ML081820131

Netco Report No. 901-02-05, Benchmarking Computer Codes for Calculating the Reactivity State of Spent Fuel Storage Racks, Storage Casks and Transportation Casks, Page 14 Through End

ML081820131
Person / Time
Site:	Nine Mile Point
Issue date:	06/27/2008
From:	Northeast Technology Corp
To:	Office of Nuclear Reactor Regulation
References
TAC MD8434 NETCO 901-02-05
Download: ML081820131 (18)

Text

4.0 CONCLUSION

S SCALE-5 and MCNP5 have been benchmarked by modeling five (5) Babcock and Wilcox critical experiments and eight (8) CSNI critical experiments representative of fuel storage rack and fuel cask geometries. At a 95% probability / 95% confidence level, the computed bias for SCALE-5 and MCNP5 are -0.01381 and -0.01460, respectively.

CASMO-4 has also been benchmarked by modeling the five (5) Babcock and Wilcox critical experiments as infinite arrays. Best estimates of the k, for the exact same geometry were calculated using SCALE-5 and applying the mean bias reported above.

The CASMO-4 bias with respect to these values was calculated to be -0.01028

+/- 0.00198 (1 sigma). The comparison of SCALE-5 and CASMO-4 serves to verify the results of each with respect to the other.

It is therefore concluded that these calculational methods have been adequately benchmarked and validated. They may be used individually or in combination for the criticality analysis of spent fuel storage racks, fuel casks and fuel casks in close proximity to fuel storage racks, provided the appropriate biases are applied.

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5.0 REFERENCES

1. "SCALE-PC: Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations and Personal Computers, Version 5," Volumes 0 through 3, RSIC Computer Code Collection CCC-725.

Oak Ridge National Laboratory: Oak Ridge, Tennessee; Draft May 2004.

2. "MCNP - A General Monte Carlo N-Particle Transport Code, Version 5",

Volumes 1 - 3, RSICC Computer Code Collection CCC-710, X-5 Monte Carlo Team, Los Alamos National Laboratory, Los Alamos, NM, April 24, 2003.

3. Ekberg, Kim, Bengt H. Forssen and Dave Knott. "CASMO-4: A Fuel Assembly Burnup Program - User's Manual," Version 1.10 STUDSVIKlSOA-95/1. Studsvik of America: Newton, Massachusetts; September 1995.

4. NETCO Report 901-02-03: "Benchmarking of the SCALE-PC Version 4.3 Criticality and Safety Analysis Sequence Using the KENO V.a Monte Carlo Code and of the Multigroup Two-Dimensional Transport Theory CASMO-4 Code",

Northeast Technology Corporation: Kingston, New York; May 1995.

5. NETCO Report 901-02-04: "Benchmarking of the SCALE-PC Version 4.3 Criticality and Safety Analysis Sequence Using the KEN05A Monte Carlo code and of the Multigroup Two-Dimensional Transport Theory CASMO-4 Code",

Northeast Technology Corporation: Kingston, New York; January 2000.

6. Baldwin, M. N., G. S. Hoovler, R. L. Eng, and F. G. Welfare. "Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel,"

BAW-1484-7. Babcock & Wilcox: Lynchburg, Virginia; July 1979.

7. Mancioppi, S., and G. F. Gualdrini. "Standard Problem Exercise on Criticality Codes for Spent-Fuel-Transport Containers," CNEN-RT/FI(81 )25. Comitato Nazionale Energia Nucleare: Rome; October 1981, Performed by CNEN for Committee for Safety of Nuclear Installations (CSNI).

8. "Quality Assurance Manual, Rev. 0, Northeast Technology Corp: Kingston, NY; 27 June 2001.

9. "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," ANSIIANS-8.1-1983, Revision of ANSI/N 16.1-1975. American Nuclear Society: La Grange Park, Illinois ;

Approved 7 October 1983.

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10. "Quality Assurance Requirements of Computer Software for Nuclear Facility Applications," Part 2.7 of "Quality Assurance Requirements for Nuclear Facility Applications," ASME NQA-2-1989, Revision of ANSI/ASME NQA-2-1986.

American Society of Mechanical Engineers: New York; Issued 30 September 1989.

11. Natrella, Mary Gibbons. Experimental Statistics, National Bureau of Standards Handbook 91. U.S. Government Printing Office: Washington, D.C.; 1 Aug. 1963.

12. Cooney, B. F., T. R Freeman, and M. H. Lipner. "Comparison of Experiments and Calculations for LWR Storage Geometries." Transactions of the American Nuclear Societv: Vol. 39, pp. 531-532; November 1981 .

13. Westfall, R M., and J. R Knight. "SCALE System Cross Section Validation with Shipping Cask Critical Experiments." Transactions of the American Nuclear Societv: Vol. 33, pp. 368-370; 1979.

14. McCamis, R H. "Validation of KENO V.a for Criticality Safety Calculations of Low-Enriched Uranium-235 Systems," AECL-10146-1 . Atomic Energy of Canada Limited, Whiteshell Laboratories, Pinawa, Manitoba; February 1991.

15. Bierman, S. R, E. D. Clayton, and B. M. Durst. "Critical Separation Between Subcritical Clusters of 2.35 Wt% 235U Enriched U02 Rods in Water with Fixed Neutron Poisons," PNL-2438. Battelle Pacific Northwest Laboratories: Richland, Washington; October 1977.

16. Bierman , S. R, E. D. Clayton, and B. M. Durst. "Critical Separation Between Subcritical Clusters of 4.29 Wt% 235U Enriched U02 Rods in Water with Fixed Neutron Poisons," NUREG/CR-0073 RC. Battelle Pacific Northwest Laboratories: Richland, Washington; May 1978.

17. "Quality Assurance Program", SOA/REV 2. Studsvik of America: Newton, Massachusetts; Approved 16 August 1991.

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Appendix 2:

Response to NRC Acceptance Review Questions

Appendix 2 NRC Comment Response

1) The licensee 's evaluation of the 1) The byproducts from degradation of the change indicates that eliminating the of the Boraflex polymer are boron Boraflex credit will not adversely affect carbide, amorphous silica (soluble) and various aspects such as material very fine crystalline silica (insoluble), all considerations, solid radioactive waste of which can leave the Boraflex panels generation, or occupational exposure. and enter the spent fuel pool water over However, there is no discussion on the time .

indefinite exposure of the SFP &

supporting components to the The silica concentration is sampled byproducts associated with the monthly per procedure. The normal Boraflex degradation. There is no mode of the spent fuel pool cleanup discussion concerning the potential system keeps the silica level in the pool increase of liquid radioactive waste low. The insoluble silica and boron generation or occupational exposure carbide deposits on horizontal surfaces associated with removing those in the spent fuel pool, primarily on the byproducts from the SFP. There is no spent fuel pool floor. If the deposition of discussion as to whether or not the these substances impacts visibility (e.g. ,

indefinite presence of those visibility of bundle handle identification byproducts materially affects any labels), vacuuming can be performed explicit or implicit assumption in a using a standard tri-nuke vacuum . The previous licensing activity. silica and boron carbide are essentially inert materials and do not chemically interact with other spent fuel pool materials.

It should be noted that elimination of the Boraflex credit in the criticality analyses requested by the license amendment request will not affect material considerations, solid radioactive waste generation, or occupational exposure.

The effects of Boraflex degradation on spent fuel pool chemistry are known and addressed by current processes. The proposed changes do not change or increase the Boraflex degradation byproducts or the current processes used to address this issue (other than in the criticality analyses).

2) In the Section 2.0, Analytical Methods 2) The CASMO and KENO methods used and Assumptions, the criticality in this analysis have been recommended analysis indicates CASMO-4 was used for criticality analysis per the Kopp 1

for both the standard cold core Guidance of 8/19/1998. In the report geometry (SCCG) and SFP storage this Guidance has been moved from rack analysis. CASMO-4 is unable to reference 8 to the list of regulations in explicitly model the asymmetries of the Section 1.2.

NMP U1 SFP storage racks. The criticality analysis indicates NETCO's use of these methods have approximations are made in CASMO-4 been previously subject to detailed to accommodate the asymmetry and review by the NRC and approved.

that the results are only used in a Reference 8 has been added to the

'relative' basis. However, there is no report and is an SER issued by the discussion on how the CASMO-4 NRC.

approximations affect the results.

Therefore, there is no assurance that CASMO has been used in most criticality the results are appropriate for use on analyses supporting reracking and spent any basis, even a 'relative' basis. The fuel storage expansion activities since staff has only found one licensing the 1980's . CASMO has been used in activity on the docket regarding numerous spent fuel rack design CAS MO-4. The staff has licensed analyses including the re-rack Duke Energy Corporation (DEC) installation of BORAL racks at Nine Mile topical report DPC-NE-1005P for core Point 1. [Letter from Daryl Hood, reload analyzes for Catawba Nuclear USNRC, to John Mueller, NMPC, Station Units 1 and 2 and McGuire 6/17/1999, "Issuance of Amendment for Nuclear Station Units 1 (Reference 4). Nine Mile Point Nuclear Unit No.1 to Since the CASMO-4 user's manual, Reflect a Planned Modification to which is referenced in the NMP U1 Increase the Storage Capacity of the SFP LAR, is not publicly available , the Spent Fuel Pool (TAC No. MA1945)."]

DEC topical report documents will have to be reviewed to determine if Referring to SCALE (KENO V.a) and limitations on the use of CASMO-4 in CASMO , the above reference (TAC No.

asymmetric configurations were MA1945) concludes, "These codes have identified or even considered. been widely used for the analyses of fuel Additionally, there is currently a rack reactivity and have been methodology change associated with bench marked against the results from the DEC topical report being reviewed numerous critical experiments."

by the staff. Those documents will also have to be reviewed to determine The KENO V.a verification of the whether or not the use of CASMO-4 is CASMO-4 approximation is given in appropriate in this manner. There is Section 2.0, page 9.

insufficient information in the LAR to begin a review to determine whether or not it is reasonable to use CASMO-4 to model SFP storage racks.

3) Section 2.0 lists eleven 3) For the reference analysis (per Kopp

".. .conservative assumptions have Guidance 81/9/981[ 5.A.3d) the been applied to the analyses." temperature and density of the 2

Whether or not they are conservative moderator for the in-rack condition was is not supported with any additional set at 150°F(O.98 gm/cm3 ) which justification. In the case of SFP bounds the maximum abnormal moderator density assumption it condition temperature of 140°F for a states: "The moderator is assumed to full core offload as stated in Section be demineralized water at full water H.1.0 of the NMP1 UFSAR.

density (t.Oqm/cm")." Subsequent Additionally, for the limiting lattice k., in information indicates this assumption the standard cold core geometry to be non-conservative. Water at full (SCCG), the moderator temperature is density (1.0gm/cm3 ) corresponds to a at 68°F (1.0 qm/crn"),

temperature of approximately 39°F.

This temperature is well below the SFP Resolution: This is clarified in normal and maximum operating Assumption 4 on Page 10 of NET-290-temperatures. Staff guidance in the 01.

Kopp Letter (Reference 2) and the NRC letter, OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, (Reference

3) is that the analysis is to be performed at the temperature which yields the maximum keff. The LAR's accident analysis indicates abnormally high temperature is the worst case accident. This indicates that higher temperatures are more reactive than lower, a condition fairly common in un-poisoned SFP storage racks. This indicates that performing the normal conditions portion of the analysis "... at full water density (1.0gm/cm3 )" is non-conservative. Therefore, a portion of the reactivity increase associated with the abnormally high temperature accident should be included in the normal conditions. It is possible, perhaps probable, that if the normal conditions analysis was performed at the most reactive temperature, the NMP U1 SFP analysis would indicate that the proposed LAR does not meet the regulatory requirement. There is insufficient information in the LAR to begin a review to determine why it is acceptable to use a non-conservative temperature for the criticality analysis.

4) Additionally, two of the assumptions 4) The end fittings are above and below appear to be incongruent. Both the the active fuel lattice and have no 3

assumption that it is conservative to influence in the H/U ratio in the lattice model the fuel assembly channeled region. There is a slight positive and the assumption that it is reactivity effect due to the water conservative to not model the fuel displaced by the low volume spacers.

assembly grids and end fittings appear This represents a very small relative to be balancing the absorption of volume of the lattice and is more than structural components against offset by negative reactivity effect of increasing the amount of over parasitic neutron capture in the Inconel moderation. It is not intuitively obvious grid spacer springs.

why both assumptions are conservative at the same time. There is insufficient information in the LAR to begin a review to determine why both are conservative at the same time.

5) In Section 3.1, CASMO-4 and KENO 5) The biases in Section 3.1 are best V.a Reactivity Calculation Comparison, estimate biases and are -

of the criticality analysis, the results of 0.00782+/-0.00361 for Keno V.a and-different fuel assembly configurations 0.01028+/-0.00198 for CASMO-4.

at zero burnup are compared for both These have been revised in the report CASMO-4 and KENO V.a in the to five decimal places and the one SCCG. In that discussion a bias for sigma standard deviation has been each computer code is cited. added. The 95/95 uncertainty of However, those biases are not the 0.00963 is the 1-sigma uncertainty same as those calculated in the (0.00361) in the bias multiplied by the Appendix, 'Benchmarking Computer 1-sided 95/95 tolerance factor of 2.67 Codes for Calculating the Reactivity for n=13 criticals. This has been State of Spent Fuel Storage Racks, included in the statistical treatment of Storage Casks and Transportation tolerances and uncertainties.

Casks.' These values are not stated or described elsewhere in the analysis. These appear in Table 3-2 and Table Therefore, the use of the biases cited 3-3 in the Appendix for KENO V.a and in this section is unsupported by the CASMO-4 respectively.

analysis.

Distinction must be made between best estimate biases and biases at a 95% probability with a 95% confidence level. The discussion in Section 3.1 uses best-estimate biased eigenvalues to provide the reader with a comparison of the two methods on a best estimate basis. Later when the maximum reactivity of the racks is determined (Le. Table 4 in NET-290-

01) the maximum reactivity of the fuel and racks is computed at a 95%

probability with a 95% confidence level.

4

6) In Section 3.2, Reactivity Calculations, 6) An iterative process is used to of the criticality analysis, the reactivity determine the REFFE for each fuel equivalent fresh fuel enrichment type. In this process the best estimate (REFFE) is determined. The subject KENO V.a bias (-0.00782) is added to fuel assembly designs are depleted in the KENO V.a calculated eigenvalue.

CASMO-4 in the core geometry and a Similarly, the best estimate CASMO-4 SCCG k, is determined. The REFFE bias of -0.01028 is applied to the is determined by iterative KENO V.a CASMO-4 calculated eigenvalue.

cases until an equivalent zero gadolinium fresh fuel enrichment is This discussion has been added to found, the REFFE. There is no Section 3.2.1 of the report.

discussion of how or if the computer code biases and uncertainties are " The REFFE was determined by applied. There is insufficient modeling each limiting (with respect to information in the LAR to begin a maximum uranium enrichment and review to determine whether or not the minimum gadolinia loading) fuel limiting REFFE has been determined. bundle lattice in the standard cold core geometry and in-rack geometry. The U235 enrichment was varied until the bias-corrected k, of the Keno rack model slightly exceeded the CASMO k., (normalized to Keno V.a) at the point of peak reactivity during depletion. The k., of the Keno V.a model at the REFFE always exceeds the k., at peak reactivity during depletion."

7) In Section 3.3, Effect of Tolerances and Uncertainties, of the criticality analysis, there are several informational deficiencies which must be rectified before a technical review can begin.

a) Numerous tolerances are a) The tolerance values used in the

'assumed' based on other work analysis are conservative for performed by the vendor. Whether the following reasons :

or not those assumptions are conservative or not is 1. U235 Enrichment - The indeterminate, given the following information was information provided. provided by Niagara Mohawk Power Corporation in letter dated December 21,1995 from Rocco Bianchi (GE Nuclear Energy) to Bart Franey (NMPC):

5

"Enrichment Variation:

For enrichment variations ~ 2.0%

U235 : Nominal +/- 0.088w/o" Furthermore, the letter states" Process history shows an overall process standard deviation of 0.023w/o U235 for enrichments ~

3.95% U235 ."

Resolution: The above explanation has been added to Section 3.3.1.

2. Pellet Density-Item 2 in email dated January 26, 2001 from J.

Winklebleck (NMPC) to K.

Lindquist (NETCO), subject:

Evaluation of NMP2 Spent Fuel Racks for GE 9x9 and 1Ox1 0 Fuel" states:

"As you can see by the attached email from GE, we are currently loading pellets with 97%

theoretical U density. The tolerance on this range is -2% to

+1% meaning that the Maximum density range of theoretical densities is from 95% to 98%."

Resolution: The above reference to the report and explanation has been added to Section 3.3.1.

3. Pellet Diameter: The conservative value of +/- 0.002 inches was assumed.

Examination of Bundle Announcement Reports in[9. 101 revealed that the maximum tolerance for the 7x7, 8x8, or 9x9 fuel types is +/- 0.001 inches.

Resolution: The above explanation has been added to Section 3.3.1.

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4. Clad Inside Diameter - The conservative value of +/- 0.002 inches was assumed.

Examination of Bundle Announcement Reports[9,10l revealed that the maximum tolerance for the 7x7, 8x8, or 9x9 fuel types is +/- 0.001 inches.

Resolution: The above explanation has been added to Section 3.3.1.

5. Clad Thickness - The value of

+/- 0.004 inches was assumed in Reference 12. Examination of Bundle Announcement Reports[9.1ol revealed that the maximum tolerance for the 7x7, 8x8, or 9x9 fuel types is +/-0.004 inches.

Resolution: The above explanation has been added to Section 3.3.1.

6. Gd203 Loading - From the same reference listed in (7.a.2) above, the letter contains the following pertaining to gadolinia loading:

" ...the 95% confidence limits on the mean gadolinia loading content shall be within +/-7.5%

(relative) of nominal. Individual pellet gadolinia content is limited to +/-1O% (relative) of Nominal."

Therefore, the tolerance value of

+/-O. 5 wlo gadolinia (10%

relative) employed in this analysis is conservative.

Resolution: The above explanation has been added to Section 3.3.1.

7

b) There is an uncertainty for the flux b) The tolerance effect for the stainless trap that is created by replacing the steel structural material associated Boraflex with SFP water. However, with the Boraflex material has been the description of that uncertainty is assessed. The reactivity effect was insufficient to determine whether found to be negligible.

that includes the tolerance for the stainless steel structural material associated with the Boraflex.

c) The Methodology Uncertainty is c) The Nominal Keff value in Table 4 listed as 0.0078 ~l<eff. However, contains the bias correction of this value is not included in Table 0.00782.

4, which lists the other uncertainties. Additionally, this Resolution: Table 4 has been value is not that determined in the revised to explicitly list the bias and Appendix, 'Benchmarking Keno V.a calculated reference keff Computer Codes for Calculating separately.

the Reactivity State of Spent Fuel Storage Racks, Storage Casks and Transportation Casks' for either CASMO-4 or KENO V.a. However, it is the same magnitude as the KENO V.a bias used in the CASMO-4 and KENO V.a Reactivity Calculation Comparison section.

d) An estimate of the depletion d) Additional methods for determining uncertainty is based on an the reactivity effect due to assumed function based on uncertainties in burnup have been burnup. The assumed function is proposed based on 5% of the not validated. The depletion reactivity decrement from 0 burnup uncertainty is not included in Table to the point of peak reactivity

4. And not including the depletion (without gadolinia) [Kopp, L1, uncertainty is justified by citing "Guidance on the Regulatory unquantified conservatisms from Requirements for Criticality Analysis the eleven assumptions in the of Fuel Storage at Light Water Analytical Methods and Reactor Power Plants", Proceedings Assumptions section. of a Consultancy on Implementation of Burnup Credit in Spent Fuel Management Systems, IAEA, July 28-31, 1998.]. Using this approach, the reactivity effect for depletion uncertainties would be 0.05 x 0.1347Ak =0.00674Ak. For the limiting bundle 9x9 bundle with peak 8

reactivity at 14GWD/MTU, assuming a O.02~k uncertainty at 40GWD/MTU results in a reactivity uncertainty of is +O.0070~k at 14GWD/MTU and is bounding.

Resolution: The reactivity uncertainty of O .0070~k is included in the statistical treatment of tolerances and uncertainties in Table 4.

8) In Section 3.4, Summary of Reactivity 8) The uncertainty in the methodology Calculations, of the criticality analysis, bias (at a 95% probability with a 95%

the reactivity uncertainty associated confidence level) is +O.00963~k and is with the tolerances is listed in Table 4. included in the statistical treatment of The reactivity uncertainties are tolerances and uncertainties.

statistically combined in root mean sum of the squares (RMSS) method. Resolution: The label in Table 4 has Included in the RMSS combination is a been revised to clarify the uncertainty term labeled 'Methodology Bias.' It is in the methodology benchmark bias.

inappropriate to include a bias in the RMSS combination. If that term is removed from the RMSS and applied directly, as is appropriate for a bias, the regulatory requirement is not met.

Whether or not this term is actually a bias or uncertainty is indeterminate from the information provided as neither the term nor the value is repeated elsewhere in the analysis.

The value does not correspond to any value in the Appendix, 'Benchmarking Computer Codes for Calculating the Reactivity State of Spent Fuel Storage Racks, Storage Casks and Transportation Casks.' There is insufficient information to ascertain what the 'Methodology Bias' represents or how it is determined.

9) In Section 3.5, Abnormal/Accident 9) The reference case was analyzed at Conditions, of the criticality analysis, the reduced moderator density two abnormal/accident conditions are corresponding to a pool temperature of 9

determined to have a significant 150°F (0.98 gm/cm3 ) to bound the impact: the misloading of an assembly maximum pool temperature and reduced moderator density corresponding to a full core offload of (increased temperature and voiding 140°F per Section H.1.0 of the NMP1 due to boiling) . The criticality analysis UFSAR. The reactivity effects of a determines that the reduced moderator misloaded or dropped bundle were density is the larger of the two effects. also analyzed at this temperature (and However, as noted previously a portion density) and found to be bounded by of the reduced moderator density the worst case accident scenario of a should be included in the normal loss of pool cooling and boiling .

condition. With a reduced effect of the moderator density abnormal condition, The results are summarized in the misloading may become the Section 3.5 of NET-290-01.

limiting condition, which when combined with an increased nominal

!<eft may result in the regulatory requirement not being met.

10) The criticality analysis includes the Appendix, 'Benchmarking Computer Codes for Calculating the Reactivity State of Spent Fuel Storage Racks, Storage Casks and Transportation Casks.' In this Appendix benchmark calculations are documented which determine computer code biases and uncertainties for KENO V.a, MCNP5, and CASMO-4. KENO V.a and MCNP5 are benchmarked to five criticality experiments which are common to other SFP criticality analyzes and eight which are not.

CASMO-4 is indirectly bench marked to five criticality experiments which are common to other SFP criticality analyzes by comparing the CASMO-4 results to the KENO V.a 'best estimate' results. There are several issues with the analysis in this appendix.

a) The vintage of the reference cited a) The critical experiments can be for the eight unfamiliar criticality found in the following references:

experiments makes determining their validation as appropriate "International Handbook of difficult. Evaluated Criticality Safety Benchmarks," NENNSC/DOC (95)03, September 2001 Edition.

Nuclear Energy Agency; September 2001.

10

The table below cross-references t he critical experiment with experiment as listed in the above Reference.

Experiment IHECSBE (Ref.6)

CSNI1-1 LEU-Therm-001 CSNI1 1 LEU-Therm-007 (case 1)

CSNI1-2-2 LEU-Therm-007 (case 2)

CSNI2-1 LEU-Therm-16 (case 12)

CSNI2-2 LEU-Therm-034 (case 14)

CSNI3-A-1 LEU-Therm-017 (case 3)

CSNI3-A-2 LEU-Therm-040 (case 6)

CSNI3-B-1 LEU-Therm-017 (case 13)

B&WXIII LEU-Therm-051 (case 10)

B&WXIV LEU-Therm-051 (case 12)

B&WXV LEU-Therm-051 (case 13)

B&WXVII LEU-Therm-051 (case 15)

B&WXIX LEU-Therm-051 (case 17)

The NRC approved the same 13 critical benchmarks in the SER of License Amendment 227 for Indian Point 2 on 5/29/2002.

"This experimental data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions that include close proximity storage and strong neutron absorbers."

b) The statistical method used for b) The results of the 13 critical determining the biases and benchmark experiments passed uncertainties is dependent on a the tests for normality using the normal distribution and a large Anderson-Darling, Cramer-Von sample size . Neither has been Mises and Kolmogorov-Smirnoff t confirmed for the analysis ests for normality. Table A-7 of presented. That the KENO V.a Experimental Statistics (Mary neutron histories result in a normal Gibbons Natrella, Experimental distribution for a particular Statistics, National Bureau of computer run, does not necessarily 11

mean that the comparison of the Standards Handbook 91, August different experiments is also a 1, 1963) contains one-sided normal distribution. Even is the tolerance factors for normal comparisons follow a normal distributions for as few as 3 distribution, sample sizes of 13 and samples.

five are too small to use the simple variance weighted mean and variance weighted standard deviation for determining the biases and uncertainties.

c) Using the data provided in Tables c) This cannot be specifically 3-2 and 3-3 and the description of addressed without reviewing the the treatment of the data, a staff staff Excel spreadsheet.

verification using an MS Excel spreadsheet produces different results than those indicated by the Tables .

d) In the Appendix Section 3.1, d) The values contained in Section 4.0 Benchmarking of SCALE-5 and of Appendix 1 are 95%

MCNP5, it states "For SCALE-5, probability/95% confidence level the resulting mean bias for this biases which is the best estimate library is -0.00782 +/- 0.00361. For bias plus the standard deviation with MCNP5, using the continuous the 1-sided tolerance factor at the energy cross-section library based 95% probability/95% confidence on ENDF/B-VI, the resulting level applied.

variance weighted mean bias is -

0.00574 +/- 0.00509. " In the Appendix Section 4.0, Conclusions, it states "At a 95% probability /

95% confidence level, the computed bias for SCALE-5 and MCNP5 are -0.01381 and -

0.01460, respectively." It is not apparent what changed between Section 3.1 and Section 4.0.

e) There are proprietary Appendices e) Appendices A, B, C and 0 of to the Appendix that were not Appendix A (NET-901-02-05) to provided . The proprietary NET-290-1 are not necessary for Appendices are required for the verification of the Benchmarking technical review. biases for the computer codes used in this analysis. Appendices A and B of NET-901-02-05 are directory listings of the as received files (executables, sample input/output, libraries, etc.) of SCALE5 and 12

MCNP5 as received on CD-ROM from RSICC. Appendix C contains the input and output files of the 13 criticals. Appendix 0 contains the CD-ROMs with the as-received versions of the code. These can be obtained from the Radiation Science Information Computation Center.

MCNP files are not relevant as the code was not used in this analysis.

13

References

1. Constellation Energy, Nine Mile Point Nuclear Station, Keith J. Poison, Vice President-Nine Mile Point, to USNRC document control desk, re: ANine Mile Point Nuclear Station Unit No.1, Docket No. 50-220, License Amendment Request Pursuant to 10 CFR 50.90:

Elimination of Credit for Boraflex in Spent Fuel Pool Criticality Analyses - Technical Specification Section 5.5, Storage of Unirradiated and Spent Fuel,@ April 3, 2008. (ADAMS ML081050374).

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

August 19, 1998. (ADAMS ML003728001)

3. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees from B. K.

Grimes. OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications. April 14, 1978, as amended by letter dated January 18,1979.

4. U.S. NRC letter to Duke Energy Corporation "FINAL SAFETY EVALUATION FOR DUKE TOPICAL REPORT DPC-NE-1005P, "NUCLEAR DESIGN METHODOLOGY USING CASM0-4/SIMULATE-3 MOX" dated August 20,2004 (ADAMS ML042370178) 14