ML081820126
| ML081820126 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/27/2008 |
| From: | Northeast Technology Corp |
| To: | Constellation Nuclear, Office of Nuclear Reactor Regulation |
| References | |
| TAC MD8434 NET-290-01 | |
| Download: ML081820126 (19) | |
Text
NET-290-01 3.0 RESULTS OF THE CRITICALITY ANALYSES 3.1 CASMO-4 and KENO V.a Reactivity Calculation Comparison As a check of the two independent methods used for these analyses, the reactivity of each fuel type in the standard cold-core geometry (SCCG) at cold temperature conditions (68°F) has been calculated both with KENO V.a and with CASMO-4 at zero burnup. Both models are exact renderings of the assemblies in core geometry. Table 2 contains the k, for each fuellatlice assembly with Gdz0 3 rods. The reported koo values include model biases which have been determined using benchmark calculations.
These model biases are ~~ =
2
-0.00782 and ~~=-0.01028 for KENO V.a and CASMO-4, respectively.
Table 2 CASM0-41 KENO V.a Reactivity Comparison in Standard Cold-Core Geometry at Zero Burnup Gd203 k~
(rods, Enrichment Fuel Type (w/o U-235) I ICASMO-4 w/o)
KENOV.a 7x7 2@1.0, 1.17551 2@0.5 2.50 1.17469 +/- 0.0004 8x8 (2 water rods) 7@4.0 3.20 1.13409 +/- 0.0004 1.13403 8x8 (4 water rods) 8@4.0 3.42 1.12932 +/-0.0004 1.12873 9x9 12@5.0 4.00 1.08419 +/- 0.0004 1.08418 9x9 12@5.0 4.21 1.10256 +/- 0.0004 1.10127 9x9 14@6.0 4.60 1.07399 +/- 0.0004 1.07273 The maximum difference between the Keno V.a and CASMO-4 eigenvalues is less than 0.0015 ~koo.
13 2
NET-290-01 3.2 Reactivity Calculations 3.2.1 CASMO-4 Depletion Calculations CASMO-4 was utilized to compute the reactivity of a limiting reactivity fuel lattice as a function of burnup for each fuel type. The limiting reactivity lattice with respect to planar average enrichment, number of gadolinia rods and gadolinia loading was determined from fuel assembly design reports[9,101. Sensitivity analyses demonstrated that average power density, average fuel temperature, saturation temperature of the moderator, and zero void results in the most reactive condition (peak bundle reactivity). Accordingly, the depletion calculations were conducted under these average conditions with 0% void.
For subsequent analyses of the Boraflex modules in finite three-dimensional models, the reactivity equivalentfresh fuel enrichment (REFFE)was first determined andsubsequently utilized in the calculations. The REFFE was determined by modeling each limiting fuel bundle lattice (with respect to maximum enrichment and minimum gadolinia loading) in the standard cold core and in-rack geometries. The U235enrichment was varied until the bias-2 corrected koo calculated using the KenoV.a rack model at the REFFE slightly exceededthe koo value calculated at the point of peak reactivity during depletion. The k., of the Keno V.a model at the REFFE always exceeds the koo at peak reactivity during depletion. Table 3 contains the k, values for each lattice type for CASMO-4 and Keno V.a.
Table 3 Reactivity Equivalent Fresh Fuel Enrichments and Limiting Lattice k, at Peak Reactivity w/o No. of k.-SCCG k.-SCCG Array Gd203 REFFE U235 Gadolinia Rods (CASMO)
(KENOV.a) 7x7 2.5 2,2 1.0,0.5 1.2428 1.2465 2.15 8x8 3.2 7
4.0 1.2164 1.2169 2.10 8x8 3.42 8
4.0 1.2254 1.2258 2.20 8x8 3.6 9
4.0 1.2368 1.2379 2.30 9x9 4.0 12 5.0 1.2340 1.2478 2.35 9x9 4.21 12 5.0 1.2441 1.2478 2.35 9x9 4.6 14 6.0 1.2323 1.2478 2.35 14 2
NET-290-01 3.2.2 Reference Keno V.a Model A reference Keno V.a model was created based upon the actual fuel assemblies loaded in the "South" BorafJex module under the tooling table. All assemblies in the "South" module are 8x8 lattices and are conservatively assumed to be at a REFFE corresponding to a burnup of peak assembly reactivity.
To simplify modeling of the South module, several assemblies were conservatively modeled at higher enrichments.
With the following exceptions, all "South" module cells contain 8x8 fuel at 3.2 peak planar enrichment:
Cells 2A55 thru 2A58 and 2B72 thru 2C72 actually contain 8x8 assemblies at 2.82 wlo U-235 peak planar enrichment, however, these cells are conservatively modeled as 8x8 assemblies at 3.2 wlo peak planar enrichment.
Cells 2B56, 2L56, 2B71 and 2L71 are empty cells containing tooling table support legs, nevertheless are conservatively modeled as 8x8 fuel assemblies at 3.2 w/o.
Cell 2A71 and 2M72 contain 8x8 assemblies at 3.2 w/o; these cells are conservatively modeled as 3.42 wlo enrichment assemblies.
Cells 2071 thru 2F71 contain 8x8 fuel assemblies at 3.2 w/o; these cells are conservatively modeled as 3.42 wlo fuel assemblies.
Cells 2G72 thru 2K72 contain 8x8 fuel assemblies at 2.82 w/o.
These cells are conservatively modeled as 8x8 fuel at 3.2 wlo enrichment.
Cells 2M59 thru 2M71 and 2K55 thru 2L55 contain 8x8 fuel assemblies at 3.42 wlo as currently loaded in the "South" module.
Figure 5 shows the fuel initial enrichment, gadolinia loading, and limiting kx, (SCCG) of assemblies modeled in the "South" module.
In the "North" module there are two non-fuel components residing in cells 2037 and 2L53.
For conservatism, all cells are assumed to include 9x9 fuel at peak reactivity. The analysis for the "North" module accounted for possible future reload enrichments up to 4.60 wlo U-235 with a minimum number of gadolinia rods at the minimum loading based upon reload assembly design reports[9,10l.
In modeling the 9x9 assemblies in the "North" module, several conservatisms were included in the model. These include:
15
NET-290-01 The number of gadolinia rods was taken at the minimum number in any zone (e.g.
vanished zones typically had one less gadolinia rod than did the dominant zones).
For assemblies with split gadolinia loadings, the minimum loading was used.
The neighboring fuel racks to the East and North of the Boraflex modules contain BORAL as the neutron absorber material. Additional arrays of BORAL modules containing 10x10 fuel assemblies were added to the Boraflex modules to create a full pool model as shown in Figure 6. Although 1Ox1 0 assemblies are not currently loaded in the BORAL modules, the future use of this array type is possible. However, 1Ox1 0 assemblies may NOT be stored in the Boraflex modules. The following conservative assumptions were used to model the additional BORAL modules:
As previous analyses had shown the 1Ox1 0 fuel type is more reactive than the 9x9 fuel type[2], the 10x10 fuel assemblies were modeled in the BORAL modules.
All fuel is at a REFFE of 2.55 wloU-235, corresponding to the tech specification limit koo::; 1.31 in SCCG. [5]
The areal density of the BORAL absorber is assumed to be at the minimum certified value of 0.015 gms B-10/cm2. [11]
The reference case is a full fuel height model with water albedoes in the axial directions.
The South and West boundary conditions both incorporate a 24-inch concrete albedo and the North and East boundaries along the BORAL modules incorporate a water albedo boundary condition.
The reference case Keno V.a model was executed using 3050 neutron generations and 5,000 neutrons per generation for a total of 15 million neutron histories.
The first fifty neutron generations were omitted to attain source convergence.
16
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.r-t-wlo U-235 Figure 5: U-235 Enrichment and Gadolinia Distribution of 8x8 Assemblies as Modeled in the NMP1 "South" Boraflex Module.
Key:
Gad rods. loading limiting Ism 17
NET-290-0 1 N
1 Figure 6: Reference Case Keno V.a Generated Plot of the NMP1 Boraflex Modules Loaded with 8x8 and 9x9 Fuel Assemblies and BORAL Modules Loaded with 1Ox10 Fuel Assemblies.
18
NET-290-01 3.3 Effect of Tolerances and Uncertainties 3.3.1 Tolerances and Calculational Uncertainties To evaluate the reactivity effects of fuel and rack manufacturing tolerances in the "North" and "South" modules, CASMO-4 and Keno V.a perturbation calculations were performed.
The limiting reactivity fuel assembly 9x9 at peak burnup was used.
This is conservative with respect to the tolerance effects from the lower reactivity assemblies. Additionally, tolerances were determined based on an infinite array of storage cells.
The following tolerance and uncertainty components are addressed:
U-235 Enrichment:
The enrichment tolerance of +/- 0.088 wlo U-235 variation about the nominal reference value of 4.60 wlo U-235 was considered. [12] Reference 12 states:
"Enrichment Variation: For enrichment variations ~ 2.0% U235.
Nominal +/- 0.088 w/o" Furthermore, the letter states "Process history shows an overall process standard deviation of 0.023 wlo U235for enrichments ~ 3.95% U235."
UOz Pellet Density:
A variation of -2.0%/+1.0 (absolute) about the nominal reference theoretical density of 97%)13] Reference 13 states:
" As you can see by the attached email from GE, we are currently loading pellets with 97% theoretical U density. The tolerance on this range is -2%
to +1% meaning that the maximum density range of theoretical densities is 2
from 95% to 98%."
Pellet Dishing:
The pellets were assumed to be undished.
This is a conservative assumption in that it maximizes the U-235 loading per axial centimeter of the fuel stack. No sensitivity analyses were completed with respect to the variations in the pellet dishing factor.
Pellet Diameter: The tolerance value of +/-0.002 inches was used. Bundle Announcement Reports [9.10] state that the maximum tolerance for the 7x7, ax8, or 9x9 fuel types is +/-0.001 inches.
19
NET-290-01 Clad Inside Diameter:
The tolerance value of +/-0.002 inches was used.
Bundle Announcement Reports [9.10] state that the maximum tolerance for the 7x7, 8x8, or 9x9 fuel types is +/-0.001 inches.
Clad Outside Diameter:
Clad 00 is bounded by the combination of wall thickness tolerance and inner diameter tolerance. Each of these is addressed separately, thus no further analysis is required.
Clad Thickness:
The value of +/-0.004 inches was used in Reference 12.
Bundle Announcement Reports [9.10] state that the maximum tolerance for the 7x7, 8x8, or9x9 fuel types is +/-0.004 inches.
Gdz03 Loading:
In the maximum reactivity assembly at 14.0 GWD/MTU the Gd203 is depleted; however, the burnup at peak reactivity depends on the initial Gd20310ading. The 2
tolerance of +/-10% (relative) in the Gd203 loading has been used[12]. Reference 12 further states:
"...the 95% confidence limits on the mean gadolinia loading content shall be within +/-7.5% (relative) of nominal. Individual pellet gadolinia content is limited to +/-10% (relative) of nominal."
Therefore, the tolerance value of to. 5 w/o gadolinia (10% relative) employed in this analysis is conservative.
Cell Inside Dimension: The manufacturing tolerance of +/- 0.030 inch for the variations in cell wall inside dimensions was used[15].
Stainless Steel Thickness: A stainless steel sheet tolerance of +/- 0.004 inches was used.[15]
Flux Trap Width: The manufacturing tolerance on the flux trap width of -0.038 inch was used.l15]
Cell-to-Cell Pitch:
Cell-to-cell pitch is determined by the cell wall thickness, cell inside dimensions and flux trap width in the NMP1 fuel rack.
Each of these is addressed separately and no further allowances are required.
20
NET-290-01 Assembly Location: The reference KENO V.a reactivity calculations are based on a model with each assembly symmetrically positioned in each storage cell.
The effect of four adjacent assemblies with minimum separation distance has been considered.
Calculational Uncertainty:
The 95% probability I 95% confidence level uncertainty associated with the reference KENO V.a calculation has been applied.
Methodology Uncertainty:
The 95% probability I 95% confidence level uncertainty of 0.0078 as determined from benchmark calculations (see Appendix) has been considered.
3.3.2 Uncertainty Introduced by Depletion Calculations Critical experiment data are generally not available for spent fuel and, accordingly, some judgment must be used to assess those uncertainties introduced by the depletion calculations.
CASMO-4 and the 70 group cross section library used for these analyses have been used extensively to generate assembly average cross sections for core follow calculations and reload fuel design in both BWRs and PWRs. Any significant error in those depletion calculations would be detectable either by incore instrumentation measurements of core power distribution or cycle energy output or both. Significant deviations between the predicted and actual fuel cycle lengths and core power distributions using CASMO-4 generated cross sections are not observed.
For the purpose of assessing the effects of uncertainties introduced by depletion calculations, it is useful to estimate the magnitude of depletion uncertainties in k, and compare this uncertainty with margins inherent in the present calculation. It is assumed that depletion calculations introduce an uncertainty in k, which is a linear function of burnup such that at a burnup of 40,000 MWD/MTU the ~kunc due to depletion effects is 0.02. So that for the limiting reactivity assembly at 14 GWD/MTU, the uncertainty introduced by depletion is 0.00700 in ~koo.
This uncertainty is included in the statistical treatment of 2
tolerances and uncertainties in Table 4. Additional methods for determining the uncertainty in depletion calculations have been proposeo.P" These methods result in a ~k due to burnup of - +0.0067, therefore the 0.00700 ~k used is bounding.
21
NET-290-01 3.4 Summary of Reactivity Calculations 3.4.1 Reference Loading The explicit model developed in Section 3.2 was executed to determine the reactivity of the two remaining NMP1 Boraflex spent fuel racks while taking no credit for Boraflex.
For the reference case with the "South" module conservatively modeled as described in Section 3.2 and the "North" module loaded with 9x9 fuel at maximum reactivity, the calculated keff was 0.92217.
Additional cases were run with various U02 11 2
enrichmentlgadolinia combinations to confirm that the reference case keff is bounding.
Table 4 contains a summary of the criticality analyses results for the NMP1 spent fuel racks.
Table 5 contains the required minimum gadolinia loading as a function of enrichment for the 9x9 fuel resident in the "North" module.
1 2 22
NET-290-01 Table 4 Summary of Criticality Calculation Results for the NMP1 Boraflex Spent Fuel Racks with North Module Containing Peak Reactivity 9x9 Fuel at a REFFE of 2.35 1 1 w/o (corresponding to 4.21 w/o, 12 Gadolinia Rods @ 5 w/o, 150 0 F) 12 Reference Case keft Methodology Bias Nominal keft (best estimate)
Tolerances and Uncertainties:
Fuel Material
- U-235 Enrichment
- U02 Density
-Gd203 Loading
-Pellet Diameter
-Clad Diameter/thickness Rack Construction
- Flux Trap Width Pitch
- CelllD
- Cell Wall Thickness Burnup Uncertainty Assembly Placement Methodology Bias Uncertainty (95/95)
Calculational Uncertainty (95/95)
Square Root of Sum of Squares:
Maximum k.. (95x 95)
Effect of Worst Case Accident Maximum k.. (95x 95, Including Accident)
Margin 23 0.91435
+0.00786 0.92217
+0.00412
+0.00024
+0.00995
+0.00160
+0.00514
+0.00594
+0.00237
+0.00000
+0.00700
+0.00062
+0.00963
+0.00032
+0.01812 0.94028
+0.00901 0.94930
+0.00070 2
NET-290-01 The reactivity effects of tolerances and uncertainties were evaluated with an infinite array of 9x9 assemblies at peak reactivity and when combined in a root-mean-square sense yield
~k = 0.01812. The difference between these values and the 0.95 design limit represents I 2 margin, which would be available to accommodate reactivity increases as may be the result of postulated accidents.
Table 5 Minimum Gadolinia Loading as a Function of Initial Peak Planar Enrichment for 9x9 Fuel[9,10J w/o U-235 Number of Gadolinia Rods w/o Gadolinia Limiting koo (SCCG) 4.0 12 5.0 1.2340 4.21*
12 5.0 1.2441 4.6+
14 6.0 1.2323
- Reference Case/Limiting koo
+ Predicted for Future Reloads 24 2
NET-290-01 3.5 Abnormal/Accident Conditions The reactivity effects of the following abnormal/accident conditions have been conservatively evaluated:
Fuel Assembly Drop Fuel Assembly Inadvertent Positioning Alongside Rack Fuel Assembly Misload Moderator Temperature Variations The drop of a 1Ox1 0 reload fuel assembly assumed to come to rest in a horizontal position on top of the "North" module has been evaluated with all assemblies in place as shown in Figure 6 (on page 19). The reactivity effect is negligible (~keff < 0.00021).
1 The inadvertent positioning or drop of a fuel assembly alongside of the Boraflex modules in the corner of the "North" module and above the "South" module and the pool wall as shown in Figure 7 has been evaluated. The increase in rack reactivity as determined by KENO V.a is negligible (~keff < 0.00082).
I 1 12 For both the assembly drop and inadvertent positioning, the reactivity effect is well within the margin inherent in the design of the NMP1 spent fuel racks assuming 100% Boraflex loss.
The misloading of a 1Ox1 0 fuel assembly in the "North" Boraflex module has been evaluated for multiple positions within the Boraflex module. The maximum reactivity effect was determined to occur when the 1Ox1 0 reload assembly is centered in the "North" module shown in Figure 8, with the resulting reactivity effect ~keff =0.00090. Under the I 1 conservative assumptions of these analyses, the maximum fuel rack keff (at a 95%
probability with a 95% confidence level) has been determined to be 0.94118.
I 1 2
The effect of variations in moderator density and temperature on the reactivity of the NMP1 fuel storage racks has been analyzed'". These analyses were performed at 220°F, the point of boiling at the depth of the fuel racks and with approximately 20% void.
The maximum reactivity effect is +0.00901~k.
For these conditions, the maximum keff is 1 1 2
0.94930 (at 95% probability/95% confidence level). Therefore, within the moderator temperature variations analyzed, adequate subcritical margin is maintained.
25
NET-290-01 Figure 7: Keno V.a Generated Plot of a Dropped Assembly Resting on Top of "North" Boraflex Module.
26
NET-290-01 Dropped Bundle Figure 8: Keno V.a Generated Plot of a Dropped Fuel Assembly Alongside of the "North" and "South" Boraflex Modules.
27
NET-290-01 Misloaded Bundle Figure 9: Keno V.a Generated Plot of a Reload Assembly (10x10)
Misloaded Adjacent to the "North" BORAL Modules.
28
NET-290-01
4.0 CONCLUSION
S The reactivity state of the NMP1 spent fuel storage pool has been analyzed and keff has been conservatively evaluated. This analysis is based on the following:
- 1) no reactivity credit for Boraflex, 2) the existing fuel loading configuration below the tooling table, and 3) bounding reactivity (9x9) fuel loaded in the "North" module. With respect to the "North" module, the bounding 9x9 fuel type is characterized by an initial enrichment of 4.21 wlo U-2 235, a minimum Gd20 310ading of 12 rods at 5.0 w/o. Analyses have demonstrated that for the NMP1 spent fuel racks the maximum keft is less than 0.95, after conservatively including the reactivity effects of tolerances, uncertainties, code biases and the effects of postulated accidents.
Based upon the analyses described above, the maximum keft of the NMP1 spent fuel racks 1 2 is shown to satisfy the 0.95 limit provided that:
- 1. The "South" Boraflex module is loaded with 8x8 assemblies containing U-235 I2 enrichments and gadolinia contents as shown in Figure 5.
- 2. The "North" Boraflex module is loaded:
with existing 7x7 or 8x8 fuel assemblies, or 1 2 with future 9x9 fuel type assemblies at peak planar enrichment with a minimum number of Gd20 3bearing rods and at a minimum gadolinia loading as specified in Table 5.
Provided that these conditions are met, the total loss of the Boraflex concurrent with the 1 2 worst case accident scenarios can be safely accommodated in the NMP1 spent fuel Boraflex racks.
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5.0 REFERENCES
- 1. "Nine Mile Point 1 Spent Fuel Pool Modification, Supplemental Submittal," Docket No. 50-220, DPR-63, Niagara Mohawk Power Corporation; June 1983.
- 2. "Evaluation of the Nine Mile Point 1 Boraflex Spent Fuel Racks for the General Electric 9x9 and 10x10 Fuel Types", NET-110-01, Northeast Technology Corp.;
March 26, 1996.
- 4. Email from W. Carter (NMP) to M. Harris (NETCO), dated 6/29/07 subject: "NMP1 Boraflex per April 8, 2007 Shuffleworks Fig File.MDI".
- 5. Email from W., Carter (NMP) to M. Harris (NETCO), dated 9/25/2007,
Subject:
"Words from NMP1 Technical Specifications."
- 6. "SCALE-PC: Modular Code System for Performing Criticality Safety Analyses for Licensing Evaluation, for Workstations and Personal Computers", Version 5, Parts 0 thru 3, RSIC Computer Code Collection CCC-545. Oak Ridge National Laboratory:
Oak Ridge, Tennessee; May 2004.
- 7. Edenius, Malte and Bengt H. Forssen. "CASMO-4:
A Fuel Assembly Bumup Program - User's Manual," Version 2.05, Rev 3, SSP-01/400. Studsvik of America:
Newton, Massachusetts; July 2003.
- 8. Letter from P.O. Milano (NRC) to M.R. Kansler (Entergy), Indian Point Nuclear 2
Generating, Unit No.2 - Amendment Re: Credit for soluble boron and burnup in Spent Fuel Pit (TAC No. MB2989), May 29,2002.
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- 9. Letter from W. Carter (NMP) to K. Lindquist (NETCO) w/Attachments: Letter from Ed Gibbs (GE) to Paul Netusil dated 7/14/1997 w/Attached Bundle Design Report for NMP1 from Cycle 1 through Reload 12 for GE8B Fuel and Applicable Sections of Bundle Design Reports for Reloads 9 through 12.
- 10. Email from W. Carter (NMP) to M. Harris (NETCO) dated 10/2/2007,
Subject:
"Bundle Announcement Reports for R13 to R19 (Current Cycle)."
11.Email from Mr. Kristopher Cummings (Holtec) to W. Carter (NMP), dated 9/11/07,
Subject:
"Proprietary Information on Holtec Racks at Nine Mile Point Nuclear Station,"
12.Letter from Rocco Bianchi (GE Nuclear Energy) to Bart Franey (Niagara Mohawk Power Corporation (NMPC) dated 12/21/1995 regarding standard deviations in gadolinia content and enrichment.
- 13. Email from J. Winklebleck (Niagara Mohawk Power Corporation) to K. Lindquist dated 1/26/2001.
- 14. Kopp, L.I., "Guidance on the Regulatory Requirements for Criticality Safety Analysis 2
of Fuel Storage at Light Water Reactor Power Plants," Proceedings of a Consultancy on Implementation of Burnup Credit in Spent Fuel Management Systems; IAEA, July 28-31, 1998.
- 15. "Criticality Analysis for Nine Mile Point Unit 1, Phase II Spent Fuel Storage Racks,"
Niagara Mohawk Power Corporation, 8202-00-0072, performed by Pickard, Lowe &
Garrick, Inc., May 1982, Revision 1, July 1982.
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