ML020530458
| ML020530458 | |
| 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 HI-2012620, Rev 3 | |
| Download: ML020530458 (68) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 FEB 9 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 Holtec 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.
Printed on recycled paper
U.S. Nuclear Regulatory Commission Page 2 FEB 1 9 2002 There are no regulatory commitments made by this letter. If 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 q1 day of "2-"
Notary Pum i e M Commission expires aco:*05*
U.S. Nuclear Regulatory Commission Page 3 FEB 19 2002 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 Pike 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 INTERNATIONAL REPORT NUMBER HI-2012620, REVISION 3
- HOLTEC, NON-PROPRIETARY VERSION E-1
.o:' Center z55 Linco1n Drive Wes.. Marton, Ni 3-C53 0900 797-
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"retephone Telephone 1856) 797- 0900 Fax Fax (615 797 - 0905 I N T E R N -- T I 0 ','. -
EVALUATION OF THE EFFECT OF THE USE OF TRITIUM PRODUCING BURNABLE ABSORBER RODS (TPBARS) ON FUEL STORAGE REQUIREMENTS FOR
%0WA7T$BAR UNIT I (TVA)
Holtec Report No: HI-2012620 (NP), R3 0
z Holtec Project No: 90941
-ý Report Class :SAFETY RELATED
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d Le=- No. 30M431 q.t.: - Januaiv 29 2002 SOE'. (") SY 0.L.Lam*
NON-PROPRIETARY VERSION RIMS, WTC A-K 11 1.J . d 0909
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Summary of Revisions Report M1-2012620 Revision 1: The document is revised to incorporate client comments transmitted to Holtec International by TVA via Letter 3 0M4 10 dated April 10, 2001. There are no changes to the oonclusions 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 Project 90941 Fage SR-i
TABLE OF CONTENTS
1.0 INTRODUCTION
AND
SUMMARY
............................. 1 2.0 ANALYSIS CRITERIA AND ASSUMPTIONS ....................... 5 3.0 ACCEPTANCE CRITERIA ........................................ 6 4.0 DESIGN AND INPUT DATA ....................................... 7 5.0 M ETHODOLOGY ............................................... 8 6.0 ANALYSIS RESULTS .................................................................. 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 REFERENCES
................................................. 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 acks ........................ ;* .................. .................... ........ ... ... ... ....... 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 Fuel R acks ..................................................................................... 20 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 Rod Types ......................................................................... 26 6.11 Comparison of the Reactivity Effects of Depletion with Different Poison M aterials .................................................................................... 27 Report HI-2012620 Project 90941 ii
LIST OF FIGURES Figure 1 Minimum Burnup o 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 Objectives 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 TI9BARs, 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 burnup, 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 238-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.
- 1. 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
- 3. 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 IFBA 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 burnup 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 burnups for other initial enrichments are shown in Figure 1.
Report HI-2012620 Project 90941 2
Arrangement 3 In this arrangement, 3 fresh fuel assemblies are checker boarded with I 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-eff< 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 3
Accident Condition Evaluation of postulated accident conditions demonstrate that 55 ppm of soluble boron in the spent fuel pool is sufficient to maintain keff < 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.
- Criticality 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 (Li-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-eff) 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, standards, 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 Specifications Two different fuel assembly designs were considered in the analyses; the Westinghouse 17xl 7 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 or 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 moael, 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 1 million neutron histories were accumulated in each calculation, generally resulting in a statistical uncertainty of about +/-0.0003 Ak (la). 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 I % in k is included by the equivalent B-10 concentration in the fuel.
Report HI-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 17xl 7 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.
W17x17 W17x17 Burnup, GWD/MTU 11 ROBUST V5H 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 as The uncertainty in depletion calculations is part of the methodology uncertainty and was taken 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 kdff 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-2012620 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 Temperature and Void Effects Temperature effects were also evaluated, using CASMO4, in the temperature range from 4°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 Burnup Distribution Initially, fuel loaded into the reactor will bum with a slightly skewed cosine power distribution. As burnup progresses, the bumup 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 bumup 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 burnups 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 11
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 1
"* 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 keff 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 kefr 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 12
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 burnups, 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 13
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% 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 U 235 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 burnup enrichment combinations in Figure 2 (minimum of 6.75 MWD/Kg-U burnup 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 Project 90941 14
10.0 REFERENCES
- 1. R.M. Westfall, et. al., "NITAWL-S: Scale System Module for Performing Resonance Shielding and Working Library Production" in SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation., NUREG/CR- 0200, 1979.
L.M. Petrie and N.F. Landers, "KENO Va. An improved Monte Carlo Criticality Program with Subgrouping" in SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation., NUREG/V-0200, 1979.
- 2. M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963. I
- 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 Bumup Program," 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/95/1.
- 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 Bumup 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 HI-2012620 Project 90941 15
Table 4.1 Design Basis Fuel Assembly Specifications 181 W 17x17 W 17x17 Fuel Assembly V5H ROBUST Clad O.D., in 0.374 0.374 Clad I.D., in 0.329 0.329 Clad Material Zircaloy-4 Zirlo Pellet Diameter, in 0.3225 0.3225 Density, g/cc 10.631 10.631 Maximum Enrichment % 4.95+/-0.05 w/o 4.95+/-0.05 w/o Active Fuel Length, in 144 144 Number Fuel Rods 264 264 Fuel Rod Pitch 0.496 0.496 Number of Thimbles 25 25 Thimble O.D. 0.474 0.482 Thimble I.D. 0.442 0.442 Report 2012620 Project 9094 1 16
Table 6. 1. Reactivity Effects of Density Tolerance in the Watts Bar Spent Fuel Racks.
BURNUP, REFERENCE FUEL DENSITY GWD/MTU kinf kilf Ak 0 0.9776 0.9795 0.0019 10 0.9107 0.9122 o 0.0015 20 0.8606 0.8621 0.0015 30 0.8173 0.8189 0.0016 35 0.7971 0.7989 0.0018 Report HI-2012620 Project 90941 17
Table 6.2 Reactivity Effects of Temperature and Void in Watts Bar Spent Fuel Racks.
BURNUP, T=4 .C T= 20 °C T= 120 °C T = 120 °C + VOID GWD/MT kinf kinf Ak+ kinf Ak* kinf 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.&619 -0.0276 20 0.8629 0.8606 -0.0023 0.8404 -0.0202 0.8134 -0.0270 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 results @ 4 °C difference with results @ 20 "C difference with results at 120 TC and no void Report HI-2012620 Project 9094 1 18
Table 6.3 Reactivity Effects of Fuel Enrichment Tolerance in Watts Bar Spent Fuel Racks.
BURNUP, REFERENCE ENRICHMENT TOLERANCE GWD/MTU kinf kinf Ak 0 0.9776 0.9793 9.0017 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 Report 111-2012620 Project 9094 1 19
Table 6.4 Reactivity Effects of Abnormal and Accident Conditions in Watts Bar Spent Fuel Racks.
ACCIDENT/ABNORMAL CONDITIONS REACTIVITY EFFECT 0
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 Report 111-2012620 Project 90941 20
Table 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).
Reference k,,- 0.9233 Required Burnup of the Spent Fuel Assemblies 20 GWD/MTU Keno5a Bias 0.0030 Temperature Correction to 4 TC 0.0023 Axial Bumup Distribution Penalty Not Applicable KENO5a Bias Uncertainty 0.0012 KENO Statistics (95/95) Uncertainty (1.7 *oy) 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 Maximum kerf 0.9375 Regulatory Limiting keff 0.9500 Report 2012620 Project 9094 I 21
Table 6.6 Summary of the Criticality Safety Analyses for Face Adjacent Storage of Spent Fuel Assemblies In Watts Bar Racks (Arrangement 1).
Reference kff 0.9271 Required Bumup 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 Burnup Distribution Penalty Not Applicatle KENO5a Bias Uncertainty 0.0012 KENO Statistics (95/95) Uncertainty (1.7 *cF) 0.0009 Mechanical Tolerance Uncertainty 0.0059 Density Tolerance Uncertainty 0.0016 Enrichment Tolerance Uncertainty 0.0017 Depletion Uncertainty 0.0023 Fuel Eccentricity Uncertainty Negative Statistical Combination of Uncertainties 0.0069 Maximum keff 0.9392 Regulatory Limiting ken-f 0.9500 Reactivity dominated by once-burned assemblies, which suppresses the axial burnup penalty.
Report HI-2012620 Project 9094 1 22
Table 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).
Reference kf-t. 0.9131 Keno5a Bias 0.0030 Temperature Correction to 4 TC 0.0016 Axial Bumup Distribution Penalty Not Applicable KENOSa Bias Uncertainty 0.0012 KENO Statistics (95/95) Uncertainty (1.7 *a) 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.ff 0.9243 Regulatory Limiting keff 0.9500 Report 111-2012620 Plroject 9094 1 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 krf 0.9365 Keno5a Bias 0.0030 Temperature Correction to 4 "C 0.0016 Axial Bumup Distribution Penalty Not Applicable KENO5a Bias Uncertainty 0.0012 KENO Statistics (95/95) Uncertainty (1.7 *ca) 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 keff 0.9477 Regulatory Limiting keff 0.9500 Report 111-2012620 Project 90941 24
Table 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 k~f 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 0.9432 loading)
Fresh fuel assembly misloaded in the location of a 0.9435 0.0112 0.9547 water cell in Arrangement 3 Fresh fuel assembly, without IFBA rods, misloaded in the location of a fresh fuel assembly, with 32 IFBA 0.9375 0.0112 0.9487 rods, in Arrangement 4 Reporl 11-2012620 Project 90941 25
Table 6. 10 Comparison of the Reactivity of Fuel Assemblies Depleted with Different Burnable Poison Rod Types.
BURNUP, W-V5H with W-V5H with W-V5H WITH GWD/MTU TPBAR BPRA WABA kilf kinf kilf 0 0.9792 0.9792 (.9792 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 11-20 12620 Project 9094 1 26
Table 6.11 Comparison of the Reactivity Effects of Depletion with Different Poison Materials k.,rf in rack 1 T r IFI3A No No Yes Yes Bumup, Yes Yes No TPBAR No MWD/KgU WABA Yes No No Yes 0 0.9792 0.9792 0.9402 0.9402 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 f 111-2012620 P~roject 90941 27
20 123 10
=..
8 UNACC PTABLE 6/ BU NUP 4
2 0
2.50 3.00 3.50 4.00 4.50 5.00 InItfll Enrichment, wf% U-235 Figure 1 Minimum Burnup of Spent Fuel in Arramgemenf of Spent and Fresh Fuel 2x2 Checkerboard of 4.95% Enrichment (Arrangement 2)
"28 Report HI-2012620 Project 90941
7.0 CD4.5 S4.0
- 2.5 23.
2.5 1.0 0.5 0.0 3.50 4.00 4,50 Inrfial Enrlchmenf, wfz U-235 Figure 2 Mlnimum Burnup for Unrestricfed Fuel of Various Initial Enrichmenfs Storage of Spent (Arrangement 1) 29 eport HI-20126-20 -" Project 90941
0.980 0.960 0.940 0.920 v 0.900 0
0.880 00
._ 0.860 I- 0.840'E 0.820 0.800 0.780 0.760 Burnup, MWD/KgU Fig. 3 Comparison of the Reactivity of Fuel Assemblies Depleted with Different Burnable Poison Rod Types 30
- eport HI-2012620 Project 90941
8.62 0.10" THK 0.02 I s s' sKK Shekh 0,036" C!*
Waier Gap 0.973" I.
0.090" ýS BOX 4
I -
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 HI-2012620 Project 90941 A-1
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" Repotrt HI-2012620 Project 90941 B-I
APPENDIX 4A: BENCHMARK CALCULATIONS 4A.1 INTRODUCTION AND
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 KZEN05a [4A.21 uses group-dependent cross sections. For the KENO5a analyses reported here, the 238 group library was chosen, procssed through the NITAWL-ll [4A.2] program to create a working libra-y and to account for resonance self-shielding in uranium-23 8 (Nordheim integral treatment). The 238 group library was chosen to avoid or minimize the errorst (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 '0'B 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 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 I
Figures 4A. 1 and 4A.2 show the calculated kt for the benchmark critical experiments as a function of the EALF for MCNP4a and KENO5a, respectively (UO, 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.
are no Linear regression analysis of the data in Figures 4A. 1 and 4A.2 show that there trends, as evidenced by very low values of the correlation coefficient (0. 13 for MCN'P4a a
and 0.21 for KENO5a). The total bias (systematic error, or mean of the deviation from kr of exactly 1.000) for the two methods of analysis are shown in the table below.
Calculational Bias of MCNP4a and KENO5a MCNP4a . 0.0009+/-0.0011 KENO5a 0.0030+/-0.0012 The bias and standard error of the bias were derived directly from the calculated k.,T 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 than 91 (4A.181 (for the number of cases analyzed, the K-factor is -2.05 or slightly more 2).
/_- i(4A.1) the PNL A classical example of experimental error is the corrected enrichment in values in experiments, first as an addendum to the initial report and, secondly, by revised subsequent reports for the same fuel rods.
These equations may be found in any standard text on statistics, for example, reference in
[4A.6] (or the MCNP4a manual) and is the same methodology used in MCNP4a and KEN05a.
Appendix 4A, Page 2
- kk2 ,(4A.2) 02_ I- *-LI k n (n-1)
Bias = (1-k) t K a*- (4A.3) the unbiased where k, are the calculated reactivities of n critical experiments; ot is called the standard error of the bias estimator of the standard deviation of the mean (also
% confidence level (mean)); K is the one-sided multiplier for 95 % probability at the 95 (NBS Handbook 91 [4A.18])."
of Standards (now Formula 4.A.3 is based on the methodology of the National Bureau first portion of the NIST) and is used to calculate the values presented on page 4.A-2. The KENO5a results.
equation, ( 1- k ), is the actual bias which is added to the MCNP4a and with the bias. The K The second term, Koa:, is the uncertainty or standard error associated 91 and are for values used were obtained from the National Bureau of Standards Handbook level. The one-sided statistical tolerance limits for 95% probability at the 95% confidence the 53 critical actual K values for the 56 critical experiments evaluated with MCNP4a and experiments evaluated with KENO5a are 2.04 and 2.05, respectively.
designs.
The bias values are used to evaluate ýhe maximum kf. values for the rack result in greater KENO5a has a slightly larger systematic error than MCNP4a, but both cross precision than published data (4A.3 through 4A.5] would indicate for collapsed section sets in KENO5a (SCALE) calculations.
4A.2 Effect of Enrichment from 2.46 w/o The benchmark critical experiments include those with enrichments ranging and to 5.74 w/o and therefore span the enrichment range for rack designs. Figures 4A.3 4A.4 show the calculated k,,f values (Table 4A. 1) as a function of the fuel enrichment reported for the critical experiments. Linear regression analyses for these data confirms (0.03 for that there are no trends, as indicated by low values of the correlation coefficients the various MCN-P4a and 0.38 for KENO5a). Thus, there are no corrections to the bias for 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 KENO5a 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 k., 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 %°RLoading 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 made performed by B&W are the most representative of the rack designs. PNL has also some measurements with absorber plates, but, with one exception (a flux-trap experiment),
errors the reactivity worth of the absorbers in the PNL tests is very lokW and any significant that might exist in the treatment of strong thin absorbers could not be revealed.
4A. 1)
Table 4A.3 lists the subset of experiments using thin neutron absorbers (from Table and shows the reactivity worth (Ak) of the absorber.!
No trends with reactivity worth of the absorber are evident, although based on the have calculations shown in Table 4A.3. some of the B&W critical experiments seem to 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 "tBconcentration in the absorber, a cross-comparison was made with MCNP4a and KENO5a (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 0 line, within an expected 95 % probability limit).
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.
Appendix 4A, Page 4
4A.4 4Miscellaneous and Minor Parameters 4A.4. 1 Reflector Material and Spacings lead reflectors.'
PNL has performed a number of critical experiments with thick steel and in Table in Table 4A.5 (subset of data Analysis of these critical experiments are listed overprediction of ker at the lower 4A. 1). There appears to be a small tendency toward data points in each series to allow a spacing, although there are an insufficient number of tendency toward overprediction at close quantitative determination of any trends. The be slightly more conservative than otherwise.
spacing means that the rack calculations may 4A.4.2 Fuel Pellet Diameter and Lattice Pitch cover a range of fuel pellet diameters from The critical experiments selected for analysis from 0.476 to 1.00 inches. In the rack designs, 0.311 to 0.444 inches, and lattice spacings fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.496 to 0.580 inch the to 0.494 inches O.D. (0.488 to 0.740 inch lattice spacing) for PWR fuel and from 0.3224 experiments analyzed provide a reasonable lattice spacing) for BWR fuel. Thus, the critical on the data in Table 4A.1, there does not representation of power reactor fuel. Based fuel pellet diameter or lattice pitch, at least appear to be any observable trend with either applicable to rack designs.
over the range of the critical experiments 4A.4.3 Soluble Boron Concentration Effects in the B&W series of critical experiments Various soluble boron concentrations were used ranging up to 2550 ppm. Results of and in one PNL experiment, with boron concentrations Table 4A.6. Analyses of the very are shown in MCNP4a (and one KENO5a) calculations ppm) show a tendency to slightly high boron concentration experiments (> 1300 exceeding 1300 ppm. In turn, this would overpredict reactivity for the three experiments higher soluble boron concentrations could be suggest that the evaluation of the racks with slightly conservative.
reflector were also performed but not Parallel experiments with a depleted uranium are not pertinent to the Holtec rack design.
included in the present analysis since they Appendix 4A, Page 5
4A.5 MsX Fue The number of critical experiments with PuO, bearing fuel (MOX) is more limited than for UO 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 ktr of 1.00, indicating that when Pu is present, both MCNP4a 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 spacings, the KENO5a calculated reactivities are below 1.00, suggesting that a smaU trend may exist with KENO5a. It is also possible that the overprediction in k,, 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 jpnsistency in calculated kl, over a wide range of the spectral index (energy of the average lethargy causing fission).
Appendix 4A, Page 6
4A.6 Rýern
[4A. 11 J.F. Briesmeister, Ed., "MCN-P4a - A General Monte Carlo N Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1993).
[4A.21 SCALE 4.3, "A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", NUYREG-0200 (ORNL-NIIREG-CSD-2fU2/R5, 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 ENDFJB-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.aY, 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 Analysis", 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.71 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-18 10, Babcock and Wilcox Company, April 1984.
Appendix 4A, Page 7
[4A. 1.0] S.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 wlo and 4.31 wlo "U Enriched UO2 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 UO Rods in Water wýth Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December, 1981.
[4A. 1.3] S.R. Bierman et al., Critical Separation Between Subcritical Clusters of 4.31 w/o "U Enriched UO2 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.151 B.M. Durst et al., Critical Experiments with 4.31 wt % m"U Enriched UO 2 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. 171 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, Ex-perimental Statistics, National Bureau of Standards, Handbook 91, August 1963.
Appendix 4A, Page 8 I
Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k ' (cV KENOSa MCNP4a KENO5a Enrich. MCNP4a Reference Identification 0.9898+/- 0.0006 0.1759 0.1753 2.46 0.9964 +/- 0.0010 1 B&W-1484 (4A.7) Core 1 2-'o260 00------0-0----0 0.2446 1.0015 +/- 0.0005 0.2553 0.2936 2.46 1.0008 +/- 0.0011 0.1999 013 2 B&W-1484 (4A.7) Core U 2.46 1.0010 +/- 0.0012 1.0005 +/- 0.0005 3 D&W-1484 (4A.7) Core 111 0.1422 0.1426 0."956 +/- 0.0012 0.9901 +/- 0.0006 CrIX2.46 4 RW18(4.)
0.1513 0.1499 2.46 0.9980 +/- 0.0014 0.9922 +/- 0.0006 5 B&W-1484 (4A.7) Core X 0.2031 014 2.46 0.9978 +/- 0.0012 1.:0005 +/- 0.00)05 6 B&W-1494 (4A.7) Core XI 0.1718 0.1662 2.46 0.9988 +/- 0.0011 0.9978 +/- 0.0006 7 B&W-1484 (4A.7) Core X11 0.9952 +/- 0.0006 0.1988 0.1708 2.46 1.0020 +/- 0.0010 8 B&W-1484 (4A.7) Core XVI A
0.2022 0 9 2.46 0,53 0.0011 0.pd28 +/- 0.0006 0.2092 021 9 &W-1484 (4A.7) Core XIV 2.46 0.9910"+/- 0.0011 0.9909 +/- 0.0006 10 B&W-1494 (4A.7) Core XV
- 0. 1757 0.1713 2.4 0.935 +/- 0.0010 o.9889 +/-+~o 11 B&W-1484 (4A.7 Core XVI 0.2083 0.2021 2.46 0.9"62 +/- 0.0012 0.9942 +/- 0.005 12 B&W-1484 (4A.7) Core XVU 0.17081 1.03 0.0012 10.9931 +/- 0.000 :0.17ý051
- -- -T 246 1.03 13 B&W-1484 (4A.7) Core XVII Appendix 4A, Page 9
Table 4A. 1 Calculations Summary of Criticality Benchmark
- UEI.F (On KENO5a MCNP4a KUENO5R MCNP4a 0.2103 0.2011 IdentificatiOD Reference Core XIX 0.1774 0.1701 14 B&W-1484 (4A.7)
Core XX o.1544 0. 1536 15 B&W-1484 (4A.7)
Core XXI 1.4475 1.4680 16 B&W-1484 (4A.7) 86 S-type Fuel, w/8 ppm B I.W43 1.5660 64 5 (4A.8) 17 B&W-1 46 S-type Fuel, w1. ppm B 0.4241 0.4331 64 5 (4A.8) 18 B&W-1 B
64 5 So-type Fuel, w/1156 ppm 0.1531 NC 19 B&W-1 (4A.8) case 1 1337 ppm B 0,4493 NC 20 B&W-1810 (4A.9)
Case 12 1899 ppm B 0.2172 NC 1
21 B&W-18 0 (4A.9)
Water Moderator 0 gap 0.1778 NC 22 French (4A.10) gap Water Moderator 2.5 cm 0.1677 NC 23 French (4A.10)
Water Moderator 5 cm gap 0.1736 NC 24 French (4A.10)
Witter Moderator 10 c" gaP NC 0.1018 25 French (4A.10) 26 PNL-3602 (4A.1l) Steel Reflector, o separation
___________________Appendix 4A, Page 10(
Table 4A.1 Summary of Criticality Benchmark Calculations Calculated"fr EALF (eV)
KENO5a I MCNP4a KENO~a Identificatioii Enrich. MCNP4a I I 0.1000 Reference I I 0.9992 +/- 0.0006 0.0909 0.9992 +L .O-OO 0.1000 27 PNL-3602 (4A.11) Steel Reflector, 1.321 cm sepn. 2.35 0.9980 +/- 0.0009 0.9964 +/- 0.0006 0.0981 0.0975 Steel Reflector, 2.616 cm sepn 2.35 0.9968 +/- 0.0009 28 PNI-3602 (4A.11) 0.0976 0.0970 0.9974 +/- 0.0010 0.9980 + 0.0006 29 PNL-3602 (4A.11) Steel Reflector, 3.912 cm sepn. 2.35 0.9939 +/- 0.0006 0.0973 0.0968 0.9962 +/- 0.0005 Steel Reflector, Infinite sepn. 2.35 30 PNL-3602 (4A.1I) NC 0,3282 NC 1.0003 - 0.0007 31 PNL-3602 (4A.11) Steel Reflector, 0 ctm sepn. 4.306 1.0012 +/- 0.0007 0.3016 0.3039 0.9997 +/- 0.0010 32 PNL-3602 (4A. 11) Steel Reflector, 1.321 cm sepn. 4.306 0.9974 +/- 0.0007 0.2911 0.2927 0.9994 +/- 0.0012 33 PNL-3602 (4A. 11) Steel Reflector, 2.616 cm sepn. 4.306 0.9951 +/- 0.0007 0.2828 0.2860 0.9969 +/- 0.0011 Steel Reflector, 5.405 an sepn. 4.306 34 PNL-3602 (4A.11) 0.2851 0.2864 0.9910 +/- 0.0020 0.9947 + 0.0007 Steel Reflector, infinite sepn. tt 4.306 35 PNL-3602 (4A, 11) 0.3135 0.3150 0.9941 +/- 0.0011 0.9970 " 0.0007 Steel Reflector, with Boral Sheets 4.306 36 PNI-3602 (4A.11) NC 0.3159 NC 1.0003 +/- 0.0007 Lead Reflector, 0 cm sepn. 4.306 37 PNL-3926 (4A.12) 0.3030 0.3044 1,0025 +/- 0.0011 0.9997 +/- 0.0007 Lead Reflector, 0.55 cm sepn. 4.306 38 PNL-3926 (4A.12) 0.2883 0.2930 1.0000 +/- 0.0012 0.9985 - 0.0007 Lead Reflector, 1.956 cm sepn. 4.306 39 1PNL-3926 (4A.12)
Appendix 4A, Page 11
Table 4A.1 Calculations
!-,ummarv of Criticality Benchmark tJ *Q ........
. "Adf_
Cac _EALE ' NY)
KENO5a MCNP4a ItMtN-u0 Identification Enrich. MCNP4a Reference 0.9946 +/-:0.0007 0.2845 o.2831 Lead Reflector, 5.405 cm sepn. 4.306 0.9971 :1: 0.0012 PNL-3926 (4A.12) 0.9950 +/- 0.0007 0.1155 40 0.1159 0."925 : 0.0012 Experiment 0041032 - no absorber 4.306 41 PNL-2615 (4A.13) 0.9:71 + 0.0007 NC NC 0.1154 Experiment 030 -Zr plates 4.306 NC 42 PNL-2615 (4A.13) NC 0.9965 +/- 0.0007 NC 0.1164 Experiment 013 -Steel plates 4.306 43 PNL-2615 (4A.13) 0.9972 +/- 0.0007 NC 0.1164 Experiment 014 - Steel plates 4.306 44 PNL-2615 (4A.13) 0.1172 0.9982 +/- 0.0010 0.9981 +/- 0.0007 0.1162 PNL-2615 (4A.13) Exp. 009 1.05% Boron-Steel plates 4.306 45 0.9982 +/- 0.0007 0.1161 0.9996 +/- 0,0012 0.1173 PNL-2615 (4A. 13) Exp. 012 1.62% Boron-Steel plates 4.306 46 0.9969 + 0.O007 0.1165 0. 117lt 0.9994 +/- 0.0012 PNL,-2615 (4A.13) Exp. 031 - Boral plates 4.306 47 0.9956 +/- 0.0007 0.3722 0.3812 0.9991 +/- 0.0011 PNI,-7167 (4A. 14) Experiment 214R - with flux trap 4.306 0.9963 +/- 0.0007 0.3742 0.3826 0.9969 + 0.0011 PNL-7167 (4A. 14) Experiment 214V3 - with flux trap 4.306 49 0.2893 NC 0.9974 +/- 0.0012 NC Case 173 - 0 ppm B 4.306 pN"-267 (4A.15) 0.5509 NC 1.0057 +/- 0.0010 NC Case 177 - 2550 ppm B 4.306
?NL-4A267 (4A. IS) 0.9171 0.8868 1.0041 £ 0.0011 1.0046 +/- 0.0006 rPN L-5 8 03 (4k.16)) MOX Fuel - Type 3.2 Exp. 21 Appendix( 4A, Page 12
Table 4A.1 Calculations Summary of Criticality Benchmark E.LF (eV Ccuated MvCNI'4a KENO5a MCN1P4a KENO5a T.1t ifh-,to/nn Enrich. 0.2968 0.2944 Reference en 1.0036 +/- 0.0006 20% Pu 43 1.0058 +/- 0.0012 0.1665 0.1706 PNL5803 (4A. 16) MOX Fuel - Type 3.2 Exp. 0.9989 +/- 0.0006 53 20% Pu 1.0083 +/- 0.0011 0.1139 0.1165 PNL-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 13 0.9?66 :t 0.0006 54 20% Pu 1.0079 -I 0.0011 0.8417 55 PNL-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 32 0.9996 +/- 0.0011 1.0005 +/- 0.0006 0.8665 6.6% Pu 0.4476 0.4580 56 WCAP-3385 (4A.17) Saxton Case 52 PuO2 0.52" pitch 1.0000 +/- 0.0010 0.9956 +/- 0.0007 5.74 0.5289 0.5197 WCAP-3385 (4A.17) Saxton Case 52 U 0.52" pitch 1.0036 +/- 0.0011 1.0047 +/- 0.OO06 57 6.6% Pu NC 58 WCAP-3385 (4A.17) Saxton Case 56 Pu02 0.56" pitch 1.0008 +/- 0.0010 NC 0.6389 6.6% Pu 13 O.2923 0.2954 WCAIP-3385 (4A.17) Saxton Case 56 borated Pu02 0.9994 +/- 0.0011 0.9967 +/- 0.0007 59 5.74 0.1520 0.1555 WCAP-3385 (4A.17) Saxton Case 56 U 0.56" pitch 1.0063 +/- 0.0011 60 6.6% Pu 1.0133 +/-:0.000 Page 2
0.79" pitch 0.1036 0.1047 WCAP-3385 (4A.17) Saxton Case 79 PuO 61 5.74 1.0039 +/- 0.0011 1,0008 +/- 0.0006 WCAi-3385 (4A.17) Saxton Case 79 U 0.79" pitch 62 4A, Notes: NC stands for not calculated.
lethargy causing fission. large experimental i EALF is the energy of the average statistical outliers (> 3o) suggesting the possibility of unusually to be in determining the calculational it These experimental results appear for conservatism, they were retained justifiably be excluded, error. Although they could basis.
Appendix Appendix 4A, 13
-Page'
Table 4A.2 t
REACTIVITIES COMPARISON OF MCNP4a AND KENO5a CALCULATED FOR VARIOUS ENRICHMENTS Calculated k., +/- la MCNP4a KENO5a Enrichment 3.0 0.8465 _ 0.0011 0.8478 +/- 0.0004 3.5 0.8820 + 0.0011 0.8841 +/- 0.0004 3.75 0.9019 +/- 0.0011 0.8987 - 0.0004 4.0 0.9132 - 0.0010 0.9140 +/- 0.0004 0.9276 +/- 0.0011 0.9237 - 0.0004 4.2 4.5 0.9400 +/-'0.0011 0.9388 +/- 0.0004 Based on the GE 8xSR fuel assembly.
Appendix 4A, Page 14
Table 4A.3 MCNP4a CALCULATED REACTiVITIS FOR CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS tik MCNP4,a Worth of Calculated EALF t 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 XIX 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 XVIU 0.0519 0.9962+/-0.0012 0.2083 4A.11 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.
Appendix 4A, Page 15 I
Table 4A.4 COMNEARISON OF MCNP4a AND KENO5a CALCULATED R.EACTViTIESt FOR VARIOUS '°B LOADINGS Calculated k, +/- lo 1'B, g/cm2 . MCN'P4a 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 +/- 0.0011 0.9251 +/- 0.0004 0.04 0.9173 +/- 0.0011 0.9179 - 0.0004 Based on a 4.5% enriched GE 8xSR fuel assembly.
Appendix 4A, Page 16
Table 4A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH TIUCK LEAD AND STEEL REFLECTORSt Separation, Ref. Case E, V4% cm MCNP4a k, K.ENO5a k.,
4A. 11 Steel 2.35 1.321 0.9980+/-0.0009 0.9992 +/-0.00*6 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 m 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 Arranged in order of increasing reflector-fuel spacing.
Appendix 4A, Page 17
Table 4A-6 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS I.
Calculated kc Boron Concentration, Reference Experiment ppm MCNP4a I
4A.15 PNL-4267 0 0.9974 + 0.0012 4A.8 B&W-1645 886 0.9970 +/- 0.0010 4A.9 B&W-1810 1337 1.0023 +/- 0.0010 4A.9 B&W-1810 1899 1.0060 + 0.0009 4A. I5 1 PNL-4267 2550 1.0057 +/- 0.0010 Appendix 4A, Page 18
Table 4A.7 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNP4a KENOSa Refereice Caser k FALEALF'1 PNL-5903 MOX Fuel - Exp. No. 21 1.0()41 +/-0.0011 0.9171 1.0046+/-0.0006 0.8868 (4A. 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 Saxton 4 0.52' pitch 0.9996+/-0.0011 0.8665 1.0005 +/-0.0006 0.8417 WCAP-3385-54
[4A.17] 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 0.6389 NC NC Saxton @ 0.79" pitch 1.0063 +/-0.0011 0.1520 1.0133 +/-0.0006 0.1555 Note: NC stands for not calculatrd Arranged in order of increasing lattice spacing.
EALF is the energy of the average lethargy causing fission.
Appendix 4A, Page 19
with Correlation Coefficient of 0.13 Linear Regression 1.010 1.005 w
41) 4-1.000 0
.4 0
o) 0D 0.995 0.990 I 1 0.1 Energy of Average Lethargy Causing Fission (Log Scale)
FIGURE 4A.1 MCNP CALCULATED k-eff VALUES SPECTRAL for INDEX VARIOUS VALUES OF THE
Linear Regression with Correlation Coefficient of 0.21 1.010 1.005
°,
CD 1.000
"-U
%e
-.2 0.995 "U
0 0.990 0.985 1
0.1 Energy of Average Lethargy Causing Fission (Log Scale)
KENO5a CALCULATED k-eff VALUES FOR INDEX FIGURE 4A.2 VALUES OF THE SPECTRAL VARIOUS
- - - Linear Regression with Correlation Coefficient of 0.03 1.010 1.005 0
. 9
"- 1.000 0 0
0.9 95 0.99O 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 1.000 "20.995
_o0 C-,
0.990 0.985 6.0 2.0 2.E 3.0 3.5 4.
Enrichment, w/o U-235 FIGURE 4A.4 KENO CALCULATED k-eff VALUES AT VARIOUS U-235 ENRICHMENTS
0.94 -
E 0.92 V)
C 0.90 0
C w
I Lii 0.88 0
z 0.86 0.84 ,
0.84 MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KENO5A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS
1 .014 1'.03 1.02 1.01 zZ-1.00 0.99
-tJ 0.98
-4 0.97 C
C C-)
0.96
-4
.4 0.95 C) a a, 0.94 0.93 0.92 0.91 Reactivity Calculated with KENO5a FIGURE 4A.6 COMPARISON OF MCNP AND KENO5a CALCULATIONS F(DR VARIOUS BORON-IO AREAL DENSITIES
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-1