A1203-B4C rods 4.3486 -447 LEU-COMP-THERM-082-005 668 pins, A1203-B4C rods 4.3486 -448 LEU-COMP-THERM-082-006 668 pins, A1203-B4C rods 4.3486 -449 LEU-COMP-THERM-090-001 664 pins, 16 steel rods 4.3486 -450 LEU-COMP-THERM-090-002 662 pins, 18 steel rods 4.3486 -451 LEU-COMP-THERM-090-003 658 pins, 14 steel rods 4.3486 -452 LEU-COMP-THERM-090-004 660 pins, 12 steel rods 4.3486 -453 LEU-COMP-THERM-090-005 660 pins, 12 steel rods 4.3486 -454 LEU-COMP-THERM-090-006 661 pins, 17 steel rods 4.3486 -455 LEU-COMP-THERM-090-007 662 pins, 16 steel rods 4.3486 -456 LEU-COMP-THERM-090-008 634 pins, 12 steel rods 4.3486 -457 LEU-COMP-THERM-090-009 620 pins, 26 steel rods 4.3486 -458 LEU-COMP-THERM-091-001 668 pins, 0 steel rods, 4 Gd203 rods 4.3486 -459 LEU-COMP-THERM-091-002 648 pins, 0 steel rods, 8 Gd203 rods 4.3486 -460 LEU-COMP-THERM-091-003 672 pins, 0 steel rods, 4 Gd203 rods 4.3486 -461 LEU-COMP-THERM-091-004 646 pins, 4 steel rods, 4 Gd203 rods 4.3486 -462 LEU-COMP-THERM-091-005 656 pins, 4 steel rods, 4 Gd2O3 rods 4.3486 -463 LEU-COMP-THERM-091-006 664 pins, 4 steel rods, 2 Gd203 rods 4.3486 -464 LEU-COMP-THERM-091-007 670 pins, 2 steel rods, 2 Gd203 rods 4.3486 -465 LEU-COMP-THERM-091-008 664 pins, 2 steel rods, 2 Gd203 rods 4.3486 -466 LEU-COMP-THERM-091-009 656 pins, 0 steel rods, 2 Gd203 rods 4.3486 -467 MIX-COMP-THERM-004-001 23x23, 1.825 cm pitch 0.72 3.01468 MIX-COMP-THERM-004-002 23x23, 1.825 cm pitch 0.72 3.011469 MIX-COMP-THERM-004-003 23x23, 1.825 cm pitch 0.72 3.011470 MIX-COMP-THERM-004-004 21x21, 1.956 cm pitch 0.72 3.011471 MIX-COMP-THERM-004-005 21x21, 1.956 cm pitch 0.72 3.011472 MIX-COMP-THERM-004-006 21x21, 1.956 cm pitch 0.72 3.011473 MIX-COMP-THERM-004-007 20x20, 2.225 cm pitch 0.72 3.011REPORT HI-2104790 B-21 Case Reference Identification U, wt% Pu,474 MIX-COMP-THERM-004-008 20x20, 2.225 cm pitch 0.72 3.01475 MIX-COMP-THERM-004-009 20x20, 2.225 cm pitch 0.72 3.011476 MIX-COMP-THERM-004-010 21x21, 2.474 cm pitch 0.72 3.011477 MIX-COMP-THERM-004-011 21x21, 2.474 cm pitch 0.72 3.011478 MIX-COMP-THERM-006-007 8 wt% 240Pu 1.05" pitch, Al Rods 0.72 2.0479 MIX-COMP-THERM-006-013 8 wt% 240Pu 1.05" pitch, B4 Rods 0.72 2.0480 MIX-COMP-THERM-006-014 8 wt% 240Pu 1.05" pitch, B3 Rods 0.72 2.0481 MIX-COMP-THERM-006-015 8 wt% 240Pu 1.05" pitch, B2 Rods 0.72 2.0482 MIX-COMP-THERM-006-016 8 wt% 240Pu 1.05" pitch, BI Rods 0.72 2.0483 MIX-COMP-THERM-006-017 8 wt% 240Pu 1.05" pitch, AI+Cd Rods 0.72 2.0484 MIX-COMP-THERM-006-023 8 wt% 240Pu 1.05" pitch, B4+Cd Rods 0.72 2.0485 MIX-COMP-THERM-006-024 8 wt% 240Pu 1.05" pitch, B3+Cd Rods 0.72 2.0486 MIX-COMP-THERM-006-025 8 wt% 240Pu 1.05" pitch, B2+Cd Rods 0.72 2.0487 MIX-COMP-THERM-006-026 8 wt% 240Pu 1.05" pitch, B 1+Cd Rods 0.72 2.0488 MIX-COMP-THERM-006-027 8 wt% 240Pu 1.05" pitch, Air+Cd Rods 0.72 2.0489 MIX-COMP-THERM-006-028 8 wt% 240Pu 1.05" pitch, H20+Cd Rods 0.72 2.0490 MIX-COMP-THERM-006-029 8 wt% 240Pu 1.32" pitch, Al Rods 0.72 2.0491 MIX-COMP-THERM-006-035 8 wt% 240Pu 1.32" pitch, B4 Rods 0.72 2.0492 MIX-COMP-THERM-006-036 8 wt% 240Pu 1.32" pitch, B3 Rods 0.72 2.0493 MIX-COMP-THERM-006-037 8 wt% 240Pu 1.32" pitch, B2 Rods 0.72 2.0494 MIX-COMP-THERM-006-038 8 wt% 240Pu 1.32" pitch, B I Rods 0.72 2.0495 MIX-COMP-THERM-006-039 8 wt% 240Pu 1.32" pitch, AI+Cd Rods 0.72 2.0496 MIX-COMP-THERM-006-045 8 wt% 240Pu 1.32" pitch, B4+Cd Rods 0.72 2.0497 MIX-COMP-THERM-006-046 8 wt% 240Pu 1.32" pitch, B3+Cd Rods 0.72 2.0498 MIX-COMP-THERM-006-047 8 wt% 240Pu 1.32" pitch, B2+Cd Rods 0.72 2.0499 MIX-COMP-THERM-006-048 8 wt% 240Pu 1.32" pitch, B 1 +Cd Rods 0.72 2.0500 MIX-COMP-THERM-006-049 8 wt% 240Pu 1.32" pitch, Air+Cd Rods 0.72 2.0501 MIX-COMP-THERM-006-050 8 wt%/o 240Pu 1.32" pitch, H20+Cd Rods 0.72 2.0502 MIX-COMP-THERM-007-005 16 wt%/o 240Pu 1.386" pitch 0.72 2.0503 MIX-COMP-THERM-007-006 16 wt% 240Pu 1.05" pitch, Al Rods 0.72 2.0504 MIX-COMP-THERM-007-012 16 wt% 240Pu 1.05" pitch, B4 Rods 0.72 2.0505 MIX-COMP-THERM-007-013 16 wt% 240Pu 1.05" pitch, B3 Rods 0.72 2.0506 MIX-COMP-THERM-007-014 16 wt% 240Pu 1.05" pitch, B2 Rods 0.72 2.0507 MIX-COMP-THERM-007-015 16 wt% 240Pu 1.05" pitch, B I Rods 0.72 2.0508 MIX-COMP-THERM-007-016 16 wt% 240Pu 1.05" pitch, AI+Cd Rods 0.72 2.0509 MIX-COMP-THERM-007-022 16 wt% 240Pu 1.05" pitch, B4+Cd Rods 0.72 2.0510 MIX-COMP-THERM-007-023 16 wt% 240Pu 1.05" pitch, B3+Cd Rods 0.72 2.0511 MIX-COMP-THERM-007-024 16 wt% 240Pu 1.05" pitch, B2+Cd Rods 0.72 2.0512 MIX-COMP-THERM-007-025 16 wt% 240Pu 1.05" pitch, BI+Cd Rods 0.72 2.0513 MIX-COMP-THERM-007-026 16 wt% 240Pu 1.05" pitch, Air+Cd Rods 0.72 2.0 IREPORT HI-2104790 B-22 Case Reference Identification U, wt% Pu,514 MIX-COMP-THERM-007-027 16 wt% 240Pu 1.05" pitch, H20+Cd Rods 0.72 2.0515 MIX-COMP-THERM-008-007 24 wt% 240Pu 1.05" pitch, Al Rods 0.72 2.0516 MIX-COMP-THERM-008-013 24 wt% 240Pu 1.05" pitch, B4 Rods 0.72 2.0517 MIX-COMP-THERM-008-014 24 wt% 240Pu 1.05" pitch, B3 Rods 0.72 2.0518 MIX-COMP-THERM-008-015 24 wt% 240Pu 1.05" pitch, B2 Rods 0.72 2.0519 MIX-COMP-THERM-008-016 24 wt% 240Pu 1.05" pitch, B I Rods 0.72 2.0520 MIX-COMP-THERM-008-017 24 wt% 240Pu 1.05" pitch, AL+Cd Rods 0.72 2.0521 MIX-COMP-THERM-008-023 24 wt% 240Pu 1.05" pitch, B4+Cd Rods 0.72 2.0522 MIX-COMP-THERM-008-024 24 wt% 240Pu 1.05" pitch, B3+Cd Rods 0.72 2.0523 MIX-COMP-THERM-008-025 24 wt% 240Pu 1.05" pitch, B2+Cd Rods 0.72 2.0524 MIX-COMP-THERM-008-026 24 wt% 240Pu 1.05" pitch, B 1 +Cd Rods 0.72 2.0525 MIX-COMP-THERM-008-027 24 wt% 240Pu 1.05" pitch, Air+Cd Rods 0.72 2.0526 MIX-COMP-THERM-008-028 24 wt% 240Pu 1.05" pitch, H20+Cd Rods 0.72 2.0527 MIX-COMP-THERM-009-001 8 wt% 240Pu 0.55" pitch 0.16 1.5528 MIX-COMP-THERM-009-002 8 wt% 240Pu 0.60" pitch 0.16 1.5529 MIX-COMP-THERM-009-003 8 wt% 240Pu 0.71" pitch 0.16 1.5530 MIX-COMP-THERM-009-004 8 wt% 240Pu 0.80" pitch 0.16 1.5531 MIX-COMP-THERM-009-005 8 wt% 240Pu 0.90" pitch 0.16 1.5532 MIX-COMP-THERM-009-006 8 wt% 240Pu 0.93" pitch 0.16 1.5REPORT HI-2 104790 B-23REPORT HI-2104790 B-23 Appendix CBenchmark of MCNP5-1.51 with ENDF[B-V(total number of pages: 27 including this page)REPORT HI-2104790 C-1 C.1 Introduction This Appendix presents the analysis of the validation results for MCNP5-1.51 code and includesthe results of the calculations, normality test, the detailed statistical trending
analysis, calculation bias and bias uncertainty for each distinct area of applicability of the parameters of interest.

C.2 Computer Code Parameter DataThe computer code MCNP5-1.51

[C. 1] is the continuous energy Monte Carlo codes and treats anarbitrary three-dimensional configuration of materials in geometric cells bounded by first- andsecond-degree surfaces and fourth-degree elliptical tori. Thermal neutrons are described by boththe free gas and S(a,3) models. All calculations were performed using the default data libraries provided with the code: the default continuous energy neutron transport data predominantly based on ENDF/B-V.

The list of ZAIDs that were used in the analysis is presented in Table C.2-1. The criticality source card was set to accumulate a total of 1.8 million neutron histories forevery individual run. The neutrons start from an arbitrary distribution, causing a generally verylarge variance of results from the first cycles in comparison with the following cycles. Therefore, the results from the first 50 cycles were skipped when calculating the average keff. The calculated keff values have associated uncertainties due to the statistical nature of the Monte Carlo codes.C.3 Analysis of MCNP5-1.51 Validation ResultsC.3.1. Calculational ResultsThe calculation results for the 156 HTC critical experiments and for the 135 selected criticalexperiments described in Appendix B are presented and discussed in this section.

The calculation results are summarized by grouping the experiments in terms of the categories as set forth inAppendix B. Calculation

results, including keff, Ucaic-, and EALF, measurement uncertainties (trp) and the calculation and measurement combined uncertainty (ar) are shown in Table C.3-1through Table C.3-5.Figure C.3-1 and C.3-2 are histograms showing the frequency of calculated k, and EALF for all291 benchmarks.

The nominal calculated k, values range from A. The EALFresults values show a range betweenDescriptive statistics for the different group of experiments is summarized in Table C.3-6.C.3.2. Normality TestIn order to assess the normality assumption, Shapiro and Wilk [5] test has been used for groupswith fewer than 50 samples while the Pearson's chi-square (X2) test [4] has been used for sampleslarger than 20 samples.

The tests are applied to the group of experiments in terms of thecategories as set forth in Appendix B.For the Shapiro and Wilk test, Table C.3-7 shows the computed Wtest value, and W value thatcan be obtained for the number of experiments from [5] to accept the normality hypothesis.

If WREPORT HI-2104790 C-2 is less than the test statistic, Wtest, then the data is considered normally distributed.

For the X2test, it is concluded normal for xf < n, where n is a number of bins for the group of experiments.

The probability Pd(,2 > ,0 2) of obtaining a value of 2 _> in an experiment with d degreesof freedom to confirm quantitatively that the agreement is satisfactory was taken or interpolated, if necessary, from Appendix D in Reference

[4]. Thus, if Pd(k2 >_ 1o2) is large, the obtained andexpected distributions are consistent; if it is small, they probably disagree.

In particular, ifPd(,k2 > i02) is less than 5%, we say that the disagreement is significant and reject the assumeddistributions at the 5% level. If it is less than 1%, the disagreement is called highly significant, and we reject the assumed distributions at the 1% level.As it is shown in Table C.3-7, all cases except Phase 1 test normal. Nevertheless, the group withall 291 experiments shows an agreement with the assumed normal distribution with theprobability Pd = 7.36%.C.3.3. Trending AnalysisTrends are determined through the use of regression fits to the calculated results.

The equations used to identify trends are given below:Y(x) = a + bx (7-1)1L 1X fVX 1'1 'X y1\(7-2)b= ( Zxi- X- YZ T)(7-3)(7-4)The squared term of the linear correlation factor r defined below (from Reference

[5]) is used toquantitatively measure the degree to which a linear relationship exist between two variables.

1-U2 (xi -:0) (yi -Y)r =(7-5)The closer r2 approaches the value of 1, the better the fit of the data to the linear equation.

Amore quantitative measure of the fit can be found by using Appendix C in Reference

[4]. Theinterpolation was applied, if necessary.

For any given observed value ro, PN(Irl -- Irol) is theprobability that N measurements of two uncorrelated variables would give a coefficient r as largeas ro. Thus, if we obtain a coefficient ro for which PN(IrI > Irol) is small, it is correspondingly unlikely that our variables are uncorrelated; that is, a correlation indicated.

In particular, ifREPORT HI-2104790 C-3 PN(IrI > Irol) < 5%, the correlation is called significant; if it is less than 1%, the correlation iscalled highly significant.

The validation results are analyzed by grouping the experiments in terms of the categories as setforth in Appendix B. Independent variables used in the trending analysis by group, correlation coefficients and trending analysis results are summarized in Table C.3-8. The linear regression equations for the independent parameter with the significant correlation of keff were presented inTable C.3-8.C.3.4. Bias and Bias Uncertainty In this section, benchmark results are analyzed using the statistical method described in section2.2.The first step is to evaluate whether the four HTC phases and selected ex eriments, should bereduced to a single set. Th ofe Phase 1 data set is , thof the Phase 2 data set is , the mean ke of the Phase 3 data set is, the mean keff of the Phase 4 data set is and the mean keff of theselected ex eriments data is .The maximum difference between the meansis just ý which is less than the uncertainty.

These sets are water moderated uranium ormixed plutonium-uranium dioxide lattices.

The addition of a absorber rods, separator plates orreflector plates is not introducing a significant increase in the ability to calculate keff. The Phase 1through Phase 4 sets and the selected experiments are considered one large set of 291experiments from now on.The analysis of the correlation coefficient in Table C.3-8 (combined set) and the plot of datatrend (Figure C.3-3) show that there is a significant trend a function of the rod pitch. This isdiscussed in the Section C.3.5.2.The total bias (systematic error or mean of the deviation from a keff of exactly 1.000) of theMCNP5-1.51 code is shown in the table belowCalculational Bias of the MCNP5-1.51 codeDescription Total Bias Bias Uncertainty HTC and Selected Experiments C.3.5. Applicability of MCNP5-1.51 Validation ResultsThis subsection contains a more detailed evaluation of the set of critical experiments.

Regarding the selected experiments, the following subjects are discussed:

" Neutron absorber and neutron reflector materials

" Fuel rod pitch trend" Neutron absorber geometryREPORT HI-2104790 C-4

  • Fuel bumup" Unborated and borated water.The general focus is to justify that using the full set of critical experiments is appropriate.

Insome cases, subsets of full set of experiments are established.

For those subsets, statistical evaluations are performed to determine bias, bias uncertainty, normality and trends. Trends areevaluated for fuel rod outer diameter, fuel rod pitch, fuel density, and EALF.C.3.5. 1. Neutron Absorber and Neutron Reflector Materials The HTC and Selected Experiments consider the following neutron absorbers and reflectors:

  • Absorbers o Boron, in the form of soluble boron in the water, boron in solid form (B4C), andboron in borated steelo Soluble gadolinium in watero Cadmium" Reflectors o Steelo Leado WaterSome typical configurations do not contain gadolinium or cadmium neutron absorbers or leadreflectors.

To verify that including those materials does not have a significant effect on theresults of the benchmarking

analyses, a subset without those experiments containing thosematerials was analyzed.

The comparison with the full set is presented in Table C.3-9 and showsno significant differences when those materials are excluded.

However, in both cases, asignificant trend is observed, as a function of the rod pitch in the experiment.

This is discussed inthe next section.C.3.5.2.

Fuel Rod Pitch TrendTo better understand the observed rod pitch trend, the results for all 291 experiments are shownin Figure C.3-4 as a function of rod pitch. It appears that the trend is due to the experiments athigher rod pitch value (> 2 cm), which consistently show keff values well above 1.0. To evaluatethe impact of those experiments at larger rod pitches, the Table C.3-10 shows a comparison ofresults with and without those experiments.

When results above 2 cm rod pitch are excluded, aslightly higher absolute bias is observed, in this case with a lower uncertainty, and no significant rod pitch trend. Based on those results it could be concluded that the trend is only caused by theexperiments at higher rod pitch values. To ensure that a potential trend would not be ignored, allfollowing evaluations are performed for the two conditions used above, i.e. for all rod pitchvalues, and for experiments with rod pitch values limited to no more than 2 cm.C.3.5.3.

Absorber GeometryREPORT HI-2 104790 C-5REPORT HI-2104790 C-5 The criticality experiments analyzed in this report include experiments with Boron in the form ofplates, absorber rods and soluble boron in water. No trend relating to these experiments isobserved.

C.3.5.4.

Fuel BurnupThe full set of critical experiments contains experiments with fresh U02 fuel, with simulated spent fuel (37.5 GWd/MTU),

and MOX fuel with Pu content between 2 and 20%, which is evenhigher than typically found in spent fuel. The experiments are therefore reasonably representative of burned fuel at different burnup levels. To verify that the experiments cover theburnup range sufficiently, the experiments are subdivided into fresh U02 fuel, HTC experiments and MOX experiments, and compared to the results of the entire set. The comparison is shown inTable C.3-1 1. The comparison shows no significant differences between the entire set and theU02 and HTC subsets, but for MOX the bias is now positive (i.e. truncated bias of 0.0), with alarger uncertainty, and some trends. However, this is based on relatively small sets ofexperiments.

Bias values are comparable between sets with and without rod pitch values above 2cm, with a maximum absolute value of .C.3.5.5.

Unborated and Borated WaterThe full set of critical experiments contains both experiments with and without soluble boron.The entire set of analyses shows no significant trend when analyzed as a function of the solubleboron level. Nevertheless, sets with and without soluble boron are analyzed and compared to thefull set that contains all experiments.

The results are shown in Table C.3-12. Similar to theprevious subsection, the comparison shows no significant differences between those subsets.Bias values are comparable between sets with and without rod pitch values above 2 cm, with amaximum absolute value of .C.4 SummaryA set of 291 critical experiments has been selected and has been used for the validation of theHoltec International criticality safety methodology.

The similarity between the chosenexperiments and the actual systems has been based on a set of screening criteria as is stated in theNUREG/CR-6698

[5]. Experiments have been categorized by common features as Phase 1through Phase 4 and selected experiments and parameterized by key variables such as latticepitch / assembly pitch, absorber solution concentration, number of fuel rods, rod outer diameter, fuel density, screen array distance, fuel enrichment and EALF. Benchmark calculations havebeen performed using the Monte Carlo code MCNP5-1.51.

It was determined that Phase 1through Phase 4 and selected experiments are in sufficient agreement that this sets are lumpedtogether as a single set of 291 experiments.

The bias and bias uncertainty are presented in sectionC.3.4. The applicability of validation results is considered in section C.3.5.The range of key parameters for the design application, benchmarks and validated AOA aresummarized in Table C.3-13. A point by point comparison between design application andbenchmarks shows that the experimental range covers all the parameters.

The soluble boronREPORT HI-2104790 C-6 concentration is extrapolated generously since 1°B is a 1/v absorber (as permitted on Table 2.3 of[5]).As for the fuel density, Table 2.3 of Reference

[5] states there is "no requirement" and that"experiments should be as close to the desired concentration as possible".

Since the experiment fuel density is 9.2 -10.4 g/cm3 and the design application one is around 10.0 -10.7 g/cm3, it isconsidered that the values are very close so the validated AOA covers the design application range.The fuel enrichment can be up to 5%. The experiments used go up to 5.74 wt% 235U. Therefore, it is considered that the validated AOA covers the design application range.C.5 References

[C. 1] "MCNP -A General Monte Carlo N-Particle Transport Code, Version 5"; Los AlamosNational Laboratory, LA-UR-03-1987 (Revised 2/1/2008).

REPORT HI-2104790 C-7 Table C.2-1 MCNP5-1.51 ZAIDs Used for EachNuclideNuclide ZAIID1H 1001.50c1°B 5010.50c11B 5011.55cC 6000.50c1IN 7014.50c160 8016.50c23Na 11023.51c Mg 12000.50c 27A1 13027.50c Si 14000.51c 31p 15031.50c 32s 16032.51c Ca 20000.51c Ti 22000.50c Cr 24000.50c 55Mn 25055.51c Fe 26000.55c 59Co 27059.50c Ni 28000.50c Cu 29000.50c Zn 30000.40c Zr 40000.56c Mo 42000.50c Cd 48000.50c Sn 50000.40c Gd 64000.35c Pb 82000.50c 234U 92234.50c 235u 92235.50c 236u 92236.50c 238 U 92238.50c 238Pu 94238.50c 239pu 94239.50c 240pu 94240.50c 241pu 94241.50c 242Pu 94242.50c 241, mln 95241.50c REPORT HI-2104790 C-8 Table C.3-1 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase1 Critical Experiments:

Water-Moderated and Reflected ArraysCase Evaluation Identification 1 MIX-COMP-THERM-HTC-001 2 MIX-COMP-THERM-HTC-002 3 MIX-COMP-THERM-HTC-003 4 MIX-COMP-THERM-HTC-004 5 MIX-COMP-THERM-HTC-005 6 MIX-COMP-THERM-HTC-006 7 MIX-COMP-THERM-HTC-007 8 MIX-COMP-THERM-HTC-008 9 MIX-COMP-THERM-HTC-009 10 MIX-COMP-THERM-HTC-010 11 MIX-COMP-THERM-HTC-0 1112 MIX-COMP-THERM-HTC-012 13 MIX-COMP-THERM-HTC-013 14 MIX-COMP-THERM-HTC-014 15 MIX-COMP-THERM-HTC-015 16 MIX-COMP-THERM-HTC-016 17 MIX-COMP-THERM-HTC-017 18 MIX-COMP-THERM-HTC-018 File- l~ri +oaci +/-Ox i !EALFname '~ cclt ~ --UU U --U UU U-REPORT HI-2 104790 C-9REPORT HI-2104790 C-9 Table C.3-2 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 2 CriticalEx riments:

Reflected Simple Arrays Moderated by Poisoned Water with Gadolinium or BoronFile-Case Evaluation Identification name kefi -0cai -Gexp -Gi EALF (eV)19 MIX-COMP-THERM-HTC-019 U UIU UM T20 MIX-COMP-THERM-HTC-020 UU21 MIX-COMP-THERM-HTC-021 UEUV22 MIX-COMP-THERM-HTC-022 U111U[U23 MIX-COMP-THERM-HTC-023 U = U U0 i24 MIX-COMP-THERM-HTC-024 U U U I25 MIX-COMP-THERM-HTC-025 U U U[U i26 MIX-COMP-THERM-HTC-026 U27 MIX-COMP-THERM-HTC-027 U U.U.U..28 MIX-COMP-THERM-HTC-028 U29 MIX-COMP-THERM-HTC-029 U30 MIX-COMP-THERM-HTC-030 U UU31 MIX-COMP-THERM-HTC-031 U32 MIX-COMP-THERM-HTC-032 U33 MIX-COMP-THERM-HTC-033 U UUU34 MIX-COMP-THERM-HTC-034 U U ..U.U.. .35 MIX-COMP-THERM-HTC-035 U36 MIX-COMP-THERM-HTC-036 U lA U U U37 MIX-COMP-THERM-HTC-037 U38 MIX-COMP-THERM-HTC-038 U39 MIX-COMP-THERM-HTC-039 U U.U.U..40 MIX-COMP-THERM-HTC-040 U U U.U.W..41 MIX-COMP-THERM-HTC-041 U42 MIX-COMP-THERM-HTC-042 U UUU-43 MIX-COMP-THERM-HTC-043 U UUU44 MIX-COMP-THERM-HTC-044 U[45 MIX-COMP-THERM-HTC-045 U UUU46 MIX-COMP-THERM-HTC-046 U11UU47 MIX-COMP-THERM-HTC-047 U U[U U i48 MIX-COMP-THERM-HTC-048 U U49 MIX-COMP-THERM-HTC-049 U U.U.U.U..

50 MIX-COMP-THERM-HTC-050 U U[ U W51 MIX-COMP-THERM-HTC-051 U U UUU52 MIX-COMP-THERM-HTC-052 U[U UU U53 MIX-COMP-THERM-HTC-053 U U UUW54 MIX-COMP-THERM-HTC-054 U U UUUW55 MIX-COMP-THERM-HTC-055 U U U UREPORT HI-2104790 C-10 File-Case Evaluation Identification name keff-i + gcalc-i + gexp + ai EALF (eV)56 MIX-COMP-THERM-HTC-056 U UUU57 MIX-COMP-THERM-HTC-057 U[U58 MIX-COMP-THERM-HTC-058 U U WUU59 MIX-COMP-THERM-HTC-059 U"J _"___"REPORT HI-2 104790 c-ilREPORT HI-2104790 C-I11 Table C.3-3 The MCNP5-1.51 Calculational Results and Measurements Uncertainties forPhase 3 Critical Experiments:

Pool StorageCase Evaluation Identification 60 MIX-COMP-THERM-HTC-060 61 MIX-COMP-THERM-HTC-061 62 MIX-COMP-THERM-HTC-062 63 MIX-COMP-THERM-HTC-063 64 MIX-COMP-THERM-HTC-064 65 MIX-COMP-THERM-HTC-065 66 MIX-COMP-THERM-HTC-066 67 MIX-COMP-THERM-HTC-067 68 MIX-COMP-THERM-HTC-068 69 MIX-COMP-THERM-HTC-069 70 MIX-COMP-THERM-HTC-070 71 MIX-COMP-THERM-HTC-071 72 MIX-COMP-THERM-HTC-072 73 MIX-COMP-THERM-HTC-073 74 MIX-COMP-THERM-HTC-074 75 MIX-COMP-THERM-HTC-075 76 MIX-COMP-THERM-HTC-076 77 MIX-COMP-THERM-HTC-077 78 MIX-COMP-THERM-HTC-078 79 MIX-COMP-THERM-HTC-079 80 MIX-COMP-THERM-HTC-080 81 MIX-COMP-THERM-HTC-081 82 MIX-COMP-THERM-HTC-082 83 MIX-COMP-THERM-HTC-083 84 MIX-COMP-THERM-HTC-084 85 MIX-COMP-THERM-HTC-085 File- ~, + ~ ~ aexp + a, EALF (eV)name-U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U ---U U --~-U U ---U U ---U U --= ----- ~----- ~----- ~------ ~----- ~REPORT HI-2 104790 c-i 2REPORT HI-2104790 C-12 Table C.3-4 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 4 CriticalExperiments:

Shipping CaskCase Evaluation Identification 86 MIX-COMP-THERM-HTC-086 87 MIX-COMP-THERM-HTC-087 88 MIX-COMP-THERM-HTC-088 89 MIX-COMP-THERM-HTC-089 90 MIX-COMP-THERM-HTC-090 91 MIX-COMP-THERM-HTC-091 92 MIX-COMP-THERM-HTC-092 93 MIX-COMP-THERM-HTC-093 94 MIX-COMP-THERM-HTC-094 95 MIX-COMP-THERM-HTC-095 96 MIX-COMP-THERM-HTC-096 97 MIX-COMP-THERM-HTC-097 98 MIX-COMP-THERM-HTC-098 99 MIX-COMP-THERM-HTC-099 100 MIX-COMP-THERM-HTC-100101 MIX-COMP-THERM-HTC-101102 MIX-COMP-THERM-HTC-102103 MIX-COMP-THERM-HTC-103File- k~ff, +/- aca!c~ +/- ~ + a, (eV)name-m -m m m--m m m--m m m---m m m-U -U U -~-U m m m--m m m-U -U U -~-m m m m-m -m m m-~ m m m-m -m m m--U U U -~--m m m--U U U -~--U U U -~---U U -~--m m -104MIX-COMP-THERM-HTC-104 I I I I II105MIX-COMP-THERM-HTC-105106 MIX-COMP-THERM-HTC-106107 MIX-COMP-THERM-HTC-107108 MIX-COMP-THERM-HTC-108 109 MIX-COMP-THERM-HTC-109110 MIX-COMP-THERM-HTC-I 10111 MIX-COMP-THERM-HTC-11112 MIX-COMP-THERM-HTC-1 12113 MIX-COMP-THERM-HTC-113114 MIX-COMP-THERM-HTC-1 14115 MIX-COMP-THERM-HTC-1 15116 MIX-COMP-THERM-HTC-116117 MIX-COMP-THERM-HTC-117---m m m--U U U ~--U U U ~--m m m-m m m-m m m-U U U U -~-m m m-U U U U -~-U U U U ~-U U U U -~U U U U -~U U U U118MIX-COMP-THERM-HTC-1 118I IN I II119 MIX-COMP-THERM-HTC-119120 MIX-COMP-THERM-HTC-120120 MIX-COMP-THERM-HTC-121121 MIX-COMP-THERM-HTC-121 U U U U U122 MIX-COMP-THERM-HTC-122 123 MIX-COMP-THERM-HTC-123 REPORT HI-2104790 C-13 CaseEvaluation Identification 124MIX-COMP-THERM-HTC-124125 MIX-COMP-THERM-HTC-125126 MIX-COMP-THERM-HTC-126127 MIX-COMP-THERM-HTC-127128 MIX-COMP-THERM-HTC-128129 MIX-COMP-THERM-HTC-129130 MIX-COMP-THERM-HTC-130131 MIX-COMP-THERM-HTC-131File-name kef-132u calc- G~x (y +1EALF JIMD(-COMP-ThERM-HTC-132 I I_32 ... .... .. .... ..... ... .2 --U133 MIX-COMP-THERM-HTC-133 134 MIX-COMP-THERM-HTC-134 135 MIX-COMP-THERM-HTC-135 136 MIX-COMP-THERM-HTC-136 137MIX-COMP-THERM-HTC-137IM 138MIX-COMP-THERM-HTC-138139 MIX-COMP-THERM-HTC-139 140 MIX-COMP-THERM-HTC-140141 MIX-COMP-THERM-HTC-141142 MIX-COMP-THERM-HTC-142143 MIX-COMP-THERM-HTC-143144 MIX-COMP-THERM-HTC-144 145 MIX-COMP-THERM-HTC-145146 MIX-COMP-THERM-HTC-146147 MIX-COMP-THERM-HTC-147148 MIX-COMP-THERM-HTC-148149 MIX-COMP-THERM-HTC-149150 MIX-COMP-THERM-HTC-150151 MIX-COMP-THERM-HTC-151152 MIX-COMP-THERM-HTC-152153 MIX-COMP-THERM-HTC-153 F-F-F-F--F--1F--F--F--1F--1F-1F--1F-1U U U ~ IU U -~ IU U U ~ IU U U ~ IU-UU U U -~ Im m m m I-m m m IU U U ~ IU U U ~ IU U U ~ IU U U -~ IU U U -~ IU U -~ IU U -~ IU U --~ Im -m I--- m IU -- m I-U -m Im U -m I--- m I--m IU --m I-m m I--m m I-m m I-m m I-m m I-m m I-m m I-m mU ~ U ~ I154MIX-COMP-THERM-HTC-154 155MIX-COMP-THERM-HTC-155156 MIX-COMP-THERM-HTC-156 REPORT HI-2 104790 C-14REPORT HI-2104790 C-14 Table C.3-5 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for SelectedCritical Experiments CaseEvaluation Identification File- klff. + ralc +/- 0expname+/- (yiEALF(eV)157 Core I__158 Core II159 Core III I160 Core IX161 Core X162 Core XI163 Core XII164 Core XIII I165 Core XIV I166 Core XV167 Core XVI I168 Core XVII169 Core XVIH I170 Core XIX I171 Core XX172 Core XXI I173 S-type Fuel, w/886 ppm B174 S-type Fuel, w/746 ppm B175 SO-type Fuel, w/1 156 ppm B176 Case 1 1337 ppm B177 Case 12 1899 ppmB B178 Water Moderator 0 gap179 Water Moderator 2.5 cm gap180 Water Moderator 5 cm gap181 Water Moderator 10 cm gap182 Steel Reflector, 1.321 cm separation 183 Steel Reflector, 2.616 cm separation 184 Steel Reflector, 3.912 cm separation 185 Steel Reflector, Infinite separation 186 Steel Reflector, 1.321 cm separation 187 Steel Reflector, 2.616 cm separation 188 Steel Reflector, 5.405 cm separation 189 Steel Reflector, Infinite separation 190 Steel Reflector, with Boral Sheets191 Lead Reflector, 0.55 cm sepn.192 Lead Reflector, 1.956 cm sepn.193 Lead Reflector, 5.405 cm sepn.194 Experiment 004/032 -no absorberREPORT 111-2104790 c-ISREPORT HI-2104790 C-15 CaseEvaluation Identification File-"CIM'nPke~ff-i-acalc-+ (Yexp-aiEALF195 Exp. 009 1.05% Boron Steel plates196 Exp. 009 1.62% Boron Steel plates197 Exp. 031 -Boral plates198 Experiment 214R -with flux traps199 Experiment 214V3 -with flux trap200 Case 173 -0 ppm B201 Case 177 -2550 ppm B202 MOX Fuel -Type 3.2 Exp. 21203 MOX Fuel -Type 3.2 Exp. 43204 MOX Fuel -Type 3.2 Exp. 13205 MOX Fuel -Type 3.2 Exp. 32206 Saxton Case 52 PuO2 0.52" pitch207 Saxton Case 52 U 0.52" pitch208 Saxton Case 56 PuO2 0.56" pitch209 Saxton Case 56 borated PuO2210 Saxton Case 56 U 0.56" pitch211 Saxton Case 79 PuO2 0.79" pitch212 Saxton Case 79 U 0.79" pitch213 0.700-in.

pitch 0 ppm B214 0.700-in.

pitch 688 ppm B215 0.870-in.

pitch 0 ppm B216 0.870-in.

pitch 1090 ppm B217 0.990-in.

pitch 0 ppm B218 0.990-in.

pitch 767 ppm B219 Saxton Case PuO2 0.735" pitch220 Saxton Case PuO2 1.04" pitch221 8 wt% 240Pu 0.80" pitch222 8 wt% 240Pu 0.93" pitch223 8 wt% 240Pu 1.05" pitch224 8 wt% 240Pu 1.143" pitch225 8 wt% 240Pu 1.32" pitch226 8 wt% 240Pu 1.386" pitch227 16 wt% 240Pu 0.93" pitch228 16 wt% 240Pu 1.05" pitch229 16 wt% 240Pu 1.143" pitch230 16 wt% 240Pu 1.32" pitch231 24 wt% 240Pu 0.80" pitch232 24 wt% 240Pu 0.93" pitch233 24 wt% 240Pu 1.05" pitch234 24 wt% 240Pu 1.143" pitchREPORT HI-2104790 C-16 I 1 7 1 T rCaseEvaluation Identification File-keff-i+ Gcalc.I Oexp-oiEALF(eV)L235 24 wt% 240Pu 1.32" pitch236 24 wt% 240Pu 1.386" pitch237 18 wt% 240Pu 0.85" pitch238 18 wt% 240Pu 0.93" pitch239 18 wt% 240Pu 1.05" pitch240 18 wt% 24OPu 1. 143" pitch241 18 wt% 240Pu 1.386" pitch242 18 wt% 240Pu 1.60" pitch243 18 wt% 240Pu 1.70" pitch317 Core XI, 1182 ppm, 36 Pyrex Rods318 Core XI, 1182 ppm, 36 Pyrex Rods319 Core XI, 1032.5 ppm, 72 Pyrex Rods320 Core XI, 1032.5 ppm, 72 Pyrex Rods321 Core XI, 794 ppm, 144 Pyrex Rods322 Core XI, 779 ppm, 144 Pyrex Rods323 Core XI, 1245 ppm, 72 Vicor Rods360 Core IV361 Core V362 Core VI363 Core VII364 Core VIII445 670 pins, A1203-B4C rods446 672 pins, A1203-B4C rods447 668 pins, A1203-B4C rods448 668 pins, A1203-B4C rods479 8 wt% 240Pu 1.05" pitch, B4 Rods480 8 wt% 240Pu 1.05" pitch, B3 Rods481 8 wt% 240Pu 1.05" pitch, B2 Rods482 8 wt% 240Pu 1.05" pitch, BI RodsUUUUUUU m Um -m miU ~ -MiUUUMIm -Urn'm -rn mlm -U MiUUUMIUUUMI~UUMiUUUMIUUUMIUUUMiUUUMIUUUMiUUUMI4848 wt% 240Pu 1.05" pitch, B4+CdP -i485 8 wt% 240Pu Rods105" pitch, B3+Cd U U U ..U -U486 ~RodsII486 I8 wt% 240Pu 1.05" pitch, B2+Cd~Rods4878 wt% 240Pu 1.05" pitch, B 1 +CdRods-i491 8 wt% 240Pu 1.32" pitch, B4 Rods492 8 wt% 240Pu 1.32" pitch, B3 Rods493 8 wt% 240Pu 1.32" pitch, B2 Rods494 , 8 wt% 240Pu 1.32" pitch, B 1 RodsU U ~ U UUUUUMIUUUUMIUUUUMIUUUUMIREPORT HI-2104790 C-17 Case Evaluation Identification File- ke i caic-- (exp 4-j EALFname ýfj(V496 8 wt% 240Pu 1.32" pitch, B4+Cd496 Rods497 8 wt% 240Pu 1.32" pitch, B3+CdRods498 8 wt% 240Pu 1.32" pitch, B2+Cd498 Rods5_49 16 wt% 240Pu 1.32" pitch, B4 I+CdRods504 16 wt% 240Pu 1.05" pitch, B4 Rods ____ m U ___505 16 wt% 240Pu 1.05" pitch, B3 Rods _____ m506 16 wt% 240Pu 1.05" pitch, B2 Rods ______ m507 16 wt% 240Pu 1.05" pitch, Bl Rods _____ M509 16 wt% 240Pu 1.05" pitch, B4+Cd U509 Rods510 16 wt% 240Pu 1.05" pitch, B3+Cd U510 Rods511 16 wt% 240Pu 1.05" pitch, B2+Cd U~Rods512 16wt%240Pu 1.05" pitch, Bl+CdoU5162 Rods516 24 wt% 240Pu 1.05" pitch, B4 Rods A 517 24 wt% 240Pu 1.05" pitch, B3 Rods_______

518 24 wt% 240Pu 1.05" pitch, B2 Rods ___ _ M519 24 wt% 240Pu 1.05" pitch, BI Rods _ _ _ _l521 24 wt% 240Pu 1.05" pitch, B4+Cd U521 Rods522 24 wt% 240Pu 1.05" pitch, B3+Cd U U U52 Rods -_ M_00 -o523 24 wt% 240Pu 1.05" pitch, B2+Cd U U _524 Rods524] 24 wt% 240Pu 1.05" pitch, Bl+Cýd _ _ __ R odsREPORT HI-2104790 C-i 8REPORT HI-2104790 C-18 Table C.3-6 Descriptive Statistics of the MCNP5-1.51 Calculational ResultsExperiment Description No. ofexp.keff rangeEALF (eV) rangePhase 1 18Phase 2 41Phase 3 26Phase 4 71Selected Experiments 135All experiments 291Table C.3-7 Normality Test Results for the MCNP5-1.51 calculations Experiment Description No. ofexp.Shapiro-Wilk Pearson's chi-square (x2)WtPePtwX2nPd(x2;d)NormalPhase 1 18Phase 2 41Phase 3 26Phase 4 71HTC Experiments 156Selected Experiments 135All experiments 291-_M I I=REPORT HI-2104790 C-19REPORT HI-2104790 C-19 Table C.3-8 Trending Analysis Results for the MCNP5-1.51 calculations Experiment Description No. ofexp.Correlated Parameter, xCorrelation Coefficient, 2rProbability, Pd(N;r)Correlation Regression

Equation, k(x)EALFPhase 1 18 PitchNumber of RodsEALFPitchPhase 2 41 Number of RodsGadolinium Conc.Boron Conc.EALFPhase 3 26 Water GapNumber of RodsEALFPhase 4 71 Water GapScreen Array DistanceEALFPitchSelected Rod ODExperiments Fuel DensityU Enrichment Pu Enrichment EALFPitchAll experiments 291 RodchRod ODFuel DensityII* Im II.IUIREPORT HI-2104790 C-20REPORT HI-2104790 C-20 Table C.3-9 Analysis of Neutron Absorbers and Reflector Materials for the MCNP5-1.

51calculations Experiment No. of Bias Normality Desriptint ex. o Bias Uncerta X2 Significant TrendsDescription exp. inty (Pd(X2;d))Allexperiments 291All exceptthose withGadolinium, 201 -Cadmium andLeadTable C.3-10 Analysis of Fuel Rod Pitch Trend for the MCNP5-1.

51 calculations Bias Normality Experiment Rod Bias Uncerta X2 Significant TrendsDescription Pitch inty (Pd(X2;d))All (291All totaln Lexperiments

<2 cm -total)All except All (201 m -those with total) -Gadolinium,

_Cadmium and <=2 cm -Lead (144 -total)REPORT HI-2 104790 C-2 1REPORT HI-2104790 C-21 Table C.3-11 Analysis of Fuel Burnup for the MCNP5-1.

51 calculations Experiment Description RodPitchBiasBiasUncertaintyNormality X2(Pd(X2;d))Significant Trends-I- 4 4All exceptthose withGadolinium, Cadmium andLeadtAll (201total)--<=2 cm(144total)-mAll (61Fresh U02 total)Fuel <=2 cm(52 total) ____All (85HTC total)Experiments

<=2 cm(82 total) _ __ ]All (55total)-LL-4 + 4MOXExperiments

<=2 cm(10 total)M MUtNote: Critical experiments with Gadolinium, Cadmium and Lead were excluded from all subsequent subsets.REPORT HI-2104790 C-22 Table C.3-12 Analysis of the Unborated and Borated Water for the MCNP5-1.51 calculations Experiment Description RodPitchBiasBiasUncertaintyNormality X2(Pd(X ;d))Significant TrendsAll exceptthose withGadolinium, Cadmium andLeadtAll (201total)-m<=2 cm(144total)ýImAll (149total)-I-All with FreshWateri i<=2 cm(94 total)E--71All (52All with total) _Borated Water <=2 cm(50 total) _tNote: Critical experiments with Gadolinium, Cadmium and Lead were excluded from all subsequent subsets.REPORT HI-2104790 C-23 Table C.3-13 Comparison of Key Parameters and Definition of Validated AOAParameter Design Benchmarks Validated Application Fissionable Material 235U, 239Pu, 241Pu 235U, 239Pu, 241Pu 235U, 239Pu, 241PuIsotopic Composition 235u/ut < 5.Owt% 1.57-5.74%

< 5wt%Pu/(U+Pu)

< 20wt% 1.104-20%

< 20wt%Physical Form U02 MOX U02 MOX U02 MOXFuel Density (g/cm3) 10.0- 10.7 9.2 -10.4 9.2 -10.7Moderator Material (coolant)

H H HPhysical Form H20 H20 H20Density (g/cm3) around 1.0 g/cm3 around 1.0 g/cm3 around 1.0 g/cm3Reflector Material H H HPhysical Form H20 H20 H20Density (g/cm3) around 1.0 g/cm3 around 1.0 g/cm3 around 1.0 g/cm3Interstitial Reflector MaterialPlate Steel or Lead Steel or Lead Steel or LeadAbsorber MaterialNone, Boron (89 -None, Boron (0 -Soluble None, Boron or 595 ppm) or 1000 ppm) orGadolinium Gadolinium (49.2 -Gadolinium (0 to199.7 ppm) 1000 ppm)Rods Boron Pyrex , Vicor' or BoronB-AlSeparating MaterialWater, B-SS, Water, B-SS, Boral Water, B-SS, BoralPlate Boral orCadmim oor Cadmium or CadmiumCadmiumGeometryLattice type Square Square, Triangle Square, Triangle1.26-1.47 Lattice Pitch (cm) (PWR) 0.968 to 4.318 0.968 to 4.318Lattce Ptch cm) 1.24 -1.88(BWR)ThermalNeutron Energy Thermal Thermal spectrum Thermal spectrumspectrum IIREPORT HI-2104790 C-24REPORT HI-2104790 C-24 Figure Proprietary Figure C.3-1 Frequency Chart for Calculated keff of the Selected 243 Benchmarks for theMCNP5-1.51 codeFigure Proprietary Figure C.3-2 Frequency Chart for Calculated EALF (eV) of the Selected 243 Benchmarks for theMCNP5-1.51 codeREPORT HI-2104790 C-25 Figure Proprietary Figure C.3-3 MCNP5-1.51 Calculated kff Values for Various Values of the Spectral Index (AllExperiments)

REPORT HI-2 104790 C-26REPORT HI-2104790 C-26 Figure Proprietary Figure C.3-4 MCNP5-1.51 Calculated k.f Values as a Function of Rod Pitch (All Experiments)

REPORT HI-2104790 C-27 Appendix DBenchmark of MCNP5-1.51 with ENDF/B-VII (total number of pages: 51 including this page)REPORT HI-2 104790 D- IREPORT HI-2104790 D-1 D.1 Introduction This Appendix presents the analysis of the validation results for MCNP5-1.51 code and includesthe results of the calculations, normality test, the detailed statistical trending

analysis, calculation bias and bias uncertainty for each distinct area of applicability of the parameters of interest.

D.2 Computer Code Parameter DataThe computer code MCNP5-1.51

[D. 1] is the continuous energy Monte Carlo codes and treats anarbitrary three-dimensional configuration of materials in geometric cells bounded by first- andsecond-degree surfaces and fourth-degree elliptical tori. Thermal neutrons are described by boththe free gas and S(a,3) models. All calculations were performed using the default data libraries provided with the code: the default continuous energy neutron transport data based on ENDF/B-VII. The list of ZAIDs that were used in the analysis is presented in Table D.2-1. The neutronsstart from an arbitrary distribution, causing a generally very large variance of results from thefirst cycles in comparison with the following cycles. Therefore, all MCNP5-1.51 calculations areperformed with 12,000 histories per cycle, 50 skipped cycles before averaging, and 100 cyclesthat are accumulated.

The calculated kff values have associated uncertainties due to the statistical nature of the Monte Carlo codes.D.3 Analysis of MCNP5-1.51 Validation ResultsD.3.1. Calculational ResultsThe calculation results for the 156 HTC critical experiments and for the 376 selected criticalexperiments described in Appendix B are presented and discussed in this section.

The calculation results are summarized by grouping the experiments in terms of the categories as set forth inAppendix B. Calculation

results, including keg-fc, orcatc-i and EALF, measurement uncertainties (u,,p) and the calculation and measurement combined uncertainty (au) are shown in Table D.3-1through Table D.3-5.Figure D.3-1 and D.3-2 are histograms showing the frequency of calculated ke and EALF for all532 benchmarks.

The nominal calculated k, values range from O i .The EALFresults values show a range betweenDescriptive statistics for the different group of experiments is summarized in Table D.3-6.D.3.2. Normality TestIn order to assess the normality assumption, Shapiro and Wilk [5] test has been used for groupswith fewer than 50 samples while the Pearson's chi-square (X2) test [4] has been used for sampleslarger than 20 samples.

The tests are applied to the group of experiments in terms of thecategories as set forth in Appendix B.For the Shapiro and Wilk test, Table D.3-7 shows the computed Wtest value, and W value thatcan be obtained for the number of experiments from [5] to accept the normality hypothesis.

If WREPORT HI-2104790 D-2 is less than the test statistic, Wtest, then the data is considered normally distributed.

For the X,test, it is concluded normal for x(2 < n, where n is a number of bins for the group of experiments.

The probability Pd(.k' > ,o02) of obtaining a value of,;2 > 102 in an experiment with d degreesof freedom to confirm quantitatively that the agreement is satisfactory was taken or interpolated, if necessary, from Appendix D in Reference

[4]. Thus, if Pd(&2 > 102) is large, the obtained andexpected distributions are consistent; if it is small, they probably disagree.

In particular, ifPd(X2 > ,k02) is less than 5%, we say that the disagreement is significant and reject the assumeddistributions at the 5% level. If it is less than 1%, the disagreement is called highly significant, and we reject the assumed distributions at the 1% level.D.3.3. Trending AnalysisTrends are determined through the use of regression fits to the calculated results.

The equations used to identify trends are given below:Y(x) = a + bx (7-1)(7-2)(7-3)A= zz #- t .-(7-4)The squared term of the linear correlation factor r defined below (from Reference

[5]) is used toquantitatively measure the degree to which a linear relationship exist between two variables.

1 (X, -2)(y,- y)2(7-5)The closer r2 approaches the value of 1, the better the fit of the data to the linear equation.

Amore quantitative measure of the fit can be found by using Appendix C in Reference

[4]. Theinterpolation was applied, if necessary.

For any given observed value ro, PN(IrI >- Irol) is theprobability that N measurements of two uncorrelated variables would give a coefficient r as largeas ro. Thus, if we obtain a coefficient ro for which PN(IrJ > Irol) is small, it is correspondingly unlikely that our variables are uncorrelated; that is, a correlation indicated.

In particular, ifPN(IrI -- Irol) 5 5%, the correlation is called significant; if it is less than 1%, the correlation iscalled highly significant.

REPORT HI-2 104790 D-3REPORT HI-2104790 D-3 The validation results are analyzed for the group of all experiments.

Independent variables usedin the trending

analysis, correlation coefficients and trending analysis results are summarized inTable D.3-8.D.3.4. Bias and Bias Uncertainty In this section, benchmark results are analyzed using the statistical method described in section2.2.The first step is to evaluate whether the HTC experiments and selected ex eriments, should bereduced to a single set. The mean, of the HTC data set is and the mean kffof the selected experiments data is .The difference between the means is just0.0010 which is less than the uncertainty.

These sets are water moderated uranium or mixedplutonium-uranium dioxide lattices.

The HTC sets of experiments and the selected experiments are considered one large set of 532 experiments from now on.The normality test in Table D.3-7 and in Figure D.3-1 shows that the data is not normallydistributed.

Therefore, the distribution free approach

[6] is used for all subsets with the rejectednormality distribution.

The lower tolerance limit with 95% probability and 95% confidence levelis determined for order data [6] and the difference between weighted average keff and this lowertolerance limit is used to determine the bias uncertainty.

This is conservative since the data isclose to the normal distribution.

The distribution free bias uncertainty is also provided in allsubsequent tables for the subsets with the rejected normality assumption.

The analysis of the correlation coefficient in Table D.3-8 and the plot of data trend (Figure D.3-3) show that there is not a clear trend in the data.The total bias (systematic error or mean of the deviation from a keff of exactly 1.000) of theMCNP5-1.51 code is shown in the table belowCalculational Bias of the MCNP5-1.51 codeDescription Total Bias Bias Uncertainty HTC and Selected Experiments D.3.5. Applicability of MCNP5-1.51 Validation ResultsThis subsection contains a more detailed evaluation of the set of critical experiments.

Regarding the selected experiments, the following subjects are discussed:

  • Neutron absorber and neutron reflector materials

" Neutron absorber geometry" Fuel burnup" Unborated and borated waterREPORT HI-2104790 D-4

  • Various Combinations of Fuel Bumup and Unborated/Borated Water.The general focus is to justify that using the full set of critical experiments is appropriate.

Insome cases, subsets of full set of experiments are established.

For those subsets, statistical evaluations are performed to determine bias, bias uncertainty, normality and trends. Trends areevaluated for fuel rod outer diameter, fuel rod pitch, fuel density, boron content, U or Puenrichment and EALF. To estimate a significance of observed trend, the residuals from the trendequation were tested for a normal distribution

[D.2]. If residuals are normally distributed thenthere is a significant trend, otherwise there is no linear trend as this violets the basic assumptions of linear regression.

For each significant linear correlation, the bias and bias uncertainty werecalculated as a function of the independent parameter.

D.3.5. 1. Neutron Absorber and Neutron Reflector Materials The HTC and Selected Experiments consider the following neutron absorbers and reflectors:

  • Absorbers o Boron, in the form of soluble boron in the water, boron in solid form (B4C), andboron in borated steel, Pyrex, Boroflex and borated aluminumo Soluble gadolinium in water and Gd203 rodso Cadmium" Reflectors o Steelo Leado WaterSome typical configurations do not contain gadolinium or cadmium neutron absorbers or leadreflectors.

To verify that including those materials does not have a significant effect on theresults of the benchmarking

analyses, a subset without those experiments containing thosematerials was analyzed.

In addition, according to recommendations of NUREG-6979

[B. 13], thefollowing HTC experiments were also excluded:

61, 65, 67, 86, 97, 98, 99, 102, 124, 135, and137. The comparison with the full set is presented in Table D.3-9 and shows no significant differences when those materials are excluded.

However, a significant correlation as a functionof EALF was determined by the residuals normality test. This correlation is presented in theFigure D.3-4. The bias and bias uncertainty as a function of the EALF were calculated for thistrend and shown in Table D.3-10, with a maximum absolute value of .D.3.5.2.

Absorber GeometryThe criticality experiments analyzed in this report include experiments with Boron in the form ofplates, absorber rods and soluble boron in water. No trend relating to these experiments isobserved.

D.3.5.3.

Fuel BurnupThe full set of critical experiments contains experiments with fresh U02 fuel, with simulated spent fuel (37.5 GWd/MTU),

and MOX fuel with Pu content between 1.5 and 20%, which isREPORT HI-2104790 D-5 even higher than typically found in spent fuel. The experiments are therefore reasonably representative of burned fuel at different burnup levels. To verify that the experiments cover theburnup range sufficiently, the experiments are subdivided into fresh U02 fuel and spent fuel withHTC and MOX experiments, and compared to the results of the entire set. The comparison isshown in Table D.3-1 1. The comparison shows no significant differences between the entire setand the fresh and spent fuel subsets.

However, in some cases, the correlations are observed.

Thesignificant trends as a function of EALF and Pu enrichment were determined in the spent fuelsubset by the residuals normality test. These correlations are presented in the Figure D.3-5 andFigure D.3-6. The bias and bias uncertainty as a function of the EALF and Pu enrichment werecalculated for these trends and shown in Table D.3-12, with a maximum absolute value of -D.3.5.4.

Unborated and Borated WaterThe full set of critical experiments contains both experiments with and without soluble boron.The entire set of analyses shows no significant trend when analyzed as a function of the solubleboron level. Nevertheless, sets with and without soluble boron are analyzed and compared to thefull set that contains all experiments.

The results are shown in Table D.3-13. Similar to theprevious subsection, the comparison shows no significant differences between those subsets.However, there are significant trends in the fresh water subset as a function of EALF and Uenrichment and in the borated water subset as a function of fuel density that were determined bythe residuals normality test. These correlations are presented in the Figure D.3-7 through FigureD.3-9. The bias and bias uncertainty as a function of the EALF, U enrichment and fuel densitywere calculated for these trends and shown in Table D.3-14, with a maximum absolute value of-D.3.5.5.

Various Combinations of Fuel Burnup and Unborated/Borated WaterTo perform more detailed evaluation of the set of critical experiments, the additional four subsetswith different combinations of fuel bumup and unborated/borated water were analyzed.

Theresults are shown in Table D.3-15. There are significant EALF trend in the subset of fresh U02fuel with fresh water, significant trends in the subset of spent fuel with fresh water as a functionof EALF and Pu enrichment and significant trends in the subset of spent fuel with borated wateras a function of rod OD and fuel density, that were determined by the residuals normality test.These correlations are presented in the Figure D.3-10 through Figure D.3-14.The bias and biasuncertainty as a function of the EALF, Pu enrichment, rod OD and fuel density were calculated for these trends and shown in Table D.3-16, with a maximum absolute value ofD.4 SummaryA set of 532 critical experiments has been selected and has been used for the validation of theHoltec International criticality safety methodology.

The similarity between the chosenexperiments and the actual systems has been based on a set of screening criteria as is stated in theNUREG/CR-6698

[5]. Experiments have been categorized by fuel burnup as fresh U02 fuel andspent fuel with HTC and MOX experiments or by unborated and borated water condition andREPORT HI-2104790 D-6 parameterized by key variables such as lattice pitch / assembly pitch, absorber solutionconcentration, number of fuel rods, rod outer diameter, fuel density, screen array distance, fuelenrichment and EALF. Benchmark calculations have been performed using the Monte Carlocode MCNP5-1.51.

It was determined that HTC experiments and selected experiments are insufficient agreement that this sets are lumped together as a single set of 532 experiments.

Thebias and bias uncertainty are presented in section D.3.4. The applicability of validation results isconsidered in section D.3.5.The range of key parameters for the design application, benchmarks and validated AOA aresummarized in Table D.3-17. A point by point comparison between design application andbenchmarks shows that the experimental range covers all the parameters.

The soluble boronconcentration is extrapolated generously since °B is a 1/v absorber (as permitted on Table 2.3 of[5]).As for the fuel density, Table 2.3 of Reference

[5] states there is "no requirement" and that"experiments should be as close to the desired concentration as possible".

Since the experiment fuel density is 9.2 -10.4 g/cm3 and the design application one is around 10.0 -10.7 g/cm3, it isconsidered that the values are very close so the validated AOA covers the design application range.The fuel enrichment can be up to 5%. The experiments used go up to 5.74 wt% 235U. Therefore, it is considered that the validated AOA covers the design application range.D.5 References

[D. 1] "MCNP -A General Monte Carlo N-Particle Transport Code, Version 5"; Los AlamosNational Laboratory, LA-IJR-03-1987 (Revised 2/1/2008).

[D.2] J. W. Barnes, "Statistical Analysis for Engineers and Scientists",

McGraw-Hill Inc., 1988REPORT HI-2104790 D-7 Table D.2-1 ZAIDs Used for Each NuclideMCNP5.1.51 MCNP5.1.51 MCNP5.1.51 Nuclide ZAD Nuclide ZAD Nuclide ZIZAID ZAID ZAID'H 1001.70c 48Ti 22048.70c

'l°Mo 42100.70c 2 H 1002.70c 49Ti 22049.70c 107Ag 47107.70c 4He 2004.70c 5°Ti 22050.70c 109Ag 47109.70c 10B 5010.70c 50Cr 24050.70c 106Cd 48106.70c "1B 5011.70c 52Cr 24052.70c 108Cd 48108.70c C 6000.70c 53Cr 24053.70c l"0Cd 48110.70c 14N 7014.70c 54Cr 24054.70c 111Cd 48111.70c 160 8016.70c 55Mn 25055.70c 112Cd 48112.70c 2°Ne 10020.42c 54Fe 26054.70c 113Cd 481 13.70c23Na 11023.70c 56Fe 26056.70c 114Cd 48114.70c 24Mg 12024.70c 57Fe 26057.70c 116Cd 48116.70c 25Mg 12025.70c 58Fe 26058.70c 1131n 49113.70c 26Mg 12026.70c 59Co 27059.70c 1151n 49115.70c 27A1 13027.70c 58Ni 28058.70c 12 Sn 50112.70c 28Si 14028.70c 6°Ni 28060.70c 114Sn 50114.70c 29Si 14029.70c 61Ni 28061.70c "15Sn 50115.70c 3°Si 14030.70c 62Ni 28062.70c 116Sn 50116.70c 31p 15031.70c 64Ni 28064.70c 17TSn 50117.70c 32s 16032.70c 63Cu 29063.70c 118Sn 50118.70c 36Ar 18036.70c 65Cu 29065.70c 119Sn 501 19.70c38Ar 18038.70c Zn 30000.70c 120Sn 50120.70c 4°Ar 18040.70c 90Zr 40090.70c 122Sn 50122.70c 39K 19039.70c 91Zr 40091.70c 124Sn 50124.70c 40K 19040.70c 92Zr 40092.70c 144Sm 62144.70c 41K 19041.70c 94Zr 40094.70c 147Sm 62147.70c 4°Ca 20040.70c 96Zr 40096.70c 148Sm 62148.70c 42Ca 20042.70c 93Nb 41093.70c 149sm 62149.70c 43Ca 20043.70c 92Mo 42092.70c

°50Sm 62150.70c 44Ca 20044.70c 94Mo 42094.70c I52Sm 62152.70c 46Ca 20046.70c 95Mo 42095.70c

'54Sm 62154.70c 48Ca 20048.70c 96Mo 42096.70c 152Gd 64152.70c 46Ti 22046.70c 97Mo 42097.70c 154Gd 64154.70c 47Ti 22047.70c 98Mo 42098.70c 155Gd 64155.70c REPORT HI-2104790 D-8 Nuclide MCNP5.1.51 MCNP5.1.51 Nlid MCNP5.1.51 ZAID ZAID ZAID156Gd 64156.70c 179Hf 72179.70c 236U 92236.70c 157Gd 64157.70c 180Hf 72180.70c 238u 92238.70c 158Gd 64158.70c 204Pb 82204.70c 238Pu 94238.70c 16°Gd 64160.70c 206Pb 82206.70c 239pu 94239.70c 174Hf 72174.70c 207Pb 82207.70c 240Pu 94240.70c 176Hf 72176.70c 208Pb 82208.70c 241Pu 94241.70c 177Hf 72177.70c 234u 92234.70c 242Pu 94242.70c 178Hf 72178.70c 235u 92235.70c 241AnM 95241.70c REPORT HI-2 104790 D-9REPORT HI-2104790 D-9 Table D.3-1 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase1 Critical Experiments:

Water-Moderated and Reflected ArraysFile-CaseEvaluation Identification kDif--I- Ccalc-i+/- Gexp+/- (yiEALF(All)I MIX-COMP-THERM-HTC-001 2 MIX-COMP-THERM-HTC-002 3 MIX-COMP-THERM-HTC-003 4 MIX-COMP-THERM-HTC-004 5 MIX-COMP-THERM-HTC-005 6 MIX-COMP-THERM-HTC-006 7 MIX-COMP-THERM-HTC-007 8 MIX-COMP-THERM-HTC-008 9 MIX-COMP-THERM-HTC-009 10 MIX-COMP-THERM-HTC-010 11 MIX-COMP-THERM-HTC-0 1112 MIX-COMP-THERM-HTC-012 13 MIX-COMP-THERM-HTC-013 14 MIX-COMP-THERM-HTC-014 15 MIX-COMP-THERM-HTC-015 16 MIX-COMP-THERM-HTC-016 17 MIX-COMP-THERM-HTC-017 18 MIX-COMP-THERM-HTC-018 REPORT HI-2104790 D-10 Table D.3-2 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 2Experiments:

Reflected Simple Arrays Moderated by Poisoned Water with Gadolinium or BoronCriticalCaseEvaluation Identification File-namekeff-i1 acalc-i-(Yexp+/- "3iEALF (eV)19 MIX-COMP-THERM-HTC-019 20 MIX-COMP-THERM-HTC-020 21 MIX-COMP-THERM-HTC-021 22 MIX-COMP-THERM-HTC-022 23 MIX-COMP-THERM-HTC-023 24 MIX-COMP-THERM-HTC-024 25 MIX-COMP-THERM-HTC-025 26 MIX-COMP-THERM-HTC-026 27 MIX-COMP-THERM-HTC-027 28 MIX-COMP-THERM-HTC-028 29 MIX-COMP-THERM-HTC-029 30 MIX-COMP-THERM-HTC-030 31 MIX-COMP-THERM-HTC-031 32 MIX-COMP-THERM-HTC-032 33 MIX-COMP-THERM-HTC-033 34 MIX-COMP-THERM-HTC-034 35 MIX-COMP-THERM-HTC-035 36 MIX-COMP-THERM-HTC-036 37 MIX-COMP-THERM-HTC-037 38 MIX-COMP-THERM-HTC-038 39 MIX-COMP-THERM-HTC-039 40 MIX-COMP-THERM-HTC-040 41 MIX-COMP-THERM-HTC-041 42 MIX-COMP-THERM-HTC-042 43 MIX-COMP-THERM-HTC-043 44 MIX-COMP-THERM-HTC-044 45 MIX-COMP-THERM-HTC-045 46 MIX-COMP-THERM-HTC-046 47 MIX-COMP-THERM-HTC-047 48 MIX-COMP-THERM-HTC-048 49 MIX-COMP-THERM-HTC-049 50 MIX-COMP-THERM-HTC-050 51 MIX-COMP-THERM-HTC-051 52 MIX-COMP-THERM-HTC-052 53 MIX-COMP-THERM-HTC-053 54 MIX-COMP-THERM-HTC-054 55 MIX-COMP-THERM-HTC-055 REPORT HI-2 104790 D- 11REPORT HI-2104790 D-11 File-nameCaseEvaluation Identification

+- Gcalc-i+/- crexp-(yiEALF (eV)56 MIX-COMP-THERM-HTC-056 57 MIX-COMP-THERM-HTC-057 58 MIX-COMP-THERM-HTC-058 59 MIX-COMP-THERM-HTC-059 IREPORT HI-2 104790 D-12REPORT HI-2104790 D-12 Table D.3-3 The MCNP5-1.51 Calculational Results and Measurements Phase 3 Critical Experiments:

Pool StorageUncertainties forCaseEvaluation Identification File- kff-inn~mp kI fI Gcalc-iI eFXP I+/- c7iEALF (eV)60 MIX-COMP-THERM-HTC-060 61 MIX-COMP-THERM-HTC-061 62 MIX-COMP-THERM-HTC-062 63 MIX-COMP-THERM-HTC-063 64 MIX-COMP-THERM-HTC-064 65 MIX-COMP-THERM-HTC-065 66 MIX-COMP-THERM-HTC-066 67 MIX-COMP-THERM-HTC-067 68 MIX-COMP-THERM-HTC-068 69 MIX-COMP-THERM-HTC-069 70 MIX-COMP-THERM-HTC-070 71 MIX-COMP-THERM-HTC-071 72 MIX-COMP-THERM-HTC-072 73 MIX-COMP-THERM-HTC-073 74 MIX-COMP-THERM-HTC-074 75 MIX-COMP-THERM-HTC-075 76 MIX-COMP-THERM-HTC-076 77 MIX-COMP-THERM-HTC-077 78 MIX-COMP-THERM-HTC-078 79 MIX-COMP-THERM-HTC-079 80 MIX-COMP-THERM-HTC-080 81 MIX-COMP-THERM-HTC-081 82 MIX-COMP-THERM-HTC-082 83 MIX-COMP-THERM-HTC-083 84 MIX-COMP-THERM-HTC-084 85 MIX-COMP-THERM-HTC-085 II-REPORT HI-2 104790 D- 13REPORT HI-2104790 D-13 Table D.3-4 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 4 CriticalExperiments:

Shipping CaskFile-CaseEvaluation Identification keff.i :+/- GcaxpI+/- (7iEALF(eV)86 MDI-COMP-THERM-HTC-086 87 MIX-COMP-THERM-HTC-087 88 MIX-COMP-THERM-HTC-088 89 MIX-COMP-THERM-HTC-089 90 MIX-COMP-THERM-HTC-090 91 MIX-COMP-THERM-HTC-091 92 MIX-COMP-THERM-HTC-092 93 MIX-COMP-THERM-HTC-093 94 MIX-COMP-THERM-HTC-094 95 MIX-COMP-THERM-HTC-095 96 MIX-COMP-THERM-HTC-096 97 MIX-COMP-THERM-HTC-097 98 MIX-COMP-THERM-HTC-098 99 MIX-COMP-THERM-HTC-099 100 MIX-COMP-THERM-HTC-100 101 MDI-COMP-THERM-HTC-101102 MIX-COMP-THERM-HTC-102 103 MIX-COMP-THERM-HTC-103 104 MIX-COMP-THERM-HTC-104 105 MIX-COMP-THERM-HTC-105 106 MIX-COMP-THERM-HTC-106 107 MIX-COMP-THERM-HTC-107 108 MIX-COMP-THERM-HTC-108 109 MIX-COMP-THERM-HTC-109 110 MIX-COMP-THERM-HTC-1 10111 MIX-COMP-THERM-HTC-1 111112 MIX-COMP-THERM-HTC-1 12113 MIX-COMP-THERM-HTC-1 13114 MIX-COMP-THERM-HTC-1 14115 MIX-COMP-THERM-HTC-1 15116 MIX-COMP-THERM-HTC-1 16117 MIX-COMP-THERM-HTC-117118 MIX-COMP-THERM-HTC-1 18119 MIX-COMP-THERM-HTC-119120 MIX-COMP-THERM-HTC-120 121 MIX-COMP-THERM-HTC-121 122 MIX-COMP-THERM-HTC-122 123 MIX-COMP-THERM-HTC-123 REPORT HI-2 104790 D- 14REPORT HI-2104790 D-14 File--I- Ucalc-CaseEvaluation Identification keff-i4- (FexpEALF(,=X'l124 MIX-COMP-THERM-HTC-124125 MIX-COMP-THERM-HTC-125126 MIX-COMP-THERM-HTC-126127 MIX-COMP-THERM-HTC-127128 MIX-COMP-THERM-HTC-128129 MIX-COMP-THERM-HTC-129130 MIX-COMP-THERM-HTC-130131 MIX-COMP-THERM-HTC-131132 MIX-COMP-THERM-HTC-132133 MIX-COMP-THERM-HTC-133 134 MIX-COMP-THERM-HTC-134 135 MIX-COMP-THERM-HTC-135136 MIX-COMP-THERM-HTC-136 137 MIX-COMP-THERM-HTC-137138 MIX-COMP-THERM-HTC-138 139 MIX-COMP-THERM-HTC-139 140 MIX-COMP-THERM-HTC-140 141 MIX-COMP-THERM-HTC-141142 MIX-COMP-THERM-HTC-142 143 MIX-COMP-THERM-HTC-143 144 MIX-COMP-THERM-HTC-144 145 MIX-COMP-THERM-HTC-145 146 MIX-COMP-THERM-HTC-146 147 MIX-COMP-THERM-HTC-147148 MIX-COMP-THERM-HTC-148 149 MIX-COMP-THERM-HTC-149 150 MIX-COMP-THERM-HTC-150 151 MIX-COMP-THERM-HTC-151 152 MIX-COMP-THERM-HTC-152153 MIX-COMP-THERM-HTC-153 154 MIX-COMP-THERM-HTC-154 155 MIX-COMP-THERM-HTC-155 156 MIX-COMP-THERM-HTC-156 REPORT HI-2 104790 D- 15REPORT HI-2104790 D-15 Table D.3-5 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for SelectedCritical Experiments r r 1CaseEvaluation Identification File-keff-iI (cale-EALFI Gexp-+/-' (i157 Core l158 Core II159 Core HII160 Core IX161 Core X162 Core XI163 Core XII164 Core XIII165 Core XIV166 Core XV167 Core XVI168 Core XVII169 Core XVIII170 Core XIX171 Core XX172 Core XXI173 S-type Fuel, w/886 ppm B174 S-type Fuel, w/746 ppm B175 SO-type Fuel, w/1 156 ppm B176 Case 1 1337 ppm B177 Case 12 1899 ppmB B178 Water Moderator 0 gap179 Water Moderator 2.5 cm gap180 Water Moderator 5 cm gap181 Water Moderator 10 cm gap182 Steel Reflector, 1.321 cm separation 183 Steel Reflector, 2.616 cm separation 184 Steel Reflector, 3.912 cm separation 185 Steel Reflector, Infinite separation 186 Steel Reflector, 1.321 cm separation 187 Steel Reflector, 2.616 cm separation 188 Steel Reflector, 5.405 cm separation 189 Steel Reflector, Infinite separation 190 Steel Reflector, with Boral Sheets191 Lead Reflector, 0.55 cm sepn.192 Lead Reflector, 1.956 cm sepn.193 Lead Reflector, 5.405 cm sepn.194 Experiment 004/032 -no absorberREPORT HI-2104790 D- 16 File-CaseEvaluation Identification keff-i-Gcalc._+/-- (yexp+/- GiEALF(P17)195 Exp. 009 1.05% Boron Steel plates196 Exp. 009 1.62% Boron Steel plates197 Exp. 031 -Boral plates198 Experiment 214R- with flux traps199 Experiment 214V3 -with flux trap200 Case 173 -0 ppm B201 Case 177 -2550 ppm B202 MOX Fuel -Type 3.2 Exp. 21203 MOX Fuel -Type 3.2 Exp. 43204 MOX Fuel -Type 3.2 Exp. 13205 MOX Fuel -Type 3.2 Exp. 32206 Saxton Case 52 PuO2 0.52" pitch207 Saxton Case 52 U 0.52" pitch208 Saxton Case 56 PuO2 0.56" pitch209 Saxton Case 56 borated PuO2210 Saxton Case 56 U 0.56" pitch211 Saxton Case 79 PuO2 0.79" pitch212 Saxton Case 79 U 0.79" pitch213 0.700-in.

pitch 0 ppm B214 0.700-in.

pitch 688 ppm B215 0.870-in.

pitch 0 ppm B216 0.870-in.

pitch 1090 ppm B217 0.990-in.

pitch 0 ppm B218 0.990-in.

pitch 767 ppm B219 Saxton Case PuO2 0.735" pitch220 Saxton Case PuO2 1.04" pitch221 8 wt% 240Pu 0.80" pitch222 8 wt% 240Pu 0.93" pitch223 8 wt% 240Pu 1.05" pitch224 8 wt% 240Pu 1.143" pitch225 8 wt% 240Pu 1.32" pitch226 8 wt% 240Pu 1.386" pitch227 16 wt% 240Pu 0.93" pitch228 16 wt% 240Pu 1.05" pitch229 16 wt% 240Pu 1.143" pitch230 16 wt% 240Pu 1.32" pitch231 24 wt% 240Pu 0.80" pitch232 24 wt% 240Pu 0.93" pitch233 24 wt% 240Pu 1.05" pitch234 24 wt% 240Pu 1.143" pitchREPORT HI-2 104790 D- 17REPORT HI-2104790 D-17 CaseEvaluation Identification File-namekeiffi 1 1 acalc- 1 1 exp 1+- CjEALF235 24 wt% 240Pu 1.32" pitch236 24 wt% 240Pu 1.386" pitch237 18 wt% 240Pu 0.85" pitch238 18 wt% 240Pu 0.93" pitch239 18 wt% 240Pu 1.05" pitch240 18 wt%240Pu 1.143" pitch241 18 wt% 240Pu 1.386" pitch242 18 wt% 240Pu 1.60" pitch243 18 wt% 240Pu 1.70" pitch244 1 Cluster245 3 Clusters, Separation 11.92 cm246 3 Clusters, Separation 8.41 cm247 3 Clusters, Separation 10.05 cm248 3 Clusters, Separation 6.39 cm249 3 Clusters, Separation 9.01 cm250 3 Clusters, Separation 4.46251 1 Cluster, l0xll.51252 1 Cluster, 9x13.35253 1 Cluster, 8x16.37254 3 Clusters, Separation 7.11 cmI1 Cluster, 614.4 Rods, Gd water255 impurity256 1 Cluster, 529.3 Rods257 1 Cluster, 523.9 Rods258 1 Cluster, 525.3 Rods259 1 Cluster, 595.4 Rods260 1 Cluster, 485.8 Rods261 1 Cluster, 523.8 Rods262 1 Cluster, 505.4 Rods263 4 Clusters, Separation 2.59 cm264 2 Clusters, Separation 1.68 cm265 4 Clusters, Separation 4.27 cm266 4 Clusters, Separation 5.95 cm267 4 Clusters, Separation 5.11 cm268 4 Clusters, Separation 6.66 cm269 4 Clusters, Separation 7.53 cm270 4 Clusters, Separation 9.00 cm271 4 Clusters, Separation 9.97 cm272 4 Clusters, Separation 11.45 cm273 4 Clusters, Separation 13.87 cm--- --- -REPORT HI-2104790 D-18 CaseEvaluation Identification File-* 1 -'pkeff. + calc-274 -3 Clusters, Separation 9.88 cm275 -3 Clusters, Separation 6.78 cm276 3 Clusters, Separation 6.176 cm277 1 Cluster, 225.8 Rods, Gd waterimpurity278 1 Cluster, 216.2 Rods279 1 Cluster, 216.6 Rods280 1 Cluster, 218.6 Rods281 1 Cluster, 167.85 Rods282 1 Cluster, 203 Rods283 1 Cluster, 173.5 Rods284 2 Clusters, Separation 2.83 cm285 3 Clusters, Separation 12.27 cm286 3 Clusters, Separation 12.493 cm287 4 Clusters, Separation 4.72 cm288 4 Clusters, Separation 8.38 cm289 4 Clusters, Separation 10.86 cm290 4 Clusters, Separation 11.29 cm291 4 Clusters, Separation 12.02 cm292 4 Clusters, Separation 13.64 cm293 4 Clusters, Separation 14.98 cm294 4 Clusters, Separation 19.81 cm295 4 Clusters, Separation 8.50 cm296 19xl9, RodPitchi-1.849 cm297 20x20, Rod Pitch -1.849 cm298 2 1 x2C1, RodPitchi-1.849 cm299 17x 17, RodPitchi-1.956 cm300 18x 18, RodPitchi-1.956 cm301 19x19, Rod Pitch -1.956 cm302 20x20, Rod Pitch -1.956 cm303 21x21, Rod Pitch -1.956 cm304 16x 16, Rod Pitch -2.15 cm305 17x17, RodPitch-2.156cm306 18x18, RodPitch-2.156cm307 19x19, Rod Pitch -2.56 cm308 20x20, Rod Pitch -2.15 cm309 15x215, Rod Pitch -2.293 cm310 16x 16, Rod Pitch -2.293 cm311 17xl7, Rod Pitch -2.293 cm312 18x18, Rod Pitch -2.293 cm-7.+ Gx----T7IREPORT 111-2104790 D- 19REPORT HI-2104790 D- 19 CaseEvaluation Identification File-keff-i I- (ycalc- -+/- Gexp+/- (YiEALF313 19x19, Rod Pitch -2.293 cm314 Core XI, 1511 ppm315 Core XI, 1335.5 ppm316 Core XI, 1335.5 ppm317 Core XI, 1182 ppm, 36 Pyrex Rods318 Core XI, 1182 ppm, 36 Pyrex Rods319 Core XI, 1032.5 ppm, 72 Pyrex Rods320 Core XI, 1032.5 ppm, 72 Pyrex Rods321 Core XI, 794 ppm, 144 Pyrex Rods322 Core XI, 779 ppm, 144 Pyrex Rods323 Core XI, 1245 ppm, 72 Vicor Rods324 Core XI, 1384 ppm, 144 A1203 Rods325 Core XI, 1348 ppm, 36 A1203 Rods326 Core XI, 1348 ppm, 36 A1203 Rods327 Core XI, 1363 ppm, 72 A1203 Rods328 Core XI, 1362 ppm, 72 A1203 Rods329 Core XI, 1158 ppm330 Core XI, 921 ppm331 0% Boron Steel plates, dist. 0.245 cm332 0% Boron Steel plates, dist. 3.277 cm333 0% Boron Steel plates, dist. 0.428 cm334 0% Boron Steel plates, dist. 3.277 cm3351.05% Boron Steel plates, dist. 3.277cm--mmmm-3361.62% Boron Steel plates, dist. 3.277cm--- mlm -337 Al plates, dist. 0.105 cm338 Al plates, dist. 3.277 cm339 Zircaloy-4 plates, dist. 0.078 cm340 Zircaloy-4 plates, dist. 3.277 cm341 Lead Reflector, 0 cm separation 342 Lead Reflector, 0.660 cm separation 343 Lead Reflector, 1.321 cm separation 344 Lead Reflector, 5.405 cm separation 345 Steel Reflector, 0 cm separation 346 Steel Reflector, 0.660 cm separation 347 Steel Reflector, 1.321 cm separation 348 Steel Reflector, 2.616 cm separation 349 Steel Reflector, 5.405 cm separation 350 Steel Reflector, 0 cm separation REPORT HI-2 104790 D-20REPORT HI-2104790 D-20 File-CaseEvaluation IdentificationF-i 'calcE.4 Gexp-(;iEALF351 Steel Reflector, 0.660 cm separation 352 Steel Reflector, 1.956 cm separation 353 Lead Reflector, 0 cm separation 354 Core lilA355 Core IIC356 Core HID357 Core IIIE358 Core IIIF359 Core IIG360 Core IV361 Core V362 Core VI363 Core VII364 Core VIII0% Boron Steel plate, Gd waterimpurity366 1.1% Boron Steel plate367 1.6% Boron Steel plate368 Boral B plate369 Boral C plate370 Boroflex, 1.84 cm separation 371 Boroflex, 1.73 cm separation 372 Steel Reflector, 0% Boron Steel plate.- --- -373Steel Reflector, 1.1% Boron Steelplate--- -I374Steel Reflector,

Boroflex, 8.37 cmseparation

-=-I375 Borated Water, 490 ppm376 Unborated Water377 Borated Water, 1030 ppm378 0% Boron Steel plates, dist. 0.645 cm379 0% Boron Steel plates, dist. 2.732 cm380 0% Boron Steel plates, dist. 4.042 cm381 0% Boron Steel plates, dist. 0.645 cm382 0% Boron Steel plates, dist. 4.042 cm383 0% Boron Steel plates, dist. 0.645 cm384 0% Boron Steel plates, dist. 4.042 cm3851.05% Boron Steel plates, dist. 0.645cm-m-m--- m386 1.05% Boron.Steel plates, dist. 4.0421___ cm J m_____ J ________=_____________________

REPORT HI-2 104790 D-2 1REPORT HI-2104790 D-21 Case Evaluation Identification File- ( L EALFname ___ (V387 1.62% Boron Steel plates, dist. 0.645cm388 1.62% Boron Steel plates, dist. 4.042cm389 Boral plates, dist. 0.645 cm390 Boral plates, dist. 4.442 cm391 Boral plates, dist. 0.645 cm392 Al plates, dist. 0.645 cm393 Al plates, dist. 4.042 cm394 Al plates, dist. 4.442 cm395 Zircaloy-4 plates, dist. 0.645 cm396 Zircaloy-4 plates, dist. 4.042 cm397 Hex, 621 Rods, Temperature 20.1C398 Hex, 889 Rods, Temperature 231A.4C399 Hex, 1951 Rods, Temperature 19.3C400 Hex, 2791 Rods, Temperature 206.OC401 Hex, 325/680 Rods, Temperature 20.8C402 Hex, 325/912 Rods, Temperature 212.1C403 Core XIA ---404 Core XIC ---405 Core XID ---406 Core XIE407 Core XIF408 Core XIG409 Core XIIIA410 No Boron Steel plates411 0% Boron Steel plates, 3 mm, dist. 0412 0% Boron Steel plates, 6 mm, dist. 0413 0% Boron Steel plates, 6 mm, dist. 0.5414 0% Boron Steel plates, 6 mm, dist. 1415 0.67% Boron Steel plates, 3 mm, dist.0416 0.67% Boron Steel plates, 6 mm, dist.0417 0.67% Boron Steel plates, 3 mm, dist.0.5418 0.67% Boron Steel plates, 6 mm, dist.0.5419 0.67% Boron Steel plates, 3 mm, dist.1420 0.67% Boron Steel plates, 6 mm, dist. _____ ______REPORT HI-2104790 D-22 Case Evaluation Identification 14210.98% Boron Steel plates, 3 mm, dist.0-m- -nm m422 0.98% Boron Steel plates, 6 mm, dist.0423 0.98% Boron Steel plates, 6 mm, dist._____ ~~~~0.5 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _4240.98% Boron Steel plates, 6 mm, dist.1---mm-425 No Boron Steel plates426 0% Boron Steel plates, dist. 0427 0.67% Boron Steel plates, dist. 0428 0.98% Boron Steel plates, dist. 0429 No Boron Steel plates430 0% Boron Steel plates, dist. 0431 0% Boron Steel plates, dist. 0.5432 0% Boron Steel plates, dist. 0433 0% Boron Steel plates, dist. 0.5434 0.67% Boron Steel plates, dist. 0435 0.67% Boron Steel plates, dist. 0.5436 0.67% Boron Steel plates, dist. 0437 0.67% Boron Steel plates, dist. 0.5438 0.98% Boron Steel plates, dist. 0439 0.98% Boron Steel plates, dist. 0.5.440 0.98% Boron Steel plates, dist. 0441 0.98% Boron Steel plates, dist. 0.5442 Otto Hahn, ZrB2 and B4C rods443 f'EN/MB-0l1 (580 pins)444 IPEN/MB-01 (560 pins)445 670 pins, A1203-B4C rods446 672 pins, A1203-B4C rods447 668 pins, A1203-B4C rods448 668 pins, A1203-B4C rods449 664 pins, 16 steel rods450 662 pins, 18 steel rods451 658 pins, 14 steel rods452 660 pins, 12 steel rods453 660 pins, 12 steel rods454 661 pins, 17 steel rods455 662 pins, 16 steel rods456 634 pins, 12 steel rodsREPORT HI-2 104790 D-23REPORT HI-2104790 D-23 File-CaseEvaluation Identification keff-i+ Gcalc- +/- exp+/- cTiEALF(ANA457 620 pins, 26 steel rods458 668 pins, 0 steel rods, 4 Gd203 rods459 648 pins, 0 steel rods, 8 Gd203 rods460 672 pins, 0 steel rods, 4 Gd203 rods461 646 pins, 4 steel rods, 4 Gd203 rods462 656 pins, 4 steel rods, 4 Gd203 rods463 664 pins, 4 steel rods, 2 Gd203 rods464 670 pins, 2 steel rods, 2 Gd203 rods465 664 pins, 2 steel rods, 2 Gd203 rods466 656 pins, 0 steel rods, 2 Gd203 rods467 23x23, 1.825 cm pitch468 23x23, 1.825 cm pitch469 23x23, 1.825 cm pitch470 21x21, 1.956 cm pitch471 21x21, 1.956 cm pitch472 21x21, 1.956 cm pitch473 20x20, 2.225 cm pitch474 20x20, 2.225 cm pitch475 20x20, 2.225 cm pitch476 21 x21,2.474 cm pitch477 21x21, 2.474 cm pitch478 8 wt% 240Pu 1.05" pitch, Al Rods479 8 wt% 240Pu 1.05" pitch, B4 Rods480 8 wt% 240Pu 1.05" pitch, B3 Rods481 8 wt% 240Pu 1.05" pitch, B2 Rods482 8 wt% 240Pu 1.05" pitch, B I Rods4838 wt% 240Pu 1.05" pitch, AI+CdRods---r -- -484 8 wt% 240Pu 1.05" pitch, B4+CdRods485 8 wt% 240Pu 1.05" pitch, B3+CdRods486 8 wt% 240Pu 1.05" pitch, B2+CdRods487 8 wt% 240Pu 1.05" pitch, B I +CdRods488 8 wt%/o 240Pu 1.05" pitch, Air+CdRods489 8 wt% 240Pu 1.05" pitch, H20+CdRods490 8 wt% 240Pu 1.32" pitch, Al Rods _ __ _491 8 wt% 240Pu 1.32" pitch, B4 Rods _ _ _ _REPORT HI-2 104790 D-24REPORT HI-2104790 D-24 File- Irai EALFCase Evaluation Identification name kf alc : GeV +/- (eV)492 8 wt% 240Pu 1.32" pitch, B3 Rods493 8 wt% 240Pu 1.32" pitch, B2 Rods494 8 wt% 240Pu 1.32" pitch, BI Rods495 8 wt% 240Pu 1.32" pitch, AI+CdRods _____496 8 wt% 240Pu 1.32" pitch, B4+CdRods497 8 wt% 240Pu 1.32" pitch, B3+CdRods498 8 wt% 240Pu 1.32" pitch, B2+CdRods499 8 wt% 240Pu 1.32" pitch, B 1 +CdRods _______500 8 wt% 240Pu 1.32" pitch, Air+CdRods501 8 wt% 240Pu 1.32" pitch, H20+CdRods502 16 wt% 240Pu 1.386" pitch503 16 wt% 240Pu 1.05" pitch, Al Rods504 16 wt% 240Pu 1.05" pitch, B4 Rods505 16 wt% 240Pu 1.05" pitch, B3 Rods506 16 wt% 240Pu 1.05" pitch, B2 Rods507 16wt%240Pu1.05"pitch, B1Rods508 16 wt% 240Pu 1.05" pitch, AI+CdRods _______509 16 wt% 240Pu 1.05" pitch, B4+CdRods510 16 wt% 240Pu 1.05" pitch, B3+CdRods511 16 wt% 240Pu 1.05" pitch, B2+CdRods512 16 wt% 240Pu 1.05" pitch, BI+CdRods _____513 16 wt% 240Pu 1.05" pitch, Air+CdRods514 16 wt% 240Pu 1.05" pitch, H20+CdRods515 24 wt% 240Pu 1.05" pitch, Al Rods516 24 wt% 240Pu 1.05" pitch, B4 Rods517 24 wt% 240Pu 1.05" pitch, B3 Rods518 24 wt% 240Pu 1.05" pitch, B2 Rods519 24 wt% 240Pu 1.05" pitch, B1 Rods520 24 wt% 240Pu 1.05" pitch, AI+CdRods _ _ __521 24 wt% 240Pu 1.05" pitch, B4+Cd _ ___REPORT HI-2104790 D-25 CaseEvaluation Identification Rods52224 wt% 240Pu 1.05" pitch, B3+Cd---n--mRods523 24 wt% 240Pu 1.05" pitch, B2+CdRods524 24 wt% 240Pu 1.05" pitch, Bl+CdRods525 24 wt% 240Pu 1.05" pitch, Air+CdRods52624 wt% 240Pu 1.05" pitch, H20+CdRods----m-527 8 wt% 240Pu 0.55" pitch528 8 wt% 240Pu 0.60" pitch529 8 wt% 240Pu 0.71" pitch530 8 wt% 240Pu 0.80" pitch531 8 wt% 240Pu 0.90" pitch532 8 wt% 240Pu 0.93" pitchREPORT HI-2104790 D-26 Table D.3-6 Descriptive Statistics of the MCNP5-1.51 Calculational ResultsExperiment Description No. ofexp.keff rangeHTC Experiments 156Selected Experiments 376All experiments 532EALF (eV) rangeTable D.3-7 Normality Test Results for the MCNP5-1.51 calculations Experiment Description No. ofexp.Shapiro-Wilk Pearson's chi-square (X2)WtestWHTC Experiments 156 N/A N/ASelected Experiments 376 N/A N/AAll experiments 532 N/A N/AnmmmPd(X2;d)NormalmmmTable D.3-8 Trending Analysis Results for the MCNP5-1.51 calculations Experiment Description No. ofexp.Correlated Parameter, xCorrelation Coefficient, 14Probability, Pd(N;r)Correlation Regression

Equation, k(x)+ 4-Allexperiments 532EALFPitchRod ODFuel Density* m* ImREPORT HI-2104790 D-27REPORT HI-2104790 D-27 Table D.3-9 Analysis of Neutron Absorbers and Reflector Materials for the MCNP5-1.51 calculations Experiment Description No.ofexp.BiasBiasUncertainty Normality X2(Pd(X2;d))

Linear Correlation Residuals Normality, (Pd(X2;d))All 532 -experiments__53 II -IIAll exceptthose withGadolinium, Cadmium andLead365REPORT HI-2104790 D-28 Table D.3-10 Bias and Bias Uncertainty as a Function of Independent REPORT HI-2104790 D-29REPORT HI-2104790 D-29 Table D.3-11 Analysis of Fuel Burnup for the MCNP5-1.51 calculations Normality Residuals xperimen of Bias Bias N i Linear Correlation Normality, Description Uncertainty X2 2exp. (Pd(xE;d))

________________

(Pd(X2;d))

All exceptthose withGadolinium, 365 mmCadmium andLeadtFresh U02 207 mIN mm m m_______m

-Fuel--HT m+MOX 158 mExperiments REPORT HI-2 104790 D-30REPORT HI-2104790 D-30 Table D.3-12 Bias and Bias Uncertainty as a Function of Independent Parameter Indepen Bias Independ BiasExperiment dent Calculated Bias Uncerta ent Calculated Bias UncertaiDescription Paramet kF inty aramete kf ntyer, x r, xEALFPu Enrichment HTC + MOXExperiments IIIIREPORT HI-2 104790 D-3 IREPORT HI-2104790 D-31 Table D.3-13 Analysis of the Unborated and Borated Water for the MCNP5-1.51 calculations No. Normality Residuals Eerimen of Bias Bias Linear Correlation Normality, Description Uncertainty22 exp. (Pd(x';d))

(Pd(x ;d))All except thosewithGadolinium, 365 M Cadmium andLeadtAll with Fresh 287 _ _ _ _ _ MWater 287 mAll with 78Borated Water 7EMO "tNote: Critical experiments with Gadolinium, Cadmium and Lead were excluded from all subsequent subsets.REPORT HI-2104790 D-32 Table D.3-14 Bias and Bias Utas a Function ofParameter All with FreshWaterEALFmElmEimEimEimEimEimEimEimEimEimElmElmEimElmEimElmElmElmElmEimE-DensitymElmEimEiU Enrichment M IAll withBorated WaterN/AREPORT HI-2 104790 D-33REPORT HI-2104790 D-33 Experiment Description IndependentParameta1- "'CalculatedBiasBiasUncertaintyIndepend Biasent Calculated Bias UncertaiParamete kyr, xREPORT HI-2 104790 D-34REPORT HI-2104790 D-34 Table D.3-15 Analysis of the Combinations of Fuel Burnup and Unborated/Borated Water for the MCNP5-1.51 calculations No. Bias Normality Residuals Experiment of Bias Uncertain x2 Linear Correlation Normality, exp. ty (Pd(x;d))

(Pd(X ;d))All except thosewithGadolinium, 365 1 1___ 1ma nCadmium andLeadtFresh U02 Fuelwith Fresh 154 -__ mWaterFresh U02 Fuelwith Borated 53 aIWaterHTC + MOXAFuel with Fresh 133 AWater MIEH-TC + MOXU.Fuel with 25 1__ __Borated WatertNote: Criticalsubsets.experiments with Gadolinium, Cadmium and Lead were excluded from all subsequent REPORT HI-2104790 D-35REPORT HI-2104790 D-35 Table D.3-16 Bias and Bias Uncertainty as a Function of Independent Parameter Indepen Bias Independ BiasExperiment dent Calculated ent Calculated Bias UncertaiDescription Paramet kBfia inct Paramete kefter, x r, xntyEALFmIImEimiImIImElmElmIImEimIIFresh U02Fuel withFresh WaterN/AmmmmmmmmmEALFmiImPu Enrichment HTC + MOXFuel withFresh WaterREPORT HI-2104790 D-36 Experiment Description IndependentParametCalculated keffBiasBiasUncertaintyIndependentParameter. xCalculated keffBiasBiasUncertaintymEmmEmmEmmEmmEmRod ODmElmElE lm UImIIHTC + MOXFuel withBorated WaterDensitymElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElmElEE==mM==MREPORT HI-2104790 D-37 Table D.3-17 Comparison of Key Parameters and Definition of Validated AOAParameter Design Benchmarks Validated Application Fissionable Material 235U, 239Pu, 241Pu 235U, 239Pu, 241Pu 235U, 239Pu, 241PuIsotopic Composition 235u/ut < 5.Owt% 1.57 -5.74% < 5wt%Pu/(U+Pu)

< 20wt% 1.104-20%

< 20wt%Physical Form UO2,MOX U02,MOX UO2, MOXFuel Density (g/cm3) 10.0 -10.7 9.2 -10.4 9.2 -10.7Moderator Material (coolant)

H H HPhysical Form H20 H20 H20Density (g/cm3) around 1.0 g/cm3 around 1.0 g/cm3 around 1.0 g/cm3Reflector Material H H HPhysical Form H20 H20 H20Density (g/cm3) around 1.0 g/cm3 around 1.0 g/cm3 around 1.0 g/cm3Interstitial Reflector MaterialPlate Steel or Lead Steel or Lead Steel or LeadAbsorber MaterialNone, Boron (15 -None, Boron (0 -Soluble None, Boron or 2550 ppm) or 2550 ppm) orGadolinium Gadolinium (48 -Gadolinium (48 to197 ppm) 197 ppm)Rods Boron Pyrex', Vicor BoronSteel or B-AlSeparating MaterialWater, B-SS, Water, B-SS, Boral, Water, B-SS, Boral,Plate Boral or Boroflex, Zircaloy or Boroflex, Zircaloy orCadmium Cadmium CadmiumGeometryLattice type Square Square, Triangle Square, Triangle1.26-1.47 Lattice Pitch (cm) (PWR) 0.968 to 4.318 0.968 to 4.318LaticePith (m) 1.24 -1.88(BWR)ThermalNeutron Energy spectrum Thermal spectrum Thermal spectrumREPORT HI-2104790 D-38 Figure Proprietary Figure D.3-1 Frequency Chart for Calculated keff of the Selected 532 Benchmarks for theMCNP5-1.51 codeFigure Proprietary Figure D.3-2 Frequency Chart for Calculated EALF (eV) of the Selected 532 Benchmarks for theMCNP5-1.51 codeREPORT HI-2104790 D-39 Figure Proprietary Figure D.3-3 MCNP5-1.51 Calculated keg Values for Various Values of the Spectral Index (AllExperiments)

REPORT HI-2104790 D-40 Figure Proprietary Figure D.3-4 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral IndexREPORT HI-2 104790 D-4 1REPORT HI-2104790 D-41 Figure Proprietary Figure D.3-5 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral IndexREPORT HI-2104790 D-42 Figure Proprietary Figure D.3-6 MCNP5-1.51 Calculated keff Values for Various Values of the Pu Enrichment REPORT HI-2 104790 D-43REPORT HI-2104790 D-43 Figure Proprietary Figure D.3-7 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral IndexREPORT H-Ii-2104790 D-44 Figure Proprietary Figure D.3-8 MCNP5-1.51 Calculated keff Values for Various Values of the U Enrichment REPORT HI-2104790 D-45 Figure Proprietary Figure D.3-9 MCNP5-1.51 Calculated keff Values for Various Values of the Fuel DensityREPORT HI-2104790 D-46REPORT HI-2104790 D-46 Figure Proprietary Figure D.3-10 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral IndexREPORT HI-2 104790 D-47REPORT HI-2104790 D-47 Figure Proprietary Figure D.3-11 MCNP5-1.51 Calculated ke, Values for Various Values of the Spectral IndexREPORT HI-2104790 D-48 Figure Proprietary Figure D.3-12 MCNP5-1.51 Calculated keff Values for Various Values of the Pu Enrichment REPORT HI-2 104790 D-49REPORT HI-2104790 D -49 Figure Proprietary Figure D.3-13 MCNP5-1.51 Calculated keff Values for Various Values of the Rod ODREPORT HI-2 104790 D-50REPORT HI-2104790 D-50 Figure Proprietary Figure D.3-14 MCNP5-1.51 Calculated keff Values for Various Values of the Fuel DensityREPORT HI-2 104790 D-5 1REPORT HI-2104790 D-51 ATTACHMENT 5Holtec International Report No. HI-2125245, Revision 4,"Licensing Report for Quad Cities Criticality Analysis for Inserts -Non Proprietary Version" Eu.'HOLTINTERN AT IINECONALHoltec Center, 555 Lincoln Drive West, Marlton, NJ 08053Telephone (856) 797- 0900Fax (856) 797 -0909Licensing Report for Quad Cities Criticality Analysis for Inserts -Non Proprietary VersionFORExelonHoltec Report No: HI-2125245 Holtec Project No: 2127Sponsoring Holtec Division:

HTSReport Class : SAFETY RELATED Summary of Revisions:

Revision 0: Original IssueRevision 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 markedby revision bars.Revision 4: Minor editorial changes to page 4 description of Table 7.1(c) and Table 2.1(c) wasmove up one line. Neither change marked by revision bar. All Revision 3 revision barsremoved.Project No. 2127Report No. H1-2125245 Page i Table of Contents1. INTRODUCTION

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

102. METHODOLOGY

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

112.1 GENERAL, APPROACH

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

...........

II2.2 COMPUTER CODES AND CROSS SECTION LIBRARIES

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

112 .2 .1 M C N P 5 -1.5 1 .......................................................

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

1 12 .2.1 .1 M C N P 5-1.5 1 V alidation

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

2.2.1.1.1 122 .2 .2 C A S M O -4 .............................

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

..132.3 ANALYSIS ME-THODS

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

132.3.1 Design Basis Fuel Assembly

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

132 .3 ,1.1 P eak R eactiv ity ......................................................................................................................................................

142.3,1.1.1 P ak R eactivity and Fuel A ssem b B urnup .............................................................

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

142.3.1.1.2.......

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

..142.3.1.2 J- ii i i ; .......................................

152.3.1.3 Determination of the Design Basis Fuel Assembly Lattice ...........................................................................

162.3.1.4 Optima2 CASMO-4 Model Simplification Effect ..........................................................................................

162 .3 .1.5 C ore O perating Param eters ...................................................................................................................................

182.3.1.5.1 R eacto r P ow eer U prate ......................................................................................................................................

182.3.1.5.2 Integral R eactivity C ontrol D evices ...........................................................................................

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

192.3.1.5.3 A xial and Planar Enrichm ent V ariations

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

192.3.1.5.4 Fuel A ssem bly D c-Channeling

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

192 .3 .1 .6 .....................

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

2 02.3.2 Reactivity Effect of Spent Fuel Pool Water Temperature

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

202.3.3 Fuel Depletion Calculation Uncertainy

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

212.3.4 Fuel and Storage Rack Manufacturing Tolerances

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

222.3.4.1 Fuel M anufacturing T olerances

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

222.3.4.2 SEP Storage Rack Manufacturing Tolerances

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

232.3.5 Radial Positioning

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

232.3.5.1 Fuel Assembly Orientation in the Core ...........................................................................................................

232.3.5.2 Fuel R adial Positioning in the R ack ......................................................................................................................

232.3.5 .3 Inserts R adial Positioning

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

252.3.5.4 Fuel Orientation in SFP Rack Cell ..............................................................................................................

252.3.6 Insert. Co po. Measrement...

rtain.. ............

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

........

262.3.6.18

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

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

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

262.3.64 MA 2G .. E ... ...........................................................................................

272.3.6 .2.......

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

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

272. .1 Te peatr ad atrD siyEf cs.............................................

......2 ..6.2 D o p d ss b y -H o i o t l........................I.........................................

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

272.3.7 Insert Coupon Measurement Uncertainty

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

......7.. 22.3.8 Maximum kff Calculation for Normal Conditions..............................................

272.4 MARGIN EVALUATIONstortion.........................................................................................................

282.5 FUEL MOVEMENT, INSPECTION AND RECONSTITUTION OPERATIONS

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

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

2.6 ACCID NT CONDITION

..........

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

292.6.1 Temperature and WaF er Density Effects .......................................

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

302.6.2 Droppeda As Fem sbly -Horizontal

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

302.6.3 Dropped A ssembly -Vertical into a Storage Cell..............................................

302.6.4 Storage Cell Distortion...........................................................................

312.6.5 Misloaded Fuel Assembly/Missing Insert.......................................................

312.6.6 Mislocated Fuel Assembly

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

322.6.6.1 Mislocation of a Fuel Assembly in the Water (jai) between the Racks and Pool Wall ...........................

322.6.6.2 Mislocation of a Fuel Assembly in the Comner between Two Racks................................................

322.6.6.3 Mislocation of a Fuel Assembly between the SFP Rack and the Inspection Platform

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

322.6.7 Mis-installment of an Insert on Wrong Side of a Cell .....................................................

33Project No. 2127 Report No. HI-2125245 Page 1 2.6.8 Insert M echanical Wear .....................................................................................................

332.6.9 Rack M ovement ...................................................................................................................

332.7 .. .....................................................................

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

332.8 SPENT FUEL, RACK INTERFACES

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

342.9 RECONSTITtr TED FuEl, ASSEMBLIES

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

353. ACCEPTANCE CRITERIA

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

363.1 AIPPICABLE CODES, STANDARDS AND GUIDANCE'S

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

364. ASSUM PTIONS ..............................................................................................................................

375. INPUT DATA ..................................................................................................................................

385.1 FUEL.. ASSEMBLY SPECIFICATION

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

385.2 REACTOR PARAMEIERS

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

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

385.3 SPEN'T FUEL POOL PARAMETERS

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

385.4 STORAGE RACK SPECIFICATION

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

395.4.1 M aterial Compositions

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

396. CO M PUTER CO DES .........................................................................................................

.. 407. ANALYSIS

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

417.1 DESIGN BASIS AND UNCERTAINTY EVALUATIONS

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

417.1.1 .........

.. 417.1.2 Determination of the Design Basis Fuel Assembly Lattice ..............................................

417.1.2.1 Fuel Assembly De-Channeling

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

417.1.3 Optima2 CASM O-4 M odel Simplification Efl&ct ............................................................

417.1.4 Core Operating Parameters

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

427.1.4.1 R eactor Pow er U prate ............................................................................................................................................

427.1.5 Water Temperature and Density Effect ............................................................................

427.1.6 Depletion Uncertainty

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

427.1.7 Fuel and Rack M anzjfacturing Tolerances

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

437.1.7.1 Fuel Assembly Tolerances

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

437.1.7.2 SFP Rack Tolerances

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

437.1.8 Radial Positioning

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

437,1.8.1 Fuel Assembly Radial Positioning in SFP Rack ............................................................................................

437.1.8.2 Fuel Orientation in SFP Rack ................................................................................................................................

437.1.9 Fu.. .............

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

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

447.1.9 .1 .........................................................................................................

...447.1.10 ; 447.2 MAXIMUM KFF CALCULATIONS FOR NORMAL CONDITIONS

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

447.3 M ARGIN EVALUATION

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

447.4 ABNORMAL AND ACCIDENT CONDITIONS

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

457.5 MAXIMUM K3:F,: CALCUi ATIONS FOR ABNORMAL AND ACCIDENT CONDITIONS

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

457.6 ........................................................................................................................

457.7 SPENT FUEL RACK INTERFACES

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

458. CONCLUSION

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

479. REFERENCES

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

48Project No. 2127 Report No. HI-2125245 Page 2 Supplement 1: Additional Calculations to Support the Revised NETCO-SNAP-IN Rack InsertD esign ..............................................................................................

.... Si-IProJect No. 2127Report No. F11-2125245 Page 3 List of TablesTableDescription Table 7.2(a)Table 7.2(b)Table 7.3Table 7.4Table 7.5Table 7.6(a)Table 7.7Table 7.8Page505152535455565758596061626364656667687071727374Results of the MCNP5-1.51 Calculations for SVEA-96 Optima2 Q122LatticesResults of the MCNP5-1.51 Calculations for GEI 4 Lattice Type 5Results of the MCNP5-1.51 Calculations for Design Basis andSimplified Model of SVEA-96 Optima2 Q122 Lattice Type 146Results of the MCNP5-1.51 Calculations for Core Operating Parameters Results of the MCNP5-1.51 Calculations for the Effect of WaterTemperature and DensityResults of the MCNP5-1.51 Calculations for the Depletion Uncertainty Results of the MCNP5-1.51 Calculations for Fuel Tolerances Results of the MCNP5-1.51 Calculations for Rack Tolerances 757677ProJect No. 2127Report No. 1-11-2125245 Page 4 TableTable 7.9(a)Table 7.9(b)Table 7.12(a)Table 7.12(b)Table 7.12(c)Table 7.13(b)Description Results of the MCNP5-1.51 Calculations for Fuel Radial Positioning inSFP RacksResults of the MCNP5-1.51 Calculations for Fuel Orientation in SFPRAAk--Page78798081Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluatethe Effect of Nominal Values Instead of Using Minimum B4C Loadingand Minimum Insert Thickness on Reactivity Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluatethe Effect of the Actual Optima2 Q122 Fuel AssemblyMargin Evaluation Summary of the Margin Evaluation Results of the MCNP5-1.51 Calculations for the Empty Storage RackCell without InsertTable 7.16Table 7.17Results of the MCNP5-i.51 Calculations for Axially Infinite Optima2Q122 LatticesResults of the MCNP5-1.51 Calculations for SFR Interface 828384858687888990lmmmmmmmProject No. 2127Report No. HI-2125245 Page 5 TableDescrintion Table SI-ITable S 1-2Table S 1-3Table S 1-4Fuel Rack Insert Revised Dimensions Results of the MCNP5 Calculations for Revised Rack Tolerances Results of the MCNP5-1.51 Calculations for Revised Fuel RadialPositioning in SFP RacksResults of the MCNP5-1.51 Calculations for Revised Fuel Orientation in SFP RacksPageUS1-5SI-6SI-7S1-8S1-9SIl-10Project No. 2127Report No. 1-11-2125245 Page 6 List of FiguresDescrintion FigurePage9293949596979899100101102103104Project No. 2127Report No. 1H1-2125245 Page 7 FigureDescription Project No. 2127Page105106107108109110III112113114115116117118.... -24, Report No. 1-11-2125245 Page 8 FigureDescription PageSI-I1Project No. 2127Report No. HI-2125245 Page 9

1. INTRODUCTION This report documents the criticality safety evaluation for the storage of spent BWR fuel in theUnit I and Unit 2 spent fuel pools (SFPs) at Quad Cities Station operated by Exelon. The Unit Iand Unit 2 SFP racks are identical and are designed to accommodate BWR fuel. Currently, theSFPI racks credit BORAFLEX for reactivity control.

This new analysis will not credit theBORAFLEX but will instead credit new NETCO-SNAP-IN rack inserts, which are new toQuad Cities but not new relative to their use for spent fuel pool reactivity control.

This analysiswill demonstrate that with credit for the inserts the effective neutron multiplication factor (kerf) inthe 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 toassure that under all credible abnormal and accident conditions, the reactivity will not exceed theregulatory limit.Criticality control in the SFP, as credited in this analysis, relies on the following:

  • Fixed neutron absorbers o NETCO-SNAP-1N0 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.

Project No. 2127Report No. 1.11-2125245 Page 10

2. METHODOLOGY 2.1 General ApproachThe analysis is performed consistent with regulatory requirements and guidance.

Thecalculations 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 NationalLaboratory

[1]. MCNP5-1.51 calculations use continuous energy cross-section data based onENDF/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 theanalysis 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 totalnumber 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 beforeaveraging, and a minimum of 150 cycles that are accumulated, The initial source is specified asnniform cnver the fineled re.oinnq (t-enmhIie0 I2.2.1.1 MCNP5-1.51 Validation ProJect No. 2127Report No. 1-I1-2125245 Page I11 Project No. 2127Report No. HI-2125245 Page 12 U2.2.2 CASMO-4Fuel 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 reactoranalysis (depletion) when providing reactivity data for specific 3D simulator codes. CASMO-4 isa two-dimensional multigroup transport theory code based on the Method of Characteristics andit 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 thefuel (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 theCASMO-4 sensitivity studies are not used as input into the kcff calculations.

However, the codeauthors have validated CASMO-4 against MCNP and various critical experiments

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

The extracted isotopes were fuirther combined and used in MCNP5-1.51 calculations.

2.3 Analysis Methods2.3.1 Design Basis Fuel AssemblyThere are various fuel designs stored in the Quad Cities SFP. For the purpose of this analysis, thereactivity of each design is evaluated and the most reactive fuel bundle lattice is determined foruse as the design basis fuel assembly to determine keff at the 95/95 level. This approach followsthe guidance in [2] and [6], and is further described below.Project No. 2127Report No. HI-2125245 Page 1 3 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, ascore depletion occurs, the Gd begins to burnout until it is essentially fully depleted.

As the Gddepletes 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 composedof various axial lattices (including blankets) that can have different axial lengths, uraniumloadings (also mixed oxide loading, for MOX fuel), fuel pin arrangements including partial orpart-length rods, Gd pin locations and loading, etc. These various lattice components can alleffect at what burnup the peak reactivity occurs and the magnitude of the peak reactivity.

TheMax ial lattices w ith in a sin g le fu el asse bl ca th r f e a l h ve d f r nt p k r a ti ty2.3.1.1.1 Peak Reactivity and Fuel Assembly BurnupTypically, a spent fuel assembly is characterized by its assembly average burnup (over all latticesor nodes). In this analysis methodology the fuel assembly average burnup is of no concern and isnot credited for reactivity control.

Rather, the methodology credits the residual Gd and otherdepletion isotopic compositions at the fuel assembly peak reactivity (most reactive lattice peakreactivity).

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) areperformed for every lattice of the SVEA-96 Optima2 fuel assembly (as will be seen in Section 7,this is the fuel assembly with the design basis lattice) over a burnup range to determine theburnup 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 burnupor fuel assembly burnup profile is not applicable because the analysis already considers eachlattice at its most reactive composition, independent of the fuel assembly average burnup.2.3.1.1.2 Project No. 2127Report No. HI-2125245 Page 14 2.3.1.2ProJect, No. 2127Report No. I-11-2125245 Page 15 2.3.1.3 Determination of the Design Basis Fuel Assembly Lattice2.3.1.4 Optima2 CASMO-4 Model Simplification EffectAs previously discussed in Section 2.3.1.2, various fuel designs were provided.

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

The Optima2 model inCASMO-4 is described as the SVEA-96 model provided in the CASMO-4 manual [4]. ThisCASMO-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 theCASMO-4 model simplifications inherent in the code. The CASMO-4 model geometry of theSVEA-96 Optimna2 fuel differs from the SVEA-96 Optirna2 fuel as follows:Project No. 2127Report No. HI-2125245 Page 16 With respect to the fuel assembly geometry models, the amount of zirconium (and therefore theamount of water) in the CASMO-4 model of the SVEA-96 Optima2 fuel is reasonably similar tothat 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 onreactivity, an applicable set of code-to-code comparisons is performed.

The following cases areevaluated.

For the purpose of showing that the two codes calculate an equivalent reactivity the following comparisons are made:Project No. 2127Report No. 1-11-2125245 Page 17

" Case 2.3.1.4.1 is compared to Case 2.3.1.4.2 at 0 GWD/MTU to show that the two codescalculate 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 thereactivity difference due to the effect of the spent fuel. The two codes use different crosssection library versions and calculation sequences.

The main calculation sequencedifference between the two codes is that CASMO-4 uses a thermal expansion of spentfuel pellet which effects the fuel density [4]. The actual density is conservatively used inMCNP5-1.51.

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

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

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

Any non-conservative resultwould be treated as a bias.Case 2.3.1.4.3 is compared to the result of the actual design basis results (similar to Case2.3.1.4.3 but with the bounding insert orientation) to show that the design basis model isconservative.

2.3.1.5 Core Operating Parameters As previously discussed, CASMO-4 is used to perform depletion calculations to determine thespent 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 onlBWR 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 Section2.3.1.5.2, 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 theirworst case.2.3.1.5.1 Reactor Power UprateTo determine the effect of the power uprate on the reactivity ofassemblies in the SFP racks, the following evaluations are performed.

ProJect No. 2127Report No. HI-2125245 Page !18 2.3.1.5.2 Integral Reactivity Control DevicesThe 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 latticewhere all rods contain fuel and no Gd. As discussed in Section 2.3.1.1, the Gd in the fuelassembly holds down the fresh fuel assembly reactivity and then, as core depletion occurs, theGd begins to burnout until it is essentially fully depleted.

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

Note that integrated absorbers do notchange 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 inSection 7) cannot be de-channeled for storage in the SFP because of its specific design.However, GE14 (the most second reactive fuel assembly, as will be shown in Section 7) may bede-channeled.

Studies are performed to evaluate the effect of storage of GE14 without the Zrchannel 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.

TheMCNP5-1.51 model used herein is a 2x2 array with the cell centered fuel assembly thatincludes 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 GEl4 fuel assembly lattice 5 (themost 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 assembliesare 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 fuelassemblies 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 fuelassemblies are eccentric away from the corner where the insert wings connect, as shownin Figure 2.13(d).Project No. 2127 Report No. 1-11-2125245 Page 19 2.3.1.62.3.2 Reactivity Effect of Spent Fuel Pool Water Temperature The Quad Cities Station SFP has a normal pool water temperature operating range below 1 50 'F.For the nominal condition, the criticality analyses are to be performed at the most reactivetemperature and density [2]. Also, there are temperature-dependent cross section effects inMCNP5-1.51 that need to be considered.

In general, both density and cross section effects maynot have the same reactivity effect for all storage rack scenarios, since configurations with strongneutron absorbers typically show a higher reactivity at lower water temperature, whileconfigurations without such neutron absorbers typically show a higher reactivity at a higherwater temperature.

For the SFP racks which credit inserts, the most reactive SFP watertemperature 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 alsoavailable at other temperatures;

however, not usually at the desired temperature for SFPcriticality analysis.

MCNP5-1.51 has the ability to automatically adjust the cross sections to thespecified temperature when using the TMP card. Furthermore, MCNP5-1.51 has the ability tomake a molecular energy adjustment for select materials (such as water) by using the S(a,p) card.The S(a,13) card is provided for certain fixed temperatures which are not always applicable toSFP 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(C,3) card fortemperatures as it does for the TMP card discussed above. Therefore, additional studies areperformed 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(U,[) calculations are performed for both upper and lower bounding S(a,f3) values, if needed.The studies mentioned above are performed for the following eases for the single cellMCNP5-1.51 SFP model (with periodic boundary conditions through the centerline of thesurrounding water 2):Project No. 2127Report No. 111-2125245 Page 20

" Case 2.3.2.1 (reference case): Temperature of 39.2 IF (277.15 K) and a density of 1.0g/cc are used to determine the reactivity at the low end of the temperature range. TheS(a,o3) card corresponds to a temperature of 68.81 IF (293.6 K).* Case 2.3.2.2:

Temperature of 150 IF (338.71 K) and a corresponding density of 0.98026g/cc are used to determine the reactivity at the high end of the temperature range. TheS(a,j3) card corresponds to a temperature of 68.81 °F (293.6 K).* Case 2.3.2.3:

Temperature of 150 IF and a corresponding density of 0.98026 g/cc. TheS(@,p) card corresponds to a temperature of 170.33 IF (350 K).The bounding water temperature and density (the temperature and its corresponding densitywhich 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 use the same MCNP5-l.51 models used in the design basis calculation.M 2.3.3 Fuel Depletion Calculation Uncertainty To account for the uncertainty of the number densities in the depletion calculations erformed inThe depletion uncertainty is applied by multiplying it with the reactivity difference (at95%/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 Gd2O3).Calculations are performed for the single cell model of design basis fuel assembly.

The uncertainty is determined by the following:

Uncertaintytopic

= [ (kcale-2 -kcale-1) + 2 * ((J .12 + Gcalc.22)

]

  • 0.05withkeac-1 = kcalc with spent fuel= kal. with fresh fuelGcal-l =Standard deviation of 0~ceal-2=

Standard deviation of kcale-2'The result of the MCNP5-1.51 calculation for the fuel depletion calculation uncertainty isstatistically combined with other uncertainties to determine keff.Project No. 2127Report No. 2125245 Page 21 2.3.4 Fuel and Storage Rack Manufacturing Tolerances In order to determine the kenr 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 arecombined statistically

[2]. The evaluations use the same MCNP5-1.51 models used in the designbasis calculation.

2.3.4.1 Fuel Manufacturing Tolerances The BWR fuel tolerances for Optima2 Q122 fuel (which is the most reactive fuel designevaluated herein) are presented in Table 5.1(a). Fuel tolerance calculations are petformed usingthe design basis fuel assembly

lattice, and therefore only the tolerances applicable to that latticeare 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 rackcalculations.

Pin specific compositions are used. The MCNP5-1.51 tolerance calculation iscompared to the MCNP5-1.51 reference case (nominal parameter values) at the 95% probability at a 95% confidence level using the following equation:

delta-kale

= (ka&2 -kclcl) +/- 2

  • q (a12 + a22)The following fuel tolerances are considered in this analysis:

" Fuel enrichment

" Gd loading* Fuel pellet density (U02 and U02+Gd2O3 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 4o Water wing canal inner widtho Channel outer square widtho Channel comer inner radiuso Central water canal inner square width* Combination of 4o channel wall thickness 3 For fuel sub-bundle pitch uncertainty calculation, the fuel hardware (channel, central water channel andwater wings) is fixed. The fuel lattices are moved only.4 Conservatively, the various tolerances are considered together.

The tolerance limits that result in anincrease 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 togetherin another set of uncertainty calculations.

Pr(ýject No. 2127Report No. H--2125245 Page 22 o Water cross wall thickness The maximum positive reactivity effect of the MCNP5-1.51 calculations for each tolerance isstatistically combined with the other tolerance

results, and this result is then statistically combined with other uncertainties when determining the kff value.2.3.4.2 SFP Storage Rack Manufacturing Tolerances The SFP rack tolerances are presented in Tables 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 thetolerance is applied for each case. The MCNP5-1.51 tolerance calculation is compared to theMCNIP5-1.51 reference case with a 95% probability at a 95% confidence level using thefollowing equation:

delta-kcac

= (kc.Ic2 -kcalcl) +/- 2 * (G12 + C22)The following SFP rack manufacturing tolerances are considered in this analysis:

" Storage cells:o Cell ID and cell pitcho Cell wall thickness

  • Rack inserts (poison)o WidthThe maximum positive reactivity effect of the MCNP5-1.51 calculations for each tolerance isstatistically combined with the other tolerance
results, and this result is then statistically combined with other uncertainties when determining the keff value.The evaluations use the same MCNP5-1.51 models used in the design basis calculation.

Theisotopic 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 biasinstead of uncertainty, for conservatism and simplification.

2.3.5 Radial Positioning 2.3.5.1 Fuel Assembly Orientation in the CoreThe fuel assembly orientation in the core with respect to its control blade does not change andtherefore the design basis calculations consider the only possible configuration.

2.3.5.2 Fuel Radial Positioning in the RackThe BWR fuel that is loaded in the SFP racks may not rest exactly in the center of the storagecell. Evaluations are performed to determine the most limiting fuel radial location.

The following eccentric fuel positioning cases are analyzed:

Project No. 2127Report No. HI-2125245 Page 23

" Case 2.3.5.2.1:

This is the reference for Case 2.3.5.2.2 through Case 2.3.5.2.5.

TheMCNP5-1.51 model used herein is a 2x2 array which is the same as the primary singlebundle MCNP5-1.51 model used elsewhere in this analysis.

In both models the fuel iscentered 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, asshown in Figure 2.7(b)." Case 2.3.5.2.3:

Every fuel assembly is positioned toward the corner where the insertwings 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 insertwings 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 2x2array, 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.

TheMCNP5-1.51 model used herein is an 8x8 array which is the same as the primary singlebundle MCNP5-1.51 model used elsewhere in this analysis.

In both models the fuel iscentered in the rack cell.* Case 2.3.5.2.7:

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

Every fuel assembly is positioned toward the corner where the insertwings connect, for the 8x8 array." Case 2.3.5.2.9:

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

Every fuel assembly is centered between insert and cell walls, for the8x8 array.* Case 2.3.5.2.11:

This is the reference for Case 2.3.5.2.12.

The MCNP5-1.51 model usedherein 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 thesingle rack cell.The maximum positive reactivity effect of the MCNP5-1.51 calculations for the fuel radialpositioning is added as the bias and the corresponding 95/95 uncertainty is statistically combinedwith other uncertainties to determine kr.fProject No. 2127Report No. HI-2 125245Page 24 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 isotopiccompositions 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 inthe positioning for inserts is evaluated.

The water gap between rack wall and insert is notassumed, since it may provide a small flux trap effect. Nevertheless, the orientation of fuelassembly with respect to position of insert is considered in Section 2.3.5.4.2.3.5.4 Fuel Orientation in SFP Rack CellAs 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 thedepletion in the core. As a result, the fuel depletion is not uniform (more discussion is providedin Section 2.3.1.1.2) and one fuel assembly corner may be more reactive than other corners andtherefore 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 themost 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 shownin Figure 2.9(a). The MCNP5.1.51 models of the other four cases are the same as that of thereference 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 SFPcells 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 asthose of the design basis fuel assembly.

2.3.6Project No. 2127Report No. HI-2125245 Page 25 2.3.6.12.3.6.1.1 Prqject No. 2127Report No. H1-2125245 Page 26 2.3.6.1.2

-2.3.6.22.3.7 Insert Coupon Measurement Uncertainty There is a measurement uncertainty associated with the B-! 0 content in the poison test coupons.

Inthis analysis, the minimum B-10 loading and the minimum insert thickness are conservatively usedfor criticality calculations.

Therefore, the coupon measurement uncertainty is not evaluated furtherin the analysis.

2.3.8 Maximum klf Calculation for Normal Conditions The calculation of the maximum kff of the SFP storage racks fully loaded with design basis fuelassemblies at their maximum reactivity is determined by adding all uncertainties and biases to thecalculated reactivity.

Note that the insert thickness and its B-I 0 loading are taken at their worst casevalues.Project No. 2127Report No. 1-11-2125245 Page 27 kefr is determined by the following equation:

keff = k¢1e + uncertainty

+ biaswhere uncertainty includes:

and the bias includesNote that each uncertainty is statistically combined with other uncertainties, while biases areadded together in order to determine keff.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 B4C loading* Minimum insert thickness

  • Minimum amount ofB-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 andCase 2.4.3.Project No. 2127Report No. 1H1-2125245 Page 28

  • Case 2.4.2: This case is the same as Case 2.4.1, except the nominal insert B4C loading,nominal insert thickness and nominal amount of B-I10 in boron are used." Case 2.4.3: This case is the same as Case 2.4.1, except the model includes each Optima2Q122 fuel lattice in the appropriate axial position.
However, the top and bottom blanketswere conservatively replaced by adjacent fuel lattices.

The peak reactivity burnup foreach individual Optima2 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 peakreactivity).

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 ofthe design basis fuel assembly.

2.5 Fuel Movement, Inspection 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 movementProject No. 2127Report No. 1-1-2125245 Page 29 Those are briefly discussed in the following sections.

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."

Thisprinciple precludes the necessity of considering the simultaneous occurrence of multiple accidentconditions.

The kfl" calculations perfomaed 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 from the designbasis cas.2.6.1 Temperature and Water Density EffectsThe SFP water temperature accident conditions for consideration are the increase in SFP watertemperature above the maximum SFP operating temperature of 150 'F. The decrease in temperature was already considered for 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 isevaluated:

Case 2.6.1: This case uses a temperature of 255 'F (397.04 K) and a density of 0.84591g/cc. The S(a,3) card corresponds to a temperature of 260.33 'F (400 K). In this model, itis assumed that the water modeled includes 10% void. Void is modeled as 10% decreasein 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 asinserts, show a higher reactivity at a lower water temperature.

The case evaluated above isperformed 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 assemblywill 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 from the fuel assemblies in the rack. This accidentis also bounded by the mislocated case, where the mislocated assembly is closer to the assemblyin the racks. Therefore, the horizontally dropped fuel assembly is not evaluated further in thereport.2.6.3 Dropped Assembly

-Vertical into a Storage CellIt is also possible to vertically drop an assembly into a location that might be occupied by anotherassembly 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.Project No. 2127Report No. 1-11-2125245 Page 30 These deformations could potentially increase reactivity.

However, the reactivity increase would besmall 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 verticaldrop is therefore bounded by this misload accident and no separate calculation is performed for thisdrop accident.

2.6.4 Storage Cell Distortion A storage cell distortion or altered geometry as a result of fuel handling equipment uplift forces ispossible.

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 theinsert 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 isperformed for the storage cell distortion accident.

As a result of significant distortion, the storage cell for whatever reason may not be able to containthe insert and also it will be therefore unacceptable for storage of a fuel assembly.

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

However to show that it isacceptable 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 8x8array, is evaluated.

Two cases with a cell centered and eccentric position of the fuel assemblies areanalyzed.

2.6.5 Misloaded Fuel Assembly/Missing InsertThe 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 aninsert.However, there are a few 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 fuelassembly is misloaded into a cell with a missing insert. To evaluate the effect, the following casesare evaluated:

" Case 2.6.5.1:

The MCNP5-1.51 model includes an 8x8 array. One cell near the center of therack 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 fuelassemblies are also eccentric toward the misloaded fuel assembly.

The periodic boundaryconditions 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 itscorresponding water density and S(a,3) card. These temperature and density are boundingfor 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 fuelassemblies centered in the rack cells. See Figure 2.10(b).Project No. 2127Report No. I-1-2125245 Page 31 2.6.6 Mislocated Fuel AssemblyThe Quad Cities SFP layout was reviewed to determine the possible worst case locations for amislocated 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 WallA fuel assembly may be mislocated in the water gap between the racks and the pool wall. Due to theneutron leakage to the outside the storage rack area, the effect of this mislocation is bounded by thatof'mislocation of a fuel assembly between the SFP rack and the inspection platform'

accident, asdiscussed 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 RacksThere are some places in the SFP, but outside of the racks, where the mislocated fuel assembly maybe in the corner between two racks (thus the mislocated fuel assembly would be adjacent to the fuelassemblies in racks from two sides). To evaluate the effect of the mislocation of a fuel assembly inthe 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 SFP rack cells. Themisplaced fuel assembly is in the corner between two racks. The fuel assemblies in the rackare eccentric toward the mislocated fuel assembly.

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

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

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 fuelassemblies are centered.

See Figures 2.1 1(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 thetemperature 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 thetemperature 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 PlatformAs discussed in Section 2.5, the fuel handling/inspection/reconstitution platform may have onefuel assembly in it at a time. There is a possibility that a fuel assembly is mislocated between theProject No. 2127Report No. HI-2125245 Page 32 SFP racks and the fuel assembly in the platform.

To evaluate the effect of the mislocation of a fuelassembly between the SFP Rack and the Inspection

Platform, the following cases are evaluated:

" 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 inthe platform is lined up with the mislocated fuel assembly.

The fuel assemblies in the rackare eccentric toward the mislocated fuel assembly.

The misplaced fuel assembly is placed asclose 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 doesnot have any fuel has at least 12 inches of water. Figure 2.12(a) shows the MCNP5-1.51 model used for 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 fuelassemblies 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 thetemperature of the model is set to the maximum (150 TF)." Case 2.6.6.3.4:

The MCNP5-1.51 model is the same as Case 2.6.6.3.2, except thetemperature of the model is set to the maximum (150 'F).2.6.7 Mis-installment of an Insert on Wrong Side of a CellThere is a small possibility that an insert is installed on wrong sides of the cell. In this case, theremay not be a poison between a fuel assembly placed in that cell and a fuel assembly in anadjacent cell. However, the effect of this mis-installment is bounded by that of 'misloaded fuelassembly/missing insert' accident that does not include the insert in one rack cell, as discussed inSection 2.6.5. No separate calculation is performed for this accident.

2.6.8 Insert Mechanical WearHanding accidents and other environmental damage may cause scratches and local wear ofinserts.

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 MovementIn the event of seismic activity, there is a hypothetical possibility that the storage rack arraysmay move and come closer to each other. Since there is no water gap modeled between cells of astorage rack, the reactivity of the rack movement case is bounded by the reactivity of the designbasis calculation.

2.7Project No. 2127Report No. HI-2125245 Page 33 2.8 Spent Fuel Rack Interfaces The spent fuel pool includes a single type of Region I spent fuel racks, which are loaded with theneutron absorbing inserts in every storage cell as well as a uniform fuel assembly loading pattern.Therefore, any possible water gaps and interfaces between the racks are bounded by the infinitearray used in the design basis calculations.

However, since the neutron absorbing inserts are locatedin the same corners 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 ofthe spent fuel pool (correspondingly, north-east corner) has two sides that are not adjacent to theinsert. Due to the neutron leakage on the periphery of the spent fuel pool the reactivity increase isnot 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 allfuel assemblies are shifted to the fuel assembly in the corner, which is discussed above, wereevaluated.

Project No. 2127 Report No. HI-2125245 Page 34 2.9 Reconstituted Fuel Assemblies The SFP contains various reconstituted assemblies which were examined and determined to berelatively old and low reactivity designs.

The reconstitution of these fuel assemblies removed fuelrods and replaced them by either fuel rods that are of the same or less initial enrichment and equalor 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 assemblymore reactive than the bounding lattice.

Therefore, reconstituted assemblies are covered by thedesign basis Optima2 Q122 lattice 146. Future reconstituted assemblies will replace fuel rods withstainless steel rods.Project No. 2127Report No. 111-2125245 Page 35

3. ACCEPTANCE CRITERIA3.1 Applicable Codes, Standards and Guidance's Codes, standard, and regulations or pertinent sections thereof that are applicable to theseanalyses 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 Freshand Spent Fuel Storage and Handling, Revision 3 -March 2007.* L. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of FuelStorage 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 andTransportation of LWR Fuel Outside Reactors (withdrawn in 2004).* USNRC, NUREG/CR-6698, Guide for Validation of Nuclear Criticality SafetyCalculational Methodology, January 2001." DSS-ISG-2010-01, Revision 0, Staff Guidance Regarding the Nuclear Criticality SafetyAnalysis for Spent Fuel Pools.Project No. 2127Report No. HI-2125245 Page 36

4. ASSUMPTIONS The analyses apply a number of assumptions, either for conservatism or to simplify thecalculation 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 theselected inputs and parameters are in fact 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 becausethis 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 tomoderator ratio (Section 7.1.7.1);

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

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

3. The neutron absorber length in the rack is more than the active region of the fuel, but it ismodeled 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 andchamfering.

This is acceptable since the amount of fuel modeled is more than the actualamount.5. For the inserts, only the worst case bounding material specifications are used (minimumB-I 0 loading and minimum thickness).

6. All models are laterally infinite arrays of the respective configuration, neglecting lateralleakage.

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 claddingconsists of one of several zirconium alloys. This is acceptable since the model neglectsthe 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. The watergaps between the spent fuel racks were conservatively neglected.

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

5. INPUT DATA5.1 Fuel Assembly Specification The SFP racks are designed to accommodate the following fuel assembly types used in the QuadCities Unit I and Unit 2, which are presented in a chronologic order along with the initialmaximum 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 Figure 5.4.The fuel rod, cladding and channel are explicitly modeled.,
Axially, the design basis MCNP modelconsiders the bounding lattice along the entire length and uses water reflectors at the top andbottom. The MCNP model for the margin evaluation calculations discussed in Section 2.4 differfrom the design basis model in that the active length specifically considers each actual lattice inits actual axial configuration (i.e. all the lattices from the Q122 bundle are modeled in the sameMCNP mrNAA1\5.2 Reactor Parameters The reactor core parameters are provided in Table 5.2(a). The reactor control blade data areprovided in Table 5.2(b). The reactor control parameters used in CASMO-4 screening and designbasis calculations are provided in Table 5.2(c).5.3 Spent Fuel Pool Parameters The spent fuel pool parameters are provided in Table 5.2(a).

5.4 Storage Rack Specification The storage rack specifications that are used in the criticality analysis are summarized in Tables5.3(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.

model consists of a single rack cell with periodic boundary conditions through the centerline ofthe water (BORAFLEX replacement),

thus simulating an infinite array of storage cells. Thestorage 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 (theminimum 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 wallthickness of the boundary is thicker than the inner wall thickness.

The SF1P model uses the innercell wall thickness only, as given in Table 5.3(a), because it decreases the amount of steel in themodel, which acts a neutron absorber.

The MCNI15-1.51 SFP rack cell model is shown in Figure 5.4.5.4.1 Material Compositions The MCNP5-1.51 material specification is provided in Table 5.4(a) for non-fuel materials, and inTable 5.4(b) for fuel materials.

Project No. 2127Report No. 1-I1-2125245 Page 39

6. COMPUTER CODESThe following computer codes were used in this analysis.
  • MCNP5-1.51

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

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

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. 2127Report No. f11-2125245 Page 40

7. ANALYSIS7.1 Design Basis and Uncertainty Evaluations 7.1.17.1.2 Determination of the Design Basis Fuel Assembly LatticeAs discussed in Section 2.3.1.3, MCNP5-1.51 calculations were performed to determine thedesign basis lattice.

The results for the SVEA-96 Optima2 Q122 lbel assembly are presented inTable 7.2(a) .The results for the GE 14 lattice type 5 are presented in Table 7.2(b), along with thebounding result of the SVEA-96 Optima2 Q122. As can be seen, the SVEA-96 Optima2 Q122lattice type 146 is bounding, and thus it is selected as the design basis lattice.

The CASMO-4model of the SVEA-96 Optima2 bundle Q122 lattice 146 used for depletion calculations is shownin Figure 5.1.7.1.2.1 Fuel Assembly De-Channeling As discussed in Section 2.3.1.5.4, the reactivity of the second most reactive assembly with no Zrchannel at various radial positioning was evaluated.

The results are provided in Table 7.2(b) andcompared 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 offuel assemblies without channels is acceptable.

7.1.3 Optima2 CASMO-4 Model Simplification EffectAs 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 inTable 7.3. As can be seen, the reactivity of the simplified model is comparable to that of thecomplete model of SVEA-96 Optima2 Q122 lattice 146 (essentially within the 95/95 uncertainty between the two calculations).

Therefore, the results show that the CASMO-4 model simplification Project No. 2127Report No. HI-2125245 Page 41 does not have a significant impact on the analysis conclusions regarding the determination of thedesign 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 wereevaluated.

The results are provided in Table 7.4. The results show that the two dominant coreoperating 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 UprateAs discussed in Section 2.3.1.5.1, the effect of the MUR on the reactivity was evaluated.

The resultsare 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 EffectAs discussed in Section 2.3.2, the effects of water temperature, and the corresponding waterdensity and temperature adjustments (S(a,3))

were evaluated for SFP racks. The results of thesecalculations are presented in Table 7.5.The results of the SFP temperature and density calculations show that as expected (for poisonedracks) the most reactive water temperature and density for the SFP racks is a temperature of39.2 'F 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 inTable 7.11 and Table 7.14.Project No. 2127Report No. 111-2125245 Page 42 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 wasdetermined.

The results of these calculations are presented in Table 7.7. The maximum positivedelta-k value for each tolerance is statistically combined.

The maximum statistical combination of fuel assembly tolerances is used to determine keff inTable 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 theSFP racks with inserts was determined.

The results of these calculations are presented in Table7.8. 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 kerr inTable 7.11 and Table 7.14.7.1.8 Radial Positioning 7.1.8.1 Fuel Assembly Radial Positioning in SFP RackAs discussed in Section 2.3.5.2, twelve fuel assembly radial positioning cases in racks wereevaluated.

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 resultsshow that most cases show a negative reactivity effect, however some delta k, 1 values arepositive.

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

Theresults of these calculations are presented in Table 7.9(b). The result for the reference case is alsoincluded.

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

Nevertheless, a maximum positive delta value is applied as abias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties inTable 7.11 and Table 7.14.Project No. 21.27Report No. HI-2125245 Page 43 7.1.9 Fuel Rod Geometry Change7.1.9.1i ne results are presenteo in i aoe /. 1 v.The maximum 'ke,,ac -cIcrercnce is added as a bias, and the '2 *

+ G3cf"2cn, (95/95uncertainty) is added as an uncertainty to determine kerr in Table 7.11 and Table 7.14.7.1.9.27.1.107.2 Maximum kff Calculations/for Normal Conditions As discussed in Section 2.3.8, the maximum kfrf for normal conditions is calculated.

The results aretabulated in Table 7.11. The results show that the maximum keff for the normal conditions in theSFP 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 forpoison thickness and loading, as well as the actual lattice configuration of the Optima2 Q122 fuelassembly.

The results of calculations are provided in Table 7.12(a) and Table 7.12(b).

As can beseen and is expected, the reactivity of design basis is larger. The use of a minimum B-10 loadingrelative 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. 2127Report No. HI-2125245 Page 44 The summary of the margin evaluation is presented in Table 7.12(c).

The result shows thatquantified margin remains in the analysis to offset potential effects not already considered in themodel.7.4 Abnormal and Accident Conditions As discussed in Section 2.6, the effects of empty storage cell, increased temperature, misloaded fuelassembly/missing insert, and mislocated fuel assembly accidents on reactivity were evaluated.

Theresults 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 increasein reactivity.

For the SFP racks, the effect onl reactivity of the missing insert is the limiting case.Thus, its calculated MCNP5-1.51 k.lc is used for maximum keff calculations for abnormal andaccident conditions, discussed in Section 7.5.The condition with the empty storage cell without insert in the spent fuel rack shows a lowerreactivity than a design basis case, therefore, it is acceptable to have the empty storage cell withoutinsert in the spent fuel pool.7.5 Maximum keff Calculations for Abnormal and Accident Conditions As discussed in Section 2.6, the maximum keff 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 andaccident conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95%confidence level.7.67.7 Spent Fuel Rack Interfaces As discussed in Sections 2.8, the interface between SFRs and pool walls, i.e. effect on reactivity ofthe peripheral fuel assemblies, that have a side non-adjacent to the insert, was evaluated.

The resultsare provided in Table 7.17. As can be seen, this condition will not result in an increase of SFRreactivity.

This result is expected because the infinite array design basis model is an infinite array ofProject No. 2127Report No. HI-2125245 Page 45 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. noadditional cell with an insert).

Therefore, this water gap edge allows for neutron leakage and as thecalculations show result in statistically equivalent results.Project No. 2127Report No. HI-2125245 Page 46

8. CONCLUSION The criticality analysis for the storage of BWR assemblies in the Quad Cities SFP racks withNETCO-SNAP-IN 6 inserts has been performed.

The results for the normal condition show thatl~ff is = 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 highestreactivity.

The results for the accident condition show that k',T is M with the storage racksfully loaded with fuel of the highest anticipated reactivity, which is SVEA-96 Optima2, at a temperature corresponding to the highest reactivity.

The maximum calculated reactivity for both normal and accident conditions includes a margin for uncertainty in reactivity calculations with a 95% probability at a 95% confidence level. Reactivity effects of abnormaland accident conditions have been evaluated to assure that under all credible abnormal andaccident conditions, the reactivity will not exceed the regulatory limit of 0.95.Pro.ject No. 2127Report No. HI-2125245 Page 47

9. REFERENCES

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

[2] L.I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of FuelStorage 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 HI-2104790 Revision 1.[4] M. Edenius, K. Ekberg, B.H. Forss6n, and D. Knott, "CASMO-4 A Fuel AssemblyBurnup Program User's Manual,"

Studsvik/SOA-95/1; and J. Rhodes, K Smith,"CASMO-4 A Fuel Assembly Burnup Program User's Manual,"

SSP-01/400, Revision 5,Studsvik of America, Inc. and Studsvik Core Analysis AB3 (proprietary).

[5] D. Knott, "CASMO-4 Benchmark Against Critical Experiments,"

SOA-94/13, Studsvikof 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 forSpent Fuel Pools, Revision 0.[7] Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001.[8] HI-2002444, Latest Revision, "Final Safety Analysis Report for the HI-STORM 100Cask System",

USNRC Docket 72-1014.[9] "Sensitivity Studies to Support Criticality Analysis Methodology,"

H1-2104598 Rev. 1,October 2010.[10] "Atlas of Neutron Resonances",

S.F. Mughabghab, 5th Edition, National Nuclear DataCenter, 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 (kif) Predictions, NUREG/CR-7109, April 2012.[13] OECD / NEA Data Bank, Java-based Nuclear Information

Software, Janis version 3.3.Project No. 2127Report No. HI-2125245 Page 48

[14] EPRI 1003222, "Poolside Examination Results and Assessment, GEl I BWR FuelExposed to 52 to 65 GWd/MTU at the Limerick 1 and 2 Reactors,"

December 2002.Project No. 2127Report No. HI-2125245 Page 49

-] -Iim I m-mI j --* I.m Immm I m-f f-+ 4-n

,,m Ii+/- 4-mmUmt -t I --_________

a. _______________________________

+/- 4 L ..Project No. 2127Report No. HI-2125245 Page 50 mJ6I NmIUmProject No. 2127Report No. HI-2125245 Page 51 Project No. 2127 Report No. HI-2125245 Page 52

-i -Project No. 2127Report No. .HI-2125245 Page 53 El9......-.-

IIIi-I-I-lIl---li-

=I I I I I I -I I I I is -Project No. 2127Report No. 111-2125245 Page 54 ALL----IrIElIIIII II' I I I Iii1I I 4 II -I IM I -IMIMMlProject No. 2127Report No. 1H1-2125245 Page 55 Project No. 2127Report No. HI-2 125245Page 56 IProject No. 2127Report No. HI-2125245 Page 57 Project No. 2127Report No. HI-2125245 Page 58 Project No. 2127Report No. HI-2125245 Page 59 Project No. 2127Report No. HI-2125245 Page 60 Project No. 2127 Report No. H1-2125245 Page 61 I1 -1Ur357LIZLIZ~IIIIPr(ýIect No. 2127Report No. HI-2125245 Page 62 Project No. 2127 Report No. 1.tl-2125245 Page 63 Project No. 2127Report No. H1-2125245 Page 64 I J I3I I IU;JIProject No. 2127Report No. Hf-2125245 Page 65

-~ I. I-mIlV-I-I-I-I.a'IILI::::Project No. 2127Report No. HI-2125245 Page 66

! MillLUIN 011 IN m11 IN 0mi IN m1111, IN 0M",ImmI M--7m- --MMM-m-----Mil 101JM mProject No. 2127Report No. HI-2125245 Page 67 Table 7.2(a)Results of the MCNP5-1.51 Calculations for SVEA-96 Optima2 Q122 LatticesNote 2: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta k,,,, may occur atan exposure which differs from that shown above.Project No. 2127Report No. HI1-2125245 Page 68 Table 7.2(a) Continued BurnupDescription (GWd/mtU) 151617Lattice 149 18(void) 1920211516Lattice 149 17(water) ,t 18192021151617Lattice 150 18192021151617Lattice 151 18192021itaC sigma Maxdelta k,,,,Uncert.(95/95)-0.0207 0.0016-0.0189 0.0016-0.0154 0.0016--0.0111 0.0016Note 2: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta k,01Imay occur at an exposure which differs from that shown above.Project No. 2127Report No. 2125245 Page 69 Table 7.2(b)Results of the MCNP5-1.51 Calculations for GE14 Lattice Type 5Description Burnup k Uncert.(GWd/mtU) kdV sigma delta k (95/95)SVEA-96 Optima2 Q122 15.5 Reference Reference lattice type 146Single GEl4 13 -0.0543 0.0016Single GE14 13.5 -0.0509 0.0015Single GEl4 14 -0.0491 0.0016Single GEI4 14.5 1 -0.0469 0.0015Single GEl4 15 -0.0473 0.0015Single GE14 15.5 -0.0479 0.0015Single GE14 16 -0.0485 0.0015Single GE14 16.5 1___l_-0.0482 0.0015Single GEl4 17 -0.0500 0.00152x2 GE14 -with channel (cell 14.5 Reference Reference centered)

(Case 2.3.1.5.4.1) 14.5 __Rfec Rern2x2 GE 14 -no channel 14.5 -0.0044 0.0016_(Case 2.3.1.5.4.2) 2x2 GE14 -no channel /eccentric 14.5 -0.0173 0.0015center (Case 2.3.1.5.4.3) 2x2 GE1 4 -no channel /eccentric 14.5 -0.0238 0.0015out (Case 2.3.1.5.4.4)

Note 2: The result of the SVEA-96 Optima2 Q122 lattice type 146 is provided as the reference.

Note 3: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kcalmay occur at an exposure which differs from that shown above.Project No. 2127Report No. HI-2125245 Page 70 Table 7.3Results of the MCNP5-1.51 Calculations for Design Basis and Simplified Model of SVEA-96 Optima2 Q122 Lattice Type 146Burnup Code kalc sigmaDescription (GWd/mtU)

Simplified model of SVEA-96Optima2 Q122 lattice 146 15.5 CASMO-4 1(Case 2.3.1.4.1)

Simplified model of SVEA-96Optima2 Q122 lattice 146 15.5 MCNP5-1.51 (Case 2.3.1.4.2)

Model of SVEA-96 Optima2 Q122lattice 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. 2127Report No. HI-2125245 Page 71 Table 7.4Results of the MCNP5-1.51 Calculations for Core Operating Parameters Power Fuel Moderat Void BurnupDescription Density Temp. or Temp. Fraction (GWd/ k,,,, sigma delta Uncert.(W/gU) Blade .K) (°FL ) mtU) kcAc (95/95)Design basis 23.688 Yes 1176 547 0 15.5 l l(reference)

Fuel temperature 23.688 Yes 588 547 0 16decreasing Moderator temperature 23.688 Yes 1176 528.8 0 15.5decreasing Void fraction 23.688 Yes 1176 547 94 22increasing Un-rodded operation 23.688No1176547017-il I/-24.1617 Yes 1276 547 0 15.5 1I 2IIL______ Yes_376_47_0

_5. ____ ____20.1348Yes1176547015.5-lIIl-lNote 1: The burnup calculations for core operating parameters were perfonrmed from 14GWd/mtU to 24 GWd/mtU.

For each core operating parameter, only reactivity of the burnup inthis 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,,,may occur at an exposure which differs from that shown above.Project No. 2127Report No. 1-11-2125245 Page 72 Table 7.5Results of the MCNP5-1.51 Calculations for the Effect of Water Temperature and DensityWater Water Temperature Description Burnup Temp. Density Adjustment, Unrtsigma delta k,, (95/5.(GWd/mtU)

(OF) (g/cc) S(aF) (95/95)(OF)

Reference:

lowerbound temperature 15.5 39.2 1 68.81 Reference Ref.(Case 2.3.2.1)Upper boundtemperature fornormal operation, low 15.5 150 0.98026 68.81 -0.0041 0.0015S(a4)(Case 2.3.2.2)Upper boundtemperature fornormal operation, 15.5 150 0.98026 170.33 -0.0066 0.0015high S(a,3)(Case 2.3.2.3)Note 1: The maximum calculation uncertainty (sigma) used tomay occur at an exposure which differs from that shown above.determine the 95/95 delta k.,,Project No. 2127Report No. HI-2125245 Page 73 Table 7.6(a)Results of the MCNP5-1.51 Calculations for the Depletion Uncertainty Depletion Description kcilc sigma Uncertainty (5%)Design basis Reference Fresh fuel, no Gd 0.0064Project No. 2127Report No. 141-2125245 Page 74 m-- I~1 ~ 1~-4 -4 4-= I =Project No. 2127Report No. .11-2125245 Page 75 Table 7.7Results of the MCNP5-1.51 Calculations for Fuel Tolerances Description PeakReactivity Burnup(GWd/mtU) k.1,sigmadelta kcnie(95/95)Max delta kca,,(95/95)Design basis (reference) 15.5Max fuel enrichment 16Min fuel enrichment 15.5Max Gd loading 16Min Gd loading 15.5Max pellet density 16Min pellet density 15.5Max pellet OD 15.5Min pellet OD 16Max clad ID 16Min clad ID 16Max clad OD 15.5Min clad OD 15.5Max sub-bundle pitch 15Min sub-bundle pitch 16.5Max pin pitch 15.5Mill pin pitch 15.5Reference Reference 0.0026* 0.0026-0.0009-0.00130.00380.0038 _____0.00000.00 120.00o 12 0 .0010.00150.00150.00110.00 100 0.00100 .0008 0 .0098-0.00020.00270.00270.00980.0098-0.0089 _____0.01220.0122-0.0086 _____Max combined water wing canalinner width, channel outer squarewidth, channel corner inner radiusand central water canal innersquare width15FI -0.00310.0031Min combined water wing canalinner width, channel outer squarewidth, channel corner inner radiusand central water canal innersouare width15.5-I-0.0011Max combination of channel wallthickness and water cross wall 16 l 0.0008thickness 0.0019Min combination of channel wallthickness and water cross wall 15.5 0.0019thickness Statistical combination of fuel tolerances 0.0171Note 1: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kclcmay occur at an exposure which differs from that shown above.Project No. 2127Report No. HI-2125245 Page 76 Table 7.8Results of the MCNP5-1.51 Calculations for Rack Tolerances Burnup delta kai, Max deltaDescription (GWd/mtU) k,,& sigma (95/95) kcle(95/95) (95/95)Design basis 15.5 Reference Reference (reference)

Max cell ID 15.5 -0.0093 N/AMax cell pitchMax wall thickness 15.5 0.0025Min wall thickness 15.5 _ 0.0008Max insert width 15.5 -0.0005 0.0004Min insert width 15.5 0.0004Statistical combination of rack tolerances 0.0026Project No. 2127Report No. 1-11-2125245 Page 77 Table 7.9(a)Results of the MCNP5-1.51 Calculations for Fuel Radial Positioning in SFP RacksBurnupUnc.

Description BurnU) kualc sigma delta kc.I, (95/95)(GWd/mtU) 95/95)_2x2 reference 15.5 Reference Ref.(Case 2.3.5.2.1) 15.5 _lRenc e2x2 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 15.5 1 --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)

15. ___efeenc Ref.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)

........1 .__0 0 5 .18x8 insert/cell center 15.5 0.0016 0.0014(Case 2.3.5.2.10)

.......I x I reference 15.5 Reference Ref(Case 2.3.5.2.11)

IxI insert/cell center 15.5 0.0000 0.0015(Case 2.3.5.2.12)

Project No. 2127Report No. 1-11-2125245 Page 78 Table 7.9(b)Results of the MCNP5-1.51 Calculations for Fuel Orientation in SFP RacksBurnup IUnc.Description Burnup k,,,, sigma delta Iellc (95/95)(GWd/mtU) 9595Reference (Shown in 15.5 Reference Ref.Figure 2.9(a)) ........Rotated fuel assembly(shown in Figure 2.9(b)) 15.5 -0.0008 0.0014Rotated fuel assembly(shown in Figure 2.9(c)) 15.5 -0.0007 0.0014Rotated fuel assembly(shown in Figure 2.9(d)) 15.5 -0.0013 0.0013Rotated fuel assembly(shown in Figure 2.9(e)) , 5 -0.0007 0.0013Project No. 2127Report No. 1-11-2125245 Page 79 S-,. --mWm nm Mým m mnmmm mmn UUm -- nProject No. 2127Report No. 1-11-2125245 Page 80 I miI. _________________

I ..."mI- ___________________

Project No. 2127Report No. HI-2125245 Page 81 Table 7.12(a)Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate the Effect of Nominal Values Instead ofUsing Minimum B4C Loading and Minimum Insert Thickness on Reactivity B-10 ArealDescription Burnup Density k,,, sigma delta kIc(GWd/mtU)

(g/cm2)Reference (designbasis) 15.5 0.0116 Reference (Case 2.4.1)Rack with nominalvalues for 134Cloading and insert 15.5 0.0133 1 -0.0103thickness (Case 2.4.2)Project No. 2127Report No. 111-2125245 Page 82 Table 7.12(b)Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate the Effect of theActual Optima2 Q122 Fuel AssemblyDescription lurnu) kenic sigma Max kcille delta k,,,,Desripion (GWd/mtU)

Optima2 Q122Lattice 146 (Designbasis)(Case 2.4.1)15.5= I = I Reference IReference 15Optima2 Q122 155Lattice 147'1616Optima2 Q122 16.5Lattice 1481714Optirna2 Q122 14Lattice 149 14.51514_Optima2 Q122 14.5Lattice 1501514Optima2 Q122 14.5Lattice 151150.88730.8843U, 0 J0.88630.88760.8925Optima2 Q122Fuel Assemblyt(Case 2.4.3)PeakReactivity Burnups(bolded)~1--0.0066t The toplattice.and bottom naturalblankets were conservatively neglected and replaced by adjacentProject No. 2127Report No. HI-2125245 Page 83 Table 7.12(c)Margin Evaluation Summary of the Margin Evaluation Description ValueInsert Composition Margin, from Table 7.12(a) -0.0103Actual Optima2 Fuel Assembly Margin, from -0.0066Table 7.12(b)Calculated Margin -0.0169Prqjject No. 2127Report No. 2125245 Page 84 V'mImU I m I Im" I mm--mmm m --F~ m mm-nU m --P~ m m --m- m --mmmmm0mm. m ImmProject No. 2127Report No. 1-11-2125245 Page 85 Table 7.13(b)Results of the MCNP5-1.51 Calculations for the Empty Storage Rack Cell without InsertDescription Burnup keli sigma delta kcal Uncertainty

_________(GWd/mtU)

(95195)Design basis 15.5 Reference Reference (8x8 array)Empty storagecell (cell 15.5 -0.0041 0.0016centered)

Empty storage 15.5 -0,0081 0.0014cell (eccentric) 15.5,_ I -0.0081 _0.0014Note 1: The design basis fuelcalculations.

assembly (Optima2 Q122 Lattice Type146) is used for theseProject No. 2127Report No. HI-2125245 Page 86

-m~IhhEI~im m11EE1~Ihh~n~~m

~ -r .rnuim m_ __ _mEl-_m_I~.m-m-m_Project No. 2127Report No. HI-2125245 Page 87 Project No. 2127Report No. 1.11-2125245 Page 88 Table 7.16Results of the MCNP5-1.51 Calculations for Axially Infinite Optima2 Q122 LatticesBurnup kcai kc.f Delta-K Uncertainty Description (GWd/mtU)

(reference)

(infinite)

Optima2 Q122Lattice 146 15.5 0.0013 0.0015(Design basis) .....Optima2QI22 15.5 0.0018 0.0015Lattice 147Optima2 QI22 16.5 0.0008 0.0014Lattice 148Optima2QI22 14.5 0.0011 0.0015Lattice 149Optima2 Q122 14.5 0.0027 0.0014Lattice 150Optima2 Q122 14.5 0.0010 0.0015Lattice 151 -F 14.5 0.0010 0.0015Note: The difference between the MCNP models under the "reference" column and the MCNP models under the "infinite" column isdescribed in Section 5.1.Project No. 2127Report No. HI-2125245 Page 89 Table 7.17Results of the MCNP5-1.51 Calculations for SFR Interface Description Burnup k,,, sigma delta k,.,, Uncertainty (GWd/mtU) a(95/95)Design basis 15.5 Reference Reference Full SFP (cell 15.5 -0.0008 0.0016centered) 15.5_1 _0.08 .01Full SFP(eccentric to 15.5 -0.0053 0.0015SFP corner)Prqject No. 2127Report No. 1H11-2125245 Page 90 at*1*Project No. 2127Report No. 2125245 Page 91 Figure Proprietary Project No. 2127Report No. HI-2125245 Page 92 Figure Proprietary Project No. 2127Report No. HI-2 125245Page 93 Figure Proprietary ProJect No. 2127Report No. 1-11-2125245 Page 94 Figure Proprietary Project No. 2127Report No. 1-11-2125245 Page 95 Figure Proprietary Project No. 2127Report No. -1.1-2125245 Page 96 Figure Proprietary Project No. 2127Report No. 1-11-2125245 Page 97 Figure Proprietary Project No. 2127Report No. 111-2125245 Page 98 Figure Proprietary Project No. 2127Report No. 1-1-2125245 Page 99 Figure Proprietary Project No. 2127Report No. 1-11-2125245 Page 100 Figure Proprietary Project No. 2127Report No. HI-2 125245Page 1. 0 1 Figure Proprietary Project No. 2127 Report No. 2125245 Page 102 Figure Proprietary Project No. 2127Report No. 1-1-2125245 Page 103 Figure Proprietary Project No. 2127 Report No. HI-2125245 Page 104 Figure Proprietary Project No. 2127Report No. HI-2125245 Page 105 Figure Proprietary ProJect No. 2127Report No. 1-11-2125245 Page 106 Figure Proprietary Project No. 2127Report No. 2125245 Page 107 Figure Proprietary Project No. 2127Report No. HI-2125245 Page 108 Figure Proprietary Project No. 2127Report No. HI-2125245 Page 109 Figure Proprietary Project No. 2127Report No. 2125245 Page 110 Figure Proprietary Project No. 2127Report No. HI-2125245 Page I I I Figure Proprietary Project No. 2127Report No. 1-11-2125245 Page 112 Figure Proprietary Project No. 2127Report No. 111-2125245 Page 113 Figure Proprietary Project No. 2127Report No. H-ti-2125245 Page 114 Figure Proprietary ProJect No. 2127Report No. HIl -2125245Page 1 15 Figure Proprietary Project No. 2127Report No. HI-2 125245Page 116 Figure Proprietary Project No. 2127Report No. 1-11-2125245 Page 117 Figure Proprietary Project No. 2127Report No. HI-2125245 Page H18 Appendix AProprietary Appendix BProprietary Project No. 2127Report No. HI-2125245 Page B-1 Appendix CProprietary Project No. 2127Report No. 1-H-2125245 Page C- I Supplement 1Additional Calculations to Support the Revised NETCO-SNAP-IN Rack Insert Design(I 1 pages including this page)Project No. 2127Report No. HI-2125245 Page S I -I1 SI.1 Introduction This Supplement documents the criticality safety evaluation for the storage of spent BWR fuel inthe Unit I and Unit 2 spent fuel pools (SFPs) at Quad Cities Station operated by Exelon. Thepurpose of this analysis is to justify that the specified changes in the NETCO-SNAP-IN rackinsert design [Sl.1 j are acceptable and bounded by the current analysis, presented in the mainpart of the report.S1.2 Methodology See Section 2 of the main report and as otherwise discussed below.S1.3 Acceptance CriteriaSee Section 3 of the main report.S1.4 Assumptions See Section 4 of the main report and as otherwise discussed below.S1.5 Input DataSee Section 5 of the main report. The revised dimensions of the NETCO-SNAP-INO rack insertare presented in Table SI -I and Figure SI -I.S1.6 Computer CodesSee Section 6 of the main report.S1.7 AnalysisThe comparison of the revised insert parameters presented in Table SI-1 with the previous insertdesign in Table 5.3(b) shows that changes are minor and therefore a significant impact on theconclusions made in the main part of the report is not expected.

Nevertheless, to verify thenegligible or minor impact of the revised insert design on results presented in the main part ofthe 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 rackThese cases are selected because the NETCO-SNAP-IN rack insert design change may impactthe reactivity in the rack. All other calculations from the main report are not affected by theNETCO-SNAP-INO rack insert design change and the results of the unaffected calculations areProject No. 2127Report No. 1-11-2125245 Page S 1 -2 used in this Supplement where applicable.

This approach is considered for both normal andaccident conditions.

S 1.7.1 SFP Rack Tolerances As discussed in Section S1.7, the effect of the manufacturing tolerances on reactivity of the SFPracks with revised inserts was determined.

The results of these calculations are presented inTable S 1-2. 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 kcfr inTFable S 1-5 and Table S1-6.S 1.7.2 Fuel Assembly Radial Positioning in the SFP RackAs discussed in Section S1.7, twelve fuel assembly radial positioning cases in the racks wereevaluated.

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 resultsshow that most cases show a negative reactivity effect, however some delta k,,,, values arepositive.

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

Theresults of these calculations are presented in Table S1-4. The result for the reference case is alsoincluded.

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

Nevertheless, a maximum positive delta k,1, value is applied as abias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties inTFable S1-5 and Table S1-6.S 1.7.4 Maximum krff Calculations for Normal Conditions The calculations of the maximum kef. for normal conditions are described in Section 2.3.8 of themain part of the report. The results for the revised NETCO-SNAP-IN rack insert design and theresults from the main part of the report are tabulated in Table S1-5. The results show that themaximum ktfr 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-IN rack insert design and arebounded by the results from the main part of the report.Project No. 2127Report No. HI-2125245 Page S 1 -3 S 1.7.5 Maximum kdr Calculations for Abnormal and Accident Conditions The calculations of the maximum k.fr for accident conditions are described in Section 2.6 of themain part of the report. The bounding accident case from the main report is recalculated using therevised NETCO-SNAP-IN rack insert design. The results for the revised NETCO-SNAP-INO rack insert design and the results from the main part of the report are tabulated in Table S1 -6. Theresults show that the maximum k~ff for abnormal and accident conditions in the SFP racks is lessthan 0.95 at a 95% probability and at a 95% confidence level for the revised NETCO-SNAP-IN rack insert design and are bounded by the results from the main part of the report.S1.8 References

[S .1] Transmittal of Design Information NF1 100434, Revision 1, "Quad Cities SFP Rack InsertDesign Information",

dated 09/i 1/2012.S1.9 Conclusions The criticality analysis for the storage of BWR assemblies in the Quad Cities SFP racks withrevised NETCO-SNAP-INO inserts has been performed.

The results show that kefr is M withthe stora racks full loaded with fuel of the highest anticipated reactivity, which is SVEA-96Optima2 , at a temperature corresponding to the highest reactivity.

The maximumcalculated 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 abnomial and accident conditions, thereactivity will not exceed the regulatory limit of 0.95.The results show that the specified changes in the insert desthe current analvsis.

oresented in the main Dart of the report. ITherefore, any insert width dimension between the value used in themain report including the specified manufacturing tolerances and the value evaluated in thisSupplement is acceptable.

Project No. 2127Report No. 1I1-2125245 Page S 1 -4 Table SI-IFuel Rack Insert Revised Dimensions

[S 1.1]-1For the details of the insert dimensions, see Figure S I-i.t See 'Table 5.3(b)Project No. 2127Report No. HI-21 25245Page S 1-5 Table S 1-2Results of the MCNP5 Calculations for Revised Rack Tolerances Revised Reference Burnup delta kac Max delta Max deltaDescription (GWd/mtU)

Filename kclc sigma (95/95) kclc kteit(95/95) (95/95)Design basis 15.5 op146-rt201155r Reference Reference Reference (reference)

Max cell ID 15.5 op146-rt202155r

-0.0091 0.0000 0.0000Max cell pitchMax wall thickness 15.5 op146-rt203155r 0.0017 0.0017 0.0025Min wall thickness 15.5 op146-rt204155r

!_0.0011_

Max insert width 15.5 op146-rt206155r 1_ 1 0.0016 0.0030 0.0004Min insert width 15.5 op]46-rt207155r 0.0030Statistical combination of rack tolerances 0.0035 0.00261 See Table 7.8Note 1: The CASMO depletion calculation filenames are op 146-dbc(-ac).

Project No. 2127Report No. 111-2125245 Page S 1 -6 Table S 1-3Results of the MCNP5-1.51 Calculations for Revised Fuel Radial Positioning in SFP RacksRevised Reference Burnup Revised Refernc.e ncDescription (GWd/mtU)

Filename kCHIC sigma delta kcaic Un(9 delta kcal, 9 5nc.(~95/95)

(95/95)2x2 reference 15.5 2x2dbrot0l55r Ref Ref. Ref. Ref.(Case 2.3.5.2.1) 2W2 eccentric center 15.5 2x2ecnt 155r1 -0.0028 0.0015 -0.0053 0.0015(Case 2.3.5.2.2) 2x2 eccentric in 15.5 2x2ein155r

-0.0054 0.0015 -0.0081 0.0013(Case 2.3.5.2.3)

.0.0.802x2 eccentric out 15.5 2x2eoutl55r

-0.0014 0.0015 -0.0047 0.0014(Case 2.3.5.2.4)

...2x2 insert/cell center 15.5 2x2icnt]

55r 0.0001 0.0016 0.0002 0.0013(Case 2.3.5.2.5) 8x8 reference 15.5 8x8dbc155r 1 Ref Ref Ref. RefCase 2.3.5.2.61R 8x8 eccentric center 15.5 8x8ecntl55r

-0.0032 0.0015 -0.0023 0.0014(Case 2.3.5.2.7)

....8x8 eccentric in 15.5 8einI55r

-0.0071 0.0015 -0.0080 0,0016(Case 2.3.5.2.8) 8x8 eccentric out 15.5 8x8eoutl55r

-0.0035 0.0016 -0.0035 0.0014(Case 2.3.5.2.9) 8x8 insert/cell center 15.5 8xSicnt 155r 0.0009 0.0014 0.0016 0.0014(Case 2.3.5.2.10)

Ix reference Ref Ref. Ref. Ref.(Case 2.3.5.2.11) 5bc 1555rIx] insert/cell center 15.5 lxlicntl55r1 0.0004 0.0015 0.0000 0.0015(Case 2.3.5.2.12) 1 ___ I i I It See Table 7.9(a)Note 1: The CASMO depletion calculation filenames are opl46-dbc(-ac).

Project No. 2127Report No. HI-2125245 Page S 1 -7 Table S1-4Results of the MCNP5-1.51 Calculations for Revised Fuel Orientation in SFP RacksRevised Reference Burnup Revised Reference unc.Description (GWd/mtU)

Filename k,,,ý sigma delta (9/5 delta (9/5___________(95/95) dlak. (95/95)Reference (Shown in 15.5 2x2dbrotOl55r Ref Ref. Ref RefFigure 2.9(a)) 5 d t 51fe.Rotated fuel assembly 155 2x2dbrotl 155r 00004 0.0014 -0.0008 0.0014(shown in Figure 2.9(b))Rotated fuel assembly 155 2x2dbrot2lS5r 0.0011 0.0015 -0.0007 0.0014(shown in Figure 2.9(c))Rotated fuel assembly 15.5 2x2dbrot3l55r 0.0016 0,0014 -0.0013 0.0013(shown in Figure 2.9(d)) 5.Rotated fuel assembly 155 2x2dbrot4l55r 0.0024 0.0016 -0.0007 0.0013(shown in Figure 2.9(e))t See Table 7.9(b)Note 1: The CASMO depletion calculation filenames are opl46-dbc(-ac).

Project No. 2127Report No. HI-2125245 Page S 1-8 ZJI!-4 -__ __ -_Im_I-If-________________________________________________

I _______________________

I.I--IProject No. 2127Report No. 11I-2125245 Page S 1 -9 4 -4_____1 =1I_ _ _ _ _ _ 11 .. ....m-II =m -m--_ _ _ _ _ _ -_Project No. 2127Report No. HI-2125245 Page S I -10 Figure Proprietary Project No. 2127Report No. H1-2125245 Page SI -I I

Retrieved from "https://kanterella.com/w/index.php?title=ML14024A562&oldid=796473"