ML020580082
| ML020580082 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 02/19/2002 |
| From: | Pace P Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| -nr, TAC MB1884 HI-2012620 Rev 3 | |
| Download: ML020580082 (68) | |
Text
Tennessee Valley Authority, Post Office E'ox 20C0. Sprng C ty. Ternessee 3331 -2000 FFS 1,1 2WJ2 10 CFR 50.9 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:
In the Matter of ) Docket No.50-390 Tennessee Valley Authority
SUBJECT:
WATTS BAR NUCLEAR PLANT - REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING TRITIUM PRODUCTION - HOLTEC ANALYSIS (TAC NO. MB1884)
On November 21, 2001, TVA submitted Holtec International analysis requested in NRC RAI letter dated October 2, 2001. That November letter provided both a proprietary and non-proprietary versions of Holtec, International report HI-2012620.
After submittal of these documents, an email was received from the WBN NRC Project Manager which requested that a clarification of the non-proprietary version of the'IToltec International analysis be made to clearly identify the proprietary information that had been removed. The enclosure to this letter provides a new non proprietary version (Revision 3) which will supercede the previous revision provided in the November 21, 2001, letter. This new version revises the document to identify Appendices A and C to be withheld in accordance with paragraph (b) (4) of 10 CFR 2.790.
Please refer to the November 21, 2001, letter for the original withholding of proprietary information request, the associated Holtec International's affidavit, and Holtec International contact information.
U.S. Nuclear Regulatory Commission Page 2 If There are no regulatory commitments made by this letter.
you have any questions about this letter, please contact me at (423) 365-1824.
Since ely, P. L. Pace Manager, Site Licensing and Industry Affairs Enclosures cc: See page 3 Subscribed and sworn to before me on this ]d day of j *2. . 0O.L" My Commission expires 1 0
U.S. Nuclear Regulatory Commission Page 3 cc (Enclosure):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. L. Mark Padovan, Senior Project Manager U.S. Nuclear Regulatory Commission MS 08G9 One White Flint North 11555 Rockville Pik'e Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 E-1
ENCLOSURE TENNESSEE VALLEY AUTHORITY WATTS NUCLEAR PLANT (WBN)
UNIT 1 DOCKET NO. 390 HOLTEC, INTERNATIONAL REPORT NUMBER HI-2012620, REVISION 3 NON-PROPRIETARY VERSION E-1
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m EVALUATION OF THE EFFECT OF THE USE OF TRITIUM PRODUCING BURNABLE ABSORBER RODS (TPBARS) ON FUEL STORAGE REQUIREMENTS FOR WA TS BAR UNIT 1 (TVA)
Holtec Report No. HI-2012620 (NP), R3 z Holtec Project No: 90941 Report Class SAFETY RELATED
-- 0APPROVED Thk* l-pp-a A~~m .01 C.- 0-q Le=tcr No. 30M431 Cot.: Janaaafv 29 2002 7E.SSEE VALEY AA%.._
SO". ) BY 0. L LVA NON-PROPRIETARY VERSION
- 1. 1RMS, WTC A-K L
Summary of Revisions Report M-2012620 Revision 1: The document is revised to incorporate client comments transmitted to Holtec International by TVA via Letter 30M410 dated April 10, 2001. There are no changes to the conclusions of the report.
Revision 2: The document is revised to incorporate client comments transmitted to Holtec International by TVA via Letter 30M414 dated July, 2001. There are no changes to the conclusions of the report.
Revision 3: The document is revised to incorporate client comments transmitted to Holtec International by TVA via Letter 30M430 dated January 11, 2002. There are no changes to the conclusions of the report.
Report HI-2012620 Projcci 90941 Page SR-I
TABLE OF CONTENTS 1,0 INTRODUCTION AND SUN2MARY ............................. 1 2.0 ANALYSIS CRITERIA AND ASSUMPTIONS ....................... 5 3.0 ACCEPTANCE CRITERIA ........................................ 6 4.0 DESIGN AND INPUT DATA ....................................... 7 5.0 M ETHO DO LO GY ......................... ...................... 8 6.0 A N A LY SIS RESU LTS .................................................................. 9 7.0 ACCIDENT CONDITIONS AND SOLUBLE BORON REQUIREMENTS... 12 8.0 OTHER BURNABLE POISON ROD INSERTS IN THE FUEL ASSEMBLIES.. 13 9.0 CRITICALITY ANALYSES RESULTS AND CONCLUSIONS................. 14 10.0 REFEREN CES ................................................. 15 TABLES (Total 11)
FIGURES (Total 4)
APPENDIX A List of Holtec's QA Approved Computer Codes List APPENDIX B Benchmark Calculations APPENDIX C List of CASMO4 and KENO-Va Input Files Report HI-2012620 Project 90941
LIST OF TABLES 4.1 Design Basis Fuel Assembly Specifications ........................................... 16 6.1 Reactivity Effects of Density Tolerance in the Watts Bar Spent Fuel Racks .......... 17 6.2 Reactivity Effects of Temperature and Void in Watts Bar Spent Fuel R ack s ........................ P.................................................................. 18 6.3 Reactivity Effects of Fuel Enrichment Tolerance in Watts Bar Spent Fuel Racks... 19 6.4 Reactivity Effects of Abnormal and Accident Conditions in Watts Bar Spent F uel Racks ............................... ..................................................... 2 0 6.5 Summary of the Criticality Safety Analyses for Checkerboard Storage of 2 Fresh and 2 Spent Fuel Assemblies In Watts Bar Racks (Arrangement 2)......21 6.6 Summary of the Criticality Safety Analyses for Face Adjacent Storage of Spent Fuel Assemblies In Watts Bar Racks (Arrangement 1) ............................... 22 6.7 Summary of the Criticality Safety Analyses for Checkerboard Storage of 3 Fresh Fuel Assemblies and I Water Cell in Watts Bar Racks (Arrangement 3) ............. 23 6.8 Summary of the Criticality Safety Analyses for Storage of Fresh Fuel Assemblies, Containing 32 IFBA rods, in Watts Bar Racks (Arrangement 4) ...... 24 6.9 Summary of the Analyses of the Postulated Accidents in the Watts Bar Spent Fuel Storage Racks ................................................................. 25 6.10 Comparison of the Reactivity of Fuel Assemblies Depleted with Different Burnable Poison R od T ypes ........................................................ . . ........... 26 6.11 Comparison of the Reactivity Effects of Depletion with Different Poison M aterials ..................................................................................... 27 Report HI-2012620 Project 90941
LIST OF FIGURES Figure 1 Minimum Burnup cf Spent Fuel in 2x2 Checkerboard Arrangement of Spent and Fresh Fuel of 4.95% Enrichment (Arrangement 2) .......................... 28 Figure 2 Minimum Burnup for Unrestricted Storage of Spent Fuel of Various Initial Enrichments (Arrangement 1) ............................................. 29 Figure 3 Comparison of the Reactivity of Fuel Assemblies Depleted with Different Burnable Poison Rod Types ...................................................... 30 Figure 4 Fuel Storage Cell Cross-Section ................................................. 31 Report HI-2012620 Project 90941 iii
1.0 INTRODUCTION
AND
SUMMARY
1.1 Obiectives and General Description The objective of the criticality safety analysis documented in this report is to evaluate the safe storage configuration of fresh and spent fuel assemblies in the Watts Bar Nuclear Plant spent fuel storage racks. This new analysis is performed with fuel assemblies containing tritium producing burnable absorber rods (TPBARs). Previous analysis performed by Holtec International [9]
determined the safe storage patterns for spent fuel in the racks for fuel containing no burnable poison rods. In addition to the ThBARs, the presence of other burnable poison rods such as WABAs and IFBA rods in the fuel assemblies has also been addressed in the present analysis.
Credit is taken for integral fuel burnable absorber (IFBA) rods and fuel bumup, where appropriate.
Soluble boron in pool water is used to protect against a mis-loaded assembly accident, where necessary. The analysis uses the KENO5a Monte Carlo code with the 23 8-group cross-section library developed by the Oak Ridge National Laboratory as the primary code for the calculations.
CASMO4 was used for calculation of fuel depletion effects and manufacturing tolerances. As permitted in the USNRC guidelines, parametric evaluations were performed for each of the manufacturing tolerances and the associated reactivity uncertainties were combined statistically. All calculations were made for an explicit modeling of the fuel and storage cell geometries to define the enrichment-burnup combinations for spent fuel configurations that assure a safe storage of fresh and spent fuel in the pool.
The following configurations of fresh and spent fuel storage in the Watts Bar racks have been analyzed in this report. The fuel was assumed to have initially contained Integral Fuel Burnable Absorber (IFBA) Rods and TPBARs, which are removed at the time the assemblies are placed in storage.
I. Storage of spent fuel with credit for burnup only.
- 2. Checkerboard of two fresh Fuel (initial enrichment of 4.95+/-0.05 wt%) and two spent fuel assemblies.
Report HI-2012620 Project 90941
- 1. Checkerboard storage of three fresh fuel assemblies (initial enrichment of 4.95=0.05 wt%)
and one cell containing only water or water and non-fuel materials.
- 4. Storage of fresh fuel, containing IFBA rods, in the racks with no other restrictions, other than that the assemblies contain at least 32 LFBA rods (1 .25x).
Postulated accident conditions, where a fresh fuel assembly without IFBA rods, is inadvertently placed into a cell intended to remain empty or to contain a spent fuel or fresh fuel assemblies with IFBA rods, have also been evaluated.
1.2 Summary of Results Arrangement I Previous analyses performed [Reference 9] showed that the required burnup for the spent fuel (initial enrichment of 4 .95+/-0.05wt%) in this configuration was 6.75 GWD/MTU. The required burnup for the fuel assemblies containing TPBARs remain the same. A summary of the calculations, for fuel with an initial enrichment of 4.95+0.05wt%, is given in Table 6.6. The required bumup for other initial enrichment is shown in Figure 2.
Arrangement 2 Previous analyses performed in [Reference 9] for a 2x2 checkerboard arrangement showed that the required burnup for the spent fuel (initial enrichment of 4.95+/-0.05 wt %) in this configuration was 20 GWD/MTU. In the present analysis, the required burnup for the fuel assemblies containing TPBARs remain the same. A summary of the calculations, for fuel with an initial enrichment of 4.95+/-0.05 wt%, is given in Table 6.5. The required bumups for other initial enrichments are shown in Figure 1.
Report HI-2012620 Project 90941
Arrangement 3 In this arrangement, 3 fresh fuel assemblies are checker boarded with 1 water cell in a 2x2 array.
This arrangement was found to be acceptable for fresh fuel storage without any additional restriction. A summary of the calculations, for fuel with an initial enrichment of 4.95+/--0.05wt%, is given in Table 6.7. Analyses were also performed to determine the limiting amount of water that can be displaced in order to checkerboard non-fissile bearing components (such as a boral coupon tree, thimble plug etc.) with fresh fuel. It was conservatively determined that 75% of water can be safely displaced in empty cells by~non-fissile bearing components. Cells containing items such as TPBAR consolidation baskets and baskets containing discarded materials may be considered water cells, as long as the material is non-fissile and no more than 75% of the water is displaced. These analyses also confirm that non-fuel bearing assembly components (i.e. thimble plugs, rod cluster control assemblies (RCCAs) etc.) may be stored in the fuel assemblies without affecting the storage requirements for the assemblies.
Arrangement 4 In this arrangement, fresh fuel assemblies containing integral burnable absorber rods (IFBA) are stored face adjacent to each other. The fuel assemblies were assumed to contain 16, 32 and 48 IFBA rods. Calculations show that the fuel assemblies containing a minimum of 32 IFBA rods can be stored in the storage cells without any credit for burnup, with a maximum k-,ffE* 0.95 including bias and uncertainties. A summary of the calculations, for fuel with an initial enrichment of 4.95+/-0.05wt% and containing 32 IFBA rods (at 1.25x), is given in Table 6.8. Assemblies with a greater number of IFBA rods would exhibit a lower reactivity.
Interface Requirements When arrangements 2 and 3 are placed adjacent to each other in the pool, there should be a barrier row of empty cells between the two arrays to prevent fresh fuel assemblies from being adjacent to each other in these arrays.
Report HI-2012620 Project 90941
Accident Condition Evaluation of postulated accident conditions demonstrate that 55 ppm of soluble boron in the spent fuel pool is sufficient to maintain kff <- 0.95, including calculational biases and all uncertainties under the most serious postulated fuel handling or mis-loading accident. Recent USNRC guidelines allow partial credit for soluble boron, and this would be more than adequate to protect against the most serious fuel handling accident. Normal soluble boron levels are maintained above 2000 ppm in the spent fuel pool.
Report HI-2012620 Project 90941 4
2.0 ANALYSIS CRITERIA AND ASSUMPTIONS To assure the true reactivity will always be less than the calculated reactivity, the following conservative analysis criteria or assumptions were used.
Criticalitv safety analyses were based upon an infinite radial array of cells. i.e., no credit was taken for radial neutron leakage. except for evaluating the rack boundaries accident conditions where neutron leakage is inherent.
Minor structural materials were neglected: i.e., spacer grids were conservatively assumed to be replaced by water.
The analyses assumed a temperature of 4 'C, which is the temperature of highest water density and highest reactivity in poisoned racks.
The analyses assumed a Westinghouse V5H 17x17 fuel assembly, which was found to be the most reactive of the fuel assembly types in use at Watts Bar Nuclear Plant, for the burnup appropriate to the analysis.
- The density of the fuel was assumed to be 97% of the nominal theoretical density, with a tolerance of+/- 2%.
" Boron-10 was used to simulate the Li-6 in the TPBARs, since CASMO-4 does not include Li-6 in the cross-section library. To accomplish this, the number density of B-10 was adjusted to give the same absorption cross section as the Li-6 by KENO-Va calculations.
This is a conservative assumption since the B-10 (1i-6) was not depleted.
"* No credit is taken for the presence of the Uranium-236 isotope in the fuel for this analysis.
"* No axial blankets were assumed to be present in the fuel rods. The entire active fuel length was assumed to have the same enrichment.
"* WABAs or TPBARs and IFBA rods were assumed to be present during the operating life of the fuel assemblies. This penalty is bounding for the fuel assemblies, which operate without poison rods.
Report HI-2012620 Project 90941 5
3.0 ACCEPTANCE CRITERIA The primary acceptance criterion is that the effective multiplication factor (k-,ff) of the racks shall remain less than or equal to 0.95, under normal conditions. The maximum k-.eff includes calculation uncertainties and reactivity effects of mechanical tolerances, under the postulated accident of the loss of all soluble boron. Applicable codes, standaras, and regulations, or pertinent sections thereof, include the following:
"* General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling.
"* USNRC Standard Review Pran, NUREG-0800, Section 9.1.2, Spent Fuel Storage.
" 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.
"* USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed), December, 1981.
"* ANSI-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.
" L. Kopp, "Guidance On The Regulatory Requirements For Criticality Analysis Of Fuel Storage At Light-Water Reactor Power Plants", USNRC Internal Memorandum L. Kopp to Timothy Collins, August 19, 1998.
Code of Federal Regulation 10CFR50.68, "Criticality Accident Requirements" Report HI-2012620 Project 90941 6
4.0 DESIGN AND INPUT DATA 4.1 Fuel Assembly and Component Design Soecifications Two different fuel assembly designs were considered in the analyses; the Westinghouse 17x 17 V5H and Robust designs. Table 4.1 provides the pertinent design details for the fuel assembly types.
Calculations were performed for fuel operating with both the TPBARs and the WABA components.
Design specifications for the TPBARs are obtained from Reference 7. The compositions of the fuel assemblies containing either IFBA Qr WABA rods were obtained from Reference 8.
4.2 Storage Racks The storage rack design is described in detail in Reference 9. A schematic of the fuel storage cell model, used in this analysis, is shown in Figure 4. The tolerances in the dimensions are also presented in Reference 9, and have been used in the present analysis.
4.3 Operating Parameters The core operating parameters for performing the depletion calculations were obtained from Reference 8. The principal core operating parameters, used in this study, are summarized in the table below.
Core Operating Parameters Value Fuel Temperature (°F) 1370 Moderator Temperature (F) 592 Average Soluble Boron in Moderator 700 (ppm)
Report HI-2012620 Project 90941 7
5.0 METHODOLOGY The criticality analyses were performed principally with the three-dimensional NITAWL-KENO5a Monte Carlo code package [1]. NITAWL was used with the 238-group SCALE-4.3 cross-section library and the Nordheim integral treatment for resonance shielding effects. Benchmark calculations, presented in Appendix A, indicate a bias of 0.0030 0.0012 (95%/95%) [2].
CASMO4., a two-dimensional deterministic code [4] using transmission probabilities, was used for depletion (burnup) calculations and to evaluate the small (differential) reactivity effects of manufacturing tolerances. Validity of the CASMO4 code was established by comparison with KENO5a calculations for comparable rack cases.
In the geometric model used in the calculations, each fuel rod, and associated cladding and each fuel assembly were explicitly described. Reflecting boundary conditions effectively defined an infinite radial array of storage cells. In the axial direction, a 30-cm water reflector was used to conservatively describe axial neutron leakage. Each stainless steel box and the water gaps [8, 9]
were described in the calculational model. The fuel cladding material was assumed to be zirconium.
Monte Carlo (KENO5a) calculations inherently include a statistical uncertainty due to the random nature of neutron tracking. To minimize the statistical uncertainty of the KENO5a calculated reactivities, a minimum of I million neutron histories were accumulated in each calculation, generally resulting in a statistical uncertainty of about +/-0.0003 Ak (Ica). Three-dimensional KENO5a calculations were necessary to describe the geometry of the checkerboard cases. However, KENO5a cannot perform depletion calculations. Depletion calculations were performed with CASMO4 with explicit description of the fission product nuclide concentration. To compensate for those fission product nuclides, which cannot be described in KENO5a, an equivalent boron-10 in the fuel was determined which produced the same reactivity in KENO5a as the CASMO4 result.
This methodology incorporates approximately 40 of the most important fission products, accounting for all but about 1% in k. The remaining 1 % in k is included by the equivalent B-10 concentration in the fuel.
Report 1[1-2012620 Project 90941 8
6.0 ANALYSIS RESULTS 6.1 Bounding Fuel Assembly Calculations were done, using CASMO4, to evaluate the reactivity of the fuel assemblies currently in use or anticipated for storage in Watts Bar spent fuel racks. Calculations show that the Westinghouse 17x17 V5H fuel assembly exhibits the highest reactivity at the burnups of interest in this analysis (from 0 to 35 GWD/MTU) and were used in all the subsequent calculations. Beyond 35 GWD/MTU burnup, the Westinghouse Robust fuel design becomes slightly more reactive, but this does not affect the present analyses.
W 17x17 W 17x17 Burnup, GWD/MTU 11 ROBUST VSH ROBUST 0 0.9792 0.9776 10 0.9132 0.9109 20 0.8629 0.8608 30 0.8193 0.8174 35 0.7990 0.7973 6.2 Evaluation of Manufacturing Tolerance Uncertainties CASMO4 calculations were made to determine the uncertainties in reactivity associated with density and enrichment tolerances. The uncertainties associated with the other mechanical tolerances have been assumed to be the same as that reported in the earlier analysis [9]. The reactivity effects of each independent tolerance were combined statistically. All fuel and rack dimensions and their dimensional tolerances are obtained from References 8 and 9. The reactivity effects of the tolerances are listed in Tables 6.5-6.8.
For estimating the reactivity uncertainties associated with tolerances in fuel enrichment and density, conservative tolerances of + 0.05% in enrichment and +/-2% in UO 2 density were assumed. The Report HI-2012620 Project 90941 9
reactivity uncertainty associated with the fuel density tolerance is listed in Table 6.1. The reactivity uncertainties for the tolerance in fuel enrichment are listed in Table 6.3.
6.3 Uncertainty in Depletion Calculations The uncertainty in depletion calculations is part of the methodology uncertainty and was taken as 5% of the reactivity decrement from beginning-of-life to the burnup of concern for the spent fuel
[5]. This methodology uncertainty is included in the calculations of the final k,ff in Table 6.5 and 6.6.
6.4 Uncertainty in TPBAR Loading Since CASMO4 does not include Li-6 (as used in the TPBARs), an equivalent boron was used to stimulate the absorption in Li-6. Since this approximation could introduce some uncertainty, a sensitivity analysis was made by doubling the boron concentration in the simulated TPBAR's*.
Results of this analysis showed that the effect on the residual reactivity was virtually negligible.
For the most sensitive storage configuration (checkerboard of 2 fresh assemblies with 2 assemblies burned to 20 GWD/MTU), doubling the TPBAR absorption resulted in only a 0.0005 Ak increase in reactivity, and would not affect the other configurations.
6.5 Eccentric Location of Fuel Assemblies The fuel assemblies are nominally stored in the center of the storage cells. Eccentric positioning of fuel assemblies in the cells normally results in a reduction in reactivity for poisoned racks.
Calculations have been made confirming negative reactivity effect of the eccentric positioning fuel assemblies at the position of closest approach. These calculations confirm that the normal centered position is the most reactive.
Report HI-2012O2 Project 90941 10
- The TPBARs are removed when the assembly is placed in storage. Therefore, the TPBAR composition only affects the residual reactivity after TPBAR removal 6.6 Temoerature and Void Effects Temperature effects were also evaluated, using CASNMO4, in the temperature range from 40 C to 120°C and the results are listed in Table 6.2. These results show that the temperature coefficient of reactivity is negative. The void coefficient of reactivity (boiling conditions) was also found to be negative for the Watts Bar racks.
6.7 Reactivity Effect of the Axial Burnuo Distribution Initially, fuel loaded into the reactor will burn 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 high neutron leakage.
Consequently, it would be expected that over most of the burnup history, distributed burnup 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.
The effect of the axial burnup distribution on reactivity was studied, on a generic basis, in detail by Turner [6]. Reference 6 indicates that below 30 GWD/MTU, the axial burnup penalty is negative.
Since all the required bumups in this analysis are substantially lower than about 30 GWD/MTU, an axial burnup penalty was not required in any of the four different storage patterns investigated.
Report HI-2012620 Project 90941 II
7.0 ACCIDENT CONDITIONS AND SOLUBLE BORON REQUIREMENTS Soluble boron is required to protect against the accident of a mis-loaded fuel assembly. The accident analyses corresponding to all the storage configurations investigated in this analysis are summarized below:
"* Fresh fuel assembly misloaded into a cell intended to store a spent fuel assembly in Arrangement I
"* Fresh fuel assembly misloaded into a cell intended to store a spent fuel assembly in Arrangement 2
"* Fresh fuel assembly misloaded into a location intended to be a water cell in the Arrangement 3
- . Fresh fuel assembly, without any IFBA, misloaded into a cell intended to store a fresh fuel assembly with 32 IFBA rods, in Arrangement 4 Table 6.9 summarizes the kerr for each of these accident analyses. The results show that the most serious postulated accident condition with the misplacement of a fresh fuel assembly occurs in arrangement 3. In this case, a fresh fuel assembly is misplaced in the location of a water cell.
Calculations show that 55 ppm of soluble boron would be required to maintain the keff in the rack below the regulatory requirement of 0.95, including bias and uncertainties. Misplacement of a fuel assembly outside the periphery of a storage module, or a dropped assembly lying on top of the rack would have a smaller reactivity effect.
Report HI-2012620 Project 90941
,1
8.0 OTHER BURNABLE POISON ROD INSERTS IN THE FUEL ASSEMBLIES The fuel assemblies used at the Watts Bar may contain poison rods other than the TPBARs.
Analyses show that the fuel assemblies containing TPBARs are more reactive than those containing BPRAs and are essentially the same as those with WABAs, at the burnups of interest. At higher bumups, the fuel assemblies with TPBARs exhibit higher reactivity. The results are summarized in Table 6.10, and illustrated in Figure 3.
With IFBA rods present, similar calculations show that the WABA case yields a slightly higher reactivity than the TPBAR case. These results are tabulated in Table 6.11. The difference does not affect the results given in this report. The spent fuel calculated without IFBA present bounds all other cases.
Report HI- 2012620 Project 90941
9.0 CRITICALITY ANALYSES RESULTS AND CONCLUSIONS Four different storage configurations of fresh and spent fuel assemblies in the Watts Bar spent fuel pool have been evaluated in this analysis. The results indicate that these storage patterns of fresh fuel assemblies (4.95--'0.05 wt%0 enrichment) and spent fuel assemblies meets the regulatory requirements. The results for the different arrangements are summarized in Tables 6.5 to 6.8.
Results show that the burnup requirements for the storage arrangements 1 and 2 remain the same as those reported in Reference 9. A summary of the conditions evaluated and the conclusions are given below:
- Spent fuel assemblies may ;e stored in unrestricted locations provided that they satisfy the burnup-enrichment combinations identified in Figure 2 (minimum of 6.75 MWD/Kg-U burnup for fuel of 4.95+/-0.05 wt% initial enrichment). Fuel of 3.8 wt% or less U235 may be also stored without restrictions.
- Storage of two fresh fuel assemblies (4.95+/-0.05 wt% initial enrichment) in a 2x2 checkerboard array with two spent fuel assemblies, whose burnup-enrichment combination is shown in Figure 1 (minimum of 20 MWD/Kg-U burnup for fuel of 4.95+/-0.05 wt% initial enrichment), satisfy the regulatory requirements.
- Checkerboard arrangement of 3 fresh fuel assemblies and 1 empty cell satisfy the regulatory requirements for fuel storage in the racks.
- Fresh fuel assemblies, of 4.95-0.05 wt% initial enrichment, containing a minimum of 32 (1.25x) IFBA rods may be stored face adjacent to each other in the spent fuel storage racks.
These may also be stored face adjacent to spent fuel assemblies satisfying bumup enrichment combinations in Figure 2 (minimum of 6.75 MWD/Kg-U bumup for fuel of 4.95+/-0.05 wt% initial enrichment).
- A water cell will always be less reactive than an irradiated fuel assembly. Conservatively, 75% of the water may be safely displaced from a cell by non-fissile materials and the cell may still be considered a water cell.
- Accident analysis show that only 55 ppm of soluble boron is required to mitigate the effects of the most serious postulated fuel misplacement and maintain the keff below 0.95, including all uncertainties and biases.
Report HI-2012620 Proiect 909411
. . vd .........
14
10.0 REFERENCES
- 1. R.M. Westfall, et. al., "NITAWL-S: Scale System Module for Performin2 Resonance Shielding and Working Library Production" in SCALE: A Modular Code System for Performing Standardized Comouter Analyses for LicensinR Evaluation., NUREG/CR- 0200, 1979.
L.M. Petrie and N.F. Landers., "KIENO Va. An improved Monte Carlo Criticality Program with Subgrouping" in SCALE: A Modular Code System for Performing Standardized Comouter Analyses for Licensing2 Evaluation., NUREG/V-0200, 1979.
- 2. M.G. Natrella, Exoerimental Statistics, National Bureau of Standards, Handbook 91, August 1963.
- 3. J.F. Briesmeister, Ed., "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A", Los Alamos National Laboratory, LA- 12625-M (1993).
- 4. A. Ahlin, M. Edenius, H. Haggblom, "CASMO- A Fuel Assembly Burnup Progam," AE RF-76-4158, Studsvik report (proprietary).
A. Ahlin and M. Edenius, "CASMO- A Fast Transport Theory Depletion Code for LWR Analysis," ANS Transactions, Vol. 26, p. 604, 1977.
D. Knott, "CASMO4 Benchmark Against Critical Experiments", Studsvik Report SOA 94/13 (Proprietary).
M. Edenius et al., "CASMO4, A Fuel Burnup Program, Users Manual" Studsvik Report SOA/95/l.
- 5. L. Kopp, "Guidance On The Regulatory Requirements For Criticality Analysis Of Fuel Storage At Light-Water Reactor Power Plants", USNRC Internal Memorandum, L. Kopp to Timothy Collins, August 19,1998.
- 6. S. E. Turner, "Uncertainty Analysis - Burnup Distributions," Presented at the 1988 DOE/SANDIA Technical Meeting on Fuel Burnup Credit
dated August 21, 2000.
- 8. TVA Letter 30M400 to Holtec International, dated January 8, 2001
- 9. Holtec Report HI-961513, Revision 2, Holtec Project Number 10371 Report -I-2012620 Project 90941 t5
Table 4.1 Design Basis Fuel Assembly Specifications 181 W 17xI7 W 17x17 Fuel Assembly V511 ROBUST Clad O.D., in 0.374 0.374 Clad I.D., 11 0.329 0.329 Clad Material Zircaloy-4 Zirlo Pellet Diametcr, in 0.3225 0.3225 00 Density, g/cc 10.631 10.631 Maximum Enrichment /0 4.95+/-0.05 w/o 4.95+/-0.05 w/o Active Fuel Length, 1n 144 144 Ntumber Fuel Rods 264 264 Fuel Rod Pitch 0.496 0,496 Ntumber of Thimbles 25 25 Thimble O.D. 0.474 0.482 Thimble I.D. 0.442 0.442 l*,epoI111-20 12620 Ploicot 90)9,11 16
Table 6. 1. Reactivity Effects of Density Tolerance in the Watts Bar Spent Fuel Racks.
BURNUP, REFERENCE FUEL I)ENSITY GWi)/l\TII ki, r ki,,f Ak 0 0.9776 0.9795 0.0019 10 0.9107 0.9122 o' 0.00 15 20 0.8606 0.8621 0.0(0 15 30 0.8173 0.8189 0.(1016 35 0,7971 0.7989 (0.0018 Report 111-2012620 loiccl .
90)9A I 17
Table 6.2 Reactivity Effects of Temperature and Void in Watts Bar Spent Fuel Racks.
BURNUP, T=4 C 'r 20 °C T= 1200 C T= 121) 'C + VOII)
GWD/NIT ki,, kilf Ak' kif Ak* kilt Ak**
0 0.9792 0.9776 -0.0016 0.9548 -0.0228 0.9262 -0.0286 10 0.9132 0.9107 -0.0025 0.8895 -0.0212 0.* 19 -0.0276 20 0.8629 0.8606 -0,0023 0.8404 -0.0202 0.8134 -0.0)270 30 0,8193 0.8173 -0.0020 0.7984 -0.0189 0.7720 -0.0264 35 0.7990 0.7971 -0.0019 0.7790 -0.0181 0.7529 -0.0261 difference with rusults (6 4 TC difference with results (5) 20 T(
- difference with results at 120 TC mnd no void Report 111-2012620 l oject )(9)94 I 19
Table 6.3 Reactivity Effects of Fuel Enrichment Tolerance in Watts Bar Spent Fuel Racks.
BURNUP, REFERENCE ENRICHMENT TOLERANCE G\VI)/NTU kil,* killf Ak 0 0.9776 0.9793 P.00 17 10 0.9107 0.9124 0.0017 20 0.8606 0.8624 0.0018 30 0.8173 0.8191 0.0018 35 0,7971 0.7990 0.0019
,cport 111-2012620} h)oicc! 90{4 11
Table 6.4 Reactivity Effects of Abnormal and Accident Conditions in Watts Bar Spent Fuel Racks ACCIDENT/ABINORMAI*ACONDITIONS REACTIVITY EFFECT Temperature increase (See Table 6.2) Negative Negative Void (Boiling) (See Table 6.2)
Misplacement of a fresh fuel assembly Positive: most serious misplacement accident requires 55 ppm soluble boron Eccentric Positioning of Fuel Assemblies Negative I111-2012620 ojlCct 9()9'01 20
Table 6.5 Summary of the Criticality Safety Analyses for Checkeiboard Storage of 2 Fresh and 2 Spent Fuel Assemblies In Watts Bar Racks (Arrangement 2).
Reference k.,, 0.9233 Required Burnup of the Spent Fuel Assemblies 20 GWD/MTU Keno5a Bias 0.0030 Temperature Correction to 4 " 0.00230 Axial Burnup Distribution Penalty Not Applicable KENOSa Bias Uncertainty 0.0012 KENO Statistics (95/95) Uncertainty (1.7 *a) 0.0009 Mechanical Tolerance Uncertainty 0,0059 Density Tolerance Uncertainty 0.0019 Enrichment Tolerance Uncertainty 0.0018 Depletion Uncertainty 0.0059 Fuel Eccentric Positioning Uncertainty Negative Statistical Combination of Uncertainties 0.0089 Maximurm k 1ir 0.9375 Regulatory Limiting kyn. 0.9500 Repoit 111-2012620 locuIt 909)41 21
Watts Bar Racks Table 6.6 Summary of the Criticality Safety Analyses for Face Adjacent Storage of Spent Fuel Assemblies In (Arrangement I Reference k,,rr 0.9271 Required Burnup of the Spent Fuel Assemblies 6.75 GWD/MTU for 4.95+/-0.05 wt% initial enrichment Keno5a Bias 0.0030 Temperature Correction to 4 TC 0.0022 Axial Burnupt Distribution Penalty Not Applicable KENOSa Bias Uncertainty 0.0012 KENO Statistics (95/95) Uncertainty (1.7 *(5) 0.0009 Mecharilcal Tolerance Uncertainty 0.0059 Density Trolerace Uncertainty 0.0016 Enrichment Tolerance Uncertainty 0.0017 Depletion Uncertainty 0.0023 Fuiel Eccentricity Uncertainty Negative Statistical Combination of Uncertainties 0.0069 Maximumfn k~ý 0.9392 Regulatory Limiting k,1r 0.9500 Reactivity dominated Ib once-buined assemblies, which suippresses the axial bturnup pcnlty.
Itepoit 111-2012620 P[ iolcct 909-11 22
I Water Cell Table 6.7 Summary of the Criticality Safety Analyses for Checkerboard Storage of 3 Fresh Fuel Assemblies and in Watts Bar Racks (Arranigemnent 3)
Reference k,11. 0.9131 Keno5a Bills 0.0030 Temperature Correction to 4 `C 0.0016 Axial BLurnup Distribution Penalty Not Applicable KENO5a Bias Uncertainty 0.0012 "
KENO Statistics (95/95) Uncertainty 1.7 cy) 0.0010 Mechanical Tolerance Uncertainty 0.0059 Density Tolerance Uncertainty 0.0019 Enrichment Tolerance Uncertainty 0.0017 Depletion Uncertainty Not Applicable Fuel Eccentricity Uncertainty Negative Statistical Combination of Uncertainties 0.0066 Maxinun k,1. 0.9243 Regulatory Limiting kerr 0.9500 Report 111-2012620 I'l l*ccot 909,11 23
Table 6.8 Summary of the Criticality Safety Analyses for Storage of Fresh Fuel Assemblies, containing 32 IFBA rods, in Watts Bar Racks (Arrangement 4).
Reference k,11 0.9365 Keno5a Bias 0.0030 Temperature Correction to 4 "C 0.0016 Axial B urnup Distribution Penalty Not Applicable KENO5a Bias Uncertainty 0.0012 KENO Statistics (95/95) Uncertainty (1.7 *'h) 0.0010 Mechanical Tolerance Uncertainty 0.0059 Density Tolerance Uncertainty 0.0019 Enrichment Tolerance Uncertainty 0.0017 Depletion Uncertainty Not Applicable Fuel Eccentricity Uncertainty Negative Statistical Combination of Uncertainties 0.0066 Maximum k~rt 0.9477 Regulatory Limiting kf1- 0.9500
.cpoi t 11i -2012620 Project 9(09,11 24
Tablc 6.9 Summary of the Analyses of the Postulated Accidents in the Watts Bar Spent Fuel Storage Racks.
Description of Accident K,.ff Calculation Biases, Total (,drf Penalty and Uncertainties Fresh fuel assembly misloaded in the location of a spent fuel assembly in Arrangement I (face 0.9292 0.0121 0.9413 adjacent storage)
Fresh fuel assembly misloaded in the location of a spent fuel assembly in Arrangement 2 (checkerboard 0.9292 0.0140 0.94-32 loading)
Fresh fuel assembly misloaded in the location of a 0.9435 0.0112 0.9547 3
Arranrgement water cell in Fresh fuel assembly, withou~t IFBA rods, misloaded in the location of a fresh fuel assembly, with 32 IF13A 0.9375 0.0112 0.9487 rods, 'in Arrangement 4
'*.c ))),I RepOit 111-2012620 25b
Table 6. 10 Comparison of thie Reactivity of Fuel Assemblies Depleted with Different Burnable Poison Rod Types.
BURNUP, W-V511 with W-V511 with W-V511 WITH G'VDI)/NYI[J 'I'l 1 A R IPRA W ABA kill, kill, kii~r 0 0 9792 0.9792 (1*97)2 10 0.9132 0.9120 0.9137 20 0.8629 0.8581 0.8634 30 0,8193 0.8087 0.8177 35 0,7990 0.7833 0.7926 Report 111-2012620 Iloject 9()9.11 26
Tab Ie 6.11 {Comparison of the Reactivity Effects of Depletion with Different Poison Materials k-,rt in rack I I I II"IA No No Yes Yes B1Urn up, TPBIAR No Yes No MWD/KgU WA13A Yes No No Yes 0,9792 0.9792 0.9402 0.9,102 10} 0.9137 0.9132 0.9016 0.9018 15 0. 8876 0.8869 0.8808 0.8813 20 0.8634 0.8629 0.8600 0.8604 25 0.8403 0.8405 0.8392 0 8390 30 0.8177 0.8193 0.8188 0.8173 35 0.7926 0.7990 0.7990 0.7951 Report 111-20)12620 I ojoct 90)9'1 27
20 18 1
1 I'I
_ _ __ /
8-UNACCiPTABLE BUR NUP 4i 21 2.50 3.00 3.50 4.00 4.50 5.00 Initial Enrichment, w! U-235 Figure 1 Minimum Burnup of Spent Fuel in 2x2 Checkerboard Arrcmgement of Spent and Fresh Fuel of 4.95% Enrichment (Arrangement 2) 28 Prolect 90941 Report HI-2012620
7.0 6.5 6.0 5.5 5.0 U) 4.5 0.
4.0 3.5 I.-
3.0 2.5 2.0 1.5 1.0 0.5 0.0 In~fial Enrichment, wz U-235 Figure 2 Minimum Burnup for Unrestricted Storage of Spent Fuel of Various Initial Enrichments (Arrangement 1) 29 Project 90941 eport H1-201220 '
0.980 0.920 -
0 0 0.900 0.880 - _ _
0 0.860" 0.840 4
N 0.820 0.800 TPA_
WAB 0.780 ___ B 0.760 ,- ,, ,,,,-n-,T ,1 ,,,7-7 ,, 1r- - I,,
0 5 10 15 20 25 30 35 4C Burnup, MvWD/KgU Fig. 3 Comparison of fhe Reacfivity of Fuel Assemblies Depleted with Different Burnable Poison Rod Types 30 Project 90941
- eDort HI-2012620
8.625" BORAL 0.10" THK 0.0233 m -1/ 2 SFUE.L OD iu~Lo ~.
Z2 IN, QZZ25 he~h0.036" DENSITY 10.631 g/cc 0 NUHRER 264 (17 X 17) 0 0CLAb Iu 0,3290 IN. Water Gap OQ0 " ROd 0D 0.374 IN. - - - - - -
0000 PITdH 0,496 IN. o.973.
OOOOOOQ*
000000' THIOBLEES ID*so*o**
00 0.7 IN. N OO00 90.Q NO. 25 000000000, OOOOOlOO'OOO > ,o 00 0 0000kO@ 0O, O000OOo0.0 Q~, ,. , .* I O00OOO@OOO00OOC 00000000000000c, O00030OO(O 8,ooO 1:0 i
-- - 0.090" ýS EOX PITCH NOT TO SCALE 10.375" FIGURE 4 FUEL STORAGE CELL CROSS SECTION eport HI-2012620 31 Project 90941
APPENDIX A List of Holtec's QA Approved Computer Codes List The list of Holtec's QA approved computer codes consists of all the codes that have been developed or verified by Holtec International for its use in nuclear safety-related applications.
This information, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, analysis and licensing of a similar product. This list is, therefore, deemed proprietary and is not presented in this non-proprietary version of HI-2012620.
Report HI1-2012620 Project 9094!
A-I
"41, APPENDIX B: BENCHMARK CALCULATIONS (Total of 26 Pages Including This Page)
Note: This appendix was taken from a different report. Hence, the next page is labeled "Appendix 4A, Page 1".
Report H 1-20 12620 Project 90941 B-1
APPENDIX 4A: BENCHMARK CALCULATIONS 4A. I INTRODUCTION AND SUMMARy Benchmark calculations have betn made on selected critical exeriments, 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.11 is a continuous energy Mot-te Carlo code and KENO5a [4A.21 uses group-dependent cross sections. For the KENO5a analyses reported here, the 238 group library was chosen, proqssed 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 errors*
(trends) that have been reported (e.g., [4A.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 "ýB 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 analysez.
Table 4A. I 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 sections 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. KENO5a 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 KENO5a. the number of fissions in each group may be collected and the EALF determined (post-processing).
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.
Appendix 4A, Page 1
Figuares 4A. 1 and 4A.2 show the calculated k., for the benchma1rk critical exeriments as a function of the PA.\LF for MCNP4a and KENO5a, respecively (UO, fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental error 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 Figtres 4A. 1 and 4A.2 show that there are no trends, as evidenced by very low values of the correlation coefficient (0.13 for MCN?4a and 0.21 for KENO5a). The'total bias (systematic error, or mean of the deviation from a lyr of exactly 1.000) for the tw7o methods of analysis are shown in the table below.
Calculational Bias of MCNP4a and KENO5a MCNP4a 0.0009 +0.0011 KENOSa. 0.0030+/-0.0012 The bias and standard error of the bias were derived directly from the calculated k~r values in Table 4A. I 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. 181 (for thenumber of cases analyzed, the K-factor is -2.05 or slightly more than 2).
k-I k, (4A.1)
A classical example of experimental error is the corrected enrichment in the PNL experir-ncnts, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods.
These equations may be found in any standard text on statistics, for cxaample, reference
[4A.61 (or the MCNP0a manual) and is the same methodology used in MCNP4a and in KENOSa.
Appendix 4A, Page 2
___ ___ ___ __- ___(-*A.2)
O- = _ _ _ _ _ _
Bias (I- k) K a7 (4-A.3) where k, are the calculated reactivities of n critical experiments; 0, 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 naltiplier for 957%probability at the 95 % confidence level (NBS Handbook 91 [4A.181)."
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, ( I- R), is the actual bias which is added to the MCNP4a and KENOSa results.
The second term, Kaý, 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 951% 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 +/-he maximum k-,, 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 k,, 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 KENOSa). Thus, there are no corrections to the bias for the various enrichments.
Appendix 4A, Page 3
As further confirmation of the absence of any trends with enrichment, a typical configuration was calculated with both MCNP4a and KENOSa for various er'ichments.
The cross-comDarison of calculations with codes of comparable sophistication is suggestd 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 k, for the two independent codes as evidenced by the 450 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 '0 B 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 erron 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.!
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 "°B concentration in the absorber, a cross-comparison was made with MCNP4a and KENOSa (as suggested in Reg.
Guide 3.41). Results art 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 450 line, within an expected 95% probability limit).
The reactivity worth of the absorber panels was deterrmined by repeating the calculation with the absorber analytically removed and calculating the incremental (Ak) change in reactivity due to the absorber.
Appendix 4A, Page 4
4A.4 SMiscellaineous and Mincr Parameters 4A.4. 1 Rtflector Material and Spacings PN,\L has pcrformed a number of critical experiments with thick: steel and lead rfle- ors Analysis of these critical experiments are listed in Table 4A.5 (subset of data in Table lower 4A. 1). There aprpears to be a small tendency toward overprediction of k._, at the each series to allow a soacing, although there are an insufficient number of data points in at close quantitative determination of any trends. The tendency toward overprediction than otherwise.
spacing means that the rack calculations may be slightly more conservative 4A.4.2 Fuel Pellet Diameter and Latce Pitch from The critical experiments selected for analysis cover a range of fuel pellet diameters the rack designs, 0.311 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches- In to 0.580 inch the fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.496 to 0.740 inch lattice spacing) for PWR fuel and from 0.3224 to 0.494 inches O.D. (0.488 a reasonable lattice spacing) for BWR. fuel. Thus, the critical experiments analyzed provide does not representation of power reactor fuel. Based on the data in Table 4A.1, there pitch, at least appear to be any observable trend with either fuel pellet diameter or lattice over the range of the critical experiments applicable to rack designs.
4,A.*4.3 Soluble Boron Concentration Effects experiments Various solubie boron concentrations were used in the B&W series of critical up to 2550 ppm. Results of and in one PNL experiment, with boron concentrations ranging MCNIP4a (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 would overpredict reactivity for the three experiments exceeding 1300 ppm. In turn, this be suggest that the evaluation of the racks with higher soluble boron concentrations could slightly conservative.
but not Parallel experiments with a depleted uranium reflector were also performed rack design.
included in the present analysis since they are not pertinent to the Holtec Appendix 4A, Page 5
4A.5 Tae number of critical experiments with PuO, beaing fuel EIOX') 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 k- of 1.00, indicating that when Pu is present, both MCNTP4a and KENO5a 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 sMacings.
the KENO5a calculated rmactivities are below 1.00, suggesting that a small trend may exist with KENO5a. It is also possible that the overprediction in km for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and km-241 growtn. This possibility is supported by the -fonsistency in calculated k-, over a wide range of the s~ectral index (energy of the average lethargy causing fission).
Appendix 4A, Page 6
4A. 6 R, fere S
[4A. 1] J.F. Briesmeister, Ed., "MCN-P4a - A General Monte Carlo N Particle Transport Code, Version 4A; Los Alamos National Laboratcry, LA-12625-M (1993).
[4A.21 SCALE 4.3, "A Modular Code System for Performing Standardized Computzr Analyses for Licensing Evaluation", N"REG-0200 (ORNL-NUREG-CSD-2/U2fR5. Revision 5, Oak Ridge National Laboratory, September 1995.
[!4A.31 M.D. DeHart and S.M. Bowman, "Validation of the SCALE, Broad Structuft 44-G Group ENDFfB-Y Cross-Section Library for Use in Criticality Safety Analyses", NREG/iCR-6102 (ORNLITM-12460)
Oak Ridge National Laboratory, September 1994.
[4A.41 W.C. Jordan et al., 'Validation of KENOV.a", CSDFTM-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 Analysis", ORNL-TM-,12667. Oak Ridge National Laboratory, undated.
[4A.6] R.J. Larsen and M.L. Marx, An Introduction to Mathematcal Statistics and its A.plications, 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 Gadoli.a: Nuclear Model Development and Critical Experiment Benchmark, BAW-18 10, Babcock and Wilcox Company, April 1984.
Appendix 4A, Pagc 7
[4A. 10] J.C. Manm-anche 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. l1] S.R. Bierman and E.D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o "U Enriched UO.
Rods 'MWater with Steel Reflecting Walls, PNL-3 60-2, Bartelle 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 2U Enriched UO, Rods in Water wjth Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December, 198 1.
[4A. 131 S.R. Bierman et al., Critical Separation Between Subcritical Clusters of 4.31 wlo "U Enriched U0 2 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratry, October 1977.
[4A. 41 S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990.
[4A.151 B.M. Durst et al., Critical Experiments with 4.31 wt % "'U Enriched UO, 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. 17] 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. NatreIla, Ex'._erimental Statistics, National Bureau of Standards, Handbook 91, August 1963.
Appendix 4A, Page 3
Table 4A.1 Summary of Criticality Benchmark Calculations c ,akculat*hl.,
MvCNP4a KEN05ii MCNP4a KENO5a Identlficatlon Enrich.
Reference 0.1759 0.1753 2.46 0.9964 +/- 0.0010 0.9898+/- 0.0(
B&W-1484 (4A.7) Core 1 0.2553 0.2446 2.46 1,0008 +/- 0.0011 1.0015 t 0.0i 2 B&W-1484 (4A.7) Core 1I 0.1939 0.1999 2.46 1.0010 +/- 0.0012 1.000FV+/- 0.04 3 B&W-1484 (4A.7) Core 111 0.1422 0.1426 2.46 0.9956 +/- 0.0012 0.9901 +/- 0.0i 4 B&%V-1484 (4A.7) Core IX o.1513 0.1499 2.46 0.9980 +/- 0.0014 0.9922 +/- 0.04 5 B&W-1484 (4A.7) Core X 0.2031 0.1947 2.46 0.9978 +/- 0.0012 1.D0005 +/- 0.0(
6 B&W-1484 (4A.7) Core XI 0.1719 0,1662 2.46 ().99"8 +/- 0.0011 0.9978 +/- 0.0D 7 Core XH B&W-1484 (4A.7) 0.1988 o.1965 2.46 1.0020 +/- 0.0010 0.9952 +/- 0.0(
8 B&W-1484 (4A.7) Core XIII 0.1986 0.2022 2.46 0.9953 +/- 0.0011 0,9928 +/- 0.0(
9 B&W-1484 (4A.7) Core XIV 0.2092 0.2014 2.46 0.9910'+/- 0.0011 0.9909 +/- 0.0(
10 B&W-1484 (4A.7) Core XV "
0,1713 0.1757 2.46 0.9935 +/- 0.0010 0.9889 +/- 0.0(
11 B&W-1484 (4A.7) Core XVI " 0.2021 0.2083 2.46 0.9962 +/- 0.0012 0.9942 +/- 0.0(
12 B&W-1484 (4A.7) Core XVH 0.1705 0.1708 2.46 1.0036 +/- 0.0012 0.9931 +/- 0.0(
13 B&W-1484 (4A.7) Core XVIII
- -Appendix 4A, I'af, 9
Table 4A. 1 Calculations Summary of Criticality Benclhmark
,tCNI'4a KENO5a KENO5 Dnrich. MC?{P4a 2.46 0.9961 +/- 0.0012 0.9971 +/- 0.0005 0.2103 0.2011 Reference Core XIX 0.1724 0.1701 14 B&W-1484 (4A.7) 0.9932 +/- 0.0006
- - 2.46
1.0008 +/- 0.0011 0.9918 f 0 ,o00M 0 1 44 01 3 Core XX B&W-148 4
(4A.7) 0 .1544 0.1536 15 2.46 0.9994 :L 0.0010 Core XX1 1.4680 16 BI&W-1484 (4A.7) 0.9924 +/- 0.0006 1.4475 2.46 0.9970 +/- 0.0010 S-type Fu 1.5660 B&W-1645 (4A.8) 0.9913 +/- 0.0006 1.5463 17 0.9990 : 0.0010 2.46 S-type Fu Fuel, winl 56 ppm B 2.46 0.9972 +/- 0.0009 0.9949 j 0.0005 0.4241 0.4331 18 1B&W-1645 (4A.8)
ýuejpp w/ iS --- -- 01 3 NC _
19 B&W,-1645 (4A.8) SO-type NC 2.46 1.0023 +/- 0.0010 Case 1 1:337 ppm B 2o B&W-1810 (4A.9) N(
NC 0.4493 Case 12 1 2.46/4.02 1.0060. +/- 0.0009 Ne 899 ppi B 02172 21 II&\V-1810 (4A.9) S----------o 0_ ___2 NC 0 . 7 8N Water M o d er ut ur 0 ga p 4.75 0.9966 +/- 0.0013 22 French (4A.10) NC 0.1778 NC Water M oderalor 2.5 cm gap 4.75 0.9952 +/- 0.0012 French (4A.10) c-a -dertor 4.75 0.9943 +/- 0.0010 NC 0.1677 NC 23 gap ---------- ._.__3. N Water M o)derator 5 cn' 24 French (4A.10) 0.1736 NC NC Water M 4.75 0.9979 +/- 0.0010 oderaflr 10 c"I gap 25 French (4A.10) NC 0.1018 NC 1.0004 +/- 0.0006 Steel Ref 2.35 lector, 0 separation 26 PNL-3602 (4A.1I1)
Appecndix ,4A, 1,1w. I(0
Table 4A.1 Summary of Criticality Benchiuark Calculations C "I1di.uhrrl ,._ EA _IrL_)-
KENO5a ¢ICNI14a KFNO5a Enrich. MCNP4a Identification 0.100:I 0.0909 Reference 0.9992 +/- 0.0006 2.35 0.9980 +/- 0.0009 27 Steel Reflector, 1.321 cm sepn.
PNL-3602 (4A.11) 0.9964 +/- 0.0006 0.0981 0.0975 28 PNI-3602 (4A. 11) Steel Reflector, 2.616 cm sepn 2.35 0.9968 +/- 0.0069 0.9974 +/- 0.0010 0.998,0 +/- 0.0006 0.0976 29 Steel Reflector, 3.912 cm sepn. 2.35 0 OO*Q + 0.0006 0.0973 0.0968 PNL-3602 (4A.11) ....
Steel Reflector, infinite set)1. 2.35 O.MLO/ +- U.-.
30 PNL-3602 (4A.11)
NC 1.0013 :1 0.0007 NC 0.3282 Steel Reflector, 0 cm sepn. q.. tW I0.3039 31 PNLI-3602 (4A. 11) 0.3039 32 PNL-3602 (4A. 11) Steel Reflector, 1.321 cm sepn. 4.306 0.9997 +/- 0.0010 4.300.94+/-001 1.0012
....... +/-t 0.0007 u u011 1 0.3016 0.2927 4.306 0.9994 +/- 0-0012 0.99,74 1: o.-r ...
33 PNIl3602 (4A. 11) Steel Reflector, 2.616 cm scin.
0.9969 +/- 0.0011 0.9951 +/- 0.0007 0.2828 0.2860
. , A I" 1111
"(nl t3aI Steel Reflector, 5.405 cm sepi. 4.306 34' 0.9947 +/- 0.0007 0.2851 0.2864 4.306 0.9910 +/- 0.0020 0.3135 0.3150 35 1PNL-3602 (4A.] 1) Steel Reflector, Infinite sepn. -. - ."" A -n W7 1n 4.306 0.9941 +/- O.o 11 . . .
36 PNM13602 (4A.11) Steel Reflector, with Boral Sheets n -' - 0 0007 1* NC 0.3159
-.3.,I.
Lead Reflector, 0 cm sepn. 4.306 Nt. . . ,-.. . 0.3044 37 PNL-3926 (4A.12) 4.306 1.0025 +/- 0.0011 0.9997 +/- 0.0007 0.3030 38 PNl-3926 (4A.12) Lead Reflector, 0.55 cm sepn. 0.2930 0.0012 0.9985 +/- 0.0007 0.2883 4.306 1.0000-39 IPNI3926 (4A.12) Lead Reflector, 1.956 cm sepn.
Appendix ,4A. Page I I
Table 4A.1 Calculations Summary of Criticality Benchmark NICNi'4a KENOSR MCNP4a KENOSa Enrich.
Identiflcation 0,2854 Reference 0.2831 0.9971 +/- 0.0012 0,9946 +/- 0.0007 4.306 40 IPNI-3926 (4A.12) l.ad Reflector, 5.405 cm sepn.
0.9950 +/- 0,0007 0.1155 0.1159 4.306 0.9925 +/- 0.0012 41 i'N1,2615 (4A.13) Experiment 004/032 - no absorber 0.911 +/- 0.0007 NC 0.1154 4.306 NC Experiment 030 - Zr plates 412 IINL-2615 (4A. 13) NC 0.9965 +/- 0.00(17
-Steel plates 4.306 NC 0.1164 Experiment 013 43 PNL-2615 (4A.13) NC 0.9972 +/- 0.000
-Steel plates 4.306 NC 0.1164 E-xperimuent 014 44 PNL-2615 (4A.13) 0.9982 :L:0.0010 0.9981 +/- 0.0007 4.306 0.1162 Exp. 009 1.05% Boron-Steel plates 0.1172 45 PNL-2615 (4A.13) 0.9982 +/- 0.0007 0.1161 0.1173 plates 4.306 0.9996 +/- 0.0012 46 PNL-2615 (4A. 13) Exp. 012 1.62% Boron-Steel 0.1171 o.1165 0.9994 +/- 0.0012 0.9969 +/- U.0(Y7 4.306 47 PNI-2615 (4A. 13) Exp. 031 - Boral plates 0.3722 0.3812 0.9991 +/- 0.0011 0.9956 +/- 0.0007 trap 4.306 48 PNL,-7167 (4A.14) Experiment 214R - with flux 0.3826 0.3742 0.9969 +/- 0.0011 0.9963 +/- 0.0007 trap 4.306 49 PNL-7167 (4A. 14) Experiment 214V3 - with flux NC 0,2893 NC 4.306 0.9974 t 0.0012 5o PNL-4267 (4A.15) Case 173 - 0 ppm B NC 0.5509 NC 4.306 1.0057 +/- 0.0010 51 pNL-4267 (4A. I5) Case 177 - 2550 ppm B 0.8868 0.9171 1.0041 +/- 0,0011 1.0046 +/- 0.0006 21 20% Pu 52 PNI,-5803 (4A.16) MIOX Fuel - Type 3.2 Exp.
Appendix 4A, Pagt: 12
"Table 4A.1 Calculations Summary of Criticality Benchmark I kuwt"fl- V,a !w I (PV) hCNP4a KENO5U MCNP4a KENOSa Iderntification tEnrich. -- .294---
0.2944 4--
1 0016 + 0 .0006 0.2968 lReference ... PU. ...
20% j .
MOX Fuel - Type 3.2 Exp. 43 20% Pu 1.0)058 +/-j U.trUU , .... .
53 PNL-5803 (4A. 16) 20% Pu 1.0083 +/- 0.0011 0.9989 +/- O.006 .1665 0.170,6 3 MOX Fud - Type 3.2 Exp. 13 .09+/-001 096 .06j~ 0.1165 54 PNI:580 (4A.16) 20% PO
,*AVi*' lei - Type 3.2 Exp. 32 20% Pu 1.0079 +/- 0.0011 0.9?66 t 0.0006 l.*
0.8417 SS PNI-5803 (4A. 10J .. _ 00
. . .P, C'. p ,O0 20.52" nitch 6.6 %Pu----1------6 Pu 00 . ".99 1 0011 1.0005 +/- 0.ooo6 0.8665 S6 WCAP-3385 (4A.17) Saxtun -aLC .'c.. ...
5.74 1.0000 +/- 0,0010 0.9956 1k 0.0007 0.4476 0.4580 O,-- -Case 52 U 0.52" pitch 6,% P 1.0036 t 0.0011 1 .0047 -! 0.0006 0.5289 0 ,51 W7 57 WCAP-3385 (4A. I o-,.....
pitch 0.6389 NC" WCAP-33 85 (4A.17) Saxton Case 56 PuO2 0.56" -- '--'------- 0 N 58 6.6% Pu .00+/-0.10N 2
WCAP-3385 (4A.17) Saxton Case 76 borated lIU0 5.74 0.9994 +/- 0.0011 0.9967 +/- 0.0007 0.2923 0.2954 69 60 IVCAP-3385 (4 A. 17) Saxton Caie 56 U 0.56" pitch
--0.0006 .1520 0.1555 1.0063 +/- 0.0011 1.0133 pitch 6.6% Pu WCAP-3385 (4A.17) Saxton Case 79 PuO2 0.79"
__.-.61 0.1036 0.1047 1.0039 +/- 0.0011 1.0008 +/- 0.0006 5.74 62 WCAi'-33-95 (4A,17M Saxton Cast 79 U 0.79" pitch calculated.
Notes: NC stands for not lethargy causing fission.
I EALF is the energy of the average possibility of unusually large expericcltail statistical outliers ( > 30) suggesting the Thcse experimental results appear to be It the calct1Iational they could justifiably be excluded, for conservatism, they were retained in determining error. Although basis.
_Appicndix 4A. Page 13
Table 4A.2 t
CONIPARISON OF MCNP4a AND KENO5a CALCULATED REACTIVTES FOR VARIOUS ENRICHMENTS Calculated ka +/- Ia MCNP4*a KEN05a Earichment 3.0 0.8465 + 0. 0011 0.81dV78 _a 0.0004 0.8820 00. 0011 0.83341 - 0.0004 I 3.5 3.75 0.9019 - 0 .0011 0.8{987 +/- 0.0004 4.0 0.9132 0 .0010 0.9 140 + 0.0004 4.2 0.9276 0 .0011 0.9 237 + 0.0004 4.5 0.9400 + 0 .0011 0.9 '388 +/--0.0004 Based on the GE 3x3R fucl assembly.
Appendix 4A, Page 14
Table 4A.3 MCNP4a CALCULATED REACTITFTES FOR CRITICAL EXPERLMENTS WITH -NEUTRON ABSORBERS Ak MCNP4a Worth of Calculated EALF Ref. Experiment Absorber Absorber 4A. 13 PNL-2615 Boral Sheet 0.9994+/-0.00)12 0. 1165 0.0139 4A.7 B&W-1484 Core XX 0.0165 1.000894-0.0011 0.1724 4A. 13 PNL-2615 1.62% Boron-steel 0.0165 0.9996+/-0.0012 0.1161 4A.7 B&W- 1484 Core XIX F0.0202 0.9961 +/-0.0012 0.2103 4A.7 B&W-1484 Core XXI 0.9994+0.0010 I0.G243 0..1544 4A.7 B&W-1484 Core X'VII 0.0519 0.9962 +/-0.0012 0.2083 4A. 11 PNL-3602 Boral Sheet 4A.7 0.0708 0.9941 +/-0.0011 0.3135 B&W-1484 Core XV 4A.7 0.0786 0.9910+/-0.0011 0.2092 B&W-1484 Care XVI 4A.7 B&W-1484 0.0845 0.9935+/-.0.0010 0.1757 Core XIV 4A.7 B&W-1484 0.1575 0.9953 -0.0011 Core XIII 0.2022
- 0. 1738 L.0020+/-0.0011 0.1988 4A.14 PNL-7167 Expt 214R flux trap 0.1931 0.9991 +/-0.0011 0.3722
- EXLU is the energy of the average icthargy causing fission.
Appendix 4A, Page 15
Table 4A.4 CO-APJSON OF MCNP4a AND KENO5a CALCULATED RE*CTIMr=St FOR VARIOUS 'OB LOADINGS Calculated kT +/- lo "GB, g/cr: MCN?4a . KENO5a 0.005 1.0381 +/- 0.0012 1.0340 +/- 0.0004 0.010 0.9960 +/- 0.0010 0.9941 +/- 0.0004 0.015 0.9727 - 0.0009 0.9713 +/- 0.0004 0.020 0.9541 +/- 0.0012 0.9560 +/- 0.0004 0.025 0.9433 : 0.0011 0.9428 +/- 0.0004 0.03 0.9325 +/- 0.0011 0.9338 - 0.0004 0.035 0.9234 4- 0.0011 0.9251 - 0.0004 0.04 0.9173 +/- 0.0011 0.9179 - 0.0004 Based on a 4.5% cnrichcd GE gxSR fuel asscmbly.
Appendix 4A, Page 16
Table 4A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH TIl=CK LEAD ANID STEEL REFLECTORS' Arranged in order of increasing reflector-fuel spacing.
Appendix 4A, Page 17
Table 4A.6 CALCULATIONS FOR CRITICAL EXPERIvfENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS Calculated k,f MCNP4a 0.9974 +/- 0.0012 0.9970 +/- 0.0010 1.0023 +/- 0.0010 1.0060 +/- 0.0009 1.0057 +/- 0.0010 Appendix 4A, Page 18
Table 4A.7 CALCULATIONS FOR CRITICAL E=ERMENTS WITH MOX FUEL MC'P4a K.N 05a Referece t CaSe IcFa kI EALF" PNL-5903 MOX Fuel - Exp. No. 21 1.0041 +/-0.0011 0.9171 1.0046 +/-0.0006 0.384
[4A. 161 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 =0.001 1 0. [139 0.9966 +/-0.0006 0.1165 WCA.P- Saxton @ 0.52' pit,,5 0.99,96+/-0.001 L 0.8665 1.0005 t0.0006 0.3417 3385-54 (4A. 171 Saxton @ 0.56' pitch 1.0036 +/-0.0011 0.5239 1.0047 +/-0.0006 0.5197 Saxton @ 0.56' pitch borated 1.0008+/-0.0010 0.5389 NC NC Saxton Q 0.79' pitch 1.0063 +/-0.0011 0.1520 1.0133 +/-0.0006 0.1555 Note: NC stand.s for no( calculated Arranged in order of increasing lattice spacing.
EALF is the energy of the average lcthargy causing Fission.
Appendix 4A, Page 19
Coefficient of 0.1 3
- - - Linear Regression with Correlation 1.010 1.005 0
1.000 U
1]
0 0
0.995 0.990 1
0.1 Energy of Average Lethargy Causing Fission (Log Scale)
FIGURE 4A.1 MCNP CALCULATED k-eff VALUES for INDEX VARIOUS VALUES OF THE SPECTRAL
with Correlation Coefficient of 0.21 Linear Regression 1.010 1.005 w
u 1.000 a,
-2 0.995 0
0.990 0.985 0.1 Energy of Average Lethargy Causing Fission (Log Scale)
FIGURE 4A.2 KENO5a CALCULA TED k-eftSPECTRAL VALUES FOR INDEX VARIOUS VALUES OF THE
- - - Linear Regression with Correlation Coefficient of 0.03 1.010 1.005 4
S1.000 0
U 0.995 0.990 -
Enrichment, w/o U-235 FIGURE 4A.3 MCNP CALCULATED k-eff VALUES AT VARIOUS U-235 ENRICHMENTS
of 0.38 Linear Regression with Correlation Coefficient 1.010 1.005 S1.000 2 0.995 0
0.990
-T-F- T--FI - -*-T T IF D 5.5 6.0 0.985 2,0 Enrichment, w/o U-235 FIGURE 4A.4 KENO CALCULATED k-eff VALUES AT VARIOUS U-235 ENRICHMENTS
0.94 0.92 0
o 0.90
°U 03 LU 0.88 0.86 0.84 MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KENO5A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS
S-..3 1.01 z
C-)
-4 0.59
- 8. C/9 0.98
.4 a,
0.97 U
C C-)
0.96
-4 0
C a,
0.93 0.92 0.91 Reactivity Calculated with KENO5a FIGURE 4A.6 COMPARISON OF MCNP AND KENO5 CALCULATIONS FOR VARIOUS BORON-1O AREAL DENSITIES
to APPENDIX C List of CASMO4 and KENO-Va Input Files The list of computer files consists of those computer code input files that were used in the analysis and a brief description of each of the input files. This information provides details on the method of analysis and if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, analysis and licensing of a similar product. This list is, therefore, deemed proprietary and is not presented in this in this non-proprietary version of HI-2012620.
Report HI-2012620 Project 90941 C-I