ML052230371
| ML052230371 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/08/2005 |
| From: | Holtec |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 1104, 2CAN080501 HI-2043262NP, Rev 0 | |
| Download: ML052230371 (72) | |
Text
Attachment 2 2CAN080501 Part 50 Criticality Analysis of the MPC-32 for ANO Unit 2 - Non-Proprietary
U..'
H O LT I N T E R N AT I Holtec Center. 555 Lincoln Drive West. Mariton. NJ 08053 E C 0 N A L Telephone (856) 797- 0900 Fax (856) 797 - 0909 Part 50 Criticality Analysis of The MPC-32 for ANO Unit 2 FOR ANO Holtec Report No: HI-2043262NP Rev 0 Holtec Project No: 1104 Report Class: SAFETY RELATED
Table of Contents
- 1. Introduction.....................................................................................................................1 1.1 Statement of Purpose..........................................
1 1.2 About This Document........................................
l1
- 2. Methodology.........................................
3
- 3. Acceptance Criteria.........................................
3
- 4. Assumptions.........................................
4
- 5. Input Data.........................................
6 5.1 Fuel Assembly and Operational Data..........................................
6 5.2 MPC-32 Basket Data..........................................
6 5.3 Geometric Description of the 3-D MCNP Model.......................................
6
- 6. Computer Codes..........................................
7
- 7. Analysis..........................................
7 7.1 Uncertainties.........................................
9 7.11 Fuel Assembly.........................................
9 7.12 MPC Manufacturing Tolerances
.9 7.13 Fuel Enrichment Tolerance
.9 7.14 Uncertainty in Depletion Calculations
.9 7.15 Eccentric Fuel Assembly Positioning.10 7.16 MCNP Statistical Uncertainties.10 7.2 Off-Normal and Accident Conditions
............................ 10 7.21 Temperature and Density Effects.10 7.22 Dropped Assembly - Horizontal
.1 7.23 Dropped Assembly - Vertical.1 7.24 Accident Resulting in Misalignment of Active Fuel With Poison Material. 1 1 7.25 Accident of a Missloaded Fresh Fuel Assembly.1 7.3 Bumup vs Enrichment Requirements 12
- 8. Computer Files......................................
12
- 9. Conclusion.....................................
13
- 10. References.....................................
13 Appendix A: Benchmark Calculations...................................
A-I Appendix B: List of Approved Computer Programs...................................
B-1 : Letter Dated November 9, 2001 Project No. 1104 Report No. HI-2043262NP Page i Holtec International
- 1. Introduction 1.1 Statement of Purpose This report documents the criticality safety evaluation for the storage of CE 16 x 16 spent nuclear fuel assemblies in a Holtec MPC-32 basket in the ANO spent fuel pool without taking credit for soluble boron under normal operating conditions. The scope of the analysis is limited to fuel assemblies with an initial nominal enrichment of no more than 4.95 v% 235U.
The objective of this analysis is to determine the minimum bumup requirements for initial fuel enrichments necessary to meet the requirements of 10CFR50.68. Since the analysis will not take any credit for soluble boron under normal conditions, it is necessary to demonstrate that keff is less than 0.95 with the MPC flooded with fresh water. The maximum calculated reactivities include a margin for uncertainty in reactivity calculations, including manufacturing tolerances, and are calculated with a 95%
probability at a 95% confidence level [1].
Reactivity effects of accident conditions have also been evaluated to assure that under all credible conditions, the reactivity will not exceed the regulatory limit of 0.95, considering the presence of an acceptable soluble boron level.
1.2 About This Document This work product has been labeled a safety-significant document in Holtec's QA System.
In order to gain acceptance as a safety-significant document in the company's quality assurance system, this document is required to undergo a prescribed review and concurrence process that requires the preparer and reviewer(s) of the document to answer a long list of questions crafted to ensure that the document has been purged of all errors of any material significance. A record of the review and verification activities is maintained in electronic form within the company's network to enable future retrieval and recapitulation of the programmatic acceptance process leading to the acceptance and release of this document under the company's QA system. Among the numerous requirements that a document of this genre must fulfill to muster approval within the company's QA program are:
- The preparer(s) and reviewer(s) are technically qualified to perform their activities per the applicable Holtec Quality Procedure (HQP).
- The input information utilized in the work effort must be drawn from referencable sources. Any assumed input data is so identified.
- All significant assumptions, as applicable, are stated.
Project No. 1104 Report No. HI-2043262NP Page 1 Holtec International
- The analysis methodology is consistent with the physics of the problem.
- Any computer code and its specific versions that may be used in this work have been formally admitted for use within the company's QA system.
- The format and content of the document is in accordance with the applicable Holtec quality procedure.
- The material content of this document is understandable to a reader with the requisite academic training and experience in the underlying technical disciplines.
Once a safety significant document produced under the company's QA System completes its review and certification cycle, it should be free of any materially significant error and should not require a revision unless its scope of treatment needs to be altered. Except for regulatory interface documents (i.e., those that are submitted to the NRC in support of a license amendment and request), revisions to Holtec safety-significant documents to amend grammar, to improve diction, or to add trivial calculations are made only if such editorial changes are warranted to prevent erroneous conclusions from being inferred by the reader. In other words, the focus in the preparation of this document is to ensure accuracy of the technical content rather than the cosmetics of presentation.
In accordance with the foregoing, this Calculation Package has been prepared pursuant to the provisions of Holtec Quality Procedures HQP 3.0 and 3.2, which require that all analyses utilized in support of the design of a safety-related or important-to-safety structure, component, or system be fully documented such that the analyses can be reproduced at any time in the future by a specialist trained in the discipline(s) involved.
HQP 3.2 sets down a rigid format structure for the content and organization of Calculation Packages that are intended to create a document which is complete in terms of the exhaustiveness of content. Calculation Packages, however, may lack the narrational smoothness of a Technical Report, and are not intended to serve as a Technical Report.
Because of its function as a repository of all analyses performed on the subject of its scope, this document is typically revised only if an error is discovered in the computations or the equipment design is modified. Additional analyses in the future may be added as numbered supplements to this Package. Each time a supplement is added or the existing material is revised, the revision status of this Package is advanced to the next number and the Table of Contents is amended. They are shared with a client only under strict controls on their use and dissemination.
This Calculation Package will be saved as a Permanent Record under the company's QA System.
Project No. 1104 Report No. HI-2043262NP Page 2 Holtec International
- 2. Methodology
- 3. Acceptance Criteria The objective of this analysis is to ensure that the effective neutron multiplication factor (kcff) is less than or equal to 0.95 with the MPC-32 basket fully loaded with fuel of the highest permissible reactivity. For normal conditions, the MPC-32 basket is analyzed as flooded with fresh water, and for accident conditions, credit for soluble boron is Project No. 1104 Report No. HI-2043262NP Holtec International Page 3
permitted. The maximum calculated reactivities include a margin for uncertainty in reac-tivity calculations, manufacturing tolerances, and temperature effects, and are calculated with a 95% probability at a 95% confidence level [1].
Applicable codes, standards, and regulations or pertinent sections thereof, 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."
- USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev. 3 - July 1981.
- USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979.
- 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.
- USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed), December 1981.
- ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.
- Code of Federal Regulation I OCFR50.68, Criticality Accident Requirements (for soluble boron)
- 4. Assumptions The following assumptions were employed in the analysis:
- 1) All depletion calculations are performed with 3 years of cooling time credited, which bounds the minimum cooling time allowable for spent fuel storage in the MPC-32.
- 2) Neutron absorption in minor structural members is neglected, i.e., spacer grids are replaced by water.
Project No. 1104 Report No. HI-2043262NP Page 4 Holtec International
- 3) {
}
- 4) {
}
- 5) {
}
- 6) The MPC-32 and HI-STAR or HI-TRAC is located in the cask pit area, which is assumed to be neutronically isolated from the rest of the spent fuel pool because the loaded fuel will be at least 12 inches from fuel stored in the adjacent racks.
Therefore interfaces need not be considered.
- 7) {
}
- 8) {
}
- 9) {
- 10) {
}
}
Additional assumptions regarding the computer models are provided in Section 5.4 of this report.
Project No. 1104 Report No. HI-2043262NP Holtec International Page 5
- 5. Input Data 5.1 Fuel Assembly and Operational Data The core operating temperatures (fuel and moderator), average soluble boron level, and power density in the depletion calculations use the same values as the depletion calculation (CASMO) inputs used in [10].
The fuel assembly dimensions and axial burnup profile are taken from [9] with the appropriate assumptions in Section 4 applied.
5.2 MPC-32 Basket Data The MPC-32 basket geometry model is based on [11] with the appropriate assumptions from Section 4 applied and the geometric modeling approach described in Section 5.3 below. Material composition of the stainless steel material is taken from [8]. The modeled composition of the MPC-32 basket poison material is described in assumption 7 in Section 4 of this report.
5.3 Geometric Description of the 3-D MCNP Model The criticality calculation adequately represents an MPC-32 loaded into either a HI-TRAC or HI-STAR overpack, however, the system is not modeled exactly. {
I) {
}
- 2) {
}
- 3) {
}
- 4) {
}
- 5) {
I Project No. 1104 Report No. HI-2043262NP Holtec International Page 6
- 6) {
}
- 7) {
}
- 6. Computer Codes The following computer codes were used during this analysis.
- MCNP4a [2] 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 MPC. MCNP4a was run on the PCs at Holtec.
- CASMO-4, Version 2.05.03 [3-5] is a two-dimensional multigroup transport theory code developed by Studsvik of Sweden. CASMO-4 performs cell criticality calculations and burnup. CASMO-4 has the capability of analytically restarting burned fuel assemblies in an infinite representation of the MPC-32 configuration.
This code was used to determine the isotopic composition of the fuel, the reactivity effects of tolerances and fuel depletion, and was used in various studies. The CASMO-4 code was run on a PC at Holtec.
- 7. Analysis This section describes the calculations that were used to determine the acceptable storage criteria for the MPC-32. In addition, this section discusses the possible abnormal and accident conditions.
Unless otherwise stated, all calculations assumed nominal characteristics for the fuel and the fuel storage cells. {
}
}
{
Project No. 1104 Report No. HI-2043262NP Holtec International Page 7
}
Initially, fuel loaded into the reactor will bum with a slightly skewed cosine power distribution. As burnup progresses, the burnup distribution will tend to flatten, becoming more highly burned in the central regions than in the upper and lower ends. At high burnup, the more reactive fuel near the ends of the fuel assembly (less than average burnup) occurs in regions of lower reactivity worth due to neutron leakage.
Consequently, it would be expected that over most of the burnup history, distributed bumup fuel assemblies would exhibit a slightly lower reactivity than that calculated for the average burnup. As burnup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced burnup.
Generic analytic results of the axial burnup effect for assemblies without axial blankets have been provided by Turner [7] based upon calculated and measured axial bumup distributions. These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnup, becoming positive at burnups greater than about 30 GWD/MTU. The trends observed in [7] suggest the possibility of a small positive reactivity effect above 30 GWD/MTU increasing to slightly over 1% Ak at 40 GWD/MTU. {
Pool water temperature effects on reactivity have been calculated with CASMO-4 and the results are presented in Table 1. The results show that the spent fuel pool temperature coefficient of reactivity is negative, i.e. a lower temperature results in a higher reactivity.
The MCNP code bias effect determined in Appendix A is also applied directly in each of the final kff calculations as a bias.
The maximum allowable value for the calculated keff at each enrichmentlburnup combination is summarized in Table 7 and is determined by the following:
Max keff = Regulatory Limit - biases - statistically combined uncertainties Where the regulatory limit = 0.95 for all cases.
Project No. 1104 Report No. HI-2043262NP Page 8 Holtec International
7.1 Uncertainties The uncertainties described in the subsections below are statistically combined via square root of the sum of the squares. The statistical combination of uncertainties plus the temperature and code biases are added to the MCNP calculated keff at each bumup/enrichment combination to determine the maximum kff. The minimum required burnup at each enrichment interval that meets the acceptance criteria is then used to generate the bumup vs enrichment relationship.
7.11 Fuel Assembly ANO Unit 2 only uses a CE 16 x 16 assembly, which is explicitly modeled in this analysis. {
}
7.12 MPC Manufacturing Tolerances 7.13 Fuel Enrichment Tolerance
{
}
7.14 Uncertainty in Depletion Calculations Since critical experiment data with spent fuel is not available for determining the uncertainty in bumup-dependent reactivity calculations, an allowance for uncertainty in reactivity was assigned based upon other considerations. Assuming the uncertainty in depletion calculations is less than 5% of the total reactivity decrement, a bumup dependent uncertainty in reactivity for bumup calculations may be assigned. {
Project No. 1104 Report No. HI-2043262NP Holtec International Page 9
}
7.15 Eccentric Fuel Assembly Positioning The base criticality calculations assume that the fuel assemblies are centered within their respective basket cells. To account for shifting of assemblies within the cell that results in a more reactive configuration, two additional MCNP calculations are performed to obtain an eccentric positioning uncertainty. {
7.16 MCNP Statistical Uncertainties
}
7.2 Off-Normal andAccident Conditions The effects on reactivity of credible abnormal and accident conditions are examined in this section. None of the abnormal or accident conditions that have been identified as credible cause the reactivity of the storage racks to exceed the limiting reactivity value of kerr = 0.95, considering the presence of soluble boron. The double contingency principle of ANSI N16.1-1975 (and the USNRC letter of April 1978) specifies that it shall require at least two unlikely independent and concurrent events to produce a criticality accident.
This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.
7.21 Temperature and Density Effects Water density and temperature effects on reactivity have been calculated {
}. The results show that the spent fuel pool temperature coefficient of reactivity is negative, and that introducing voids in the water internal to the storage cell (to simulate boiling) further decreases reactivity. Therefore the maximum density of water (1.0 g/cc) is used in this analysis.
Project No. 1104 Report No. HI-2043262NP Holtec International Page 10
7.22 Dropped Assembly - Horizontal For the case in which a fuel assembly is assumed to be dropped on top of an MPC-32, {
7 7.23 Dropped Assembly - Vertical An event of a vertical drop accident resulting in an assembly leaning immediately adjacent to the HI-TRAC or HI-STAR would have an insignificant effect on reactivity
{}
7.24 Accident Resulting in Misalignment of Active Fuel With Poison Material Any event resulting in misalignment of the active fuel region with the basket poison material is bound by an extremely conservative accident analysis {
7.25 Accident of a Missloaded Fresh Fuel Assembly The misplacement of a fresh unburned fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (kJff of 0.95). This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (4.95 wt%) were to be inadvertently misloaded into a storage cell intended for spent fuel in the center of the basket.
{
Project No. 1104 Report No. HI-2043262NP Holtec International Page 11
7.3 Burnup vs Enrichment Requirements Calculations were performed to determine acceptable minimum burnups for selected initial 235U enrichments from 2.5wt% to 4.95wt%. The burnup vs enrichment values are tabulated in Table 8, and are ploted in Figure 1. Each result is acceptable per the limit in Table 7. A bounding linear equation was established based on this data yielding the following:
BU = 9E-16 Where BU = The minimum required burnup in GWd/MTU And E = Initial assembly enrichment in wt% 235U
- 8. Computer Files All input files for the calculations are stored in the directory {
}
and its subdirectories on the Holtec server. The input file names and descriptions of the
{
} calculations are listed in the table below.
File name Code Description
{J
{2.2%
enriched fresh fuel
{
}
l{
74.95% enriched fresh fuel, used for depletion uncertainty
{l
{
7Accident- {
) case with all poison replaced with 950 ppm sb water X
l7Accident - {
) case with missload of fresh assembly 400 ppm sb
{
}
{
}
Eccentric positioning- {
) case with assemblies shifted towards center of basket
{
}
Eccentric positioning- {
} case with assemblies shifted towards periphery of basket.
These files all use the convention {
}
Where {
}=enrichment (x 0.1%)*
} {
{
}=bunup (in GNVd/MTU)
{
}=cooling time (in years)
Tolerance and temperature reactivity effects
...A.....
4.Depletion 2.5% enriched fuel
{~ }Depletion 3.0% enriched fuel Depletion 3.5% enriched fuel
{
{j}
f Depletion 4.0% enriched fuel 4
L....
4Depletion 4.5% enriched fuel I
4 1{
}Depletion 4.95% enriched fuel ProjectNo. 1104.
Report No. HI-2043262NP Holtec International Page 12
- 9. Conclusion This report documents the criticality analysis for the storage of CE 16x16 PWR spent nuclear fuel in the Holtec MPC-32 in the ANO Unit 2 spent fuel pool with initial enrichments up to 4.95 wt% 235U. The calculation that determines the maximum permissible kerr from MCNP is provided in Table 7. MCNP results and corresponding minimum burnup requirements were determined at incremental enrichments and are provided in Table 8. Each enrichment case meets the defined acceptance criteria, as summarized in Table 7. A plot and corresponding bounding linear equation was established based on the results in Table 8 and is shown in Figure 1.
The following bumup vs enrichment relationship may be used:
BU= 9E - 16 Where BU = The minimum required burnup in GWd/MTU And E = Initial assembly enrichment in wt% 235U The effects of postulated accident scenarios were also evaluated, where the most limiting case of all poison plates being replaced with water yielded a required a soluble boron concentration of 950 ppm. This is acceptable since it is well below the normal operating level in the ANO Unit 2 spent fuel pool. Results for this accident case and a less limiting case of a missloaded fresh assembly are summarized in Tables 9 and 10 respectively.
1 0. References 1
- 1. M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.
- 2. J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-12625, Los Alamos National Laboratory (1993).
Note: The revision status of Holtec documents cited above is subject to updates as the project progresses.
This document will be revised if a revision to any of the above-referenced Holtec work materially affects the instructions, results, conclusions or analyses contained in this document. Otherwise, a revision to this document will not be made and the latest revision of the referenced Holtec documents shall be assumed to supercede the revision numbers cited above. The Holtec Project Manager bears the undivided responsibility to ensure that there is no intra-document conflict with respect to the information contained in all Holtec generated documents on a safety-significant project. The latest revision number of all documents produced by Holtec International in a safety significant project is readily available from the company's electronic network.
Project No. 1104 Report No. HI-2043262NP Page 13 Holtec International
- 3. M. Edenius, K. Ekberg, B.H. Forssen, and D. Knott, "CASMO-4 A Fuel Assembly Burnup Program User's Manual," Studsvik/SOA-95/1, Studsvik of America, Inc. and Studsvik Core Analysis AB (proprietary).
- 4. D. Knott, "CASMO-4 Benchmark Against Critical Experiments", SOA-94/13, Studsvik of America, Inc., (proprietary).
- 5. D. Knott, "CASMO-4 Benchmark Against MCNP," SOA-94/12, Studsvik of America, Inc., (proprietary).
- 6. Not Used.
- 7. S.E. Turner, "Uncertainty Analysis - Burnup Distributions", presented at the DOE/SANDIA Technical Meeting on Fuel Burnup Credit, Special Session, ANS/ENS Conference, Washington, D.C., November 2, 1988.
- 8.
I11-2012771 Rev 6, HI-STAR 100 and HI-STORM 10O Additional Criticality Calculations, Holtcc International.
- 9. CEO 2001-00284, File No: 104-35, 204-35. Letter from F.H. Smith to J.S.
Rowe dated November 9, 2001. (Included as Attachment I in this report.)
- 10. IHI-2022864 Rev 5, Criticality Safety Evaluation of the ANO Unit 2 Spent Fuel Pool Storage Racks, Holtec International.
- 11. Holtec Drawing 3752 Rev. 9, MPC-32 Fuel Basket, Holtec International.
- 12. Table Al.18, SA 480, ASME Section II, 1998.
Project No. 1104 Report No. HI-2043262NP Holtec International Page 14
Table 1: Temperature Effects ({
})
Temp (K) ki.,r 273 1.15221 277 1.15187 300 1.14917 313 1.14712 343 7
1.14153 373 7
1.13470 :
Table 2: MPC Manufacturing Tolerance Results ({
})
Case kinf A kinf Reference 1.14917 n/a min cell ID, red pitch 1.15249
-0.00332 max cell ID, Inc pitch 1.14525 0.00392 inc box thick, red cell ID 1.15316
-0.00399 red box thick, inc cell ID 1.14483 0.00434 Note: Negative values indicate an increase in reactivity condition since Alqnt= Ref-Tolerance Case Table 3: Water Temperature And Density Effects ({
})
Temp (K)
Void Fraction kinf 300 0.0 1.14917 373 0.1 1.11284 Table 4: Fuel Enrichment Uncertainty ({
})
I kinf (5.0 wt/o... U)
I kinf (4.95wt% 1j5U)
I A knfl 1
1.15118 1
1.14917 1
0.00201 Table 5: Depletion Uncertainty ({
})
Maximum Maximum Fresh Fresh Depletion Bumup Burnup Fuel Fuel Uncertainty Case krff Case kff (5% Akwfr)
{
0.93180
{
1.12600 0.00971 ProjectNo. 1104 Report No. HI-2043262NP Holtec International Page 15
Table 6: Eccentric Positioning ({
})
Reference Reference Eccentric Eccentric Akff Case ker ffCase kerffr
{
0.93180 0.93683 0.00503 Table 7: Maximum Allowable Calculated keff Regulaory Limiting keff 0.95 Uncertainties Bias Uncertainty (95%/95%)
0.00 11 Calculational Statistics (95%/95%, 2.Oxc) 0.0016 Min Cell ID, Reduced Pitch 0.0033 Increased box thickness, Reduced ID 0.0040 Fuel Enrichment Tolerance 0.0020 Depletion 0.0097 Eccentric Positioning 0.0050 Statistical Combination of Uncertainties 0.0124 Biases Calculational Bias (see Appendix A) 0.0009 Temperature Ak (From 300K to 273K) 0.0030 Max Allowable keff (Regulatory Limiting 0.9337 kerf-biases-combined uncertainties)
Project No. 1104 Report No. HI-2043262NP Holtec International Page 16
Table 8: {
} Burnup vs Enrichment Results File Name keff Initial Enrichment Burnup (235U wt%)
(GWd/MTU)
{
}
0.93299 2.2 0
{
0.92027 2.5 5
{
}
0.92200 3.0 10
( {
0.92485 3.5 15
{
}
0.93106 4.0 19
{
}
0.92891 4.5 24
{
}
0.93180 4.95 28 Table 9: Summary of Poison Replaced With Borated Water Accident File Name
{
}
Initial Enrichment (wt% o.U) 4.95 Bumup (GWd/MTU) 28 Soluble Boron Level (ppm) 950 Calculated klrf ({
1)0.93234 Table 10: Summary of Missloaded Fresh Assembly Accident File Name
{
}
Initial Enrichment (wt%.. U) 4.95 Burnup (GWd/MTU) 28 Soluble Boron Level (ppm) 400 Calculated keff ({
})
0.91704 Project No. 1104 Report No. HI-2043262NP Holtec International Page 17
3 I.-
1 CG
!5 5-5 -
0 1
2 3
Initial Enrichment (wt% U.235) 4 5
6 Figure 1: Plot of Burnup vs Enrichment With Bounding Linear Equation Project No. 1104 Report No. HI-2043262NP Holtec International Page 18
Appendix A Benchmark Calculations (total number of pages: 26 including this pagW)
Note: because this appendix was taken from a different report, the next page is labeled "Appendix 4A, Page 1".
Project No. 1104 Report No. HI-2043262NP Page A-I Holtec International
i APPENDIX 4A: BENCHMARK CALCULATIONS 4A.1 INTRODUCfl N
MAND
SUMMARY
Benchmark calculations have been made on selected critical experiments, chosen, in so far as possible, to bound the range of variables in the rack designs. Two independent methods of analysis were used, differing in cross section libraries and in the treatment of the cross sections. MCNP4a [4A.1] is a continuous energy Monte Carlo code and KEN05a [4A.2]
uses group-dependent cross sections. For the KEN05a analyses reported here, the 238-group library was chosen, processed through the NITAWL-Il [4A.2] program to create a working library and to account for resonance self-shielding in uranium-238 (Nordheim integral treatment). The 238 group library was chosen to avoid or minimize the errorst (trends) that have been reported (e.g., E4A.3 through 4A.5]) for calculations with collapsed cross section sets.
In rack designs, the three most significant parameters affecting criticality are (1) the fuel enrichment, (2) the `0B loading in the neutron absorber, and (3) the lattice spacing (or water-gap thickness if a flux-trap design is used). Other parameters, within the normal range of rack and fuel designs, have a smaller effect, but are also included in the analyses.
Table 4A.1 summarizes results of the benchmark calculations for all cases selected and analyzed, as referenced in the table. The effect of the major variables are discussed in subsequent iections below. It is important to note that there is obviously considerable overlap in parameters since it is not possible to vary a single parameter and maintain criticality; some other parameter or parameters must be concurrently varied to maintain criticality.
One possible way of representing the data is through a spectrum index that incorporates all of the variations in parameters. KEN05a computes and prints the "energy of the average lethargy causing fission" (EALF). In MCNP4a, by utilizing the tally option with the identical 238-group energy structure as in KEN05a, the number of fissions in each group may be collected and the EALF determined (post-processing).
t Small but observable trends (errors) have been reported for calculations with the 27-group and 44-group collapsed libraries. These errors are probably due to the use of a single collapsing spectrum when the spectrum should be different for the various cases analyzed, as evidenced by the spectrum indices.
Holtec International.
Appendix 4A, Page 1
Figures 4A. 1 and 4A.2 show the calculated kh for the benchmark critical experiments as a function of the EALF for MCNP4a and KENO5a, respectively (U02 fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental errort in performing the critical experiments within each laboratory, as well as between the various testing laboratories. The B&W critical experiments show a larger experimental error than the PNL criticals. This would be expected since the B&W criticals encompass a greater range of critical parameters than the PNL criticals.
Linear regression analysis of the data in Figures 4A. 1 and 4A.2 show that there are no trends, as evidenced by very low values of the correlation coefficient (0.13 for MCNP4a and 0.21 for KEN05a). The total bias (systematic error, or mean of the deviation from a kff of exactly 1.000) for the two methods of analysis are shown in the table below.
Calculational Bias of MCNP4a and KEN05a MCNP4a
- 0.0009i0.0011 KEN05a 0.0030+/-0.0012 l
The bias and standard error of the bias were derived directly from the calculated Kff values in Table 4A. 1 using the following equations", with the standard error multiplied by the one-sided K-factor for 95 % probability at the 95% confidence level from NBS Handbook 91 [4A. 18] (for the number of cases analyzed, the K-factor is -2.05 or slightly more than 2).
k 1
ki (4A.1) n I
t A classical example of experimental error is the corrected enrichment in the PNL experiments, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods.
tt These equations may be found in any standard text on statistics, for example, reference
[4A.6] (or the MCNP4a manual) and is the same methodology used in MCNP4a and in KENO5a.
Holtec International Appendix 4A, Page 2 Holtec International Appendix 4A, Page 2
-2 (Y k,2 I (4A.2)
=
n (n-1)
Bias = (1-k') i K a; (4A.3) where k; are the calculated reactivities of n critical experiments; oa is the unbiased estimator of the standard deviation of the mean (also called the standard error of the bias (mean)); K is the one-sided multiplier for 95 % probability at the 95% confidence level (NBS Handbook 91 [4A.18]).
Formula 4.A.3 is based on the methodology of the National Bureau of Standards (now NIST) and is used to calculate the values presented on page 4.A-2. The first portion of the equation, ( 1-I ), is the actual bias which is added to the MCNP4a and KENO5a results.
The second term, Kai, is the uncertainty or standard error associated with the bias. The K values used were obtained from the National Bureau of Standards Handbook 91 and are for one-sided statistical tolerance limits for 95 % probability at the 95 % confidence level. The actual K values for the 56 critical experiments evaluated with MCNP4a and the 53 critical experiments evaluated with KENO5a are 2.04 and 2.05, respectively.
The bias values are used to evaluate lte maximum k2rt values for the rack designs.
KENO5a has a slightly larger systematic error than MCNP4a, but both result in greater precision than published data [4A.3 through 4A.5] would indicate for collapsed cross section sets in KENO5a (SCALE) calculations.
4A.2 Effect of Enrichment The benchmark critical experiments include those with enrichments ranging from 2.46 w/o to 5.74 w/o and therefore span the enrichment range for rack designs. Figures 4A.3 and 4A.4 show the calculated kff values (Table 4A.1) as a function of the fuel enrichment reported for the critical experiments. Linear regression analyses for these data confirms that there are no trends, as indicated by low values of the correlation coefficients (0.03 for MCNP4a and 0.38 for KENO5a). Thus, there are no corrections to the bias for the various enrichments.
Holtec International.
Appendix 4A, Page 3 Holtec International Appendix 4A, Page 3
As further confirmation of the absence of any trends with enrichment, a typical configuration was calculated with both MCNP4a and KEN05a for various enrichments.
The cross-comparison of calculations with codes of comparable sophistication is suggested in Reg. Guide 3.41. Results of this comparison, shown in Table 4A.2 and Figure 4A.5, confirm no significant difference in the calculated values of kff for the two independent codes as evidenced by the 45° slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence of an enrichment effect (trend) in the bias.
4A.3 Effect of 10B Loading Several laboratories have performed critical experiments with a variety of thin absorber panels similar to the Boral panels in the rack designs. Of these critical experiments, those performed by B&W are the most representative of the rack designs. PNL has also made some measurements with absorber plates, but, with one exception (a flux-trap experiment),
the reactivity worth of the absorbers in the PNL tests is very loW and any significant errors that might exist in the treatment of strong thin absorbers could not be revealed.
Table 4A.3 lists the subset of experiments using thin neutron absorbers (from Table 4A. 1) and shows the reactivity worth (Ak) of the absorber.t No trends with reactivity worth of the absorber are evident, although based on the calculations shown in Table 4A.3, some of the B&W critical experiments seem to have unusually large experimental errors. B&W made an effort to report some of their experimental errors. Other laboratories did not evaluate their experimental errors.
To further confirm the absence of a significant trend with `0B concentration in the absorber, a cross-comparison was made with MCNP4a and KENOSa (as suggested in Reg.
Guide 3.41). Results are shown in Figure 4A.6 and Table 4A.4 for a typical geometry.
These data substantiate the absence of any error (trend) in either of the two codes for the conditions analyzed (data points fall on a 45° line, within an expected 95% probability limit).
t The reactivity worth of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental (Ak) change in reactivity due to the absorber.
Holtec Intemational Appendix 4A, Page 4 Holtec International' Appendix 4A, Page 4
4A.4 Miscellaneous and Minor Parameters 4A.4. 1 Reflector Material and Spacings PNL has performed a number of critical experiments with thick steel and lead reflectors.t Analysis of these critical experiments are listed in Table 4A.5 (subset of data in Table 4A. 1). There appears to be a small tendency toward overprediction of K. at the lower spacing, although there are an insufficient number of data points in each series to allow a quantitative determination of any trends. The tendency toward overprediction at close spacing means that the rack calculations may be slightly more conservative than otherwise.
4A.4.2 Fuel Pellet Diameter and Lattice Pitch The critical experiments selected for analysis cover a range of fuel pellet diameters from 0.3 11 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches. In the rack designs, the fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.496 to 0.580 inch lattice spacing) for PWR fuel and from 0.3224 to 0.494 inches O.D. (0.488 to 0.740 inch lattice spacing) for DWR fuel. Thus, the critical experiments analyzed provide a reasonable representation of power reactor fuel. Based on the data in Table 4A.1, there does not appear to be any observable trend with either fuel pellet diameter or lattice pitch, at least over the range of the critical experiments applicable to rack designs.
4A.4.3 Soluble Boron Concentration Mffects Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KENO5a) calculations are shown in Table 4A.6. Analyses of the very high boron concentration experiments (> 1300 ppm) show a tendency to slightly overpredict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the racks with higher soluble boron concentrations could be slightly conservative.
t Parallel experiments with a depleted uranium reflector were also performed but not included in the present analysis since they are not pertinent to the Holtec rack design.
Holtec International Appendix 4A, Page 5 Holtec International:
Appendix 4A, Page 5
4A;5 MOX Fuel The number of critical experiments with PuO2 bearing fuel (MOX) is more limited than for U0 2 fuel. However, a number of MOX critical experiments have been analyzed and the results are shown in Table 4A.7. Results of these analyses are generally above a kff of 1.00, indicating that when Pu is present, both MCNP4a and KEN05a overpredict the reactivity. This may indicate that calculation for MOX fuel will be expected to be conservative, especially with MCNP4a. It may be noted that for the larger lattice spacings, the KEN05a calculated reactivities are below 1.00, suggesting that a small trend may exist with KEN05a. It is also possible that the overprediction in kr, for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and Am-241 growth. This possibility is supported by the consistency in calculated kf over a wide range of the spectral index (energy of the average lethargy causing fission).
Holtec International-Appendix 4A, Page 6
4A.6 References (4A.1] J.F. Briesmeister, Ed., 'MCNP4a - A General Monte Carlo N-Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1993).
[4A.2] SCALE 4.3, "A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", NUREG-0200 (ORNL-NUREG-CSD-21U21R5, Revision 5, Oak Ridge National Laboratory, September 1995.
[4A.3] M.D. DeHart and S.M. Bowman, "Validation of the SCALE Broad Structure 44-G Group ENDF/B-Y Cross-Section Library for Use in Criticality Safety Analyses", NUREGICR-6102 (ORNL/TM-12460)
Oak Ridge National Laboratory, September 1994.
[4A.4] W.C. Jordan et al., 'Validation of KENOV.a", CSD/TM-238, Martin Marietta. Energy Systems, Inc., Oak Ridge National Laboratory, December 1986.
[4A.5]
O.W. Hermann et al., 'Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analysist, ORNL-TM-12667, Oak Ridge National Laboratory, undated.
[4A.6] R.J. Larsen and M.L. Marx, An Introduction to Mathematical Statistics and its Applications, Prentice-Hall, 1986.
[4A.7] M.N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, Babcock and Wilcox Company, July 1979.
[4A.8]
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.
[4A.9]
L.W. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-1810, Babcock and Wilcox Company, April 1984.
Holtec International Appendix 4A, Page 7 Holtec International Appendix 4A, Page 7
14A.10] J.C. Manaranche et al., 'Dissolution and Storage Experimental Program with 4.75 w/o Enriched Uranium-Oxide Rods,' Trans.
Am. Nucl. Soc. 33: 362-364 (1979).
[4A.11] S.R. Bierman and E.D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o 2`U Enriched UO, Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981.
[4A.12] S.R. Bierman et al., Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o 'U Enriched U02 Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December, 1981.
[4A. 13] S.R. Bierman et al., Critical Separation Between Subcritical Clusters of 4.31 w/o 235U Enriched U02 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977. *
[4A.14] S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990,
[4A.15] B.M. Durst et al., Critical Experiments with 4.31 wt % 35U Enriched U02 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982.
[4A. 16] S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1981.
[4A.17J 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.
[4A. 18] M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.
Holtec Intemational Appendix 4A, Page 8
Table 4A.1 Summary of Criticality Benchmark Calculations Cankilted k -
VALF t (eVI
__ i.. %,
.^t,,
REN05a MCNP4a Reference IdentifIcation Enrich.
MCNP4a KENOSa 1
B&W-1484 (4A.7)
Core I 2.46 0.9964 +/- 0.0010 0.9898+/- 0.0006 0.1759 0.1753 2
B&W-1484 (4A.7)
Core II 2.46 1.0008 +/- 0.0011 1.0015 i 0.0005 0.2553 0.2446 3
B&W-1484 (4A.7)
Core m 2.46 1.0010 +/- 0.0012 1.0005 i 0.0005 0.199 0.1939 4
B&W-1484 (4A.7)
Core IX 2.46 0.9956 +/- 0.0012 0.9901 +/- 0.0006 0.1422 0.1426 5
B&W-1484 (4A.7)
Core X 2.46 0.9980 +/- 0.0014 0.9922 +/- 0.0006 0.1513 0.1499 6
B&W-1484 (4A7)
Core XI 2.46 0.9978 +/- 0.0012 1.ooS +/- 0.0005i.0.2031 0.1947 7
B&W-1484 (4A.7)
Core XII 2.46 0.9988 +/- 0.0011 0.9978 +/- 0.0006 0.1718 0.1662 8
B&W-1484 (4A-n Core XM 2.46 1.0020 +/- 0.0010 0.9952 +/- 0.0006 0.1988 0.1965 9
B&W-1484 (4OA Core XIV 2.46 0.9953 +/- 0.0011 0.9928 +/- 0.0006 0.2022 0.1986 10 B&W-1484 (4A.7)
Core XV" 2.46 0.9910%i 0.0011 0.9909 +/- 0.0006 0.2092 0.2014 11 B&W-1484 (4A.7)
Core XVI 2.46 0.9935 +/- 0.0010 0.9889 +/- 0.0006 0.17S7 0.1713 12 D&W-1484 (4A.7)
Core XVII 2.46 0.9962 +/- 0.0012 0.9942 +/- 0.0005 0.20S3 0.2021 13 B&W-I484 (4A.7)
Core XVm 2.46 1.0036 +/- 0.0012 0.9931 +/- 0.0006 0.1705 0.1708 Holtec International Appendix 4A, Page 9 Holtec International Appendix 4A, Page 9
Table 4A.1 Summary of Criticality Benchmark Calculations Calculatedl -
EAUF' (eV)
MCNP4a KENOSa Reference Identiricatlon Enrich.
MCNP4a KENO5a 14 B&W-1484 (4A.7)
Core XI 2.46 0.9961 +/- 0.0012 0.9971 +/- 0.0005 0.2103 0.2011 15 B&W-1484 (4A.7)
Core XX 2.46 1.0008 +/- 0.0011 0.9932 +/- 0.0006 0.1724 0.1701 16 B&W-1484 (4A.7)
Core XXI 2.46 0.9994 +/- 0.0010 0.9918 +/- 0.0006 0.1544 0.1536 17 B&W-1645 (4A.8)
S-type Fuel, w1886 ppm B 2.46 0.9970 +/- 0.0010 0.9924 +/- 0.0006 1.4475 1.4680 18 B&W-1645 (4A.8)
S-type Fuel, W1746 ppm B 2.46 0.9990 0.0010 0.9913 +/- 0.0006 1.5463 1.5660 19 B&W-1645 (4A.8)
SO-type Fuel, w/11S6 ppm B 2.46 0.9972 +/- 0.0009 0.9949 +/- 0.0005 0.4241 0.4331 20 B&W-1810 (4A.9)
Case 1 1337 ppm B 2.46 1.0023 +/- 0.0010 NC 0.1531 NC 21 B&W-1810 (4A.9)
Case 12 1899 ppm B 2.46/4.02 1.0060 +/- 0.0009 NC 0.4493 NC 22 French (4A.10)
Water Moderator 0gp 4.75 0.9966 +/- 0.0013 NC 0.2172 NC 23 French (4A.10)
Water Moderator 2.5 cm gap 4.75 0.9962 +/- 0.0012 NC 0.1778 NC 24 French (4A.10)
Water Moderator 5 cm gap 4.75 0.9943 +/- 0.0010 NC 0.1677 NC 25 French (4A.10)
Water Moderator 10 cm gap 4.75 0.9979 +/- 0.0010 NC 0.1736 NC 26 PNL-3602 (4A.11)
Steel Reflector, 0 separatIon 2.35 NC 1.0004 +/- 0.0006 NC 0.1018 Holtec International Appendix 4A, Page 10
Table 4A.1 Summary of Criticality Benchmark Calculations Cplc"lated k......
EAL t
)eV Reference Identification Enrich.
MCNP4a REN05a MCNP4a KENO5a 27 PNL-3602 (4A.11)
Steel Reflector, 1.321 cm sepn.
2.35 0.9980 +/- 0.0009 0.9992 +/- 0.0006 0.1000 0.0909 28 PNL-3602 (4A.11)
Steel Reflector, 2.616 cm sepn 2.35 0.9968 i 0.0009 0.9964 +/- 0.0006 0.0981 0.0975 29 PNL-3602 (4A.11)
Steel Reflector, 3.912 cm sepn.
2.35 0.9974 +/- 0.0010 0.9980 +/- 0.0006 0.0976 0.0970 30 PNL-3602 (4A.11)
Steel Reflector, Inflnlte sepn.
2.35 0.9962 i 0.0008 0.9939 +/- 0.0006 0.0973 0.0968 31 PNL-3602 (4A.11)
Steel Reflector, 0 cm sepn.
4.306 NC 1.0003 t 0.0007 NC 0.3282 32 PNL-3602 (4A.11)
Steel Reflector, 1.321 cm sepn.
4.306 0.9997 +/- 0.0010 1.0012 +/- 0.0007 0.3016 0.3039 33 PNL-3602 (4A.11)
Steel Reflector, 2.616 cm sepn.
4.306 0.9994 +/- 0.0012 0.9974 +/- 0.0007 0.2911 0.2927 34 PNL-3602 (4A.11)
Steel Reflector, 5.405 cm sepn.
4.306 0.9969 +/- 0.0011 0.9951 +/- 0.0007 0.2828 0.2860 35 PNL-3602 (4A.11)
Steel Reflector, Infinite sepn. tt 4.306 0.9910 +/- 0.0020 0.9947 +/- 0.0007 0.28S51 0.2864 36 PNL-3602 (4A.11)
Steel Reflector, with Boral Sheets 4.306 0.9941 +/- 0.0011 0.9970 +/- 0.0007 0.3135 0.3150 37 PNL3926 (4A.12)
Lead Reflector, 0 cm sepn.
4.306 NC 1.0003 +/- 0.0007 NC 0.3159 38 PNL-3926 (4A.12)
Lead Reflector, 0.55 cm sepn.
4.306 1.0025 +/- 0.0011 0.9997 +/- 0.0007 0.3030 0.3044 39 PNIL3926 (4A.12)
Lead Reflector, 1.956 cm sepn.
4.306 1.0000 +/- 0.0012 0.9985 1 0.0007 0.2883 0.2930 Holtec International Appendix 4A, Page 11 Holtee Intemnational Appendix 4A, Page I I
Table 4A.1 Summary of Criticality Benchmark Calculations Is VALIYt (eVI Rererence Identification Enrich.
MCNP4a KEN05a MCNP4a KENOSa 40 PNI,3926 (4A.12)
Lead Reflector, 5.405 cm sepn.
4.306 0.9971 +/- 0.0012 0.9946 +/- 0.0007 0.2831 0.2854 41 PNIr2615 (4A.13)
ExperIment 0041032 - no absorber 4.306 0.9925 +/- 0.0012 0.9950 +/- 0.0007 0.1155 0.1159 42 PNL-2615 (4A.13)
Experiment 030
- Zr plates 4.306 NC 0.9971 +/- 0.0007 NC 0.1154 43 PNL-25 (4A.13)
Experiment 013
- Steel plates 4.306 NC 0.9965 +/- 0.0007 NC 0.1164 44 PNL-2615 (4A.13)
ExperIment 014
- Steel plates 4.306 NC 0.9972 +/- 0.0007 NC 0.1164 45 PNL-2615 (4A.13)
Exp. 009 1.05% Boron-Steel plates 4.306 0.9982 +/- 0.0010 0.9981 +/- 0.0007 0.1172 0.1162 46 PNL-2615 (4A.13)
Exp. 012 1.62% Boron-Steel plates 4.306 0.9996 +/- 0.0012 0.9982 +/- 0.0007 0.1161 0.1173 47 PNI-2615 (4A.13)
Exp. 031 - Boral plates 4.306 0.9994 +/- 0.0012 0.9969 +/- 0.0007 0.1165 0.1171 48 PNL7167 (4A.14)
Experiment 214R - with flux trap 4.306 0.9991 +/- 0.0011 0.9956 +/- 0.0007 0.3722 0.3812 49 PNI7167 (4A.14)
Experiment 214V3 - wi flux trap 4.306 0.9969 +/- 0.0011 0.9963 +/- 0.0007 0.3742 0.3826 50 PNL-4267 (4A.15)
Case 173 - 0 ppm B 4.306 0.9974 +/- 0.0012 NC 0.2893 NC 51 PNL-4267 (4A.15)
Case 177 - 2550 ppm B 4.306 1.0057 +/- 0.0010 NC 0.5509 NC 52 PNL5803 (4A.16)
MOX Fuel - Type 3.2 Exp. 21 20% Pu 1.0041 +/- 0.0011 1.0046 +/- 0.0006 0.9171 0.8868 Holtec International Appendix 4A, Page 12
Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k.M RALF t teV)
Reference Identirication Enrich.
MCNP4a KEN05a MCNP4a KENOSa 53 PNL-5803 (4A.16)
MOX Fuel - Type 3.2 Exp. 43 20% Pu 1.0058 +/- 0.0012 1.0036 i 0.0006 0.2968 0.2944 54 PNL-5803 (4A.16)
MOX Fuel - Type 3.2 Exp. 13 20% Pu 1.0083 +/- 0.0011 0.9989 +/- 0.0006 0.1665 0.1706 55 PNL5803 (4A.16)
MOX Fuel - Type 3.2 Exp. 32 20% Pu 1.0079 +/- 0.0011 0.9966 i 0.0006 0.1139 0.1165 56 WCAP-3385 (4A.17)
Saxton Case 52 PuO2 0.52" pitch 6.6% Pu 0.9996 +/- 0.0011 1.0005 i 0.0006 0.8665 0.8417 57 WCAP-3385 (4A.17)
Saxton Case 52 U 0.52" pitch 5.74 1.0000 +/- 0.0010 0.9956 +/- 0.0007 0.4476 0.4580 58 WCAP-3385 (4A.17)
Saxton Case 56 PuO2 0.56" pitch 6.6% Pu 1.0036 +/- 0.0011 1.0047 +/- 0.0006 0.5289 0.5197 59 WCAP-3385 (4A.17)
Saxton Case 56 borated PuO2 6.6% Pu 1.0008 +/- 0.0010 NC 0.6389 NC.
60 WCAP-3385 (4A.17)
Saxton Case 56 U 0.56" pitch 5.74 0.9994 +/- 0.0011 0.9967 +/- 0.0007 0.2923 0.2954 61 WCAP-3385 (4A.17)
Saxton Case 79 PuO2 0.79" pitch 6.6% Pu 1.0063 +/- 0.0011 1.0133 +/- 0.0006 0.1520 0.1555 62 WCAP-3385 (4A&17)
Saxton Case 79 U 0.79" pitch 5.74 1.0039 +/- 0.0011 1.0008 L
0.0006 0.1036 0.1047 Notes: NC stands for not calculated.
t EALF is the energy of the average lethargy causing fission.
tt These experimental results appear to be statistical outliers (>3n) suggesting the possibility of unusually large experimental error. Although they could justifiably be excluded, for conservatism, they were retained in determining the calculational.
basis.
I~olec I ter atio al A pen ix 4, Pa e 1 Holtec International Appendixt 4A, Page 13
Table 4A.2 COMPARISON OF MCNP4a AND KENO5a CALCULATED REACTIVITIES' FOR VARIOUS ENRICHMENTS Calculated 1Xr E lo Enrichment MCNP4a KENO5a 3.0 0.8465 +/- 0.0011 0.8478 +/- 0.0004 3.5 0.8820 +/- 0.0011 0.8841 +/- 0.0004 3.75 X.9019 +/- 0.0011 0.8987 +/- 0.0004 4.0 0.9132 +/- 0.0010 0.9140 +/- 0.0004 4.2 0.9276 +/- 0.0011 0.9237 i 0.0004 4.5 0.9400 +/- 0.0011 0.9388 +/- 0.0004 t
Based on the GE 8xWR fuel assembly.
Holtec Intemational Appendix 4A, Page 14
Table 4A.3 MCNP4a CALCULATED REACTIVITIES FOR CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS
/k MCNP4a Worth of Calculated EALFt Ref.
Experiment Absorber kw (eV) 4A.13 PNL-2615 Boral Sheet 0.0139 0.9994+/-0.0012 0.1165 4A.7 B&W-1484 Core XX 0.0165 1.0008+/-0.0011 0.1724 4A.13 PNL,2615 1.62% Boron-steel 0.0165 0.9996+/-0.0012 0.1161 4A.7 B&W-1484 Core XX 0.0202 0.9961+/-0.0012 0.2103 4A.7 B&W-1484 Core XXI 0.0243 0.9994+/-0.0010 0.1544 4A.7 B&W-1484 Core XVII 0.0519 0.9962+/-0.0012 0.2083 4A. I I PNL-3602 Boral Sheet 0.0708 0.9941+/-0.0011 0.3135 4A.7 B&W-1484 Core XV 0.0786 0.9910+/-0.0011 0.2092 4A.7 B&W-1484 Core XVI 0.0845 0.9935+/-0.0010 0.1757 4A.7 B&W-1484 Core XIV 0.1575 0.9953+/-0.0011 0.2022 4A.7 B&W-1484 Core XIII 0.1738 1.0020+/-0.0011 0.1988 4A.14 PNL-7167 Expt 214R flux trap 0.1931 0.9991+/-0.0011 0.3722 tEALF is the energy of the average lethargy causing fission.
HotcItrainlApni A ae1 Holtec International Appendix 4A, Page 15
Table 4A.4 COMPARISON OF MCNP4a AND KENO5a CALCULATED RBEACTVITIESt FOR VARIOUS "'B LOADINGS Calculated Ki +/- la
'0B, glcOm 2
MCNP4a EKENO5a 0.005 1.0381 i 0.0012 1.0340 i 0.0004 0.010 0.9960 i 0.0010 0.9941 +/- 0.0004 0.015 0.9727 i 0.0009 0.9713 i 0.0004 0.020 0.9541 i 0.0012 0.9560 i 0.0004 0.025 0.9433 i 0.0011 0.9428 i 0.0004 0.03 0.9325 i 0.0011 0.9338 +/- 0.0004 0.035 0.9234 i 0.0011 0.9251 i 0.0004 0.04 0.9173 i 0.0011 0.9179 i 0.0004 t
Based on a 4.5% enriched GE 8x8R fuel assembly.
Holtec: International, mn Appendix 4A, Page l16
Table 4A.5 CALCULATIONS FOR CRUTICAL EXPERIMENTS WITH THICK LEAD) AND STEEL REFLECTORSt Separation, Ref.
Case E, wt%
cm MCNP4a kff KEN05a kf 4A.11 Steel 2.35 1.321 0.9980+/-0.0009 0.9992+/-0.0006 Reflector 2.35 2.616 0.9968+/-0.0009 0.9964+/-0.0006 2.35 3.912 0.9974+/-0.0010 0.9980+/-0.0006 2.35 0.9962+/-0.0008 0.9939+/-0.0006 4A.11 Steel 4.306 1.321 0.9997+/-0.0010 1.0012+/-0.0007 Reflector 4.306 2.616 0.9994+/-0.0012 0.9974+/-0.0007 4.306 3.405 0.9969+/-0.0011 0.9951 +/-0.0007 4.306 co 0.9910+/-0.0020 0.9947+/-0.0007 4A.12 Lead 4.306 0.55 1.0025+/-0.0011 0.9997+/-0.0007 Reflector 4.306 1.956 1.0000+/-0.0012 0.9985+/-0.0007 4.306 5.405 0.9971+/-0.0012 0.9946+/-0.0007 t
Arranged in order of increasing riflector-fuel spacing.
Holtec; International '
n Appendix 4A, Page 17
Table 4A.6 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS Calculated kn Boron Concentration, Reference.
Experiment ppm MCNP4a KENO5a 4A.15 PNLA4267 0
0.9974 +/- 0.0012 4A.8 B&W-1645 886 0.9970 +/- 0.0010 0.9924 +/- 0.0006 4A.9 B&W-1810 1337 1.0023 +/- 0.0010 4A.9 B&W-1810 1899 1.0060 +/- 0.0009 4A.15 PNL-4267 2550 1.0057 +/- 0.0010 1-loltec International Appendix 4A, Page lB Holtec International Appendix 4A, Page IS
Table 4A.7 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNP4a KENO5a Reference Caset k
EALFt kff EALF" PNL-5803 MOX Fuel - Exp. No. 21 1.0041 +/-0.0011 0.9171 1.0046+/-0.0006 0.8868 14A. 16]
MOX Fuel - Exp. No. 43 1.0058+/-0.0012 0.2968 1.0036+/-0.0006 0.2944 MOX Fuel - Exp. No. 13 1.0083+/-0.0011 0.1665 0.9989*0.0006 0.1706 MOX Fuel -Exp. No. 32 1.0079+/-0.0011 0.1139 0.9966+/-0.0006 0.1165 WCAP-Saxton a 0.52" pitch 0.9996+/-0.0011 0.8665 1.0005+/-0.0006 0.8417 3385-54
[4A. 17]
Saxton @ 0.56" pitch 1.0036+/-0.0011 0.5289 1.0047+/-0.0006 0.5197.
Saxton 0 0.56' pitch borated 1.0008*0.0010 0.6389 NC NC Saxton @ 0.79" pitch 1.0063+/-0.0011 0.1520 1.0133+/-0.0006 0.1555 Note: NC stands for not calculated t
Arranged in order of increasing lattice spacing.
tt EALF is the energy of the average lethargy causing fission.
Holtc Inerntionl Apendx 4, Pae 1 Holtec Intemnational Appendix 4A, Page 19
Linear Regression with Correlation Coefficient of 0.13 1.010 1.005 0
a a) 4-0
-X 1.000 0
. 0
-3 0
0 0.995 0.990
- 0.
Energy of Average Lethargy Causing Fission (Log Scale)
FIGURE 4A.1 MCNP CALCULATED k-eff
- VARIOUS VALUES OF THE VALUES for SPECTRAL INDEX
Linear Regression with Correlation Coefficient of 0.21 1.010 1.005 0
Q 1.000 4-a, o.
0 0
0.990I 0.985 -
0.1 1
Energy of Average Lethargy Causing Fission (Log Scale)
FIGURE 4A.2. KEN05a VARIOUS CALCU LATED VALUES OF k-eff VALUES THE SPECTRAL FOR INDEX
Linear Regression with Correlation Coefficient of 0.03 1.010 1.005 4-0 1.000 I
t___
- 1 0.995 0.990 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 6.0 Enrichment, w/o U-235 FIGURE 4A.3.
MCNP CALCULATED k-eff VALUES AT VARIOUS U-235 ENRICHMENTS
Linear Regression with Correlation Coefficient of 0.38 1.010 1.005
, 1.000 4-
'4-a) x eX-0 CD 4-0 o 0.995 73 0
0.990 0.985 1
^1 1
1 1
l
§ W
l~~
s-'
111111111_
llg}filllXr_
r-l I;lsL l " l f i*l_
Z.U Z.5 3.0 3.5 Enrichment, w/o 4.0 4.5 5.0 5.5 b.U U-235 FIGURE 4A.4.
KENO CALCULATED k-eff VALUES AT VARIOUS U-235 ENRICHMENTS
0.94 0.92 E
W 0
0 75 0
4-I4-m 0Ln 0
LiJ 3N 0.90 0.88 0.86 0.84 -f--
0.84 MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KEN05A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS
zC.,
3:
4-a
._-I 0-C-, ID C,
Reactivity Calculated with KEN05a 4...
FIGURE 4A.6.
COMPARISON OF MCNP AND KEN05a CALCULATIONS FOR VARIOUS BORON-10 AREAL DENSITIES
Appendix B Approved Computer Program List (Total Pages: 6 including this page)
Project No. 1104 Report No. HI-2043262NP Page B-I Holtec International
IIOLTEC APPROVED COMPUTER PROGRAM LIST REV. 72 June 29, 2004 PROGRAM VERSION CERTIFIED USERS OPERATIN REMARKS CODE (Category)
GSYSTEM USED ANSYS (A) 5.7,7.0 JZ, ER, Windows PK, CWVB, SPA, AIS, IR, SP, AK. SJ, RW, VRP AIRCOOL 5.21,6.1 Windows BACKFILL 2.0 DOS/
Windows BONAMI (Scale) 4.3, 4.4 Windows BULKTEM 3.0 DOS/
Windows CASMO-4 (A) 1.13.04 ERD, SPA, DMM, KC, UNIU Version 1.13.04 2.05.03 (UNIX),
ST,VJB Windows should not be used 2.05.03 for new projects (WINDOWS) and should only be used when necessary for additional calculations on previous projects.
The user should refer to the error notice documented in c4ser.04-results.pdf located in \\generic\\library\\
nucleaierror notices\\
concerning the use of version 1.13.04.
Library N should be used with version 2.05.03 for all new reports issued after June I", 2003.
Revisions to reports issued prior to June I",
2003 may continue to use the old Library L.
CASMO-3 (A) 4.4, 4.7 ERD, SPA, DMM, KC, ST UNIX CELLDAN 4.4.1 Windows CHANBP6 (A) 1.0 SJ, PK, CWB, AIS, SP,AK DOS/Windo
-ws Project No. 1104 Report No. HI-2043262NP Holtec International Page B-2
HOLTEC APPROVED COMPUTER PROGRAM LIST REV. 72 June 29, 2004 PROGRAM VERSION CERTIFIED USERS OPERATIN REMARKS CODE (Category)
G SYSTEM USED CHAPO8 1.0 Windows (CHAPLS 10)
CONPRO 1.0 DOS/Windo ws CORRE 1.3 DOS/Windo ws DECAY 1.4, 1.5 DOS/Windo ws DECOR 1.0 DOS/Windo ws DR.BEAMPRO 1.0.5 Windows DR.FRAME 2.0 Windows DYNAMO (A) 2.51 AIS, SP, CWB, PK, SJ DOS/Windo Personnel ws qualified to use MR216 are automatically qualified to use DYNAMO.
DYNAPOST 2.0 DOS/Windo
'ws FIMPACT 1.0 DOS/Windo ws FLUENT (A) 4.32, 4.56, ER, IR, DMM, SPA Windows Do not use porous 6.1.18 medium with zero velocity.
FTLOAD 1.4 DOS GENEQ 1.3 DOS HXFLOW 1.0 DOS/Windo I
'ws INSYST 2.01 Windows KENO-SA (A) 4.3, 4.4 ERD, SPA, DMM, KC, Windows ST,VJB__
LONGOR 1.0 DOS/Windo ws LNSMTH2 1.0 DOS/Windo ws LS-DYNA3D (A) 936, 940, 950, JZ, AIS, SPA, SP, Windows 960,970 KPSVRP MAXDISP8 1.8 DOS/Windo ws MAXDIS16 1.0 DOS/Windo ws ProjectNo. 1104 Report No. HI-2043262NP Holtec International Page B-3
HOLTEC APPROVED COMPUTER PROGRAM LIST REV. 72 June 29, 2004 PROGRAM VERSION CERTIFIED USERS OPERATIN REMARKS CODE (Categori)
G SYSTEM USED MCNP (A) 4A, 4B ERD, SPA, KC,STDMM, Windows/
CASMO-4 4A VJB, MAP UNIX Lumped Fission Products (IDs 401 and 402) and Isotope Pml48M (ID 61248) can be modeled in MCNP 4A using the cross sections documented in Hl-2033031. Use of these cross sections is restricted to MCNP 4A, and to material specifications in atom densities.
MASSINV 1.4, 1.5,2.1 DOS/Windo Ws MR2 1.7 AIS, SP, CWB, PK, SJ DOS/Windo For use in wet ws storage analysis only.
MR216 (A) 1.0, 2.0, AIS, SP, CWB, PK, SJ,AK DOS/Windo Versions 2.2 and 2.2,2.4 ws 2.4 for use in dry storage analyses only. Use DYNAMO for liquefaction problems.
MSREFINE 1.2,1.3, 2.1 DOS/Windo MULPOOLD 2.1 DOS/Windo Ws MULTIl 1.3,1.4,1.5, Windows 1.54, 1.55 NITAWL (Scale) 4.3, 4.4 Windows NASTRAN 6.2, Windows DESKTOP 2001,6.4,2002 (WORKING
,2003 MODEL) l ONEPOOL 1.4.1, 1.5, 1.6 DOS/Windo ws ORIGENS (Scale) 4.3, 4.4 Windows PD16 1.1, 1.0,2.1 Windows Project No. 1104 Report No. HI-2043262NP Holtec International Page B-4
IIOLTEC APPROVED COMPUTER PROGRAM LIST REV. 72 June 29, 2004 PROGRAM VERSION CERTIFIED USERS OPERATIN REMARKS CODE (Category)
G SYSTEM USED PREDYNAI 1.5,1.4 DOS/Windo ws PREMULT8 1.0 DOS/Windo ws PRESPRG8 1.0 DOS/Windo ws PSDI 1.0 DOS/Windo Ws QAD CGGP DOS/Windo Ws SAS2H (Scale) 4.3, 4.4 Windows SFMR2A 1.0 DOS/Windo SHAPEBUILDER 3.0 DOS/Windo SIFATIG 1.0 DOS/Windo Ws SOLIDWORKS
- 200iPLUS, DOSJWindo This program may 2003 ws be used to calculate Weight, Volume, Centroid and Moment of Inertia.
As a precaution, user should avoid keeping more than one drawing files open at any given time during a Solidworks session.
If there is a need for multiples drawing files to be open at once, user should ensure that the part names for all open files are uniquely named (i.e. no two parts have the same name.)
Project No. 1104 Report No. HI-2043262NP Holtec International Page B-5
HOLTEC APPROVED COMPUTER PROGRAM LIST REV. 72 June 29, 2004 PROGRAM VERSION CERTIFIED USERS OPERATIN REMARKS CODE (Categorv)
G SYSTEM USED DOS/Windo SPG16 1.0, 2.0, 3.0 ws SHAKE2000 1.1.0, 1.4.0 DOS/Windo Ws STARDYNE (A) 4.4, 4.5 SP Windows STER 5.04 Windows TBOIL 1.7, 1.9 DOS/Windo See HI-92832 for ws restriction on vl.7.
THERPOOL 1.2,1.2A DOS/Windo TRIEL 2.0 DOS/Windo ws VERSUP 1.0 DOS VIBIDOF 1.0 DOS/Windo VMCHANGE 1.4, 1.3 Windows WEIGHT 1.0 Windows NOTES:
- 1. XXXX = ALPHANUMERIC COMBINATION
- 2.
GENERAL PURPOSES UTILITY CODES (MATHCAD, EXCEL, ETC.) MAYBE USED ANYTIME.
Project No. 1104 Report No. HI-2043262NP Holtec International Page B-6 Letter from F.H. Smith to J.S. Rowe dated November 9, 2001 (13 Pages, including this one)
Project No. 1104 Report No. HI-2043262NP Holtec International
-- Entergy Inter-Office To: J. S. Rowe Correspondence From: F. H. Smith Date: November 9, 2001
Subject:
Inputs for ANO SFP Criticality Analysis, Phase 2 CEO 2001-00284 File No.: 104-35, 204-35 This letter provides transmittal of data for phase two of the ANO spent fuel pool rack analysis.
On the attached pages, information on the following topics can be found:
Fuel parameters (Unit 2)
Details of axial blankets (Unit 2)
Reactor specific power (Unit 2)
I Core soluble boron letdown curves (Unit 2)
Core operating temperatures (Unit 2)
Radial and Axial Peaking Factors (both units)
An axial burnup profile for spent fuel (Unit 2)
Details of integral poison materials in the fuel (Unit 2)
This information has been prepared and documented in accordance with the Headquarters Nuclear Quality Assurance Program governing nuclear safety related analyses.
Please feel free tontt require any additional information.
F. H. Smith Supervisor, Core Design FHS/wbb cc:
W. B. Bird (M-ECH-36)
C. D. Walker (N-GSB)
Corporate File [12)
Fuel Parameters Table I presents fuel parameters data. For additional information on guide tubes and instrument tubes for ANO-2 fuel assemblies, please refer to Entergy letter CEXO 2000-00464.
Table 1. Fuel Parameters Data ROD PARAMETER ANO-2 VALUE Assembly tyypeCE 16xl6 Fuel pellet outside diameter (in.)
0.325 Cladding thickness (in.)
0.025 Cladding outside diameter (in.)
0.382 Cladding MaterialZir-4 Maximum stack density (g/cc) 10.412
- Maximum enrichment. wvt% U-235 5.0%
ASSEMBLY PARAMIETER Array size l6x16 Number of fuel rods 236 Assembly Width (in.)
8.130 - 8.149 Assembly Pitch (in.)
8.180 Fuel rod pitch (in.)
0.506 Number of control rod guide tubes and instrument tubes 5 guide tubes (2x2)
Guide tubes outside diameter (in.)
0.980 Guide tubes inside diameter (in.)
0.900 Active fuel length (in.)
149.61 - 150.0
- A stack density that is 95% of theoretical density should bound future reloads, as well as all previous fuel designs.
Details of Axial Blankets ANO-2 fuel designs do not utilize axial enrichment blankets.
Core Soluble Boron Concentration Core soluble boron concentration data is provided in Figure I in the form of letdown curves for four recent cycles. These curves are representative of recent cycle operations. However, in order to bound future core designs, please adjust these values to a BOC concentration of 2000 ppm.
Boron Letdown vs. Exposure CB 0
I 0C u
a In
-cli
-&-C12
-C14 0
2 4
6 8
10 12 14 16 18 20 UApasure (GWdWMT)
Figure 1. Representative Boron Letdown Curves Core Operating Conditions Table 2 documents Core Operating temperatures for ANO-2. Table 3 documents design values for axial and radial peaking at both units.
Table 2. Core Operating Temperatures PARAMETER NOMINAL VALUE Average T&,,1000-1040 OF Moderator TW, 553.5 "F (C1-9) 545.0 "F (CI 0-14) 549.0 OF (Cl 5) 55 1.0 "F (C16)
Moderator T.,,
573-578 "F IModerator T,,,,,,
-604 OF Table 3. Maximum Radial and Axial Peaking for TIH Analysis UNIT I UNIT 2 l Axial 1.65 1.248 l Radial F,%FI
= 1.80 F. = 1.65
The core specific power ranges from approximately 37.0 to 40.0 MW/MTU.
Axial Burnup Profile EOI has developed an EOC axial burnup distribution for use in the ANO-2 criticality analysis.
This burnup profile is shown in Table 4.
Table 4. Generic Axial Bumup Distribution AXIAL SEGMENT (CNI)
RELATIVE BURINUP 0 to 15.24 0.55 15.24 to 30.48 0.82 30.48 to 60.96 1.01 60.96 to 121.92 1.105 121.92 to 182.88 1.105 182.88 to 243.84 1.075 243.84 to 304.80 1.07 304.80 to 335.28 1.02 335.28 to 350.52 0.92 350.52 to 365.76
> 0.72 365.76 to 381 0.47 Integral Poison Materials For the purposes of the criticality analysis, the presence of a stronger poison during the assembly's core lifetime is conservative. However, since some poison material may remain in the fuel after discharge, a lesser poison loading may result in a more reactive bundle in the SFP. The most limiting poison configuration should be assumed for each burnup requirement. For the criticality analysis, low enriched pins should be modeled as high enriched. This section will summarize the three types of poison material in use at ANO-2.
B4C Boron Carbide was used as an integral poison material for cycles 1-12 (fuel batches A-P). These poison rods consist of B4C in an A120, matrix. Boron-10 concentrations vary from 0.004 glinch to 0.028 glinch. This corresponds to 0.403 wt% to 4.632 Nt/o B4C, with boron-10 comprising approximately 18.3% of the boron present. Between 0 and 16 B4C rods were used for each lattice. Loading patterns are shown in Figure 2.
0 Burnable Absorbers 4 Burnable Absnrhers L
L L
I IL L
L_
_L L
LLL L
S L
I_
L L _
L I
LL L
I I_
L II 1
I L L
___LLSL L
L L-__H L
L I
I L
I I
L IL I
L__
L L
I L
I LI L I
_ L I
I I
I LILSI I
I SLI L
LI I
I I
L
_N LI L
L L
LILI L
L I
SILI L
LT Eftf L
L L
L I
I.
I
_ L I
L r _
L L
- ISIII I
7S LI LI l
l l
LI Sl S
L
_~~
_Y 7_
L1 LL L
L I I T L l
L_
L I L
=
I l
I I
_ I I.T II11 L I L
I L
L_
L 1j I L 12 Burnable Absorbers (A)
L L
ISI I
I I
IlL I
I SI I
L L 1.1 I
rI I
I I
I SIL IS
{
-m 1
L-11 I~s s
lL
_1X 1I I *LI L I I T I LI *T1 I11 L
IL S
L L
L I L
L L
LLI I
L S
- -LI L S r L
L_
L II L
I I I
I LI L
L L
I L
L S I f. LSSLS I L I
L IL L
Figure 2a. B4C Loading Patterns
12 Burnable Absorbers (B) s_
Si _
S _
S S
_S
-S
~~ _
L 5__
S_--5 S
S 16 Burnable Absorbers (B)
L L I
I I L L L
s S
_I S
S L LI L L L L
L L
L L
L L
r
=
LL I _
IL LlIL
=
_ S I LIL S= = Sl Ll L lSl L
L L L L
IL I
S _
L S
I I
IS L
L I
= l L L LLT I
-1
--- L Lll 1 Rivrnnhln Ahcrrhdkrc fA%
I.
l I
L Isl I.
IS S
I L
S SS S
S
_S S
L S
I S
S {
S Ll lL 1
LL KEY:
L = Low enrichment pin (convert to high enriched for this analysis)
S = B,C pin Figure 2b. B4C Loading Patterns Gadolinium Cycles 1 3-I 5 (fuel batches R, S. and T) utilize Gd203 poison at 6.0% weight percent. with either 2.311i or 2.5% enriched U20.
The Gadolinia loading patterns are shown on the following pages.
i i
I R2 RI R3 O HIGH ENRICHMENT (4.40 W/O)
B LOW ENRICHMENT (3.90 W/OJ GADOLINIA ROD (GAOOLINIA ENRICHMENT:G W/O, CARRIER ENRICHMENT:2.3 W/O)
SAR FIGURE NO. 4.3-IC AMENDMENT-16 NUCLEAR ONE SCALE f NONE ARKANSAsNCLAOE DRAWN I CHARLEY RANKIN wMff 2
a ENTEMRY DESIGN t ENTERGY RUSSELLVILLE. ARKANSAS CAD NO s 4f3-01c.iar INTEGRAL BURNABLE POISON SHIM BASED ON ORAWING NO SHEET IEV ANO ENRICHMENT ZONING PATTERNS FOR BATCH R FUEL ASSEMBLIES II III Figure 3. Batch R Gad Loading
52 Si
]
S4 o U02 Pin. High Enrichment 4.50 w/o) 0 U02 Pin. Low Enrichment 14.10 w/ol Gd2O3 -U02 Pin (Godolinia Enrichment.
6 w/o; Carrier-Enrichment: 2.5 w/o)
I i
SAR FIGURE NO. 4.3-18 AMENDMENT-16
^
^SCALE 3 NONE ARKANSAs NUCLEAR ONE DRAWN ICHARLEY RANKIN UJUT 2
-ENTERaY DESIGN JENTERGY RUSSELLVILLE.
ARKANSAS CAD NO a 4f3-01bsar-INTEGRAL BURNABLE POISON SHIM BASED ON DRAWING NO SHEET REV ANO ENRICHMENT ZONING PATTERNS FOR BATCH S FUEL ASSEMBLIES Figure 4. Batch S Gad Loading
Tl r2 T3 Gd2 03'UO2 Pin o U02 Pin. High Enrichment El Woterhole a U02 Pin, Low Enrichment T4 SAR FIGURE NO.4.3-IA AMENDMENT-16 I
I SCALE
' NONE ARKANSAS NUCLEAR ONE a
DRAWN JCHARLEY RANKIN UNIT 2
ENTERMY DESIGN S ENTERGY RUSSELLVILLE. ARKANSAS CAD NO s 403 01.
5 ar INTEdRAL BURNABLE POISON SHIM BASED ON DRAWING NO SHEET FEy AND ENRICHMENT ZONING PATTERNS I
FOR BATCH T FUEL ASSEMBLIES I
I I
Figure 5. Batch TGad Loading
Erbium Erbium will be utilized as a poison material for future cycles at ANO-2, at 2.1 wI% Er.O.
Figure 6 shows the erbium loading patterns that may be used for ANO-2 fuel.
O Frhinm Rnoid 24 Erbium Rods LIL I I I
I I I LIL LI I-I I I
I III Ir TL ILI L I
I I
L L
LI I
_I I
L L
L IL I I I
ILI L I I I
L LI I IL L
_L 32 ErirnRd I -_
L I L
LL L
L L
I I LI I
I_
I I
I I
L L 32 Erbiumn Rods L
LI I I I
I LI LI
_EI L
L I L
L E
I I
E _
E E
E I E _ I E
I LIEl L
I I
LILIElL I
I LIElE L I_
I_
I E
E I
I E
E L
E E L _
E_
EL EL I
L E LLE L
L I E lI LI LI I
IIE I
L L
I I
I I
I I
I I
I L
L LLI I _
I _
I I I I L LL ILI LI El L LI I I 1 ILI LI El LI L1 E
I l
El I
E IL 1 ETL I EI E
IL In I I E
I I
El I
I I
L E
EI L I
I LE E E E
El_
L L E
LI I
L E LI LI I
E E
L E
I 1
I E
L E 1 L I
L I
I Li E Li L I
l lI I L L
48 Erbium Rods IL L
- 1. L L
L E
EI ELI E L EL II_
L ElE LE E
E
'IE E
Ll E E
IL I El E l
L I
I I
E E l E
E LE E
E I EL I
E l E I
I_ __
E L.
E E
_E E-
£E E
_+
E E
I E IL E I E L
- 1. E E
- 1. I.
LLE EE Figure 6a. Erbium Loading Patterns
E E I_
IE l
I El I _I_
L ELLE
_ L El El ELLI I
I I
LF EI l E I I
I I E I E T E
E I
E E
E I E
E L E E
E
_LE E
E E
L EE L E
E E
E E
E I
I L ElEL
_E l_
IE EE I
E E L
E E
rbim E
IoE E
E T
T E
E
_ L El ElWE E I
L E ElE El L
lLlET lEI I
_T LE T TE1E I L
E T
E EI LI EI TETEI I
T IL 80 Erbium Rods 1 El E I1 El I
1 EEL 1 l ElET I
TIX E
l El E E E I
I 7 E
I I
I E
I T I E E
I I
El I El El E E
I E
El El E E E
l E ElE E
I I
E I
I I
E E
I I
I I
E l E
I_
E IE I E E
E I
E IE E I EI E
_ E EI E 1
E I
E
_ _ EI I I E EIEIE IEEE E IEIE IEEE L
EF1E I I
E._
EEL IL I_
I_
III IL I
72 Erbium Rods I
LI I
I I
I I
L IL L I E
I E
E E L E E Ir E l '- I I
I I
E F
E E I E E
I _
I l E I E E
E E
E I E E E E I I EI E
E I
_ E E
I E
E E EEE E I L E E E I
I E
E El I
I E
El I
I E
E I
I l
E l I l
E E
E E E EE l
I_
E I
E E
88 Erbium Rods E
lE1 lE I E I I
T I
I Er E
lLZ L L E
II E
I E I E' E 1
E E E
I I
EI IE E E
F 1I I
E 7 E E
EI I
I EI E
E El EE F
E7 E E EI E I I
E E E E I
El I
EE l
I I
I E
El E
I E
E I
E E
EEEE I
El El E
E I
E I El E
EE E
El I
E 7 E E ElE E I E
E E E
E E L
l I
I IF E
F LI E L I
I I
_LL v
Figure 6b. Erbium Loading Patterns
100 Erbium Rods E
EL E
E I L
E E
L E
E _
El E
I I
E E E I
L E
El E E
I E E IE I EEL El E
I E
E El E
E E
E I El EE El EEl EE E
E II E ElEE II E
I I E
E I_
I I
El E
I_
E I
I ElElEE II E
E El ElEEE I
I E ElElE E
E E
E E
E E
E E
E L EEJEJEE EE E EE L
EE I
I EIEE E EL KEY:
L - Low enriched pin (model as high enriched for the criticality analysis.)
E - Erbiun pin (2.1% Er2O3, varying uranium enrichments)
Figure 6c. Erblum Loading Patterns 2CAN080501 Holtec Affidavit Regarding Withholding from Public Disclosure
AFFIDAVIT PURSUANT TO 10CFR2.390 I, Vince Bilovsky, being duly sworn, depose and state as follows:
(1)
I am the Holtec International Project Manager for Holtec Project 1104 (ANO Dry Storage) and have reviewed the information described in paragraph (2) which is sought to be withheld, and am authorized to apply for its withholding.
(2)
The information sought to be withheld is Holtec Report HI-2043262 Rev 0 with the associated computer files.
(3)
In making this application for withholding of proprietary information of which it is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 1 OCFR Part 9.17(a)(4), 2.390(a)(4), and 2.390(b)(1) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential commercial information", and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992),
and Public Citizen Health Research Group v. FDA, 704F2dl280 (DC Cir.
1983).
1 of 5
AFFIDAVIT PURSUANT TO 10CFR2.390 (4)
Some examples of categories of information which fit into the definition of proprietary information are:
- a.
Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Holtec's competitors without license from Holtec International constitutes a competitive economic advantage over other companies;
- b.
Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.
- c.
Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies of Holtec International, its customers, or its suppliers;
- d.
Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holtec International;
- e.
Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraph 4.a and 4.b, above.
(5)
The information sought to be withheld is being submitted to the NRC in confidence. The information (including that compiled from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to 2of5
AFFIDAVIT PURSUANT TO 10CFR2.390 prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.
(6)
Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge.
Access to such documents within Holtec International is limited on a "need to know" basis.
(7)
The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation.
Disclosures outside Holtec International are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8)
The information classified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is classified as proprietary because it contains detailed descriptions of analytical approaches and methodologies not available elsewhere. This information would provide other parties, including competitors, with information from Holtec International's technical database and the results of evaluations performed by Holtec International. A substantial effort has been expended by Holtec International to develop this information. Release of this information would improve a competitor's position because it would enable Holtec's competitor to copy our technology and offer it for sale in competition with our company, causing us financial injury.
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AFFIDAVIT PURSUANT TO 10CFR2.390 (9)
Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Holtec International's comprehensive spent fuel storage technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology, and includes development of the expertise to determine and apply the appropriate evaluation process.
The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.
The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.
Holtec International's competitive advantage will be lost if its competitors are able to use the results of the Holtec International experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.
The value of this information to Holtec International would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.
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AFFIDAVIT PURSUANT TO 10CFR2.390 STATE OF NEW JERSEY
)
)
ss:
COUNTY OF BURLINGTON )
Mr. Vince Bilovsky, being duly sworn, deposes and says:
That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.
Executed at Marlton, New Jersey, this 2nd day of August, 2005.
Vince Bilovsky Holtec International Subscribed and sworn before me this _
day of 2005.
~C MARIA C. MASSI NOTARY PUBLIC OF NEW JERSEY
'lAy Commission Expires April 25, 2010 5 of5