ML081820131
| 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.
14
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.
15
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.
16
Appendix 2:
Response to NRC Acceptance Review Questions
NRC Comment Appendix 2
Response
- 1) The licensee's evaluation of the change indicates that eliminating the Boraflex credit will not adversely affect various aspects such as material considerations, solid radioactive waste generation, or occupational exposure.
However, there is no discussion on the indefinite exposure of the SFP &
supporting components to the byproducts associated with the Boraflex degradation. There is no discussion concerning the potential increase of liquid radioactive waste generation or occupational exposure associated with removing those byproducts from the SFP. There is no discussion as to whether or not the indefinite presence of those byproducts materially affects any explicit or implicit assumption in a previous licensing activity.
2)
In the Section 2.0, Analytical Methods and Assumptions, the criticality analysis indicates CASMO-4 was used
- 1) The byproducts from degradation of the of the Boraflex polymer are boron carbide, amorphous silica (soluble) and very fine crystalline silica (insoluble), all of which can leave the Boraflex panels and enter the spent fuel pool water over time.
The silica concentration is sampled monthly per procedure. The normal mode of the spent fuel pool cleanup system keeps the silica level in the pool low. The insoluble silica and boron carbide deposits on horizontal surfaces in the spent fuel pool, primarily on the spent fuel pool floor. If the deposition of these substances impacts visibility (e.g.,
visibility of bundle handle identification labels), vacuuming can be performed using a standard tri-nuke vacuum. The 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) The CASMO and KENO methods used in this analysis have been recommended for criticality analysis per the Kopp 1
for both the standard cold core geometry (SCCG) and SFP storage rack analysis. CASMO-4 is unable to explicitly model the asymmetries of the NMP U1 SFP storage racks. The criticality analysis indicates approximations are made in CASMO-4 to accommodate the asymmetry and that the results are only used in a
'relative' basis. However, there is no discussion on how the CASMO-4 approximations affect the results.
Therefore, there is no assurance that the results are appropriate for use on any basis, even a 'relative' basis. The staff has only found one licensing activity on the docket regarding CASMO-4. The staff has licensed Duke Energy Corporation (DEC) topical report DPC-NE-1005P for core reload analyzes for Catawba Nuclear Station Units 1 and 2 and McGuire Nuclear Station Units 1 (Reference 4).
Since the CASMO-4 user's manual, which is referenced in the NMP U1 SFP LAR, is not publicly available, the DEC topical report documents will have to be reviewed to determine if limitations on the use of CASMO-4 in asymmetric configurations were identified or even considered.
Additionally, there is currently a methodology change associated with the DEC topical report being reviewed by the staff. Those documents will also have to be reviewed to determine whether or not the use of CASMO-4 is appropriate in this manner. There is 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
"...conservative assumptions have been applied to the analyses."
Guidance of 8/19/1998. In the report this Guidance has been moved from reference 8 to the list of regulations in Section 1.2.
NETCO's use of these methods have been previously subject to detailed review by the NRC and approved.
Reference 8 has been added to the report and is an SER issued by the NRC.
CASMO has been used in most criticality analyses supporting reracking and spent fuel storage expansion activities since the 1980's. CASMO has been used in numerous spent fuel rack design analyses including the re-rack installation of BORAL racks at Nine Mile Point 1. [Letter from Daryl Hood, USNRC, to John Mueller, NMPC, 6/17/1999, "Issuance of Amendment for Nine Mile Point Nuclear Unit No.1 to Reflect a Planned Modification to Increase the Storage Capacity of the Spent Fuel Pool (TAC No. MA1945)."]
Referring to SCALE (KENO V.a) and CASMO, the above reference (TAC No.
MA1945) concludes, "These codes have been widely used for the analyses of fuel rack reactivity and have been benchmarked against the results from numerous critical experiments."
The KENO V.a verification of the CASMO-4 approximation is given in Section 2.0, page 9.
3)
For the reference analysis (per Kopp Guidance 81/9/981[ 5.A.3d) the temperature and density of the 2
Whether or not they are conservative is not supported with any additional justification. In the case of SFP moderator density assumption it states: "The moderator is assumed to be demineralized water at full water density (t.Oqm/cm")." Subsequent information indicates this assumption to be non-conservative. Water at full density (1.0gm/cm3 ) corresponds to a temperature of approximately 39°F.
This temperature is well below the SFP normal and maximum operating temperatures. Staff guidance in the 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 appear to be incongruent. Both the 3
moderator for the in-rack condition was set at 150°F(O.98 gm/cm3) which bounds the maximum abnormal condition temperature of 140°F for a full core offload as stated in Section H.1.0 of the NMP1 UFSAR.
Additionally, for the limiting lattice k., in the standard cold core geometry (SCCG), the moderator temperature is at 68°F (1.0 qm/crn"),
Resolution: This is clarified in Assumption 4 on Page 10 of NET-290-01.
4)
The end fittings are above and below the active fuel lattice and have no
assumption that it is conservative to model the fuel assembly channeled and the assumption that it is conservative to not model the fuel assembly grids and end fittings appear to be balancing the absorption of structural components against increasing the amount of over moderation. It is not intuitively obvious 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 V.a Reactivity Calculation Comparison, of the criticality analysis, the results of different fuel assembly configurations at zero burnup are compared for both CASMO-4 and KENO V.a in the SCCG. In that discussion a bias for each computer code is cited.
However, those biases are not the same as those calculated in the Appendix, 'Benchmarking Computer Codes for Calculating the Reactivity State of Spent Fuel Storage Racks, Storage Casks and Transportation Casks.' These values are not stated or described elsewhere in the analysis.
Therefore, the use of the biases cited in this section is unsupported by the analysis.
4 influence in the H/U ratio in the lattice region. There is a slight positive reactivity effect due to the water displaced by the low volume spacers.
This represents a very small relative volume of the lattice and is more than offset by negative reactivity effect of parasitic neutron capture in the Inconel grid spacer springs.
5)
The biases in Section 3.1 are best estimate biases and are -
0.00782+/-0.00361 for Keno V.a and-0.01028+/-0.00198 for CASMO-4.
These have been revised in the report to five decimal places and the one sigma standard deviation has been added. The 95/95 uncertainty of 0.00963 is the 1-sigma uncertainty (0.00361) in the bias multiplied by the 1-sided 95/95 tolerance factor of 2.67 for n=13 criticals. This has been included in the statistical treatment of tolerances and uncertainties.
These appear in Table 3-2 and Table 3-3 in the Appendix for KENO V.a and CASMO-4 respectively.
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.
6)
In Section 3.2, Reactivity Calculations, of the criticality analysis, the reactivity equivalent fresh fuel enrichment (REFFE) is determined. The subject fuel assembly designs are depleted in CASMO-4 in the core geometry and a SCCG k, is determined. The REFFE is determined by iterative KENO V.a cases until an equivalent zero gadolinium fresh fuel enrichment is found, the REFFE. There is no discussion of how or if the computer code biases and uncertainties are applied. There is insufficient information in the LAR to begin a review to determine whether or not the limiting REFFE has been determined.
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
'assumed' based on other work performed by the vendor. Whether or not those assumptions are conservative or not is indeterminate, given the information provided.
5 6)
An iterative process is used to determine the REFFE for each fuel type. In this process the best estimate KENO V.a bias (-0.00782) is added to the KENO V.a calculated eigenvalue.
Similarly, the best estimate CASMO-4 bias of -0.01028 is applied to the CASMO-4 calculated eigenvalue.
This discussion has been added to Section 3.2.1 of the report.
" The REFFE was determined by modeling each limiting (with respect to maximum uranium enrichment and minimum gadolinia loading) fuel 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."
a)
The tolerance values used in the analysis are conservative for the following reasons :
- 1. U235 Enrichment - The following information was provided by Niagara Mohawk Power Corporation in letter dated December 21,1995 from Rocco Bianchi (GE Nuclear Energy) to Bart Franey (NMPC):
6 "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.
7
- 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) ofNominal."
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.
b) There is an uncertainty for the flux trap that is created by replacing the Boraflex with SFP water. However, the description of that uncertainty is insufficient to determine whether that includes the tolerance for the stainless steel structural material associated with the Boraflex.
c)
The Methodology Uncertainty is listed as 0.0078 ~l<eff. However, this value is not included in Table 4, which lists the other uncertainties. Additionally, this value is not that determined in the Appendix, 'Benchmarking Computer Codes for Calculating 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 uncertainty is based on an assumed function based on burnup. The assumed function is not validated. The depletion uncertainty is not included in Table
- 4. And not including the depletion uncertainty is justified by citing unquantified conservatisms from the eleven assumptions in the Analytical Methods and Assumptions section.
8 b)
The tolerance effect for the stainless steel structural material associated with the Boraflex material has been assessed. The reactivity effect was found to be negligible.
c)
The Nominal Keff value in Table 4 contains the bias correction of 0.00782.
Resolution: Table 4 has been revised to explicitly list the bias and Keno V.a calculated reference keff separately.
d)
Additional methods for determining the reactivity effect due to uncertainties in burnup have been proposed based on 5% of the reactivity decrement from 0 burnup to the point of peak reactivity (without gadolinia) [Kopp, L1, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants", Proceedings 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
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 impact: the misloading of an assembly and reduced moderator density (increased temperature and voiding due to boiling). The criticality analysis determines that the reduced moderator density is the larger of the two effects.
However, as noted previously a portion of the reduced moderator density should be included in the normal condition. With a reduced effect of the moderator density abnormal condition, the misloading may become the 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 benchmarked 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 for the eight unfamiliar criticality experiments makes determining their validation as appropriate difficult.
10 150°F (0.98 gm/cm3) to bound the maximum pool temperature corresponding to a full core offload of 140°F per Section H.1.0 of the NMP1 UFSAR. The reactivity effects of a misloaded or dropped bundle were also analyzed at this temperature (and density) and found to be bounded by the worst case accident scenario of a loss of pool cooling and boiling.
The results are summarized in Section 3.5 of NET-290-01.
a) The critical experiments can be found in the following references:
"International Handbook of Evaluated Criticality Safety Benchmarks," NENNSC/DOC (95)03, September 2001 Edition.
Nuclear Energy Agency; September 2001.
b) The statistical method used for determining the biases and uncertainties is dependent on a normal distribution and a large sample size. Neither has been confirmed for the analysis presented. That the KENO V.a neutron histories result in a normal distribution for a particular computer run, does not necessarily 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 results of the 13 critical benchmark experiments passed the tests for normality using the Anderson-Darling, Cramer-Von Mises and Kolmogorov-Smirnoff t ests for normality.
Table A-7 of Experimental Statistics (Mary Gibbons Natrella, Experimental Statistics, National Bureau of 11
mean that the comparison of the different experiments is also a normal distribution. Even is the comparisons follow a normal distribution, sample sizes of 13 and 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 3-2 and 3-3 and the description of the treatment of the data, a staff verification using an MS Excel spreadsheet produces different results than those indicated by the Tables.
d)
In the Appendix Section 3.1, Benchmarking of SCALE-5 and MCNP5, it states "For SCALE-5, the resulting mean bias for this library is -0.00782 +/- 0.00361. For MCNP5, using the continuous energy cross-section library based on ENDF/B-VI, the resulting 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 to the Appendix that were not provided. The proprietary Appendices are required for the technical review.
12 Standards Handbook 91, August 1, 1963) contains one-sided tolerance factors for normal distributions for as few as 3 samples.
c)
This cannot be specifically addressed without reviewing the staff Excel spreadsheet.
d)
The values contained in Section 4.0 of Appendix 1 are 95%
probability/95% confidence level biases which is the best estimate bias plus the standard deviation with the 1-sided tolerance factor at the 95% probability/95% confidence level applied.
e)
Appendices A, B, C and 0 of Appendix A (NET-901-02-05) to NET-290-1 are not necessary for verification of the Benchmarking 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
13 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.
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