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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 Version.
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ATTACHMENT 4 Holtec International Report No. HI-2104790, Revision 1, "Nuclear Group Computer Code Benchmark Calculations"

U. HEW Holtec Center. 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797- 0900 HO LTEC INTER NATIO0NAL Fax (856) 797 - 0909 Nuclear Group Computer Code Benchmark Calculations FOR GENERIC: NON-PROPRIETARY VERSION Holtec Report No: HI-2104790 Holtec Project No: GENERIC Sponsoring Holtec Division: HTS Report Class: SAFETY RELATED

Summary of Revisions Revision 0 Original Issue Revision 1 Additional criticality experiments were added to Appendix B. The index numbers of criticality experiments in Appendix C were updated to be consistent with a new revision of Appendix B. The benchmark of MCNP5-1.51 with ENDF/B-VII was added in Appendix D.

REPORT HI-2104790 i

Table of Contents 1.0 Introdu ction ......................................................................................................................... 3 2.0 M ethodology ....................................................................................................................... 3 2.1 Determination of Bias and Bias Uncertainty ................................................................ 3 2.2 Statistical Methods ...................................................................................................... 4 2.2.1 Single Sided Tolerance Limit Method .................................................................. 4 2.2.2 Confidence Band with Administrative Margin Method ........................................ 5 2.2.3 Non-parametric Statistical Treatment Method ...................................................... 6 2.3 Area of Applicability .............................................. 8 2.3.1 Key Parameters Identification ................................................................................ 8 2.3.2 Screening Area of Applicability .......................................................................... 9 3.0 A ssu m ption s ........................................................................................................................ 9 4.0 Computer Files ............................................................................................................ 9 5.0 Su mm ary ............................................................................................................................. 9 6 .0 R eferences ......................................................................................................................... 10 Appendix A: Holtec Approved Computer Program List ............................................. A-1 Appendix B: Description of the Critical Experiments ............................................. B-1 Appendix C: Benchmark of MCNP5-1.51 with ENDF/B-V ..................................... C-1 Appendix D: Benchmark of MCNP5-1.51 with ENDF/B-VII ...................................... D-1 ii HI-2 104790 REPORT HI-2104790 ii

1.0 Introduction This report documents the criticality experiment benchmark validation calculations for the following computer codes and libraries combinations and establishes the criticality code bias and bias 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 as the Haut Taux de Combustion (HTC) experiments and to the selected critical, presented in Appendix B, with geometric and material characteristics similar to that of spent fuel storage and transport casks. The simulated fuel rods used in these experiments contained uranium or mixture of uranium and plutonium oxides. In the HTC experiments the plutonium-to-uranium ratio and the isotopic compositions of both the uranium and plutonium were designed to be similar to what would be found in a typical pressurized-water reactor (PWR) fuel assembly that initially had an enrichment 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 with standards 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 method used to calculate criticality safety margins and that the validation must be documented in a written report describing the method, computer program and cross section libraries used, the experimental 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 safety calculation has been performed following reference [5] methodology. The validation allows the understanding of the accuracy of the calculational methodology to predict subcriticality.

Validation includes identification of the difference between calculated and experimental neutron effective multiplication factor (keff), called the bias. A set of appropriate critical experiments are selected 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 of applicability (AOA), which establishes the limits of the systems that can be analyzed using the validated 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 that incorporates 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 the calculation uncertainty. Then, the weighted mean multiplication factor keffand the bias uncertainty (Sp) are given by:

keff --

So1 ar1 (2-2)

S,= N[s 2 + j 2 (2-3) where s2 is the variance about the mean and j2 is the average total uncertainty, given by:

2 2 1 1 (2-4)

-2_ n1 (2-5) where n is the number of critical experiments used in the validation and keg.i is the ith value of the multiplication 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 keff is greater than one.

Statistical Methods Single Sided Tolerance Limit Method If the benchmark calculated neutron multiplication factor does not exhibit trends with the parameters, the lower tolerance limit or single sided tolerance limit method can be used. A weighted lower limit tolerance (KL) is a single lower limit above which a defined fraction of the population of keg is expected to lie, with a prescribed confidence and within the area of the applicability. The term "weighted" refers to a specific statistical technique where the REPORT HI-2104790 4

uncertainties in the data are used to weight the data point. Data with high uncertainties will have less "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 sided lower 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 when applying the single sided tolerance limit for a normally distributed data and U is the single sided lower tolerance factor, determined from the following equations [6]. Note that for groups with larger than 50 samples, the single sided lower tolerance factor for 50 samples was conservatively used.

z 1 Pz+ z - ab U =

a (2-9) a = z 12-Y 2(N-1)

(2-10) b =1- 2 pz-N 2 N

(2-11) where zj-p is the critical value from the normal distribution that is exceeded with probability i-p and z1 .y is the critical value from the normal distribution that is exceeded with probability 1-y.

Confidence Band with Administrative Margin Method If 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 method applies a statistical calculation of the bias and its uncertainty plus an administrative margin to a linear 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) where REPORT HI-2104790 5

w(x) = tj~ x sp 1+ -n+ (X-*nG --)

(2-13) and n 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 S 2k(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 n 1 i i=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 Method Data that do not follow a normal distribution can be analyzed by non-parametric techniques. The analysis results in a determination of the degree of confidence that a fraction of the true population of data lies above the smallest observed value. The more data is available in the sample, the higher the degree of confidence.

The following equation determines the percent confidence that a fraction of the population is above the lowest observed value:

6 HI-2 104790 REPORT HI-2104790 6

m-I fl= 1 - j! (n --j)! (1 -q)Iqn-j (2-17) where q 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 smallest sample; m=2 for the second smallest sample, etc.). Non-parametric techniques do not require reliance upon distributions, but are rather an analysis of ranks. Therefore, the samples 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), the equation reduces to:

= l-q" = 1-0.95" (2-18)

This information is then used to determine the Non-parametric Margin from Table 2.2 in Reference [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 Method When a relationship between a calculated keff and an independent variable can be determined, a single-sided lower tolerance band may be used. This is a conservative method that provides a fitted curve above which the true population of keff is expected to lie. The tolerance band equation is actually a calibration curve relation.

The equation for the single-sided lower tolerance band is KL = Kfit(x) - Spfit I2F(a+n +-Z) (n-2 a(x)x yn 2 (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. The degree 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, Z 2 P.l is the symmetric percentile of the Gaussian or normal distribution that contains the P fraction, y = (I -p)/2, (2-21) 2 X 1-y,n-2 is the upper Chi-square percentile,

_- S 2 5 2 Spfit f=t + (2-22)

Slit = = n- -2 To.2 [keff.

1 1 - fitt(Xi)] 21 (2-23)

Area of Applicability The area(s) of applicability refers to the key physical parameter(s) that define a particular fissile configuration. 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 approach used 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 critical experiments.

Steps i. and ii. are presented in subsections 2.3.1 and 2.3.2, respectively. Step iii. is presented in Appendix B. Steps iv. and v. are presented in Appendix C and D.

Key Parameters Identification 8

HI-2 104790 REPORT HI-2104790 8

This validation will cover a number of designs but all the designs will consider the same key parameters 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 pellets clad in a zirconium alloy. The moderator and reflector is water which in some cases has dissolved boron. or gadolinium solutions. Absorber plates made of borated steel, Boral, Zircaloy Boroflex or cadmium and absorber rods made of steel, aluminum, Gd 20 3, Pyrex, Vicor or borated aluminum will be included in this validation. Some experiments were performed with steel 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 in references [B.6] through [B.12]. The fuel assemblies may be separated by water, water and an absorber 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 0.07 and 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 critical experiments, 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, fuel rod arrays were modeled as lattices, all fuel rod details were modeled, and the water between the rods was modeled as specified in the experiment description. However, structures further away from the experiment, such as building walls and foundations, were not included in the models.

4.0 Computer Files All computer files to support this analysis are provided on the Holtec server in

\Projects\0\Reports\HI-2104790 and its subdirectories.

5.0 Summary The criticality experiment benchmark validation calculations for the computer codes and libraries 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 bias uncertainty for these codes are presented in appropriate appendices. The similarity between the chosen experiments and the actual systems has been based on a set of screening criteria as is stated in the NUREG/CR-6698 [5].

The summary of biases and bias uncertainties for the validated computer codes is shown in Table 5.1.

6.0 References

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

[2] ANSI/ANS-8.17, "American National Standard for Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors," American Nuclear Society, La Grange Park, Illinois.

[3] Criticality Benchmark Guide for Light Water Reactor Fuel in Transportation and Storage Packages, 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 Designs Parameter Critical Experiment Requirement Range of Key Parameters 2 35 23 9 241 23 5 23 9 241 Fissionable Material U, Pu, Pu U, Pu, Pu Isotopic Composition 235U/Ut

< 5.Owt% 0. 16wt% to 5.74wt%

Pu/(U+Pu) < 20wt% 1.104wt% to 20wt%/o Physical Form UO 2,MOX UO 2,MOX Moderator Material (coolant) H H Physical Form H20 H20 Density Normal pressure & temperature around 1.0 g/cm 3 condition Reflector Material H H Physical Form H20 H20 Density Normal pressure & temperature around 1.0 g/cm 3 condition Interstitial Reflector Material Plate Steel or Lead Steel or Lead Absorber Material None, Boron (0 to 2550 ppm) or Soluble None, Boron or Gadolinium Gadolinium (0 to 197 ppm)

Rods Boron Pyrex, Vicor, Steel or B-Al I Separating Material Water, B-SS, Boral, Boroflex, Plate Water, B-SS, Boral or Cadmium Zircaloy or Cadmium Geometry Fuel Square/Triangle lattice of fuel Square/Triangle lattice of fuel pins pins Neutron Energy Thermal spectrum Thermal spectrum 11 HI-2104790 REPORT HI-2104790 I1I

Table 5-1 Summary of Biases and Bias Uncertainties for the Validated Computer Codes Computer Code Total Bias Bias Uncertainty MCNP5-1.51 with ENDF/B-V (Appendix C)

MCNP5-1.51 with ENDF/B-VII (Appendix D) 12 HI-2 104790 REPORT HI-2104790 12

Appendix A Holtec Approved Computer Program List (total number of pages: 5 including this page)

Appendix Proprietary A-I HI-2 104790 REPORT HI-2104790 A-1

Appendix B Description of the Critical Experiments (total number of pages: 16 including this page)

B-i HI-2104790 REPORT HI-2104790 B-I

B.1. Introduction and Purpose The purpose of this Appendix is to document the description of the full set of critical experiments 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) at the experimental criticality facility in Valduc, France, between 1988 and 1990. The fuel rods were fabricated specifically for this set of experiments. The fuel consisted of 1-cm-long pellets contained within Zircaloy-4 cladding. The plutonium-to-uranium ratio and the isotopic compositions of both the uranium and plutonium used in the simulated fuel rods were designed to be similar to what would be found in a typical pressurized-water reactor fuel assembly that initially had an enrichment of 4.5 wt % 235U and was burned to 37,500 MWd/MTU. The fuel material also includes 241Am, which is present due to the decay of 24 1Pu. The fuel rods were held in place by an upper and a lower grid and were contained in one or four assemblies placed into a rectangular tank. The critical approach was accomplished by varying the water or solution level in the tank containing the fuel pin arrays. The critical condition was extrapolated from a subcritical 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 156 critical configurations. Detailed descriptions of the critical experiments are presented in references [B.1] through [B.4]. The HTC experiments include configurations designed to simulate fuel handling activities, pool storage, and transport in casks constructed of thick lead or steel and were categorized into four phases.

B.2.1. Phase 1: Water-Moderated and Reflected Arrays The first phase included 18 configurations, each involving a single square-pitched array of rods with rod pitch varying from 1.3 to 2.3 cm.

The tank was incrementally filled with water at room temperature, water being injected at the bottom of the tank. A measurement needle provided water height. Therefore, the water was used as core moderator and as reflector beneath the fuel and around the array on four sides. The critical 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.

B-2 HI-2 104790 REPORT HI-2104790 B-2

B.2.2. Phase 2: Reflected Simple Arrays Moderated by Poisoned Water with Gadolinium or Boron The second phase included 41 configurations that were similar to the first phase except that the water used as moderator and reflector included either boron or gadolinium in solution at various concentrations.

The tank was incrementally filled with poisoned solution at room temperature, this solution being pumped in the bottom of the tank. A measurement needle provided solution height. The critical 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 with boron solutions.

The experiments key physical parameters are summarized in Table B-2 through Table B-3.

B.2.3. Phase 3: Pool Storage The 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 tank which was itself located on the floor in the middle (approximately) of a large room. The spacing between assemblies was varied, and some of the assemblies had B-SS, Boral, or cadmium plates attached to the sides of the four assemblies.

The tank was incrementally filled with water at room temperature, water being pumped in at the bottom of the tank. A measurement needle provided water height. Therefore, the water was used as core moderator and as reflector beneath the fuel and around the array on four sides. The critical 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 four arrays, and the others without any.

The experiments key physical parameters are summarized in Table B-4.

B.2.4. Phase 4: Shipping Cask The fourth phase simulated cask conditions and included 71 configurations similar to the Phase 3 configurations except thick steel or lead shields were placed around the outside of the 2 x 2 array of fuel assemblies. These assemblies with, in some cases, canisters, were placed on a pedestal centered 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 screen varied from one case to another.

B-3 REPORT HI-2104790 B-3

The tank was incrementally filled with water at room temperature, water being pumped in at the bottom of the tank. A measurement needle provided water height. Therefore, the water was used as core moderator and as reflector beneath the fuel and around the array on four sides behind the reflector 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 thirty three with steel reflector screens. Twenty six among the former and twenty one among the latter used 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 and spent fuel storage analysis. This section provides a summary description of the materials and physical 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 U0 2 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-irradiated234mixed-oxide 235 (MOX) 238 fuel of similar characteristics to spent fuel.

28 The239U0 2 24 experiments 24 address 24 U,1 U and U. The MOX critical experiments address 2 38Pu, Pu, 24°pu, 241pu, 242pu and Am. 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 et gadolinium) (Reflected simple arrays moderated by poisoned water with gadolinium or boron) Rddvaluation des experiences," DSU/SEC/T/2005-38/D.R., Institut de Radioprotection et de Sfiretd Nuclraire, 2008.

[B.3] F. Fernex, "Programme HTC - Phase 3 : Configurations "stockage en piscine" (Pool storage) R66valuation des expdriences," DSU/SEC/T/2005-37/D.R., Institut de Radioprotection 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 de Radioprotection et de Sfiret6 Nuclraire, 2008.

REPORT HI-2104790 B-4

[B.5] C. Portella, C. Woillard "Programme "HTC" - Experiences de criticit6 avec des crayons combustibles HTC (type REP A haut taux de combustion) - Rdsultats de l'6tude paramrtrique 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 Tightly Packed 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 Critical Experiment 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 Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1986.

[B. 12] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium 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.

B-5 HI-2 104790 REPORT 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 water number (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 + 0.06 2 MIX-COMP-THERM-HTC-002 2335 2.3 38 x 37 1406 06/06/88 21.1 87.68 + 0.06 3 MIX-COMP-THERM-HTC-003 2336 37 x 37 1369 06/07/88 21.0 90.38 + 0.06 4 MIX-COMP-THERM-HTC-004 2337 27 x 27 729 06/09/88 20.7 63.77 + 0.06 5 MIX-COMP-THERM-HTC-005 2339 1.9 25 x 25 625 06/13/88 20.5 81.95 + 0.08 6 MIX-COMP-THERM-HTC-006 2340 25 x 24 600 06/14/88 20.7 90.22 + 0.06 7 MIX-COMP-THERM-HTC-007 2341 26 x 26 676 06/15/88 20.2 65.11 +/- 0.07 8 MIX-COMP-THERM-HTC-008 2342 1.7 25 x 25 625 06/16/88 21.0 74.86 +/- 0.06 9 MIX-COMP-THERM-HTC-009 2343 25 x 24 600 06/16/88 20.8 82.25 + 0.06 10 MIX-COMP-THERM-HTC-010 2345 29 x 29 841 06/26/88 21.1 59.92 + 0.06 11 MIX-COMP-THERM-HTC-01 1 2347 1.5 27 x 27 729 06/23/88 21.3 76.72 +/- 0.06 12 MIX-COMP-THERM-HTC-012 2348 27 x 26 702 06/23/88 21.1 84.57 +/- 0.06 13 MIX-COMP-THERM-HTC-013 2349 39 x 39 1521 06/29/88 21.3 53.77 +/- 0.06 14 MIX-COMP-THERM-HTC-014 2352 1.3 34 x 34 1156 07/05/88 21.3 80.16 +/- 0.06 15 MIX-COMP-THERM-HTC-015 2353 34 x 33 1122 07/06/88 21.3 86.35 +/- 0.06 16 MIX-COMP-THERM-HTC-016 2355 50 x 18 900 07/19/88 21.0 69.07 +/- 0.06 17 MIX-COMP-THERM-HTC-017 2357 1.7 50 x 17 850 07/21/88 21.4 83.15 + 0.08 18 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 array REPORT 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 Date of Temperature Critical water Experiment Pitch Case 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 0.052 20 MIX-COMP-THERM-HTC-020 2406 38 x 37 1406 01/23/89 19.7 87.16 +/- 0.04 0.052 21 MIX-COMP-THERM-HTC-021 2407 42 x 42 1764 01/23/89 20.1 80.13 +/- 0.04 0.100 22 MIX-COMP-THERM-HTC-022 2408 42 x 41 1722 01/25/89 19.7 84.38 +/- 0.04 0.099 23 MIX-COMP-THERM-HTC-023 2409 1.3 41 x 41 1681 01/25/89 19.6 89.54 + 0.04 0.099 24 MIX-COMP-THERM-HTC-024 2410 46 x 46 2116 01/26/89 20.1 81.33 +/- 0.04 0.151 25 MIX-COMP-THERM-HTC-025 2411 45 x 45 2025 01/27/89 20.0 89.49 +/- 0.04 0.148 26 MIX-COMP-THERM-HTC-026 2412 50 x 50 2500 01/30/89 20.7 85.83 +/- 0.04 0.200 27 MIX-COMP-THERM-HTC-027 2415 50 x 49 2450 02/01/89 19.6 90.03 + 0.05 0.197 28 MIX-COMP-THERM-HTC-028 2417 50 x 50 2500 02/09/89 19.6 89.67 + 0.04 0.196 29 MIX-COMP-THERM-HTC-029 2419 42 x 42 1764 02/14/89 21.4 85.88 +/- 0.05 0.147 30 MIX-COMP-THERM-HTC-030 2420 42 x 41 1722 02/15/89 21.0 90.51 +/-0.05 0.147 31 MIX-COMP-THERM-HTC-031 2422 1.5 36 x 36 1296 02/21/89 22.1 83.86 +/--0.05 0.098 32 MIX-COMP-THERM-HTC-032 2423 36 x 35 1260 02/21/89 22.6 89.85 +/- 0.04 0.098 33 MIX-COMP-THERM-HTC-033 2425 32 x 32 1024 02/24/89 20.9 73.60 +/- 0.05 0.048 34 MIX-COMP-THERM-HTC-034 2427 31 x 31 961 02/27/89 20.6 84.14 +/- 0.04 0.048 35 MIX-COMP-THERM-HTC-035 2430 1.7 31 x 30 930 03/01/89 21.1 85.87 +/- 0.05 0.048 36 MIX-COMP-THERM-HTC-036 2434 1.9 35 x 35 1225 03/08/89 21.7 89.61 +/- 0.04 0.048 37 MIX-COMP-THERM-HTC-037 2436 39 x 39 1521 03/13/89 22.5 85.86 + 0.05 0.097 1.7 38 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], not retained REPORT 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 nedge Total experiment (°C) water (cm)height (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 +/- 0.001 40 MIX-COMP-THERM-HTC-040 2438 37 x 36 1332 04/18/89 22.8 83.84 +/- 0.04 0.106 +/- 0.001 41 MIX-COMP-THERM-HTC-041 2441 39 x 39 1521 04/20/89 23.5 84.04 + 0.04 0.205 +/- 0.002 42 MIX-COMP-THERM-HTC-042 2444 42 x 41 1722 04/26/89 23.0 85.40 + 0.05 0.299 +/- 0.003 1.3 43 MIX-COMP-THERM-HTC-043 2446 45 x 44 1980 05/09/89 24.2 84.14 + 0.04 0.400 +/- 0.004 44 MIX-COMP-THERM-HTC-044 2447 44 x 44 1936 05/10/89 24.7 88.63 +/- 0.05 0.399 +/- 0.004 45 MIX-COMP-THERM-HTC-045 2448 47 x 47 2009 05/11/89 26.3 88.44 +/- 0.04 0.486 +/- 0.005 46 MIX-COMP-THERM-HTC-046 2449 50 x 50 2500 05/17/89 25.1 90.64 +/- 0.04 0.587 +/- 0.006 47 MIX-COMP-THERM-HTC-047 2459 49 x 49 2401 06/05/89 24.7 88.88 + 0.04 0.595 +/- 0.006 48 MIX-COMP-THERM-HTC-048 2468 43 x 43 1849 06/15/89 22.7 89.46 +/- 0.04 0.499 +/- 0.005 49 MIX-COMP-THERM-HTC-049 2470 1.5 39 x 39 1521 06/19/89 23.6 85.37 +/- 0.05 0.393 + 0.004 50 MIX-COMP-THERM-HTC-050 2471 35 x 35 1225 06/21/89 23.6 88.90 +/- 0.04 0.295 +/- 0.003 51 M1X-COMP-THERM-HTC-051 2473 32 x 32 1024 06/27/89 23.5 87.02 +/- 0.04 0.200 + 0.002 52 MIX-COMP-THERM-HTC-052 2475 30 x 29 870 07/03/89 23.6 82.48 +/- 0.04 0.089 + 0.001 53 MIX-COMP-THERM-HTC-053 2478 28 x 28 784 07/06/89 23.8 85.10 +/- 0.04 0.090 + 0.001 54 MIX-COMP-THERM-HTC-054 2483 32 x 32 1024 07/19/89 24.2 87.06 +/- 0.04 0.194 +/- 0.002 55 MIX-COMP-THERM-HTC-055 2485 1.7 37 x 37 1369 07/21/89 24.5 89.65 + 0.04 0.286 + 0.003 56 MIX-COMP-THERM-HTC-056 2487 45 x 44 1980 08/09/89 23.8 88.72 +/- 0.04 0.415 +/- 0.004 57 MIX-COMP-THERM-HTC-057 2482 50 x 21 1050 07/17/89 24.0 77.74 +/- 0.04 0.100 +/- 0.001 58 MIX-COMP-THERM-HTC-058 2490 1.9 39 x 38 1482 09/08/89 22.9 88.41 +/- 0.04 0.220 +/- 0.002

/ 59 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]

Canister Number of Rods Date of Temperature Critical water Water Experiment Case 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 3.5 61 MIX-COMP-THERM-HTC-061 2520 25 x 24 600 01/09/90 18.7 49.55 +/- 0.34 0.0 62 MIX-COMP-THERM-HTC-062 2521 Borated 25 x 24 600 01/10/90 18.8 71.45 +/- 0.34 2.0 Steel 63 MIX-COMP-THERM-HTC-063 2522 25 x 24 600 01/10/90 19.0 89.96 +/- 0.34 3.0 64 MIX-COMP-THERM-HTC-064 2523 25 x 24 600 01/12/90 18.9 58.23 +/- 0.34 1.0 65 MIX-COMP-THERM-HTC-065 2514 Boral 25 x 25 625 12/28/89 20.6 90.03 +/- 0.34 0.0 66 MIX-COMP-THERM-HTC-066 2511 25 x 25 625 12/21/89 21.1 82.16 +/- 0.34 2.0 67 MIX-COMP-THERM-HTC-067 2524 25 x 24 600 01/15/90 18.7 55.33 +/- 0.34 0.0 68 MIX-COMP-THERM-HTC-068 2525 m 25 x 24 600 01/16/90 19.0 67.95 +/- 0.34 1.0 69 MIX-COMP-THERM-HTC-069 2526 25 x 24 600 01/17/90 19.1 79.83 + 0.34 1.5 70 MIX-COMP-THERM-HTC-070 2527 25 x 24 600 01/18/90 19.1 58.66 +/- 0.34 0.5 71 MIX-COMP-THERM-HTC-071 2509 25 x 25 625 12/19/89 20.9 84.75 +/- 0.34 18.0 72 MIX-COMP-THERM-HTC-072 2531 25 x 24 600 01/23/90 19.0 88.2 +/- 0.34 14.5 73 MIX-COMP-THERM-HTC-073 2532 24 x 24 576 01/24/90 19.1 81.18 -0.34 11.0 74 MIX-COMP-THERM-HTC-074 2533 24 x 23 552 01/25/90 19.3 82.12 +/-0.34 10.0 75 MIX-COMP-THERM-HTC-075 2534 23 x 23 529 01/26/90 19.4 81.2 +/- 0.34 9.0 76 MIX-COMP-THERM-HTC-076 2535 22 x 22 484 01/30/90 19.7 86.17 +/- 0.34 8.0 77 MIX-COMP-THERM-HTC-077 2536 20 x 20 400 01/31/90 19.7 82.08 +/- 0.34 6.0 78 MIX-COMP-THERM-HTC-078 2537 17 x 17 289 02/01/90 19.9 77.92 +/- 0.34 4.0 79 MIX-COMP-THERM-HTC-079 2538 17 x 16 272 02/02/90 20.0 90.28 +/- 0.34 4.0 80 MIX-COMP-THERM-HTC-080 2539 14 x 14 196 02/05/90 20.2 75.99 +/- 0.34 2.0 81 MIX-COMP-THERM-HTC-081 2541 13 x 13 169 02/06/90 20.0 83.17 +/- 0.34 1.0 82 MIX-COMP-THERM-HTC-082 2544 13 x 13 169 02/07/90 20.4 79.46 +/- 0.34 0.0 83 MIX-COMP-THERM-HTC-083 2547 25 x 25 625 02/19/90 20.9 29.46 +/- 0.34 0.0 84 MIX-COMP-THERM-HTC-084 2548 25 x 25 625 02/20/90 20.9 37.96 +/- 0.34 4.0 85 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 cm pitch) [B.4]

Date of Temperature Water Gap Screen array Critical water Case Reference Experiment Canister height (cm) number Type experiment (°C) (cm) (a) distance (cm)

(b) (c) 86 MIX-COMP-THERM-HTC-086 2562 03/16/90 22.8 0.0 0.0 42.53 + 0.34 87 MIX-COMP-THERM-HTC-087 2563 03/19/90 23.1 0.5 0.0 44.79 +/- 0.34 88 MIX-COMP-THERM-HTC-088 2564 03/20/90 23.3 1.0 0.0 47.86 + 0.34 89 MIX-COMP-THERM-HTC-089 2565 03/21/90 23.1 1.5 0.0 51.3 + 0.34 90 MIX-COMP-THERM-HTC-090 2566 03/22/90 23.3 2.0 0.0 54.65 + 0.34 91 MIX-COMP-THERM-HTC-091 2567 Borated 03/22/90 23.4 3.0 0.0 62.04 -0.34 Steel 92 MIX-COMP-THERM-HTC-092 2568 03/23/90 23.6 3.5 0.0 66.10 - 0.34 93 MIX-COMP-THERM-HTC-093 2569 03/26/90 23.5 2.0 0.5 55.87 - 0.34 94 MIX-COMP-THERM-HTC-094 2570 03/27/90 23.1 2.0 1.0 57.33 + 0.34 95 MIX-COMP-THERM-HTC-095 2571 03/27/90 23.0 2.0 1.5 58.68 + 0.34 96 MIX-COMP-THERM-HTC-096 2572 03/28/90 22.9 2.0 2.0 59.78 +/- 0.34 97 MIX-COMP-THERM-HTC-097 2586 04/23/90 21.9 0.0 0.0 72.47 +/- 0.34 98 MIX-COMP-THERM-HTC-098 2587 04/24/90 22.0 0.0 0.0 72.49 +/- 0.34 99 MIX-COMP-THERM-HTC-099 2588 Boral 04/24/90 22.2 0.0 0.5 74.70 +/- 0.34 100 MIX-COMP-THERM-HTC-100 2624 07/13/90 21.6 1.0 0.0 86.06 +/- 0.34 101 MIX-COMP-THERM-HTC-101 2625 07/18/90 22.4 0.5 0.0 76.69 + 0.34 102 MIX-COMP-THERM-HTC-102 2577 04/05/90 22.7 0.0 0.0 46.13 +/- 0.34 103 MIX-COMP-THERM-HTC-103 2578 04/05/90 22.6 1.0 0.0 52.89 +/- 0.34 104 MIX-COMP-THERM-HTC-104 2579 04/06/90 22.6 2.0 0.0 63.52 +/- 0.34 105 MIX-COMP-THERM-HTC-105 2580 Cadmium 04/09/90 22.4 2.5 0.0 69.83 +/- 0.34 106 MIX-COMP-THERM-HTC-106 2581 04/11/90 22.5 2.0 0.5 65.84 +/- 0.34 107 MIX-COMP-THERM-HTC-107 2582 04/11/90 22.5 2.0 1.0 68.63 + 0.34 108 MIX-COMP-THERM-HTC-108 2583 04/12/90 22.4 2.0 1.5 71.21 +/- 0.34 REPORT HI-2104790 B-10

Screen array Critical water Case Reference Experiment Canister Date of Temperature Water Gap Sce (cm) heit(cm) w number 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+/- 0.34 110 MIX-COMP-THERM-HTC-1 10 2621 07/03/90 22.3 3.0 0.0 76.25 +/- 0.34 111 MIX-COMP-THERM-HTC-1 11 2622 07/04/90 22.3 3.5 0.0 83.38 +/- 0.34 112 MIX-COMP-THERM-HTC-1 12 2550 02/23/90 21.4 0.0 0.0 27.45 + 0.34 113 MIX-COMP-THERM-HTC-1 13 2551 02/26/90 22.1 1.0 0.0 28.00 +/- 0.34 114 MIX-COMP-THERM-HTC-114 2552 02/28/90 21.8 2.0 0.0 29.37 +/- 0.34 115 MIX-COMP-THERM-HTC-115 2553 03/01/90 21.8 4.0 0.0 34.65 +/- 0.34 116 MIX-COMP-THERM-HTC-1 16 2554 03/02/90 21.3 6.0 0.0 41.60 +/- 0.34 117 MIX-COMP-THERM-HTC-117 2555 03/05/90 20.7 8.0 0.0 48.65 +/- 0.34 No 118 MIX-COMP-THERM-HTC-118 2556 03/06/90 20.7 10.0 0.0 54.74 +/- 0.34 119 MIX-COMP-THERM-HTC-119 2557 03/07/90 20.9 12.0 0.0 59.57 +/-0.34 120 MIX-COMP-THERM-HTC-120 2558 03/09/90 21.3 2.0 0.5 29.43 +/- 0.34 121 MIX-COMP-THERM-HTC-121 2559 03/12/90 21.7 2.0 1.0 29.46 +/- 0.34 122 MIX-COMP-THERM-HTC-122 2560 03/13/90 21.9 2.0 1.5 29.55 +/-0.34 123 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 cm pitch) [B.4]

Experiment Canister Date of Temperature Water Gap Screen array Critical water Case 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 + 0.34 125 MIX-COMP-THERM-HTC-125 2603 05/21/90 23.4 0.5 0.0 44.14 +/- 0.34 126 MIX-COMP-THERM-HTC-126 2604 05/22/90 22.9 1.0 0.0 46.96 +/- 0.34 127 MIX-COMP-THERM-HTC-127 2605 05/29/90 20.4 1.5 0.0 50.16 + 0.34 128 MIX-COMP-THERM-HTC-128 2606 05/30/90 20.1 2.0 0.0 53.43 +/- 0.34 129 MIX-COMP-THERM-HTC-129 2607 Borated 05/31/90 20.0 2.0 0.5 54.71 +/--0.34 Steel 130 MIX-COMP-THERM-HTC-130 2608 06/05/90 20.2 2.0 1.0 56.32 +/- 0.34 131 MIX-COMP-THERM-HTC-131 2609 06/05/90 20.1 2.0 1.5 57.96 +/- 0.34 132 MIX-COMP-THERM-HTC-132 2610 06/06/90 19.7 2.0 2.0 59.16 +-0.34 133 MIX-COMP-THERM-HTC-133 2611 06/08/90 19.5 3.0 0.0 60.38 +/- 0.34 134 MIX-COMP-THERM-HTC-134 2612 06/12/90 20.1 3.5 0.0 64.19 + 0.34 135 MIX-COMP-THERM-HTC-135 2589 04/26/90 22.4 0.0 0.0 69.82 + 0.34 Boral 136 MIX-COMP-THERM-HTC-136 2626 07/19/90 22.6 0.5 0.0 73.44 + 0.34 137 MIX-COMP-THERM-HTC-137 2613 06/13/90 20.5 0.0 0.0 44.70 +/- 0.34 138 MIX-COMP-THERM-HTC-138 2614 06/13/90 20.6 1.0 0.0 51.00 + 0.34 139 MIX-COMP-THERM-HTC-139 2615 06/14/90 20.6 2.0 0.0 60.26 +/- 0.34 140 MIX-COMP-THERM-HTC-140 2616 06/15/90 20.7 2.0 0.5 62.54 + 0.34 141 MIX-COMP-THERM-HTC-141 2617 Cadmium 06/18/90 21.0 2.0 1.0 65.85 + 0.34 142 MIX-COMP-THERM-HTC-142 2618 06/19/90 21.3 2.0 1.5 68.70 +/- 0.34 143 MIX-COMP-THERM-HTC-143 2619 06/20/90 21.5 2.0 2.0 71.00 +/- 0.34 144 MIX-COMP-THERM-HTC-144 2620 06/21/90 21.7 2.5 0.0 65.76 +/- 0.34 145 MIX-COMP-THERM-HTC-145 2590 04/27/90 22.4 0.0 0.0 27.77 + 0.34 146 MIX-COMP-THERM-HTC-146 2591 No 05/09/90 24.4 1.0 0.0 28.34 + 0.34 REPORT HI-2104790 B-12

Experiment Canister Date of Temperature Water G Screen array Critical water Case 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 +/- 0.34 148 MIX-COMP-THERM-HTC-148 2593 05/10/90 24.3 2.0 0.5 29.68 +/- 0.34 149 MIX-COMP-THERM-HTC-149 2594 05/11/90 24.5 2.0 1.0 29.66 +/- 0.34 150 MIX-COMP-THERM-HTC-150 2595 05/11/90 24.4 2.0 1.5 29.68 +/--0.34 151 MIX-COMP-THERM-HTC-151 2596 05/14/90 24.7 2.0 2.0 29.76 +/- 0.34 152 MIX-COMP-THERM-HTC-152 2597 05/15/90 24.6 4.0 0.0 35.33 +/- 0.34 153 MIX-COMP-THERM-HTC-153 2598 05/15/90 24.6 6.0 0.0 43.24 +/- 0.34 154 MIX-COMP-THERM-HTC-154 2599 05/16/90 24.7 8.0 0.0 51.30 +/- 0.34 155 MIX-COMP-THERM-HTC-155 2600 05/17/90 24.7 10.0 0.0 58.73 +/- 0.34 156 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% 'wt 157 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 4.306 201 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 20 203 PNL-5803 [B.1 1] MOX Fuel - Type 3.2 Exp. 43 0.71 20 204 PNL-5803 [B.1 1] MOX Fuel - Type 3.2 Exp. 13 0.71 20 205 PNL-5803 [B.11] MOX Fuel - Type 3.2 Exp. 32 0.71 20 206 MIX-COMP-THERM-003-001 Saxton Case 52 PuO2 0.52" pitch 0.72 6.6 207 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 6.6 209 MIX-COMP-THERM-003-003 Saxton Case 56 borated PuO2 0.72 6.6 210 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 6.6 212 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 2.0 214 MIX-COMP-THERM-002-031 0.700-in. pitch 688 ppm B 0.72 2.0 215 MIX-COMP-THERM-002-032 0.870-in. pitch 0 ppm B 0.72 2.0 216 MIX-COMP-THERM-002-033 0.870-in. pitch 1090 ppm B 0.72 2.0 217 MIX-COMP-THERM-002-034 0.990-in. pitch 0 ppm B 0.72 2.0 218 MIX-COMP-THERM-002-035 0.990-in. pitch 767 ppm B 0.72 2.0 219 MIX-COMP-THERM-003-004 Saxton Case PuO2 0.735" pitch 0.72 6.6 220 MIX-COMP-THERM-003-006 Saxton Case PuO2 1.04" pitch 0.72 6.6 221 MIX-COMP-THERM-006-001 8 wt% 240Pu 0.80" pitch 0.71 2.0 222 MIX-COMP-THERM-006-002 8 wt% 240Pu 0.93" pitch 0.71 2.0 223 MIX-COMP-THERM-006-003 8 wt% 240Pu 1.05" pitch 0.71 2.0 224 MIX-COMP-THERM-006-004 8 wt% 240Pu 1.143" pitch 0.71 2.0 225 MIX-COMP-THERM-006-005 8 wt% 240Pu 1.32" pitch 0.71 2.0 226 MIX-COMP-THERM-006-006 8 wt% 240Pu 1.386" pitch 0.71 2.0 227 MIX-COMP-THERM-007-001 16 wt% 240Pu 0.93" pitch 0.72 2.0 228 MIX-COMP-THERM-007-002 16 wt% 240Pu 1.05" pitch 0.72 2.0 229 MIX-COMP-THERM-007-003 16 wt% 240Pu 1.143" pitch 0.72 2.0 230 MIX-COMP-THERM-007-004 16 wt%/o 240Pu 1.32" pitch 0.72 2.0 231 MIX-COMP-THERM-008-001 24 wt0 /o 240Pu 0.80" pitch 0.72 2.0 232 MIX-COMP-THERM-008-002 24 wt%/o 240Pu 0.93" pitch 0.72 2.0 233 MIX-COMP-THERM-008-003 24 wt%/o 240Pu 1.05" pitch 0.72 2.0 234 MIX-COMP-THERM-008-004 24 wt%/o 240Pu 1.143" pitch 0.72 2.0 235 MIX-COMP-THERM-008-005 24 wt0 o 240Pu 1.32" pitch 0.72 2.0 REPORT HI-2104790 B-15

Case Reference Identification U, wt% Pu, 236 MIX-COMP-THERM-008-006 24 wt% 240Pu 1.386" pitch 0.72 2.0 237 MIX-COMP-THERM-005-001 18 wt% 240Pu 0.85" pitch 0.72 4.0 238 MIX-COMP-THERM-005-002 18 wt% 240Pu 0.93" pitch 0.72 4.0 239 MIX-COMP-THERM-005-003 18 wt% 240Pu 1.05" pitch 0.72 4.0 240 MIX-COMP-THERM-005-004 18 wt% 240Pu 1.143" pitch 0.72 4.0 241 MIX-COMP-THERM-005-005 18 wt%/o 240Pu 1.386" pitch 0.72 4.0 242 MIX-COMP-THERM-005-006 18 wt% 240Pu 1.60" pitch 0.72 4.0 243 MIX-COMP-THERM-005-007 18 wt% 240Pu 1.70" pitch 0.72 4.0 244 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% Put 276 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 2.459 317 LEU-COMP-THERM-008-004 Core XI, 1182 ppm, 36 Pyrex Rods 2.459 318 LEU-COMP-THERM-008-005 Core XI, 1182 ppm, 36 Pyrex Rods 2.459 319 LEU-COMP-THERM-008-006 Core XI, 1032.5 ppm, 72 Pyrex Rods 2.459 320 LEU-COMP-THERM-008-007 Core XI, 1032.5 ppm, 72 Pyrex Rods 2.459 321 LEU-COMP-THERM-008-008 Core XI, 794 ppm, 144 Pyrex Rods 2.459 322 LEU-COMP-THERM-008-009 Core XI, 779 ppm, 144 Pyrex Rods 2.459 323 LEU-COMP-THERM-008-010 Core XI, 1245 ppm, 72 Vicor Rods 2.459 324 LEU-COMP-THERM-008-011 Core XI, 1384 ppm, 144 A120 3 Rods 2.459 325 LEU-COMP-THERM-008-012 Core XI, 1348 ppm, 36 A1203 Rods 2.459 326 LEU-COMP-THERM-008-013 Core XI, 1348 ppm, 36 A1203 Rods 2.459 327 LEU-COMP-THERM-008-014 Core XI, 1363 ppm, 72 A1203 Rods 2.459 328 LEU-COMP-THERM-008-015 Core XI, 1362 ppm, 72 A1203 Rods 2.459 329 LEU-COMP-THERM-008-016 Core XI, 1158 ppm 2.459 330 LEU-COMP-THERM-008-017 Core XI, 921 ppm 2.459 331 LEU-COMP-THERM-009-001 0% Boron Steel plates, dist. 0.245 cm 4.306 332 LEU-COMP-THERM-009-002 0% Boron Steel plates, dist. 3.277 cm 4.306 333 LEU-COMP-THERM-009-003 0% Boron Steel plates, dist. 0.428 cm 4.306 334 LEU-COMP-THERM-009-004 0% Boron Steel plates, dist. 3.277 cm 4.306 335 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 2.35 369 LEU-COMP-THERM-012-005 Boral C plate 2.35 370 LEU-COMP-THERM-012-006 Boroflex, 1.84 cm separation 2.35 371 LEU-COMP-THERM-012-007 Boroflex, 1.73 cm separation 2.35 372 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

Reference Identification U, wt% Pu, Case 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 -__j REPORT 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 B4 C 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, A120 3-B 4C rods 4.3486 -

446 LEU-COMP-THERM-082-004 672 pins, A120 3-B4C rods 4.3486 -

447 LEU-COMP-THERM-082-005 668 pins, A120 3-B4C rods 4.3486 -

448 LEU-COMP-THERM-082-006 668 pins, A120 3-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 Gd 20 3 rods 4.3486 -

459 LEU-COMP-THERM-091-002 648 pins, 0 steel rods, 8 Gd2 0 3 rods 4.3486 -

460 LEU-COMP-THERM-091-003 672 pins, 0 steel rods, 4 Gd2 0 3 rods 4.3486 -

461 LEU-COMP-THERM-091-004 646 pins, 4 steel rods, 4 Gd 2 0 3 rods 4.3486 -

462 LEU-COMP-THERM-091-005 656 pins, 4 steel rods, 4 Gd2 O 3 rods 4.3486 -

463 LEU-COMP-THERM-091-006 664 pins, 4 steel rods, 2 Gd 20 3 rods 4.3486 -

464 LEU-COMP-THERM-091-007 670 pins, 2 steel rods, 2 Gd 20 3 rods 4.3486 -

465 LEU-COMP-THERM-091-008 664 pins, 2 steel rods, 2 Gd 2 0 3 rods 4.3486 -

466 LEU-COMP-THERM-091-009 656 pins, 0 steel rods, 2 Gd 20 3 rods 4.3486 -

467 MIX-COMP-THERM-004-001 23x23, 1.825 cm pitch 0.72 3.01 468 MIX-COMP-THERM-004-002 23x23, 1.825 cm pitch 0.72 3.011 469 MIX-COMP-THERM-004-003 23x23, 1.825 cm pitch 0.72 3.011 470 MIX-COMP-THERM-004-004 21x21, 1.956 cm pitch 0.72 3.011 471 MIX-COMP-THERM-004-005 21x21, 1.956 cm pitch 0.72 3.011 472 MIX-COMP-THERM-004-006 21x21, 1.956 cm pitch 0.72 3.011 473 MIX-COMP-THERM-004-007 20x20, 2.225 cm pitch 0.72 3.011 REPORT HI-2104790 B-21

Case Reference Identification U, wt% Pu, 474 MIX-COMP-THERM-004-008 20x20, 2.225 cm pitch 0.72 3.01 475 MIX-COMP-THERM-004-009 20x20, 2.225 cm pitch 0.72 3.011 476 MIX-COMP-THERM-004-010 21x21, 2.474 cm pitch 0.72 3.011 477 MIX-COMP-THERM-004-011 21x21, 2.474 cm pitch 0.72 3.011 478 MIX-COMP-THERM-006-007 8 wt% 240Pu 1.05" pitch, Al Rods 0.72 2.0 479 MIX-COMP-THERM-006-013 8 wt% 240Pu 1.05" pitch, B4 Rods 0.72 2.0 480 MIX-COMP-THERM-006-014 8 wt% 240Pu 1.05" pitch, B3 Rods 0.72 2.0 481 MIX-COMP-THERM-006-015 8 wt% 240Pu 1.05" pitch, B2 Rods 0.72 2.0 482 MIX-COMP-THERM-006-016 8 wt% 240Pu 1.05" pitch, BI Rods 0.72 2.0 483 MIX-COMP-THERM-006-017 8 wt% 240Pu 1.05" pitch, AI+Cd Rods 0.72 2.0 484 MIX-COMP-THERM-006-023 8 wt% 240Pu 1.05" pitch, B4+Cd Rods 0.72 2.0 485 MIX-COMP-THERM-006-024 8 wt% 240Pu 1.05" pitch, B3+Cd Rods 0.72 2.0 486 MIX-COMP-THERM-006-025 8 wt% 240Pu 1.05" pitch, B2+Cd Rods 0.72 2.0 487 MIX-COMP-THERM-006-026 8 wt% 240Pu 1.05" pitch, B 1+Cd Rods 0.72 2.0 488 MIX-COMP-THERM-006-027 8 wt% 240Pu 1.05" pitch, Air+Cd Rods 0.72 2.0 489 MIX-COMP-THERM-006-028 8 wt% 240Pu 1.05" pitch, H20+Cd Rods 0.72 2.0 490 MIX-COMP-THERM-006-029 8 wt% 240Pu 1.32" pitch, Al Rods 0.72 2.0 491 MIX-COMP-THERM-006-035 8 wt% 240Pu 1.32" pitch, B4 Rods 0.72 2.0 492 MIX-COMP-THERM-006-036 8 wt% 240Pu 1.32" pitch, B3 Rods 0.72 2.0 493 MIX-COMP-THERM-006-037 8 wt% 240Pu 1.32" pitch, B2 Rods 0.72 2.0 494 MIX-COMP-THERM-006-038 8 wt% 240Pu 1.32" pitch, B I Rods 0.72 2.0 495 MIX-COMP-THERM-006-039 8 wt% 240Pu 1.32" pitch, AI+Cd Rods 0.72 2.0 496 MIX-COMP-THERM-006-045 8 wt% 240Pu 1.32" pitch, B4+Cd Rods 0.72 2.0 497 MIX-COMP-THERM-006-046 8 wt% 240Pu 1.32" pitch, B3+Cd Rods 0.72 2.0 498 MIX-COMP-THERM-006-047 8 wt% 240Pu 1.32" pitch, B2+Cd Rods 0.72 2.0 499 MIX-COMP-THERM-006-048 8 wt% 240Pu 1.32" pitch, B1 +Cd Rods 0.72 2.0 500 MIX-COMP-THERM-006-049 8 wt% 240Pu 1.32" pitch, Air+Cd Rods 0.72 2.0 501 MIX-COMP-THERM-006-050 8 wt%/o 240Pu 1.32" pitch, H20+Cd Rods 0.72 2.0 502 MIX-COMP-THERM-007-005 16 wt%/o 240Pu 1.386" pitch 0.72 2.0 503 MIX-COMP-THERM-007-006 16 wt% 240Pu 1.05" pitch, Al Rods 0.72 2.0 504 MIX-COMP-THERM-007-012 16 wt% 240Pu 1.05" pitch, B4 Rods 0.72 2.0 505 MIX-COMP-THERM-007-013 16 wt% 240Pu 1.05" pitch, B3 Rods 0.72 2.0 506 MIX-COMP-THERM-007-014 16 wt% 240Pu 1.05" pitch, B2 Rods 0.72 2.0 507 MIX-COMP-THERM-007-015 16 wt% 240Pu 1.05" pitch, B I Rods 0.72 2.0 508 MIX-COMP-THERM-007-016 16 wt% 240Pu 1.05" pitch, AI+Cd Rods 0.72 2.0 509 MIX-COMP-THERM-007-022 16 wt% 240Pu 1.05" pitch, B4+Cd Rods 0.72 2.0 510 MIX-COMP-THERM-007-023 16 wt% 240Pu 1.05" pitch, B3+Cd Rods 0.72 2.0 511 MIX-COMP-THERM-007-024 16 wt% 240Pu 1.05" pitch, B2+Cd Rods 0.72 2.0 512 MIX-COMP-THERM-007-025 16 wt% 240Pu 1.05" pitch, BI+Cd Rods 0.72 2.0 513 MIX-COMP-THERM-007-026 16 wt% 240Pu 1.05" pitch, Air+Cd Rods 0.72 2.0 I REPORT HI-2104790 B-22

Reference Identification U, wt% Pu, Case 514 MIX-COMP-THERM-007-027 16 wt% 240Pu 1.05" pitch, H20+Cd Rods 0.72 2.0 515 MIX-COMP-THERM-008-007 24 wt% 240Pu 1.05" pitch, Al Rods 0.72 2.0 516 MIX-COMP-THERM-008-013 24 wt% 240Pu 1.05" pitch, B4 Rods 0.72 2.0 517 MIX-COMP-THERM-008-014 24 wt% 240Pu 1.05" pitch, B3 Rods 0.72 2.0 518 MIX-COMP-THERM-008-015 24 wt% 240Pu 1.05" pitch, B2 Rods 0.72 2.0 519 MIX-COMP-THERM-008-016 24 wt% 240Pu 1.05" pitch, B I Rods 0.72 2.0 520 MIX-COMP-THERM-008-017 24 wt% 240Pu 1.05" pitch, AL+Cd Rods 0.72 2.0 521 MIX-COMP-THERM-008-023 24 wt% 240Pu 1.05" pitch, B4+Cd Rods 0.72 2.0 522 MIX-COMP-THERM-008-024 24 wt% 240Pu 1.05" pitch, B3+Cd Rods 0.72 2.0 523 MIX-COMP-THERM-008-025 24 wt% 240Pu 1.05" pitch, B2+Cd Rods 0.72 2.0 524 MIX-COMP-THERM-008-026 24 wt% 240Pu 1.05" pitch, B 1+Cd Rods 0.72 2.0 525 MIX-COMP-THERM-008-027 24 wt% 240Pu 1.05" pitch, Air+Cd Rods 0.72 2.0 526 MIX-COMP-THERM-008-028 24 wt% 240Pu 1.05" pitch, H20+Cd Rods 0.72 2.0 527 MIX-COMP-THERM-009-001 8 wt% 240Pu 0.55" pitch 0.16 1.5 528 MIX-COMP-THERM-009-002 8 wt% 240Pu 0.60" pitch 0.16 1.5 529 MIX-COMP-THERM-009-003 8 wt% 240Pu 0.71" pitch 0.16 1.5 530 MIX-COMP-THERM-009-004 8 wt% 240Pu 0.80" pitch 0.16 1.5 531 MIX-COMP-THERM-009-005 8 wt% 240Pu 0.90" pitch 0.16 1.5 532 MIX-COMP-THERM-009-006 8 wt% 240Pu 0.93" pitch 0.16 1.5 B-23 REPORT HI-2 104790 REPORT HI-2104790 B-23

Appendix C Benchmark 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 includes the 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 Data The computer code MCNP5-1.51 [C. 1] is the continuous energy Monte Carlo codes and treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Thermal neutrons are described by both the 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 for every individual run. The neutrons start from an arbitrary distribution, causing a generally very large 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 Results C.3.1. Calculational Results The calculation results for the 156 HTC critical experiments and for the 135 selected critical experiments 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 in Appendix B. Calculation results, including keff, Ucaic-, and EALF, measurement uncertainties (trp) and the calculation and measurement combined uncertainty (ar) are shown in Table C.3-1 through Table C.3-5.

Figure C.3-1 and C.3-2 are histograms showing the frequency of calculated k, and EALF for all 291 benchmarks. The nominal calculated k, values range from A. The EALF results values show a range between Descriptive statistics for the different group of experiments is summarized in Table C.3-6.

C.3.2. Normality Test In order to assess the normality assumption, Shapiro and Wilk [5] test has been used for groups with fewer than 50 samples while the Pearson's chi-square (X2) test [4] has been used for samples larger than 20 samples. The tests are applied to the group of experiments in terms of the categories as set forth in Appendix B.

For the Shapiro and Wilk test, Table C.3-7 shows the computed Wtest value, and W value that can be obtained for the number of experiments from [5] to accept the normality hypothesis. If W REPORT HI-2104790 C-2

is less than the test statistic, Wtest, then the data is considered normally distributed. For the X2 test, 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 _>*o2 in an experiment with d degrees of 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 >_1o 2 ) is large, the obtained and expected distributions are consistent; if it is small, they probably disagree. In particular, if Pd(,k2 > i02) is less than 5%, we say that the disagreement is significant and reject the assumed distributions 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 with all 291 experiments shows an agreement with the assumed normal distribution with the probability Pd = 7.36%.

C.3.3. Trending Analysis Trends 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 1 ' 1X fVX 1 'X y 1 \

(7-2) b= ( Zxi- X- YZT)

(7-3)

(7-4)

The squared term of the linear correlation factor r defined below (from Reference [5]) is used to quantitatively 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. A more quantitative measure of the fit can be found by using Appendix C in Reference [4]. The interpolation was applied, if necessary. For any given observed value ro, PN(Irl -- Irol) is the probability that N measurements of two uncorrelated variables would give a coefficient r as large as 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, if REPORT HI-2104790 C-3

PN(IrI > Irol) < 5%, the correlation is called significant; if it is less than 1%, the correlation is called highly significant.

The validation results are analyzed by grouping the experiments in terms of the categories as set forth 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 in Table C.3-8.

C.3.4. Bias and Bias Uncertainty In this section, benchmark results are analyzed using the statistical method described in section 2.2.

The first step is to evaluate whether the four HTC phases and selected ex eriments, should be reduced to a single set. Th ofe Phase 1 data set is , th of 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 the selected ex eriments data is . The maximum difference between the means is just ý which is less than the uncertainty. These sets are water moderated uranium or mixed plutonium-uranium dioxide lattices. The addition of a absorber rods, separator plates or reflector plates is not introducing a significant increase in the ability to calculate keff. The Phase 1 through Phase 4 sets and the selected experiments are considered one large set of 291 experiments from now on.

The analysis of the correlation coefficient in Table C.3-8 (combined set) and the plot of data trend (Figure C.3-3) show that there is a significant trend a function of the rod pitch. This is discussed 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 the MCNP5-1.51 code is shown in the table below Calculational Bias of the MCNP5-1.51 code Description Total Bias Bias Uncertainty HTC and Selected Experiments C.3.5. Applicability of MCNP5-1.51 Validation Results This 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 geometry REPORT 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. In some 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 are evaluated 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 (B4 C), and boron in borated steel o Soluble gadolinium in water o Cadmium

" Reflectors o Steel o Lead o Water Some typical configurations do not contain gadolinium or cadmium neutron absorbers or lead reflectors. To verify that including those materials does not have a significant effect on the results of the benchmarking analyses, a subset without those experiments containing those materials was analyzed. The comparison with the full set is presented in Table C.3-9 and shows no significant differences when those materials are excluded. However, in both cases, a significant trend is observed, as a function of the rod pitch in the experiment. This is discussed in the next section.

C.3.5.2. Fuel Rod Pitch Trend To better understand the observed rod pitch trend, the results for all 291 experiments are shown in Figure C.3-4 as a function of rod pitch. It appears that the trend is due to the experiments at higher rod pitch value (> 2 cm), which consistently show keff values well above 1.0. To evaluate the impact of those experiments at larger rod pitches, the Table C.3-10 shows a comparison of results with and without those experiments. When results above 2 cm rod pitch are excluded, a slightly 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 the experiments at higher rod pitch values. To ensure that a potential trend would not be ignored, all following evaluations are performed for the two conditions used above, i.e. for all rod pitch values, and for experiments with rod pitch values limited to no more than 2 cm.

C.3.5.3. Absorber Geometry C-5 HI-2 104790 REPORT HI-2104790 C-5

The criticality experiments analyzed in this report include experiments with Boron in the form of plates, absorber rods and soluble boron in water. No trend relating to these experiments is observed.

C.3.5.4. Fuel Burnup The full set of critical experiments contains experiments with fresh U0 2 fuel, with simulated spent fuel (37.5 GWd/MTU), and MOX fuel with Pu content between 2 and 20%, which is 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 the burnup range sufficiently, the experiments are subdivided into fresh U0 2 fuel, HTC experiments and MOX experiments, and compared to the results of the entire set. The comparison is shown in Table C.3-1 1. The comparison shows no significant differences between the entire set and the U0 2 and HTC subsets, but for MOX the bias is now positive (i.e. truncated bias of 0.0), with a larger uncertainty, and some trends. However, this is based on relatively small sets of experiments. Bias values are comparable between sets with and without rod pitch values above 2 cm, with a maximum absolute value of .

C.3.5.5. Unborated and Borated Water The 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 soluble boron level. Nevertheless, sets with and without soluble boron are analyzed and compared to the full set that contains all experiments. The results are shown in Table C.3-12. Similar to the previous 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 a maximum absolute value of .

C.4 Summary A set of 291 critical experiments has been selected and has been used for the validation of the Holtec International criticality safety methodology. The similarity between the chosen experiments and the actual systems has been based on a set of screening criteria as is stated in the NUREG/CR-6698 [5]. Experiments have been categorized by common features as Phase 1 through Phase 4 and selected experiments and parameterized by key variables such as lattice pitch / assembly pitch, absorber solution concentration, number of fuel rods, rod outer diameter, fuel density, screen array distance, fuel enrichment and EALF. Benchmark calculations have been performed using the Monte Carlo code MCNP5-1.51. It was determined that Phase 1 through Phase 4 and selected experiments are in sufficient agreement that this sets are lumped together as a single set of 291 experiments. The bias and bias uncertainty are presented in section C.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 are summarized in Table C.3-13. A point by point comparison between design application and benchmarks shows that the experimental range covers all the parameters. The soluble boron REPORT 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/cm 3 and the design application one is around 10.0 - 10.7 g/cm 3, it is considered 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% 235 U. 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 Alamos National Laboratory, LA-UR-03-1987 (Revised 2/1/2008).

REPORT HI-2104790 C-7

Table C.2-1 MCNP5-1.51 ZAIDs Used for Each Nuclide Nuclide ZAIID 1H 1001.50c 1°B 5010.50c 11B 5011.55c C 6000.50c 1IN 7014.50c 160 8016.50c 23 Na 11023.51c Mg 12000.50c 27 A1 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 59 Co 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 234 U 92234.50c 235u 92235.50c 236u 92236.50c 238 U 92238.50c 238 Pu 94238.50c 239pu 94239.50c 240pu 94240.50c 241pu 94241.50c 242 Pu 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 Phase 1 Critical Experiments: Water-Moderated and Reflected Arrays Case Evaluation Identification File- l~ri +oaci +/-Ox i !EALF 1 MIX-COMP-THERM-HTC-001 name '~ cclt ~ ep(eV*

2 MIX-COMP-THERM-HTC-002 3 MIX-COMP-THERM-HTC-003 - UU - U -

4 MIX-COMP-THERM-HTC-004 5 MIX-COMP-THERM-HTC-005 6 MIX-COMP-THERM-HTC-006 -U UU U 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 11 12 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 C-9 REPORT HI-2 104790 REPORT HI-2104790 C-9

Table C.3-2 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 2 Critical Ex riments: Reflected Simple Arrays Moderated by Poisoned Water with Gadolinium or Boron File-Case Evaluation Identification name kefi - 0 cai - Gexp - Gi EALF (eV) 19 MIX-COMP-THERM-HTC-019 U UIU UM T 20 MIX-COMP-THERM-HTC-020 UU 21 MIX-COMP-THERM-HTC-021 UEUV 22 MIX-COMP-THERM-HTC-022 U111U[U 23 MIX-COMP-THERM-HTC-023 U = U U0 i 24 MIX-COMP-THERM-HTC-024 U U U I 25 MIX-COMP-THERM-HTC-025 U U U[U i 26 MIX-COMP-THERM-HTC-026 U 27 MIX-COMP-THERM-HTC-027 U U.U.U..

28 MIX-COMP-THERM-HTC-028 U 29 MIX-COMP-THERM-HTC-029 U 30 MIX-COMP-THERM-HTC-030 U UU 31 MIX-COMP-THERM-HTC-031 U 32 MIX-COMP-THERM-HTC-032 U 33 MIX-COMP-THERM-HTC-033 U UUU 34 MIX-COMP-THERM-HTC-034 U U U.U.. ...

35 MIX-COMP-THERM-HTC-035 U 36 MIX-COMP-THERM-HTC-036 U lA U U U 37 MIX-COMP-THERM-HTC-037 U 38 MIX-COMP-THERM-HTC-038 U 39 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 U 42 MIX-COMP-THERM-HTC-042 U UUU-43 MIX-COMP-THERM-HTC-043 U UUU 44 MIX-COMP-THERM-HTC-044 U[

45 MIX-COMP-THERM-HTC-045 U UUU 46 MIX-COMP-THERM-HTC-046 U11UU 47 MIX-COMP-THERM-HTC-047 U U[U U i 48 MIX-COMP-THERM-HTC-048 U U 49 MIX-COMP-THERM-HTC-049 U U.U.U.U..

50 MIX-COMP-THERM-HTC-050 U U[ U W 51 MIX-COMP-THERM-HTC-051 U U UUU 52 MIX-COMP-THERM-HTC-052 U[U UU U 53 MIX-COMP-THERM-HTC-053 U U UUW 54 MIX-COMP-THERM-HTC-054 U U UUUW 55 MIX-COMP-THERM-HTC-055 U U U U REPORT HI-2104790 C-10

File-Case Evaluation Identification name keff-i + gcalc-i + gexp + ai EALF (eV) 56 MIX-COMP-THERM-HTC-056 U UUU 57 MIX-COMP-THERM-HTC-057 U[U 58 MIX-COMP-THERM-HTC-058 U U WUU 59 MIX-COMP-THERM-HTC-059 Ul*U"UW U"J _"___"

c-il HI-2 104790 REPORT HI-2104790 REPORT C-I11

Table C.3-3 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 3 Critical Experiments: Pool Storage File- ~, + ~ ~ aexp + a, EALF (eV)

Case Evaluation Identification name 60 MIX-COMP-THERM-HTC-060 - U U - -

61 MIX-COMP-THERM-HTC-061 - U U - -

62 MIX-COMP-THERM-HTC-062 - U U - -

63 MIX-COMP-THERM-HTC-063 - U U - -

64 MIX-COMP-THERM-HTC-064 - U U - -

65 MIX-COMP-THERM-HTC-065 - U U - -

66 MIX-COMP-THERM-HTC-066 - U U - -

67 MIX-COMP-THERM-HTC-067 - U U - -

68 MIX-COMP-THERM-HTC-068 - U U - -

69 MIX-COMP-THERM-HTC-069 - U U - -

70 MIX-COMP-THERM-HTC-070 - U U - -

71 MIX-COMP-THERM-HTC-071 - U U - -

72 MIX-COMP-THERM-HTC-072 - U U - -

73 MIX-COMP-THERM-HTC-073 - U U - -

74 MIX-COMP-THERM-HTC-074 - U U - -

75 MIX-COMP-THERM-HTC-075 - U U - -

76 MIX-COMP-THERM-HTC-076 - U U - -

77 MIX-COMP-THERM-HTC-077 - U U - - ~

78 MIX-COMP-THERM-HTC-078 - U U - -

79 MIX-COMP-THERM-HTC-079 - U U - -

80 MIX-COMP-THERM-HTC-080 - U U - -

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 HI-2 104790 c-i C-12 2

REPORT HI-2104790

Table C.3-4 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 4 Critical Experiments: Shipping Cask File- k~ff, +/- aca!c~ +/-~ + a, (eV)

Case Evaluation Identification name 86 MIX-COMP-THERM-HTC-086 - m - m m m 87 MIX-COMP-THERM-HTC-087 - - m m m 88 MIX-COMP-THERM-HTC-088 - - m m m 89 MIX-COMP-THERM-HTC-089 - - - m m m 90 MIX-COMP-THERM-HTC-090 - U - U U

- m m m

-~

91 MIX-COMP-THERM-HTC-091 U 92 MIX-COMP-THERM-HTC-092 - - m m m 93 MIX-COMP-THERM-HTC-093 - U - U U

- m m m m

-~

94 MIX-COMP-THERM-HTC-094 95 MIX-COMP-THERM-HTC-095 - m - m m m 96 MIX-COMP-THERM-HTC-096 - ~ m m m 97 MIX-COMP-THERM-HTC-097 - m - m m m 98 MIX-COMP-THERM-HTC-098 - - U U U

- - m m m

-~

99 MIX-COMP-THERM-HTC-099 100 MIX-COMP-THERM-HTC- 100 - - U U U

- - U

-~

101 MIX-COMP-THERM-HTC- 101 U U

-~

102 MIX-COMP-THERM-HTC- 102 U U

- - m m -

-~

103 MIX-COMP-THERM-HTC- 103 104 MIX-COMP-THERM-HTC-104 I I I I I I 105 106 MIX-COMP-THERM-HTC- 105 MIX-COMP-THERM-HTC- 106

-- -- -U m U

m U

m ~

107 MIX-COMP-THERM-HTC- 107 - - U U U ~

108 MIX-COMP-THERM-HTC-108 - - m m m 109 MIX-COMP-THERM-HTC- 109 - m m m 110 MIX-COMP-THERM-HTC-I 10 - m m m 111 MIX-COMP-THERM-HTC- 11 - U U U U

- m m m

-~

112 MIX-COMP-THERM-HTC-1 12 113 MIX-COMP-THERM-HTC- 113 - U U U U

-~

114 MIX-COMP-THERM-HTC-1 14 U U U U ~

115 MIX-COMP-THERM-HTC-1 15 - U U U U

- ~

-~

116 MIX-COMP-THERM-HTC- 116 U U U U 117 MIX-COMP-THERM-HTC- 117 U U U U 118 MIX-COMP-THERM-HTC-1 118 I IN I I I 119 MIX-COMP-THERM-HTC- 119 120 MIX-COMP-THERM-HTC- 120 120 MIX-COMP-THERM-HTC- 121 121 MIX-COMP-THERM-HTC-121 U U U U U 122 MIX-COMP-THERM-HTC-122 123 MIX-COMP-THERM-HTC-123 REPORT HI-2104790 C-13

Case Evaluation Identification File-name kef ucalc- G~x + (y 1 EALF I J 124 MIX-COMP-THERM-HTC- 124 U U U ~ I 125 MIX-COMP-THERM-HTC- 125 U U - ~ I 126 MIX-COMP-THERM-HTC- 126 - U U U ~ I 127 MIX-COMP-THERM-HTC- 127 U U U ~ I 128 MIX-COMP-THERM-HTC- 128 U-U 129 MIX-COMP-THERM-HTC- 129 U U U I m m m m

-~

130 MIX-COMP-THERM-HTC- 130 I 131 132 MIX-COMP-THERM-HTC- 131 MD(-COMP-ThERM-HTC-132 I I

- m m m I

_32 . .. . . .. .. . ... . . ... . .. .2 - - U U U U ~ I 133 MIX-COMP-THERM-HTC-133 U U U ~ I 134 MIX-COMP-THERM-HTC-134 U U U ~ I 135 MIX-COMP-THERM-HTC-135 U U U -~ I 136 MIX-COMP-THERM-HTC-136 U U U I

- ~

-~

137 MIX-COMP-THERM-HTC-137IM U U I 138 MIX-COMP-THERM-HTC- 138 U U - ~ I 139 MIX-COMP-THERM-HTC-139 U U - I m - m

-~

140 MIX-COMP-THERM-HTC- 140 F- I 141 MIX-COMP-THERM-HTC- 141 - -- m I 142 MIX-COMP-THERM-HTC- 142 U -- m I 143 MIX-COMP-THERM-HTC- 143 F- - U - m I 144 MIX-COMP-THERM-HTC-144 F- m U - m I 145 MIX-COMP-THERM-HTC- 145 F-- - -- m I 146 MIX-COMP-THERM-HTC- 146 F--1 - - m I 147 MIX-COMP-THERM-HTC- 147 F-- U - - m I 148 MIX-COMP-THERM-HTC- 148 F-- - m m I 149 MIX-COMP-THERM-HTC- 149 F- - m m I 150 MIX-COMP-THERM-HTC- 150 F--1 - m m I 151 MIX-COMP-THERM-HTC- 151 F-1 - m m I 152 MIX-COMP-THERM-HTC- 152 F--1 - m m I 153 MIX-COMP-THERM-HTC-153 F-1 - m m I 154 MIX-COMP-THERM-HTC-154 - m m I 155 MIX-COMP-THERM-HTC- 155 - m m 156 MIX-COMP-THERM-HTC-156 U ~ U ~ I C-14 REPORT HI-2 104790 REPORT HI-2104790 C-14

Table C.3-5 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Selected Critical Experiments File- klff. + ralc +/- 0 exp EALF Case Evaluation Identification name

+/- (yi (eV) 157 Core I__

158 Core II 159 Core III I 160 Core IX 161 Core X 162 Core XI 163 Core XII 164 Core XIII I 165 Core XIV I 166 Core XV 167 Core XVI I 168 Core XVII 169 Core XVIH I 170 Core XIX I 171 Core XX 172 Core XXI I 173 S-type Fuel, w/886 ppm B 174 S-type Fuel, w/746 ppm B 175 SO-type Fuel, w/1 156 ppm B 176 Case 1 1337 ppm B 177 Case 12 1899 ppmB B 178 Water Moderator 0 gap 179 Water Moderator 2.5 cm gap 180 Water Moderator 5 cm gap 181 Water Moderator 10 cm gap 182 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 Sheets 191 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 absorber c-IS REPORT 111-2104790 REPORT HI-2104790 C-15

File- - acalc- EALF Case Evaluation Identification "CIM'nP ke~ff-i + (Yexp - ai 195 Exp. 009 1.05% Boron Steel plates 196 Exp. 009 1.62% Boron Steel plates 197 Exp. 031 - Boral plates 198 Experiment 214R - with flux traps 199 Experiment 214V3 -with flux trap 200 Case 173 - 0 ppm B 201 Case 177 - 2550 ppm B 202 MOX Fuel - Type 3.2 Exp. 21 203 MOX Fuel - Type 3.2 Exp. 43 204 MOX Fuel - Type 3.2 Exp. 13 205 MOX Fuel - Type 3.2 Exp. 32 206 Saxton Case 52 PuO2 0.52" pitch 207 Saxton Case 52 U 0.52" pitch 208 Saxton Case 56 PuO2 0.56" pitch 209 Saxton Case 56 borated PuO2 210 Saxton Case 56 U 0.56" pitch 211 Saxton Case 79 PuO2 0.79" pitch 212 Saxton Case 79 U 0.79" pitch 213 0.700-in. pitch 0 ppm B 214 0.700-in. pitch 688 ppm B 215 0.870-in. pitch 0 ppm B 216 0.870-in. pitch 1090 ppm B 217 0.990-in. pitch 0 ppm B 218 0.990-in. pitch 767 ppm B 219 Saxton Case PuO2 0.735" pitch 220 Saxton Case PuO2 1.04" pitch 221 8 wt% 240Pu 0.80" pitch 222 8 wt% 240Pu 0.93" pitch 223 8 wt% 240Pu 1.05" pitch 224 8 wt% 240Pu 1.143" pitch 225 8 wt% 240Pu 1.32" pitch 226 8 wt% 240Pu 1.386" pitch 227 16 wt% 240Pu 0.93" pitch 228 16 wt% 240Pu 1.05" pitch 229 16 wt% 240Pu 1.143" pitch 230 16 wt% 240Pu 1.32" pitch 231 24 wt% 240Pu 0.80" pitch 232 24 wt% 240Pu 0.93" pitch 233 24 wt% 240Pu 1.05" pitch 234 24 wt% 240Pu 1.143" pitch REPORT HI-2104790 C-16

I 1 7 1 T r File- + Gcalc. EALF Case Evaluation Identification keff-i I O

exp - oi (eV)L 235 24 wt% 240Pu 1.32" pitch 236 24 wt% 240Pu 1.386" pitch 237 18 wt% 240Pu 0.85" pitch 238 18 wt% 240Pu 0.93" pitch 239 18 wt% 240Pu 1.05" pitch 240 18 wt% 24OPu 1. 143" pitch 241 18 wt% 240Pu 1.386" pitch 242 18 wt% 240Pu 1.60" pitch U

243 18 wt% 240Pu 1.70" pitch 317 Core XI, 1182 ppm, 36 Pyrex Rods U

318 Core XI, 1182 ppm, 36 Pyrex Rods U 319 Core XI, 1032.5 ppm, 72 Pyrex Rods U U 320 Core XI, 1032.5 ppm, 72 Pyrex Rods U 321 Core XI, 794 ppm, 144 Pyrex Rods U m U 322 Core XI, 779 ppm, 144 Pyrex Rods m - m mi 323 360 Core XI, 1245 ppm, 72 Vicor Rods Core IV U ~ - Mi UUUMI 361 Core V m - Urn' 362 Core VI m - rn ml 363 Core VII m - U Mi 364 Core VIII UUUMI 445 670 pins, A1203-B4C rods UUUMI 446 672 pins, A1203-B4C rods ~UUMi 447 668 pins, A1203-B4C rods UUUMI 448 668 pins, A1203-B4C rods UUUMI 479 8 wt% 240Pu 1.05" pitch, B4 Rods UUUMi 480 8 wt% 240Pu 1.05" pitch, B3 Rods UUUMI 481 8 wt% 240Pu 1.05" pitch, B2 Rods UUUMi 482 8 wt% 240Pu 1.05" pitch, BI Rods UUUMI 8 wt% 240Pu 1.05" pitch, B4+Cd 484 P nd.* -i 485 8 wt% 240Pu Rods105" pitch, B3+Cd U U U .. U -U 486 I8 1.05" pitch, B2+Cd 486 wt% 240Pu ~RodsII

~Rods 8 wt% 240Pu 1.05" pitch, B 1+Cd 487 Rods U U ~ U U

-i 491 8 wt% 240Pu 1.32" pitch, B4 Rods UUUUMI 492 8 wt% 240Pu 1.32" pitch, B3 Rods UUUUMI 493 8 wt% 240Pu 1.32" pitch, B2 Rods UUUUMI 494 , 8 wt% 240Pu 1.32" pitch, B 1 Rods UUUUMI REPORT HI-2104790 C-17

Case Evaluation Identification File- ke i caic-- (exp 4-j EALF name ýfj(V 496 8 wt% 240Pu 1.32" pitch, B4+Cd 496 Rods 497 8 wt% 240Pu 1.32" pitch, B3+Cd Rods 498 8 wt% 240Pu 1.32" pitch, B2+Cd 498 Rods 5_49 16 wt% 240Pu 1.32" pitch, B4I+Cd Rods 504 16 wt% 240Pu 1.05" pitch, B4 Rods m U m

505 16 wt% 240Pu 1.05" pitch, B3 Rods _____

506 16 wt% 240Pu 1.05" pitch, B2 Rods ______ m 507 16 wt% 240Pu 1.05" pitch, Bl Rods _____ M 509 16 wt% 240Pu 1.05" pitch, B4+Cd U 509 Rods 510 16 wt% 240Pu 1.05" pitch, B3+Cd U 510 Rods 511 16 wt% 2 4 0Pu 1.05" pitch, B2+Cd

~Rods U

512 16wt%240Pu 1.05" pitch, Bl+CdoU 5162 Rods 516 24 wt% 240Pu 1.05" pitch, B4 Rods

  • A wUl*

517 24 wt% 240Pu 1.05" pitch, B3 Rods_______

518 24 wt% 240Pu 1.05" pitch, B2 Rods ___ _ M 519 24 wt% 240Pu 1.05" pitch, BI Rods _ _ _ _l 521 24 wt% 240Pu 1.05" pitch, B4+Cd U 521 Rods 522 52 24 wt% 240Pu 1.05" pitch, B3+Cd Rods -_

U

  • M_00 U -o U

523 24 wt% 240Pu 1.05" pitch, B2+Cd U U _

524 Rods

  • _ _** ._*_* _
  • 524] 24 wt% 240Pu 1.05" pitch, Bl+Cýd

__ Rods C-i 8 HI-2104790 REPORT HI-2104790 C-18

Table C.3-6 Descriptive Statistics of the MCNP5-1.51 Calculational Results No. of Experiment Description keff range EALF (eV) range exp.

Phase 1 18 Phase 2 41 Phase 3 26 Phase 4 71 Selected Experiments 135 All experiments 291 Table C.3-7 Normality Test Results for the MCNP5-1.51 calculations Experiment No. of Shapiro-Wilk Pearson's chi-square (x2 )

Description exp. WtPePt w X2 n Pd(x 2 ;d) Normal Phase 1 18 Phase 2 41 Phase 3 26 Phase 4 71 HTC Experiments 156 Selected Experiments 135 -_M I I=

All experiments 291 C-19 REPORT HI-2104790 C-19

Table C.3-8 Trending Analysis Results for the MCNP5-1.51 calculations Correlation Experiment No. of Probability, Correlated Parameter, x Coefficient, Correlation Regression Equation, k(x)

Description exp. 2 Pd(N;r) r EALF Phase 1 18 Pitch Number of Rods I EALF Pitch Phase 2 41 Number of Rods Gadolinium Conc.

Boron Conc.

EALF Phase 3 26 Water Gap I Number of Rods EALF

  • I Phase 4 71 Water Gap m I Screen Array Distance EALF I

Selected Pitch Rod OD

. I Experiments Fuel Density U U Enrichment Pu Enrichment I

EALF Pitch 291 Rodch All experiments Rod OD Fuel Density C-20 REPORT HI-2104790 C-20

Table C.3-9 Analysis of Neutron Absorbers and Reflector Materials for the MCNP5-1. 51 calculations No. of Bias Normality Experiment Desriptint ex. o Bias Uncerta X2 Significant Trends Description exp. inty (Pd(X2 ;d))

All experiments 291 All except those with Gadolinium, 201 -

Cadmium and Lead Table C.3-10 Analysis of Fuel Rod Pitch Trend for the MCNP5-1. 51 calculations Bias Normality Experiment Rod Bias Uncerta X2 Significant Trends Description Pitch inty 2 (Pd(X ;d))

All (291 All totaln L experiments <2 cm total)

All except All (201 m -

those with total) -

Gadolinium, _

Cadmium and <=2 cm -

Lead (144 -

total)

C-2 1 HI-2 104790 REPORT HI-2104790 REPORT C-21

Table C.3-11 Analysis of Fuel Burnup for the MCNP5-1. 51 calculations Bias Normality Experiment Rod X2 Bias Uncerta Significant Trends Description Pitch 2 inty (Pd(X ;d))

-I- 4 4 All except All (201 those with total) - -

Gadolinium, Cadmium and <=2 cm Leadt (144 - m total)

All (61 Fresh U0 2 total)

Fuel <=2 cm (52 total) ____

All (85 HTC total)

Experiments

<=2 cm (82 total) _ __ ]

All (55 total) -

LL 4 + 4 MOX Experiments <=2 cm (10 total) M M U tNote: 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 Bias Normality Experiment Rod X2 Bias Uncerta Significant Trends Description Pitch inty (Pd(X ;d))

All except All (201 those with total) - m Gadolinium, Cadmium and <=2 cm Leadt (144 ýIm total)

All (149 total) -I-All with Fresh i i Water

<=2 cm (94 total) E--71 All (52 All 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 AOA Parameter Design Benchmarks Validated Application 2 35 2 39 24 1 23 5 23 9 241 2 35 2 39 24 1 Fissionable Material U, Pu, Pu U, Pu, Pu U, Pu, Pu Isotopic Composition 235 u/ut < 5.Owt% 1.57-5.74% < 5wt%

Pu/(U+Pu) < 20wt% 1.104-20% < 20wt%

Physical Form U0 2 MOX U0 2 MOX U0 2 MOX Fuel Density (g/cm 3) 10.0- 10.7 9.2 - 10.4 9.2 - 10.7 Moderator Material (coolant) H H H Physical Form H 20 H 20 H 20 Density (g/cm 3) around 1.0 g/cm 3 around 1.0 g/cm 3 around 1.0 g/cm 3 Reflector Material H H H Physical Form H2 0 H 20 H2 0 Density (g/cm 3) around 1.0 g/cm 3 around 1.0 g/cm 3 around 1.0 g/cm 3 Interstitial Reflector Material Plate Steel or Lead Steel or Lead Steel or Lead Absorber Material None, Boron (89 - None, Boron (0 -

Soluble None, Boron or 595 ppm) or 1000 ppm) or Gadolinium Gadolinium (49.2 - Gadolinium (0 to 199.7 ppm) 1000 ppm)

Rods Boron Pyrex , Vicor' or Boron B-Al Separating Material Water, B-SS, Water, B-SS, Boral Water, B-SS, Boral Plate Boral or Cadmim oor Cadmium or Cadmium Cadmium Geometry Lattice type Square Square, Triangle Square, Triangle 1.26-1.47 LatticeLattce Ptch cm)

Pitch (cm) 1.24 (PWR) -1.88 0.968 to 4.318 0.968 to 4.318 (BWR)

Thermal Neutron Energy Thermal spectrum II Thermal spectrum Thermal spectrum C-24 REPORT HI-2104790 REPORT HI-2104790 C-24

Figure Proprietary Figure C.3-1 Frequency Chart for Calculated keff of the Selected 243 Benchmarks for the MCNP5-1.51 code Figure Proprietary Figure C.3-2 Frequency Chart for Calculated EALF (eV) of the Selected 243 Benchmarks for the MCNP5-1.51 code REPORT HI-2104790 C-25

Figure Proprietary Figure C.3-3 MCNP5-1.51 Calculated kff Values for Various Values of the Spectral Index (All Experiments)

C-26 HI-2 104790 REPORT 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 D Benchmark of MCNP5-1.51 with ENDF/B-VII (total number of pages: 51 including this page)

D- I REPORT HI-2 104790 REPORT HI-2104790 D-1

D.1 Introduction This Appendix presents the analysis of the validation results for MCNP5-1.51 code and includes the 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 Data The computer code MCNP5-1.51 [D. 1] is the continuous energy Monte Carlo codes and treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Thermal neutrons are described by both the 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 neutrons start from an arbitrary distribution, causing a generally very large variance of results from the first cycles in comparison with the following cycles. Therefore, all MCNP5-1.51 calculations are performed with 12,000 histories per cycle, 50 skipped cycles before averaging, and 100 cycles that 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 Results D.3.1. Calculational Results The calculation results for the 156 HTC critical experiments and for the 376 selected critical experiments 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 in Appendix 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.3-1 through Table D.3-5.

Figure D.3-1 and D.3-2 are histograms showing the frequency of calculated ke and EALF for all 532 benchmarks. The nominal calculated k, values range from results values show a range between O i . The EALF Descriptive statistics for the different group of experiments is summarized in Table D.3-6.

D.3.2. Normality Test In order to assess the normality assumption, Shapiro and Wilk [5] test has been used for groups with fewer than 50 samples while the Pearson's chi-square (X2) test [4] has been used for samples larger than 20 samples. The tests are applied to the group of experiments in terms of the categories as set forth in Appendix B.

For the Shapiro and Wilk test, Table D.3-7 shows the computed Wtest value, and W value that can be obtained for the number of experiments from [5] to accept the normality hypothesis. If W REPORT 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 degrees of 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 > 10 2 ) is large, the obtained and expected distributions are consistent; if it is small, they probably disagree. In particular, if Pd(X2 > ,k 0 2 ) is less than 5%, we say that the disagreement is significant and reject the assumed distributions 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 Analysis Trends 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) zz t A=

(7-4)

The squared term of the linear correlation factor r defined below (from Reference [5]) is used to quantitatively 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. A more quantitative measure of the fit can be found by using Appendix C in Reference [4]. The interpolation was applied, if necessary. For any given observed value ro, PN(IrI >- Irol) is the probability that N measurements of two uncorrelated variables would give a coefficient r as large as 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, if PN(IrI -- Irol) 5 5%, the correlation is called significant; if it is less than 1%, the correlation is called highly significant.

D-3 HI-2 104790 REPORT HI-2104790 REPORT D-3

The validation results are analyzed for the group of all experiments. Independent variables used in the trending analysis, correlation coefficients and trending analysis results are summarized in Table D.3-8.

D.3.4. Bias and Bias Uncertainty In this section, benchmark results are analyzed using the statistical method described in section 2.2.

The first step is to evaluate whether the HTC experiments and selected ex eriments, should be reduced to a single set. The mean, of the HTC data set is and the mean kff of the selected experiments data is . The difference between the means is just 0.0010 which is less than the uncertainty. These sets are water moderated uranium or mixed plutonium-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 normally distributed. Therefore, the distribution free approach [6] is used for all subsets with the rejected normality distribution. The lower tolerance limit with 95% probability and 95% confidence level is determined for order data [6] and the difference between weighted average keff and this lower tolerance limit is used to determine the bias uncertainty. This is conservative since the data is close to the normal distribution. The distribution free bias uncertainty is also provided in all subsequent 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 the MCNP5-1.51 code is shown in the table below Calculational Bias of the MCNP5-1.51 code Description Total Bias Bias Uncertainty HTC and Selected Experiments D.3.5. Applicability of MCNP5-1.51 Validation Results This 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 water REPORT 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. In some 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 are evaluated for fuel rod outer diameter, fuel rod pitch, fuel density, boron content, U or Pu enrichment and EALF. To estimate a significance of observed trend, the residuals from the trend equation were tested for a normal distribution [D.2]. If residuals are normally distributed then there 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 were calculated 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 (B4 C), and boron in borated steel, Pyrex, Boroflex and borated aluminum o Soluble gadolinium in water and Gd 20 3 rods o Cadmium

" Reflectors o Steel o Lead o Water Some typical configurations do not contain gadolinium or cadmium neutron absorbers or lead reflectors. To verify that including those materials does not have a significant effect on the results of the benchmarking analyses, a subset without those experiments containing those materials was analyzed. In addition, according to recommendations of NUREG-6979 [B. 13], the following HTC experiments were also excluded: 61, 65, 67, 86, 97, 98, 99, 102, 124, 135, and 137. 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 function of EALF was determined by the residuals normality test. This correlation is presented in the Figure D.3-4. The bias and bias uncertainty as a function of the EALF were calculated for this trend and shown in Table D.3-10, with a maximum absolute value of .

D.3.5.2. Absorber Geometry The criticality experiments analyzed in this report include experiments with Boron in the form of plates, absorber rods and soluble boron in water. No trend relating to these experiments is observed.

D.3.5.3. Fuel Burnup The full set of critical experiments contains experiments with fresh U0 2 fuel, with simulated spent fuel (37.5 GWd/MTU), and MOX fuel with Pu content between 1.5 and 20%, which is REPORT 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 the burnup range sufficiently, the experiments are subdivided into fresh U0 2 fuel and spent fuel with HTC and MOX experiments, and compared to the results of the entire set. The comparison is shown in Table D.3-1 1. The comparison shows no significant differences between the entire set and the fresh and spent fuel subsets. However, in some cases, the correlations are observed. The significant trends as a function of EALF and Pu enrichment were determined in the spent fuel subset by the residuals normality test. These correlations are presented in the Figure D.3-5 and Figure D.3-6. The bias and bias uncertainty as a function of the EALF and Pu enrichment were calculated for these trends and shown in Table D.3-12, with a maximum absolute value of -

D.3.5.4. Unborated and Borated Water The 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 soluble boron level. Nevertheless, sets with and without soluble boron are analyzed and compared to the full set that contains all experiments. The results are shown in Table D.3-13. Similar to the previous 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 U enrichment and in the borated water subset as a function of fuel density that were determined by the residuals normality test. These correlations are presented in the Figure D.3-7 through Figure D.3-9. The bias and bias uncertainty as a function of the EALF, U enrichment and fuel density were 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 Water To perform more detailed evaluation of the set of critical experiments, the additional four subsets with different combinations of fuel bumup and unborated/borated water were analyzed. The results are shown in Table D.3-15. There are significant EALF trend in the subset of fresh U0 2 fuel with fresh water, significant trends in the subset of spent fuel with fresh water as a function of EALF and Pu enrichment and significant trends in the subset of spent fuel with borated water as 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 bias uncertainty 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 of D.4 Summary A set of 532 critical experiments has been selected and has been used for the validation of the Holtec International criticality safety methodology. The similarity between the chosen experiments and the actual systems has been based on a set of screening criteria as is stated in the NUREG/CR-6698 [5]. Experiments have been categorized by fuel burnup as fresh U0 2 fuel and spent fuel with HTC and MOX experiments or by unborated and borated water condition and REPORT HI-2104790 D-6

parameterized by key variables such as lattice pitch / assembly pitch, absorber solution concentration, number of fuel rods, rod outer diameter, fuel density, screen array distance, fuel enrichment and EALF. Benchmark calculations have been performed using the Monte Carlo code MCNP5-1.51. It was determined that HTC experiments and selected experiments are in sufficient agreement that this sets are lumped together as a single set of 532 experiments. The bias and bias uncertainty are presented in section D.3.4. The applicability of validation results is considered in section D.3.5.

The range of key parameters for the design application, benchmarks and validated AOA are summarized in Table D.3-17. A point by point comparison between design application and benchmarks shows that the experimental range covers all the parameters. The soluble boron concentration 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/cm 3 and the design application one is around 10.0 - 10.7 g/cm 3, it is considered 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 Alamos National Laboratory, LA-IJR-03-1987 (Revised 2/1/2008).

[D.2] J. W. Barnes, "Statistical Analysis for Engineers and Scientists", McGraw-Hill Inc., 1988 REPORT HI-2104790 D-7

Table D.2-1 ZAIDs Used for Each Nuclide Nuclide MCNP5.1.51 ZAD Nuclide MCNP5.1.51 ZAD Nuclide MCNP5.1.51 ZI ZAID ZAID ZAID

'H 1001.70c 48Ti 22048.70c 'l°Mo 42100.70c 2H 1002.70c 49Ti 22049.70c 107 Ag 47107.70c 4He 5°Ti 2004.70c 22050.70c 109Ag 47109.70c 10B 5010.70c 50 Cr 24050.70c 106 Cd 48106.70c "1B 5011.70c 52Cr 24052.70c 108Cd 48108.70c C 6000.70c 53 Cr 24053.70c l"0Cd 48110.70c 14N 7014.70c 54 111Cd Cr 24054.70c 48111.70c 160 8016.70c 55Mn 25055.70c 112Cd 48112.70c 2°Ne 54 10020.42c Fe 26054.70c 113Cd 481 13.70c 23 Na 11023.70c 56Fe 26056.70c 4 11 Cd 48114.70c 24 Mg 12024.70c 57Fe 26057.70c 116Cd 48116.70c 25 Mg 12025.70c 58Fe 26058.70c 1131n 49113.70c 26Mg 59 1151n 12026.70c Co 27059.70c 49115.70c 27A1 13027.70c 58Ni 28058.70c 1 2 Sn 50112.70c 28Si 14028.70c 6°Ni 28060.70c 114Sn 50114.70c 29 61Ni Si 14029.70c 28061.70c "15 Sn 50115.70c 3°Si 62Ni 14030.70c 28062.70c 116Sn 50116.70c 31p 15031.70c 64Ni 28064.70c 1 7TSn 50117.70c 32s 16032.70c 63 Cu 29063.70c 118Sn 50118.70c 65 36Ar 18036.70c Cu 29065.70c 119Sn 501 19.70c 38Ar 18038.70c Zn 30000.70c 120Sn 50120.70c 18040.70c 90 4°Ar Zr 40090.70c 122Sn 50122.70c 39K 19039.70c 91 Zr 40091.70c 124 Sn 50124.70c 40 92Zr K 19040.70c 40092.70c 144Sm 62144.70c 41K 19041.70c 94 Zr 40094.70c 47 1 Sm 62147.70c 4°Ca 20040.70c 96 Zr 40096.70c 148Sm 62148.70c 42 Ca 20042.70c 93 Nb 41093.70c 149 sm 62149.70c 43Ca 20043.70c 92 50 Mo 42092.70c ° Sm 62150.70c 44Ca 20044.70c 94 Mo 42094.70c I52Sm 62152.70c 46Ca 20046.70c 95 Mo 42095.70c '54Sm 62154.70c 48Ca 20048.70c 96 Mo 42096.70c 152Gd 64152.70c 46Ti 22046.70c 97 Mo 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 ZAID 156Gd 64156.70c 179 236 Hf 72179.70c U 92236.70c 157 18 0 Gd 64157.70c Hf 72180.70c 238u 92238.70c 58 204 1 Gd 64158.70c Pb 82204.70c 238 Pu 94238.70c 16°Gd 64160.70c 206 239 Pb 82206.70c pu 94239.70c 74 207 1 Hf 72174.70c Pb 82207.70c 24 0 Pu 94240.70c 76 2 08 1 Hf 72176.70c Pb 82208.70c 241Pu 94241.70c 177 Hf 72177.70c 234u 92234.70c 242 Pu 94242.70c 78 1 Hf 72178.70c 235u 92235.70c 24 1 AnM 95241.70c D-9 REPORT HI-2 104790 REPORT HI-2104790 D-9

Table D.3-1 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 1 Critical Experiments: Water-Moderated and Reflected Arrays File- EALF Case Evaluation Identification kDif- -I- Ccalc-i +/- Gexp +/- (yi (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 11 12 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 2 Critical Experiments: Reflected Simple Arrays Moderated by Poisoned Water with Gadolinium or Boron File-Case Evaluation Identification name keff-i 1 acalc-i - (Yexp +/-" 3 i EALF (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 D- 11 HI-2 104790 REPORT HI-2104790 D-11

File-Case Evaluation Identification name +- Gcalc-i +/- crexp - (yi EALF (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 I

D-12 REPORT HI-2 104790 REPORT HI-2104790 D-12

Table D.3-3 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 3 Critical Experiments: Pool Storage File- kff-i Case Evaluation Identification nn~mp kI f I Gcalc-i I eFXP I +/- c7i EALF (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 85 MIX-COMP-THERM-HTC-084 MIX-COMP-THERM-HTC-085 I I-D- 13 HI-2 104790 REPORT HI-2104790 D-13

Table D.3-4 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Phase 4 Critical Experiments: Shipping Cask I File- EALF Case Evaluation Identification keff.i Gcaxp

+/- +/- (7i (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- 101 102 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 10 111 MIX-COMP-THERM-HTC-1 111 112 MIX-COMP-THERM-HTC-1 12 113 MIX-COMP-THERM-HTC-1 13 114 MIX-COMP-THERM-HTC-1 14 115 MIX-COMP-THERM-HTC-1 15 116 MIX-COMP-THERM-HTC-1 16 117 MIX-COMP-THERM-HTC- 117 118 MIX-COMP-THERM-HTC-1 18 119 MIX-COMP-THERM-HTC- 119 120 MIX-COMP-THERM-HTC-120 121 MIX-COMP-THERM-HTC-121 122 MIX-COMP-THERM-HTC-122 123 MIX-COMP-THERM-HTC-123 D-14 REPORT HI-2 104790 REPORT HI-2104790 D-14

File- -I- Ucalc- EALF Case Evaluation Identification keff-i 4- (Fexp (,=X'l 124 MIX-COMP-THERM-HTC- 124 125 MIX-COMP-THERM-HTC- 125 126 MIX-COMP-THERM-HTC- 126 127 MIX-COMP-THERM-HTC- 127 128 MIX-COMP-THERM-HTC- 128 129 MIX-COMP-THERM-HTC- 129 130 MIX-COMP-THERM-HTC- 130 131 MIX-COMP-THERM-HTC- 131 132 MIX-COMP-THERM-HTC- 132 133 MIX-COMP-THERM-HTC-133 134 MIX-COMP-THERM-HTC-134 135 MIX-COMP-THERM-HTC- 135 136 MIX-COMP-THERM-HTC-136 137 MIX-COMP-THERM-HTC- 137 138 MIX-COMP-THERM-HTC-138 139 MIX-COMP-THERM-HTC-139 140 MIX-COMP-THERM-HTC-140 141 MIX-COMP-THERM-HTC- 141 142 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- 147 148 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- 152 153 MIX-COMP-THERM-HTC-153 154 MIX-COMP-THERM-HTC-154 155 MIX-COMP-THERM-HTC-155 156 MIX-COMP-THERM-HTC-156 D- 15 HI-2 104790 REPORT HI-2104790 D-15

Table D.3-5 The MCNP5-1.51 Calculational Results and Measurements Uncertainties for Selected Critical Experiments r r 1 File- I (cale- EALF Case Evaluation Identification keff-i I Gexp -+/-'(i 157 Core l 158 Core II 159 Core HII 160 Core IX 161 Core X 162 Core XI 163 Core XII 164 Core XIII 165 Core XIV 166 Core XV 167 Core XVI 168 Core XVII 169 Core XVIII 170 Core XIX 171 Core XX 172 Core XXI 173 S-type Fuel, w/886 ppm B 174 S-type Fuel, w/746 ppm B 175 SO-type Fuel, w/1 156 ppm B 176 Case 1 1337 ppm B 177 Case 12 1899 ppmB B 178 Water Moderator 0 gap 179 Water Moderator 2.5 cm gap 180 Water Moderator 5 cm gap 181 Water Moderator 10 cm gap 182 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 Sheets 191 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 absorber REPORT HI-2104790 D- 16

File- - Gcalc._ EALF Case Evaluation Identification keff-i +/-- (yexp +/- Gi (P17) 195 Exp. 009 1.05% Boron Steel plates 196 Exp. 009 1.62% Boron Steel plates 197 Exp. 031 - Boral plates 198 Experiment 214R- with flux traps 199 Experiment 214V3 -with flux trap 200 Case 173 - 0 ppm B 201 Case 177 - 2550 ppm B 202 MOX Fuel - Type 3.2 Exp. 21 203 MOX Fuel - Type 3.2 Exp. 43 204 MOX Fuel - Type 3.2 Exp. 13 205 MOX Fuel - Type 3.2 Exp. 32 206 Saxton Case 52 PuO2 0.52" pitch 207 Saxton Case 52 U 0.52" pitch 208 Saxton Case 56 PuO2 0.56" pitch 209 Saxton Case 56 borated PuO2 210 Saxton Case 56 U 0.56" pitch 211 Saxton Case 79 PuO2 0.79" pitch 212 Saxton Case 79 U 0.79" pitch 213 0.700-in. pitch 0 ppm B 214 0.700-in. pitch 688 ppm B 215 0.870-in. pitch 0 ppm B 216 0.870-in. pitch 1090 ppm B 217 0.990-in. pitch 0 ppm B 218 0.990-in. pitch 767 ppm B 219 Saxton Case PuO2 0.735" pitch 220 Saxton Case PuO2 1.04" pitch 221 8 wt% 240Pu 0.80" pitch 222 8 wt% 240Pu 0.93" pitch 223 8 wt% 240Pu 1.05" pitch 224 8 wt% 240Pu 1.143" pitch 225 8 wt% 240Pu 1.32" pitch 226 8 wt% 240Pu 1.386" pitch 227 16 wt% 240Pu 0.93" pitch 228 16 wt% 240Pu 1.05" pitch 229 16 wt% 240Pu 1.143" pitch 230 16 wt% 240Pu 1.32" pitch 231 24 wt% 240Pu 0.80" pitch 232 24 wt% 240Pu 0.93" pitch 233 24 wt% 240Pu 1.05" pitch 234 24 wt% 240Pu 1.143" pitch D- 17 HI-2 104790 REPORT HI-2104790 REPORT D-17

File- EALF Case Evaluation Identification keiffi 11 acalc- 1 1 exp 1 +- Cj name 235 24 wt% 240Pu 1.32" pitch 236 24 wt% 240Pu 1.386" pitch 237 18 wt% 240Pu 0.85" pitch 238 18 wt% 240Pu 0.93" pitch 239 18 wt% 240Pu 1.05" pitch 240 18 wt%240Pu 1.143" pitch 241 18 wt% 240Pu 1.386" pitch 242 18 wt% 240Pu 1.60" pitch 243 18 wt% 240Pu 1.70" pitch 244 1 Cluster 245 3 Clusters, Separation 11.92 cm 246 3 Clusters, Separation 8.41 cm 247 3 Clusters, Separation 10.05 cm 248 3 Clusters, Separation 6.39 cm 249 3 Clusters, Separation 9.01 cm 250 3 Clusters, Separation 4.46 251 1 Cluster, l0xll.51 252 1 Cluster, 9x13.35 253 1 Cluster, 8x16.37 254 3 Clusters, Separation 7.11 cm 255 256 I1Cluster, 614.4 Rods, Gd water impurity 1 Cluster, 529.3 Rods 257 1 Cluster, 523.9 Rods 258 1 Cluster, 525.3 Rods 259 1 Cluster, 595.4 Rods 260 1 Cluster, 485.8 Rods 261 1 Cluster, 523.8 Rods 262 1 Cluster, 505.4 Rods 263 4 Clusters, Separation 2.59 cm 264 2 Clusters, Separation 1.68 cm 265 4 Clusters, Separation 4.27 cm 266 4 Clusters, Separation 5.95 cm 267 4 Clusters, Separation 5.11 cm 268 4 Clusters, Separation 6.66 cm 269 4 Clusters, Separation 7.53 cm 270 4 Clusters, Separation 9.00 cm 271 4 Clusters, Separation 9.97 cm 272 4 Clusters, Separation 11.45 cm 273 4 Clusters, Separation 13.87 cm REPORT HI-2104790 D-18

File-Case Evaluation Identification

  • 1 -'p keff. + calc- .+ Gx 274 -3Clusters, Separation 9.88 cm 275 -3 Clusters, Separation 6.78 cm 276 3 Clusters, Separation 6.176 cm -7 277 1 Cluster, 225.8 Rods, Gd water impurity - --

278 1 Cluster, 216.2 Rods 279 1 Cluster, 216.6 Rods 280 1 Cluster, 218.6 Rods 281 1 Cluster, 167.85 Rods 282 1 Cluster, 203 Rods -T 283 1 Cluster, 173.5 Rods 284 2 Clusters, Separation 2.83 cm 285 3 Clusters, Separation 12.27 cm 286 3 Clusters, Separation 12.493 cm 287 4 Clusters, Separation 4.72 cm 288 4 Clusters, Separation 8.38 cm 289 4 Clusters, Separation 10.86 cm 290 4 Clusters, Separation 11.29 cm 7 291 4 Clusters, Separation 12.02 cm 292 4 Clusters, Separation 13.64 cm 293 4 Clusters, Separation 14.98 cm 294 4 Clusters, Separation 19.81 cm 295 4 Clusters, Separation 8.50 cm 296 19xl9, RodPitchi- 1.849 cm 297 20x20, Rod Pitch - 1.849 cm 298 2 1x2C1, RodPitchi- 1.849 cm 299 17x 17, RodPitchi- 1.956 cm 300 18x 18, RodPitchi- 1.956 cm 301 19x19, Rod Pitch - 1.956 cm 302 20x20, Rod Pitch - 1.956 cm 303 21x21, Rod Pitch - 1.956 cm 304 16x 16, Rod Pitch -2.15 cm 305 17x17, RodPitch- 2.156cm 306 18x18, RodPitch- 2.156cm 307 19x19, Rod Pitch - 2.56 cm 308 20x20, Rod Pitch - 2.15 cm 309 15x215, Rod Pitch - 2.293 cm 310 16x 16, Rod Pitch - 2.293 cm 311 17xl7, Rod Pitch - 2.293 cm 312 18x18, Rod Pitch - 2.293 cm I

D- 19 REPORT 111-2104790 REPORT HI-2104790 D- 19

File- keff-i I- (ycalc- -+/-Gexp +/- (Yi EALF Case Evaluation Identification 313 19x19, Rod Pitch - 2.293 cm 314 Core XI, 1511 ppm 315 Core XI, 1335.5 ppm 316 Core XI, 1335.5 ppm 317 Core XI, 1182 ppm, 36 Pyrex Rods 318 Core XI, 1182 ppm, 36 Pyrex Rods 319 Core XI, 1032.5 ppm, 72 Pyrex Rods 320 Core XI, 1032.5 ppm, 72 Pyrex Rods 321 Core XI, 794 ppm, 144 Pyrex Rods 322 Core XI, 779 ppm, 144 Pyrex Rods 323 Core XI, 1245 ppm, 72 Vicor Rods 324 Core XI, 1384 ppm, 144 A120 3 Rods 325 Core XI, 1348 ppm, 36 A1203 Rods 326 Core XI, 1348 ppm, 36 A1203 Rods 327 Core XI, 1363 ppm, 72 A1203 Rods 328 Core XI, 1362 ppm, 72 A1203 Rods 329 Core XI, 1158 ppm 330 Core XI, 921 ppm 331 0% Boron Steel plates, dist. 0.245 cm 332 0% Boron Steel plates, dist. 3.277 cm 333 0% Boron Steel plates, dist. 0.428 cm 334 0% Boron Steel plates, dist. 3.277 cm 1.05% Boron Steel plates, dist. 3.277 335 cm - -mmmm-336 1.62% Boron Steel plates, dist. 3.277 cm - -- mlm -

337 Al plates, dist. 0.105 cm 338 Al plates, dist. 3.277 cm 339 Zircaloy-4 plates, dist. 0.078 cm 340 Zircaloy-4 plates, dist. 3.277 cm 341 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 D-20 HI-2 104790 REPORT HI-2104790 D-20

File- F-i 'calcE EALF Case Evaluation Identification k*_i- .4 Gexp - (;i 351 Steel Reflector, 0.660 cm separation 352 Steel Reflector, 1.956 cm separation 353 Lead Reflector, 0 cm separation 354 Core lilA 355 Core IIC 356 Core HID 357 Core IIIE 358 Core IIIF 359 Core IIG 360 Core IV 361 Core V 362 Core VI 363 Core VII 364 Core VIII 0% Boron Steel plate, Gd water 366 impurity 1.1% Boron Steel plate 367 1.6% Boron Steel plate 368 Boral B plate 369 Boral C plate 370 Boroflex, 1.84 cm separation 371 Boroflex, 1.73 cm separation 372 Steel Reflector, 0% Boron Steel plate Steel Reflector, 1.1% Boron Steel 373 plate - -- - I 374 Steel Reflector, Boroflex, 8.37 cm separation - -I

=

375 Borated Water, 490 ppm 376 Unborated Water 377 Borated Water, 1030 ppm 378 0% Boron Steel plates, dist. 0.645 cm 379 0% Boron Steel plates, dist. 2.732 cm 380 0% Boron Steel plates, dist. 4.042 cm 381 0% Boron Steel plates, dist. 0.645 cm 382 0% Boron Steel plates, dist. 4.042 cm 383 0% Boron Steel plates, dist. 0.645 cm 384 0% Boron Steel plates, dist. 4.042 cm 1.05% Boron Steel plates, dist. 0.645 385 cm -m-m--- m 386 1.05% Boron.Steel plates, dist. 4.042 1___ cm J m_____ J________=_____________________

D-2 1 HI-2 104790 REPORT HI-2104790 D-21

Evaluation Identification File- L (

EALF Case name ___ (V 387 1.62% Boron Steel plates, dist. 0.645 cm 388 1.62% Boron Steel plates, dist. 4.042 cm 389 Boral plates, dist. 0.645 cm 390 Boral plates, dist. 4.442 cm 391 Boral plates, dist. 0.645 cm 392 Al plates, dist. 0.645 cm 393 Al plates, dist. 4.042 cm 394 Al plates, dist. 4.442 cm 395 Zircaloy-4 plates, dist. 0.645 cm 396 Zircaloy-4 plates, dist. 4.042 cm 397 Hex, 621 Rods, Temperature 20.1C 398 Hex, 889 Rods, Temperature 231A.4C 399 Hex, 1951 Rods, Temperature 19.3C 400 Hex, 2791 Rods, Temperature 206.OC 401 Hex, 325/680 Rods, Temperature 20.8C 402 Hex, 325/912 Rods, Temperature 212.1C 403 Core XIA - - -

404 Core XIC - - -

405 Core XID - - -

406 Core XIE 407 Core XIF 408 Core XIG 409 Core XIIIA 410 No Boron Steel plates 411 0% Boron Steel plates, 3 mm, dist. 0 412 0% Boron Steel plates, 6 mm, dist. 0 413 0% Boron Steel plates, 6 mm, dist. 0.5 414 0% Boron Steel plates, 6 mm, dist. 1 415 0.67% Boron Steel plates, 3 mm, dist.

0 416 0.67% Boron Steel plates, 6 mm, dist.

0 417 0.67% Boron Steel plates, 3 mm, dist.

0.5 418 0.67% Boron Steel plates, 6 mm, dist.

0.5 419 0.67% Boron Steel plates, 3 mm, dist.

1 420 0.67% Boron Steel plates, 6 mm, dist. _____ ______

REPORT HI-2104790 D-22

Case Evaluation Identification 1

0.98% Boron Steel plates, 3 mm, dist.

421 0 -m- -nm m 422 0.98% Boron Steel0 plates, 6 mm, dist.

423 0.98% Boron Steel plates, 6 mm, dist.

_____ ~~~~0.5 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

424 0.98% Boron Steel plates, 6 mm, dist.

1 - -- mm-425 No Boron Steel plates 426 0% Boron Steel plates, dist. 0 427 0.67% Boron Steel plates, dist. 0 428 0.98% Boron Steel plates, dist. 0 429 No Boron Steel plates 430 0% Boron Steel plates, dist. 0 431 0% Boron Steel plates, dist. 0.5 432 0% Boron Steel plates, dist. 0 433 0% Boron Steel plates, dist. 0.5 434 0.67% Boron Steel plates, dist. 0 435 0.67% Boron Steel plates, dist. 0.5 436 0.67% Boron Steel plates, dist. 0 437 0.67% Boron Steel plates, dist. 0.5 438 0.98% Boron Steel plates, dist. 0 439 0.98% Boron Steel plates, dist. 0.5.

440 0.98% Boron Steel plates, dist. 0 441 0.98% Boron Steel plates, dist. 0.5 442 Otto Hahn, ZrB2 and B 4C rods 443 f'EN/MB-0l1 (580 pins) 444 IPEN/MB-01 (560 pins) 445 670 pins, A120 3-B 4C rods 446 672 pins, A120 3-B4C rods 447 668 pins, A120 3-B4C rods 448 668 pins, A120 3-B4C rods 449 664 pins, 16 steel rods 450 662 pins, 18 steel rods 451 658 pins, 14 steel rods 452 660 pins, 12 steel rods 453 660 pins, 12 steel rods 454 661 pins, 17 steel rods 455 662 pins, 16 steel rods 456 634 pins, 12 steel rods D-23 REPORT HI-2 104790 REPORT HI-2104790 D-23

File- +Gcalc- +/- exp EALF Case Evaluation Identification keff-i +/- cTi (ANA 457 620 pins, 26 steel rods 458 668 pins, 0 steel rods, 4 Gd 20 3 rods 459 648 pins, 0 steel rods, 8 Gd 20 3 rods 460 672 pins, 0 steel rods, 4 Gd 20 3 rods 461 646 pins, 4 steel rods, 4 Gd 20 3 rods 462 656 pins, 4 steel rods, 4 Gd 20 3 rods 463 664 pins, 4 steel rods, 2 Gd 20 3 rods 464 670 pins, 2 steel rods, 2 Gd 20 3 rods 465 664 pins, 2 steel rods, 2 Gd 20 3 rods 466 656 pins, 0 steel rods, 2 Gd 20 3 rods 467 23x23, 1.825 cm pitch 468 23x23, 1.825 cm pitch 469 23x23, 1.825 cm pitch 470 21x21, 1.956 cm pitch 471 21x21, 1.956 cm pitch 472 21x21, 1.956 cm pitch 473 20x20, 2.225 cm pitch 474 20x20, 2.225 cm pitch 475 20x20, 2.225 cm pitch 476 21 x21,2.474 cm pitch 477 21x21, 2.474 cm pitch 478 8 wt% 240Pu 1.05" pitch, Al Rods 479 8 wt% 240Pu 1.05" pitch, B4 Rods 480 8 wt% 240Pu 1.05" pitch, B3 Rods 481 8 wt% 240Pu 1.05" pitch, B2 Rods 482 8 wt% 240Pu 1.05" pitch, B I Rods 8 wt% 240Pu 1.05" pitch, AI+Cd 483 Rods - -- --r -

484 8 wt% 240Pu 1.05" pitch, B4+Cd Rods 485 8 wt% 240Pu 1.05" pitch, B3+Cd Rods 486 8 wt% 240Pu 1.05" pitch, B2+Cd Rods 487 8 wt% 240Pu 1.05" pitch, B I +Cd Rods 488 8 wt%/o 240Pu 1.05" pitch, Air+Cd Rods 489 8 wt% 240Pu 1.05" pitch, H20+Cd Rods 490 8 wt% 240Pu 1.32" pitch, Al Rods _ __ _

491 8 wt% 240Pu 1.32" pitch, B4 Rods _ _ _ _

D-24 HI-2 104790 REPORT HI-2104790 D-24

Case Evaluation Identification File-name kf Irai alc : GeV +/- EALF (eV) 492 8 wt% 240Pu 1.32" pitch, B3 Rods 493 8 wt% 240Pu 1.32" pitch, B2 Rods 494 8 wt% 240Pu 1.32" pitch, BI Rods 495 8 wt% 240Pu 1.32" pitch, AI+Cd Rods _____

496 8 wt% 240Pu 1.32" pitch, B4+Cd Rods 497 8 wt% 240Pu 1.32" pitch, B3+Cd Rods 498 8 wt% 240Pu 1.32" pitch, B2+Cd Rods 499 8 wt% 240Pu 1.32" pitch, B 1 +Cd Rods _______

500 Rods 8 wt% 240Pu 1.32" pitch, Air+Cd 501 Rodspitch, H20+Cd 8 wt% 240Pu 1.32" 502 16 wt% 240Pu 1.386" pitch 503 16 wt% 240Pu 1.05" pitch, Al Rods 504 16 wt% 240Pu 1.05" pitch, B4 Rods 505 16 wt% 240Pu 1.05" pitch, B3 Rods 506 16 wt% 240Pu 1.05" pitch, B2 Rods 507 16wt%240Pu1.05"pitch, B1Rods 508 16 wt% 240Pu 1.05" pitch, AI+Cd Rods _______

509 16 wt% 240Pu 1.05" pitch, B4+Cd Rods 510 16 wt% 240Pu 1.05" pitch, B3+Cd Rods 511 16 wt% 240Pu 1.05" pitch, B2+Cd Rods 512 16 wt% 240Pu 1.05" pitch, BI+Cd Rods _____

513 16 wt% 240Pu 1.05" pitch, Air+Cd Rods 514 16 wt% 240Pu 1.05" pitch, H20+Cd Rods 515 24 wt% 240Pu 1.05" pitch, Al Rods 516 24 wt% 240Pu 1.05" pitch, B4 Rods 517 24 wt% 240Pu 1.05" pitch, B3 Rods 518 24 wt% 240Pu 1.05" pitch, B2 Rods 519 24 wt% 240Pu 1.05" pitch, B1 Rods 520 24 wt% 240Pu 1.05" pitch, AI+Cd Rods _ _ __

521 24 wt% 240Pu 1.05" pitch, B4+Cd _ ___

REPORT HI-2104790 D-25

Case Evaluation Identification Rods 522 24 wt% 240Pu 1.05" pitch, B3+Cd Rods - -- n--m 523 24 wt% 240Pu 1.05" pitch, B2+Cd Rods 524 24 wt% 240Pu 1.05" pitch, Bl+Cd Rods 525 24 wt% 240Pu 1.05" pitch, Air+Cd Rods 526 24 wt% 240Pu 1.05" pitch, H20+Cd Rods - -- -m-527 8 wt% 240Pu 0.55" pitch 528 8 wt% 240Pu 0.60" pitch 529 8 wt% 240Pu 0.71" pitch 530 8 wt% 240Pu 0.80" pitch 531 8 wt% 240Pu 0.90" pitch 532 8 wt% 240Pu 0.93" pitch REPORT HI-2104790 D-26

Table D.3-6 Descriptive Statistics of the MCNP5-1.51 Calculational Results No. of Experiment Description keff range EALF (eV) range exp.

HTC Experiments 156 Selected Experiments 376 All experiments 532 Table D.3-7 Normality Test Results for the MCNP5-1.51 calculations Experiment No. of Shapiro-Wilk Pearson's chi-square (X2)

Description exp. Wtest W n Pd(X2 ;d) Normal HTC Experiments 156 N/A N/A m m Selected Experiments 376 N/A N/A m m All experiments 532 N/A N/A m m Table D.3-8 Trending Analysis Results for the MCNP5-1.51 calculations Correlation Regression Experiment No. of Correlated Probability, Coefficient, Correlation Equation, Description exp. Parameter, x Pd(N;r) 14 k(x)

+ 4-EALF *m All 532 Pitch

  • I experiments Rod OD m Fuel Density D-27 REPORT HI-2104790 REPORT HI-2104790 D-27

Table D.3-9 Analysis of Neutron Absorbers and Reflector Materials for the MCNP5-1.51 calculations No. Normality Residuals Experiment Bias X2 of Bias Linear Correlation Normality, Description Uncertainty exp. (Pd(X2;d)) (Pd(X2 ;d))

All experiments__53 All except 532 -

II - II those with Gadolinium, 365 Cadmium and Lead REPORT HI-2104790 D-28

Table D.3-10 Bias and Bias Uncertainty as a Function of Independent D-29 REPORT 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 exp. Uncertainty X2 (Pd(xE;d)) ________________ (Pd(X2;d))

2 All except those with Gadolinium, 365 mm Cadmium and Leadt Fresh U0 Fuel--

2 207 mIN mm m m_______m -

HT m

+MOX 158 m Experiments D-30 HI-2 104790 REPORT HI-2104790 D-30

Table D.3-12 Bias and Bias Uncertainty as a Function of Independent Parameter Indepen Bias Independ Bias Experiment dent Calculated Bias Uncerta ent Calculated Bias Uncertai Description Paramet kF inty aramete kf nty er, x r, x EALF Pu Enrichment HTC + MOX Experiments I I I I D-3 I HI-2 104790 REPORT 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 those with Gadolinium, 365 M ==*m m&*

Cadmium and Leadt All with Fresh 287 _ _ M m

Water 287 All with 78 Borated 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 Ut as a Function of Parameter EALF U Enrichment mEl M I mEi mEi mEi mEi mEi mEi mEi mEi All with Fresh mEi Water mEl mEl mEi mEl mEi mEl mEl mEl mEl mEi mE-Density mEl mEi mEi All with N/A Borated Water D-33 REPORT HI-2 104790 REPORT HI-2104790 D-33

Indepen Independ Bias Bias Experiment dent Calculated ent Calculated Bias Uncertai Bias Uncerta Description Paramet k*ff Paramete ky inty a1- "'

r, x D-34 REPORT HI-2 104790 REPORT 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 those with Gadolinium, 365 1 1___ 1ma n Cadmium and Leadt Fresh U0 2 Fuel with Fresh Water 154 - __ m Fresh U0 2 Fuel with Borated 53 aI Water HTC + MOXA Fuel with Fresh 133 A Water MIE H-TC + MOXU.

Fuel with 25 1__ __

Borated Water tNote: Critical experiments with Gadolinium, Cadmium and Lead were excluded from all subsequent subsets.

D-35 HI-2104790 REPORT HI-2104790 D-35

Table D.3-16 Bias and Bias Uncertainty as a Function of Independent Parameter Indepen Bias Independ Bias Experiment dent Calculated ent Calculated Bias Uncertai Description Paramet kBfia inct Paramete keft er, x r, xnty EALF mII mEi miI mII mEl mEl mII mEi mII Fresh U0 2 Fuel with N/A Fresh Water m m

m m

m m

m m

m EALF Pu Enrichment miI m

HTC + MOX Fuel with Fresh Water REPORT HI-2104790 D-36

Indepen Independ Bias Bias Experiment dent Calculated ent Calculated Bias Uncerta Bias Uncertai Description Paramet keff Paramete keff inty nty

r. x mEm mEm mEm mEm mEm I m I I Rod OD Density mEl mEl mEl mEl mEl mEl mEl mEl E l mEl mEl mEl HTC + MOX m U mEl Fuel with mEl Borated Water mEl mEl mEl mEl mEl mEl EE==m mEl mEl mEl M==M mEl REPORT HI-2104790 D-37

Table D.3-17 Comparison of Key Parameters and Definition of Validated AOA Parameter Design Benchmarks Validated Application 23 5 239 24 1 235 2 39 24 1 23 5 239 241 Fissionable Material U, Pu, Pu U, Pu, Pu U, Pu, Pu Isotopic Composition 23 5 u/ut < 5.Owt% 1.57 - 5.74% < 5wt%

Pu/(U+Pu) < 20wt% 1.104-20% < 20wt%

Physical Form UO 2,MOX U02,MOX UO 2, MOX Fuel Density (g/cm 3) 10.0 - 10.7 9.2 - 10.4 9.2 - 10.7 Moderator Material (coolant) H H H Physical Form H20 H20 H20 Density (g/cm 3) around 1.0 g/cm 3 around 1.0 g/cm 3 around 1.0 g/cm 3 Reflector Material H H H Physical Form H20 H20 H 20 Density (g/cm 3) around 1.0 g/cm 3 around 1.0 g/cm 3 around 1.0 g/cm 3 Interstitial Reflector Material Plate Steel or Lead Steel or Lead Steel or Lead Absorber Material None, Boron (15 - None, Boron (0 -

Soluble None, Boron or 2550 ppm) or 2550 ppm) or Gadolinium Gadolinium (48 - Gadolinium (48 to 197 ppm) 197 ppm)

Rods Boron Pyrex', Vicor Boron Steel or B-Al Separating Material Water, B-SS, Water, B-SS, Boral, Water, B-SS, Boral, Plate Boral or Boroflex, Zircaloy or Boroflex, Zircaloy or Cadmium Cadmium Cadmium Geometry Lattice type Square Square, Triangle Square, Triangle 1.26-1.47 LatticeLaticePith Pitch (m) (cm) 1.24 (PWR)-1.88 0.968 to 4.318 0.968 to 4.318 (BWR)

Thermal Neutron Energy spectrum Thermal spectrum Thermal spectrum REPORT HI-2104790 D-38

Figure Proprietary Figure D.3-1 Frequency Chart for Calculated keff of the Selected 532 Benchmarks for the MCNP5-1.51 code Figure Proprietary Figure D.3-2 Frequency Chart for Calculated EALF (eV) of the Selected 532 Benchmarks for the MCNP5-1.51 code REPORT HI-2104790 D-39

Figure Proprietary Figure D.3-3 MCNP5-1.51 Calculated keg Values for Various Values of the Spectral Index (All Experiments)

REPORT HI-2104790 D-40

Figure Proprietary Figure D.3-4 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral Index D-4 1 HI-2 104790 REPORT HI-2104790 D-41

Figure Proprietary Figure D.3-5 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral Index REPORT HI-2104790 D-42

Figure Proprietary Figure D.3-6 MCNP5-1.51 Calculated keff Values for Various Values of the Pu Enrichment D-43 REPORT HI-2 REPORT 104790 HI-2104790 D-43

Figure Proprietary Figure D.3-7 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral Index REPORT 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 Density D-46 HI-2104790 REPORT HI-2104790 D-46

Figure Proprietary Figure D.3-10 MCNP5-1.51 Calculated keff Values for Various Values of the Spectral Index D-47 HI-2 104790 REPORT HI-2104790 D-47

Figure Proprietary Figure D.3-11 MCNP5-1.51 Calculated ke, Values for Various Values of the Spectral Index REPORT HI-2104790 D-48

Figure Proprietary Figure D.3-12 MCNP5-1.51 Calculated keff Values for Various Values of the Pu Enrichment D-49 HI-2 104790 REPORT HI-2104790 REPORT D -49

Figure Proprietary Figure D.3-13 MCNP5-1.51 Calculated keff Values for Various Values of the Rod OD D-50 HI-2 104790 REPORT HI-2104790 D-50

Figure Proprietary Figure D.3-14 MCNP5-1.51 Calculated keff Values for Various Values of the Fuel Density D-5 1 HI-2 104790 REPORT HI-2104790 D-51

ATTACHMENT 5 Holtec International Report No. HI-2125245, Revision 4, "Licensing Report for Quad Cities Criticality Analysis for Inserts - Non Proprietary Version"

Eu.' IN Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797- 0900 HOLT EC INTERN AT I ONAL Fax (856) 797 - 0909 Licensing Report for Quad Cities Criticality Analysis for Inserts - Non ProprietaryVersion FOR Exelon Holtec Report No: HI-2125245 Holtec Project No: 2127 Sponsoring Holtec Division: HTS Report Class : SAFETY RELATED

Summary of Revisions:

Revision 0: Original Issue Revision 1: Supplement I was added to cover a new revision of NETCO-SNAP-INO rack insert.

Revision 2: All Revision I revision bars were removed. No other changes were made.

Revision 3: Sections 2.3.8, 2.7, 7.6, 8.0 and Appendix B were revised. All changes were marked by revision bars.

Revision 4: Minor editorial changes to page 4 description of Table 7.1(c) and Table 2.1(c) was move up one line. Neither change marked by revision bar. All Revision 3 revision bars removed.

Project No. 2127 Report No. H1-2125245 Page i

Table of Contents

1. INTRODUCTION ................................................................................................. 10
2. METHODOLOGY ............................................................................................. 11 2.1 GENERAL, APPROACH ....................................................................................... . .......... II 2.2 COMPUTER CODES AND CROSS SECTION LIBRARIES ................................................................. 11 2.2 .1 MC NP5 -1.5 1 ....................................................... ............................................................ 11 2 .2.1 .1 M CN P 5-1.5 1 V alidation ........................................................................................................................................

2.2.1.1.1 12 2 .2 .2 CAS MO -4 ............................. . . ............................................................................. . . 13 2.3 ANALYSIS ME-THODS .................................................................................................................. 13 2.3.1 Design Basis Fuel Assembly ............................................................................................. 13 2 .3,1.1 P eak R eactiv ity ...................................................................................................................................................... 14 2.3,1.1.1 P ak R eactivity and Fuel A ssemb B urnup ............................................................. .................................... 14 2.3.1.1.2....... ...................................................................................... . . 14 2.3.1.2 J- ii i i  ; ....................................... 15 2.3.1.3 Determination of the Design Basis Fuel Assembly Lattice ........................................................................... 16 2.3.1.4 Optima2 CASMO-4 Model Simplification Effect .......................................................................................... 16 2 .3 .1.5 C ore O perating Param eters ................................................................................................................................... 18 2.3.1.5.1 R eacto r P oweerU prate ...................................................................................................................................... 18 2.3.1.5.2 Integral R eactivity C ontrol Devices ........................................................................................... ................... 19 2.3.1.5.3 A xial and Planar Enrichm ent V ariations .................................................................................................... 19 2.3.1.5.4 Fuel A ssem bly D c-Channeling .............................................................................................................. 19 2 .3 .1 .6 ..................... ......................................................................... 20 2.3.2 Reactivity Effect of Spent Fuel Pool Water Temperature................................................ 20 2.3.3 Fuel Depletion CalculationUncertainy......................................................................... 21 2.3.4 Fuel and Storage Rack Manufacturing Tolerances ......................................................... 22 2.3.4.1 Fuel M anufacturing Tolerances ............................................................................................................................ 22 2.3.4.2 SEP Storage Rack Manufacturing Tolerances ............................................................................................... 23 2.3.5 Radial Positioning........................................................................................................... 23 2.3.5.1 Fuel Assembly Orientation in the Core ........................................................................................................... 23 2.3.5.2 Fuel R adial Positioning in the R ack ...................................................................................................................... 23 2.3.5 .3 Inserts R adial Positioning ...................................................................................................................................... 25 2.3.5.4 Fuel Orientation in SFP Rack Cell .............................................................................................................. 25 2.3.6 Insert. Co po. Measrement... rtain.. ............ . ..... I ................ ........ 26 2.3.6.18 .... .

.. *........... . ....................... ......................................... . .................... 26 2.3.64 2G ..

MA E . .. ........................................................................................... 27 2.3.6 .2....... .. . .......... ..... ................................................ 27 2.3.7 Insert Coupon Measurement Uncertainty............................................. ...... 27..

2.3.8 Maximum kff CalculationforsiyEf Normal

...... Conditions..............................................

cs............................................. 27

2. .1 Te peatr ad atrD 2.4 2 ..6.2 MARGIN D o EVALUATIONstortion.........................................................................................................

p d ss b y - Ho i o t l........................I......................................... ........................................ 28 27 2.5 FUEL MOVEMENT, INSPECTION AND RECONSTITUTION OPERATIONS .............................. ............

2.6 ACCID NT CONDITION .......... ................................................................................................... 29 2.6.1 Temperature and WaFer Density Effects ....................................... ............... 30 2.6.2 Droppeda As Fem- Horizontal sbly ........................................................................................ 30 2.6.3 DroppedAssembly - Vertical into a Storage Cell.............................................. 30 2.6.4 Storage Cell Distortion........................................................................... 31 2.6.5 Misloaded Fuel Assembly/Missing Insert....................................................... 31 2.6.6 Mislocated Fuel Assembly ....................................................................... 32 2.6.6.1 Mislocation of a Fuel Assembly in the Water (jai) between the Racks and Pool Wall ........................... 32 2.6.6.2 Mislocation of a Fuel Assembly in the Comner between Two Racks................................................ 32 2.6.6.3 Mislocation of a Fuel Assembly between the SFP Rack and the Inspection Platform ................................ 32 2.6.7 Mis-installment of an Insert on Wrong Side of a Cell..................................................... 33 Project No. 2127 Report No. HI-2125245 Page 1

2.6.8 Insert Mechanical Wear ..................................................................................................... 33 2.6.9 Rack Movement ................................................................................................................... 33 2.7 .. ..................................................................... ............................................... 33 2.8 SPENT FUEL, RACK INTERFACES .............................................................................................. 34 2.9 RECONSTITtr TED FuEl, ASSEMBLIES ....................................................................................... 35

3. ACCEPTANCE CRITERIA ........................................................................................................... 36 3.1 AIPPICABLE CODES, STANDARDS AND GUIDANCE'S ............................................................ 36
4. ASSUM PTIONS .............................................................................................................................. 37
5. INPUT DATA .................................................................................................................................. 38 5.1 FUEL.. ASSEMBLY SPECIFICATION ............................................................................................. 38 5.2 REACTOR PARAMEIERS ..... ........................ .............................. 38 5.3 SPEN'T FUEL POOL PARAMETERS ............................................................................................ 38 5.4 STORAGE RACK SPECIFICATION .............................................................................................. 39 5.4.1 MaterialCompositions................................................................................................... 39
6. CO M PUTER CO DES ......................................................................................................... .. 40
7. ANALYSIS ....................................................................................................................................... 41 7.1 DESIGN BASIS AND UNCERTAINTY EVALUATIONS ................................................................. 41 7.1.1 ......... .. 41 7.1.2 Determination of the Design Basis Fuel Assembly Lattice .............................................. 41 7.1.2.1 Fuel Assembly De-Channeling ............................................................................................................................. 41 7.1.3 Optima2 CASM O-4 Model Simplification Efl&ct ............................................................ 41 7.1.4 Core OperatingParameters............................................................................................ 42 7.1.4.1 Reactor Pow er Uprate ............................................................................................................................................ 42 7.1.5 Water Temperature and Density Effect ............................................................................ 42 7.1.6 Depletion Uncertainty...................................................................................................... 42 7.1.7 Fuel and Rack Manzjfacturing Tolerances ..................................................................... 43 7.1.7.1 Fuel Assembly Tolerances .................................................................................................................................... 43 7.1.7.2 SFP Rack Tolerances ............................................................................................................................................ 43 7.1.8 Radial Positioning........................................................................................................... 43 7,1.8.1 Fuel Assembly Radial Positioning in SFP Rack ............................................................................................ 43 7.1.8.2 Fuel Orientation in SFP Rack ................................................................................................................................ 43 7.1.9 Fu.. ............. ............... ................................................ 44 7.1.9 .1 ......................................................................................................... . . . 44 7.1.10 ; 44 7.2 MAXIMUM KFF CALCULATIONS FOR NORMAL CONDITIONS .................................................. 44 7.3 M ARGIN EVALUATION ................................................................................................................ 44 7.4 ABNORMAL AND ACCIDENT CONDITIONS .............................................................................. 45 7.5 MAXIMUM K3:F,: CALCUi ATIONS FOR ABNORMAL AND ACCIDENT CONDITIONS ...................... 45 7.6. ....................................................................................................................... 45 7.7 SPENT FUEL RACK INTERFACES ............................................................................................. 45
8. CONCLUSION ................................................................................................................................ 47
9. REFERENCES ................................................................................................................................ 48 Project No. 2127 Report No. HI-2125245 Page 2

Supplement 1: Additional Calculations to Support the Revised NETCO-SNAP-IN Rack Insert Design .............................................................................................. . . .. Si-I ProJect No. 2127 Report No. F11-2125245 Page 3

List of Tables Table Description Page 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 Results of the MCNP5-1.51 Calculations for SVEA-96 Optima2 Q122 Table 7.2(a) 68 Lattices Table 7.2(b) Results of the MCNP5-1.51 Calculations for GEI 4 Lattice Type 5 70 Results of the MCNP5-1.51 Calculations for Design Basis and Table 7.3 71 Simplified Model of SVEA-96 Optima2 Q122 Lattice Type 146 Table 7.4 Results of the MCNP5-1.51 Calculations for Core Operating Parameters 72 Results of the MCNP5-1.51 Calculations for the Effect of Water Table 7.5 73 Temperature and Density Table 7.6(a) Results of the MCNP5-1.51 Calculations for the Depletion Uncertainty 74 75 Table 7.7 Results of the MCNP5-1.51 Calculations for Fuel Tolerances 76 Table 7.8 Results of the MCNP5-1.51 Calculations for Rack Tolerances 77 ProJect No. 2127 Report No. 1-11-2125245 Page 4

Table Description Page Results of the MCNP5-1.51 Calculations for Fuel Radial Positioning in Table 7.9(a) 78 SFP Racks Results of the MCNP5-1.51 Calculations for Fuel Orientation in SFP Table 7.9(b) RAAk-- 79 80 81 Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate Table 7.12(a) the Effect of Nominal Values Instead of Using Minimum B4C Loading 82 and Minimum Insert Thickness on Reactivity Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate 83 Table 7.12(b) the Effect of the Actual Optima2 Q122 Fuel Assembly Table 7.12(c) Margin Evaluation Summary of the Margin Evaluation 84 85 Results of the MCNP5-1.51 Calculations for the Empty Storage Rack 86 Table 7.13(b)

Cell without Insert 87 88 Results of the MCNP5-i.51 Calculations for Axially Infinite Optima2 89 Table 7.16 Q122 Lattices Table 7.17 Results of the MCNP5-1.51 Calculations for SFR Interface 90 l

m m

m m

m m

m Project No. 2127 Report No. HI-2125245 Page 5

Table Descrintion Page U

Table SI-I Fuel Rack Insert Revised Dimensions S1-5 Table S 1-2 Results of the MCNP5 Calculations for Revised Rack Tolerances SI-6 Results of the MCNP5-1.51 Calculations for Revised Fuel Radial Table S1-3 SI-7 Positioning in SFP Racks Results of the MCNP5-1.51 Calculations for Revised Fuel Orientation Table S1-4 S1-8 in SFP Racks S1-9 SIl-10 Project No. 2127 Report No. 1-11-2125245 Page 6

List of Figures Figure Descrintion Page 92 93 94 95 96 97 98 99 100 101 102 103 104 Project No. 2127 Report No. 1H1-2125245 Page 7

Figure Description Page 105 106 107 108 109 110 III 112 113 114 115 116 117 118

....-24,,*

B*.6 Project No. 2127 Report No. 1-11-2125245 Page 8

Figure Description Page SI-I1 Project No. 2127 Report No. HI-2125245 Page 9

1. INTRODUCTION This report documents the criticality safety evaluation for the storage of spent BWR fuel in the Unit I and Unit 2 spent fuel pools (SFPs) at Quad Cities Station operated by Exelon. The Unit I and Unit 2 SFP racks are identical and are designed to accommodate BWR fuel. Currently, the SFPI racks credit BORAFLEX for reactivity control. This new analysis will not credit the BORAFLEX but will instead credit new NETCO-SNAP-IN rack inserts, which are new to Quad Cities but not new relative to their use for spent fuel pool reactivity control. This analysis will demonstrate that with credit for the inserts the effective neutron multiplication factor (kerf) in the SFP racks fully loaded with fuel of the highest anticipated reactivity, at a temperature corresponding to the highest reactivity, is less than 0.95 with a 95% probability at a 95%

confidence level. Reactivity effects of abnormal and accident conditions are also evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit.

Criticality control in the SFP, as credited in this analysis, relies on the following:

  • Fixed neutron absorbers o NETCO-SNAP-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. 2127 Report No. 1.11-2125245 Page 10

2. METHODOLOGY 2.1 GeneralApproach The analysis is performed consistent with regulatory requirements and guidance. The calculations are performed using either the worst case bounding approach or the statistical analysis approach with respect to the various calculation parameters. The approach considered for each parameter is discussed below.

2.2 Computer Codes and Cross Section Libraries 2.2.1 MCNP5-1.51 MCNP5-1.51 is a three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory [1]. MCNP5-1.51 calculations use continuous energy cross-section data based on ENDF/B-VII. MCNP is selected because it has history of successful use in fuel storage criticality analyses and has most of the necessary features (except for fuel depletion analysis) for the analysis to be performed for Quad Cities Station SFP.

The convergence of a Monte Carlo criticality problem is sensitive to the following parameters:

(1) number of histories per cycle, (2) the number of cycles skipped before averaging, (3) the total number of cycles and (4) the initial source distribution. All MCNP5 calculations are performed with a minimum of 12,000 histories per cycle, a minimum of 150 skipped cycles before averaging, and a minimum of 150 cycles that are accumulated, The initial source is specified as nniform cnver the fineled re.oinnq (t-enmhIie0 I 2.2.1.1 MCNP5-1.51 Validation ProJect No. 2127 Report No. 1-I1-2125245 Page I11

Project No. 2127 Report No. HI-2125245 Page 12 U

2.2.2 CASMO-4 Fuel depletion analyses during core operation are performed with CASMO-4 Version 2.05.14 (using the 70-group cross-section library), which has been approved by the NRC for reactor analysis (depletion) when providing reactivity data for specific 3D simulator codes. CASMO-4 is a two-dimensional multigroup transport theory code based on the Method of Characteristics and it is developed by Studsvik of Sweden [4]. CASMO-4 is used to perform depletion calculations and to perform various sensitivity studies. The uncertainty on the isotopic composition of the fuel (i.e., the number density) is considered as discussed below (see Section 2.3.3). A validation for CASMO-4 to develop a bias and bias uncertainty is not necessary because the results of the CASMO-4 sensitivity studies are not used as input into the kcff calculations. However, the code authors 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 number of isotopes that may be extracted for specific pins. Therefore, two independent CASMO-4 depletion calculations were performed to separately extract the actinides and fission products.

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

2.3 Analysis Methods 2.3.1 Design Basis Fuel Assembly There are various fuel designs stored in the Quad Cities SFP. For the purpose of this analysis, the reactivity of each design is evaluated and the most reactive fuel bundle lattice is determined for use as the design basis fuel assembly to determine keff at the 95/95 level. This approach follows the guidance in [2] and [6], and is further described below.

Project No. 2127 Report No. HI-2125245 Page 13

2.3.1.1 Peak Reactivity The BWR fuel designs used at the Quad Cities Station use Gd as an integral burnable absorber.

Initially, the Gd in the fuel assembly holds down the fresh fuel assembly reactivity and then, as core depletion occurs, the Gd begins to burnout until it is essentially fully depleted. As the Gd depletes the reactivity of the fuel assembly increases until it reaches a peak. This peak reactivity is the fuel assembly's most reactive condition. Note that most BWR fuel designs are composed of various axial lattices (including blankets) that can have different axial lengths, uranium loadings (also mixed oxide loading, for MOX fuel), fuel pin arrangements including partial or part-length rods, Gd pin locations and loading, etc. These various lattice components can all effect at what burnup the peak reactivity occurs and the magnitude of the peak reactivity.ti The k r a ty gle fu el asse bl ca th r f e a l h ve d f r nt p Max ial lattices w ith in a sin 2.3.1.1.1 Peak Reactivity and Fuel Assembly Burnup Typically, a spent fuel assembly is characterized by its assembly average burnup (over all lattices or nodes). In this analysis methodology the fuel assembly average burnup is of no concern and is not credited for reactivity control. Rather, the methodology credits the residual Gd and other depletion isotopic compositions at the fuel assembly peak reactivity (most reactive lattice peak reactivity). While the peak reactivity occurs at some specific lattice burnup, the peak reactivity lattice burnup varies from lattice to lattice within a fuel design. Therefore, independent calculations with MCNP5-1.51 using pin specific compositions (see Section 2.3.1.1.2) are performed for every lattice of the SVEA-96 Optima2 fuel assembly (as will be seen in Section 7, this is the fuel assembly with the design basis lattice) over a burnup range to determine the burnup at peak reactivity for every lattice. Since each lattice is considered at its peak reactivity (and therefore the lattice or nodal burnup at which that occurs), the fuel assembly average burnup or fuel assembly burnup profile is not applicable because the analysis already considers each lattice at its most reactive composition, independent of the fuel assembly average burnup.

2.3.1.1.2 Project No. 2127 Report No. HI-2125245 Page 14

2.3.1.2 ProJect, No. 2127 Report No. I-11-2125245 Page 15

2.3.1.3 Determination of the Design Basis Fuel Assembly Lattice 2.3.1.4 Optima2 CASMO-4 Model Simplification Effect As previously discussed in Section 2.3.1.2, various fuel designs were provided. Of these fuel designs, the SVEA-96 Optima2 designs were specified to be bounding. The Optima2 model in CASMO-4 is described as the SVEA-96 model provided in the CASMO-4 manual [4]. This CASMO-4 internal model is slightly different from the actual fuel assembly geometry.

Therefore, it is important to evaluate and if necessary quantify the reactivity effect of the CASMO-4 model simplifications inherent in the code. The CASMO-4 model geometry of the SVEA-96 Optimna2 fuel differs from the SVEA-96 Optirna2 fuel as follows:

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With respect to the fuel assembly geometry models, the amount of zirconium (and therefore the amount of water) in the CASMO-4 model of the SVEA-96 Optima2 fuel is reasonably similar to that of the actual SVEA-96 Optima2 fuel and therefore these built-in CASMO-4 simplifications are acceptable. However, to evaluate the CASMO-4 model geometry simplification effect on reactivity, an applicable set of code-to-code comparisons is performed. The following cases are evaluated.

For the purpose of showing that the two codes calculate an equivalent reactivity the following comparisons are made:

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" Case 2.3.1.4.1 is compared to Case 2.3.1.4.2 at 0 GWD/MTU to show that the two codes calculate similar results with respect to the fuel assembly and storage rack geometry.

" Case 2.3.1.4.1 is compared to Case 2.3.1.4.2 at peak reactivity burnup to quantify the reactivity difference due to the effect of the spent fuel. The two codes use different cross section library versions and calculation sequences. The main calculation sequence difference between the two codes is that CASMO-4 uses a thermal expansion of spent fuel pellet which effects the fuel density [4]. The actual density is conservatively used in MCNP5-1.51. The results are expected to show that the MCNP5-1.51 code is conservative with respect to the CASMO-4 code. Any non-conservative result would be treated as a bias.

" Case 2.3.1.4.3 is compared to Case 2.3.1.4.2 to show the reactivity difference between the simplified MCNP5-1.51 model and the design basis model that is slightly modified to be similar to the CASMO-4 insert orientation. This case is expected to show that the design basis model with respect to the fuel pin pitch (and subsequent sub-bundle pitch) is conservative. This is expected to be conservative because the design basis model fuel compositions are taken from the average fuel pin pitch CASMO-4 calculations and used in the MCNP5-1.51 design basis actual fuel pin locations. Any non-conservative result would be treated as a bias.

Case 2.3.1.4.3 is compared to the result of the actual design basis results (similar to Case 2.3.1.4.3 but with the bounding insert orientation) to show that the design basis model is conservative.

2.3.1.5 Core Operating Parameters As previously discussed, CASMO-4 is used to perform depletion calculations to determine the spent fuel isotopic composition. The operating parameters for spent fuel depletion calculations are discussed in this Section. The operating parameters which may have a significant impact onl BWR spent fuel isotopic composition are void fraction, control blade history, moderator temperature, fuel temperature, and power density. Other parameters such as axial enrichment distribution and effect of burnable absorbers are discussed in Section 2.3.1.5.3 and Section 2.3.1.5.2, respectively. Sensitivity studies are performed to show the effect of each individual parameter, and to confirm that the selected values are in fact appropriate when combined at their worst case.

2.3.1.5.1 Reactor Power Uprate To determine the effect of the power uprate on the reactivity of assemblies in the SFP racks, the following evaluations are performed.

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2.3.1.5.2 Integral Reactivity Control Devices The only type of burnable absorber used for the fuel assemblies covered in this analysis is Gd.

The use of Gd does not increase the reactivity of the assembly, compared to an assembly lattice where all rods contain fuel and no Gd. As discussed in Section 2.3.1.1, the Gd in the fuel assembly holds down the fresh fuel assembly reactivity and then, as core depletion occurs, the Gd begins to burnout until it is essentially fully depleted. As the Gd depletes the reactivity of the fuel assembly increases until it reaches a peak. This peak reactivity is the fuel assembly's most reactive condition, which is used for design basis condition. Note that integrated absorbers do not change the amount of water in the assembly, which is a large part of the effect of non-integral absorbers.

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

However, GE14 (the most second reactive fuel assembly, as will be shown in Section 7) may be de-channeled. Studies are performed to evaluate the effect of storage of GE14 without the Zr channel at various radial positioning in the storage cells. The following cases are evaluated.

" Case 2.3.1.5.4.1: This is the reference for Case 2.3.1.5.4.2 through Case 2.3.1.5.4.4. The MCNP5-1.51 model used herein is a 2x2 array with the cell centered fuel assembly that includes the Zr channel, as shown in Figure 2.13(a).

" Case 2.3.1.5.4.2: The MCNP5-1.51 is a 2x2 array of GEl4 fuel assembly lattice 5 (the most reactive lattice of GEI4, as will be shown in Section 7). The Zr channel is removed, as shown in Figure 2.13(b). The fuel 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 fuel assemblies are eccentric toward the center, as shown in Figure 2.13(c).

" Case 2.3.1.5.4.4: The MCNP5-1.51 is the same as that of Case 2.3.1.5.4.2, except the fuel assemblies are eccentric away from the corner where the insert wings connect, as shown in Figure 2.13(d).

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2.3.1.6 2.3.2 Reactivity Effect of Spent Fuel Pool Water Temperature The Quad Cities Station SFP has a normal pool water temperature operating range below 150 'F.

For the nominal condition, the criticality analyses are to be performed at the most reactive temperature and density [2]. Also, there are temperature-dependent cross section effects in MCNP5-1.51 that need to be considered. In general, both density and cross section effects may not have the same reactivity effect for all storage rack scenarios, since configurations with strong neutron absorbers typically show a higher reactivity at lower water temperature, while configurations without such neutron absorbers typically show a higher reactivity at a higher water temperature. For the SFP racks which credit inserts, the most reactive SFP water temperature and density is expected to be at 39.2 TF and I g/cc, respectively.

The standard cross section temperature in MCNP5-1.51 is 293.6 K. Cross sections are also available at other temperatures; however, not usually at the desired temperature for SFP criticality analysis. MCNP5-1.51 has the ability to automatically adjust the cross sections to the specified temperature when using the TMP card. Furthermore, MCNP5-1.51 has the ability to make a molecular energy adjustment for select materials (such as water) by using the S(a,p) card.

The S(a,13) card is provided for certain fixed temperatures which are not always applicable to SFP criticality analysis. Rather, there are limited temperature options, i.e., 293.6 K and 350 K, etc. Additionally, MCNP5-1.51 does not have the ability to adjust the S(C,3) card for temperatures as it does for the TMP card discussed above. Therefore, additional studies are performed to show the impact of the S(a,3) card at the two available temperatures.

To determine the water temperature and density which result in the maximum reactivity, MCNP5-1.51 calculations are run using the bounding values. Additionally, S(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 cell MCNP5-1.51 SFP model (with periodic boundary conditions through the centerline of the surrounding water 2):

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" Case 2.3.2.1 (reference case): Temperature of 39.2 IF (277.15 K) and a density of 1.0 g/cc are used to determine the reactivity at the low end of the temperature range. The S(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 0.98026 g/cc are used to determine the reactivity at the high end of the temperature range. The S(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. The S(@,p) card corresponds to a temperature of 170.33 IF (350 K).

The bounding water temperature and density (the temperature and its corresponding density which result in the maximum reactivity) of the above cases are applied to all further calculations so that the most reactive water temperature and density is considered. Note that the evaluations 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 in The depletion uncertainty is applied by multiplying it with the reactivity difference (at 95%/95%) between the MCNP5-1.51 calculation with spent fuel at peak reactivity (includes residual Gd) and a corresponding MCNP5-1.51 calculation with fresh fuel (without Gd2 O3).

Calculations are performed for the single cell model of design basis fuel assembly.

The uncertainty is determined by the following:

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

  • 0.05 with keac- 1 = kcalc with spent fuel kw*.2 = kal. with fresh fuel Gcal-l =Standard deviation of kcalc*i 0

~ceal-2= Standard deviation of kcale-2

'The result of the MCNP5-1.51 calculation for the fuel depletion calculation uncertainty is statistically combined with other uncertainties to determine keff.

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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 are combined statistically [2]. The evaluations use the same MCNP5-1.51 models used in the design basis calculation.

2.3.4.1 Fuel Manufacturing Tolerances The BWR fuel tolerances for Optima2 Q122 fuel (which is the most reactive fuel design evaluated herein) are presented in Table 5.1(a). Fuel tolerance calculations are petformed using the design basis fuel assembly lattice, and therefore only the tolerances applicable to that lattice are applicable. Separate CASMO-4 depletion calculations are performed for each fuel tolerance and the full value of the tolerance is applied for each case in both the depletion and in rack calculations. Pin specific compositions are used. The MCNP5-1.51 tolerance calculation is compared to the MCNP5-1.51 reference case (nominal parameter values) at the 95% probability at a 95% confidence level using the following equation:

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

  • q (a1 2 + a22)

The following fuel tolerances are considered in this analysis:

" Fuel enrichment

" Gd loading

  • Fuel pellet density (U0 2 and U0 2+Gd2O 3 fuel rods)
  • Fuel pellet outer diameter (OD)

" Fuel cladding inner diameter (ID)

" Fuel cladding OD

" Fuel pin pitch

  • Fuel sub-bundle pitch 3

" Combination of 4 o Water wing canal inner width o Channel outer square width o Channel comer inner radius o Central water canal inner square width

  • Combination of 4 o channel wall thickness 3 For fuel sub-bundle pitch uncertainty calculation, the fuel hardware (channel, central water channel and water wings) is fixed. The fuel lattices are moved only.

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

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o Water cross wall thickness The maximum positive reactivity effect of the MCNP5-1.51 calculations for each tolerance is statistically combined with the other tolerance results, and this result is then statistically combined with other uncertainties when determining the 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 the tolerance is applied for each case. The MCNP5-1.51 tolerance calculation is compared to the MCNIP5-1.51 reference case with a 95% probability at a 95% confidence level using the following 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 pitch o Cell wall thickness

  • Rack inserts (poison) o Width The maximum positive reactivity effect of the MCNP5-1.51 calculations for each tolerance is statistically combined with the other tolerance results, and this result is then statistically combined with other uncertainties when determining the keff value.

The evaluations use the same MCNP5-1.51 models used in the design basis calculation. The isotopic compositions of the fuel rods are the same as those of the design basis fuel assembly.

The poison thickness and loading are used at their minimum values; i.e., they are treated as a bias instead of uncertainty, for conservatism and simplification.

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

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

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" Case 2.3.5.2.1: This is the reference for Case 2.3.5.2.2 through Case 2.3.5.2.5. The MCNP5-1.51 model used herein is a 2x2 array which is the same as the primary single bundle MCNP5-1.51 model used elsewhere in this analysis. In both models the fuel is centered in the rack cell. See Figure 2.7(a).

" Case 2.3.5.2.2: Every fuel assembly is positioned toward the center, for the 2x2 array, as shown in Figure 2.7(b).

" Case 2.3.5.2.3: Every fuel assembly is positioned toward the corner where the insert wings connect, for the 2x2 array, as shown in Figure 2.7(c).

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

" Case 2.3.5.2.5: Every fuel assembly is centered between insert and cell walls, for the 2x2 array, as shown in Figure 2.7(e).

  • Case 2.3.5.2.6: This is the reference for Case 2.3.5.2.7 through Case 2.3.5.2.10. The MCNP5-1.51 model used herein is an 8x8 array which is the same as the primary single bundle MCNP5-1.51 model used elsewhere in this analysis. In both models the fuel is centered in the rack cell.
  • Case 2.3.5.2.7: Every fuel assembly is positioned toward the center, for the 8x8 array, as shown in Figure 2.8.
  • Case 2.3.5.2.8: Every fuel assembly is positioned toward the corner where the insert wings connect, for the 8x8 array.

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

" Case 2.3.5.2.10: Every fuel assembly is centered between insert and cell walls, for the 8x8 array.

  • Case 2.3.5.2.11: This is the reference for Case 2.3.5.2.12. The MCNP5-1.51 model used herein is a single rack cell where the fuel is centered.

" Case 2.3.5.2.12: The fuel assembly is centered between insert and cell walls, for the single rack cell.

The maximum positive reactivity effect of the MCNP5-1.51 calculations for the fuel radial positioning is added as the bias and the corresponding 95/95 uncertainty is statistically combined with other uncertainties to determine kr.f Project No. 2127 Report No. HI-2 125245 Page 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 isotopic compositions of the fuel rods are the same as those of the design basis fuel assembly.

2.3.5.3 Inserts Radial Positioning Since the insert width and SFR cell inner diameter are comparable, and each insert is installed into the rack cell such that the insert becomes an integral part of the fuel rack, no uncertainty in the positioning for inserts is evaluated. The water gap between rack wall and insert is not assumed, since it may provide a small flux trap effect. Nevertheless, the orientation of fuel assembly with respect to position of insert is considered in Section 2.3.5.4.

2.3.5.4 Fuel Orientation in SFP Rack Cell As described in Section 5.1, fuel assemblies have various radial fuel enrichments and gadolinium distribution. Also, one corner of each fuel assembly is adjacent to the control blade during the depletion in the core. As a result, the fuel depletion is not uniform (more discussion is provided in Section 2.3.1.1.2) and one fuel assembly corner may be more reactive than other corners and therefore the fuel assembly orientation in the SFP storage cell may have an impact on reactivity.

Five cases are analyzed to assess the fuel assembly orientation variations and to determine the most limiting fuel orientation in SFP rack cell with respect to the insert.

The MCNP5-1.51 model of the reference case is the design basis fuel in the 2x2 array, as shown in Figure 2.9(a). The MCNP5.1.51 models of the other four cases are the same as that of the reference case, except with different orientation of fuel assemblies with respect to the inserts.

Figure 2.9(b) through Figure 2.9(e) show the configurations of the fuel assemblies in the SFP cells for the evaluated cases.

Note that the evaluations use the same MCNP5-1.51 models with periodic boundary conditions used in the design basis calculation. The isotopic compositions of the fuel rods are the same as those of the design basis fuel assembly.

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2.3.6.1 2.3.6.1.1 Prqject No. 2127 Report No. H1-2125245 Page 26

2.3.6.1.2 -

2.3.6.2 2.3.7 Insert Coupon Measurement Uncertainty There is a measurement uncertainty associated with the B-! 0 content in the poison test coupons. In this analysis, the minimum B-10 loading and the minimum insert thickness are conservatively used for criticality calculations. Therefore, the coupon measurement uncertainty is not evaluated further in the analysis.

2.3.8 Maximum klf Calculation for Normal Conditions The calculation of the maximum kff of the SFP storage racks fully loaded with design basis fuel assemblies at their maximum reactivity is determined by adding all uncertainties and biases to the calculated reactivity. Note that the insert thickness and its B-I 0 loading are taken at their worst case values.

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kefr is determined by the following equation:

keff = k¢1e + uncertainty + bias where uncertainty includes:

and the bias includes Note that each uncertainty is statistically combined with other uncertainties, while biases are added 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 B4 C 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 and Case 2.4.3.

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  • 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 Optima2 Q122 fuel lattice in the appropriate axial position. However, the top and bottom blankets were conservatively replaced by adjacent fuel lattices. The peak reactivity burnup for each individual 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 peak reactivity). Therefore, the model represents a conservative maximum but unrealistic reactivity of the actual Optima2 fuel assembly.

The differences between the reactivity of Cases 2.4.2 and 2.4.3 and the reactivity of reference Case 2.4.1 provide a quantified margin.

Note that the evaluations use the same MCNP5-1.51 models used in the design basis calculation.

The isotopic compositions of the fuel rods of Case 2.4.1 and Case 2.4.2 are the same as those of the design basis fuel assembly.

2.5 Fuel Movement, 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 movement Project No. 2127 Report 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." This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions. 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 design basis cas.

2.6.1 Temperature and Water Density Effects The SFP water temperature accident conditions for consideration are the increase in SFP water temperature above the maximum SFP operating temperature of 150 'F. The decrease in temperature was already considered 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 is evaluated:

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

The evaluation use the same MCNP5-1.51 model used in the design basis calculation.

Note that as discussed in Section 2.3.2, SFP storage racks with strong neutron absorbers, such as inserts, show a higher reactivity at a lower water temperature. The case evaluated above is performed to confirm this statement.

2.6.2 Dropped Assembly - Horizontal For the case in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a separation distance more than 12 inches.

Also, the length of the inserts (as indicated in Table 5.3(b)) covers this separation distance. Thus, the horizontally dropped assembly is decoupled from the fuel assemblies in the rack. This accident is also bounded by the mislocated case, where the mislocated assembly is closer to the assembly in the racks. Therefore, the horizontally dropped fuel assembly is not evaluated further in the report.

2.6.3 Dropped Assembly - Vertical into a Storage Cell It is also possible to vertically drop an assembly into a location that might be occupied by another assembly or that might be empty. Such a vertical impact would at most cause a small compression of the stored assembly, if present, or result in a small deformation of the baseplate for an empty cell.

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These deformations could potentially increase reactivity. However, the reactivity increase would be small compared to the reactivity increase created by the 'misloaded fuel assembly/missing insert' accident (discussed in Section 2.6.5) that does not include the insert in one rack cell. The vertical drop is therefore bounded by this misload accident and no separate calculation is performed for this drop accident.

2.6.4 Storage Cell Distortion A storage cell distortion or altered geometry as a result of fuel handling equipment uplift forces is possible. However, the reactivity increase would be small compared to the possible reactivity increase created by the 'misloaded fuel assembly/missing insert" accident that does not include the insert in one rack cell, as discussed in Section 2.6.5. The storage cell distortion is therefore bounded by the 'misloaded fuel assembly/missing insert' accident and no separate calculation is performed for the storage cell distortion accident.

As a result of significant distortion, the storage cell for whatever reason may not be able to contain the insert and also it will be therefore unacceptable for storage of a fuel assembly. This condition is bounded by the 'misloaded fuel assembly/missing insert' accident. However to show that it is acceptable for normal operation and that the empty storage cell decreases the reactivity of the SFR, the model with an empty storage cell, i.e. without a fuel assembly and insert, in the center of a 8x8 array, is evaluated. Two cases with a cell centered and eccentric position of the fuel assemblies are analyzed.

2.6.5 Misloaded Fuel Assembly/Missing Insert The fuel storage racks are qualified for storage of fuel assembly with the highest anticipated reactivity; thus it is not possible to misload a fuel assembly if every cell with a fuel assembly has an insert.

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 fuel assembly is misloaded into a cell with a missing insert. To evaluate the effect, the following cases are evaluated:

" Case 2.6.5.1: The MCNP5-1.51 model includes an 8x8 array. One cell near the center of the rack does not have the insert. The misloaded fuel assembly is the design basis fuel assembly.

This fuel assembly is eccentric toward the walls that are not covered by inserts. Other fuel assemblies are also eccentric toward the misloaded fuel assembly. The periodic boundary conditions are used through the centerline of the surrounding water (BORAFLEX replacement). The temperature of the model is set to the minimum (39.2 TF) with its corresponding water density and S(a,3) card. These temperature and density are bounding for the SFP racks. See Figure 2.10(a).

  • Case 2.6.5.2: The MCNP5-1.51 model is the same as Case 2.6.5.1, except with all fuel assemblies centered in the rack cells. See Figure 2.10(b).

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2.6.6 Mislocated Fuel Assembly The Quad Cities SFP layout was reviewed to determine the possible worst case locations for a mislocated fuel assembly. Three hypothetical locations where a fuel assembly may be mislocated are:

" In the water gap between the racks and the pool wall

  • In the corner between two racks

" Between the SFP rack and the inspection platform.

The three cited scenarios are evaluated, as follows.

2.6.6.1 Mislocation of a Fuel Assembly in the Water Gap between the Racks and Pool Wall A fuel assembly may be mislocated in the water gap between the racks and the pool wall. Due to the neutron leakage to the outside the storage rack area, the effect of this mislocation is bounded by that of'mislocation of a fuel assembly between the SFP rack and the inspection platform' accident, as discussed in Section 2.6.6.3. No separate calculation is performed for this accident.

2.6.6.2 Mislocation of a Fuel Assembly in the Corner between Two Racks There are some places in the SFP, but outside of the racks, where the mislocated fuel assembly may be in the corner between two racks (thus the mislocated fuel assembly would be adjacent to the fuel assemblies in racks from two sides). To evaluate the effect of the mislocation of a fuel assembly in the corner between two racks, the following cases are evaluated:

  • Case 2.6.6.2.1: The MCNP5-1.51 model is three 8x8 arrays of SFP rack cells. The misplaced fuel assembly is in the corner between two racks. The fuel assemblies in the rack are eccentric toward the mislocated fuel assembly. The misplaced fuel assembly is placed as close to the racks as possible. All fuel assemblies in the model are the design basis fuel assembly. Figures 2.1 1(a) and 2.11 (b) show the MCNP5-1.51 model used for this analysis.

" Case 2.6.6.2.2: The MCNP5-1.51 model is the same as Case 2.6.6.2.1, except with all fuel assemblies are centered. See Figures 2.1 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 the temperature of the model is set to the maximum (150 'F).
  • Case 2.6.6.2.4: The MCNP5-1.51 model is the same as Case 2.6.6.2.2, except the temperature of the model is set to the maximum (150 'F).

2.6.6.3 Mislocation of a Fuel Assembly between the SFP Rack and the Inspection Platform As discussed in Section 2.5, the fuel handling/inspection/reconstitution platform may have one fuel assembly in it at a time. There is a possibility that a fuel assembly is mislocated between the Project No. 2127 Report No. HI-2125245 Page 32

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

" Case 2.6.6.3.1: The MCNP5-1.51 model is an 8x8 array of SFP rack cells. The misplaced fuel assembly is adjacent to the SFP rack and the inspection platform. The fuel assembly in the platform is lined up with the mislocated fuel assembly. The fuel assemblies in the rack are eccentric toward the mislocated fuel assembly. The misplaced fuel assembly is placed as close to the rack and fuel assembly in the inspection station as possible. All fuel assemblies in the model are design basis fuel assembly. The side of the fuel in the platform which does not have any fuel has at least 12 inches of water. Figure 2.12(a) shows the MCNP5-1.51 model used 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 fuel assemblies are centered. See Figure 2.12(b).
  • Case 2.6.6.3.3: The MCNP5-1.51 model is the same as Case 2.6.6.3.1, except the temperature of the model is set to the maximum (150 TF).

" Case 2.6.6.3.4: The MCNP5-1.51 model is the same as Case 2.6.6.3.2, except the temperature of the model is set to the maximum (150 'F).

2.6.7 Mis-installment of an Insert on Wrong Side of a Cell There is a small possibility that an insert is installed on wrong sides of the cell. In this case, there may not be a poison between a fuel assembly placed in that cell and a fuel assembly in an adjacent cell. However, the effect of this mis-installment is bounded by that of 'misloaded fuel assembly/missing insert' accident that does not include the insert in one rack cell, as discussed in Section 2.6.5. No separate calculation is performed for this accident.

2.6.8 Insert Mechanical Wear Handing accidents and other environmental damage may cause scratches and local wear of inserts. The effect of this accident is bounded by that of 'misloaded fuel assembly/missing insert' accident, as discussed in Section 2.6.5.

2.6.9 Rack Movement In the event of seismic activity, there is a hypothetical possibility that the storage rack arrays may move and come closer to each other. Since there is no water gap modeled between cells of a storage rack, the reactivity of the rack movement case is bounded by the reactivity of the design basis calculation.

2.7 Project No. 2127 Report 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 the neutron 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 infinite array used in the design basis calculations. However, since the neutron absorbing inserts are located in 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 of the spent fuel pool (correspondingly, north-east corner) has two sides that are not adjacent to the insert. Due to the neutron leakage on the periphery of the spent fuel pool the reactivity increase is not expected. Nevertheless, to evaluate the effect of such conditions, the full spent fuel pool model (74x74 array) loaded with the cell centered design basis fuel assemblies and the model where all fuel assemblies are shifted to the fuel assembly in the corner, which is discussed above, were evaluated.

Project No. 2127 Report No. HI-2125245 Page 34

2.9 ReconstitutedFuel Assemblies The SFP contains various reconstituted assemblies which were examined and determined to be relatively old and low reactivity designs. The reconstitution of these fuel assemblies removed fuel rods and replaced them by either fuel rods that are of the same or less initial enrichment and equal or greater Gd loading (with burnup similar to the rod they replaced) or solid stainless steel rods.

The reactivity effect of this reconstitution is not sufficient to make the reconstituted fuel assembly more reactive than the bounding lattice. Therefore, reconstituted assemblies are covered by the design basis Optima2 Q122 lattice 146. Future reconstituted assemblies will replace fuel rods with stainless steel rods.

Project No. 2127 Report No. 111-2125245 Page 35

3. ACCEPTANCE CRITERIA 3.1 Applicable Codes, Standardsand Guidance's Codes, standard, and regulations or pertinent sections thereof that are applicable to these analyses include the following:
  • Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criterion 62, "Prevention of Criticality in Fuel Storage and Handling."
  • Code of Federal Regulations, Title 10, Part 50.68, "Criticality Accident Requirements."
  • USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Revision 3 - March 2007.
  • L. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T.

Collins, August 19, 1998.

" ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors (withdrawn in 2004).

  • USNRC, NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, January 2001.

" DSS-ISG-2010-01, Revision 0, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools.

Project No. 2127 Report No. HI-2125245 Page 36

4. ASSUMPTIONS The analyses apply a number of assumptions, either for conservatism or to simplify the calculation approach. Important aspects of applying those assumptions are as follows:
1. Bounding or sufficiently conservative inputs and assumptions are used essentially throughout the entire analyses, and as necessary studies are presented to show that the selected inputs and parameters are in 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 because this spent fuel pool contains no soluble boron, the region around the fuel rods is under-moderated, as confirmed by the fuel tolerances calculations that change the fuel to moderator ratio (Section 7.1.7.1); therefore, neglecting the spacer grid places more water within the calculation model. In addition, the inconel springs within the spacer are a stronger neutron absorber than water. The active fuel region repeats periodically in the vertical direction. Therefore, neutron absorption in upper and lower tie plates, fuel plenums, etc. is neglected.

3. The neutron absorber length in the rack is more than the active region of the fuel, but it is modeled to be the same length.
4. The fuel density is assumed to be equal to the pellet density, and is conservatively modeled as a solid right cylinder over the entire active length, neglecting dishing and chamfering. This is acceptable since the amount of fuel modeled is more than the actual amount.
5. For the inserts, only the worst case bounding material specifications are used (minimum B-I 0 loading and minimum thickness).
6. All models are laterally infinite arrays of the respective configuration, neglecting lateral leakage. The exception is where the model boundaries are water, as specified.
7. All fuel cladding materials are modeled as pure zirconium, while the actual fuel cladding consists of one of several zirconium alloys. This is acceptable since the model neglects the trace elements in the alloy which provide additional neutron absorption.

8.

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

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

5. INPUT DATA 5.1 Fuel Assembly Specification The SFP racks are designed to accommodate the following fuel assembly types used in the Quad Cities Unit I and Unit 2, which are presented in a chronologic order along with the initial maximum planar average enrichment (IMPAE):

The specifications for the most reactive fuel assemblies from the fuel product lines discussed above are presented in Table 5.1. The additional specifications for other fuel design variations are presented in Appendix A.

The fuel assembly MCNP model used for the design basis calculations is presented in Figure 5.4.

The fuel rod, cladding and channel are explicitly modeled.,

Axially, the design basis MCNP model considers the bounding lattice along the entire length and uses water reflectors at the top and bottom. The MCNP model for the margin evaluation calculations discussed in Section 2.4 differ from the design basis model in that the active length specifically considers each actual lattice in its actual axial configuration (i.e. all the lattices from the Q122 bundle are modeled in the same MCNP mrNAA1\

5.2 Reactor Parameters The reactor core parameters are provided in Table 5.2(a). The reactor control blade data are provided in Table 5.2(b). The reactor control parameters used in CASMO-4 screening and design basis calculations are provided in Table 5.2(c).

5.3 Spent Fuel PoolParameters 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 Tables 5.3(a) and 5.3(b). The Quad Cities Unit I and Unit 2 SFP are shown in Figures 5.2(a) and 5.2(b),

respectively.

model consists of a single rack cell with periodic boundary conditions through the centerline of the water (BORAFLEX replacement), thus simulating an infinite array of storage cells. The storage rack cell is modeled the same length as the active fuel and all other storage rack materials are neglected. The neutron absorber is modeled with the worst case bounding values (the minimum B-10 loading and the minimum thickness) provided in Table 5.3(b) and Figure 5.3.

The cell wall thickness of the boundary is different from that of inner walls. The cell wall thickness of the boundary is thicker than the inner wall thickness. The SF1P model uses the inner cell wall thickness only, as given in Table 5.3(a), because it decreases the amount of steel in the model, which acts a neutron absorber.

The 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 in Table 5.4(b) for fuel materials.

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

6. COMPUTER CODES The 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 full three dimensional calculations for the loaded storage racks. MCNP5-1.51 was run on the PCs at Holtec.

" CASMO-4 [4] is a two-dimensional multigroup transport theory code developed by Studsvik. CASMO-4 is used to perform the depletion calculation for the pin-specific approach, and for various studies. CASMO-4 was run on the PCs at Holtec.

Project No. 2127 Report No. f11-2125245 Page 40

7. ANALYSIS 7.1 Design Basis and Uncertainty Evaluations 7.1.1 7.1.2 Determination of the Design Basis Fuel Assembly Lattice As discussed in Section 2.3.1.3, MCNP5-1.51 calculations were performed to determine the design basis lattice. The results for the SVEA-96 Optima2 Q122 lbel assembly are presented in Table 7.2(a) . The results for the GE 14 lattice type 5 are presented in Table 7.2(b), along with the bounding result of the SVEA-96 Optima2 Q122. As can be seen, the SVEA-96 Optima2 Q122 lattice type 146 is bounding, and thus it is selected as the design basis lattice. The CASMO-4 model of the SVEA-96 Optima2 bundle Q122 lattice 146 used for depletion calculations is shown in Figure 5.1.

7.1.2.1 Fuel Assembly De-Channeling As discussed in Section 2.3.1.5.4, the reactivity of the second most reactive assembly with no Zr channel at various radial positioning was evaluated. The results are provided in Table 7.2(b) and compared with the reactivity of the design basis lattice (SVEA-96 Optima2 Q122 lattice type 146).

As can be seen, the SVEA-96 Optima2 Q122 lattice type 146 is bounding. Therefore, storage of fuel assemblies without channels is acceptable.

7.1.3 Optima2 CASMO-4 Model Simplification Effect As discussed in Section 2.3.1.4, the effect of CASMO-4 model simplifications on the calculated reactivity of the SVEA-96 Optima2 Q122 lattice 146 was evaluated. The results are provided in Table 7.3. As can be seen, the reactivity of the simplified model is comparable to that of the complete model of SVEA-96 Optima2 Q122 lattice 146 (essentially within the 95/95 uncertainty between the two calculations). Therefore, the results show that the CASMO-4 model simplification Project No. 2127 Report No. HI-2125245 Page 41

does not have a significant impact on the analysis conclusions regarding the determination of the design basis lattice.

7.1.4 Core Operating Parameters As discussed in Section 2.3.1.5, the effects of the core operating parameters on the reactivity were evaluated. The results are provided in Table 7.4. The results show that the two dominant core operating parameters are the control blade insertion and void fraction. The other core operating parameters have an insignificant impact. Therefore, the design basis (bounding) core operating parameters are: control blades inserted, 0% void fraction, maximum fuel and moderator temperature and maximum specific power.

7.1.4.1 Reactor Power Uprate As discussed in Section 2.3.1.5.1, the effect of the MUR on the reactivity was evaluated. The results are provided in Table 7.4. The most important core operating parameters are rodded operation (control blades) and void fraction. Other parameters have relatively negligible effects on reactivity.

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

7.1.5 Water Temperature and Density Effect As discussed in Section 2.3.2, the effects of water temperature, and the corresponding water density and temperature adjustments (S(a,3)) were evaluated for SFP racks. The results of these calculations are presented in Table 7.5.

The results of the SFP temperature and density calculations show that as expected (for poisoned racks) the most reactive water temperature and density for the SFP racks is a temperature of 39.2 '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 in Table 7.11 and Table 7.14.

Project No. 2127 Report 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 was determined. The results of these calculations are presented in Table 7.7. The maximum positive delta-k value for each tolerance is statistically combined.

The maximum statistical combination of fuel assembly tolerances is used to determine keff in Table 7.11 and Table 7.14.

7.1.7.2 SFP Rack Tolerances As discussed in Section 2.3.4.2, the effect of the manufacturing tolerances on reactivity of the SFP racks with inserts was determined. The results of these calculations are presented in Table 7.8. The maximum positive delta-k value for each tolerance is statistically combined.

The maximum statistical combination of the SFP rack tolerances is used to determine kerr in Table 7.11 and Table 7.14.

7.1.8 Radial Positioning 7.1.8.1 Fuel Assembly Radial Positioning in SFP Rack As discussed in Section 2.3.5.2, twelve fuel assembly radial positioning cases in racks were evaluated. The results of these calculations are presented in Table 7.9(a). For each eccentric position case, the result for similar but cell centered case is considered as a reference. The results show that most cases show a negative reactivity effect, however some delta k, 1 values are positive. Therefore, a maximum delta k.,, value is applied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in Table 7.11 and Table 7.14.

7.1.8.2 Fuel Orientation in SFP Rack As discussed in Section 2.3.5.4, five fuel assembly orientation cases in racks were evaluated. The results of these calculations are presented in Table 7.9(b). The result for the reference case is also included. The results show that all cases are statistically equivalent and the reactivity effect of fuel orientation is negligible. Nevertheless, a maximum positive delta k,*,* value is applied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in Table 7.11 and Table 7.14.

Project No. 21.27 Report No. HI-2125245 Page 43

7.1.9 Fuel Rod Geometry Change 7.1.9.1 i ne results are presenteo in i aoe /.1 v.

The maximum 'ke,,ac -cIcrercnce is added as a bias, and the '2 * (*/('calc2

+ G3cf"2cn, (95/95 uncertainty) is added as an uncertainty to determine kerr in Table 7.11 and Table 7.14.

7.1.9.2 7.1.10 7.2 Maximum kff Calculations/forNormal Conditions As discussed in Section 2.3.8, the maximum kfrf for normal conditions is calculated. The results are tabulated in Table 7.11. The results show that the maximum keff for the normal conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95% confidence level.

7.3 Margin Evaluation As discussed in Section 2.4, the margin analyses were performed using the nominal values for poison thickness and loading, as well as the actual lattice configuration of the Optima2 Q122 fuel assembly. The results of calculations are provided in Table 7.12(a) and Table 7.12(b). As can be seen and is expected, the reactivity of design basis is larger. The use of a minimum B-10 loading relative to use of a nominal B-10 loading with tolerance uncertainty provide an additional -1%

reactivity margin to the regulatory limit with a 95% probability at a 95% confidence level.

Project No. 2127 Report No. HI-2125245 Page 44

The summary of the margin evaluation is presented in Table 7.12(c). The result shows that quantified margin remains in the analysis to offset potential effects not already considered in the model.

7.4 Abnormal andAccident Conditions As discussed in Section 2.6, the effects of empty storage cell, increased temperature, misloaded fuel assembly/missing insert, and mislocated fuel assembly accidents on reactivity were evaluated. The results are provided in Table 7.13(a) and Table 7.13(b).

As can be seen, the increased water temperature will not result in an increase in reactivity.

Both misloaded fuel assembly/missing insert and mislocated fuel accidents may result in an increase in reactivity. For the SFP racks, the effect 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 and accident conditions, discussed in Section 7.5.

The condition with the empty storage cell without insert in the spent fuel rack shows a lower reactivity than a design basis case, therefore, it is acceptable to have the empty storage cell without insert in the spent fuel pool.

7.5 Maximum keff Calculationsfor Abnormal andAccident 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 and accident conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95%

confidence level.

7.6 7.7 Spent Fuel Rack Interfaces As discussed in Sections 2.8, the interface between SFRs and pool walls, i.e. effect on reactivity of the peripheral fuel assemblies, that have a side non-adjacent to the insert, was evaluated. The results are provided in Table 7.17. As can be seen, this condition will not result in an increase of SFR reactivity. This result is expected because the infinite array design basis model is an infinite array of Project No. 2127 Report 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. no additional cell with an insert). Therefore, this water gap edge allows for neutron leakage and as the calculations show result in statistically equivalent results.

Project No. 2127 Report No. HI-2125245 Page 46

8. CONCLUSION The criticality analysis for the storage of BWR assemblies in the Quad Cities SFP racks with NETCO-SNAP-IN 6 inserts has been performed. The results for the normal condition show that l~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 highest reactivity. The results for the accident condition show that k',T is M with the storage racks fully 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 abnormal and accident conditions have been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95.

Pro.ject No. 2127 Report No. HI-2125245 Page 47

9. REFERENCES

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

[2] L.I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T.

Collins, August 19, 1998.

[3] "Nuclear Group Computer Code Benchmark Calculations," Holtec Report HI-2104790 Revision 1.

[4] M. Edenius, K. Ekberg, B.H. Forss6n, and D. Knott, "CASMO-4 A Fuel Assembly Burnup 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, Studsvik of America, Inc., (proprietary); and D. Knott, "CASMO-4 Benchmark Against MCNP,"

SOA-94/12, Studsvik of America, Inc., (proprietary).

[6] DSS-ISG-2010-01, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, Revision 0.

[7] Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001.

[8] HI-2002444, Latest Revision, "Final Safety Analysis Report for the HI-STORM 100 Cask 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 Data Center, Brookhaven National Laboratory, Upton, USA.

[11] "Spent Nuclear Fuel Burnup Credit Analysis Validation", ORNL Presentation to NRC, September 21, 2010.

[12] An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (kif) Predictions, NUREG/CR-7109, April 2012.

[13] OECD / NEA Data Bank, Java-based Nuclear Information Software, Janis version 3.3.

Project No. 2127 Report No. HI-2125245 Page 48

[14] EPRI 1003222, "Poolside Examination Results and Assessment, GEl I BWR Fuel Exposed to 52 to 65 GWd/MTU at the Limerick 1 and 2 Reactors," December 2002.

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

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Table 7.2(a)

Results of the MCNP5-1.51 Calculations for SVEA-96 Optima2 Q122 Lattices Note 2: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta k,,,, may occur at an exposure which differs from that shown above.

Project No. 2127 Report No. HI1-2125245 Page 68

Table 7.2(a) Continued Burnup sigma Max Uncert.

Description (GWd/mtU) itaC delta k,,,,

(95/95) 15 16 17 18 Lattice 149 19 -0.0207 0.0016 (void) 20 21 15 16 17 Lattice 149 18 -0.0189 0.0016 (water) ,t 19 20 21 15 16 17 Lattice 150 18 -0.0154 0.0016 19 20 21 15 16 17 Lattice 151 18 - -0.0111 0.0016 19 20 21 Note 2: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta k,01I may occur at an exposure which differs from that shown above.

Project No. 2127 Report No. 2125245 Page 69

Table 7.2(b)

Results of the MCNP5-1.51 Calculations for GE14 Lattice Type 5 Burnup k Uncert.

Description (GWd/mtU) kdV sigma delta k (95/95)

SVEA-96 Optima2 Q122 15.5 Reference Reference lattice type 146 Single GEl4 13 -0.0543 0.0016 Single GE14 13.5 -0.0509 0.0015 Single GEl4 14 -0.0491 0.0016 Single GEI4 14.5 1 -0.0469 0.0015 Single GEl4 15 -0.0473 0.0015 Single GE14 15.5 -0.0479 0.0015 Single GE14 16 -0.0485 0.0015 Single GE14 16.5 1___l_-0.0482 0.0015 Single GEl4 17 -0.0500 0.0015 2x2 GE14 - with channel (cell 14.5 Reference Reference centered) (Case 2.3.1.5.4.1) 14.5 __Rfec Rern 2x2 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 0.0015 center (Case 2.3.1.5.4.3) 2x2 GE14 - no channel /eccentric 14.5 -0.0238 0.0015 out (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 kcal may occur at an exposure which differs from that shown above.

Project No. 2127 Report No. HI-2125245 Page 70

Table 7.3 Results of the MCNP5-1.51 Calculations for Design Basis and Simplified Model of SVEA-96 Optima2 Q122 Lattice Type 146 Burnup Code kalc sigma Description (GWd/mtU)

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

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

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

Note 1: These calculations were performed using the design basis core operating parameters as indicated in Table 5.2(c).

Project No. 2127 Report No. HI-2125245 Page 71

Table 7.4 Results of the MCNP5-1.51 Calculations for Core Operating Parameters Power Fuel Moderat Void Burnup Description 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 16 decreasing Moderator temperature 23.688 Yes 1176 528.8 0 15.5 decreasing Void fraction 23.688 Yes 1176 547 94 22 increasing Un-rodded operation 23.688 No 1176 547 0 17

-il I/-

24.1617 Yes 1276 547 0 15.5 1 I 2 II L______ Yes_376_47_0 _5. ____ ____

20.1348 Yes 1176 547 0 15.5 -lIIl-l Note 1: The burnup calculations for core operating parameters were perfonrmed from 14 GWd/mtU to 24 GWd/mtU. For each core operating parameter, only reactivity of the burnup in this range which results in the largest reactivity is reported.

Note 2: The bounding case is bolded.

Note 3: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta k,,,

may occur at an exposure which differs from that shown above.

Project No. 2127 Report No. 1-11-2125245 Page 72

Table 7.5 Results of the MCNP5-1.51 Calculations for the Effect of Water Temperature and Density Water Water Temperature Temp. Density Adjustment, delta k,, Unrtsigma (95/5.

Description Burnup (GWd/mtU) (OF) (g/cc) S(aF) (95/95)

(OF)

Reference:

lower bound temperature 15.5 39.2 1 68.81 Reference Ref.

(Case 2.3.2.1)

Upper bound temperature for normal operation, low 15.5 150 0.98026 68.81 -0.0041 0.0015 S(a4)

(Case 2.3.2.2)

Upper bound temperature for normal operation, 15.5 150 0.98026 170.33 -0.0066 0.0015 high S(a,3)

(Case 2.3.2.3)

Note 1: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta k.,,

may occur at an exposure which differs from that shown above.

Project No. 2127 Report No. HI-2125245 Page 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 0.0064 Project No. 2127 Report No. 141-2125245 Page 74

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Project No. 2127 Report No. .11-2125245 Page 75

Table 7.7 Results of the MCNP5-1.51 Calculations for Fuel Tolerances Peak Reactivity delta kcnie Max delta kca,,

Description k.1, sigma Burnup (95/95) (95/95)

(GWd/mtU)

Design basis (reference) 15.5 Reference Reference Max fuel enrichment 16 0.0026

  • 0.0026 Min fuel enrichment 15.5 -0.0009 Max Gd loading 16 -0.0013 0.0038 0.0038 Min Gd loading 15.5 Max pellet density 16 0.0000 0.00 12 Min pellet density 15.5 0.00o 12 0 .001 Max pellet OD 15.5 0.0015 0.0015 Min pellet OD 16 0.0011 Max clad ID 16 0.00 10 0 0.0010 Min clad ID 16 0 .0008 0 .0098 Max clad OD 15.5 -0.0002 0.0027 Min clad OD 15.5 0.0027 Max sub-bundle pitch 15 0.0098 0.0098 Min sub-bundle pitch 16.5 -0.0089 _____

Max pin pitch 15.5 0.0122 0.0122 Mill pin pitch 15.5 -0.0086 _____

Max combined water wing canal inner width, channel outer square width, channel corner inner radius and central water canal inner 15 -

FI 0.0031 square width 0.0031 Min combined water wing canal inner width, channel outer square width, channel corner inner radius 15.5 - I -0.0011 and central water canal inner souare width Max combination of channel wall thickness and water cross wall 16 l 0.0008 thickness 0.0019 Min combination of channel wall thickness and water cross wall 15.5 0.0019 thickness Statistical combination of fuel tolerances 0.0171 Note 1: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kclc may occur at an exposure which differs from that shown above.

Project No. 2127 Report No. HI-2125245 Page 76

Table 7.8 Results of the MCNP5-1.51 Calculations for Rack Tolerances Burnup delta kai, Max delta Description (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/A Max cell pitch Max wall thickness 15.5 0.0025 Min wall thickness 15.5 _ 0.0008 Max insert width 15.5 -0.0005 0.0004 Min insert width 15.5 0.0004 Statistical combination of rack tolerances 0.0026 Project No. 2127 Report No. 1-11-2125245 Page 77

Table 7.9(a)

Results of the MCNP5-1.51 Calculations for Fuel Radial Positioning in SFP Racks BurnupUnc.

BurnU) kualc sigma delta kc.I, (95/95)

Description (GWd/mtU) 95/95)_

2x2 reference 15.5 Reference Ref.

(Case 2.3.5.2.1) 15.5 _lRence 2x2 eccentric center 15.5 -0.0053 0.0015 (Case 2.3.5.2.2) 2x2 eccentric in 15.5 -0.0081 0.0013 (Case 2.3.5.2.3) ......

2x2 eccentric out 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 . 1 8x8 insert/cell center 15.5 0.0016 0.0014 (Case 2.3.5.2.10) .......

Ix 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. 2127 Report No. 1-11-2125245 Page 78

Table 7.9(b)

Results of the MCNP5-1.51 Calculations for Fuel Orientation in SFP Racks Burnup IUnc. (95/95)

Description Burnup k,,,, sigma delta Iellc (GWd/mtU) 9595 Reference (Shown in 15.5 Reference Ref.

Figure 2.9(a)) ........

Rotated fuel assembly (shown in Figure 2.9(b)) 15.5 -0.0008 0.0014 Rotated fuel assembly (shown in Figure 2.9(c)) 15.5 -0.0007 0.0014 Rotated fuel assembly (shown in Figure 2.9(d)) 15.5 -0.0013 0.0013 Rotated fuel assembly (shown in Figure 2.9(e)) , 5 -0.0007 0.0013 Project No. 2127 Report No. 1-11-2125245 Page 79

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Project No. 2127 Report 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 of Using Minimum B4C Loading and Minimum Insert Thickness on Reactivity B-10 Areal Description Burnup Density k,,, sigma delta kIc (GWd/mtU) (g/cm2 )

Reference (design basis) 15.5 0.0116 Reference (Case 2.4.1)

Rack with nominal values for 134 C loading and insert 15.5 0.0133 1 -0.0103 thickness (Case 2.4.2)

Project No. 2127 Report No. 111-2125245 Page 82

Table 7.12(b)

Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate the Effect of the Actual Optima2 Q122 Fuel Assembly Desripion Description (GWd/mtU) kenic lurnu) sigma Max kcille delta k,,,,

Optima2 Q122 Lattice 146 (Design basis) 15.5 = I = IReference IReference (Case 2.4.1) 15 Optima2 Q122 155 0.8873 Lattice 147' 16 16 16.5 Optima2 Q122 0.8843 Lattice 148 17 14 14 Optirna2 Q122 Lattice 149 14.5 U, 0 (*l.,,, J 15 14_

14.5 Optima2 Q122 0.8863 Lattice 150 15 14 14.5 Optima2 Q122 0.8876 Lattice 151 15 Peak Optima2 Q122 t Reactivity Fuel Assembly (Case 2.4.3)

Burnups ~1- 0.8925 -0.0066 (bolded) t The top and bottom natural blankets were conservatively neglected and replaced by adjacent lattice.

Project No. 2127 Report No. HI-2125245 Page 83

Table 7.12(c)

Margin Evaluation Summary of the Margin Evaluation Description Value Insert Composition Margin, from Table 7.12(a) -0.0103 Actual Optima2 Fuel Assembly Margin, from -0.0066 Table 7.12(b)

Calculated Margin -0.0169 Prqjject No. 2127 Report No. 2125245 Page 84

V' mIm U Im I I m"mm m --

F~ mm--mI m mm-nU m --

P~ m m --

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m m m m m 0 m

m. m Imm Project No. 2127 Report No. 1-11-2125245 Page 85

Table 7.13(b)

Results of the MCNP5-1.51 Calculations for the Empty Storage Rack Cell without Insert Burnup keli sigma delta kcal Uncertainty Description

_________(GWd/mtU) (95195)

Design basis 15.5 Reference Reference (8x8 array)

Empty storage cell (cell 15.5 -0.0041 0.0016 centered)

Empty storage 15.5 -0,0081 0.0014 cell (eccentric) 15.5,_ -0.0081 _0.0014 I Note 1: The design basis fuel assembly (Optima2 Q122 Lattice Type 146) is used for these calculations.

Project No. 2127 Report No. HI-2125245 Page 86

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Project No. 2127 Report No. HI-2125245 Page 87

Project No. 2127 Report No. 1.11-2125245 Page 88 Table 7.16 Results of the MCNP5-1.51 Calculations for Axially Infinite Optima2 Q122 Lattices Burnup kcai kc.f Delta-K Uncertainty Description (GWd/mtU) (reference) (infinite)

Optima2 Q122 Lattice 146 15.5 0.0013 0.0015 (Design basis) .....

Optima2QI22 15.5 0.0018 0.0015 Lattice 147 Optima2 QI22 16.5 0.0008 0.0014 Lattice 148 Optima2QI22 14.5 0.0011 0.0015 Lattice 149 Optima2 Q122 14.5 0.0027 0.0014 Lattice 150 Optima2 Q122 Lattice 151 -F 14.5 14.5 0.0010 0.0010 0.0015 0.0015 Note: The difference between the MCNP models under the "reference" column and the MCNP models under the "infinite" column is described in Section 5.1.

Project No. 2127 Report No. HI-2125245 Page 89

Table 7.17 Results 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 0.0016 centered) 15.5_1 _0.08 .01 Full SFP (eccentric to 15.5 -0.0053 0.0015 SFP corner)

Prqject No. 2127 Report No. 1H11-2125245 Page 90

at *1*

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Figure Proprietary Project No. 2127 Report No. HI-2125245 Page 92

Figure Proprietary Project No. 2127 Report No. HI-2 125245 Page 93

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Figure Proprietary Project No. 2127 Report No. HI-2 125245 Page 1.0 1

Figure Proprietary Page 102 Report No. 2125245 Project No. 2127

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Proprietary Figure Page 104 Report No. HI-2125245 Project No. 2127

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Figure Proprietary ProJect No. 2127 Report No. HIl -2125245 Page 1 15

Figure Proprietary Project No. 2127 Report No. HI-2 125245 Page 116

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Figure Proprietary Project No. 2127 Report No. HI-2125245 Page H18

Appendix A Proprietary

Appendix B Proprietary Project No. 2127 Report No. HI-2125245 Page B-1

Appendix C Proprietary Project No. 2127 Report No. 1-H-2125245 Page C- I

Supplement 1 Additional Calculations to Support the Revised NETCO-SNAP-IN Rack Insert Design (I1 pages including this page)

Project No. 2127 Report No. HI-2125245 Page S I-I1

SI.1 Introduction This Supplement documents the criticality safety evaluation for the storage of spent BWR fuel in the Unit I and Unit 2 spent fuel pools (SFPs) at Quad Cities Station operated by Exelon. The purpose of this analysis is to justify that the specified changes in the NETCO-SNAP-IN rack insert design [Sl.1 j are acceptable and bounded by the current analysis, presented in the main part of the report.

S1.2 Methodology See Section 2 of the main report and as otherwise discussed below.

S1.3 Acceptance Criteria See Section 3 of the main report.

S1.4 Assumptions See Section 4 of the main report and as otherwise discussed below.

S1.5 Input Data See Section 5 of the main report. The revised dimensions of the NETCO-SNAP-INO rack insert are presented in Table SI - I and Figure SI - I.

S1.6 Computer Codes See Section 6 of the main report.

S1.7 Analysis The comparison of the revised insert parameters presented in Table SI-1 with the previous insert design in Table 5.3(b) shows that changes are minor and therefore a significant impact on the conclusions made in the main part of the report is not expected. Nevertheless, to verify the negligible or minor impact of the revised insert design on results presented in the main part of the report additional calculations are presented in this Supplement. The additional calculations presented in this Supplement are similar to those in report for the following cases:

  • SFP rack tolerances
  • Fuel assembly radial positioning in the SFP rack
  • Fuel orientation in the SFP rack These cases are selected because the NETCO-SNAP-IN rack insert design change may impact the reactivity in the rack. All other calculations from the main report are not affected by the NETCO-SNAP-INO rack insert design change and the results of the unaffected calculations are Project No. 2127 Report No. 1-11-2125245 Page S 1-2

used in this Supplement where applicable. This approach is considered for both normal and accident conditions.

S 1.7.1 SFP Rack Tolerances As discussed in Section S1.7, the effect of the manufacturing tolerances on reactivity of the SFP racks with revised inserts was determined. The results of these calculations are presented in Table S1-2. 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 in TFable S 1-5 and Table S1-6.

S 1.7.2 Fuel Assembly Radial Positioning in the SFP Rack As discussed in Section S1.7, twelve fuel assembly radial positioning cases in the racks were evaluated. The results of these calculations are presented in Table S1-3. For each eccentric position case, the result for similar but cell centered case is considered as a reference. The results show that most cases show a negative reactivity effect, however some delta k,,,, values are positive. Therefore, a maximum delta kcIc value is applied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in 'able S1-5 and Table S1-6.

S1.7.3 Fuel Orientation in the SFP Rack As discussed in Section S1.7, five filel assembly orientation cases in racks were evaluated. The results of these calculations are presented in Table S1-4. The result for the reference case is also included. The results show that all cases are statistically equivalent and the reactivity effect of fuel orientation is negligible. Nevertheless, a maximum positive delta k, 1 , value is applied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in TFable 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 the main part of the report. The results for the revised NETCO-SNAP-IN rack insert design and the results from the main part of the report are tabulated in Table S1-5. The results show that the maximum 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 are bounded by the results from the main part of the report.

Project No. 2127 Report 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 the main part of the report. The bounding accident case from the main report is recalculated using the revised NETCO-SNAP-IN rack insert design. The results for the revised NETCO-SNAP-INO rack insert design and the results from the main part of the report are tabulated in Table S1 -6. The results show that the maximum k~ff for abnormal and accident conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95% confidence level 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 Insert Design Information", dated 09/i 1/2012.

S1.9 Conclusions The criticality analysis for the storage of BWR assemblies in the Quad Cities SFP racks with revised NETCO-SNAP-INO inserts has been performed. The results show that kefr is M with the stora racks full 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 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, the reactivity will not exceed the regulatory limit of 0.95.

The results show that the specified changes in the insert des the current analvsis. oresented in the main Dart of the report. I Therefore, any insert width dimension between the value used in the main report including the specified manufacturing tolerances and the value evaluated in this Supplement is acceptable.

Project No. 2127 Report No. 1I1-2125245 Page S 1-4

Table SI-I Fuel Rack Insert Revised Dimensions [S 1.1]

- 1 For the details of the insert dimensions, see Figure S I-i.

t See 'Table 5.3(b)

Project No. 2127 Report No. HI-21 25245 Page S1-5

Table S1-2 Results of the MCNP5 Calculations for Revised Rack Tolerances Revised Reference Burnup delta kac Max delta Max delta Description (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 Max cell pitch 0.0000 Max wall thickness 15.5 op146-rt203155r 0.0017 0.0017 0.0025 Min wall thickness 15.5 op146-rt204155r !_0.0011_

Max insert width 15.5 op146-rt206155r 1_ 1 0.0016 0.0030 0.0004 Min insert width 15.5 op]46-rt207155r 0.0030 Statistical combination of rack tolerances 0.0035 0.0026 1 See Table 7.8 Note 1: The CASMO depletion calculation filenames are op 146-dbc(-ac).

Project No. 2127 Report No. 111-2125245 Page S 1-6

Table S 1-3 Results of the MCNP5-1.51 Calculations for Revised Fuel Radial Positioning in SFP Racks Revised Reference Burnup Revised Refernc.e Description (GWd/mtU) Filename kCHIC sigma delta kcaic Un(9 delta kcal, 9 nc 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) .0.0.80 2x2 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. Ref Case 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 1555r Ix] insert/cell center 15.5 lxlicntl55r1 0.0004 0.0015 0.0000 0.0015 (Case 2.3.5.2.12) 1 ___ I i I I t See Table 7.9(a)

Note 1: The CASMO depletion calculation filenames are opl46-dbc(-ac).

Project No. 2127 Report No. HI-2125245 Page S1-7

Table S1-4 Results of the MCNP5-1.51 Calculations for Revised Fuel Orientation in SFP Racks Revised Revised Reference Reference unc.

Burnup Description (GWd/mtU) Filename k,,,ý sigma delta kl* (9/5 delta kI* (9/5

___________(95/95) dlak. (95/95)

Reference (Shown in 15.5 2x2dbrotOl55r Ref Ref. Ref Ref Figure 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. 2127 Report No. HI-2125245 Page S 1-8

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