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#REDIRECT [[NL-17-144, Enclosure 2 - NET-28091-003-01NP, Rev. 0, Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit]]
{{Adams
| number = ML17354A015
| issue date = 11/28/2017
| title = Enclosure 2 - NET-28091-003-01NP, Rev. 0, Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit
| author name =
| author affiliation = Entergy Nuclear Operations, Inc
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000247
| license number = DPR-026
| contact person =
| case reference number = NL-17-144
| document report number = NET-28091-003-01NP, Rev. 0
| document type = Final Safety Analysis Report (FSAR)
| page count = 247
}}
 
=Text=
{{#Wiki_filter:ENCLOSURE 2 TO NL-17-144 Curtiss-Wright Nuclear Division, NETCO Report NET-28091-003-01, Revision 0 (Non-Proprietary Version) Entergy Nuclear Operations , Inc. Indian Point Unit 2 Docket No. 50-24 7 NET-28091-003-0lNP, Rev. 0 Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit R e v: Date: 0 11/2 8/J.?~ Prepared by: Curtiss-Wright Nuclear Division, NETCO 44 Shelter Rock Rd. Danbury , CT 06810 Prepared for: Entergy Nuclear Operations
-Indian Point Energy Center under Contract No. 10502876 Prepared By: n Reviewed By: \ Approved By: ~L.-..., , *t'~~~ -'7 ,-Y:tlitlt-t *Jk< \ .Q 1'r, '-y ..
This Page Intentionally Left Blank Table of Contents 1 Introduction
.................................................................................................
1 1.1 Background
.............................................................................................................................
1 1.2 Description of the Analysis .....................................................................................................
2 1.3 Acceptance Criteria ................................................................................................................
5 2 Methodology
.................................................................................................
6 2.1 Computer Codes ......................................................................................................................
7 3 Input Data ..................................................................................................
11 3.1 SFP and Storage Rack Specifications
..................................................................................
11 3.2 Fuel Assembly Designs ..........................................................................................................
14 3.3 Fuel Assembly Insert Designs ..............................................................................................
17 3.4 Plant Operation Data ............................................................................................................
20 4 Validation
...................................................................................................
23 4.1 U02, Structural Materials, and Absorbers Validation
......................................................
24 4.2 MOX Validation
....................................................................................................................
25 4.3 Critical Experiments Effect on the Final k9s 1 9s ...................................................................
26 5 Depletion Calculations
..............................................................................
28 5.1 Limiting Depletion Parameters
-Temperatures
................................................................
29 5.1.1 Averaged Assembly Radial Peaking Factor ...........................................................
29 5.1.3 Moderator Temperature
..........................................................................................
31 5.1.4 Fuel Temperature
.....................................................................................................
34 5.1.5 Selection of Bounding Model and Temperatures
...................................................
37 5.2 Limiting Depletion Parameters
-Burnable Absorbers
.....................................................
42 5.3 Limiting Depletion Parameters
-Soluble Boron ...............................................................
43 5.4 Limiting Depletion Parameters
-Specific Power ...............................................................
44 5.5 Limiting Depletion Parameters
-Control Rod Operation
................................................
45 5.6 Depletion Analysis Model .....................................................................................................
51 NET-28091-0003-01, Revision 0 iii 
: 5. 7 Special Case Depletions
........................................................................................................
55 5.8 Reduced Power Operation at End of Life and Fission Gases ...........................................
57 5.9 Production of Atom Density Sets .........................................................................................
58 5.10 Summary of Limiting Depletion Conditions
.......................................................................
60 6 Rack Model ................................................................................................
62 6.1 SCALE 2x2 Radial Models ...................................................................................................
62 6.2 Axial Model. ...........................................................................................................................
65 6.2.1 Axial Burn up Distribution
.......................................................................................
65 6.3 Dimensional Changes with Irradiation
...............................................................................
72 6.3.1 Clad Creep .................................................................................................................
72 6.3.2 Grid Growth ...........................
...................................................................................
79 6.4 Averaged Assembly Peaking Factor Interpolation
............................................................
82 6.5 Convergence of the 2x2 Infinite Model Calculations
.........................................................
83 6.6 Full Pool Models ....................................................................................................................
84 6.6.1 Sensitivity of the Full Pool Model to Modeling Assumptions
................................
87 6.6.2 Convergence of the Full Pool Model... .....................................................................
88 6.7 Summary of Modeling Assumptions
...................................................................................
93 7 Sensitivity Analysis ....................................................................................
94 7.1 Manufacturing Tolerances
...................................................................................................
94 7.2 Burn up Dependent Biases and Uncertainties
.....................................................................
96 7.3 Eccentricity
............................................................................................................................
99 7.4 Additional Biases and Uncertainties
..................................................................................
102 7.5 Biases and Uncertainties Rack-up .....................................................................................
104 7.6 Interface Uncertainty Treatment.
......................................................................................
107 8 Results .......................................................................................................
I 08 8.1 Temperature Effects ...........................................................................................................
108 8.2 Region 1 Fuel Categories 1 and 2 .......................................................................................
109 NET-28091-0003-01, Revision 0 iv 
 
===8.3 Region===
2 Category 4 Batch Grouping Z -Current and Future Fuel .............................
111 8.3.1 Curve Fit ..................................................................................................................
113 8.3.2 Confirmation Calculations for Category 4 ...........................................................
114 8.4 Determination of Burnup Requirements for Categories 3 and 5 ....................................
115 8.4.1 Cell Category Layout in Region 2 ..........................................................................
115 8.4.2 Additional Burnup Requirements for Fuel Categories 3 and 5 ..........................
116 8.4.3 Confirmation of k9s 1 9s for Full Pool (includes Category 3 and 5) ........................
120 8.5 Alternate Arrangements for Region 1 ...............................................................................
124 8.6 Calculations for Discharged Fuel (IP2 A-X and IP3 A-AA) ...........................................
127 8.7 Cell Blockers ........................................................................................................................
135 8.8 Region 2 Checkerboard
......................................................................................................
136 8.9 Burnup Penalty for Hafnium Flux Suppression Inserts ..................................................
136 8.10 Failed Fuel Containers
........................................................................................................
136 8.11 Fuel Rod Storage Basket ....................................................................................................
138 8.12 Assemblies with Missing Fuel Rods ...................................................................................
139 8.13 Storage of Miscellaneous Materials
...................................................................................
141 8.14 Borated Conditions
.............................................................................................................
141 8.15 Burnup Penalty for High Soluble Boron Conditions
.......................................................
143 9 Normal Operations and Accident Analysis ...........................................
144 9.1 Normal Operations
.............................................................................................................
145 9.2 Misplaced Assembly ............................................................................................................
146 9.3 Dropped Assembly ..............................................................................................................
149 9.4 Over Temperature
..............................................................................................................
150 9.5 Multiple Misloads ................................................................................................................
151 9.6 Boron Dilution Accident .....................................................................................................
152 9.7 Seismic Event .......................................................................................................................
153 10 Summary ..................................................................................................
154 NET-28091-0003-01 , Revision 0 V J 10.1 Review of DSS-ISG-2010-01
...............................................................................................
154 10.2 Fuel Reactivity Categorization
..........................................................................................
158 10.3 Allowable SFP Cells for Each Fuel Category ...................................................................
160 10.4 Fuel and Operating Requirements
....................................................................................
163 References
.......................................
................................................................
166 Appendix A: Validation of SCALE 6.1.2 for Criticality Analysis Using Laboratory Critical Experiments
.............................................
A-1 A.1. Overview ..................................................................................................
A-1 A.2. U02 Laboratory Critical Experiments
.................................................
A-1 A.2.1 Introduction
.........................................................................................................................
A-1 A.2.2 Definition of the Range of Parameters to Be Validated
...................................................
A-2 A.2.3 Selection of the Fresh U02 Critical Benchmark Experiments
........................................
A-2 A.2.4 Computer Analysis of the U02 Benchmark Critical Experiments
...............................
A-10 A.2.5 Statistical Analysis of the Fresh U02 Critical Benchmark Results ..............................
A-20 A.2.6 Establishing the Bias and the Uncertainty
......................................................................
A-28 A.2.7 Subcritical Margin ............................................................................................................
A-29 A.2.8 Area of Applicability (Benchmark Applicability)
..........................................................
A-29 A.2.9 Summary of U02 Laboratory Critical Experiment Analysis ........................................
A-32 A.3. HTC and MOX Critical Experiments
................................................
A-33 A.3.1 HTC Critical Experiments
...............................................................................................
A-33 A.3.2 MOX Critical Experiments
................
..............................................................................
A-39 A.3.3 Bias and Uncertainty from the MOX/HTC Critical Experiments
...............................
A-43 A.4. Temperature Dependent Critical Experiments
.................................
A-44 A.5. Summary of Validation Using Laboratory Critical Experiments
... A-48 A.6. Appendix References
............................................................................
A-49 Appendix B: Fuel Categorization for Unit 2 Batches A Through X and Unit 3 A through AA ...................................................................
B-1 NET-28091-0003-01, Revision 0 VI List of Tables Table 2.1: 185 Isotopes Used in the Analysis .....................................................................................
8 Table 3.1: Region 1 and 2 Storage Rack Dimensions
[8, 9) ...........................................................
14 Table 3.2: Fuel Assembly Dimensions
[11, 12) ................................................................................
17 Table 3.3: Control Rod and Hafnium Rod Descriptions
[11) ........................................................
19 Table 3.4: Pyrex and Wet Annular Burnable Absorber Descriptions
[11, 12, 15] ......................
19 Table 3.5: Key Operating Features by Cycle Used in IP2 .............................................................
21 Table 3.6: Key Operating Features by Cycle Used in IP3 .............................................................
22 Table 5.1: Moderator Exit Temperature, Texit , versus Peaking Factor for Batch Groups ..........
31 Table 5.2: Moderator Exit Density versus Peaking Factor for Batch Groups .............................
32 Table 5.3: Enthalpy Node Factor versus Axial Burnup Shape .....................................................
33 Table 5.4: Moderator Temperature (K) at each Node versus Burnup Profile ............................
33 Table 5.5: Fuel Temperature (K) at each Node versus Burnup Profile .......................................
36 Table 5.6: Fit Coefficients for Top Node Moderator Temperature and Density .........................
40 Table 5.7: Fit Coefficients for 3rd Node Moderator Temperature and Density ...........................
40 Table 5.8: Burnable Absorbers versus Batch Grouping ................................................................
43 Table 5.9: Soluble Boron versus Batch Grouping ..........................................................................
43 Table 5.10: Assemblies under D-Bank for the First 21 Cycles of IP2 ..........................................
.45 Table 5.11: Effect of Modeling the Bite Position rather than Burnable Absorbers
....................
4 7 Table 5.12: Burn up Penalty for Assem. with Burnable Absorbers followed by Bite D-bank ..... 48 Table 5.13: Assemblies with BA Inserts plus under D-Bank in Non-Bite Cycles ........................
48 Table 5.14: Assemblies under D-Bank for the first 11 Cycles ofIP3 ............................................
50 Table 5.15: SCALE/TRITON minus CASM0-5 Ak of Depletion at 100 Hours Cooling ............
53 Table 5.16: SCALE/TRITON minus CASM0-5 Ak of Depletion at 5 Years Cooling ................
53 Table 5.17: SCALE/TRITON minus CASM0-5 Ak of Depletion at 15 Years Cooling ..............
54 NET-28091-0003-01 , Revision 0 VII Table 5.18: Percent Difference in the Ak of Depletion at 100 Hours Cooling ..............................
54 Table 5.19: Percent Difference in the Ak of Depletion at 5 Years Cooling ...................................
54 Table 5.20: Percent Difference in the Ak of Depletion at 15 Years Cooling .................................
55 Table 5.21: Special Case Depletion Parameters
.............................................................................
55 Table 5.22: Verification of Cooling Time Model in the Interpolation Program ..........................
60 Table 6.1: Axial Burn up Profile vs. Burnup Bin [27) .....................................................................
67 Table 6.2: Axial Relative Burnups for Blanketed Discharged Fuel.. ............................................
70 Table 6.3: Axial Relative Burnups for Batch Z Fuel.. ....................................................................
71 Table 6.4: Calculated k versus Number of Nodes Modeled ...........................................................
72 Table 6.5: Full Pool Model Sensitivity Tests ...................................................................................
88 Table 6.6: k e rr Changes With Start Source ......................................................................................
90 Table 7.1: Tolerance Reactivity Effects ...........................................................................................
94 Table 7.2: Eccentricity Results .......................................................................................................
101 Table 7.3: Total Bias and Uncertainties for Region 1 , Categories 1, 2, 3 ...................................
105 Table 7.4: Sample Category 4 and 5 Bias and Uncertainty Rack-up .........................................
106 Table 7.5: Total Bias and Uncertainty for Fresh Fuel in Region 2 .......................
......................
106 Table 8.1: Calculated k e rr as a Function of Temperature
.............................................................
109 Table 8.2: Confirmation of Region 1 Requirements for Category 1 and 2 Fuel.. ......................
110 Table 8.3: Change in k err with Burn up and number of IFBA Rods ............................................
110 Table 8.4: Minimum Burnup Requirements (GWd/T) for Category 4 Batch Grouping Z ...... 111 Table 8.5: Curve Fit Coefficients for Category 4 Fuel.. ...............................................................
113 Table 8.6: Calculated k e rr Values at each Category 4 Batch Z Burnup Point ............................
114 Table 8.7: Total Bias and Uncertainty at each Category 4 Batch Z Burnup Point.. .................
114 Table 8.8: k 9s19s for each Category 4 Batch Z Burnup Point .......................................................
115 Table 8.9: Region 2 Models at Loading Curve (Cat 5 is Cat 4 plus 11 GWd/T) ........................
121 NET-28091-0003-01, Revi s ion 0 V lll Table 8.10: Eccentric Options for Region 1 ...............
................
...................................................
122 Table 8.11: Maximum Full Pool k9s 1 9s assuming Various Cycle Lengths ...................................
124 Table 8.12: Dependence of k e rr on the Region 1 Arrangement
....................................................
126 Table 8.13: Batch A-D Minimum Burnup Requirements (GWd/T) for Category 4 ................
128 Table 8.14: Batch E-F Minimum Burn up Requirements (GWd/T) for Category 4 .................
128 Table 8.15: Batch G-L Minimum Burnup Requirements (GWd/T) for Category 4 ................
129 Table 8.16: Batch M-P Minimum Burnup Requirements (GWd/T) for Category 4 ................
129 Table 8.17: Batch Q-S Minimum Burnup Requirements (GWd/T) for Category 4 .................
130 Table 8.18: Batch T-V Minimum Burnup Requirements (GWd/T) for Category 4 .................
130 Table 8.19: Batch W Minimum Burnup Requirements (GWd/T) for Category 4 ....................
131 Table 8.20: Batch X Minimum Burnup Requirements (GWd/T) for Category 4 .....................
131 Table 8.21: Batch A-U (IP3) Minimum Burnup Requirements (GWd/T) for Category 4 ....... 132 Table 8.22: Batch V-X (IP3) Minimum Burnup Requirements (GWd/T) for Category 4 ....... 132 Table 8.23: Individual Assembly Analysis for Category 3 .....................................
......................
133 Table 8.24: Individual Assembly Analysis for Category 4 ...........................................................
134 Table 8.25: Failed Fuel Container Pin Analysis ............
.............................
..................................
138 Table 8.26: Normal Operations with Boron Dilution ppm (Full Pool Model) ...........................
142 Table 8.27: Burnup Penalty Results at 1200 ppm ........................................................................
143 Table 9.1: Misplaced 5.0 w/o 64 IFBA Assemblies with 2000 ppm .............................................
148 Table 10.1: DSS-ISG-2010-01 Checklist
..................
...............
.......................................................
154 Table 10.2: Summary of Loading Requirements for Fuel Batch Z ....................
........................
159 Table 10.3: Fuel Design Requirements for Batch Z assemblies
..................................................
164 Table 10.4: Fuel Assembly Operating Requirements
.......................................
............................
165 Table A.l: Selection Review ofOECD/NEA Criticality Benchmarks
........................................
A-3 Table A.2: Critical Experiment Results with SCALE 6.1.2 and ENDF/B-VII
........................
A-11 Table A.3: Summary of Critical Experiments Containing Boron ............................................
A-18 NET-28091-0003-01, Revision 0 I X Table A.4: Wilk-Shapiro Test Results Output From DATAPLOT (4) ....................................
A-21 Table A.5: Area of Applicability (Benchmark Applicability)
...................................................
A-29 Table A.6: HTC Phase 1 Results ..................................................................................................
A-34 Table A.7: HTC Phase 2a, Gadolinium Solutions, Results ........................................................
A-35 Table A.8: HTC Phase 2b, Boron Solutions, Results .................................................................
A-36 Table A.9: HTC Phase 3 Results -Water Reflected Assemblies
..............................................
A-37 Table A.10: HTC Phase 4 Results -Steel Reflected Assemblies
...............................................
A-38 Table A.11: Results of MOX Critical Benchmarks (SCALE 6.1.2, ENDF/B-VII)
..................
A-40 Table A.12: LCT-46 with Full Thermal Expansion
...................................................................
A-44 Table A.13: LCT-46 with No Thermal Expansion of Solids ......................................................
A-47 Table B.1: Fuel Assembly Reactivity Categorization for Assembly IDs A -X for Unit 2 .........
B-2 Table B.2: Fuel Assembly Reactivity Categorization for Fuel Assembly IDs A -AA for IP3 .. B-5 NET-28091-0003-01 , Revision 0 X List of Figures Figure 1.1: Fuel Category Placement in the IP2 SFP (base case) ...................................................
3 Figure 3.1: IP2 SFP Taken From Holtec Drawing #397 [35) ........................................................
12 Figure 3.2: Small Section of the Region 1 Rack [8] ........................................................................
13 Figure 3.3: Region 2 Rack Showing Cell Boxes and Resultant Cells [9] ......................................
13 Figure 5.1: Averaged Assembly Peaking Factors of Assemblies in the IP2 SFP .........................
30 Figure 5.2: Fuel Temperature Change with Burnup and Relative Power ...................................
35 Figure 5.3: Fuel Temperature (K) at 25 GWd/T vs. Peaking Factor (PF) at the Top Node ....... 36 Figure 5.4: Top Node Moderator Temp (K) vs. Average Assembly Peaking Factor ..................
39 Figure 5.5: Top Node Moderator Density vs. Avg Assembly Peaking Factor ............................
.40 Figure 6.1: Region 1 KENO Model. .............
...........
.......................
..............................
....................
63 Figure 6.2: Region 2 KENO Model ...................................................................................
...............
64 Figure 6.3: Comparison of Creep-down for ZIRLO&#x17d; and Zircaloy-4
[37] ................................
73 Figure 6.4: Diameter Decrease versus Exposure Time [39) ............................
............................
... 74 Figure 6.5: Clad Creep Down for Vandellos 2 Nuclear Power Plant [40) ....................................
75 Figure 6.6: Axial Distribution of the Fuel Rod Diameter at 50.5 GWd/T [41] ............................
76 Figure 6.7: Oxide Layer thickness with Burnup [42) .....................................................................
77 Figure 6.8: Density of Fuel Pellet as a Function of Pellet Burnup [43) .........................................
77 Figure 6.9: ZIRLO&#x17d; Grid Growth [42) ..............................................
.................
..........................
80 Figure 6.10: Zircaloy-4 and MS Grid growth versus burnup [44) ................................................
80 Figure 6.11: Grid Growth of ZIRLO&#x17d; and Zircaloy-4 versus Elevation
[45) ............................
81 Figure 6.12: Calcu lat ed kerr versus Assembly Average Peaking Factor ........................................
83 Figure 6.13: Full Pool Model ......................
......................................................................................
85 Figure 6.14: Model of Module H Showing Control Rods .............................................................. 86 Figure 6.15: Locations of the Start Sources for the Convergence Tests .......................................
91 N ET-28091-0003-01, Revi s ion 0 XI Figure 6.16: Change in Average k err with Progressing Generations
.............................................
92 Figure 7.1: Category 4 Region 2 with 16 Assemblies Eccentrically Placed ................................
100 Figure 7.2: Eccentric Model for Category 2 with Central Row Shifted Down ..........................
102 Figure 8.1: Calculated kerr as a Function of Category 5 Burnup Using 5.0 w/o Fuel... .............. 118 Figure 8.2: kerr as a Function of Category 5 Burnup Using 4.2 w/o Enriched Fuel ...................
119 Figure 8.3: Calculated k e rr as a Function of Category 3 Burnup Using 5.0 w/o Fuel... ..............
120 Figure 8.4: Refueling Arrangement
..........
.....................................................
................................
125 Figure 8.5: No Cat 2 Arrangement
................................................................................................
125 Figure 8.6: Max Cat 1 Arrangement
.............................................................................................
126 Figure 8.7: Example Odd Arrangement
........................................................................................
126 Figure 8.8: Cell Blocker Region 2 Model ....................
.........................................
.........................
135 Figure 8.9: Failed Fuel Container Pin Model .............
..........
.................
..................................
..... 137 Figure 8.10: Model for the Fuel Rod Storage Basket.. ...............
...................
..........
.....................
138 Figure 8.11: k err versus Number of Missing Fuel Rods ................................................................
140 Figure 8.12: Model for Assemblies with 36 Missing Fuel Rods [1] ..............
...............................
140 Figure 9.1: Misplaced Assembly at the Cask Area Corner .....................................
.................... 147 Figure 9.2: Misplaced Assembly between the Fuel Elevator and the Rack ...............................
148 Figure 9.3: Full Pool Model with Dropped Assembly ................................................................
.. 150 Figure 10.1: Fuel Category Location Requirements (Base Case) ...............................................
161 Figure 10.2: Refueling Arrangement
.............................................................................................
162 Figure 10.3: Max Cat 1 Arrangement
..........................
.................................................................
162 Figure 10.4: Example Odd Arrangement
......................................................................................
162 Figure A.1: Distribution of the Calculated k err values Around the Mean .................................
A-22 Figure A.2: k err as a Function of the Energy of the Average Lethargy Causing Fission .........
A-24 Figure A.3: kerr as a Function of the Pin Diameter .....................................................................
A-25 Figure A.4: k e rr as a Function of the Lattice Pitch ......................................................................
A-26 NET-28091-0003-01, R evision 0 X II Figure A.5: k.rr as a Function of the Fuel Enrichment
...............................................................
A-27 Figure A.6: k.rr as a Function of the Soluble Boron Content..
...................................................
A-28 Figure A.7: k 0 rr as a Function of the EALF for the HTC Experiments
.....................................
A-39 Figure A.8: Predicted k.rr as a Function of the Plutonium Content..
........................................
A-42 Figure A.9: Predicted k.rr as a Function of the Am-241 Content .............................................. A-43 Figure A.10: LCT-046 Corrected Calculated k.rr per Case .......................................................
A-45 Figure A.11: LCT-046 Corrected Calculated k 0 rr Versus Temperature
...................................
A-46 NET-28091-0003-01 , Revision 0 Xlll 1 Introduction This report s ummari zes th e 20 1 7 criticality safe ty a n alysis (CSA) for the Indian P oint U nit 2 (IP2) spe nt fuel pool (SF P) takin g no cre dit for a b sorber pan e l s. The current 200 l CSA of record [36] t akes partial credit for Boraflex TM panels which h ave degrad e d an d co ntinu e to d egra d e. In order to remo ve the dep e ndenc e on the Borafl ex TM panels thi s new 20 17 CSA credits empty ce ll s , control rods , and leak age a lon g th e outer two storage rows of the SFP. In 2 015 , a CSA to remo ve cre dit for the Boraflex TM pan e l s which used ne w metal-matrix-composite absorber in se rt s was previously submitte d to the Nuclear Regulatory Conunission (NRC) an d was r eviewe d [l , 2 , an d 3]. How ever , the approach t ake n in this 2017 CSA i s ex p ecte d to re s ult in a more tim ely resolution of the Borafl ex TM degrad ation i ss ue. Since th e Indian Point Energy Ce nt er (IPEC) utili zes the Un it 2 (IP2) SFP for t emporary storage of U nit 3 (IP3) fuel prior to pl aceme nt into dry s torage casks , this 2 017 CSA a llow s storage in the IP2 SF P of all fuel asse mb l ies dis c h arge d from b oth IP 2 and IP 3. 1.1 Background The IP 2 SFP rac k s currently c redit Bo raflex TM as the n e utron absorber, w hich is known to degr ade over time. Du e to this fact , E nt ergy (the operator of IP EC) w ill no lon ger take c redit for the Borafl ex TM for r eac ti v it y h o ld-down. In early 2015, IP EC s ubmitted a 20 15 criticality a n a ly s is which used neutron a bsorber insert s to replace th e negative reactivity of th e Bo raflex TM [ 1]. During this process , the NRC re qu este d additional infonn ation in June 2015 a nd IP EC i ss u ed a r es pon se in August 2015 [2]. In November 2015 th e NRC i ssue d a staff review of this criticality a naly s i s , conc ludin g th at " The NRC staff fi nd s that the CSA methodology is acce ptabl e for u se at IP 2" [3]. Subsequent to thi s 2015 CSA re v i ew, scopi n g s tudies determined th at it wo uld be more timely and l ess c hall e n ging to refuelin g outages to l oa d a dditional fuel assemblies int o casks for dr y cask s tora ge and use e mpt y cells a nd control rods for criticality control. NET-28091-0003-01, Revi s ion 0 IP2 and IP3 are both 4-loop Westinghouse power p l ants that uti l ize the l Sx 15 fuel assemb l y design. The physical dimension requirements of the fuel of both units are the same , as both units have had all of their fuel assemblies manufactured by Westinghouse.
IP3 does not have the capability to load dry storage casks, so fuel from IP3 is moved to the IP2 SFP for temporary storage prior to l oading into dry cask storage. Placement of IP3 fuel is currently restricted to Region 1-2 in the IP2 SFP. This 2017 CSA allows the IP3 fuel to be placed anywhere in the IP2 SFP , so long as it meets the reactivity requirements outlined herein. 1.2 Description of the Analys i s This 2017 CSA determines the loading criteria for storage of fuel assemblies in the IP2 SFP by taking credit for empty cell locations , control rods , and the periphery (outer two rows) of the SFP. The analysis does not credit any Boraflex TM neutron absorber tha t might remain in the racks. Taking credit for empty cells and control rods can accommodate the current and future spent fuel inventory.
The analysis defines five reactivity categories for the fuel and defines storage locations for each reactivity category.
The categories are numbered from one to five with a Category 1 fuel assembly being the most reactive and a Category 5 fuel assembly being the least reactive.
Similarly , each cell in the SFP has also been assigned a category number with a Category 1 cell being able to accept the most reactive fuel whi l e a Category 5 cell can only accept the least reactive fuel. For example, a Category 1 cell can accept all categories of fuel while a Category 5 cell can only accept a Category 5 fuel assembly because all other (lower numbered) categories are more react ive. Figure 1.1 below shows the base case arrangement of the fue l categories in the IP2 SFP. Note that the base case arrangement only shows four reactivity categories since it is the most l imiting reac t ivity arrangement.
Category 1 fuel , which is missing on Figure 1.1, is only needed for fresh fuel* or l ow burned fuel, which should not be present after IP2 shuts down. Category l fuel will be controlled by two rules for placement in Region 1. The
* Throughout this document , fre s h fuel is used to describe fuel that has never been in the core. NET-28091-0003-01, Revision 0 2 assemb l y's enrichment , bu mup , coo lin g time , and averaged assemb l y peaking factor* are u sed to detennine the reactivity category. For permanently di scharged fuel, full use of operating data i s u sed to precisely obtain the change in react i vity with bumup. For fue l that m ay sti ll b e placed in the core , bounding d ep l etion va lu es are used. I 2 3
* S ' 7 I 9 10 11 U U 14 15 16 17 11 1' 20 Z1 ll lJ 24 25 2fi 27 21 29 30 31 H I--+-+-+--+--+--+--+--+--+-, ~--1--1---,l--of--f--+-+-+-+--+--+--+--+--+-+--+-+-+--l--l
--,I G D I--+-+-+--!--+--+--+--+--+---<
>-t-t-+--+--+--+--+-+-+--+--+--+--+--+---+---+--+---+--t---t
....... C I--+-+-+--+--+--+--+--+--+-, ~--1--1-f--f--l--+-+-+-+--+--+--+--+--+-+-+-+--+--1
--1-8 Key: 0 ~OP 0 ON OM Ol DI( OJ OH DG Of OE CP HH-+-+-+--+-+--1-+-+-l OwaterHole D S0%W ate r Hole
* category t Fuel D category 2 Fuel 0 category 3 Fuel .. CP CN CH """"'--fl
....... --+-+-+-+--+--+-
+-+-1-1
* category4Fuel ca t egory 5 Fu e l category s Fuel with a requi r ed full length RCCA [!) Blo t ked Cell Cl Cl( CJ QI CG a as 8l .. BJ 8H 8f 8( Al ... AJ AH AG Al ... AO AC CM HH-+-+-+--+-+--+-+--+-l CL 1--1,--i--+-+-+-t--+--+-+-+-i CK HH-+-+-+--+-+--1-+--+-l CJ HH-+-+-+--+-+-+-+-+-l CH i--1,--+--+-+-+-t--+-+""+-+-i CG HH-+-+-+--t--+--1-+-+-l Cf HH-+-+-+-++-+-+-+-l C, 1--1,--i--+-+-+-t--+--+--+-+-i co F9'=\=*-"p\,={=4=!
"'*'~ 1-HH-+-+-+--++-+-+-'
8H BM Bl .. BJ 8H BG Bf 8{ B O H-t-+-1-+,..-:~rt-H BC j Cask Area l--l,--l--t--t--t--+--t--t-+-r-lt-11-1--t--t--t--r-t--t--t---t-+-'l'-1 AB ...... ..,....~~~
........ ...._.._ ....... ..._ ................
_. ...... _._~--~-~~ Figure 1.1: Fuel Category Placement in the IP2 SFP (base case)
* The averaged a s sembly peaking factor is the assemb l y burn up divided by the sum of the cycle burn ups for the cycles the assembly was in th e core. T hrou g h out this document, this average d assemb l y p ealci n g factor i s often a bb reviate d as pealcing factor or PF. NET-28091-0003-01, Re v i sion 0 3 For most permanently discharged fuel from Units 2 and 3, a table of the fuel categorization for each assembly ID is provided.
This approach is taken since there are mu l tiple groupings of assemblies (with s imilar operational characteristics) due to changes in c ore power (temperatures), burnabl e absorbers , and axial blanket de signs. This CSA provides the discharged assembly categories in Appendix B which will be added to the Technical Specifications. Fuel assemblies that may still be placed in the core are categorized by a set of simple equations to determine the fuel reactivity category.
The fuel categories are numbered from most reactive fuel (Category 1 fuel) to lea st reactive fuel (Category 5 fuel). The storage loc atio ns in the SFP that can accommodate each category of fuel are limited by Figure 1.1 and modifications to this arrangement are specified and analyzed in this CSA. Details of the analysis methodology are provided i n the following sections of thi s report. Section 2 contains a summary of the methodology.
Section 3 provides the input data for the rack designs, fuel assembly design s, fuel assembly inserts, and plant operation data. Section 4 describes the computer code validation and determination of co de bia s and uncertainty.
Section 5 describes the depl etio n analysis and the selection of bounding input parameters.
Section 6 de scribes the basic fuel rack models , while Section 7 contains the sensitivity analysis of manufacturing tolerances and additional biases and un certaint ie s. Section 8 contains the result s of the reactivity calculations. Section 9 contains a summary of normal operation s and the accident analysi s. Section 10 provides a summary of the analysis including the limits of the analysis.
Appendix A contains the detailed results of the validation of SCALE 6.1.2 and its applicability to this analysis, while Appendix B con t ains the categorization of fuel assemblies previous l y discharged from IP2 and IP3. This new CSA for the IP2 SFP follows the most recent methods. The Nuclear Energy Institute (NEI) has been working with the NRC to produce guidance for SFP analysis [ 4]. This CSA closely follows the NEI guidance.
The NEI guidance started with the NRC draft Interim Staff Guidance (ISG) DSS-ISG-2010-01 [5]. All of the requirements set in DSS-ISG-2010-01 are met and are reviewed in Section 10.1. Additional guidance was provided by the Kopp Memo [56] for depletion of atom densities , NUREG/CR-NET-28091-0003-01, Revision 0 4 7109 [22] for the worth of minor actinides and fission products, and NUREG/CR-6998 [26] for the burnup uncertainty. 1.3 Acceptance Criteria The acceptance criteria of the analysis are to ensure compliance with lOCFRS0.68
[6]. Specifically, the analysis demonstrates that:
* the k 9s 1 9s of the SFP is less than 1.0 after accounting for a ll biases and uncertainties when not taking credit for soluble boron (with a 95% probability at a 95% confidence level) [6], and
* the k9 5 1 9s of the SFP is less than 0.95 after accounting for all biases and uncertainties when taking credit for solub l e boron (with a 95% probability at a 95% confidence level) [6]. In addition to meeting the above criteria , an eng i neering safety margin is provided to cover unanticipated issues. The engineering safety margin used is 1 %, so that the k 9 5 1 9 5 target value is 0.99 for no soluble boron and 0.94 with solub l e boron. NET-28091-0003-01, Revision 0 5 2 Methodology The CSA performed in this report uses a method that is comprised of the following steps. Each step refers to a section in this report where further information is provided. 1. Review the current IP2 SFP rack design (Section 3.1). 2. Review the historical and projected fuel designs and inserts for use in IP2 and IP3. Ensure that the analysis covers all of the designs (Sections 3.2 and 3.3). 3. Review the historical and projected operating cycles of IP2 and IP3 (Section 3.4). 4. Validate the computer codes for the application (Section 4). 5. Deplete the fuel using a two-dimensional (2-D) lattice representation of the core using bounding depletion values (including bounding burnab l e absorbers) for sets of fuel assemblies (down to individual assemblies) (Section 5). 6. Develop a radia ll y infinite three-dimensiona l (3-D) Monte Carlo mode l of the Region 1 and Region 2 racks using periodic boundary conditions. The axial modeling height is finite, including conservative modeling of the axial bumup distribution (Sections
 
===6.1 through===
6.5). 7. Develop full pool models to take advantage of leakage at the boundaries of the SFP as well as control rods at specific locations.
Use this model in checking interfaces between category cells (Section 6.6). 8. Based on the radially infinite 3-D Monte Carlo model , determine the reactivity effects associated with the manufacturing and fuel tolerances (Section 7.1). 9. Determine the bias and uncertainty associated with bumup (Section 7.2). 10. Determine the bias due to eccentric placemen t of fuel assemblies in the rack cell (Section 7.3). 11. Ascertain through analysis the most limiting SFP temperature by Region (Section 8.1). 12. Use the radially infinite 3-D Monte Carlo model with the combined biases and uncertainties to detennine the minimum bumup as a function of enrichment, averaged assembly peaking factor , NET-2809 1-0003-01, Revision 0 6 and cooling time for Categories 1, 2, and 4 at the most l imiting SFP temperature. This analysis is perfonned with no soluble boron (Sections 8.2 and 8.3). 13. Determine the fuel Category 3 and 5 additional burnup requirement and test the Region and category cell interfaces using a full pool 3-D Monte Carlo model (Section 8.4). 14. Perform accident analyses (dropped assembly, misplaced assembly, over temperature (boiling SFP water), boron dilution , seismic , and multiple assembly misloads) with the appropriate models (Section 9). 15. Summarize the resulting loading requirements and the limits of the ana l ysis (Section 10). 2. 1 Computer Codes This analysis uses the t5-depl TRJTON module of SCALE 6.1.2 [7] for the depletion analys i s and the CSAS5 module for the criticality ana l ysis. All of the analyses are perfonned using the 238 group ENDF/B-VII.O library (v7-238).
The CSAS5 mod ul e utilizes CENTRM and BONAM! for the resonance self-shielding calculations and KENO V.a for the Monte Carlo calculation of k.rr*. Unless noted, all of the CSAS5 computer runs use a Monte Carlo sampling of at least 8000 generations and 8000 neutrons per generation to achieve a statistical uncertainty in k err of less than 0.0001. The t5-depl sequence of TRJTON utilizes CENTRM and BONAM I for the resonance treatment and then uses KENO V.a for the collapsing of the cross-sections from 238 groups to one group for use in ORJGEN. The input parameter, parm=(a d dnux=4), is used in the analysis which tracks the maximum number of problem specific collapsed isotopes (388). At the end of the depletion ana l ysis , the OPUS module is used to output atom densities for use in the criticality model. In the OPUS input , 185 isotopes are specified, as shown in Table 2.1. The isotopes that are not included have low atom densit i es (less than l E-12), combined with small cross-sections, in the spent fuel composition. In other words, the e l iminated isotopes do not impact the reactivity of the spent fuel and consequently will not impact the criticality
* Throughout this document , ke rr is used as a short hand notation fork-effective. NET-2809 1-0003-01, Revision 0 7 analysis.
Immediately after shutdown, there is an increase in reactivity in the first few days due to the decay of Xe-135 and Np-239 (poison is being removed and fissile Pu-239 is being added). Rather than follow this change in reactivity and to assure that the peak reactivity occurs at 72 hours , all of the Xe-135 is converted to Cs-135 and all of the Np-239 i s converted to Pu-239. As pre v iously mentioned , atom densities les s than lE-12 are eliminated.
Table 2.1: 185 Isotopes Used in the Analysis Isotope Isotope Isotope Isotope Isotope Isotope Isotol!e Ag-109 Cm-243 Gd-160 Nd-145 Rb-85 Sm-153 Te-130 Ag-llOm Cm-244 Ge-73 Nd-146 Rb-86 Sm-154 Te-132 Ag-111 Cm-245 Ge-76 Nd-147 Rb-87 Sn-1 15 U-234 Am-241 Cm-246 Ho-165 Nd-148 Rh-103 Sn-116 U-235 Am-242m Cs-133 I-127 Nd-150 Rh-105 Sn-117 U-236 Am-243 Cs-134 I-129 Np-237 Ru-100 Sn-118 U-237 As-75 Cs-135 I-131 Np-238 Ru-101 Sn-119 U-238 Ba-134 Cs-136 I-135 Np-239 Ru-102 Sn-120 Xe-12 8 Ba-135 Cs-137 In-115 0-16 Ru-103 Sn-122 Xe-129 Ba-136 Dy-160 Kr-82 Pd-104 Ru-104 Sn-123 Xe-130 Ba-137 Dy-161 Kr-83 Pd-1 05 Ru-105 Sn-124 Xe-131 Ba-138 Dy-162 Kr-84 Pd-106 Ru-106 Sn-125 Xe-132 Ba-140 Dy-16 3 Kr-85 Pd-1 07 Ru-99 Sn-126 Xe-133 Br-81 Dy-164 Kr-86 Pd-108 Sb-121 Sr-86 Xe-134 Cd-110 Er-166 La-138 Pd-110 Sb-123 Sr-88 Xe-135 Cd-111 Eu-151 La-139 Pm-147 Sb-124 Sr-89 Xe-136 Cd-112 Eu-15 2 La-140 Pm-148 Sb-125 Sr-90 Y-89 Cd-113 Eu-153 Mo-100 Pm-148m Se-76 Tb-159 Y-90 Cd-114 Eu-154 Mo-95 Pm-1 49 Se-77 Tb-160 Y-91 Cd-115m Eu-155 Mo-96 Pm-1 51 Se-80 Tc-99 Zr-91 Cd-116 Eu-156 Mo-97 Pr-141 Se-82 Te-122 Zr-93 Ce-140 Gd-152 Mo-98 Pr-143 Sm-147 Te-124 Zr-95 Ce-141 Gd-154 Mo-99 Pu-23 8 Sm-148 Te-125 Zr-96 Ce-142 Gd-155 Nb-95 Pu-239 Sm-149 Te-126 Ce-143 Gd-156 Nd-142 Pu-240 Sm-150 Te-127m Ce-144 Gd-157 Nd-143 Pu-241 Sm-151 Te-128 Cm-242 Gd-158 Nd-144 Pu-242 Sm-152 Te-129m NET-28091-0003-01 , Revision 0 8 In addition to using SCALE, a FORTRAN code (INTRPND) is u sed to interpolate between bumup s from the OPUS outpu t and also to decay the isotopic content to the de sired coo ling time. The INTRPND code, which has been verified and va lidated [ 1 O], reads an axial burn up profile to get the shape of the bumup axially, so multiple atom density sets can be made quickly. The code was validated by comparing the k eff calculated with the code-interpolated number den sities to the k e tr calcu lated with number densitie s directly from SCALE/TRITON, in which n o interpo l ation is used. Furthermore , SCALE/TRITON is used to decay to a given cooling time and similar comparisons were made. A ll of the differences between the k e ff va lue s based upon the interpolated isotopics an d the SCALE direct isotopics are with in the statistical unc ertainty of the k e ff calculations (see Section 5.9). The INTRPND FORTRAN program is contro lled under NETCO's quality assurance program that meets the requirements of lOCFRSO , Appendix B, 10CFR21 , and ASME NQA-1. The program ha s been audited by NUPIC. NETCO maintains documented procedures and assigned responsibilities to control the engineer ing activities relative to the acquisition, classification, d eve l opment , testing, eval uation , modification, use, maintenance, retirement, and user notification of computer software utilized by NETCO for applicatio n s that are s af e ty r e lat e d or important to saf e ty. Software i s controlled under NETCO Standard Operating Procedures (SOPs) in the Standard Operating Procedures Handbook. Specifically, SOP 2.4, Software Contro l , provides procedures for software acquisition, software de s i gn, E rror Notification , Configuration Control, User Documentation, Verification/Validation, Software Testing/Benchmarking and Run Log maintenan ce. T h ese features are subject to further procedural control as provided for in NETCO Software Control Procedures , SCP-001, Proc e dur e for C l assification of NETCO Softwar e Us e d for En g in ee ring Calculations, and SCP-002 , Proc e dur e for NETCO Comput e r Identification and Installed Softwar e In v e ntory. Additionally , SOP 2.10 , Control of Manual and Computerized Calculations , provides procedures for documentation of the accuracy, traceability and verifia bility of computerized calculations.
NET-28091-0003-01, Revision 0 9 Unless otherwise specified , all of the k err values reported in this document are raw calculated k eff values with no adjustment for bias and uncertainty. The final values to be compared to the criticality criteria are the calculated values plus the total bias and uncertainty (notated as " k 9s 1 9s"). NET-28091-0003-01 , Revision 0 10 3 Input Data For the criticality analysis , input data is needed for the SFP and storage racks (Section 3 .1 ), the fuel assemblies (Section 3.2), the fuel assembly inserts (Section 3.3) and the plant operating data (Section 3.4). 3. 1 SFP and Storage Rack Specification s The IP2 SFP is shown in Figure 3.1. It is lined with a 0.25 inch stainless steel plate covering the concrete walls. In this SFP are three Region 1 (flux t rap) modules and nine Region 2 modules. The southwest corner (bottom right in Figure 3.1) is an empty area for placement and l oading of storage and transport casks. The southeast corner ( top right in Figure 3 .1) contains two large cylinders for containing failed fuel and the new fuel elevator.
The placement of the modules is shown on Figure 3 .1. The Region 1 fuel racks (flux trap design) contain Boraflex TM in sheaths. The Boraflex &#x17d; is not credited in this analysis.
Figure 3.2 shows the general arrangement of the cells in a Region 1 module [8]. Table 3.1 contains the dimensions and to l erances from the manufacturer
's draw i ng for Regio n I [8]. Region 2 is an egg-crate design where square storage tubes with Boraflex TM sheaths are joined at the corners via spacer rods creating "resultant" cells between the tubes. Figure 3.3 shows two complete cells in the Region 2 type rack [9]. The cell on th e left wou l d be called the " resultant" cell. Notice the fuel in the resultant cell is not bounded by four flat walls but rather by the Boraflex TM sheaths. The dimensions for the Region 2 rack are also shown in Table 3.1 [9]. Two rack positions will be filled with cell blockers. The cell blockers will be made of stainless steel and will not displace more than 50% of the water at the elevations containing fuel rods. NET-2809 1-0003-0l , Revision 0 II 
'J' * *~ Ce:LL OAT (GIOU 11 CI Q.\C S. l.!f.91 Ct'. LS P1 CH-I&:* 1, :* ~lGICN 2: 1-,1 RAC-<S 11 ~IC U.!
* PJTC -9.C 4 TOTAL: ;. 1121 QAC S 03 1 1 El.l'i ,. .. u , I. PAlfO r fl ,o .. u., ..* , !.* .._ ____ _ Figure 3.1: IP2 SFP Taken From Holtec Drawing #397 [35) N ET-28091-0003-01 , R ev i s ion 0 1 2 Figure 3.2: Small Section of the Region 1 Rack [8] Figure 3.3: Region 2 Rack Showing Cell Boxes and Resultant Cells [9] NET-28091-0003-01 , Revision 0 13 Proprietary Information Removed Ta bl e 3.1: R eg ion 1 an d 2 St o rage R ack Di mens io ns (8, 9] Va lu e Att ribu te (i nc h es) To l era n ce Region 1 Rack f 81 Vertical cell pitch 10.545 Horizontal cell pitch 10.76 5 Cell ID 8.75 Cell wall thickness 0.075 +/- 0.007 Boraflex TM sheat hing width 7.70 Minimum Boraflex TM sheathing thicknes s 0.035* +/- 0.003 Boraflex TM sheathing distance 0.112 --from cell wall Connecting plates thickness 0.09375 Not used Regio n 2 R ack f 91 Cell pitch 9.04 I I Cell ID 8.80 Cell wall thickness 0.075 +/- 0.007 Boraflex TM sheat hing width 7.70 Minimum Boraflex TM sheathing thickness 0.035 +/- 0.003 Boraflex TM shea thing distance 0.092 --from cell wall *0.0235" and 0.035" wra ppers allowed per drawing The racks are constructed of Type 304 stainless steel [8, 9]. The Boraflex TM is mod e led as water. If any Boraflex TM remained , it would contain 1 0 B, which would decrease rea ct ivity compared to water. The E lectric Power Re sea rch Institute (EPRI) Technical R eport 103300 Section 2 (55] concludes the rate of boron carbide loss is proportional to the rate of the silica loss. Therefore, a water displacing material without boron is not credible. 3.2 Fuel Assembly Designs IP2 and IP3 have used a number of fuel de s igns. All of the fue l assemblies for both units were purchased from Westinghouse and each of the fuel design s used the same clad and pellet dimensions for each of the 15 x 15 designs. The guide tube material and dim e nsions ha ve changed as IPEC transitioned from HIP AR and LOP AR fuel to OF A fuel. The only difference relevant to criticality between Standard (LOP AR) and OF A fuel is that the guide tube outer diameter for OF A fuel is s lightly sma ller but the t hickness of the guide tube is unchanged. NET-2809 1-0003-01, Revision 0 14 The following Westinghouse fuel designs have been used at IPEC: 1. HIP AR: This was the initial fuel design and used stainless steel guide tubes and Inconel grids (Cycles 1 through 4 feed assemblies for IP2). HIPAR was ne v er used in IP3. 2. Standard Fuel (LOPAR): Changed from stainless steel to Zircalo y guide tubes (Cycles 5 through 9 feed assemblies for IP2 , Cycles 1 through 4 feed a s semblies for IP3). 3. OFA Fuel: Changed to Zircaloy grids and a small reduction in the guide tube diameter (Cycles 10 , 11, and 12 feed assemblies for IP2 , Cycles 5 and 6 feed assemblies for IP3). 4. Vantage 5 Fuel (VS): Added six inch natural uranium axial blankets and use of integral fuel burnable absorber (IFBA) which incorporates a thin coating of ZrB2 on the outside cylindrical surface area of the fuel pellet (Cycles 7 , 8, and 9 for IP3 , not used in IP2). 5. Vantage+ Fuel with Zircaloy-4 Grids (V+ Z4): Added intermediate flow mixing grids , changed the fuel clad to ZIRLO&#x17d;, and added enriched axial blankets (Cycles 13 and 14 for IP2 , IP3 did not use this fuel design). 6. Vantage+ Fuel with ZIRLO&#x17d; Grids (V+ Zlo): Same as V+ Z4 but changed the grid material to ZIRLO&#x17d; (Cycle 15 for IP2 , C y cle 10 for IP3). 7. Performance
+ Fuel (P+): Added a protective bottom grid and longer bottom end plugs. (IP2 Cycle 16, IP3 Cycles 11 , 12 and 13). 8. 15X15 Upgrade Fuel (15Up): Changed the grid design and modified the guide tube dashpot by using a tube-in-tube (more water displacement
-less reactive). (IP2 Cycle 17 to present , IP3 Cycle 14 to present).
Most of the fuel assembly design changes have been related to the grids. The grids are ignored (modeled as water) in the criticality modeling, since they displace water in fuel assemblies.
PWR fuel de signs are under moderated so that there is a negative moderator coefficient of reactivity.
Even at cold temperatures the fuel assembly designs are under moderated. With high soluble boron concentrations , it NET-2809 1-0003-01 , Revision 0 15 can be non-conservative to ignore grids , but for borated cases, there is a large margin to the criticality limits (see Section 8.14), so ignoring the grids is still acceptable.
Independent from the 8 designs listed above , Westinghouse has also changed the pellet theoretical density , chamfering and dishing. Only the stack density i s modeled which is the average density of the U0 2 inside the pellet OD. These historical stack densitie s ha ve been determined by taking the measured assembly U0 2 mass* divided by the fuel volume ((15* 15-21)* 144*rc*(pellet OD)2/4). The maximum stack density of any of the historical fuel is 95% of the theoretical density. The maximum stack density of any assembly of current fuel (Batch Wand beyond for both units)t is 94.83% of the theoretical density. A value of 95% of the U0 2 theoretical density i s used for all of the analysis ( except for some individual assembly depletions for which the as-built stack density is known). An uncertainty of 0.0035 in the fraction of the theoretical density is added to the uncertainty rack up for the future reload fuel. This is the difference between the average fraction of the theoretical density and the maximum observed for IP2 fuel Batches 2A through 2D (the average fraction of the theoretical density for all of these assemblies is 0.9448). Finally, several axial blanket designs have been used at the Indian Point Units. This analysis credits the lower enrichment in the axial blankets.
For the axial blanket types used , see Tables 3.5 and 3.6. The fuel dimensions and tolerances are taken from References
[11] and [12] and are provided on Table 3.2. The dimensions used in the analysis are the nominal dimensions and the reacti vity effects of tolerances are combined statistically which is the square root of the sum of the squares of the individual uncertainties ( denoted as RSS). The reactivity effect of the guide tube tolerance has been shown to be insignificant to the reactivity
[ 13].
* For blanketed fuel , the ma ss ofU02 is the unblanketed ma ss and the length is adjusted to the unblanketed length t See Tables 3.5 and 3.6 for a description of fuel batches NET-28091-0003-01, Revision 0 16 Westinghouse Non-Proprietary Class 3 The assembly pitch in the core for Westinghouse 15xl5 plants is 8.466 inches [11]. The uniform pin pitch could only increase (8.466-15 x 0.563)/15 = 0.0014 inches b efore the core inter-assembly gap is gone. A pin pitch tolerance of0.0014 inches is used for the analysis (see Table 7.1). Table 3.2: Fuel Assembly Dimensions
[11, 12] Value Attribute (inches) Tolerance Used Fue l pellet U0 2 stack density 95.0%TD 0.35% Fuel pellet OD 0.3659 r l"*c Fuel clad OD 0.4220 [ ]"*c Fue l clad ID 0.3734 r 1 3.C F u e l pin pitch 0.5630 +/- 0.0014 (using assembly pitch) Active fuel length 144 None (insi1mificant effect) HIPARFuel Stainless stee l guide tube OD 0.545 Insignificant effect on k eff [ 13] Stainless stee l guide tube ID 0.515 Insignificant effect on k eff [ 13] LOPARFuel Guide tube OD 0.546 Insignificant effect on k e ff f 13 l Guide tube ID 0.512 Insi 1mificant effect on k e m 131 Current Fuel (OF A and beyond) Guide tube OD 0.533 Insignificant effect on k e m13] Guide tube ID 0.499 Insignificant effect on k e m 13 l The fuel clad and guide tube material is Zircaloy-4 or ZIRLO&#x17d; except for HIPAR fuel , which used stainless steel for the guide tubes. For fuel pellets that are coated with ZrB 2 (IFBA), the 1 0 B loading is ]"*c mg 1 0 B/inch (IX) for IFBA rods in fresh fuel assemblies in the calculation of the SFP k eff and a maximum of[ ]"*c mg 1 0 B!inch ( 1.5X) for depletion calculations
[ 12]. 3.3 Fuel Assembly Insert Designs The fuel assemblies used in Indian Point Units 2 and 3 have contained a number of different types of inserts in the guide tubes during operation.
They are: 1. Pyrex burnable absorbers
: 2. Wet Annular Burnable Absorbers (W ABA) 3. Unclad hafnium flux suppression assemb li es 4. Primary source assemb li es NET-28091-0003-01 , Revision 0 17 L 5. Secondary source assemblies
: 6. Full-Length Control Rods 7. Part Length Control Rods The criticality calculations credit control rods inserted in assemblies at specified locations in the SFP as well as the use of additional control rods to reduce the reactivity of assemblies that fail to meet the loading requirements. No credit is taken for any other inserts in the criticality calculations.
However , depletion calculations are performed with inserts in order to harden the spectrum and m a ximize the reactivity of burned fuel. The depletion analysis is performed with groupings of assemblies.
The groups may be as small as one assembly or as big as all of the feed assemblies from multiple cycles. Section 5 describes the assembly groupings.
In each group, the effect of the inserts is maximized by using the highest boron content and the most water displacement.
For groups where the boron loading for the burnable absorbers has varied, the maximum boron loading has been used in the analysis.
A Pyrex burnable absorber displaces more water and has a higher 1 0 B loading than a W ABA, so for groups which have both burnable absorbers , the Pyrex burnable absorber is conservatively selected. Primary and secondary sources displace les s water than burnable absorber inserts. Therefore they can be ignored in the depletion analysis.
Indian Point Units 2 and 3 had 8 part length control rods in Cycle l. The part length control rods were in assemblies Al 1, A24, A47 , A49, ASO, AS 1, A54 , and ASS in IP2 and assemblies A38 , A43, A44 , A45 , A51 , A59 , A63, and A64 in IP3. These assemblies have an enrichment of2.25 wt% 235 U-235 (w/o)* and burnups greater than 16 GWd/MTU. Due to reactivity margin after burnup for these assemblies , part
* Throughout the document , th e s ymbol w/o m e an s we i ght percent 135 U NET-28091-0003-01 , Revision 0 18 length control rods were not part of the analysis of this CSA. Details on the disposition of these assemblies are given in Section 8.6. Thim bl e plugs have a l so been used at the Indian Point Units , but since these do not exten d into the active fuel region they can be ignored. Tab les 3.3 and 3.4 provide the input data used in th e analysis.
Note that the Pyrex burnable a b sorber u s ed at Indian Point had a range of 1 0 B l oadings. T h e hi ghest 1 0 B l oadin g is conservative for the depletion analysis a nd that value is in Table 3.4. Further, the l argest borosilicate glass dimensions are utilized which is a lso conservative.
Table 3.3: Control Rod and Hafnium Rod Descriptions
[11] Parameter Control Rod Hafnium Number of R odlets er assembl 20 20 or l ess . Absorber OD in 0.3975 0.3810 Absorbe r Material Ag-In-Cd Hf 80-1 5-5 wt% Absorber densit cc 10.17 13.3P Clad OD in 0.4390 0.38 10 t Clad ID (in) 0.4006 none Clad Materia l SS304 none Table 3.4: Pyrex and Wet Annular Burnable Absorber Descriptions
[11, 12, 15] Parameter Material Inside inner clad C lad material Inner C l ad ID in Inner C l ad OD in) A b sorber ID (in) A b sorber OD in Absorber Material P rex Void SS304 0.2235 0.2365 0.2430 0.3960 0.4005 0.4390 B 2 0 3-Si0 2 (18.1 wt% B 2 0 3) Densit = 2.23 cc t WABA Water Zr 0.225 0.267 0.278 0.318 0.329 0.381 Ah0 3-B4C (0.00603 gm 1 0 B/cm)
* 20 ro dl ets were u s e d in the analysis.
To date the maximum rodlets u sed i s 16. t From the SCALE manual , Ref. [7]. i T hi s i s the OD of the unclad hafnium. NET-2809 1-000 3-01 , R evision 0 19 Westinghouse Non-Proprietary Class 3 3.4 Plant Operation Data Plant operational data is needed for the depletion analysis , the axial bumup profile used for the SFP k calculations, and for the reactivity categorization of discharged fuel. Tables 3.5 and 3.6 show the power and temperatures for each operating cycle for IP2 and IP3. In addition , Tables 3.5 and 3.6 identify for each cycle, which burnable absorber design was used and the highest cycle bumup for any assemb l y containing a burnable absorber.
The tables also provides the fuel feed enrichment and axial blanket design , enr ichm ent and l ength for eac h cycle. The end-of-cycle (EOC) burnup is a ls o included on the tables. Fina ll y, the cycle-average solub l e boron concentration (ppm) is given. On Tab l es 3.5 and 3.6, T;n is the core inlet temperature and Ta ve is the average moderator temperature i n the active fuel (not the vessel average temp erature). In genera l the vessel average temperature is approximately 3 &deg;Flower than the core average temperature. For example in Cycle 21 of IP2 , the core average temperature is [ &deg;F]a , c (see Table 3.5) but the vessel average temperature is 565 &deg;F. Throughout this report , T ave refers to the core average temperature.
In addition to the information on Tables 3.5 and 3.6 the following operational data was used: 1. The enrichment, discharge bumup, and discharge date of all assemblies.
This is used for the reactivity categor ization of the fuel assemblies as well as the determination of the average peaking factor for the assembly. 2. All of the assemb ly l ocations and their inserts for every cycle. This is used for adjustments for depletion with control rods and hafnium inserts. It is also used for assemb l y specific analysis when the assemb ly bumup is slightly less than a batch grouping burnup requirement.
: 3. The end-of-cyc l e (EOC) bumup for each assembly.
This is u sed to weight the cycle average ppm to assure that no assembly exceeds the solub l e boron concentration (ppm) used in the depletion analysis. 4. The bumup by axial node for all blanketed fuel assemb li es. This data is used to determine bounding axial bumup distributions.
N ET-28091-0003-01, Revision 0 20 Westinghouse Non-Proprietary Class 3 Table 3.5: Ke y Operating Features by Cycle Used i n IP2 Avg Feed Cycle Feed Blanket Max BA Soluble Batch Power Tav e Tiu Burnup Enrich. Enrich. BA* Max BA Burnup Fuel Boron Cycle ID (MWt) (OF) (OF) (GWd/T) (w/o) Blanket (w/o) Type Loading (GWd/T) Design (ppm) 1 A, B ,C 2758 -a , c 1 6.4 2.2/2.8/3.3 n o n e P yrex 20 R od l ets 1 8.5 HIPAR 545 -2 D 2758 10.7 3.1 none -P yrex 20 R od l ets 13.0 HIPAR 576 3 E 2758 10.8 3.2 none -P yrex 7 R od l ets 12.0 HIPAR 497 4 F 2758 9.8 3.35 none -Pyrex 12 R odlets 12.2 HIPAR 575 5 G 2758 12.2 3.3 none -P yrex 12 R odlets 15.0 LOPAR 486 6 H 2758 1 3.2 3.2 none -P yrex 20 R od l ets 16.7 LO PAR 464 7 J 2758 12.4 3.44 none -Pyrex 16 R odlets 15.5 LOPAR 491 8 K 2758 13.8 3.2/3.44 none -WABA 12 R odlets 17.4 LOPAR 511 9 L 2758 11.4 3.4/3.7 none -WABA 1 6 Rodlet s 14.6 LOPAR 587 10 M 2758 13.3 3.6/4.2 WABA 20 Rod l e t s 17.3 OFA 7 1 6 307 1.4 non e -11 N 307 1.4 18.1 3.75/4.0 5 IFBA 116(1 X) 24.5 OFA 615 none -WABA 16 R od l ets 12 p 3071.4 20.7 3.6/4.2/4.6 IFBA 116(1X) 28.1 OFA 725 none -WABA 20 R od let s 1 3 Q 3071.4 20.9 4.4/4.8 6" ann ul a r 2.6 IFB A 148(1.5X) 28.8 V+Z4 797 WABA 12 Rodlets 14 R 3071.4 19.0 4.6/4.95 6" arurn l ar 2.6 IFBA 148(1.5X) 26.7 V+Z4 876 WABA 20 Rodlets 15 s 3071.4 22.1 4.8 6" an nul a r 2.6 IFBA 148(1.25X) 29.9 V+Zlo 804 WABA 16 Rodlet s 16 T 3 114.4 23.9 4.6/4.95 8" an nul a r 3.2 IFBA 1 48(1.5X) 32.4 P+ 791 WABA 20 Rodlets 1 7 u 3216 18.7 4.0/4.4 8" annu l ar 3.2 IFBA 148(1.25X) 24.9 15Up 670 WABA 16 Rodlets 18 V 3216 24.5 4.6/4.95 8" so lid 3.2 IFBA 148(1.25X) 33.1 15Up 8 1 2 except IFBA WABA 20 Rodl ets 19 w 3216 24.8 4.6/4.95 8" so lid 3.2 Bot IFBA 148(1.25X) 32.6 15Up 847 except IFBA 3.4 Top WABA 20 Rodlet s 20 X 3216 23.7 4.8/4.95 8" so lid 3.2 B ot IFB A 148(1.25X) 32.3 15Up 890 except IFBA 3.6 Top WABA 20 Rodl e t s 2 1 2A 3216 25.6 4.6/4.95 8" so lid 3.6 B ot IFB A 14 8(1.25X) 33.8 15Up 851 --ex c e pt IFBA 4.0 Top WABA 20 Rodlets
* BA is burnabl e a b so rb e r NET-2 8 091-0003-01, Re vision 0 2 1 Westinghouse Non-Proprietary Class 3 Table 3.6: Ke y Operating Features b y C y cle U sed in IP3 Avg F e e d Cy cl e Fee d Blank e t M a x B A S olubl e Batch Pow e r Tave T i n Burnup Enrich. Enrich. B A* Ma x B A Burnup F uel Boron Cy cle ID (MWt) ("F) ("F) (G W d/T) (w/o) Bl a nk e t (w/o) T v o e Loa din2 (GW d/T) D es i2n (Dom) 1 A,B,C 3025 ~.c 17.3 2.2 5/2.8/3.3 none -Pyrex 20 Rodlets 19.4 LOPAR 515 2 p 3025 11.3 3.1 none -Pyrex 12 Rodlets 10.6 LOPAR 528 3 R 3025 12.8 3.3 none -Pyrex 12 Rodlets 16.2 LOPAR 464 4 s 3025 14.1 3.2/3.4 none -Pyrex 20 Rodlets 17.3 LOPAR 534 5 T 3025 14.3 3.2/3.4 none -WABA 16 Rodlets 17.9 OFA 578 6 u 3025 14.8 3.2/3.6 none -WABA 12 Rodlets 19.5 OFA 576 7 V 3025 13.4 3.4/3.8 6" so lid 0.72 IFBA 60(1X) 18.2 V5 513 WABA 20 Rodlets 8 w 3025 14.1 3.8/4.2 6" so l id 0.74 WABA 12 Rodlets 18.9 V5 642 9 X 3025 19.2 4.0/4.4 6" so l id 0.74 IFBA l 16{1.5X) 25.9 V5 694 WABA 20 Rodlets 10 y 3025 22.4 4.4/4.6 6" annular 2.6 IFBA 80(1.5X) 31.1 V+Zlo 777 WABA 20 Rodlet s 11 AA 3025 18.7 4.3/4.6 6" annular 2.6 IFBA 80(1.25X) 25.9 P+ 676 WABA 20 Rod l ets 12 BB 3025/ 23.0 4.5/4.9 5 8" annular 3.2 IFBA 100(1.5X) 30.3 P+ 832 3067 WABA 20 Rodlets 13 cc 3067.4 23.7 4.95 8" annular 3.2 IFBA 100(1.5X) 32.1 P+ 828 WABA 20 Rodlets 1 4 DD 3180/ 25.2 4.6/4.95 8" an nul ar 3.2 IFBA 148(1.25X) 33.1 15Up 843 3188 WABA 20 Rodlets 15 EE 3188.4 24.9 4.6/4.95 8" solid 3.2 IFBA 148(1.2 5X) 32.0 15Up 840 except IFBA WABA 20 Rodlet s 16 FF 3188.4 24.1 4.6/4.9 5 8" solid 3.2 Bot IFBA 148{1.25X) 31.6 15Up 813 except IFBA 3.6 Top WABA 20 Rodlets 17 GG 3188.4 25.0 4.95 8" solid 3.6 Bot IFBA 148(1.25X) 33.2 15Up 844 except IFBA 3.8 Top WABA 20 Rodlets 18 3D 3188.4 24.6 4.6/4.95 8" so lid 3.6 Bot IFBA 148(1.25X) 32.0 15Up 844 except IFBA 4.0 Top WABA 20 Rodlets --* BA is burnable absorbe r NET-28091-0003-01 , Revision 0 22 4 Validation The validation of the SCALE 6.1.2, TRITON (t5-depl) and CSAS5 models requires a number of steps. First, validation for the U0 2 , structural materials, and control rods is performed by modeling the U0 2 critical benchmarks from the OECD/NEA handbook , NUREG/CR-6361, and BA W-1810 [ 16 , 17, and 1 8]. Second, si nce burned fuel contains plutonium and Am-241, Mixed Uranium/Plutonium Oxide (MOX) critical experiments including the HTC critical experiments are added. Since burned fuel has a range of plutonium content from close to zero wt% to about 1.5 wt%, the bias and uncertainty is found by taking the most limiting bias and uncertainty set from either the U0 2 or MOX critical experiments sets. The above validation steps assume that the fuel composition is known. The third step is to account for the bias and uncertainty in the determination of the isotopic content (depletion analysis).
It is assumed that t he bias in the calculated k etr is zero and the uncertainty in the calcu l ated k etT resulting from the u ncertainty in the isotopic content is 5% of the reactivity decrement (L\k) to the bumup of interest.
This assumption is acceptable in DSS-ISG-2010-01 (5] and i s supported by analyses by Oak Ridge National L aboratory (ORNL) and EPRI (19 , 20, and 21]. The final step for burned fuel is to add a bias and uncertainty for th e possib l e error in the fission products and minor actinides cross sections. ORNL studied the uncertainty in the cross-sections and used TSUNAMI to propagate this uncertainty to a cask system (22]. ORNL concluded that a bias of 1.5% of the reactivity worth of the fission products and minor actinides is conservative.
On page 106 of the ORNL report , the la st sente nce states, "A n upp er va lue of 1.5% of the worth is also applicable for SNF isotopic compositions consisting of all nuclides in the SFP configuration." For this analysis, the bias and uncertaint y to cover the fission products and minor actinides cross sections is a bias of 1.5% of the worth of the fission products and minor actinides.
The details of the validation based on critical experiments a re found in Appendix A. This section briefly describes the method and summarizes the results. NET-2809 1-0003-01, Revision 0 23 
: 4. 1 U02 , Structural Materials , and Absorbers Va li datio n The validation for U0 2 , s tructural materi a l s , a nd absorbers follows NUREG/C R-669 8 [23]. Thre e hundred and tw enty eight (328) c ritical experiments w e r e se l ecte d from the OECD/NEA h a ndbook , NUREG/CR-6361, an d BW-1810 that m atch the co ndition s of the IP2 SFP. T h ese experi ments were analyzed with SCALE 6.1.2 using the 238-group EN DF/B-VII.O cross-section lib rary. The resulting pr edicte d k err values were th en teste d for trend s on th e k ey parameters that influe n ce k. Usi n g these trends, the single most limitin g bias and unc erta inty in th e area of ap pli cabi lit y fro m all of the tr en d s is d etermi ned. Although so m e of the tre nd s m ay not be sta ti stica ll y s ignifi cant , it i s conservative to u se a ll of the tre nd s in determining the limiting bi as and un certainty. Ta bl e A.5 is the area of app licability for t he va lid at ion. The IP2 SFP is covered b y the area of a pplic ability of the validation. Specifica ll y , 1. E nri chmen t: The b ench mark s se l ecte d spa n a range of e nrichm e nt s fro m 2.35 to 6.90 w/o. The fuel in th e SFP ranges from 2.2 1 to 5.0 w/o. The bias d ecreases with lo weri n g e nrichm e nt and the slope i s small so extrapo l ating to 2.21 i s appropr i ate (see Ta bl e A.5). 2. Spectrum:
The selected b enc hm arks cover a wide range of the neutron e n ergy spect rum by varying the pin pitch. T he Energy of the Average Lethargy causing Fission (EALF) of the benchmarks ranges from 0.0605 to 0.8485 eV. The ca lcul ated EALF in the SFP ran ges from 0.1 to 0.65 eV. 3. Fuel Pin Pitch: The fuel pin pitch of the selected benchmarks ranges from 1.075 to 2.54 cm. The Indian Point fuel pin pitch i s 1.4 3 c m. 4. Flux Trap: T he b enchmar ks includ e spaci n g between assemblies ofO to 15.4 cm. The flux trap design for R egion 1 is 3.4 to 4.0 c m. 5. Ag-In-Cd control rods: 51 critical expe rim e nt s we r e se l ected that includ e Ag-I n-Cd control rods. Ag-In-C d control rods are cre dited in th e SFP a n a l ysis. NET-2809 1-0003-01 , R ev i s ion 0 24 
: 6. Soluble Boron: The benchmarks have soluble boron concentrations up to 5030 ppm. The maximum ppm used in this analysis is the minimum Technical Specification limit of 2000 ppm. Details on the area of applicability are found in Appendix A. The most limiting bias and uncertainty from this validation is due to a trend in the neutron energy spectrum, EALF. From this trend , a bias of 0.0024 6k for storage systems with an EALF up to 0.4 eV and 0.0036 6k for storage systems with an EALF from 0.4 to 0.65 eV has been determined.
Fuel storage systems without soluble boron are covered by the first range of EALF and use 0.0024 6k for the bias. Heavily borated storage configurations typically have an EALF greater than 0.4 eV and therefore would use the 0.0036 6k bias. The 95/95 uncertainty is 0.0035 6k for all of the analyses. 4.2 MOX Validation The critical experiments that most closely match spent nuclear fuel are the HTC critical experiments
[24]. These experiments are Mixed Uranium/Plutonium Oxide (MOX) experiments that were designed to match the uranium and plutonium isotopic content of 4.5 w/o fuel burned to 37.5 GWd/T. These experiments were analyzed for this validation. Since t here are fuel assemblies in the SFP that have less burnup , as well as fuel assemblies that have more burnup, additional critical experiments a re needed. MOX experiments from the International Handbook [16] were added to cover the higher burned conditions and the fresh U0 2 critical experiments from Section 4.1 cover the lower burned conditions.
The maximum plutonium content in spent fuel is less than 2 wt% Pu. The Internation a l Handbook [ 16] has 39 critical experiments with less than 2 wt% Pu. There are 117 HTC critical experiments applicable to this analysis [24]. The HTC critical experiments are all 1.1 wt% Pu. Combined there are 156 MOX critical experiments used for the validation. NET-28091-0003-01 , Revision 0 25 Because of the limited number of independent experiments, only a trend on EALF was detennined. The bias from this MOX set (including HTC) at 0.4 eV for the EALF is only 0.0021 ~k but due to the small set of experiments and larger variation in calcu lat ed k, the uncertaint y in the bias is 0.0087 ~k. Although the bias for the MOX cases is slightly sma ll er than the U0 2 set , the uncertainty is much larger. As pointed out in a previous RAJ [2], the higher uncertainty may result in requiring the use of the MOX bia s and uncertainty set. Indeed , although calculations are performed using both the U0 2 bias and uncertainty and the MOX bia s and uncertainty , the MOX bia s and uncertainty is more limiting in all cases with burned fuel. The U0 2 bias and uncertainty is u s ed for fre s h fuel. A s with the U0 2 v alidation set , the MOX set has a higher bias and uncertainty for the harder neutron s pectrum. For EALF between 0.4 and 0.65 eV , the MOX bia s is 0.0027 ~k and the uncertainty is 0.0112 ~k. 4.3 Critical Experiments Effect on the Final k9s19s The final k 9s 1 9s that is compared to the criticality limits uses the more limiting bias and uncertainty from the U0 2 or MOX/HTC sets. The two set s of bias and uncertainty are: 1. Based on the U0 2 experiments
: For EALF less than 0.4 eV the bias is 0.0024 ~k. For EALF between 0.4 eV and 0.65 eV the bias i s 0.0036 ~k. The uncertainty for the entire range of EALF is 0.0035 ~k. 2. Based on the MOX/HTC experiments:
For EALF less than 0.4 eV the bias is 0.0021 ~k. For EALF between 0.4 eV and 0.65 eV the bias is 0.0027 ~k. The uncertainty for the range ofEALF l ess than 0.4 eV is 0.0087 ~k. The uncertainty for the range of E ALF 0.4 to 0.65 eV i s 0.0112 &. For all burned fuel , the MOX/HTC bias and uncertainty is limiting and used to determine the k 9s 1 9s. NET-2809 1-00 03-01, Revision 0 26 A set of experiments was used to determine a !ik bias for high temperature (but less than boiling).
If higher temperature cases are more limiting (high ppm or strong dependence on water density in empty cells), then a temperature bias of 8.6E-6 !ik/&deg;C multiplied by the difference in temperature between the desired temperature and 20 &deg;C is applied. Also, an uncertainty in the temperature bias due to the spread in the calculated k's is 0.0013 !ik. This is included in the statistical treatment of the uncertainties. NET-28091-0003-01 , Revision 0 27 5 Depletion Calculations The objective of the depletion ana l ysis is to generate atom densities that are used in SFP analysis.
The dep l etion analysis depends on the physical characteristics of the fuel and the operating history. The physical characteristics of the fuel and the operating history are described in Section 3. Since IPEC has changed the fuel design and operating history , the dep l etion analysis is performed separately for groupings of assemblies that have similar characteristics.
For every fuel cycle , the core is loaded with fresh fue l ca ll ed a "batc h" of assemb l ies (each assemb l y having an ID that begins with the same l etter). A batch grouping consists of more than o n e batch that has s i milar characteristics.
These groupi n gs are defined in this sub-section.
In the following sub-sections, limiting depletion parameters are determined for each batch grouping. The depletion parameters are the moderator and fuel temperatures (Section 5.1 ), the burnable absorbers present d u r i ng depletion (Section 5.2), the soluble boro n (Section 5.3), the specific power (Section 5.4), and presence of contro l rods (Section 5.5). The moderator and fuel temperatures are different at different core heights and also depend on the assembly peaking factor so sets of atom densities are generated as a function of the axial node and the assemb l y peaking factor. Since the bumup and temperatures are d i fferent at different ax i a l nodes, the atom densities are different for each axia l node. Section 5.6 descr i bes t h e computer model for the dep l etion ana l ys i s. The dep l etion analys i s i s performed for each batch grouping using the bounding depletion parameters determined in Sections 5.1 through 5.5. For some permanently discharged assemblies , the bounding parameters for the batch are , overly conservative.
For such assemb l ies, the actual depletion parameters can be used. These special cases are described in Section 5.7. Section 5.8 descri b es adjustments to certain atom densities to account for reduced power operation and fission gas release. Section 5.9 describes how the atom densities are retrieved an d input into the SFP model , an d Section 5.1 0 summarizes the limiting depletion conditions. NET-2809 1-0003-0 l, Revision 0 28 The I ndian Point fuel management and operating approach changed from annual to longer refueling cycles res u lting in higher feed enrichments and increased use of b u rnable absorbers.
In order to not overly increase the bumup requirements for older fue l , the depletion analysis is subdivided into e l even fuel batch groupings.
T h e nine batch groupings for IP2 fuel are A-D , E-F, G-L , M-P , Q-S , T-V, W , X, and Z*, which i s future fuel and includes Batch 2A and subsequent batches. The 10th b atc h grouping, labeled " IP3 (A-U)", app l ies to IP3 Fuel Batches A through U. The 11 th batch grouping , labeled " IP3 X)" app l ies to IP3 fue l Batches V, W , and X. For newer fuel at IP3 (higher enrichment Batches X and above), Batch Grouping Z (future fue l) is used. With a thorough know l edge of the Indian Point Units 2 and 3 operating history provided in Section 3.4, it is possible to select depletion parameters that bound all fuel within a particular b atc h grouping.
 
===5.1 Limiting===
Depletion Parameters
-Temperatures Higher fuel and moderator temperatures during depletion produce more reactive fuel after depletion. The follow i ng section describe s a simple but conservative method for detennining appropriate fuel and moderator temperatures used in the dep l etion calcu l ations. 5.1.1 Averaged Assembly Radial Peaking Factor Historically , analyses used the core exit temperature for the mo d erator temperature during depletion ca l culations. Howe ve r , this was non-conservative.
As a re s ult of radial po we r peaking , some assemblies are burned at a higher re l ative power than the core average, thus the moderator and fuel temperatures are higher. Ignoring the radia l power peaking would be n on-conservative if groups of high power assemblies are placed together in the SFP. As with most parameters used for depletion calculations, the time ordering of the radial power effect has a very minor i mpact on the fina l reactivity ( confirmed at th e end of Section 5.1.4), therefore b umup averagi n g of the fue l and moderato r temperature is appropriate.
* No fue l batche s labe l ed Z were used in Indian Point so Z i s a convenient label for all future fuel. NET-2809 1-0003-01, Revision 0 29 Furthermore , since critica li ty requires a vo lu me of lo w enriched fuel th at is much greater than the vo lum e of a fuel pin , the peaking fac to r of inter est is the asse mbl y radial p ea kin g factor. Calc ulati o n of the averaged asse mbl y radial peaking factor is tri v i a l since it is s impl y the assem bl y di sc h arge bumup divided b y t h e sum of the core average bumups for th e cycles that the fuel assemb l y was resident in the core. The highest averaged assem bl y radial peaking factor for a ll assemb li es with a bumup of2 1 GWd/T (or more) has a l ways been l ess than 1.4. Figure 5.1 shows the averaged assembly ra di a l peaking factor as a function of dis c har ge bumup for the projected fue l in ve nto ry in the IP2 fuel after refueling for Cyc l e 24. Althoug h Figure 5.1 s how s a maximum peaking factor of l ess than 1.30 , once burned fuel has a higher averaged assembly rad i a l peaking factor and can r each close to 1.4. 1.350 1.250 ...
* 0 .... ... u -r IV 1.1 50 --+ .... tlO C ::ii: : 1.050 '
* Q. > * :E E 0.950 , QI VI V, <! -c 0.8 50 QI
* tlO IV ... 0.7 50 :> * <!
* 0.650 -* 0.550 -----~-----~-----, 5000 15000 25000 35000 45000 55000 65000 Burnup (MWd/MTU)
Figure 5.1: Averaged Assembly Peaking Factors of Assemblies in the IP2 SFP NET-28091-0003-01, Re v ision 0 30 Westinghouse Non-Proprietary Class 3 5.1.3 Moderator Temperature The moderator temperature increases as the water rises through the core. For the top node, using the core exit temperature at the averaged assembly radial peakin g factor is conservative (the actual node temp era ture is somewhat l ess since the temperature become s the core exit temperature only at the top of the top node). The average core exit temperature (T co r eex it) is T coreex it= T;n + 2 x (T ave -T;n) = 2 x Tav e -T;n where Ta ve and T;n are the average and inlet temperatures for each cycle provided in Tables 3.5 and 3.6. The enthalpy at T;n and T coreex it can be obtained from steam tables at a pressure of 2235 psia (the minimum allowed pressure).
For an averaged assembly radial peakin g factor of PF , the delta entha lp y is: 6ha sse mbl y = PF X (h co r e ex it -h;n) The assemb l y exit enthalpy is then h 0 , + ti.ha sse mbl y. The temperature corresponding to this assemb l y exit enthalpy are obtained from steam tables and converted to Kelvin. The exit temperatures for various peaking factors using the T ave and T;n data from Tab l es 3.5 and 3.6 are shown in Table 5.1 , below. Table 5.1: Moderator Exit Temperature, T exit, versus Peaking Factor for Batch Groups Texit at 1.4 Tcxit at 1.2 Tcxit at 1.0 Texit at 0.8 Texit at 0.6 Batch Tave (&deg;F) T;u (&deg;F) (K) (K) (K) (K) (K) A , B ,C, D a , c 601.9 596.2 590.3 584.1 577.6 E thru L 589.9 584.1 578.1 571.9 565.6 M,N , P 598.8 592.5 585.9 579.1 572.0 0 , R , S 598.8 592.5 585.9 579.1 572.0 T , U , V 603.2 596.8 590.1 582.9 5 75.5 w,x 603.2 596.8 590.1 582.9 575.5 Z (after X) 603.5 597.7 591.5 585.0 578.3 All IP3 603.5 597.7 591.5 585.0 578.3 -The densities corresponding to these exit temperatures are calcu l ated from steam tables and are presented in Table 5.2: NET-2809 1-000 3-01 , Revision 0 31 Westinghou se Non-Propriet a r y Clas s 3 Table 5.2: Moderator Exit Density versus Peaking Factor for Batch Groups densit y at densit y at densit y at densit y at density at Batch Tav e (&deg;F) Tin (&deg;F) 1.4 fa/cc) 1.2 (e:/cc) 1.0 (e:/cc) 0.8 (e:/cc) 0.6 fa/c c) A, B ,C, D a , c 0.6552 0.67 1 6 0.6874 0.7024 0.7169 E thru L 0.6884 0.7024 0.7158 0.7288 0.7414 M , N,P 0.6645 0.6816 0.6980 0.71 37 0.7287 Q , R , S 0.6645 0.6816 0.698 0 0.7 1 37 0.7287 T , U , V 0.6510 0.6700 0.6879 0.7 050 0.72 14 w,x 0.6 510 0.6700 0.6879 0.7050 0.72 14 Z (after X) 0.6502 0.6676 0.6842 0.700 1 0.7154 All IP 3 -0.6502 0.6676 0.6842 0.700 1 0.7154 At the second node from the top , the mod erato r temp erature i s lo wer bec a u se of th e heat adde d b y th e top node. The second node enthalpy is whe re hin h co r e ex it P F NF h in + NF*PF*(h co re ex it -h in) inl et entha lp y exit ent h a lp y (core average) averaged assemb l y p ea king factor enthalpy nod e factor The e nth alpy nod e factor in the eq u a tion above d epends on h ow mu c h heat i s added by the top n ode a nd this depend s on th e r e l ative axia l po we r of the top node. T h e refor e, a n ax ial po wer sha p e i s need ed. Since this is a d ep letion calcul atio n , the ax i a l power s h ape over the li fe of the assem bl y i s the axia l bumu p profile. The DOE axi a l bumup profil es a re u sed for the axia l bu mup shape [27]. These profile s a re a function of the burnup and are di sc u ssed in Section 6. To illustrat e an exa mple calculation, the p rofile from th e 46+ GWd/T burnup bin i s u sed. T h e relative burnups at the top five nod es a r e 1.0 72 , 1.050 , 0.992 , 0.833, and 0.515 where 0.515 i s the top node , 0.833 is the seco nd node from th e top , etc. The e nth a lp y node fac tor for the top of the se cond nod e i s (1 8 -0.5 1 5)/1 8 = 0.971 si nce t h e r e are 1 8 node s and th e h eat a dded b y the top node i s 0.5 15/18 tim es the tota l h eat added. Similarly , the node factor for the bottom of the secon d node is (18 -0.833 -0.515)/1 8 = 0.925. The average node factor for the seco nd node i s the average of these two values or 0.94 8. NET-28 091-000 3-01 , R ev i sion 0 32 The average entha lp y node factor for the third node is (36 -0.992 -2x0.833 -2x0.5 l 5) / 36 = 0.898. The exit ent h a lpi e s ca n b e con v erted to temperature and density u si n g steam tables. Table 5.3 summarizes the node factors for the top five nodes as a function of the axial bumup shape (represe nt ed in the ta bl e as the bumup bin). Table 5.3: Entha lp y Node Factor versus Axial Burnup Shape Node 18-22 22-26 26-30 30-38 38-46 46+ Too 1.000 1.000 1.000 1.000 1.00 0 1.000 2nd 0.96 1 0.956 0.955 0.955 0.949 0.948 3r d 0.9 1 6 0.913 0.911 0.907 0.898 0.898 4111 0.86 1 0.860 0.858 0.850 0.841 0.841 5111 0.801 0.800 0.798 0.7 91 0.781 0.78 2 The exit enthalpy is co n servative l y used for the top node ent h a lp y so the ent h a lp y node factor is a l ways 1.00 and there is no bumup shape dependence. In genera l , the node factor increases with lo wer re l ative powers at the top b ecause more of the heat is b e in g adde d at the lower nodes. From t he a bo ve n ode factors and using a peaking factor of 1.4 0 , the moderator temperature (K) for B atch Gro upin g Z (future fuel) as a function of the bumup shape is shown in Table 5.4. Table 5.4: Moderator Temperature (K) at each Node versus Burnup Profile Node 18-22 22-26 26-30 30-38 38-46 46+ Too 603.5 603.5 603.5 603.5 603.5 603.5 2n d 60 1.9 601.7 60 1.7 601.7 601.4 601.4 3rd 600.1 600.0 599.9 599.7 599.3 599.3 4th 597.8 597.8 59 7.7 597.4 597.0 597.0 5th 595.3 595.2 595.1 594.8 594.4 594.5 NET-2809 1-000 3-01 , Re v ision 0 33 5.1.4 Fuel Temperature As with the moderator temperature , the fuel temperature in the top nodes is a function o f the axial profile at the top. The fuel temperature in the top node i s the temperature corresponding to a tota l peaking factor of PF x AF where PF is the averaged assembly radial peaking factor and AF is the axial peaking factor (for example , AF would be 0.515 using the 46+ bumup bin). Fuel temperatures were calcu l ated by INTERPIN-3
[28] (provided in Ref. [29]) for various total peaking factors and are shown on F i gure 5.2. The fue l temperature starts at a high va lu e and then decreases for a whi l e before starting to increase again. Since SCALE allow s the input of multiple temperatures for a material during depletion as a function ofbumup , six data s ets are input to SCALE (at 0 , 2 , 15 , 25, 40 , and 70 GWd!T). For a radia l PF of 1.4 , the relat i ve power to use for the top node is 1.4 x 0.515 = 0. 721. From Figure 5.2 , the top node (relati v e power of0.721) fuel temperatures a t 0, 2 , 15 , 25 , 40 , and 70 GWd!T are 837.5 , 812.9, 7 8 9.5, 781.6 , 786.8, and 8 1 5.1 K , respectively. At a PF of 1.2, the top node fue l temperatures are 799.2, 778.1 , 758.1 , 751.2, 752.3 , and 774.9 K. At a PF of 1.0, the top node fuel temperatures are 760.9, 743.4, 726.7, 720.8 , 717.8 , and 734.8 Kand at a PF of 0.8 , the top node fuel temperatures are 722.6 , 708.6, 695.2 , 690.4 , 683.3, and 694.7 K. NET-2809 1-0003-01 , Revision 0 34 1300 1200 1100 E ::, 1000 a. E i!!-.; ::, u.. 900 800 700 --INTERPIN-3 Average Fuel Temperature versus Burnup at Various Relative Powers Typical IP-2 Cycle 20 --___. ' .---, r-6-, 1,---"' -* ____. -I t&.,_~ th ,i.-t~ -p-8-j 3--S--~ --a :: , ia--i I}-I It-i:r-~. ; 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 Burnup (GWD/MTU) 05 -a-1 0 Figure 5.2: Fuel Temperature Change with Burn up and Relative Power The fuel temperature at the top node as a function of averaged assembly radia l peaking factor is linear and can be expres s ed as: Top Node Fuel Temperature at O GWd/T = 19l.5*PF + 569.4 Top Node Fuel Temperature at 2 GWd/T = l 73.8*PF + 569.5 Top Node Fuel Temperature at 15 GWd/T = 157.2*PF + 569.5 Top Node Fuel Temperature at 25 GWd/T = 152.0*PF + 56 8.8 Top Node Fuel Temperature at 40 GWd/T = l 72.5*PF + 545.3 Top Node Fuel Temperature at 70 GWd/T = 200.7*PF + 534.2 Figure 5.3 demonstrates that the top node fuel temperature at 25 GWd/T is indeed a linea r function of the averaged assembly peaking factor (PF). NET-2809 1-000 3-01, Revision 0 35 8 00 ~--------------------
-78 0 +------------------
-=----z 76 0 +--------------=c:::,'"""-----
'-' -Lin ear 7 40 +-----;: F:-:-: i t---------:;;,tl'IE:--------72 0 +-----------:
..li
: a. [ 7 00 +-------~""'-------------68 0 E-, -66 0 +--:::a a=-
<1,1 64 0 +---------------------
62 0 +---------------------
6 00 +-----,-----,------..---...........,r------, 0.5 0.7 0.9 1.1 1.3 1.5 Top N ode P e aking Factor Figure 5.3: Fuel Temperature (K) at 25 GWd/T v s. Peaking Factor (PF) at the Top Node As w ith th e mo d e ra to r t e mp era tur e , th e fuel te mp e ratur e in th e 2 11 ct a n d l owe r nod es i s a fun c tion of th e axia l burnup profi l e. T a bl e 5.5 p rov id es th e fu e l t e mp erat ur es a t 25 GWd/T at a ra di a l P F of 1.40 u s in g t h e ran ge of D OE a x i a l bu rn up s h a p es (r e pr ese n t e d as t h e burnup b in). Table 5.5: Fuel Temperature (K) at each Node versus Burnup Profile Node 18-22 22-26 26-30 30-38 38-46 46+ Too 723.0 753.9 7 5 7.3 753.5 78 0.4 7 81.6 2" d 845.3 8 5 3.5 8 5 8.5 8 7 8.2 9 1 2.0 922.9 3rd 972.5 9 41.8 9 4 7.2 992.9 1011. 1002. 4th 1 0 30. 10 26. 10 29. 1 034. 1040. 103 1. 5th 1 058. 10 60. 10 63. 1 046. 1 04 7. 1042. No t e th a t fo r fue l t e mpe ratures , th e t e mp erat ur e d ec r e a ses w i t h lo wer re l at i ve p owe r s at th e t o p w h ich i s t h e o ppo s it e of w h a t happ e n s wit h th e mod era t o r t e mp erature. N ET-28 091-000 3-01 , R ev i s i o n 0 36 
 
====5.1.5 Selec====
t ion of Bounding Model and Temperatures Historically , many criticality analyses used the moderator exit temperature and a single bounding high fuel temperature for all axial nodes. Early analyses non-conservatively ignored radial peaking and used the core exit temperature for the moderator temperature.
This non-conservatism is removed by using the averaged assembly radial peaking factor. Using a single set of depletion parameters for all axial nodes is a conservatism that is not needed. The top node is unique since:
* IFBA absorbers are cut short and do not extend into the top node;
* Control rods , if at the bite position , are present in the top node;
* Enrichment is lower for axially blanketed fuel;
* Fuel temperature is lower due to the sharp decrease in power at the end of the fuel. For these reasons, the top node is depleted with i ts own set of depletion parameters.
All of the nodes below the top node are depleted with a separate set o f temperatures which is conservative for a ll of the lower nodes. It is possib l e to use the temperatures in Tables 5.4 and 5.5 and perfonn the depletion analysis using separate depletion s for each axial node and burnup bin , but although this approach has been applied in the past [30], reducing the number of depletion parameters greatly simplifies the analysis.
To simp l ify the analysis , it is practica l to select bo un ding temperature conditions that cover all burnup bins. Since the fuel and moderator temperature do not change the same amount from burnup bin to burnup bin, a sensitivity study was performed for the fuel and moderator temperatures. It was determined t hat a change in moderator temperature of 10 K during depletion causes a change in reactivity of 0.0052 t.k while a change in fuel temperature of 100 K causes a change in reactivity of 0.0026 t.k. Using Tables 5.4 and 5.5 it can be seen that for the top node, the 46+ burnup bin conservatively r epresents both the moderator and fuel temperatures (the moderator temperature variation is independent of shape but the fuel temperature is highest for the 46+ burnup bin). For the second node , the 46+ burnup b in should also be used because the moderator temperature is a weak function of the shape but the fuel NET-2809 1-0003-01 , Revision 0 37 temperature is a strong function of the shape. For examp l e, using the 30-38 burnup bin , the moderator temperature at the second node is 0.3 K higher than the 46+ burnup bin but the fuel temperature is 44.7 K lower. The reactivity effect of the 0.3 K higher moderator temperature is only +0.0002 in keff while the reactivity effect of the 44.7 K lower fuel temperature is -0.0012, so using the 46+ burnup bin for the second node is conservative for all burnup bin s. Since the second node fuel temperature is l ess than the fuel temperature at the lower nodes, a test is required to determine if the third or second node is more limiting. For the third node , it would appear that using the 38-46 burnup bin is more limitin g (the fuel temperature is higher using the 38-46 bin compared to the 46+ bin and the moderator temperature is the same). However, using the 38-46 bin for the top node , the fuel temperature is 1.2 K lower , and for the second node, the fuel temperature is 10.9 K lower , while the 3rd node fuel temperature is 9 K higher. The net effect is that u sing the 46+ bin is conservative compared to the 38-46 bin when all nodes are considered. However , to check this , a specia l depletion using the 38-46 shape to obtain fuel and moderator temperatures in the top 5 nodes was perfonned and the result compared to u s ing the 46+ s h a pe for the top and 3'd node s and then using the 3rd nod e temperatures for the 2"d, 3'd, 4t\ and 5th node s (the method that is used for the final selected depletion analysis).
The res ults s ho w that using the 46+ shape to determine fuel and moderator temperatures in thi s manner is more conservative than using the 38-46 s h ape by 0.0006 ~k. This difference was obtained by using the b urnup shape at 38 GWd/T in the keffca l culation.
For additiona l confirmation, the reacti v ity effects using t hree other axial burnup profiles (at bumups of 22, 30, and 45 GWd/T) are 0.0003 ~k, 0.0 005 ~k, and 0.0005 ~k, respectively , with the 46+ temperatures a l ways being more conservative.
For the 4th node , using the 3rd node temperatures means that the moderator temperature being used i s 2.3 K higher than the 4 th node, while the fuel temperature is 29 K l ower than the 4th node. The reactivity effect of the 2.3 K higher moderator temperature is +0.0012 ~k while the reactivity effect of the 29 K lo wer fuel temperature is -0.0008 ~k. So using the 3'd node temperatures is conservative for the 4th node. NET-2809 1-00 03-01, Revision 0 38 The sa m e reasoning applies for th e 5th and l ower nod es. T h e rationale for this is that the moderator temperature is decreasing faster than the fuel temperature is increasing (in terms of net reactivity), so u sing th e J<d nod e temperatures is co ns e r vative for th e 4th and a ll low er nodes. For convenience, the moderator temperature at the top node can be conservatively fit with a straight l ine as illustrated in Figure 5.4 for Batch Grouping Z (future fuel). T h e straight line values shown on Figure 5.4 are always the same or conservative with res pect to the points. 610 605 z 600 +----------------:
.-,,:.=-----.._, -Linear Fit 595 =
* Points 590 +----------:
..,.=
I. c. 585 8 E-580 +---:~"'-----------------
575 570 +----~---~---~---~---~
0.5 0.7 0.9 1.1 1.3 1.5 Peakin g Fac tor Figure 5.4: Top Node Moderator Temp (K) vs. Average Assembly Peaking Factor Figu r e 5.5 shows th e top node moderator density as a functio n of the averaged assembly peaking factor. NET-28091-0003-01, R ev i s ion 0 39 0.73 ---0.72 CJ CJ oli 0.7 1 '-" .f' 0.7 "' C 0.69 ... 0.68 0 .... 0.67 c,: ... -Linear Fit "O 0.66 0
* Point s 0.65 "O 0.64 0 z 0.63 Q. 0 0.5 0.7 0.9 1.1 1.3 1.5 f,-, Peaking Factor Figure 5.5: Top Node Moderator Density vs. Avg Assembly Peaking Factor Tab l e 5.6 s ummari zes th e lin ea r fits for th e moderator temperature and density for each batch grouping (top node) u s ing the equations:
Exit Temp era ture (K) =C l+ C2 x PF Ex it den sity (glee)= C3 + C4 x P F Table 5.6: Fit Coefficients for Top Node Moderator Temperature and Density Batch Cl C2 C3 C4 A,B,C,D 560.8 29.5 0.7 632 -0.0 77 1 E t hru L 548.1 30.0 0.78 1 2 -0.0 663 M , N , P 552.9 33.0 0.7769 -0.0803 O , R , S 552.9 33.0 0.7769 -0.0803 T,U , V 556.6 33.5 0.77 42 -0.0880 w , x 556.6 33.5 0.7742 -0.088 0 Z (after X) 560.5 31.0 0.76 43 -0.0 8 15 A ll IP3 560.5 31.0 0.7643 -0.0 8 1 5 Table 5.7 s ummari zes the lin ea r fit s for the 3rd node from the top mod erato r temperature a nd d e n sity. Table 5.7: Fit Coefficients for 3rd Node Moderator Temperature and Density (Use d for all nodes except the top nod e) Batch Cl C2 C3 C4 A , B , C,D 559.6 27.5 0.7619 -0.06 78 NET-28091-0003-01 , R evision 0 40 E thru L 547.5 27.5 0.7803 -0.0585 M , N , P 552.5 30.0 0.7755 -0.0705 Q , R , S 552.5 30.0 0.7755 -0.0705 T , U,V 555.5 31.0 0.7726 -0.0771 W , X 555.5 31.0 0.7726 -0.0771 Z (after X) 559.7 28.5 0.7628 -0.0715 A ll IP3 559.7 28.5 0.7628 -0.0715 The fuel temperatures at the 3'd from the top node (used for all nodes except the top node) are: 3rd Node Fuel Temperature at O GW d/T = 351.0*PF + 587 .2 3'd Node Fuel Temperature at 2 GWd/T = 319.5*PF + 584.8 3'd Node Fuel Temperature at 15 GWd/T = 291.5*PF + 580.8 3rd Node Fuel Temperature at 25 GWd/T = 33 l.2*PF + 538.2 3'd Node Fuel Temperature at 40 GWd/T = 371.0*PF + 514.4 3'd Node Fuel Temperature at 70 GWd/T = 437.8*PF + 493.0 For modeling simplicity , the 3rd node temperatures are used for the 2nd and l ower nodes because this is conservative.
For axial blankets, the burnup profile at the top will have relative burnups that are smaller than the DOE profile for full-length fuel. As a result of the smaller relative burnups , the moderator temperature increases slightly but the fuel temperature decrease is more significant.
For the same reasons as discussed above regarding using a lo wer burnup bin, using the DOE profile for the last burnup bin is conservative for axial blanket fuel. The use of a burnup averaged assembly radial peaking factor assumes that the impact of the temperatures is independent of the power as a function of burnup. In fuel core loading designs , assemblies are depleted with a peaking factor greater than 1.0 during its first cycle (fresh asse mbl y with burnab l e absorbers).
After the burnab l e absorber is removed , the assembly is moved and the peaking factor during the second cycle is genera ll y less than 1.0. To show that depletion using the average pea king factor over the life of the assemb l y is appropriate , a specia l depletion was performed in which the first 25 GW d/T was depleted at a peaking factor of 1.20 and the second 25 GW d/T was depleted at a peaking factor of0.80. The ke ff for this case at 50 GWd/T (5.0 w/o fuel) is 0.9371. The k e ffusing a peaking factor of 1.00 throughout the depletion is 0.9377. This demonstrates that using the average NET-28091-0003-01, Revision 0 41 peaking factor for the assembly is ap propri ate and s li g htl y conservat i ve. As further confirmat ion , the depletion was repeated u si n g peaking factors of 1.40 an d 0.60. T h e k e ff for this case is 0.9361, which is even s m a ller. 5.2 Limiting Depletion Parameters
-Burnable Absorbers Burnable absorbers h arden the spectrum during depletion , which result in more plutonium production and l ess U-235 cons umption for a give n bumup [31 , 32]. The spect rnm hardening comes from th e absorptio n of thermal n e utron s by the a b sor b er and di sp l acement of th e water in the guide tubes. The effect on reactivity a l so depends on how l ong (in terms of GW dff) the burnable absorber remains in the fuel before b eing removed. Therefore, th e burnab l e a b sorbers that m aximize the 1 0 B l oading and water di s placement should be used. For each batch grouping, the most limiting actual burnable absorbers are used. IP2 and IP 3 have u sed thr ee types of burn ab l e a b sorbers: Pyrex , Wet Ann ul ar Burnable A b sorbers (WABA) and Integral Fue l Burnable Absorbers (IFBA). Table 3.4 provides the dimensions and material details of Pyr ex and W ABA inserts. IFBA rods are di scussed in Section 3.2. The Pyrex and W ABA de signs consist of rodlets m ounted to a base plate which sits on the top of the fuel assembly.
The number ofrod l ets varies by position in the core to h elp contro l power peaking. The most limiting design h as 20 rodlets. Table 5.8 provides the wors t case burnable absorbers for each batch grouping. It s hould be noted that the IFBA a nd th e poison part of the W ABA does not extend to the top of the active fuel. The IFBA starts at least 6 in c h es from the top (8 inches with 8 inch ax i a l blankets) and the poison part of the W ABA starts at l east 5 inches fro m the top (6 in ches with 6 inch axial blankets a nd 8 inches with 8 inch axia l bl ankets). For all b atc h gro upin gs excep t M -P with IFBA/W ABA , th e top node can be depleted wit h no IFBA an d a WABA that h as no boron. For M -P , the 8 inch top node is NET-28 091-000 3-01 , Re v i sion 0 42 depl e t e d w ith IFBA that h as one quart e r 1 0 B (2 inche s of the 8 inches h as IFBA) a nd a W A BA with one h a l f the 1 0 B ( conservative l y mode l s th e W A B A poison i s 4 inches from the top). Table 5.8: Burnable Absorbers versus Batch Grouping BA Max BA Max B A Batch Tvoe Loadin2 Burnuo A , B , C , D P y r ex 20 ro dl e t s 18.5 E thru F P y rex 1 2 ro dl ets 12.2 G thru L Pyrex 20 ro dl ets 16.7 M,N , P IFBA 11 6 (I .OX) 28.1 WABA 20 rodlets Q , R , S IF BA 14 8 (1.5X) 26.7 WABA 20 rod l e t s T , U , V IFBA 14 8 (1.5X) 33.8 WABA 20 ro dl e t s W , X IFBA 14 8 (1.25X) 32.6 WABA 20 rod l ets Z (after X, and IFBA 148 (1.25X) 33.2 IP3 afte r U) WABA 20 ro dlet s IP3 (A-U) Pyrex 20 rod l ets 19.4 5.3 Limiting Depletion Parameters
-Soluble Boron Soluble boron h ar d e n s th e neutron s p ectrum, m a kin g the fue l mor e reactive for a g i ve n burnup. It has been s h ow n that performi n g d ep l e tion calculations at th e burnup averaged so luble boron concentrat i on i s acceptab l e (rather than u s in g a time-dependent so lubl e boron l et down curve) [57]. Tables 3.5 a nd 3.6 s ho w the cycl e average so lubl e boro n concentration for eac h cyc l e. Since nearly eve r y asse mbl y i s burn ed at l east two cy cl es, th e solu bl e boron to u se for the depletion ana l ysis is the multi-cycle burnup average d so lubl e boron for each asse mbl y. This multi-cycl e burnup averaged so lubl e boron i s calcu l ate d u s in g the assembly cycle burnup s to we i g ht the cyc l e average ppm. Table 5.9 s ummari zes t h e so lubl e boron used in the depletion analysis for eac h b atch grouping.
Table 5.9: Soluble Boron versus Batch Grouping Boron Used Batch in Depletion A , B ,C, D 570 E and F 580 NET-2809 1-000 3-01 , R evision 0 43 G thru L 660 M , N,P 820 Q,R , S 850 T,U , V 880 W , X 880 Z (after X and IP3 after U) 950 IP3 (A-U) 560 5.4 Limiting Depletion Parameters
-Specific Power ORNL performed a study of the sensitivity of bumup credit to specific power and determined it is a small effect [33]. For b u rnup credit using all isotopes, a lower specific power is slightly more reactive. However, the reactivity effect of moderator temperature and fuel temperature increases with higher specific power. The reactivity effect of higher temperatures is larger than the reactivity effect of a lower specific power. Since the fuel can operate at only one specific power, the specific power used during depletion is determined to match the relative power used for the other depletion parameters.
This approac h is consistent with DSS-ISG-2010-01 for SFP analysis [5]. The average specific power is the core power divided by the total initial heavy metal mass. Using the stack density (see Section 3.2) of 0.95 multiplied by t he U0 2 theoretical density , the initial mass of Uranium metal is then 89.66 metric tons for the Indian Point cores. The specific power (Watts/g U) is then where SP= Power x PF x RAB/ 89.66 Power= Total core power (MW) (from Tables 3.5 and 3.6) PF= Averaged assembly radial peaking factor RAB = Relative Axial Bumup from the DOE shapes For the top node (RAB= 0.515) the specific power ranges from 15.8 to 18.5 W/g U multiplied by the peaking factor for all of the batch groupings ( due to changes in the total core power). Since the specific power has a small effect , the specific power of 16 W/gU multiplied by the averaged assembly radial peaking factor is used for the top node for all fuel batch groupings. The balance of the axial nodes use temperatures developed using the 3rc1 node relative burnup (RAB= 0.992) from the DOE axial burnup profiles. The specific power for the nodes below the top node ranges from 25.6 to 35.6 multiplied by the NET-28091-0003-01, Revision 0 44 peaking factor. Again, a simple single specific power value of26 W/gU multip l ied by the peaking factor is used for all fuel batch groupings.
 
===5.5 Limiting===
Depletion Parameters
-Control Rod Operation IP2 and IP3 have 193 fuel assemblies in the core. Nine of these assembly locations are under the Control Bank D (less than 5% o f the number of assemblies). Control Bank Dis the only control bank that may be inserted during power operation if the power is greater than 70% of the rated power. IP2 and IP3 have not operated with Control Bank D in the core for any significant bumup except at the "bite" position.
The bite position is set as the location where the worth of the lead control bank is 2 pcm per step. The bite position changes from cycle to cycle and during cyc l e operation but is typica ll y between 207 to 217 steps withdrawn , which corresponds to the rod being inserted 8. 7 or less inches into the core. Operation with Control Bank D at the bite position occurred in IP2 durin g the first 17 c y cles. Cycle s 1 8 and beyond for IP2 and a ll cycles for IP3 operated with all of the control rods fu ll y withdrawn from the core. Table 5.10 s hows the fuel a s sembly ID s that were under control bank D for IP2 for the cycle s containing Batches A through X Tab l e 5.10: Asse mbli es under D-B ank for t h e First 21 Cycles of I P 2 C o re Locat i o n Feed o fD-B6 BlO F2 F14 H S K2 K14 P6 P lO Batc h B a nk Cy cl e Asse mb h I D 1 A17 A09 A30 A39 A44 A34 AlO A26 A33 A , B,C 2 BOl B 23 B64 B57 B06 B54 B60 B18 B27 D 3 C45 C51 C55 C59 B53 C54 C60 C42 C40 E NET-28091-0003-01 , Revision 0 45 4 D47 D69 D09 Dl0 D25 D49 D72 D32 D46 F 5 E57 E32 E53 E06 D71 El3 EOl E27 E25 G 6 F06 Fl4 F46 F03 F45 F23 F60 Fl6 F05 H 7 G49 G50 G68 G67 G70 G65 G69 G52 G61 1 8 Jl8 132 122 107 H27 167 130 133 129 K 9 Kl 8 Kl9 K09 Kl2 165 K29 K27 K20 Kl7 L 10 L61 L47 L54 L60 K25 L41 L55 L34 L57 M 11 M54 M63 M69 M60 L09 M65 M62 M53 M46 N 12 P06 P09 P08 P03 MOS P 11 PIO P05 P04 p 1 3 Q25 Ql7 Q22 Q21 N24 Ql4 Q27 Q26 Ql5 Q 14 R 82 R76 R79 R 78 R 7 1 R 75 R77 R72 R 15 S35 S40 S33 S39 S38 S37 S36 S34 s 16 T67 T69 T68 T63 S77 T7 1 T 70 T65 T62 T 17 U65 U73 U58 U56 U61 U60 U59 U57 u 1 8 V32 V51 V46 V52 V44 V30 V49 V31 V 19 W89 W75 W93 W82 W21 W77 W76 W71 W64 w 20 X24 X25 X21 Xl3 W21 X42 X34 X35 X23 X 21 2A64 2A84 2A58 2A55 X03 2A66 2A38 2A70 2A29 2A The reference depletion analyses for a ll bu t the future cycles (Batch Groupi n g Z) are modeled with no control rods inserted. For the asse mbli es li sted in Table 5.10 that are s h aded blue or ye ll ow, the burnup r equireme nt for storage i s increased b y a n appropriate burnup penalty. The asse mbli es marked w ith green shading on Table 5.10 did not require a burnup penalty s inc e they did not contai n W ABAs a nd the r eactivity effect of WABAs in cluded in the standar d d e pl etion a nal ys i s is l arger than the reactivity effect of the sho rt tim e that Co ntrol Bank D may h ave been in these asse mblie s. Note that af t er Cycle 17 , Control Bank D was maintained in a n all-out position (no bite). T h e asse mbli es in Table 5.10 sha d e d in pink a nd yellow did not contain burnable absorber in serts durin g ac tual core ope ration s but did h ave the Co ntrol B ank D rods in th e bit e position. A n a l ys i s s ho wed th at assem bli es depleted wit h 20 rodlet P yrex burnable a b sorbers conservatively bound s asse mblie s that were operated with Co ntrol B ank D at the bit e po s ition. No burnup p e nalt y is needed for the pink assem blie s since they are mod e l ed wit h 20 rodlet Pyrex burn ab l e absorbers. The yellow asse mblie s are d e pl ete d with a 20 rodlet WAB A. S ince W ABA's do not h arde n th e s pe c trum as much as Pyr ex and since W ABA absorber materi a l i s not in the top nod e, a s m all burnup correction is requir ed. For the se NET-28091-0003-01, Revision 0 46 yellow assemblies , the burnup requirements are increased by 0.5 GWd/T. Tab l e 5.11 shows the cases analyzed to confirm the bumup requirement increase (or shortened to "penalty")
for assembl i es where the D-bank was at the bite position and the assembly did not contain a burnable absorber.
Note that the L'.k is converted to a L'.GWdfT by use of the sensitivity of ketf to burnup that is discussed in Section 7. Table 5.11: Effect of Modeling the Bite Position rather than Burnable Absorbers Batch Group, Fuel Burnable Calculated Calculated kcff A kcrr with D-with burnable Ak Burnup Enrichment and Burnup Absorber bank Bite absorber (GWd/T) G-L, 3.0 w/o, 24.23 GWd/T 20 Pyrex 0.9620 0.9684 -0.0064 -0.90 M-P, 4.2 w/o, 39.28 GWd/T 20WABA 0.9570 0.9554 0.0016 0.30 M-P, 4.6 w/o, 41.31 GWd/T 20WABA 0.9616 0.9608 0.0008 0.20 T -V, 5.0 w/o, 41.94 GWd/T 20 WABA 0.9597 0.9590 0.0007 0.11 The assemblies in the blue shaded portion of Table 5.10 are depleted during the first cycle with a b urnable absorber and then contained a control rod in serted to the bite position.
For these assemb li es the required bumup is increased by 2 GWd/T. The 2 GWdfT penalty is determined by running a separate set of depletions with D-bank at the bite position. The bite depletion is performed as follows: 1. For the top node , the depletion analysis is perfonned with the control rod fully inserted for the entire depletion (this is conservative since this fuel was under D-bank for only one cycle). 2. For the lower nodes , the burnable absorber is in the fuel until the maximum burnable absorber burnup is reached and then the control rod is placed in the guide tube for 2 GWd/T and then removed. Contro l rod bumup of 2 GWd/T is considerably more bumup than what has been experienced at IP2. Calcu lati ons of k e ff using the standard depletions and the bite depletions are performed using the Region 2 three-out-of-four model (see Section 6). Table 5.12 shows the results of this analysis.
The L'.k values shown in the fourth column are converted to delta burnups u sing the bumup measurement NET-2809 1-000 3-01, Revision 0 47 uncertainty calculations given in Section 7. The maximum delta burnup on Ta bl e 5.12 i s ro und ed up to 2 GWd/T which is used for the assem bli es in the blue s h aded ce ll s on Table 5.10. Table 5.12: Burnup Penalty for Assembles with Burnable Absorbers followed b y Bite D-bank Batch Group, Fuel Enrichment, and Burnup. k-dbnk k-standard Ak ABU GWd/T A-D , 3.0 w/o, 22.77 GWd/T 0.9721 0.968 7 0.0034 0.68 G -L, 3.0 w/o, 24.23 GWd/T 0.9735 0.9684 0.0051 1.0 2 M-P, 4.2 w/o, 39.28 GWd/T 0.9653 0.9552 0.0101 1.87 M -P , 4.6 w/o, 41.31 GWd/T 0.9708 0.9608 0.0100 1.83 T -V, 4.2 w/o, 36.45 GWd/T 0.9703 0.9621 0.0082 1.64 T -V, 5.0 w/o, 41.94 GWd/T 0.9677 0.9590 0.0087 1.74 The assembly shaded orange (X 03 *) represents a separate group. The orange group is for assemblies that contai n ed a burnable absorber insert for their first cyc l e and the n were subseq u ently place d und er D-bank in a later cycle but wi th out " bite" operation. These assem bli es may ha ve been burned a s h ort time under D-b ank in cyc l es without bite o p eration but are mode l e d as h aving no D-bank opera tion. To cover some D-bank operat ion , 1 GW d/T is added to the burnup requirement of these assemblies.
The analysis to s upp ort a 1 GWd/T pena lt y modified the s t a nd ard depletion analysis by adding 1 and 2 GWd/T ofburnup wit h co ntrol rods to the top node an d the lower n odes , respective l y, after the burnable a b sorber is r emoved. Ta bl e 5.13 shows the cases run to confirm this penalty. The result s on Table 5.13 are rounde d up to arrive at the 1 GWd/T penalty to th e burnu p requirement
: s. Table 5.13: Assemblies with B A Inserts plus under D-Bank in Non-Bite Cycles Batch Group , Fuel Enrichment, and Burn up k-dbnk k-standard Ak ABU GWd/T W, 4.6 w/o , 40.07, GWd/T 0.9629 0.9598 0.0031 0.51 X, 4.6 w/o, 40.07 GWd/T 0.9627 0.9597 0.0030 0.49 IP3 (A-U), 3.4 w/o, 29.77 GWd/T 0.9698 0.9666 0.0032 0.5 4
* X04 was pl aced unde r a D-Bank location in Cyc l e 22 a nd i s also in c lud ed in this group. N ET-2809 1-000 3-01, R ev i sion 0 48 There are two assemblies that are shaded red on Table 5.10 because they are unique. Assembly R08 was located under D-bank at the bite position for two cycles. This assembly did not contain W ABAs in i t. If R08 had been under D-bank for only one cycle t hen it would also be a " yellow" assembly but it was operated under D-bank for a 2" ct cycle. This assembly has already been casked, and its bumup is 6. 7 GW d/T above the requirements for its assigned category (Category 4 fuel) if it were returned to the SFP. It is concluded that Category 4 is the proper assignment for R08. Assembly U41 is like assembly R08 except that the second cycle of operation for assembly U41 did not have the D-bank at the bite position since bite operation ended with Cycle 17. Thus this assembly would need the yellow bumup penalty of 0.5 GWd/T plus some additional margin to cover some operational use of D-bank in Cycle 18. U41 is categorized as Category 5 fuel and exceeds the Category 5 minimum by over 6 GW d/T so the categorization of U4 l is appropriate. For future cycles (Batch Z) (white cells in Table 5.10) it is not known which assemblies will be under D-bank. To cover power operation with some control rods inserted, the top node for D-bank assemblies is depleted for 1 GWd/T with a control rod and lower nodes are depleted for 2 GWd/T with a control rod. The rest of the depletion is with a 20 finger W ABA w h ich is never removed plus 148 IFBA ( l .5X). This is the same method that was used in the previous CSA (1]. This approach eliminates the need to check fu ture assemblies for rodded operation under D-bank or when the W ABA was pulled. IP3 fuel is covered by three depletion batch groupings:
: 1. Batch Grouping IP3 (A-U) which covers Batches A through U 2. Batch Grouping IP3 (V-X) which covers Batches V , W, and X , and 3. The Z Batch Grouping which covers IP2 and IP3 assemblies beyond X. NET-28091-0003-01 , Revision 0 49 T h e IP3 (A-U) b a t c h gro u p in g i s d e pl e t ed w ith a 2 0 ro dl et P yrex bu rna bl e a b so rb er w hich i s r e m ove d a t 2 0 G Wd/f. No co ntrol rod s a r e in th e b ase case d e pl e ti o n. T a bl e 5.1 4 s h ows th e a sse mbli e s in IP 3 w h ic h we r e und er D-b a nk (asse mbli es in B a t c h es A throu g h V we r e not in t h e IP 3 co r e afte r Cy cl e 11). T h e c ol o r c o din g o n Ta bl e 5.1 4 is t h e sa m e as p r ev i o u s l y d i sc u sse d , so th ere i s no burn up pe n a lt y for th e green s had e d assemb li es , a nd t h e ora n ge s h a d e d asse mbli es h ave a burnu p req ui re m e n t that h as b een in c r ease d b y 1 GWd/T. Table 5.14: Assemblies under D-Bank for the first 11 Cycles ofIP3 Core Feed Location B6 BIO F2 FI4 HS K2 KI4 P6 PIO Batch ofD-Bank Cvcle Assemb l y ID I A24 A21 AOI A23 A30 Al8 Al7 A06 A02 A , B , C 2 B14 BOS B19 B25 A24 B03 B41 B37 B56 p 3 C45 C64 cso C58 A24 C54 C44 C06 C53 R 4 R32 R53 R71 R25 P l 1 R06 R43 R47 Rl7 s 5 S04 S3 5 S2 1 S28 R03 S09 Sl3 S l4 S 11 T 6 T74 T41 T70 T46 TSO T61 T76 T64 T65 u 7 U46 U75 U53 U56 TSO U73 U7 1 USO U74 V 8 V38 V37 V40 V35 T53 V34 V36 V33 V39 w 9 W39 W30 WI2 W06 V42 W I S W03 W20 W37 X 1 0 YI7 Y39 Y28 Y54 V44 Y38 Y40 Y41 Y23 y 11 AA28 AA22 AAS! AA23 U02 AA30 AA24 AA31 AA19 AA N ET-2 8 091-0003-01 , R ev ision 0 5 0 Westinghou s e Non-Proprietary Class 3 5.6 Depletion Analysis Model The dep l etion ana l ysis uses the SCALE 6.1.2 , t5-depl sequence of TRITON with the ENDF/B-VII.O 238 group cross section l ibrary. The model is a simple 2D 15xl5 array of pins centered in an 8.466 x 8.466 inch (assembly pitch in core) square of water (11). The 15x15 array contains 20 guide tube s and a voided instrumentation tube. In the core l ocations for the incore flux monitoring system , there is a gas-filled tube in which the detectors travel. Rather than model this tube in detail and/or separate o u t the assemb l ies that were in instrument l ocations, the instrument tube of all assemblies are modeled with void inside the tube inner diameter.
The 20 guide tubes contain the burnable absorber inserts or control rods as needed for the batch grouping.
Some of the fuel rods have ZrB 2 coatings (IFBA rods). This is modeled as a ring next to the pellet outer diameter of 0.001 cm thick of ZrB 2 meeting the 1 0 B linear density ([ ]a ,c mg 1 0 B/i nch except Batch Grouping M-P , which uses [ ]a ,c mg 1 0 B/inch) (20). All of the fuel rods are a single material.
Therefore, there is no variation in atom densities across the assembly with depleted fuel. Further , the resonance se l f-shielding (performed through lattice cell cards in SCALE) is the same for all pins including IFBA pins. A study performed to answer a question from the NRC on the previous CSA showed that ignoring the IFBA in the lattice cell calcu l ation was not significant (2). The depletion problems use 4000 neutrons per generation and 1000 generations.
This n umber of neutrons per generation and number of generations was shown to be adequate by a c onvergence study detailed in Chapter 4 of reference (2 1]. Small depletion time steps are needed to accurate l y account for the spectral changes due to Xe and Sm and the initial build in of Pu and other fission products. The initia l time steps (MWd/T) are 150 , 350 , 500 , 500, and 500 , followed by steps of 2000 MWd/T until the maximum bumup is achieved. For the Z fuel batch, the depletion contains control rods for the first 2 GWd/T (1 GWd/T in the top node) and then the problem is restarted u sing the StdCmpMixOOOxx fi l e where xx i s the material n u mber. The fuel temperatures use the "timetable" block to input the bumup dependent temperatures shown below Figure 5.2 and Table 5.7. Due to the restart , the bumups in the t i metable block are adjusted for the bumup NET-28091-0003-01, Revision 0 51 before the restart. For the other fuel groupings, a restart i s also used , but to remove the burnable absorber not the control rod. The restart burnup depends on the maximum burnup for any assembly with a burnable absorber insert. This burnup is conservatively rounded up to the next 1 GWd/T. The depletion block uses the default for the fuel mixture but constant flux option for the burnable absorber materials. Using the above model inputs , U02 fuel is depleted at fuel enrichments of2.0, 2.2, 2.6, 3.0, 3.4, 3.8, 4.2, 4.6, and 5.0 w/o U-235 at peaking factors of 0.6 0 , 0.80 , 1.00, 1.20, and 1.40. Enrichments for axial bl ankets are 2.6, 3.2, 3.4, 3.6, and 4.0 w/o with the same range of peaking factors. The burnup points at which the isotopic data is collected are 0.15 , 0.50, 1.0 , 1.5 , 2.0, and then every 2.0 GW d/T after that. Although the depletion is carried out with a full set of 388 nuclides, the nuclides used in the SFP model are a reduced set (185 nuclides found on Table 2.1). In order to confirm the TRITON modeling is adequate, comparisons were made with CASM0-5 benchmarks.
The change in k err as a function of burnup derived from using CASM0-5 and SCALE/TRITON depletion is provided for 198 cases. The CASMO t.k values are published in Reference
[20] for specific benchmark conditions
.* T h ese include cases with W ABA and IFBA. Reference
[20] is currently under review by the NRC, but the only values used here are the pure CASMO result s (not the bias and uncertainty that is under review). Tables 5.15, 5.16, and 5.17 provide the difference in the t.k as a function of depletion between CASM0-5 and SCALE/TRITON. Notice that a negative value indicates that SCALE/TRITON conservatively under predicts the reactivity of depletion predicted by CASM0-5. Further, note that the maximum deviation is less than 0.0030. Tables 5.18 , 5.19, and 5.20, show the percent difference in the
* The CASM0-5 ilk of depletion is not given directly in Reference
[20]. Tables C-3 to C-5 of Reference
[20] provide the CASM0-5 ilk of depletion plus the CASMO 5 bia s given in Table 10-1 of Reference
[20]. For thi s application, the CASM0-5 bias is subtracted from Tables C-3 to C-5 to yield the CASM0-5 ilk of depletion.
NET-28091-0003-01, Revision 0 52 i:1k of de pl etion. Th e max i mum perce n t d iffer e nce i s 1.37%. At t h ese sm a ll d i ffere n ce s, it i s un clear w h ich i:1k o f d ep l etio n is c o rrect so th e u t il izatio n of th e 5% unce rt a in ty a l lowe d b y DSS-ISG-2 010-0 1 i s appropria t e. In add i tion , R efe ren c e [21] ha s s hown ex ce ll e nt a gre e m e nt between T RI T ON/NE WT , whic h w as u se d for t h e anal ys i s o f c h emi ca l a ss ay s and th e TRITON/KENO appro ac h u s ed i n thi s anal ys i s. Table 5.15: SCA LE/TRI T O N minus CA SM0-5 ~k of Depl e tion at 100 Hours Coolin g Ca se de s cription Ca se 10 2 0 30 40 50 60 3.2 5% e nrichm e nt d e pl e ti o n 1 -0.000 7 -0.0 0 1 2 -0.0015 -0.0025 -0.00 22 -0.00 20 5.00% e nr ic h ment d e p l etio n 2 0.0 001 0.000 2 0.0003 0.0000 0.0004 0.0006 4.2 5% enr i c h m e nt d e p l e tion 3 0.0004 0.0000 -0.000 2 -0.0005 -0.0008 -0.0005 off-no min a l pin d e pl e tion 4 -0.000 6 -0.001 2 -0.0015 -0.001 6 -0.00 1 9 -0.00 22 20 W ABA d ep l e ti o n 5 0.0002 0.000 7 0.0 003 -0.0002 -0.000 I -0.0002 I 04 IF B A dep l et i o n 6 0.001 2 0.0011 0.0004 -0.000 7 -0.0009 -0.001 8 1 04 IFBA , 20 WABA dep l et i on 7 0.0009 0.001 6 0.0 0 08 0.0000 -0.0001 -0.000 8 hi g h boron d e p l e t io n= 1500 ppm 8 0.0 004 -0.0001 -0.000 3 -0.0004 -0.0 001 -0.0001 branch t o h ot ra c k= 3 38.7 K 9 -0.0002 -0.0003 0.0000 -0.0005 -0.000 I -0.0001 b ra n c h t o ra c k bor o n = 1500 ppm 1 0 -0.0008 -0.00 15 -0.001 9 -0.0 0 2 3 -0.0 02 6 -0.00 2 5 hi g h power d e n s it y d e pl e tion 11 0.0 000 -0.000 7 -0.000 7 -0.0009 -0.0 008 -0.000 8 Table 5.16: S CALE/TRITON minu s CA S M0-5 ~k of Depletion at 5 Y ears Coolin g C ase description Ca se 10 20 30 40 50 60 3.2 5% enr i c hm ent d e pl e t i o n I 0.0000 -0.0005 -0.0009 -0.00 1 4 -0.0008 -0.0004 5.00% e nric h m e nt d e pl e t i o n 2 0.0 00 8 0.0 0 11 0.000 8 0.000 8 0.00 10 0.001 3 4.2 5% e nri c hm e nt d e p l e t i on 3 0.0010 0.0005 0.0003 0.0001 0.0002 0.0003 o ff-nominal pin d e pl e tion 4 -0.0 0 0 1 -0.00 03 -0.00 I 0 -0.0011 -0.0 0 I 0 -0.000 9 20 W ABA dep l etion 5 0.000 8 0.0011 0.0009 0.000 8 0.0008 0.0010 104 IFBA d e pl e tion 6 0.00 2 0 0.0014 0.0009 0.000 1 0.000 1 -0.0005 I 04 IFBA , 20 W ABA dep l e t ion 7 0.0020 0.00 2 4 0.0015 0.00 1 2 0.0006 0.0004 h ig h boro n d e p letion = 1 500 p pm 8 0.000 7 0.0008 0.000 7 0.0003 0.0007 0.0010 branch to h ot rack= 33 8.7 K 9 0.0004 0.000 6 0.0003 0.0004 0.0009 0.0010 b ranc h t o rack b oron= 1 500 p pm 10 0.0002 -0.000 7 -0.00 1 0 -0.00 1 7 -0.0015 -0.00 1 4 h i g h po w er d e n s it y d e p l e tion 11 0.00 0 6 0.000 4 0.0003 0.0004 0.0 0 0 2 0.000 8 NET-28 0 9 1-00 03-01 , R evis i on 0 53 Table 5.17: SCALE/TRITON minus CASM0-5 8.k of Depletion at 15 Years Cooling Case description Case 10 20 30 40 50 3.25% e nri c hm ent dep l et i o n 1 0.0007 -0.0004 -0.00 1 4 -0.00 1 6 -0.00 1 5 5.00% e nrichm ent depletion 2 0.00 11 0.0014 0.0007 0.0008 0.0008 4.2 5% e nri chment dep l et i on 3 0.00 14 0.0010 0.0002 0.0000 0.0000 off-no min a l pin depletion 4 0.0006 -0.000 4 -0.0006 -0.00 14 -0.00 1 3 20 W ABA d e pl etion 5 0.00 14 0.0018 0.0010 0.0005 0.0006 104 IFBA depletion 6 0.0025 0.00 19 0.0010 0.0004 -0.0002 104 IFBA, 2 0 W ABA d e pl et i o n 7 0.0027 0.0028 0.001 7 0.0010 0.0008 h ig h b oro n d e pl etion= 1500 pp m 8 0.00 11 0.0009 0.0006 0.000 1 0.0006 br anc h t o hot rack= 338.7K 9 0.0005 0.0004 0.0001 0.0003 0.0006 branch to rack boron = 1 500 ppm 10 0.0004 -0.0007 -0.0013 -0.00 1 6 -0.00 1 9 hi gh power density depletion 11 0.00 11 0.0008 0.0001 0.0000 0.00 04 Table 5.18: Percent Difference in the 8.k of Depletion at 100 Hours Cooling (SCALE/TRITON minus CASM0-5 Llk of D epletio n over the Llk of Depletion)
Case description Case 10 20 30 40 50 3.25% e nrichm ent dep l et i o n 1 -0.50 -0.52 -0.48 -0.62 -0.48 5.00% e nri c hm e n t depletion 2 0.06 0.08 0.10 0.00 0.10 4.25% enric hm ent depletion 3 0.3 1 0.02 -0.06 -0.14 -0.18 off-nominal pin depletion 4 -0.49 -0.54 -0.48 -0.42 -0.40 20 W ABA depleti o n 5 0.09 0.31 0.09 -0.04 -0.02 104 IFBA d e pl et i on 6 0.71 0.51 0.1 2 -0.19 -0.20 104 IFBA , 20 WABA d e pl et i o n 7 0.37 0.65 0.29 -0.01 -0.01 hi gh boron d ep l et i o n = 1500 ppm 8 0.36 -0.06 -0.09 -0.1 2 -0.02 b ranc h to hot rack= 338.7K 9 -0.16 -0.13 0.00 -0.13 -0.02 branch to rack boron = 1 500 pp m 10 -0.79 -0.83 -0.76 -0.72 -0.68 hi gh power d e n s it y dep l e ti o n 11 -0.02 -0.34 -0.25 -0.25 -0.19 Table 5.19: Percent Difference in the 8.k of Depletion at 5 Years Cooling (SCALE/T RITON minus CASM0-5 Ll k of D epletio n over the Llk of D e ple tion) Case description Case 10 20 30 40 50 3.25% e nrichm ent depletion 1 -0.01 -0.22 -0.26 -0.32 -0.17 5.00% e nrichm ent depletion 2 0.68 0.51 0.28 0.20 0.21 4.25% e nrichm e n t d ep l et i o n 3 0.79 0.20 0.10 0.01 0.04 off-no minal pin d eplet ion 4 -0.09 -0.15 -0.30 -0.25 -0.20 20 W ABA d e pl etion 5 0.38 0.44 0.29 0.1 9 0.16 104 IFBA d e pl etion 6 1.16 0.61 0.30 0.03 0.01 10 4 IFBA, 20 WABA d ep l e tion 7 0.78 0.95 0.48 0.30 0.12 hi gh boron d ep l e ti on= 1500 ppm 8 0.5 6 0.34 0.23 0.0 6 0.14 b ra n c h to hot rack= 338.7K 9 0.31 0.25 0.1 0 0.09 0.18 branch to rack b oron = 1500 ppm 10 0.19 -0.4 0 -0.37 -0.48 -0.37 hi g h p owe r d e n s it y depletion 11 0.47 0.16 0.10 0.09 0.04 N ET-28091-0003-01 , Revisi on 0 60 -0.0009 0.0012 0.0003 -0.0010 0.0005 -0.0004 0.0004 0.0006 0.0008 -0.00 14 0.0005 60 -0.41 0.12 -0.09 -0.41 -0.03 -0.36 -0.17 -0.01 -0.02 -0.58 -0.16 60 -0.08 0.25 0.05 -0.16 0.18 -0.10 0.0 8 0.19 0.18 -0.3 1 0.1 5 54 Table 5.20: Percent Difference in the Llk of Depletion at 15 Years Cooling (SCALE(f RITO N minus CASM0-5.c.k of Depletion o ve r the t.k of Depletion)
Case description Case 10 20 30 40 50 3.25% e nrichm e n t depl e tion 1 0.49 -0.17 -0.37 -0.33 -0.28 5.00% enrichment depletion 2 0.92 0.62 0.2 3 0.1 8 0.16 4.2 5% enr i c hm e nt deplet i on 3 1.09 0.40 0.06 -0.0 1 -0.01 off-no minal pin d ep l et ion 4 0.47 -0.18 -0.17 -0.30 -0.24 20 W ABA depl etion 5 0.66 0.69 0.30 0.10 0.1 1 104 IFBA depl etion 6 1.37 0.77 0.28 0.09 -0.03 104 IFBA , 20 W ABA d e pl e ti on 7 1.03 1.07 0.52 0.24 0.1 7 high boron d e pl e ti o n = 1 500 ppm 8 0.86 0.37 0.18 0.01 0.11 branch to h ot rack= 338.7K 9 0.38 0.15 0.03 0.06 0.11 branch t o rack b oron = 1500 ppm 10 0.38 -0.38 -0.45 -0.42 -0.43 hi g h power d e n s it y depletion 11 0.84 0.32 0.03 -0.01 0.07 5. 7 Special Case Depletions 60 -0.15 0.21 0.05 -0.1 6 0.08 -0.07 0.06 0.10 0.1 3 -0.28 0.08 Due to th e l imited s pace in Region 1 , assem bli es th a t ha ve been disch a rg e d and do not meet th e requirements for R egi on 2 were re v iewed to remove conservatisms in the depletion anal ys is. The lar ges t conservatism i s genera ll y th e depletion condition that all assemb li es co nt ain th e m aximum burnable absorbers of the batch gro upin g. For se l ected assem b l i es, a n a l ys is i s performed usin g th e fu ll avai l ab l e in fonnatio n on the assemb l y. T hi s sec tion di sc u sses the change in d ep l et ion co ndition s. Sectio n 8 pro vi d es the result s of th e a n a l ysis for special asse mblie s. Tab l e 5.21 pro vides the d e pletion p ara met ers t hat are u se d for s p ecia l d e pl e tion a n a lysi s. Table 5.21: Special Case Depletion Parameters Limiting Enrichment Fraction of PPM Peaking Assembly Fuel Type Theoretical Burnable Absorber (soluble ID (w/o) Density boron) Factor AI O HIPAR 2.2 1 0.943 None 570 0.92 F44 HIPAR 3.35 0.933 None 540 1.05 L48 LOPAR 3.69 0.944 1 6WABA 660 0.69 W52 OFA 4.96 0.946 20 W AB A/10 0 IFBA 880 0.84 XIS OFA 4.9 5 0.950 20 W AB A/11 6 IFBA 880 0.8 7 U12 (IP3) LOPAR 3.2 1 0.950 1 2 WABA 560 0.90 V43 (IP3) vs 3.80 0.950 20 W AB A/6 0 IFB A 6 50 1.1 2 NET-28091-0003-01, R ev i s ion 0 55 AlO represents 4 other assemblies (A09, A26, A33, and A34) which have the same characteristics but slightly higher burnups. These assemblies were burned for one cycle under the D-bank (therefore, had no burnable absorber).
The top node is depleted with a control rod for the entire depletion.
The lower nodes were depleted with a control rod for 2 GWd/T. Since the assembly was under D-bank in the bite position , t he DOE axial burnup profile between 14 and 18 GWd/T i s u se d. F44 was in Cycles 4 and 5 and did not contain burnable absorber inserts. With burnup during only two cycles, F44 did not meet the Category 4 fuel requirement by a small amount. When analyzed with its actual burnable absorber (none), it easily made the requirements for Category 4. Of the eight symmetric s isters to F44, six were in the core for thr ee cycles. The remaining sister, F52 has been casked but if it is returned to the SFP, it too meets the Category 4 requirement using this special case analysis. L48 and its sisters (L37, L38, L39, L44 , L51, L52, and L64) spent two cycles on the outside corner of the core. Because of this placement in the core, the burnup after three cycles was too low for Category 4 fuel by about 0. 7 GW d/T if the standard depletion condition for burnable absorbers is used (20 rodlet P yrex). Since this group of assemblies actually had a 16 rodlet W ABA in sert , the analysis of this set of assemblies using the W ABA instead of Pyrex , showed these assemblies meet the Category 4 reactivity requirements.
W52 and X18 are the lowest bumup assemblies of two sets of eight symmetric sister assemblies (W47 , WSO , W52, W53, W54 , W55, W59 , W60 and X09, Xl 1 , X12, X14, X16, X18, X44, X45). Whichever category W52 and X18 qualifies for, then the other seven will also qualify. W52 misses the Category 4 fuel burnup requirement by about 1 GW d/T. X 18 misses the Category 4 requirements by only 0.2 GWd/T. The only benefit in the depletion analysis for these two sets comes from reducing the IFBA rods from 148 to 100 and 116 for the Wand X set respectively. A special depletion is performed for these two sets with the reduced IFBA. In addition to the improved depletion, the actual axial burnup NET-28091-0003-01, Revision 0 56 profile for these assemblies is used. With these adjustments, these two groups of eight assemblies meet the Category 4 requirements.
Finally, a set of fuel assemblies from IP3 did not meet the Category 4 fuel requirements using the standard depletion analysis. The assemblies Ul2 , U2 l, U3 l , and U4 l from IP3 actually contained 12 WABA rather than the 20 Pyrex used in the standard depletion for Batch Group IP3 (A-U). Modeling t hese assemblies with the correct burnable absorber inserts allow them to make the Category 4 fuel r equirements. V43 and V48 did not initially meet the Category 3 fuel requirement but with special depletions , these two assemblies qualify for Category 3. 5.8 Reduced Power Operation at End of Life and Fission Gase s DeHart [34] demonstrated that operating history has a small effect on spent fuel reactivity.
However, at the end-of-cycle (EOC), the reactor power may be reduced (for example , a planned coast down) and this can cause a small reactivity change. One of the key absorbing fission products, Sm-149, reaches an equilibrium concentration during power operation tha t is independent of power. However , its precursor, Pm-149 , is directly proportional to power. At a reduced power , there is less Pm-149. Pm-149 decays into Sm-149 with a 2.2 day half-life.
Thus, if a reactor reduces power at end of cycle, there would be less Sm-149 in the cooled fue l , which is a positive reactivity effect. Therefore, ignoring low power operation during the last month is non-conservative. To account for this effect, the amount of Pm-149 can be reduced to one half of the full power content for all criticality calculations (which results in a penalty of about 100 pcm). This covers coast downs to 50% power and covers all past operating experience and anticipated future operation at the Indian Point plants. Furthermore , a significant fraction of once or twice burned assemblies are placed on the core periphery in the last cycle of the assembly's depletion. So a high peaking factor does not reflect a possible very low peaking factor (0.5) at the end of depletion.
As discussed above, it is at the end of depletion that the amount of Pm-149 is important.
To account for both the "end of depletion" effect (0.5) NET-28091-0003-01, Revision 0 57 and the coast down effect (0.5), t h e amoun t of Pm-14 9 i s re du ce d t o on l y 0.5 x 0.5 = 0.25 o f t h e full p ower i s otop ic conte n t in a ll criticality calculations.
This results in a reactivity penalty of about 250 pcm in the final k err calculations.
To conservati v ely account for fission gases escaping the fuel and migrating to the plenum , a ll krypton and xenon isotopes are reduced by 32%, all rubidium isotopes are reduced by 44%, and all iodine and bro mine isotopes are reduced by 10%. These adjustment factors are justified in the response to an RAI documented in Reference
[2]. 5.9 Production of Atom Density Sets SCALE TRITON outputs atom densities in two ways , through an output file , StdCmpMixOOOXX where XX is the material number , or OPUS pit files. The StdCmpMixOOOXX supplie s the atom densitie s for the end of the run wherea s the OPUS pit files provides atom densities for selected isotopes as a function ofb u rnup and coo l ing time. Rather than reru n SCALE for each desired burnup and coo l ing time, OPUS pit files are saved. For burnups between the SCALE time step s , the atom den s ities are l i nearly interpolated between the time steps. The SCALE time decayed atom densities are on l y valid for t he last time step s o in order to get the atom densities after cooling, the atom densities are decayed outside of SCALE. The cooling time decay , burnup interpo l ation , Pm-149 correction , and the fission gas corrections are perfonned using a sma ll FORTRAN code developed by this critica li ty team ca ll ed INTRPND. With isotopic s from the depletion calculation s recorded e v ery 2 GWd ff (see Section 5.6), the isotopics at any particular burnup can be interpolated.
Since the burnup delta is small between burnup points , linear interpolation can be used. To validate this approach , the isotopics at 40 GWd/T were interpolated from the OPUS p i t files at 38 GWd/T and 42 GWdff at 5.0 w/o enrichment.
Using the interpolated isotopics the k err for the previous CSA Region 2 model was 0.96653+/-0.00015. Taking the isotopics directly from the OPUS plot files at 40 GW d/T , the calculated k err was 0.96651 +/- 0.000 1 5. The difference is well within the Monte Carlo statistics. A similar verificat i on was perfonned at 11 GWd/T , NET-2809 1-0003-01 , Revision 0 58 where an interpolation between 9 and 13 was compared to the direct calculation at 11 GW d/T. The calculated k e ff using the direct isotopics was 1.1366 +/- 0.0002 , while the interpolated case was 1.1362 +/- 0.0002 , a difference of only 0.0004 , which is within the expected Monte Carlo variation. Each isotope is decayed using decay constants from the CRC Handbook 85th Edition [58]. Each iso tope is decayed into its daughter product which also may be radioactive (the decay is e-I i where 11. is th e decay constant). To ensure that the correct isotopics are obtained , the decay time desired is divided into 10 sub-intervals. The first nine sub-intervals are decayed and the tenth interval is then divided into 100 sub-intervals. The first 99 of these sub-inter v als are decayed and the last sub-interval is again divided into 100 sub-inter v als. This corrects for the fact that some nuclides are decaying into something else that is also radioactive.
If the decay time is not divided in to these very fine sub-intervals, the final concentration at the end will not be correct. It was found that this division of the decay time is fine enough such that any finer division resulted in no discernable difference in the concentrations. The correct concentration for all nuclides at the end of the decay time is thus obtained. To check the cooling time model used in the interpolation program , a special depletion was performed at 5.0 w/o to a bumup of 40 GWd/T. Then SCALE was used to decay the isotopes for 72 hours , 1 year , 5 years , and 25 years. The interpolation program was also used to decay the isotopes to the same cooling times. Table 5.22 shows the results of the verification of the cooling time. The differences are within the Monte Carlo statistics (2 sigma of+/- 0.0004) except for the c a se at 25 years. The calculated k e ff from the interpolation program at 25 years is conservative as it produces a higher k eff-NET-28091-0003-01 , Revision 0 59 Table 5.22: Verification of Cooling Time Model in the Interpolation Program Interpolation Cooling Time SCALE/ORIGEN k Program k Difference 72 hours 1.002 3 1.0023 -1 year 0.9993 0.9996 0.0003 5 years 0.9847 0.9848 0.0 001 25 years 0.9449 0.9457 0.0 008 5.10 Summary of Limiting Depletion Conditions This section h as provided the details on the limiting depl etion conditions. The d ep letion ana l ysis i s p erfo rmed for 11 batch groupings and six sets of s pecial assemblies.
For the temperature calculations , several batch groupings were combined, so there are on l y eight sets of temperatures. For each batch grouping, the depletion analysis is perfonned w ith five different temperature sets for the fuel and moderator that correspond to the burnup avera ge d assembly peaking factors of 0.60, 0.80 , 1.00, 1.2 0 , and 1.40. For each b atch grouping and p ea king factor, a depletion ana l ysis is performed for limiting top node conditions and conditions lim iting for a ll node s below th e top node. The depletion analysis is perform e d ove r a range of enrichments appropriate for each batch grouping. Atom d ensit ies were generated at each 2 GWd/T bumup step. Atom densities for burnups in between these points were determined by interpolation using the small FORTRAN code, INTRPND, d esc ribed in Section 5.9. The INTRPND code deca ye d the isotopic concentrations to any cooling time. Finally , the INTRPND code corrected th e Pm-149 atom densities for low power operation and corrected the fission gas fractions for their release rates for long-term storage of fuel. The limiting parameters for each batch grouping a re found using: 1. Tab l es 5.6 and 5.7 for moderator temperature and density, 2. Equations under Figure 5.2 and under Table 5.7 for the fuel temperature , NET-28091-0003-01 , Revision 0 60 Westinghouse Non-Proprietary Class 3 3. Table 5.8 for the burnable absorber design and bumup at which the burnable absorber is removed (except for Batch Grouping Z for w hi ch the burnabl e absorbe r i s never removed), and 4. Table 5.9 for the soluble boron concentration. Specific powers of 16 and 26 W/g multiplied by the peaking factor are used for the top and lower nodes respectively for a ll groupings. 95% of the U0 2 theoretical density is used for the stack density except for the special cases given on Table 5.21. For future fuel (Batch Grouping Z), the fuel assem bl y is depleted with a contro l rod inserted for 2 GW d/T. Then the assembly is depleted with a 20 rod l et W ABA which is never removed. The initial control rod depletion is to cover future extended part power operation with contro l rods inserted.
The bumup of2 GWd/T burnup requires operation of approximately
 
===1.4 effective===
 
fu ll power months. In addition to the WABA, the fuel is mode l ed as contai nin g 1 48 IFBA pins in all but the top node. The IFBA 1 0 B lo ading is l .5X [( mg 1 0 B/inch)] a.c to cover future designs (residua l poison from IFBA is not credited in the criticality model). For a ll other batch gro upin gs the co ntrol rods are n ot included in the standard depletion.
However assemblies under D-bank are id entifie d and the burnup requirements for these assem bli es are increased as specified in Section 5.5. NET-28091-0003-0 l, Revision 0 6 1 6 Rack Model This section describes the Keno models used in the analysis. Two-by-two (2x2) storage cell array models are used for the analysis of the three out of four areas in Region 1 (Category 2 cells) and Region 2 (Category 4 cells). A 2x2 model is also used to analyze the checker board arrangements in Region 1 and Regio n 2 (Category 1 cells). A full pool model is created to confirm the burnup requirements for the Category 3 and Category 5 fuel assemb l ies , the category cell interfaces, and to perform analyses of the Misplaced Assembly and multiple misload accidents.
 
===6.1 SCALE===
2x2 Radial Model s The rack and fuel dimensions are given in Section 3. The nominal dimensions are used in the models with the exceptions mentioned in this section. The Boraflex T M is modeled as water. As pointed out in Section 3, if any Boraflex TM remains it would still have some 1 0 B so modeling it as water is conservative. The Boraflex TM sheathing is a plate with the outside edge bent down at a 45 degree angle creating a 0.112-inch (Region 1) or 0.092-inch (Region 2) pocket for the Boraflex TM sheet (8 , 9]. The SCALE model preserves the minimum sheet material for Region 2 (which is less than Region 1) by modeling the sheathing as a squared off box with a width of 7.70 inches (8 , 9]. The same sheathing width is used for both regions. In Section 7 it is shown that using the minimum sheath material is conservative but not s i gnificant.
The connecting steel between Region I cells is modeled as an extension of the cell wall rather than a separate piece of steel. The connecting steel is slightly thicker (0.09375 inch) than the cell wall thickness (0.075 inch) [8]. This model was confirmed by accurate l y modeling the connector steel in a Region 1 checkerboard model of water holes and fresh assemb l ies of 5.0 w/o enrichment with 64 IFBA rods. The d i fference calculated is 0.0005 L\k in the conservative direction (the Monte Carlo uncertainty for these nms is only 0.00006 L\k). NET-2809 1-0003-01 , Revision 0 62 The Region 1 rack modules are separated by 1.625 inches and the Boraflex TM sheathing on the outside of the module is 0.075 inches thick rather than the nominal 0.0235 inches [8,35]. The normal flux t rap is 1.351 inches in the East-West (vertical) direction and 1.571 inches in the North-South (horizontal) direction
[8]. The greater separation between cells at the module interfaces and thicker sheathing assures that the infinite model is conservative for the finite racks consisting of three module s. The Region 2 rack modules are separated by at least 1.25 inches [35]. The re s ultant cells on the outside row of the module have a 0.075 inch wall [9]. Even witho ut the module separation, the infinite model i s conservative since, the Region 2 module interfaces possess an additiona l steel plate. Since Region 2 is an arrangement of cell boxes with resultant cells, a 2x2 model with a periodic boundary condition is required.
Since Region l storage patterns include a checkerboard and a 3-out-of-4 arrangement, a 2x2 model i s also required for Region 1. Figures 6.1 and 6.2 are co l or plots from the KENO models for Region 1 and 2, respectively.
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Figure 6.1: Region 1 KENO Model NET-28091-0003-01, Revision 0 63 
* * * * * * * *
* Fig u re 6.2: R eg ion 2 KEN O M od e l The 2x2 model s are two cell pitches wide in the x and y directions.
In creat ing the R egio n 2 model , t h e bottom l eft is modeled as a complete ce ll box with its Boraflex &#x17d; sheathing. T h is requires that the cell w a ll be sp l it between the top and bottom of th e model (likewi se for the l ef t an d right). When the periodic boundar y condition i s applied, th e two ce ll wall pieces fit together to precisely match the actua l rack d i mensions. NET-2809 1-0003-01 , Revis i on 0 64 
 
===6.2 Axial===
Model All of the in finite model s discussed in this section are finite axia ll y. Above the active fuel , the models extend the clad 7.3 inches (except for Batches A-F where the plenum is 5.2 inches) creating a plenum composed of a stainless steel spring and void (the spring is 8% of the plenum volume). On top of the plenu m r egion is a 3 1.4 3 cm of a homogeneous mixture of 50% water /50% steel (to simulate the end fitting). Below the fue l , the reflector is 50 cm of water. Outside of the reflectors , there is a zero flux bo undary c o ndition. For the axial distribution of dep l etion isotopics , the fuel is modeled as nine discrete axial nodes. For non-blanketed fuel , the top eight nodes are 8 inches each and the bottom node i s 80 i nches. For assemblies with 6 inch blankets , the top eig ht nodes are 6 inches and the bottom node is 96 inches. For assemblies with 8 inch blankets , the top node is 6 inches , the second node is 2 inches, the third node is 4 inches, the 4u1, 5th, 6th, 7 t h, and 3th nodes are 6 inches and the 9th node is 102 inches. For axial burnup models using 13 nodes, the top node is 6 in ches , the second node is 2 inches, the third node is 4 inches, t he fourth through 12 t h nodes are 6 inches, and the 13 t h node is 78 inches. The bumup distribution that determines the fuel atom densities for eac h of these n odes is discussed below in Section 6.2.1. Some of the axial blanket designs used annular pe ll ets and these are conservative l y modeled u s in g solid pe ll ets (i.e., there i s more fuel in t h e model than reality).
 
====6.2.1 Axial====
Burnup Distribution To model the axia l variation of a fuel assembly isotopic content, an axial burnup profile is needed. For this ana l ysis , the limiting profiles from NUREG/CR-6801 [27] are used for the full-length bl anketed) fuel. Table 6.1 , below , is reproduced fro m NUREG/CR-6801. The initi a l burnup distribution is approximately cos in e shaped sh i fted down a litt l e due to l ower temperatures at the bottom of the core. With increasing burnu p, the center reactivity decreases, so the flux moves t owar d the e nd and the burnup di stribut ion flatt ens. In spec tion of Table 6.1 for burn up bin s 1 through 9 , as expected, shows the top node relative power generally decreases as the bumup decreases.
However, burnup bins 3 and 5 actua ll y ha ve hi gher relative burnups th an bins 2 and 4. This does not NET-2809 1-0 003-01, Revision 0 65 make physical sense. NUREG/CR-6801 is based on end of cycle data collected by the DOE. The shapes used for burnup bins 2 and 4 come from assemblies that experienced some feature that suppressed the b urnup at the top of the core such as a transition to axial blankets or perhaps control rod insertion. If the cycle length for these limiting assemblies had been shorter , then these assemblies would probably have had similar relative burnups and would have been counted in the lower bumup bin. It is concluded that the higher relative power seen from going from bin 2 to 3 or from bin 4 to 5 is actually an artifact of the data used to create the database and not due to a phys i ca l process. To eliminate this artificial increase in relative burnup with decreasing burnup, shapes 3 and 5 are eliminated and shapes 2 and 4 are used to cover the 2/3 and the 4/5 bins , respectively.
The top node (97 .22% of the axial height) for bins 6 and 7 is greater than the top node for bin 4 , but when the top two or three nodes are a v eraged , the expected decrease in relative bumup at the top of the core is observed , so bins 6 and 7 are not eliminated.
In summary , only bum up groups 1, 2, 4, 6, 7, 8 , and 9 from Table 6.1 are used. Burn up profile groups 10, 1 1 , and 12 are not used because no non-blanketed assembly has a burnup less than 14 GWd/T (an exception is F65 which required special analysi s, see Section 8.6). The profiles for these low burnup bins were actually selected to match a center peaked flux because the reactivity is not dominated by the top nodes un t il about 14 GWd/T. NET-28091-0003-01 , Revision 0 66 Table 6.1: Axial Burnup Profile vs. Burnup Bin (27) Burnup gro u p 1' 2' 3 .j 5 6 7 9 1 0 I J' 1 _t ial Burnu ran es (GWd/MT height C 0 ,) 42-46 38 2 34 38 <6 27 0.666 0 66 0 064 06 I 3 0.944 0.9 6 0924 I 007 13 9 1.0-t I 0-iS I 0 56 I 135 19.44 I 080 I 1 04 I 097 I 133 -.00 I 091 I 11 2 I 10 I 09 0. l.()<)3 I. I 06 I IO I I 0 .I I 1.092 1.102 I 10 3 I I L I 53 4 1 69 1.0 I 090 I 097 I 112 I 11 9 I 0 47 47.22 1.09-t 1.12 5 I 1 26 1.0 50 1.094 1.136 I 1 32 I 060 I 07 I 077 69 44 I L4 I 79 75.0 0 1.077 I. t:!O 1.0 73 5 I o-o 1.069 1.05 7 I 056 I 041 I I O I I 0 52 6.1 1 0.992 1.010 0.996 0.974 0. 71 I O-i5 0.996 91.6 0. 33 0. l I 0. 23 0.743 0.7 0.689 0.669 0.894 0.845 97 22 0 447 0 448 0. 73 0 '69 0 --T he burnup profiles are a step funct ion of burnup, so they are di sco ntinu ous at th e burnup bin bound a ri es. To e liminat e th ese discontinuities in a co ns erva ti ve mann e r , t h e s h a p e in a n y bin i s cons e r vat ivel y assumed to occur at the maximum burnup in th e bin. For any burnup in b etwee n the se burnup point s, the s h a p e i s lin ea rl y interpo l ated (th e s h a p es are not changing rapid l y b etwee n burnup bin s). For example , s uppo se a n asse mbly h as a burnup of39 GWd/T. U s in gNUREG/CR-6801 dir ect l y , th e top nod e wo uld have a rel ative burnup of 0.525. In thi s analysis , ho wever, the top n o d e ha s a rel ative burnup of onl y 0.467 (linearly interpolating b etwee n 0.447 at 38 GW d/T and 0.525 at 42 GWd/T). This ensures no di sco ntinuiti es and all burnup s h a p es a r e conservative.
NET-2 8 091-0003-01 , Re v ision 0 67 For analysis of full-length fuel , the l ower 10 nodes are averaged into one node. This does not affect the calcu l ation of k , since the top half of the assemb l y dominates the reactivity.
In fact, averaging the lower 10 nodes effective l y brings the bottom lower b u rned fuel toward the more reactive top, so the approach is conservative
.* There are only a few assemblies in the IP2 or IP3 SFPs that have less than 18 GW d (T ( 11 Batch A assemblies from IP2 , 30 Batch A assemblies from IP3, IP2 assembly F65 , and IP3 assemblies V43 and V48). If there are any future assemblies with a burnup of less than 18 GWd/T, they must be assigned as Category l fuel , which does not credit burnup. The burnup profile for the 14-18 GWd fT burnup bin on Table 6.1 comes from a unique profile produced by considerable burn up under a control rod. The 15 assemblies having less than 18 GWdfT burnup that were burned under D bank are analyzed with the 14-1 8 GWdfT burnup profile from NUREG/CR-6801. A ll of the Batch A assemblies in this group have significant margin since they were a ll depleted with a 20 rodlet Pyrex burnable absorber , but none of them actually had a burnable absorber.
The other assemb l ies are analyzed using the 18-22 profile (the 18-22 profile is more reactive than the 10-14 profile). For analysis of a xial blanket fuel , a conservative burnup shape is obtained by using the smallest relative burnup at each node from all assemblies in the group. The relative burnups are not re-normalized , so the assembly burnup in the analys i s is reduced. Since the reactivity is controlled by the top nodes , however , compensating burnup increases in the lower nodes have little effect on k. The ninth node from the top is used for the ninth and all lower nodes. This has been shown to be conservative when compared to using all of the nodes. This ensures a conservati v e profi l e when the reactivity is dominated by either the top or the center (the reactivity is never dominated by the bottom because the bottom nodes always have burnups that are higher than the corresponding top nodes).
* For more discussion on using only the top 8 nodes and an averaged bottom node see the respon s e to the NRC RAI number 16 performed for the 2015 CSA (2]. NET-2809 1-0003-01 , Revision 0 68 For blanketed assemblies, all ax i a l bumup profiles from the plant were reviewed to detennine a limitin g ax i a l bumup profile for each of five axia l blanket designs: 6 inch annular* 2.6 w/o (Batch Q , R, S ofIP2) 8 inch annular 3.2 w/o (Batch T, U, V of IP2) 8 inch so lid 3.4 w/o (Batch W of IP2) 8 inch solid 3.6 w/o (Batch X of IP2) 8 inch solid 4.0 w/o (Batch 2A+ ofIP2 and Batch GG+ ofIP3) Assemb li es in Batch X are segregated into assemb li es that were depleted with no W ABA and those t hat were depleted with W ABA. The reason for this i s that the axia l bumup profile for non-W ABA assemblies is more limiting (lower bumup at the top) than for WABA assemblies , because the WABA pushes power toward the ends. In order to not penalize the WABA assemblies with the non-WABA profile, a separate depletion is perfonned by using no W ABA depletion with 148 IFBA. For the WABA group, the most limiting axia l profile in the WABA group is used, and ke rr is 0.9593. The most limiting axial profile in the non-WABA group is used to ana l yze the non-W ABA group, and k e tr is 0.9532. So , the loading curve for Batch X (based on the W ABA ana l ysis) can be used for all X assemblies.
Table 6.2 shows the axial bumup profiles detennined from the plant data.
* The designs which used annular pellets are conservative l y modeled using solid pellets (more fuel). NET-28091-0003-01, Revision 0 69 Ta bl e 6.2: Ax ial Rela t i ve Burnup s fo r Blanket e d Di sc har ge d F u e l IP2 Batch: Q, R, S T,U,V w X X noWABA Blanket Length: 6 8 8 8 8 (inches) Blanket Enrich: 2.6 3.2 3.4 3.6 3.6 (w/o) Top Node 0.420 0.448 0.471 0.505 0.519 2nd Node 0.74 3 0.59 8 0.622 0.663 0.6 77 3rd Node 0.9 0 6 0.76 2 0.772 0.779 0.8 0 5 4th Node 0.9 89 0.866 0.8 8 3 0.884 0.8 66 5th Node 1.0 2 9 0.97 5 0.98 6 0.990 0.9 87 6th Node 1.04 8 1.020 1.03 0 1.033 1.0 27 7th Nod e 1.0 59 1.0 41 1.05 0 1.052 1.0 48 8th Nod e 1.0 64 1.0 51 1.060 1.061 1.0 57 9th Nod e 1.06 8 1.0 5 7 1.06 6 1.067 1.0 62 For the Batch group Z axial burnup profile , the profiles from Batch 2A assemblies are used. This was the first batch to use 4.0 w/o blankets. The previous batch used 3.6 w/o blankets.
Since the top nodes of the 2A fue l assemblies were surrounded by the 3.6 w/o blankets , the 4.0 w/o blanket would be s l ightly underburned by the presence of the 3.6 w/o blanket s. This " transition" from 3.6 to 4.0 w/o blankets bounds all future 4.0 w/o blanket fuel (designated as Batch Group Z). The Batch 2A assemblies are di vided into two groups -assemblies that had a WABA insert plus 148 IFBA and all other assemblies.
Since the WAB N 14 8 IFBA depletion condition is used , only the assemblies that had a WABA insert plu s 1 48 IFBA should be used to find the l imiting axial profi l e. Re s ults from the Batch X analysis showed that the dep l etion effect from a reduced amount of burnab l e absorber is worth more than the effect of the axial shape with reduced absorbers. As furth er confirmation of this effect, a s pecial depletion was performed for Batch Zin which the fuel was depleted with a W ABA plus 116 IFBA compared to the s t andard depletion with a W ABA and 148 IFBA. The k err for the 1 16 IFBA depletion using the correspo n di n g profile is 0.9569 while the k e rr for the 1 48 IFBA dep l etion using the 1 48 IFBA profi l e is 0.9584. The limiting profile is the minimum relative power in each n ode for a ll 2A assemblies that had a WABA inser t plus 148 IFBA. The limiting profile fo u nd at three different bumups is in Tab l e 6.3. NET-2809 1-0003-01, Rev i sion 0 70 Table 6.3: Axial Relative Burnups for Batch Z Fuel Burnup: 48 28.5 21 GWd/T GWd/T GWd/T Top Node 0.562 0.560 0.540 2 0.727 0.746 0.729 3 0.815 0.787 0.754 4 0.893 0.826 0.787 5 0.984 0.950 0.916 6 1.020 1.001 0.976 7 1.039 1.029 1.017 8 1.047 1.042 1.041 9 1.051 1.050 1.061 10 1.055 1.057 1.074 11 1.056 1.052 1.076 12 1.065 1.067 1.095 13 1.061 1.059 1.088 14 1.065 1.061 1.092 15 1.070 1.072 1.102 16 1.069 1.064 1.089 17 1.075 1.071 1.093 18 1.080 1.083 1.095 19 1.081 1.086 1.088 20 1.081 1.085 1.075 21 1.074 1.079 1.054 22 1.047 1.045 1.003 23 0.976 0.971 0.918 24 0.875 0.890 0.848 25 0.713 0.738 0.718 Bottom Node 0.534 0.545 0.520 To s impli fy the a n a l ys i s , only the top nod es n ee d t o b e m o d e l e d w ith the l as t nod e r eprese ntin g a ll of t h e n o d es b e l ow the l as t n o d e. The fewe r th e numb e r o f n o d es mod e l e d , th e more co n serva ti ve th e r es u lt b eca u se a ll o f th e n o d es b e lo w th e l as t on e ar e m o d e l e d w ith a l owe r burnu p. I t was found th a t u s in g ni ne n o d es fo r th e hi g h e r burn e d fu e l i s n o t ove rl y co n se r v ati ve (th e k at 4 8.2 GWd/T u s in g nin e n o d es i s 0.9 5 86 w hil e th e k u s in g a ll 26 n o d es i s 0.9585). Fo r th e ana l ys i s at 2 1 a nd 28 G Wd/T (Z fue l), 13 n o d es we r e u se d in s t ea d of nin e, in o rd er to r e du ce some o f th e co n serva ti s m. Ta bl e 6.4 s umm arizes th e k c a lc u l a tion s (th e M o nt e C arlo s t a t is ti c al un ce rt ai nt y i s onl y+/- 0.00003 t.k). NET-2 8 091-000 3-0 1 , R ev i s i o n 0 7 1 Ta bl e 6.4: Ca lcul ate d k ve r s u s Numbe r o f N o des M od eled Burnu p 9 n o d es 13 n o d es 26 no d es (GWd/T) 21.0 (3 of 4 i n R eg i on 1) 0.9689 0.9686 0.9686 28.5 ( 4 of 4 in R egion 1) 1.0117 1.0115 1.0114 4 8.2 (3 of 4 in R egion 2) 0.9586 0.9585 0.9585 6.3 Dimensional Changes with Irradiation The fue l assembly dimensions change a small amo u nt with irradiation.
This creates a change in r eactivity.
This change in reactivity is real but too small to include in fuel management analysis.
With irradiation , the fuel pellet densifies and then expands , the clad grows and creeps down to the pellet , and the assembly grid expands. These are changes to the dimensions not the mass. The changes in the dimensions of the fuel pellet (with mass constant) create an insignificant reactivity ch a nge. The change in the dimensions of the clad and grid , however, result in more water relative to the fuel. Since the fuel is deliberately designed to be under moderated (to ensure a negative power coefficient), adding more water to the fuel assembly is a positive reactivity effect. 6.3.1 Clad Creep The clad outer diameter initially decreases due to creep caused by the pressure difference between the core pressure (2000+ psi) and the He fill gas (200+ psi) in the rod. At some point the clad contacts the pellet and then the clad outer diameter increases as t h e pe ll et swells. When manufactured, per Tab l e 3.2 , the difference between the pellet diameter and clad inner diameter is 0.0075 inches or 190.5 microns (l0*6 meters). If the clad were to creep down to the initial pellet diameter , then this would be a 1.8% decrease in the clad outer diameter (it is assumed that the clad thickness is constant).
The clad creep rate depends on the clad material.
However, for all of the c l ad materials, the pellet densities and then grows back past its initial outer diameter before 190 microns of creep occ u rs. NET-28091-0003-01, Revision 0 72 Figure 6.3 s ho ws clad creep after one cycle at North Anna [37]. As see n from Figure 6.3, Z irc a l oy-4 creeps down faster tha n Z IRL O&#x17d;. T h e m aximum c r eep down af t er one cycle is abo u t 90 microns for Zircaloy-4 and 70 microns for ZIRLO&#x17d;. The creep do w n data i s only for one cyc l e, s in ce b y the e nd of the seco nd cycle the c l ad h ad reached th e pellet OD. Sabo l , et al, confinns this as fo ll ows: " Th e profllom etry data obtain e d aft e r on e cy cl e of ir radiation has b een used to e valuat e th e diff e r e n ce s in th e alloys' in-r eac tor c r ee p b e ha v ior. Fu e l-clad c onta ct over most of th e rod l e n g th occurred during th e seco nd cy cl e, and th e pr ofl l o m e tr y data obta in e d on th e two-cycle rod s ar e co ntroll e d b y th e fuel p e ll e t swe llin g rath e r than th e claddin g c r ee p. Th erefo r e, th e two-cycle data c annot be u sed fo r c r eep analys i s". [3 8] F igure 6.4 pro v ides mor e data on the creep of Z ir ca l oy-4 [39]. From this plot , the creep down is ba s ically lin ear with burnup and can r eac h about -0.8% of the initial clad outer diamet er (abo ut 80 mi cro n s which agrees with th e creep d a ta from Fig ur e 6.3). Craepdown (microns) 0 20 40 60 80 100 120 0 so Zircaloy-4 ,oo 1 SO 200 2SO 300 Location (cm from bottom of ro(1) Figure 6.3: Comparison of Creep-down for ZIRLO&#x17d; and Zircalo y-4 [37) (One cycle of irradiation at North Anna Unit 1) NET-28091-0003-01, R ev ision 0 73 
-1,0 -0 , 9
* o,e ll P.t -0,7 iii e 01 0 , 64 B0 -0,6 02 0,40) ct, A -63 0,59 dPA: !.. -0,5 & nelof rods -O -04 <i , G) [i) 'l {l -0,1 normalized with J0.85 -02 I , 200 300 400 soo 600 800 1000 1200 Exposur* Tim* (Oaysl Figure 6.4: Diameter Decrease versus Exposure Time [39) Figure 6.5 shows the clad creep for severa l a llo ys for fuel u se d at the Vandellos 2 Nuclear Power Plant [40]. T he initial clad OD for the Vandellos plant is 9.5 nun. NET-28091-0003-01 , Re vision 0 74 o ,,r"-----------------------------------.
-20 ,-.. '&sect;: -40 a ..t: <.) -60 .I 'O ... -80 0 ell *= 'O 'O --100 -120 Segment Type Cladding X : SS Conventional Zr'y-4 0 : WI Low-tin Zr'y-4 6: \\Z ZIRLO 0: M.\1 MDA * : !\fit Low-rui with te.xture control A : ~t MDA with texture control * : 12 1 ZIRLO with texture control D o* *
* 0 ... A A A A * * * ... ... ...... 0 * -140 ________________________________
...._ _____ ... 0 10 20 30 40 50 60 Segment a\*en,ge bumup (G cLt) Figure 6.5: Clad Creep Down for Vandellos 2 Nuclear Power Plant [40) Fina ll y, Figure 6.6 shows the clad creep for a fuel assembly from the Ulchin Unit 2 PWR [41]. Thi s assembly uses a low tin Zircalo y-4 clad. The fue l as s embly ha s a burnup of 50.5 GWd/T. The initial clad OD i s 9.5 mm. A s can be seen from F i gure 6.6 , the clad OD with oxide i s greater than 9.5 mm in the hi g h r ea c ti v i ty zone. Thi s i s due to s ignificant oxid e build up and fuel pellet e x pansion. For this as s embly at t his burnup i t is conservative to ignore creep. NET-2809 1-00 03-01 , Revision 0 75 I-Q) +-' Q.) E cu 0 9.65------------------------, 9.60 9.55 9.50 9.45 9.40 -----Average Diameter --Diameter(oxide corrected) 0 500 1000 1500 2000 2500 3000 3500 4000 Distance from Bottom End, mm Figure 6.6: Axial Distribution of the Fuel Rod Diameter at 50.5 GWd/T [41] Wit h bumup , the clad a l so builds up an oxide coating. This oxide coating displaces water so from a reac tivity p o in t of view , it is similar to in creasing the clad outside di a m eter. Fig ur e 6.7 shows a l arge data base of the oxide layer growth with bumup [ 42). After the clad creeps d own to the fuel pellet , the clad OD increases due to pellet swelling. Figure 6.8 shows the n ormal fue l pellet swelling (v i a pellet density) a s a function ofbumup [4 3). NET-28 091-000 3-01, Re vision 0 76 160 Z ir calo y-4 140 *ZIRLO O ZIRLO -P l ant C O lo w T in Z I RLO
* P l ant C 120 Cl ZIRLO
* P l ant D C low T i n ZIRLO -P l ant D 100 t.Op!tm1 zed ZIRLO
* Plant E &#xa2;Optimized ZIRLO Plan t s F & G 80 60 40 20 10 20 30 40 so 60 70 80 B umup (G W d U) Figure 6.7: Oxide La y er thickne s s with Burnup [42] 100 98 96 0 f-94 --.. .. --: --'::!?. 0 .&#xa3; 92 (/) ----.. C (l) 0 90 ----.. ---,------' 88 --.. .. --0 20 40 60 80 100 Pe ll et Bu rn up MW d/kg U Figure 6.8: Density of Fuel Pellet as a Function of Pellet Burn up [43] N E T-28091-000 3-01 , Revision 0 77 With the data presented in these six figures , a creep model has been approximated. Since the reactivity effect is small , the model is not meant to be bounding but to have a slightly conservative mean reactivity effect. The deviation about the mean is an uncertainty that could be statistically combined with the other uncertainties.
However, since the entire effect is small, the statistical combination of the uncertainty would be negligible , so it is ignored for this CSA. The modeling is as follows: 1. Per Figure 6.8, the fuel pellet returns to its original density at 30 GWd/T. 2. The pellet density changes are translated to c h anges in the pellet outer diameter (no axial swelling). Axial swelling is a planar loss of mass that decreases
: k. 3. Using the slope of the change in pellet density given on Figure 6.8 ( l % density reduction per 10 GWd/T), the expanded pellet outer diameter is detennined as a function ofburnup. 4. Using Figures 6.4 and 6.5, assume the creep is linear with burnup. The slope for conventional Zircaloy-4 is approximated as 2.5 microns per GWd/T. 5. Using this data, determine the bumup where the clad creeps down to touching the pellet. 6. Use Figure 6.7 to estimate an oxide thickness gain of0.5 micron per GWd/T. This data is used to dete rmine when the clad returns to its original outer diameter.
Using this approach the maximum creep (the point the clad reaches the pellet) for Zircaloy-4 is at 46 GWd/T and is 115 microns. However, at that bumup the oxide layer is 23 microns thick (46 microns diametrical effect) making the net reduction of the clad outer diameter only 69 microns. Upon reaching a burnup of 5 6 GWd/T, the clad plus the oxide layer is back to its original outer diameter (this agrees with the measurements shown on Figure 6.6). The clad creep model used here simplifies the effects in a slightly conservative manner. This CSA ass umes the clad outer diameter decreases linearly from O to l 00 microns over the burnup range from NET-28091-0003-0 l, Revision 0 78 zero to 40 GWdff. As th e bumup increases from 40 GWdff to 56 GWdff, the c l ad outer diameter increases lin early to the point where it is th e same as the initial c l a d outer diameter.
Any furt h er in crease in the c lad outer diameter is con s ervat i ve l y ignored. The creep down as a function of bumup was determined to cover both Zircaloy-4 a nd ZIRLO&#x17d; clad fuel. ZIRLO&#x17d; clad however , creeps down less , so the analysis is conservative for ZIRLO&#x17d;. ZIRLO&#x17d; fuel has been u s ed since Cycle 13 (IP2) and Cycle 9 (IP3). 6.3.2 Grid Growth Zirca lo y grids grow a small amount with irradiation. Figure 6.9 shows measured grid growth in ZIRLO&#x17d; and Optimized ZIRLO&#x17d; [42]. Figure 6.10 shows the gr id growth in Zircaloy-4 (and M5 which is not used at Indian Point) [ 44]. RXA on Figure 6.10 stands for fabricated in stress-relief annea l ed (SRA) and recry s tallized (RXA) conditions.
Figure 6.11 shows the grid growth from VC Summer and Wolf Cree k measurements
[ 45]. The Wolf Creek asse mbli es are burn ed to 50 and 51 GW d/T [ 46]. ZIRLO&#x17d; grows l e s s than Zircaloy-4, but they both grow more at higher temperatures than at l ower t emperatures. The Inconel grids do not have a growth problem [ 4 7]. King , et. al. states , " Mo s t We s tinghouse fuel designs u se Inconel top grids and Inconel bottom grids , and the Inconel grids are designed to maintain a pre-load on t he fuel rod until end of life" [ 48]. NET-2809 1-000 3-01 , Revision 0 79 
! ,, cS .c lJr1o Grid GrOV;ih Data Base 0.80 ------------------------------.
+ Plant A -Plan!B 060-t---1
* Plant C X GF Opbllllled ZIRLO Data t & 40 ---1 , Plant E Op . ed ZIRLO Da a
<'5
* t ' "O *c (!) + + ! i -. ' . t I * . t_t : ** ** ; : * ' t X i f + X X t ; X 000+----,,----,----,---...----,.---,---.,----,--...,..-----i 20000 25000 30000 35000 40000 45000 50000 55000 60000 65000 70000 Bumup ( ,1WD ITU) Figure 6.9: ZlRLO&#x17d; Grid Growth r42] 0.9 0.8 RXA Zircaloy-4 0.7 0.6 0 0.5 0 0.4 0.3 t 0.2
* 0.1 0 0 0 8 0 0 0
* t,.
* j ... MS&#x17d; 0.0 0 10 20 30 40 50 FA burnup (GW d l t U) Figure 6.10: Zircaloy-4 and MS Grid growth versus burnup [44] NET-28091-0003-01 , Revision 0 60 8 0 1.2 15 ~ZIRLO Grid Envelope Zr-4 Grid Envelope (Assembly
: 1) 1.0 --tr-Zr-4 Grid Enve l ope (Assembly 2} 12 --ZIRLO Fluence -:;-0.8 -* * * * *
* Z ircal oy-4 Fluence C 9 ., u ... ., E: 0.6 .s:. .. *6 0 ... C) 0.4 3 0.2 0.0
* 0 0 1 000 2000 3000 4000 Ax i al Elevation (mm) Figure 6.11: Grid Growth of ZIRLO&#x17d; and Zircalo y-4 ve rsus Elevation
[45) There a re practical limit s to grid growt h. Excessive or unexpected dimensional changes of guide tube s or s p acer grids of a fuel assemb l y can re su lt in operationa l i ssues s uch as incomplete (contro l) rod in sertion (IRI) [ 42] or potential fuel assemb l y interact i ons and h a ndlin g concerns due to increases in the fuel asse mbl y enve l ope r es ultin g from th e l atera l growt h of the gr id s. > CD ::e ... I\ w .. E c 11'1 .. 0 ! C CD ::i iL For Ind ia n Point, if the fuel pin pitc h expands s uch that it i s uniform for the reactor asse mbly pitch , closing the i nter-a sse mbl y gap , then the fuel pin pitc h expands on l y 0.25%. Note that the outer ce ll s of the grid do n ot ha ve the fuel pin centere d b etween gr i d straps so there i s some add iti onal space even if the grid expan d ed to uniformly fit the asse mbl y pitch. If the grid stra p s were touc hin g for two e qu a ll y expande d assemblies the grid growt h wou ld be 0.69%. NET-28 091-000 3-01, Re v i sion 0 8 1 The grid growth clearly increases with bumup and the increase is more pronounced for higher bumups. A cubic fit has been generated to estimate the grid growth. The fit is roughly drawn on Figures 6.9 and 6.10. It is: Grid Growth(%)=
4.3E-6*BU 3 -0.000 1 3*BU 2 +0.0051 *BU where BU is the bumup in GWd/T. Although not all measurements of grid growth lie beneath this curve, using this grid growt h model as a unifonn pitch expans i on is expected to be conservative.
The Inconel grids above the active fuel will hold the fuel pins in their original pitch, since the Inconel growth i s in significant (from a mater i a l point of view as well as seeing less fluence , since they are outside of the active fuel region). The lower grids expand l ess. Therefore , in order to get the expansion in the pitch , the fuel rods wou ld have to bow. This analysis conservatively assumes the expa nd ed pitch is uni form over the entire l engt h of the fuel. For simp l icit y, a single fit is used to cover Zirca l oy-4 and ZIRLO&#x17d;. However, Zircaloy-4 grows more tha n ZIRLO&#x17d;. Zircaloy-4 was used as a grid for only Batches M , N, P, Q , and R in IP2 and Batches T, U , V, Wand X in IP3. All of these assemb li es have margin to their assigned reactivity category. In fact, if the grid growth were increased from 0.44% to 1.0% at 50 GWd/T , none of the assigned ca t egories for the Zircaloy-4 grid fuel assemb li es wou ld c h a n ge. 6.4 Averaged Assembly Peaking Factor Interpolation The depletion analysis is perfonned at five discrete averaged assembly peaking factors; 0.6, 0.8, 1.0, 1.2 , and 1.4. In order to simplify the dependence on this peaking factor, it is desirable to fit these five points with a single straig ht line ( one for each burn up , enrichment , cooling time , and batch grouping). It was fo und th at interpo l a tin g between peaking factors of 0.6 and 1.4 is slightly non-conservative at a PF=l.00. To ensure conservatism , depletions are performed at 0.8 and 1.2 and are li nearly int erpo l ated NET-2809 1-0003-01, R evision 0 82 and extrapolated using these two points. It was found that extrapo lati on to 0.6 and 1.4 is conservative. A representative graph at 50 GWd/T (Batch Z at 72 hour cooling) is shown on Figure 6.12. 0.960 0.955 0.95 0 0.945 k 0.940 0.935 0.930 0.925 0.5
* Points -Linear between 0.8 and 1.2 ' 0.7 0.9 T 1.1 Peaking Factor -,---1.3 1.5 Figure 6.12: Calculated ken versus Assembly Average Peaking Factor 6.5 Convergence of the 2x2 Infinite Model Calculations The convergence of the 2x2 reflected model k e ff calculation is generally achieved after only a few hundred generations.
However , all of the CSASS computer run s use a Monte Carlo sampling of at least 8000 generations and 8000 neutrons per generation.
Convergence could have been a problem in the past, when v ery few neutron histories were run (300 , 000 total neutron histories), but due to increasing the number of histories to 64 million , convergence is no l onger an issue. For both the number of generations skipped and the starting source , the SCALE default is used. For t he number of generations skipped , the default is 3. However , SCALE calculates the number of generations to skip that gives the minimum uncertain ty in the final result. Th e k eff reported for all of the calculations is the k eff With the optimum generations skipped. The number of generations skipped spans a wide range but is genera ll y between 100 and 200. When running a large number of generations , such as 8000, the input number of generations skipped is not significant to the final results. The default start NET-28091-0003
-01, Revision 0 83 source is a unifonn source over all of the fissile materials in the model. The number of neutrons per generation is a l ways 8000 or greater and for the four assembly models this sampling is enough to find the most reactive portion of the model. 6.6 Full Pool Models The full pool model is created by taking the 2x2 models for Region 1 and 2 described in Sections 6.1 and 6.2 and using them as units that are reproduced in arrays. The model has 4 large arrays (see Figure 3.1 for module identification):
: 1. Region 1 module A (10x8), 2. Region 1 modules Band C (combined as 21x9), 3. Region 2 modules D, E-1 , F-1, F-2 , G-1 , and G-2 (combined as 24x32), and 4. Region 2 modules E-2 , E-3 , and H (combined as l lx32). Modules E-2, E-3, and H are 11 ce ll s across (north to south), but since the modeling is using 2x2 units , a new lx2 unit was made and added to the right hand side of the model. This lx2 model correctly removed the Boraflex TM box on the outside of the rack module near the SFP wall but did not add the steel plate that is used to close the resultant cell on the outside of the module. The full pool model does not model the gap between the rack modules. This means that Module H is p laced next to Module C and lowered so that the bottom of Module H is the same as the bottom of Module C. Directly below Module H without any additional space are Modules E-2 and E-3. To the left of Modules E-2 and E-3 are the other Region 2 modules without any space between modules. This means there is more water on the outside than in the real SFP , because the inside dimensions of the SFP in the model are the actual dimensions of the SFP. The separation in the model between the rack and the SFP liner on the to p and left side of Region 1 is the actua l separation.
Since the separation between the rack modules has been removed in the model, there is more water on the right hand side of the model than is really in the SFP. Similar l y , at the bottom of the model there is more water than is actually there. Since NET-28091-0003-01, Revision 0 84 th e 2x2 unit fr o m Sec t i on 6 was a pp l i e d , th e ex t e r i o r of th e rac k modul es g en e ra ll y h as l ess s t ee l th a n th e o ut s id e wall o f a rac k modul e, b ut it i s clo se. In Sec t ion 6.6.1 , a nal ys i s of th e se n s iti v it y t o th e S F P e d ge i s p e r fo rmed , w hi c h s how s s m a ll se n s iti v it y. T h e s ep arat i o n b e t wee n th e R egi on 1 ra c k a nd th e t op wa ll i s 2.1 25 in c h es. T h e se p ara ti o n b e t wee n R egio n 1 and R egio n 2 ra ck s from th e wa ll on t h e l eft i s 1.25 i nc h es T h e SF P h as a 0.25 in c h s t a inl ess stee l lin e r and a co n c r e t e w a ll o u ts id e th e lin er. Figu re 6.1 3 is a SC AL E ge n era t e d dra wi n g of th e full p oo l m o d e l. F i g u re 6.13: F ull Pool M od e l N E T-2 8 091-000 3-01 , R ev i s ion 0 ll.C(M) O wrn fWHCRUll 2 -l'WHLIUl'll.
J -MlUtlAL" -l'VIJ[IUAL 7 0 Ml[IUAL I -Ml[RlR.10 C)MIOUAL J t -fWHDUAL 6' -M I UUnl '3 M I UUAI. 1 10 M l (RIAL 2 1 1 flHlUUAL 2 1 2 ~Ml(IUtlt.2 1 3 (::J mmnAL 2 1 4 Q M1&#xa3;11tlAL 2 1!S WHOUAL 2 1 8 MIUUAl.217 M lUUAl. 211 MTiltlAI.
312 CJ MlUWll 3 1 3 -Ml[RIAI 3 1 4 Ill Ml[IW'll 3 t!S -Mf(IUnt.Jl6 -MfUUfll.3 1 7 -Ml[RlfL 318 -M J Utltll.,1 0 -Ml(RIAL'lll
-P!AllRIAL41 2 -Mf[MIAL<II J -PWU UIIAI "11'4 -MJ[RIAL"115
-MllJUf:11'41 6 -Ml[IUtll.<1 1 7 -Ml(RIAL*H 8 -Ml(RIAL!UI -MfU UAL!S l 2 -Ml[MIIW..!SIJ -M1CRIAL!il 4 -Mll.RIAL!S l!S MIIM U II. 5 1 8 -WHUtlAL!Sl7 -MIUUAl.!S 1 8 O iwHUllnt.9etl 85 -----_j The pink squares in Figure 6.13 are empt y cells which can contain up to 50% water displacement with non-fuel compone nt s. These ce ll s are modeled w ith a void fraction of 50%. The water holes (white) are modeled as pure water. However, Section 8.13 provides ana ly sis that shows that up to 50% volume fraction of stainless stee l may be in these water holes. Figure 1.1 shows that contro l rods are in specified l ocations.
These contro l rods are in the full pool model but cannot be seen in Figure 6.13 due to poor reso lution. F igure 6.14 is a blow up of Module Hand the contro l rods can be seen. ........................................................................................................................................................................
:.::.:*:.::.::
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Figure 6.14: Model of Modu l e H Showing Control Rods NET-28091-0003-01, Revision 0 """" O vo10 -Ml[IIIAL l -Mrt:RIH..4
-MJ[Rlnl.7 IW'l1&#xa3;111nL 8 -MT[RIALIO c:J MlERIAl. 3 1 -PIAll:IUlll..
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21 5 MIERJnt 2 1 6 MT(ll: l AL 3 10 MTCIUl'W..
3 1 1 MrEIUl'll 3 12 MTEll:1,-_
313 -MTU U nL l l<I -MICRIAI 3 1 5 -MT&#xa3;RIN 3 1 6 -MT[RIAl..3 1 7 -Mf[Rll'll 3 1 1 -Mf(ll:lnL<II O -MTERIAL<lll
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'5 H -Ml[NIN..t ll'5 l'IAT(ll:fnL 5 U l -l'IATEll:Hl.
5 1 7 -Mf(ll:Hl.5 11 D"'HENHI. Ne 86 There could be a concern regarding the content of control rods that have been used in the core. Due to end effec t s the reactivity is dominated at the top of the active fue l. However, the top third of t h e control rods is not allowed in the core if the power is above 50% of full power. Any depletion of the control rods is insignificant.
However, to confirm that content is not important, a run was performed with the atom densities of the control rod reduced 20%. The calculated k e ff is within the Monte Carlo uncertainty of the reference case. 6. 6. 1 Sensitivity of the Full Pool Model to Modeling Assumptions A number of model sensitivity cases were performed to determine if the modeling of the outside of the racks and the SFP wall is adequate.
Table 6.5 provides the results of these sensitivity cases. The reference cases use the nominal separation from the wall found on the SFP layout drawing for the top and left side of the model [35]. The reference cases are at the limiting enrichment (5.0 w/o) and bumups (21, 27.7, 48.19, and 57.89) p l us asymmetry.
For this sensitiv i ty study, the linear dimensional changes are l arge (tota l e l imination and a doubling of the width). Even with these large changes , the maximum reactivity is only five times the Monte Carlo uncertainty of the cases. The models are slightly more sensitive to the separation from the wall but with moving the racks as close to the wall as possib l e, the reactivity effect is at most 0.0005 in k. The model uses the standard regulatory concrete which is a s t andard mix t ure in SCALE. An EPRI study of a conservative minimum water concrete is used to find the sensitivity
[13]. This very conservative concrete only increased k by 0.0008 for Region 2 and 0.0004 for Region 1. Since all of the extreme changes caused 0.0008 or less change in reactivity , the full pool model is adequate.
The maximum k for the analysis is 1 % less than the regulatory requirement.
NET-28091-0003-01, Revision 0 87 Table 6.5: Full Pool Model Sensitivity Tests Case k Sigma '1k Region 1 Reference 0.9687 0.00007 Eliminated the SFP steel liner 0.9689 0.00006 -0.0002 Increase the SFP steel liner from .25 to .5 inches 0.9686 0.00006 0.0001 Decreased rack/liner separation from 2.125 to 0.8225 inches (Top) 0.9691 0.00006 -0.0004 Increased rack/liner separation from 2.125 to 4.125 inches (Top) 0.9686 0.00006 0.0002 Changed Concrete from reg-concrete to EPRI -minimum water 0.9691 0.00006 -0.0004 Region 2 Reference 0.9584 0.00005 Eliminated the SFP steel liner 0.9585 0.00006 -0.0001 Increase the SFP steel liner from .25 to .5 inches 0.9587 0.00006 -0.0003 Moved rack to meet the SFP steel liner (left side) 0.9589 0.00006 -0.0005 Increased rack/liner separation from 1.25 to 2.98 inches (left) 0.9584 0.00006 0.0000 Changed Concrete from reg-concrete to EPRI -minimum water 0.9592 0.00006 -0.000 8 6.6.2 Convergence of the Full Pool Model There i s a c la ss ic problem for Monte Carlo c onvergence known as the "k-E ffecti ve of the world." T hi s problem wa s introduced by E lliot White s ides in 19 71 (50] and more r ece nt paper s such as the Brian C. Kiedrow s ki and Forre st B. Brown pap e r at ICNC 2011 (51]. Concern has been raised that a large full pool model could face these probl e m s. First, in 1971 when White s id es rais e d thi s issue it was common pra c tice t o run a few hundred neutron s per generation. The pap er b y Kiedrow ski a nd Brown u se d a lot mor e n e utron s per generatio n but considered solutions with 10 , 000 n e utron s p er ge neration o r l ess as typical. For this CSA, employing computers with mor e powerful CPUs , 16 , 000 n e utrons are s tarted per generation. This analysis a lso u ses 8, 000 generations.
The problem posed b y Kiedrowski and Brown used a cadmium wrapper to isolate th e reacti ve locati o n. The full-pool-model employed in th e current ana ly s i s i s more neutronic a ll y coupled than the model u se d in the Kiedrowski and Brown problem so a t the curr e nt number of neutron s per ge n eratio n , so urc e convergence i s not a problem. Kiedrow s ki and Brown pointed out "s upp ose th e source s p ec ification is modified to incorporate only th e c e ntral sphere. Will thi s yie ld more reliable results? The answer is yes, and remarkably so. Eve n wit h a batch size of lk , the value ofk e ff is a l ways predict e d co rrectl y for each of th e 100 trials." In th e current NET-2809 1-000 3-01, Revision 0 88 analysis, when there is an isolated high reactivity area, the start source is specified at that location, thereby removing the convergence concern. Many articles are available relating to the topic of source convergence. This issue , however, is not of primary concern for this CSA, as it seeks to determine the k err, not the flux. In the Kiedrowski and Brown paper , they say , "'sloshing' of the fission source has an observable impact on k 0 rr .... however, the impact o n ke ff itself is small." This sloshing of the fission source increases the uncertainty in the calculated k err which is included in the bias and uncertainties. The l arge number of generations used in this CSA assures t he mean k err is correctly predicted within the uncerta i nty. To confirm that the full pool model is sufficiently coupled , six different start sources were used to analyze one of the final full-pool cases. The calculated k 0 rr is dominated by Region I. For cases including Region I , this CSA uses a start source that covered most of Region I. In order to challenge the convergence issue , the source was started at four boxes as far from the dominate k err area as possible. A final case was run with the SCALE default source , which is uniform over all fission materials in the problem. Figure 6.15 shows where the four start sources are located for the model. Table 6.6 shows the calculated k err values. All of the calculated k 0 rr values are within two standard deviations. The cases with the worst start source have a higher reported sigma. It is concluded that the neutrons per generation and t he number of generations used for the full models for this CSA is sufficient for convergence.
Figure 6.16 shows how the average k err behaves with progressing generations.
It is clear that the neutron population is moving away from the start source when the start sources are located in the comers of Region 2. All of the cases in this CSA are converged.
In some engineering applications, such as checking interface interaction , all that is needed is to determine that the location of the start source is not the most reactive location in the model. When this is the case , then convergence is not required. For example, in this case , it is clear that the three Region 2 start sources are not at the most reactive location in the model. NET-28091-0003-0l , Revision 0 89 This ana l ysis uses the k etr produced after a certain number of generations skipped. The number of generations to skip is determined by SCALE, whic h minimizes the standard deviation. It has been found that this almost always yields the best estimate of k. Table 6.6 shows that the number of ge nerations skipped can vary greatly. Clearly, for this convergence test , if a fixed number of generations to skip were used, the results would be le ss accurate.
Reviewing Figure 6.16 makes it clear that this SCALE feature is correct and needed if the start source is poorly se l ected. The differences in k err seen at the right hand side of Figure 6.16 are greater than that seen in Table 6.6 since the variation in the number of generations skipped is n ot part of the average k e tr shown on Figure 6.16. Table 6.6: kerr Changes With Start Source (Full Pool Model -8000 Generations, 16000 Neutrons per Generation)
Source Location Calculated k Reported Sigma Generations Skiooed Region I only 0.968708 0.00006 7 662 Uniform for Problem (Region l and 0.968724 0.000066 372 2) Region I Bottom Left 0.968704 0.000066 445 Top Right 0.968663 0.000066 525 Bottom Ri g ht to Cask Area 0.968809 0.000075 2066 Bottom Far Right Above Cask Area 0.968682 0.00006 8 1282 NET-28091-0003-01, Revision 0 90 F igure 6.15: Locations of the Start Sources for the Convergence Tests NET-28091-0003-01 , R ev ision 0 Lftl:NO O vo10 MT[AIM.2 -M T[Rl._.l -MAT O II.--, '4 O w ou:11..._a O M 1 u11V...1 1 -Ml&#xa3;Rl(I,_
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-l'Wl lUW'l.*U 5 -M'JUtlAL ... 1 6 -IW H(fllAL'41 7 -IW H&#xa3;RUll. *11 8 c:J ,-nEIUnL SIO -l'W'H[l!IAL 5 11 -IVI I UHl'W..11 1 2 -IWIT11UAL ll 1 3 -MT(RIAl.5 1<1 -11 M (l!IAL 5 t5 c::J MU:: RUW .. 5 1 6 l'WHl:: H l nl 5 17 -M!E fUN ti 18 9 1 0.97 0.96 0.95 ..:,: "C CII 0.94 .. "S ... iv u CII t,Q 0.93 ... ------CII ::, er: 0.92 0.91 0.90 T 0 1000 2000 ' 3000 4000 5000 6000 7000 Number of Generations 8000 Start Scource Location
* Region 1
* Uniform
* Region 1 Left Bottom Top Right , Far R i ght Bott om
* Bottom Right Cask A rea Figure 6.16: Change in Average kcrr with Progressing Generations NET-28091-0003-01 , Revision 0 92 
: 6. 7 Summary of Modeling Assumption s The following is a summar y of the modeling assumptions
: 1. Bounding fuel stack density and nominal dimensions for pellet OD , clad OD/ID , and guide tube OD/ID are used. 2. Axial blankets are modeled for five different axial blanket designs. A conservative bumup profile for each blanket design is created by using the smallest relative power at each node from all assemblies having the same blanket design. 3. Grids are ignored (grids displace water between the fuel pins which causes k e rr to decrease).
: 4. No Boraflex TM in the Boraflex TM sheathing and the Boraflex &#x17d; is replaced with water (if any Boraflex T M remained , it would be less reactive than water). 5. NUREG/CR-6801 bounding axial bumup profiles are used for all full-length fuel. To addre s s discontinuities , the shape in any bin is conservatively assumed to occur at the maximum bumup in the bin. For any bumup between points , the shape is linearly interpolated. 6. Top of fuel assembly models a plenum (length dependent on batch) filled with 8% volume fraction of stainless steel. Above the plenum is a 50/50 mixture of stainless steel and water to simulate the end fitting. Bottom reflector is 50 cm of water. 7. Periodic boundary conditions are used to represent an infinite array. 8. The dimensional changes with bumup are modeled as described in Section 6.3. a) In reference to clad creep , clad thickne ss i s con s tant, and creep occurs linearly , in two phases (creep inwards and growth ou twa rds). b) In reference to grid growth , pin pitch expand s uniformly.
The water temperature in Region 1 of the SFP is mode l ed at 180 &deg;F which is the SFP design basis maximum temperature.
In Region 2 , the modeled temperature is 70 &deg;C which is the temperature that maximizes k e rr in Region 2. NET-28091-0003-01, Revision 0 93 P roprieta r y In formatio n R e m ove d Westing h o u se N on-P ro pri e t ary C l ass 3 7 Sensitivity Analysis T hi s sec ti o n prese nt s a n a l ysis of th e se n sitivity of th e m o d e l s t o th e ma nu fac turin g t o l era n ces. After t h e se n s iti v i t y i s d e t e rmin e d , th e rac k up o f th e un ce rt a inti es a nd bia ses i s pr ese nt e d. 7.1 Manufacturing Tolerances Ca lculation s a r e p e rform e d to qu a nti fy th e reac ti v ity effec t of ch a n ges du e to m a nu fac turin g t o l era n ces. For R egio n 2, th e t o l era n ce ca lcul at i o n s are p erfo rm e d a t th e h ig h es t cre d i t ed bumup co n ditio n s (5 w/o at 49.5 G Wd/T, P F=l.4) and a l ow b urn e d co nd i ti o n (4.2 w/o at 30.0 GWd/T, 2 5 year coo lin g, P F=.6). Ta bl e 7.1 pr ese nt s t h e ca lcul ate d tol e r a n ce reac ti v iti es. Table 7.1: Tolerance Reactivity Effects Re2ion 1 1 2 2 2 Fuel Reactivity Cate2ory 2 3 4 4 5 Arran2ement 3 of 4 4 of 4 3 of 4 3 of 4 4 of 4 Enrichment (w/o) 5 5 4.2 5 5 Burnup (GWd/T) 20.5 28 30 49.5 60 Tolerance Calculated Ak (inches) (no Monte Carlo Uncertainty Adjustment)
Pellet Density +0.35% 0.0003 0.0004 0.0003 0.0005 0.0005 Pellet OD [ ] a.c 0.0002 0.0001 0.0001 0.0000 0.000 1 Clad ID [ ] a.c 0.0002 0.0002 0.0000 0.0001 0.000 2 Clad OD [ ] a.c 0.0008 0.0008 0.0009 0.0009 0.0007 Pin Pitch +0.0014 0.0028 0.0032 0.0016 0.0015 0.0013 Vertical Cell 0.00 32 0.0049 0.0016 0.0014 0.0015 Pitch --Horizontal 0.00 3 5 0.0050 0.0016 0.0014 0.0015 Cell Pitch --Wall Thick -0.007 0.0037 0.0043 0.0026 0.0025 0.0025 Cell ID 0.0002 0.0002 0.0003 0.000 2 0.0002 BoraflexTM Sheathing
-0.003 0.0016 0.0016 0.0010 0.0009 0.0010 Thickness RSS 0.0069 0.0090 0.0040 0.0038 0.0038 N E T-28 091-000 3-0 1 , R ev i s ion 0 94 The sign of the tolerance on Table 7.1 shows which direction increases
: k. No tolerance reactivity effect is calculated for Category 1 fuel , the Category 3 tolerance is applied for Category 1. A checkerboard of Category 1 fuel (see Section 8.1) has a large margin to the criticality safety limit so an approximate tolerance is appropriate. PWR fuel assemblies are designed to be under moderated at power , so the moderator temperature coefficient is negative to prevent large power excursions.
Therefore, increasing water between the fuel rods (and ignoring grids) increases
: k. This is demonstrated by calculations of the reactivity from varying the pin pitch and the fuel clad outer diameter (shown i n Table 7.1). The grids are conservatively ignored since they displace water around the fuel pins. The fuel pin pitch tolerance (0.0014 inch) used in this a nalysis is the maximum pin separation possible before the assembly gap becomes zero and all pins in the core are separated by a single enlarged pin pitch. The fuel enrichment used for determining if the loading requirements are met is the as-built enrichment for each assembly. The uncertainty in the as-built enrichment is +/-0.02%. Note that the uncertainty in the as built enrichment is less than the traditional uncertainty of 0.05% which is based on the nominal enrichment.
The reactivity of the fuel enrichment uncertainty is larger at low enrichments.
Calculations show that the reactivity due to enrichment uncertainty for Category 4 fuel is 0.0028 Llk at 2.0 w/o and is 0.0008 L'lk at 5.0 w/o. Fuel Categories 1 , 2, and 3 have fixed burnup requirements at an enrichment of 5.0 w/o , so there is no enrichment uncertainty.
For Category 4 fuel, the enrichment uncertainty reactivity effect is linearly interpolated using the two points 0.0028 L'lk at 2.0 w/o and 0.0008 L'lk at 5.0 w/o. For Category 5 fuel, the enrichment uncertainty at 5.0 w/o is used because Category 5 fuel has a fixed bumup differential which would have the least margin at 5.0 w/o. A tighter (smaller) rack cell pitch increases k etr of the SFP for both regions because the fuel assemblies are closer together.
The Region 1 change in ke rr for reducing the cell pitch is much larger due NET-28091-0003-01 , Revision 0 95 to the decrease in the flux trap. In ca l culating the reactivity effect of decreasing the cell ID , the cell pitch is maintained , so the effect on k e ff is small. A conservative minimum width of the Boraflex TM sheathing is used for the mode l. Calculations are performed where the sheathing width is increased from 7.7 to 8.0 inches in Region 1 and Region 2. The k e ff decreases by 0.0005 t.k and 0.0004 t.k for Regions 1 and 2 , respectively. Calculations show that the highest reactivity occurs with elevated temperatures (see Section 8.1). This is due to the water hole in the 2x2 model holding down reactivity. A higher temperature in the water hole results i n a lower water density and the reactivity hold down is reduced. For Region 2 , the peak reactivity occurs at a temperature of 70 &deg;C. Above 70 &deg;C , the reactivity begins to decrease because the reduced mo deration from lower density water within the fuel array is dominated by the reactivity hold down of the water in the water hole. For Region 1 , the reactivity increases with increasing temperature all of the way to boiling. Therefore , for Region 1 , the water temperature is modeled at 180 &deg;F, the SFP design basis maximum temperature during normal operation. Boiling conditions are analyzed as an accident where soluble boron credit can be used. No calculations of tolerances or sensitivities were made with borated water. The borated conditions have excess margin , which covers any differences in sensitivity with borated water. 7.2 Burnup Dependent Biases and Uncertainties Bumup i ncreases the uncertainty in the analysis of k. To account for thi s there are se v eral biases and uncertainties that are bumup dependent.
They are the depletion uncertainty , the minor actinide and fission produ c t bias , the bumup uncertainty , the clad creep bias , and grid growth bias. The first b umup dependent bias or uncertainty is the depletion uncertainty in the atom densities.
This is accounted for by an uncertainty of 5% of the t.k between the zero bumup case and the case at the desired bumup. The 5% has been supported by a number of studies mentioned in Section 4 and is NET-28091-0003-01 , Revision 0 96 recommended via DSS-ISG-2010-01
[5]. For fuel Categories 2 , 3 , 4 , and 5, the zero burnup k e fffor the enrichment of interest is calculated and used with the calculated k e ff of the burned case to determine the worth of the depletion uncertainty.
Note that no bias is applicable. As an example, at 49.5 GWd/T, the delta-k of depletion is 0.3257 ~kin the 3-out-of-4 arrangement of Region 2 (Category 4 cell). This makes the depletion uncertainty 0.05 x 0.3257 ~k = 0.0163 t.k. The second burnup dependent bias or uncertainty is the minor actinid es and fission product worth bias. This bias, previously mentioned in Section 4, covers the bias and uncertainty due to the lack of criticality data for the minor actinides and fission products. This bia s is detennined by calculating ke rr in the appropriate model with the 1ninor actinide s and fission products remo v ed. The difference in reactivity between the calculations with and without these isotopes is multiplied by 1.5% and included as a bias. This approach was suggested in NUREG/CR-7109 and conservatively co v ers the uncertainty
[22). As an example, at 49.5 GWd/T for Category 4 fuel , a calculation determined that the actinide and fission product worth is 0.1371 ~k. Therefore the bias at 49.5 GWd/T is 0.0021 ~k. The third burnup dependent bias or uncertainty is the uncertainty in the declared burnup from the reactor records (shortened to burnup uncertainty). The burnup uncertainty from the reactor records is assumed to be 5% of the burnup [26]. This value is based on comparisons presented in Section 7 .2 of NUREG/CR-6998 [26] of in-core measured burnups that demonstrate that the uncertainty in assigned burnup values is less than 5%. The effect on reactivity is calculated by comparing the kerr calculated for the same case at two different burnups. For example, at 5.0 w/o the k e ffat 49.5 GWd/T i s 0.9560. The k e ff at 0.9x49 .5=44.55 GW d/T is 0.9813 (10% less burnup ). So the~ due to a 5% burn up uncertainty at 49.5 GWd/T is: (0.9813 -0.9560) I 2 = 0.0127 ~k The bumup uncertainty changes as a function ofbumup and enrichment because the delta-k between two low bumups can be larger than the delta-k between two high bumups. The burnup uncertainty is NET-28091-0003-01 , Revision 0 97 calculated for each loading curve point by the same procedure (us in g the L'lk between two different burnups with all other parameters stayi n g the same). As developed in Section 6.3.1, the clad outer diameter reduction (in microns) due to clad creep is a function ofburnup , starting at Omicrons at O GWd/T , lin early increasing to 100 microns at 40 GWd/T , then linearly decreasing to zero at 58 GWd/T. To de t ermine the reactivity effect at the maximum creep of 100 microns, a special depletion was run with the c l ad outer diameter reduced by 50 microns (the average reduction between O and 40 GWd/T). A case was then run at 40 GWd/T with the c lad OD reduced by 100 microns u sing the specia l depletion for the number densities. When compared to the nominal case at 40 GWd/T , the delta-k at 100 microns is 0.0012. The c l ad creep bias (L'lk) is expressed as a function of b urnup as fo ll ow s: C l ad creep bias (L'lk) = 0.00 1 2 x BU/40 = 0.0012 x (58-BU)/1 8 BU::S 40 BU>40 As can be seen in Table 7 .1 , the sensitivity to the clad OD is similar for all categories of fuel so the same clad creep bias formulation is used for all categories.
Finally, as discussed in Section 6.3.2 , the grid growth as a percent of the grid cross section is a function of burnup. The tolerance calcu l ation for pin pitch uses 0.0014 inch for the pin pitch to l erance which is 0.25% of the pitch. The grid growth bias is: Grid Growth Bias (L'lk) = (0.0000043 x BU 3 -0.00013 x BU 2 + 0.0051 x BU) x 4 x pp where pp= pin pitch tolerance worth from Tab l e 7.1. NET-28091-0003-01, R evision 0 98 
 
===7.3 Eccentricity===
 
Generally the plant intends to place the fuel assembly in the center of the ce ll. However , it is acceptab l e to have the assembly in any lo cation with in the cell. A study performed for the Millstone 2 license application showed for that plant , the placements were approximately random [ 49]. For this 2017 CSA, i t is assumed that the placement of the assemblies in the cells is random (there is nothing in the cell that would cause the assembly to be preferen t ially placed in one comer over another). As was performed for Millstone 2, the number of assemblies that is eccentrica ll y placed in particular quadrants is determined such that the probability of such placement is l ess than 5% over the lifetime of the plant. Region 2 contains a 3-out-of-4 set of Category 4 ce ll s surrounded by Category 5 cells in a 4-out-of-4 arrangement on the outer two rows and a 4-out-of-4 arrangement with checkerboarded control rods. To determine an eccentricity bias for the Category 4 ce ll arrangement, 16 assemblies are placed as c l ose as possible (u sing the standard pin pitch) to a central assemb l y. Figure 7.1 shows the placement of th e assemb li es. The proba bilit y of 16 assemb l ies being randomly p l aced in the most reactive quadrant is 0.25 1 6 = 2.3*E-10. There are 504 Category 4 cell locations. It is conservative to estimate that eac h l ocation cou l d have 100 moves. This makes the probabi li ty of getting such an arrangement 2.3*E-10*504*100=
1.2E-5 which is much less than t h e required 0.05. NET-2809 1-0 003-01 , Revision 0 99 
--------------------
-------, Figure 7.1: Category 4 Region 2 with 16 Assemblies Eccentrically Placed The model used is an 8x8 model with periodic boundary cond iti ons , where a ll assemblies except the central 16 eccentric assemblies about a central assembly are centered. This means that there are actually about eight eccentric sets separated by two rows of centered assemb li es. The enrichment and bumup used for the fue l is 5.0 w/o and 48.19 GWd/T. Table 7.2 s ho ws the results of the ana l ysis. The refere nc e calculation i s th e 8x8 model with centered assemb li es. The ecce ntri c it y bias for Region 2 is n egat i ve and therefore conservatively i gnored. This i s not unexpected s in ce wa t er hol es or contro l s rods can break up the effect of eccentricity.
NET-28091-0003-01, Revision 0 1 00 Table 7.2: Eccentricity Results Category Calculated k Sigma Ak Reference 4 0.9586 0.00004 16 Eccentric Assemblies 4 0.9585 0.00004 -0.0001 Reference 2 0.9671 0.00007 16 Eccentric Assemblies 2 0.9690 0.00006 0.0019 16 Eccentric Assemblies Shifted down 2 0.9693 0.00007 0.0022 The bumup increment for the Category 5 fuel is determined such that Category 5 fuel is not limiting (see Section 8.4.2). Therefore, the eccentricity is maximized by moving Category 5 fuel closer to the C ategory 4 cells. In the final full pool model all of the Category 5 fuel assemblies in Region 2 are moved right in their cells to be as close as possible to the left hand side of Category 4 cells. Region 1 is more complicated. First, a checkerboard area with Category 1 fuel is designed to have a low k eff-The keff is sufficiently low (0.8548) that eccentricity within Category 1 cells i s not a concern. Category 2 fuel eccentricity is analyzed with 16 assemblies placed closest to a central assembly in the same manner as was performed for Category 4 (Region 2). Due to the flux trap, however , it is possible that it is more reactive to mov e the central assembly from the ce nter of the cell toward one of the four sides. Therefore , analysis was also performed where the central row of assemblies is moved down to be closer to the row of assemblies below it. Figure 7.2 shows this arrangement of fuel assemblies.
The results of the analysis for Category 2 are shown on Table 7.2. Category 2 is showing an eccentricity bias and the bias is slightly lar ger when the central row is moved down. The full pool mod el incorporates the eccentric placement of the assemblies, so the bia s is intrinsic to the analysis.
Therefore, it does not need to be added to the final calculated
: k. As in Region 2, the bumup penalty for th e outer two rows of Region 1 (Category 3 cells) was selected to prevent Category 3 fuel from being more limiting.
In the final full pool model, both rows of Category 3 fuel assemblies are mo ve d down toward the Category 2 ( or 1) cells to maximize the NET-28091-0003-01, Revision 0 101 eccentricity
/interface effect. Similarly the Category 5 fuel assem b lies at the Reg i on 1/Region 2 i nterface are moved up and to the l eft to make them as close as possible to Category 2 ce ll s. ::.::.:::.::.::
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I ~:_;;;:::;;;_;;;;L_ I f F i g ur e 7.2: Ecc en tr i c M od e l fo r Categ or y 2 w i t h Ce ntr a l Ro w S hi fted D o w n 7.4 Additional Biases and Uncertainties The criticality validation for major actinides, absor b ers, and structural materials provides a bias and uncertainty due to validation.
As provided in Section 4 , the validation bias and uncertainty for a ll calculations except those with very hard spectra s uch as borated cases or boiling cases are 0.0021 Lik and NET-28091-0003-01 , Rev i sion 0 102 0.0087 t.k respectively (for fresh fuel the bias and uncertainty is less as given in Section 4.3). For cases with a hard spectrum (EALF greater than 0.4 eV) the bias and uncertainty are 0.0027 t.k and 0.0112 t.k respectively.
The criticality validation also revealed the need for an additional bias and uncertainty for conditions above room temperature.
Due to using water holes to control reactivity, it was found that elevated t emperatures have a higher reactivity.
For Region 2, the highest reactivity is at a temperature of 70 &deg;C (see Section 8.7). For Region 1, the highest reactivity is at the SFP design basis maximum temperature which is 180 &deg;F (82 &deg;C). From the validation section, the temperature bias at 70 &deg;C is 0.00043 t.k with an uncertainty of 0.0013 t.k while the temperature bias at 82 &deg;C is 0.00053 t.k with an uncertainty of 0.0013 t.k. A 2x2 model cannot provide adequate modeling of eccentric placement of fuel assemblies in rack cells. Therefore, a bias is required.
This bias is detennined in calculations performed in Section 7.3. It was found that eccentric loading of Category 4 cells (3-out-of-4 of Region 2) does not increase reactivity , so there is no eccentricity bias for Category 4 fuel. The Category 5 fuel bumup penalty, which is derived in Section 8, includes eccentric positioning in the large model, so it is inherently included in the bumup penalty. However, Category 2 (3-out-of-4 in the Region l flux trap design) does have an eccentricity bias of0.0022 t.k (see Section 7.3). The final additional bias and uncertainty is the reactivity effect due to Monte Carlo statistical uncertainty.
The 95/95 Monte Carlo statistical uncertainty in each tolerance calculation is where <:rb and crp are the Monte Carlo standard deviations for the base case and the perturbed case, respectively.
The base case calculation was run with 1.024 billion histories to reduce the statistical uncertainty to+/- 0.00002 t.k (1 sigma) for the base case. The perturbed calculations are run for 64 million NET-28091-0003-01, Revision 0 103 histories for an uncertainty of+/- 0.00008 ~k ( l sigma). This makes the Monte Carlo standard deviations
((2*0.00 0 02)2 + (2*0.00008) 2 )05 = 0.000 1 6 ~k. In t h e statistical combination of terms each of t h ese would be squared. Since there are 13 statistical to l erance components the sum of the se terms would be 13*(0.00016)2. The fina l step of the statistical comb i nation is taking the square root. The square root of 13*(0.00016)2 i s 0.0006 ~k. So , 0.0006 ~k is the Monte Carlo statistical uncertainty in the tolerance calculations and is combined in the total rack up of u n certainties.
 
===7.5 Biases===
and Uncertainties Rack-up Sections 7.1 through 7.3 pro vi des the biase s and uncertainties and in the case of the bumup dependent bia ses and uncertainties , how to calculate them. For Region 1 , the fuel categories all ha ve fixed bumup and are va li d up to 5.0 w/o enrichment.
Therefore , the final bias and uncertainty can be determined. Table 7.3 p rovides the total rack up of biases and uncertaintie s, inc l uding the statist i cal combination of the uncertainties for Region 1 (fuel Categories 1 , 2, a n d 3). NET-2809 1-0003-01, R evision 0 104 Table 7.3: Total Bias and Uncertainties for Region 1, Categories 1, 2, 3 Category 1 2 3 Arrangement 2 of 4 3 of 4 4 of 4 Enrichment (w/o) 5 5 5 Burnup (GWd/T) 0 21 28.5 Component Bias Uncertainty Bias Uncertainty Bias Uncertainty Validation ( critical 0.0024 0.0035 0.0021 0.0087 0.0021 0.0087 experiments)
Depletion uncertainty
---0.0067 -0.0092 Minor Actinides and Fission 0.0010 0.0013 Products Bias --Bumup uncert ainty ---0.0057 -0.0081 Manufacturing Tolerances
-0.0090 -0.0069 -0.0090 Enrichment Uncertainty
------Monte Carlo statistics
-0.0006 -0.0006 -0.0006 Ecce ntricity Bias 0.0 -0.0022 -o.o* -Clad creep bia s --0.0006 -0.0009 -Grid growth bi as --0.0010 -0.0018 -Elevated temperature 0.0005 0.0013 0.0005 0.0013 0.0005 0.00 1 3 Total Rack Up (~k = RSS) 0.0029 0.0098 0.0074 0.0142 0.0066 0.0176 Sum of Bias and 0.0127 0.0217 0.0242 Uncertainties
* Category 3 cells are located on the outer two rows of Region 1 so eccentricity in an infinite model is not relevant.
The eccentricity is part of the full pool model and therefore no bias is applied. For the two categories of fuel in Region 2, the burnup is allowed to change so the bias and unc ertainty will change. However, Table 7.4 is provided as an example of the total rack up for Category 4 and 5. NET-28091-0003-01, Revision 0 105 Table 7.4: Sample Categor y 4 and 5 Bias and Uncertainty Rack-up Cate2orv 4 5 Arrangement 3 of 4 4 of 4 Enrichment (w/o) 5 5 Burnup (GWd/T) 49.5 60.5 Component Bias Uncertainty Bias Uncertainty Va lid ation (critical experiments) 0.0021 0.0087 0.002 1 0.0087 D e pl etion uncertainty
-0.0163 -0.02 14 Minor Actinides an d Fission Products Bias 0.0021 0.0026 Burnup un certainty
-0.0124 -0.0 155 Manu facturi n g Tolerances
-0.0040 -0.00 40 E nri c hm ent Uncertainty
-0.0008 -0.0008 Mont e Carlo stat i s tic s -0.0006 -0.0006 Eccentricity Bias 0.0 -o.o* -Cla d creep bias 0.0006 -0.0000 -Gri d growth bias 0.0029 -0.0041 -E l evated temperature 0.0004 0.0013 0.000 4 0.0013 Total Rack U p (~k = RSS) 0.0081 0.0227 0.0092 0.0282 Sum of Bias and Uncertainties 0.0308 0.0374
* Catego r y 5 ce ll s are loc a ted on the outer t wo rows of R egion 2 so ecce ntr icity in an infin ite mod e l i s not relevant.
The eccentricity i s part of the full pool model and therefore n o bia s i s app li ed. Fresh 5.0 w/o fue l with 64 IFBA (and any burned fuel) can a l so be stored in a c h eckerboard pattern of assemb li es a nd water holes in Region 2 and can be stored in the 3-o u t-of-4 area of Region 2 if it contain s a contro l rod. The rack up of uncertainties that app li es to fresh 5 .0 w/o fuel in R eg ion 2 is s hown in Table 7.5. Table 7.5: Total Bias and Uncertainty for Fresh Fuel in Region 2 Component Bias Uncertainty Validation (cri tic a l experime nt s) 0.0024 0.0035 Manufacturing Tolerances 0.00 40 F uel enric hment -Monte Carlo statistics 0.0006 E le vated temperature 0.0004 0.0013 Total Rack Up (~k = RSS) 0.0028 0.0055 Sum of Bias and Uncertainties 0.0083 N ET-28091-0003-01, Revi s ion 0 106 None of the above to l erance calculations were performed under borated conditions.
The reason is because t h e borated condition has significant margi n. The boron di l ution analysis of record shows that a dilution down to 786 ppm is not credib l e [52]. However , the mini m um ppm selected for the borated analysis is 700 ppm. Any small increase in the tolerance uncertainties would be covered by this 86 ppm margin in addition to the l arge margin from 0.9 5 reported in Section 8.5. 7.6 Interface Uncertainty Treatment When analyzing a fu ll pool , the calculated k e ff wi ll come from the most reactive area of the SFP. However, when the uncertainty is not t h e same in all areas , the analysis may not correctly find the most l imiting k. In order to address this concern, the bumup of the high burnup regions are adjusted down to account for t h e difference i n the uncerta in ty. Tables 7.3 and 7.4 show the uncertainties for t h e 3-out-of-4 and the 4-out-of-4 areas i n Regions 1 and 2. For Category 3 fuel to match the bias and uncerta i nty of Category 2 fuel, a burnup reduction is required to matc h the 0.0048 6k difference in bias and uncertainty. The 0.0047 6k is the sum of0.0025 6k between Category 2 and Category 3 fuel (see bottom of Table 7.3) plus the 0.0022 6k eccentric i ty effect for Category 2 fuel which wi ll be included in the full poo l model. To account for thi s additional uncertainty , the burnup for Category 3 fuel is decreased by 0.8 GWd/T in the full pool calculations (the reactivity due to burnup at 28.5 GWd/T is 0.6% in kefffor every 1 GWd/T burnup ). Similarly for analysis of Region 2, the Category 5 fuel bumup must be decreased to match the reactivity d i fference in the bias and uncertainty between Category 4 and Category 5 fuel. From Table 7.4 this difference is 0.0066 6k. To account for this additional uncertainty , the burnup for Category 5 fuel is decreased by 1.3 GWd/T in the full pool calculations (t h e reactivity change due to burnup at 60.5 GWd ff is 0.51 % in k e ff for every 1 GWdff bumup). NET-2809 1-0003-01, Revision 0 107 8 Results With the biases and uncertainties determined, the minimum loading requirements can be ca l cu l ated. These minimum loading requirements meet the 10CFR50.68 requirements. Specifically , k 95;9 5 must be less than 1.0 with no soluble boron credit and less than 0.95 with credit for soluble boron. For this analysis, these limits are met while maintaining about a 1 % margin in k. It has been demonstrated that for all unborated cases k 95;95 is less than 0.99 and for the borated cases k 95;95 is less than 0.94 after adding biases and uncertainties. 8. 1 Temperature Effects The criticality analysis must cover the full range of temperatures allowed in the SFP. Rather than perform the criticality ana l ysis at a reference temperature and add a bias , the criticality analysis is perfonned at the most limiting temperatures.
Table 8.1 summarizes the Region 1 and 2 (3-out-of-4 area) calculations at 12 different temperatures (4, 10 , 20, 30 , 40, 50, 60, 70 , 80 , 90 , 95 , and 99 &deg;C). From the SCALE validation , there is a temperature bias of 0.0000086 for each &deg;C above 20 &deg;C. For example, the bias at 60 &deg;C is 0.00034, while the bias at 70 &deg;C is 0.00043 , and the appropriate temperature bias is added to the calculated k etr values. The results demonstrate that the bias-corrected reactivity is largest at 70 &deg;C for Region 2 and at 99 &deg;C for Region 1 under unborated conditions.
Except for Table 8.1 and the temperature accident, a ll ca l culated k e tr va l ues for Region 1 are performed at 180 &deg;F (82 &deg;C), which corresponds to the SFP design basis maximum temperature for the IP2 SFP. In developing the loading curves for Region 2 , all of the calculated k etr values are performed at 70 &deg; C which is the most reactive temperature.
NET-2809 1-0003-01, Revision 0 108 Ta bl e 8.1: Ca lcu lated keff as a Funct io n of Tem p erat u re Te mp e r a tu re D e n s i ty R eg ion 1 R egio n 1 R egio n 2 R eg i o n 2 (OC) (glee) ealc. k a dj. k ealc. k ad j. k 4 1.0000 0.96386 0.96386 0.95569 0.95569 10 0.9997 0.96388 0.96388 0.95575 0.95575 2 0 0.9982 0.96372 0.96372 0.95545 0.95545 30 0.9957 0.96508 0.965 17 0.95582 0.95591 4 0 0.9922 0.96628 0.966 46 0.95595 0.95612 5 0 0.9880 0.96745 0.96771 0.95610 0.95636 6 0 0.9832 0.968 38 0.96873 0.95606 0.95641 70 0.9778 0.9693 1 0.9697 4 0.95599 0.95642 80 0.9718 0.97000 0.97 052 0.95577 0.95629 90 0.9653 0.97089 0.97149 0.95564 0.95624 95 0.9619 0.97123 0.97187 0.95552 0.95616 99 0.9591 0.97156 0.97224 0.95544 0.95612 8.2 Region 1 Fuel Categories 1 and 2 Using the 2x2 mode l described in Section 6 with the a tom densities de velo ped in Section 5 and with the bias and uncertainties established in Section 7, the l oading requirements for Region 1 fuel Categories 1 and 2 are determined. Table 8.2 shows the fuel requirements and the calculated k -for a checkerboard arrangement of fuel in Region 1 (Category 1 cells) and a three out of four arrangement of fuel in Region 1 (Category 2 cells). Category 1 fuel is designed to require 64 IFBA rods in 5.0 w/o fuel assemblies.
To provide flexibility in fue l design , the number of IFBA for fresh fue l can be reduced for lower enrichments
*. The number of minimum IFBA for fuel less than or equal to 5.0 , 4.5 , 4.0, 3.5, and 3.0 w/o is 64, 48 , 32, 16 , and 0 , respective l y. The k e fffor these cases are all les s than the k e ffwith 64 IFBA 5.0 w/o. No credit for burnup is taken. For Category 2 fuel , the fuel must be burned at l east 21 GW d/T and the maximum enrichment is 5.0 w/o. The Category 2 bumup requirement of 21 GW d/T is based on Batch Z fuel having an 8 inc h 4.0 w/o U-235 axial blanket where the relative burnup distribution is from Tab l e 6.3. An axia l blanket that is l ess than 4.0 w/o U-235 or more than 8 inches long is bounded by this ana l ysis, but a shorter or higher enriched blanket is not.
* The IFBA requirement is only for Batch Z. NET-2809 1-0003-01, Revision 0 109 Table 8.2: Confirmation of Region 1 Requirements for Categor y 1 and 2 Fuel Fuel Catee.or v 1 2 Arrane.ement 2-out-of-4 3-out-of-4 Maximum Enrichment (w/o) 5.0 5.0 Minimum Burnup (GWdrf) 0 21 Minimum IFB A Rod s 64 -Calculated k 0.8548 0.9686 Bias and U ncertain ty 0.0127 0.0217 k 9S/95 0.8675 0.9903* For burn ed fuel, no c r e dit is take n for IFBA or a n y in sert in the guide tubes , w ith th e excep ti o n of full l e ngth RCCAs in designated areas. Ca lcul ations show th at 5.0 w/o fuel with 64 IFBA rods h as l ower reactivity at a ll bumups compared to th e BOC eq uilibriu m Xe val u e. Table 8.3 s ho ws the va lu es of kerrin t he core geometry as a function of bumup for fue l w ith various IFBA l oadings. T h e orange s h a d e d bl ocks are the bumups where kerr ha s increased over its initial eq uilib ri um va lu e. The 64 IFBA case is a lw ays l ess than the initial kerr(w ith equilibri um Xe) of 1.2109. Table 8.3: Change in k ett with Burn up and number oflFBA Rods (Analysis performed at core co ndi tions for 5.0 w/o fuel) N umber of IFBA Rods+ 0 32 64 80 116 Burnup (GWd/T}. Calculated keff at core conditions 0.15 1.2989 1.2526 1.2109 1.1924 1.1 523 0.50 1.2925 1.2489 1.2086 1.1903 1.151 3 1.00 1.2855 1.24 41 1.2064 1.1890 1.1 526 1.50 1.280 1 1.2417 1.2069 1.1905 1.1 5 5 9 2.00 1.2757 1.24 03 1.2068 1.1913 1.1 593 3.00 1.268 4 1.2360 1.2062 1.1930 1.1 632 4.00 1.252 1 1.2249 1.1 992 1.1874 1.1 6 1 0 5.00 1.2423 1.2 1 84 1.1963 1.1848 1.1 625 6.00 1.23 1 9 1.2 11 7 1.1 929 1.1827 1.1 631 8.00 1.2 1 67 1.2013 1.1 850 1.1782 1.1 609 10.00 1.1974 1.1 862 1.1 75 0 1.1693 1.15 63
* The infinite k err exceeds 0.99 but the ac tu a l finite k err for thi s r eg ion i s l ess th an 0.99 (0.9881). NET-2809 1-0003-01, Re v ision 0 110 Reactivity decreases with increasing category number. Therefore, fuel from any higher numbered category can be placed in any location that allows for a lower numbered category.
For examp l e, a fuel assembly categorized as Category 5 can be placed anywhere in the SFP. However, a cell in the SFP that requires Category 5 fuel may not contain a lower category fuel assembly. All of the historical fuel through Batch X of IP2 and Batch AA of IP3 has been categorized.
 
===8.3 Region===
2 Category 4 Batch Grouping Z -Current and Future Fuel The minimum burnup requirements (loading curve) for Category 4 fuel for Batch Grouping Z (current design and future fuel assemblies) are presented in Table 8.4. The SFP cells where Category 4 (or above) fuel is required is shown on Figure 1.1 as the green shaded cells in Region 2. The other batch groupings are analyzed separately, and the results are presented in Section 8.6. Table 8.4: Minimum Burnup Requirements (GWd/T) for Category 4 Batch Grouping Z Enrichment
* Cooling Time (years) PF=l.2 (w/o) o t 1 2 5 10 15 25 t 4.2 40.27 39.69 38.92 37.23 35.13 33.75 32.20 4.6 44.27 43.60 42.83 40.9 9 38.71 37.25 35.52 5.0 48.19 47.52 46.61 44.67 42.30 40.71 38.85 Enrichment
* Cooling Time (years) PF=0.80 (w/o) ot 1 2 5 10 15 25t 4.2 38.67 38.11 37.48 36.02 34.12 32.96 31.59 4.6 42.60 41.97 41.31 39.70 37.72 36.45 34.90 5.0 46.52 45.8 4 45.16 43.39 41.14 39.77 38.05 Table 8.4 provides the burnup requirement in GWd/T as a function of initial U-235 enrichment and cooling time for two different peaking factors. For each assembly, the peaking factor i s known, and the
* The enrichment to be used is the enrichment of the center section between the blanket material.
t O years coo lin g is actua ll y 72 hours. This i s the cooling time that maximize s k. t Fue l coo l ed to more than 25 years must use the 25 year burnup requirement.
NET-28091-0003-01 , Revision 0 111 bumup requirement for that assembly can be interpo la ted bet w een 0.8 0 and 1.20. O ver 95% of the fuel inventory ha s peaking factors between 0.80 and 1.2 0. Extrapolation abo ve 1.20 and below 0.8 is also acceptable because this has been shown to be conservative (see Section 6.4). After adjusting for an assembly's peaking factor, if an assembly fails the lo ad in g curve, it can b e stored anywhere in Region I as long as the bumup is greater than 28.5 GWd/T. Table 8.4 can be linearl y interpolated to find the re quired bumup at any enrichment
/cooling time combination but it is recommended that the curve fit be used instead, as described in Section 8.2.1. As discussed lat er, the bumup requir e ments are adjusted if the assemb l y contained a h a fnium insert or has any fuel pins remo ve d. Fresh assemblie s with at least 64 IFBA rods and an inserted control rod are Category 4. Analysis of an infinite system of Region 2 cells in a 3-out-of-4 arrangement using 5 .0 w/o fresh fuel with 64 IFBA rods and a control rod in every assembly produced a k e ff of 0.96 03. To be conservative , the control rod a tom densities were reduced by 10%. The bias and un certainty for this case is 0.0083 (see Tab le 7.5), so k 9s 1 9s is 0.9686, which is well below 0.99. To provide flexibility in fuel de sign, the number of IFBA for fresh fuel can be reduced for lower enrichments.
The number of minimum IFBA for fuel less than or equal to 5.0 , 4.5 , 4.0 , 3.5, and 3.0 w/o is 64 , 48 , 32, 16 , and 0 , respectively. The k e ff for these cases are all l ess than the k eff with 64 IFBA. As discussed earlier, the reactivity of fuel with 64 IFBA decreases with bumup so a fresh assemb l y is more limiting than the same assembly havin g a sma ll amount ofbumup. NET-28091-0003-01 , Revision 0 112 
 
====8.3.1 Curve====
Fit The data points of Table 8.4 have b een fit with a nine parameter curve having the following form: Minimum Bumup Requirement w h ere E = U-235 initial enrichment (w/o) CT = cooling time (years) a 1 -a 9 = fitting coefficients No extrapolation is a ll owed , so fuel at enrichments l ess than 4.2 w/o must use 4.2 for the enric hm ent , and fuel cooled more t h an 25 years must use 25 for the cooling time. The cur v e is purposefully conservative in that the minimum burnup requiremen t generated from th e curve is always equa l to or greater than the bumup s ho wn in Tab l e 8.4. The coefficients are s h own in Table 8.5. Severa l curve fits were attempted but this curve fit matched the data with the least amount of conservatism while being we ll b e haved b etween the data points of Table 8.4. A spreadsheet for the fit was created to en s ure the int ermediate points follow the expected behavior.
The exponentia l term in the fit is needed to mimic the physics of radioactive decay. Table 8.5: Curve Fit Coefficients for Category 4 Fuel (Gro up Z -Current a nd Future Fue l) Coefficient PF= o.so* PF= 1.20 a1 15.1405 -6.26824 az -4.8 11 33 5.29367 a3 0.753855 -0.37154 a4 0.121 252 0.129582 as -0.0150991
-0.0204918 a6 0.00 1 27009 0.00 205596 a 1 -16.2293 -0.1 3331 as 14.0159 6.9037 a9 -0.687 054 0.122068
* Only two peaking factors are needed (0.8 and 1.2). Ca l c ul ations show that e x trapo l at i o n belo w 0.8 a nd abo v e 1.2 is conserva ti ve for all ot h er peaking facto r s. NET-2809 1-000 3-01 , Revision 0 113 
 
====8.3.2 Confirmati====
o n Calculations for Category 4 To e n sure that all bumup/enrich m ent/cooling time combinations given in the loadin g curve meet the criticality requirements , each loading curve bumup/enrichment
/coo l ing time point was run in the 2x2 KENO model to verify that each point meets the criticality requirements. The calculated keff va lu es are shown in Table 8.6. Table 8.6: Calculated k cff Values at each Category 4 Batch Z Burnup Point Enrichment Cooling Time (years) PF=l.20 (w/o) 0 1 2 5 10 15 25 4.2 0.9614 0.9611 0.9613 0.9606 0.9608 0.9608 0.9604 4.6 0.9600 0.9599 0.9597 0.9592 0.9594 0.9593 0.9590 5.0 0.95 86 0.9581 0.9585 0.9579 0.9577 0.9576 0.9 573 Enrichment Cooling Time (years) PF=0.80 (w/o) 0 1 2 5 10 15 25 4.2 0.96 12 0.961 3 0.9611 0.9605 0.96 11 0.960 8 0.9602 4.6 0.9599 0.9600 0.9596 0.9591 0.9590 0.95 88 0.9585 5.0 0.9584 0.9585 0.9578 0.9577 0.9584 0.9579 0.9579 The total un certainty i s the bias plus a statistica l combination of a ll of the unc ertainties (see Section 7). These total un certainties are shown in Table 8.7. Table 8.7: Total Bias and Uncertainty at each Category 4 Batch Z Burnup Point Enrichment PF=l.20 w/o 0 1 10 15 25 4.2 0.0278 0.0277 0.0277 0.0274 0.0273 0.0272 4.6 0.0292 0.0 292 0.0 293 0.0292 0.0291 0.0289 0.0288 5.0 0.0307 0.030 8 0.0307 0.0306 0.0306 0.0306 0.0304 Enrichment PF=0.80 w/o 0 1 5 10 15 25 4.2 0.0277 0.0276 0.0276 0.0275 0.0274 0.0273 0.0272 4.6 0.0292 0.0292 0.0293 0.0293 0.0292 0.0290 0.0290 5.0 0.0307 0.0307 0.0308 0.0307 0.0306 0.0306 0.0 305 NET-2809 1-000 3-01, R evis ion 0 114 A ft e r addin g t h e t ot a l un certai nt y t o th e calc ulat e d ke ff va lu es , a ll p oints are l ess th a n 0.99 a s s h ow n in Tab l e 8.8. Table 8.8: k9 s t 9s for each Category 4 Batch Z Burnup Point Enrichment Coolin!?:
Time (vears* PF=l.20 (w/o) 0 1 2 5 10 15 25 4.2 0.9892* 0.9889 0.9 8 90 0.9 882 0.9882 0.988 1 0.9 876 4.6 0.9892 0.9892 0.9 8 90 0.9 88 5 0.988 5 0.9882 0.9 877 5.0 0.9893 0.9 889 0.9 8 9 2 0.9 88 5 0.9883 0.9882 0.9 877 Enrichment Coolin!?:
Time (vears\ PF=0.80 (w/o) 0 1 2 5 10 15 25 4.2 0.9889 0.9 889 0.9887 0.988 0 0.988 4 0.9880 0.98 7 5 4.6 0.989 1 0.9892 0.9889 0.988 4 0.9882 0.9879 0.9875 5.0 0.989 1 0.989 1 0.9886 0.9 883 0.989 0 0.9 886 0.9 885 8.4 Determination of Burnup Requirements for Categories 3 and 5 Ce ll C at egories 3 a nd 5 util ize th e n e ut ron l ea k age a t th e e d ge o f th e SF P in o rd e r t o re m ove th e need fo r o n e out of four wa t er hol es use d fo r ce ll Catego ri es 2 a nd 4. S in ce th e e d ge of th e SF P n e utron l eakage i s u se d , a fu ll po o l m o d e l is r e quir e d. T hi s full p oo l m o d e l al so a ll ows for ca lcu lat i o n o f th e ecce n t ri c it y effect a nd th e i m p act of th e int e r face b e t wee n ce ll Ca te go ri es an d R egio n s. 8.4. 1 Cell Category Layout in Region 2 T h e c on ce p t fo r Re g ion 2 i s to h ave a thr ee o ut of four arra n ge m e nt of fue l w ith t wo o ut e r ro ws of fuel r e qu i rin g a n increase d bu rn u p bu t u s in g t he l ea k age at th e e d ge t o r e du ce th e in c r ease d burnup re quir e m e nt. H oweve r , IP2 h a d a se t o f s p are co ntrol rod s th a t co uld b e u se d t o re du ce t h e numb e r of wa t er hol es s u c h th at o n e l ess cas k wo uld n ee d to b e lo a d e d. T h e c ontrol ro d s in a n assem bl y do n o t h ave as l arge a n egat i ve rea cti v it y as a wa t e r h o l e so th e control rods n ee d e d t o b e t wo o ut of fo ur , rath e r th a n
* T h e va lu es in Tab l e 8.8 d o not a l ways m a t ch th e s um of Ta bl es 8.6 a nd 8.7 due to ro und off , s in ce eac h t a bl e was d eve l o p e d u s in g m o re s i g nifi cant di g it s b efo r e ro undin g fo r th e ta bl e. N E T-2 8 091-000 3-01 , R ev i s i o n 0 115 one out of four. Rather than make a separate fuel category for the cell area with control rods, this cell area is forced to be the same category as the two outer rows of assemblies ( cell Category 5). Because the checkerboard of control rods reduced k err more than the two rows on the outside of the SFP , it is possible t o reduce the amount of water required in the water ho l es. This is the reason for the l ocation of the pink cell s. Determining the location of the pink cell s and the control rods required con s iderable iteration. The control rod locations can be water holes since the negative reactivity of a water hole is greater t han a Category 5 fue l assembly with a control rod. The control rod may not be placed in or removed while the assembly is in the control rod location. The control rod must be inserted into the assembly while the assembly is in a cell not requiring a control rod, and then the assembly with control rod can be mo v ed into position. Likewise, the control rod may be removed only while the assembl y is in a cell not requiring a control rod. It is permissib l e to remove a control rod at a Category 5 cell as long as all adjacent ce ll s (eight ce ll s) are water ho l es since this iso l ates the assembly , but this is not the expected method. Category 5 fuel must be placed on both sides of the Region 1/2 interface except in SFP locations J-31 and H-31 (alternate arrangement s are allowed as discussed in Section 8.5). This eliminates the interaction between the two Regions. Placing Category 4 fuel on the Region 1 side of the interface was tried , but it drew the reactivity to the interface w i th a s li ght increase in k err and, therefore , was rejected.
 
====8.4.2 Additional====
 
Burnup Requirements for Fuel Categories 3 and 5 The objectives in setting the loading requirements for Categories 3 and 5 are fir s t to make Categories 2 and 4 more limiting than Categories 3 and 5 , and second to make them simple. The simplicity is accomplished by making the l oading requirements a constant bumup penalty (independent of enrichment, cooling time and peaking factor) to the fue l Category 2 and 4 requirements. The bumup penalty is the smallest fraction of the bumup when the bumup is highest. Therefore , the bumup penalty is determined using the highest bumup requirement s in Category 2 and 4 so it is conservative for lower NET-28091-0003-01 , Revision 0 116 bumups. To find the bumup penalty, the Category 3 and 5 bumups were changed holding the Category 2 and 4 bumups constant.
Since the design objective k e ff for Region 1 is higher than that for Region 2 ( due to lower bumup requirements), the Region 2 analysis must be performed with the fuel removed from Region 1, so that the reactivity is dominated by Region 2. The Region 1 analysis contains all of the fuel from both regions. Therefore , the Region 2 analysis is perfom1ed first so that the correct bumups in Region 2 are used while doing the Region 1 analysis.
To determine the additional bumup requirement needed for Category 5 fuel , analysis wa s performed for Region 2 where the Category 4 bumup is 49.5 GWd/T and the Category 5 bumup is varied. For these cases the enrichment for both categories is 5.0 w/o. For this model , all of the Category 5 fuel is shifted right to maximize the effect of the Category 4/5 interface.
Figure 8.1 is a plot of the variatio n in k e ff with the Category 5 bumup. As can be seen from Figure 8.1 at lower bumups (such as 54 GWd/T) the Category 5 fuel is determining
: k. By 59 GW d/T, k e ff is controlled mainly by the Category 4 fuel and by 61 GWd/T the Category 5 fuel is no longer affecting the SFP k. Based on this information the increased burnup requirement (burnup penalty) for Category 5 fuel is selected as 11 GWd/T. NET-28091-0003-01 , Revision 0 117 
"" "C ., 0.980 0.975 0.970 0.965 u a 0.960 0.955 0.950 + 54 55 56 57 58 59 60 61 Burnup of Category 5 Fuel (GWd/MTU)
Figure 8.1: Calculated k err as a Function of Category 5 Burn up Using 5.0 w/o Fuel 62 In or d er to confirm that using the burnup penalty d etermined from the highest bumup is co n servative , the analysis was also performed using a l ower bumup for Category 4 fuel. This model uses 4.2 w/o enriche d fuel , 32 GWd/T for the bumup (25 years coo l ed and a 0.8 p eaking factor). F i gure 8.2 s ho ws the results of thi s analy s is. As can be seen from Figure 8.2 , by a burnup of39 GWd/T the burnup of Category 5 n o l onger matters. This wou ld imply a penalty of7 GWd/T which mea n s that using the penalty detennined at the higher bumup (11 GWd/T) i s indeed conservati v e. NET-28091-0003-01, R ev ision 0 11 8 1.000 ----0.995 ---0.990 0.985 .:..: 0.980 ,:, ... 0.975 :::, ... iii u 0.970 0.965 0.960 F ---0.955 0.950 T T .-,-33 35 37 39 41 Burnup (GWd/MTU)
Figure 8.2: k c rr as a Function of Categor y 5 Burnup Using 4.2 w/o Enriched Fuel The sa m e p roce ss i s r e p eated for R eg i o n 1 w h ere t h e cate g ories of concern are Categor y 2 a nd Category 3. H owe v er , th e bu rn u p r e quir e m e n t for th e 3-o u t-of-4 a r ea of R egio n 1 (Category
: 2) i s a fixe d 2 1 GW d/T ind e p en d e nt of e nri c hm e nt , co o lin g t im e , o r p ea kin g fac to r. F i gure 8.3 s h ows t h e r es ult s of th e ana l ys i s t o d ete rmin e th e burnu p p e n a lt y. A s ca n b e see n fro m F i gure 8.3 , Catego r y 3 b umup greater th a n 27.7 GW d/T h a s a very sma ll i mp act o n the full poo l calc ul a t e d k. A 7.5 GWd/T a d ditio n a l bumup r e qu ire m e nt fo r Catego r y 3 i s se l ecte d. T hi s makes th e Category 3 minimum burnup requirement 28.5 GWd/T. N E T-2 8 091-000 3-01 , Re v i s ion 0 11 9 0.984 0.982 0.980 .:ii:: 0.978 "C GI .... "' 0.976 :i I.I iii 0.974 u 0.972 0.970 0.968 25 25.5 26 26.5 27 27.5 28 28.5 29 Category 3 Burnup (GWd/MTU)
Figure 8.3: Calculated ketr as a Function of Category 3 Burn up Using 5.0 w/o Fuel 8.4.3 Confirmation of k9s 1 9s for Full Pool (includes Category 3 and 5) Since the total combined uncertainty is hi gher in Region 2 , the design objective k e rr for Region 1 is higher than Region 2. This mean s that a full pool model will produce a k e rr value driven by the higher reacti vity fuel in Region 1 and not yield any infonnation for Region 2. Before determinin g the Region 1 k e rr, it is desirable to confinn the Region 2 burnup penalty for Category 5. For these cases , the full pool is modeled with water holes in Region 1 so that the reactivity is driven by Region 2 fuel. The burnup for Category 5 fuel is reduced consistent with the difference in the bias and uncertainty bet ween C ategory 4 a nd Category 5 fuel. As determined in Section 7 .6, a burnup reduction of 1.3 GW d/T for the Category 5 fuel is required. Calculations were performed for 5.0 w/o fuel at the lo ading curve (for curre nt fuel) at a peaking factor of 1.2 for zero , two and 25 years of coo lin g time. The Category 4 burnups are taken from Table 8.4. The Category 5 burnup s are 11-1.3 = 9.7 GWd/T hi g her. Table 8.9 s hows calcula t ed k's and the k 9s 195's for three 5.0 w/o cases at a peaking factor of 1.20. The bias and uncertainty used for Table 8.9 come from Table 8.7. Table 8.9 shows that the loading criteria of Category 4 plus 11 GWd/T for NET-2809 1-000 3-01 , Revision 0 120 Catego r y 5 m eets t h e k ( <0.9 9) crite ri a. As expe ct e d , th e r e is s li g htl y m ore m a r gi n for t h e 2 yea r c o o l ed case t h a n th e 72 h o u r co ol e d case a nd eve n mo r e m a r g in for t h e 2 5 year coo l e d case. This i s b e c a u se t h e burnup r e quir e m ent d ec r eases w i t h co olin g t ime a nd s o t h e fixe d burnu p pe n a l ty of 11 GW d/T i s a l a r ger fr act ion o f th e bu rn up r e quir eme n t. It i s co n ce i va bl e th a t a h ig h e r p ea kin g fac t o r mi g h t res ult in l ess m argi n. A n addit io n a l c a lcul at i o n was p e r forme d u s in g a p eaki n g fac t or of 1.3. Th e hi ghes t p ea kin g fac t or fo r fu e l in th e IP 2 SF P m eeti n g th e Ca t ego r y 4 burnu p re quir e m e nt s is 1.272. The ca lcul a t e d k 9 5 1 9 5 for a p ea kin g fac t or of 1.3 i s 0.989 7. Table 8.9: Region 2 Models at Loading Cur v e (Cat 5 i s Cat 4 plus 11 GWd/T) Categor y 4 Category 5 Burnup Burnup in Peaking Cooling Calculated Si g ma Bias and k 9 S/95 (GWd/T) Model Factor Time k U ncertaint y. (GWd/T) 48.1 9 57.89 1.2 72 ho ur s 0.9584 0.00005 0.0307 0.9891 46.6 1 56.3 1 1.2 2 yea r s 0.9578 0.00006 0.0 307 0.9885 38.8 5 48.55 1.2 25 y ea r s 0.956 1 0.00006 0.0304 0.9865 48.6 1 58.3 1 1.3 72 h o ur s 0.9588 0.00006 0.0 309 0.9897 T h e mod e l u sed fo r th e R egio n 2 a n a l ys i s h as th e Ca t ego r y 5 fu e l eccen t rica ll y pl a c ed o n th e ri g h t h a nd s id e of eac h ce ll (th e r eac ti vity effe ct of mov in g th e Categ o ry 5 fu e l t o th e ri g ht was ca lcul a ted a nd i s wo rth 0.000 8 ~k). With thi s m ode l , w hi c h pic k s up ecce nt ric it y as we ll as a n y int erface effec t s, th e m ax imum k 9s 1 9s is m e t w i t h m o r e t h a n 1 % m a r gi n t o th e r egu l atory limit. U p o n c onfirm at i o n of th e R egio n 2 l oa d i n g re quir e m e n ts, t h e a n a l ys i s of R eg ion 1 ca n p ro cee d b y addin g Ca te go r y 2 a nd 3 fu e l in R eg i o n 1 o f t h e m o d e l. In orde r t o co mp ensate fo r th e d iffe r e n ce in bi as a nd un ce rt a int y b etwee n Ca t egory 2 a nd Ca t egory 3 , th e bu rnup of th e Category 3 fu e l i s re du ce d. In S ec t io n 7.6 , th e di ffe r e nc e in un cer tainti es for 7.5 GWd/T i s 0.8 GWd/T. The bumup r e quir e m e nt for C at ego r y 2 i s 2 1 G Wd/T a nd th e b u rnup r e qu i r e m e nt for Ca t ego r y 3 i s 28.5 G Wd/T. R e du c in g 28.5 GW d/T b y 0.8 m ea ns th e modeled burnup for Category 3 is 27.7 GWd/T. C on seq u e ntl y , th e k 9s 1 9s N ET-28 091-000 3-0 1 , R ev i s i o n 0 1 2 1 for the SFP is the calculated k eff in the full pool model plus the bias and uncertainty for Category 2 (21 GWd/T). The Region 1 eccentricity can be in two forms; in the center of Category 2 cells or at the boundary of Category 2 and 3 cells. The highest keffis when the eccentricity is in the center of the Category 2 cell area of Region 1. Table 8.10 shows the results of the analysis of different eccentric options. The analysis uses t he primary arrangement of Region 1 ( shown on Figure 1.1) with all fuel at 5 .0 w/o enrichment and the burnups of Categories 2 through 5 fuel are 21 , 27.7 , 48.19, and 57.89 GWd/T. For all but the last case on Table 8.10, all of the Category 3 and 5 fuel assemblies are centered ( except the case where the eccentric positioning is at the Category 2/3 boundary where only those assemblies with the eccentric grouping are moved in). Table 8.10: Eccentric Options for Region 1 Case k Sigma .dk All Region 1 Assemblies centered in cells 0.9665 0.00007 Plus Category 3 and 5 fuel in Region 1 moved 0.9681 0.00007 0.0016 toward the center Also Plus 16 Eccentric at Category 2/3 Interface 0.9686 0.00006 0.0021 16 Eccentric in the middle of Category 2 plus 0.9687 0.00007 0.0022 Category 3 and 5 moved in toward Category 2 From Table 7.3 , the bias and uncertainty is 0.0194 Llk with the eccentricity bias removed. The full pool model contain s the eccentricity and interface effects. The final k9s 1 9s for the Figure 1.1 arrangement of Region 1 is 0.9687 + 0.0194 = 0.9881. Thus , for the primary arrangement of Region 1 the target 0.99 is satisfied, leaving more than 1 % margin to the 10CFR50.68 limit. The peaking factors used for Category 2 and Category 3 fuel are 0.9 and 1.2, respectively.
These are based on an expected cycle length for the final cycle of 23.8 GWd/T. There are no assemblies currently in the SFP that are Category 2. Until the final cycle, it is expected that low burned fuel will be returned to the core for more burnup. If the final cycle is 23.8 GWd/T , then the highest peaking factor that matches NET-28091-0003-01, Revision 0 122 t he minimum burnup requirements for Category 2 (21 GWd/T) is 21/23.8 = 0.88. Assemblies in that cycle will certainly have higher peaking factor s, but they would then exceed the minimum burnup r equirement.
For example, if an assembly had a peaking factor of 1.4 and the cycle length is 23.8 GWd/T, its burnup would be 23.8*1.4 = 33.3 GWd/T which would greatly exceed the Category 2 burnup requirement.
The reactivity effect of additional bumup is greater than the effect of higher t emperatures during depletion that is caused by a higher peaking factor. For Category 3 , the highest peaking factor that matches the minimum burnup requirements for 28.5 GWD/T and a cycle length of 23.8 GWd/T can be calculated and is 28.5/23.8 = 1.2. The calculated k using 0.9 and 1.2 peaking factors i s conservative for any final cycle burnup greater than 23.8 GWd/T. The final cycle ma y be cut short so additional calculations have been done for different cycle lengths. If the final cycle were only 20.3 GW d/T then the peaking factors would be 21/20.3 = 1.03 and 28.5/20.3 = 1.40. The calculation using peaking factors of 1.1 and 1.4 for Category 2 and 3 , respectively, resulted in a k 9s 1 9s of 0.9895. With cycle bumups of less than 20.3, no new Category 3 assemblies would be produced since the maximum assembly peaking factor is 1.4 at any point in the cycle (note that this is not bumup averaged).
The most reactive assemblies in the SFP, except for four assemblies , have a bumup of 40 GWd/T. For cycles less than 20.3 GWd/T , the maximum peaking factor of 1.4 is used for Category 2 fuel , and the Category 3 fuel i s modeled as 40 GWd/T (5.0 w/o fuel) with a peaking factor of 1.4. The calculated k 95 1 95 for this case is 0.9888. In conclusion , for all final cycle lengths , the calculated k 9 5 1 9 5 is less than the regulatory limit by more than 1 % ink. Although it is not expected that there would be multiple short cycles leaving more reacti v e fuel in Region 1 , a case was analyzed where the peaking factor of 1.4 is used for both Category 2 and 3 fuel using the minimum burnup requirements and 5.0 w/o fuel. This calculated k 9s 1 9 5 is only 0.9904. This exceeds the design objective but still provides 0.96% in k margin to the regulatory limit. These cases are summarized on Table 8.11. NET-28091-0003-01 , Revision 0 123 Table 8.11: Maxim u m Full Poo l k9s t 9s assuming Various Cycle Lengths Cycle Length Category 2 Category 3 Calculated Sigma Bias and k9S/95 (GWd/f) Peakin11:
Factor Peakin11:
Factor k Uncertainty
>23.8 0.9 1.2 0.9687 0.00007 0.0194 0.9881 20.3 -23.8 1.1 1.4 0.9701 0.00006 0.0194 0.9895 <20.3 1.4 1.4 but 40 GWd!f 0.9694 0.00007 0.0194 0.9888 Multiple Short Cycles 1.4 1.4 0.97 10 0.00007 0.0194 0.9904 8.5 Alternate Arrangements for Region 1 A checkerboard arrangement of Category 1 fuel i s much less reacti ve than the base arrangement of fuel in Region 1. The calculated k err of unburned 5.0 w/o fuel wit h 64 IFBA rods checkerboarded in Re gio n 1 is 0.8 548. Since this k is so low , it is permitted to replace an area of Region 1 with a checkerboard arrangement of Category 1 cells. In order to not create an increase in k due to interface s, there are two rules for creating the area for Category 1 storage. 1. Each Category 1 cell must be face adjacent with at least three water holes. 2. Each Category 2 cell may not ha ve more than one face adjacent to a Category 1 cell. Given these constraints, three additional arrangements of fuel in Region 1 were analyzed to confirm the reduction ink with Category 1 checkerboards present. The first additional arrangement represents the expected arrangement in Region 1 prior to lo a ding a new cycle. For this arrangeme nt , fresh fuel is generally in the new fuel vault but some Category 1 fuel is needed to be in the SFP. This arrangement will be called the " Refueling Arrangement." The seco nd additiona l arrangement covers the case where all of the Category 2 cells are removed but the Category 3 and 5 cells are still u se d. This arrangement will be called the "No Cat 2" arrangement.
The final additional arrangement maximizes the number of Category 1 cells in Region 1. This arrangement i s called "Max Cat l." Figures 8.4 through 8.6 show these arrangements of Region 1. Figure 8.7 shows an arrangement of Region 1 that is not limiting but is useful to illustrate the Category 1 rule s. The term "c heckerboard" is NET-2809 1-0003-01, Revision 0 124 used to d escribe the regu l ar array bu t t h e key req ui re m ent is t h e fi r s t rule req u iri n g at l east 3 face adjace n t water ho l es. This re qui re m ent i s illu stra t e d in Figu r e 8.7 showing o nl y one , two , or thr e e Category l cells. T h e arrangeme n t of other Ca t egories that are n ot replaced b y Category l m u st not cha n ge even with t h e pr ese n ce of a n y nu mber of Category I area s. A l t h ou g h di fferent arrange m e n ts of Ca t egory 2 cell s are pos s i ble that w o ul d not v io l ate the critic a li ty r e quir e m e nts , the rule s to a ll o w th es e types of rearr a ngement s w ou l d b e complex a n d pron e to error and th e r e fore ar e not a llow e d. In order to fairl y compare th e arrange m e nt of the base case and t h e a ddi t i onal arra n ge m ents , t h es e a n a l yses we r e perfonne d with a ll of t h e fuel c ente r e d i n i t s ce ll. Ta bl e 8.12 prov id es the calc ul ate d k e ff v a lu es for each arrangement.
2 3 4 5 6 7 8 9 W ll ll U U ll H V U U W li ll B M ll U V U ll H G F E D C B F i g u re 8.4: R e fu e lin g A rr a n ge m e n t Figure 8.5: No Cat 2 Arrangement NET-28 091-0003-0 1 , R ev i sio n 0 125 H G F E D C B H G F E D C B 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 Figure 8.6: Max Cat l Arrangement 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 ~----1------------------t= Lj -~ -~ ---*------,......,.., -------A X Figure 8.7: Example Odd Arrangement Table 8.12: Dependence of kcff on the Region 1 Arrangement Arrangement Description k Sigma ~k from Number Reference 1 Referen ce -No Category 1 0.9665 0.00007 2 Refu eling Arrangement 0.9654 0.00006 -0.0011 3 No Category 2 Arrangement 0.9651 0.00006 -0.0014 4 Maximum Ca tegory 1 Arra n ge m ent 0.9584 0.00006 -0.0081 5 Odd Arrangement 0.9655 0.00005 -0.0010 NET-28091-0003-01, Revision 0 126 
 
===8.6 Calculations===
 
for Discharged Fuel (IP2 A-X and IP3 A-AA) As noted in Section 5, depletion conditions vary for each batch. Bounding depletion conditions are used for each batch grouping or individual assembly.
Minimum burnup requirements are then determined for each batch grouping or individual assembly.
It is desirable to allow as many assemblies as possible to be stored in Region 2 due to the limited size of Region 1. The minimum burnup requirement of Region 2 is for the 3-out-of-4 area. If an assembly burnup exceeds this minimum requirement, then it is classified as Category 4 fuel. If the burnup exceeds the minimum requirement by 11 GWd/T or more, then it is classified as Category 5 fuel. There are currently a sufficient number of assemblies that meet the Category 5 criteria so that all cells requiring Category 5 can be filled. If the burnup of an assembly is less than the general Category 4 requirement for the batch grouping , its classification is further studied. Many of these cases make the requirements for Category 4 after further analysis (see Table 8.24). A few remaining assemblies do not have more than 28.5 GWd/T burnup and further analysis is performed to show that they meet the reactivity requirement for Category 3 (see Table 8.23). Tables 8.13 to 8.22 are the Category 4 loading requirements (the 3-out-of-4 area of Region 2) for each batch or batch grouping as a function of initial enrichment , cooling time, and peaking factor. NET-2809 1-0003-01, Rev i sion 0 127 
' Table 8.13: Batch A-D Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Cooling Time (years) PF=l.20 (w/o) 10 25 45 2.0 7.90 7.3 1 7.04 2.2 10.56 9.69 9.35 2.6 16.6 8 15.29 14.64 3.0 22.78 2 1.12 20.32 3.4 27.82 2 5.90 25.0 2 3.8 32.37 30.63 29.69 Enrichment Coolin!! Time (years) PF=0.80 (w/o) 10 25 45 2.0 7.62 7.09 6.85 2.2 10.18 9.44 9.15 2.6 16.2 0 14.93 14.3 7 3.0 22.27 20.74 2 0.00 3.4 27.25 25.51 24.72 3.8 31.84 30.24 29.34 Table 8.14: Batch E-F Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Cooling Time (years) PF=l.20 (w/o) 10 25 45 3.0 21.45 20.02 1 9.35 3.4 26.48 24.87 24.16 3.8 31.31 29.66 28.76 Enrichment Coolin!! Time (years) PF=0.80 (w/o) 10 25 45 3.0 21.05 19.76 19.14 3.4 26.03 24.6 1 23.96 3.8 30.93 29.36 28.53 NET-2809 1-000 3-01, R ev ision 0 1 28 Table 8.15: Batch G-L Mini mum Burnup Requirements (GWd/T) for Categor y 4 Enrichment Coolin PF=l.20 (w/o) 10 45 3.0 24.39 22.8 1 22.11 3.4 29.42 27.43 26.52 3.8 33.64 31.9 0 3 1.0 4 Enrichment Coolin PF=0.80 (w/o) 10 45 3.0 23.97 22.54 2 1.86 3.4 28.91 27.1 2 26.26 3.8 33.19 31.56 30.78 Table 8.16: Batch M-P Min imum Burnup Requirements (GWd/T) for Categor y 4 Enrichment Coolin Time ears PF=l.20 (w/o) 10 15 25 45 3.4 2 9.85 28.92 27.78 26.78 3.8 3 3.69 32.87 3 1.83 30.92 4.2 3 9.28 38.59 37.62 36.34 4.6 41 .3 1 40.56 39.66 38.91 5.0 4 5.42 43.96 42.16 41.25 Enrichment ears PF=0.80 (w/o) 10 25 45 3.4 2 8.92 28.06 27.05 26.17 3.8 3 3.40 32.64 3 1.68 30.86 4.2 3 8.85 38.30 37.17 35.99 4.6 41 .43 40.78 40.01 39.37 5.0 4 4.58 43.27 4 1.79 4 1.04 NET-2809 1-000 3-01, R ev i sion 0 129 Table 8.17: Batch Q-S Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Coolin2 Time (years) PF=l.20 (w/o) 10 15 25 3.8 2 9.75 28.6 5 27.29 4.2 33.63 3 2.43 3 0.96 4.6 37.5 0 36.2 0 34.58 5.0 4 1.3 1 39.88 38.1 4 Enrichment Coolin2 Time (years) PF=0.80 (w/o) 10 15 25 3.8 28.93 27.9 4 26.74 4.2 32.83 3 1.75 3 0.4 2 4.6 36.67 3 5.49 34.0 3 5.0 40.4 1 39.11 37.53 Table 8.18: Batch T-V Minimum Burnup Requirements (GWd/T) for Categor y 4 E nrichment Coolin2 Time (years) PF=l.20 (w/o) 5 10 15 25 3.8 32.40 3 0.65 29.44 28.07 4.2 36.4 5 3 4.51 33.1 9 3 1.63 4.6 40.41 3 8.27 36.86 35.1 8 5.0 4 4.24 41.94 40.42 38.64 E nrichment Coolin2 Time (years) PF=0.80 (w/o) 5 10 15 25 3.8 3 1.41 2 9.80 28.8 1 27.58 4.2 3 5.3 4 33.59 32.47 3 1.1 1 4.6 39.2 1 37.3 6 36.1 1 3 4.60 5.0 4 2.98 40.87 39.55 3 7.90 N ET-2 8 091-000 3-01 , R ev i s i o n 0 1 30 Table 8.19: Batch W Minimum Burnup Requirements (GWd/T) for Categor y 4 Enrichment Cooling Time (years) PF=l.2 (w/o) 2 5 10 15 25 3.8 33.72 32.2 5 3 0.43 29.2 8 27.92 4.2 37.86 3 6.2 1 34.23 32.8 9 3 1.4 0 4.6 41.8 4 40.07 37.90 36.53 34.8 7 5.0 45.7 5 43.82 4 1.5 2 4 0.04 38.2 7 Enrichment Cooling Time (y ears) PF=0.80 (w/o) 2 5 10 15 25 3.8 32.49 3 1.23 29.66 28.66 27.4 3 4.2 36.52 3 5.1 1 33.35 32.27 3 0.88 4.6 40.42 3 8.86 36.93 35.7 1 34.23 5.0 44.2 0 42.52 40.50 39.1 9 37.57 Table 8.20: Batch X Minimum Burnup Requirements (GWd/T) for Categor y 4 Enrichment Coolin!?:
Time (years) PF=l.2 (w/o) 2 5 10 15 25 3.8 3 3.73 32.3 0 30.48 29.29 27.96 4.2 37.8 5 36.2 1 34.23 32.94 3 1.4 1 4.6 41.8 1 40.0 8 37.84 36.50 34.8 6 5.0 45.69 43.82 41.4 3 39.94 38.15 Enrichment Cooling Time (years) PF=0.80 (w/o) 2 5 10 15 25 3.8 32.53 3 1.25 29.59 28.63 27.4 0 4.2 36.53 35.1 0 33.27 32.1 8 30.85 4.6 40.39 3 8.8 1 36.9 1 35.69 34.1 8 5.0 44.16 4 2.4 9 40.4 7 39.1 6 37.56 The fo ll owing loa d ing r e q uire m e nt s a r e for a ll di sc h arge d fue l from IP3 (B atc h es A t hru X). NET-28 091-000 3-0 1 , R ev i s i o n 0 1 3 1 Table 8.21: Batch A-U (IP3) Minimum Burnup Requirem e nt s (GWd/T) for C a te g or y 4 Enrichm e n t Cooling Time (y ear s) P F=l.20 (w/o) 10 25 45 2.2 12.52 11.4 1 10.92 2.6 18.83 17.25 16.52 3.0 24.71 23.0 1 22.27 3.4 29.79 27.70 26.71 3.8 33.94 32.08 31.21 Enrichm e nt Cooling Time (year s) P F=0.80 (w/o) 10 25 45 2.2 12.09 11.10 10.67 2.6 18.28 16.87 16.23 3.0 24.17 22.66 21.97 3.4 29.16 27.28 26.37 3.8 33.39 31.72 30.90 Table 8.2 2: Batch V-X (I P3) M inimum Burnup Requir e m e nts (GW d/T) for C ate g o ry 4 E nrichment C ooling Time (y ear s) P F=l.20 (w/o) 10 25 45 3.4 28.28 27.24 25.92 3.8 31.90 30.69 29.26 4.2 35.13 33.75 32.19 E nrich me nt C ooling T im e (ye ar s) P F=0.80 (w/o) 10 25 45 3.4 27.57 26.56 25.43 3.8 31.04 29.96 28.7 1 4.2 34.1 1 32.95 31.58 The IP3 fuel after Batch U u ses the Batch Z dep l etion parameters with two additional enrichment points at 3.4 and 3.8 w/o. For these l ower enrichment points, the fue l is modeled as full lengt h which is conservat i ve for blanketed fue l (Batches V , W , and X h ave six inc h natural urani u m blankets w h i l e Y and AA have six inch 2.6 w/o bl ankets). T h e b atches Y and AA (an d t h e h igher enric h ed fuel from B a t ch X) use the Z l oa d ing curve w h ich was base d on mode lin g a n 8-inch ax i a l b l anket wit h a blanket enric h ment of 4.0 w/o. If the blanket enrichment were the same, modeling a 6-i n ch blanket as 8 inches wo ul d be conservative.
Howeve r , the highest enr i ched 6-inch b l anket is 2.6 w/o. The fo ll owing cases were run to NET-28 091-0003-0 1 , R ev i s i on 0 132 show that u sing the Z l oad in g curve (8-in ch at 4.0 w/o) for the s h orter l ess enric h ed bl anket (6-inch at 2.6 w/o) i s conservative.
4.6 w/o, 42 GWd/T, Z b a t ch: 4.6 w/o, 42 GWd/T , 6-inch AA: 4.6 w/o, 46 GWd/T , Z b atch: 4.6 w/o , 46 GWd/T, 6-inch AA: ' k-eff = 0.97 1 23 +/- 0.00008 k-eff= 0.96982 +/- 0.00008 ~k = 0.0014 k-eff = 0.95008 +/- 0.00008 k-eff= 0.94884 +/- 0.00008 ~k = 0.00 1 3 Tab l e 8.23 shows a set of assem blie s th at faile d Category 4 but h ave a bumup that i s less than 28.5 GW d/T (the minim um bu mup for Ca t egory 3). To s h ow whet h er they can make Category 3 , each assemb l y was calculate d in the infinite 2x2 m odel of R egion 1 conta in ing four fuel assembl i es. T h e calcu l ate d k e rrwith 28.5 GWd/T fuel (5.0 w/o enr i c hm e nt , PF=l.4) in this model i s 1.0133 (not e that Category 3 fuel u ses l ea ka ge from the edge of the SFP to l ower its k -this is further explained in Section 8.8). If the calc ulat ed k err for a n assemb l y i s l ess th an 1.0133, then it is classified as Category 3 even tho u g h it has a burnup l ess than 28.5 GWd/T. Table 8.23: Individual Assembly Analysis for Category 3 Assembly ID Enrich Burnup Cooling PF Cale. k (MWd/T) (years) X02 4.802 28357 6.3 1.1 94 0.9956 XOl 4.802 28460 6.3 1.199 0.9956 F65 3.346 1 2 034 37.7 1.228 0.9911 V43 (IP3) 3.803 14949 26.2 1.11 6 1.0005 V48 (IP3) 3.8 05 15180 26.2 1.1 33 1.0005 Based on th ese results, a ll of these assemblies are cl assified as Category 3 even though the ir burnup is l ess than 28.5 GWd/T. The assem bli es in Ta bl e 8.24 fai l e d Category 4 by a small amo unt of bumup , so th ey were a nal yzed u s in g the d e pl et i on cond ition s for the assembly rather t han the depletion conditions for the batch. Section 5.7 d esc rib es the spec i a l depletions and the axial bumup profiles used. The atom densities for these asse mbli es were placed in the Region 2 (3-out-of-4) 2x2 model (a ll three assem bli es in t h e mod e l NET-2809 1-000 3-0 1 , R ev i s ion 0 133 are the sa m e). The calculated kerr plus th e bi as a nd unc er taint y i s le ss than 0.99 in all cases, so the y are cla ssifie d as Category 4 (stored in the 3-out-of-4 area of Re gio n 2). Table 8.24: Individual Assembly Analysis for Category 4 Assembly ID Enrich Burnup Cooling PF Calc.k Bias & Uncer. k9S/95 AlO 2.21 15038 42.3 0.92 0.94 8 9 0.0166 0.9655 F44 3.35 23017 35.8 1.05 0.9609 0.021 8 0.9827 L48 3.69 29515 25.4 0.69 0.962 4 0.0235 0.9859 W52 4.96 40641 6.3 0.8 4 0.9569 0.0307 0.9876 X18 4.95 42479 4.3 0.87 0.9566 0.0306 0.9872 U12(IP3) 3.21 24800 27.8 0.90 0.9588 0.0223 0.98 11 With the above loading curves, eve r y hi storica l assembly through B a t c h X of IP2 and Batch AA of IP 3 h as been categorized as 3, 4, or 5. This categorization is su mmari ze d in Appendix B. Fuel asse mbli es A l 1 , A24 , A47, A49, ASO, A51, A54, and ASS in IP2 an d assemblies A38, A43 , A44 , A45, A51, A59, A63, and A64 in IP3 contained part l engt h control rods. These assemblies ha ve an e nri c hment of 2.25 w t% 235 U-235 (w/o) and burnups greater than 16 GWd/MTU. The Category 4 burnup require ment for th ese assemblies is l ess than 11 GWd/T. Although the r eact i v it y effect of control rods i s significant
[ 14], it is not enough to overcome the 50% excess burnup in these assemblies. These assemblies ha ve insufficient burnup for Category 5 a nd are d ispos ition ed as Catego r y 4. N ET-28091-0003-01, Revision 0 1 34 
: 8. 7 Cell Blockers IP2 will have two cells blocked (no fuel) at location s A22 a nd BC72. One of these ce ll s (A22) is at the Region 1 and 2 interface and is not credited.
The other ce ll (BC72), however , i s on the e dge of the SFP above the cask lo a din g area. Due to th e ce ll blo cker , it i s possible to allow a Category 4 ce ll on both sides of the cell blocker a l ong the cask l oading area. Figure 8.8 shows the SCALE model w i t h the two extra Category 4 ce ll s. The k err for thi s case is 0.95844 +/- 0.00006. This can be compared to the r efere nc e k err of 0.95839 (without th e cell blocker and wit hout th e two additional Category 4 cells). The differen ce is within the Monte Ca rl o unc erta inty. Figure 8.8: Cell Blocker Region 2 Model NET-2809 1-0 003-01, Revision 0 L((;(P') Q wuo MfOIIAL 2 -MIOUAI.. J -MIUIIAl..-4 c::J ,wnuun1..a 0 MltlWl .. 3 l -N'H(RIAL M -Ml[RIAL.S J -MlfHfn.., uo -MIUIIAl411
-Ml{Nl,-_<1 12 a wmM I AL .. 1 3 -Ml(Nlnl-4 14 -11Al(RIAI 41 S -MTUIIJll41 6 -M I Ol:ll'tl ,tt 7 -MfCfitlrt."I
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r m s Ml(RfAL $1 8 MIOUAL. 5 1 7 -Ml[RIAl.11 1 1 135 
: 8. 8 Region 2 Checkerboard Figure 1.1 does not show a Region 2 checkerboard arrangement since it is unexpected that the plant will ever use a Region 2 checkerboard. However, after the removal of a significant number of assemblies, it may be convenient to create a zone of the SFP near the cask loading area where any assembly may be placed (including fresh 5.0 w/o fuel). For this contingency , no interface analysis has been performed , so t he Region 2 checkerboard zone must have a row of water holes on all sides. The edge of the SFP can be credited as a row of water holes. Using an infinite 2x2 checkerboard model of Region 2 , the calculated k eff of 5.0 w/o fuel with 64 IFBA rods is 0.9521. With the bias and uncertainty (0.0083 from Tab l e 7.5) this produces a k 9s 1 9 s of 0.9604. This easily meets the 0.99 target. All of the fresh and burned fuel from IP2 and IP3 qualify for placement in this Region 2 checkerboard. 8.9 Burnup Penalty f or Hafnium Flux Suppressio n Inserts Hafnium inserts have been used at eight corner locations in the IP3 core starting with Cycle 11 to mitigate concerns o v er Pressurized Thermal Shock. Hafnium inserts have never been used at IP2. The burnup penalty for hafnium inserts was determined in the previous CSA to be 1.31 GWd/T (worst case) [1]. To provide more than adequate margin , a penalty of2 GWd/T is added to the bumup requirement for any assembly that had a hafnium insert any time during its life. 8.10 Failed Fuel Containers The southeast comer (p l ease note that on the drawings in this report North points left) of the SFP contains two 16" circular pipes and are labeled "failed fuel containers" on Figure 3 .1. These containers have been used to store pieces of failed fuel rods, neutron sources, and fission chambers. The neutron sources and fission chambers contain too l ittle fissi l e material to be a concern. The fission chambers have less than 10 mg U-235 each [25]. The neutron sources also have a very small amount of fissile material.
The ANSI/ANS 8.1 standard [53] states that 700 grams ofU-235 in any configuration is always subcritical.
However, the failed fuel containers are not fully decoupled from the Module H of Region 2. NET-2809 1-0003-01, Rev i sion 0 136 This analysis permits 16 fuel rods in each of the failed fue l containers. Rather than model the actual container , 16 pins are placed close to the fuel modules for each failed fuel container.
Since the same SCALE unit is used in two places in the model, a third set of 16 fuel pin s is in the model. Fig ure 8.9 shows the model with the extra pins. The extra pins did not change the k eff of the model. Table 8.25 shows the re su lt s of calcu l ations where the number of pins is var i ed. A lth ough more than 16 fuel pins per failed fuel container wou ld be possible, there are current l y pieces from only one fai l ed fuel pin in the containers no w , and greater than 16 rods in each is not credible. The fuel rods are modeled as fresh bl anketed, no-IFB A , 5.0 w/o rods. The start source was placed near the failed fuel container.
Figure 8.9: Failed Fuel Container Pin Model (Blow up of the top right of the Region 2 model) NET-28091-0003-01 , R ev i sion 0 -1111101111.
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-f1A1C IU ,.S 1 7 -l'W<llltU l!Ut 137 Table 8.25: Failed Fuel Container Pin Analysis Pins oer Container Calculated k Sigma 0 0.9584 0.00005 4 0.9586 0.00006 9 0.9585 0.0 0007 1 6 0.9585 0.00006 81 0.9635 0.00007 121 0.9793 0.00008 144 0.9959 0.00008 8. 11 Fuel Rod Storage Basket The Indian Point SFP can h ave movable fuel rod storage basket s that can be used to sto re fuel rod s. These baskets can fit in a storage ce ll and the y have 52 holes for storing fuel rods. This was modeled as 52 fresh 5.0 w/o fuel rods in the 3-out-o f-4 area of Region 2 (see Figure 8.10). The calculated k efT for this configuration was 0.9584, which is we ll below the k 95 1 95 requirement of 0.99 since the bias and uncertainty for fre s h fuel is only 0.0083 t.k (from Tab l e 7.5). Therefore, the fuel rod storage basket i s cla ss ified as reactivity Category 4 (refer to Table B. l ). Figure 8.10: Model for the Fuel Rod Storage Basket NET-28091-0003-01 , Revision 0 138 
: 8. 12 Assemblies with Missing Fuel Rods Typically , when a fue l assembly has one or more failed fue l rods, the failed fuel rod is removed and replaced w ith a stainless steel rod having the same outer dimension as a fuel rod. If this is performed, there is no criticality concern since the reconstituted assembly wou ld be less reactive than the original assemb l y. However , if a failed fuel rod is removed but not replaced with a stainless steel rod, the reactivity in creases because there is more water ava il ab l e. An ana l ysis was perfonned for the previous CSA [ 1] in which one or more fuel rods are removed from an assem bl y to estimate the effect on k. This analysis was not repeated since the approach taken provided a large margin. The model that was u sed for t his ana l ysis contained absorber inserts. It was detennined that ke ff gradually increases as more fuel rods are removed up to and including 36 missing fuel rods. If more than 36 fue l rods are removed, k eff begins to decrease (see Figure 8.11 ). The change in k e ffwith 36 missing fuel ro d s (see Figure 8.1 2) was 0.0 1 84 ~k (a 2x2 array with a ll 4 assemb lie s h aving 36 missing rods). For simp li city, a bumup penalty of 4 GWd/T would cover this reactivity in crease for an assembly wit h any number of missing rods. There is on l y one fuel assembly in the SFP (assembly ID ofT67) that has a missing fue l rod. This assembly has only one missing fuel rod. The initial fuel enrichment for this assemb l y was 4.952% and the bumup is 49.81 GWd/T and the assemb l y has cooled more than 10 years. T hi s assemb l y exceeds th e Category 4 lo ading req uir ement by over 7 GWd/T, so it exceeds the requirement by more than the 4 GWd/T pena l ty for missing fuel rods. If any assemb l y in the future is re-constituted without replacing fuel rods with stainless steel rods, then 4 GWd/T wou l d have to be added to the l oading curve requirement before it could be stored. This penalty covers any number of missing fuel rods , and there is no other loading restriction (two or more fuel assemblies with missing rods could be stored next to eac h other as l ong as the 4 GWd/T is added for each assembly).
NET-2809 1-0003-01, R ev i sion 0 1 39 k 0.9850 0.9800 0.9750 0.9700 0.9650 0.9600 k versus Missing Fuel Rods 0 10 20 30 40 Number of Missing Fuel Rods Figure 8.11: k c ff versus Number of Missing Fuel Rods 000000<)00
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08*)000()00000000 000000000000000 0~0000000000000 000000000000000 000000000000000 o<x , o,x;o o>.)<Y o.) :)')'>oo~-:, o .. o~: oo:-:: o,) 00:) ,)O_>JO O,Y) ):JO 00 1 \JO 000 !ooo i.):..)*) CK). i '..X).:J :.) )O <)oo oo:J ooo : ocy_'i ,),) 0 00.>.)C) <Y) .... :JO O 00 . .)0::) 00000 0 0 0000~ 00000 0 0 GOOOO oo '. _ _-, :.) C .:) ,_-, 0:.) :J.) :-, o i .. ; o ,. ; 0.:) 00000 0 0 00000 00060 0 0 00000 oo-* <)O <J oo , -,) .:) O'.)* --oo o o*s:).:*oo ooo ,**o,) ')c::y*., j:J:) * :J.).J :-*, o,:.) 00*1 ooo ooo *oo~oo* 000 ooo *oo'oo* ooo oo. , O*Y *o ,; : ocJ .. , ,YJ o.:, .oo:*'.o
* o , , oo-:;oo 000000000600600 000000000000000 000000000000000 000000000000000 Figure 8.12: Model for Assemblies with 36 Missing Fuel Rods [1] N ET-2 8 091-0003-01 , R ev ision 0 50 140 
: 8. 13 Storage of Miscellaneous Materials Misce ll aneous non-actinide materia l s (for examp l e, empty or full trash baskets), can be stored in fuel positions of any category.
However, there are some special cases where some of the material may be stored in a water hole or 50% water ho l e. If the misce ll aneous material is any type of steel , Inconel, or absorber material (e.g., absorber coupons , stainless steel coupon trees , control rods , unburned burnable absorbers) it may di s place up to 50% of the w a ter volume at the active fuel zone ( 144 inches) of a water hole or 50% water ho l e (t h ere are no restrictions on material above or below the active fuel zone). If the miscellaneous materia l is a very low absorbing material such as a void , zirconium , a l uminum , cloth , plastic , concrete , etc., it cannot be placed in a water h ole but may be p l aced in a 50% water ho l e so long as the 50% water hole still ha s 50% water volume in the acti v e fuel zone. To confirm that a water hole i s allowed to contain 50% absorbing material , two ca s e s were run -one with 8 0% water , 20% sta inl ess stee l and another wit h 50% water, 50% stainless stee l. The diffe r ence in k eff from the reference case is -0.0100 L'ik and -0.0045 L'ik , respective l y. Other material s such as Inconel , absorber coupons , unburned burnable absorbers or control rod s absorb more neutrons than stain l ess steel and are covered by this ana l y s is. Any uranium that is not in a fuel assembly (for examp l e a removed or damaged fuel rod) must be stored in the Failed Fuel Containers (see Section 8.10) or the Fuel Rod Storage Basket (see Section 8.11). 8. 14 Borated Conditions The most l imiting acceptance criteria is for the unborated condition , so the loading criteria have been set using models that do not contain sol u b l e boron. In order to confirm that k 9s 1 9s is l es s than 0.95 at a b oron conte n t that is maintained even after a boron di lut ion accide nt (Section 9.6), a l imited num b er of additiona l calculations were performed.
The soluble boron concentration of 700 ppm is used for these calculations since this concentration ca n be easily supported by the boron dilution analysis , and i t yie l ds significant margin ink. For Category 2 fuel , at a burnup of2l GWd/f , the calculated k eFF With water at NET-2809 1-0003-0 l , Rev i sion 0 141 180 &deg;F and cont a inin g 700 ppm boron is 0.8 496. Wi t h bias and uncertaint y, this become s k 95 1 95 = 0.87 12. Due to the difference between 0.87 and the tar ge t va lu e of 0.9 4 , n o further ca l cu l ations are warranted.
For Category 4 fuel , at the loadin g curve point s for 5.0 w/o fuel at 72 hour (PF=l.20) and the 4.2 w/o fuel at 25 years (PF=0.80) with 700 ppm boron in the SFP water at 70 &deg;C, the calculated k eff at 72 hours i s 0.86 65 , w hil e the k err at 25 years is 0.8631. With t h e b ias and uncertainty app l ied , the k 9s 1 9s, to b e c ompared to the regulatory limit of 0.95 , becomes 0.8 972 at 72 hours and 0.8903 at 25 years. The bias and uncertainty u se d here for both Region 1 a nd 2 i s from th e unborated analysis.
Calculation of borated uncertaintie s is not ne e ded d u e to the lar ge margin from t h e regul ato ry l imit. Eve n if borated unc e rtainti es were calc ulated , it i s expected th a t they would b e sma ller , since the primar y uncertainty is the burnup u ncert a int y and the r eac tivity worth of burnup d ecreases with in creas ing boron concentration due to s pectral h a rdening. Furthermore, ignorin g the grids i s still conservative at 700 ppm. In addition to th e analysis wit h the 2x2 model s, a full pool case was run at 700 ppm. T he calcul ate d k eff is 0.8 600. This case had the mo s t reacti ve loadin g permitt e d for all five ca tegorie s and includes eccentricity. Using the highe st bi as a nd uncertaint y of all re gio ns , 0.0374 , the k 9s 1 9s i s 0.8974 which i s much less than the t a rget of 0.94. An a n a l ysis was a l so perform e d for Re gio n 2 only. Table 8.26 pro v ide s the re s ult s from the full pool model s. T a bl e 8.2 6: Norm a l Op e ration s w ith Boron Dilution ppm (F ull Pool M odel) R e gi o n Calc ulat e d k s igm a All Categorie s at 5.0 w/o , 72 hours cooled, Peaking B o th 0.8 600 0.0000 8 Factor of 1.2 , Loading Curve Burnups Catego ri es 4 and 5 at 5.0 w/o, 72 h ours coo l ed , Peaking 2 0.8595 0.00005 Facto r of 1.2 , Loading Curve Burnup s NET-2809 1-0003-01, Revision 0 142 8.15 Burnup Penalty for High Soluble Boron Conditions If a cycle is shut down very early, it is possible that the limiting soluble boron used in the depletion analysis (950 ppm) would not be met. The cycle would have to be shut down extremely early since the 950 ppm would be v iolated onl y if the cycle were shut down more than two months early. To cover thi s unlikely possibility , a special depletion was done at a soluble boron concentration of 1200 ppm t hroughout the depletion (this v alue exceeds the highest cycle average ppm at any burnup). If the burnup averaged ppm for any assembly exceeds 950 ppm , burnup penalties of0.2 , 0.3 , 0.6 , and 0.9 GWd/T would have to be applied to the burnup requirement for Category 2 , 3 , 4 , and 5 , respecti v ely. The following table summarizes calcul a tions to show that the s e burnup penalties are conservati v e (the Monte Carlo uncertainty is+/- 0.00003). Table 8.27: Burnup Pena l ty Results at 1200 ppm Case Calculate d k Category 2 (3 of 4 in Region 1 ): 5.0 w/o , 21 GWd/T, 950 ppm 0.96 8 9 5.0 w/o , 21.2 GWd/T , 1200 ppm 0.9686 Category 3 (4 of 4 in R e gion 1): 5.0 w/o, 28.5 GWd/T , 950 ppm 1.0117 5.0 w/o , 28.8 GWd/T , 1200 oom 1.0113 Category 4 (3 of 4 in Region 2): 5.0 w/o, 48.19 GWd/T, 950 oom 0.9586 5.0 w/o , 48.79 GWd/T , 1200 ppm 0.9583 Category 5 ( 4 of 4 in Region 2): 5.0 w/o , 59.19 GWd/T , 950 ppm 1.0187 5.0 w/o , 60.09 GWd/T , 1200 ppm 1.0183 NET-28091-0003-01, Revision 0 143 9 Normal Operations and Accident Analys i s The criticality analysis must address all conditions in the SFP that can cause criticality.
The normal operations are reviewed in Section 9.1. The accident analysis must assume the worst case conditions from the range of normal operations. The accident analysis is covered in Sections 9.2 through 9.7 and analyzes the possible upset co nditions that increase the reactivity of the SFP. The following accidents are analyzed:
* A fresh assembly misplaced outside of the fuel racks but next to the fuel racks,
* A fresh assembly dropped into an empty cell,
* An over-temperature accident (water boiling in the SFP as a result of a loss of cooling), and
* A multiple assembly misload event. Two more accident conditions are considered, but no analysis is necessary.
An assembly dropped horizonta ll y on top of other assemblies is not specifica ll y analyzed, because the assemblies are de-coupled as a result of the structure above the active fuel. The horizontal assembly would rest more than 20 cm above the top of the active fuel of the assemblies in the rack. This accident would be covered by the more severe accident of a fresh assembly dropped into an empty cell. The second accident condition would be a single misloaded assembly.
For example , an assembly that is supposed to have a control rod inserted but does not. All violations of the l oading requirements are bounded by a fresh assembly dropped into an empty cell. The last subsection of this section describes why a seismic event does not cause a criticality concern. NET-28091-0003-01, Revision 0 144 
: 9. 1 Normal Operations A single isolat e d assembly at 5.0 w/o and no IFBA will have a k 9 s 1 9 s < 0.99. An assemb l y is isol ate d if there is 20 cm of wa ter b etwee n asse mbli es [ 1 3] (a row of em pty cells is 23 cm for R egio n 2). The equ ipment in the SFP can only move one assembly at a tim e. IP 2 doe s not have a n y racks in the SFP or refu e ling canal for temporary storage of fuel. However , there are t wo l ocatio n s w h e r e it wo uld be possible to plac e two assem bli es wit hin 20 cm of each other o ut s ide of the rack: w h e n a fuel assembly i s in the n e w fuel e l eva tor and when a fuel assemb l y i s vertica l in the up ender. Indian Point's proc e dur es w ill not permit movi n g a n assembly outside the rack within 25 cm of fuel in the n ew fuel e l eva tor or up en d er. The critica lit y a n a l ysis cre dit s the leakage at the edge of th e racks. Therefore, placi n g an assem bl y within 20 c m of the side of th e rack at the act i ve fuel e l eva ti ons is not permitted. IPEC's procedure wi ll also preclude thi s. Moving fuel in a nd out of rack ce ll s do es not in c r ease k err s ince the most r eac tiv e por tion of the fuel assemb l y is at th e top of the fuel, so movin g the bottom of th e fuel pa s t the top as it is in ser ted or rem ove d do es not in crease k. Fue l in s pe ction is performed above the rack w h e r e th e fuel assem bl y is iso l ated. Any reconstitution is p e rformed w hil e the assemb l y is iso l ated. I so lation r e quire s a row of water hole s on all si de s a nd corners. The o ut s ide of th e SFP can b e considered as a row of water holes. T he SFP i s required b y it s Techn ical Specifications t o contain at l east 2000 ppm of so lubl e boron. The SF P water temperat ur e durin g normal operation ranges fro m above freezing to 1 80 &deg;F (the SFP de s i gn basis ma ximum temp erature).
NET-28091-0003-01 , Revision 0 145 
 
===9.2 Misplaced===
 
Assembly For the misplaced fuel assembly accident, a fresh 5.0 w/o fuel assembly with 64 IFBA rods is placed in the SFP next to the rack in the most reactive location.
There are two locations which could be limiting for a misplaced assembly; the inside comers of the cask area and near the new fuel elevator when the elevator has a new fuel assembly in it. For the misplacement of fuel near the new fue l assembly in the elevator the fue l elevator is modeled in the cask loading area of the SFP. The key features of the model are maintained; close to a wa ll with two rows of Category 5 cells and in a big pool of water. The array structure used in the SCALE model makes it difficult to model at the actual location. The fuel elevator is conservatively modeled as exactly one assembly away from the side of Region 2 , so one misplaced assembly can exactly fit between the elevator and the rack. Figures 9.1 and 9.2 show the location of the misp l aced assemblies analyzed. Table 9.1 presents the results of this misplaced assembly accident analysis.
As can be seen from Table 9.1 , with the Technical Specification minimum of 2000 ppm of soluble boron , the final k 9 5 1 9s is below the target of 0.94. The analyses used starting sources located near the misplaced assembly.
NET-2809 1-0003-01, Revision 0 146 Figure 9.1: Misplaced Assembly at the Cask Area Corner N ET-28 091-000 3-0 1 , R ev i s i o n 0 L[QHO O vom Ml(AIAl.2 -MTUUl'll.3
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===9.3 Dropped===
Assembly For the dropped assembly accident, a fresh 5.0 w/o assembly (no IFBA) is dropped into one of the empty cells in Region 2 with the SFP at the Technical Specification minimum of 2000 ppm. F u rther, the assembly dropped is modeled with the grids failed , which allows for full expansion of the pin pitch into the cell (this maximum expansion of the pin pitch removes any concerns about fuel grid failure after the drop). The pin pitch expansion is modeled as the maximum uniform expansion that would fit in the cell. Figure 9.3 shows the fu ll pool model for the dropped assembly analysis with the dropped assembly. Note that Region 1 cells ha ve been removed so that the k e rr for Region 2 can be found, which comes with a higher bias and uncertainty than Region 1. The other fuel assemblies are at the lo ading curve limit for 5.0 w/o, no cooling and a peaking factor of 1.2. The calculated kerr is 0.8700 with the 2000 ppm soluble boron. After adding bias and uncertainty , this would be much lower than the target of 0.94, giving 1 % in margin. With this much margin , there i s no need to reevaluate the bias and uncertainties for borated conditions.
NET-28091-0003-01, Revision 0 149 F i g u re 9.3: F ull Pool M od e l w i t h Dropp e d Assem bl y 9.4 Over Temperature Lf.(i(Nl O wuo IIATCAIAl.
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[ 4]. Using the highest ke rr full pool model , water is modeled with 2000 ppm of so luble boron. The calculated NET-28091-0003-01 , Rev i sion 0 150 k e ffis 0.7464 +/- 0.00007. This case ha s a l arge margin to the 0.94 tar get, so calcu l ation ofa bias and un certa inty for this spec ifi c case i s not n ecessary.
 
===9.5 Multiple===
Misloads The mo st limi ting accident is the multiple misload case. For th e R egion 1 multipl e misload, it is assumed th at all ce ll s are filled with the most limiting assembl i es for IP2 (5.0 w/o enriched fuel wi th no bumup and 64 IFBA rods). The calc ulat ed k eff at 2000 ppm is 0.8196 +/- 0.00008. C l early, there i s significa nt mar gin for th e complete mi s lo ad of assem bli es in Region 1. For th e multi-mi s load ana l ysis for R egion 2, a ll ce ll s that are p ermitted to contain fuel are modeled as misloaded with once burned 5.0 w/o fuel with a burnup of24 GWd/T , as unburned fuel is eas il y identifiable and expected to be loaded into Region 1. Nearly a ll of th e once burned fuel asse mbli es exceed this bumup. Fo r example, if Cycle 23 com pl e t es it s expected burnup , four assemb li es will h ave a burnup of22 GWd/T with all of the r es t above 24 GWd/T. Curre ntl y (during IP 2 Cycle 23), only four assemb li es would be in the SFP during refueling wit h a bumup l ess than 24 GWd/T. Since the numb er of assemb li es b e l ow 24 GWd/T i s so few, misloading them in a reactivity significa nt way i s not credible (note that a sing l e misload is covered b y the dropped rod analysis, and if the misloads are not clo se to each ot her , th e effec t on k i s the same as a single misload).
T h e ca lcul ated k e ff for this case with 2000 ppm i s 0.9124 +/-0.00006. The bia s and uncertainty for this case is 0.0223 Lik (from conservative interpolation using Tab l e 7.3, the manufacturing tolerance s from Table 7.4, plus the hi gher va lidation bias and uncertainty due to the h arder spec trum). Addi n g 0.0223 to 0.9124 result s in a k 9s 1 9s= 0.9347 which is b e l ow th e t arget of 0.94. The multi-mi sload analysis did not mi s lo ad fuel i nto th e water hole s or 50% water h oles. It also m ainta ined th e contro l rods at th e location s required by th e Tec hni ca l Specificat ion s. The wa ter hol es an d control rod loc ations in R egion 2 are not allowed to c h ange position.
With the impl ementa tion of thi s N ET-28091-0003-01 , Revi sion 0 151 CSA, the IPEC staff wi ll receive training with emphasis on the fixed positions in Region 2 for the contro l rods and water holes. Although it is clearly not allowed by the Technical Specifications and the staff will have had specific training to reinforce the requirements of the control rods staying in the specific locations, analysis has been performed where all of the control rods in Region 2 are removed. This case assumes that the fuel is consistent with the most limiting conditions allowed by the Technical Specifications (5.0 w/o fuel with burnups of 48 .19 and 57.89 GW d/T). The calculated k e tr with 2000 ppm is 0.7880. Again, this k e tr after the addition of an appropriate bias and uncertainty is significantly less than the target 0.94. The final multi-misload analysis assumes the nonnal loading of Region 2 but a l l water holes are filled with fresh 5.0 w/o fuel with 64 IFBA. Since the reactivity is dominated by the fresh fuel, it is appropriate to reduce the Category 4 fuel burnup by the difference in the bias and uncertainty.
From Table 8. 7, the bias and uncertainty at 48.19 GWd/T is 0.0307 t.k. The Region 2 fresh fuel bias and uncertainty is 0.0083 t.k (from Table 7.5). Therefore a delta burnup to cover 0.0307 -0.0083= 0.0224 t.k is taken. This is estimated as 4 GWd/T. The Category 4 fuel mode l ed is the loading curve 48.19 -4 = 44.19 GWd/T. The calculated ke tr is 0.9224 +/- 0.00005. The Region 2 bias and uncertainty for Category 1 fuel from Table 7.5 after adjusting for the high EALF is 0.0151 t.k. Thus , the k 9s 1 9s is 0.9224 + 0.0151 = 0.9375. Since this is l ess than 0.94, this mu l ti-mis l oad meets the requirement.
: 9. 6 Boron Dilution Accident Crediting 700 ppm of soluble boron reduces the calculated k e tr plus biases and uncertainty to well below 0.94 (see Table 8.26). The boron dilution analysis of record [52] shows that dilution from the Technical Specification required minimum of 2000 ppm to 700 ppm is not credible due to the amount of time and water needed to dilute to this l evel. NET-28091-0003-01, Revision 0 152 
: 9. 7 Seism ic Event This CSA is not crediting any absorber plates that could be affected by a seismic event. Further the space between the rack modules is not credited.
The space between the rack modu l es and the side of the SFP has a very small reactivity effect (see Table 6.5), so the soluble boron would easily cover any possible movement of the rack modules toward the wall. Any random v ariations in the cell dimensions would have to cover multiple cells to affect ke rr, and , again , the soluble boron would easily cover these variations.
The dropped assembly that could resu l t from a seismic event is covered in Section 9.3. In summary, this CSA is not sensitive to a seismic event. NET-28091-0003-01, Revision 0 153 10 Summary This CSA removes the reli a nce o n BoraflexTM in the IP2 SFP. Thi s is accomp li shed by use of water hole s, control rods , a nd l eakage at the edge of the SFP. Each fuel assembly i s categorized b y its r eactiv ity a nd spec ifi c lo cations in the SFP are reserved for each re act ivity category. The fuel categorization o f historical fuel i s acco mplish e d b y u se of batch gro uping s w ith simi l ar characteristics.
T h e categorization takes credit for lower enriched ax ial blanket s , coo lin g time , and assembly average peakin g factor. This CSA a l so categorizes IP3 fue l for storage in th e IP2 SFP. Section 10.1 contains a confirmation checklist th at th e gui dance ofDSS-ISG-2010-01
[5] is fo ll owe d. Section 10.2 de sc ribes the reacti v ity categori za tion of fuel assemb li es. Section 10.3 identifies the reactivity cate gorizati on of each cell. Section 10.4 li s ts the limitation s for fuel assemblies that ha ve n o t been categori ze d in Appendix B. 10.1 Review of DSS-ISG-2010-01 Tab l e 10.1 s how s the guidance given in DSS-ISG-2010-01 a nd how this criticality analysis follows that guida n ce. Table 10.1: DSS-ISG-2010-01 Checklist Section in this Guidance from DSS-ISG-2010-01 Implementation Report 1. Fuel Assembly Selection A ll fuel ha s co m e from th e sa me ve nd or w ith th e same cla d Section 3.2 Demon s trat e a ll fue l for a ll o ut s id e diameter.
Small des i g n c h a ng es have been conditio n s in s i g nifi cant to cr iti ca lit y a nal ys i s. 2. Depletion Ana l ys i s Thi s is fo ll owe d. Uncertainty fo r th e i soto pi c content is Sect i o n 4 a.i. 5% (Kopp M e mo) s h ould co n s id e r e d and impl emented as 5% of the d e pl e tion reactivity only be used to cover uncertaintie s (i.e., d e l ta-k of depletion). In a ddition , a bi as of 1.5% of the in i soto pic concentration worth of the minor actinides a nd fi ss i o n products i s a ppli ed to cover th eir bias a nd un certainty.
: 2. D e pletion Analysis No int egral bu rn ab l e a bs or ber s are co n side r e d for fre s h fuel Section 4 a.i i. Reacti v it y d ecre m e nt s h o uld for det er minin g the r eac ti vity decrement.
not inc lud e the integ ra l burnabl e absorbers.
: 2. D e pl e tion Analysis Boundin g va lue s within eac h b a tch grouping are u sed for a ll Sections 5.1 b.i. Bounding va lu es s hould b e paramet ers. throu g h 5.5 u sed. NET-28091-0003-01 , Revision 0 154 Section in this Guidance from DSS-ISG-2010-01 Implementation Report 2. Depletion Analysis The hi ghes t power i s used which l eads to hi g h e r mod erato r b.ii. Use the more limitin g a nd fuel tem p erat ur es , thus in creas in g k. To acco un t for Sections 5.4 bounding p arameter when a conflict low er power coast down , a Sm-149 correct i on is made. and 5.8 occ ur s. 2. Depletion Analysis Bounding va lu es within eac h batch group in g are u sed for all Sections 5 .1 b.iii. Non-bounding values are p a r ameters. th rough 5.5 outside sco p e ofISG. 2. Dep l e tion Analysis IP 2 a nd IP3 had s tand ard burnable absorbers , a nd W ABAs. Sections 5.2 c.i. All r emovab l e burnabl e IP 3 a l so had Hf flu x s uppr essors. All of these are a nd 8.9 absorbers mu st be co n s ider ed. conservatively accounted for. 2. Depletion Analys i s The analys i s in c lud es the maximum number ofIFBA rods at Section 5.2 c.ii. Limiting integra l burn ab l e th e hi ghest boron lo ading for eac h fuel batch gro upin g. absorbe r s s h ou ld be u se d. 2. Depletion Analys i s For d epletio n analysis, the maximum absor be r mat erial i s Section 5.2 c.iii. Model the burnab l e mod e l ed w ith th e m ax imum wa t er displacement.
For the p oo l abso rb e r s a oo ropriately.
a n a l ys i s, a ll burn a bl e absor b ers are removed. 2. Depletion Ana l ysis The depletion m o d e l correc tl y accounts for the in creased rate Section 5 c.i v. Consider co mp et in g effects of plutonium produ c ti on fro m in creased fast n e utron capture in U-238 2. Dep l etion Ana l ysis A ll hi storica l assemblies under D-Bank were id e ntifi ed an d Sections 5.5 d.i. Spectrum h a rden i ng from th e a pp ropr i a t e treatment i s a ppli ed. For c urr en t fuel , it i s rodded o p erat i on sho uld be assumed that a co nt rol rod was fully inserted for 2 GWd/T consi d e r e d. burnup. 2. Dep l et ion Analysis The axial profiles for asse mbli es without ax i a l blankets are Section 6.2 d.ii. Effect of contro l rods on the from NUREG/C R-6801. T h ese pro files include rodded cases. axia l burnup p rofi l e s hou l d be For bl anketed fu e l , actual burnup profiles are used. T h ese co n si d ere d profil es cove r any contro l rod e ff ects. 3. Crit i ca lit y Analysis The axial profiles for assemb li es without ax ial bl ankets are Sect i on 6.2 a. Ax ia l Burnup Profi l e from NUREG/C R-6 801 an d used in a conservat i ve manner. i. Use ofNUREG/C R-6 801 is For blanketed fu e l , actual burn up p rofiles are u se d. acceptab l e if d one properly 3. Crit icality Analysis Site-specific profi l es are used. Section 6.2 a. Axia l Burnup Profile ii. Site-specific profiles 3. Cr iticality Analysis For full l eng th fuel , results from uni form and NUREG/CR-Sect i on 6.2 a. Ax i a l Burnup Profi l e 6801 s h apes were compared at 10 GWd/T a nd the iii. Uniform profi l es NUREG/CR-680 1 shape was mor e li miting. The l owest burnup of a ny ful l l e n gth fuel i s greater than 10 GWd/T. For axia l l y blanketed fu e l , the lowest relative power at each node from a ll the blank eted fuel burn ed a t Indian Point was used to d eter min e the axial burnu p di s tributi on for eac h ax i a l blanket design. Since these relative powers were not renormalized, it covers both the top p eaked and center peaked condition.
: 3. Crit i ca l i t y Analysis The rack dimensions and m ater i a l s a r e provided b y the Section 3.1 b. Rack Model manu facture r (References 8 and 9). i. Mod e l input s s hou l d b e tracea b l e. 3. Critica lit y Analysis No c r edit i s t ake n for absorbe r p a n e l s. b. Rack Model ii. Effic i e n cy of t he n e utron absorber s h o uld be es tabli s h e d. NET-28091-0003-01, Revision 0 155 Section in this Guidance from DSS-ISG-2010
-01 Implementation Report 3. C ritica l it y A n a l ys i s No c r e di t i s t a k e n for absor b e r p a n e l s. b. R ac k M o d e l iii. C on se r va ti ve d e grad a ti o n s h o uld b e u se d. 3. C ri t i ca lit y Ana l ys i s T h e in terface a n a l ys i s adj u ste d th e bu rnu p fo r th e d iffere n ce S ec ti ons 7.6 c. Int e rfa ces -Use th e m axi m um in th e bi as a nd un ce rt a in ty. a nd 8.4.3 un ce rtainti es fr o m e i t h e r s id e. 3. C riti ca lit y Ana l ysis Th e r e a r e n o te mp ora r y storage l oca ti ons in th e IP 2 SF P. A ll Sec ti o n 9.1 d. N o rmal Cond iti o n s -All n o rm a l o p era tin g co ndit io n s a r e c ove r ed by t h e a nal ysis. n o rm a l co nditi o n s suc h as m ove m e nt of fu e l and in s p e c t i o n s s h ou ld b e co n s id ere d. 3. C riti c ality A n alysis A ll n o rm a l initi a l co ndi tio n s a r e co n s id ered. F o r exa m p l e , in Sec ti o n 9 e. Acc id e n t Con di t i o n s t h e mi sp l aced asse mbl y ana l ys i s it i s assu m ed th a t th e fuel i. S h o uld c o n s id er a ll norm a l e l e v a to r has a fr es h fu e l asse mb l y in it w h e n a n o th e r co nditi o n s a s b ase co nditi o n s. asse mbl y is mi s pl ace d n ext t o i t. 3. Crit i ca lit y A n a l ysis L a r ge m a r g in s ex i s t for a ll of t h e acc i dent co nditi o n s with t h e Sec ti o n 9 e. Acc id e n t Co ndi t ion s exce pti o n of th e multipl e mi s l oa d s c e n a r ios. Th ese sce n a ri os , i i. G rad e d a pp roac h m ay be h owever , s till m ee t th e targe t k 9 5 1 95 (see Sect i o n 9.5). t ake n w h e n c r e di t in g so l ub l e b oro n. 4. C riti ca l it y C od e Va lid a tion NURE G/C R-6698 i s fo ll owe d for th e va l i d a t io n. A pp e n d i x A NURE G/C R-6698 e nd o r sed 4. C r i ti ca lit y Co d e Va lid a ti o n T h e HT C c r it i ca l ex p e rim e n ts are in c l u d ed i n th e a n a l ys i s. A pp en di x A.3 a. Area of A pp lica bilit y i. In c l ud e t h e HTC c riti ca l s 4. Cri tic a lit y Co d e V a lid a ti o n T h e fin a l b ias a nd un ce rt ai nt y i s d e t e rmi ne d b y th e m ost A pp e n dix A a. Area o f A pp lica bilit y limitin g o f e ith e r t h e M OX a nd H TC c ri tica l s or th e U0 2 ii. Use a pp ro p riate c ri t i ca l s c riti ca l s. 4. Cr iti ca lit y Co d e Va lidation 328 fr es h fu e l c riti c al ex p eri m e nt s a r e u se d. 11 7 H TC A pp e n dix A a. Area of A ppli ca bilit y c riti ca l s are u se d as w e ll as 6 3 MO X cr it ica l s. G roupin gs o f iii. S uffi c i e n t c riti ca l s fo r c ri t i ca l sets are a n a l yze d to co nfirm w h en th ey s h o ul d be a n a l ys i s a nd app ro pri a t e g roupin g. in c l ud e d i n th e se t as a w h o l e. 4. Crit i ca lit y Co d e Va lid a tion T h e l a r ge numb e r o f c riti ca l ex p e rim e n ts u sed a nd th e l arge A pp e n d i x A.2 a. Area o f A pp l i ca bilit y va ri a ti o n in c ri t i ca l co n figura ti o n s (geometry a nd m a t e r ia l) i v. B e s ur e th e se t i s n o t hi g hl y r e du ces th e co nc e rn a b o u t b e in g c orr e l ated. Th e a n a l ys i s co r re l a t e d. u se d 37 d i ffe r e nt se t s of expe rim e nt s th at we r e p e rform ed in 7 diff ere nt cr iti ca l fa ciliti es. 4. C ri t i c ality C od e Va lid a tion T h e t re nd a n a l ys i s i s p e r for m e d o n a ll of t h e maj o r A pp e nd ix A.2.5 b. T r e nd An a l ys i s p ara m e t e r s. T h e tr e nd a n a l ys i s fo un d th e b es t lin ea r tit. No A d e qu a t e , a pp ro pri a t e, n ot tr e nd s were rejec t e d t o b e co n serva ti ve. The m os t limiti ng r e j ecte d. bi as a nd un ce rtainty for th e area of a ppli ca bilit y i s appli ed ass umin g ei th e r that a ll t rends a r e r ea l or t h e r e a r e n o t re nd s. 4. Cr iti ca lit y Co d e Va lid a ti o n Th e s t a ti s ti ca l a pproa c h r eco mm e nd e d in NURE G/C R-6698 A pp e n dix A.2.5 c. S t a ti s ti ca l Trea tment i s u se d. T hu s th e v ari a n ce o f th e p o pul a ti o n a bout th e m ea n i. Use th e varia n ce o f th e i s u se d rat h e r t h a n th e va r ia n ce o f th e m ean. p o pulati o n a b o ut th e m ea n 4. C ri t i ca lity C o d e Va lid a ti o n Th e s t a ti s ti ca l approa c h r eco mm e nd e d in NURE G/C R-6698 App e n d i x A.2.5 c. St a ti s ti c al Treatme nt i s u se d. T h e co rr ec t co nfid e n ce fac t o r s a r e u se d. ii. U se co rr ec t co nfid e n ce fac t o r s. 4. Cr i t i ca lit y C o d e Va lida t i o n No rm a li ty t es tin g i s p e r fo rm e d a nd th e a pp ro pri ate s t a ti s ti ca l A pp e ndi x A.2.5 c. Stati s ti c al T r eat m e nt t r ea tment is a ppli e d. iii. Con s id e r N o rmalit y N E T-28091-0003-01 , Revision 0 15 6 Guidance from DSS-ISG-2010-01
: 4. Crit i ca lity Code Validation
: d. Lum ed Fission Product s 4. Crit i ca lit y Code Validation
: e. Code-to-Code Co mpari sons 5. Miscellaneou s a. Precedence
: b. Reference s c. Ass um tions NET-28091-0003-01 , Revision 0 Im lementation Lumped fission products are not u sed. No code-to-code comparisons are used for va lidati on. However , CASM0-5 analysis was used to co nfirm that the ISG-2010-01 allowed 5% of the delta kerr of depletion is ade uate. Precedence is n ot u sed as a li ce n s in g basis. Reference s used were carefully chosen to be applicable to the point being made. Assum tions are identified.
Section in this Re ort 15 7 10.2 Fuel Reactivity Categorization The reactivity category for each fuel assembly must be determined prior to l oading in the IP 2 SFP. For IP2 Batches A through X, the reactivity category for each assembly is found on Table B. l and for IP3 Batches A through AA on Table B.2 of Appendix B. Fuel for IP2 and IP3 batche s beyond those in Appendix B are labeled " Batch Z". The equations to determine the categorization for Batch Z are: B 12 = (-6.26824 + 5.29367*E
-0.37 l 54*E 2) e -(0.1 29582 -002049 1 s*E + o.00205596*E*E l*C T _0_ 13 331 + 6.9037*E + 0.122068*E 2 B o.s ( 15.1405 _ 4.8 l l 33*E + 0. 753855*E 2) e -(0.1 2 1 252 -0.0 1 5099 1 *E + o.00121009*E*El*CT _ 1 6_2293 + where MRB E CT PF 14.0 159*E -0.687054*E 2 MRB = Bo.s + (PF -0.8) x (B1.2 -Bo.s) I 0.4 is the Minimum Required Burnup (GWd/T) is the U-235 initial enrichment (w/o) is the cooling time (years) Assembly Burnup I (Sum of Cycle Burn up s) If an assembly had an in serted hafnium flux suppression insert a n y time during its l ife, then 2 GW d/T bumup must be added to the MRB. If an assembly ha s any number of mi ssing fuel pins that have not been rep l aced by stain les s steel rods, then 4 GW d/T burnup must be added to the MRB. If an assembly was burned with a burnup averaged sol ubl e boron concentration of greater than 950 ppm , then 0.2, 0.3, 0.6, and 0.9 GWd/T must be added to the MRB for fuel Categories 2 , 3, 4, and 5 respectively.
If the fuel has a burn up greater than (MRB+ 11) the fuel is Category 5. If the fuel has a bum up greater than MRB but less than the Category 5 requirement the fue l is Category 4. If the fuel has a burnup greater than 28.5 GWd/T but less than the MRB the fuel is Category 3. If the fuel has a bumup greater than 2 1 GWd/T but l ess than 28.5 GWd/T the fuel is Category 2. If the fuel has l ess than NET-2809 1-000 3-01, Revision 0 158 Westinghouse Non-Proprietary Class 3 21 GW d/T or violates any of the requirements of Table 10.4 the fuel is Category 1 or Category 4 if it conta in s a control ro d. For a Batch Z assembly to b e s tor ed in the SFP , it must ha ve at l east 64, 48 , 32 , or 16 IFBA rods for enric hm ents less than or equa l to 5.0, 4.5 , 4.0, an d 3.5 w/o , respectively.
If an assembly has a control rod in it, its fuel category increases. If a Category 1 fuel assem bl y has a contro l rod in it , it is classified as Category 4 fuel. If a Category 2 , 3 , or 4 fuel assembly h as a control rod in it, t h e assemb l y becomes a Category 5 fuel assem bl y. Contro l rods that are required in the fuel l ayout (see Section 10.3) may not b e credited to raise the category of that fuel asse mbly (for exam pl e, a Category 2 assembly w ith an inserted control rod may not be placed in a cell in the control rod area that requires a control rod). The above loading requirements have been summarized below in Table 10.2. Table 10.2: Summary of Loading Requirements for Fuel Batch Z Minimum Burnup Requirement Category 1 Fresh fuel with at least 64 , 48, 32, 16, or O IFBA rods for enrichments less than or equal to 5.0 , 4.5 , 4.0 , 3.5 , and 3.0 w/o respectivel
: y. (@ [ mg 10 B/inch] a , c or 2reater).
Also, burned fuel with less than 21 GWd/T burnup is Cate2or v 1. Category 2 Burned fuel assemblies with at least 21 GWd/T burnup , initial enrichment of 5.0 w/o or less. Category 3 Burned fuel assemblies with at least 28.5 GWd/T burnup, initial enrichment of 5.0 w/o or less. Category 4 Burned fuel assemblies whose loading requirements are determined from Table 8.4 or the curve fit described in Section 8.3.1. Also, any Category 1 assembly containing a control rod is Category 4. Category 5 Category 4 burnup requirement plus 11 GWd/T. Also, any Category 2, 3, or 4 assembly containing a control rod is Categor y 5. N ET-28091-0003-01 , R ev ision 0 15 9 10.3 Allowable SFP Cells for Each Fuel Category The ana l ysis of the IP2 SFP uses a cell dependent reactivity.
That reactivity by position is defined in Figure 10.1 by use of ce ll categories, required contro l rods , and two types of water holes. Each cell in the IP2 SFP has a predetermined minimum fuel category.
However, there are a few alternative configurations of the cell categories in the SFP. Figure 10.1 shows the primary assignment of the cell categories for the SFP. Only Category 5 fuel may be placed in a Category 5 cell. Although Category 5 fuel can be stored anywhere in the SFP, the outer two rows of Region 2, the checkerboard control rod area of Region 2, and the interface between Region 1 and Region 2 must contain only Category 5 fuel or remain empty (alternate arrangements are allowed as discussed in Section 8.5). A Category 4 cell can accept Category 4 or 5 fuel. A Category 3 cell can accept Category 3 , 4 , or 5 fuel. A Category 2 cell can accept Category 2 , 3 , 4 , or 5 fuel. A Category 1 cell can accept all fuel. Category 1 cells can replace an area of Region 1 on the Figure 1 0.1 arrangement (Region 1) using the following two rules: 1. Each Category 1 cell must be face adjacent with at least three water holes. 2. Each Category 2 cell may not have more than one face adjacent to a Category 1 cell. Figures 10.2 through 10.4 show examples of allowable Category 1 cell locations in Region 1. It is a l so permitted to create a checkerboard arrangement of Category 1 cells in Region 2. In order to prevent interaction with other portions of Region 2, the checkerboard area must have a row of water holes on all sides. The water outside of the rack counts as a row of water holes. Finally, a 3 by 3 block of cells where all of the ce ll s are water h o l es except the center cell may be c rea ted anyw h ere in the SFP, and any fue l assemb l y may be placed i n the center pos i tion (see Section 9.1). NET-2809 1-0003-01, Rev i sion 0 160 1 1
* s , 1 a , 10 11 11 u 14 15 " 11 11 1, 10 u 21 n 24 zs z, l1 1a u 10 31 1.SMl1tltJ101112U1CU 09 ON OM t---t---t--t---11-+---t--+---t-+-il-+-+--+-ll-+-+-+-l-+-+-+-!ll-+-+-+-l-+-+-+-l-l l'-"1---t--t---t--t-t-,--,--,,-r, Ol E *~-+---+--t--+--+-,.-o-o~>-~
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OH C I ~+--+--+--+-+-+--l--1
"""" 1--1-4 1--+--+--+--+--t---+-+--+-!---i rt-1--r-r---r-t-t~-1rt-t-1r-r-r-t-t-1-rt-11 ~-+--+--+-++-1--1--1--1--1-1 0G B Key: D WaterHole D S0%WaterHole
* category 1Fuel D category 2 Fuel D categorylfuel
* category4Fuel category s Fuel category S Fuelw~h a r equired lull lengthRCCA
[!J Blocked Cell A >( O E '.=;=::=&#xa2;:.::&#xa2;=;;:=;::;:;:::!;:::;::=&#xa2;=;::::;:=;;;;;!;:::!;::=:=&#xa2;::;;;=:;:::=;=:;::::!;;::::!;:~:,a;;;;::::~~=:=::=::=::~~
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-~ Figure 10.1: Fuel Categor y Location Requirement s (Base Cas e) N ET-28 091-000 3-0 1 , Re v i s i o n 0 1 6 1 H G F E D C B H G F E D C B H G F E D C B 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 2 6 27 28 29 30 31 ,. Figure 10.2: Refueling Arrangement 1 2 3 4 5 6 7 8 9 10 Figure 10.3: Max Cat 1 Arrangement 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 1 7 18 1 9 20 21 22 23 24 25 26 27 28 29 30 31 I I ~--+ . oc:-~-t I I ----r --i -!----L -~----*---------*->->--I A X Figure 10.4: Example Odd Arrangement N ET-28091-0003-01 , Re v ision 0 162 There are restrictions on what ca n b e p l aced in th e water h o l es or 50% water hol es. T h e wa t e r hol e ma y not h ave anything in it (except water) in the act i ve fuel area with the fo llowin g except ion s:
* If the it e m i s m ade of s tainless stee l , Inconel or absorber mat eria l , it i s a l lo we d to di s pl ace up to 50% of the water a nd sti ll be sto red in a white cell ( 100% water ce ll).
* The 50% water holes (pink cells) a llo w di sp l acement of the water b y any non-fuel m ateria l up to 50% volume fraction.
This wo u l d a ll ow, for exa mpl e , a co mpon e nt made entire l y of z ir co nium takin g up l ess than 50% of the vo l um e to be stored in a 50% water ho l e (pink cell).
* The cell blo c k e r at location BC72 (in the first row of cells abo v e the cask area pit) ma y n ot b e moved an d is require d in order to c l assify the cells on either si d e of the ce ll blocker as Category 4 cells. 10.4 Fuel and Operating Requirements The actual fuel and operating con dition s are used in th e ana l ys i s of historical fuel (Batches A thr o u gh X fo r IP2 and B atches A through AA for IP3). Fue l b a t c h es after X for IP2 an d after AA for IP3 are called Batch Z. This sec tion de scri bes t h e requir e m e nt s for Batch Z. To m ee t the limitation s of this c riti ca lit y ana l ys i s, the fuel design mu st meet the d es i gn requirements g i ve n o n Tab l e 10.3. NET-28 091-0003-01, Revision 0 1 63 Westinghouse Non-Proprietary Class 3 I Table 10.3: Fuel Design Requirements for Batch Z assemblies Attribute Value Notes Maximum fuel pellet U0 2 stack density 95.0%TD This includes dishing a nd chamfering Fuel pellet OD (inches) 0.3659 Nominal F u e l clad OD (inc h es) 0.4220 Nomina l Fuel clad ID (inches) 0.3734 Nominal Fuel pin pitch (inches) 0.563 Nominal Guide tube OD (inches) 0.533 Nominal G uid e tube ID (inches) 0.499 Nominal Maximum enrichment ( wt% m u) 5.0 Maxim um blanket enriclunent (wt% m u) 4.0 Mi nimum bl a nket l ength (inches) 8 Requirement for enrichments Minim um number ofIFBA rods* 64 (IX loading) less than or equal to 4.5, 4.0, 3.5, and 3.0 w/o is 48, 32 , 16 , and O IFBA, resp ect ively Minim um IFBA l ength (inches -centered)* 12 8 0.00603 g 1 0 B/cm Design changes that increase Maximum W ABA l oading water displacement are not per rodlet covered. Maximum IFBA rods and 10 B l oadi n g 14 8 IFBA rods [ mg 1 0 B/inc h]"*c ( l .5X) per rod *These requirements are only for storage of fuel assemblies that h ave not been in the core. The depletion parameters are se lect ed to cover anticipated future operation, howe ver , verification is req uired. Table 10.4 lists the operating requirements from the depletion analysis for Batch Z (recently discharged and future fue l). The temperature and so lubl e boron assumptions are averages over the total burnup (mu l ti-cycle) for a given asse mbly. Ifan assemb l y is dep l eted such that any of the Table 10.4 parameters are not met, then the assemb l y would have to be classified as Category 1 fuel or classified as Category 4 fuel with a control rod inserted. NET-2809 1-000 3-01, R evision 0 164 Table 10.4: Fuel Assembly Operating Requirements Parameter Value Notes Maximum Core Inlet 542.6 O f Temperature Maximum Core 62.4 O f Delta T Maximum Operation
.::: 2 GWd/T This control rod inserted bumup covers rod s with Control Rod s inserted to any depth. Thi s is an a v era g e for all cycles in which the Maximum Bumup assem bl y was depleted.
If 950 ppm is exceeded, a s stated in Section 10.2, a bumup Averaged Soluble .:S 1200 ppm pe n a l ty of0.2 , 0.3 , 0.6, and 0.9 GWd/T must Boron be ad ded to the MRB for Categories 2 , 3, 4 , an d 5 respectively.
Average Power To cover reduced power operation at end of Durin g the Last 30 >50% c y cle prior to offload Days of Operation NET-2809 1-000 3-01 , R ev i s ion 0 165 References
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[2] " Response to Request for Additional In formation Regarding the Indi an Point uclear Generating Unit No. 2 -Spent Fuel Pool Critica lit y Analysis ," Letter L 089 , Entergy Nuclear Northeast, Buchanan , NY , August 14 , 2015. (Accession Number: ML15261A527 , Non Proprietary
 
==Attachment:==
 
ML15261A528)
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[4) Guidan ce for P e rforming Criti c ality Anal yses of Fu e l Storag e at Li g ht-Wat e r R e a c tor P owe r Plant s, Revision 2, Draft C , NEI 12-16 , Nuclear Energy Institute , Wash in gton , DC , August 2017. [5] K. Wood , "Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01 , Revision 0 , Staff G uidance Regarding th e Nuclear Criticality Safety Analysis For Spent Fuel Poo l s ," Access ion Number MLl 106 20086, Nuclear Regulatory Commiss ion , Rockvi ll e , MD , October 2011. [6] Code of Federal Regulations , Title l 0 , Part 50 , Section 68 , " Cr iti ca lit y Accident Requirements." [7] S ca l e: A Compr e h e nsiv e Mod e lin g and Simulation Suit e for Nucl e ar Saf ety Ana l ys i s and D es i g n, ORNL ff M-2005/39 , Vers i on 6.1 , June 20 11. Available from Radiation Safet y Information Computat ion a l Center at Oak Ridge National La b oratory as CCC-785. Includes 6.1.2 update dat e d 2/28/1 3. [8] " Rack Cons tru ction (SH'T. 2) R eg i o n l Storage Rack s ," Drawing Number 398 R ev 5 , March 1 9 , 1990 , Project No. 81000 , P. 0. No. 8-24470 , Holtec International , Mount Laure l , NJ. [9] " Rack Co n structio n (S H'T. I) Region 2 Storage R acks ," Drawing Number 400 R ev 5 , March 19 , 1990 , Project No. 81000, P. 0. No. 8-24470 , Holtec Int ernationa l , Mount Laure l , NJ. [ 1 OJ I N TRPN D 8: V e rification and Validation, CWN D , NETCO , Danbury , CT: November 20 1 7 [ 11] Indian Point Unit 3 Up dated Fina l Safety Analysis R eport (UFSA R), Revision 03, October 3, 2009. [12) Indian Point Unit 2 Up d ated Fina l Safety Analysis R eport (UFSAR), Revision 26 , 2016. NET-28 091-000 3-01 , Revi s ion 0 166 
[ 13] Dale Lancaster, S e nsitivity Analys i s for Sp e nt Fu e l Pool Criticality, EPRI , Palo Alto, CA: 2014 , Technical Report Number: 3002003073. [14] C. E. Sanders and J.C. Wagner , Param e tri c Study of th e Eff ec t of Control Rods for PWR Burnup Cr e dit , NUREG/CR-6759, prepared for the US Nuclear Regulatory Commission by Oak Ridge National Laboratory , Oak Rid ge, Tenn., February 2002. [15] G. Radulescu , I. C. Gauld, and G. Ila s, SCALE 5.1 Pr e di c tion s of PWR Sp e nt Nuclear Fu e l I sotopic Composition s, ORNL/TM-2010
/44 , Oak Ridge National Laboratory, Oak Ridge, Tennessee , USA , March 2010. [ 16] Int e rnational Handbook of E v aluat e d Criti ca lity Saf ety B enc hmark E x p e rim e nts , NEA/NSC/DOC(95)3, Nuclear Energy Agency , OECD , Paris , September , 2010. [ 17] J. J. Lichtenwalter, S. M. Bo w man , M. D. DeHart , and C. M. Hopper , Criti c ality B e n c hmark Guid e for Li g ht-Wat e r-R e a c tor Fu e l in Tran s portation and Storag e Pa c kag es, NUREG/CR-6361 (ORNL/TM-13211), Spent Fuel Project Office, Office of Nuclear Material Safety a nd Safeguards , U.S. Nuclear R egu latory Commission , Washington , DC 20555-0001 , March 1997. [18] L. W. Newman , et al , Urania-Gadolinia
: N ucl e ar Mod e l D eve lopm e nt and Critical Experiment B e n c hmark , BA W-1810 , Babcock & Wilcox , Utility Power Generation Divi s ion , Lynchburg , VA , April 1984. [19] G. Radulescu , I. C. Gauld , G. Ila s, and J.C. Wagner , An Approac h for Validating Actin id e and Fission Produ c t Burnup Cr e dit Criticality Saf ety Analys es -Isotopic Composition Pr e dictions, NUREG I CR-7108 , Office of Nuclear Regulatory Research , U.S. Nuclear Regulatory Commission , Washington, DC , USA , April 2012. [20] K. S. Smith , et al., B e n chmarks for Quant ifying Fuel R e a c tivity D e pl e tion Uncertainty , E PRI , Palo Alto, CA , Technical Report Number 1022909 (201 1). [21] D. B. Lancaster , Utilization of th e EPRJ D e pl e tion B e n chmarks for Burnup Credit Validation, EPRI, Palo Alto, CA, 1025203 (2012). [22] J.M. Scaglione , D. E. Mueller , J.C. Wagner and W. J. Marshall, An Approac h for Validating.
Actinide and Fi ss ion Produ c t Burnup Cr e dit Criti c ality Saf ety Ana lys esCriticality (k , JJ) Pr e di c tions , US Nuclear Regulatory Commission, NUREG/CR-7109 , Oak Ridge National Laboratory , Oak Rid ge, Tenn. (2012). [23] J.C. Dean and R.W. Tayloe, Jr., Guid e for Validation of Nuclear Criti ca l ity Safety Cal c ulational M e thodology , NUREG/CR-6698 , Nuclear Regulatory Commission , Washington , DC January 2001. [24] D. E. Mueller , K. R. Elam , and P. B. Fox , Evaluation of th e Fr e n c h Haut Taux d e Combustion (HTC) Critical Exp e rim e nt Data , NUREG/CR-6979 (ORNL/TM-2007
/0 83), prepared for the US Nuclear Regulatory Commission by Oak Ridge National Laboratory , Oak Ridge, Tenn., September 2008. NET-28091-0003-01 , Revision 0 167 (25] Design Input Record , EN-DC-149Rl4 14021 8, Indian Point , February 17, 2014 (26] B. B. Bevard , J.C. Wagner , and C. V. Parks , R ev iew of Information for Spent Nuclear Fu e l Burnup Confirmation, NUREG/CR-6998, prepared for the US Nuclear Regulatory Commission by Oak Ridge Nationa l Laboratory, Oak Ridge , Tenn., December 2009. (27] J.C. Wagner , M. D. DeHart, and , C. V. Parks , R eco mm endatio n s for Addressi ng Axial Burnup in PWR Burnup Cr e dit Analys es, US Nuclear Regulatory Commission, NUREG/CR-6801 , Oak Ridge National Laboratory, Oak Ridge , Tenn. (2003). (28] D. Hagrman , INTERPIN-3 User's Manual, SSP-01/430, Studsvik Scandpower, Inc. (2001). (29] R. E. Griffith to G. Delfini , "F uel Temperatures vs Burnup Curves for Indian Point Unit 2," Entergy lnter-Office Correspondence, CE020 13-001 07, August 15, 2013, s upp orted by Entergy Calculatio n package NEAD-SR-13
/023 , EC 46111. [30] Criticality Saf ety Evaluation of th e North Anna N ew Fu e l Storag e Area and Spent Fu e l Pool Allowing 5 wt% U-235 Enriched Fu e l , Nuclear Engineering and Fuel, Dominion Resources Services, Inc., November 2016. (Accession Number: MLl 7129A452)
(31] J.C. Wagner and C. V. Parks , Param e tric Stud y of th e Eff ec t of Burnabl e Poi so n Rod s for PWR Burnup Cr e dit, US Nuclear Regulatory Commission, NUREG/CR-676 1 , Oak Ridge Nationa l Laboratory, Oak Ri dge, Tenn. (2002). (32] C. E. Sa nd ers and J.C. Wagner , Study of the Effect of Int egra l Burnabl e Poison Rod s for PWR Burnup Cr e dit , US Nuclear Regulatory Commission, NUREG/CR-6760, Oak Ridge Nationa l Laboratory , Oak Ridge , Tenn. (2002). (33] C. V. Parks, M. D. DeHart, and J.C. Wagner , R e view and Prioriti z ation of Technical Issu es R e lat e d to Burnup Cr e dit for LWR Fu e l , US Nuclear Regulatory Commission , NUREG/CR-6665, Oak Ridge Nationa l Laboratory, Oak Rid ge, Tenn. (2000). (34] M. D. D eHart, S e nsitivity and Param e tri c Evaluations of Significant Aspects of Burnup Cr e dit for PWR Spent Fu e l Pa ckages, ORNL/TM-12973 , Lockheed Martin E n ergy Research Corp., Oak Ridge National Laboratory, May 1996. (35] "Poo l Layout Spent Fuel Storage Racks ," Drawing Numb er 397 Rev 4 , December 8, 1989 , Project No. 81000 , P. 0. No. 8-24470, Ho l tec International , Mount Laurel, NJ. (36] "Critica li ty Analysis for Soluble Boron and Burnup Credit in the Con Edison [the former licensee]
Indian Point Unit No. 2 Spe nt Fuel Storage Racks , NET-173-01, Septe mb er, 2001. (Accession Num b er: MLO 12680336)
(37] G.P. Sabo l , G. Schoenberger , and M.G. Balfour, " Impro ved PWR Fue l Claddi ng ," Materials for Adva n ce d Water Coo l ed Reactors, Proceedings of a Technical Committee Meeting , Plzen, Czechoslovakia , May 14-17 , 1991 , IAEA-TECDOC-665 , IAEA , VIENNA, 1992. NET-2809 1-000 3-01, R evis ion 0 168 
[38] Sabol, G. P , Comstock, R. J., Weiner, R. A., Larouere, E, and Stanutz , R. N., "In-R eactor Corros ion Performance of ZIRLO and Zircaloy-4," Zir con ium in th e Nuclear Indust ry: T e nth Int e rnational S ym posium , ASTM STP 1245 , A. M. Garde and E. R. Bradley , Eds., American Society for Testing and Material s, Philadelphia , 1994 , pp. 724-744. [39] Garzaro lli , F., Manze l , R., Re schke, S., and Tenckhoff, E., "Review of Co rro sio n and Dimensional Behavior of Zircaloy under Water Reactor Conditions," Zirconium in th e Nuclear Indu stry (Fourth Conf e r e nc e), ASTM STP 681, Amer i can Society for Testing and Materials , 1979 , pp. 91-106. [ 40] Y. Irisa , et a l , "Seg mented Fuel Rod Irradiation Program On Advanced Mate ri a l s For High Burnup ," An Int erna tional Topical M ee ting on Li ght Wat e r R eactor Fu e l Performanc e, Park City, Utah , Apri l 10-13 , 2000, American Nuclear Society, La Grange Park , Illin ois. [41] C. B. Lee, et al , " Post-irradiation Examination of High Bumup U02 Fuel ," Proceeding s of the 2004 International Meeting on L WR Fuel Performance , Orlando, Florida, September 19-22 , 2004, American Nuclear Society, La Grange Park , Illinois. [42] David Mitchell , Anand Gard e, and Dennis Da v is , "Optimized ZlRLO&#x17d; Fue l Perfom1ance in Westi n g hou se PWRs ," Proceedings of th e 2010 LWR Fu e l P erfo rman ce Me e ting/Top Fu e l/WRFPM, Septem b er 26-29, 20 10
* Orlando , Florida , USA, A m erican Nuclear Society, La Grange Park , Illin ois. [43] R. Manzel and C. T. Walker , " High Burnup Fue l Microstructure And lts Effect On Fuel Rod Performance
," An Int ernational Topical M ee ting on Light Wat e r R eac tor Fuel P erfo rmanc e, Park City, Utah , Apri l 10-13 , 2000, American Nuclear Society, La Grange Park, Illin ois. [44) Dennis Gottuso , Jean-Noel Canat , Pierre Mollard , " A Family of Upgraded Fuel Assemb li es for PWR ," Top Fu e l 2006 , 2006 Int e rnational M ee ting on LWR Fue l P e rformanc e, October 22-26, 2006, Sa l amanca, Spain , European Nuclear Society. [45] King , S. J., Kesterson , R. L., Yueh, K. H., Comstock, R. J., Herwig , W. M., and Ferguson, S. D., "Impact of H y drogen on Dimensional Stability of ZIRLO Fue l Assemb li es," Zirconium in th e Nuclear Indu stry: Thirt ee nth Int e rnational Symposium , ASTM STP 1423 , G.D. Moan and P. Rudling , Eds., ASTM Internation al, West Conshohocken , PA , 2002 , pp. 471-489. [ 46] R. L. Kesterson , S. J. King , and R. J. Comstock , " Imp act of Hydrogen on Dimensional Stab ilit y of Fuel Assemb li es," An Int e rnational Topical M eet ing on Light Wat e r R e actor Fu e l P e rformanc e , Park City, Utah, April 10-13 , 2000 , America n N ucl ear Society, La Grange Park , Illinois. [ 47] Morize , P., Baicry , J., and Mardon, J. P., "Effect oflrradiation at 588 Kon Mechanical Properties and Deformation Behavior of Zirconium Alloy Strip," Zirconium in th e Nuclear Indu stry: Seventh Int e rnational Symposium, ASTM STP 939. R. B. Adamson and L. F. P. Van Swam, E d s., American Society for Testing and Materials , Philadelphia , 1987 , pp. 101-119. NET-2809 1-00 03-01, Revision 0 169 j 
[ 48] S t eve n J. Kin g, Micha e l Y. Young, Fabrice M. Guerout , a nd Ni ge l J. Fisher, " FrettinWear Beha vior of Z ircaloy-4 , OPTIN , and Z IRLO Fuel Rods and Grid Supports Und er Var ious Autoclave and H y drauli c Loop E nduranc e Test Co ndition s," Zi rconium in the N ucle ar Indu stry: Fourteenth International Symposium , ASTM STP 14 67 , P. Rudlin g and Bruce Kamm enzi nd , E d s., ASTM Int e rn ationa l , W est Cons hoho cken , PA, 2006, p. 826. [49] R e spo ns e To R e qu e s t For Additional Information R e garding Pr opos e d Technical Sp e cification Chang e For Sp e nt Fu e l Storag e (Non-Pr oprietary), R espo n se to RAI-23 , Lette r from D omi nion Nuclear Co nn ect i cut to the USNRC , Jul y 2 1 , 2015. (NRC Adams Access ion Number: ML15209 A729.) [50] G.E. Whit esi d es , "A Difficult y in Comp utin g th e k-Effective of th e World," Trans. Am. Nucl. Soc., 14 , pp. 680 (1971). [51] Brian C. Ki edrowski a nd Forrest B. Bro wn , " Diffi c ulti es Com putin g kin Non-Unifonn, Multi-Region Systems with Loose, Asymmet ri c Cou plin g," Pro cee din gs of th e 9th Int e rnational Conf e r e n ce on Nucl e ar Criti c ality Saf e ty (ICNC20 1 l), Edin bur g h , Scotland, September 19-22 , 2011. [52] NET-173-02. Rev. l , " Indian Point Unit 2 Spent Fuel Pool (SF P) Boron Dilution Ana l ys i s , September 2001. (Access ion Number: ML012680336)
[53] ANSVANS-8.1-199 8 (R2007), " Nucl ear Crit i ca lit y Safety in Operations w ith Fissionable Materials Outside R eac tor s ," A m er ican Nuclear Society , La Grange Park , Illinoi s. [ 54] Calc ulation Notebook, NET-901-02-08, SCALE 6.1. 2 Validation for Critica li ty Anal ys i s -Am e ndm e nt Ca l c ulations for IP 2 , R ev.0; CWN D , D anbury , CT. [55] K. L indqui st, et a l , Guid e lin e s for B ora.fl e x U se in Sp e nt-Fu e l Storag e Racks, E PRI , Palo A lto , CA, T echn ic a l R e port Number 103300 (1983). [56] NRC Memorandum from L. Kopp t o T. Co llins , " Guidance on the Regulatory Requirements for C riticali ty A naly s i s of Fue l S tora ge at Lig ht-W a t er R eac tor Power Plants," August 1 9, 199 8. (Access ion Number: MLl 1 088A0 1 3) [57] J.C. Wagner , " Impact of Soluble Boron Modeling for PWR Burnup Cre dit Criticality Safe t y Analyses," Trans. Am. Nucl. Soc., 89 , pp. 120 (2003). [5 8] D.R. Lide , CRC Handbook of Chemistr y and Physics , 85 1 h Edition , C RC Pr ess LLC, Boca Raton , FL. (2004). NET-28091-0003-01, R ev i sio n 0 170 Appendix A: Validation of SCALE 6.1.2 for Criticality Analysis Using Laboratory Critical Experiments A. 1. Overview This appendix detennines the computer code and cross-section library bia s and uncertainty in the k eff values calculated for the Indian Point Units 2 and 3 spent fuel pools when using SCAL E 6.1.2. [ 1] The bias and uncertainties detennined in this Appendix cover the major actinide s, absorbers , and structural materials for the spent fuel pool with fresh or burned fuel. This analysis use s the CSAS5 module of SCALE 6.1.2. [ 1] All of the analyses are performed using the 238 group ENDF/B-VII library (v7-238). The CSAS5 module executes the CENTRM and BONAM! programs for the resonance self-shielding calculations and KENO V.a for the Monte Carlo calculation of k. All of the computer runs use a large Monte Carlo sampling of at least 1500 generations and 6000 neutrons per generation.
This Appendix is divided into three sections:
: 1) U0 2 critical experiments, and 2) HTC and MOX critical experiments, and 3) Temperature Dependent critica l experiments. After these three sections is a summary section. A.2. U02 Laboratory Critical Experiments A.2.1 Introduction The validation consists of modeling 328 U0 2 critica l experiments and the determination of the bias and the uncertainty in the calculation of keff for U0 2 fuel. This v alidation follows the direction of NUREG/CR-6698 , " Guide for Validat ion of Nuc l ear Criticality Safety Calculational Methodology" [2]. The guide establishes the following steps for performing the validation
: 1. Define operation/process to identify the range of parameters to be validated
: 2. Select critica l experiment data 3. Model the experiments
: 4. Analyze the data 5. Define the area of applicability of the va lid ation and limit ations It further defines the steps of "Analyze the data" as: 1. Detennine the Bias and Bias Uncertainty NET-28091-0003-01, Revision 0 A-1 
: 2. Identify Trends in Data, Inc l uding Discussion of Methods for Establishing Bias Trends 3. Test for Norma l or Other Distributions
: 4. Select the Statistical Method for Treatment of Data 5. Identify and S u pport Subcr it ica l Margin 6. Calculate the Upper Safety Limit This approach is followed for this va l idation analysis.
A.2.2 Definition of the Range of Parameters to Be Validated The validation guidance document [2] states: "Prior to th e initiation of the validation activity, th e op e rating c onditions and parameters for which th e validation is to apply must b e identifi e d. Th e fissile isotop e , e nrichm e nt of fissile isotop e, fu e l d e nsity , fu e l c h e mical form , typ e s of n e utron mod e rator s and r e fl ec tors , range of mod e rator to fissil e i s otop e , n e utron ab s orb e rs , and ph y sical c onfigurations ar e among th e param e t e rs to specify. Th e s e param e t e r s will com e to d e fine th e ar e a of applicability for th e validation e ffort. " Almost all pool applications have common neutronic characteristics and therefore can be validated together.
The racks are assumed to be flooded with water at near room temperature and below 100 &deg;C. The fuel is l ow enriched u ranium dioxide (less than or equal to 5.0 wt% U-235). The fuel is in pellets with a density of greater than 92% of the theoretica l density. The only significant neutron moderators are water and the oxygen in the fuel pellet. The neutron absorbers credited are boron (as plates, perhaps rods , or in so l ution) and Ag-In-Cd contro l rods. The reflectors are water, steel , or concrete. The fue l is in assemblies , but the ana l ysis is also valid for disassembled assemb l ies. The assembly arrangement can vary by design from totally isolated assemblies to a close packed array of assemblies. A.2.3 Selection of the Fresh U02 Critical Benchmark Experiments The U0 2 benchmarks that were se l ected met the fo ll owing criteria:
* Low enriched (5 wt% U-235 or less) U0 2 to cover the principle isotopes of concern.
* Fuel in rods to assure that the heterogeneous analysis used in SCALE also is applied in the benchmark analysis.
* Square lattices to assure the lattice features of SCALE used in the rack analysis are also modeled in the critical benchmarks selected.
* Presence of boron as soluble boron , borated steel , boron bearing rods , sheets of aluminum with boron , or Boraflex TM.
* Presence of Ag-In-Cd control rods.
* No emphasis on a feature or materia l not of importance to the rack analysis. The OECD/NEA Int e rnational Handbook of Evaluat e d Criticality Saf e ty B e nchmarks Experiments
[3] is now considered as the appropriate reference for criticality safety benchmarks.
This handbook has reviewed the available benchmarks and evaluated the uncertainties in the experiments.
The appropriate modeling is presented. All of the experiments used in this validation except some experiments for Ag-In-NET-28091-0003-01, Revision O A-2 Cd control rods were taken from this handbook.
Volume IV of the handbook is for low enriched uranium systems. The section of Volume IV of interest to this validation is the " Thermal Compound Systems." All of the experiments selected are numbered LEU-COMP-THERM-OXX.
This validation will refer to the experiments LEU-COMP-THERM-OXX as just XX where any leading zero is not included.
There are more critical experiments in the handbook that meet the requirements for this validation than would be necessary to use. However, most of the applicable available benchmarks were used. There are 95 sets of benchmarks in the 2016 version of the handbook. 25 of these were eliminated since they were for hexagonal arrays. Five more were eliminated due to high enrichments. Seven experiments were not for light water moderated U0 2 fuel rods. Four experiments were eliminated due to high uncertainties.
Five more were eliminated since they depend on features not in the spent fuel pool. This leaves 49 benchmark sets of which 35 were used for this validation.
The 14 unused benchmark sets were reviewed to be sure that there was no feature of the experimental set that was missing in the selected 35 sets. The international handbook only had two sets of experiments from one laboratory with Ag-In-Cd control rods. In order to fully ascertain if Ag-In-Cd control rods introduce a bias a search of additional criticals was performed. The first compilation searched was NUREG/CR-6361 [5], which was a report for the validation of SCALE. In that report were Ag-In-Cd critical experiments from two reports, BA W-1810 [6] and WCAP-3269 (data taken from NUREG/CR-6361).
Finally , Lawrence Livermore Laboratory compiled critical experiments from the ANS national meeting. [7] A search of this source did not provide any additional Ag-In-Cd critical experiments. The selected 35 benchmark sets from the international handbook and the 2 additional sets for Ag-In-Cd rods include critical experiments from eight different critical experiment facilities.
The fuel was mainly clad in aluminum but experiments with stainless steel and zirconium cladding were also in the set. The critical benchmark sets generally contained multiple experiments but not all cases from each critical benchmark set is used. In some sets there are experiments that emphasize features that are out of scope of this validation such as lead or copper reflectors. The 37 selected benchmark sets resulted in 328 experiments that are used for the statistical analysis.
85 experiments used boron (soluble or in absorber plates). A later section will evaluate the area of applicability provided by this selection of critical benchmarks. Table A. l provides a summary of all of the low enriched thermal experiments (non-U metal) from the OECD/NEA handbook [3] and why some experiments were not used. Table A.l: Selection Review of OECD/NEA Criticality Benchmarks (All Experiments Start With LEU-COMP-THERM-)
Benchmark Description Lab Selected?
Number WATER-MODERATED U(2.35)02 FUEL l RODS lN 2.032-C M SQUAR E-PIT C HED P N L All 8 ARR A YS WATER-M O DERATED U(4.31)0 2 FUEL 2 RODS lN 2.54-CM SQUAR E-PIT C HED PNL All 5 ARRAYS NET-28091-0003-01 , Revision 0 A-3 Benchmark Descr i ption Lab Selected?
Number WATE R-MODE R ATED U(2.35)02 FU E L No n e. Gd impmit y n o t we ll 3 R O D S IN 1.68 4-C M SQUARE-PITC H ED P N L known. No t b e n c lun a rk ARRAYS (GA D O LI N I UM WATE R q uality. IMPUR IT Y) WATE R-MODERATED U(4.31)02 FUEL No n e. Gd impu1it y n o t we ll 4 RODS I N 1.892-C M SQUARE-PITC H E D P NL kn ow n. No t b e n c hm ark ARRAYS (GA D O LI N I UM WATE R qua li ty. IMPUR I T Y) C RITl CAL EX PERIM EN TS WITH L OW-No n e. No sa mpl e SCALE EN RI C H E D U RAN I UM DI OXIDE FUE L 5 ROD S IN WATER CONTA I NING P N L decks. So lubl e G d n o t u se d in DI S SOL YE O GADOLIN I UM p oo l s. C R I T I CA L ARRAYS OF LOW-EN R I C H ED 6 U02 FUEL R O DS W I T H WA TER-TO-FUEL J AEA A ll 1 8 VO L UME RATIOS R ANGING FROM 1.5 TO 3.0 WAT ER-REF L ECTE D 4.738-WT.%-O nl y 4 cases u se d r es t ar e in 7 EN RI C H ED URAN I UM DI OXIDE FUEL-Ya ldu c RO D ARRAYS h exago n a l arra ys. C R I TI CAL LATT I CES OF U02 FUEL ROD S 8 AN D P ERTURBING RODS I N BORATED B&W A ll 1 7 W A T E R WATER-MODERATED R ECTANGU LAR C L US TER S OF U(4.3 1)02 FUEL RODS (2.54-2 1 cas es u se d. Did n ot 9 C M PI TCH) SE PARAT E D BY S T EE L , P NL includ e Co pp e r c a ses s in ce n o BORAL , CO PP ER , C ADMI UM, ALU MINUM , coppe r in p oo l s. OR ZrRC ALOY-4 P LATES WATER-MODERATED U(4.3 1)02 FUEL 22 cas es u se d. Did n o t u s e 1 0 R O D S R EF LECT E D BY TWO LEA D , P NL l ea d cases s in ce n o l ea d in URANIUM , OR S T EE L WAL LS poo l s. C RITl CAL EX P ER IM EN TS SUPPORTING II CLOSE PR OXIM I TY WATER STORAGE OF B&W A ll 1 5 POWER REACTOR FUEL (PART I -ABSO RB ER ROD S) WATER-MODERATED RECTANGULAR CLUSTE RS OF U(2.35)02 FUE L No n e. Gd impurit y n ot well 1 2 ROD S(l.6 84-C M P I T C H) SE P ARATE D BY P N L kn own. Not b e n c lun a rk STEEL , BORAL, BOROFLEX , qua li ty. CADM I UM , OR CO PP ER PLATES /G A DOLI N I UM WATER I MPUR I TY) WATER-M ODE RAT E D RE CTANGU L AR CLUSTERS OF U(4.3 1)02 FUEL ROD S 1 3 (1.892-CM PITCH) SE P ARATED BY STEEL , P NL 5 cases u se d. Did n o t u se th e BORAL , BORO FL EX , CADM I UM, OR 2 ca s e s w ith co pp e r. COP P ER PLAT ES , WITH STEE L REFLECTING WALLS WATER-REF L ECTE D ARRAYS OF No n e u s ed. Hi g h b o ron U(4.3 1)02 FUEL RODS (1.890-CM AND 14 1.7 1 5-CM SQUARE P I T C H) I N BORA T E D P NL co nt e nt un ce rtain ty. Not WATER b e n c lunark qua l it y. TH E VVER EX P E RIM ENTS: REGU L A R 1 5 AN D P ERTU RBED H EXAGONA L KFK I No n e u sed due to h ex a rr ays. L A TTI CES OF LOW-EN R I C HED U02 FU E L RODS IN LI G H T WATER WATER-MODERATE D RE CTANGU L AR C LUST E RS OF U(2.35)0 2 F UE L RODS 26 cases u se d. Did n ot u se 1 6 (2.032-CM PITCH) SE P ARATED BY S T EEL , P N L th e 6 co pp e r cases BORA L , COPPE R , CADMnJM , ALU MIN U M , OR Z rRCA L OY -4 PL ATES WATER-M ODE RAT E D U(2.35)02 FU E L 23 cases u se d. Did n o t u se 1 7 RODS REFLECTED BY TWO LEAD , P NL th e 6 cases w i th a l ead URAN I UM , OR STEE L WALLS r e fl ec t o r. NET-28091-0003-01 , Re v ision 0 A-4 Benchmark Description Lab Selected?
Number LIG H T WATER MO D ERATED AND 1 8 REFLECTED LOW ENR I CHED URAN I UM W i nfrith No n e used. Co m p l ex system. D I OXlDE (7 WT.%) ROD LATT I CE WATER-MODERATE D H EXAGONALLY 1 9 PIT CHED LATT I CES OF U(5%)02 K u rc h a t ov In stitute No n e used d u e to h ex arrays. S T AINLESS STEEL C L AD FUEL RO D S WATER-MODERATE D H EXAGONALLY 20 PIT CHED PART I A LL Y FLOODE D K u rc h a t ov In s t itute None used d u e t o h ex arrays. LATT I CES OF U(5%)02 Z I RCONIUM CLAD FUEL RODS , 1.3-CM PIT C H H EXAGONALLY P I TC H ED PART I ALLY FLOODED LATT I CES OF U(5%)02 2 1 ZIRCONIUM CLA D FUEL RODS K u rc h a t ov In s titute None u sed d u e t o h ex arrays. MO D E R ATED BY WATE R WITH BOR I C AC ID UN I FORM WATER-MODERATED 22 H EXAGONALLY P I TC H ED LATT I CES OF K u rc h atov Institute None u sed due to h ex a r rays. RODS WITH U(l 0%)02 FUEL 23 P ARTIALLY FLOO D ED UN I FORM K u rc h a t ov Institute No n e u se d du e t o h ex arrays. LATT I CES OF RODS W ITH U(l 0%)02 FUEL WATER-MODERATE D SQUARE-P I TCHED Did n ot u se e ith e r case du e t o 2 4 UN I FORM LATT I CES OF RODS W I TH K ur c h a t ov In stitute I O wt% U-235 e nri c lu nent U( I 0%)02 FUEL WATER-MODERATE D H EXAGONAL L Y 25 PIT CHED LATT I CES OF U(7.5%)02 Ku r c h atov Institute None u se d d u e t o h ex arrays. STAINLESS-STEEL-C L AD FUEL RODS WATER-MODERATED U(4.92)02 FUEL 26 RO D S IN 1.29 , 1.09 , AN D 1.01 CM P I TC H I PPE No n e used du e t o h ex a 1 rnys. H EXAGONAL LAT T I CES AT DIFFEREN T TEM P ERATURES WATER-MODERATED AND LEAD-None use d du e t o l ead 27 REFLECTED 4.738% ENR I C H ED URAN I UM Va l duc reflecto r. DIOXIDE ROD ARRAYS WATER-MODERATED U(4.3 1)02 FUEL 28 RO D S IN TRIANGU L AR LATTICES W l T H PNL None used d u e t o h ex arrays. BORON , CADMIUM AN D GADO LI N I UM AS SOLUB L E P O I SONS WATER MODERATED AND WATER No n e used. hf pl a t es cases REFLECTED 4.74% ENR I CHED URAN I UM without Hf h ave th e same 29 D I OX ID E ROD ARRAYS SURROUNDED BY Va l duc p i tc h and pin as b e n clunark 7 HAFN I UM P LATES above. No s i g nifi ca n t additio n a l va l u e. VVER Phys i cs Expe rim e nt s: Reg ul ar H exago n a l ( l.27-c m Pit c h) Latt i ces of L ow-30 E nri c h ed U(3.5 Wt.% 235U)02 F u e l Ro d s in K ur c h a t ov In st i tute No n e u sed du e t o h ex a 1 rny s. L i g h t Water at D i ffe r e nt Core Cr iti ca l D im e n s i o n s WATER-MODERATE D H EXAGONALLY 3 1 PI TCHED PARTIALLY FLOODED Ku r c h a t ov In stitute No n e u se d due t o h ex arrays. LATT I CES OF U(5%)02 ZIRCON I UM-C L AD FUEL RODS , 0.8-CM P I TC H UN I FORM WATER-MODERATED 32 LATT I CES OF RO D S W ITH U{l0%)02 F UEL K ur c h atov In stitu t e No n e used d u e t o h ex arrays. IN RANGE FROM 20&deg;c TO 274&deg;C REFLECTED AND UNREFLECTED 33 ASSEMBLIES OF 2 AN D 3%-ENR I CHED ORNL None used. NotU0 2 URAN I UM FLUOR I DE I N PARAFFIN FOU R 4.738-WT.%-EN RI CHED URAN I UM 6 cases use d. Di d n o t use D l OX lD E ROD ASSEMB L IES CONTA I NE D cases wit h gap l ess th a n 2.5 3 4 IN CADMIUM , BO R ATE D STAIN L ESS Va l du c c m du e to hi g h un ce rt a in ty. STEE L , OR BORAL SQUARE CAN I STERS , D i d not u se C d pl ate cases WATER-MODERATED AND-REFLECTED s i nce Cd pl ates n o t in poo l. N ET-28 091-0003-01 , R ev i s i o n 0 A-5 Benchmark Description Lab Selected?
Number CR I T I CAL ARRAYS OF LOW-ENR I C H E D Use d 2 cases. Did n o t u se th e 35 U02 FUEL RODS IN WATER WIT H J AEA case with di sso l ved Gd. (n ot SOLUBLE GADO LINI UM OR BORON P O I SON l ike poo l). T H E VVER EX P E RCM ENTS: REGULAR 36 AND P ERTU RB E D H EXAGONA L KFKI No n e u se d due t o h ex arr ays. LATT I CES OF LOW-ENR I C HED U02 FUEL RODS IN LIG H T WATER -P a rt 2 WATER-MODERATED AND PARTIALLY 37 CONCRETE-REFL EC T ED 4.738-WT.%-Va ldu c No n e u sed. No S i g nifi ca nt ENR I CHED URAN I UM DI OXIDE ROD Va lu e add e d. ARRAYS WATER-MODERATED 4.738-WT.%-
No n e u sed. Use d a borat ed 38 EN RI C H ED URAN I UM DI OX LD E ROD Va l duc co n crete reflec t o r (not l ike ARRAYS NEXT T O A BORA TED CONC RETE SCREEN p oo l). IN COM PL ETE ARRAYS OF WATER-39 REFLECTED 4.738-WT.%-ENR I C H E D Va ldu c Used a ll 1 7 cases. U R AN I UM DIOXIDE FUEL-ROD ARRAYS FOU R 4.738-WT.%-EN RI C H ED URAN I UM DIO X ID E ROD ASSE MBLLES CON TA I NED 4 0 IN BOR ATED STA I N L ESS STEEL OR Va ldu c Use d 4 cases. Did n o t u se BORAL SQUARE CAN I STERS , WATE R l ea d re fl ec t or cases. MODERATED AN D REFLECTED BY LEAD OR STEEL STO R AGE AR R AYS OF 3%-EN R!C H E D D id n ot u se the 5 cases du e t o 41 LWR ASSEMBLIES:
THE C R I STO II Ca d arac h e complex geo m e tr y. EX P ER I MENT IN Tl-I E EO L E REACTOR W A T ER-MO D E RATED R ECTANGU LAR C LUST ERS OF U(2.35)02 FUEL RODS 4 2 (1.684-CM P I TC H) SEPARATED BY STEEL , P NL Used 5 cases. Did n o t u se BORAL , BOROFLEX , C ADMIUM , OR cop p e r cases. COPPER PLAT ES, W I TH STEE L REFLECTING WALLS CR ITI CAL LOADING CONF!GURA TIO NS Use d o nl y o n e case. Rest of 4 3 O F TH E IPEN/MB-0 1 REA CTO R W ITH A !P EN cases were n ot s i g nifi ca ntl y H EAVY SS-304 REFLE CTO R diff e rent. C RITI CAL LOADING CON FIGURAT I ONS Use d o nl y o n e case. Rest of 44 OF Tl-I E IP EN/MB-0 1 REACTOR W I T H U02, !P EN cases were n o t s i g nifi ca ntl y S TAINL ESS STEE L AND COPPER RODS differ e nt. PL EX I GLAS OR CONCRETE-REFLECTE D No n e u se d s in ce n o t pin 45 U(4.4 6)308 W I TH H/U=0.77 AN D R oc k y F l ats INT ERS T I T I A L MODERATION geo m etry. C RITI CAL LOAD ING CONF I GURAT I ONS 4 6 OF T H E I PEN/MB-0 1 REA CTO R IP EN Use d 1 7 cases th a t did n ot CONSIDERING TEMPERATURE h ave co oper p i n s. VAR I AT I ON FROM 14&deg;C TO 85&deg;C FUEL TRANSPO RT FLASK C RITI CA L 4 7 BENCHMARK EXPE RCM ENTS WITH LOW-W infr ith No n e u se d. 3 co mpl ex cases. EN RI CHED URAN I U M DI OX ID E FUEL LI G HT WATE R M O D ERATED AN D 48 R E FL ECTED LOW-ENR I C H ED (3 WT.% W infrith A II 5 c a ses u s ed 235U) URANIUM DIOXID E ROD LATT I CES MARACAS PROGR AMME: P OLYT H ENE-REFLECTED C RITICAL CONF I GURA TIO NS No n e u se d. P ow d e r rather 49 W ITH LOW-ENR I C H E D AND LOW-Va ldu c than p e ll e t s. No t s imil ar to MODERATED URAN I UM D I OX ID E p oo l s. P OWDER, U/5)02 14 9SM SOLUTION TANK I N T H E M IDDL E 7 cas es u se d. Did n ot u se 50 OF WATER-MODE R ATED 4.738-W T.%-V aldu c cases w ith di sso l ved Sm. EN RI C H E D U R AN I UM DI OXI D E ROD Thi s i s n o t typica l o f p oo l s. ARRAYS NET-28091-0003-01 , R evision 0 A-6 Benchmark Description Lab Se lected? Number CR I T I CAL EXPERrMEN TS SUPPORT I NG 9 cases used. Di d n o t u se cases wit h the b ora t ed Al 5 1 CLOSE PROXrMITY WATER STORAGE OF B&W p l ates since primai y so ur ce POWER REACTOR FUEL (PART II -I SOLAT I NG PL ATES) li s t ed a high uncertainty in th e boron co nt e nt. URAN I UM DIOX I D E (4.738-WT.%-
52 ENR I CHED) FUEL ROD ARRAYS Val du c None used due to h ex arrays. MODERATED AND REFLECTED BY GADOLIN I UM N I TRATE SOLUT I ON VVER PHY S I CS EXPE R[MENTS: REGULAR HEXAGONAL ( 1.27 CM PITCH) LA TT I CES 53 OF LOW-ENRICHED U(4.4 WT.% 235U)02 K ur c h atov In stitute No n e u sed due t o h ex arrays. FUEL RODS IN UGI-IT WATER AT D I FFERENT CO R E CR ITI CAL D I MENS I ONS CR I T I CAL LOADING CONF I GURA TION S Use d on l y o n e case. Rest of 54 OF T H E IP EN/MB-0 1 R EACTOR W I TH U02, I P EN cases were not s i g nifi ca ntl y AND U02-G d 203 RODS differe nt. U G I-I T-WATER MODERATED AN D Neit h er case u sed. Comp l ex 55 REFLECTED LOW-ENRICHED URAN I UM W infrith (3 wt.% 235U) D I OX ID E ROD LATT I CES geo metry. CR I T I CAL EXPE RIM ENT WITI-I BORAX-V No n e u s ed. Co mpl ex BWR 56 BO I LING WATER REACTOR TYPE FUEL !NL ASSEMB LI ES geo m e try. 4.738-WT.%-EN R I C H E D URAN I UM 57 D I OX I DE FUEL ROD ARRAYS REFLECTED Va ld uc No n e u se d. No S i g nifi ca nt BY WATER fN A DRY STORAGE V a lu e a dd ed. C ONF I GURAT I ON C R I T I CAL LOAD I NG CON FIG URA TI ONS None u se d. No S i g ni ficant 58 OF Tl-I E fPEN/MB-01 REACTOR W!TI-I I P EN L ARGE VO ID I N Tl-IE R EFLECTOR Va lu e ad d ed. 59 Not included in 20 I O Handb ook RBMKGRAPI-I IT E R EACTOR: UN I FORM CONF I GURAT I ONS OF U(I.8 , 2.0 , o r 2.4% 235U)02 FUEL ASSEMBLIES , A D 60 CONF I GURAT IO NS OF U(2.0% 235U)02 Kurc h atov In st i tute None used. RBMK-n ot ASSEMBUES W I T I-I EM PTY C H ANNELS , typ i ca l ofLWRs WATER COLUMNS , A D BORON OR THORIUM ABSORBERS , W ITI-I OR W I THOUT WATER I N C H ANNE L S VVER PHYS I CS EX PERIM ENTS: HEXAGONAL (1.27-CM PI TCH) LATT I CES OF U(4.4 WT.% 235U)02 FUEL RODS I N 6 1 LI G HT WATER, P E RTURB E D BY BORON , K ur c h atov ln s titute No n e u s ed d u e t o h ex a rr ays. H AFN I UM , OR DYSPROSIUM ABSO RB ER RODS , O R BY WATER GAP WITI-I/WITI-IOUT EM PTY ALUMIN I UM TUBES 2.6%-ENR I CI-IED U02 ROD S IN UGI-IT-62 WATER MOD ERA TOR W I TI-I BORA TED JAEA None u s ed. No S i g ni ficant STAINLESS STEEL P LATE: SINGL E Va lu e adde d. ARRAYS UGI-I T-WATE R MODERATED AN D 63 REFLECTED LOW-ENR I C H ED URAN I UM Winfrith No n e u s ed. No S i g n ificant (3 wt.% 235U) DIOXID E R OD LATT I CES Va lu e a dd e d. W ITH DIS CRETE P O I SON-ROD AR R AYS VVER PH YSICS EX PERrME NTS: R EGU LAR H EXAGONAL (1.2 7 CM PIT C H) LATTICES 64 OF LOW-ENR I C HED U(2.4 WT.% 235U)02 K ur c h atov In stitute No n e u sed due t o h ex arrays FUEL RODS fN UGI-IT WATER AT D I FFERENT CORE CRITIC AL DIMENS I ONS NET-28091-0003-01 , R evision 0 A-7 Benchmark Descript i on Lab S el e cted? Number CR I T I CAL CONFIGURAT I ONS OF 2.6%-ENR I C H ED U02 ROD ARRAYS I N LIG H T-No n e u se d. No Sig ni ficant 65 WATER MODERATOR WITH BORA TED JAEA STArNLESS STEEL PLATE: COU PLED Value added. ARRAYS PLEX I GLAS-REF L ECTED, CONCRETE-66 R EF LECTED , OR TH I N STEEL-REFLECTED R ocky Flats None u se d. No t a n array of U(4.46)308 WITH H/U=0.77 AND HEU rods. DR I VERS CRITICAL LOADrNG CONFIGURAT I ONS 67 OF TH E IPEN/MB-0 1 REA CTO R !P EN None u sed s ince Mo l y rods COMPOSED OF FUEL AND are not us e d in pool. MOLYBDENUM RODS PLEX I GLAS-REFLECTED, CONCRETE-68 REFLECTED , OR TH I N STEEL-REFLECTED Rocky Flats None u se d. Not an array of U(4.48)308 WITH H/U=l.25 OR H/U=2.03 rods. AND HEU DR I VERS PLEX I GLAS-REFLECTED U(4.48)308 W I T H None us ed. Not a n array of 69 H/U=l.25 OR H/U=2.03 AND INTERS TI TIAL Rocky Flat s MODERATION rods. VVER P HYS I CS EXPERCMENTS:
REGULAR H EXAGONAL (1.l O-C M P I TCH) LATT I CES 70 OF LOW-ENRI C H E D U(6.5 WT.% 235U)02 K u r c hatov I n s titute None u se d due t o h ex arrays. FUEL RODS rN LIG H T WATER AT D LF FE RENT CORE CRITIC AL D I MENS I ONS LOW MODERATED 4.738-WT.%-
7 1 ENR I C HED URAN I UM DIOXID E FUEL Valduc All 4 cases u sed. ROD ARRAYS UNDER-MODERATED 4.738-WT.%-72 ENR I CHED URAN I UM DIOXID E FUEL Va l du c Used 3 cases. D i d not u se ROD ARRAYS REFLECTED BYWATER OR Po l yethylene re fl ecto r cases. POLYE TH YLENE UNDER-MODERATED 4.738-WT.%-73 ENR I C HED URANIUM DIOXID E FUEL Va l du c None u se d. No S i g n ificant ROD ARRAYS REFLECTED BY WATER Value added. W I TH HETEROGENEITIES MIRTE PROGRAM FOUR 4.738-WT.%-ENR I CHED URAN I UM-D I OXIDE FUEL-None u sed. 2 cases w ith out 74 ROD ARRAYS rN WATER SEPARATED BY Va l du c A CROSS-SHAPED SCREEN OF T I TAN I UM Ti scree n co u l d be u sed. (5 MM AND 1 0 MM THI C K) VVER PHYS I CS EXPER I MENTS: HEXAGONAL (1.1 0 CM PIT C H) LATT I CES 75 OF LOW-ENRI C HED U(6.5 WT.% 235U)02 Kurchatov In s titute None u s ed due to h ex arrays. FUEL RODS rN LIGHT WATER, PERTURB E D BY BORON ABSORBER RODS AND WATER HOL ES LI GHT WATER MODERATED AND 76 R E FLECTED LOW ENR I C HED URAN I UM W in frith No n e u se d. No S i gnifi ca nt (3 WT.% 235U) DIOX I DE ROD LATTICES Value added. W ITH EX-CORE DETECTOR FEA TVRE On l y one case u se d. Rest of 77 CR I T I CAL LOADrNG CON FIG URA T I ONS !PE N cases sa me m ater i a l s with OF THE IPEN/MB-01 REA CTO R sma ll modificat i o n of arrays. Not sufficie ntl v ind e p e n de n t. WATER-MOD E RATED SQUARE-P I TC H ED 78 U(6.90)02 FUEL ROD LA TT I CES WITH 0.52 Sandia None u se d. No S i g n ificant FUEL-TO-WATER VOLUME RAT I O Va l ue adde d. (0.855 CM P I TCH) WATER-MODERATED U(4.31)02 FUEL 79 ROD LATT I CES CONTA I NING RHOD I UM Sandia None u se d d u e to h ex arrays. FO I LS NET-28 0 9 1-0003-01, R ev i s i on 0 A-8 Benchmark Descr ip tion Lab Selected?
Number WATER-MODERATED SQUARE-P I TC H ED None used. No add iti o n a l 80 U(6.90)02 FUEL RO D LATT I CES W ITH 0.67 Sa n d i a S i g ni fica n t Va l u e a dd ed. FUEL TO WATER VOLUME RA TI O P WR TYPE U02 FUEL RODS W ITH ENR I CHMENTS OF 3.5 AND 6.6 WT.% S in g l e case n ot u se. U nu sual 8 1 W I T H BURNABLE ABSORBER ("OTTO ANEX HAHN" NUCLEAR S HI P PROGRAM , case. SECOND CORE) CR ITI CAL LOADING CONF I GURAT I ONS Use d on l y o n e case. Res t of OF TH E IP EN/MB-0 1 REACTOR W I TH LOW 82 ENRICHED FUEL AND BURNABLE !PEN cases were n ot s i g ni fica n t l y POISON RODS differe n t. CR I T I CAL LOADING CONF I GURAT I ONS Used on l y one case. Res t of 83 OF T H E I PEN/MB-0 1 REACTOR W ITH A I PEN cases were n o t s i g nifi ca n t l y BIG CENTRAL VOID diffe r e nt. CR I T I CAL LOADING CONF I GURAT I ONS 84 OF TH E I PEN/MB-0 1 REACTOR W ITH A I PEN Used t h e s in g l e case. CENTRAL C R UC I FORM ROD VVER P H YS I CS EX P ER[MENTS:
REGULAR H EXAGONAL ( 1.27 CM PI TCH) LA TT I CES 85 OF L OW-ENR I C H E D U(6.5 WT.% 235U)02 Kur c h a t ov In st i tute No n e u sed du e t o h ex a rr ays. FUEL RODS IN LI G H T WATER AT D IFFERENT CORE CR I T I C AL D I MENS I ONS VVER P H YS I CS EX P ER I MENTS: H EXAGONAL LATT I CES (1.275 CM PI TC H) 86 O F LOW ENR I C H ED U(3.6 , 4.4 WT.% NR I None used due t o h ex a rr ays. 235U)02 FUEL ASSE M BLIES I N LI G H T WATE R WITH H 3B03 VVER PHYS I CS EXPER I MENTS: H EXAGONAL LATT I CES (1.22-CM PI TC H) 87 OF L OW-ENR I C H ED U(3.6 , 4.4 WT.% NR I No n e u sed d u e t o h ex arrays. U235)02 FUEL ASSE M B L IES I N LI G H T WATER W ITH VAR I ABLE FUEL-ASSEMBLY PI TCH CR I T I CAL LOADING CONF I GURAT I ONS 88 OF TH E [PEN/M B-0 1 R EACTOR WITH I PEN Used a ll 35. HEAVY REFLECTORS COMPOSE D OF CARBON STEEL AND N I CKEL C RI T I CAL LOADING CONF I GURAT I ONS Use d o nl y o n e case. R est of OF TH E IPEN/MB-0 I REACTOR W ITH U02 89 AN D BORATE D STAINLESS STEEL I PEN cases were n o t s i g nifi ca ntl y P LATES differe nt. CR ITI CAL LOADING CONF I GURA TI ONS Use d on l y o n e case. R es t of 90 OF TH E IPEN/MB-0 I REACTOR W ITH U02 I PEN cases we r e n o t s i g nifi ca n t l y AND STA I N L ESS STEEL RODS diffe r e n t. CR ITI CAL LOAD I NG CON FI GURAT I ONS Used on l y o n e case. Res t of 9 1 OF TH E IP EN/MB-0 1 REACTOR W I T H U02 , I PEN cases were n o t s i g nifi ca ntl y STA I N L ESS STEE L AND GD203 R O D S di ffe r e nt. CR I T I CAL LOADING CONF I GURAT I ONS 92 OF T H E IPEN/MB-0 I REACTOR W ITH I PEN Used a ll 6. SOLUB L E BO R ON D EUTE R fUM CR ITI CA L ASSEMBLY W ITl-1 93 1.2% ENR I CHE D U R ANIUM VAR Y I NG P NC Not u sed since cases u se D20 COOLANT VO I D FRACT I ON AN D rathe r th a n H 20 LATT I CE PIT C H VVER PH YS I CS EX P ER[MENTS:
REGULAR H EXAGONAL (1.1 0 CM P I TC H) TWO-9 4 RE GI ON LATT I CES OF L OW-ENR I C H ED K u rc h a t ov In s t itu t e No n e u sed d u e t o h ex arrays. U(6.5 AND 4.4 W T.% 235U)02 FUE L R O D S IN LI GHT WATE R AT DI FFERENT CORE CR I T I CAL DI MENS I ONS 95 No t in c lud e d in th e 20 1 6 H a ndb oo k N E T-28091-000 3-01 , Re v i s i o n 0 A-9 B e nchm ark D escr iptio n La b Se l e c te d? N umb er PARTI A LL Y-R E FLE C TED WA T E R-Us ed a ll 1 9 (Ev e n thou g h MOD E R A T E D SQUARE-PIT C HED 96 U(6.90)02 F UE L ROD L A TI!CE S WITH 0.6 7 Sandi a h igh e nri c hment , adds an FU E L TO WA TER V OLUM E RA TIO indep e nd e nt l ab and E ALF (0.8 00 C M PIT C H) cove ra ge) TITANIUM A ND/OR ALUMTN U M ROD-R E PL ACE M EN T E XP E R I M E NT S I N FULLY-N on e u se d. Hig h enri c hment 9 7 R E FL E C TE D W A T E R-MOD E R A TED S andi a and Ti in m a n y c a ses. No S QUAR E-PIT C H E D U(6.9 0)0 2 FUE L ROD additi o n a l S ignifi ca nt V alu e LATTI C ES WITH 0.67 FUEL TOW AT E R a dd e d. V OL U M E R A T I O (0.8 00 C M P I T C H) A.2.4 Computer Analysis of the U02 Benchmark Critical Experiments SCALE input decks exist on the OECD/NEA handbook [3] disc for many of the critical experiments.
In general, t h ese input decks were used with minor modifications.
None of the decks ( except LCT-96) were for SCALE 6.1.2 or the ENDF/B-VII library. The number of neutrons per generation and the number of generations were, in general , too low. All of the decks were modified to 6000 neutrons per generation and 1500 generations or more. This was sufficient to make the Monte Carlo uncertainty to be 0.0002 or about one tenth the experimental uncertainty. The input decks matched the isotopic content given in the handbook but this was confirmed.
The geometric modeling in the decks also matched the descriptions in the handbook but this too was confirmed.
In short, a l though there was considerab l e help by starting with the input fi l es given in the handbook, the ownership of the files was taken , as required by NUREG/CR-6698 [2] and as stated in section 2.3: For s p ec ific c ritical ex p e rim e nt s, th e fa c ility or s it e ma y choo se to u se input fil es g e n e rat e d e ls e wh e r e to e xp e dit e th e validation pro ce ss. Th e s it e ha s th e r es pon s ibility for e nsurin g that input fil e s and th e options sel ec t e d ar e appropriat e for us e. R eg ardl e ss of th e s ource of the input fil e , th e s it e must hav e r ev i e w e d th e d esc ription of e a c h c riti c al ex p e rim e nt and d e t e rmin e d that th e r e pr e s e ntation of th e e xp e rim e nt , including simplifying assumptions and options, ar e consi s t e nt with th e int e nd e d us e. In oth e r words , th e s it e must a ss um e own e r s hip of th e input fil e. For LCT-8 the input decks were actually 2D mode l s. As part of the Internationa l Handbook independent review of LCT-08 eva l uation , Virgi n ia Dean, perfonned 3D ana l ysis, found a 0.002 bias , and declared it not " s i gnificant." (Appendix D of the evaluation in the International Handbook.)
0.002 is not insignificant when the bias from a ll of the critical experiments given in Reference l is only 0.0024. The LCT-08 evaluation provided all of the detail for a 3D analysis.
It was chosen to reanalyze LCT-08 with a 3D model. With the 3 D modeling the average calc ul ated ke rr of the LCT-08 cases is 0.9978 (the 2D mode l given on the In t ernational Handbook files yie l ded an average k e rr of 0.9970). BAW-1810 [6] reports 23 critical cores. This analysis uses 9 of these cores. Since these cores are not part of t h e i n t ernationa l h andbook, t h e cases were li mited to those re l ated to Ag-In-Cd cases. A ll of the Ag-In-Cd cores were se l ected as well as the cores that were the c l osest match where water holes replaced the Ag-I n-Cd rods. T h e B&W faci li ties were used for 3 evaluations in the International Handbook, LCT-08 , LCT-1 1 , and LCT-51. All three of t hese sets are used in the va l idation. Some BA W-1 8 1 0 cases are NET-28091-0003-01, Revision 0 A-10 used in NUREG/CR-6361 [5]. The input decks for the cases started with the NUREG/CR-6361 decks but changed the Al alloy clad to match the Al alloy atom densities reported in Table 18 of LCT -08. The input ignore s the bottom and top grids. The original source, BA W-1810, was carefully re viewe d to be sure that there was good agreement with the International Handbook and NUREG/CR-6361 input decks. Cases 1-2, 3-4, 5-6, 5A-6A and 8-9 are pairin gs where the only change was interchanging 16 Ag-In-Cd rods for water holes. The ma ximu m difference between any two cases is 0.00075. The mean differenc e is 0.00002 where the uncertaint y in each case i s 0.0000 8. It is clear that there is no significant difference in the ability to predict k e ff when there are Ag-In-Cd control rods pre sent. BA W-1810 do es not assign an experi mental uncertainty.
The mean uncertainties assigned to LCT-08, LCT-11, and LCT-51 are 0.0012 , 0.00251, and 0.00207 respectively. An uncertainty of 0.0025 (the larg est of the three) was assigned.
The WCAP-3269 input deck s are directly from NUREG/CR-6361 with modifications for the newer cross section librar y, sma ll changes in SCALE input format caused by a new er versio n of the code, and mor e neutrons per generation and generations. The uncertaint y assigned to th ese cases i s 0.004. This uncertainty estimate is one of the largest for the entire set of experiments.
The average uncertainty of all of the experiments is 0.0019. Table A.2 shows the results of the analysis of the 328 critical experiments, along with parameters th at are used to check for trends in the results. The spectra l ind ex, the E nergy of the Average Lethargy of the neutrons causing Fission (EALF) is a calculated va lue from the SCALE output. Table A.2: Critical Experiment Results with SCALE 6.1.2 and ENDF/B-VII Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. keff ID No. (wt% U-Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (~ k) LCT-1 1 2.350 1.2 70 2.032 0.09 64 0.003 0.9981 2 2.350 1.2 70 2.032 0.0957 0.00 3 0.9977 3 2.350 1.2 70 2.032 0.0950 0.003 0.9970 4 2.350 1.270 2.032 0.0955 0.00 3 0.9976 5 2.350 1.2 70 2.032 0.0942 0.003 0.9956 6 2.350 1.2 70 2.032 0.0952 0.003 0.9978 7 2.350 1.270 2.032 0.0934 0.0031 0.9974 8 2.350 1.2 70 2.032 0.0945 0.003 0.9964 LCT-2 1 4.310 1.415 2.540 0.11 32 0.002 0.99 71 2 4.310 1.415 2.540 0.1129 0.002 0.9987 3 4.310 1.415 2.540 0.1129 0.002 0.9984 4 4.310 1.415 2.540 0.1119 0.00 1 8 0.9979 5 4.310 1.415 2.540 0.1103 0.001 9 0.9962 LCT-6 1 2.596 1.417 1.849 0.23 66 0.002 0.9977 2 2.596 1.417 1.849 0.2432 0.002 0.9987 3 2.596 1.41 7 1.849 0.2495 0.002 0.9987 4 2.596 1.41 7 1.956 0.181 8 0.002 0.9984 5 2.596 1.41 7 1.956 0.1 871 0.002 0.9986 6 2.596 1.41 7 1.956 0.19 27 0.002 0.9983 7 2.596 1.41 7 1.956 0.19 77 0.002 0.9989 NET-28091-0003-01, Revision 0 A-11 Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. k ,rr ID No. (wt% U-Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (Ll k) 8 2.596 1.41 7 1.956 0.2028 0.002 0.9986 9 2.596 1.417 2.150 0.1359 0.002 0.9988 10 2.596 1.417 2.150 0.13 94 0.002 0.9988 11 2.596 1.417 2.150 0.142 7 0.002 0.9985 12 2.596 1.41 7 2.150 0.1462 0.002 0.9982 13 2.596 1.41 7 2.150 0.1497 0.002 0.9981 14 2.596 1.41 7 2.293 0.114 7 0.002 0.9988 15 2.596 1.417 2.293 0.11 74 0.002 0.9983 16 2.596 1.417 2.293 0.1200 0.002 0.9991 17 2.596 1.417 2.293 0.1228 0.002 0.9987 18 2.596 1.41 7 2.293 0.1254 0.002 0.9985 LCT-7 1 4.738 0.940 1.260 0.2411 0.0014 0.9959 2 4.738 0.940 1.600 0.1 090 0.000 8 0.9980 3 4.738 0.940 2.100 0.0708 0.0007 0.9976 4 4.738 0.940 2.520 0.0605 0.0008 0.9983 LCT-8 1 2.459 1.206 1.636 0.2845 0.0012 0.99 76 2 2.459 1.206 1.636 0.2502 0.001 2 0.9984 3 2.459 1.206 1.636 0.2502 0.0012 0.9990 4 2.459 1.206 1.636 0.2506 0.0012 0.9980 5 2.459 1.206 1.636 0.2506 0.0012 0.9976 6 2.459 1.20 6 1.636 0.2502 0.0012 0.9977 7 2.459 1.206 1.636 0.2502 0.001 2 0.9971 8 2.459 1.206 1.636 0.2486 0.0012 0.9960 9 2.459 1.20 6 1.636 0.2479 0.0012 0.9963 10 2.459 1.206 1.636 0.2 534 0.0012 0.9978 11 2.459 1.206 1.636 0.2586 0.0012 0.9985 12 2.459 1.206 1.636 0.2524 0.0012 0.9985 13 2.459 1.206 1.636 0.2523 0.0012 0.9985 14 2.459 1.206 1.636 0.2547 0.0012 0.9982 15 2.459 1.20 6 1.636 0.2546 0.001 2 0.9980 16 2.459 1.206 1.636 0.2315 0.0012 0.9981 17 2.459 1.20 6 1.636 0.2017 0.001 2 0.9974 LCT-9 1 4.310 1.415 2.540 0.11 27 0.00 21 0.9980 2 4.310 1.4 15 2.540 0.11 22 0.0021 0.9986 3 4.310 1.415 2.540 0.1125 0.00 21 0.99 79 4 4.310 1.415 2.540 0.11 21 0.0021 0.9981 5 4.310 1.415 2.540 0.1136 0.0021 0.9993 6 4.310 1.415 2.540 0.1 127 0.0021 0.9985 7 4.310 1.415 2.540 0.11 37 0.0021 0.9994 8 4.310 1.415 2.540 0.1130 0.0021 0.998 1 9 4.310 1.415 2.540 0.11 35 0.00 21 0.9986 16 4.310 1.415 2.540 0.11 35 0.0021 0.9987 17 4.310 1.415 2.540 0.112 7 0.0021 0.9991 18 4.310 1.415 2.540 0.1138 0.0021 0.9977 19 4.310 1.415 2.540 0.112 9 0.00 21 0.9986 20 4.3 10 1.415 2.540 0.11 37 0.0021 0.9982 21 4.310 1.415 2.540 0.11 29 0.0021 0.9988 22 4.310 1.415 2.540 0.113 8 0.0021 0.9984 23 4.3 10 1.415 2.540 0.1130 0.0021 0.9994 NET-28091-0003
-01 , R evision 0 A-1 2 Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. k err ID No. (wt% U-Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (Li k) 24 4.310 1.415 2.540 0.1122 0.0021 0.9979 25 4.310 1.415 2.540 0.1120 0.0021 0.9983 26 4.310 1.415 2.540 0.1121 0.0021 0.9987 27 4.310 1.415 2.540 0.1119 0.0021 0.998 5 LCT-10 5 4.310 1.415 2.540 0.3547 0.0021 1.0000 6 4.310 1.415 2.540 0.26 15 0.0021 1.0003 7 4.310 1.415 2.540 0.2092 0.0021 1.000 6 8 4.310 1.415 2.540 0.1844 0.0021 0.99 79 9 4.3 10 1.415 2.540 0.1221 0.0021 1.00 07 10 4.310 1.415 2.540 0.1183 0.0021 1.001 3 11 4.310 1.415 2.540 0.1154 0.0021 1.0006 12 4.310 1.415 2.540 0.1122 0.0021 1.000 0 13 4.310 1.415 2.540 0.1105 0.0021 0.9968 14 4.310 1.415 1.892 0.307 1 0.0028 1.0014 15 4.310 1.415 1.892 0.29 50 0.0028 1.0018 16 4.310 1.415 1.892 0.2853 0.0028 1.0021 1 7 4.310 1.415 1.892 0.2 787 0.0028 1.0021 18 4.310 1.415 1.892 0.2749 0.0028 1.0010 19 4.310 1.415 1.892 0.2677 0.0028 1.0008 24 4.310 1.415 1.892 0.5990 0.0028 0.9994 25 4.310 1.41 5 1.892 0.553 6 0.002 8 1.001 0 26 4.310 1.415 1.892 0.512 2 0.0028 1.001 0 27 4.310 1.41 5 1.892 0.4780 0.002 8 1.001 7 28 4.3 10 1.415 1.892 0.4485 0.0028 1.0017 29 4.310 1.415 1.892 0.4232 0.0028 1.0016 30 4.3 10 1.415 1.892 0.36 79 0.0028 0.9996 LCT-1 1 1 2.459 1.206 1.636 0.1 685 0.0018 0.9968 2 2.459 1.206 1.636 0.2450 0.0032 0.9967 3 2.459 1.206 1.636 0.19 20 0.003 2 0.99 7 1 4 2.459 1.206 1.636 0.1 927 0.0032 0.99 72 5 2.459 1.206 1.636 0.1 935 0.0032 0.9970 6 2.459 1.206 1.636 0.1951 0.003 2 0.99 70 7 2.459 1.206 1.636 0.1959 0.0032 0.9967 8 2.459 1.206 1.636 0.1972 0.0032 0.9974 9 2.459 1.206 1.636 0.19 84 0.0032 0.9975 10 2.459 1.2 06 1.636 0.1866 0.0017 0.9945 11 2.459 1.206 1.6 36 0.1628 0.0017 0.9940 12 2.459 1.206 1.636 0.16 70 0.0017 0.9950 13 2.459 1.206 1.636 0.1 475 0.001 7 0.9943 14 2.459 1.206 1.636 0.150 8 0.001 7 0.9946 15 2.459 1.206 1.636 0.1 387 0.0018 0.9959 L C T-13 1 4.3 10 1.415 1.892 0.2862 0.0018 1.0005 2 4.310 1.415 1.892 0.2939 0.0018 1.0004 3 4.310 1.415 1.892 0.297 4 0.0018 1.0003 4 4.310 1.415 1.892 0.2969 0.0018 1.0007 5 4.310 1.415 1.892 0.2961 0.0032 1.0003 LCT-16 l 2.350 1.270 2.032 0.0957 0.0031 0.99 73 2 2.350 1.27 0 2.032 0.0954 0.0031 0.9962 3 2.350 1.270 2.032 0.0954 0.0031 0.9967 NET-28091-0003-01 , Revi sion 0 A-13 B e nchmark C ase E nri c hm e nt F u e l Pin F u e l P i n EALF Meas. k err ID No. (wt% U-Diam ete r Pit c h (c m) (eV) U n ce rt a in ty 23 5) (cm) ,~ k) 4 2.350 1.2 70 2.032 0.095 6 0.0031 0.9960 5 2.350 1.2 70 2.032 0.0952 0.0031 0.9970 6 2.350 1.270 2.032 0.0961 0.0031 0.9971 7 2.350 1.270 2.032 0.0959 0.0031 0.9973 8 2.350 1.270 2.032 0.0969 0.0031 0.9972 9 2.350 1.270 2.032 0.0961 0.0031 0.9977 10 2.350 1.270 2.032 0.09 70 0.0031 0.9971 11 2.350 l.270 2.032 0.0962 0.0031 0.9978 12 2.350 1.270 2.032 0.0974 0.003 1 0.9972 13 2.350 1.270 2.032 0.0965 0.0031 0.9979 14 2.350 1.270 2.032 0.0975 0.0031 0.9974 21 2.350 1.270 2.032 0.0971 0.0031 0.9977 22 2.350 1.270 2.032 0.0968 0.003 1 0.9974 23 2.350 1.270 2.032 0.0963 0.0031 0.9977 24 2.350 1.270 2.032 0.0967 0.0031 0.9970 25 2.350 1.270 2.032 0.0963 0.0031 0.9972 26 2.350 1.270 2.032 0.0969 0.0031 0.9976 27 2.350 1.270 2.032 0.0963 0.0031 0.9979 28 2.350 1.270 2.032 0.0951 0.0031 0.9972 29 2.350 1.270 2.032 0.0950 0.0031 0.9969 30 2.350 1.270 2.032 0.0949 0.0031 0.9965 31 2.350 1.270 2.032 0.0950 0.0031 0.9979 32 2.350 1.2 70 2.032 0.0949 0.0031 0.9972 LCT-1 7 4 2.350 1.270 2.032 0.201 7 0.0031 0.9983 5 2.350 1.270 2.032 0.1779 0.0031 0.9994 6 2.350 1.2 70 2.032 0.1 685 0.0031 0.9989 7 2.350 1.270 2.032 0.1597 0.0031 0.9994 8 2.350 1.270 2.032 0.1333 0.0031 0.9972 9 2.350 1.270 2.032 0.1092 0.0031 0.997 3 10 2.350 1.270 2.032 0.0998 0.0031 0.9973 11 2.350 1.270 2.032 0.09 79 0.0031 0.9979 12 2.350 1.270 2.032 0.0968 0.0031 0.9977 13 2.350 l.270 2.032 0.0953 0.0031 0.99 76 14 2.350 1.270 2.032 0.0946 0.0031 0.9985 15 2.350 1.270 1.684 0.1777 0.0028 0.996 1 16 2.350 1.270 1.684 0.1711 0.0028 0.9983 17 . 2.350 1.270 1.684 0.1665 0.0028 0.9987 18 2.350 1.27 0 1.68 4 0.1 648 0.0028 0.9974 19 2.350 1.270 1.684 0.1622 0.0028 0.9978 20 2.350 1.27 0 1.684 0.1607 0.0028 0.9971 21 2.350 1.270 1.684 0.1592 0.0028 0.9966 22 2.350 1.270 1.684 0.1584 0.0028 0.9959 26 2.350 1.270 1.68 4 0.3741 0.0028 0.9958 27 2.350 1.270 1.68 4 0.3203 0.0028 0.9972 28 2.350 l.270 1.684 0.2806 0.0028 0.9974 29 2.350 1.270 1.684 0.2505 0.0028 0.9984 LCT-34 4 4.738 0.940 1.600 0.1367 0.0039 1.0003 5 4.738 0.940 1.600 0.1 330 0.0039 0.9999 6 4.738 0.940 1.600 0.1298 0.0039 1.0017 NET-2809 1-0003-01 , R evision 0 A-14 Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. keff ID No. (wt% U-Diameter Pitch (cm) (eV) Uncertainty 235) (cm) ,~ k) 7 4.738 0.940 1.600 0.1279 0.0039 1.0002 8 4.738 0.940 1.600 0.1258 0.0039 0.9992 15 4.738 0.940 1.600 0.1348 0.0043 0.9947 LCT-35 1 2.596 1.417 1.956 0.2086 0.0018 0.9983 2 2.596 1.417 1.956 0.2126 0.0019 0.9976 LCT-39 1 4.738 0.940 1.260 0.22 18 0.0014 0.9953 2 4.738 0.940 1.260 0.2119 0.0014 0.9969 3 4.738 0.940 1.260 0.1923 0.0014 0.9965 4 4.738 0.940 1.260 0.1836 0.0014 0.9961 5 4.738 0.940 1.260 0.1393 0.0009 0.9978 6 4.738 0.940 1.260 0.1452 0.0009 0.9977 7 4.738 0.940 1.260 0.2132 0.0012 0.9962 8 4.738 0.940 1.260 0.2031 0.0012 0.9963 9 4.738 0.940 1.260 0.19 76 0.0012 0.9969 10 4.738 0.940 1.260 0.1732 0.0012 0.9970 11 4.738 0.940 1.260 0.22 18 0.0013 0.99 53 12 4.738 0.940 1.260 0.2 166 0.0013 0.9951 13 4.738 0.940 1.260 0.2146 0.0013 0.9951 14 4.738 0.940 1.260 0.2 124 0.0013 0.9954 15 4.738 0.940 1.260 0.2 112 0.0013 0.9959 16 4.738 0.940 1.260 0.2 104 0.0013 0.9967 17 4.738 0.940 1.260 0.2099 0.0013 0.9960 LCT-40 1 4.738 0.940 1.600 0.1427 0.0039 0.9966 5 4.738 0.9 40 1.6 00 0.1377 0.0042 0.9951 9 4.738 0.940 1.600 0.1470 0.0046 0.9993 10 4.738 0.940 1.600 0.1419 0.0046 0.9931 LCT-42 1 2.350 1.270 1.684 0.1690 0.0016 0.9971 2 2.350 1.270 1.684 0.1753 0.0016 0.9968 3 2.350 1.270 1.684 0.1819 0.0016 0.9981 4 2.350 1.270 1.684 0.1804 0.0017 0.9980 5 2.350 1.270 1.684 0.1775 0.0033 0.9981 LCT-43 2 4.349 0.980 1.500 0.1553 0.0010 1.0007 LCT-44 1 4.349 0.980 1.500 0.1474 0.0010 0.9993 LCT-46 1 4.349 0.981 1.500 0.1488 0.00044 0.9991 2 4.349 0.981 1.500 0.1525 0.00044 0.9989 3 4.349 0.981 1.500 0.1542 0.00044 0.9988 4 4.349 0.98 1 1.500 0.1556 0.00044 0.9989 5 4.349 0.981 1.500 0.1573 0.00044 0.9986 6 4.349 0.981 1.500 0.1595 0.00044 0.9987 7 4.349 0.981 1.5 00 0.1479 0.00044 0.9991 8 4.349 0.981 1.500 0.1550 0.00044 0.9988 9 4.349 0.981 1.500 0.1 594 0.00044 0.9987 10 4.349 0.98 1 1.500 0.1621 0.00044 0.9988 11 4.349 0.98 1 1.500 0.1672 0.00044 0.9988 12 4.3 49 0.98 1 1.500 0.15 39 0.00044 0.9986 13 4.349 0.981 1.500 0.1570 0.00044 0.9986 14 4.3 49 0.981 1.500 0.1596 0.00044 0.9986 15 4.349 0.981 1.500 0.1618 0.00044 0.9984 16 4.349 0.981 1.500 0.1655 0.00044 0.9983 NET-28091-0003-01 , R evision 0 A-15 B e nchmark Case E nrichm e nt F u e l Pin F u e l Pin EALF Meas. k.rr ID N o. (wt% U-Diam e t e r Pi t ch (c m) (eV) U n ce rt a in ty 2 35) (c m) (~ k) 17 4.349 0.981 1.500 0.1724 0.00044 0.9983 LCT-48 1 3.005 1.094 1.320 0.6771 0.0025 0.9990 2 3.005 1.094 1.320 0.6508 0.0025 0.9983 3 3.005 1.094 1.320 0.6824 0.0025 0.9984 4 3.005 1.094 1.320 0.6838 0.0025 0.9988 5 3.005 1.094 1.320 0.6736 0.0025 0.9983 LCT-50 1 4.738 0.940 1.300 0.1998 0.0010 0.9983 2 4.738 0.940 1.300 0.1907 0.0010 0.99 78 3 4.738 0.940 1.300 0.2075 0.0010 0.997 8 4 4.738 0.940 1.300 0.1977 0.0010 0.9972 5 4.738 0.940 1.300 0.2230 0.0010 0.9983 6 4.738 0.940 1.300 0.2 1 41 0.0010 0.9991 7 4.738 0.940 1.300 0.2095 0.0010 0.9992 LCT-51 1 ClO 2.459 1.206 1.636 0.1472 0.0020 0.9965 2 cl la 2.459 1.206 1.636 0.1968 0.0024 0.9972 3 cl 1 b 2.459 1.206 1.636 0.1964 0.0024 0.9972 4 cl le 2.459 1.2 06 1.636 0.1979 0.0024 0.9975 5 cl ld 2.459 1.206 1.636 0.1989 0.0024 0.9970 6 cl le 2.459 1.206 1.636 0.1998 0.0024 0.9972 7 cl If 2.459 1.206 1.636 0.2000 0.0024 0.9973 8 cl lg 2.459 1.206 1.636 0.20 11 0.0024 0.9971 9 cl2 2.459 1.206 1.636 0.1669 0.0019 0.9969 LCT-54 I 4.349 0.980 1.500 0.1508 0.0005 0.9996 LCT-71 1 4.738 0.949 1.100 0.7592 0.00076 0.9955 2 4.738 0.949 1.100 0.6972 0.00076 0.9954 3 4.738 0.949 1.100 0.6610 0.000 76 0.9948 4 4.738 0.949 1.075 0.8485 0.0008 0.9 951 LCT-72 l 4.738 0.949 1.600 0.111 7 0.0012 0.9990 2 4.738 0.949 1.600 0.1077 0.0012 0.9985 3 4.738 0.949 1.600 0.1099 0.0012 0.9988 LCT-77 3 4.349 0.980 1.500 0.1621 0.0010 1.000 6 LCT-82 3 4.349 0.980 1.500 0.1497 0.0010 1.0005 LCT-83 1 4.349 0.980 1.500 0.1516 0.0010 1.0001 LCT-84 I 4.349 0.980 1.500 0.1541 0.0010 1.0008 LCT-88 1 4.349 0.98 1 1.500 0.1543 0.00044 0.9993 2 4.349 0.981 1.500 0.1556 0.00044 0.9992 3 4.349 0.981 1.500 0.1561 0.00044 0.9992 4 4.349 0.98 1 1.500 0.1560 0.00044 0.999 7 5 4.349 0.981 1.500 0.1560 0.00044 0.9996 6 4.349 0.981 1.500 0.1560 0.00044 0.9998 7 4.349 0.98 1 1.500 0.1559 0.00044 0.9999 8 4.349 0.981 1.500 0.1560 0.00044 0.9994 9 4.349 0.981 1.500 0.1560 0.00044 0.9994 10 4.349 0.981 1.500 0.1 561 0.00044 0.9989 11 4.349 0.981 1.500 0.1561 0.00044 0.9986 12 4.349 0.981 1.500 0.1563 0.00044 0.9980 13 4.349 0.981 1.500 0.1565 0.00044 0.9975 14 4.349 0.981 1.500 0.1564 0.00044 0.9971 15 4.3 49 0.981 1.500 0.1 566 0.00044 0.9967 NET-28091-0003-01 , R evision 0 A-16 Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. k err ID No. (wt% U-Diam ete r Pitch (cm) (eV) Uncertainty 235) (cm) (d k) 16 4.349 0.981 1.500 0.1 566 0.00044 0.9963 17 4.349 0.981 1.500 0.156 7 0.00044 0.9957 18 4.349 0.981 1.500 0.15 68 0.00044 0.9954 19 4.349 0.981 1.500 0.15 60 0.00044 0.9994 20 4.349 0.981 1.500 0.15 67 0.00044 0.9990 21 4.349 0.981 1.500 0.15 71 0.00044 0.9992 22 4.349 0.981 1.500 0.15 73 0.00044 0.9991 23 4.349 0.981 1.500 0.15 75 0.00044 0.9991 24 4.349 0.981 1.500 0.1576 0.0 0044 0.9991 25 4.349 0.981 1.500 0.1578 0.00044 0.9991 26 4.349 0.981 1.500 0.1578 0.00044 0.9992 27 4.349 0.981 1.500 0.15 79 0.00044 0.9992 28 4.349 0.981 1.500 0.15 80 0.00044 0.9991 29 4.349 0.981 1.500 0.1 582 0.00044 0.9992 30 4.349 0.981 1.500 0.1582 0.00044 0.9995 31 4.349 0.981 1.500 0.1583 0.00044 0.9995 32 4.349 0.981 1.500 0.1 584 0.00044 0.9995 33 4.349 0.9 81 1.500 0.1585 0.00044 0.9994 34 4.349 0.981 1.500 0.1584 0.00044 0.9996 35 4.349 0.981 1.500 0.1 584 0.00044 0.9996 LCT-89 1 4.349 0.980 1.500 0.1 530 0.0010 1.0000 LCT-90 1 4.349 0.980 1.500 0.1 459 0.0010 0.9994 LCT-91 4 4.349 0.980 1.500 0.1 508 0.00 10 0.9999 LCT-92 1 4.349 0.981 1.500 0.1543 0.00044 0.9996 2 4.349 0.981 1.500 0.1545 0.00044 0.9994 3 4.349 0.981 1.500 0.15 45 0.00044 0.9996 4 4.349 0.981 1.500 0.154 9 0.00044 0.9994 5 4.349 0.981 1.500 0.1555 0.00046 0.9988 6 4.349 0.981 1.500 0.1 559 0.00 055 0.9994 LCT-96 1 6.903 0.635 0.800 0.5690 0.00095 0.9973 2 6.903 0.635 0.800 0.5674 0.00095 0.99 72 3 6.903 0.635 0.800 0.4191 0.00095 0.9993 4 6.903 0.635 0.800 0.5704 0.00095 0.99 71 5 6.903 0.635 0.800 0.561 7 0.00095 0.997 1 6 6.903 0.635 0.800 0.549 2 0.00095 0.9968 7 6.903 0.635 0.800 0.5304 0.00095 0.9965 8 6.903 0.635 0.800 0.5068 0.00095 0.9966 9 6.903 0.635 0.800 0.492 9 0.00095 0.9965 10 6.903 0.635 0.800 0.4929 0.00 095 0.9963 11 6.903 0.635 0.800 0.4630 0.00095 0.9974 12 6.903 0.6 35 0.800 0.43 17 0.00095 0.9975 13 6.903 0.635 0.800 0.4032 0.00095 0.9978 14 6.903 0.635 0.800 0.3800 0.00095 0.99 77 15 6.903 0.635 0.800 0.3604 0.00095 0.9979 16 6.903 0.635 0.800 0.4320 0.00095 0.9978 17 6.903 0.635 0.800 0.3756 0.00095 0.9984 1 8 6.903 0.635 0.800 0.33 1 7 0.00095 0.9986 19 6.903 0.635 0.800 0.2997 0.00095 0.9989 BAW-1810 1 2.460 1.206 1.636 0.2477 0.00250 0.9990 NET-2809 1-0003-01, R ev i sion 0 A-17 Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Mea s. k err ID No. (wt% U-Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (d k) 2 2.460 1.206 1.636 0.2470 0.00250 0.99 82 3 2.460 1.20 6 1.636 0.2465 0.00250 0.9 983 4 2.460 1.206 1.636 0.2452 0.00250 0.9991 5 2.460 1.206 1.636 0.24 61 0.00250 0.9982 5A 2.460 1.206 1.636 0.2457 0.00250 0.998 1 6 2.460 1.20 6 1.636 0.2 453 0.00250 0.9981 6A 2.460 1.206 1.636 0.2450 0.002 50 0.9982 8 2.460 1.206 1.636 0.2453 0.00250 0.9982 9 2.460 1.206 1.636 0.244 8 0.00250 0.998 1 WCAP-3269 2.7 2.720 1.18 9 1.524 0.2599 0.00400 0.9988 5.7 5.700 0.993 1.422 0.3006 0.00400 0.9978 3.7-12 3.700 0.860 1.1 05 0.4309 0.00400 0.9985 3.7-24 3.700 0.860 1.105 0.42 88 0.00400 0.9979 3.7-48 3.700 0.860 1.1 05 0.455 8 0.00400 0.9969 Since b oro n credit is u sed it is important validate boron with critica l experiments.
Table A.3 s how s the boron information on b oron-containing benchmarks , alon g with the calculated
: k. Table A.3: Summary of Critical Experiments Containing Boron Benchmark Case No. Soluble Separator Plate No. of kerr ID Boron 10 B Areal Densit y Boron (ppm) (gm/cm 2) Rods LCT-8 1 1511 0.9976 2 1 334 0.9984 3 1 337 0.9990 4 11 83 36 0.9980 5 11 81 36 0.9976 6 10 34 72 0.9977 7 10 31 72 0.997 1 8 794 144 0.9960 9 779 144 0.9963 10 1245 72 0.9978 11 1 384 0.998 5 12 1 348 0.998 5 13 134 8 0.9985 14 1 363 0.9982 15 1 362 0.9980 16 115 8 0.998 1 17 921 0.9974 LCT-9 5 0.004549 0.9993 6 0.00 4549 0.9985 LCT-9 7 0.006904 0.9994 8 0.006904 0.998 1 NET-2809 1-000 3-01, R ev i sion 0 A-1 8 Benchmark Ca se N o. So lubl e Separator Plate N o. of k err ID Boron 10 B Area l Densit y Boron (ppm) (gm/cm 2) Rod s 9 0.066946 0.9986 LCT-11 2 1037 0.9967 3 769 0.997 1 4 764 0.9972 5 762 0.9970 6 753 0.9970 7 739 0.9967 8 721 0.9974 9 702 0.9975 10 84 0.9945 11 64 0.9940 12 64 0.9950 13 34 0.9943 14 34 0.9946 LCT-13 2 0.004549 1.0004 3 0.030173 1.0003 4 0.056950 1.0007 LCT-16 8 0.004549 0.9972 9 0.004549 0.9977 10 0.006904 0.997 1 11 0.006904 0.9978 12 0.066946 0.9972 13 0.066946 0.9979 14 0.066946 0.9974 LCT-34 4 0.002521 1.0003 5 0.002521 0.9999 6 0.00252 1 1.0017 7 0.00252 1 1.0002 8 0.00252 1 0.9992 15 0.04601 1 0.9947 LCT-35 1 70 0.9983 2 147.7 0.99 7 6 LCT-40 1 0.002521 0.9966 5 0.046011 0.9951 9 0.04601 1 0.9993 10 0.046011 0.9931 LCT-42 2 0.004549 0.9968 3 0.030 1 73 0.998 1 4 0.056950 0.9980 LCT-50 3 822 0.9978 4 822 0.9972 5 5030 0.9983 6 5030 0.999 1 7 5030 0.9992 LCT-51 1 ClO 14 3 0.9965 NET-2809 1-00 03-01 , Revision 0 A-19 Benchmark Case No. So lu ble Separator P l ate No. of kerr ID Boron 10 B Area l Densit y Boron (p p m) (gm/cm 2) Ro d s 2 cl la 510 0.9 972 3 cl lb 514 0.9 972 4 cl le 501 0.9975 5 cl ld 493 0.9 970 6 cl le 474 0.9972 7 cl lf 462 0.9973 8 cl lg 432 0.9971 9 cl2 217 0.9969 LCT-77 3 4 1.0006 LCT-82 3 6 1.0005 LCT-92 1 0.1 0.9996 2 6 0.9994 3 11 0.9996 4 22 0.9994 5 43 0.9988 6 95 0.9994 BAW-1 8 10 1 1337 0.9990 2 1250 0.9982 3 12 39 0.9983 4 1171 0.9991 5 120 8 0.9982 SA 1191 0.9981 6 1156 0.9981 6A 1136 0.9982 8 11 71 0.9982 9 1131 0.9981 A.2.5 Statist ic al Ana l ys i s of t he Fresh U0 2 Cri t i cal Ben c h m a r k Res ults The stat i s tical treatment u se d follows th e guidance pro v ided in NUREG/C R-669 8 [2]. The NUREG approach weights the calculated kerr values by the experimental uncertainty.
This approach means the hi gher quality experiments (i.e.: lo we r uncert ai nty-see Table A.2) affect the results more than the lo w quality (i.e.: higher uncertainty) experiments. The uncertainty weighting is used for the analysis of the set of experiments as a whole , as we ll as for th e ana lysi s for trends. Before see king trends the 328 critical benchm ar ks set are reviewed as a whole. The unw eig hted mean k e rr of the 328 samples is 0.9981 with a standard de v iation of0.0015. The weighted mean is 0.9985 and the weighted standard deviation i s 0.0015. The average uncert ai nt y of the experiments (interpreted as one sigma) is 0.0019. Since the tot a l one sigma standard de viatio n i s only 0.0015, this suggests that the experimental uncertainty dominates the uncertainty and there is little to be gained with improved methods. Unles s sta ted otherwi se all of the r esu lt s pre se nted will come from the weighted analysis.
The bias of the set as a who l e is 0.0015. The uncertainty is the standard deviation multiplied by the sided lower tolerance factor (taken as 2.065 from Referenc e 2 for more than 50 samples) so it is 0.0031. NET-28091-0003-01, Revi s ion 0 A-20 As recommended by NUREG/CR-6698, the resu lt s of the validatio n are checked for normality.
T h e Nationa l In stitute of Sta nd ards and Tec hnol ogy (N I ST) h as made publicly avai l ab l e a st a tistical pack a ge , DATAPLOT [4). The 328 critica l experiments were tested with the Wilk-Shapiro normality test and were found to a dh ere to a normal distribution at the 90% l evel. The test results are s h own in Tab l e A.4. A histogram plot of the data is shown on Figure A. l. Table A.4: Wilk-Shapiro Test Results Output From DATAPLOT [4] NET-2809 1-00 03-0 1 , Revision 0 A-21 60 so C iii 40 .l: "' ::.: -a "' 8 30 b .. Ill .a E 20 10 0 Calculated keff Distribution Versus a Normal Distribution Figure A.l: Distribution of the Calculated kcrr values Around the Mean If a feature of a subset of the critical experiments creates a statistically subset , this feature needs to be corrected before combining all of the critical results. There are 89 experiments that have boron in them. The average k eff of the boron containing cases is 0.9978 which is very close to the average of all cases (0.9981 ). Similarly, there are 17 cases that used pure Cadmium absorbers.
The mean of these cases is 0.9982. There are 51 cases that use Ag-In-Cd control rods with a mean of 0.9987. Since the standard deviation of the set as a whole is 0.0015 (unweighted) it is clear these features are not skewing the results. The next step in the analysis is to look for trends in the data. The math will always find a trend but only the real or statistically significant trends are of interest.
Section 3.2.2 of the DOE/RW technical report in support of validation for burnup credit [8] describes an appropriate trend test. In this test, the null hypothesis is that the slope of the trend is zero (no trend) and it tests to determine if there is confidence that the calculated slope is a more accurate representation than a zero slope. The equations for this test are presented here. NET-28091-0003-01, Revision 0 A-22 Let the regression fit be of the form: k= a+ b X Let x-bar be the average value ofx for then cases and define: and define: then the test statistic is: sxx = ~)x; -x)2 i=l,n (n -2)
* Sxx T=\bl* This test statistic is then compared to the Student's t-distribution at the desired confidence level and n-2 degrees of freedom. In the past it was assumed that unless there is a high confidence level (95%) that the slope was non-zero, the analysis would assume a zero slope (no trend) on the given parameter.
Since the analysis will include consideration of the data as non-trended, it is more conservative to assume there is also a trend. Inverting the statistical test to requiring a high confidence that the slope is zero will result in all cases having a trend. At this time, although a test on the confidence of the trend is performed, the analysis assumes all calculated trends are real. For this work the weighted k.,ffvalues are used to determine the fit to a straight line. Refer to NUREG/CR-6698
[2] equations 10 through 13. NUREG/CR-6698
[2] describes the appropriate tolerance band for criticality validation.
This work simply applies the equations (equations 23 to 30) given in the NUREG. Note that the tolerance band is found using the weighted experimental data. The width of the tolerance band is the uncertainty.
In the final analysis, the calculated keff of the system must be less than the minimum of k(x) minus the uncertainty minus the administrative safety margin. The uncertainty in k.,ff from other independent uncertainties, such as the manufacturing tolerances, burnup, and depletion uncertainties can be statistically combined with the uncertainty in the criticality validation.
The rest of this section will evaluate the trends in keff as a function of trending parameters using the methods described above. Historically, an Upper Subcriticality Limit (USL) was assigned from the criticality validation analysis.
This is not done here, since the other uncertainties (e.g., manufacturing tolerances of the rack, depletion uncertainty, etc.) are not known at this time. NET-28091-0003-01, Revision 0 A-23 Neutron spectrum Trends in the calculated k e rr of the benchmarks were so u ght as a function of the neutron spect rum. Since a l arge number of things can affect the spectrum , a s in g l e index calc ul ated by SCALE is used. This index is the E n ergy (eV) of the Average Let har gy caus in g F i ssion (EALF). Figure A.2 s h ows the distribution of k err va lu es around the me an k , which is s h own as the red lin e. Visual inspection of the graph a nd the statistica l analysis of the results of the statistical analysis suggest that there is a statistically s ignific ant trend on neutron spectrum.
Using NUREG/CR-6698 [2] equations 10 through 13 and the data from Table A.2 , the predicted mean ke rr as a function of EALF i s: k(EALF) = 0.999406 -0.00459
* EALF T h e units for EALF are eV. The bias at 0.4 eV is 0.0024. The unc e rtain ty at 0.4 eV i s 0.0030. T h e bias at 0.65 eV i s 0.0036. T h e un certainty at 0.65 is 0.0034. 1.003 -------------1.002 *** 1.001 *
* 1.000 :i: 0.999 QI :li:: -o 0.998 * * ---.----,-----~---------QI ... 1 0.997 .., u 0.996 0.995 0.994 0.993 0.992 ,----T 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 Energy of the Average Lethargy Causing Fission (eV) Figure A.2: k eff as a Function of the Energy of the Average Lethargy Causing Fission NET-28091-0003-01 , Revision 0 A-24 Geometry tests Two trend tests were performed to determine if lattice/geometric parameters are adequately treated by SCALE 6.1.2. The first parameter is the fuel pin diameter.
A small, statistically insignificant trend was found when the critical experiment analysis results were correlated to the fuel pin diameter.
The second lattice parameter tested is the lattice pitch. A statistically significant trend on lattice pitch was found. The trend on pitch or pin diameter could be caused by the spectral trend found in the previous subsection.
Using NUREG/CR-6698 [2] equations 10 through 13 and the data from Table A.2 , the predicted mean ke rr as a function of pin diameter is: k(Pin Diameter)
= 0.997805 + ( 7.43E-04)*Pin Diameter where the pin diameter is in cm. The predicted mean k e rr as a function of pitch is: k(Pitch) = 0.997098 + ( 9.646E-04)*Pitch where lattice pitch is in cm. The tolerance band widths are 3.4E-03 and 3.SE-03 for the pin diameter and pitch respectively. Figures A.3 and A.4 graphically present k e rr as a function of the pin diameter and the lattice pitch. 1.003 l 1.002
* 1.001 1.000 ---I :i: 0.999 o/ ; 0.998 41 .. 3 0.997 * * ... 0.996
* 0.995 i 0.994 0.99 3
* 0.992 0.8 0.9 1 1.1 1.2 1.3 1.4 1.5 Pin Diameter (cm) Figure A.3: kerr as a Function of the Pin Diameter NET-28091-0003-01 , Revision 0 A-25 1.003 1.002 1.001 1.000 := 0.999 QI =-:: 0.998 "C QI .... Ill 0.997 :i V iv 0.996 u 0.995 0.994 0.993 0.992 1 Enrichment 1.2 1.4 -+-----1.6 1.8 Pitch (cm) 2 2.2 Figure A.4: keff as a Function of the Lattice Pitch 2.4 2.6 The fuel to be stored in the racks ranges in enrichment from 1.6 wt% 235 U to 5 wt% m u. It was determined that there is no statistically significant trend on enrichment.
Although not statistically significant, the trend in the mean k eff is: k(Enrichment) 0.998824 -( 6.4E-05)*Enrichment where Enrichment is wt% m u. The tolerance band width is 3.4E-03. Figure A.5 graphically presents the results. NET-28091-0003-01 , Revision 0 A-26 1.003 ----------------1.002 1.001 1 0.999
* 41 :.::: "O 0.998 41 .. I'll 0.997 :i V 7ii 0.996 u 0.995 0.994 0.993 ------------* 0.992 2 2.5 3 3.5 4 4.5 5 5.5 6 6.5 7 Enrichment (wt% U-235) Figure A.5: keff as a Function of the Fuel Enrichment Boron Content A trend test was performed to determine whether the calculated k e rr values of the benchmark experiments co ntain a statistically significant trend as a function of the soluble boron ppm. No statistically significant t rend was found. However, it is conservatively assumed that the trend s are real. The follo w ing equation is the best fit of the data for k err v ersus soluble boron ppm. Figure A.6 shows the results of the analyses. The uncertainty around the mean v alues given in the followin g equation s i s 0.0029 at O ppm and 0.0035 at 2000 ppm. k(ppm soluble boron) 0.99856 + ( 1.56E-07)*ppm NET-28091-0003-01 , Revision 0 A-27 0.9995 -------0.9990 ** ----------*
* 0.9 98 5 :i: 0.9980 "O ... "' * :i 0.9975 ... ia u 0.9970 * ... 0.9965
* 0.9960 0 1 000 2000 3000 4 000 5000 6000 Soluble Boron (ppm) Figure A.6: kcff as a Function of the Soluble Boron Content A.2.6 Establishing the Bias and the Uncertainty To make the incorporation of the bias and bias uncertainty in the criticality analysis conservative , the most limiting bias and bias un certainty from the trends in the range of interest is u sed. At the lattic e pitch for Westinghouse 15xl5 fuel (1.43 cm) the bias and unc ertainty are 0.0015 and 0.0029 respectively.
At the Westinghouse 15xl5 fuel pin diameter (1.072 cm) the bias and unc ertainty are 0.0014 and 0.0029 respectively. Thus the bia s as a function of pitch is more limiting.
The bias as a function of enrichment is greatest at 5 wt% and is 0.0015. The uncertainty over the range of enrichments is 0.0034. The maximum bias and uncertainty as a function of soluble boron ppm occurs at 2000 ppm and is 0.0018 and 0.0035 respectively. The spectrum as measured by the EALF in the pool with no solub l e boron is genera ll y between 0.2 and 0.4 eV. The bias increases as the spectrum hardens and the bias at 0.4 eV is 0.0024. This is the most limitin g bias. For heavily borat ed cases the EALF can get almost as high as 0.65 eV. At 0.65 eV the bias is 0.0036. For the criticality analysis a bias of 0.0024 is used for all EALF less than 0.4 (limiting cases for no boron credit) and 0.0036 for EALF values between 0.4 and 0.65 eV (heavily borated cases). The maximum uncertaint y for any trends is 0.0035 which comes from the so lubl e boron analysis.
In order to mak e the analysis simple 0.0035 is selected for the uncertaint y in the bias. NET-2809 1-000 3-01 , Revision 0 A-28 The uncertainty of the set as a who l e is 0.0031. The uncertainty for the trended analysis is generally less since taking advantage of the trend reduces the difference between the experimental value and the predicted value. A.2. 7 Subcritical Margin In the USA , the NRC has established subcritical margins for rack analysis. The subcritical margin for borated spent fuel poo ls , casks , and fully flooded dry storage racks is O when the analysis is perfonned with unborated water. This is actually saying the subcritical margin is contained in the uncredited soluble boron. To make sure there is sufficient so lubl e boron, analysis is a l so performed with soluble boron and a subcritica l margin of 5% in k e rr is required.
For dry storage racks analyzed with optimum moderation, the subcritical margin is 2% and 5% with full moderation. In the ana l ysis of 32 8 critical experiments, which generously cover the range of expected conditions , the lowest calculated k e rr was 0.9931. This supports the position that the subcritical margin is more than sufficient.
A.2.BArea of Applicability (Benchmark Applicability)
The critica l benchmarks selected cover all commercia l light water reactor fuel storage racks or casks. To summarize the range of the benchmark applicability ( or area of applicability), Table A.5 is provided below. Table A.5: Area of Applicability (Benchmark Applicability)
Parameter Range Comments Fissionable Materia l/Physical U02 Form Enrichment (wt% U-235) 2.35 to 6.903 Some extrapolation of the bias to lower enrichments may be needed. Enrichments less than 2.35 are rarely limiting and generally only used in 1 s t cores. The bias is becoming smaller at low enrichments.
Using the maximum bias and uncertainty for all of the trends easily covers the small extrapolation needed. The maximum enrichment of 5% is within the range of experiments.
NET-28091-0003-01 , Revision 0 A-29 Parameter Range Comments Spectrum Expected range in applications:
-EALF (eV) 0.0605 to 0.1 to 0.6 0.8485 The experiments easily cover the entire expected range of limiting conditions. Lattice Characteristics Type Square Hex lattices have been excluded Pin Pitch ( cm) 1.075 to 2.54 Pin pitch of 1.43 cm is within the range. Assembly Spacing in Racks This covers all spacing. Neutron Distance between Assemblies 0 to 15.4 transport through larger than 15 .4 cm has a small effect on k. Note that the (cm) spacing is assumed to be filled with full density water. If the water density is less this separation effectively increases.
Therefore , optimum moderation cases of wide spaced racks are covered. Absorbers Ag-In-Cd control rods Contained in No significant difference in bias 51 critical between Ag-In-Cd critical experiments experiments and those that did not contain the control rods. Absorbers Soluble Boron 0 to 5030 ppm All designs are within this range. Concentration Absorbers Cd bearing experiments showed no Cd (component of Ag-In-Cd Absorber dependence on the number of rods. Credit for these rods is acceptable.
Cd rods) panels NET-28091-0003-01 , Revision 0 A-30 Parameter Range C o mments Reflector Experiments included water Reflectors Most rack analysis will assume an and steel adequate l y infinite system. Fu ll pool mode l covered reflectors are adeq u ate l y covered. Temperature Room This temperature range covers a ll Temperature norma l operating temperatures. Over to 358 K temperature accident conditions have significant margin due to ppm boron. Moderating mater i al water The moderator in a ll benchmark experiments are water , t herefore water as a moderating material is covered NET-28091-0003-01 , Revision 0 A-3 1 A.2.9 Summary of U02 Laboratory Critical Experiment Analysis This va lidati on follows the guidance ofNUREG/CR-6 698. Key aspects of the guidance are the selection of expe rim ents, ana l ysis of the experiments, statistica l treatment, determination of the bias and the bias uncertainty , and finally id entification of the area of applicability.
328 U0 2 critical experiments have been selected that cover the range of conditions for rack ana l ysis. The experime nt s h ave been a nal yzed using SCALE 6.1.2 a nd the EN DF/B-VII 238 gro up cross sect ion s and the resu ltin g bias in keff is very small. The results of the criticality ana l ysis were tested for trends against 5 different p arameters import ant to reactivity.
It was conservatively assumed that the any trend found was significant.
Using the trends, the most l imiting bi as and bias un certainty is determined to b e 0.0024 for the bias for EALF up to 0.4 eV and 0.0036 for EALF's in the range of 0.4 and 0.65 eV and the uncertainty is 0.0035 for all analysis.
The area of applicabi lit y is found in Ta bl e A.5. NET-2809 1-000 3-01, R evision 0 A-32 A.3. HTC and MOX Critical Experiments Burned fuel contains a low concentration of plutonium (about 1 wt%), as well as the uranium and thu s i s actually Mixed Oxide (MOX) fuel. Most classical MOX experiments ha ve plutonium concentrations at least twice as hi gh as that contained in burned fuel. A series of experiments were performed in France and purchased by the US for domestic use , which model the uranium and plutonium concentration, which matches 4.5 wt% U-235 fuel burned to 37.5 GWd/T [12]. This fuel has 1.1 wt% plutonium and 1.57 wt% U-235. Both the HTC critical experiments and a l arge series of c l assical MOX experiments were analyzed.
A.3. 1 HTC Critical Experiments All of the HTC critical experiments used the same fuel pins. The criticality of the se experiments was controlled by adjusting the critical water height. The fuel pin s were used in 156 critical arrangements.
117 of these were relevant to spent fuel pool analysis.
The experiments were performed in four pha ses. Phase 1 [ 13] consists of 1 7 cases where the pin pitch was varied from 1.3 cm to 2.3 cm and different quantities of pins were used to change the critical height. An 18 111 case was done where the array was mo ve d to the edge of the tank , so the boundary was the steel tank followed by void. This condition i s not typical of a spent fuel pool , so this case was not analyzed.
The average k eff of the Phase 1 cases was 0.99910. Phase 2 [ 14] consisted of 20 cases where gadolini u m of various concentrations was dissolved in the water (Phase 2a) and 21 cases where boron was dissolved in the water (Phase 2b). These experiments also varie d the pitch ( 1.3 to 1.9 cm) and the number of pins. The average k eff of the gadolinium cases was 0.998 15 and the average for the boron cases was 0.99897. Phase 3 [ 1 5] consists of 26 experiments where the p i ns were arra n ged as 4 "asse mb l ies." Eac h assembly used a 1.6 cm pin pitch. The assembly separation was varied, as well as the number of pins in each assembly.
Finally , eleven cases boxed the assemb l ies with an absorber (borated steel, boral, or cadmium).
The average k eff of these 26 cases was 0.99890. Finally, Phase 4 [ 16] consisted of redoing the same type of experiments as Phase 3, except with reflector screens. The 38 experiments which used the lead reflector screen were not included in this ana l ysis, since lead reflectors are not common in spent fuel pools. The 33 steel reflector experiments were included. The average k eff of these cases was 0.99858. References 13 through 16 provided a ll of the detai l s for the analysis.
The mode l ing was straight forward. The references gave a simp l e model and a detailed model. The mode l created for this work fo ll owed the detai l ed m odel , except t h at the top grid outside of the array and the basket supports were not mo d e l ed. Both of these assumptions were part of the simplified model and have a negligible impact on k. The model used actually exceeded the detailed model , since the spring above the fuel was modeled by homogenizing it with the void. NET-28091-0003-01, Revision O A-33 Tables A.6 through A.10 present the results of the analysis.
A statistical analysis of the HTC set as a whole was performed consistent with the method provided in NUREG/CR-6698 , where the experimental uncertainties were taken from References 13 through 16. The mean uncertainty weighted k e ff is 0.99878 and the uncertainty is 0.00590. This makes th e bias 0.00122. Since all of the pins are the same , trend analysis on the pin diameter and enrichment are not possible. The pin pitch changes are made to adju s t the spectrum, so the only trend analysis performed is on the spectrum (EALF). The trend analysis on the HTC set (performed consistent with NUREG/CR-6698) on EALF yielded the followin g function:
k(EALF) = 0.999541 -0.00548
* EALF The units for EALF are eV. The uncertainty about the trending k e ff is 0.0076 ink. Figure A.7 shows the results of the HTC analysis.
Table A.6: HTC Phase 1 Results Case No. kerr Mo n te EALF Pitch Car l o (eV) (cm) S i g m a 1 0.99913 0.00015 0.069486 2.3 2 0.99893 0.00016 0.066544 2.3 3 0.99892 0.00016 0.066412 2.3 4 0.99974 0.00017 0.084957 1.9 5 0.99983 0.00017 0.082795 1.9 6 0.99946 0.00020 0.082123 1.9 7 0.99977 0.00019 0.102248 1.7 8 0.99962 0.00018 0.100654 1.7 9 0.99903 0.00019 0.099687 1.7 10 0.99991 0.00019 0.140669 1.5 11 0.99898 0.00020 0.135753 1.5 12 0.99906 0.00019 0.133996 1.5 13 0.99813 0.00021 0.256212 1.3 14 0.99776 0.00019 0.234183 1.3 15 0.99812 0.00022 0.230564 1.3 16 0.99952 0.00020 0.101408 1.7 17 0.99882 0.00019 0.099384 1.7 NET-28091-0003-01 , Revision 0 A-34 Table A.7: HTC Phase 2a, Gadolinium Solutions, Results Case No. k.rr Monte EALF Pitch Gadolinium Carlo (eV) (cm) Concentration Sigma (Q:/1) 1 0.99784 0.00020 0.25279 1.3 0.0520 2 0.99792 0.00021 0.24946 1.3 0.0520 3 0.99777 0.00019 0.27074 1.3 0.1005 4 0.99771 0.0001 8 0.26 756 1.3 0.1005 5 0.99784 0.0001 8 0.2 6333 1.3 0.1005 6 0.99683 0.0001 8 0.28513 1.3 0.1505 7 0.99684 0.00019 0.27847 1.3 0.1505 8 0.99623 0.00016 0.29552 1.3 0.1997 9 0.99608 0.00018 0.29253 1.3 0.1997 10 0.99689 0.00017 0.169 82 1.5 0.1997 11 0.99766 0.00019 0.16252 1.5 0.1495 12 0.99771 0.00018 0.16101 1.5 0.1495 13 0.99868 0.00017 0.15392 1.5 0.1000 14 0.99861 0.00018 0.15223 1.5 0.1000 15 0.99983 0.00020 0.14727 1.5 0.0492 16 0.99976 0.00019 0.14432 1.5 0.0492 17 1.00053 0.00018 0.106 3 1 1.7 0.0492 18 1.00070 0.00017 0.087 83 1.9 0.0492 19 0.99707 0.00016 0.11369 1.7 0.1010 20 1.00050 0.00019 0.10648 1.7 0.0492 NET-28091-0003-01, Revi s ion 0 A-35 Table A.8: HTC Phase 2b, Boron Solutions, Results Case No. kerr Monte EALF Pitch Boron Carlo (eV) (cm) Concentration Si2ma {l?:/1) 1 0.99835 0.00020 0.24780 1.3 0.100 2 0.99760 0.00020 0.24450 1.3 0.106 3 0.99816 0.00020 0.25528 1.3 0.205 4 0.99904 0.00020 0.26400 1.3 0.299 5 0.99886 0.00019 0.27475 1.3 0.400 6 0.99852 0.00019 0.27125 1.3 0.399 7 0.99933 0.00018 0.27977 1.3 0.486 8 0.99894 0.00019 0.28781 1.3 0.587 9 0.99952 0.00016 0.16627 1.5 0.595 10 0.99811 0.00019 0.16087 1.5 0.499 11 0.99990 0.00017 0.15663 1.5 0.393 12 0.99987 0.00018 0.15007 1.5 0.295 13 0.99887 0.00018 0.14559 1.5 0.200 14 1.00192 0.00018 0.14024 1.5 0.089 15 1.00338 0.00018 0.10325 1.7 0.090 16 1.00202 0.00017 0.10717 1.7 0.194 17 1.00313 0.00017 0.11049 1.7 0.286 18 0.99367 0.00017 0.11577 1.7 0.415 19 1.00021 0.00021 0.10473 1.7 0.100 20 0.99251 0.00017 0.08965 1.9 0.220 2 1 0.99642 0.00017 0.08611 1.9 0.110 NET-28091-0003-01, Revision 0 A-36 Table A.9: HTC Phase 3 Results -Water Reflected Assemblies
* (1.6 cm pin pitch) Case No. k e rr Monte EALF Absorber Assembly Carlo (eV) Box Separation Si2ma Material (cm) 1 0.99 774 0.00022 0.12377 Borated SS 3.5 2 0.99986 0.00019 0.14095 Borated SS 0 3 0.99710 0.00019 0.12939 Borated SS 2 4 0.99715 0.0001 8 0.12391 Borated SS 3 5 0.9969 9 0.00018 0.13503 Borated SS 1 6 0.99987 0.00019 0.1 2974 Bora) 0 7 0.99614 0.00019 0.12866 Cd 2 8 1.00381 0.0001 8 0.13904 Cd 0 9 0.99 646 0.00017 0.1 3345 Cd 1 10 0.99672 0.0001 8 0.12952 Cd 1.5 11 0.99571 0.00019 0.13726 Cd 0.5 12 0.999 01 0.00017 0.11 277 none 1 8 13 0.999 15 0.0001 8 0.11167 none 14.5 14 0.99934 0.0001 8 0.111 83 none 11 15 0.99910 0.00019 0.11093 none 10 16 0.999 61 0.00019 0.110 3 0 none 9 17 0.99930 0.0001 8 0.10842 non e 8 18 0.9 9980 0.00017 0.1065 6 none 6 19 1.00016 0.0001 8 0.10421 none 4 20 1.00044 0.0001 8 0.10206 none 4 21 0.99976 0.0001 8 0.10470 none 2 22 1.00047 0.00019 0.10714 none 1 23 0.998 93 0.0001 8 0.11506 non e 0 24 0.99949 0.00020 0.15073 none 0 25 0.999 96 0.0001 8 0.12672 none 4 26 0.99937 0.00020 0.11550 non e 10 NET-28091-0003-01, Revisi o n 0 A-37 Case No. 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 1 8 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 Table A.10: HTC Phase 4 Results -Steel Reflected Assemblies (1.6 cm pin pitch) k c rr Monte EALF Absorber Assembly Separation Carlo (eV) Box Separation From Reflector Sigma Material (cm) (cm) 1.00157 0.00019 0.15363 Borat ed SS 0 0.0 0.99845 0.00018 0.15069 Borat ed SS 0.5 0.0 0.99797 0.00018 0.14674 Bo rated SS 1 0.0 0.99826 0.00018 0.14 227 Borat ed SS 1.5 0.0 0.99839 0.00019 0.13923 Borated SS 2 0.0 0.99712 0.00018 0.13820 Borat ed SS 2 0.5 0.99634 0.00018 0.13705 Borated SS 2 1.0 0.99650 0.00018 0.13598 Borat ed SS 2 1.5 0.99658 0.00018 0.13518 Borated SS 2 2.0 0.99834 0.00018 0.13430 Borated SS 3 0.0 0.99821 0.00018 0.13234 Borated SS 3.5 0.0 1.00095 0.00018 0.13558 Bora! 0 0.0 0.99653 0.00018 0.13386 Bora! 0.5 0.0 1.00431 0.00017 0.14979 Cd 0 0.0 0.99818 0.00020 0.14323 Cd 1 0.0 0.99769 0.00017 0.13683 Cd 2 0.0 0.99615 0.00018 0.13568 Cd 2 0.5 0.99536 0.00019 0.13423 Cd 2 1.0 0.99513 0.00018 0.13315 Cd 2 1.5 0.99465 0.00018 0.13235 Cd 2 2.0 0.99869 0.00018 0.13390 Cd 2.5 0.0 1.00060 0.00018 0.17427 none 0 0.0 1.00057 0.00018 0.16641 non e 1 0.0 0.99973 0.00018 0.15852 non e 2 0.0 0.99935 0.00018 0.15709 none 2 0.5 0.99946 0.00018 0.15559 none 2 1.0 0.99939 0.00018 0.15431 none 2 1.5 0.99937 0.00019 0.15351 none 2 2.0 0.99941 0.00019 0.14426 non e 4 0.0 0.99964 0.00018 0.13456 none 6 0.0 0.99953 0.00018 0.12 886 none 8 0.0 0.99947 0.00017 0.12537 none 10 0.0 0.99940 0.00018 0.12333 non e 12 0.0 NET-28091-0003-01, Revi s ion 0 A-3 8 
.:,: ',:J 41 1.006 1.004 1.002 1 * ------.-** * ---.
* 0.998 +---------
* ... ;; u
* 0.996 +-----------
0.994 +-------* *:I I * * * .. # 0.992 +-----------------------------------
0.99 +-----,-------,------r-----,-----.,------,-------, 0.000000 0.050000 0.1 00000 0.1 50000 0.200000 0.250000 0.300000 0.350000 Energy of the Average Lethargy of Fission (EALF) (ev) Figure A.7: k eff as a Function of the EALF for the HTC Experiments A.3.2 MOX Critical Experiments The se l ect ion of the MOX critical exper im ents was limi ted to the l ow enriched MOX lattice cri ti cal experime nt s. All 63 of the low enric h ed MOX pin critical experiments documented in the OECD handbook [1 7] were utili zed. The ac tu a l inpu t decks were initiated from avai l ab l e decks found in NUREG/CR-6102 [18] and the International Handbook.
[1 7] The decks were modified to update to the new cross-section library and changes in the SCALE inpu t format. Table A.11 presents the results of t h e 63 se l ected MOX critical experiments.
The Reference column has the eva lu a ti on number from the International Hand book. [17] For example, OECD-7 refers to the OECD International Handbook case MIX-COMP-THERM-07. Trends were in vestigate d as a function of EALF, plutonium conte nt , and the Am-241/U-238 ratio. As the spectrum h a rd ens (hig h er EALF), t h ere is a small trend to higher k. With more plutonium conte nt , k e ff increases. This is seen in Figure A.8. NET-28 091-0003-01 , R evision 0 A-39 The change in ke rr with cooling time is dominated by the reactivity of the decay of Pu-241 to Am-241. By plotting kerr versus the Am-241/U-238 ratio , it i s possible to determine if the bia s should be chang e d for cooling. Figure A.9 shows that with increasing Am-241 c ontent , the calculated k err of th e critical experiments increases. This obser v ation show s that the z ero cooling time bias conser v ati v ely covers the cooling time. Table A.11: Results of MOX Critical Benchmarks (SCALE 6.1.2, ENDF/B-VII)
Case ID Reference k e ff sigma EALF Pu Pu Am241/U238 (eV) wt% 240% 093array OECD-7 1.0009 0.00025 0.1903 2.00 16 6.82E-05 105al.in OECD-7 0.9942 0.00027 0.1 3 69 2.00 16 7.55E-05 105array OECD-7 0.9960 0.00025 0.1377 2.00 16 7.55E-05 105bl OECD-7 0.9914 0.00026 0.1379 2.00 16 7.55E-05 105b2 OECD-7 0.9921 0.00024 0.1377 2.00 16 7.55E-05 105b3 OECD-7 0.9933 0.00025 0.1373 2.00 16 7.55E-05 105b4 OECD-7 0.9940 0.00026 0.1371 2.00 16 7.55E-05 l 143arra OECD-7 0.9980 0.00026 0.1166 2.00 16 8. l 3E-05 132array OECD-7 0.9971 0.00022 0.0953 2.00 16 8.13E-05 1386arra OECD-7 0.9942 0.00023 0.0906 2.00 16 6.97E-05 epri70b OECD-2 0.9992 0.00025 0.7209 2.00 7.8 7.29E-05 epri70un OECD-2 0.9974 0.00027 0.5409 2.00 7.8 7.29E-05 epri87b OECD-2 1.0019 0.00022 0.2710 2.00 7.8 7.29E-05 epri87un OECD-2 0.9981 0.00032 0.1852 2.00 7.8 7.29E-05 epri99b OECD-2 1.0012 0.00024 0.1772 2.00 7.8 7.29E-05 epri99un OECD-2 1.0007 0.00027 0.1333 2.00 7.8 7.29E-05 klmct009 OECD-9 0.9994 0.00024 0.5169 1.50 8 l.06E-05 k2mct009f OECD-9 0.9941 0.00027 0.2 9 43 1.50 8 9.77E-06 k3mct009 OECD-9 0.9934 0.00024 0.152 8 1.50 8 8.96E-06 K4mct009 OECD-9 0.9921 0.00024 0.1155 1.50 8 8.96E-06 K5mct009 OECD-9 0.9925 0.00021 0.0947 1.50 8 8.96E-06 K6mct009 OECD-9 0.9937 0.00024 0.0905 1.50 8 9.77E-06 omct61 OECD-6 0.9954 0.00026 0.3570 2.00 8 2.24E-05 omct62 OECD-6 0.9990 0.00029 0.1 8 85 2.00 8 2.24E-05 omct63 OECD-6 0.9943 0.00027 0.1374 2.00 8 2 , 24E-05 omct64 OECD-6 0.9982 0.00025 0.1167 2.00 8 2.24E-05 omct65 OECD-6 0.9994 0.00025 0.0956 2.00 8 2.24E-05 omct66 OECD-6 0.9956 0.00024 0.0907 2.00 8 2.24E-05 mct8cl OECD-8 0.9978 0.00029 0.3776 2.00 24 7.93E-05 mct8c2 OECD-8 0.9977 0.00028 0.1922 2.00 24 7.27E-05 mct8c3 OECD-8 0.9967 0.00024 0.13 8 3 2.00 24 8.59E-05 mct8c4 OECD-8 1.0006 0.00027 0.1170 2.00 24 9.88E-05 mct8c5 OECD-8 1.0000 0.00026 0.0955 2.00 24 9.56E-05 mct8c6 OECD-8 0.9992 0.00023 0.0905 2.00 24 7.27E-05 mct8cal OECD-8 0.9967 0.00025 0.1375 2.00 24 8.59E-05 mct8cbl OECD-8 0.9931 0.00024 0.1387 2.00 24 8.59E-05 mct8cb3 OECD-8 0.9941 0.00025 0.1381 2.00 24 8.59E-05 NET-28091-0003-01, Revision 0 A-40 Case ID Reference kerr sigma EALF Pu Pu Am241/U238 (eV) wt 0/o 240% meteb2 OECD-8 0.9937 0.00024 0.1385 2.00 24 8.59E-05 meteb4 OECD-8 0.9942 0.00026 0.1378 2.00 24 8.59E-05 mixo25lk OECD-5 1.0011 0.00032 0.3732 4.00 18 l.59E-04 mixo252k OECD-5 0.9985 0.00027 0.2476 4.00 18 l.59E-04 mixo253k OECD-5 1.0044 0.00027 0.1712 4.00 18 l.59E-04 mixo254k OECD-5 1.0004 0.00029 0.1425 4.00 18 l.59E-04 mixo255k OECD-5 1.0034 0.00028 0.1058 4.00 18 l.59E-04 mixo256k OECD-5 1.0023 0.00024 0.0917 4.00 18 l .59E-04 mixo257k OECD-5 1.0036 0.00024 0.0 8 75 4.00 18 l .59E-04 saxtnl04 OECD-3 1.00044 0.00027 0.0987 6.60 8.6 8.43E-05 s axtn56b OECD-3 0.99962 0.00028 0.6133 6.60 8.6 8.43E-05 saxtn735 OECD-3 0.99999 0.00031 0.1 8 20 6.60 8.6 8.43E-05 saxtn792 OECD-3 0.99951 0.00031 0.1505 6.60 8.6 8.43E-05 Saxton52 OECD-3 0.99977 0.00028 0.8 517 6.60 8.6 8.43E-05 Saxton56 OECD-3 1.00018 0.0003 0.5177 6.60 8.6 8.43E-05 teal OECD-4 0.99572 0.00027 0.1418 3.01 22 l.04E-04 tealO OECD-4 0.9988 0.00024 0.0792 3.01 22 9.3 lE-05 teal 1 OECD-4 0.99886 0.00023 0.0788 3.01 22 2.06E-04 tea2 OECD-4 0.9964 0.0003 0.1409 3.01 22 l.99E-04 tea3 OECD-4 0.99665 0.00028 0.1403 3.01 22 2.96E-04 tea4 OECD-4 0.99644 0.00026 0.1172 3.01 22 9.88E-05 tea5 OECD-4 0.9974 0.00027 0.1167 3.01 22 2.02E-04 tea6 OECD-4 0.99848 0.00025 0.1156 3.01 22 3.90E-04 tea7 OECD-4 0.99753 0.00025 0.0917 3.01 22 8.88E-05 tea8 OECD-4 0.99801 0.00025 0.0913 3.01 22 2.03E-04 tea9 OECD-4 0.99864 0.00025 0.0909 3.01 22 3.02E-04 NET-28091-0003-01 , Revision 0 A-41 1.0060 -----------*
* 1.0040 -----* 1.0020 * *
* 1.0000 * --------;: I
* 0.9980 .x
* 0.9960 ' --.--------0.9940 ---.11-* t
* 0.9920 -----------* 0.9900 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 P u wt% Figure A.8: Predicted k err as a Function of the Plutonium Content NET-28091-0003-01 , R evision 0 A-42 
:i: QI 1.0060 1.0040 *
* 1.0020
* t * *
* 1.0000 i * * ,.
* 0.99 8 0 * ---* ** *
* 0.9960
* 0.99 4 0
* 0.9900 +---~--~---.----~--~---,-----r---~----,,---
0.0 E +OO 5.0E-05 1.0E-04 1.SE-04 2.0E-04 2.SE-04 3.0E-04 3.SE-04 4.0E-04 4.SE-04 5.0E-04 Ratio of Am-241 to U-238 Figure A.9: Predicted kerr as a Function of the Am-241 Content A.3.3 Bias and Uncertainty from the MOX/HTC Critical Experiments The bias and uncertainty of burned fuel depends on the amount of plutonium in the burned fuel. As shown on Figure A.9 the bias decreases with plutonium content. However , the uncertainty increases with plutonium content. In order to determine appropriate biase s and uncertaintie s the HTC and MOX critical benchmarks are combined. The MOX experiments with plutonium content above 2 wt% were useful for confirmation that the bias decreases with plutonium content , but the maximum plutonium content in spent nuclear fuel about 1.5 wt% plutonium , so using experiments above 2 wt% plutonium needl e ss l y increa s e s the uncertainty. The bias and uncertainty in the bias for the HTC/MOX (2 wt% PU or less) set is controlled by the EALF trend. For EALF's less than 0.4 eV the maximum bias and uncertainty are 0.0021 and 0.0087 respectively.
For EALF's from 0.4 to 0.65 eV the maximum bias and uncertaint y are 0.0027 and 0.0112 respectivel
: y. NET-28091-0003-01 , Revision 0 A-4 3 A.4. Temperature Dependent Critical Experiments Since the criticality analysis for spent fuel pools must consider the full range of temperatures a ll owed for the poo l the LCT-46 set of critical experiments are needed to assure the correct bias and uncertainty is used for conditions where the pool is at its highest temperatures. The suite of critical experiments other than LCT-46 contains a range of fuel to moderator ratios that shou l d adequately cover the impact of the density change in the water as the pool temperature rises but no other experiments test the Doppler broadening of the cross sections or the change in the thennal scattering.
LCT-46 consists of 22 experiments but the last 5 experiments contain copper rods. Since copper is not norm a lly in spent fuel pools only the fir s t 17 experiments are analyzed here. Section 3 of LCT-046 specifies the critical benchmark and the SCALE models used follow that specification.
The specification has a couple of minor ambiguities related to the thermal expansion given as Table 29 of LCT-046. For this ana l ysis all of the expansion factors from Table 46 were app l ied to all of the x-y dimensions.
That mean s that the same SS component expansion factor was applied to pitch and the inner and outer diameter of the clad. This is consistent with the MCNP samples given in the Appendix of LCT-046. For the axial expansion only the fuel was expanded.
As with the MCNP sample input the same expansion factor was used for the radius and the axial dire ct ion. Table A.12 shows the corrected SCALE 6.1.2 ENDF/B-VII results for the 17 critical experiments.
Corrected r es ults in this case means they were divided by the kerr of the benchmark which was not quite 1.0. Table A.12: LCT-46 w i t h Full Thermal Expansi o n Case Temperature (K) Corrected SCALE k SCALE s i i!ma 1 297.05 0.999082 0.000065 2 310.41 0.998902 0.000071 3 315.43 0.998817 0.000067 4 319.96 0.998908 0.000073 5 324.93 0.998629 0.000067 6 332.53 0.998746 0.000067 7 287.22 0.999148 0.000067 8 3 15.91 0.998819 0.000066 9 330.27 0.998696 0.000068 10 337.44 0.998804 0.000065 11 351.99 0.998829 0.000068 12 303.6 0.998649 0.000065 13 312.95 0.998641 0.000069 14 321.1 6 0.998556 0.000067 15 328.24 0.998401 0.000068 16 338.26 0.998318 0.000067 17 358.3 1 0.99825 6 0.000065 NET-28091-0003-01, Revision 0 A-44 Figure A.10 plots the results of the analysis as a function of case. As can be seen from this plot there does seem to be a trend wit h t emperature. Figure A.11 i s the dat a plotted against temperature wit h the l ea s t squares lin ear fit. T h e fit is statistica ll y significant.
The slope is -8.6E-6 deltak/&deg;C. The uncertainty arou nd th e fit is 0.0013. The bias is d etermined b y multiplying th e change from room tempera tur e in &deg;C b y 8.6E-6. The uncertainty of 0.00 1 3 is an independent un certai nt y that can be statistically co mbin ed with the other uncertainties.
k versus Case (three sets with increasing temperature in each set) 0.999200 0.999100 X 0.999000 0.998900 0.998800 -i, QI ... .!! 0.998700 :I u a o.998600 0.998500 0.998400 0.998300 Rectangu l a r Set X X X Ro u nded Set X X X -xx 4 Gd Rods Set X 0.998200 T ---,-----, T ,--T -,-----, 0 2 4 6 8 10 12 14 16 18 Case Number Figure A.10: LCT-046 Corrected Calculated k eff per Case NET-28091-0003-01 , R evision 0 A-45 0.999200 t 0.999 10 0 0.999000 --:,r. 0.998900 ... ., .1! 0.998800 :, u a o.998700 ...
* 0.998600 -j----&sect; 0.998500 0.99 84 0 0 0.998300 0.998200 + 290 300 *
* 310 * * * * * *
* 320 330 340 350 360 Temperature (K} Figure A.11: LCT-046 Corrected Calculated keffVersus Temperature 370 It is common practice not to thermally expand the solids when doing analysis of elevated temperatures in criticality analysis. Table A.13 shows the re s ults of the analysis repeated where the temperatures of the materials were increased and the density of the water decreased but no thermal expansion of the solids (fuel , pitch, etc.). As can be s e en in Table A.13 the difference between just expanding the water (lowering the density) and full thennal expansion is similar to the Monte Carlo uncertainty. The maximum difference is 0.00027 which is less than 4 times the Monte Carlo one sigma uncertainty of one of the two calculations used in the difference.
NET-28091-0003-01, Revision 0 A-46 Table A.13: LCT-46 with No Thermal Expansion of Solids Difference in k e rt From Case Temperature (K) Corrected SCALE k SCALE si!!.ma Full Thermal Expansion 1 297.05 0.999114 0.000069 -0.00003 2 3 10.4 1 0.998899 0.000068 0.00000 3 3 15.43 0.998736 0.000067 0.00008 4 3 1 9.96 0.998666 0.000074 0.0 00 24 5 324.93 0.998491 0.000069 0.00014 6 332.53 0.998 743 0.000069 0.00000 7 287.22 0.999302 0.000067 -0.00015 8 3 15.9 1 0.9988 74 0.00 00 70 -0.00 006 9 330.27 0.99859 7 0.000068 0.00010 10 337.44 0.998564 0.000067 0.00024 11 351.99 0.998558 0.000070 0.00027 1 2 303.6 0.998633 0.000070 0.00002 1 3 312.95 0.998651 0.0000 7 0 -0.00001 1 4 32 1.1 6 0.998469 0.000067 0.00009 15 328.24 0.998 420 0.000067 -0.00002 1 6 338.26 0.998500 0.000067 -0.00018 1 7 358.3 1 0.998042 0.000067 0.00021 Since the hi g h er temperatures have a h ar d er spect rum , th e effect of the hi g her temperatures could h ave already b ee n capture d in the trend on spec trum (EALF). This was t es t ed by using the slope of th e chan ge in k e ff wit h EALF from th e full set of cri ti ca l experiments. Using the EALF biased k s th e lin ear fit was reanalyzed.
The maximum bi as (0 to 100 C) cha n ged from 0.00086 to 0.00071. The spectrum i s a small amount of th e t empera tur e effect a nd i s therefore ignored for the final conclusion.
The ana l ys i s of the on l y set of thermal cr iti ca l expe rim ents in the Int e rn ationa l Handbook that u ses e l evate d temperatures b e lo w b oiling h as s h own a s m a ll in crease in th e bi as wit h t e mp erature. The bi as is d e t ermi n e d b y multiplying the change from room temperature in &deg;C b y 8.6E-6. The un certa inty of 0.0013 i s an independent uncertainty that can b e statistica ll y combined with the other unc e rt ainties. NET-2809 1-000 3-01, R ev i sion 0 A-47 A.5. Summary of Validation Using Laboratory Critical Experiments Nuclear fuel starts as U0 2 and as it bums it becomes a mixture of U0 2 and Pu 0 2. SCALE 6.1.2 with the ENDF/B-VII.O cross sections calculates a slightly higher k eff as the Pu0 2 content increases.
The correct bias and uncertainty shou l d be a function of the plutonium weight percent but this would be overly complicated for a small effect. The bias for the initial condition from U0 2 critical experiments would be conservative for spent nuclear fuel. However , the uncertaint y from the U0 2 only set is smaller than the uncertainty from the MOX set. To conservatively cover the all of the conditions of the fuel the final 95/95 ke ff is calculated twice , once using the U0 2 critica l experiments bias and uncertainty and once using the MOX/HTC bias and certainty.
The higher final 95/95 k e ff is used for comparison to the k e ff criteria. The two bias and uncertainty sets are: 1. Based on the U0 2 experiments
: For EALF's less than 0.4 eV the bias is 0.0024. For EALF's between 0.4 ev and 0.65 eV the bias is 0.0036. The uncertainty for the entire range of EALF is 0.0035. 2. Based on the MOX/HTC experiments
: For EALF's less than 0.4 eY the bias is 0.0021. For EALF's between 0.4 ev and 0.65 eV the bias is 0.0027. The uncertainty for the range of EALF 0 to 0.4 eV is 0.0087. The uncertainty for the range ofEALF 0.4 to 0.65 eV is 0.0112 For all burned fuel the MOX/HTC bias and uncertainty actually determine the 95/95 k. For most cases in the pool analysis, the most dense water conditions are most limiting.
However , if higher temperature cases are more lirniting , then a temperature bias of 8.6E-6 multiplied by the change from room temperature in &deg;C is applied. In addition the uncertainty in this bias, 0.0013 , needs to be included in the uncertainty rack up. NET-28091-0003-01, Revision 0 A-48 A. 6. Appendix References
[ 1] Scal e: A Compreh e nsiv e Mod e ling and Simulation Suite for Nucl e ar Saf e ty Analysis and D e sign , ORNL/TM-2005
/39, Version 6.1 , June 2011. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-785. [2] J.C. Dean and R.W. Tayloe, Jr., Guid e for Validation of Nucl e ar Criti c ality Saf e ty Cal c ulational M e thodology , NUREG/CR-6698 , Nuclear Regulatory Commission, Washington , DC January 2001. [3] Int e rnational Handbook of Evaluat e d Criti c ali ty Safety B e n c hmark Exp e rim e nts, NEA/NSC/DOC(95)3, Volume IV, Nuclear Energy Agency, OECD , Paris, September , 2016. [ 4] DAT APLOT is statistical software supported by the National Institute of Standards and Technology.
It can be down loaded at: http://www.itl.nist.gov
/div898/software/dataplot/ [5] J. J. Lichtenwalter , S. M. Bowman, M. D. DeHart , and C. M. Hopper, Criti c ality B e n c hmark Guid e for Light-Wat e r-R e actor Fu e l in Tran s portation and Storag e Packa ges , NUREG/CR-6361 (ORNL/TM-13211), Spent Fuel Project Office, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission , Washington , DC 20555-0001, March 1997. [6] L. W. Newman , et al, Urania-Gadolinia:
Nucl e ar Model D e v e lopm e nt and Critical Exp e rim e nt B e n c hmark, BA W-1810 , Babcock & Wilcox , Utility Power Generation Division , Lynchburg , VA , April 1984. [7] Brian L. Koponen and Viktor E. Hampel, Nucl e ar Criticality Saf e ty Exp e rim e nts , Cal c ulations, and Anal y s e s-1958 to 1982 , UCRL-53369 , Lawrence Livermore Laboratory , University of California , Livermore , California , October 21 , 1982. [8] M. Rahimi , E. Fuentes, and D. Lancaster, I s otopi c and Criti c ality Validation/or PWR Actinid e-OnlyBurnup Credit, DOE/RW-0497, U.S. Department of Energy , Office of Civilian Rad i oactive Waste Management, Washington, DC, May,1997. [9] [NOT USED] [10] [NOT USED] [11] [NOT USED] [12] D. E. Muel l er, K. R. Elam, and P. B. Fox , Evaluation of th e Fr e n c h Haut Tau x d e Combustion (HTC) Critical Exp e rim e nt Data , NUREG/CR-6979 (ORNL/TM-2007
/083), prepared for the US Nuc l ear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., September 2008. [13] F. Femex , "Programme HTC -Phase 1 : Reseaux de crayons dans l'eau pure moderated and reflected simple arrays) Reevaluation des experiences
," DSU/SEC/T/2005-33/D.R., Institut de Radioprotection et de Surete Nucleaire, 2008. NET-28091-0003-01 , Rev i sion 0 A-49 
[14] F. Fernex , Pro g ramm e HTC-Pha se 2: R ese au x s impl e s e n e au e mpoi s onn ee (bor e e t gad o linium) (R e fl ecte d simpl e arra y s mod e rat e d b y poison e d wa t e r with g adolinium or boron) R ee valuation d e s e xp e ri e n ces, DSU/SEC/T/2005-38/D.R., lnstitut de Radioprotection et de Surete Nucleaire , 2008. [ 15] F. Fernex , Pro g ramm e HTC -Pha se 3 : Confi g urations "s to c ka ge e n pi sc in e" (Pool stora ge) R eev aluation d es ex p e ri e nc e s , DSU/SEC!T/2005-37/D.R., lnstitut de Radioprotection et de Surete Nucleaire , 2008. [16] F. Fernex, Pro g ramm e HTC-Pha se 4: Confi g urations "c h a t e aux d e tran s port" (Shippin g c a s k) -R eev aluation d e s e xp e ri e n ces , DSU/SEC!T/2005-36/D.R., Institut de Radioprotection et de Surete Nucleaire , 2008. [17] Int e rnational Handbook of E v aluat e d Criti c ali ty Saf e ty B e n c hmark E x p er im e nt s, NEA/NSC/D0C(95)3, Volume VI , Nuclear Energy Agency , OECD , Paris , September , 2010. [18] M. D. DeHart and S. M. Bowman , Validation o f th e SCALE B roa d Stru c tur e 44-Group E N DF I B-V Cro ss-S ect ion Libra ry for U se in Criti ca li ty Saf ety A nal yse s , NUREG/CR-6102 (ORNL!T M-12460), Oak Ridge National Laboratory , Oak Ridg e, TN , September 1994. NET-28091-0003-01 , Revision 0 A-50 Appendix B: Fuel Categorization for Unit 2 Batches A Through X and Unit 3 A through AA All of the early discharged fuel has been categorized. Nearly all of the fuel is either Category 4 or Category 5. The table has been color coded to quickly identify the Category.
Category 3 is yellow , Category 4 is green, and Category 5 is b l ue. A range of assembly IDs that have the same Category are grouped together to reduce the length of the table. All but two assemblies for historical fuel at Unit 3 ha v e been categorized as Category 4 even though about ha l f of them could have been Category 5. Since the spent fuel pool at Unit 2 is used only temporarily for Unit 3 fue l while it is being casked, the lower reactivity Category is not needed. NET-2809 1-0003-01, Rev i sion 0 B-1 Table B.1: F u e l A s s embl y R e acti v i ty C ate g ori za tion for Asse mbl y ID s A t hrou g h X for Unit 2 Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category A01-A65 4 E43-ESS 4 K01-Kl3 4 E56 3 K14-Kl5 5 801-807 4 E57-E60 4 K16-K57 4 808-813 5 K58 5 814-823 4 FOl 3 K59-K68 4 B24-B26 5 F02-F20 4 827-B64 4 F21 3 L01-L07 4 F22-F30 4 L08-Ll0 5 C01-C04 4 F31-F34 5 Lll-L63 4 COS-C06 5 F35 4 L64 3 C07-C12 4 F36 3 L65-L68 4 C13 5 F37-F39 4 (14 4 F40 3 M01-M04 4 C15-C18 5 F41-F49 4 MOS 5 C19-C28 4 FSO 3 M06-M08 4 C29 5 F51-F60 4 M09 5 C30-C64 4 F61 3 M10-M12 4 F62-F64 4 M13-M14 5 D01-D25 4 F65 3 M15-M20 4 D26 5 F66 4 M21 5 D27-D60 4 F67-F68 5 M22-M23 4 D61-D68 5 M24 5 D69-D72 4 GOl-GOS 4 M25-M27 4 G06 5 M28 5 E01-E14 4 G07-G37 4 M29-M30 4 ElS 3 G38 5 M31 5 E16-E19 5 G39-G72 4 M32-M34 4 E20 4 M35 5 E21-E24 5 H01-H38 4 M36-M37 4 E25-E27 4 H39-H51 5 M38-M44 5 E28-E31 5 H52-H54 4 M45 3 E32-E33 4 HSS 5 M46 4 E34-E35 5 H56 4 M47-M48 5 E36-E40 4 M49-MSO 4 E41-E42 5 J01-J68 4 M51-M52 5 NET-2809 1-0003-01 , Revision 0 B-2 Table B.1: F u e l Asse mbl y Reactivit y Categorization for Assembl y IDs A throu g h X for U nit 2 (Continued)
Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category M53-M54 4 Q71-Q73 4 T42-T43 4 MSS-MSG 5 Q74-Q76 5 T44-T46 5 M57 4 Q77 4 T47 4 M58-M59 5 Q78 5 T48 5 MGO 4 Q79-Q80 4 T49-T51 4 M61 3 T52-T53 5 M62-M63 4 R01-R07 5 T54 4 M64 3 ROB 4 TSS 5 MGS 4 R09-R38 5 T56-T72 4 M66 5 R39 4 T73-T80 5 M67 3 R40-R43 5 M68 5 R44-RSO 4 M69-M71 4 R51-R69 5 U01-U04 5 M72 5 R70 4 uos 4 R71-R72 5 U06-U13 5 N01-N08 4 R73-R74 4 U14 4 N09-N12 5 R75-R79 5 U15-U16 5 N13-N14 4 R80-R81 4 U17-U21 4 N15-N16 5 R82 5 U22 5 N17-N23 4 R83-R85 4 U23 4 N24-N32 5 U24-U49 5 N33-N47 4 S01-S44 5 USO 4 N48 5 S45 4 USl 5 N49-N80 4 S46-S47 5 U52 4 S48 4 U53-U61 5 P01-P02 4 S49-S61 5 U62-U64 4 P03 3 S62 4 UGS 5 P04-P47 4 S63-S65 5 U66-U68 4 P48 5 S66 4 U69-U73 5 P49-P60 4 S67-S77 5 P61-P72 5 V01-V16 5 V17-V29 4 QOl-QGS 5 T01-T32 5 V30-V35 5 QGG 4 T33-T34 4 V36 4 Q67-Q68 5 T35-T36 5 V37-V38 5 Q69 4 T37 3 V39 4 Q70 5 T38-T41 5 V40-V41 5 N ET-28 091-000 3-0 1 , Re v i s ion 0 B-3 Ta bl e B.1: Fue l Assembly React i vity Categor i zation for Assembly IDs A through X for Unit 2 (Co n t inu ed) Assembly ID Category V42-V43 4 V44-V49 5 vso 4 V51-V54 5 VSS-V57 4 V58-V61 5 V62 4 V63 5 V64-V65 4 V66-V67 5 V68 4 V69-V77 5 V78-V79 4 V80-V81 5 V82 4 V83 5 V84 4 V85 5 V86 4 V87-V88 5 V89 4 V90-V91 5 V92 4 WOl-WlO 4 Wll 5 W 1 2-W15 4 W16 5 W17 4 W18-W19 5 W20 4
* FRSB is the fuel rod storage ba s ket NET-28091-0003-01 , Revision 0 Indian Point Uni t 2 F uel Assembly ID Category Assembly ID Category W21 5 X01-X02 3 W22 4 X03-X04 5 W23 5 XOS-X37 4 W24 4 X38 5 W25 5 X39-X49 4 W26 4 XSO-XSl 5 W27 5 X52-X53 4 W28-W34 4 X54-XSS 5 W35 5 X56-X58 4 W36-W38 4 X59-X60 5 W39 5 X61-X62 4 W40 4 X63 5 W41-W43 5 X64-X65 4 W44-W45 4 X66 5 W46 5 X67 4 W47 4 X68-X69 5 W48-W49 5 X70-X73 4 wso 4 X74 5 WSl 5 X75 4 W52-WSS 4 X76 5 W56-W58 5 X77 4 W59-W60 4 X78 5 W61 5 X79 4 W62 4 X80-X93 s W63-W67 5 X94-X95 4 W68 4 X96 5 W69-W71 5 W72 4 FRss* 4 W73-W83 5 W84 4 W85-W93 5 B-4 Table B.2: F u e l Assembl y Reactivity Categ o rizat io n for Fuel Assembl y IDs A through AA fo r Unit 3 Indian Point Unit 3 Fue l Assembly ID I Category I Assembly ID I Category I Assembly ID I Category V43 I 3 I V48 I 3 I I All Other Indian Po i nt 3 fuel (Batches A through AA) are Category 4 N ET-28 091-000 3-01, R ev i s i o n 0 B-5 ENCLOSURE 3 TO NL-17-144 Indian Point Unit 2 NEI 12-16 Draft Revision 2c Checklist Entergy Nuclear Operations , Inc. Indian Point Unit 2 Docket No. 50-247 Criticality Analysis Checklist
-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation
 
===1.0 Introduction===
 
and Overview Purpose of submittal YES Section 1. Remove credit for Boraflex&#x17d; Changes requested YES Section 1 and 10 Summary of physical changes YES Section 1. Boraflex&#x17d; loss Summary of Tech Spec changes YES Section 10 Summary of analytical scope YES Section 1.2 2.0 Acceptance Criteria and Regulatory Guidance Summary of requirements and guidance YES Requirements documents referenced YES Section 1.3 Guidance documents referenced YES Section 1.2 and 10.1 Acceptance criteria described YES Section 1.3 3.0 Reactor and Fuel Design Description Describe reactor operating parameters YES Section 3.4 Describe all fuel in pool YES Section 3.2 Geometric dimensions (Nominal and YES Section 3.2 Tolerance)
Schematic of guide tube patterns YES Figure 6.1 Material compositions YES Section 3 Describe future fuel to be covered YES Section 3.2 Geometric dimensions (nomina l and YES Section 3.2 to l erance) Schematic of guide tube patterns YES Section 3.2 Material compositions YES Figu re 6.1 Describe all fuel inserts YES Section 3.3. Geometric dimens io ns (nominal and YES Section 3.3 Tolerances are not used tolerance) since Depletion An alysis uses nominal dimensions. Schematic (axia l/cross section) NO Standard Westinghouse Designs Material compositions YES Sect ion 3.3 Describe non-standard fuel YES Section 8.10-8.12 Geometric dimens io ns YES Section 8.11 Describe non-fuel items in fuel cells NO Limits given in Section 8.13 Nominal and tolerance dimensions NO Limit s given in Section 8.13 4.0 Spent Fuel Pool/Storage Rack Description New fuel vault and Storage rack description N/A No change i n current license needed Nominal and tolerance dimensions N/A Schematic (axia l/cro ss section) N/A Material compositions N/A Spent fuel pool, Storage rack description YES Section 3.1 Nominal and tolerance dimensions YES Section 3.1, Tab l e 3.1 Page 1 of 7 Criticality Analysis Checklist
-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes/ Explanation Schematic (axial/cross section) YES/NO Cross sections in Figures 3.2 and 3.3. No axial details given since there is no variation in the over relevant axial heights Material compositions YES Section 3.1 (SCALE elemental brea k down of SS 304 is u sed but not specified)
Other Reactivity Contro l Devices {Inserts)
N/A There are no rack inserts. Control rods inserted in the fuel assemb li es are credited and covered with the fuel inserts. Nom inal and tolerance dimensions N/A Schematic (axial/cross section) N/A Material compositions N/A 5.0 Overview of the Method of Analysis New fuel rack analys is description N/A Storage geometries N/A Bounding assemb ly design(s)
N/A Integral absorber cred it N/A Accident analysis N/A Spent fuel storage rack analy sis description YES Section 2.0 Storage geometries YES Figure 1.1 and Section 8.5 Bounding assembly design(s) YES Batch Groupings are used. Introduced in Section 5. Fuel designs given in Section 3.2 So lub le boron credit YES Soluble boron credit is taken by inference with the criteria se l ected in Section 1.3. Boron dilution analysis YES Section 9.6 which references the current approved analysis. Burnup credit YES Section 2.0 Decay/cooling time credit YES Section 10.2 Integral absorber cred it YES Section 10.2 Other credit YES Figure 1.1 shows the credited control rods in specific locat ions Fixed neutron absorbers N/A Not taking credit for any fixed neutron absorbers Aging management program N/A Accident analysis YES Section 9 Temperature increase YES Section 9.4 Assembly drop YES Section 9.3 Single Assembly misload YES Section 9.3 Multiple misload YES Section 9.5 Page 2 of 7 Criticality Analysis Checklist
-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Boron dilution YES Sect i on 9.6 which references the current approved analysis. Other YES Sect i on 9.2 (Misplaced Assembly), Sect i on 9.7 (Seism i c) Fuel out of rack analysis (Normal Operations)
Yes Section 9.1 Handling YES Section 9.1 Movement YES Section 9.1 Inspection YES Sect i on 9.1 6.0 Computer Codes, Cross Sections, and Validation Overview Code/Modules Used for Calculation of kett YES Sect i on 2.1 Cross section library YES Section 2.1 Descript i on of nuclides used YES Table 2.1 Convergence checks YES Section 6.5 and Section 6.6.2 Code/Module Used for Depletion Calculation YES Section 2.1 and Section 5.6 Cross sect i on library YES Section 2.1 and Section 5.6 Description of nuclides used YES Section 2.1 Convergence checks YES Section 5.6 Validation of Code and Library YES Sect i on 4 and Appendix A Major Actinides and Structural Materials YES Sections 4.1 and 4.2 Minor Actinides and F i ss i on Products YES 1.5% bias (NUREG/CR-7109) -Section 4 Absorbers Credited YES Sections 4.1 and A.2.3 7.0 Criticality Safety Analysis of the New Fuel Rack Not part of License Application Rack model N/A Boundary conditions N/A Source distribution N/A Geometry restrictions N/A Limiting fuel design N/A Fuel dens i ty N/A Burnable Poisons N/A Fuel dimens i ons N/A Axial blankets N/A Limiting rack model N/A Storage vault dimensions and materials N/A Temperature N/A Multiple regions/configurations N/A Flooded N/A Low density moderator N/A Eccentric fuel placement N/A Tolerances N/A Fuel geometry N/A Fuel pin pitch N/A Page 3 of 7 Criticality Analysis Checklist
-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Fuel pellet OD N/A Fuel clad OD N/A Fuel content N/A Enrichment N/A Density N/A Integral Absorber N/A Rack geometry N/A Rack pitch N/A Cell wall thickness N/A Storage vault dimensions/materials N/A Code uncertainty N/A Biases N/A Temperature N/A Code bias N/A Moderator Conditions N/A Fully flooded and optimum density N/A 8.0 Depletion Analysis for Spent Fuel Section 5 Depletion Model Considerations Time step verification YES Section 5.6 Convergence verification YES Section 5.6 Simplifications YES Section 5.6 Non-uniform enrichments YES Axial Blankets, Sections 5.6 and 6.2 Post depletion nucl i de ad j ustments YES Section 5.9 Cooling time YES Sect i on 5.9 Depletion Parameters YES Sections 5.1-5.5 Burnable Absorbers YES Section 5.2 Integral absorbers Yes Section 5.2 Soluble Boron YES Section 5.3 Fuel and Moderator Temperature YES Section 5.1 Specific power YES Section 5.4 Control rod insertion YES Section 5.5 Atypical Cycle Operating H i story YES Section 5.3 utilized full details of cycle lengths (some short cycles) for determining average ppm. Section 5.1 utilized most limit i ng temperatures in cycles where power changed (IP2 cycle 10 and IP3 cycle 12) 9.0 Critica l ity Safety Analysis of Spent Fuel Pool Storage Racks Rack model YES Boundary conditions YES Section 6.1 Source distribution Yes Section 6.5, Section 6.6.2 Geometry restrictions YES Sect i on 10.3 Page 4 of 7 Criticality Analysis Checklist
-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Design Basis Fuel Description NO Mul tiple batch groupings as well as assemb l y specific analyses are utilized.
The secti ons given be low for this to pic are where the batch va lu es are identified. Fue l dens ity YES Sect i on 3.2 Burnable Poisons YES Section 5.2 Fuel assembly inserts YES Section 5.2 Fuel dimensions YES Section 3.2 Axial blankets YES Section 3.2 C onfi gurations conside red Bo r ated YES Section 8.14 Un borated YES All but Section 8.14 Multiple rack designs YES Sections 8.2, 8.3 , and 8.4 Alternate storage geometry YES Section 8.5 Reactivity Control Devices Fuel Assembly Insert s YES Control Rod Credit , Sections 6.6 and 8.3 Storage Cell Inserts N/A No used. Storage Cell Blocking Devices YES Sect ion 8.7 Axial burnup shapes Uniform/Distributed YES Section 6.2.1 Nodalizat i on YES Section 6.2. Blankets mode l ed YES Section 6.2 Tolerances/Uncertainties Fuel geometry Fuel rod pin pitch YES Section 7.1 Fuel pellet OD YES Section 7.1 Cladding OD YES Section 7.1 Axial fuel pos ition NO Insig nificant reactiv i ty since the rack has no axial variation (Applies to racks crediting absorbers that are not full l ength which doe s not apply to Indian Point.) Fuel content En rich ment YES Section 7.1 Dens it y YES Sect ion 7.1 Assembly inser t dimens ions and mater ials NO Deplet i on uses nominal dimensions , Control rod densities are reduced a bounding 20% (Section 6.6) Rack geometry Flux trap size (width) NO Reduction of the cell pitch reduced the flux trap width Rack cel l pitch YES Section 7.1 Rack wall thickness YES Section 7.1 Page 5 of 7 \J Criticality Analysis Checklist
-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes/ Explanation Neutron absorber dimensions N/A No credit taken for rack absorbers Rack insert dimensions and materials N/A No rack inserts used Code val i dation uncerta i nty YES Section 7.5 Criticality case uncerta i nty (statistical)
YES Section 7.5 Depletion uncertainty YES Sect i on 7.2 Burnup uncertainty YES Section 7.2 Biases Design basis fuel design NO Used most limiting fuel in each batch grouping Fuel geometry Clad creep YES Section 7.2 Grid growth (pin pitch) YES Sect i on 7.2 Minimum grid volume NO Conservatively ignored grids. Minor actinides and fission product worth YES Section 7.2 Code bias YES Section 7.5 Temperature NO Analysis was performed at most limiting temperature. See Section 8.1 for determining most limiting temperature. Eccentric fuel placement YES Section 7.4. Include in full pool analysis rather than a bias. See Section 8.4.3. lncore thimble depletion effect NO Included in the analy s is rather than a bias. See Section 5.6. NRC administrative margin NO Rather than specify a bias for the NRC administrative margin , the k 95;59 is calculated showing at least 1% margin. Calculated k 95;59 in Sections 8.3.2 and 8.4.3 Modeling simplifications Identified and described YES Section 6.2 10.0 Interface Analysis Interface configurations analyzed Between dissimilar racks Section 6.6 Between storage configurations within a rack Section 6.6 Interface restrictions NO Categorization of cell rather than an interface restriction. 11.0 Normal Conditions Fuel handling equipment NO Fuel handling equipment can only handle one assembly at a time and therefo r e do not pose a criticality concern. Fuel handling operations are in Section 9.1. Administrative controls YES Section 9.1 Fuel inspection equipment or processes YES Section 9.1 Page 6 of 7 Criticality Analysis Checklist
-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Fuel reconstitution YES Section 9.1 12.0 Accident Analysis Boron dilution YES Sect i on 9.6 which references the cur r ent approved analysis. Norm al conditions YES Sect i on 9.1. Accident conditions YES Section 9.6 which references the current approved analysis. Single assembly mislead YES Sect i on 9.3 Fuel assembly misplacement YES Sect i on 9.2 Neutron absorber insert mislead YES Sect i on 9.5 addresse s withdrawa l of required control rods. Multiple fuel mislead YES Section 9.5 Dropped assembly YES Sect i on 9.3 Temperature YES Section 9.4 Seismic event or other natural phenomena YES Section 9.7 13.0 Analysis Results and Conclusions Summary of results YES Sect i onlO Burnup curve(s) YES Section 10.2 Intermediate decay time treatment YES Section 10.2 New administrative co ntrols YES Section 9.1 Techn i cal Spec i fication markup covered Technical Specification markups YES in an a ttachment separate from the CSA report. 14.0 References Appendix A Computer Code Validation
: Code validation methodology and bases YES Appendix A New Fuel YES Section A.2 Dep l eted Fuel YES Section A.3 MOX crit i cal YES Section A.3.2 HTC critical YES Section A.3.1 H i gh temperature crit i cals YES Section A.4 Convergence NO Convergence of the Cr i tical Exper i ments is cove r ed by the same discussion of convergence for all the analysis. See Section 6.5 Trends YES Sect i on A.2.5 Bias and uncertainty YES Section A.2.6 Range of applicability YES Section A.2.8 Analysis of area of appl i cability coverage YES Section A.2.8 Page 7 of 7}}

Revision as of 19:46, 15 March 2019

Enclosure 2 - NET-28091-003-01NP, Rev. 0, Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit
ML17354A015
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 11/28/2017
From:
Entergy Nuclear Operations
To:
Office of Nuclear Reactor Regulation
References
NL-17-144 NET-28091-003-01NP, Rev. 0
Download: ML17354A015 (247)
v • d • e


Text

ENCLOSURE 2 TO NL-17-144 Curtiss-Wright Nuclear Division, NETCO Report NET-28091-003-01, Revision 0 (Non-Proprietary Version) Entergy Nuclear Operations , Inc. Indian Point Unit 2 Docket No. 50-24 7 NET-28091-003-0lNP, Rev. 0 Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit R e v: Date: 0 11/2 8/J.?~ Prepared by: Curtiss-Wright Nuclear Division, NETCO 44 Shelter Rock Rd. Danbury , CT 06810 Prepared for: Entergy Nuclear Operations

-Indian Point Energy Center under Contract No. 10502876 Prepared By: n Reviewed By: \ Approved By: ~L.-..., , *t'~~~ -'7 ,-Y:tlitlt-t *Jk< \ .Q 1'r, '-y ..

This Page Intentionally Left Blank Table of Contents 1 Introduction

.................................................................................................

1 1.1 Background

.............................................................................................................................

1 1.2 Description of the Analysis .....................................................................................................

2 1.3 Acceptance Criteria ................................................................................................................

5 2 Methodology

.................................................................................................

6 2.1 Computer Codes ......................................................................................................................

7 3 Input Data ..................................................................................................

11 3.1 SFP and Storage Rack Specifications

..................................................................................

11 3.2 Fuel Assembly Designs ..........................................................................................................

14 3.3 Fuel Assembly Insert Designs ..............................................................................................

17 3.4 Plant Operation Data ............................................................................................................

20 4 Validation

...................................................................................................

23 4.1 U02, Structural Materials, and Absorbers Validation

......................................................

24 4.2 MOX Validation

....................................................................................................................

25 4.3 Critical Experiments Effect on the Final k9s 1 9s ...................................................................

26 5 Depletion Calculations

..............................................................................

28 5.1 Limiting Depletion Parameters

-Temperatures

................................................................

29 5.1.1 Averaged Assembly Radial Peaking Factor ...........................................................

29 5.1.3 Moderator Temperature

..........................................................................................

31 5.1.4 Fuel Temperature

.....................................................................................................

34 5.1.5 Selection of Bounding Model and Temperatures

...................................................

37 5.2 Limiting Depletion Parameters

-Burnable Absorbers

.....................................................

42 5.3 Limiting Depletion Parameters

-Soluble Boron ...............................................................

43 5.4 Limiting Depletion Parameters

-Specific Power ...............................................................

44 5.5 Limiting Depletion Parameters

-Control Rod Operation

................................................

45 5.6 Depletion Analysis Model .....................................................................................................

51 NET-28091-0003-01, Revision 0 iii

5. 7 Special Case Depletions

........................................................................................................

55 5.8 Reduced Power Operation at End of Life and Fission Gases ...........................................

57 5.9 Production of Atom Density Sets .........................................................................................

58 5.10 Summary of Limiting Depletion Conditions

.......................................................................

60 6 Rack Model ................................................................................................

62 6.1 SCALE 2x2 Radial Models ...................................................................................................

62 6.2 Axial Model. ...........................................................................................................................

65 6.2.1 Axial Burn up Distribution

.......................................................................................

65 6.3 Dimensional Changes with Irradiation

...............................................................................

72 6.3.1 Clad Creep .................................................................................................................

72 6.3.2 Grid Growth ...........................

...................................................................................

79 6.4 Averaged Assembly Peaking Factor Interpolation

............................................................

82 6.5 Convergence of the 2x2 Infinite Model Calculations

.........................................................

83 6.6 Full Pool Models ....................................................................................................................

84 6.6.1 Sensitivity of the Full Pool Model to Modeling Assumptions

................................

87 6.6.2 Convergence of the Full Pool Model... .....................................................................

88 6.7 Summary of Modeling Assumptions

...................................................................................

93 7 Sensitivity Analysis ....................................................................................

94 7.1 Manufacturing Tolerances

...................................................................................................

94 7.2 Burn up Dependent Biases and Uncertainties

.....................................................................

96 7.3 Eccentricity

............................................................................................................................

99 7.4 Additional Biases and Uncertainties

..................................................................................

102 7.5 Biases and Uncertainties Rack-up .....................................................................................

104 7.6 Interface Uncertainty Treatment.

......................................................................................

107 8 Results .......................................................................................................

I 08 8.1 Temperature Effects ...........................................................................................................

108 8.2 Region 1 Fuel Categories 1 and 2 .......................................................................................

109 NET-28091-0003-01, Revision 0 iv

8.3 Region

2 Category 4 Batch Grouping Z -Current and Future Fuel .............................

111 8.3.1 Curve Fit ..................................................................................................................

113 8.3.2 Confirmation Calculations for Category 4 ...........................................................

114 8.4 Determination of Burnup Requirements for Categories 3 and 5 ....................................

115 8.4.1 Cell Category Layout in Region 2 ..........................................................................

115 8.4.2 Additional Burnup Requirements for Fuel Categories 3 and 5 ..........................

116 8.4.3 Confirmation of k9s 1 9s for Full Pool (includes Category 3 and 5) ........................

120 8.5 Alternate Arrangements for Region 1 ...............................................................................

124 8.6 Calculations for Discharged Fuel (IP2 A-X and IP3 A-AA) ...........................................

127 8.7 Cell Blockers ........................................................................................................................

135 8.8 Region 2 Checkerboard

......................................................................................................

136 8.9 Burnup Penalty for Hafnium Flux Suppression Inserts ..................................................

136 8.10 Failed Fuel Containers

........................................................................................................

136 8.11 Fuel Rod Storage Basket ....................................................................................................

138 8.12 Assemblies with Missing Fuel Rods ...................................................................................

139 8.13 Storage of Miscellaneous Materials

...................................................................................

141 8.14 Borated Conditions

.............................................................................................................

141 8.15 Burnup Penalty for High Soluble Boron Conditions

.......................................................

143 9 Normal Operations and Accident Analysis ...........................................

144 9.1 Normal Operations

.............................................................................................................

145 9.2 Misplaced Assembly ............................................................................................................

146 9.3 Dropped Assembly ..............................................................................................................

149 9.4 Over Temperature

..............................................................................................................

150 9.5 Multiple Misloads ................................................................................................................

151 9.6 Boron Dilution Accident .....................................................................................................

152 9.7 Seismic Event .......................................................................................................................

153 10 Summary ..................................................................................................

154 NET-28091-0003-01 , Revision 0 V J 10.1 Review of DSS-ISG-2010-01

...............................................................................................

154 10.2 Fuel Reactivity Categorization

..........................................................................................

158 10.3 Allowable SFP Cells for Each Fuel Category ...................................................................

160 10.4 Fuel and Operating Requirements

....................................................................................

163 References

.......................................

................................................................

166 Appendix A: Validation of SCALE 6.1.2 for Criticality Analysis Using Laboratory Critical Experiments

.............................................

A-1 A.1. Overview ..................................................................................................

A-1 A.2. U02 Laboratory Critical Experiments

.................................................

A-1 A.2.1 Introduction

.........................................................................................................................

A-1 A.2.2 Definition of the Range of Parameters to Be Validated

...................................................

A-2 A.2.3 Selection of the Fresh U02 Critical Benchmark Experiments

........................................

A-2 A.2.4 Computer Analysis of the U02 Benchmark Critical Experiments

...............................

A-10 A.2.5 Statistical Analysis of the Fresh U02 Critical Benchmark Results ..............................

A-20 A.2.6 Establishing the Bias and the Uncertainty

......................................................................

A-28 A.2.7 Subcritical Margin ............................................................................................................

A-29 A.2.8 Area of Applicability (Benchmark Applicability)

..........................................................

A-29 A.2.9 Summary of U02 Laboratory Critical Experiment Analysis ........................................

A-32 A.3. HTC and MOX Critical Experiments

................................................

A-33 A.3.1 HTC Critical Experiments

...............................................................................................

A-33 A.3.2 MOX Critical Experiments

................

..............................................................................

A-39 A.3.3 Bias and Uncertainty from the MOX/HTC Critical Experiments

...............................

A-43 A.4. Temperature Dependent Critical Experiments

.................................

A-44 A.5. Summary of Validation Using Laboratory Critical Experiments

... A-48 A.6. Appendix References

............................................................................

A-49 Appendix B: Fuel Categorization for Unit 2 Batches A Through X and Unit 3 A through AA ...................................................................

B-1 NET-28091-0003-01, Revision 0 VI List of Tables Table 2.1: 185 Isotopes Used in the Analysis .....................................................................................

8 Table 3.1: Region 1 and 2 Storage Rack Dimensions

[8, 9) ...........................................................

14 Table 3.2: Fuel Assembly Dimensions

[11, 12) ................................................................................

17 Table 3.3: Control Rod and Hafnium Rod Descriptions

[11) ........................................................

19 Table 3.4: Pyrex and Wet Annular Burnable Absorber Descriptions

[11, 12, 15] ......................

19 Table 3.5: Key Operating Features by Cycle Used in IP2 .............................................................

21 Table 3.6: Key Operating Features by Cycle Used in IP3 .............................................................

22 Table 5.1: Moderator Exit Temperature, Texit , versus Peaking Factor for Batch Groups ..........

31 Table 5.2: Moderator Exit Density versus Peaking Factor for Batch Groups .............................

32 Table 5.3: Enthalpy Node Factor versus Axial Burnup Shape .....................................................

33 Table 5.4: Moderator Temperature (K) at each Node versus Burnup Profile ............................

33 Table 5.5: Fuel Temperature (K) at each Node versus Burnup Profile .......................................

36 Table 5.6: Fit Coefficients for Top Node Moderator Temperature and Density .........................

40 Table 5.7: Fit Coefficients for 3rd Node Moderator Temperature and Density ...........................

40 Table 5.8: Burnable Absorbers versus Batch Grouping ................................................................

43 Table 5.9: Soluble Boron versus Batch Grouping ..........................................................................

43 Table 5.10: Assemblies under D-Bank for the First 21 Cycles of IP2 ..........................................

.45 Table 5.11: Effect of Modeling the Bite Position rather than Burnable Absorbers

....................

4 7 Table 5.12: Burn up Penalty for Assem. with Burnable Absorbers followed by Bite D-bank ..... 48 Table 5.13: Assemblies with BA Inserts plus under D-Bank in Non-Bite Cycles ........................

48 Table 5.14: Assemblies under D-Bank for the first 11 Cycles ofIP3 ............................................

50 Table 5.15: SCALE/TRITON minus CASM0-5 Ak of Depletion at 100 Hours Cooling ............

53 Table 5.16: SCALE/TRITON minus CASM0-5 Ak of Depletion at 5 Years Cooling ................

53 Table 5.17: SCALE/TRITON minus CASM0-5 Ak of Depletion at 15 Years Cooling ..............

54 NET-28091-0003-01 , Revision 0 VII Table 5.18: Percent Difference in the Ak of Depletion at 100 Hours Cooling ..............................

54 Table 5.19: Percent Difference in the Ak of Depletion at 5 Years Cooling ...................................

54 Table 5.20: Percent Difference in the Ak of Depletion at 15 Years Cooling .................................

55 Table 5.21: Special Case Depletion Parameters

.............................................................................

55 Table 5.22: Verification of Cooling Time Model in the Interpolation Program ..........................

60 Table 6.1: Axial Burn up Profile vs. Burnup Bin [27) .....................................................................

67 Table 6.2: Axial Relative Burnups for Blanketed Discharged Fuel.. ............................................

70 Table 6.3: Axial Relative Burnups for Batch Z Fuel.. ....................................................................

71 Table 6.4: Calculated k versus Number of Nodes Modeled ...........................................................

72 Table 6.5: Full Pool Model Sensitivity Tests ...................................................................................

88 Table 6.6: k e rr Changes With Start Source ......................................................................................

90 Table 7.1: Tolerance Reactivity Effects ...........................................................................................

94 Table 7.2: Eccentricity Results .......................................................................................................

101 Table 7.3: Total Bias and Uncertainties for Region 1 , Categories 1, 2, 3 ...................................

105 Table 7.4: Sample Category 4 and 5 Bias and Uncertainty Rack-up .........................................

106 Table 7.5: Total Bias and Uncertainty for Fresh Fuel in Region 2 .......................

......................

106 Table 8.1: Calculated k e rr as a Function of Temperature

.............................................................

109 Table 8.2: Confirmation of Region 1 Requirements for Category 1 and 2 Fuel.. ......................

110 Table 8.3: Change in k err with Burn up and number of IFBA Rods ............................................

110 Table 8.4: Minimum Burnup Requirements (GWd/T) for Category 4 Batch Grouping Z ...... 111 Table 8.5: Curve Fit Coefficients for Category 4 Fuel.. ...............................................................

113 Table 8.6: Calculated k e rr Values at each Category 4 Batch Z Burnup Point ............................

114 Table 8.7: Total Bias and Uncertainty at each Category 4 Batch Z Burnup Point.. .................

114 Table 8.8: k 9s19s for each Category 4 Batch Z Burnup Point .......................................................

115 Table 8.9: Region 2 Models at Loading Curve (Cat 5 is Cat 4 plus 11 GWd/T) ........................

121 NET-28091-0003-01, Revi s ion 0 V lll Table 8.10: Eccentric Options for Region 1 ...............

................

...................................................

122 Table 8.11: Maximum Full Pool k9s 1 9s assuming Various Cycle Lengths ...................................

124 Table 8.12: Dependence of k e rr on the Region 1 Arrangement

....................................................

126 Table 8.13: Batch A-D Minimum Burnup Requirements (GWd/T) for Category 4 ................

128 Table 8.14: Batch E-F Minimum Burn up Requirements (GWd/T) for Category 4 .................

128 Table 8.15: Batch G-L Minimum Burnup Requirements (GWd/T) for Category 4 ................

129 Table 8.16: Batch M-P Minimum Burnup Requirements (GWd/T) for Category 4 ................

129 Table 8.17: Batch Q-S Minimum Burnup Requirements (GWd/T) for Category 4 .................

130 Table 8.18: Batch T-V Minimum Burnup Requirements (GWd/T) for Category 4 .................

130 Table 8.19: Batch W Minimum Burnup Requirements (GWd/T) for Category 4 ....................

131 Table 8.20: Batch X Minimum Burnup Requirements (GWd/T) for Category 4 .....................

131 Table 8.21: Batch A-U (IP3) Minimum Burnup Requirements (GWd/T) for Category 4 ....... 132 Table 8.22: Batch V-X (IP3) Minimum Burnup Requirements (GWd/T) for Category 4 ....... 132 Table 8.23: Individual Assembly Analysis for Category 3 .....................................

......................

133 Table 8.24: Individual Assembly Analysis for Category 4 ...........................................................

134 Table 8.25: Failed Fuel Container Pin Analysis ............

.............................

..................................

138 Table 8.26: Normal Operations with Boron Dilution ppm (Full Pool Model) ...........................

142 Table 8.27: Burnup Penalty Results at 1200 ppm ........................................................................

143 Table 9.1: Misplaced 5.0 w/o 64 IFBA Assemblies with 2000 ppm .............................................

148 Table 10.1: DSS-ISG-2010-01 Checklist

..................

...............

.......................................................

154 Table 10.2: Summary of Loading Requirements for Fuel Batch Z ....................

........................

159 Table 10.3: Fuel Design Requirements for Batch Z assemblies

..................................................

164 Table 10.4: Fuel Assembly Operating Requirements

.......................................

............................

165 Table A.l: Selection Review ofOECD/NEA Criticality Benchmarks

........................................

A-3 Table A.2: Critical Experiment Results with SCALE 6.1.2 and ENDF/B-VII

........................

A-11 Table A.3: Summary of Critical Experiments Containing Boron ............................................

A-18 NET-28091-0003-01, Revision 0 I X Table A.4: Wilk-Shapiro Test Results Output From DATAPLOT (4) ....................................

A-21 Table A.5: Area of Applicability (Benchmark Applicability)

...................................................

A-29 Table A.6: HTC Phase 1 Results ..................................................................................................

A-34 Table A.7: HTC Phase 2a, Gadolinium Solutions, Results ........................................................

A-35 Table A.8: HTC Phase 2b, Boron Solutions, Results .................................................................

A-36 Table A.9: HTC Phase 3 Results -Water Reflected Assemblies

..............................................

A-37 Table A.10: HTC Phase 4 Results -Steel Reflected Assemblies

...............................................

A-38 Table A.11: Results of MOX Critical Benchmarks (SCALE 6.1.2, ENDF/B-VII)

..................

A-40 Table A.12: LCT-46 with Full Thermal Expansion

...................................................................

A-44 Table A.13: LCT-46 with No Thermal Expansion of Solids ......................................................

A-47 Table B.1: Fuel Assembly Reactivity Categorization for Assembly IDs A -X for Unit 2 .........

B-2 Table B.2: Fuel Assembly Reactivity Categorization for Fuel Assembly IDs A -AA for IP3 .. B-5 NET-28091-0003-01 , Revision 0 X List of Figures Figure 1.1: Fuel Category Placement in the IP2 SFP (base case) ...................................................

3 Figure 3.1: IP2 SFP Taken From Holtec Drawing #397 [35) ........................................................

12 Figure 3.2: Small Section of the Region 1 Rack [8] ........................................................................

13 Figure 3.3: Region 2 Rack Showing Cell Boxes and Resultant Cells [9] ......................................

13 Figure 5.1: Averaged Assembly Peaking Factors of Assemblies in the IP2 SFP .........................

30 Figure 5.2: Fuel Temperature Change with Burnup and Relative Power ...................................

35 Figure 5.3: Fuel Temperature (K) at 25 GWd/T vs. Peaking Factor (PF) at the Top Node ....... 36 Figure 5.4: Top Node Moderator Temp (K) vs. Average Assembly Peaking Factor ..................

39 Figure 5.5: Top Node Moderator Density vs. Avg Assembly Peaking Factor ............................

.40 Figure 6.1: Region 1 KENO Model. .............

...........

.......................

..............................

....................

63 Figure 6.2: Region 2 KENO Model ...................................................................................

...............

64 Figure 6.3: Comparison of Creep-down for ZIRLOŽ and Zircaloy-4

[37] ................................

73 Figure 6.4: Diameter Decrease versus Exposure Time [39) ............................

............................

... 74 Figure 6.5: Clad Creep Down for Vandellos 2 Nuclear Power Plant [40) ....................................

75 Figure 6.6: Axial Distribution of the Fuel Rod Diameter at 50.5 GWd/T [41] ............................

76 Figure 6.7: Oxide Layer thickness with Burnup [42) .....................................................................

77 Figure 6.8: Density of Fuel Pellet as a Function of Pellet Burnup [43) .........................................

77 Figure 6.9: ZIRLOŽ Grid Growth [42) ..............................................

.................

..........................

80 Figure 6.10: Zircaloy-4 and MS Grid growth versus burnup [44) ................................................

80 Figure 6.11: Grid Growth of ZIRLOŽ and Zircaloy-4 versus Elevation

[45) ............................

81 Figure 6.12: Calcu lat ed kerr versus Assembly Average Peaking Factor ........................................

83 Figure 6.13: Full Pool Model ......................

......................................................................................

85 Figure 6.14: Model of Module H Showing Control Rods .............................................................. 86 Figure 6.15: Locations of the Start Sources for the Convergence Tests .......................................

91 N ET-28091-0003-01, Revi s ion 0 XI Figure 6.16: Change in Average k err with Progressing Generations

.............................................

92 Figure 7.1: Category 4 Region 2 with 16 Assemblies Eccentrically Placed ................................

100 Figure 7.2: Eccentric Model for Category 2 with Central Row Shifted Down ..........................

102 Figure 8.1: Calculated kerr as a Function of Category 5 Burnup Using 5.0 w/o Fuel... .............. 118 Figure 8.2: kerr as a Function of Category 5 Burnup Using 4.2 w/o Enriched Fuel ...................

119 Figure 8.3: Calculated k e rr as a Function of Category 3 Burnup Using 5.0 w/o Fuel... ..............

120 Figure 8.4: Refueling Arrangement

..........

.....................................................

................................

125 Figure 8.5: No Cat 2 Arrangement

................................................................................................

125 Figure 8.6: Max Cat 1 Arrangement

.............................................................................................

126 Figure 8.7: Example Odd Arrangement

........................................................................................

126 Figure 8.8: Cell Blocker Region 2 Model ....................

.........................................

.........................

135 Figure 8.9: Failed Fuel Container Pin Model .............

..........

.................

..................................

..... 137 Figure 8.10: Model for the Fuel Rod Storage Basket.. ...............

...................

..........

.....................

138 Figure 8.11: k err versus Number of Missing Fuel Rods ................................................................

140 Figure 8.12: Model for Assemblies with 36 Missing Fuel Rods [1] ..............

...............................

140 Figure 9.1: Misplaced Assembly at the Cask Area Corner .....................................

.................... 147 Figure 9.2: Misplaced Assembly between the Fuel Elevator and the Rack ...............................

148 Figure 9.3: Full Pool Model with Dropped Assembly ................................................................

.. 150 Figure 10.1: Fuel Category Location Requirements (Base Case) ...............................................

161 Figure 10.2: Refueling Arrangement

.............................................................................................

162 Figure 10.3: Max Cat 1 Arrangement

..........................

.................................................................

162 Figure 10.4: Example Odd Arrangement

......................................................................................

162 Figure A.1: Distribution of the Calculated k err values Around the Mean .................................

A-22 Figure A.2: k err as a Function of the Energy of the Average Lethargy Causing Fission .........

A-24 Figure A.3: kerr as a Function of the Pin Diameter .....................................................................

A-25 Figure A.4: k e rr as a Function of the Lattice Pitch ......................................................................

A-26 NET-28091-0003-01, R evision 0 X II Figure A.5: k.rr as a Function of the Fuel Enrichment

...............................................................

A-27 Figure A.6: k.rr as a Function of the Soluble Boron Content..

...................................................

A-28 Figure A.7: k 0 rr as a Function of the EALF for the HTC Experiments

.....................................

A-39 Figure A.8: Predicted k.rr as a Function of the Plutonium Content..

........................................

A-42 Figure A.9: Predicted k.rr as a Function of the Am-241 Content .............................................. A-43 Figure A.10: LCT-046 Corrected Calculated k.rr per Case .......................................................

A-45 Figure A.11: LCT-046 Corrected Calculated k 0 rr Versus Temperature

...................................

A-46 NET-28091-0003-01 , Revision 0 Xlll 1 Introduction This report s ummari zes th e 20 1 7 criticality safe ty a n alysis (CSA) for the Indian P oint U nit 2 (IP2) spe nt fuel pool (SF P) takin g no cre dit for a b sorber pan e l s. The current 200 l CSA of record [36] t akes partial credit for Boraflex TM panels which h ave degrad e d an d co ntinu e to d egra d e. In order to remo ve the dep e ndenc e on the Borafl ex TM panels thi s new 20 17 CSA credits empty ce ll s , control rods , and leak age a lon g th e outer two storage rows of the SFP. In 2 015 , a CSA to remo ve cre dit for the Boraflex TM pan e l s which used ne w metal-matrix-composite absorber in se rt s was previously submitte d to the Nuclear Regulatory Conunission (NRC) an d was r eviewe d [l , 2 , an d 3]. How ever , the approach t ake n in this 2017 CSA i s ex p ecte d to re s ult in a more tim ely resolution of the Borafl ex TM degrad ation i ss ue. Since th e Indian Point Energy Ce nt er (IPEC) utili zes the Un it 2 (IP2) SFP for t emporary storage of U nit 3 (IP3) fuel prior to pl aceme nt into dry s torage casks , this 2 017 CSA a llow s storage in the IP2 SF P of all fuel asse mb l ies dis c h arge d from b oth IP 2 and IP 3. 1.1 Background The IP 2 SFP rac k s currently c redit Bo raflex TM as the n e utron absorber, w hich is known to degr ade over time. Du e to this fact , E nt ergy (the operator of IP EC) w ill no lon ger take c redit for the Borafl ex TM for r eac ti v it y h o ld-down. In early 2015, IP EC s ubmitted a 20 15 criticality a n a ly s is which used neutron a bsorber insert s to replace th e negative reactivity of th e Bo raflex TM [ 1]. During this process , the NRC re qu este d additional infonn ation in June 2015 a nd IP EC i ss u ed a r es pon se in August 2015 [2]. In November 2015 th e NRC i ssue d a staff review of this criticality a naly s i s , conc ludin g th at " The NRC staff fi nd s that the CSA methodology is acce ptabl e for u se at IP 2" [3]. Subsequent to thi s 2015 CSA re v i ew, scopi n g s tudies determined th at it wo uld be more timely and l ess c hall e n ging to refuelin g outages to l oa d a dditional fuel assemblies int o casks for dr y cask s tora ge and use e mpt y cells a nd control rods for criticality control. NET-28091-0003-01, Revi s ion 0 IP2 and IP3 are both 4-loop Westinghouse power p l ants that uti l ize the l Sx 15 fuel assemb l y design. The physical dimension requirements of the fuel of both units are the same , as both units have had all of their fuel assemblies manufactured by Westinghouse.

IP3 does not have the capability to load dry storage casks, so fuel from IP3 is moved to the IP2 SFP for temporary storage prior to l oading into dry cask storage. Placement of IP3 fuel is currently restricted to Region 1-2 in the IP2 SFP. This 2017 CSA allows the IP3 fuel to be placed anywhere in the IP2 SFP , so long as it meets the reactivity requirements outlined herein. 1.2 Description of the Analys i s This 2017 CSA determines the loading criteria for storage of fuel assemblies in the IP2 SFP by taking credit for empty cell locations , control rods , and the periphery (outer two rows) of the SFP. The analysis does not credit any Boraflex TM neutron absorber tha t might remain in the racks. Taking credit for empty cells and control rods can accommodate the current and future spent fuel inventory.

The analysis defines five reactivity categories for the fuel and defines storage locations for each reactivity category.

The categories are numbered from one to five with a Category 1 fuel assembly being the most reactive and a Category 5 fuel assembly being the least reactive.

Similarly , each cell in the SFP has also been assigned a category number with a Category 1 cell being able to accept the most reactive fuel whi l e a Category 5 cell can only accept the least reactive fuel. For example, a Category 1 cell can accept all categories of fuel while a Category 5 cell can only accept a Category 5 fuel assembly because all other (lower numbered) categories are more react ive. Figure 1.1 below shows the base case arrangement of the fue l categories in the IP2 SFP. Note that the base case arrangement only shows four reactivity categories since it is the most l imiting reac t ivity arrangement.

Category 1 fuel , which is missing on Figure 1.1, is only needed for fresh fuel* or l ow burned fuel, which should not be present after IP2 shuts down. Category l fuel will be controlled by two rules for placement in Region 1. The

  • Throughout this document , fre s h fuel is used to describe fuel that has never been in the core. NET-28091-0003-01, Revision 0 2 assemb l y's enrichment , bu mup , coo lin g time , and averaged assemb l y peaking factor* are u sed to detennine the reactivity category. For permanently di scharged fuel, full use of operating data i s u sed to precisely obtain the change in react i vity with bumup. For fue l that m ay sti ll b e placed in the core , bounding d ep l etion va lu es are used. I 2 3
  • S ' 7 I 9 10 11 U U 14 15 16 17 11 1' 20 Z1 ll lJ 24 25 2fi 27 21 29 30 31 H I--+-+-+--+--+--+--+--+--+-, ~--1--1---,l--of--f--+-+-+-+--+--+--+--+--+-+--+-+-+--l--l

--,I G D I--+-+-+--!--+--+--+--+--+---<

>-t-t-+--+--+--+--+-+-+--+--+--+--+--+---+---+--+---+--t---t

....... C I--+-+-+--+--+--+--+--+--+-, ~--1--1-f--f--l--+-+-+-+--+--+--+--+--+-+-+-+--+--1

--1-8 Key: 0 ~OP 0 ON OM Ol DI( OJ OH DG Of OE CP HH-+-+-+--+-+--1-+-+-l OwaterHole D S0%W ate r Hole

  • category t Fuel D category 2 Fuel 0 category 3 Fuel .. CP CN CH """"'--fl

....... --+-+-+-+--+--+-

+-+-1-1

  • category4Fuel ca t egory 5 Fu e l category s Fuel with a requi r ed full length RCCA [!) Blo t ked Cell Cl Cl( CJ QI CG a as 8l .. BJ 8H 8f 8( Al ... AJ AH AG Al ... AO AC CM HH-+-+-+--+-+--+-+--+-l CL 1--1,--i--+-+-+-t--+--+-+-+-i CK HH-+-+-+--+-+--1-+--+-l CJ HH-+-+-+--+-+-+-+-+-l CH i--1,--+--+-+-+-t--+-+""+-+-i CG HH-+-+-+--t--+--1-+-+-l Cf HH-+-+-+-++-+-+-+-l C, 1--1,--i--+-+-+-t--+--+--+-+-i co F9'=\=*-"p\,={=4=!

"'*'~ 1-HH-+-+-+--++-+-+-'

8H BM Bl .. BJ 8H BG Bf 8{ B O H-t-+-1-+,..-:~rt-H BC j Cask Area l--l,--l--t--t--t--+--t--t-+-r-lt-11-1--t--t--t--r-t--t--t---t-+-'l'-1 AB ...... ..,....~~~

........ ...._.._ ....... ..._ ................

_. ...... _._~--~-~~ Figure 1.1: Fuel Category Placement in the IP2 SFP (base case)

  • The averaged a s sembly peaking factor is the assemb l y burn up divided by the sum of the cycle burn ups for the cycles the assembly was in th e core. T hrou g h out this document, this average d assemb l y p ealci n g factor i s often a bb reviate d as pealcing factor or PF. NET-28091-0003-01, Re v i sion 0 3 For most permanently discharged fuel from Units 2 and 3, a table of the fuel categorization for each assembly ID is provided.

This approach is taken since there are mu l tiple groupings of assemblies (with s imilar operational characteristics) due to changes in c ore power (temperatures), burnabl e absorbers , and axial blanket de signs. This CSA provides the discharged assembly categories in Appendix B which will be added to the Technical Specifications. Fuel assemblies that may still be placed in the core are categorized by a set of simple equations to determine the fuel reactivity category.

The fuel categories are numbered from most reactive fuel (Category 1 fuel) to lea st reactive fuel (Category 5 fuel). The storage loc atio ns in the SFP that can accommodate each category of fuel are limited by Figure 1.1 and modifications to this arrangement are specified and analyzed in this CSA. Details of the analysis methodology are provided i n the following sections of thi s report. Section 2 contains a summary of the methodology.

Section 3 provides the input data for the rack designs, fuel assembly design s, fuel assembly inserts, and plant operation data. Section 4 describes the computer code validation and determination of co de bia s and uncertainty.

Section 5 describes the depl etio n analysis and the selection of bounding input parameters.

Section 6 de scribes the basic fuel rack models , while Section 7 contains the sensitivity analysis of manufacturing tolerances and additional biases and un certaint ie s. Section 8 contains the result s of the reactivity calculations. Section 9 contains a summary of normal operation s and the accident analysi s. Section 10 provides a summary of the analysis including the limits of the analysis.

Appendix A contains the detailed results of the validation of SCALE 6.1.2 and its applicability to this analysis, while Appendix B con t ains the categorization of fuel assemblies previous l y discharged from IP2 and IP3. This new CSA for the IP2 SFP follows the most recent methods. The Nuclear Energy Institute (NEI) has been working with the NRC to produce guidance for SFP analysis [ 4]. This CSA closely follows the NEI guidance.

The NEI guidance started with the NRC draft Interim Staff Guidance (ISG) DSS-ISG-2010-01 [5]. All of the requirements set in DSS-ISG-2010-01 are met and are reviewed in Section 10.1. Additional guidance was provided by the Kopp Memo [56] for depletion of atom densities , NUREG/CR-NET-28091-0003-01, Revision 0 4 7109 [22] for the worth of minor actinides and fission products, and NUREG/CR-6998 [26] for the burnup uncertainty. 1.3 Acceptance Criteria The acceptance criteria of the analysis are to ensure compliance with lOCFRS0.68

[6]. Specifically, the analysis demonstrates that:

  • the k 9s 1 9s of the SFP is less than 1.0 after accounting for a ll biases and uncertainties when not taking credit for soluble boron (with a 95% probability at a 95% confidence level) [6], and
  • the k9 5 1 9s of the SFP is less than 0.95 after accounting for all biases and uncertainties when taking credit for solub l e boron (with a 95% probability at a 95% confidence level) [6]. In addition to meeting the above criteria , an eng i neering safety margin is provided to cover unanticipated issues. The engineering safety margin used is 1 %, so that the k 9 5 1 9 5 target value is 0.99 for no soluble boron and 0.94 with solub l e boron. NET-28091-0003-01, Revision 0 5 2 Methodology The CSA performed in this report uses a method that is comprised of the following steps. Each step refers to a section in this report where further information is provided. 1. Review the current IP2 SFP rack design (Section 3.1). 2. Review the historical and projected fuel designs and inserts for use in IP2 and IP3. Ensure that the analysis covers all of the designs (Sections 3.2 and 3.3). 3. Review the historical and projected operating cycles of IP2 and IP3 (Section 3.4). 4. Validate the computer codes for the application (Section 4). 5. Deplete the fuel using a two-dimensional (2-D) lattice representation of the core using bounding depletion values (including bounding burnab l e absorbers) for sets of fuel assemblies (down to individual assemblies) (Section 5). 6. Develop a radia ll y infinite three-dimensiona l (3-D) Monte Carlo mode l of the Region 1 and Region 2 racks using periodic boundary conditions. The axial modeling height is finite, including conservative modeling of the axial bumup distribution (Sections

6.1 through

6.5). 7. Develop full pool models to take advantage of leakage at the boundaries of the SFP as well as control rods at specific locations.

Use this model in checking interfaces between category cells (Section 6.6). 8. Based on the radially infinite 3-D Monte Carlo model , determine the reactivity effects associated with the manufacturing and fuel tolerances (Section 7.1). 9. Determine the bias and uncertainty associated with bumup (Section 7.2). 10. Determine the bias due to eccentric placemen t of fuel assemblies in the rack cell (Section 7.3). 11. Ascertain through analysis the most limiting SFP temperature by Region (Section 8.1). 12. Use the radially infinite 3-D Monte Carlo model with the combined biases and uncertainties to detennine the minimum bumup as a function of enrichment, averaged assembly peaking factor , NET-2809 1-0003-01, Revision 0 6 and cooling time for Categories 1, 2, and 4 at the most l imiting SFP temperature. This analysis is perfonned with no soluble boron (Sections 8.2 and 8.3). 13. Determine the fuel Category 3 and 5 additional burnup requirement and test the Region and category cell interfaces using a full pool 3-D Monte Carlo model (Section 8.4). 14. Perform accident analyses (dropped assembly, misplaced assembly, over temperature (boiling SFP water), boron dilution , seismic , and multiple assembly misloads) with the appropriate models (Section 9). 15. Summarize the resulting loading requirements and the limits of the ana l ysis (Section 10). 2. 1 Computer Codes This analysis uses the t5-depl TRJTON module of SCALE 6.1.2 [7] for the depletion analys i s and the CSAS5 module for the criticality ana l ysis. All of the analyses are perfonned using the 238 group ENDF/B-VII.O library (v7-238).

The CSAS5 mod ul e utilizes CENTRM and BONAM! for the resonance self-shielding calculations and KENO V.a for the Monte Carlo calculation of k.rr*. Unless noted, all of the CSAS5 computer runs use a Monte Carlo sampling of at least 8000 generations and 8000 neutrons per generation to achieve a statistical uncertainty in k err of less than 0.0001. The t5-depl sequence of TRJTON utilizes CENTRM and BONAM I for the resonance treatment and then uses KENO V.a for the collapsing of the cross-sections from 238 groups to one group for use in ORJGEN. The input parameter, parm=(a d dnux=4), is used in the analysis which tracks the maximum number of problem specific collapsed isotopes (388). At the end of the depletion ana l ysis , the OPUS module is used to output atom densities for use in the criticality model. In the OPUS input , 185 isotopes are specified, as shown in Table 2.1. The isotopes that are not included have low atom densit i es (less than l E-12), combined with small cross-sections, in the spent fuel composition. In other words, the e l iminated isotopes do not impact the reactivity of the spent fuel and consequently will not impact the criticality

  • Throughout this document , ke rr is used as a short hand notation fork-effective. NET-2809 1-0003-01, Revision 0 7 analysis.

Immediately after shutdown, there is an increase in reactivity in the first few days due to the decay of Xe-135 and Np-239 (poison is being removed and fissile Pu-239 is being added). Rather than follow this change in reactivity and to assure that the peak reactivity occurs at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> , all of the Xe-135 is converted to Cs-135 and all of the Np-239 i s converted to Pu-239. As pre v iously mentioned , atom densities les s than lE-12 are eliminated.

Table 2.1: 185 Isotopes Used in the Analysis Isotope Isotope Isotope Isotope Isotope Isotope Isotol!e Ag-109 Cm-243 Gd-160 Nd-145 Rb-85 Sm-153 Te-130 Ag-llOm Cm-244 Ge-73 Nd-146 Rb-86 Sm-154 Te-132 Ag-111 Cm-245 Ge-76 Nd-147 Rb-87 Sn-1 15 U-234 Am-241 Cm-246 Ho-165 Nd-148 Rh-103 Sn-116 U-235 Am-242m Cs-133 I-127 Nd-150 Rh-105 Sn-117 U-236 Am-243 Cs-134 I-129 Np-237 Ru-100 Sn-118 U-237 As-75 Cs-135 I-131 Np-238 Ru-101 Sn-119 U-238 Ba-134 Cs-136 I-135 Np-239 Ru-102 Sn-120 Xe-12 8 Ba-135 Cs-137 In-115 0-16 Ru-103 Sn-122 Xe-129 Ba-136 Dy-160 Kr-82 Pd-104 Ru-104 Sn-123 Xe-130 Ba-137 Dy-161 Kr-83 Pd-1 05 Ru-105 Sn-124 Xe-131 Ba-138 Dy-162 Kr-84 Pd-106 Ru-106 Sn-125 Xe-132 Ba-140 Dy-16 3 Kr-85 Pd-1 07 Ru-99 Sn-126 Xe-133 Br-81 Dy-164 Kr-86 Pd-108 Sb-121 Sr-86 Xe-134 Cd-110 Er-166 La-138 Pd-110 Sb-123 Sr-88 Xe-135 Cd-111 Eu-151 La-139 Pm-147 Sb-124 Sr-89 Xe-136 Cd-112 Eu-15 2 La-140 Pm-148 Sb-125 Sr-90 Y-89 Cd-113 Eu-153 Mo-100 Pm-148m Se-76 Tb-159 Y-90 Cd-114 Eu-154 Mo-95 Pm-1 49 Se-77 Tb-160 Y-91 Cd-115m Eu-155 Mo-96 Pm-1 51 Se-80 Tc-99 Zr-91 Cd-116 Eu-156 Mo-97 Pr-141 Se-82 Te-122 Zr-93 Ce-140 Gd-152 Mo-98 Pr-143 Sm-147 Te-124 Zr-95 Ce-141 Gd-154 Mo-99 Pu-23 8 Sm-148 Te-125 Zr-96 Ce-142 Gd-155 Nb-95 Pu-239 Sm-149 Te-126 Ce-143 Gd-156 Nd-142 Pu-240 Sm-150 Te-127m Ce-144 Gd-157 Nd-143 Pu-241 Sm-151 Te-128 Cm-242 Gd-158 Nd-144 Pu-242 Sm-152 Te-129m NET-28091-0003-01 , Revision 0 8 In addition to using SCALE, a FORTRAN code (INTRPND) is u sed to interpolate between bumup s from the OPUS outpu t and also to decay the isotopic content to the de sired coo ling time. The INTRPND code, which has been verified and va lidated [ 1 O], reads an axial burn up profile to get the shape of the bumup axially, so multiple atom density sets can be made quickly. The code was validated by comparing the k eff calculated with the code-interpolated number den sities to the k e tr calcu lated with number densitie s directly from SCALE/TRITON, in which n o interpo l ation is used. Furthermore , SCALE/TRITON is used to decay to a given cooling time and similar comparisons were made. A ll of the differences between the k e ff va lue s based upon the interpolated isotopics an d the SCALE direct isotopics are with in the statistical unc ertainty of the k e ff calculations (see Section 5.9). The INTRPND FORTRAN program is contro lled under NETCO's quality assurance program that meets the requirements of lOCFRSO , Appendix B, 10CFR21 , and ASME NQA-1. The program ha s been audited by NUPIC. NETCO maintains documented procedures and assigned responsibilities to control the engineer ing activities relative to the acquisition, classification, d eve l opment , testing, eval uation , modification, use, maintenance, retirement, and user notification of computer software utilized by NETCO for applicatio n s that are s af e ty r e lat e d or important to saf e ty. Software i s controlled under NETCO Standard Operating Procedures (SOPs) in the Standard Operating Procedures Handbook. Specifically, SOP 2.4, Software Contro l , provides procedures for software acquisition, software de s i gn, E rror Notification , Configuration Control, User Documentation, Verification/Validation, Software Testing/Benchmarking and Run Log maintenan ce. T h ese features are subject to further procedural control as provided for in NETCO Software Control Procedures , SCP-001, Proc e dur e for C l assification of NETCO Softwar e Us e d for En g in ee ring Calculations, and SCP-002 , Proc e dur e for NETCO Comput e r Identification and Installed Softwar e In v e ntory. Additionally , SOP 2.10 , Control of Manual and Computerized Calculations , provides procedures for documentation of the accuracy, traceability and verifia bility of computerized calculations.

NET-28091-0003-01, Revision 0 9 Unless otherwise specified , all of the k err values reported in this document are raw calculated k eff values with no adjustment for bias and uncertainty. The final values to be compared to the criticality criteria are the calculated values plus the total bias and uncertainty (notated as " k 9s 1 9s"). NET-28091-0003-01 , Revision 0 10 3 Input Data For the criticality analysis , input data is needed for the SFP and storage racks (Section 3 .1 ), the fuel assemblies (Section 3.2), the fuel assembly inserts (Section 3.3) and the plant operating data (Section 3.4). 3. 1 SFP and Storage Rack Specification s The IP2 SFP is shown in Figure 3.1. It is lined with a 0.25 inch stainless steel plate covering the concrete walls. In this SFP are three Region 1 (flux t rap) modules and nine Region 2 modules. The southwest corner (bottom right in Figure 3.1) is an empty area for placement and l oading of storage and transport casks. The southeast corner ( top right in Figure 3 .1) contains two large cylinders for containing failed fuel and the new fuel elevator.

The placement of the modules is shown on Figure 3 .1. The Region 1 fuel racks (flux trap design) contain Boraflex TM in sheaths. The Boraflex Ž is not credited in this analysis.

Figure 3.2 shows the general arrangement of the cells in a Region 1 module [8]. Table 3.1 contains the dimensions and to l erances from the manufacturer

's draw i ng for Regio n I [8]. Region 2 is an egg-crate design where square storage tubes with Boraflex TM sheaths are joined at the corners via spacer rods creating "resultant" cells between the tubes. Figure 3.3 shows two complete cells in the Region 2 type rack [9]. The cell on th e left wou l d be called the " resultant" cell. Notice the fuel in the resultant cell is not bounded by four flat walls but rather by the Boraflex TM sheaths. The dimensions for the Region 2 rack are also shown in Table 3.1 [9]. Two rack positions will be filled with cell blockers. The cell blockers will be made of stainless steel and will not displace more than 50% of the water at the elevations containing fuel rods. NET-2809 1-0003-0l , Revision 0 II

'J' * *~ Ce:LL OAT (GIOU 11 CI Q.\C S. l.!f.91 Ct'. LS P1 CH-I&:* 1, :* ~lGICN 2: 1-,1 RAC- * :E E 0.950 , QI VI V, <! -c 0.8 50 QI

  • tlO IV ... 0.7 50 :> * <!
  • 0.650 -* 0.550 -----~-----~-----, 5000 15000 25000 35000 45000 55000 65000 Burnup (MWd/MTU)

Figure 5.1: Averaged Assembly Peaking Factors of Assemblies in the IP2 SFP NET-28091-0003-01, Re v ision 0 30 Westinghouse Non-Proprietary Class 3 5.1.3 Moderator Temperature The moderator temperature increases as the water rises through the core. For the top node, using the core exit temperature at the averaged assembly radial peakin g factor is conservative (the actual node temp era ture is somewhat l ess since the temperature become s the core exit temperature only at the top of the top node). The average core exit temperature (T co r eex it) is T coreex it= T;n + 2 x (T ave -T;n) = 2 x Tav e -T;n where Ta ve and T;n are the average and inlet temperatures for each cycle provided in Tables 3.5 and 3.6. The enthalpy at T;n and T coreex it can be obtained from steam tables at a pressure of 2235 psia (the minimum allowed pressure).

For an averaged assembly radial peakin g factor of PF , the delta entha lp y is: 6ha sse mbl y = PF X (h co r e ex it -h;n) The assemb l y exit enthalpy is then h 0 , + ti.ha sse mbl y. The temperature corresponding to this assemb l y exit enthalpy are obtained from steam tables and converted to Kelvin. The exit temperatures for various peaking factors using the T ave and T;n data from Tab l es 3.5 and 3.6 are shown in Table 5.1 , below. Table 5.1: Moderator Exit Temperature, T exit, versus Peaking Factor for Batch Groups Texit at 1.4 Tcxit at 1.2 Tcxit at 1.0 Texit at 0.8 Texit at 0.6 Batch Tave (°F) T;u (°F) (K) (K) (K) (K) (K) A , B ,C, D a , c 601.9 596.2 590.3 584.1 577.6 E thru L 589.9 584.1 578.1 571.9 565.6 M,N , P 598.8 592.5 585.9 579.1 572.0 0 , R , S 598.8 592.5 585.9 579.1 572.0 T , U , V 603.2 596.8 590.1 582.9 5 75.5 w,x 603.2 596.8 590.1 582.9 575.5 Z (after X) 603.5 597.7 591.5 585.0 578.3 All IP3 603.5 597.7 591.5 585.0 578.3 -The densities corresponding to these exit temperatures are calcu l ated from steam tables and are presented in Table 5.2: NET-2809 1-000 3-01 , Revision 0 31 Westinghou se Non-Propriet a r y Clas s 3 Table 5.2: Moderator Exit Density versus Peaking Factor for Batch Groups densit y at densit y at densit y at densit y at density at Batch Tav e (°F) Tin (°F) 1.4 fa/cc) 1.2 (e:/cc) 1.0 (e:/cc) 0.8 (e:/cc) 0.6 fa/c c) A, B ,C, D a , c 0.6552 0.67 1 6 0.6874 0.7024 0.7169 E thru L 0.6884 0.7024 0.7158 0.7288 0.7414 M , N,P 0.6645 0.6816 0.6980 0.71 37 0.7287 Q , R , S 0.6645 0.6816 0.698 0 0.7 1 37 0.7287 T , U , V 0.6510 0.6700 0.6879 0.7 050 0.72 14 w,x 0.6 510 0.6700 0.6879 0.7050 0.72 14 Z (after X) 0.6502 0.6676 0.6842 0.700 1 0.7154 All IP 3 -0.6502 0.6676 0.6842 0.700 1 0.7154 At the second node from the top , the mod erato r temp erature i s lo wer bec a u se of th e heat adde d b y th e top node. The second node enthalpy is whe re hin h co r e ex it P F NF h in + NF*PF*(h co re ex it -h in) inl et entha lp y exit ent h a lp y (core average) averaged assemb l y p ea king factor enthalpy nod e factor The e nth alpy nod e factor in the eq u a tion above d epends on h ow mu c h heat i s added by the top n ode a nd this depend s on th e r e l ative axia l po we r of the top node. T h e refor e, a n ax ial po wer sha p e i s need ed. Since this is a d ep letion calcul atio n , the ax i a l power s h ape over the li fe of the assem bl y i s the axia l bumu p profile. The DOE axi a l bumup profil es a re u sed for the axia l bu mup shape [27]. These profile s a re a function of the burnup and are di sc u ssed in Section 6. To illustrat e an exa mple calculation, the p rofile from th e 46+ GWd/T burnup bin i s u sed. T h e relative burnups at the top five nod es a r e 1.0 72 , 1.050 , 0.992 , 0.833, and 0.515 where 0.515 i s the top node , 0.833 is the seco nd node from th e top , etc. The e nth a lp y node fac tor for the top of the se cond nod e i s (1 8 -0.5 1 5)/1 8 = 0.971 si nce t h e r e are 1 8 node s and th e h eat a dded b y the top node i s 0.5 15/18 tim es the tota l h eat added. Similarly , the node factor for the bottom of the secon d node is (18 -0.833 -0.515)/1 8 = 0.925. The average node factor for the seco nd node i s the average of these two values or 0.94 8. NET-28 091-000 3-01 , R ev i sion 0 32 The average entha lp y node factor for the third node is (36 -0.992 -2x0.833 -2x0.5 l 5) / 36 = 0.898. The exit ent h a lpi e s ca n b e con v erted to temperature and density u si n g steam tables. Table 5.3 summarizes the node factors for the top five nodes as a function of the axial bumup shape (represe nt ed in the ta bl e as the bumup bin). Table 5.3: Entha lp y Node Factor versus Axial Burnup Shape Node 18-22 22-26 26-30 30-38 38-46 46+ Too 1.000 1.000 1.000 1.000 1.00 0 1.000 2nd 0.96 1 0.956 0.955 0.955 0.949 0.948 3r d 0.9 1 6 0.913 0.911 0.907 0.898 0.898 4111 0.86 1 0.860 0.858 0.850 0.841 0.841 5111 0.801 0.800 0.798 0.7 91 0.781 0.78 2 The exit enthalpy is co n servative l y used for the top node ent h a lp y so the ent h a lp y node factor is a l ways 1.00 and there is no bumup shape dependence. In genera l , the node factor increases with lo wer re l ative powers at the top b ecause more of the heat is b e in g adde d at the lower nodes. From t he a bo ve n ode factors and using a peaking factor of 1.4 0 , the moderator temperature (K) for B atch Gro upin g Z (future fuel) as a function of the bumup shape is shown in Table 5.4. Table 5.4: Moderator Temperature (K) at each Node versus Burnup Profile Node 18-22 22-26 26-30 30-38 38-46 46+ Too 603.5 603.5 603.5 603.5 603.5 603.5 2n d 60 1.9 601.7 60 1.7 601.7 601.4 601.4 3rd 600.1 600.0 599.9 599.7 599.3 599.3 4th 597.8 597.8 59 7.7 597.4 597.0 597.0 5th 595.3 595.2 595.1 594.8 594.4 594.5 NET-2809 1-000 3-01 , Re v ision 0 33 5.1.4 Fuel Temperature As with the moderator temperature , the fuel temperature in the top nodes is a function o f the axial profile at the top. The fuel temperature in the top node i s the temperature corresponding to a tota l peaking factor of PF x AF where PF is the averaged assembly radial peaking factor and AF is the axial peaking factor (for example , AF would be 0.515 using the 46+ bumup bin). Fuel temperatures were calcu l ated by INTERPIN-3

[28] (provided in Ref. [29]) for various total peaking factors and are shown on F i gure 5.2. The fue l temperature starts at a high va lu e and then decreases for a whi l e before starting to increase again. Since SCALE allow s the input of multiple temperatures for a material during depletion as a function ofbumup , six data s ets are input to SCALE (at 0 , 2 , 15 , 25, 40 , and 70 GWd!T). For a radia l PF of 1.4 , the relat i ve power to use for the top node is 1.4 x 0.515 = 0. 721. From Figure 5.2 , the top node (relati v e power of0.721) fuel temperatures a t 0, 2 , 15 , 25 , 40 , and 70 GWd!T are 837.5 , 812.9, 7 8 9.5, 781.6 , 786.8, and 8 1 5.1 K , respectively. At a PF of 1.2, the top node fue l temperatures are 799.2, 778.1 , 758.1 , 751.2, 752.3 , and 774.9 K. At a PF of 1.0, the top node fuel temperatures are 760.9, 743.4, 726.7, 720.8 , 717.8 , and 734.8 Kand at a PF of 0.8 , the top node fuel temperatures are 722.6 , 708.6, 695.2 , 690.4 , 683.3, and 694.7 K. NET-2809 1-0003-01 , Revision 0 34 1300 1200 1100 E ::, 1000 a. E i!!-.; ::, u.. 900 800 700 --INTERPIN-3 Average Fuel Temperature versus Burnup at Various Relative Powers Typical IP-2 Cycle 20 --___. ' .---, r-6-, 1,---"' -* ____. -I t&.,_~ th ,i.-t~ -p-8-j 3--S--~ --a :: , ia--i I}-I It-i:r-~. ; 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 Burnup (GWD/MTU) 05 -a-1 0 Figure 5.2: Fuel Temperature Change with Burn up and Relative Power The fuel temperature at the top node as a function of averaged assembly radia l peaking factor is linear and can be expres s ed as: Top Node Fuel Temperature at O GWd/T = 19l.5*PF + 569.4 Top Node Fuel Temperature at 2 GWd/T = l 73.8*PF + 569.5 Top Node Fuel Temperature at 15 GWd/T = 157.2*PF + 569.5 Top Node Fuel Temperature at 25 GWd/T = 152.0*PF + 56 8.8 Top Node Fuel Temperature at 40 GWd/T = l 72.5*PF + 545.3 Top Node Fuel Temperature at 70 GWd/T = 200.7*PF + 534.2 Figure 5.3 demonstrates that the top node fuel temperature at 25 GWd/T is indeed a linea r function of the averaged assembly peaking factor (PF). NET-2809 1-000 3-01, Revision 0 35 8 00 ~--------------------

-78 0 +------------------

-=----z 76 0 +--------------=c:::,'"""-----

'-' -Lin ear 7 40 +-----;: F:-:-: i t---------:;;,tl'IE:--------72 0 +-----------:

..li

a. [ 7 00 +-------~""'-------------68 0 E-, -66 0 +--:::a a=-

<1,1 64 0 +---------------------

62 0 +---------------------

6 00 +-----,-----,------..---...........,r------, 0.5 0.7 0.9 1.1 1.3 1.5 Top N ode P e aking Factor Figure 5.3: Fuel Temperature (K) at 25 GWd/T v s. Peaking Factor (PF) at the Top Node As w ith th e mo d e ra to r t e mp era tur e , th e fuel te mp e ratur e in th e 2 11 ct a n d l owe r nod es i s a fun c tion of th e axia l burnup profi l e. T a bl e 5.5 p rov id es th e fu e l t e mp erat ur es a t 25 GWd/T at a ra di a l P F of 1.40 u s in g t h e ran ge of D OE a x i a l bu rn up s h a p es (r e pr ese n t e d as t h e burnup b in). Table 5.5: Fuel Temperature (K) at each Node versus Burnup Profile Node 18-22 22-26 26-30 30-38 38-46 46+ Too 723.0 753.9 7 5 7.3 753.5 78 0.4 7 81.6 2" d 845.3 8 5 3.5 8 5 8.5 8 7 8.2 9 1 2.0 922.9 3rd 972.5 9 41.8 9 4 7.2 992.9 1011. 1002. 4th 1 0 30. 10 26. 10 29. 1 034. 1040. 103 1. 5th 1 058. 10 60. 10 63. 1 046. 1 04 7. 1042. No t e th a t fo r fue l t e mpe ratures , th e t e mp erat ur e d ec r e a ses w i t h lo wer re l at i ve p owe r s at th e t o p w h ich i s t h e o ppo s it e of w h a t happ e n s wit h th e mod era t o r t e mp erature. N ET-28 091-000 3-01 , R ev i s i o n 0 36

5.1.5 Selec

t ion of Bounding Model and Temperatures Historically , many criticality analyses used the moderator exit temperature and a single bounding high fuel temperature for all axial nodes. Early analyses non-conservatively ignored radial peaking and used the core exit temperature for the moderator temperature.

This non-conservatism is removed by using the averaged assembly radial peaking factor. Using a single set of depletion parameters for all axial nodes is a conservatism that is not needed. The top node is unique since:

  • IFBA absorbers are cut short and do not extend into the top node;
  • Control rods , if at the bite position , are present in the top node;
  • Enrichment is lower for axially blanketed fuel;
  • Fuel temperature is lower due to the sharp decrease in power at the end of the fuel. For these reasons, the top node is depleted with i ts own set of depletion parameters.

All of the nodes below the top node are depleted with a separate set o f temperatures which is conservative for a ll of the lower nodes. It is possib l e to use the temperatures in Tables 5.4 and 5.5 and perfonn the depletion analysis using separate depletion s for each axial node and burnup bin , but although this approach has been applied in the past [30], reducing the number of depletion parameters greatly simplifies the analysis.

To simp l ify the analysis , it is practica l to select bo un ding temperature conditions that cover all burnup bins. Since the fuel and moderator temperature do not change the same amount from burnup bin to burnup bin, a sensitivity study was performed for the fuel and moderator temperatures. It was determined t hat a change in moderator temperature of 10 K during depletion causes a change in reactivity of 0.0052 t.k while a change in fuel temperature of 100 K causes a change in reactivity of 0.0026 t.k. Using Tables 5.4 and 5.5 it can be seen that for the top node, the 46+ burnup bin conservatively r epresents both the moderator and fuel temperatures (the moderator temperature variation is independent of shape but the fuel temperature is highest for the 46+ burnup bin). For the second node , the 46+ burnup b in should also be used because the moderator temperature is a weak function of the shape but the fuel NET-2809 1-0003-01 , Revision 0 37 temperature is a strong function of the shape. For examp l e, using the 30-38 burnup bin , the moderator temperature at the second node is 0.3 K higher than the 46+ burnup bin but the fuel temperature is 44.7 K lower. The reactivity effect of the 0.3 K higher moderator temperature is only +0.0002 in keff while the reactivity effect of the 44.7 K lower fuel temperature is -0.0012, so using the 46+ burnup bin for the second node is conservative for all burnup bin s. Since the second node fuel temperature is l ess than the fuel temperature at the lower nodes, a test is required to determine if the third or second node is more limiting. For the third node , it would appear that using the 38-46 burnup bin is more limitin g (the fuel temperature is higher using the 38-46 bin compared to the 46+ bin and the moderator temperature is the same). However, using the 38-46 bin for the top node , the fuel temperature is 1.2 K lower , and for the second node, the fuel temperature is 10.9 K lower , while the 3rd node fuel temperature is 9 K higher. The net effect is that u sing the 46+ bin is conservative compared to the 38-46 bin when all nodes are considered. However , to check this , a specia l depletion using the 38-46 shape to obtain fuel and moderator temperatures in the top 5 nodes was perfonned and the result compared to u s ing the 46+ s h a pe for the top and 3'd node s and then using the 3rd nod e temperatures for the 2"d, 3'd, 4t\ and 5th node s (the method that is used for the final selected depletion analysis).

The res ults s ho w that using the 46+ shape to determine fuel and moderator temperatures in thi s manner is more conservative than using the 38-46 s h ape by 0.0006 ~k. This difference was obtained by using the b urnup shape at 38 GWd/T in the keffca l culation.

For additiona l confirmation, the reacti v ity effects using t hree other axial burnup profiles (at bumups of 22, 30, and 45 GWd/T) are 0.0003 ~k, 0.0 005 ~k, and 0.0005 ~k, respectively , with the 46+ temperatures a l ways being more conservative.

For the 4th node , using the 3rd node temperatures means that the moderator temperature being used i s 2.3 K higher than the 4 th node, while the fuel temperature is 29 K l ower than the 4th node. The reactivity effect of the 2.3 K higher moderator temperature is +0.0012 ~k while the reactivity effect of the 29 K lo wer fuel temperature is -0.0008 ~k. So using the 3'd node temperatures is conservative for the 4th node. NET-2809 1-00 03-01, Revision 0 38 The sa m e reasoning applies for th e 5th and l ower nod es. T h e rationale for this is that the moderator temperature is decreasing faster than the fuel temperature is increasing (in terms of net reactivity), so u sing th e J<d nod e temperatures is co ns e r vative for th e 4th and a ll low er nodes. For convenience, the moderator temperature at the top node can be conservatively fit with a straight l ine as illustrated in Figure 5.4 for Batch Grouping Z (future fuel). T h e straight line values shown on Figure 5.4 are always the same or conservative with res pect to the points. 610 605 z 600 +----------------:

.-,,:.=-----.._, -Linear Fit 595 =

  • Points 590 +----------:

..,.=

I. c. 585 8 E-580 +---:~"'-----------------

575 570 +----~---~---~---~---~

0.5 0.7 0.9 1.1 1.3 1.5 Peakin g Fac tor Figure 5.4: Top Node Moderator Temp (K) vs. Average Assembly Peaking Factor Figu r e 5.5 shows th e top node moderator density as a functio n of the averaged assembly peaking factor. NET-28091-0003-01, R ev i s ion 0 39 0.73 ---0.72 CJ CJ oli 0.7 1 '-" .f' 0.7 "' C 0.69 ... 0.68 0 .... 0.67 c,: ... -Linear Fit "O 0.66 0

  • Point s 0.65 "O 0.64 0 z 0.63 Q. 0 0.5 0.7 0.9 1.1 1.3 1.5 f,-, Peaking Factor Figure 5.5: Top Node Moderator Density vs. Avg Assembly Peaking Factor Tab l e 5.6 s ummari zes th e lin ea r fits for th e moderator temperature and density for each batch grouping (top node) u s ing the equations:

Exit Temp era ture (K) =C l+ C2 x PF Ex it den sity (glee)= C3 + C4 x P F Table 5.6: Fit Coefficients for Top Node Moderator Temperature and Density Batch Cl C2 C3 C4 A,B,C,D 560.8 29.5 0.7 632 -0.0 77 1 E t hru L 548.1 30.0 0.78 1 2 -0.0 663 M , N , P 552.9 33.0 0.7769 -0.0803 O , R , S 552.9 33.0 0.7769 -0.0803 T,U , V 556.6 33.5 0.77 42 -0.0880 w , x 556.6 33.5 0.7742 -0.088 0 Z (after X) 560.5 31.0 0.76 43 -0.0 8 15 A ll IP3 560.5 31.0 0.7643 -0.0 8 1 5 Table 5.7 s ummari zes the lin ea r fit s for the 3rd node from the top mod erato r temperature a nd d e n sity. Table 5.7: Fit Coefficients for 3rd Node Moderator Temperature and Density (Use d for all nodes except the top nod e) Batch Cl C2 C3 C4 A , B , C,D 559.6 27.5 0.7619 -0.06 78 NET-28091-0003-01 , R evision 0 40 E thru L 547.5 27.5 0.7803 -0.0585 M , N , P 552.5 30.0 0.7755 -0.0705 Q , R , S 552.5 30.0 0.7755 -0.0705 T , U,V 555.5 31.0 0.7726 -0.0771 W , X 555.5 31.0 0.7726 -0.0771 Z (after X) 559.7 28.5 0.7628 -0.0715 A ll IP3 559.7 28.5 0.7628 -0.0715 The fuel temperatures at the 3'd from the top node (used for all nodes except the top node) are: 3rd Node Fuel Temperature at O GW d/T = 351.0*PF + 587 .2 3'd Node Fuel Temperature at 2 GWd/T = 319.5*PF + 584.8 3'd Node Fuel Temperature at 15 GWd/T = 291.5*PF + 580.8 3rd Node Fuel Temperature at 25 GWd/T = 33 l.2*PF + 538.2 3'd Node Fuel Temperature at 40 GWd/T = 371.0*PF + 514.4 3'd Node Fuel Temperature at 70 GWd/T = 437.8*PF + 493.0 For modeling simplicity , the 3rd node temperatures are used for the 2nd and l ower nodes because this is conservative.

For axial blankets, the burnup profile at the top will have relative burnups that are smaller than the DOE profile for full-length fuel. As a result of the smaller relative burnups , the moderator temperature increases slightly but the fuel temperature decrease is more significant.

For the same reasons as discussed above regarding using a lo wer burnup bin, using the DOE profile for the last burnup bin is conservative for axial blanket fuel. The use of a burnup averaged assembly radial peaking factor assumes that the impact of the temperatures is independent of the power as a function of burnup. In fuel core loading designs , assemblies are depleted with a peaking factor greater than 1.0 during its first cycle (fresh asse mbl y with burnab l e absorbers).

After the burnab l e absorber is removed , the assembly is moved and the peaking factor during the second cycle is genera ll y less than 1.0. To show that depletion using the average pea king factor over the life of the assemb l y is appropriate , a specia l depletion was performed in which the first 25 GW d/T was depleted at a peaking factor of 1.20 and the second 25 GW d/T was depleted at a peaking factor of0.80. The ke ff for this case at 50 GWd/T (5.0 w/o fuel) is 0.9371. The k e ffusing a peaking factor of 1.00 throughout the depletion is 0.9377. This demonstrates that using the average NET-28091-0003-01, Revision 0 41 peaking factor for the assembly is ap propri ate and s li g htl y conservat i ve. As further confirmat ion , the depletion was repeated u si n g peaking factors of 1.40 an d 0.60. T h e k e ff for this case is 0.9361, which is even s m a ller. 5.2 Limiting Depletion Parameters

-Burnable Absorbers Burnable absorbers h arden the spectrum during depletion , which result in more plutonium production and l ess U-235 cons umption for a give n bumup [31 , 32]. The spect rnm hardening comes from th e absorptio n of thermal n e utron s by the a b sor b er and di sp l acement of th e water in the guide tubes. The effect on reactivity a l so depends on how l ong (in terms of GW dff) the burnable absorber remains in the fuel before b eing removed. Therefore, th e burnab l e a b sorbers that m aximize the 1 0 B l oading and water di s placement should be used. For each batch grouping, the most limiting actual burnable absorbers are used. IP2 and IP 3 have u sed thr ee types of burn ab l e a b sorbers: Pyrex , Wet Ann ul ar Burnable A b sorbers (WABA) and Integral Fue l Burnable Absorbers (IFBA). Table 3.4 provides the dimensions and material details of Pyr ex and W ABA inserts. IFBA rods are di scussed in Section 3.2. The Pyrex and W ABA de signs consist of rodlets m ounted to a base plate which sits on the top of the fuel assembly.

The number ofrod l ets varies by position in the core to h elp contro l power peaking. The most limiting design h as 20 rodlets. Table 5.8 provides the wors t case burnable absorbers for each batch grouping. It s hould be noted that the IFBA a nd th e poison part of the W ABA does not extend to the top of the active fuel. The IFBA starts at least 6 in c h es from the top (8 inches with 8 inch ax i a l blankets) and the poison part of the W ABA starts at l east 5 inches fro m the top (6 in ches with 6 inch axial blankets a nd 8 inches with 8 inch axia l bl ankets). For all b atc h gro upin gs excep t M -P with IFBA/W ABA , th e top node can be depleted wit h no IFBA an d a WABA that h as no boron. For M -P , the 8 inch top node is NET-28 091-000 3-01 , Re v i sion 0 42 depl e t e d w ith IFBA that h as one quart e r 1 0 B (2 inche s of the 8 inches h as IFBA) a nd a W A BA with one h a l f the 1 0 B ( conservative l y mode l s th e W A B A poison i s 4 inches from the top). Table 5.8: Burnable Absorbers versus Batch Grouping BA Max BA Max B A Batch Tvoe Loadin2 Burnuo A , B , C , D P y r ex 20 ro dl e t s 18.5 E thru F P y rex 1 2 ro dl ets 12.2 G thru L Pyrex 20 ro dl ets 16.7 M,N , P IFBA 11 6 (I .OX) 28.1 WABA 20 rodlets Q , R , S IF BA 14 8 (1.5X) 26.7 WABA 20 rod l e t s T , U , V IFBA 14 8 (1.5X) 33.8 WABA 20 ro dl e t s W , X IFBA 14 8 (1.25X) 32.6 WABA 20 rod l ets Z (after X, and IFBA 148 (1.25X) 33.2 IP3 afte r U) WABA 20 ro dlet s IP3 (A-U) Pyrex 20 rod l ets 19.4 5.3 Limiting Depletion Parameters

-Soluble Boron Soluble boron h ar d e n s th e neutron s p ectrum, m a kin g the fue l mor e reactive for a g i ve n burnup. It has been s h ow n that performi n g d ep l e tion calculations at th e burnup averaged so luble boron concentrat i on i s acceptab l e (rather than u s in g a time-dependent so lubl e boron l et down curve) [57]. Tables 3.5 a nd 3.6 s ho w the cycl e average so lubl e boro n concentration for eac h cyc l e. Since nearly eve r y asse mbl y i s burn ed at l east two cy cl es, th e solu bl e boron to u se for the depletion ana l ysis is the multi-cycle burnup average d so lubl e boron for each asse mbl y. This multi-cycl e burnup averaged so lubl e boron i s calcu l ate d u s in g the assembly cycle burnup s to we i g ht the cyc l e average ppm. Table 5.9 s ummari zes t h e so lubl e boron used in the depletion analysis for eac h b atch grouping.

Table 5.9: Soluble Boron versus Batch Grouping Boron Used Batch in Depletion A , B ,C, D 570 E and F 580 NET-2809 1-000 3-01 , R evision 0 43 G thru L 660 M , N,P 820 Q,R , S 850 T,U , V 880 W , X 880 Z (after X and IP3 after U) 950 IP3 (A-U) 560 5.4 Limiting Depletion Parameters

-Specific Power ORNL performed a study of the sensitivity of bumup credit to specific power and determined it is a small effect [33]. For b u rnup credit using all isotopes, a lower specific power is slightly more reactive. However, the reactivity effect of moderator temperature and fuel temperature increases with higher specific power. The reactivity effect of higher temperatures is larger than the reactivity effect of a lower specific power. Since the fuel can operate at only one specific power, the specific power used during depletion is determined to match the relative power used for the other depletion parameters.

This approac h is consistent with DSS-ISG-2010-01 for SFP analysis [5]. The average specific power is the core power divided by the total initial heavy metal mass. Using the stack density (see Section 3.2) of 0.95 multiplied by t he U0 2 theoretical density , the initial mass of Uranium metal is then 89.66 metric tons for the Indian Point cores. The specific power (Watts/g U) is then where SP= Power x PF x RAB/ 89.66 Power= Total core power (MW) (from Tables 3.5 and 3.6) PF= Averaged assembly radial peaking factor RAB = Relative Axial Bumup from the DOE shapes For the top node (RAB= 0.515) the specific power ranges from 15.8 to 18.5 W/g U multiplied by the peaking factor for all of the batch groupings ( due to changes in the total core power). Since the specific power has a small effect , the specific power of 16 W/gU multiplied by the averaged assembly radial peaking factor is used for the top node for all fuel batch groupings. The balance of the axial nodes use temperatures developed using the 3rc1 node relative burnup (RAB= 0.992) from the DOE axial burnup profiles. The specific power for the nodes below the top node ranges from 25.6 to 35.6 multiplied by the NET-28091-0003-01, Revision 0 44 peaking factor. Again, a simple single specific power value of26 W/gU multip l ied by the peaking factor is used for all fuel batch groupings.

5.5 Limiting

Depletion Parameters

-Control Rod Operation IP2 and IP3 have 193 fuel assemblies in the core. Nine of these assembly locations are under the Control Bank D (less than 5% o f the number of assemblies). Control Bank Dis the only control bank that may be inserted during power operation if the power is greater than 70% of the rated power. IP2 and IP3 have not operated with Control Bank D in the core for any significant bumup except at the "bite" position.

The bite position is set as the location where the worth of the lead control bank is 2 pcm per step. The bite position changes from cycle to cycle and during cyc l e operation but is typica ll y between 207 to 217 steps withdrawn , which corresponds to the rod being inserted 8. 7 or less inches into the core. Operation with Control Bank D at the bite position occurred in IP2 durin g the first 17 c y cles. Cycle s 1 8 and beyond for IP2 and a ll cycles for IP3 operated with all of the control rods fu ll y withdrawn from the core. Table 5.10 s hows the fuel a s sembly ID s that were under control bank D for IP2 for the cycle s containing Batches A through X Tab l e 5.10: Asse mbli es under D-B ank for t h e First 21 Cycles of I P 2 C o re Locat i o n Feed o fD-B6 BlO F2 F14 H S K2 K14 P6 P lO Batc h B a nk Cy cl e Asse mb h I D 1 A17 A09 A30 A39 A44 A34 AlO A26 A33 A , B,C 2 BOl B 23 B64 B57 B06 B54 B60 B18 B27 D 3 C45 C51 C55 C59 B53 C54 C60 C42 C40 E NET-28091-0003-01 , Revision 0 45 4 D47 D69 D09 Dl0 D25 D49 D72 D32 D46 F 5 E57 E32 E53 E06 D71 El3 EOl E27 E25 G 6 F06 Fl4 F46 F03 F45 F23 F60 Fl6 F05 H 7 G49 G50 G68 G67 G70 G65 G69 G52 G61 1 8 Jl8 132 122 107 H27 167 130 133 129 K 9 Kl 8 Kl9 K09 Kl2 165 K29 K27 K20 Kl7 L 10 L61 L47 L54 L60 K25 L41 L55 L34 L57 M 11 M54 M63 M69 M60 L09 M65 M62 M53 M46 N 12 P06 P09 P08 P03 MOS P 11 PIO P05 P04 p 1 3 Q25 Ql7 Q22 Q21 N24 Ql4 Q27 Q26 Ql5 Q 14 R 82 R76 R79 R 78 R 7 1 R 75 R77 R72 R 15 S35 S40 S33 S39 S38 S37 S36 S34 s 16 T67 T69 T68 T63 S77 T7 1 T 70 T65 T62 T 17 U65 U73 U58 U56 U61 U60 U59 U57 u 1 8 V32 V51 V46 V52 V44 V30 V49 V31 V 19 W89 W75 W93 W82 W21 W77 W76 W71 W64 w 20 X24 X25 X21 Xl3 W21 X42 X34 X35 X23 X 21 2A64 2A84 2A58 2A55 X03 2A66 2A38 2A70 2A29 2A The reference depletion analyses for a ll bu t the future cycles (Batch Groupi n g Z) are modeled with no control rods inserted. For the asse mbli es li sted in Table 5.10 that are s h aded blue or ye ll ow, the burnup r equireme nt for storage i s increased b y a n appropriate burnup penalty. The asse mbli es marked w ith green shading on Table 5.10 did not require a burnup penalty s inc e they did not contai n W ABAs a nd the r eactivity effect of WABAs in cluded in the standar d d e pl etion a nal ys i s is l arger than the reactivity effect of the sho rt tim e that Co ntrol Bank D may h ave been in these asse mblie s. Note that af t er Cycle 17 , Control Bank D was maintained in a n all-out position (no bite). T h e asse mbli es in Table 5.10 sha d e d in pink a nd yellow did not contain burnable absorber in serts durin g ac tual core ope ration s but did h ave the Co ntrol B ank D rods in th e bit e position. A n a l ys i s s ho wed th at assem bli es depleted wit h 20 rodlet P yrex burnable a b sorbers conservatively bound s asse mblie s that were operated with Co ntrol B ank D at the bit e po s ition. No burnup p e nalt y is needed for the pink assem blie s since they are mod e l ed wit h 20 rodlet Pyrex burn ab l e absorbers. The yellow asse mblie s are d e pl ete d with a 20 rodlet WAB A. S ince W ABA's do not h arde n th e s pe c trum as much as Pyr ex and since W ABA absorber materi a l i s not in the top nod e, a s m all burnup correction is requir ed. For the se NET-28091-0003-01, Revision 0 46 yellow assemblies , the burnup requirements are increased by 0.5 GWd/T. Tab l e 5.11 shows the cases analyzed to confirm the bumup requirement increase (or shortened to "penalty")

for assembl i es where the D-bank was at the bite position and the assembly did not contain a burnable absorber.

Note that the L'.k is converted to a L'.GWdfT by use of the sensitivity of ketf to burnup that is discussed in Section 7. Table 5.11: Effect of Modeling the Bite Position rather than Burnable Absorbers Batch Group, Fuel Burnable Calculated Calculated kcff A kcrr with D-with burnable Ak Burnup Enrichment and Burnup Absorber bank Bite absorber (GWd/T) G-L, 3.0 w/o, 24.23 GWd/T 20 Pyrex 0.9620 0.9684 -0.0064 -0.90 M-P, 4.2 w/o, 39.28 GWd/T 20WABA 0.9570 0.9554 0.0016 0.30 M-P, 4.6 w/o, 41.31 GWd/T 20WABA 0.9616 0.9608 0.0008 0.20 T -V, 5.0 w/o, 41.94 GWd/T 20 WABA 0.9597 0.9590 0.0007 0.11 The assemblies in the blue shaded portion of Table 5.10 are depleted during the first cycle with a b urnable absorber and then contained a control rod in serted to the bite position.

For these assemb li es the required bumup is increased by 2 GWd/T. The 2 GWdfT penalty is determined by running a separate set of depletions with D-bank at the bite position. The bite depletion is performed as follows: 1. For the top node , the depletion analysis is perfonned with the control rod fully inserted for the entire depletion (this is conservative since this fuel was under D-bank for only one cycle). 2. For the lower nodes , the burnable absorber is in the fuel until the maximum burnable absorber burnup is reached and then the control rod is placed in the guide tube for 2 GWd/T and then removed. Contro l rod bumup of 2 GWd/T is considerably more bumup than what has been experienced at IP2. Calcu lati ons of k e ff using the standard depletions and the bite depletions are performed using the Region 2 three-out-of-four model (see Section 6). Table 5.12 shows the results of this analysis.

The L'.k values shown in the fourth column are converted to delta burnups u sing the bumup measurement NET-2809 1-000 3-01, Revision 0 47 uncertainty calculations given in Section 7. The maximum delta burnup on Ta bl e 5.12 i s ro und ed up to 2 GWd/T which is used for the assem bli es in the blue s h aded ce ll s on Table 5.10. Table 5.12: Burnup Penalty for Assembles with Burnable Absorbers followed b y Bite D-bank Batch Group, Fuel Enrichment, and Burnup. k-dbnk k-standard Ak ABU GWd/T A-D , 3.0 w/o, 22.77 GWd/T 0.9721 0.968 7 0.0034 0.68 G -L, 3.0 w/o, 24.23 GWd/T 0.9735 0.9684 0.0051 1.0 2 M-P, 4.2 w/o, 39.28 GWd/T 0.9653 0.9552 0.0101 1.87 M -P , 4.6 w/o, 41.31 GWd/T 0.9708 0.9608 0.0100 1.83 T -V, 4.2 w/o, 36.45 GWd/T 0.9703 0.9621 0.0082 1.64 T -V, 5.0 w/o, 41.94 GWd/T 0.9677 0.9590 0.0087 1.74 The assembly shaded orange (X 03 *) represents a separate group. The orange group is for assemblies that contai n ed a burnable absorber insert for their first cyc l e and the n were subseq u ently place d und er D-bank in a later cycle but wi th out " bite" operation. These assem bli es may ha ve been burned a s h ort time under D-b ank in cyc l es without bite o p eration but are mode l e d as h aving no D-bank opera tion. To cover some D-bank operat ion , 1 GW d/T is added to the burnup requirement of these assemblies.

The analysis to s upp ort a 1 GWd/T pena lt y modified the s t a nd ard depletion analysis by adding 1 and 2 GWd/T ofburnup wit h co ntrol rods to the top node an d the lower n odes , respective l y, after the burnable a b sorber is r emoved. Ta bl e 5.13 shows the cases run to confirm this penalty. The result s on Table 5.13 are rounde d up to arrive at the 1 GWd/T penalty to th e burnu p requirement

s. Table 5.13: Assemblies with B A Inserts plus under D-Bank in Non-Bite Cycles Batch Group , Fuel Enrichment, and Burn up k-dbnk k-standard Ak ABU GWd/T W, 4.6 w/o , 40.07, GWd/T 0.9629 0.9598 0.0031 0.51 X, 4.6 w/o, 40.07 GWd/T 0.9627 0.9597 0.0030 0.49 IP3 (A-U), 3.4 w/o, 29.77 GWd/T 0.9698 0.9666 0.0032 0.5 4
  • X04 was pl aced unde r a D-Bank location in Cyc l e 22 a nd i s also in c lud ed in this group. N ET-2809 1-000 3-01, R ev i sion 0 48 There are two assemblies that are shaded red on Table 5.10 because they are unique. Assembly R08 was located under D-bank at the bite position for two cycles. This assembly did not contain W ABAs in i t. If R08 had been under D-bank for only one cycle t hen it would also be a " yellow" assembly but it was operated under D-bank for a 2" ct cycle. This assembly has already been casked, and its bumup is 6. 7 GW d/T above the requirements for its assigned category (Category 4 fuel) if it were returned to the SFP. It is concluded that Category 4 is the proper assignment for R08. Assembly U41 is like assembly R08 except that the second cycle of operation for assembly U41 did not have the D-bank at the bite position since bite operation ended with Cycle 17. Thus this assembly would need the yellow bumup penalty of 0.5 GWd/T plus some additional margin to cover some operational use of D-bank in Cycle 18. U41 is categorized as Category 5 fuel and exceeds the Category 5 minimum by over 6 GW d/T so the categorization of U4 l is appropriate. For future cycles (Batch Z) (white cells in Table 5.10) it is not known which assemblies will be under D-bank. To cover power operation with some control rods inserted, the top node for D-bank assemblies is depleted for 1 GWd/T with a control rod and lower nodes are depleted for 2 GWd/T with a control rod. The rest of the depletion is with a 20 finger W ABA w h ich is never removed plus 148 IFBA ( l .5X). This is the same method that was used in the previous CSA (1]. This approach eliminates the need to check fu ture assemblies for rodded operation under D-bank or when the W ABA was pulled. IP3 fuel is covered by three depletion batch groupings:
1. Batch Grouping IP3 (A-U) which covers Batches A through U 2. Batch Grouping IP3 (V-X) which covers Batches V , W, and X , and 3. The Z Batch Grouping which covers IP2 and IP3 assemblies beyond X. NET-28091-0003-01 , Revision 0 49 T h e IP3 (A-U) b a t c h gro u p in g i s d e pl e t ed w ith a 2 0 ro dl et P yrex bu rna bl e a b so rb er w hich i s r e m ove d a t 2 0 G Wd/f. No co ntrol rod s a r e in th e b ase case d e pl e ti o n. T a bl e 5.1 4 s h ows th e a sse mbli e s in IP 3 w h ic h we r e und er D-b a nk (asse mbli es in B a t c h es A throu g h V we r e not in t h e IP 3 co r e afte r Cy cl e 11). T h e c ol o r c o din g o n Ta bl e 5.1 4 is t h e sa m e as p r ev i o u s l y d i sc u sse d , so th ere i s no burn up pe n a lt y for th e green s had e d assemb li es , a nd t h e ora n ge s h a d e d asse mbli es h ave a burnu p req ui re m e n t that h as b een in c r ease d b y 1 GWd/T. Table 5.14: Assemblies under D-Bank for the first 11 Cycles ofIP3 Core Feed Location B6 BIO F2 FI4 HS K2 KI4 P6 PIO Batch ofD-Bank Cvcle Assemb l y ID I A24 A21 AOI A23 A30 Al8 Al7 A06 A02 A , B , C 2 B14 BOS B19 B25 A24 B03 B41 B37 B56 p 3 C45 C64 cso C58 A24 C54 C44 C06 C53 R 4 R32 R53 R71 R25 P l 1 R06 R43 R47 Rl7 s 5 S04 S3 5 S2 1 S28 R03 S09 Sl3 S l4 S 11 T 6 T74 T41 T70 T46 TSO T61 T76 T64 T65 u 7 U46 U75 U53 U56 TSO U73 U7 1 USO U74 V 8 V38 V37 V40 V35 T53 V34 V36 V33 V39 w 9 W39 W30 WI2 W06 V42 W I S W03 W20 W37 X 1 0 YI7 Y39 Y28 Y54 V44 Y38 Y40 Y41 Y23 y 11 AA28 AA22 AAS! AA23 U02 AA30 AA24 AA31 AA19 AA N ET-2 8 091-0003-01 , R ev ision 0 5 0 Westinghou s e Non-Proprietary Class 3 5.6 Depletion Analysis Model The dep l etion ana l ysis uses the SCALE 6.1.2 , t5-depl sequence of TRITON with the ENDF/B-VII.O 238 group cross section l ibrary. The model is a simple 2D 15xl5 array of pins centered in an 8.466 x 8.466 inch (assembly pitch in core) square of water (11). The 15x15 array contains 20 guide tube s and a voided instrumentation tube. In the core l ocations for the incore flux monitoring system , there is a gas-filled tube in which the detectors travel. Rather than model this tube in detail and/or separate o u t the assemb l ies that were in instrument l ocations, the instrument tube of all assemblies are modeled with void inside the tube inner diameter.

The 20 guide tubes contain the burnable absorber inserts or control rods as needed for the batch grouping.

Some of the fuel rods have ZrB 2 coatings (IFBA rods). This is modeled as a ring next to the pellet outer diameter of 0.001 cm thick of ZrB 2 meeting the 1 0 B linear density ([ ]a ,c mg 1 0 B/i nch except Batch Grouping M-P , which uses [ ]a ,c mg 1 0 B/inch) (20). All of the fuel rods are a single material.

Therefore, there is no variation in atom densities across the assembly with depleted fuel. Further , the resonance se l f-shielding (performed through lattice cell cards in SCALE) is the same for all pins including IFBA pins. A study performed to answer a question from the NRC on the previous CSA showed that ignoring the IFBA in the lattice cell calcu l ation was not significant (2). The depletion problems use 4000 neutrons per generation and 1000 generations.

This n umber of neutrons per generation and number of generations was shown to be adequate by a c onvergence study detailed in Chapter 4 of reference (2 1]. Small depletion time steps are needed to accurate l y account for the spectral changes due to Xe and Sm and the initial build in of Pu and other fission products. The initia l time steps (MWd/T) are 150 , 350 , 500 , 500, and 500 , followed by steps of 2000 MWd/T until the maximum bumup is achieved. For the Z fuel batch, the depletion contains control rods for the first 2 GWd/T (1 GWd/T in the top node) and then the problem is restarted u sing the StdCmpMixOOOxx fi l e where xx i s the material n u mber. The fuel temperatures use the "timetable" block to input the bumup dependent temperatures shown below Figure 5.2 and Table 5.7. Due to the restart , the bumups in the t i metable block are adjusted for the bumup NET-28091-0003-01, Revision 0 51 before the restart. For the other fuel groupings, a restart i s also used , but to remove the burnable absorber not the control rod. The restart burnup depends on the maximum burnup for any assembly with a burnable absorber insert. This burnup is conservatively rounded up to the next 1 GWd/T. The depletion block uses the default for the fuel mixture but constant flux option for the burnable absorber materials. Using the above model inputs , U02 fuel is depleted at fuel enrichments of2.0, 2.2, 2.6, 3.0, 3.4, 3.8, 4.2, 4.6, and 5.0 w/o U-235 at peaking factors of 0.6 0 , 0.80 , 1.00, 1.20, and 1.40. Enrichments for axial bl ankets are 2.6, 3.2, 3.4, 3.6, and 4.0 w/o with the same range of peaking factors. The burnup points at which the isotopic data is collected are 0.15 , 0.50, 1.0 , 1.5 , 2.0, and then every 2.0 GW d/T after that. Although the depletion is carried out with a full set of 388 nuclides, the nuclides used in the SFP model are a reduced set (185 nuclides found on Table 2.1). In order to confirm the TRITON modeling is adequate, comparisons were made with CASM0-5 benchmarks.

The change in k err as a function of burnup derived from using CASM0-5 and SCALE/TRITON depletion is provided for 198 cases. The CASMO t.k values are published in Reference

[20] for specific benchmark conditions

.* T h ese include cases with W ABA and IFBA. Reference

[20] is currently under review by the NRC, but the only values used here are the pure CASMO result s (not the bias and uncertainty that is under review). Tables 5.15, 5.16, and 5.17 provide the difference in the t.k as a function of depletion between CASM0-5 and SCALE/TRITON. Notice that a negative value indicates that SCALE/TRITON conservatively under predicts the reactivity of depletion predicted by CASM0-5. Further, note that the maximum deviation is less than 0.0030. Tables 5.18 , 5.19, and 5.20, show the percent difference in the

  • The CASM0-5 ilk of depletion is not given directly in Reference

[20]. Tables C-3 to C-5 of Reference

[20] provide the CASM0-5 ilk of depletion plus the CASMO 5 bia s given in Table 10-1 of Reference

[20]. For thi s application, the CASM0-5 bias is subtracted from Tables C-3 to C-5 to yield the CASM0-5 ilk of depletion.

NET-28091-0003-01, Revision 0 52 i:1k of de pl etion. Th e max i mum perce n t d iffer e nce i s 1.37%. At t h ese sm a ll d i ffere n ce s, it i s un clear w h ich i:1k o f d ep l etio n is c o rrect so th e u t il izatio n of th e 5% unce rt a in ty a l lowe d b y DSS-ISG-2 010-0 1 i s appropria t e. In add i tion , R efe ren c e [21] ha s s hown ex ce ll e nt a gre e m e nt between T RI T ON/NE WT , whic h w as u se d for t h e anal ys i s o f c h emi ca l a ss ay s and th e TRITON/KENO appro ac h u s ed i n thi s anal ys i s. Table 5.15: SCA LE/TRI T O N minus CA SM0-5 ~k of Depl e tion at 100 Hours Coolin g Ca se de s cription Ca se 10 2 0 30 40 50 60 3.2 5% e nrichm e nt d e pl e ti o n 1 -0.000 7 -0.0 0 1 2 -0.0015 -0.0025 -0.00 22 -0.00 20 5.00% e nr ic h ment d e p l etio n 2 0.0 001 0.000 2 0.0003 0.0000 0.0004 0.0006 4.2 5% enr i c h m e nt d e p l e tion 3 0.0004 0.0000 -0.000 2 -0.0005 -0.0008 -0.0005 off-no min a l pin d e pl e tion 4 -0.000 6 -0.001 2 -0.0015 -0.001 6 -0.00 1 9 -0.00 22 20 W ABA d ep l e ti o n 5 0.0002 0.000 7 0.0 003 -0.0002 -0.000 I -0.0002 I 04 IF B A dep l et i o n 6 0.001 2 0.0011 0.0004 -0.000 7 -0.0009 -0.001 8 1 04 IFBA , 20 WABA dep l et i on 7 0.0009 0.001 6 0.0 0 08 0.0000 -0.0001 -0.000 8 hi g h boron d e p l e t io n= 1500 ppm 8 0.0 004 -0.0001 -0.000 3 -0.0004 -0.0 001 -0.0001 branch t o h ot ra c k= 3 38.7 K 9 -0.0002 -0.0003 0.0000 -0.0005 -0.000 I -0.0001 b ra n c h t o ra c k bor o n = 1500 ppm 1 0 -0.0008 -0.00 15 -0.001 9 -0.0 0 2 3 -0.0 02 6 -0.00 2 5 hi g h power d e n s it y d e pl e tion 11 0.0 000 -0.000 7 -0.000 7 -0.0009 -0.0 008 -0.000 8 Table 5.16: S CALE/TRITON minu s CA S M0-5 ~k of Depletion at 5 Y ears Coolin g C ase description Ca se 10 20 30 40 50 60 3.2 5% enr i c hm ent d e pl e t i o n I 0.0000 -0.0005 -0.0009 -0.00 1 4 -0.0008 -0.0004 5.00% e nric h m e nt d e pl e t i o n 2 0.0 00 8 0.0 0 11 0.000 8 0.000 8 0.00 10 0.001 3 4.2 5% e nri c hm e nt d e p l e t i on 3 0.0010 0.0005 0.0003 0.0001 0.0002 0.0003 o ff-nominal pin d e pl e tion 4 -0.0 0 0 1 -0.00 03 -0.00 I 0 -0.0011 -0.0 0 I 0 -0.000 9 20 W ABA dep l etion 5 0.000 8 0.0011 0.0009 0.000 8 0.0008 0.0010 104 IFBA d e pl e tion 6 0.00 2 0 0.0014 0.0009 0.000 1 0.000 1 -0.0005 I 04 IFBA , 20 W ABA dep l e t ion 7 0.0020 0.00 2 4 0.0015 0.00 1 2 0.0006 0.0004 h ig h boro n d e p letion = 1 500 p pm 8 0.000 7 0.0008 0.000 7 0.0003 0.0007 0.0010 branch to h ot rack= 33 8.7 K 9 0.0004 0.000 6 0.0003 0.0004 0.0009 0.0010 b ranc h t o rack b oron= 1 500 p pm 10 0.0002 -0.000 7 -0.00 1 0 -0.00 1 7 -0.0015 -0.00 1 4 h i g h po w er d e n s it y d e p l e tion 11 0.00 0 6 0.000 4 0.0003 0.0004 0.0 0 0 2 0.000 8 NET-28 0 9 1-00 03-01 , R evis i on 0 53 Table 5.17: SCALE/TRITON minus CASM0-5 8.k of Depletion at 15 Years Cooling Case description Case 10 20 30 40 50 3.25% e nri c hm ent dep l et i o n 1 0.0007 -0.0004 -0.00 1 4 -0.00 1 6 -0.00 1 5 5.00% e nrichm ent depletion 2 0.00 11 0.0014 0.0007 0.0008 0.0008 4.2 5% e nri chment dep l et i on 3 0.00 14 0.0010 0.0002 0.0000 0.0000 off-no min a l pin depletion 4 0.0006 -0.000 4 -0.0006 -0.00 14 -0.00 1 3 20 W ABA d e pl etion 5 0.00 14 0.0018 0.0010 0.0005 0.0006 104 IFBA depletion 6 0.0025 0.00 19 0.0010 0.0004 -0.0002 104 IFBA, 2 0 W ABA d e pl et i o n 7 0.0027 0.0028 0.001 7 0.0010 0.0008 h ig h b oro n d e pl etion= 1500 pp m 8 0.00 11 0.0009 0.0006 0.000 1 0.0006 br anc h t o hot rack= 338.7K 9 0.0005 0.0004 0.0001 0.0003 0.0006 branch to rack boron = 1 500 ppm 10 0.0004 -0.0007 -0.0013 -0.00 1 6 -0.00 1 9 hi gh power density depletion 11 0.00 11 0.0008 0.0001 0.0000 0.00 04 Table 5.18: Percent Difference in the 8.k of Depletion at 100 Hours Cooling (SCALE/TRITON minus CASM0-5 Llk of D epletio n over the Llk of Depletion)

Case description Case 10 20 30 40 50 3.25% e nrichm ent dep l et i o n 1 -0.50 -0.52 -0.48 -0.62 -0.48 5.00% e nri c hm e n t depletion 2 0.06 0.08 0.10 0.00 0.10 4.25% enric hm ent depletion 3 0.3 1 0.02 -0.06 -0.14 -0.18 off-nominal pin depletion 4 -0.49 -0.54 -0.48 -0.42 -0.40 20 W ABA depleti o n 5 0.09 0.31 0.09 -0.04 -0.02 104 IFBA d e pl et i on 6 0.71 0.51 0.1 2 -0.19 -0.20 104 IFBA , 20 WABA d e pl et i o n 7 0.37 0.65 0.29 -0.01 -0.01 hi gh boron d ep l et i o n = 1500 ppm 8 0.36 -0.06 -0.09 -0.1 2 -0.02 b ranc h to hot rack= 338.7K 9 -0.16 -0.13 0.00 -0.13 -0.02 branch to rack boron = 1 500 pp m 10 -0.79 -0.83 -0.76 -0.72 -0.68 hi gh power d e n s it y dep l e ti o n 11 -0.02 -0.34 -0.25 -0.25 -0.19 Table 5.19: Percent Difference in the 8.k of Depletion at 5 Years Cooling (SCALE/T RITON minus CASM0-5 Ll k of D epletio n over the Llk of D e ple tion) Case description Case 10 20 30 40 50 3.25% e nrichm ent depletion 1 -0.01 -0.22 -0.26 -0.32 -0.17 5.00% e nrichm ent depletion 2 0.68 0.51 0.28 0.20 0.21 4.25% e nrichm e n t d ep l et i o n 3 0.79 0.20 0.10 0.01 0.04 off-no minal pin d eplet ion 4 -0.09 -0.15 -0.30 -0.25 -0.20 20 W ABA d e pl etion 5 0.38 0.44 0.29 0.1 9 0.16 104 IFBA d e pl etion 6 1.16 0.61 0.30 0.03 0.01 10 4 IFBA, 20 WABA d ep l e tion 7 0.78 0.95 0.48 0.30 0.12 hi gh boron d ep l e ti on= 1500 ppm 8 0.5 6 0.34 0.23 0.0 6 0.14 b ra n c h to hot rack= 338.7K 9 0.31 0.25 0.1 0 0.09 0.18 branch to rack b oron = 1500 ppm 10 0.19 -0.4 0 -0.37 -0.48 -0.37 hi g h p owe r d e n s it y depletion 11 0.47 0.16 0.10 0.09 0.04 N ET-28091-0003-01 , Revisi on 0 60 -0.0009 0.0012 0.0003 -0.0010 0.0005 -0.0004 0.0004 0.0006 0.0008 -0.00 14 0.0005 60 -0.41 0.12 -0.09 -0.41 -0.03 -0.36 -0.17 -0.01 -0.02 -0.58 -0.16 60 -0.08 0.25 0.05 -0.16 0.18 -0.10 0.0 8 0.19 0.18 -0.3 1 0.1 5 54 Table 5.20: Percent Difference in the Llk of Depletion at 15 Years Cooling (SCALE(f RITO N minus CASM0-5.c.k of Depletion o ve r the t.k of Depletion)

Case description Case 10 20 30 40 50 3.25% e nrichm e n t depl e tion 1 0.49 -0.17 -0.37 -0.33 -0.28 5.00% enrichment depletion 2 0.92 0.62 0.2 3 0.1 8 0.16 4.2 5% enr i c hm e nt deplet i on 3 1.09 0.40 0.06 -0.0 1 -0.01 off-no minal pin d ep l et ion 4 0.47 -0.18 -0.17 -0.30 -0.24 20 W ABA depl etion 5 0.66 0.69 0.30 0.10 0.1 1 104 IFBA depl etion 6 1.37 0.77 0.28 0.09 -0.03 104 IFBA , 20 W ABA d e pl e ti on 7 1.03 1.07 0.52 0.24 0.1 7 high boron d e pl e ti o n = 1 500 ppm 8 0.86 0.37 0.18 0.01 0.11 branch to h ot rack= 338.7K 9 0.38 0.15 0.03 0.06 0.11 branch t o rack b oron = 1500 ppm 10 0.38 -0.38 -0.45 -0.42 -0.43 hi g h power d e n s it y depletion 11 0.84 0.32 0.03 -0.01 0.07 5. 7 Special Case Depletions 60 -0.15 0.21 0.05 -0.1 6 0.08 -0.07 0.06 0.10 0.1 3 -0.28 0.08 Due to th e l imited s pace in Region 1 , assem bli es th a t ha ve been disch a rg e d and do not meet th e requirements for R egi on 2 were re v iewed to remove conservatisms in the depletion anal ys is. The lar ges t conservatism i s genera ll y th e depletion condition that all assemb li es co nt ain th e m aximum burnable absorbers of the batch gro upin g. For se l ected assem b l i es, a n a l ys is i s performed usin g th e fu ll avai l ab l e in fonnatio n on the assemb l y. T hi s sec tion di sc u sses the change in d ep l et ion co ndition s. Sectio n 8 pro vi d es the result s of th e a n a l ysis for special asse mblie s. Tab l e 5.21 pro vides the d e pletion p ara met ers t hat are u se d for s p ecia l d e pl e tion a n a lysi s. Table 5.21: Special Case Depletion Parameters Limiting Enrichment Fraction of PPM Peaking Assembly Fuel Type Theoretical Burnable Absorber (soluble ID (w/o) Density boron) Factor AI O HIPAR 2.2 1 0.943 None 570 0.92 F44 HIPAR 3.35 0.933 None 540 1.05 L48 LOPAR 3.69 0.944 1 6WABA 660 0.69 W52 OFA 4.96 0.946 20 W AB A/10 0 IFBA 880 0.84 XIS OFA 4.9 5 0.950 20 W AB A/11 6 IFBA 880 0.8 7 U12 (IP3) LOPAR 3.2 1 0.950 1 2 WABA 560 0.90 V43 (IP3) vs 3.80 0.950 20 W AB A/6 0 IFB A 6 50 1.1 2 NET-28091-0003-01, R ev i s ion 0 55 AlO represents 4 other assemblies (A09, A26, A33, and A34) which have the same characteristics but slightly higher burnups. These assemblies were burned for one cycle under the D-bank (therefore, had no burnable absorber).

The top node is depleted with a control rod for the entire depletion.

The lower nodes were depleted with a control rod for 2 GWd/T. Since the assembly was under D-bank in the bite position , t he DOE axial burnup profile between 14 and 18 GWd/T i s u se d. F44 was in Cycles 4 and 5 and did not contain burnable absorber inserts. With burnup during only two cycles, F44 did not meet the Category 4 fuel requirement by a small amount. When analyzed with its actual burnable absorber (none), it easily made the requirements for Category 4. Of the eight symmetric s isters to F44, six were in the core for thr ee cycles. The remaining sister, F52 has been casked but if it is returned to the SFP, it too meets the Category 4 requirement using this special case analysis. L48 and its sisters (L37, L38, L39, L44 , L51, L52, and L64) spent two cycles on the outside corner of the core. Because of this placement in the core, the burnup after three cycles was too low for Category 4 fuel by about 0. 7 GW d/T if the standard depletion condition for burnable absorbers is used (20 rodlet P yrex). Since this group of assemblies actually had a 16 rodlet W ABA in sert , the analysis of this set of assemblies using the W ABA instead of Pyrex , showed these assemblies meet the Category 4 reactivity requirements.

W52 and X18 are the lowest bumup assemblies of two sets of eight symmetric sister assemblies (W47 , WSO , W52, W53, W54 , W55, W59 , W60 and X09, Xl 1 , X12, X14, X16, X18, X44, X45). Whichever category W52 and X18 qualifies for, then the other seven will also qualify. W52 misses the Category 4 fuel burnup requirement by about 1 GW d/T. X 18 misses the Category 4 requirements by only 0.2 GWd/T. The only benefit in the depletion analysis for these two sets comes from reducing the IFBA rods from 148 to 100 and 116 for the Wand X set respectively. A special depletion is performed for these two sets with the reduced IFBA. In addition to the improved depletion, the actual axial burnup NET-28091-0003-01, Revision 0 56 profile for these assemblies is used. With these adjustments, these two groups of eight assemblies meet the Category 4 requirements.

Finally, a set of fuel assemblies from IP3 did not meet the Category 4 fuel requirements using the standard depletion analysis. The assemblies Ul2 , U2 l, U3 l , and U4 l from IP3 actually contained 12 WABA rather than the 20 Pyrex used in the standard depletion for Batch Group IP3 (A-U). Modeling t hese assemblies with the correct burnable absorber inserts allow them to make the Category 4 fuel r equirements. V43 and V48 did not initially meet the Category 3 fuel requirement but with special depletions , these two assemblies qualify for Category 3. 5.8 Reduced Power Operation at End of Life and Fission Gase s DeHart [34] demonstrated that operating history has a small effect on spent fuel reactivity.

However, at the end-of-cycle (EOC), the reactor power may be reduced (for example , a planned coast down) and this can cause a small reactivity change. One of the key absorbing fission products, Sm-149, reaches an equilibrium concentration during power operation tha t is independent of power. However , its precursor, Pm-149 , is directly proportional to power. At a reduced power , there is less Pm-149. Pm-149 decays into Sm-149 with a 2.2 day half-life.

Thus, if a reactor reduces power at end of cycle, there would be less Sm-149 in the cooled fue l , which is a positive reactivity effect. Therefore, ignoring low power operation during the last month is non-conservative. To account for this effect, the amount of Pm-149 can be reduced to one half of the full power content for all criticality calculations (which results in a penalty of about 100 pcm). This covers coast downs to 50% power and covers all past operating experience and anticipated future operation at the Indian Point plants. Furthermore , a significant fraction of once or twice burned assemblies are placed on the core periphery in the last cycle of the assembly's depletion. So a high peaking factor does not reflect a possible very low peaking factor (0.5) at the end of depletion.

As discussed above, it is at the end of depletion that the amount of Pm-149 is important.

To account for both the "end of depletion" effect (0.5) NET-28091-0003-01, Revision 0 57 and the coast down effect (0.5), t h e amoun t of Pm-14 9 i s re du ce d t o on l y 0.5 x 0.5 = 0.25 o f t h e full p ower i s otop ic conte n t in a ll criticality calculations.

This results in a reactivity penalty of about 250 pcm in the final k err calculations.

To conservati v ely account for fission gases escaping the fuel and migrating to the plenum , a ll krypton and xenon isotopes are reduced by 32%, all rubidium isotopes are reduced by 44%, and all iodine and bro mine isotopes are reduced by 10%. These adjustment factors are justified in the response to an RAI documented in Reference

[2]. 5.9 Production of Atom Density Sets SCALE TRITON outputs atom densities in two ways , through an output file , StdCmpMixOOOXX where XX is the material number , or OPUS pit files. The StdCmpMixOOOXX supplie s the atom densitie s for the end of the run wherea s the OPUS pit files provides atom densities for selected isotopes as a function ofb u rnup and coo l ing time. Rather than reru n SCALE for each desired burnup and coo l ing time, OPUS pit files are saved. For burnups between the SCALE time step s , the atom den s ities are l i nearly interpolated between the time steps. The SCALE time decayed atom densities are on l y valid for t he last time step s o in order to get the atom densities after cooling, the atom densities are decayed outside of SCALE. The cooling time decay , burnup interpo l ation , Pm-149 correction , and the fission gas corrections are perfonned using a sma ll FORTRAN code developed by this critica li ty team ca ll ed INTRPND. With isotopic s from the depletion calculation s recorded e v ery 2 GWd ff (see Section 5.6), the isotopics at any particular burnup can be interpolated.

Since the burnup delta is small between burnup points , linear interpolation can be used. To validate this approach , the isotopics at 40 GWd/T were interpolated from the OPUS p i t files at 38 GWd/T and 42 GWdff at 5.0 w/o enrichment.

Using the interpolated isotopics the k err for the previous CSA Region 2 model was 0.96653+/-0.00015. Taking the isotopics directly from the OPUS plot files at 40 GW d/T , the calculated k err was 0.96651 +/- 0.000 1 5. The difference is well within the Monte Carlo statistics. A similar verificat i on was perfonned at 11 GWd/T , NET-2809 1-0003-01 , Revision 0 58 where an interpolation between 9 and 13 was compared to the direct calculation at 11 GW d/T. The calculated k e ff using the direct isotopics was 1.1366 +/- 0.0002 , while the interpolated case was 1.1362 +/- 0.0002 , a difference of only 0.0004 , which is within the expected Monte Carlo variation. Each isotope is decayed using decay constants from the CRC Handbook 85th Edition [58]. Each iso tope is decayed into its daughter product which also may be radioactive (the decay is e-I i where 11. is th e decay constant). To ensure that the correct isotopics are obtained , the decay time desired is divided into 10 sub-intervals. The first nine sub-intervals are decayed and the tenth interval is then divided into 100 sub-intervals. The first 99 of these sub-inter v als are decayed and the last sub-interval is again divided into 100 sub-inter v als. This corrects for the fact that some nuclides are decaying into something else that is also radioactive.

If the decay time is not divided in to these very fine sub-intervals, the final concentration at the end will not be correct. It was found that this division of the decay time is fine enough such that any finer division resulted in no discernable difference in the concentrations. The correct concentration for all nuclides at the end of the decay time is thus obtained. To check the cooling time model used in the interpolation program , a special depletion was performed at 5.0 w/o to a bumup of 40 GWd/T. Then SCALE was used to decay the isotopes for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> , 1 year , 5 years , and 25 years. The interpolation program was also used to decay the isotopes to the same cooling times. Table 5.22 shows the results of the verification of the cooling time. The differences are within the Monte Carlo statistics (2 sigma of+/- 0.0004) except for the c a se at 25 years. The calculated k e ff from the interpolation program at 25 years is conservative as it produces a higher k eff-NET-28091-0003-01 , Revision 0 59 Table 5.22: Verification of Cooling Time Model in the Interpolation Program Interpolation Cooling Time SCALE/ORIGEN k Program k Difference 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1.002 3 1.0023 -1 year 0.9993 0.9996 0.0003 5 years 0.9847 0.9848 0.0 001 25 years 0.9449 0.9457 0.0 008 5.10 Summary of Limiting Depletion Conditions This section h as provided the details on the limiting depl etion conditions. The d ep letion ana l ysis i s p erfo rmed for 11 batch groupings and six sets of s pecial assemblies.

For the temperature calculations , several batch groupings were combined, so there are on l y eight sets of temperatures. For each batch grouping, the depletion analysis is perfonned w ith five different temperature sets for the fuel and moderator that correspond to the burnup avera ge d assembly peaking factors of 0.60, 0.80 , 1.00, 1.2 0 , and 1.40. For each b atch grouping and p ea king factor, a depletion ana l ysis is performed for limiting top node conditions and conditions lim iting for a ll node s below th e top node. The depletion analysis is perform e d ove r a range of enrichments appropriate for each batch grouping. Atom d ensit ies were generated at each 2 GWd/T bumup step. Atom densities for burnups in between these points were determined by interpolation using the small FORTRAN code, INTRPND, d esc ribed in Section 5.9. The INTRPND code deca ye d the isotopic concentrations to any cooling time. Finally , the INTRPND code corrected th e Pm-149 atom densities for low power operation and corrected the fission gas fractions for their release rates for long-term storage of fuel. The limiting parameters for each batch grouping a re found using: 1. Tab l es 5.6 and 5.7 for moderator temperature and density, 2. Equations under Figure 5.2 and under Table 5.7 for the fuel temperature , NET-28091-0003-01 , Revision 0 60 Westinghouse Non-Proprietary Class 3 3. Table 5.8 for the burnable absorber design and bumup at which the burnable absorber is removed (except for Batch Grouping Z for w hi ch the burnabl e absorbe r i s never removed), and 4. Table 5.9 for the soluble boron concentration. Specific powers of 16 and 26 W/g multiplied by the peaking factor are used for the top and lower nodes respectively for a ll groupings. 95% of the U0 2 theoretical density is used for the stack density except for the special cases given on Table 5.21. For future fuel (Batch Grouping Z), the fuel assem bl y is depleted with a contro l rod inserted for 2 GW d/T. Then the assembly is depleted with a 20 rod l et W ABA which is never removed. The initial control rod depletion is to cover future extended part power operation with contro l rods inserted.

The bumup of2 GWd/T burnup requires operation of approximately

1.4 effective

fu ll power months. In addition to the WABA, the fuel is mode l ed as contai nin g 1 48 IFBA pins in all but the top node. The IFBA 1 0 B lo ading is l .5X [( mg 1 0 B/inch)] a.c to cover future designs (residua l poison from IFBA is not credited in the criticality model). For a ll other batch gro upin gs the co ntrol rods are n ot included in the standard depletion.

However assemblies under D-bank are id entifie d and the burnup requirements for these assem bli es are increased as specified in Section 5.5. NET-28091-0003-0 l, Revision 0 6 1 6 Rack Model This section describes the Keno models used in the analysis. Two-by-two (2x2) storage cell array models are used for the analysis of the three out of four areas in Region 1 (Category 2 cells) and Region 2 (Category 4 cells). A 2x2 model is also used to analyze the checker board arrangements in Region 1 and Regio n 2 (Category 1 cells). A full pool model is created to confirm the burnup requirements for the Category 3 and Category 5 fuel assemb l ies , the category cell interfaces, and to perform analyses of the Misplaced Assembly and multiple misload accidents.

6.1 SCALE

2x2 Radial Model s The rack and fuel dimensions are given in Section 3. The nominal dimensions are used in the models with the exceptions mentioned in this section. The Boraflex T M is modeled as water. As pointed out in Section 3, if any Boraflex TM remains it would still have some 1 0 B so modeling it as water is conservative. The Boraflex TM sheathing is a plate with the outside edge bent down at a 45 degree angle creating a 0.112-inch (Region 1) or 0.092-inch (Region 2) pocket for the Boraflex TM sheet (8 , 9]. The SCALE model preserves the minimum sheet material for Region 2 (which is less than Region 1) by modeling the sheathing as a squared off box with a width of 7.70 inches (8 , 9]. The same sheathing width is used for both regions. In Section 7 it is shown that using the minimum sheath material is conservative but not s i gnificant.

The connecting steel between Region I cells is modeled as an extension of the cell wall rather than a separate piece of steel. The connecting steel is slightly thicker (0.09375 inch) than the cell wall thickness (0.075 inch) [8]. This model was confirmed by accurate l y modeling the connector steel in a Region 1 checkerboard model of water holes and fresh assemb l ies of 5.0 w/o enrichment with 64 IFBA rods. The d i fference calculated is 0.0005 L\k in the conservative direction (the Monte Carlo uncertainty for these nms is only 0.00006 L\k). NET-2809 1-0003-01 , Revision 0 62 The Region 1 rack modules are separated by 1.625 inches and the Boraflex TM sheathing on the outside of the module is 0.075 inches thick rather than the nominal 0.0235 inches [8,35]. The normal flux t rap is 1.351 inches in the East-West (vertical) direction and 1.571 inches in the North-South (horizontal) direction

[8]. The greater separation between cells at the module interfaces and thicker sheathing assures that the infinite model is conservative for the finite racks consisting of three module s. The Region 2 rack modules are separated by at least 1.25 inches [35]. The re s ultant cells on the outside row of the module have a 0.075 inch wall [9]. Even witho ut the module separation, the infinite model i s conservative since, the Region 2 module interfaces possess an additiona l steel plate. Since Region 2 is an arrangement of cell boxes with resultant cells, a 2x2 model with a periodic boundary condition is required.

Since Region l storage patterns include a checkerboard and a 3-out-of-4 arrangement, a 2x2 model i s also required for Region 1. Figures 6.1 and 6.2 are co l or plots from the KENO models for Region 1 and 2, respectively.

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Figure 6.1: Region 1 KENO Model NET-28091-0003-01, Revision 0 63

  • * * * * * * *
  • Fig u re 6.2: R eg ion 2 KEN O M od e l The 2x2 model s are two cell pitches wide in the x and y directions.

In creat ing the R egio n 2 model , t h e bottom l eft is modeled as a complete ce ll box with its Boraflex Ž sheathing. T h is requires that the cell w a ll be sp l it between the top and bottom of th e model (likewi se for the l ef t an d right). When the periodic boundar y condition i s applied, th e two ce ll wall pieces fit together to precisely match the actua l rack d i mensions. NET-2809 1-0003-01 , Revis i on 0 64

6.2 Axial

Model All of the in finite model s discussed in this section are finite axia ll y. Above the active fuel , the models extend the clad 7.3 inches (except for Batches A-F where the plenum is 5.2 inches) creating a plenum composed of a stainless steel spring and void (the spring is 8% of the plenum volume). On top of the plenu m r egion is a 3 1.4 3 cm of a homogeneous mixture of 50% water /50% steel (to simulate the end fitting). Below the fue l , the reflector is 50 cm of water. Outside of the reflectors , there is a zero flux bo undary c o ndition. For the axial distribution of dep l etion isotopics , the fuel is modeled as nine discrete axial nodes. For non-blanketed fuel , the top eight nodes are 8 inches each and the bottom node i s 80 i nches. For assemblies with 6 inch blankets , the top eig ht nodes are 6 inches and the bottom node is 96 inches. For assemblies with 8 inch blankets , the top node is 6 inches , the second node is 2 inches, the third node is 4 inches, the 4u1, 5th, 6th, 7 t h, and 3th nodes are 6 inches and the 9th node is 102 inches. For axial burnup models using 13 nodes, the top node is 6 in ches , the second node is 2 inches, the third node is 4 inches, t he fourth through 12 t h nodes are 6 inches, and the 13 t h node is 78 inches. The bumup distribution that determines the fuel atom densities for eac h of these n odes is discussed below in Section 6.2.1. Some of the axial blanket designs used annular pe ll ets and these are conservative l y modeled u s in g solid pe ll ets (i.e., there i s more fuel in t h e model than reality).

6.2.1 Axial

Burnup Distribution To model the axia l variation of a fuel assembly isotopic content, an axial burnup profile is needed. For this ana l ysis , the limiting profiles from NUREG/CR-6801 [27] are used for the full-length bl anketed) fuel. Table 6.1 , below , is reproduced fro m NUREG/CR-6801. The initi a l burnup distribution is approximately cos in e shaped sh i fted down a litt l e due to l ower temperatures at the bottom of the core. With increasing burnu p, the center reactivity decreases, so the flux moves t owar d the e nd and the burnup di stribut ion flatt ens. In spec tion of Table 6.1 for burn up bin s 1 through 9 , as expected, shows the top node relative power generally decreases as the bumup decreases.

However, burnup bins 3 and 5 actua ll y ha ve hi gher relative burnups th an bins 2 and 4. This does not NET-2809 1-0 003-01, Revision 0 65 make physical sense. NUREG/CR-6801 is based on end of cycle data collected by the DOE. The shapes used for burnup bins 2 and 4 come from assemblies that experienced some feature that suppressed the b urnup at the top of the core such as a transition to axial blankets or perhaps control rod insertion. If the cycle length for these limiting assemblies had been shorter , then these assemblies would probably have had similar relative burnups and would have been counted in the lower bumup bin. It is concluded that the higher relative power seen from going from bin 2 to 3 or from bin 4 to 5 is actually an artifact of the data used to create the database and not due to a phys i ca l process. To eliminate this artificial increase in relative burnup with decreasing burnup, shapes 3 and 5 are eliminated and shapes 2 and 4 are used to cover the 2/3 and the 4/5 bins , respectively.

The top node (97 .22% of the axial height) for bins 6 and 7 is greater than the top node for bin 4 , but when the top two or three nodes are a v eraged , the expected decrease in relative bumup at the top of the core is observed , so bins 6 and 7 are not eliminated.

In summary , only bum up groups 1, 2, 4, 6, 7, 8 , and 9 from Table 6.1 are used. Burn up profile groups 10, 1 1 , and 12 are not used because no non-blanketed assembly has a burnup less than 14 GWd/T (an exception is F65 which required special analysi s, see Section 8.6). The profiles for these low burnup bins were actually selected to match a center peaked flux because the reactivity is not dominated by the top nodes un t il about 14 GWd/T. NET-28091-0003-01 , Revision 0 66 Table 6.1: Axial Burnup Profile vs. Burnup Bin (27) Burnup gro u p 1' 2' 3 .j 5 6 7 9 1 0 I J' 1 _t ial Burnu ran es (GWd/MT height C 0 ,) 42-46 38 2 34 38 <6 27 0.666 0 66 0 064 06 I 3 0.944 0.9 6 0924 I 007 13 9 1.0-t I 0-iS I 0 56 I 135 19.44 I 080 I 1 04 I 097 I 133 -.00 I 091 I 11 2 I 10 I 09 0. l.()<)3 I. I 06 I IO I I 0 .I I 1.092 1.102 I 10 3 I I L I 53 4 1 69 1.0 I 090 I 097 I 112 I 11 9 I 0 47 47.22 1.09-t 1.12 5 I 1 26 1.0 50 1.094 1.136 I 1 32 I 060 I 07 I 077 69 44 I L4 I 79 75.0 0 1.077 I. t:!O 1.0 73 5 I o-o 1.069 1.05 7 I 056 I 041 I I O I I 0 52 6.1 1 0.992 1.010 0.996 0.974 0. 71 I O-i5 0.996 91.6 0. 33 0. l I 0. 23 0.743 0.7 0.689 0.669 0.894 0.845 97 22 0 447 0 448 0. 73 0 '69 0 --T he burnup profiles are a step funct ion of burnup, so they are di sco ntinu ous at th e burnup bin bound a ri es. To e liminat e th ese discontinuities in a co ns erva ti ve mann e r , t h e s h a p e in a n y bin i s cons e r vat ivel y assumed to occur at the maximum burnup in th e bin. For any burnup in b etwee n the se burnup point s, the s h a p e i s lin ea rl y interpo l ated (th e s h a p es are not changing rapid l y b etwee n burnup bin s). For example , s uppo se a n asse mbly h as a burnup of39 GWd/T. U s in gNUREG/CR-6801 dir ect l y , th e top nod e wo uld have a rel ative burnup of 0.525. In thi s analysis , ho wever, the top n o d e ha s a rel ative burnup of onl y 0.467 (linearly interpolating b etwee n 0.447 at 38 GW d/T and 0.525 at 42 GWd/T). This ensures no di sco ntinuiti es and all burnup s h a p es a r e conservative.

NET-2 8 091-0003-01 , Re v ision 0 67 For analysis of full-length fuel , the l ower 10 nodes are averaged into one node. This does not affect the calcu l ation of k , since the top half of the assemb l y dominates the reactivity.

In fact, averaging the lower 10 nodes effective l y brings the bottom lower b u rned fuel toward the more reactive top, so the approach is conservative

.* There are only a few assemblies in the IP2 or IP3 SFPs that have less than 18 GW d (T ( 11 Batch A assemblies from IP2 , 30 Batch A assemblies from IP3, IP2 assembly F65 , and IP3 assemblies V43 and V48). If there are any future assemblies with a burnup of less than 18 GWd/T, they must be assigned as Category l fuel , which does not credit burnup. The burnup profile for the 14-18 GWd fT burnup bin on Table 6.1 comes from a unique profile produced by considerable burn up under a control rod. The 15 assemblies having less than 18 GWdfT burnup that were burned under D bank are analyzed with the 14-1 8 GWdfT burnup profile from NUREG/CR-6801. A ll of the Batch A assemblies in this group have significant margin since they were a ll depleted with a 20 rodlet Pyrex burnable absorber , but none of them actually had a burnable absorber.

The other assemb l ies are analyzed using the 18-22 profile (the 18-22 profile is more reactive than the 10-14 profile). For analysis of a xial blanket fuel , a conservative burnup shape is obtained by using the smallest relative burnup at each node from all assemblies in the group. The relative burnups are not re-normalized , so the assembly burnup in the analys i s is reduced. Since the reactivity is controlled by the top nodes , however , compensating burnup increases in the lower nodes have little effect on k. The ninth node from the top is used for the ninth and all lower nodes. This has been shown to be conservative when compared to using all of the nodes. This ensures a conservati v e profi l e when the reactivity is dominated by either the top or the center (the reactivity is never dominated by the bottom because the bottom nodes always have burnups that are higher than the corresponding top nodes).

  • For more discussion on using only the top 8 nodes and an averaged bottom node see the respon s e to the NRC RAI number 16 performed for the 2015 CSA (2]. NET-2809 1-0003-01 , Revision 0 68 For blanketed assemblies, all ax i a l bumup profiles from the plant were reviewed to detennine a limitin g ax i a l bumup profile for each of five axia l blanket designs: 6 inch annular* 2.6 w/o (Batch Q , R, S ofIP2) 8 inch annular 3.2 w/o (Batch T, U, V of IP2) 8 inch so lid 3.4 w/o (Batch W of IP2) 8 inch solid 3.6 w/o (Batch X of IP2) 8 inch solid 4.0 w/o (Batch 2A+ ofIP2 and Batch GG+ ofIP3) Assemb li es in Batch X are segregated into assemb li es that were depleted with no W ABA and those t hat were depleted with W ABA. The reason for this i s that the axia l bumup profile for non-W ABA assemblies is more limiting (lower bumup at the top) than for WABA assemblies , because the WABA pushes power toward the ends. In order to not penalize the WABA assemblies with the non-WABA profile, a separate depletion is perfonned by using no W ABA depletion with 148 IFBA. For the WABA group, the most limiting axia l profile in the WABA group is used, and ke rr is 0.9593. The most limiting axial profile in the non-WABA group is used to ana l yze the non-W ABA group, and k e tr is 0.9532. So , the loading curve for Batch X (based on the W ABA ana l ysis) can be used for all X assemblies.

Table 6.2 shows the axial bumup profiles detennined from the plant data.

  • The designs which used annular pellets are conservative l y modeled using solid pellets (more fuel). NET-28091-0003-01, Revision 0 69 Ta bl e 6.2: Ax ial Rela t i ve Burnup s fo r Blanket e d Di sc har ge d F u e l IP2 Batch: Q, R, S T,U,V w X X noWABA Blanket Length: 6 8 8 8 8 (inches) Blanket Enrich: 2.6 3.2 3.4 3.6 3.6 (w/o) Top Node 0.420 0.448 0.471 0.505 0.519 2nd Node 0.74 3 0.59 8 0.622 0.663 0.6 77 3rd Node 0.9 0 6 0.76 2 0.772 0.779 0.8 0 5 4th Node 0.9 89 0.866 0.8 8 3 0.884 0.8 66 5th Node 1.0 2 9 0.97 5 0.98 6 0.990 0.9 87 6th Node 1.04 8 1.020 1.03 0 1.033 1.0 27 7th Nod e 1.0 59 1.0 41 1.05 0 1.052 1.0 48 8th Nod e 1.0 64 1.0 51 1.060 1.061 1.0 57 9th Nod e 1.06 8 1.0 5 7 1.06 6 1.067 1.0 62 For the Batch group Z axial burnup profile , the profiles from Batch 2A assemblies are used. This was the first batch to use 4.0 w/o blankets. The previous batch used 3.6 w/o blankets.

Since the top nodes of the 2A fue l assemblies were surrounded by the 3.6 w/o blankets , the 4.0 w/o blanket would be s l ightly underburned by the presence of the 3.6 w/o blanket s. This " transition" from 3.6 to 4.0 w/o blankets bounds all future 4.0 w/o blanket fuel (designated as Batch Group Z). The Batch 2A assemblies are di vided into two groups -assemblies that had a WABA insert plus 148 IFBA and all other assemblies.

Since the WAB N 14 8 IFBA depletion condition is used , only the assemblies that had a WABA insert plu s 1 48 IFBA should be used to find the l imiting axial profi l e. Re s ults from the Batch X analysis showed that the dep l etion effect from a reduced amount of burnab l e absorber is worth more than the effect of the axial shape with reduced absorbers. As furth er confirmation of this effect, a s pecial depletion was performed for Batch Zin which the fuel was depleted with a W ABA plus 116 IFBA compared to the s t andard depletion with a W ABA and 148 IFBA. The k err for the 1 16 IFBA depletion using the correspo n di n g profile is 0.9569 while the k e rr for the 1 48 IFBA dep l etion using the 1 48 IFBA profi l e is 0.9584. The limiting profile is the minimum relative power in each n ode for a ll 2A assemblies that had a WABA inser t plus 148 IFBA. The limiting profile fo u nd at three different bumups is in Tab l e 6.3. NET-2809 1-0003-01, Rev i sion 0 70 Table 6.3: Axial Relative Burnups for Batch Z Fuel Burnup: 48 28.5 21 GWd/T GWd/T GWd/T Top Node 0.562 0.560 0.540 2 0.727 0.746 0.729 3 0.815 0.787 0.754 4 0.893 0.826 0.787 5 0.984 0.950 0.916 6 1.020 1.001 0.976 7 1.039 1.029 1.017 8 1.047 1.042 1.041 9 1.051 1.050 1.061 10 1.055 1.057 1.074 11 1.056 1.052 1.076 12 1.065 1.067 1.095 13 1.061 1.059 1.088 14 1.065 1.061 1.092 15 1.070 1.072 1.102 16 1.069 1.064 1.089 17 1.075 1.071 1.093 18 1.080 1.083 1.095 19 1.081 1.086 1.088 20 1.081 1.085 1.075 21 1.074 1.079 1.054 22 1.047 1.045 1.003 23 0.976 0.971 0.918 24 0.875 0.890 0.848 25 0.713 0.738 0.718 Bottom Node 0.534 0.545 0.520 To s impli fy the a n a l ys i s , only the top nod es n ee d t o b e m o d e l e d w ith the l as t nod e r eprese ntin g a ll of t h e n o d es b e l ow the l as t n o d e. The fewe r th e numb e r o f n o d es mod e l e d , th e more co n serva ti ve th e r es u lt b eca u se a ll o f th e n o d es b e lo w th e l as t on e ar e m o d e l e d w ith a l owe r burnu p. I t was found th a t u s in g ni ne n o d es fo r th e hi g h e r burn e d fu e l i s n o t ove rl y co n se r v ati ve (th e k at 4 8.2 GWd/T u s in g nin e n o d es i s 0.9 5 86 w hil e th e k u s in g a ll 26 n o d es i s 0.9585). Fo r th e ana l ys i s at 2 1 a nd 28 G Wd/T (Z fue l), 13 n o d es we r e u se d in s t ea d of nin e, in o rd er to r e du ce some o f th e co n serva ti s m. Ta bl e 6.4 s umm arizes th e k c a lc u l a tion s (th e M o nt e C arlo s t a t is ti c al un ce rt ai nt y i s onl y+/- 0.00003 t.k). NET-2 8 091-000 3-0 1 , R ev i s i o n 0 7 1 Ta bl e 6.4: Ca lcul ate d k ve r s u s Numbe r o f N o des M od eled Burnu p 9 n o d es 13 n o d es 26 no d es (GWd/T) 21.0 (3 of 4 i n R eg i on 1) 0.9689 0.9686 0.9686 28.5 ( 4 of 4 in R egion 1) 1.0117 1.0115 1.0114 4 8.2 (3 of 4 in R egion 2) 0.9586 0.9585 0.9585 6.3 Dimensional Changes with Irradiation The fue l assembly dimensions change a small amo u nt with irradiation.

This creates a change in r eactivity.

This change in reactivity is real but too small to include in fuel management analysis.

With irradiation , the fuel pellet densifies and then expands , the clad grows and creeps down to the pellet , and the assembly grid expands. These are changes to the dimensions not the mass. The changes in the dimensions of the fuel pellet (with mass constant) create an insignificant reactivity ch a nge. The change in the dimensions of the clad and grid , however, result in more water relative to the fuel. Since the fuel is deliberately designed to be under moderated (to ensure a negative power coefficient), adding more water to the fuel assembly is a positive reactivity effect. 6.3.1 Clad Creep The clad outer diameter initially decreases due to creep caused by the pressure difference between the core pressure (2000+ psi) and the He fill gas (200+ psi) in the rod. At some point the clad contacts the pellet and then the clad outer diameter increases as t h e pe ll et swells. When manufactured, per Tab l e 3.2 , the difference between the pellet diameter and clad inner diameter is 0.0075 inches or 190.5 microns (l0*6 meters). If the clad were to creep down to the initial pellet diameter , then this would be a 1.8% decrease in the clad outer diameter (it is assumed that the clad thickness is constant).

The clad creep rate depends on the clad material.

However, for all of the c l ad materials, the pellet densities and then grows back past its initial outer diameter before 190 microns of creep occ u rs. NET-28091-0003-01, Revision 0 72 Figure 6.3 s ho ws clad creep after one cycle at North Anna [37]. As see n from Figure 6.3, Z irc a l oy-4 creeps down faster tha n Z IRL OŽ. T h e m aximum c r eep down af t er one cycle is abo u t 90 microns for Zircaloy-4 and 70 microns for ZIRLOŽ. The creep do w n data i s only for one cyc l e, s in ce b y the e nd of the seco nd cycle the c l ad h ad reached th e pellet OD. Sabo l , et al, confinns this as fo ll ows: " Th e profllom etry data obtain e d aft e r on e cy cl e of ir radiation has b een used to e valuat e th e diff e r e n ce s in th e alloys' in-r eac tor c r ee p b e ha v ior. Fu e l-clad c onta ct over most of th e rod l e n g th occurred during th e seco nd cy cl e, and th e pr ofl l o m e tr y data obta in e d on th e two-cycle rod s ar e co ntroll e d b y th e fuel p e ll e t swe llin g rath e r than th e claddin g c r ee p. Th erefo r e, th e two-cycle data c annot be u sed fo r c r eep analys i s". [3 8] F igure 6.4 pro v ides mor e data on the creep of Z ir ca l oy-4 [39]. From this plot , the creep down is ba s ically lin ear with burnup and can r eac h about -0.8% of the initial clad outer diamet er (abo ut 80 mi cro n s which agrees with th e creep d a ta from Fig ur e 6.3). Craepdown (microns) 0 20 40 60 80 100 120 0 so Zircaloy-4 ,oo 1 SO 200 2SO 300 Location (cm from bottom of ro(1) Figure 6.3: Comparison of Creep-down for ZIRLOŽ and Zircalo y-4 [37) (One cycle of irradiation at North Anna Unit 1) NET-28091-0003-01, R ev ision 0 73

-1,0 -0 , 9

  • o,e ll P.t -0,7 iii e 01 0 , 64 B0 -0,6 02 0,40) ct, A -63 0,59 dPA: !.. -0,5 & nelof rods -O -04 <i , G) [i) 'l {l -0,1 normalized with J0.85 -02 I , 200 300 400 soo 600 800 1000 1200 Exposur* Tim* (Oaysl Figure 6.4: Diameter Decrease versus Exposure Time [39) Figure 6.5 shows the clad creep for severa l a llo ys for fuel u se d at the Vandellos 2 Nuclear Power Plant [40]. T he initial clad OD for the Vandellos plant is 9.5 nun. NET-28091-0003-01 , Re vision 0 74 o ,,r"-----------------------------------.

-20 ,-.. '§: -40 a ..t: <.) -60 .I 'O ... -80 0 ell *= 'O 'O --100 -120 Segment Type Cladding X : SS Conventional Zr'y-4 0 : WI Low-tin Zr'y-4 6: \\Z ZIRLO 0: M.\1 MDA * : !\fit Low-rui with te.xture control A : ~t MDA with texture control * : 12 1 ZIRLO with texture control D o* *

  • 0 ... A A A A * * * ... ... ...... 0 * -140 ________________________________

...._ _____ ... 0 10 20 30 40 50 60 Segment a\*en,ge bumup (G cLt) Figure 6.5: Clad Creep Down for Vandellos 2 Nuclear Power Plant [40) Fina ll y, Figure 6.6 shows the clad creep for a fuel assembly from the Ulchin Unit 2 PWR [41]. Thi s assembly uses a low tin Zircalo y-4 clad. The fue l as s embly ha s a burnup of 50.5 GWd/T. The initial clad OD i s 9.5 mm. A s can be seen from F i gure 6.6 , the clad OD with oxide i s greater than 9.5 mm in the hi g h r ea c ti v i ty zone. Thi s i s due to s ignificant oxid e build up and fuel pellet e x pansion. For this as s embly at t his burnup i t is conservative to ignore creep. NET-2809 1-00 03-01 , Revision 0 75 I-Q) +-' Q.) E cu 0 9.65------------------------, 9.60 9.55 9.50 9.45 9.40 -----Average Diameter --Diameter(oxide corrected) 0 500 1000 1500 2000 2500 3000 3500 4000 Distance from Bottom End, mm Figure 6.6: Axial Distribution of the Fuel Rod Diameter at 50.5 GWd/T [41] Wit h bumup , the clad a l so builds up an oxide coating. This oxide coating displaces water so from a reac tivity p o in t of view , it is similar to in creasing the clad outside di a m eter. Fig ur e 6.7 shows a l arge data base of the oxide layer growth with bumup [ 42). After the clad creeps d own to the fuel pellet , the clad OD increases due to pellet swelling. Figure 6.8 shows the n ormal fue l pellet swelling (v i a pellet density) a s a function ofbumup [4 3). NET-28 091-000 3-01, Re vision 0 76 160 Z ir calo y-4 140 *ZIRLO O ZIRLO -P l ant C O lo w T in Z I RLO

  • P l ant D C low T i n ZIRLO -P l ant D 100 t.Op!tm1 zed ZIRLO
  • Plant E ¢Optimized ZIRLO Plan t s F & G 80 60 40 20 10 20 30 40 so 60 70 80 B umup (G W d U) Figure 6.7: Oxide La y er thickne s s with Burnup [42] 100 98 96 0 f-94 --.. .. --: --'::!?. 0 .£ 92 (/) ----.. C (l) 0 90 ----.. ---,------' 88 --.. .. --0 20 40 60 80 100 Pe ll et Bu rn up MW d/kg U Figure 6.8: Density of Fuel Pellet as a Function of Pellet Burn up [43] N E T-28091-000 3-01 , Revision 0 77 With the data presented in these six figures , a creep model has been approximated. Since the reactivity effect is small , the model is not meant to be bounding but to have a slightly conservative mean reactivity effect. The deviation about the mean is an uncertainty that could be statistically combined with the other uncertainties.

However, since the entire effect is small, the statistical combination of the uncertainty would be negligible , so it is ignored for this CSA. The modeling is as follows: 1. Per Figure 6.8, the fuel pellet returns to its original density at 30 GWd/T. 2. The pellet density changes are translated to c h anges in the pellet outer diameter (no axial swelling). Axial swelling is a planar loss of mass that decreases

k. 3. Using the slope of the change in pellet density given on Figure 6.8 ( l % density reduction per 10 GWd/T), the expanded pellet outer diameter is detennined as a function ofburnup. 4. Using Figures 6.4 and 6.5, assume the creep is linear with burnup. The slope for conventional Zircaloy-4 is approximated as 2.5 microns per GWd/T. 5. Using this data, determine the bumup where the clad creeps down to touching the pellet. 6. Use Figure 6.7 to estimate an oxide thickness gain of0.5 micron per GWd/T. This data is used to dete rmine when the clad returns to its original outer diameter.

Using this approach the maximum creep (the point the clad reaches the pellet) for Zircaloy-4 is at 46 GWd/T and is 115 microns. However, at that bumup the oxide layer is 23 microns thick (46 microns diametrical effect) making the net reduction of the clad outer diameter only 69 microns. Upon reaching a burnup of 5 6 GWd/T, the clad plus the oxide layer is back to its original outer diameter (this agrees with the measurements shown on Figure 6.6). The clad creep model used here simplifies the effects in a slightly conservative manner. This CSA ass umes the clad outer diameter decreases linearly from O to l 00 microns over the burnup range from NET-28091-0003-0 l, Revision 0 78 zero to 40 GWdff. As th e bumup increases from 40 GWdff to 56 GWdff, the c l ad outer diameter increases lin early to the point where it is th e same as the initial c l a d outer diameter.

Any furt h er in crease in the c lad outer diameter is con s ervat i ve l y ignored. The creep down as a function of bumup was determined to cover both Zircaloy-4 a nd ZIRLOŽ clad fuel. ZIRLOŽ clad however , creeps down less , so the analysis is conservative for ZIRLOŽ. ZIRLOŽ fuel has been u s ed since Cycle 13 (IP2) and Cycle 9 (IP3). 6.3.2 Grid Growth Zirca lo y grids grow a small amount with irradiation. Figure 6.9 shows measured grid growth in ZIRLOŽ and Optimized ZIRLOŽ [42]. Figure 6.10 shows the gr id growth in Zircaloy-4 (and M5 which is not used at Indian Point) [ 44]. RXA on Figure 6.10 stands for fabricated in stress-relief annea l ed (SRA) and recry s tallized (RXA) conditions.

Figure 6.11 shows the grid growth from VC Summer and Wolf Cree k measurements

[ 45]. The Wolf Creek asse mbli es are burn ed to 50 and 51 GW d/T [ 46]. ZIRLOŽ grows l e s s than Zircaloy-4, but they both grow more at higher temperatures than at l ower t emperatures. The Inconel grids do not have a growth problem [ 4 7]. King , et. al. states , " Mo s t We s tinghouse fuel designs u se Inconel top grids and Inconel bottom grids , and the Inconel grids are designed to maintain a pre-load on t he fuel rod until end of life" [ 48]. NET-2809 1-000 3-01 , Revision 0 79

! ,, cS .c lJr1o Grid GrOV;ih Data Base 0.80 ------------------------------.

+ Plant A -Plan!B 060-t---1

  • Plant C X GF Opbllllled ZIRLO Data t & 40 ---1 , Plant E Op . ed ZIRLO Da a

<'5

  • t ' "O *c (!) + + ! i -. ' . t I * . t_t : ** ** ; : * ' t X i f + X X t ; X 000+----,,----,----,---...----,.---,---.,----,--...,..-----i 20000 25000 30000 35000 40000 45000 50000 55000 60000 65000 70000 Bumup ( ,1WD ITU) Figure 6.9: ZlRLOŽ Grid Growth r42] 0.9 0.8 RXA Zircaloy-4 0.7 0.6 0 0.5 0 0.4 0.3 t 0.2
  • 0.1 0 0 0 8 0 0 0
  • t,.
  • j ... MSŽ 0.0 0 10 20 30 40 50 FA burnup (GW d l t U) Figure 6.10: Zircaloy-4 and MS Grid growth versus burnup [44] NET-28091-0003-01 , Revision 0 60 8 0 1.2 15 ~ZIRLO Grid Envelope Zr-4 Grid Envelope (Assembly
1) 1.0 --tr-Zr-4 Grid Enve l ope (Assembly 2} 12 --ZIRLO Fluence -:;-0.8 -* * * * *
  • Z ircal oy-4 Fluence C 9 ., u ... ., E: 0.6 .s:. .. *6 0 ... C) 0.4 3 0.2 0.0
  • 0 0 1 000 2000 3000 4000 Ax i al Elevation (mm) Figure 6.11: Grid Growth of ZIRLOŽ and Zircalo y-4 ve rsus Elevation

[45) There a re practical limit s to grid growt h. Excessive or unexpected dimensional changes of guide tube s or s p acer grids of a fuel assemb l y can re su lt in operationa l i ssues s uch as incomplete (contro l) rod in sertion (IRI) [ 42] or potential fuel assemb l y interact i ons and h a ndlin g concerns due to increases in the fuel asse mbl y enve l ope r es ultin g from th e l atera l growt h of the gr id s. > CD ::e ... I\ w .. E c 11'1 .. 0 ! C CD ::i iL For Ind ia n Point, if the fuel pin pitc h expands s uch that it i s uniform for the reactor asse mbly pitch , closing the i nter-a sse mbl y gap , then the fuel pin pitc h expands on l y 0.25%. Note that the outer ce ll s of the grid do n ot ha ve the fuel pin centere d b etween gr i d straps so there i s some add iti onal space even if the grid expan d ed to uniformly fit the asse mbl y pitch. If the grid stra p s were touc hin g for two e qu a ll y expande d assemblies the grid growt h wou ld be 0.69%. NET-28 091-000 3-01, Re v i sion 0 8 1 The grid growth clearly increases with bumup and the increase is more pronounced for higher bumups. A cubic fit has been generated to estimate the grid growth. The fit is roughly drawn on Figures 6.9 and 6.10. It is: Grid Growth(%)=

4.3E-6*BU 3 -0.000 1 3*BU 2 +0.0051 *BU where BU is the bumup in GWd/T. Although not all measurements of grid growth lie beneath this curve, using this grid growt h model as a unifonn pitch expans i on is expected to be conservative.

The Inconel grids above the active fuel will hold the fuel pins in their original pitch, since the Inconel growth i s in significant (from a mater i a l point of view as well as seeing less fluence , since they are outside of the active fuel region). The lower grids expand l ess. Therefore , in order to get the expansion in the pitch , the fuel rods wou ld have to bow. This analysis conservatively assumes the expa nd ed pitch is uni form over the entire l engt h of the fuel. For simp l icit y, a single fit is used to cover Zirca l oy-4 and ZIRLOŽ. However, Zircaloy-4 grows more tha n ZIRLOŽ. Zircaloy-4 was used as a grid for only Batches M , N, P, Q , and R in IP2 and Batches T, U , V, Wand X in IP3. All of these assemb li es have margin to their assigned reactivity category. In fact, if the grid growth were increased from 0.44% to 1.0% at 50 GWd/T , none of the assigned ca t egories for the Zircaloy-4 grid fuel assemb li es wou ld c h a n ge. 6.4 Averaged Assembly Peaking Factor Interpolation The depletion analysis is perfonned at five discrete averaged assembly peaking factors; 0.6, 0.8, 1.0, 1.2 , and 1.4. In order to simplify the dependence on this peaking factor, it is desirable to fit these five points with a single straig ht line ( one for each burn up , enrichment , cooling time , and batch grouping). It was fo und th at interpo l a tin g between peaking factors of 0.6 and 1.4 is slightly non-conservative at a PF=l.00. To ensure conservatism , depletions are performed at 0.8 and 1.2 and are li nearly int erpo l ated NET-2809 1-0003-01, R evision 0 82 and extrapolated using these two points. It was found that extrapo lati on to 0.6 and 1.4 is conservative. A representative graph at 50 GWd/T (Batch Z at 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cooling) is shown on Figure 6.12. 0.960 0.955 0.95 0 0.945 k 0.940 0.935 0.930 0.925 0.5

  • Points -Linear between 0.8 and 1.2 ' 0.7 0.9 T 1.1 Peaking Factor -,---1.3 1.5 Figure 6.12: Calculated ken versus Assembly Average Peaking Factor 6.5 Convergence of the 2x2 Infinite Model Calculations The convergence of the 2x2 reflected model k e ff calculation is generally achieved after only a few hundred generations.

However , all of the CSASS computer run s use a Monte Carlo sampling of at least 8000 generations and 8000 neutrons per generation.

Convergence could have been a problem in the past, when v ery few neutron histories were run (300 , 000 total neutron histories), but due to increasing the number of histories to 64 million , convergence is no l onger an issue. For both the number of generations skipped and the starting source , the SCALE default is used. For t he number of generations skipped , the default is 3. However , SCALE calculates the number of generations to skip that gives the minimum uncertain ty in the final result. Th e k eff reported for all of the calculations is the k eff With the optimum generations skipped. The number of generations skipped spans a wide range but is genera ll y between 100 and 200. When running a large number of generations , such as 8000, the input number of generations skipped is not significant to the final results. The default start NET-28091-0003

-01, Revision 0 83 source is a unifonn source over all of the fissile materials in the model. The number of neutrons per generation is a l ways 8000 or greater and for the four assembly models this sampling is enough to find the most reactive portion of the model. 6.6 Full Pool Models The full pool model is created by taking the 2x2 models for Region 1 and 2 described in Sections 6.1 and 6.2 and using them as units that are reproduced in arrays. The model has 4 large arrays (see Figure 3.1 for module identification):

1. Region 1 module A (10x8), 2. Region 1 modules Band C (combined as 21x9), 3. Region 2 modules D, E-1 , F-1, F-2 , G-1 , and G-2 (combined as 24x32), and 4. Region 2 modules E-2 , E-3 , and H (combined as l lx32). Modules E-2, E-3, and H are 11 ce ll s across (north to south), but since the modeling is using 2x2 units , a new lx2 unit was made and added to the right hand side of the model. This lx2 model correctly removed the Boraflex TM box on the outside of the rack module near the SFP wall but did not add the steel plate that is used to close the resultant cell on the outside of the module. The full pool model does not model the gap between the rack modules. This means that Module H is p laced next to Module C and lowered so that the bottom of Module H is the same as the bottom of Module C. Directly below Module H without any additional space are Modules E-2 and E-3. To the left of Modules E-2 and E-3 are the other Region 2 modules without any space between modules. This means there is more water on the outside than in the real SFP , because the inside dimensions of the SFP in the model are the actual dimensions of the SFP. The separation in the model between the rack and the SFP liner on the to p and left side of Region 1 is the actua l separation.

Since the separation between the rack modules has been removed in the model, there is more water on the right hand side of the model than is really in the SFP. Similar l y , at the bottom of the model there is more water than is actually there. Since NET-28091-0003-01, Revision 0 84 th e 2x2 unit fr o m Sec t i on 6 was a pp l i e d , th e ex t e r i o r of th e rac k modul es g en e ra ll y h as l ess s t ee l th a n th e o ut s id e wall o f a rac k modul e, b ut it i s clo se. In Sec t ion 6.6.1 , a nal ys i s of th e se n s iti v it y t o th e S F P e d ge i s p e r fo rmed , w hi c h s how s s m a ll se n s iti v it y. T h e s ep arat i o n b e t wee n th e R egi on 1 ra c k a nd th e t op wa ll i s 2.1 25 in c h es. T h e se p ara ti o n b e t wee n R egio n 1 and R egio n 2 ra ck s from th e wa ll on t h e l eft i s 1.25 i nc h es T h e SF P h as a 0.25 in c h s t a inl ess stee l lin e r and a co n c r e t e w a ll o u ts id e th e lin er. Figu re 6.1 3 is a SC AL E ge n era t e d dra wi n g of th e full p oo l m o d e l. F i g u re 6.13: F ull Pool M od e l N E T-2 8 091-000 3-01 , R ev i s ion 0 ll.C(M) O wrn fWHCRUll 2 -l'WHLIUl'll.

J -MlUtlAL" -l'VIJ[IUAL 7 0 Ml[IUAL I -Ml[RlR.10 C)MIOUAL J t -fWHDUAL 6' -M I UUnl '3 M I UUAI. 1 10 M l (RIAL 2 1 1 flHlUUAL 2 1 2 ~Ml(IUtlt.2 1 3 (::J mmnAL 2 1 4 Q M1£11tlAL 2 1!S WHOUAL 2 1 8 MIUUAl.217 M lUUAl. 211 MTiltlAI.

312 CJ MlUWll 3 1 3 -Ml[RIAI 3 1 4 Ill Ml[IW'll 3 t!S -Mf(IUnt.Jl6 -MfUUfll.3 1 7 -Ml[RlfL 318 -M J Utltll.,1 0 -Ml(RIAL'lll

-P!AllRIAL41 2 -Mf[MIAL<II J -PWU UIIAI "11'4 -MJ[RIAL"115

-MllJUf:11'41 6 -Ml[IUtll.<1 1 7 -Ml(RIAL*H 8 -Ml(RIAL!UI -MfU UAL!S l 2 -Ml[MIIW..!SIJ -M1CRIAL!il 4 -Mll.RIAL!S l!S MIIM U II. 5 1 8 -WHUtlAL!Sl7 -MIUUAl.!S 1 8 O iwHUllnt.9etl 85 -----_j The pink squares in Figure 6.13 are empt y cells which can contain up to 50% water displacement with non-fuel compone nt s. These ce ll s are modeled w ith a void fraction of 50%. The water holes (white) are modeled as pure water. However, Section 8.13 provides ana ly sis that shows that up to 50% volume fraction of stainless stee l may be in these water holes. Figure 1.1 shows that contro l rods are in specified l ocations.

These contro l rods are in the full pool model but cannot be seen in Figure 6.13 due to poor reso lution. F igure 6.14 is a blow up of Module Hand the contro l rods can be seen. ........................................................................................................................................................................

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5 1 7 -Mf(ll:Hl.5 11 D"'HENHI. Ne 86 There could be a concern regarding the content of control rods that have been used in the core. Due to end effec t s the reactivity is dominated at the top of the active fue l. However, the top third of t h e control rods is not allowed in the core if the power is above 50% of full power. Any depletion of the control rods is insignificant.

However, to confirm that content is not important, a run was performed with the atom densities of the control rod reduced 20%. The calculated k e ff is within the Monte Carlo uncertainty of the reference case. 6. 6. 1 Sensitivity of the Full Pool Model to Modeling Assumptions A number of model sensitivity cases were performed to determine if the modeling of the outside of the racks and the SFP wall is adequate.

Table 6.5 provides the results of these sensitivity cases. The reference cases use the nominal separation from the wall found on the SFP layout drawing for the top and left side of the model [35]. The reference cases are at the limiting enrichment (5.0 w/o) and bumups (21, 27.7, 48.19, and 57.89) p l us asymmetry.

For this sensitiv i ty study, the linear dimensional changes are l arge (tota l e l imination and a doubling of the width). Even with these large changes , the maximum reactivity is only five times the Monte Carlo uncertainty of the cases. The models are slightly more sensitive to the separation from the wall but with moving the racks as close to the wall as possib l e, the reactivity effect is at most 0.0005 in k. The model uses the standard regulatory concrete which is a s t andard mix t ure in SCALE. An EPRI study of a conservative minimum water concrete is used to find the sensitivity

[13]. This very conservative concrete only increased k by 0.0008 for Region 2 and 0.0004 for Region 1. Since all of the extreme changes caused 0.0008 or less change in reactivity , the full pool model is adequate.

The maximum k for the analysis is 1 % less than the regulatory requirement.

NET-28091-0003-01, Revision 0 87 Table 6.5: Full Pool Model Sensitivity Tests Case k Sigma '1k Region 1 Reference 0.9687 0.00007 Eliminated the SFP steel liner 0.9689 0.00006 -0.0002 Increase the SFP steel liner from .25 to .5 inches 0.9686 0.00006 0.0001 Decreased rack/liner separation from 2.125 to 0.8225 inches (Top) 0.9691 0.00006 -0.0004 Increased rack/liner separation from 2.125 to 4.125 inches (Top) 0.9686 0.00006 0.0002 Changed Concrete from reg-concrete to EPRI -minimum water 0.9691 0.00006 -0.0004 Region 2 Reference 0.9584 0.00005 Eliminated the SFP steel liner 0.9585 0.00006 -0.0001 Increase the SFP steel liner from .25 to .5 inches 0.9587 0.00006 -0.0003 Moved rack to meet the SFP steel liner (left side) 0.9589 0.00006 -0.0005 Increased rack/liner separation from 1.25 to 2.98 inches (left) 0.9584 0.00006 0.0000 Changed Concrete from reg-concrete to EPRI -minimum water 0.9592 0.00006 -0.000 8 6.6.2 Convergence of the Full Pool Model There i s a c la ss ic problem for Monte Carlo c onvergence known as the "k-E ffecti ve of the world." T hi s problem wa s introduced by E lliot White s ides in 19 71 (50] and more r ece nt paper s such as the Brian C. Kiedrow s ki and Forre st B. Brown pap e r at ICNC 2011 (51]. Concern has been raised that a large full pool model could face these probl e m s. First, in 1971 when White s id es rais e d thi s issue it was common pra c tice t o run a few hundred neutron s per generation. The pap er b y Kiedrow ski a nd Brown u se d a lot mor e n e utron s per generatio n but considered solutions with 10 , 000 n e utron s p er ge neration o r l ess as typical. For this CSA, employing computers with mor e powerful CPUs , 16 , 000 n e utrons are s tarted per generation. This analysis a lso u ses 8, 000 generations.

The problem posed b y Kiedrowski and Brown used a cadmium wrapper to isolate th e reacti ve locati o n. The full-pool-model employed in th e current ana ly s i s i s more neutronic a ll y coupled than the model u se d in the Kiedrowski and Brown problem so a t the curr e nt number of neutron s per ge n eratio n , so urc e convergence i s not a problem. Kiedrow s ki and Brown pointed out "s upp ose th e source s p ec ification is modified to incorporate only th e c e ntral sphere. Will thi s yie ld more reliable results? The answer is yes, and remarkably so. Eve n wit h a batch size of lk , the value ofk e ff is a l ways predict e d co rrectl y for each of th e 100 trials." In th e current NET-2809 1-000 3-01, Revision 0 88 analysis, when there is an isolated high reactivity area, the start source is specified at that location, thereby removing the convergence concern. Many articles are available relating to the topic of source convergence. This issue , however, is not of primary concern for this CSA, as it seeks to determine the k err, not the flux. In the Kiedrowski and Brown paper , they say , "'sloshing' of the fission source has an observable impact on k 0 rr .... however, the impact o n ke ff itself is small." This sloshing of the fission source increases the uncertainty in the calculated k err which is included in the bias and uncertainties. The l arge number of generations used in this CSA assures t he mean k err is correctly predicted within the uncerta i nty. To confirm that the full pool model is sufficiently coupled , six different start sources were used to analyze one of the final full-pool cases. The calculated k 0 rr is dominated by Region I. For cases including Region I , this CSA uses a start source that covered most of Region I. In order to challenge the convergence issue , the source was started at four boxes as far from the dominate k err area as possible. A final case was run with the SCALE default source , which is uniform over all fission materials in the problem. Figure 6.15 shows where the four start sources are located for the model. Table 6.6 shows the calculated k err values. All of the calculated k 0 rr values are within two standard deviations. The cases with the worst start source have a higher reported sigma. It is concluded that the neutrons per generation and t he number of generations used for the full models for this CSA is sufficient for convergence.

Figure 6.16 shows how the average k err behaves with progressing generations.

It is clear that the neutron population is moving away from the start source when the start sources are located in the comers of Region 2. All of the cases in this CSA are converged.

In some engineering applications, such as checking interface interaction , all that is needed is to determine that the location of the start source is not the most reactive location in the model. When this is the case , then convergence is not required. For example, in this case , it is clear that the three Region 2 start sources are not at the most reactive location in the model. NET-28091-0003-0l , Revision 0 89 This ana l ysis uses the k etr produced after a certain number of generations skipped. The number of generations to skip is determined by SCALE, whic h minimizes the standard deviation. It has been found that this almost always yields the best estimate of k. Table 6.6 shows that the number of ge nerations skipped can vary greatly. Clearly, for this convergence test , if a fixed number of generations to skip were used, the results would be le ss accurate.

Reviewing Figure 6.16 makes it clear that this SCALE feature is correct and needed if the start source is poorly se l ected. The differences in k err seen at the right hand side of Figure 6.16 are greater than that seen in Table 6.6 since the variation in the number of generations skipped is n ot part of the average k e tr shown on Figure 6.16. Table 6.6: kerr Changes With Start Source (Full Pool Model -8000 Generations, 16000 Neutrons per Generation)

Source Location Calculated k Reported Sigma Generations Skiooed Region I only 0.968708 0.00006 7 662 Uniform for Problem (Region l and 0.968724 0.000066 372 2) Region I Bottom Left 0.968704 0.000066 445 Top Right 0.968663 0.000066 525 Bottom Ri g ht to Cask Area 0.968809 0.000075 2066 Bottom Far Right Above Cask Area 0.968682 0.00006 8 1282 NET-28091-0003-01, Revision 0 90 F igure 6.15: Locations of the Start Sources for the Convergence Tests NET-28091-0003-01 , R ev ision 0 Lftl:NO O vo10 MT[AIM.2 -M T[Rl._.l -MAT O II.--, '4 O w ou:11..._a O M 1 u11V...1 1 -Ml£Rl(I,_

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  • Region 1
  • Uniform
  • Region 1 Left Bottom Top Right , Far R i ght Bott om
  • Bottom Right Cask A rea Figure 6.16: Change in Average kcrr with Progressing Generations NET-28091-0003-01 , Revision 0 92
6. 7 Summary of Modeling Assumption s The following is a summar y of the modeling assumptions
1. Bounding fuel stack density and nominal dimensions for pellet OD , clad OD/ID , and guide tube OD/ID are used. 2. Axial blankets are modeled for five different axial blanket designs. A conservative bumup profile for each blanket design is created by using the smallest relative power at each node from all assemblies having the same blanket design. 3. Grids are ignored (grids displace water between the fuel pins which causes k e rr to decrease).
4. No Boraflex TM in the Boraflex TM sheathing and the Boraflex Ž is replaced with water (if any Boraflex T M remained , it would be less reactive than water). 5. NUREG/CR-6801 bounding axial bumup profiles are used for all full-length fuel. To addre s s discontinuities , the shape in any bin is conservatively assumed to occur at the maximum bumup in the bin. For any bumup between points , the shape is linearly interpolated. 6. Top of fuel assembly models a plenum (length dependent on batch) filled with 8% volume fraction of stainless steel. Above the plenum is a 50/50 mixture of stainless steel and water to simulate the end fitting. Bottom reflector is 50 cm of water. 7. Periodic boundary conditions are used to represent an infinite array. 8. The dimensional changes with bumup are modeled as described in Section 6.3. a) In reference to clad creep , clad thickne ss i s con s tant, and creep occurs linearly , in two phases (creep inwards and growth ou twa rds). b) In reference to grid growth , pin pitch expand s uniformly.

The water temperature in Region 1 of the SFP is mode l ed at 180 °F which is the SFP design basis maximum temperature.

In Region 2 , the modeled temperature is 70 °C which is the temperature that maximizes k e rr in Region 2. NET-28091-0003-01, Revision 0 93 P roprieta r y In formatio n R e m ove d Westing h o u se N on-P ro pri e t ary C l ass 3 7 Sensitivity Analysis T hi s sec ti o n prese nt s a n a l ysis of th e se n sitivity of th e m o d e l s t o th e ma nu fac turin g t o l era n ces. After t h e se n s iti v i t y i s d e t e rmin e d , th e rac k up o f th e un ce rt a inti es a nd bia ses i s pr ese nt e d. 7.1 Manufacturing Tolerances Ca lculation s a r e p e rform e d to qu a nti fy th e reac ti v ity effec t of ch a n ges du e to m a nu fac turin g t o l era n ces. For R egio n 2, th e t o l era n ce ca lcul at i o n s are p erfo rm e d a t th e h ig h es t cre d i t ed bumup co n ditio n s (5 w/o at 49.5 G Wd/T, P F=l.4) and a l ow b urn e d co nd i ti o n (4.2 w/o at 30.0 GWd/T, 2 5 year coo lin g, P F=.6). Ta bl e 7.1 pr ese nt s t h e ca lcul ate d tol e r a n ce reac ti v iti es. Table 7.1: Tolerance Reactivity Effects Re2ion 1 1 2 2 2 Fuel Reactivity Cate2ory 2 3 4 4 5 Arran2ement 3 of 4 4 of 4 3 of 4 3 of 4 4 of 4 Enrichment (w/o) 5 5 4.2 5 5 Burnup (GWd/T) 20.5 28 30 49.5 60 Tolerance Calculated Ak (inches) (no Monte Carlo Uncertainty Adjustment)

Pellet Density +0.35% 0.0003 0.0004 0.0003 0.0005 0.0005 Pellet OD [ ] a.c 0.0002 0.0001 0.0001 0.0000 0.000 1 Clad ID [ ] a.c 0.0002 0.0002 0.0000 0.0001 0.000 2 Clad OD [ ] a.c 0.0008 0.0008 0.0009 0.0009 0.0007 Pin Pitch +0.0014 0.0028 0.0032 0.0016 0.0015 0.0013 Vertical Cell 0.00 32 0.0049 0.0016 0.0014 0.0015 Pitch --Horizontal 0.00 3 5 0.0050 0.0016 0.0014 0.0015 Cell Pitch --Wall Thick -0.007 0.0037 0.0043 0.0026 0.0025 0.0025 Cell ID 0.0002 0.0002 0.0003 0.000 2 0.0002 BoraflexTM Sheathing

-0.003 0.0016 0.0016 0.0010 0.0009 0.0010 Thickness RSS 0.0069 0.0090 0.0040 0.0038 0.0038 N E T-28 091-000 3-0 1 , R ev i s ion 0 94 The sign of the tolerance on Table 7.1 shows which direction increases

k. No tolerance reactivity effect is calculated for Category 1 fuel , the Category 3 tolerance is applied for Category 1. A checkerboard of Category 1 fuel (see Section 8.1) has a large margin to the criticality safety limit so an approximate tolerance is appropriate. PWR fuel assemblies are designed to be under moderated at power , so the moderator temperature coefficient is negative to prevent large power excursions.

Therefore, increasing water between the fuel rods (and ignoring grids) increases

k. This is demonstrated by calculations of the reactivity from varying the pin pitch and the fuel clad outer diameter (shown i n Table 7.1). The grids are conservatively ignored since they displace water around the fuel pins. The fuel pin pitch tolerance (0.0014 inch) used in this a nalysis is the maximum pin separation possible before the assembly gap becomes zero and all pins in the core are separated by a single enlarged pin pitch. The fuel enrichment used for determining if the loading requirements are met is the as-built enrichment for each assembly. The uncertainty in the as-built enrichment is +/-0.02%. Note that the uncertainty in the as built enrichment is less than the traditional uncertainty of 0.05% which is based on the nominal enrichment.

The reactivity of the fuel enrichment uncertainty is larger at low enrichments.

Calculations show that the reactivity due to enrichment uncertainty for Category 4 fuel is 0.0028 Llk at 2.0 w/o and is 0.0008 L'lk at 5.0 w/o. Fuel Categories 1 , 2, and 3 have fixed burnup requirements at an enrichment of 5.0 w/o , so there is no enrichment uncertainty.

For Category 4 fuel, the enrichment uncertainty reactivity effect is linearly interpolated using the two points 0.0028 L'lk at 2.0 w/o and 0.0008 L'lk at 5.0 w/o. For Category 5 fuel, the enrichment uncertainty at 5.0 w/o is used because Category 5 fuel has a fixed bumup differential which would have the least margin at 5.0 w/o. A tighter (smaller) rack cell pitch increases k etr of the SFP for both regions because the fuel assemblies are closer together.

The Region 1 change in ke rr for reducing the cell pitch is much larger due NET-28091-0003-01 , Revision 0 95 to the decrease in the flux trap. In ca l culating the reactivity effect of decreasing the cell ID , the cell pitch is maintained , so the effect on k e ff is small. A conservative minimum width of the Boraflex TM sheathing is used for the mode l. Calculations are performed where the sheathing width is increased from 7.7 to 8.0 inches in Region 1 and Region 2. The k e ff decreases by 0.0005 t.k and 0.0004 t.k for Regions 1 and 2 , respectively. Calculations show that the highest reactivity occurs with elevated temperatures (see Section 8.1). This is due to the water hole in the 2x2 model holding down reactivity. A higher temperature in the water hole results i n a lower water density and the reactivity hold down is reduced. For Region 2 , the peak reactivity occurs at a temperature of 70 °C. Above 70 °C , the reactivity begins to decrease because the reduced mo deration from lower density water within the fuel array is dominated by the reactivity hold down of the water in the water hole. For Region 1 , the reactivity increases with increasing temperature all of the way to boiling. Therefore , for Region 1 , the water temperature is modeled at 180 °F, the SFP design basis maximum temperature during normal operation. Boiling conditions are analyzed as an accident where soluble boron credit can be used. No calculations of tolerances or sensitivities were made with borated water. The borated conditions have excess margin , which covers any differences in sensitivity with borated water. 7.2 Burnup Dependent Biases and Uncertainties Bumup i ncreases the uncertainty in the analysis of k. To account for thi s there are se v eral biases and uncertainties that are bumup dependent.

They are the depletion uncertainty , the minor actinide and fission produ c t bias , the bumup uncertainty , the clad creep bias , and grid growth bias. The first b umup dependent bias or uncertainty is the depletion uncertainty in the atom densities.

This is accounted for by an uncertainty of 5% of the t.k between the zero bumup case and the case at the desired bumup. The 5% has been supported by a number of studies mentioned in Section 4 and is NET-28091-0003-01 , Revision 0 96 recommended via DSS-ISG-2010-01

[5]. For fuel Categories 2 , 3 , 4 , and 5, the zero burnup k e fffor the enrichment of interest is calculated and used with the calculated k e ff of the burned case to determine the worth of the depletion uncertainty.

Note that no bias is applicable. As an example, at 49.5 GWd/T, the delta-k of depletion is 0.3257 ~kin the 3-out-of-4 arrangement of Region 2 (Category 4 cell). This makes the depletion uncertainty 0.05 x 0.3257 ~k = 0.0163 t.k. The second burnup dependent bias or uncertainty is the minor actinid es and fission product worth bias. This bias, previously mentioned in Section 4, covers the bias and uncertainty due to the lack of criticality data for the minor actinides and fission products. This bia s is detennined by calculating ke rr in the appropriate model with the 1ninor actinide s and fission products remo v ed. The difference in reactivity between the calculations with and without these isotopes is multiplied by 1.5% and included as a bias. This approach was suggested in NUREG/CR-7109 and conservatively co v ers the uncertainty

[22). As an example, at 49.5 GWd/T for Category 4 fuel , a calculation determined that the actinide and fission product worth is 0.1371 ~k. Therefore the bias at 49.5 GWd/T is 0.0021 ~k. The third burnup dependent bias or uncertainty is the uncertainty in the declared burnup from the reactor records (shortened to burnup uncertainty). The burnup uncertainty from the reactor records is assumed to be 5% of the burnup [26]. This value is based on comparisons presented in Section 7 .2 of NUREG/CR-6998 [26] of in-core measured burnups that demonstrate that the uncertainty in assigned burnup values is less than 5%. The effect on reactivity is calculated by comparing the kerr calculated for the same case at two different burnups. For example, at 5.0 w/o the k e ffat 49.5 GWd/T i s 0.9560. The k e ff at 0.9x49 .5=44.55 GW d/T is 0.9813 (10% less burnup ). So the~ due to a 5% burn up uncertainty at 49.5 GWd/T is: (0.9813 -0.9560) I 2 = 0.0127 ~k The bumup uncertainty changes as a function ofbumup and enrichment because the delta-k between two low bumups can be larger than the delta-k between two high bumups. The burnup uncertainty is NET-28091-0003-01 , Revision 0 97 calculated for each loading curve point by the same procedure (us in g the L'lk between two different burnups with all other parameters stayi n g the same). As developed in Section 6.3.1, the clad outer diameter reduction (in microns) due to clad creep is a function ofburnup , starting at Omicrons at O GWd/T , lin early increasing to 100 microns at 40 GWd/T , then linearly decreasing to zero at 58 GWd/T. To de t ermine the reactivity effect at the maximum creep of 100 microns, a special depletion was run with the c l ad outer diameter reduced by 50 microns (the average reduction between O and 40 GWd/T). A case was then run at 40 GWd/T with the c lad OD reduced by 100 microns u sing the specia l depletion for the number densities. When compared to the nominal case at 40 GWd/T , the delta-k at 100 microns is 0.0012. The c l ad creep bias (L'lk) is expressed as a function of b urnup as fo ll ow s: C l ad creep bias (L'lk) = 0.00 1 2 x BU/40 = 0.0012 x (58-BU)/1 8 BU::S 40 BU>40 As can be seen in Table 7 .1 , the sensitivity to the clad OD is similar for all categories of fuel so the same clad creep bias formulation is used for all categories.

Finally, as discussed in Section 6.3.2 , the grid growth as a percent of the grid cross section is a function of burnup. The tolerance calcu l ation for pin pitch uses 0.0014 inch for the pin pitch to l erance which is 0.25% of the pitch. The grid growth bias is: Grid Growth Bias (L'lk) = (0.0000043 x BU 3 -0.00013 x BU 2 + 0.0051 x BU) x 4 x pp where pp= pin pitch tolerance worth from Tab l e 7.1. NET-28091-0003-01, R evision 0 98

7.3 Eccentricity

Generally the plant intends to place the fuel assembly in the center of the ce ll. However , it is acceptab l e to have the assembly in any lo cation with in the cell. A study performed for the Millstone 2 license application showed for that plant , the placements were approximately random [ 49]. For this 2017 CSA, i t is assumed that the placement of the assemblies in the cells is random (there is nothing in the cell that would cause the assembly to be preferen t ially placed in one comer over another). As was performed for Millstone 2, the number of assemblies that is eccentrica ll y placed in particular quadrants is determined such that the probability of such placement is l ess than 5% over the lifetime of the plant. Region 2 contains a 3-out-of-4 set of Category 4 ce ll s surrounded by Category 5 cells in a 4-out-of-4 arrangement on the outer two rows and a 4-out-of-4 arrangement with checkerboarded control rods. To determine an eccentricity bias for the Category 4 ce ll arrangement, 16 assemblies are placed as c l ose as possible (u sing the standard pin pitch) to a central assemb l y. Figure 7.1 shows the placement of th e assemb li es. The proba bilit y of 16 assemb l ies being randomly p l aced in the most reactive quadrant is 0.25 1 6 = 2.3*E-10. There are 504 Category 4 cell locations. It is conservative to estimate that eac h l ocation cou l d have 100 moves. This makes the probabi li ty of getting such an arrangement 2.3*E-10*504*100=

1.2E-5 which is much less than t h e required 0.05. NET-2809 1-0 003-01 , Revision 0 99

, Figure 7.1: Category 4 Region 2 with 16 Assemblies Eccentrically Placed The model used is an 8x8 model with periodic boundary cond iti ons , where a ll assemblies except the central 16 eccentric assemblies about a central assembly are centered. This means that there are actually about eight eccentric sets separated by two rows of centered assemb li es. The enrichment and bumup used for the fue l is 5.0 w/o and 48.19 GWd/T. Table 7.2 s ho ws the results of the ana l ysis. The refere nc e calculation i s th e 8x8 model with centered assemb li es. The ecce ntri c it y bias for Region 2 is n egat i ve and therefore conservatively i gnored. This i s not unexpected s in ce wa t er hol es or contro l s rods can break up the effect of eccentricity.

NET-28091-0003-01, Revision 0 1 00 Table 7.2: Eccentricity Results Category Calculated k Sigma Ak Reference 4 0.9586 0.00004 16 Eccentric Assemblies 4 0.9585 0.00004 -0.0001 Reference 2 0.9671 0.00007 16 Eccentric Assemblies 2 0.9690 0.00006 0.0019 16 Eccentric Assemblies Shifted down 2 0.9693 0.00007 0.0022 The bumup increment for the Category 5 fuel is determined such that Category 5 fuel is not limiting (see Section 8.4.2). Therefore, the eccentricity is maximized by moving Category 5 fuel closer to the C ategory 4 cells. In the final full pool model all of the Category 5 fuel assemblies in Region 2 are moved right in their cells to be as close as possible to the left hand side of Category 4 cells. Region 1 is more complicated. First, a checkerboard area with Category 1 fuel is designed to have a low k eff-The keff is sufficiently low (0.8548) that eccentricity within Category 1 cells i s not a concern. Category 2 fuel eccentricity is analyzed with 16 assemblies placed closest to a central assembly in the same manner as was performed for Category 4 (Region 2). Due to the flux trap, however , it is possible that it is more reactive to mov e the central assembly from the ce nter of the cell toward one of the four sides. Therefore , analysis was also performed where the central row of assemblies is moved down to be closer to the row of assemblies below it. Figure 7.2 shows this arrangement of fuel assemblies.

The results of the analysis for Category 2 are shown on Table 7.2. Category 2 is showing an eccentricity bias and the bias is slightly lar ger when the central row is moved down. The full pool mod el incorporates the eccentric placement of the assemblies, so the bia s is intrinsic to the analysis.

Therefore, it does not need to be added to the final calculated

k. As in Region 2, the bumup penalty for th e outer two rows of Region 1 (Category 3 cells) was selected to prevent Category 3 fuel from being more limiting.

In the final full pool model, both rows of Category 3 fuel assemblies are mo ve d down toward the Category 2 ( or 1) cells to maximize the NET-28091-0003-01, Revision 0 101 eccentricity

/interface effect. Similarly the Category 5 fuel assem b lies at the Reg i on 1/Region 2 i nterface are moved up and to the l eft to make them as close as possible to Category 2 ce ll s. ::.::.:::.::.::

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I ~:_;;;:::;;;_;;;;L_ I f F i g ur e 7.2: Ecc en tr i c M od e l fo r Categ or y 2 w i t h Ce ntr a l Ro w S hi fted D o w n 7.4 Additional Biases and Uncertainties The criticality validation for major actinides, absor b ers, and structural materials provides a bias and uncertainty due to validation.

As provided in Section 4 , the validation bias and uncertainty for a ll calculations except those with very hard spectra s uch as borated cases or boiling cases are 0.0021 Lik and NET-28091-0003-01 , Rev i sion 0 102 0.0087 t.k respectively (for fresh fuel the bias and uncertainty is less as given in Section 4.3). For cases with a hard spectrum (EALF greater than 0.4 eV) the bias and uncertainty are 0.0027 t.k and 0.0112 t.k respectively.

The criticality validation also revealed the need for an additional bias and uncertainty for conditions above room temperature.

Due to using water holes to control reactivity, it was found that elevated t emperatures have a higher reactivity.

For Region 2, the highest reactivity is at a temperature of 70 °C (see Section 8.7). For Region 1, the highest reactivity is at the SFP design basis maximum temperature which is 180 °F (82 °C). From the validation section, the temperature bias at 70 °C is 0.00043 t.k with an uncertainty of 0.0013 t.k while the temperature bias at 82 °C is 0.00053 t.k with an uncertainty of 0.0013 t.k. A 2x2 model cannot provide adequate modeling of eccentric placement of fuel assemblies in rack cells. Therefore, a bias is required.

This bias is detennined in calculations performed in Section 7.3. It was found that eccentric loading of Category 4 cells (3-out-of-4 of Region 2) does not increase reactivity , so there is no eccentricity bias for Category 4 fuel. The Category 5 fuel bumup penalty, which is derived in Section 8, includes eccentric positioning in the large model, so it is inherently included in the bumup penalty. However, Category 2 (3-out-of-4 in the Region l flux trap design) does have an eccentricity bias of0.0022 t.k (see Section 7.3). The final additional bias and uncertainty is the reactivity effect due to Monte Carlo statistical uncertainty.

The 95/95 Monte Carlo statistical uncertainty in each tolerance calculation is where <:rb and crp are the Monte Carlo standard deviations for the base case and the perturbed case, respectively.

The base case calculation was run with 1.024 billion histories to reduce the statistical uncertainty to+/- 0.00002 t.k (1 sigma) for the base case. The perturbed calculations are run for 64 million NET-28091-0003-01, Revision 0 103 histories for an uncertainty of+/- 0.00008 ~k ( l sigma). This makes the Monte Carlo standard deviations

((2*0.00 0 02)2 + (2*0.00008) 2 )05 = 0.000 1 6 ~k. In t h e statistical combination of terms each of t h ese would be squared. Since there are 13 statistical to l erance components the sum of the se terms would be 13*(0.00016)2. The fina l step of the statistical comb i nation is taking the square root. The square root of 13*(0.00016)2 i s 0.0006 ~k. So , 0.0006 ~k is the Monte Carlo statistical uncertainty in the tolerance calculations and is combined in the total rack up of u n certainties.

7.5 Biases

and Uncertainties Rack-up Sections 7.1 through 7.3 pro vi des the biase s and uncertainties and in the case of the bumup dependent bia ses and uncertainties , how to calculate them. For Region 1 , the fuel categories all ha ve fixed bumup and are va li d up to 5.0 w/o enrichment.

Therefore , the final bias and uncertainty can be determined. Table 7.3 p rovides the total rack up of biases and uncertaintie s, inc l uding the statist i cal combination of the uncertainties for Region 1 (fuel Categories 1 , 2, a n d 3). NET-2809 1-0003-01, R evision 0 104 Table 7.3: Total Bias and Uncertainties for Region 1, Categories 1, 2, 3 Category 1 2 3 Arrangement 2 of 4 3 of 4 4 of 4 Enrichment (w/o) 5 5 5 Burnup (GWd/T) 0 21 28.5 Component Bias Uncertainty Bias Uncertainty Bias Uncertainty Validation ( critical 0.0024 0.0035 0.0021 0.0087 0.0021 0.0087 experiments)

Depletion uncertainty

---0.0067 -0.0092 Minor Actinides and Fission 0.0010 0.0013 Products Bias --Bumup uncert ainty ---0.0057 -0.0081 Manufacturing Tolerances

-0.0090 -0.0069 -0.0090 Enrichment Uncertainty

Monte Carlo statistics

-0.0006 -0.0006 -0.0006 Ecce ntricity Bias 0.0 -0.0022 -o.o* -Clad creep bia s --0.0006 -0.0009 -Grid growth bi as --0.0010 -0.0018 -Elevated temperature 0.0005 0.0013 0.0005 0.0013 0.0005 0.00 1 3 Total Rack Up (~k = RSS) 0.0029 0.0098 0.0074 0.0142 0.0066 0.0176 Sum of Bias and 0.0127 0.0217 0.0242 Uncertainties

  • Category 3 cells are located on the outer two rows of Region 1 so eccentricity in an infinite model is not relevant.

The eccentricity is part of the full pool model and therefore no bias is applied. For the two categories of fuel in Region 2, the burnup is allowed to change so the bias and unc ertainty will change. However, Table 7.4 is provided as an example of the total rack up for Category 4 and 5. NET-28091-0003-01, Revision 0 105 Table 7.4: Sample Categor y 4 and 5 Bias and Uncertainty Rack-up Cate2orv 4 5 Arrangement 3 of 4 4 of 4 Enrichment (w/o) 5 5 Burnup (GWd/T) 49.5 60.5 Component Bias Uncertainty Bias Uncertainty Va lid ation (critical experiments) 0.0021 0.0087 0.002 1 0.0087 D e pl etion uncertainty

-0.0163 -0.02 14 Minor Actinides an d Fission Products Bias 0.0021 0.0026 Burnup un certainty

-0.0124 -0.0 155 Manu facturi n g Tolerances

-0.0040 -0.00 40 E nri c hm ent Uncertainty

-0.0008 -0.0008 Mont e Carlo stat i s tic s -0.0006 -0.0006 Eccentricity Bias 0.0 -o.o* -Cla d creep bias 0.0006 -0.0000 -Gri d growth bias 0.0029 -0.0041 -E l evated temperature 0.0004 0.0013 0.000 4 0.0013 Total Rack U p (~k = RSS) 0.0081 0.0227 0.0092 0.0282 Sum of Bias and Uncertainties 0.0308 0.0374

  • Catego r y 5 ce ll s are loc a ted on the outer t wo rows of R egion 2 so ecce ntr icity in an infin ite mod e l i s not relevant.

The eccentricity i s part of the full pool model and therefore n o bia s i s app li ed. Fresh 5.0 w/o fue l with 64 IFBA (and any burned fuel) can a l so be stored in a c h eckerboard pattern of assemb li es a nd water holes in Region 2 and can be stored in the 3-o u t-of-4 area of Region 2 if it contain s a contro l rod. The rack up of uncertainties that app li es to fresh 5 .0 w/o fuel in R eg ion 2 is s hown in Table 7.5. Table 7.5: Total Bias and Uncertainty for Fresh Fuel in Region 2 Component Bias Uncertainty Validation (cri tic a l experime nt s) 0.0024 0.0035 Manufacturing Tolerances 0.00 40 F uel enric hment -Monte Carlo statistics 0.0006 E le vated temperature 0.0004 0.0013 Total Rack Up (~k = RSS) 0.0028 0.0055 Sum of Bias and Uncertainties 0.0083 N ET-28091-0003-01, Revi s ion 0 106 None of the above to l erance calculations were performed under borated conditions.

The reason is because t h e borated condition has significant margi n. The boron di l ution analysis of record shows that a dilution down to 786 ppm is not credib l e [52]. However , the mini m um ppm selected for the borated analysis is 700 ppm. Any small increase in the tolerance uncertainties would be covered by this 86 ppm margin in addition to the l arge margin from 0.9 5 reported in Section 8.5. 7.6 Interface Uncertainty Treatment When analyzing a fu ll pool , the calculated k e ff wi ll come from the most reactive area of the SFP. However, when the uncertainty is not t h e same in all areas , the analysis may not correctly find the most l imiting k. In order to address this concern, the bumup of the high burnup regions are adjusted down to account for t h e difference i n the uncerta in ty. Tables 7.3 and 7.4 show the uncertainties for t h e 3-out-of-4 and the 4-out-of-4 areas i n Regions 1 and 2. For Category 3 fuel to match the bias and uncerta i nty of Category 2 fuel, a burnup reduction is required to matc h the 0.0048 6k difference in bias and uncertainty. The 0.0047 6k is the sum of0.0025 6k between Category 2 and Category 3 fuel (see bottom of Table 7.3) plus the 0.0022 6k eccentric i ty effect for Category 2 fuel which wi ll be included in the full poo l model. To account for thi s additional uncertainty , the burnup for Category 3 fuel is decreased by 0.8 GWd/T in the full pool calculations (the reactivity due to burnup at 28.5 GWd/T is 0.6% in kefffor every 1 GWd/T burnup ). Similarly for analysis of Region 2, the Category 5 fuel bumup must be decreased to match the reactivity d i fference in the bias and uncertainty between Category 4 and Category 5 fuel. From Table 7.4 this difference is 0.0066 6k. To account for this additional uncertainty , the burnup for Category 5 fuel is decreased by 1.3 GWd/T in the full pool calculations (t h e reactivity change due to burnup at 60.5 GWd ff is 0.51 % in k e ff for every 1 GWdff bumup). NET-2809 1-0003-01, Revision 0 107 8 Results With the biases and uncertainties determined, the minimum loading requirements can be ca l cu l ated. These minimum loading requirements meet the 10CFR50.68 requirements. Specifically , k 95;9 5 must be less than 1.0 with no soluble boron credit and less than 0.95 with credit for soluble boron. For this analysis, these limits are met while maintaining about a 1 % margin in k. It has been demonstrated that for all unborated cases k 95;95 is less than 0.99 and for the borated cases k 95;95 is less than 0.94 after adding biases and uncertainties. 8. 1 Temperature Effects The criticality analysis must cover the full range of temperatures allowed in the SFP. Rather than perform the criticality ana l ysis at a reference temperature and add a bias , the criticality analysis is perfonned at the most limiting temperatures.

Table 8.1 summarizes the Region 1 and 2 (3-out-of-4 area) calculations at 12 different temperatures (4, 10 , 20, 30 , 40, 50, 60, 70 , 80 , 90 , 95 , and 99 °C). From the SCALE validation , there is a temperature bias of 0.0000086 for each °C above 20 °C. For example, the bias at 60 °C is 0.00034, while the bias at 70 °C is 0.00043 , and the appropriate temperature bias is added to the calculated k etr values. The results demonstrate that the bias-corrected reactivity is largest at 70 °C for Region 2 and at 99 °C for Region 1 under unborated conditions.

Except for Table 8.1 and the temperature accident, a ll ca l culated k e tr va l ues for Region 1 are performed at 180 °F (82 °C), which corresponds to the SFP design basis maximum temperature for the IP2 SFP. In developing the loading curves for Region 2 , all of the calculated k etr values are performed at 70 ° C which is the most reactive temperature.

NET-2809 1-0003-01, Revision 0 108 Ta bl e 8.1: Ca lcu lated keff as a Funct io n of Tem p erat u re Te mp e r a tu re D e n s i ty R eg ion 1 R egio n 1 R egio n 2 R eg i o n 2 (OC) (glee) ealc. k a dj. k ealc. k ad j. k 4 1.0000 0.96386 0.96386 0.95569 0.95569 10 0.9997 0.96388 0.96388 0.95575 0.95575 2 0 0.9982 0.96372 0.96372 0.95545 0.95545 30 0.9957 0.96508 0.965 17 0.95582 0.95591 4 0 0.9922 0.96628 0.966 46 0.95595 0.95612 5 0 0.9880 0.96745 0.96771 0.95610 0.95636 6 0 0.9832 0.968 38 0.96873 0.95606 0.95641 70 0.9778 0.9693 1 0.9697 4 0.95599 0.95642 80 0.9718 0.97000 0.97 052 0.95577 0.95629 90 0.9653 0.97089 0.97149 0.95564 0.95624 95 0.9619 0.97123 0.97187 0.95552 0.95616 99 0.9591 0.97156 0.97224 0.95544 0.95612 8.2 Region 1 Fuel Categories 1 and 2 Using the 2x2 mode l described in Section 6 with the a tom densities de velo ped in Section 5 and with the bias and uncertainties established in Section 7, the l oading requirements for Region 1 fuel Categories 1 and 2 are determined. Table 8.2 shows the fuel requirements and the calculated k -for a checkerboard arrangement of fuel in Region 1 (Category 1 cells) and a three out of four arrangement of fuel in Region 1 (Category 2 cells). Category 1 fuel is designed to require 64 IFBA rods in 5.0 w/o fuel assemblies.

To provide flexibility in fue l design , the number of IFBA for fresh fue l can be reduced for lower enrichments

  • . The number of minimum IFBA for fuel less than or equal to 5.0 , 4.5 , 4.0, 3.5, and 3.0 w/o is 64, 48 , 32, 16 , and 0 , respective l y. The k e fffor these cases are all les s than the k e ffwith 64 IFBA 5.0 w/o. No credit for burnup is taken. For Category 2 fuel , the fuel must be burned at l east 21 GW d/T and the maximum enrichment is 5.0 w/o. The Category 2 bumup requirement of 21 GW d/T is based on Batch Z fuel having an 8 inc h 4.0 w/o U-235 axial blanket where the relative burnup distribution is from Tab l e 6.3. An axia l blanket that is l ess than 4.0 w/o U-235 or more than 8 inches long is bounded by this ana l ysis, but a shorter or higher enriched blanket is not.
  • The IFBA requirement is only for Batch Z. NET-2809 1-0003-01, Revision 0 109 Table 8.2: Confirmation of Region 1 Requirements for Categor y 1 and 2 Fuel Fuel Catee.or v 1 2 Arrane.ement 2-out-of-4 3-out-of-4 Maximum Enrichment (w/o) 5.0 5.0 Minimum Burnup (GWdrf) 0 21 Minimum IFB A Rod s 64 -Calculated k 0.8548 0.9686 Bias and U ncertain ty 0.0127 0.0217 k 9S/95 0.8675 0.9903* For burn ed fuel, no c r e dit is take n for IFBA or a n y in sert in the guide tubes , w ith th e excep ti o n of full l e ngth RCCAs in designated areas. Ca lcul ations show th at 5.0 w/o fuel with 64 IFBA rods h as l ower reactivity at a ll bumups compared to th e BOC eq uilibriu m Xe val u e. Table 8.3 s ho ws the va lu es of kerrin t he core geometry as a function of bumup for fue l w ith various IFBA l oadings. T h e orange s h a d e d bl ocks are the bumups where kerr ha s increased over its initial eq uilib ri um va lu e. The 64 IFBA case is a lw ays l ess than the initial kerr(w ith equilibri um Xe) of 1.2109. Table 8.3: Change in k ett with Burn up and number oflFBA Rods (Analysis performed at core co ndi tions for 5.0 w/o fuel) N umber of IFBA Rods+ 0 32 64 80 116 Burnup (GWd/T}. Calculated keff at core conditions 0.15 1.2989 1.2526 1.2109 1.1924 1.1 523 0.50 1.2925 1.2489 1.2086 1.1903 1.151 3 1.00 1.2855 1.24 41 1.2064 1.1890 1.1 526 1.50 1.280 1 1.2417 1.2069 1.1905 1.1 5 5 9 2.00 1.2757 1.24 03 1.2068 1.1913 1.1 593 3.00 1.268 4 1.2360 1.2062 1.1930 1.1 632 4.00 1.252 1 1.2249 1.1 992 1.1874 1.1 6 1 0 5.00 1.2423 1.2 1 84 1.1963 1.1848 1.1 625 6.00 1.23 1 9 1.2 11 7 1.1 929 1.1827 1.1 631 8.00 1.2 1 67 1.2013 1.1 850 1.1782 1.1 609 10.00 1.1974 1.1 862 1.1 75 0 1.1693 1.15 63
  • The infinite k err exceeds 0.99 but the ac tu a l finite k err for thi s r eg ion i s l ess th an 0.99 (0.9881). NET-2809 1-0003-01, Re v ision 0 110 Reactivity decreases with increasing category number. Therefore, fuel from any higher numbered category can be placed in any location that allows for a lower numbered category.

For examp l e, a fuel assembly categorized as Category 5 can be placed anywhere in the SFP. However, a cell in the SFP that requires Category 5 fuel may not contain a lower category fuel assembly. All of the historical fuel through Batch X of IP2 and Batch AA of IP3 has been categorized.

8.3 Region

2 Category 4 Batch Grouping Z -Current and Future Fuel The minimum burnup requirements (loading curve) for Category 4 fuel for Batch Grouping Z (current design and future fuel assemblies) are presented in Table 8.4. The SFP cells where Category 4 (or above) fuel is required is shown on Figure 1.1 as the green shaded cells in Region 2. The other batch groupings are analyzed separately, and the results are presented in Section 8.6. Table 8.4: Minimum Burnup Requirements (GWd/T) for Category 4 Batch Grouping Z Enrichment

  • Cooling Time (years) PF=l.2 (w/o) o t 1 2 5 10 15 25 t 4.2 40.27 39.69 38.92 37.23 35.13 33.75 32.20 4.6 44.27 43.60 42.83 40.9 9 38.71 37.25 35.52 5.0 48.19 47.52 46.61 44.67 42.30 40.71 38.85 Enrichment
  • Cooling Time (years) PF=0.80 (w/o) ot 1 2 5 10 15 25t 4.2 38.67 38.11 37.48 36.02 34.12 32.96 31.59 4.6 42.60 41.97 41.31 39.70 37.72 36.45 34.90 5.0 46.52 45.8 4 45.16 43.39 41.14 39.77 38.05 Table 8.4 provides the burnup requirement in GWd/T as a function of initial U-235 enrichment and cooling time for two different peaking factors. For each assembly, the peaking factor i s known, and the
  • The enrichment to be used is the enrichment of the center section between the blanket material.

t O years coo lin g is actua ll y 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This i s the cooling time that maximize s k. t Fue l coo l ed to more than 25 years must use the 25 year burnup requirement.

NET-28091-0003-01 , Revision 0 111 bumup requirement for that assembly can be interpo la ted bet w een 0.8 0 and 1.20. O ver 95% of the fuel inventory ha s peaking factors between 0.80 and 1.2 0. Extrapolation abo ve 1.20 and below 0.8 is also acceptable because this has been shown to be conservative (see Section 6.4). After adjusting for an assembly's peaking factor, if an assembly fails the lo ad in g curve, it can b e stored anywhere in Region I as long as the bumup is greater than 28.5 GWd/T. Table 8.4 can be linearl y interpolated to find the re quired bumup at any enrichment

/cooling time combination but it is recommended that the curve fit be used instead, as described in Section 8.2.1. As discussed lat er, the bumup requir e ments are adjusted if the assemb l y contained a h a fnium insert or has any fuel pins remo ve d. Fresh assemblie s with at least 64 IFBA rods and an inserted control rod are Category 4. Analysis of an infinite system of Region 2 cells in a 3-out-of-4 arrangement using 5 .0 w/o fresh fuel with 64 IFBA rods and a control rod in every assembly produced a k e ff of 0.96 03. To be conservative , the control rod a tom densities were reduced by 10%. The bias and un certainty for this case is 0.0083 (see Tab le 7.5), so k 9s 1 9s is 0.9686, which is well below 0.99. To provide flexibility in fuel de sign, the number of IFBA for fresh fuel can be reduced for lower enrichments.

The number of minimum IFBA for fuel less than or equal to 5.0 , 4.5 , 4.0 , 3.5, and 3.0 w/o is 64 , 48 , 32, 16 , and 0 , respectively. The k e ff for these cases are all l ess than the k eff with 64 IFBA. As discussed earlier, the reactivity of fuel with 64 IFBA decreases with bumup so a fresh assemb l y is more limiting than the same assembly havin g a sma ll amount ofbumup. NET-28091-0003-01 , Revision 0 112

8.3.1 Curve

Fit The data points of Table 8.4 have b een fit with a nine parameter curve having the following form: Minimum Bumup Requirement w h ere E = U-235 initial enrichment (w/o) CT = cooling time (years) a 1 -a 9 = fitting coefficients No extrapolation is a ll owed , so fuel at enrichments l ess than 4.2 w/o must use 4.2 for the enric hm ent , and fuel cooled more t h an 25 years must use 25 for the cooling time. The cur v e is purposefully conservative in that the minimum burnup requiremen t generated from th e curve is always equa l to or greater than the bumup s ho wn in Tab l e 8.4. The coefficients are s h own in Table 8.5. Severa l curve fits were attempted but this curve fit matched the data with the least amount of conservatism while being we ll b e haved b etween the data points of Table 8.4. A spreadsheet for the fit was created to en s ure the int ermediate points follow the expected behavior.

The exponentia l term in the fit is needed to mimic the physics of radioactive decay. Table 8.5: Curve Fit Coefficients for Category 4 Fuel (Gro up Z -Current a nd Future Fue l) Coefficient PF= o.so* PF= 1.20 a1 15.1405 -6.26824 az -4.8 11 33 5.29367 a3 0.753855 -0.37154 a4 0.121 252 0.129582 as -0.0150991

-0.0204918 a6 0.00 1 27009 0.00 205596 a 1 -16.2293 -0.1 3331 as 14.0159 6.9037 a9 -0.687 054 0.122068

  • Only two peaking factors are needed (0.8 and 1.2). Ca l c ul ations show that e x trapo l at i o n belo w 0.8 a nd abo v e 1.2 is conserva ti ve for all ot h er peaking facto r s. NET-2809 1-000 3-01 , Revision 0 113

8.3.2 Confirmati

o n Calculations for Category 4 To e n sure that all bumup/enrich m ent/cooling time combinations given in the loadin g curve meet the criticality requirements , each loading curve bumup/enrichment

/coo l ing time point was run in the 2x2 KENO model to verify that each point meets the criticality requirements. The calculated keff va lu es are shown in Table 8.6. Table 8.6: Calculated k cff Values at each Category 4 Batch Z Burnup Point Enrichment Cooling Time (years) PF=l.20 (w/o) 0 1 2 5 10 15 25 4.2 0.9614 0.9611 0.9613 0.9606 0.9608 0.9608 0.9604 4.6 0.9600 0.9599 0.9597 0.9592 0.9594 0.9593 0.9590 5.0 0.95 86 0.9581 0.9585 0.9579 0.9577 0.9576 0.9 573 Enrichment Cooling Time (years) PF=0.80 (w/o) 0 1 2 5 10 15 25 4.2 0.96 12 0.961 3 0.9611 0.9605 0.96 11 0.960 8 0.9602 4.6 0.9599 0.9600 0.9596 0.9591 0.9590 0.95 88 0.9585 5.0 0.9584 0.9585 0.9578 0.9577 0.9584 0.9579 0.9579 The total un certainty i s the bias plus a statistica l combination of a ll of the unc ertainties (see Section 7). These total un certainties are shown in Table 8.7. Table 8.7: Total Bias and Uncertainty at each Category 4 Batch Z Burnup Point Enrichment PF=l.20 w/o 0 1 10 15 25 4.2 0.0278 0.0277 0.0277 0.0274 0.0273 0.0272 4.6 0.0292 0.0 292 0.0 293 0.0292 0.0291 0.0289 0.0288 5.0 0.0307 0.030 8 0.0307 0.0306 0.0306 0.0306 0.0304 Enrichment PF=0.80 w/o 0 1 5 10 15 25 4.2 0.0277 0.0276 0.0276 0.0275 0.0274 0.0273 0.0272 4.6 0.0292 0.0292 0.0293 0.0293 0.0292 0.0290 0.0290 5.0 0.0307 0.0307 0.0308 0.0307 0.0306 0.0306 0.0 305 NET-2809 1-000 3-01, R evis ion 0 114 A ft e r addin g t h e t ot a l un certai nt y t o th e calc ulat e d ke ff va lu es , a ll p oints are l ess th a n 0.99 a s s h ow n in Tab l e 8.8. Table 8.8: k9 s t 9s for each Category 4 Batch Z Burnup Point Enrichment Coolin!?:

Time (vears* PF=l.20 (w/o) 0 1 2 5 10 15 25 4.2 0.9892* 0.9889 0.9 8 90 0.9 882 0.9882 0.988 1 0.9 876 4.6 0.9892 0.9892 0.9 8 90 0.9 88 5 0.988 5 0.9882 0.9 877 5.0 0.9893 0.9 889 0.9 8 9 2 0.9 88 5 0.9883 0.9882 0.9 877 Enrichment Coolin!?:

Time (vears\ PF=0.80 (w/o) 0 1 2 5 10 15 25 4.2 0.9889 0.9 889 0.9887 0.988 0 0.988 4 0.9880 0.98 7 5 4.6 0.989 1 0.9892 0.9889 0.988 4 0.9882 0.9879 0.9875 5.0 0.989 1 0.989 1 0.9886 0.9 883 0.989 0 0.9 886 0.9 885 8.4 Determination of Burnup Requirements for Categories 3 and 5 Ce ll C at egories 3 a nd 5 util ize th e n e ut ron l ea k age a t th e e d ge o f th e SF P in o rd e r t o re m ove th e need fo r o n e out of four wa t er hol es use d fo r ce ll Catego ri es 2 a nd 4. S in ce th e e d ge of th e SF P n e utron l eakage i s u se d , a fu ll po o l m o d e l is r e quir e d. T hi s full p oo l m o d e l al so a ll ows for ca lcu lat i o n o f th e ecce n t ri c it y effect a nd th e i m p act of th e int e r face b e t wee n ce ll Ca te go ri es an d R egio n s. 8.4. 1 Cell Category Layout in Region 2 T h e c on ce p t fo r Re g ion 2 i s to h ave a thr ee o ut of four arra n ge m e nt of fue l w ith t wo o ut e r ro ws of fuel r e qu i rin g a n increase d bu rn u p bu t u s in g t he l ea k age at th e e d ge t o r e du ce th e in c r ease d burnup re quir e m e nt. H oweve r , IP2 h a d a se t o f s p are co ntrol rod s th a t co uld b e u se d t o re du ce t h e numb e r of wa t er hol es s u c h th at o n e l ess cas k wo uld n ee d to b e lo a d e d. T h e c ontrol ro d s in a n assem bl y do n o t h ave as l arge a n egat i ve rea cti v it y as a wa t e r h o l e so th e control rods n ee d e d t o b e t wo o ut of fo ur , rath e r th a n

  • T h e va lu es in Tab l e 8.8 d o not a l ways m a t ch th e s um of Ta bl es 8.6 a nd 8.7 due to ro und off , s in ce eac h t a bl e was d eve l o p e d u s in g m o re s i g nifi cant di g it s b efo r e ro undin g fo r th e ta bl e. N E T-2 8 091-000 3-01 , R ev i s i o n 0 115 one out of four. Rather than make a separate fuel category for the cell area with control rods, this cell area is forced to be the same category as the two outer rows of assemblies ( cell Category 5). Because the checkerboard of control rods reduced k err more than the two rows on the outside of the SFP , it is possible t o reduce the amount of water required in the water ho l es. This is the reason for the l ocation of the pink cell s. Determining the location of the pink cell s and the control rods required con s iderable iteration. The control rod locations can be water holes since the negative reactivity of a water hole is greater t han a Category 5 fue l assembly with a control rod. The control rod may not be placed in or removed while the assembly is in the control rod location. The control rod must be inserted into the assembly while the assembly is in a cell not requiring a control rod, and then the assembly with control rod can be mo v ed into position. Likewise, the control rod may be removed only while the assembl y is in a cell not requiring a control rod. It is permissib l e to remove a control rod at a Category 5 cell as long as all adjacent ce ll s (eight ce ll s) are water ho l es since this iso l ates the assembly , but this is not the expected method. Category 5 fuel must be placed on both sides of the Region 1/2 interface except in SFP locations J-31 and H-31 (alternate arrangement s are allowed as discussed in Section 8.5). This eliminates the interaction between the two Regions. Placing Category 4 fuel on the Region 1 side of the interface was tried , but it drew the reactivity to the interface w i th a s li ght increase in k err and, therefore , was rejected.

8.4.2 Additional

Burnup Requirements for Fuel Categories 3 and 5 The objectives in setting the loading requirements for Categories 3 and 5 are fir s t to make Categories 2 and 4 more limiting than Categories 3 and 5 , and second to make them simple. The simplicity is accomplished by making the l oading requirements a constant bumup penalty (independent of enrichment, cooling time and peaking factor) to the fue l Category 2 and 4 requirements. The bumup penalty is the smallest fraction of the bumup when the bumup is highest. Therefore , the bumup penalty is determined using the highest bumup requirement s in Category 2 and 4 so it is conservative for lower NET-28091-0003-01 , Revision 0 116 bumups. To find the bumup penalty, the Category 3 and 5 bumups were changed holding the Category 2 and 4 bumups constant.

Since the design objective k e ff for Region 1 is higher than that for Region 2 ( due to lower bumup requirements), the Region 2 analysis must be performed with the fuel removed from Region 1, so that the reactivity is dominated by Region 2. The Region 1 analysis contains all of the fuel from both regions. Therefore , the Region 2 analysis is perfom1ed first so that the correct bumups in Region 2 are used while doing the Region 1 analysis.

To determine the additional bumup requirement needed for Category 5 fuel , analysis wa s performed for Region 2 where the Category 4 bumup is 49.5 GWd/T and the Category 5 bumup is varied. For these cases the enrichment for both categories is 5.0 w/o. For this model , all of the Category 5 fuel is shifted right to maximize the effect of the Category 4/5 interface.

Figure 8.1 is a plot of the variatio n in k e ff with the Category 5 bumup. As can be seen from Figure 8.1 at lower bumups (such as 54 GWd/T) the Category 5 fuel is determining

k. By 59 GW d/T, k e ff is controlled mainly by the Category 4 fuel and by 61 GWd/T the Category 5 fuel is no longer affecting the SFP k. Based on this information the increased burnup requirement (burnup penalty) for Category 5 fuel is selected as 11 GWd/T. NET-28091-0003-01 , Revision 0 117

"" "C ., 0.980 0.975 0.970 0.965 u a 0.960 0.955 0.950 + 54 55 56 57 58 59 60 61 Burnup of Category 5 Fuel (GWd/MTU)

Figure 8.1: Calculated k err as a Function of Category 5 Burn up Using 5.0 w/o Fuel 62 In or d er to confirm that using the burnup penalty d etermined from the highest bumup is co n servative , the analysis was also performed using a l ower bumup for Category 4 fuel. This model uses 4.2 w/o enriche d fuel , 32 GWd/T for the bumup (25 years coo l ed and a 0.8 p eaking factor). F i gure 8.2 s ho ws the results of thi s analy s is. As can be seen from Figure 8.2 , by a burnup of39 GWd/T the burnup of Category 5 n o l onger matters. This wou ld imply a penalty of7 GWd/T which mea n s that using the penalty detennined at the higher bumup (11 GWd/T) i s indeed conservati v e. NET-28091-0003-01, R ev ision 0 11 8 1.000 ----0.995 ---0.990 0.985 .:..: 0.980 ,:, ... 0.975 :::, ... iii u 0.970 0.965 0.960 F ---0.955 0.950 T T .-,-33 35 37 39 41 Burnup (GWd/MTU)

Figure 8.2: k c rr as a Function of Categor y 5 Burnup Using 4.2 w/o Enriched Fuel The sa m e p roce ss i s r e p eated for R eg i o n 1 w h ere t h e cate g ories of concern are Categor y 2 a nd Category 3. H owe v er , th e bu rn u p r e quir e m e n t for th e 3-o u t-of-4 a r ea of R egio n 1 (Category

2) i s a fixe d 2 1 GW d/T ind e p en d e nt of e nri c hm e nt , co o lin g t im e , o r p ea kin g fac to r. F i gure 8.3 s h ows t h e r es ult s of th e ana l ys i s t o d ete rmin e th e burnu p p e n a lt y. A s ca n b e see n fro m F i gure 8.3 , Catego r y 3 b umup greater th a n 27.7 GW d/T h a s a very sma ll i mp act o n the full poo l calc ul a t e d k. A 7.5 GWd/T a d ditio n a l bumup r e qu ire m e nt fo r Catego r y 3 i s se l ecte d. T hi s makes th e Category 3 minimum burnup requirement 28.5 GWd/T. N E T-2 8 091-000 3-01 , Re v i s ion 0 11 9 0.984 0.982 0.980 .:ii:: 0.978 "C GI .... "' 0.976 :i I.I iii 0.974 u 0.972 0.970 0.968 25 25.5 26 26.5 27 27.5 28 28.5 29 Category 3 Burnup (GWd/MTU)

Figure 8.3: Calculated ketr as a Function of Category 3 Burn up Using 5.0 w/o Fuel 8.4.3 Confirmation of k9s 1 9s for Full Pool (includes Category 3 and 5) Since the total combined uncertainty is hi gher in Region 2 , the design objective k e rr for Region 1 is higher than Region 2. This mean s that a full pool model will produce a k e rr value driven by the higher reacti vity fuel in Region 1 and not yield any infonnation for Region 2. Before determinin g the Region 1 k e rr, it is desirable to confinn the Region 2 burnup penalty for Category 5. For these cases , the full pool is modeled with water holes in Region 1 so that the reactivity is driven by Region 2 fuel. The burnup for Category 5 fuel is reduced consistent with the difference in the bias and uncertainty bet ween C ategory 4 a nd Category 5 fuel. As determined in Section 7 .6, a burnup reduction of 1.3 GW d/T for the Category 5 fuel is required. Calculations were performed for 5.0 w/o fuel at the lo ading curve (for curre nt fuel) at a peaking factor of 1.2 for zero , two and 25 years of coo lin g time. The Category 4 burnups are taken from Table 8.4. The Category 5 burnup s are 11-1.3 = 9.7 GWd/T hi g her. Table 8.9 s hows calcula t ed k's and the k 9s 195's for three 5.0 w/o cases at a peaking factor of 1.20. The bias and uncertainty used for Table 8.9 come from Table 8.7. Table 8.9 shows that the loading criteria of Category 4 plus 11 GWd/T for NET-2809 1-000 3-01 , Revision 0 120 Catego r y 5 m eets t h e k ( <0.9 9) crite ri a. As expe ct e d , th e r e is s li g htl y m ore m a r gi n for t h e 2 yea r c o o l ed case t h a n th e 72 h o u r co ol e d case a nd eve n mo r e m a r g in for t h e 2 5 year coo l e d case. This i s b e c a u se t h e burnup r e quir e m ent d ec r eases w i t h co olin g t ime a nd s o t h e fixe d burnu p pe n a l ty of 11 GW d/T i s a l a r ger fr act ion o f th e bu rn up r e quir eme n t. It i s co n ce i va bl e th a t a h ig h e r p ea kin g fac t o r mi g h t res ult in l ess m argi n. A n addit io n a l c a lcul at i o n was p e r forme d u s in g a p eaki n g fac t or of 1.3. Th e hi ghes t p ea kin g fac t or fo r fu e l in th e IP 2 SF P m eeti n g th e Ca t ego r y 4 burnu p re quir e m e nt s is 1.272. The ca lcul a t e d k 9 5 1 9 5 for a p ea kin g fac t or of 1.3 i s 0.989 7. Table 8.9: Region 2 Models at Loading Cur v e (Cat 5 i s Cat 4 plus 11 GWd/T) Categor y 4 Category 5 Burnup Burnup in Peaking Cooling Calculated Si g ma Bias and k 9 S/95 (GWd/T) Model Factor Time k U ncertaint y. (GWd/T) 48.1 9 57.89 1.2 72 ho ur s 0.9584 0.00005 0.0307 0.9891 46.6 1 56.3 1 1.2 2 yea r s 0.9578 0.00006 0.0 307 0.9885 38.8 5 48.55 1.2 25 y ea r s 0.956 1 0.00006 0.0304 0.9865 48.6 1 58.3 1 1.3 72 h o ur s 0.9588 0.00006 0.0 309 0.9897 T h e mod e l u sed fo r th e R egio n 2 a n a l ys i s h as th e Ca t ego r y 5 fu e l eccen t rica ll y pl a c ed o n th e ri g h t h a nd s id e of eac h ce ll (th e r eac ti vity effe ct of mov in g th e Categ o ry 5 fu e l t o th e ri g ht was ca lcul a ted a nd i s wo rth 0.000 8 ~k). With thi s m ode l , w hi c h pic k s up ecce nt ric it y as we ll as a n y int erface effec t s, th e m ax imum k 9s 1 9s is m e t w i t h m o r e t h a n 1 % m a r gi n t o th e r egu l atory limit. U p o n c onfirm at i o n of th e R egio n 2 l oa d i n g re quir e m e n ts, t h e a n a l ys i s of R eg ion 1 ca n p ro cee d b y addin g Ca te go r y 2 a nd 3 fu e l in R eg i o n 1 o f t h e m o d e l. In orde r t o co mp ensate fo r th e d iffe r e n ce in bi as a nd un ce rt a int y b etwee n Ca t egory 2 a nd Ca t egory 3 , th e bu rnup of th e Category 3 fu e l i s re du ce d. In S ec t io n 7.6 , th e di ffe r e nc e in un cer tainti es for 7.5 GWd/T i s 0.8 GWd/T. The bumup r e quir e m e nt for C at ego r y 2 i s 2 1 G Wd/T a nd th e b u rnup r e qu i r e m e nt for Ca t ego r y 3 i s 28.5 G Wd/T. R e du c in g 28.5 GW d/T b y 0.8 m ea ns th e modeled burnup for Category 3 is 27.7 GWd/T. C on seq u e ntl y , th e k 9s 1 9s N ET-28 091-000 3-0 1 , R ev i s i o n 0 1 2 1 for the SFP is the calculated k eff in the full pool model plus the bias and uncertainty for Category 2 (21 GWd/T). The Region 1 eccentricity can be in two forms; in the center of Category 2 cells or at the boundary of Category 2 and 3 cells. The highest keffis when the eccentricity is in the center of the Category 2 cell area of Region 1. Table 8.10 shows the results of the analysis of different eccentric options. The analysis uses t he primary arrangement of Region 1 ( shown on Figure 1.1) with all fuel at 5 .0 w/o enrichment and the burnups of Categories 2 through 5 fuel are 21 , 27.7 , 48.19, and 57.89 GWd/T. For all but the last case on Table 8.10, all of the Category 3 and 5 fuel assemblies are centered ( except the case where the eccentric positioning is at the Category 2/3 boundary where only those assemblies with the eccentric grouping are moved in). Table 8.10: Eccentric Options for Region 1 Case k Sigma .dk All Region 1 Assemblies centered in cells 0.9665 0.00007 Plus Category 3 and 5 fuel in Region 1 moved 0.9681 0.00007 0.0016 toward the center Also Plus 16 Eccentric at Category 2/3 Interface 0.9686 0.00006 0.0021 16 Eccentric in the middle of Category 2 plus 0.9687 0.00007 0.0022 Category 3 and 5 moved in toward Category 2 From Table 7.3 , the bias and uncertainty is 0.0194 Llk with the eccentricity bias removed. The full pool model contain s the eccentricity and interface effects. The final k9s 1 9s for the Figure 1.1 arrangement of Region 1 is 0.9687 + 0.0194 = 0.9881. Thus , for the primary arrangement of Region 1 the target 0.99 is satisfied, leaving more than 1 % margin to the 10CFR50.68 limit. The peaking factors used for Category 2 and Category 3 fuel are 0.9 and 1.2, respectively.

These are based on an expected cycle length for the final cycle of 23.8 GWd/T. There are no assemblies currently in the SFP that are Category 2. Until the final cycle, it is expected that low burned fuel will be returned to the core for more burnup. If the final cycle is 23.8 GWd/T , then the highest peaking factor that matches NET-28091-0003-01, Revision 0 122 t he minimum burnup requirements for Category 2 (21 GWd/T) is 21/23.8 = 0.88. Assemblies in that cycle will certainly have higher peaking factor s, but they would then exceed the minimum burnup r equirement.

For example, if an assembly had a peaking factor of 1.4 and the cycle length is 23.8 GWd/T, its burnup would be 23.8*1.4 = 33.3 GWd/T which would greatly exceed the Category 2 burnup requirement.

The reactivity effect of additional bumup is greater than the effect of higher t emperatures during depletion that is caused by a higher peaking factor. For Category 3 , the highest peaking factor that matches the minimum burnup requirements for 28.5 GWD/T and a cycle length of 23.8 GWd/T can be calculated and is 28.5/23.8 = 1.2. The calculated k using 0.9 and 1.2 peaking factors i s conservative for any final cycle burnup greater than 23.8 GWd/T. The final cycle ma y be cut short so additional calculations have been done for different cycle lengths. If the final cycle were only 20.3 GW d/T then the peaking factors would be 21/20.3 = 1.03 and 28.5/20.3 = 1.40. The calculation using peaking factors of 1.1 and 1.4 for Category 2 and 3 , respectively, resulted in a k 9s 1 9s of 0.9895. With cycle bumups of less than 20.3, no new Category 3 assemblies would be produced since the maximum assembly peaking factor is 1.4 at any point in the cycle (note that this is not bumup averaged).

The most reactive assemblies in the SFP, except for four assemblies , have a bumup of 40 GWd/T. For cycles less than 20.3 GWd/T , the maximum peaking factor of 1.4 is used for Category 2 fuel , and the Category 3 fuel i s modeled as 40 GWd/T (5.0 w/o fuel) with a peaking factor of 1.4. The calculated k 95 1 95 for this case is 0.9888. In conclusion , for all final cycle lengths , the calculated k 9 5 1 9 5 is less than the regulatory limit by more than 1 % ink. Although it is not expected that there would be multiple short cycles leaving more reacti v e fuel in Region 1 , a case was analyzed where the peaking factor of 1.4 is used for both Category 2 and 3 fuel using the minimum burnup requirements and 5.0 w/o fuel. This calculated k 9s 1 9 5 is only 0.9904. This exceeds the design objective but still provides 0.96% in k margin to the regulatory limit. These cases are summarized on Table 8.11. NET-28091-0003-01 , Revision 0 123 Table 8.11: Maxim u m Full Poo l k9s t 9s assuming Various Cycle Lengths Cycle Length Category 2 Category 3 Calculated Sigma Bias and k9S/95 (GWd/f) Peakin11:

Factor Peakin11:

Factor k Uncertainty

>23.8 0.9 1.2 0.9687 0.00007 0.0194 0.9881 20.3 -23.8 1.1 1.4 0.9701 0.00006 0.0194 0.9895 <20.3 1.4 1.4 but 40 GWd!f 0.9694 0.00007 0.0194 0.9888 Multiple Short Cycles 1.4 1.4 0.97 10 0.00007 0.0194 0.9904 8.5 Alternate Arrangements for Region 1 A checkerboard arrangement of Category 1 fuel i s much less reacti ve than the base arrangement of fuel in Region 1. The calculated k err of unburned 5.0 w/o fuel wit h 64 IFBA rods checkerboarded in Re gio n 1 is 0.8 548. Since this k is so low , it is permitted to replace an area of Region 1 with a checkerboard arrangement of Category 1 cells. In order to not create an increase in k due to interface s, there are two rules for creating the area for Category 1 storage. 1. Each Category 1 cell must be face adjacent with at least three water holes. 2. Each Category 2 cell may not ha ve more than one face adjacent to a Category 1 cell. Given these constraints, three additional arrangements of fuel in Region 1 were analyzed to confirm the reduction ink with Category 1 checkerboards present. The first additional arrangement represents the expected arrangement in Region 1 prior to lo a ding a new cycle. For this arrangeme nt , fresh fuel is generally in the new fuel vault but some Category 1 fuel is needed to be in the SFP. This arrangement will be called the " Refueling Arrangement." The seco nd additiona l arrangement covers the case where all of the Category 2 cells are removed but the Category 3 and 5 cells are still u se d. This arrangement will be called the "No Cat 2" arrangement.

The final additional arrangement maximizes the number of Category 1 cells in Region 1. This arrangement i s called "Max Cat l." Figures 8.4 through 8.6 show these arrangements of Region 1. Figure 8.7 shows an arrangement of Region 1 that is not limiting but is useful to illustrate the Category 1 rule s. The term "c heckerboard" is NET-2809 1-0003-01, Revision 0 124 used to d escribe the regu l ar array bu t t h e key req ui re m ent is t h e fi r s t rule req u iri n g at l east 3 face adjace n t water ho l es. This re qui re m ent i s illu stra t e d in Figu r e 8.7 showing o nl y one , two , or thr e e Category l cells. T h e arrangeme n t of other Ca t egories that are n ot replaced b y Category l m u st not cha n ge even with t h e pr ese n ce of a n y nu mber of Category I area s. A l t h ou g h di fferent arrange m e n ts of Ca t egory 2 cell s are pos s i ble that w o ul d not v io l ate the critic a li ty r e quir e m e nts , the rule s to a ll o w th es e types of rearr a ngement s w ou l d b e complex a n d pron e to error and th e r e fore ar e not a llow e d. In order to fairl y compare th e arrange m e nt of the base case and t h e a ddi t i onal arra n ge m ents , t h es e a n a l yses we r e perfonne d with a ll of t h e fuel c ente r e d i n i t s ce ll. Ta bl e 8.12 prov id es the calc ul ate d k e ff v a lu es for each arrangement.

2 3 4 5 6 7 8 9 W ll ll U U ll H V U U W li ll B M ll U V U ll H G F E D C B F i g u re 8.4: R e fu e lin g A rr a n ge m e n t Figure 8.5: No Cat 2 Arrangement NET-28 091-0003-0 1 , R ev i sio n 0 125 H G F E D C B H G F E D C B 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 Figure 8.6: Max Cat l Arrangement 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 ~----1------------------t= Lj -~ -~ ---*------,......,.., -------A X Figure 8.7: Example Odd Arrangement Table 8.12: Dependence of kcff on the Region 1 Arrangement Arrangement Description k Sigma ~k from Number Reference 1 Referen ce -No Category 1 0.9665 0.00007 2 Refu eling Arrangement 0.9654 0.00006 -0.0011 3 No Category 2 Arrangement 0.9651 0.00006 -0.0014 4 Maximum Ca tegory 1 Arra n ge m ent 0.9584 0.00006 -0.0081 5 Odd Arrangement 0.9655 0.00005 -0.0010 NET-28091-0003-01, Revision 0 126

8.6 Calculations

for Discharged Fuel (IP2 A-X and IP3 A-AA) As noted in Section 5, depletion conditions vary for each batch. Bounding depletion conditions are used for each batch grouping or individual assembly.

Minimum burnup requirements are then determined for each batch grouping or individual assembly.

It is desirable to allow as many assemblies as possible to be stored in Region 2 due to the limited size of Region 1. The minimum burnup requirement of Region 2 is for the 3-out-of-4 area. If an assembly burnup exceeds this minimum requirement, then it is classified as Category 4 fuel. If the burnup exceeds the minimum requirement by 11 GWd/T or more, then it is classified as Category 5 fuel. There are currently a sufficient number of assemblies that meet the Category 5 criteria so that all cells requiring Category 5 can be filled. If the burnup of an assembly is less than the general Category 4 requirement for the batch grouping , its classification is further studied. Many of these cases make the requirements for Category 4 after further analysis (see Table 8.24). A few remaining assemblies do not have more than 28.5 GWd/T burnup and further analysis is performed to show that they meet the reactivity requirement for Category 3 (see Table 8.23). Tables 8.13 to 8.22 are the Category 4 loading requirements (the 3-out-of-4 area of Region 2) for each batch or batch grouping as a function of initial enrichment , cooling time, and peaking factor. NET-2809 1-0003-01, Rev i sion 0 127

' Table 8.13: Batch A-D Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Cooling Time (years) PF=l.20 (w/o) 10 25 45 2.0 7.90 7.3 1 7.04 2.2 10.56 9.69 9.35 2.6 16.6 8 15.29 14.64 3.0 22.78 2 1.12 20.32 3.4 27.82 2 5.90 25.0 2 3.8 32.37 30.63 29.69 Enrichment Coolin!! Time (years) PF=0.80 (w/o) 10 25 45 2.0 7.62 7.09 6.85 2.2 10.18 9.44 9.15 2.6 16.2 0 14.93 14.3 7 3.0 22.27 20.74 2 0.00 3.4 27.25 25.51 24.72 3.8 31.84 30.24 29.34 Table 8.14: Batch E-F Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Cooling Time (years) PF=l.20 (w/o) 10 25 45 3.0 21.45 20.02 1 9.35 3.4 26.48 24.87 24.16 3.8 31.31 29.66 28.76 Enrichment Coolin!! Time (years) PF=0.80 (w/o) 10 25 45 3.0 21.05 19.76 19.14 3.4 26.03 24.6 1 23.96 3.8 30.93 29.36 28.53 NET-2809 1-000 3-01, R ev ision 0 1 28 Table 8.15: Batch G-L Mini mum Burnup Requirements (GWd/T) for Categor y 4 Enrichment Coolin PF=l.20 (w/o) 10 45 3.0 24.39 22.8 1 22.11 3.4 29.42 27.43 26.52 3.8 33.64 31.9 0 3 1.0 4 Enrichment Coolin PF=0.80 (w/o) 10 45 3.0 23.97 22.54 2 1.86 3.4 28.91 27.1 2 26.26 3.8 33.19 31.56 30.78 Table 8.16: Batch M-P Min imum Burnup Requirements (GWd/T) for Categor y 4 Enrichment Coolin Time ears PF=l.20 (w/o) 10 15 25 45 3.4 2 9.85 28.92 27.78 26.78 3.8 3 3.69 32.87 3 1.83 30.92 4.2 3 9.28 38.59 37.62 36.34 4.6 41 .3 1 40.56 39.66 38.91 5.0 4 5.42 43.96 42.16 41.25 Enrichment ears PF=0.80 (w/o) 10 25 45 3.4 2 8.92 28.06 27.05 26.17 3.8 3 3.40 32.64 3 1.68 30.86 4.2 3 8.85 38.30 37.17 35.99 4.6 41 .43 40.78 40.01 39.37 5.0 4 4.58 43.27 4 1.79 4 1.04 NET-2809 1-000 3-01, R ev i sion 0 129 Table 8.17: Batch Q-S Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Coolin2 Time (years) PF=l.20 (w/o) 10 15 25 3.8 2 9.75 28.6 5 27.29 4.2 33.63 3 2.43 3 0.96 4.6 37.5 0 36.2 0 34.58 5.0 4 1.3 1 39.88 38.1 4 Enrichment Coolin2 Time (years) PF=0.80 (w/o) 10 15 25 3.8 28.93 27.9 4 26.74 4.2 32.83 3 1.75 3 0.4 2 4.6 36.67 3 5.49 34.0 3 5.0 40.4 1 39.11 37.53 Table 8.18: Batch T-V Minimum Burnup Requirements (GWd/T) for Categor y 4 E nrichment Coolin2 Time (years) PF=l.20 (w/o) 5 10 15 25 3.8 32.40 3 0.65 29.44 28.07 4.2 36.4 5 3 4.51 33.1 9 3 1.63 4.6 40.41 3 8.27 36.86 35.1 8 5.0 4 4.24 41.94 40.42 38.64 E nrichment Coolin2 Time (years) PF=0.80 (w/o) 5 10 15 25 3.8 3 1.41 2 9.80 28.8 1 27.58 4.2 3 5.3 4 33.59 32.47 3 1.1 1 4.6 39.2 1 37.3 6 36.1 1 3 4.60 5.0 4 2.98 40.87 39.55 3 7.90 N ET-2 8 091-000 3-01 , R ev i s i o n 0 1 30 Table 8.19: Batch W Minimum Burnup Requirements (GWd/T) for Categor y 4 Enrichment Cooling Time (years) PF=l.2 (w/o) 2 5 10 15 25 3.8 33.72 32.2 5 3 0.43 29.2 8 27.92 4.2 37.86 3 6.2 1 34.23 32.8 9 3 1.4 0 4.6 41.8 4 40.07 37.90 36.53 34.8 7 5.0 45.7 5 43.82 4 1.5 2 4 0.04 38.2 7 Enrichment Cooling Time (y ears) PF=0.80 (w/o) 2 5 10 15 25 3.8 32.49 3 1.23 29.66 28.66 27.4 3 4.2 36.52 3 5.1 1 33.35 32.27 3 0.88 4.6 40.42 3 8.86 36.93 35.7 1 34.23 5.0 44.2 0 42.52 40.50 39.1 9 37.57 Table 8.20: Batch X Minimum Burnup Requirements (GWd/T) for Categor y 4 Enrichment Coolin!?:

Time (years) PF=l.2 (w/o) 2 5 10 15 25 3.8 3 3.73 32.3 0 30.48 29.29 27.96 4.2 37.8 5 36.2 1 34.23 32.94 3 1.4 1 4.6 41.8 1 40.0 8 37.84 36.50 34.8 6 5.0 45.69 43.82 41.4 3 39.94 38.15 Enrichment Cooling Time (years) PF=0.80 (w/o) 2 5 10 15 25 3.8 32.53 3 1.25 29.59 28.63 27.4 0 4.2 36.53 35.1 0 33.27 32.1 8 30.85 4.6 40.39 3 8.8 1 36.9 1 35.69 34.1 8 5.0 44.16 4 2.4 9 40.4 7 39.1 6 37.56 The fo ll owing loa d ing r e q uire m e nt s a r e for a ll di sc h arge d fue l from IP3 (B atc h es A t hru X). NET-28 091-000 3-0 1 , R ev i s i o n 0 1 3 1 Table 8.21: Batch A-U (IP3) Minimum Burnup Requirem e nt s (GWd/T) for C a te g or y 4 Enrichm e n t Cooling Time (y ear s) P F=l.20 (w/o) 10 25 45 2.2 12.52 11.4 1 10.92 2.6 18.83 17.25 16.52 3.0 24.71 23.0 1 22.27 3.4 29.79 27.70 26.71 3.8 33.94 32.08 31.21 Enrichm e nt Cooling Time (year s) P F=0.80 (w/o) 10 25 45 2.2 12.09 11.10 10.67 2.6 18.28 16.87 16.23 3.0 24.17 22.66 21.97 3.4 29.16 27.28 26.37 3.8 33.39 31.72 30.90 Table 8.2 2: Batch V-X (I P3) M inimum Burnup Requir e m e nts (GW d/T) for C ate g o ry 4 E nrichment C ooling Time (y ear s) P F=l.20 (w/o) 10 25 45 3.4 28.28 27.24 25.92 3.8 31.90 30.69 29.26 4.2 35.13 33.75 32.19 E nrich me nt C ooling T im e (ye ar s) P F=0.80 (w/o) 10 25 45 3.4 27.57 26.56 25.43 3.8 31.04 29.96 28.7 1 4.2 34.1 1 32.95 31.58 The IP3 fuel after Batch U u ses the Batch Z dep l etion parameters with two additional enrichment points at 3.4 and 3.8 w/o. For these l ower enrichment points, the fue l is modeled as full lengt h which is conservat i ve for blanketed fue l (Batches V , W , and X h ave six inc h natural urani u m blankets w h i l e Y and AA have six inch 2.6 w/o bl ankets). T h e b atches Y and AA (an d t h e h igher enric h ed fuel from B a t ch X) use the Z l oa d ing curve w h ich was base d on mode lin g a n 8-inch ax i a l b l anket wit h a blanket enric h ment of 4.0 w/o. If the blanket enrichment were the same, modeling a 6-i n ch blanket as 8 inches wo ul d be conservative.

Howeve r , the highest enr i ched 6-inch b l anket is 2.6 w/o. The fo ll owing cases were run to NET-28 091-0003-0 1 , R ev i s i on 0 132 show that u sing the Z l oad in g curve (8-in ch at 4.0 w/o) for the s h orter l ess enric h ed bl anket (6-inch at 2.6 w/o) i s conservative.

4.6 w/o, 42 GWd/T, Z b a t ch: 4.6 w/o, 42 GWd/T , 6-inch AA: 4.6 w/o, 46 GWd/T , Z b atch: 4.6 w/o , 46 GWd/T, 6-inch AA: ' k-eff = 0.97 1 23 +/- 0.00008 k-eff= 0.96982 +/- 0.00008 ~k = 0.0014 k-eff = 0.95008 +/- 0.00008 k-eff= 0.94884 +/- 0.00008 ~k = 0.00 1 3 Tab l e 8.23 shows a set of assem blie s th at faile d Category 4 but h ave a bumup that i s less than 28.5 GW d/T (the minim um bu mup for Ca t egory 3). To s h ow whet h er they can make Category 3 , each assemb l y was calculate d in the infinite 2x2 m odel of R egion 1 conta in ing four fuel assembl i es. T h e calcu l ate d k e rrwith 28.5 GWd/T fuel (5.0 w/o enr i c hm e nt , PF=l.4) in this model i s 1.0133 (not e that Category 3 fuel u ses l ea ka ge from the edge of the SFP to l ower its k -this is further explained in Section 8.8). If the calc ulat ed k err for a n assemb l y i s l ess th an 1.0133, then it is classified as Category 3 even tho u g h it has a burnup l ess than 28.5 GWd/T. Table 8.23: Individual Assembly Analysis for Category 3 Assembly ID Enrich Burnup Cooling PF Cale. k (MWd/T) (years) X02 4.802 28357 6.3 1.1 94 0.9956 XOl 4.802 28460 6.3 1.199 0.9956 F65 3.346 1 2 034 37.7 1.228 0.9911 V43 (IP3) 3.803 14949 26.2 1.11 6 1.0005 V48 (IP3) 3.8 05 15180 26.2 1.1 33 1.0005 Based on th ese results, a ll of these assemblies are cl assified as Category 3 even though the ir burnup is l ess than 28.5 GWd/T. The assem bli es in Ta bl e 8.24 fai l e d Category 4 by a small amo unt of bumup , so th ey were a nal yzed u s in g the d e pl et i on cond ition s for the assembly rather t han the depletion conditions for the batch. Section 5.7 d esc rib es the spec i a l depletions and the axial bumup profiles used. The atom densities for these asse mbli es were placed in the Region 2 (3-out-of-4) 2x2 model (a ll three assem bli es in t h e mod e l NET-2809 1-000 3-0 1 , R ev i s ion 0 133 are the sa m e). The calculated kerr plus th e bi as a nd unc er taint y i s le ss than 0.99 in all cases, so the y are cla ssifie d as Category 4 (stored in the 3-out-of-4 area of Re gio n 2). Table 8.24: Individual Assembly Analysis for Category 4 Assembly ID Enrich Burnup Cooling PF Calc.k Bias & Uncer. k9S/95 AlO 2.21 15038 42.3 0.92 0.94 8 9 0.0166 0.9655 F44 3.35 23017 35.8 1.05 0.9609 0.021 8 0.9827 L48 3.69 29515 25.4 0.69 0.962 4 0.0235 0.9859 W52 4.96 40641 6.3 0.8 4 0.9569 0.0307 0.9876 X18 4.95 42479 4.3 0.87 0.9566 0.0306 0.9872 U12(IP3) 3.21 24800 27.8 0.90 0.9588 0.0223 0.98 11 With the above loading curves, eve r y hi storica l assembly through B a t c h X of IP2 and Batch AA of IP 3 h as been categorized as 3, 4, or 5. This categorization is su mmari ze d in Appendix B. Fuel asse mbli es A l 1 , A24 , A47, A49, ASO, A51, A54, and ASS in IP2 an d assemblies A38, A43 , A44 , A45, A51, A59, A63, and A64 in IP3 contained part l engt h control rods. These assemblies ha ve an e nri c hment of 2.25 w t% 235 U-235 (w/o) and burnups greater than 16 GWd/MTU. The Category 4 burnup require ment for th ese assemblies is l ess than 11 GWd/T. Although the r eact i v it y effect of control rods i s significant

[ 14], it is not enough to overcome the 50% excess burnup in these assemblies. These assemblies ha ve insufficient burnup for Category 5 a nd are d ispos ition ed as Catego r y 4. N ET-28091-0003-01, Revision 0 1 34

8. 7 Cell Blockers IP2 will have two cells blocked (no fuel) at location s A22 a nd BC72. One of these ce ll s (A22) is at the Region 1 and 2 interface and is not credited.

The other ce ll (BC72), however , i s on the e dge of the SFP above the cask lo a din g area. Due to th e ce ll blo cker , it i s possible to allow a Category 4 ce ll on both sides of the cell blocker a l ong the cask l oading area. Figure 8.8 shows the SCALE model w i t h the two extra Category 4 ce ll s. The k err for thi s case is 0.95844 +/- 0.00006. This can be compared to the r efere nc e k err of 0.95839 (without th e cell blocker and wit hout th e two additional Category 4 cells). The differen ce is within the Monte Ca rl o unc erta inty. Figure 8.8: Cell Blocker Region 2 Model NET-2809 1-0 003-01, Revision 0 L((;(P') Q wuo MfOIIAL 2 -MIOUAI.. J -MIUIIAl..-4 c::J ,wnuun1..a 0 MltlWl .. 3 l -N'H(RIAL M -Ml[RIAL.S J -MlfHfn.., uo -MIUIIAl411

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8. 8 Region 2 Checkerboard Figure 1.1 does not show a Region 2 checkerboard arrangement since it is unexpected that the plant will ever use a Region 2 checkerboard. However, after the removal of a significant number of assemblies, it may be convenient to create a zone of the SFP near the cask loading area where any assembly may be placed (including fresh 5.0 w/o fuel). For this contingency , no interface analysis has been performed , so t he Region 2 checkerboard zone must have a row of water holes on all sides. The edge of the SFP can be credited as a row of water holes. Using an infinite 2x2 checkerboard model of Region 2 , the calculated k eff of 5.0 w/o fuel with 64 IFBA rods is 0.9521. With the bias and uncertainty (0.0083 from Tab l e 7.5) this produces a k 9s 1 9 s of 0.9604. This easily meets the 0.99 target. All of the fresh and burned fuel from IP2 and IP3 qualify for placement in this Region 2 checkerboard. 8.9 Burnup Penalty f or Hafnium Flux Suppressio n Inserts Hafnium inserts have been used at eight corner locations in the IP3 core starting with Cycle 11 to mitigate concerns o v er Pressurized Thermal Shock. Hafnium inserts have never been used at IP2. The burnup penalty for hafnium inserts was determined in the previous CSA to be 1.31 GWd/T (worst case) [1]. To provide more than adequate margin , a penalty of2 GWd/T is added to the bumup requirement for any assembly that had a hafnium insert any time during its life. 8.10 Failed Fuel Containers The southeast comer (p l ease note that on the drawings in this report North points left) of the SFP contains two 16" circular pipes and are labeled "failed fuel containers" on Figure 3 .1. These containers have been used to store pieces of failed fuel rods, neutron sources, and fission chambers. The neutron sources and fission chambers contain too l ittle fissi l e material to be a concern. The fission chambers have less than 10 mg U-235 each [25]. The neutron sources also have a very small amount of fissile material.

The ANSI/ANS 8.1 standard [53] states that 700 grams ofU-235 in any configuration is always subcritical.

However, the failed fuel containers are not fully decoupled from the Module H of Region 2. NET-2809 1-0003-01, Rev i sion 0 136 This analysis permits 16 fuel rods in each of the failed fue l containers. Rather than model the actual container , 16 pins are placed close to the fuel modules for each failed fuel container.

Since the same SCALE unit is used in two places in the model, a third set of 16 fuel pin s is in the model. Fig ure 8.9 shows the model with the extra pins. The extra pins did not change the k eff of the model. Table 8.25 shows the re su lt s of calcu l ations where the number of pins is var i ed. A lth ough more than 16 fuel pins per failed fuel container wou ld be possible, there are current l y pieces from only one fai l ed fuel pin in the containers no w , and greater than 16 rods in each is not credible. The fuel rods are modeled as fresh bl anketed, no-IFB A , 5.0 w/o rods. The start source was placed near the failed fuel container.

Figure 8.9: Failed Fuel Container Pin Model (Blow up of the top right of the Region 2 model) NET-28091-0003-01 , R ev i sion 0 -1111101111.

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-f1A1C IU ,.S 1 7 -l'W<llltU l!Ut 137 Table 8.25: Failed Fuel Container Pin Analysis Pins oer Container Calculated k Sigma 0 0.9584 0.00005 4 0.9586 0.00006 9 0.9585 0.0 0007 1 6 0.9585 0.00006 81 0.9635 0.00007 121 0.9793 0.00008 144 0.9959 0.00008 8. 11 Fuel Rod Storage Basket The Indian Point SFP can h ave movable fuel rod storage basket s that can be used to sto re fuel rod s. These baskets can fit in a storage ce ll and the y have 52 holes for storing fuel rods. This was modeled as 52 fresh 5.0 w/o fuel rods in the 3-out-o f-4 area of Region 2 (see Figure 8.10). The calculated k efT for this configuration was 0.9584, which is we ll below the k 95 1 95 requirement of 0.99 since the bias and uncertainty for fre s h fuel is only 0.0083 t.k (from Tab l e 7.5). Therefore, the fuel rod storage basket i s cla ss ified as reactivity Category 4 (refer to Table B. l ). Figure 8.10: Model for the Fuel Rod Storage Basket NET-28091-0003-01 , Revision 0 138

8. 12 Assemblies with Missing Fuel Rods Typically , when a fue l assembly has one or more failed fue l rods, the failed fuel rod is removed and replaced w ith a stainless steel rod having the same outer dimension as a fuel rod. If this is performed, there is no criticality concern since the reconstituted assembly wou ld be less reactive than the original assemb l y. However , if a failed fuel rod is removed but not replaced with a stainless steel rod, the reactivity in creases because there is more water ava il ab l e. An ana l ysis was perfonned for the previous CSA [ 1] in which one or more fuel rods are removed from an assem bl y to estimate the effect on k. This analysis was not repeated since the approach taken provided a large margin. The model that was u sed for t his ana l ysis contained absorber inserts. It was detennined that ke ff gradually increases as more fuel rods are removed up to and including 36 missing fuel rods. If more than 36 fue l rods are removed, k eff begins to decrease (see Figure 8.11 ). The change in k e ffwith 36 missing fuel ro d s (see Figure 8.1 2) was 0.0 1 84 ~k (a 2x2 array with a ll 4 assemb lie s h aving 36 missing rods). For simp li city, a bumup penalty of 4 GWd/T would cover this reactivity in crease for an assembly wit h any number of missing rods. There is on l y one fuel assembly in the SFP (assembly ID ofT67) that has a missing fue l rod. This assembly has only one missing fuel rod. The initial fuel enrichment for this assemb l y was 4.952% and the bumup is 49.81 GWd/T and the assemb l y has cooled more than 10 years. T hi s assemb l y exceeds th e Category 4 lo ading req uir ement by over 7 GWd/T, so it exceeds the requirement by more than the 4 GWd/T pena l ty for missing fuel rods. If any assemb l y in the future is re-constituted without replacing fuel rods with stainless steel rods, then 4 GWd/T wou l d have to be added to the l oading curve requirement before it could be stored. This penalty covers any number of missing fuel rods , and there is no other loading restriction (two or more fuel assemblies with missing rods could be stored next to eac h other as l ong as the 4 GWd/T is added for each assembly).

NET-2809 1-0003-01, R ev i sion 0 1 39 k 0.9850 0.9800 0.9750 0.9700 0.9650 0.9600 k versus Missing Fuel Rods 0 10 20 30 40 Number of Missing Fuel Rods Figure 8.11: k c ff versus Number of Missing Fuel Rods 000000<)00

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  • o , , oo-:;oo 000000000600600 000000000000000 000000000000000 000000000000000 Figure 8.12: Model for Assemblies with 36 Missing Fuel Rods [1] N ET-2 8 091-0003-01 , R ev ision 0 50 140
8. 13 Storage of Miscellaneous Materials Misce ll aneous non-actinide materia l s (for examp l e, empty or full trash baskets), can be stored in fuel positions of any category.

However, there are some special cases where some of the material may be stored in a water hole or 50% water ho l e. If the misce ll aneous material is any type of steel , Inconel, or absorber material (e.g., absorber coupons , stainless steel coupon trees , control rods , unburned burnable absorbers) it may di s place up to 50% of the w a ter volume at the active fuel zone ( 144 inches) of a water hole or 50% water ho l e (t h ere are no restrictions on material above or below the active fuel zone). If the miscellaneous materia l is a very low absorbing material such as a void , zirconium , a l uminum , cloth , plastic , concrete , etc., it cannot be placed in a water h ole but may be p l aced in a 50% water ho l e so long as the 50% water hole still ha s 50% water volume in the acti v e fuel zone. To confirm that a water hole i s allowed to contain 50% absorbing material , two ca s e s were run -one with 8 0% water , 20% sta inl ess stee l and another wit h 50% water, 50% stainless stee l. The diffe r ence in k eff from the reference case is -0.0100 L'ik and -0.0045 L'ik , respective l y. Other material s such as Inconel , absorber coupons , unburned burnable absorbers or control rod s absorb more neutrons than stain l ess steel and are covered by this ana l y s is. Any uranium that is not in a fuel assembly (for examp l e a removed or damaged fuel rod) must be stored in the Failed Fuel Containers (see Section 8.10) or the Fuel Rod Storage Basket (see Section 8.11). 8. 14 Borated Conditions The most l imiting acceptance criteria is for the unborated condition , so the loading criteria have been set using models that do not contain sol u b l e boron. In order to confirm that k 9s 1 9s is l es s than 0.95 at a b oron conte n t that is maintained even after a boron di lut ion accide nt (Section 9.6), a l imited num b er of additiona l calculations were performed.

The soluble boron concentration of 700 ppm is used for these calculations since this concentration ca n be easily supported by the boron dilution analysis , and i t yie l ds significant margin ink. For Category 2 fuel , at a burnup of2l GWd/f , the calculated k eFF With water at NET-2809 1-0003-0 l , Rev i sion 0 141 180 °F and cont a inin g 700 ppm boron is 0.8 496. Wi t h bias and uncertaint y, this become s k 95 1 95 = 0.87 12. Due to the difference between 0.87 and the tar ge t va lu e of 0.9 4 , n o further ca l cu l ations are warranted.

For Category 4 fuel , at the loadin g curve point s for 5.0 w/o fuel at 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (PF=l.20) and the 4.2 w/o fuel at 25 years (PF=0.80) with 700 ppm boron in the SFP water at 70 °C, the calculated k eff at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> i s 0.86 65 , w hil e the k err at 25 years is 0.8631. With t h e b ias and uncertainty app l ied , the k 9s 1 9s, to b e c ompared to the regulatory limit of 0.95 , becomes 0.8 972 at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 0.8903 at 25 years. The bias and uncertainty u se d here for both Region 1 a nd 2 i s from th e unborated analysis.

Calculation of borated uncertaintie s is not ne e ded d u e to the lar ge margin from t h e regul ato ry l imit. Eve n if borated unc e rtainti es were calc ulated , it i s expected th a t they would b e sma ller , since the primar y uncertainty is the burnup u ncert a int y and the r eac tivity worth of burnup d ecreases with in creas ing boron concentration due to s pectral h a rdening. Furthermore, ignorin g the grids i s still conservative at 700 ppm. In addition to th e analysis wit h the 2x2 model s, a full pool case was run at 700 ppm. T he calcul ate d k eff is 0.8 600. This case had the mo s t reacti ve loadin g permitt e d for all five ca tegorie s and includes eccentricity. Using the highe st bi as a nd uncertaint y of all re gio ns , 0.0374 , the k 9s 1 9s i s 0.8974 which i s much less than the t a rget of 0.94. An a n a l ysis was a l so perform e d for Re gio n 2 only. Table 8.26 pro v ide s the re s ult s from the full pool model s. T a bl e 8.2 6: Norm a l Op e ration s w ith Boron Dilution ppm (F ull Pool M odel) R e gi o n Calc ulat e d k s igm a All Categorie s at 5.0 w/o , 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cooled, Peaking B o th 0.8 600 0.0000 8 Factor of 1.2 , Loading Curve Burnups Catego ri es 4 and 5 at 5.0 w/o, 72 h ours coo l ed , Peaking 2 0.8595 0.00005 Facto r of 1.2 , Loading Curve Burnup s NET-2809 1-0003-01, Revision 0 142 8.15 Burnup Penalty for High Soluble Boron Conditions If a cycle is shut down very early, it is possible that the limiting soluble boron used in the depletion analysis (950 ppm) would not be met. The cycle would have to be shut down extremely early since the 950 ppm would be v iolated onl y if the cycle were shut down more than two months early. To cover thi s unlikely possibility , a special depletion was done at a soluble boron concentration of 1200 ppm t hroughout the depletion (this v alue exceeds the highest cycle average ppm at any burnup). If the burnup averaged ppm for any assembly exceeds 950 ppm , burnup penalties of0.2 , 0.3 , 0.6 , and 0.9 GWd/T would have to be applied to the burnup requirement for Category 2 , 3 , 4 , and 5 , respecti v ely. The following table summarizes calcul a tions to show that the s e burnup penalties are conservati v e (the Monte Carlo uncertainty is+/- 0.00003). Table 8.27: Burnup Pena l ty Results at 1200 ppm Case Calculate d k Category 2 (3 of 4 in Region 1 ): 5.0 w/o , 21 GWd/T, 950 ppm 0.96 8 9 5.0 w/o , 21.2 GWd/T , 1200 ppm 0.9686 Category 3 (4 of 4 in R e gion 1): 5.0 w/o, 28.5 GWd/T , 950 ppm 1.0117 5.0 w/o , 28.8 GWd/T , 1200 oom 1.0113 Category 4 (3 of 4 in Region 2): 5.0 w/o, 48.19 GWd/T, 950 oom 0.9586 5.0 w/o , 48.79 GWd/T , 1200 ppm 0.9583 Category 5 ( 4 of 4 in Region 2): 5.0 w/o , 59.19 GWd/T , 950 ppm 1.0187 5.0 w/o , 60.09 GWd/T , 1200 ppm 1.0183 NET-28091-0003-01, Revision 0 143 9 Normal Operations and Accident Analys i s The criticality analysis must address all conditions in the SFP that can cause criticality.

The normal operations are reviewed in Section 9.1. The accident analysis must assume the worst case conditions from the range of normal operations. The accident analysis is covered in Sections 9.2 through 9.7 and analyzes the possible upset co nditions that increase the reactivity of the SFP. The following accidents are analyzed:

  • A fresh assembly misplaced outside of the fuel racks but next to the fuel racks,
  • A fresh assembly dropped into an empty cell,
  • An over-temperature accident (water boiling in the SFP as a result of a loss of cooling), and
  • A multiple assembly misload event. Two more accident conditions are considered, but no analysis is necessary.

An assembly dropped horizonta ll y on top of other assemblies is not specifica ll y analyzed, because the assemblies are de-coupled as a result of the structure above the active fuel. The horizontal assembly would rest more than 20 cm above the top of the active fuel of the assemblies in the rack. This accident would be covered by the more severe accident of a fresh assembly dropped into an empty cell. The second accident condition would be a single misloaded assembly.

For example , an assembly that is supposed to have a control rod inserted but does not. All violations of the l oading requirements are bounded by a fresh assembly dropped into an empty cell. The last subsection of this section describes why a seismic event does not cause a criticality concern. NET-28091-0003-01, Revision 0 144

9. 1 Normal Operations A single isolat e d assembly at 5.0 w/o and no IFBA will have a k 9 s 1 9 s < 0.99. An assemb l y is isol ate d if there is 20 cm of wa ter b etwee n asse mbli es [ 1 3] (a row of em pty cells is 23 cm for R egio n 2). The equ ipment in the SFP can only move one assembly at a tim e. IP 2 doe s not have a n y racks in the SFP or refu e ling canal for temporary storage of fuel. However , there are t wo l ocatio n s w h e r e it wo uld be possible to plac e two assem bli es wit hin 20 cm of each other o ut s ide of the rack: w h e n a fuel assembly i s in the n e w fuel e l eva tor and when a fuel assemb l y i s vertica l in the up ender. Indian Point's proc e dur es w ill not permit movi n g a n assembly outside the rack within 25 cm of fuel in the n ew fuel e l eva tor or up en d er. The critica lit y a n a l ysis cre dit s the leakage at the edge of th e racks. Therefore, placi n g an assem bl y within 20 c m of the side of th e rack at the act i ve fuel e l eva ti ons is not permitted. IPEC's procedure wi ll also preclude thi s. Moving fuel in a nd out of rack ce ll s do es not in c r ease k err s ince the most r eac tiv e por tion of the fuel assemb l y is at th e top of the fuel, so movin g the bottom of th e fuel pa s t the top as it is in ser ted or rem ove d do es not in crease k. Fue l in s pe ction is performed above the rack w h e r e th e fuel assem bl y is iso l ated. Any reconstitution is p e rformed w hil e the assemb l y is iso l ated. I so lation r e quire s a row of water hole s on all si de s a nd corners. The o ut s ide of th e SFP can b e considered as a row of water holes. T he SFP i s required b y it s Techn ical Specifications t o contain at l east 2000 ppm of so lubl e boron. The SF P water temperat ur e durin g normal operation ranges fro m above freezing to 1 80 °F (the SFP de s i gn basis ma ximum temp erature).

NET-28091-0003-01 , Revision 0 145

9.2 Misplaced

Assembly For the misplaced fuel assembly accident, a fresh 5.0 w/o fuel assembly with 64 IFBA rods is placed in the SFP next to the rack in the most reactive location.

There are two locations which could be limiting for a misplaced assembly; the inside comers of the cask area and near the new fuel elevator when the elevator has a new fuel assembly in it. For the misplacement of fuel near the new fue l assembly in the elevator the fue l elevator is modeled in the cask loading area of the SFP. The key features of the model are maintained; close to a wa ll with two rows of Category 5 cells and in a big pool of water. The array structure used in the SCALE model makes it difficult to model at the actual location. The fuel elevator is conservatively modeled as exactly one assembly away from the side of Region 2 , so one misplaced assembly can exactly fit between the elevator and the rack. Figures 9.1 and 9.2 show the location of the misp l aced assemblies analyzed. Table 9.1 presents the results of this misplaced assembly accident analysis.

As can be seen from Table 9.1 , with the Technical Specification minimum of 2000 ppm of soluble boron , the final k 9 5 1 9s is below the target of 0.94. The analyses used starting sources located near the misplaced assembly.

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9.3 Dropped

Assembly For the dropped assembly accident, a fresh 5.0 w/o assembly (no IFBA) is dropped into one of the empty cells in Region 2 with the SFP at the Technical Specification minimum of 2000 ppm. F u rther, the assembly dropped is modeled with the grids failed , which allows for full expansion of the pin pitch into the cell (this maximum expansion of the pin pitch removes any concerns about fuel grid failure after the drop). The pin pitch expansion is modeled as the maximum uniform expansion that would fit in the cell. Figure 9.3 shows the fu ll pool model for the dropped assembly analysis with the dropped assembly. Note that Region 1 cells ha ve been removed so that the k e rr for Region 2 can be found, which comes with a higher bias and uncertainty than Region 1. The other fuel assemblies are at the lo ading curve limit for 5.0 w/o, no cooling and a peaking factor of 1.2. The calculated kerr is 0.8700 with the 2000 ppm soluble boron. After adding bias and uncertainty , this would be much lower than the target of 0.94, giving 1 % in margin. With this much margin , there i s no need to reevaluate the bias and uncertainties for borated conditions.

NET-28091-0003-01, Revision 0 149 F i g u re 9.3: F ull Pool M od e l w i t h Dropp e d Assem bl y 9.4 Over Temperature Lf.(i(Nl O wuo IIATCAIAl.

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[ 4]. Using the highest ke rr full pool model , water is modeled with 2000 ppm of so luble boron. The calculated NET-28091-0003-01 , Rev i sion 0 150 k e ffis 0.7464 +/- 0.00007. This case ha s a l arge margin to the 0.94 tar get, so calcu l ation ofa bias and un certa inty for this spec ifi c case i s not n ecessary.

9.5 Multiple

Misloads The mo st limi ting accident is the multiple misload case. For th e R egion 1 multipl e misload, it is assumed th at all ce ll s are filled with the most limiting assembl i es for IP2 (5.0 w/o enriched fuel wi th no bumup and 64 IFBA rods). The calc ulat ed k eff at 2000 ppm is 0.8196 +/- 0.00008. C l early, there i s significa nt mar gin for th e complete mi s lo ad of assem bli es in Region 1. For th e multi-mi s load ana l ysis for R egion 2, a ll ce ll s that are p ermitted to contain fuel are modeled as misloaded with once burned 5.0 w/o fuel with a burnup of24 GWd/T , as unburned fuel is eas il y identifiable and expected to be loaded into Region 1. Nearly a ll of th e once burned fuel asse mbli es exceed this bumup. Fo r example, if Cycle 23 com pl e t es it s expected burnup , four assemb li es will h ave a burnup of22 GWd/T with all of the r es t above 24 GWd/T. Curre ntl y (during IP 2 Cycle 23), only four assemb li es would be in the SFP during refueling wit h a bumup l ess than 24 GWd/T. Since the numb er of assemb li es b e l ow 24 GWd/T i s so few, misloading them in a reactivity significa nt way i s not credible (note that a sing l e misload is covered b y the dropped rod analysis, and if the misloads are not clo se to each ot her , th e effec t on k i s the same as a single misload).

T h e ca lcul ated k e ff for this case with 2000 ppm i s 0.9124 +/-0.00006. The bia s and uncertainty for this case is 0.0223 Lik (from conservative interpolation using Tab l e 7.3, the manufacturing tolerance s from Table 7.4, plus the hi gher va lidation bias and uncertainty due to the h arder spec trum). Addi n g 0.0223 to 0.9124 result s in a k 9s 1 9s= 0.9347 which is b e l ow th e t arget of 0.94. The multi-mi sload analysis did not mi s lo ad fuel i nto th e water hole s or 50% water h oles. It also m ainta ined th e contro l rods at th e location s required by th e Tec hni ca l Specificat ion s. The wa ter hol es an d control rod loc ations in R egion 2 are not allowed to c h ange position.

With the impl ementa tion of thi s N ET-28091-0003-01 , Revi sion 0 151 CSA, the IPEC staff wi ll receive training with emphasis on the fixed positions in Region 2 for the contro l rods and water holes. Although it is clearly not allowed by the Technical Specifications and the staff will have had specific training to reinforce the requirements of the control rods staying in the specific locations, analysis has been performed where all of the control rods in Region 2 are removed. This case assumes that the fuel is consistent with the most limiting conditions allowed by the Technical Specifications (5.0 w/o fuel with burnups of 48 .19 and 57.89 GW d/T). The calculated k e tr with 2000 ppm is 0.7880. Again, this k e tr after the addition of an appropriate bias and uncertainty is significantly less than the target 0.94. The final multi-misload analysis assumes the nonnal loading of Region 2 but a l l water holes are filled with fresh 5.0 w/o fuel with 64 IFBA. Since the reactivity is dominated by the fresh fuel, it is appropriate to reduce the Category 4 fuel burnup by the difference in the bias and uncertainty.

From Table 8. 7, the bias and uncertainty at 48.19 GWd/T is 0.0307 t.k. The Region 2 fresh fuel bias and uncertainty is 0.0083 t.k (from Table 7.5). Therefore a delta burnup to cover 0.0307 -0.0083= 0.0224 t.k is taken. This is estimated as 4 GWd/T. The Category 4 fuel mode l ed is the loading curve 48.19 -4 = 44.19 GWd/T. The calculated ke tr is 0.9224 +/- 0.00005. The Region 2 bias and uncertainty for Category 1 fuel from Table 7.5 after adjusting for the high EALF is 0.0151 t.k. Thus , the k 9s 1 9s is 0.9224 + 0.0151 = 0.9375. Since this is l ess than 0.94, this mu l ti-mis l oad meets the requirement.

9. 6 Boron Dilution Accident Crediting 700 ppm of soluble boron reduces the calculated k e tr plus biases and uncertainty to well below 0.94 (see Table 8.26). The boron dilution analysis of record [52] shows that dilution from the Technical Specification required minimum of 2000 ppm to 700 ppm is not credible due to the amount of time and water needed to dilute to this l evel. NET-28091-0003-01, Revision 0 152
9. 7 Seism ic Event This CSA is not crediting any absorber plates that could be affected by a seismic event. Further the space between the rack modules is not credited.

The space between the rack modu l es and the side of the SFP has a very small reactivity effect (see Table 6.5), so the soluble boron would easily cover any possible movement of the rack modules toward the wall. Any random v ariations in the cell dimensions would have to cover multiple cells to affect ke rr, and , again , the soluble boron would easily cover these variations.

The dropped assembly that could resu l t from a seismic event is covered in Section 9.3. In summary, this CSA is not sensitive to a seismic event. NET-28091-0003-01, Revision 0 153 10 Summary This CSA removes the reli a nce o n BoraflexTM in the IP2 SFP. Thi s is accomp li shed by use of water hole s, control rods , a nd l eakage at the edge of the SFP. Each fuel assembly i s categorized b y its r eactiv ity a nd spec ifi c lo cations in the SFP are reserved for each re act ivity category. The fuel categorization o f historical fuel i s acco mplish e d b y u se of batch gro uping s w ith simi l ar characteristics.

T h e categorization takes credit for lower enriched ax ial blanket s , coo lin g time , and assembly average peakin g factor. This CSA a l so categorizes IP3 fue l for storage in th e IP2 SFP. Section 10.1 contains a confirmation checklist th at th e gui dance ofDSS-ISG-2010-01

[5] is fo ll owe d. Section 10.2 de sc ribes the reacti v ity categori za tion of fuel assemb li es. Section 10.3 identifies the reactivity cate gorizati on of each cell. Section 10.4 li s ts the limitation s for fuel assemblies that ha ve n o t been categori ze d in Appendix B. 10.1 Review of DSS-ISG-2010-01 Tab l e 10.1 s how s the guidance given in DSS-ISG-2010-01 a nd how this criticality analysis follows that guida n ce. Table 10.1: DSS-ISG-2010-01 Checklist Section in this Guidance from DSS-ISG-2010-01 Implementation Report 1. Fuel Assembly Selection A ll fuel ha s co m e from th e sa me ve nd or w ith th e same cla d Section 3.2 Demon s trat e a ll fue l for a ll o ut s id e diameter.

Small des i g n c h a ng es have been conditio n s in s i g nifi cant to cr iti ca lit y a nal ys i s. 2. Depletion Ana l ys i s Thi s is fo ll owe d. Uncertainty fo r th e i soto pi c content is Sect i o n 4 a.i. 5% (Kopp M e mo) s h ould co n s id e r e d and impl emented as 5% of the d e pl e tion reactivity only be used to cover uncertaintie s (i.e., d e l ta-k of depletion). In a ddition , a bi as of 1.5% of the in i soto pic concentration worth of the minor actinides a nd fi ss i o n products i s a ppli ed to cover th eir bias a nd un certainty.

2. D e pletion Analysis No int egral bu rn ab l e a bs or ber s are co n side r e d for fre s h fuel Section 4 a.i i. Reacti v it y d ecre m e nt s h o uld for det er minin g the r eac ti vity decrement.

not inc lud e the integ ra l burnabl e absorbers.

2. D e pl e tion Analysis Boundin g va lue s within eac h b a tch grouping are u sed for a ll Sections 5.1 b.i. Bounding va lu es s hould b e paramet ers. throu g h 5.5 u sed. NET-28091-0003-01 , Revision 0 154 Section in this Guidance from DSS-ISG-2010-01 Implementation Report 2. Depletion Analysis The hi ghes t power i s used which l eads to hi g h e r mod erato r b.ii. Use the more limitin g a nd fuel tem p erat ur es , thus in creas in g k. To acco un t for Sections 5.4 bounding p arameter when a conflict low er power coast down , a Sm-149 correct i on is made. and 5.8 occ ur s. 2. Depletion Analysis Bounding va lu es within eac h batch group in g are u sed for all Sections 5 .1 b.iii. Non-bounding values are p a r ameters. th rough 5.5 outside sco p e ofISG. 2. Dep l e tion Analysis IP 2 a nd IP3 had s tand ard burnable absorbers , a nd W ABAs. Sections 5.2 c.i. All r emovab l e burnabl e IP 3 a l so had Hf flu x s uppr essors. All of these are a nd 8.9 absorbers mu st be co n s ider ed. conservatively accounted for. 2. Depletion Analys i s The analys i s in c lud es the maximum number ofIFBA rods at Section 5.2 c.ii. Limiting integra l burn ab l e th e hi ghest boron lo ading for eac h fuel batch gro upin g. absorbe r s s h ou ld be u se d. 2. Depletion Analys i s For d epletio n analysis, the maximum absor be r mat erial i s Section 5.2 c.iii. Model the burnab l e mod e l ed w ith th e m ax imum wa t er displacement.

For the p oo l abso rb e r s a oo ropriately.

a n a l ys i s, a ll burn a bl e absor b ers are removed. 2. Depletion Ana l ysis The depletion m o d e l correc tl y accounts for the in creased rate Section 5 c.i v. Consider co mp et in g effects of plutonium produ c ti on fro m in creased fast n e utron capture in U-238 2. Dep l etion Ana l ysis A ll hi storica l assemblies under D-Bank were id e ntifi ed an d Sections 5.5 d.i. Spectrum h a rden i ng from th e a pp ropr i a t e treatment i s a ppli ed. For c urr en t fuel , it i s rodded o p erat i on sho uld be assumed that a co nt rol rod was fully inserted for 2 GWd/T consi d e r e d. burnup. 2. Dep l et ion Analysis The axial profiles for asse mbli es without ax i a l blankets are Section 6.2 d.ii. Effect of contro l rods on the from NUREG/C R-6801. T h ese pro files include rodded cases. axia l burnup p rofi l e s hou l d be For bl anketed fu e l , actual burnup profiles are used. T h ese co n si d ere d profil es cove r any contro l rod e ff ects. 3. Crit i ca lit y Analysis The axial profiles for assemb li es without ax ial bl ankets are Sect i on 6.2 a. Ax ia l Burnup Profi l e from NUREG/C R-6 801 an d used in a conservat i ve manner. i. Use ofNUREG/C R-6 801 is For blanketed fu e l , actual burn up p rofiles are u se d. acceptab l e if d one properly 3. Crit icality Analysis Site-specific profi l es are used. Section 6.2 a. Axia l Burnup Profile ii. Site-specific profiles 3. Cr iticality Analysis For full l eng th fuel , results from uni form and NUREG/CR-Sect i on 6.2 a. Ax i a l Burnup Profi l e 6801 s h apes were compared at 10 GWd/T a nd the iii. Uniform profi l es NUREG/CR-680 1 shape was mor e li miting. The l owest burnup of a ny ful l l e n gth fuel i s greater than 10 GWd/T. For axia l l y blanketed fu e l , the lowest relative power at each node from a ll the blank eted fuel burn ed a t Indian Point was used to d eter min e the axial burnu p di s tributi on for eac h ax i a l blanket design. Since these relative powers were not renormalized, it covers both the top p eaked and center peaked condition.

3. Crit i ca l i t y Analysis The rack dimensions and m ater i a l s a r e provided b y the Section 3.1 b. Rack Model manu facture r (References 8 and 9). i. Mod e l input s s hou l d b e tracea b l e. 3. Critica lit y Analysis No c r edit i s t ake n for absorbe r p a n e l s. b. Rack Model ii. Effic i e n cy of t he n e utron absorber s h o uld be es tabli s h e d. NET-28091-0003-01, Revision 0 155 Section in this Guidance from DSS-ISG-2010

-01 Implementation Report 3. C ritica l it y A n a l ys i s No c r e di t i s t a k e n for absor b e r p a n e l s. b. R ac k M o d e l iii. C on se r va ti ve d e grad a ti o n s h o uld b e u se d. 3. C ri t i ca lit y Ana l ys i s T h e in terface a n a l ys i s adj u ste d th e bu rnu p fo r th e d iffere n ce S ec ti ons 7.6 c. Int e rfa ces -Use th e m axi m um in th e bi as a nd un ce rt a in ty. a nd 8.4.3 un ce rtainti es fr o m e i t h e r s id e. 3. C riti ca lit y Ana l ysis Th e r e a r e n o te mp ora r y storage l oca ti ons in th e IP 2 SF P. A ll Sec ti o n 9.1 d. N o rmal Cond iti o n s -All n o rm a l o p era tin g co ndit io n s a r e c ove r ed by t h e a nal ysis. n o rm a l co nditi o n s suc h as m ove m e nt of fu e l and in s p e c t i o n s s h ou ld b e co n s id ere d. 3. C riti c ality A n alysis A ll n o rm a l initi a l co ndi tio n s a r e co n s id ered. F o r exa m p l e , in Sec ti o n 9 e. Acc id e n t Con di t i o n s t h e mi sp l aced asse mbl y ana l ys i s it i s assu m ed th a t th e fuel i. S h o uld c o n s id er a ll norm a l e l e v a to r has a fr es h fu e l asse mb l y in it w h e n a n o th e r co nditi o n s a s b ase co nditi o n s. asse mbl y is mi s pl ace d n ext t o i t. 3. Crit i ca lit y A n a l ysis L a r ge m a r g in s ex i s t for a ll of t h e acc i dent co nditi o n s with t h e Sec ti o n 9 e. Acc id e n t Co ndi t ion s exce pti o n of th e multipl e mi s l oa d s c e n a r ios. Th ese sce n a ri os , i i. G rad e d a pp roac h m ay be h owever , s till m ee t th e targe t k 9 5 1 95 (see Sect i o n 9.5). t ake n w h e n c r e di t in g so l ub l e b oro n. 4. C riti ca l it y C od e Va lid a tion NURE G/C R-6698 i s fo ll owe d for th e va l i d a t io n. A pp e n d i x A NURE G/C R-6698 e nd o r sed 4. C r i ti ca lit y Co d e Va lid a ti o n T h e HT C c r it i ca l ex p e rim e n ts are in c l u d ed i n th e a n a l ys i s. A pp en di x A.3 a. Area of A pp lica bilit y i. In c l ud e t h e HTC c riti ca l s 4. Cri tic a lit y Co d e V a lid a ti o n T h e fin a l b ias a nd un ce rt ai nt y i s d e t e rmi ne d b y th e m ost A pp e n dix A a. Area o f A pp lica bilit y limitin g o f e ith e r t h e M OX a nd H TC c ri tica l s or th e U0 2 ii. Use a pp ro p riate c ri t i ca l s c riti ca l s. 4. Cr iti ca lit y Co d e Va lidation 328 fr es h fu e l c riti c al ex p eri m e nt s a r e u se d. 11 7 H TC A pp e n dix A a. Area of A ppli ca bilit y c riti ca l s are u se d as w e ll as 6 3 MO X cr it ica l s. G roupin gs o f iii. S uffi c i e n t c riti ca l s fo r c ri t i ca l sets are a n a l yze d to co nfirm w h en th ey s h o ul d be a n a l ys i s a nd app ro pri a t e g roupin g. in c l ud e d i n th e se t as a w h o l e. 4. Crit i ca lit y Co d e Va lid a tion T h e l a r ge numb e r o f c riti ca l ex p e rim e n ts u sed a nd th e l arge A pp e n d i x A.2 a. Area o f A pp l i ca bilit y va ri a ti o n in c ri t i ca l co n figura ti o n s (geometry a nd m a t e r ia l) i v. B e s ur e th e se t i s n o t hi g hl y r e du ces th e co nc e rn a b o u t b e in g c orr e l ated. Th e a n a l ys i s co r re l a t e d. u se d 37 d i ffe r e nt se t s of expe rim e nt s th at we r e p e rform ed in 7 diff ere nt cr iti ca l fa ciliti es. 4. C ri t i c ality C od e Va lid a tion T h e t re nd a n a l ys i s i s p e r for m e d o n a ll of t h e maj o r A pp e nd ix A.2.5 b. T r e nd An a l ys i s p ara m e t e r s. T h e tr e nd a n a l ys i s fo un d th e b es t lin ea r tit. No A d e qu a t e , a pp ro pri a t e, n ot tr e nd s were rejec t e d t o b e co n serva ti ve. The m os t limiti ng r e j ecte d. bi as a nd un ce rtainty for th e area of a ppli ca bilit y i s appli ed ass umin g ei th e r that a ll t rends a r e r ea l or t h e r e a r e n o t re nd s. 4. Cr iti ca lit y Co d e Va lid a ti o n Th e s t a ti s ti ca l a pproa c h r eco mm e nd e d in NURE G/C R-6698 A pp e n dix A.2.5 c. S t a ti s ti ca l Trea tment i s u se d. T hu s th e v ari a n ce o f th e p o pul a ti o n a bout th e m ea n i. Use th e varia n ce o f th e i s u se d rat h e r t h a n th e va r ia n ce o f th e m ean. p o pulati o n a b o ut th e m ea n 4. C ri t i ca lity C o d e Va lid a ti o n Th e s t a ti s ti ca l approa c h r eco mm e nd e d in NURE G/C R-6698 App e n d i x A.2.5 c. St a ti s ti c al Treatme nt i s u se d. T h e co rr ec t co nfid e n ce fac t o r s a r e u se d. ii. U se co rr ec t co nfid e n ce fac t o r s. 4. Cr i t i ca lit y C o d e Va lida t i o n No rm a li ty t es tin g i s p e r fo rm e d a nd th e a pp ro pri ate s t a ti s ti ca l A pp e ndi x A.2.5 c. Stati s ti c al T r eat m e nt t r ea tment is a ppli e d. iii. Con s id e r N o rmalit y N E T-28091-0003-01 , Revision 0 15 6 Guidance from DSS-ISG-2010-01

4. Crit i ca lity Code Validation
d. Lum ed Fission Product s 4. Crit i ca lit y Code Validation
e. Code-to-Code Co mpari sons 5. Miscellaneou s a. Precedence
b. Reference s c. Ass um tions NET-28091-0003-01 , Revision 0 Im lementation Lumped fission products are not u sed. No code-to-code comparisons are used for va lidati on. However , CASM0-5 analysis was used to co nfirm that the ISG-2010-01 allowed 5% of the delta kerr of depletion is ade uate. Precedence is n ot u sed as a li ce n s in g basis. Reference s used were carefully chosen to be applicable to the point being made. Assum tions are identified.

Section in this Re ort 15 7 10.2 Fuel Reactivity Categorization The reactivity category for each fuel assembly must be determined prior to l oading in the IP 2 SFP. For IP2 Batches A through X, the reactivity category for each assembly is found on Table B. l and for IP3 Batches A through AA on Table B.2 of Appendix B. Fuel for IP2 and IP3 batche s beyond those in Appendix B are labeled " Batch Z". The equations to determine the categorization for Batch Z are: B 12 = (-6.26824 + 5.29367*E

-0.37 l 54*E 2) e -(0.1 29582 -002049 1 s*E + o.00205596*E*E l*C T _0_ 13 331 + 6.9037*E + 0.122068*E 2 B o.s ( 15.1405 _ 4.8 l l 33*E + 0. 753855*E 2) e -(0.1 2 1 252 -0.0 1 5099 1 *E + o.00121009*E*El*CT _ 1 6_2293 + where MRB E CT PF 14.0 159*E -0.687054*E 2 MRB = Bo.s + (PF -0.8) x (B1.2 -Bo.s) I 0.4 is the Minimum Required Burnup (GWd/T) is the U-235 initial enrichment (w/o) is the cooling time (years) Assembly Burnup I (Sum of Cycle Burn up s) If an assembly had an in serted hafnium flux suppression insert a n y time during its l ife, then 2 GW d/T bumup must be added to the MRB. If an assembly ha s any number of mi ssing fuel pins that have not been rep l aced by stain les s steel rods, then 4 GW d/T burnup must be added to the MRB. If an assembly was burned with a burnup averaged sol ubl e boron concentration of greater than 950 ppm , then 0.2, 0.3, 0.6, and 0.9 GWd/T must be added to the MRB for fuel Categories 2 , 3, 4, and 5 respectively.

If the fuel has a burn up greater than (MRB+ 11) the fuel is Category 5. If the fuel has a bum up greater than MRB but less than the Category 5 requirement the fue l is Category 4. If the fuel has a burnup greater than 28.5 GWd/T but less than the MRB the fuel is Category 3. If the fuel has a bumup greater than 2 1 GWd/T but l ess than 28.5 GWd/T the fuel is Category 2. If the fuel has l ess than NET-2809 1-000 3-01, Revision 0 158 Westinghouse Non-Proprietary Class 3 21 GW d/T or violates any of the requirements of Table 10.4 the fuel is Category 1 or Category 4 if it conta in s a control ro d. For a Batch Z assembly to b e s tor ed in the SFP , it must ha ve at l east 64, 48 , 32 , or 16 IFBA rods for enric hm ents less than or equa l to 5.0, 4.5 , 4.0, an d 3.5 w/o , respectively.

If an assembly has a control rod in it, its fuel category increases. If a Category 1 fuel assem bl y has a contro l rod in it , it is classified as Category 4 fuel. If a Category 2 , 3 , or 4 fuel assembly h as a control rod in it, t h e assemb l y becomes a Category 5 fuel assem bl y. Contro l rods that are required in the fuel l ayout (see Section 10.3) may not b e credited to raise the category of that fuel asse mbly (for exam pl e, a Category 2 assembly w ith an inserted control rod may not be placed in a cell in the control rod area that requires a control rod). The above loading requirements have been summarized below in Table 10.2. Table 10.2: Summary of Loading Requirements for Fuel Batch Z Minimum Burnup Requirement Category 1 Fresh fuel with at least 64 , 48, 32, 16, or O IFBA rods for enrichments less than or equal to 5.0 , 4.5 , 4.0 , 3.5 , and 3.0 w/o respectivel

y. (@ [ mg 10 B/inch] a , c or 2reater).

Also, burned fuel with less than 21 GWd/T burnup is Cate2or v 1. Category 2 Burned fuel assemblies with at least 21 GWd/T burnup , initial enrichment of 5.0 w/o or less. Category 3 Burned fuel assemblies with at least 28.5 GWd/T burnup, initial enrichment of 5.0 w/o or less. Category 4 Burned fuel assemblies whose loading requirements are determined from Table 8.4 or the curve fit described in Section 8.3.1. Also, any Category 1 assembly containing a control rod is Category 4. Category 5 Category 4 burnup requirement plus 11 GWd/T. Also, any Category 2, 3, or 4 assembly containing a control rod is Categor y 5. N ET-28091-0003-01 , R ev ision 0 15 9 10.3 Allowable SFP Cells for Each Fuel Category The ana l ysis of the IP2 SFP uses a cell dependent reactivity.

That reactivity by position is defined in Figure 10.1 by use of ce ll categories, required contro l rods , and two types of water holes. Each cell in the IP2 SFP has a predetermined minimum fuel category.

However, there are a few alternative configurations of the cell categories in the SFP. Figure 10.1 shows the primary assignment of the cell categories for the SFP. Only Category 5 fuel may be placed in a Category 5 cell. Although Category 5 fuel can be stored anywhere in the SFP, the outer two rows of Region 2, the checkerboard control rod area of Region 2, and the interface between Region 1 and Region 2 must contain only Category 5 fuel or remain empty (alternate arrangements are allowed as discussed in Section 8.5). A Category 4 cell can accept Category 4 or 5 fuel. A Category 3 cell can accept Category 3 , 4 , or 5 fuel. A Category 2 cell can accept Category 2 , 3 , 4 , or 5 fuel. A Category 1 cell can accept all fuel. Category 1 cells can replace an area of Region 1 on the Figure 1 0.1 arrangement (Region 1) using the following two rules: 1. Each Category 1 cell must be face adjacent with at least three water holes. 2. Each Category 2 cell may not have more than one face adjacent to a Category 1 cell. Figures 10.2 through 10.4 show examples of allowable Category 1 cell locations in Region 1. It is a l so permitted to create a checkerboard arrangement of Category 1 cells in Region 2. In order to prevent interaction with other portions of Region 2, the checkerboard area must have a row of water holes on all sides. The water outside of the rack counts as a row of water holes. Finally, a 3 by 3 block of cells where all of the ce ll s are water h o l es except the center cell may be c rea ted anyw h ere in the SFP, and any fue l assemb l y may be placed i n the center pos i tion (see Section 9.1). NET-2809 1-0003-01, Rev i sion 0 160 1 1

  • s , 1 a , 10 11 11 u 14 15 " 11 11 1, 10 u 21 n 24 zs z, l1 1a u 10 31 1.SMl1tltJ101112U1CU 09 ON OM t---t---t--t---11-+---t--+---t-+-il-+-+--+-ll-+-+-+-l-+-+-+-!ll-+-+-+-l-+-+-+-l-l l'-"1---t--t---t--t-t-,--,--,,-r, Ol E *~-+---+--t--+--+-,.-o-o~>-~

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  • category 1Fuel D category 2 Fuel D categorylfuel
  • category4Fuel category s Fuel category S Fuelw~h a r equired lull lengthRCCA

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---.t-:: ,-,t-H co BN B M Bl BK &I 8H 9G .. 8( BO BC .......,.._.__._.,_ ........ _..__..__.._...., / Cask Area 1-1--+-+-+-+--+--+-+--+--+-+-lt--t--+-+-+-+---+--+-+--+---t--t--i AB ._._.._.._ .......................

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-~ Figure 10.1: Fuel Categor y Location Requirement s (Base Cas e) N ET-28 091-000 3-0 1 , Re v i s i o n 0 1 6 1 H G F E D C B H G F E D C B H G F E D C B 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 2 6 27 28 29 30 31 ,. Figure 10.2: Refueling Arrangement 1 2 3 4 5 6 7 8 9 10 Figure 10.3: Max Cat 1 Arrangement 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 1 7 18 1 9 20 21 22 23 24 25 26 27 28 29 30 31 I I ~--+ . oc:-~-t I I ----r --i -!----L -~----*---------*->->--I A X Figure 10.4: Example Odd Arrangement N ET-28091-0003-01 , Re v ision 0 162 There are restrictions on what ca n b e p l aced in th e water h o l es or 50% water hol es. T h e wa t e r hol e ma y not h ave anything in it (except water) in the act i ve fuel area with the fo llowin g except ion s:

  • If the it e m i s m ade of s tainless stee l , Inconel or absorber mat eria l , it i s a l lo we d to di s pl ace up to 50% of the water a nd sti ll be sto red in a white cell ( 100% water ce ll).
  • The 50% water holes (pink cells) a llo w di sp l acement of the water b y any non-fuel m ateria l up to 50% volume fraction.

This wo u l d a ll ow, for exa mpl e , a co mpon e nt made entire l y of z ir co nium takin g up l ess than 50% of the vo l um e to be stored in a 50% water ho l e (pink cell).

  • The cell blo c k e r at location BC72 (in the first row of cells abo v e the cask area pit) ma y n ot b e moved an d is require d in order to c l assify the cells on either si d e of the ce ll blocker as Category 4 cells. 10.4 Fuel and Operating Requirements The actual fuel and operating con dition s are used in th e ana l ys i s of historical fuel (Batches A thr o u gh X fo r IP2 and B atches A through AA for IP3). Fue l b a t c h es after X for IP2 an d after AA for IP3 are called Batch Z. This sec tion de scri bes t h e requir e m e nt s for Batch Z. To m ee t the limitation s of this c riti ca lit y ana l ys i s, the fuel design mu st meet the d es i gn requirements g i ve n o n Tab l e 10.3. NET-28 091-0003-01, Revision 0 1 63 Westinghouse Non-Proprietary Class 3 I Table 10.3: Fuel Design Requirements for Batch Z assemblies Attribute Value Notes Maximum fuel pellet U0 2 stack density 95.0%TD This includes dishing a nd chamfering Fuel pellet OD (inches) 0.3659 Nominal F u e l clad OD (inc h es) 0.4220 Nomina l Fuel clad ID (inches) 0.3734 Nominal Fuel pin pitch (inches) 0.563 Nominal Guide tube OD (inches) 0.533 Nominal G uid e tube ID (inches) 0.499 Nominal Maximum enrichment ( wt% m u) 5.0 Maxim um blanket enriclunent (wt% m u) 4.0 Mi nimum bl a nket l ength (inches) 8 Requirement for enrichments Minim um number ofIFBA rods* 64 (IX loading) less than or equal to 4.5, 4.0, 3.5, and 3.0 w/o is 48, 32 , 16 , and O IFBA, resp ect ively Minim um IFBA l ength (inches -centered)* 12 8 0.00603 g 1 0 B/cm Design changes that increase Maximum W ABA l oading water displacement are not per rodlet covered. Maximum IFBA rods and 10 B l oadi n g 14 8 IFBA rods [ mg 1 0 B/inc h]"*c ( l .5X) per rod *These requirements are only for storage of fuel assemblies that h ave not been in the core. The depletion parameters are se lect ed to cover anticipated future operation, howe ver , verification is req uired. Table 10.4 lists the operating requirements from the depletion analysis for Batch Z (recently discharged and future fue l). The temperature and so lubl e boron assumptions are averages over the total burnup (mu l ti-cycle) for a given asse mbly. Ifan assemb l y is dep l eted such that any of the Table 10.4 parameters are not met, then the assemb l y would have to be classified as Category 1 fuel or classified as Category 4 fuel with a control rod inserted. NET-2809 1-000 3-01, R evision 0 164 Table 10.4: Fuel Assembly Operating Requirements Parameter Value Notes Maximum Core Inlet 542.6 O f Temperature Maximum Core 62.4 O f Delta T Maximum Operation

.::: 2 GWd/T This control rod inserted bumup covers rod s with Control Rod s inserted to any depth. Thi s is an a v era g e for all cycles in which the Maximum Bumup assem bl y was depleted.

If 950 ppm is exceeded, a s stated in Section 10.2, a bumup Averaged Soluble .:S 1200 ppm pe n a l ty of0.2 , 0.3 , 0.6, and 0.9 GWd/T must Boron be ad ded to the MRB for Categories 2 , 3, 4 , an d 5 respectively.

Average Power To cover reduced power operation at end of Durin g the Last 30 >50% c y cle prior to offload Days of Operation NET-2809 1-000 3-01 , R ev i s ion 0 165 References

[1] Criti c ality Saf e ty Analysis of th e Indian Point Unit 2 Sp e nt Fu e l Pool w ith Cr e dit for In se rt e d N e utron Absorb e r Pan e l s , NET-300067-01, Rev. 1 , C urti ss-Wright Nuclear Division, NETCO , Danbury , CT , February 2015. (Accession Number: ML15062A200)

[2] " Response to Request for Additional In formation Regarding the Indi an Point uclear Generating Unit No. 2 -Spent Fuel Pool Critica lit y Analysis ," Letter L 089 , Entergy Nuclear Northeast, Buchanan , NY , August 14 , 2015. (Accession Number: ML15261A527 , Non Proprietary

Attachment:

ML15261A528)

[3] " Indian Point Nuclear Generating Unit No. 2 -Staff R eview of NETCO Report NET-300067-0 1 , " Criticality Safety Analysis of the Indian Point Unit 2 Spent Fuel Pool with Credit for Inserted Neutron Absorber Panels ,"" US NRC , Washington , DC, November 23 , 2015. (Accession Number: ML15292Al61)

[4) Guidan ce for P e rforming Criti c ality Anal yses of Fu e l Storag e at Li g ht-Wat e r R e a c tor P owe r Plant s, Revision 2, Draft C , NEI 12-16 , Nuclear Energy Institute , Wash in gton , DC , August 2017. [5] K. Wood , "Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01 , Revision 0 , Staff G uidance Regarding th e Nuclear Criticality Safety Analysis For Spent Fuel Poo l s ," Access ion Number MLl 106 20086, Nuclear Regulatory Commiss ion , Rockvi ll e , MD , October 2011. [6] Code of Federal Regulations , Title l 0 , Part 50 , Section 68 , " Cr iti ca lit y Accident Requirements." [7] S ca l e: A Compr e h e nsiv e Mod e lin g and Simulation Suit e for Nucl e ar Saf ety Ana l ys i s and D es i g n, ORNL ff M-2005/39 , Vers i on 6.1 , June 20 11. Available from Radiation Safet y Information Computat ion a l Center at Oak Ridge National La b oratory as CCC-785. Includes 6.1.2 update dat e d 2/28/1 3. [8] " Rack Cons tru ction (SH'T. 2) R eg i o n l Storage Rack s ," Drawing Number 398 R ev 5 , March 1 9 , 1990 , Project No. 81000 , P. 0. No. 8-24470 , Holtec International , Mount Laure l , NJ. [9] " Rack Co n structio n (S H'T. I) Region 2 Storage R acks ," Drawing Number 400 R ev 5 , March 19 , 1990 , Project No. 81000, P. 0. No. 8-24470 , Holtec Int ernationa l , Mount Laure l , NJ. [ 1 OJ I N TRPN D 8: V e rification and Validation, CWN D , NETCO , Danbury , CT: November 20 1 7 [ 11] Indian Point Unit 3 Up dated Fina l Safety Analysis R eport (UFSA R), Revision 03, October 3, 2009. [12) Indian Point Unit 2 Up d ated Fina l Safety Analysis R eport (UFSAR), Revision 26 , 2016. NET-28 091-000 3-01 , Revi s ion 0 166

[ 13] Dale Lancaster, S e nsitivity Analys i s for Sp e nt Fu e l Pool Criticality, EPRI , Palo Alto, CA: 2014 , Technical Report Number: 3002003073. [14] C. E. Sanders and J.C. Wagner , Param e tri c Study of th e Eff ec t of Control Rods for PWR Burnup Cr e dit , NUREG/CR-6759, prepared for the US Nuclear Regulatory Commission by Oak Ridge National Laboratory , Oak Rid ge, Tenn., February 2002. [15] G. Radulescu , I. C. Gauld, and G. Ila s, SCALE 5.1 Pr e di c tion s of PWR Sp e nt Nuclear Fu e l I sotopic Composition s, ORNL/TM-2010

/44 , Oak Ridge National Laboratory, Oak Ridge, Tennessee , USA , March 2010. [ 16] Int e rnational Handbook of E v aluat e d Criti ca lity Saf ety B enc hmark E x p e rim e nts , NEA/NSC/DOC(95)3, Nuclear Energy Agency , OECD , Paris , September , 2010. [ 17] J. J. Lichtenwalter, S. M. Bo w man , M. D. DeHart , and C. M. Hopper , Criti c ality B e n c hmark Guid e for Li g ht-Wat e r-R e a c tor Fu e l in Tran s portation and Storag e Pa c kag es, NUREG/CR-6361 (ORNL/TM-13211), Spent Fuel Project Office, Office of Nuclear Material Safety a nd Safeguards , U.S. Nuclear R egu latory Commission , Washington , DC 20555-0001 , March 1997. [18] L. W. Newman , et al , Urania-Gadolinia

N ucl e ar Mod e l D eve lopm e nt and Critical Experiment B e n c hmark , BA W-1810 , Babcock & Wilcox , Utility Power Generation Divi s ion , Lynchburg , VA , April 1984. [19] G. Radulescu , I. C. Gauld , G. Ila s, and J.C. Wagner , An Approac h for Validating Actin id e and Fission Produ c t Burnup Cr e dit Criticality Saf ety Analys es -Isotopic Composition Pr e dictions, NUREG I CR-7108 , Office of Nuclear Regulatory Research , U.S. Nuclear Regulatory Commission , Washington, DC , USA , April 2012. [20] K. S. Smith , et al., B e n chmarks for Quant ifying Fuel R e a c tivity D e pl e tion Uncertainty , E PRI , Palo Alto, CA , Technical Report Number 1022909 (201 1). [21] D. B. Lancaster , Utilization of th e EPRJ D e pl e tion B e n chmarks for Burnup Credit Validation, EPRI, Palo Alto, CA, 1025203 (2012). [22] J.M. Scaglione , D. E. Mueller , J.C. Wagner and W. J. Marshall, An Approac h for Validating.

Actinide and Fi ss ion Produ c t Burnup Cr e dit Criti c ality Saf ety Ana lys esCriticality (k , JJ) Pr e di c tions , US Nuclear Regulatory Commission, NUREG/CR-7109 , Oak Ridge National Laboratory , Oak Rid ge, Tenn. (2012). [23] J.C. Dean and R.W. Tayloe, Jr., Guid e for Validation of Nuclear Criti ca l ity Safety Cal c ulational M e thodology , NUREG/CR-6698 , Nuclear Regulatory Commission , Washington , DC January 2001. [24] D. E. Mueller , K. R. Elam , and P. B. Fox , Evaluation of th e Fr e n c h Haut Taux d e Combustion (HTC) Critical Exp e rim e nt Data , NUREG/CR-6979 (ORNL/TM-2007

/0 83), prepared for the US Nuclear Regulatory Commission by Oak Ridge National Laboratory , Oak Ridge, Tenn., September 2008. NET-28091-0003-01 , Revision 0 167 (25] Design Input Record , EN-DC-149Rl4 14021 8, Indian Point , February 17, 2014 (26] B. B. Bevard , J.C. Wagner , and C. V. Parks , R ev iew of Information for Spent Nuclear Fu e l Burnup Confirmation, NUREG/CR-6998, prepared for the US Nuclear Regulatory Commission by Oak Ridge Nationa l Laboratory, Oak Ridge , Tenn., December 2009. (27] J.C. Wagner , M. D. DeHart, and , C. V. Parks , R eco mm endatio n s for Addressi ng Axial Burnup in PWR Burnup Cr e dit Analys es, US Nuclear Regulatory Commission, NUREG/CR-6801 , Oak Ridge National Laboratory, Oak Ridge , Tenn. (2003). (28] D. Hagrman , INTERPIN-3 User's Manual, SSP-01/430, Studsvik Scandpower, Inc. (2001). (29] R. E. Griffith to G. Delfini , "F uel Temperatures vs Burnup Curves for Indian Point Unit 2," Entergy lnter-Office Correspondence, CE020 13-001 07, August 15, 2013, s upp orted by Entergy Calculatio n package NEAD-SR-13

/023 , EC 46111. [30] Criticality Saf ety Evaluation of th e North Anna N ew Fu e l Storag e Area and Spent Fu e l Pool Allowing 5 wt% U-235 Enriched Fu e l , Nuclear Engineering and Fuel, Dominion Resources Services, Inc., November 2016. (Accession Number: MLl 7129A452)

(31] J.C. Wagner and C. V. Parks , Param e tric Stud y of th e Eff ec t of Burnabl e Poi so n Rod s for PWR Burnup Cr e dit, US Nuclear Regulatory Commission, NUREG/CR-676 1 , Oak Ridge Nationa l Laboratory, Oak Ri dge, Tenn. (2002). (32] C. E. Sa nd ers and J.C. Wagner , Study of the Effect of Int egra l Burnabl e Poison Rod s for PWR Burnup Cr e dit , US Nuclear Regulatory Commission, NUREG/CR-6760, Oak Ridge Nationa l Laboratory , Oak Ridge , Tenn. (2002). (33] C. V. Parks, M. D. DeHart, and J.C. Wagner , R e view and Prioriti z ation of Technical Issu es R e lat e d to Burnup Cr e dit for LWR Fu e l , US Nuclear Regulatory Commission , NUREG/CR-6665, Oak Ridge Nationa l Laboratory, Oak Rid ge, Tenn. (2000). (34] M. D. D eHart, S e nsitivity and Param e tri c Evaluations of Significant Aspects of Burnup Cr e dit for PWR Spent Fu e l Pa ckages, ORNL/TM-12973 , Lockheed Martin E n ergy Research Corp., Oak Ridge National Laboratory, May 1996. (35] "Poo l Layout Spent Fuel Storage Racks ," Drawing Numb er 397 Rev 4 , December 8, 1989 , Project No. 81000 , P. 0. No. 8-24470, Ho l tec International , Mount Laurel, NJ. (36] "Critica li ty Analysis for Soluble Boron and Burnup Credit in the Con Edison [the former licensee]

Indian Point Unit No. 2 Spe nt Fuel Storage Racks , NET-173-01, Septe mb er, 2001. (Accession Num b er: MLO 12680336)

(37] G.P. Sabo l , G. Schoenberger , and M.G. Balfour, " Impro ved PWR Fue l Claddi ng ," Materials for Adva n ce d Water Coo l ed Reactors, Proceedings of a Technical Committee Meeting , Plzen, Czechoslovakia , May 14-17 , 1991 , IAEA-TECDOC-665 , IAEA , VIENNA, 1992. NET-2809 1-000 3-01, R evis ion 0 168

[38] Sabol, G. P , Comstock, R. J., Weiner, R. A., Larouere, E, and Stanutz , R. N., "In-R eactor Corros ion Performance of ZIRLO and Zircaloy-4," Zir con ium in th e Nuclear Indust ry: T e nth Int e rnational S ym posium , ASTM STP 1245 , A. M. Garde and E. R. Bradley , Eds., American Society for Testing and Material s, Philadelphia , 1994 , pp. 724-744. [39] Garzaro lli , F., Manze l , R., Re schke, S., and Tenckhoff, E., "Review of Co rro sio n and Dimensional Behavior of Zircaloy under Water Reactor Conditions," Zirconium in th e Nuclear Indu stry (Fourth Conf e r e nc e), ASTM STP 681, Amer i can Society for Testing and Materials , 1979 , pp.91-106. [ 40] Y. Irisa , et a l , "Seg mented Fuel Rod Irradiation Program On Advanced Mate ri a l s For High Burnup ," An Int erna tional Topical M ee ting on Li ght Wat e r R eactor Fu e l Performanc e, Park City, Utah , Apri l 10-13 , 2000, American Nuclear Society, La Grange Park , Illin ois. [41] C. B. Lee, et al , " Post-irradiation Examination of High Bumup U02 Fuel ," Proceeding s of the 2004 International Meeting on L WR Fuel Performance , Orlando, Florida, September 19-22 , 2004, American Nuclear Society, La Grange Park , Illinois. [42] David Mitchell , Anand Gard e, and Dennis Da v is , "Optimized ZlRLOŽ Fue l Perfom1ance in Westi n g hou se PWRs ," Proceedings of th e 2010 LWR Fu e l P erfo rman ce Me e ting/Top Fu e l/WRFPM, Septem b er 26-29, 20 10

  • Orlando , Florida , USA, A m erican Nuclear Society, La Grange Park , Illin ois. [43] R. Manzel and C. T. Walker , " High Burnup Fue l Microstructure And lts Effect On Fuel Rod Performance

," An Int ernational Topical M ee ting on Light Wat e r R eac tor Fuel P erfo rmanc e, Park City, Utah , Apri l 10-13 , 2000, American Nuclear Society, La Grange Park, Illin ois. [44) Dennis Gottuso , Jean-Noel Canat , Pierre Mollard , " A Family of Upgraded Fuel Assemb li es for PWR ," Top Fu e l 2006 , 2006 Int e rnational M ee ting on LWR Fue l P e rformanc e, October 22-26, 2006, Sa l amanca, Spain , European Nuclear Society. [45] King , S. J., Kesterson , R. L., Yueh, K. H., Comstock, R. J., Herwig , W. M., and Ferguson, S. D., "Impact of H y drogen on Dimensional Stability of ZIRLO Fue l Assemb li es," Zirconium in th e Nuclear Indu stry: Thirt ee nth Int e rnational Symposium , ASTM STP 1423 , G.D. Moan and P. Rudling , Eds., ASTM Internation al, West Conshohocken , PA , 2002 , pp. 471-489. [ 46] R. L. Kesterson , S. J. King , and R. J. Comstock , " Imp act of Hydrogen on Dimensional Stab ilit y of Fuel Assemb li es," An Int e rnational Topical M eet ing on Light Wat e r R e actor Fu e l P e rformanc e , Park City, Utah, April 10-13 , 2000 , America n N ucl ear Society, La Grange Park , Illinois. [ 47] Morize , P., Baicry , J., and Mardon, J. P., "Effect oflrradiation at 588 Kon Mechanical Properties and Deformation Behavior of Zirconium Alloy Strip," Zirconium in th e Nuclear Indu stry: Seventh Int e rnational Symposium, ASTM STP 939. R. B. Adamson and L. F. P. Van Swam, E d s., American Society for Testing and Materials , Philadelphia , 1987 , pp. 101-119. NET-2809 1-00 03-01, Revision 0 169 j

[ 48] S t eve n J. Kin g, Micha e l Y. Young, Fabrice M. Guerout , a nd Ni ge l J. Fisher, " FrettinWear Beha vior of Z ircaloy-4 , OPTIN , and Z IRLO Fuel Rods and Grid Supports Und er Var ious Autoclave and H y drauli c Loop E nduranc e Test Co ndition s," Zi rconium in the N ucle ar Indu stry: Fourteenth International Symposium , ASTM STP 14 67 , P. Rudlin g and Bruce Kamm enzi nd , E d s., ASTM Int e rn ationa l , W est Cons hoho cken , PA, 2006, p. 826. [49] R e spo ns e To R e qu e s t For Additional Information R e garding Pr opos e d Technical Sp e cification Chang e For Sp e nt Fu e l Storag e (Non-Pr oprietary), R espo n se to RAI-23 , Lette r from D omi nion Nuclear Co nn ect i cut to the USNRC , Jul y 2 1 , 2015. (NRC Adams Access ion Number: ML15209 A729.) [50] G.E. Whit esi d es , "A Difficult y in Comp utin g th e k-Effective of th e World," Trans. Am. Nucl. Soc., 14 , pp. 680 (1971). [51] Brian C. Ki edrowski a nd Forrest B. Bro wn , " Diffi c ulti es Com putin g kin Non-Unifonn, Multi-Region Systems with Loose, Asymmet ri c Cou plin g," Pro cee din gs of th e 9th Int e rnational Conf e r e n ce on Nucl e ar Criti c ality Saf e ty (ICNC20 1 l), Edin bur g h , Scotland, September 19-22 , 2011. [52] NET-173-02. Rev. l , " Indian Point Unit 2 Spent Fuel Pool (SF P) Boron Dilution Ana l ys i s , September 2001. (Access ion Number: ML012680336)

[53] ANSVANS-8.1-199 8 (R2007), " Nucl ear Crit i ca lit y Safety in Operations w ith Fissionable Materials Outside R eac tor s ," A m er ican Nuclear Society , La Grange Park , Illinoi s. [ 54] Calc ulation Notebook, NET-901-02-08, SCALE 6.1. 2 Validation for Critica li ty Anal ys i s -Am e ndm e nt Ca l c ulations for IP 2 , R ev.0; CWN D , D anbury , CT. [55] K. L indqui st, et a l , Guid e lin e s for B ora.fl e x U se in Sp e nt-Fu e l Storag e Racks, E PRI , Palo A lto , CA, T echn ic a l R e port Number 103300 (1983). [56] NRC Memorandum from L. Kopp t o T. Co llins , " Guidance on the Regulatory Requirements for C riticali ty A naly s i s of Fue l S tora ge at Lig ht-W a t er R eac tor Power Plants," August 1 9, 199 8. (Access ion Number: MLl 1 088A0 1 3) [57] J.C. Wagner , " Impact of Soluble Boron Modeling for PWR Burnup Cre dit Criticality Safe t y Analyses," Trans. Am. Nucl. Soc., 89 , pp. 120 (2003). [5 8] D.R. Lide , CRC Handbook of Chemistr y and Physics , 85 1 h Edition , C RC Pr ess LLC, Boca Raton , FL. (2004). NET-28091-0003-01, R ev i sio n 0 170 Appendix A: Validation of SCALE 6.1.2 for Criticality Analysis Using Laboratory Critical Experiments A. 1. Overview This appendix detennines the computer code and cross-section library bia s and uncertainty in the k eff values calculated for the Indian Point Units 2 and 3 spent fuel pools when using SCAL E 6.1.2. [ 1] The bias and uncertainties detennined in this Appendix cover the major actinide s, absorbers , and structural materials for the spent fuel pool with fresh or burned fuel. This analysis use s the CSAS5 module of SCALE 6.1.2. [ 1] All of the analyses are performed using the 238 group ENDF/B-VII library (v7-238). The CSAS5 module executes the CENTRM and BONAM! programs for the resonance self-shielding calculations and KENO V.a for the Monte Carlo calculation of k. All of the computer runs use a large Monte Carlo sampling of at least 1500 generations and 6000 neutrons per generation.

This Appendix is divided into three sections:

1) U0 2 critical experiments, and 2) HTC and MOX critical experiments, and 3) Temperature Dependent critica l experiments. After these three sections is a summary section. A.2. U02 Laboratory Critical Experiments A.2.1 Introduction The validation consists of modeling 328 U0 2 critica l experiments and the determination of the bias and the uncertainty in the calculation of keff for U0 2 fuel. This v alidation follows the direction of NUREG/CR-6698 , " Guide for Validat ion of Nuc l ear Criticality Safety Calculational Methodology" [2]. The guide establishes the following steps for performing the validation
1. Define operation/process to identify the range of parameters to be validated
2. Select critica l experiment data 3. Model the experiments
4. Analyze the data 5. Define the area of applicability of the va lid ation and limit ations It further defines the steps of "Analyze the data" as: 1. Detennine the Bias and Bias Uncertainty NET-28091-0003-01, Revision 0 A-1
2. Identify Trends in Data, Inc l uding Discussion of Methods for Establishing Bias Trends 3. Test for Norma l or Other Distributions
4. Select the Statistical Method for Treatment of Data 5. Identify and S u pport Subcr it ica l Margin 6. Calculate the Upper Safety Limit This approach is followed for this va l idation analysis.

A.2.2 Definition of the Range of Parameters to Be Validated The validation guidance document [2] states: "Prior to th e initiation of the validation activity, th e op e rating c onditions and parameters for which th e validation is to apply must b e identifi e d. Th e fissile isotop e , e nrichm e nt of fissile isotop e, fu e l d e nsity , fu e l c h e mical form , typ e s of n e utron mod e rator s and r e fl ec tors , range of mod e rator to fissil e i s otop e , n e utron ab s orb e rs , and ph y sical c onfigurations ar e among th e param e t e rs to specify. Th e s e param e t e r s will com e to d e fine th e ar e a of applicability for th e validation e ffort. " Almost all pool applications have common neutronic characteristics and therefore can be validated together.

The racks are assumed to be flooded with water at near room temperature and below 100 °C. The fuel is l ow enriched u ranium dioxide (less than or equal to 5.0 wt% U-235). The fuel is in pellets with a density of greater than 92% of the theoretica l density. The only significant neutron moderators are water and the oxygen in the fuel pellet. The neutron absorbers credited are boron (as plates, perhaps rods , or in so l ution) and Ag-In-Cd contro l rods. The reflectors are water, steel , or concrete. The fue l is in assemblies , but the ana l ysis is also valid for disassembled assemb l ies. The assembly arrangement can vary by design from totally isolated assemblies to a close packed array of assemblies. A.2.3 Selection of the Fresh U02 Critical Benchmark Experiments The U0 2 benchmarks that were se l ected met the fo ll owing criteria:

  • Low enriched (5 wt% U-235 or less) U0 2 to cover the principle isotopes of concern.
  • Fuel in rods to assure that the heterogeneous analysis used in SCALE also is applied in the benchmark analysis.
  • Square lattices to assure the lattice features of SCALE used in the rack analysis are also modeled in the critical benchmarks selected.
  • Presence of boron as soluble boron , borated steel , boron bearing rods , sheets of aluminum with boron , or Boraflex TM.
  • No emphasis on a feature or materia l not of importance to the rack analysis. The OECD/NEA Int e rnational Handbook of Evaluat e d Criticality Saf e ty B e nchmarks Experiments

[3] is now considered as the appropriate reference for criticality safety benchmarks.

This handbook has reviewed the available benchmarks and evaluated the uncertainties in the experiments.

The appropriate modeling is presented. All of the experiments used in this validation except some experiments for Ag-In-NET-28091-0003-01, Revision O A-2 Cd control rods were taken from this handbook.

Volume IV of the handbook is for low enriched uranium systems. The section of Volume IV of interest to this validation is the " Thermal Compound Systems." All of the experiments selected are numbered LEU-COMP-THERM-OXX.

This validation will refer to the experiments LEU-COMP-THERM-OXX as just XX where any leading zero is not included.

There are more critical experiments in the handbook that meet the requirements for this validation than would be necessary to use. However, most of the applicable available benchmarks were used. There are 95 sets of benchmarks in the 2016 version of the handbook. 25 of these were eliminated since they were for hexagonal arrays. Five more were eliminated due to high enrichments. Seven experiments were not for light water moderated U0 2 fuel rods. Four experiments were eliminated due to high uncertainties.

Five more were eliminated since they depend on features not in the spent fuel pool. This leaves 49 benchmark sets of which 35 were used for this validation.

The 14 unused benchmark sets were reviewed to be sure that there was no feature of the experimental set that was missing in the selected 35 sets. The international handbook only had two sets of experiments from one laboratory with Ag-In-Cd control rods. In order to fully ascertain if Ag-In-Cd control rods introduce a bias a search of additional criticals was performed. The first compilation searched was NUREG/CR-6361 [5], which was a report for the validation of SCALE. In that report were Ag-In-Cd critical experiments from two reports, BA W-1810 [6] and WCAP-3269 (data taken from NUREG/CR-6361).

Finally , Lawrence Livermore Laboratory compiled critical experiments from the ANS national meeting. [7] A search of this source did not provide any additional Ag-In-Cd critical experiments. The selected 35 benchmark sets from the international handbook and the 2 additional sets for Ag-In-Cd rods include critical experiments from eight different critical experiment facilities.

The fuel was mainly clad in aluminum but experiments with stainless steel and zirconium cladding were also in the set. The critical benchmark sets generally contained multiple experiments but not all cases from each critical benchmark set is used. In some sets there are experiments that emphasize features that are out of scope of this validation such as lead or copper reflectors. The 37 selected benchmark sets resulted in 328 experiments that are used for the statistical analysis.

85 experiments used boron (soluble or in absorber plates). A later section will evaluate the area of applicability provided by this selection of critical benchmarks. Table A. l provides a summary of all of the low enriched thermal experiments (non-U metal) from the OECD/NEA handbook [3] and why some experiments were not used. Table A.l: Selection Review of OECD/NEA Criticality Benchmarks (All Experiments Start With LEU-COMP-THERM-)

Benchmark Description Lab Selected?

Number WATER-MODERATED U(2.35)02 FUEL l RODS lN 2.032-C M SQUAR E-PIT C HED P N L All 8 ARR A YS WATER-M O DERATED U(4.31)0 2 FUEL 2 RODS lN 2.54-CM SQUAR E-PIT C HED PNL All 5 ARRAYS NET-28091-0003-01 , Revision 0 A-3 Benchmark Descr i ption Lab Selected?

Number WATE R-MODE R ATED U(2.35)02 FU E L No n e. Gd impmit y n o t we ll 3 R O D S IN 1.68 4-C M SQUARE-PITC H ED P N L known. No t b e n c lun a rk ARRAYS (GA D O LI N I UM WATE R q uality. IMPUR IT Y) WATE R-MODERATED U(4.31)02 FUEL No n e. Gd impu1it y n o t we ll 4 RODS I N 1.892-C M SQUARE-PITC H E D P NL kn ow n. No t b e n c hm ark ARRAYS (GA D O LI N I UM WATE R qua li ty. IMPUR I T Y) C RITl CAL EX PERIM EN TS WITH L OW-No n e. No sa mpl e SCALE EN RI C H E D U RAN I UM DI OXIDE FUE L 5 ROD S IN WATER CONTA I NING P N L decks. So lubl e G d n o t u se d in DI S SOL YE O GADOLIN I UM p oo l s. C R I T I CA L ARRAYS OF LOW-EN R I C H ED 6 U02 FUEL R O DS W I T H WA TER-TO-FUEL J AEA A ll 1 8 VO L UME RATIOS R ANGING FROM 1.5 TO 3.0 WAT ER-REF L ECTE D 4.738-WT.%-O nl y 4 cases u se d r es t ar e in 7 EN RI C H ED URAN I UM DI OXIDE FUEL-Ya ldu c RO D ARRAYS h exago n a l arra ys. C R I TI CAL LATT I CES OF U02 FUEL ROD S 8 AN D P ERTURBING RODS I N BORATED B&W A ll 1 7 W A T E R WATER-MODERATED R ECTANGU LAR C L US TER S OF U(4.3 1)02 FUEL RODS (2.54-2 1 cas es u se d. Did n ot 9 C M PI TCH) SE PARAT E D BY S T EE L , P NL includ e Co pp e r c a ses s in ce n o BORAL , CO PP ER , C ADMI UM, ALU MINUM , coppe r in p oo l s. OR ZrRC ALOY-4 P LATES WATER-MODERATED U(4.3 1)02 FUEL 22 cas es u se d. Did n o t u s e 1 0 R O D S R EF LECT E D BY TWO LEA D , P NL l ea d cases s in ce n o l ea d in URANIUM , OR S T EE L WAL LS poo l s. C RITl CAL EX P ER IM EN TS SUPPORTING II CLOSE PR OXIM I TY WATER STORAGE OF B&W A ll 1 5 POWER REACTOR FUEL (PART I -ABSO RB ER ROD S) WATER-MODERATED RECTANGULAR CLUSTE RS OF U(2.35)02 FUE L No n e. Gd impurit y n ot well 1 2 ROD S(l.6 84-C M P I T C H) SE P ARATE D BY P N L kn own. Not b e n c lun a rk STEEL , BORAL, BOROFLEX , qua li ty. CADM I UM , OR CO PP ER PLATES /G A DOLI N I UM WATER I MPUR I TY) WATER-M ODE RAT E D RE CTANGU L AR CLUSTERS OF U(4.3 1)02 FUEL ROD S 1 3 (1.892-CM PITCH) SE P ARATED BY STEEL , P NL 5 cases u se d. Did n o t u se th e BORAL , BORO FL EX , CADM I UM, OR 2 ca s e s w ith co pp e r. COP P ER PLAT ES , WITH STEE L REFLECTING WALLS WATER-REF L ECTE D ARRAYS OF No n e u s ed. Hi g h b o ron U(4.3 1)02 FUEL RODS (1.890-CM AND 14 1.7 1 5-CM SQUARE P I T C H) I N BORA T E D P NL co nt e nt un ce rtain ty. Not WATER b e n c lunark qua l it y. TH E VVER EX P E RIM ENTS: REGU L A R 1 5 AN D P ERTU RBED H EXAGONA L KFK I No n e u sed due to h ex a rr ays. L A TTI CES OF LOW-EN R I C HED U02 FU E L RODS IN LI G H T WATER WATER-MODERATE D RE CTANGU L AR C LUST E RS OF U(2.35)0 2 F UE L RODS 26 cases u se d. Did n ot u se 1 6 (2.032-CM PITCH) SE P ARATED BY S T EEL , P N L th e 6 co pp e r cases BORA L , COPPE R , CADMnJM , ALU MIN U M , OR Z rRCA L OY -4 PL ATES WATER-M ODE RAT E D U(2.35)02 FU E L 23 cases u se d. Did n o t u se 1 7 RODS REFLECTED BY TWO LEAD , P NL th e 6 cases w i th a l ead URAN I UM , OR STEE L WALLS r e fl ec t o r. NET-28091-0003-01 , Re v ision 0 A-4 Benchmark Description Lab Selected?

Number LIG H T WATER MO D ERATED AND 1 8 REFLECTED LOW ENR I CHED URAN I UM W i nfrith No n e used. Co m p l ex system. D I OXlDE (7 WT.%) ROD LATT I CE WATER-MODERATE D H EXAGONALLY 1 9 PIT CHED LATT I CES OF U(5%)02 K u rc h a t ov In stitute No n e used d u e to h ex arrays. S T AINLESS STEEL C L AD FUEL RO D S WATER-MODERATE D H EXAGONALLY 20 PIT CHED PART I A LL Y FLOODE D K u rc h a t ov In s t itute None used d u e t o h ex arrays. LATT I CES OF U(5%)02 Z I RCONIUM CLAD FUEL RODS , 1.3-CM PIT C H H EXAGONALLY P I TC H ED PART I ALLY FLOODED LATT I CES OF U(5%)02 2 1 ZIRCONIUM CLA D FUEL RODS K u rc h a t ov In s titute None u sed d u e t o h ex arrays. MO D E R ATED BY WATE R WITH BOR I C AC ID UN I FORM WATER-MODERATED 22 H EXAGONALLY P I TC H ED LATT I CES OF K u rc h atov Institute None u sed due to h ex a r rays. RODS WITH U(l 0%)02 FUEL 23 P ARTIALLY FLOO D ED UN I FORM K u rc h a t ov Institute No n e u se d du e t o h ex arrays. LATT I CES OF RODS W ITH U(l 0%)02 FUEL WATER-MODERATE D SQUARE-P I TCHED Did n ot u se e ith e r case du e t o 2 4 UN I FORM LATT I CES OF RODS W I TH K ur c h a t ov In stitute I O wt% U-235 e nri c lu nent U( I 0%)02 FUEL WATER-MODERATE D H EXAGONAL L Y 25 PIT CHED LATT I CES OF U(7.5%)02 Ku r c h atov Institute None u se d d u e t o h ex arrays. STAINLESS-STEEL-C L AD FUEL RODS WATER-MODERATED U(4.92)02 FUEL 26 RO D S IN 1.29 , 1.09 , AN D 1.01 CM P I TC H I PPE No n e used du e t o h ex a 1 rnys. H EXAGONAL LAT T I CES AT DIFFEREN T TEM P ERATURES WATER-MODERATED AND LEAD-None use d du e t o l ead 27 REFLECTED 4.738% ENR I C H ED URAN I UM Va l duc reflecto r. DIOXIDE ROD ARRAYS WATER-MODERATED U(4.3 1)02 FUEL 28 RO D S IN TRIANGU L AR LATTICES W l T H PNL None used d u e t o h ex arrays. BORON , CADMIUM AN D GADO LI N I UM AS SOLUB L E P O I SONS WATER MODERATED AND WATER No n e used. hf pl a t es cases REFLECTED 4.74% ENR I CHED URAN I UM without Hf h ave th e same 29 D I OX ID E ROD ARRAYS SURROUNDED BY Va l duc p i tc h and pin as b e n clunark 7 HAFN I UM P LATES above. No s i g nifi ca n t additio n a l va l u e. VVER Phys i cs Expe rim e nt s: Reg ul ar H exago n a l ( l.27-c m Pit c h) Latt i ces of L ow-30 E nri c h ed U(3.5 Wt.% 235U)02 F u e l Ro d s in K ur c h a t ov In st i tute No n e u sed du e t o h ex a 1 rny s. L i g h t Water at D i ffe r e nt Core Cr iti ca l D im e n s i o n s WATER-MODERATE D H EXAGONALLY 3 1 PI TCHED PARTIALLY FLOODED Ku r c h a t ov In stitute No n e u se d due t o h ex arrays. LATT I CES OF U(5%)02 ZIRCON I UM-C L AD FUEL RODS , 0.8-CM P I TC H UN I FORM WATER-MODERATED 32 LATT I CES OF RO D S W ITH U{l0%)02 F UEL K ur c h atov In stitu t e No n e used d u e t o h ex arrays. IN RANGE FROM 20°c TO 274°C REFLECTED AND UNREFLECTED 33 ASSEMBLIES OF 2 AN D 3%-ENR I CHED ORNL None used. NotU0 2 URAN I UM FLUOR I DE I N PARAFFIN FOU R 4.738-WT.%-EN RI CHED URAN I UM 6 cases use d. Di d n o t use D l OX lD E ROD ASSEMB L IES CONTA I NE D cases wit h gap l ess th a n 2.5 3 4 IN CADMIUM , BO R ATE D STAIN L ESS Va l du c c m du e to hi g h un ce rt a in ty. STEE L , OR BORAL SQUARE CAN I STERS , D i d not u se C d pl ate cases WATER-MODERATED AND-REFLECTED s i nce Cd pl ates n o t in poo l. N ET-28 091-0003-01 , R ev i s i o n 0 A-5 Benchmark Description Lab Selected?

Number CR I T I CAL ARRAYS OF LOW-ENR I C H E D Use d 2 cases. Did n o t u se th e 35 U02 FUEL RODS IN WATER WIT H J AEA case with di sso l ved Gd. (n ot SOLUBLE GADO LINI UM OR BORON P O I SON l ike poo l). T H E VVER EX P E RCM ENTS: REGULAR 36 AND P ERTU RB E D H EXAGONA L KFKI No n e u se d due t o h ex arr ays. LATT I CES OF LOW-ENR I C HED U02 FUEL RODS IN LIG H T WATER -P a rt 2 WATER-MODERATED AND PARTIALLY 37 CONCRETE-REFL EC T ED 4.738-WT.%-Va ldu c No n e u sed. No S i g nifi ca nt ENR I CHED URAN I UM DI OXIDE ROD Va lu e add e d. ARRAYS WATER-MODERATED 4.738-WT.%-

No n e u sed. Use d a borat ed 38 EN RI C H ED URAN I UM DI OX LD E ROD Va l duc co n crete reflec t o r (not l ike ARRAYS NEXT T O A BORA TED CONC RETE SCREEN p oo l). IN COM PL ETE ARRAYS OF WATER-39 REFLECTED 4.738-WT.%-ENR I C H E D Va ldu c Used a ll 1 7 cases. U R AN I UM DIOXIDE FUEL-ROD ARRAYS FOU R 4.738-WT.%-EN RI C H ED URAN I UM DIO X ID E ROD ASSE MBLLES CON TA I NED 4 0 IN BOR ATED STA I N L ESS STEEL OR Va ldu c Use d 4 cases. Did n o t u se BORAL SQUARE CAN I STERS , WATE R l ea d re fl ec t or cases. MODERATED AN D REFLECTED BY LEAD OR STEEL STO R AGE AR R AYS OF 3%-EN R!C H E D D id n ot u se the 5 cases du e t o 41 LWR ASSEMBLIES:

THE C R I STO II Ca d arac h e complex geo m e tr y. EX P ER I MENT IN Tl-I E EO L E REACTOR W A T ER-MO D E RATED R ECTANGU LAR C LUST ERS OF U(2.35)02 FUEL RODS 4 2 (1.684-CM P I TC H) SEPARATED BY STEEL , P NL Used 5 cases. Did n o t u se BORAL , BOROFLEX , C ADMIUM , OR cop p e r cases. COPPER PLAT ES, W I TH STEE L REFLECTING WALLS CR ITI CAL LOADING CONF!GURA TIO NS Use d o nl y o n e case. Rest of 4 3 O F TH E IPEN/MB-0 1 REA CTO R W ITH A !P EN cases were n ot s i g nifi ca ntl y H EAVY SS-304 REFLE CTO R diff e rent. C RITI CAL LOADING CON FIGURAT I ONS Use d o nl y o n e case. Rest of 44 OF Tl-I E IP EN/MB-0 1 REACTOR W I T H U02, !P EN cases were n o t s i g nifi ca ntl y S TAINL ESS STEE L AND COPPER RODS differ e nt. PL EX I GLAS OR CONCRETE-REFLECTE D No n e u se d s in ce n o t pin 45 U(4.4 6)308 W I TH H/U=0.77 AN D R oc k y F l ats INT ERS T I T I A L MODERATION geo m etry. C RITI CAL LOAD ING CONF I GURAT I ONS 4 6 OF T H E I PEN/MB-0 1 REA CTO R IP EN Use d 1 7 cases th a t did n ot CONSIDERING TEMPERATURE h ave co oper p i n s. VAR I AT I ON FROM 14°C TO 85°C FUEL TRANSPO RT FLASK C RITI CA L 4 7 BENCHMARK EXPE RCM ENTS WITH LOW-W infr ith No n e u se d. 3 co mpl ex cases. EN RI CHED URAN I U M DI OX ID E FUEL LI G HT WATE R M O D ERATED AN D 48 R E FL ECTED LOW-ENR I C H ED (3 WT.% W infrith A II 5 c a ses u s ed 235U) URANIUM DIOXID E ROD LATT I CES MARACAS PROGR AMME: P OLYT H ENE-REFLECTED C RITICAL CONF I GURA TIO NS No n e u se d. P ow d e r rather 49 W ITH LOW-ENR I C H E D AND LOW-Va ldu c than p e ll e t s. No t s imil ar to MODERATED URAN I UM D I OX ID E p oo l s. P OWDER, U/5)02 14 9SM SOLUTION TANK I N T H E M IDDL E 7 cas es u se d. Did n ot u se 50 OF WATER-MODE R ATED 4.738-W T.%-V aldu c cases w ith di sso l ved Sm. EN RI C H E D U R AN I UM DI OXI D E ROD Thi s i s n o t typica l o f p oo l s. ARRAYS NET-28091-0003-01 , R evision 0 A-6 Benchmark Description Lab Se lected? Number CR I T I CAL EXPERrMEN TS SUPPORT I NG 9 cases used. Di d n o t u se cases wit h the b ora t ed Al 5 1 CLOSE PROXrMITY WATER STORAGE OF B&W p l ates since primai y so ur ce POWER REACTOR FUEL (PART II -I SOLAT I NG PL ATES) li s t ed a high uncertainty in th e boron co nt e nt. URAN I UM DIOX I D E (4.738-WT.%-

52 ENR I CHED) FUEL ROD ARRAYS Val du c None used due to h ex arrays. MODERATED AND REFLECTED BY GADOLIN I UM N I TRATE SOLUT I ON VVER PHY S I CS EXPE R[MENTS: REGULAR HEXAGONAL ( 1.27 CM PITCH) LA TT I CES 53 OF LOW-ENRICHED U(4.4 WT.% 235U)02 K ur c h atov In stitute No n e u sed due t o h ex arrays. FUEL RODS IN UGI-IT WATER AT D I FFERENT CO R E CR ITI CAL D I MENS I ONS CR I T I CAL LOADING CONF I GURA TION S Use d on l y o n e case. Rest of 54 OF T H E IP EN/MB-0 1 R EACTOR W I TH U02, I P EN cases were not s i g nifi ca ntl y AND U02-G d 203 RODS differe nt. U G I-I T-WATER MODERATED AN D Neit h er case u sed. Comp l ex 55 REFLECTED LOW-ENRICHED URAN I UM W infrith (3 wt.% 235U) D I OX ID E ROD LATT I CES geo metry. CR I T I CAL EXPE RIM ENT WITI-I BORAX-V No n e u s ed. Co mpl ex BWR 56 BO I LING WATER REACTOR TYPE FUEL !NL ASSEMB LI ES geo m e try. 4.738-WT.%-EN R I C H E D URAN I UM 57 D I OX I DE FUEL ROD ARRAYS REFLECTED Va ld uc No n e u se d. No S i g nifi ca nt BY WATER fN A DRY STORAGE V a lu e a dd ed. C ONF I GURAT I ON C R I T I CAL LOAD I NG CON FIG URA TI ONS None u se d. No S i g ni ficant 58 OF Tl-I E fPEN/MB-01 REACTOR W!TI-I I P EN L ARGE VO ID I N Tl-IE R EFLECTOR Va lu e ad d ed. 59 Not included in 20 I O Handb ook RBMKGRAPI-I IT E R EACTOR: UN I FORM CONF I GURAT I ONS OF U(I.8 , 2.0 , o r 2.4% 235U)02 FUEL ASSEMBLIES , A D 60 CONF I GURAT IO NS OF U(2.0% 235U)02 Kurc h atov In st i tute None used. RBMK-n ot ASSEMBUES W I T I-I EM PTY C H ANNELS , typ i ca l ofLWRs WATER COLUMNS , A D BORON OR THORIUM ABSORBERS , W ITI-I OR W I THOUT WATER I N C H ANNE L S VVER PHYS I CS EX PERIM ENTS: HEXAGONAL (1.27-CM PI TCH) LATT I CES OF U(4.4 WT.% 235U)02 FUEL RODS I N 6 1 LI G HT WATER, P E RTURB E D BY BORON , K ur c h atov ln s titute No n e u s ed d u e t o h ex a rr ays. H AFN I UM , OR DYSPROSIUM ABSO RB ER RODS , O R BY WATER GAP WITI-I/WITI-IOUT EM PTY ALUMIN I UM TUBES 2.6%-ENR I CI-IED U02 ROD S IN UGI-IT-62 WATER MOD ERA TOR W I TI-I BORA TED JAEA None u s ed. No S i g ni ficant STAINLESS STEEL P LATE: SINGL E Va lu e adde d. ARRAYS UGI-I T-WATE R MODERATED AN D 63 REFLECTED LOW-ENR I C H ED URAN I UM Winfrith No n e u s ed. No S i g n ificant (3 wt.% 235U) DIOXID E R OD LATT I CES Va lu e a dd e d. W ITH DIS CRETE P O I SON-ROD AR R AYS VVER PH YSICS EX PERrME NTS: R EGU LAR H EXAGONAL (1.2 7 CM PIT C H) LATTICES 64 OF LOW-ENR I C HED U(2.4 WT.% 235U)02 K ur c h atov In stitute No n e u sed due t o h ex arrays FUEL RODS fN UGI-IT WATER AT D I FFERENT CORE CRITIC AL DIMENS I ONS NET-28091-0003-01 , R evision 0 A-7 Benchmark Descript i on Lab S el e cted? Number CR I T I CAL CONFIGURAT I ONS OF 2.6%-ENR I C H ED U02 ROD ARRAYS I N LIG H T-No n e u se d. No Sig ni ficant 65 WATER MODERATOR WITH BORA TED JAEA STArNLESS STEEL PLATE: COU PLED Value added. ARRAYS PLEX I GLAS-REF L ECTED, CONCRETE-66 R EF LECTED , OR TH I N STEEL-REFLECTED R ocky Flats None u se d. No t a n array of U(4.46)308 WITH H/U=0.77 AND HEU rods. DR I VERS CRITICAL LOADrNG CONFIGURAT I ONS 67 OF TH E IPEN/MB-0 1 REA CTO R !P EN None u sed s ince Mo l y rods COMPOSED OF FUEL AND are not us e d in pool. MOLYBDENUM RODS PLEX I GLAS-REFLECTED, CONCRETE-68 REFLECTED , OR TH I N STEEL-REFLECTED Rocky Flats None u se d. Not an array of U(4.48)308 WITH H/U=l.25 OR H/U=2.03 rods. AND HEU DR I VERS PLEX I GLAS-REFLECTED U(4.48)308 W I T H None us ed. Not a n array of 69 H/U=l.25 OR H/U=2.03 AND INTERS TI TIAL Rocky Flat s MODERATION rods. VVER P HYS I CS EXPERCMENTS:

REGULAR H EXAGONAL (1.l O-C M P I TCH) LATT I CES 70 OF LOW-ENRI C H E D U(6.5 WT.% 235U)02 K u r c hatov I n s titute None u se d due t o h ex arrays. FUEL RODS rN LIG H T WATER AT D LF FE RENT CORE CRITIC AL D I MENS I ONS LOW MODERATED 4.738-WT.%-

7 1 ENR I C HED URAN I UM DIOXID E FUEL Valduc All 4 cases u sed. ROD ARRAYS UNDER-MODERATED 4.738-WT.%-72 ENR I CHED URAN I UM DIOXID E FUEL Va l du c Used 3 cases. D i d not u se ROD ARRAYS REFLECTED BYWATER OR Po l yethylene re fl ecto r cases. POLYE TH YLENE UNDER-MODERATED 4.738-WT.%-73 ENR I C HED URANIUM DIOXID E FUEL Va l du c None u se d. No S i g n ificant ROD ARRAYS REFLECTED BY WATER Value added. W I TH HETEROGENEITIES MIRTE PROGRAM FOUR 4.738-WT.%-ENR I CHED URAN I UM-D I OXIDE FUEL-None u sed. 2 cases w ith out 74 ROD ARRAYS rN WATER SEPARATED BY Va l du c A CROSS-SHAPED SCREEN OF T I TAN I UM Ti scree n co u l d be u sed. (5 MM AND 1 0 MM THI C K) VVER PHYS I CS EXPER I MENTS: HEXAGONAL (1.1 0 CM PIT C H) LATT I CES 75 OF LOW-ENRI C HED U(6.5 WT.% 235U)02 Kurchatov In s titute None u s ed due to h ex arrays. FUEL RODS rN LIGHT WATER, PERTURB E D BY BORON ABSORBER RODS AND WATER HOL ES LI GHT WATER MODERATED AND 76 R E FLECTED LOW ENR I C HED URAN I UM W in frith No n e u se d. No S i gnifi ca nt (3 WT.% 235U) DIOX I DE ROD LATTICES Value added. W ITH EX-CORE DETECTOR FEA TVRE On l y one case u se d. Rest of 77 CR I T I CAL LOADrNG CON FIG URA T I ONS !PE N cases sa me m ater i a l s with OF THE IPEN/MB-01 REA CTO R sma ll modificat i o n of arrays. Not sufficie ntl v ind e p e n de n t. WATER-MOD E RATED SQUARE-P I TC H ED 78 U(6.90)02 FUEL ROD LA TT I CES WITH 0.52 Sandia None u se d. No S i g n ificant FUEL-TO-WATER VOLUME RAT I O Va l ue adde d. (0.855 CM P I TCH) WATER-MODERATED U(4.31)02 FUEL 79 ROD LATT I CES CONTA I NING RHOD I UM Sandia None u se d d u e to h ex arrays. FO I LS NET-28 0 9 1-0003-01, R ev i s i on 0 A-8 Benchmark Descr ip tion Lab Selected?

Number WATER-MODERATED SQUARE-P I TC H ED None used. No add iti o n a l 80 U(6.90)02 FUEL RO D LATT I CES W ITH 0.67 Sa n d i a S i g ni fica n t Va l u e a dd ed. FUEL TO WATER VOLUME RA TI O P WR TYPE U02 FUEL RODS W ITH ENR I CHMENTS OF 3.5 AND 6.6 WT.% S in g l e case n ot u se. U nu sual 8 1 W I T H BURNABLE ABSORBER ("OTTO ANEX HAHN" NUCLEAR S HI P PROGRAM , case. SECOND CORE) CR ITI CAL LOADING CONF I GURAT I ONS Use d on l y o n e case. Res t of OF TH E IP EN/MB-0 1 REACTOR W I TH LOW 82 ENRICHED FUEL AND BURNABLE !PEN cases were n ot s i g ni fica n t l y POISON RODS differe n t. CR I T I CAL LOADING CONF I GURAT I ONS Used on l y one case. Res t of 83 OF T H E I PEN/MB-0 1 REACTOR W ITH A I PEN cases were n o t s i g nifi ca n t l y BIG CENTRAL VOID diffe r e nt. CR I T I CAL LOADING CONF I GURAT I ONS 84 OF TH E I PEN/MB-0 1 REACTOR W ITH A I PEN Used t h e s in g l e case. CENTRAL C R UC I FORM ROD VVER P H YS I CS EX P ER[MENTS:

REGULAR H EXAGONAL ( 1.27 CM PI TCH) LA TT I CES 85 OF L OW-ENR I C H E D U(6.5 WT.% 235U)02 Kur c h a t ov In st i tute No n e u sed du e t o h ex a rr ays. FUEL RODS IN LI G H T WATER AT D IFFERENT CORE CR I T I C AL D I MENS I ONS VVER P H YS I CS EX P ER I MENTS: H EXAGONAL LATT I CES (1.275 CM PI TC H) 86 O F LOW ENR I C H ED U(3.6 , 4.4 WT.% NR I None used due t o h ex a rr ays. 235U)02 FUEL ASSE M BLIES I N LI G H T WATE R WITH H 3B03 VVER PHYS I CS EXPER I MENTS: H EXAGONAL LATT I CES (1.22-CM PI TC H) 87 OF L OW-ENR I C H ED U(3.6 , 4.4 WT.% NR I No n e u sed d u e t o h ex arrays. U235)02 FUEL ASSE M B L IES I N LI G H T WATER W ITH VAR I ABLE FUEL-ASSEMBLY PI TCH CR I T I CAL LOADING CONF I GURAT I ONS 88 OF TH E [PEN/M B-0 1 R EACTOR WITH I PEN Used a ll 35. HEAVY REFLECTORS COMPOSE D OF CARBON STEEL AND N I CKEL C RI T I CAL LOADING CONF I GURAT I ONS Use d o nl y o n e case. R est of OF TH E IPEN/MB-0 I REACTOR W ITH U02 89 AN D BORATE D STAINLESS STEEL I PEN cases were n o t s i g nifi ca ntl y P LATES differe nt. CR ITI CAL LOADING CONF I GURA TI ONS Use d on l y o n e case. R es t of 90 OF TH E IPEN/MB-0 I REACTOR W ITH U02 I PEN cases we r e n o t s i g nifi ca n t l y AND STA I N L ESS STEEL RODS diffe r e n t. CR ITI CAL LOAD I NG CON FI GURAT I ONS Used on l y o n e case. Res t of 9 1 OF TH E IP EN/MB-0 1 REACTOR W I T H U02 , I PEN cases were n o t s i g nifi ca ntl y STA I N L ESS STEE L AND GD203 R O D S di ffe r e nt. CR I T I CAL LOADING CONF I GURAT I ONS 92 OF T H E IPEN/MB-0 I REACTOR W ITH I PEN Used a ll 6. SOLUB L E BO R ON D EUTE R fUM CR ITI CA L ASSEMBLY W ITl-1 93 1.2% ENR I CHE D U R ANIUM VAR Y I NG P NC Not u sed since cases u se D20 COOLANT VO I D FRACT I ON AN D rathe r th a n H 20 LATT I CE PIT C H VVER PH YS I CS EX P ER[MENTS:

REGULAR H EXAGONAL (1.1 0 CM P I TC H) TWO-9 4 RE GI ON LATT I CES OF L OW-ENR I C H ED K u rc h a t ov In s t itu t e No n e u sed d u e t o h ex arrays. U(6.5 AND 4.4 W T.% 235U)02 FUE L R O D S IN LI GHT WATE R AT DI FFERENT CORE CR I T I CAL DI MENS I ONS 95 No t in c lud e d in th e 20 1 6 H a ndb oo k N E T-28091-000 3-01 , Re v i s i o n 0 A-9 B e nchm ark D escr iptio n La b Se l e c te d? N umb er PARTI A LL Y-R E FLE C TED WA T E R-Us ed a ll 1 9 (Ev e n thou g h MOD E R A T E D SQUARE-PIT C HED 96 U(6.90)02 F UE L ROD L A TI!CE S WITH 0.6 7 Sandi a h igh e nri c hment , adds an FU E L TO WA TER V OLUM E RA TIO indep e nd e nt l ab and E ALF (0.8 00 C M PIT C H) cove ra ge) TITANIUM A ND/OR ALUMTN U M ROD-R E PL ACE M EN T E XP E R I M E NT S I N FULLY-N on e u se d. Hig h enri c hment 9 7 R E FL E C TE D W A T E R-MOD E R A TED S andi a and Ti in m a n y c a ses. No S QUAR E-PIT C H E D U(6.9 0)0 2 FUE L ROD additi o n a l S ignifi ca nt V alu e LATTI C ES WITH 0.67 FUEL TOW AT E R a dd e d. V OL U M E R A T I O (0.8 00 C M P I T C H) A.2.4 Computer Analysis of the U02 Benchmark Critical Experiments SCALE input decks exist on the OECD/NEA handbook [3] disc for many of the critical experiments.

In general, t h ese input decks were used with minor modifications.

None of the decks ( except LCT-96) were for SCALE 6.1.2 or the ENDF/B-VII library. The number of neutrons per generation and the number of generations were, in general , too low. All of the decks were modified to 6000 neutrons per generation and 1500 generations or more. This was sufficient to make the Monte Carlo uncertainty to be 0.0002 or about one tenth the experimental uncertainty. The input decks matched the isotopic content given in the handbook but this was confirmed.

The geometric modeling in the decks also matched the descriptions in the handbook but this too was confirmed.

In short, a l though there was considerab l e help by starting with the input fi l es given in the handbook, the ownership of the files was taken , as required by NUREG/CR-6698 [2] and as stated in section 2.3: For s p ec ific c ritical ex p e rim e nt s, th e fa c ility or s it e ma y choo se to u se input fil es g e n e rat e d e ls e wh e r e to e xp e dit e th e validation pro ce ss. Th e s it e ha s th e r es pon s ibility for e nsurin g that input fil e s and th e options sel ec t e d ar e appropriat e for us e. R eg ardl e ss of th e s ource of the input fil e , th e s it e must hav e r ev i e w e d th e d esc ription of e a c h c riti c al ex p e rim e nt and d e t e rmin e d that th e r e pr e s e ntation of th e e xp e rim e nt , including simplifying assumptions and options, ar e consi s t e nt with th e int e nd e d us e. In oth e r words , th e s it e must a ss um e own e r s hip of th e input fil e. For LCT-8 the input decks were actually 2D mode l s. As part of the Internationa l Handbook independent review of LCT-08 eva l uation , Virgi n ia Dean, perfonned 3D ana l ysis, found a 0.002 bias , and declared it not " s i gnificant." (Appendix D of the evaluation in the International Handbook.)

0.002 is not insignificant when the bias from a ll of the critical experiments given in Reference l is only 0.0024. The LCT-08 evaluation provided all of the detail for a 3D analysis.

It was chosen to reanalyze LCT-08 with a 3D model. With the 3 D modeling the average calc ul ated ke rr of the LCT-08 cases is 0.9978 (the 2D mode l given on the In t ernational Handbook files yie l ded an average k e rr of 0.9970). BAW-1810 [6] reports 23 critical cores. This analysis uses 9 of these cores. Since these cores are not part of t h e i n t ernationa l h andbook, t h e cases were li mited to those re l ated to Ag-In-Cd cases. A ll of the Ag-In-Cd cores were se l ected as well as the cores that were the c l osest match where water holes replaced the Ag-I n-Cd rods. T h e B&W faci li ties were used for 3 evaluations in the International Handbook, LCT-08 , LCT-1 1 , and LCT-51. All three of t hese sets are used in the va l idation. Some BA W-1 8 1 0 cases are NET-28091-0003-01, Revision 0 A-10 used in NUREG/CR-6361 [5]. The input decks for the cases started with the NUREG/CR-6361 decks but changed the Al alloy clad to match the Al alloy atom densities reported in Table 18 of LCT -08. The input ignore s the bottom and top grids. The original source, BA W-1810, was carefully re viewe d to be sure that there was good agreement with the International Handbook and NUREG/CR-6361 input decks. Cases 1-2, 3-4, 5-6, 5A-6A and 8-9 are pairin gs where the only change was interchanging 16 Ag-In-Cd rods for water holes. The ma ximu m difference between any two cases is 0.00075. The mean differenc e is 0.00002 where the uncertaint y in each case i s 0.0000 8. It is clear that there is no significant difference in the ability to predict k e ff when there are Ag-In-Cd control rods pre sent. BA W-1810 do es not assign an experi mental uncertainty.

The mean uncertainties assigned to LCT-08, LCT-11, and LCT-51 are 0.0012 , 0.00251, and 0.00207 respectively. An uncertainty of 0.0025 (the larg est of the three) was assigned.

The WCAP-3269 input deck s are directly from NUREG/CR-6361 with modifications for the newer cross section librar y, sma ll changes in SCALE input format caused by a new er versio n of the code, and mor e neutrons per generation and generations. The uncertaint y assigned to th ese cases i s 0.004. This uncertainty estimate is one of the largest for the entire set of experiments.

The average uncertainty of all of the experiments is 0.0019. Table A.2 shows the results of the analysis of the 328 critical experiments, along with parameters th at are used to check for trends in the results. The spectra l ind ex, the E nergy of the Average Lethargy of the neutrons causing Fission (EALF) is a calculated va lue from the SCALE output. Table A.2: Critical Experiment Results with SCALE 6.1.2 and ENDF/B-VII Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. keff ID No. (wt% U-Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (~ k) LCT-1 1 2.350 1.2 70 2.032 0.09 64 0.003 0.9981 2 2.350 1.2 70 2.032 0.0957 0.00 3 0.9977 3 2.350 1.2 70 2.032 0.0950 0.003 0.9970 4 2.350 1.270 2.032 0.0955 0.00 3 0.9976 5 2.350 1.2 70 2.032 0.0942 0.003 0.9956 6 2.350 1.2 70 2.032 0.0952 0.003 0.9978 7 2.350 1.270 2.032 0.0934 0.0031 0.9974 8 2.350 1.2 70 2.032 0.0945 0.003 0.9964 LCT-2 1 4.310 1.415 2.540 0.11 32 0.002 0.99 71 2 4.310 1.415 2.540 0.1129 0.002 0.9987 3 4.310 1.415 2.540 0.1129 0.002 0.9984 4 4.310 1.415 2.540 0.1119 0.00 1 8 0.9979 5 4.310 1.415 2.540 0.1103 0.001 9 0.9962 LCT-6 1 2.596 1.417 1.849 0.23 66 0.002 0.9977 2 2.596 1.417 1.849 0.2432 0.002 0.9987 3 2.596 1.41 7 1.849 0.2495 0.002 0.9987 4 2.596 1.41 7 1.956 0.181 8 0.002 0.9984 5 2.596 1.41 7 1.956 0.1 871 0.002 0.9986 6 2.596 1.41 7 1.956 0.19 27 0.002 0.9983 7 2.596 1.41 7 1.956 0.19 77 0.002 0.9989 NET-28091-0003-01, Revision 0 A-11 Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. k ,rr ID No. (wt% U-Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (Ll k) 8 2.596 1.41 7 1.956 0.2028 0.002 0.9986 9 2.596 1.417 2.150 0.1359 0.002 0.9988 10 2.596 1.417 2.150 0.13 94 0.002 0.9988 11 2.596 1.417 2.150 0.142 7 0.002 0.9985 12 2.596 1.41 7 2.150 0.1462 0.002 0.9982 13 2.596 1.41 7 2.150 0.1497 0.002 0.9981 14 2.596 1.41 7 2.293 0.114 7 0.002 0.9988 15 2.596 1.417 2.293 0.11 74 0.002 0.9983 16 2.596 1.417 2.293 0.1200 0.002 0.9991 17 2.596 1.417 2.293 0.1228 0.002 0.9987 18 2.596 1.41 7 2.293 0.1254 0.002 0.9985 LCT-7 1 4.738 0.940 1.260 0.2411 0.0014 0.9959 2 4.738 0.940 1.600 0.1 090 0.000 8 0.9980 3 4.738 0.940 2.100 0.0708 0.0007 0.9976 4 4.738 0.940 2.520 0.0605 0.0008 0.9983 LCT-8 1 2.459 1.206 1.636 0.2845 0.0012 0.99 76 2 2.459 1.206 1.636 0.2502 0.001 2 0.9984 3 2.459 1.206 1.636 0.2502 0.0012 0.9990 4 2.459 1.206 1.636 0.2506 0.0012 0.9980 5 2.459 1.206 1.636 0.2506 0.0012 0.9976 6 2.459 1.20 6 1.636 0.2502 0.0012 0.9977 7 2.459 1.206 1.636 0.2502 0.001 2 0.9971 8 2.459 1.206 1.636 0.2486 0.0012 0.9960 9 2.459 1.20 6 1.636 0.2479 0.0012 0.9963 10 2.459 1.206 1.636 0.2 534 0.0012 0.9978 11 2.459 1.206 1.636 0.2586 0.0012 0.9985 12 2.459 1.206 1.636 0.2524 0.0012 0.9985 13 2.459 1.206 1.636 0.2523 0.0012 0.9985 14 2.459 1.206 1.636 0.2547 0.0012 0.9982 15 2.459 1.20 6 1.636 0.2546 0.001 2 0.9980 16 2.459 1.206 1.636 0.2315 0.0012 0.9981 17 2.459 1.20 6 1.636 0.2017 0.001 2 0.9974 LCT-9 1 4.310 1.415 2.540 0.11 27 0.00 21 0.9980 2 4.310 1.4 15 2.540 0.11 22 0.0021 0.9986 3 4.310 1.415 2.540 0.1125 0.00 21 0.99 79 4 4.310 1.415 2.540 0.11 21 0.0021 0.9981 5 4.310 1.415 2.540 0.1136 0.0021 0.9993 6 4.310 1.415 2.540 0.1 127 0.0021 0.9985 7 4.310 1.415 2.540 0.11 37 0.0021 0.9994 8 4.310 1.415 2.540 0.1130 0.0021 0.998 1 9 4.310 1.415 2.540 0.11 35 0.00 21 0.9986 16 4.310 1.415 2.540 0.11 35 0.0021 0.9987 17 4.310 1.415 2.540 0.112 7 0.0021 0.9991 18 4.310 1.415 2.540 0.1138 0.0021 0.9977 19 4.310 1.415 2.540 0.112 9 0.00 21 0.9986 20 4.3 10 1.415 2.540 0.11 37 0.0021 0.9982 21 4.310 1.415 2.540 0.11 29 0.0021 0.9988 22 4.310 1.415 2.540 0.113 8 0.0021 0.9984 23 4.3 10 1.415 2.540 0.1130 0.0021 0.9994 NET-28091-0003

-01 , R evision 0 A-1 2 Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. k err ID No. (wt% U-Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (Li k) 24 4.310 1.415 2.540 0.1122 0.0021 0.9979 25 4.310 1.415 2.540 0.1120 0.0021 0.9983 26 4.310 1.415 2.540 0.1121 0.0021 0.9987 27 4.310 1.415 2.540 0.1119 0.0021 0.998 5 LCT-10 5 4.310 1.415 2.540 0.3547 0.0021 1.0000 6 4.310 1.415 2.540 0.26 15 0.0021 1.0003 7 4.310 1.415 2.540 0.2092 0.0021 1.000 6 8 4.310 1.415 2.540 0.1844 0.0021 0.99 79 9 4.3 10 1.415 2.540 0.1221 0.0021 1.00 07 10 4.310 1.415 2.540 0.1183 0.0021 1.001 3 11 4.310 1.415 2.540 0.1154 0.0021 1.0006 12 4.310 1.415 2.540 0.1122 0.0021 1.000 0 13 4.310 1.415 2.540 0.1105 0.0021 0.9968 14 4.310 1.415 1.892 0.307 1 0.0028 1.0014 15 4.310 1.415 1.892 0.29 50 0.0028 1.0018 16 4.310 1.415 1.892 0.2853 0.0028 1.0021 1 7 4.310 1.415 1.892 0.2 787 0.0028 1.0021 18 4.310 1.415 1.892 0.2749 0.0028 1.0010 19 4.310 1.415 1.892 0.2677 0.0028 1.0008 24 4.310 1.415 1.892 0.5990 0.0028 0.9994 25 4.310 1.41 5 1.892 0.553 6 0.002 8 1.001 0 26 4.310 1.415 1.892 0.512 2 0.0028 1.001 0 27 4.310 1.41 5 1.892 0.4780 0.002 8 1.001 7 28 4.3 10 1.415 1.892 0.4485 0.0028 1.0017 29 4.310 1.415 1.892 0.4232 0.0028 1.0016 30 4.3 10 1.415 1.892 0.36 79 0.0028 0.9996 LCT-1 1 1 2.459 1.206 1.636 0.1 685 0.0018 0.9968 2 2.459 1.206 1.636 0.2450 0.0032 0.9967 3 2.459 1.206 1.636 0.19 20 0.003 2 0.99 7 1 4 2.459 1.206 1.636 0.1 927 0.0032 0.99 72 5 2.459 1.206 1.636 0.1 935 0.0032 0.9970 6 2.459 1.206 1.636 0.1951 0.003 2 0.99 70 7 2.459 1.206 1.636 0.1959 0.0032 0.9967 8 2.459 1.206 1.636 0.1972 0.0032 0.9974 9 2.459 1.206 1.636 0.19 84 0.0032 0.9975 10 2.459 1.2 06 1.636 0.1866 0.0017 0.9945 11 2.459 1.206 1.6 36 0.1628 0.0017 0.9940 12 2.459 1.206 1.636 0.16 70 0.0017 0.9950 13 2.459 1.206 1.636 0.1 475 0.001 7 0.9943 14 2.459 1.206 1.636 0.150 8 0.001 7 0.9946 15 2.459 1.206 1.636 0.1 387 0.0018 0.9959 L C T-13 1 4.3 10 1.415 1.892 0.2862 0.0018 1.0005 2 4.310 1.415 1.892 0.2939 0.0018 1.0004 3 4.310 1.415 1.892 0.297 4 0.0018 1.0003 4 4.310 1.415 1.892 0.2969 0.0018 1.0007 5 4.310 1.415 1.892 0.2961 0.0032 1.0003 LCT-16 l 2.350 1.270 2.032 0.0957 0.0031 0.99 73 2 2.350 1.27 0 2.032 0.0954 0.0031 0.9962 3 2.350 1.270 2.032 0.0954 0.0031 0.9967 NET-28091-0003-01 , Revi sion 0 A-13 B e nchmark C ase E nri c hm e nt F u e l Pin F u e l P i n EALF Meas. k err ID No. (wt% U-Diam ete r Pit c h (c m) (eV) U n ce rt a in ty 23 5) (cm) ,~ k) 4 2.350 1.2 70 2.032 0.095 6 0.0031 0.9960 5 2.350 1.2 70 2.032 0.0952 0.0031 0.9970 6 2.350 1.270 2.032 0.0961 0.0031 0.9971 7 2.350 1.270 2.032 0.0959 0.0031 0.9973 8 2.350 1.270 2.032 0.0969 0.0031 0.9972 9 2.350 1.270 2.032 0.0961 0.0031 0.9977 10 2.350 1.270 2.032 0.09 70 0.0031 0.9971 11 2.350 l.270 2.032 0.0962 0.0031 0.9978 12 2.350 1.270 2.032 0.0974 0.003 1 0.9972 13 2.350 1.270 2.032 0.0965 0.0031 0.9979 14 2.350 1.270 2.032 0.0975 0.0031 0.9974 21 2.350 1.270 2.032 0.0971 0.0031 0.9977 22 2.350 1.270 2.032 0.0968 0.003 1 0.9974 23 2.350 1.270 2.032 0.0963 0.0031 0.9977 24 2.350 1.270 2.032 0.0967 0.0031 0.9970 25 2.350 1.270 2.032 0.0963 0.0031 0.9972 26 2.350 1.270 2.032 0.0969 0.0031 0.9976 27 2.350 1.270 2.032 0.0963 0.0031 0.9979 28 2.350 1.270 2.032 0.0951 0.0031 0.9972 29 2.350 1.270 2.032 0.0950 0.0031 0.9969 30 2.350 1.270 2.032 0.0949 0.0031 0.9965 31 2.350 1.270 2.032 0.0950 0.0031 0.9979 32 2.350 1.2 70 2.032 0.0949 0.0031 0.9972 LCT-1 7 4 2.350 1.270 2.032 0.201 7 0.0031 0.9983 5 2.350 1.270 2.032 0.1779 0.0031 0.9994 6 2.350 1.2 70 2.032 0.1 685 0.0031 0.9989 7 2.350 1.270 2.032 0.1597 0.0031 0.9994 8 2.350 1.270 2.032 0.1333 0.0031 0.9972 9 2.350 1.270 2.032 0.1092 0.0031 0.997 3 10 2.350 1.270 2.032 0.0998 0.0031 0.9973 11 2.350 1.270 2.032 0.09 79 0.0031 0.9979 12 2.350 1.270 2.032 0.0968 0.0031 0.9977 13 2.350 l.270 2.032 0.0953 0.0031 0.99 76 14 2.350 1.270 2.032 0.0946 0.0031 0.9985 15 2.350 1.270 1.684 0.1777 0.0028 0.996 1 16 2.350 1.270 1.684 0.1711 0.0028 0.9983 17 . 2.350 1.270 1.684 0.1665 0.0028 0.9987 18 2.350 1.27 0 1.68 4 0.1 648 0.0028 0.9974 19 2.350 1.270 1.684 0.1622 0.0028 0.9978 20 2.350 1.27 0 1.684 0.1607 0.0028 0.9971 21 2.350 1.270 1.684 0.1592 0.0028 0.9966 22 2.350 1.270 1.684 0.1584 0.0028 0.9959 26 2.350 1.270 1.68 4 0.3741 0.0028 0.9958 27 2.350 1.270 1.68 4 0.3203 0.0028 0.9972 28 2.350 l.270 1.684 0.2806 0.0028 0.9974 29 2.350 1.270 1.684 0.2505 0.0028 0.9984 LCT-34 4 4.738 0.940 1.600 0.1367 0.0039 1.0003 5 4.738 0.940 1.600 0.1 330 0.0039 0.9999 6 4.738 0.940 1.600 0.1298 0.0039 1.0017 NET-2809 1-0003-01 , R evision 0 A-14 Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. keff ID No. (wt% U-Diameter Pitch (cm) (eV) Uncertainty 235) (cm) ,~ k) 7 4.738 0.940 1.600 0.1279 0.0039 1.0002 8 4.738 0.940 1.600 0.1258 0.0039 0.9992 15 4.738 0.940 1.600 0.1348 0.0043 0.9947 LCT-35 1 2.596 1.417 1.956 0.2086 0.0018 0.9983 2 2.596 1.417 1.956 0.2126 0.0019 0.9976 LCT-39 1 4.738 0.940 1.260 0.22 18 0.0014 0.9953 2 4.738 0.940 1.260 0.2119 0.0014 0.9969 3 4.738 0.940 1.260 0.1923 0.0014 0.9965 4 4.738 0.940 1.260 0.1836 0.0014 0.9961 5 4.738 0.940 1.260 0.1393 0.0009 0.9978 6 4.738 0.940 1.260 0.1452 0.0009 0.9977 7 4.738 0.940 1.260 0.2132 0.0012 0.9962 8 4.738 0.940 1.260 0.2031 0.0012 0.9963 9 4.738 0.940 1.260 0.19 76 0.0012 0.9969 10 4.738 0.940 1.260 0.1732 0.0012 0.9970 11 4.738 0.940 1.260 0.22 18 0.0013 0.99 53 12 4.738 0.940 1.260 0.2 166 0.0013 0.9951 13 4.738 0.940 1.260 0.2146 0.0013 0.9951 14 4.738 0.940 1.260 0.2 124 0.0013 0.9954 15 4.738 0.940 1.260 0.2 112 0.0013 0.9959 16 4.738 0.940 1.260 0.2 104 0.0013 0.9967 17 4.738 0.940 1.260 0.2099 0.0013 0.9960 LCT-40 1 4.738 0.940 1.600 0.1427 0.0039 0.9966 5 4.738 0.9 40 1.6 00 0.1377 0.0042 0.9951 9 4.738 0.940 1.600 0.1470 0.0046 0.9993 10 4.738 0.940 1.600 0.1419 0.0046 0.9931 LCT-42 1 2.350 1.270 1.684 0.1690 0.0016 0.9971 2 2.350 1.270 1.684 0.1753 0.0016 0.9968 3 2.350 1.270 1.684 0.1819 0.0016 0.9981 4 2.350 1.270 1.684 0.1804 0.0017 0.9980 5 2.350 1.270 1.684 0.1775 0.0033 0.9981 LCT-43 2 4.349 0.980 1.500 0.1553 0.0010 1.0007 LCT-44 1 4.349 0.980 1.500 0.1474 0.0010 0.9993 LCT-46 1 4.349 0.981 1.500 0.1488 0.00044 0.9991 2 4.349 0.981 1.500 0.1525 0.00044 0.9989 3 4.349 0.981 1.500 0.1542 0.00044 0.9988 4 4.349 0.98 1 1.500 0.1556 0.00044 0.9989 5 4.349 0.981 1.500 0.1573 0.00044 0.9986 6 4.349 0.981 1.500 0.1595 0.00044 0.9987 7 4.349 0.981 1.5 00 0.1479 0.00044 0.9991 8 4.349 0.981 1.500 0.1550 0.00044 0.9988 9 4.349 0.981 1.500 0.1 594 0.00044 0.9987 10 4.349 0.98 1 1.500 0.1621 0.00044 0.9988 11 4.349 0.98 1 1.500 0.1672 0.00044 0.9988 12 4.3 49 0.98 1 1.500 0.15 39 0.00044 0.9986 13 4.349 0.981 1.500 0.1570 0.00044 0.9986 14 4.3 49 0.981 1.500 0.1596 0.00044 0.9986 15 4.349 0.981 1.500 0.1618 0.00044 0.9984 16 4.349 0.981 1.500 0.1655 0.00044 0.9983 NET-28091-0003-01 , R evision 0 A-15 B e nchmark Case E nrichm e nt F u e l Pin F u e l Pin EALF Meas. k.rr ID N o. (wt% U-Diam e t e r Pi t ch (c m) (eV) U n ce rt a in ty 2 35) (c m) (~ k) 17 4.349 0.981 1.500 0.1724 0.00044 0.9983 LCT-48 1 3.005 1.094 1.320 0.6771 0.0025 0.9990 2 3.005 1.094 1.320 0.6508 0.0025 0.9983 3 3.005 1.094 1.320 0.6824 0.0025 0.9984 4 3.005 1.094 1.320 0.6838 0.0025 0.9988 5 3.005 1.094 1.320 0.6736 0.0025 0.9983 LCT-50 1 4.738 0.940 1.300 0.1998 0.0010 0.9983 2 4.738 0.940 1.300 0.1907 0.0010 0.99 78 3 4.738 0.940 1.300 0.2075 0.0010 0.997 8 4 4.738 0.940 1.300 0.1977 0.0010 0.9972 5 4.738 0.940 1.300 0.2230 0.0010 0.9983 6 4.738 0.940 1.300 0.2 1 41 0.0010 0.9991 7 4.738 0.940 1.300 0.2095 0.0010 0.9992 LCT-51 1 ClO 2.459 1.206 1.636 0.1472 0.0020 0.9965 2 cl la 2.459 1.206 1.636 0.1968 0.0024 0.9972 3 cl 1 b 2.459 1.206 1.636 0.1964 0.0024 0.9972 4 cl le 2.459 1.2 06 1.636 0.1979 0.0024 0.9975 5 cl ld 2.459 1.206 1.636 0.1989 0.0024 0.9970 6 cl le 2.459 1.206 1.636 0.1998 0.0024 0.9972 7 cl If 2.459 1.206 1.636 0.2000 0.0024 0.9973 8 cl lg 2.459 1.206 1.636 0.20 11 0.0024 0.9971 9 cl2 2.459 1.206 1.636 0.1669 0.0019 0.9969 LCT-54 I 4.349 0.980 1.500 0.1508 0.0005 0.9996 LCT-71 1 4.738 0.949 1.100 0.7592 0.00076 0.9955 2 4.738 0.949 1.100 0.6972 0.00076 0.9954 3 4.738 0.949 1.100 0.6610 0.000 76 0.9948 4 4.738 0.949 1.075 0.8485 0.0008 0.9 951 LCT-72 l 4.738 0.949 1.600 0.111 7 0.0012 0.9990 2 4.738 0.949 1.600 0.1077 0.0012 0.9985 3 4.738 0.949 1.600 0.1099 0.0012 0.9988 LCT-77 3 4.349 0.980 1.500 0.1621 0.0010 1.000 6 LCT-82 3 4.349 0.980 1.500 0.1497 0.0010 1.0005 LCT-83 1 4.349 0.980 1.500 0.1516 0.0010 1.0001 LCT-84 I 4.349 0.980 1.500 0.1541 0.0010 1.0008 LCT-88 1 4.349 0.98 1 1.500 0.1543 0.00044 0.9993 2 4.349 0.981 1.500 0.1556 0.00044 0.9992 3 4.349 0.981 1.500 0.1561 0.00044 0.9992 4 4.349 0.98 1 1.500 0.1560 0.00044 0.999 7 5 4.349 0.981 1.500 0.1560 0.00044 0.9996 6 4.349 0.981 1.500 0.1560 0.00044 0.9998 7 4.349 0.98 1 1.500 0.1559 0.00044 0.9999 8 4.349 0.981 1.500 0.1560 0.00044 0.9994 9 4.349 0.981 1.500 0.1560 0.00044 0.9994 10 4.349 0.981 1.500 0.1 561 0.00044 0.9989 11 4.349 0.981 1.500 0.1561 0.00044 0.9986 12 4.349 0.981 1.500 0.1563 0.00044 0.9980 13 4.349 0.981 1.500 0.1565 0.00044 0.9975 14 4.349 0.981 1.500 0.1564 0.00044 0.9971 15 4.3 49 0.981 1.500 0.1 566 0.00044 0.9967 NET-28091-0003-01 , R evision 0 A-16 Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. k err ID No. (wt% U-Diam ete r Pitch (cm) (eV) Uncertainty 235) (cm) (d k) 16 4.349 0.981 1.500 0.1 566 0.00044 0.9963 17 4.349 0.981 1.500 0.156 7 0.00044 0.9957 18 4.349 0.981 1.500 0.15 68 0.00044 0.9954 19 4.349 0.981 1.500 0.15 60 0.00044 0.9994 20 4.349 0.981 1.500 0.15 67 0.00044 0.9990 21 4.349 0.981 1.500 0.15 71 0.00044 0.9992 22 4.349 0.981 1.500 0.15 73 0.00044 0.9991 23 4.349 0.981 1.500 0.15 75 0.00044 0.9991 24 4.349 0.981 1.500 0.1576 0.0 0044 0.9991 25 4.349 0.981 1.500 0.1578 0.00044 0.9991 26 4.349 0.981 1.500 0.1578 0.00044 0.9992 27 4.349 0.981 1.500 0.15 79 0.00044 0.9992 28 4.349 0.981 1.500 0.15 80 0.00044 0.9991 29 4.349 0.981 1.500 0.1 582 0.00044 0.9992 30 4.349 0.981 1.500 0.1582 0.00044 0.9995 31 4.349 0.981 1.500 0.1583 0.00044 0.9995 32 4.349 0.981 1.500 0.1 584 0.00044 0.9995 33 4.349 0.9 81 1.500 0.1585 0.00044 0.9994 34 4.349 0.981 1.500 0.1584 0.00044 0.9996 35 4.349 0.981 1.500 0.1 584 0.00044 0.9996 LCT-89 1 4.349 0.980 1.500 0.1 530 0.0010 1.0000 LCT-90 1 4.349 0.980 1.500 0.1 459 0.0010 0.9994 LCT-91 4 4.349 0.980 1.500 0.1 508 0.00 10 0.9999 LCT-92 1 4.349 0.981 1.500 0.1543 0.00044 0.9996 2 4.349 0.981 1.500 0.1545 0.00044 0.9994 3 4.349 0.981 1.500 0.15 45 0.00044 0.9996 4 4.349 0.981 1.500 0.154 9 0.00044 0.9994 5 4.349 0.981 1.500 0.1555 0.00046 0.9988 6 4.349 0.981 1.500 0.1 559 0.00 055 0.9994 LCT-96 1 6.903 0.635 0.800 0.5690 0.00095 0.9973 2 6.903 0.635 0.800 0.5674 0.00095 0.99 72 3 6.903 0.635 0.800 0.4191 0.00095 0.9993 4 6.903 0.635 0.800 0.5704 0.00095 0.99 71 5 6.903 0.635 0.800 0.561 7 0.00095 0.997 1 6 6.903 0.635 0.800 0.549 2 0.00095 0.9968 7 6.903 0.635 0.800 0.5304 0.00095 0.9965 8 6.903 0.635 0.800 0.5068 0.00095 0.9966 9 6.903 0.635 0.800 0.492 9 0.00095 0.9965 10 6.903 0.635 0.800 0.4929 0.00 095 0.9963 11 6.903 0.635 0.800 0.4630 0.00095 0.9974 12 6.903 0.6 35 0.800 0.43 17 0.00095 0.9975 13 6.903 0.635 0.800 0.4032 0.00095 0.9978 14 6.903 0.635 0.800 0.3800 0.00095 0.99 77 15 6.903 0.635 0.800 0.3604 0.00095 0.9979 16 6.903 0.635 0.800 0.4320 0.00095 0.9978 17 6.903 0.635 0.800 0.3756 0.00095 0.9984 1 8 6.903 0.635 0.800 0.33 1 7 0.00095 0.9986 19 6.903 0.635 0.800 0.2997 0.00095 0.9989 BAW-1810 1 2.460 1.206 1.636 0.2477 0.00250 0.9990 NET-2809 1-0003-01, R ev i sion 0 A-17 Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Mea s. k err ID No. (wt% U-Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (d k) 2 2.460 1.206 1.636 0.2470 0.00250 0.99 82 3 2.460 1.20 6 1.636 0.2465 0.00250 0.9 983 4 2.460 1.206 1.636 0.2452 0.00250 0.9991 5 2.460 1.206 1.636 0.24 61 0.00250 0.9982 5A 2.460 1.206 1.636 0.2457 0.00250 0.998 1 6 2.460 1.20 6 1.636 0.2 453 0.00250 0.9981 6A 2.460 1.206 1.636 0.2450 0.002 50 0.9982 8 2.460 1.206 1.636 0.2453 0.00250 0.9982 9 2.460 1.206 1.636 0.244 8 0.00250 0.998 1 WCAP-3269 2.7 2.720 1.18 9 1.524 0.2599 0.00400 0.9988 5.7 5.700 0.993 1.422 0.3006 0.00400 0.9978 3.7-12 3.700 0.860 1.1 05 0.4309 0.00400 0.9985 3.7-24 3.700 0.860 1.105 0.42 88 0.00400 0.9979 3.7-48 3.700 0.860 1.1 05 0.455 8 0.00400 0.9969 Since b oro n credit is u sed it is important validate boron with critica l experiments.

Table A.3 s how s the boron information on b oron-containing benchmarks , alon g with the calculated

k. Table A.3: Summary of Critical Experiments Containing Boron Benchmark Case No. Soluble Separator Plate No. of kerr ID Boron 10 B Areal Densit y Boron (ppm) (gm/cm 2) Rods LCT-8 1 1511 0.9976 2 1 334 0.9984 3 1 337 0.9990 4 11 83 36 0.9980 5 11 81 36 0.9976 6 10 34 72 0.9977 7 10 31 72 0.997 1 8 794 144 0.9960 9 779 144 0.9963 10 1245 72 0.9978 11 1 384 0.998 5 12 1 348 0.998 5 13 134 8 0.9985 14 1 363 0.9982 15 1 362 0.9980 16 115 8 0.998 1 17 921 0.9974 LCT-9 5 0.004549 0.9993 6 0.00 4549 0.9985 LCT-9 7 0.006904 0.9994 8 0.006904 0.998 1 NET-2809 1-000 3-01, R ev i sion 0 A-1 8 Benchmark Ca se N o. So lubl e Separator Plate N o. of k err ID Boron 10 B Area l Densit y Boron (ppm) (gm/cm 2) Rod s 9 0.066946 0.9986 LCT-11 2 1037 0.9967 3 769 0.997 1 4 764 0.9972 5 762 0.9970 6 753 0.9970 7 739 0.9967 8 721 0.9974 9 702 0.9975 10 84 0.9945 11 64 0.9940 12 64 0.9950 13 34 0.9943 14 34 0.9946 LCT-13 2 0.004549 1.0004 3 0.030173 1.0003 4 0.056950 1.0007 LCT-16 8 0.004549 0.9972 9 0.004549 0.9977 10 0.006904 0.997 1 11 0.006904 0.9978 12 0.066946 0.9972 13 0.066946 0.9979 14 0.066946 0.9974 LCT-34 4 0.002521 1.0003 5 0.002521 0.9999 6 0.00252 1 1.0017 7 0.00252 1 1.0002 8 0.00252 1 0.9992 15 0.04601 1 0.9947 LCT-35 1 70 0.9983 2 147.7 0.99 7 6 LCT-40 1 0.002521 0.9966 5 0.046011 0.9951 9 0.04601 1 0.9993 10 0.046011 0.9931 LCT-42 2 0.004549 0.9968 3 0.030 1 73 0.998 1 4 0.056950 0.9980 LCT-50 3 822 0.9978 4 822 0.9972 5 5030 0.9983 6 5030 0.999 1 7 5030 0.9992 LCT-51 1 ClO 14 3 0.9965 NET-2809 1-00 03-01 , Revision 0 A-19 Benchmark Case No. So lu ble Separator P l ate No. of kerr ID Boron 10 B Area l Densit y Boron (p p m) (gm/cm 2) Ro d s 2 cl la 510 0.9 972 3 cl lb 514 0.9 972 4 cl le 501 0.9975 5 cl ld 493 0.9 970 6 cl le 474 0.9972 7 cl lf 462 0.9973 8 cl lg 432 0.9971 9 cl2 217 0.9969 LCT-77 3 4 1.0006 LCT-82 3 6 1.0005 LCT-92 1 0.1 0.9996 2 6 0.9994 3 11 0.9996 4 22 0.9994 5 43 0.9988 6 95 0.9994 BAW-1 8 10 1 1337 0.9990 2 1250 0.9982 3 12 39 0.9983 4 1171 0.9991 5 120 8 0.9982 SA 1191 0.9981 6 1156 0.9981 6A 1136 0.9982 8 11 71 0.9982 9 1131 0.9981 A.2.5 Statist ic al Ana l ys i s of t he Fresh U0 2 Cri t i cal Ben c h m a r k Res ults The stat i s tical treatment u se d follows th e guidance pro v ided in NUREG/C R-669 8 [2]. The NUREG approach weights the calculated kerr values by the experimental uncertainty.

This approach means the hi gher quality experiments (i.e.: lo we r uncert ai nty-see Table A.2) affect the results more than the lo w quality (i.e.: higher uncertainty) experiments. The uncertainty weighting is used for the analysis of the set of experiments as a whole , as we ll as for th e ana lysi s for trends. Before see king trends the 328 critical benchm ar ks set are reviewed as a whole. The unw eig hted mean k e rr of the 328 samples is 0.9981 with a standard de v iation of0.0015. The weighted mean is 0.9985 and the weighted standard deviation i s 0.0015. The average uncert ai nt y of the experiments (interpreted as one sigma) is 0.0019. Since the tot a l one sigma standard de viatio n i s only 0.0015, this suggests that the experimental uncertainty dominates the uncertainty and there is little to be gained with improved methods. Unles s sta ted otherwi se all of the r esu lt s pre se nted will come from the weighted analysis.

The bias of the set as a who l e is 0.0015. The uncertainty is the standard deviation multiplied by the sided lower tolerance factor (taken as 2.065 from Referenc e 2 for more than 50 samples) so it is 0.0031. NET-28091-0003-01, Revi s ion 0 A-20 As recommended by NUREG/CR-6698, the resu lt s of the validatio n are checked for normality.

T h e Nationa l In stitute of Sta nd ards and Tec hnol ogy (N I ST) h as made publicly avai l ab l e a st a tistical pack a ge , DATAPLOT [4). The 328 critica l experiments were tested with the Wilk-Shapiro normality test and were found to a dh ere to a normal distribution at the 90% l evel. The test results are s h own in Tab l e A.4. A histogram plot of the data is shown on Figure A. l. Table A.4: Wilk-Shapiro Test Results Output From DATAPLOT [4] NET-2809 1-00 03-0 1 , Revision 0 A-21 60 so C iii 40 .l: "' ::.: -a "' 8 30 b .. Ill .a E 20 10 0 Calculated keff Distribution Versus a Normal Distribution Figure A.l: Distribution of the Calculated kcrr values Around the Mean If a feature of a subset of the critical experiments creates a statistically subset , this feature needs to be corrected before combining all of the critical results. There are 89 experiments that have boron in them. The average k eff of the boron containing cases is 0.9978 which is very close to the average of all cases (0.9981 ). Similarly, there are 17 cases that used pure Cadmium absorbers.

The mean of these cases is 0.9982. There are 51 cases that use Ag-In-Cd control rods with a mean of 0.9987. Since the standard deviation of the set as a whole is 0.0015 (unweighted) it is clear these features are not skewing the results. The next step in the analysis is to look for trends in the data. The math will always find a trend but only the real or statistically significant trends are of interest.

Section 3.2.2 of the DOE/RW technical report in support of validation for burnup credit [8] describes an appropriate trend test. In this test, the null hypothesis is that the slope of the trend is zero (no trend) and it tests to determine if there is confidence that the calculated slope is a more accurate representation than a zero slope. The equations for this test are presented here. NET-28091-0003-01, Revision 0 A-22 Let the regression fit be of the form: k= a+ b X Let x-bar be the average value ofx for then cases and define: and define: then the test statistic is: sxx = ~)x; -x)2 i=l,n (n -2)

  • Sxx T=\bl* This test statistic is then compared to the Student's t-distribution at the desired confidence level and n-2 degrees of freedom. In the past it was assumed that unless there is a high confidence level (95%) that the slope was non-zero, the analysis would assume a zero slope (no trend) on the given parameter.

Since the analysis will include consideration of the data as non-trended, it is more conservative to assume there is also a trend. Inverting the statistical test to requiring a high confidence that the slope is zero will result in all cases having a trend. At this time, although a test on the confidence of the trend is performed, the analysis assumes all calculated trends are real. For this work the weighted k.,ffvalues are used to determine the fit to a straight line. Refer to NUREG/CR-6698

[2] equations 10 through 13. NUREG/CR-6698

[2] describes the appropriate tolerance band for criticality validation.

This work simply applies the equations (equations 23 to 30) given in the NUREG. Note that the tolerance band is found using the weighted experimental data. The width of the tolerance band is the uncertainty.

In the final analysis, the calculated keff of the system must be less than the minimum of k(x) minus the uncertainty minus the administrative safety margin. The uncertainty in k.,ff from other independent uncertainties, such as the manufacturing tolerances, burnup, and depletion uncertainties can be statistically combined with the uncertainty in the criticality validation.

The rest of this section will evaluate the trends in keff as a function of trending parameters using the methods described above. Historically, an Upper Subcriticality Limit (USL) was assigned from the criticality validation analysis.

This is not done here, since the other uncertainties (e.g., manufacturing tolerances of the rack, depletion uncertainty, etc.) are not known at this time. NET-28091-0003-01, Revision 0 A-23 Neutron spectrum Trends in the calculated k e rr of the benchmarks were so u ght as a function of the neutron spect rum. Since a l arge number of things can affect the spectrum , a s in g l e index calc ul ated by SCALE is used. This index is the E n ergy (eV) of the Average Let har gy caus in g F i ssion (EALF). Figure A.2 s h ows the distribution of k err va lu es around the me an k , which is s h own as the red lin e. Visual inspection of the graph a nd the statistica l analysis of the results of the statistical analysis suggest that there is a statistically s ignific ant trend on neutron spectrum.

Using NUREG/CR-6698 [2] equations 10 through 13 and the data from Table A.2 , the predicted mean ke rr as a function of EALF i s: k(EALF) = 0.999406 -0.00459

  • EALF T h e units for EALF are eV. The bias at 0.4 eV is 0.0024. The unc e rtain ty at 0.4 eV i s 0.0030. T h e bias at 0.65 eV i s 0.0036. T h e un certainty at 0.65 is 0.0034. 1.003 -------------1.002 *** 1.001 *
  • 1.000 :i: 0.999 QI :li:: -o 0.998 * * ---.----,-----~---------QI ... 1 0.997 .., u 0.996 0.995 0.994 0.993 0.992 ,----T 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 Energy of the Average Lethargy Causing Fission (eV) Figure A.2: k eff as a Function of the Energy of the Average Lethargy Causing Fission NET-28091-0003-01 , Revision 0 A-24 Geometry tests Two trend tests were performed to determine if lattice/geometric parameters are adequately treated by SCALE 6.1.2. The first parameter is the fuel pin diameter.

A small, statistically insignificant trend was found when the critical experiment analysis results were correlated to the fuel pin diameter.

The second lattice parameter tested is the lattice pitch. A statistically significant trend on lattice pitch was found. The trend on pitch or pin diameter could be caused by the spectral trend found in the previous subsection.

Using NUREG/CR-6698 [2] equations 10 through 13 and the data from Table A.2 , the predicted mean ke rr as a function of pin diameter is: k(Pin Diameter)

= 0.997805 + ( 7.43E-04)*Pin Diameter where the pin diameter is in cm. The predicted mean k e rr as a function of pitch is: k(Pitch) = 0.997098 + ( 9.646E-04)*Pitch where lattice pitch is in cm. The tolerance band widths are 3.4E-03 and 3.SE-03 for the pin diameter and pitch respectively. Figures A.3 and A.4 graphically present k e rr as a function of the pin diameter and the lattice pitch. 1.003 l 1.002

  • 1.001 1.000 ---I :i: 0.999 o/ ; 0.998 41 .. 3 0.997 * * ... 0.996
  • 0.995 i 0.994 0.99 3
  • 0.992 0.8 0.9 1 1.1 1.2 1.3 1.4 1.5 Pin Diameter (cm) Figure A.3: kerr as a Function of the Pin Diameter NET-28091-0003-01 , Revision 0 A-25 1.003 1.002 1.001 1.000 := 0.999 QI =-:: 0.998 "C QI .... Ill 0.997 :i V iv 0.996 u 0.995 0.994 0.993 0.992 1 Enrichment 1.2 1.4 -+-----1.6 1.8 Pitch (cm) 2 2.2 Figure A.4: keff as a Function of the Lattice Pitch 2.4 2.6 The fuel to be stored in the racks ranges in enrichment from 1.6 wt% 235 U to 5 wt% m u. It was determined that there is no statistically significant trend on enrichment.

Although not statistically significant, the trend in the mean k eff is: k(Enrichment) 0.998824 -( 6.4E-05)*Enrichment where Enrichment is wt% m u. The tolerance band width is 3.4E-03. Figure A.5 graphically presents the results. NET-28091-0003-01 , Revision 0 A-26 1.003 ----------------1.002 1.001 1 0.999

  • 41 :.::: "O 0.998 41 .. I'll 0.997 :i V 7ii 0.996 u 0.995 0.994 0.993 ------------* 0.992 2 2.5 3 3.5 4 4.5 5 5.5 6 6.5 7 Enrichment (wt% U-235) Figure A.5: keff as a Function of the Fuel Enrichment Boron Content A trend test was performed to determine whether the calculated k e rr values of the benchmark experiments co ntain a statistically significant trend as a function of the soluble boron ppm. No statistically significant t rend was found. However, it is conservatively assumed that the trend s are real. The follo w ing equation is the best fit of the data for k err v ersus soluble boron ppm. Figure A.6 shows the results of the analyses. The uncertainty around the mean v alues given in the followin g equation s i s 0.0029 at O ppm and 0.0035 at 2000 ppm. k(ppm soluble boron) 0.99856 + ( 1.56E-07)*ppm NET-28091-0003-01 , Revision 0 A-27 0.9995 -------0.9990 ** ----------*
  • 0.9 98 5 :i: 0.9980 "O ... "' * :i 0.9975 ... ia u 0.9970 * ... 0.9965
  • 0.9960 0 1 000 2000 3000 4 000 5000 6000 Soluble Boron (ppm) Figure A.6: kcff as a Function of the Soluble Boron Content A.2.6 Establishing the Bias and the Uncertainty To make the incorporation of the bias and bias uncertainty in the criticality analysis conservative , the most limiting bias and bias un certainty from the trends in the range of interest is u sed. At the lattic e pitch for Westinghouse 15xl5 fuel (1.43 cm) the bias and unc ertainty are 0.0015 and 0.0029 respectively.

At the Westinghouse 15xl5 fuel pin diameter (1.072 cm) the bias and unc ertainty are 0.0014 and 0.0029 respectively. Thus the bia s as a function of pitch is more limiting.

The bias as a function of enrichment is greatest at 5 wt% and is 0.0015. The uncertainty over the range of enrichments is 0.0034. The maximum bias and uncertainty as a function of soluble boron ppm occurs at 2000 ppm and is 0.0018 and 0.0035 respectively. The spectrum as measured by the EALF in the pool with no solub l e boron is genera ll y between 0.2 and 0.4 eV. The bias increases as the spectrum hardens and the bias at 0.4 eV is 0.0024. This is the most limitin g bias. For heavily borat ed cases the EALF can get almost as high as 0.65 eV. At 0.65 eV the bias is 0.0036. For the criticality analysis a bias of 0.0024 is used for all EALF less than 0.4 (limiting cases for no boron credit) and 0.0036 for EALF values between 0.4 and 0.65 eV (heavily borated cases). The maximum uncertaint y for any trends is 0.0035 which comes from the so lubl e boron analysis.

In order to mak e the analysis simple 0.0035 is selected for the uncertaint y in the bias. NET-2809 1-000 3-01 , Revision 0 A-28 The uncertainty of the set as a who l e is 0.0031. The uncertainty for the trended analysis is generally less since taking advantage of the trend reduces the difference between the experimental value and the predicted value. A.2. 7 Subcritical Margin In the USA , the NRC has established subcritical margins for rack analysis. The subcritical margin for borated spent fuel poo ls , casks , and fully flooded dry storage racks is O when the analysis is perfonned with unborated water. This is actually saying the subcritical margin is contained in the uncredited soluble boron. To make sure there is sufficient so lubl e boron, analysis is a l so performed with soluble boron and a subcritica l margin of 5% in k e rr is required.

For dry storage racks analyzed with optimum moderation, the subcritical margin is 2% and 5% with full moderation. In the ana l ysis of 32 8 critical experiments, which generously cover the range of expected conditions , the lowest calculated k e rr was 0.9931. This supports the position that the subcritical margin is more than sufficient.

A.2.BArea of Applicability (Benchmark Applicability)

The critica l benchmarks selected cover all commercia l light water reactor fuel storage racks or casks. To summarize the range of the benchmark applicability ( or area of applicability), Table A.5 is provided below. Table A.5: Area of Applicability (Benchmark Applicability)

Parameter Range Comments Fissionable Materia l/Physical U02 Form Enrichment (wt% U-235) 2.35 to 6.903 Some extrapolation of the bias to lower enrichments may be needed. Enrichments less than 2.35 are rarely limiting and generally only used in 1 s t cores. The bias is becoming smaller at low enrichments.

Using the maximum bias and uncertainty for all of the trends easily covers the small extrapolation needed. The maximum enrichment of 5% is within the range of experiments.

NET-28091-0003-01 , Revision 0 A-29 Parameter Range Comments Spectrum Expected range in applications:

-EALF (eV) 0.0605 to 0.1 to 0.6 0.8485 The experiments easily cover the entire expected range of limiting conditions. Lattice Characteristics Type Square Hex lattices have been excluded Pin Pitch ( cm) 1.075 to 2.54 Pin pitch of 1.43 cm is within the range. Assembly Spacing in Racks This covers all spacing. Neutron Distance between Assemblies 0 to 15.4 transport through larger than 15 .4 cm has a small effect on k. Note that the (cm) spacing is assumed to be filled with full density water. If the water density is less this separation effectively increases.

Therefore , optimum moderation cases of wide spaced racks are covered. Absorbers Ag-In-Cd control rods Contained in No significant difference in bias 51 critical between Ag-In-Cd critical experiments experiments and those that did not contain the control rods. Absorbers Soluble Boron 0 to 5030 ppm All designs are within this range. Concentration Absorbers Cd bearing experiments showed no Cd (component of Ag-In-Cd Absorber dependence on the number of rods. Credit for these rods is acceptable.

Cd rods) panels NET-28091-0003-01 , Revision 0 A-30 Parameter Range C o mments Reflector Experiments included water Reflectors Most rack analysis will assume an and steel adequate l y infinite system. Fu ll pool mode l covered reflectors are adeq u ate l y covered. Temperature Room This temperature range covers a ll Temperature norma l operating temperatures. Over to 358 K temperature accident conditions have significant margin due to ppm boron. Moderating mater i al water The moderator in a ll benchmark experiments are water , t herefore water as a moderating material is covered NET-28091-0003-01 , Revision 0 A-3 1 A.2.9 Summary of U02 Laboratory Critical Experiment Analysis This va lidati on follows the guidance ofNUREG/CR-6 698. Key aspects of the guidance are the selection of expe rim ents, ana l ysis of the experiments, statistica l treatment, determination of the bias and the bias uncertainty , and finally id entification of the area of applicability.

328 U0 2 critical experiments have been selected that cover the range of conditions for rack ana l ysis. The experime nt s h ave been a nal yzed using SCALE 6.1.2 a nd the EN DF/B-VII 238 gro up cross sect ion s and the resu ltin g bias in keff is very small. The results of the criticality ana l ysis were tested for trends against 5 different p arameters import ant to reactivity.

It was conservatively assumed that the any trend found was significant.

Using the trends, the most l imiting bi as and bias un certainty is determined to b e 0.0024 for the bias for EALF up to 0.4 eV and 0.0036 for EALF's in the range of 0.4 and 0.65 eV and the uncertainty is 0.0035 for all analysis.

The area of applicabi lit y is found in Ta bl e A.5. NET-2809 1-000 3-01, R evision 0 A-32 A.3. HTC and MOX Critical Experiments Burned fuel contains a low concentration of plutonium (about 1 wt%), as well as the uranium and thu s i s actually Mixed Oxide (MOX) fuel. Most classical MOX experiments ha ve plutonium concentrations at least twice as hi gh as that contained in burned fuel. A series of experiments were performed in France and purchased by the US for domestic use , which model the uranium and plutonium concentration, which matches 4.5 wt% U-235 fuel burned to 37.5 GWd/T [12]. This fuel has 1.1 wt% plutonium and 1.57 wt% U-235. Both the HTC critical experiments and a l arge series of c l assical MOX experiments were analyzed.

A.3. 1 HTC Critical Experiments All of the HTC critical experiments used the same fuel pins. The criticality of the se experiments was controlled by adjusting the critical water height. The fuel pin s were used in 156 critical arrangements.

117 of these were relevant to spent fuel pool analysis.

The experiments were performed in four pha ses. Phase 1 [ 13] consists of 1 7 cases where the pin pitch was varied from 1.3 cm to 2.3 cm and different quantities of pins were used to change the critical height. An 18 111 case was done where the array was mo ve d to the edge of the tank , so the boundary was the steel tank followed by void. This condition i s not typical of a spent fuel pool , so this case was not analyzed.

The average k eff of the Phase 1 cases was 0.99910. Phase 2 [ 14] consisted of 20 cases where gadolini u m of various concentrations was dissolved in the water (Phase 2a) and 21 cases where boron was dissolved in the water (Phase 2b). These experiments also varie d the pitch ( 1.3 to 1.9 cm) and the number of pins. The average k eff of the gadolinium cases was 0.998 15 and the average for the boron cases was 0.99897. Phase 3 [ 1 5] consists of 26 experiments where the p i ns were arra n ged as 4 "asse mb l ies." Eac h assembly used a 1.6 cm pin pitch. The assembly separation was varied, as well as the number of pins in each assembly.

Finally , eleven cases boxed the assemb l ies with an absorber (borated steel, boral, or cadmium).

The average k eff of these 26 cases was 0.99890. Finally, Phase 4 [ 16] consisted of redoing the same type of experiments as Phase 3, except with reflector screens. The 38 experiments which used the lead reflector screen were not included in this ana l ysis, since lead reflectors are not common in spent fuel pools. The 33 steel reflector experiments were included. The average k eff of these cases was 0.99858. References 13 through 16 provided a ll of the detai l s for the analysis.

The mode l ing was straight forward. The references gave a simp l e model and a detailed model. The mode l created for this work fo ll owed the detai l ed m odel , except t h at the top grid outside of the array and the basket supports were not mo d e l ed. Both of these assumptions were part of the simplified model and have a negligible impact on k. The model used actually exceeded the detailed model , since the spring above the fuel was modeled by homogenizing it with the void. NET-28091-0003-01, Revision O A-33 Tables A.6 through A.10 present the results of the analysis.

A statistical analysis of the HTC set as a whole was performed consistent with the method provided in NUREG/CR-6698 , where the experimental uncertainties were taken from References 13 through 16. The mean uncertainty weighted k e ff is 0.99878 and the uncertainty is 0.00590. This makes th e bias 0.00122. Since all of the pins are the same , trend analysis on the pin diameter and enrichment are not possible. The pin pitch changes are made to adju s t the spectrum, so the only trend analysis performed is on the spectrum (EALF). The trend analysis on the HTC set (performed consistent with NUREG/CR-6698) on EALF yielded the followin g function:

k(EALF) = 0.999541 -0.00548

  • EALF The units for EALF are eV. The uncertainty about the trending k e ff is 0.0076 ink. Figure A.7 shows the results of the HTC analysis.

Table A.6: HTC Phase 1 Results Case No. kerr Mo n te EALF Pitch Car l o (eV) (cm) S i g m a 1 0.99913 0.00015 0.069486 2.3 2 0.99893 0.00016 0.066544 2.3 3 0.99892 0.00016 0.066412 2.3 4 0.99974 0.00017 0.084957 1.9 5 0.99983 0.00017 0.082795 1.9 6 0.99946 0.00020 0.082123 1.9 7 0.99977 0.00019 0.102248 1.7 8 0.99962 0.00018 0.100654 1.7 9 0.99903 0.00019 0.099687 1.7 10 0.99991 0.00019 0.140669 1.5 11 0.99898 0.00020 0.135753 1.5 12 0.99906 0.00019 0.133996 1.5 13 0.99813 0.00021 0.256212 1.3 14 0.99776 0.00019 0.234183 1.3 15 0.99812 0.00022 0.230564 1.3 16 0.99952 0.00020 0.101408 1.7 17 0.99882 0.00019 0.099384 1.7 NET-28091-0003-01 , Revision 0 A-34 Table A.7: HTC Phase 2a, Gadolinium Solutions, Results Case No. k.rr Monte EALF Pitch Gadolinium Carlo (eV) (cm) Concentration Sigma (Q:/1) 1 0.99784 0.00020 0.25279 1.3 0.0520 2 0.99792 0.00021 0.24946 1.3 0.0520 3 0.99777 0.00019 0.27074 1.3 0.1005 4 0.99771 0.0001 8 0.26 756 1.3 0.1005 5 0.99784 0.0001 8 0.2 6333 1.3 0.1005 6 0.99683 0.0001 8 0.28513 1.3 0.1505 7 0.99684 0.00019 0.27847 1.3 0.1505 8 0.99623 0.00016 0.29552 1.3 0.1997 9 0.99608 0.00018 0.29253 1.3 0.1997 10 0.99689 0.00017 0.169 82 1.5 0.1997 11 0.99766 0.00019 0.16252 1.5 0.1495 12 0.99771 0.00018 0.16101 1.5 0.1495 13 0.99868 0.00017 0.15392 1.5 0.1000 14 0.99861 0.00018 0.15223 1.5 0.1000 15 0.99983 0.00020 0.14727 1.5 0.0492 16 0.99976 0.00019 0.14432 1.5 0.0492 17 1.00053 0.00018 0.106 3 1 1.7 0.0492 18 1.00070 0.00017 0.087 83 1.9 0.0492 19 0.99707 0.00016 0.11369 1.7 0.1010 20 1.00050 0.00019 0.10648 1.7 0.0492 NET-28091-0003-01, Revi s ion 0 A-35 Table A.8: HTC Phase 2b, Boron Solutions, Results Case No. kerr Monte EALF Pitch Boron Carlo (eV) (cm) Concentration Si2ma {l?:/1) 1 0.99835 0.00020 0.24780 1.3 0.100 2 0.99760 0.00020 0.24450 1.3 0.106 3 0.99816 0.00020 0.25528 1.3 0.205 4 0.99904 0.00020 0.26400 1.3 0.299 5 0.99886 0.00019 0.27475 1.3 0.400 6 0.99852 0.00019 0.27125 1.3 0.399 7 0.99933 0.00018 0.27977 1.3 0.486 8 0.99894 0.00019 0.28781 1.3 0.587 9 0.99952 0.00016 0.16627 1.5 0.595 10 0.99811 0.00019 0.16087 1.5 0.499 11 0.99990 0.00017 0.15663 1.5 0.393 12 0.99987 0.00018 0.15007 1.5 0.295 13 0.99887 0.00018 0.14559 1.5 0.200 14 1.00192 0.00018 0.14024 1.5 0.089 15 1.00338 0.00018 0.10325 1.7 0.090 16 1.00202 0.00017 0.10717 1.7 0.194 17 1.00313 0.00017 0.11049 1.7 0.286 18 0.99367 0.00017 0.11577 1.7 0.415 19 1.00021 0.00021 0.10473 1.7 0.100 20 0.99251 0.00017 0.08965 1.9 0.220 2 1 0.99642 0.00017 0.08611 1.9 0.110 NET-28091-0003-01, Revision 0 A-36 Table A.9: HTC Phase 3 Results -Water Reflected Assemblies

  • (1.6 cm pin pitch) Case No. k e rr Monte EALF Absorber Assembly Carlo (eV) Box Separation Si2ma Material (cm) 1 0.99 774 0.00022 0.12377 Borated SS 3.5 2 0.99986 0.00019 0.14095 Borated SS 0 3 0.99710 0.00019 0.12939 Borated SS 2 4 0.99715 0.0001 8 0.12391 Borated SS 3 5 0.9969 9 0.00018 0.13503 Borated SS 1 6 0.99987 0.00019 0.1 2974 Bora) 0 7 0.99614 0.00019 0.12866 Cd 2 8 1.00381 0.0001 8 0.13904 Cd 0 9 0.99 646 0.00017 0.1 3345 Cd 1 10 0.99672 0.0001 8 0.12952 Cd 1.5 11 0.99571 0.00019 0.13726 Cd 0.5 12 0.999 01 0.00017 0.11 277 none 1 8 13 0.999 15 0.0001 8 0.11167 none 14.5 14 0.99934 0.0001 8 0.111 83 none 11 15 0.99910 0.00019 0.11093 none 10 16 0.999 61 0.00019 0.110 3 0 none 9 17 0.99930 0.0001 8 0.10842 non e 8 18 0.9 9980 0.00017 0.1065 6 none 6 19 1.00016 0.0001 8 0.10421 none 4 20 1.00044 0.0001 8 0.10206 none 4 21 0.99976 0.0001 8 0.10470 none 2 22 1.00047 0.00019 0.10714 none 1 23 0.998 93 0.0001 8 0.11506 non e 0 24 0.99949 0.00020 0.15073 none 0 25 0.999 96 0.0001 8 0.12672 none 4 26 0.99937 0.00020 0.11550 non e 10 NET-28091-0003-01, Revisi o n 0 A-37 Case No. 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 1 8 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 Table A.10: HTC Phase 4 Results -Steel Reflected Assemblies (1.6 cm pin pitch) k c rr Monte EALF Absorber Assembly Separation Carlo (eV) Box Separation From Reflector Sigma Material (cm) (cm) 1.00157 0.00019 0.15363 Borat ed SS 0 0.0 0.99845 0.00018 0.15069 Borat ed SS 0.5 0.0 0.99797 0.00018 0.14674 Bo rated SS 1 0.0 0.99826 0.00018 0.14 227 Borat ed SS 1.5 0.0 0.99839 0.00019 0.13923 Borated SS 2 0.0 0.99712 0.00018 0.13820 Borat ed SS 2 0.5 0.99634 0.00018 0.13705 Borated SS 2 1.0 0.99650 0.00018 0.13598 Borat ed SS 2 1.5 0.99658 0.00018 0.13518 Borated SS 2 2.0 0.99834 0.00018 0.13430 Borated SS 3 0.0 0.99821 0.00018 0.13234 Borated SS 3.5 0.0 1.00095 0.00018 0.13558 Bora! 0 0.0 0.99653 0.00018 0.13386 Bora! 0.5 0.0 1.00431 0.00017 0.14979 Cd 0 0.0 0.99818 0.00020 0.14323 Cd 1 0.0 0.99769 0.00017 0.13683 Cd 2 0.0 0.99615 0.00018 0.13568 Cd 2 0.5 0.99536 0.00019 0.13423 Cd 2 1.0 0.99513 0.00018 0.13315 Cd 2 1.5 0.99465 0.00018 0.13235 Cd 2 2.0 0.99869 0.00018 0.13390 Cd 2.5 0.0 1.00060 0.00018 0.17427 none 0 0.0 1.00057 0.00018 0.16641 non e 1 0.0 0.99973 0.00018 0.15852 non e 2 0.0 0.99935 0.00018 0.15709 none 2 0.5 0.99946 0.00018 0.15559 none 2 1.0 0.99939 0.00018 0.15431 none 2 1.5 0.99937 0.00019 0.15351 none 2 2.0 0.99941 0.00019 0.14426 non e 4 0.0 0.99964 0.00018 0.13456 none 6 0.0 0.99953 0.00018 0.12 886 none 8 0.0 0.99947 0.00017 0.12537 none 10 0.0 0.99940 0.00018 0.12333 non e 12 0.0 NET-28091-0003-01, Revi s ion 0 A-3 8

.:,: ',:J 41 1.006 1.004 1.002 1 * ------.-** * ---.

  • 0.998 +---------
  • ... ;; u
  • 0.996 +-----------

0.994 +-------* *:I I * * * .. # 0.992 +-----------------------------------

0.99 +-----,-------,------r-----,-----.,------,-------, 0.000000 0.050000 0.1 00000 0.1 50000 0.200000 0.250000 0.300000 0.350000 Energy of the Average Lethargy of Fission (EALF) (ev) Figure A.7: k eff as a Function of the EALF for the HTC Experiments A.3.2 MOX Critical Experiments The se l ect ion of the MOX critical exper im ents was limi ted to the l ow enriched MOX lattice cri ti cal experime nt s. All 63 of the low enric h ed MOX pin critical experiments documented in the OECD handbook [1 7] were utili zed. The ac tu a l inpu t decks were initiated from avai l ab l e decks found in NUREG/CR-6102 [18] and the International Handbook.

[1 7] The decks were modified to update to the new cross-section library and changes in the SCALE inpu t format. Table A.11 presents the results of t h e 63 se l ected MOX critical experiments.

The Reference column has the eva lu a ti on number from the International Hand book. [17] For example, OECD-7 refers to the OECD International Handbook case MIX-COMP-THERM-07. Trends were in vestigate d as a function of EALF, plutonium conte nt , and the Am-241/U-238 ratio. As the spectrum h a rd ens (hig h er EALF), t h ere is a small trend to higher k. With more plutonium conte nt , k e ff increases. This is seen in Figure A.8. NET-28 091-0003-01 , R evision 0 A-39 The change in ke rr with cooling time is dominated by the reactivity of the decay of Pu-241 to Am-241. By plotting kerr versus the Am-241/U-238 ratio , it i s possible to determine if the bia s should be chang e d for cooling. Figure A.9 shows that with increasing Am-241 c ontent , the calculated k err of th e critical experiments increases. This obser v ation show s that the z ero cooling time bias conser v ati v ely covers the cooling time. Table A.11: Results of MOX Critical Benchmarks (SCALE 6.1.2, ENDF/B-VII)

Case ID Reference k e ff sigma EALF Pu Pu Am241/U238 (eV) wt% 240% 093array OECD-7 1.0009 0.00025 0.1903 2.00 16 6.82E-05 105al.in OECD-7 0.9942 0.00027 0.1 3 69 2.00 16 7.55E-05 105array OECD-7 0.9960 0.00025 0.1377 2.00 16 7.55E-05 105bl OECD-7 0.9914 0.00026 0.1379 2.00 16 7.55E-05 105b2 OECD-7 0.9921 0.00024 0.1377 2.00 16 7.55E-05 105b3 OECD-7 0.9933 0.00025 0.1373 2.00 16 7.55E-05 105b4 OECD-7 0.9940 0.00026 0.1371 2.00 16 7.55E-05 l 143arra OECD-7 0.9980 0.00026 0.1166 2.00 16 8. l 3E-05 132array OECD-7 0.9971 0.00022 0.0953 2.00 16 8.13E-05 1386arra OECD-7 0.9942 0.00023 0.0906 2.00 16 6.97E-05 epri70b OECD-2 0.9992 0.00025 0.7209 2.00 7.8 7.29E-05 epri70un OECD-2 0.9974 0.00027 0.5409 2.00 7.8 7.29E-05 epri87b OECD-2 1.0019 0.00022 0.2710 2.00 7.8 7.29E-05 epri87un OECD-2 0.9981 0.00032 0.1852 2.00 7.8 7.29E-05 epri99b OECD-2 1.0012 0.00024 0.1772 2.00 7.8 7.29E-05 epri99un OECD-2 1.0007 0.00027 0.1333 2.00 7.8 7.29E-05 klmct009 OECD-9 0.9994 0.00024 0.5169 1.50 8 l.06E-05 k2mct009f OECD-9 0.9941 0.00027 0.2 9 43 1.50 8 9.77E-06 k3mct009 OECD-9 0.9934 0.00024 0.152 8 1.50 8 8.96E-06 K4mct009 OECD-9 0.9921 0.00024 0.1155 1.50 8 8.96E-06 K5mct009 OECD-9 0.9925 0.00021 0.0947 1.50 8 8.96E-06 K6mct009 OECD-9 0.9937 0.00024 0.0905 1.50 8 9.77E-06 omct61 OECD-6 0.9954 0.00026 0.3570 2.00 8 2.24E-05 omct62 OECD-6 0.9990 0.00029 0.1 8 85 2.00 8 2.24E-05 omct63 OECD-6 0.9943 0.00027 0.1374 2.00 8 2 , 24E-05 omct64 OECD-6 0.9982 0.00025 0.1167 2.00 8 2.24E-05 omct65 OECD-6 0.9994 0.00025 0.0956 2.00 8 2.24E-05 omct66 OECD-6 0.9956 0.00024 0.0907 2.00 8 2.24E-05 mct8cl OECD-8 0.9978 0.00029 0.3776 2.00 24 7.93E-05 mct8c2 OECD-8 0.9977 0.00028 0.1922 2.00 24 7.27E-05 mct8c3 OECD-8 0.9967 0.00024 0.13 8 3 2.00 24 8.59E-05 mct8c4 OECD-8 1.0006 0.00027 0.1170 2.00 24 9.88E-05 mct8c5 OECD-8 1.0000 0.00026 0.0955 2.00 24 9.56E-05 mct8c6 OECD-8 0.9992 0.00023 0.0905 2.00 24 7.27E-05 mct8cal OECD-8 0.9967 0.00025 0.1375 2.00 24 8.59E-05 mct8cbl OECD-8 0.9931 0.00024 0.1387 2.00 24 8.59E-05 mct8cb3 OECD-8 0.9941 0.00025 0.1381 2.00 24 8.59E-05 NET-28091-0003-01, Revision 0 A-40 Case ID Reference kerr sigma EALF Pu Pu Am241/U238 (eV) wt 0/o 240% meteb2 OECD-8 0.9937 0.00024 0.1385 2.00 24 8.59E-05 meteb4 OECD-8 0.9942 0.00026 0.1378 2.00 24 8.59E-05 mixo25lk OECD-5 1.0011 0.00032 0.3732 4.00 18 l.59E-04 mixo252k OECD-5 0.9985 0.00027 0.2476 4.00 18 l.59E-04 mixo253k OECD-5 1.0044 0.00027 0.1712 4.00 18 l.59E-04 mixo254k OECD-5 1.0004 0.00029 0.1425 4.00 18 l.59E-04 mixo255k OECD-5 1.0034 0.00028 0.1058 4.00 18 l.59E-04 mixo256k OECD-5 1.0023 0.00024 0.0917 4.00 18 l .59E-04 mixo257k OECD-5 1.0036 0.00024 0.0 8 75 4.00 18 l .59E-04 saxtnl04 OECD-3 1.00044 0.00027 0.0987 6.60 8.6 8.43E-05 s axtn56b OECD-3 0.99962 0.00028 0.6133 6.60 8.6 8.43E-05 saxtn735 OECD-3 0.99999 0.00031 0.1 8 20 6.60 8.6 8.43E-05 saxtn792 OECD-3 0.99951 0.00031 0.1505 6.60 8.6 8.43E-05 Saxton52 OECD-3 0.99977 0.00028 0.8 517 6.60 8.6 8.43E-05 Saxton56 OECD-3 1.00018 0.0003 0.5177 6.60 8.6 8.43E-05 teal OECD-4 0.99572 0.00027 0.1418 3.01 22 l.04E-04 tealO OECD-4 0.9988 0.00024 0.0792 3.01 22 9.3 lE-05 teal 1 OECD-4 0.99886 0.00023 0.0788 3.01 22 2.06E-04 tea2 OECD-4 0.9964 0.0003 0.1409 3.01 22 l.99E-04 tea3 OECD-4 0.99665 0.00028 0.1403 3.01 22 2.96E-04 tea4 OECD-4 0.99644 0.00026 0.1172 3.01 22 9.88E-05 tea5 OECD-4 0.9974 0.00027 0.1167 3.01 22 2.02E-04 tea6 OECD-4 0.99848 0.00025 0.1156 3.01 22 3.90E-04 tea7 OECD-4 0.99753 0.00025 0.0917 3.01 22 8.88E-05 tea8 OECD-4 0.99801 0.00025 0.0913 3.01 22 2.03E-04 tea9 OECD-4 0.99864 0.00025 0.0909 3.01 22 3.02E-04 NET-28091-0003-01 , Revision 0 A-41 1.0060 -----------*

  • 1.0040 -----* 1.0020 * *
  • 1.0000 * --------;: I
  • 0.9980 .x
  • 0.9960 ' --.--------0.9940 ---.11-* t
  • 0.9920 -----------* 0.9900 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 P u wt% Figure A.8: Predicted k err as a Function of the Plutonium Content NET-28091-0003-01 , R evision 0 A-42
i: QI 1.0060 1.0040 *
  • 1.0020
  • t * *
  • 1.0000 i * * ,.
  • 0.99 8 0 * ---* ** *
  • 0.9960
  • 0.99 4 0
  • 0.9900 +---~--~---.----~--~---,-----r---~----,,---

0.0 E +OO 5.0E-05 1.0E-04 1.SE-04 2.0E-04 2.SE-04 3.0E-04 3.SE-04 4.0E-04 4.SE-04 5.0E-04 Ratio of Am-241 to U-238 Figure A.9: Predicted kerr as a Function of the Am-241 Content A.3.3 Bias and Uncertainty from the MOX/HTC Critical Experiments The bias and uncertainty of burned fuel depends on the amount of plutonium in the burned fuel. As shown on Figure A.9 the bias decreases with plutonium content. However , the uncertainty increases with plutonium content. In order to determine appropriate biase s and uncertaintie s the HTC and MOX critical benchmarks are combined. The MOX experiments with plutonium content above 2 wt% were useful for confirmation that the bias decreases with plutonium content , but the maximum plutonium content in spent nuclear fuel about 1.5 wt% plutonium , so using experiments above 2 wt% plutonium needl e ss l y increa s e s the uncertainty. The bias and uncertainty in the bias for the HTC/MOX (2 wt% PU or less) set is controlled by the EALF trend. For EALF's less than 0.4 eV the maximum bias and uncertainty are 0.0021 and 0.0087 respectively.

For EALF's from 0.4 to 0.65 eV the maximum bias and uncertaint y are 0.0027 and 0.0112 respectivel

y. NET-28091-0003-01 , Revision 0 A-4 3 A.4. Temperature Dependent Critical Experiments Since the criticality analysis for spent fuel pools must consider the full range of temperatures a ll owed for the poo l the LCT-46 set of critical experiments are needed to assure the correct bias and uncertainty is used for conditions where the pool is at its highest temperatures. The suite of critical experiments other than LCT-46 contains a range of fuel to moderator ratios that shou l d adequately cover the impact of the density change in the water as the pool temperature rises but no other experiments test the Doppler broadening of the cross sections or the change in the thennal scattering.

LCT-46 consists of 22 experiments but the last 5 experiments contain copper rods. Since copper is not norm a lly in spent fuel pools only the fir s t 17 experiments are analyzed here. Section 3 of LCT-046 specifies the critical benchmark and the SCALE models used follow that specification.

The specification has a couple of minor ambiguities related to the thermal expansion given as Table 29 of LCT-046. For this ana l ysis all of the expansion factors from Table 46 were app l ied to all of the x-y dimensions.

That mean s that the same SS component expansion factor was applied to pitch and the inner and outer diameter of the clad. This is consistent with the MCNP samples given in the Appendix of LCT-046. For the axial expansion only the fuel was expanded.

As with the MCNP sample input the same expansion factor was used for the radius and the axial dire ct ion. Table A.12 shows the corrected SCALE 6.1.2 ENDF/B-VII results for the 17 critical experiments.

Corrected r es ults in this case means they were divided by the kerr of the benchmark which was not quite 1.0. Table A.12: LCT-46 w i t h Full Thermal Expansi o n Case Temperature (K) Corrected SCALE k SCALE s i i!ma 1 297.05 0.999082 0.000065 2 310.41 0.998902 0.000071 3 315.43 0.998817 0.000067 4 319.96 0.998908 0.000073 5 324.93 0.998629 0.000067 6 332.53 0.998746 0.000067 7 287.22 0.999148 0.000067 8 3 15.91 0.998819 0.000066 9 330.27 0.998696 0.000068 10 337.44 0.998804 0.000065 11 351.99 0.998829 0.000068 12 303.6 0.998649 0.000065 13 312.95 0.998641 0.000069 14 321.1 6 0.998556 0.000067 15 328.24 0.998401 0.000068 16 338.26 0.998318 0.000067 17 358.3 1 0.99825 6 0.000065 NET-28091-0003-01, Revision 0 A-44 Figure A.10 plots the results of the analysis as a function of case. As can be seen from this plot there does seem to be a trend wit h t emperature. Figure A.11 i s the dat a plotted against temperature wit h the l ea s t squares lin ear fit. T h e fit is statistica ll y significant.

The slope is -8.6E-6 deltak/°C. The uncertainty arou nd th e fit is 0.0013. The bias is d etermined b y multiplying th e change from room tempera tur e in °C b y 8.6E-6. The uncertainty of 0.00 1 3 is an independent un certai nt y that can be statistically co mbin ed with the other uncertainties.

k versus Case (three sets with increasing temperature in each set) 0.999200 0.999100 X 0.999000 0.998900 0.998800 -i, QI ... .!! 0.998700 :I u a o.998600 0.998500 0.998400 0.998300 Rectangu l a r Set X X X Ro u nded Set X X X -xx 4 Gd Rods Set X 0.998200 T ---,-----, T ,--T -,-----, 0 2 4 6 8 10 12 14 16 18 Case Number Figure A.10: LCT-046 Corrected Calculated k eff per Case NET-28091-0003-01 , R evision 0 A-45 0.999200 t 0.999 10 0 0.999000 --:,r. 0.998900 ... ., .1! 0.998800 :, u a o.998700 ...

  • 0.998600 -j----§ 0.998500 0.99 84 0 0 0.998300 0.998200 + 290 300 *
  • 310 * * * * * *
  • 320 330 340 350 360 Temperature (K} Figure A.11: LCT-046 Corrected Calculated keffVersus Temperature 370 It is common practice not to thermally expand the solids when doing analysis of elevated temperatures in criticality analysis. Table A.13 shows the re s ults of the analysis repeated where the temperatures of the materials were increased and the density of the water decreased but no thermal expansion of the solids (fuel , pitch, etc.). As can be s e en in Table A.13 the difference between just expanding the water (lowering the density) and full thennal expansion is similar to the Monte Carlo uncertainty. The maximum difference is 0.00027 which is less than 4 times the Monte Carlo one sigma uncertainty of one of the two calculations used in the difference.

NET-28091-0003-01, Revision 0 A-46 Table A.13: LCT-46 with No Thermal Expansion of Solids Difference in k e rt From Case Temperature (K) Corrected SCALE k SCALE si!!.ma Full Thermal Expansion 1 297.05 0.999114 0.000069 -0.00003 2 3 10.4 1 0.998899 0.000068 0.00000 3 3 15.43 0.998736 0.000067 0.00008 4 3 1 9.96 0.998666 0.000074 0.0 00 24 5 324.93 0.998491 0.000069 0.00014 6 332.53 0.998 743 0.000069 0.00000 7 287.22 0.999302 0.000067 -0.00015 8 3 15.9 1 0.9988 74 0.00 00 70 -0.00 006 9 330.27 0.99859 7 0.000068 0.00010 10 337.44 0.998564 0.000067 0.00024 11 351.99 0.998558 0.000070 0.00027 1 2 303.6 0.998633 0.000070 0.00002 1 3 312.95 0.998651 0.0000 7 0 -0.00001 1 4 32 1.1 6 0.998469 0.000067 0.00009 15 328.24 0.998 420 0.000067 -0.00002 1 6 338.26 0.998500 0.000067 -0.00018 1 7 358.3 1 0.998042 0.000067 0.00021 Since the hi g h er temperatures have a h ar d er spect rum , th e effect of the hi g her temperatures could h ave already b ee n capture d in the trend on spec trum (EALF). This was t es t ed by using the slope of th e chan ge in k e ff wit h EALF from th e full set of cri ti ca l experiments. Using the EALF biased k s th e lin ear fit was reanalyzed.

The maximum bi as (0 to 100 C) cha n ged from 0.00086 to 0.00071. The spectrum i s a small amount of th e t empera tur e effect a nd i s therefore ignored for the final conclusion.

The ana l ys i s of the on l y set of thermal cr iti ca l expe rim ents in the Int e rn ationa l Handbook that u ses e l evate d temperatures b e lo w b oiling h as s h own a s m a ll in crease in th e bi as wit h t e mp erature. The bi as is d e t ermi n e d b y multiplying the change from room temperature in °C b y 8.6E-6. The un certa inty of 0.0013 i s an independent uncertainty that can b e statistica ll y combined with the other unc e rt ainties. NET-2809 1-000 3-01, R ev i sion 0 A-47 A.5. Summary of Validation Using Laboratory Critical Experiments Nuclear fuel starts as U0 2 and as it bums it becomes a mixture of U0 2 and Pu 0 2. SCALE 6.1.2 with the ENDF/B-VII.O cross sections calculates a slightly higher k eff as the Pu0 2 content increases.

The correct bias and uncertainty shou l d be a function of the plutonium weight percent but this would be overly complicated for a small effect. The bias for the initial condition from U0 2 critical experiments would be conservative for spent nuclear fuel. However , the uncertaint y from the U0 2 only set is smaller than the uncertainty from the MOX set. To conservatively cover the all of the conditions of the fuel the final 95/95 ke ff is calculated twice , once using the U0 2 critica l experiments bias and uncertainty and once using the MOX/HTC bias and certainty.

The higher final 95/95 k e ff is used for comparison to the k e ff criteria. The two bias and uncertainty sets are: 1. Based on the U0 2 experiments

For EALF's less than 0.4 eV the bias is 0.0024. For EALF's between 0.4 ev and 0.65 eV the bias is 0.0036. The uncertainty for the entire range of EALF is 0.0035. 2. Based on the MOX/HTC experiments
For EALF's less than 0.4 eY the bias is 0.0021. For EALF's between 0.4 ev and 0.65 eV the bias is 0.0027. The uncertainty for the range of EALF 0 to 0.4 eV is 0.0087. The uncertainty for the range ofEALF 0.4 to 0.65 eV is 0.0112 For all burned fuel the MOX/HTC bias and uncertainty actually determine the 95/95 k. For most cases in the pool analysis, the most dense water conditions are most limiting.

However , if higher temperature cases are more lirniting , then a temperature bias of 8.6E-6 multiplied by the change from room temperature in °C is applied. In addition the uncertainty in this bias, 0.0013 , needs to be included in the uncertainty rack up. NET-28091-0003-01, Revision 0 A-48 A. 6. Appendix References

[ 1] Scal e: A Compreh e nsiv e Mod e ling and Simulation Suite for Nucl e ar Saf e ty Analysis and D e sign , ORNL/TM-2005

/39, Version 6.1 , June 2011. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-785. [2] J.C. Dean and R.W. Tayloe, Jr., Guid e for Validation of Nucl e ar Criti c ality Saf e ty Cal c ulational M e thodology , NUREG/CR-6698 , Nuclear Regulatory Commission, Washington , DC January 2001. [3] Int e rnational Handbook of Evaluat e d Criti c ali ty Safety B e n c hmark Exp e rim e nts, NEA/NSC/DOC(95)3, Volume IV, Nuclear Energy Agency, OECD , Paris, September , 2016. [ 4] DAT APLOT is statistical software supported by the National Institute of Standards and Technology.

It can be down loaded at: http://www.itl.nist.gov

/div898/software/dataplot/ [5] J. J. Lichtenwalter , S. M. Bowman, M. D. DeHart , and C. M. Hopper, Criti c ality B e n c hmark Guid e for Light-Wat e r-R e actor Fu e l in Tran s portation and Storag e Packa ges , NUREG/CR-6361 (ORNL/TM-13211), Spent Fuel Project Office, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission , Washington , DC 20555-0001, March 1997. [6] L. W. Newman , et al, Urania-Gadolinia:

Nucl e ar Model D e v e lopm e nt and Critical Exp e rim e nt B e n c hmark, BA W-1810 , Babcock & Wilcox , Utility Power Generation Division , Lynchburg , VA , April 1984. [7] Brian L. Koponen and Viktor E. Hampel, Nucl e ar Criticality Saf e ty Exp e rim e nts , Cal c ulations, and Anal y s e s-1958 to 1982 , UCRL-53369 , Lawrence Livermore Laboratory , University of California , Livermore , California , October 21 , 1982. [8] M. Rahimi , E. Fuentes, and D. Lancaster, I s otopi c and Criti c ality Validation/or PWR Actinid e-OnlyBurnup Credit, DOE/RW-0497, U.S. Department of Energy , Office of Civilian Rad i oactive Waste Management, Washington, DC, May,1997. [9] [NOT USED] [10] [NOT USED] [11] [NOT USED] [12] D. E. Muel l er, K. R. Elam, and P. B. Fox , Evaluation of th e Fr e n c h Haut Tau x d e Combustion (HTC) Critical Exp e rim e nt Data , NUREG/CR-6979 (ORNL/TM-2007

/083), prepared for the US Nuc l ear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., September 2008. [13] F. Femex , "Programme HTC -Phase 1 : Reseaux de crayons dans l'eau pure moderated and reflected simple arrays) Reevaluation des experiences

," DSU/SEC/T/2005-33/D.R., Institut de Radioprotection et de Surete Nucleaire, 2008. NET-28091-0003-01 , Rev i sion 0 A-49

[14] F. Fernex , Pro g ramm e HTC-Pha se 2: R ese au x s impl e s e n e au e mpoi s onn ee (bor e e t gad o linium) (R e fl ecte d simpl e arra y s mod e rat e d b y poison e d wa t e r with g adolinium or boron) R ee valuation d e s e xp e ri e n ces, DSU/SEC/T/2005-38/D.R., lnstitut de Radioprotection et de Surete Nucleaire , 2008. [ 15] F. Fernex , Pro g ramm e HTC -Pha se 3 : Confi g urations "s to c ka ge e n pi sc in e" (Pool stora ge) R eev aluation d es ex p e ri e nc e s , DSU/SEC!T/2005-37/D.R., lnstitut de Radioprotection et de Surete Nucleaire , 2008. [16] F. Fernex, Pro g ramm e HTC-Pha se 4: Confi g urations "c h a t e aux d e tran s port" (Shippin g c a s k) -R eev aluation d e s e xp e ri e n ces , DSU/SEC!T/2005-36/D.R., Institut de Radioprotection et de Surete Nucleaire , 2008. [17] Int e rnational Handbook of E v aluat e d Criti c ali ty Saf e ty B e n c hmark E x p er im e nt s, NEA/NSC/D0C(95)3, Volume VI , Nuclear Energy Agency , OECD , Paris , September , 2010. [18] M. D. DeHart and S. M. Bowman , Validation o f th e SCALE B roa d Stru c tur e 44-Group E N DF I B-V Cro ss-S ect ion Libra ry for U se in Criti ca li ty Saf ety A nal yse s , NUREG/CR-6102 (ORNL!T M-12460), Oak Ridge National Laboratory , Oak Ridg e, TN , September 1994. NET-28091-0003-01 , Revision 0 A-50 Appendix B: Fuel Categorization for Unit 2 Batches A Through X and Unit 3 A through AA All of the early discharged fuel has been categorized. Nearly all of the fuel is either Category 4 or Category 5. The table has been color coded to quickly identify the Category.

Category 3 is yellow , Category 4 is green, and Category 5 is b l ue. A range of assembly IDs that have the same Category are grouped together to reduce the length of the table. All but two assemblies for historical fuel at Unit 3 ha v e been categorized as Category 4 even though about ha l f of them could have been Category 5. Since the spent fuel pool at Unit 2 is used only temporarily for Unit 3 fue l while it is being casked, the lower reactivity Category is not needed. NET-2809 1-0003-01, Rev i sion 0 B-1 Table B.1: F u e l A s s embl y R e acti v i ty C ate g ori za tion for Asse mbl y ID s A t hrou g h X for Unit 2 Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category A01-A65 4 E43-ESS 4 K01-Kl3 4 E56 3 K14-Kl5 5 801-807 4 E57-E60 4 K16-K57 4 808-813 5 K58 5 814-823 4 FOl 3 K59-K68 4 B24-B26 5 F02-F20 4 827-B64 4 F21 3 L01-L07 4 F22-F30 4 L08-Ll0 5 C01-C04 4 F31-F34 5 Lll-L63 4 COS-C06 5 F35 4 L64 3 C07-C12 4 F36 3 L65-L68 4 C13 5 F37-F39 4 (14 4 F40 3 M01-M04 4 C15-C18 5 F41-F49 4 MOS 5 C19-C28 4 FSO 3 M06-M08 4 C29 5 F51-F60 4 M09 5 C30-C64 4 F61 3 M10-M12 4 F62-F64 4 M13-M14 5 D01-D25 4 F65 3 M15-M20 4 D26 5 F66 4 M21 5 D27-D60 4 F67-F68 5 M22-M23 4 D61-D68 5 M24 5 D69-D72 4 GOl-GOS 4 M25-M27 4 G06 5 M28 5 E01-E14 4 G07-G37 4 M29-M30 4 ElS 3 G38 5 M31 5 E16-E19 5 G39-G72 4 M32-M34 4 E20 4 M35 5 E21-E24 5 H01-H38 4 M36-M37 4 E25-E27 4 H39-H51 5 M38-M44 5 E28-E31 5 H52-H54 4 M45 3 E32-E33 4 HSS 5 M46 4 E34-E35 5 H56 4 M47-M48 5 E36-E40 4 M49-MSO 4 E41-E42 5 J01-J68 4 M51-M52 5 NET-2809 1-0003-01 , Revision 0 B-2 Table B.1: F u e l Asse mbl y Reactivit y Categorization for Assembl y IDs A throu g h X for U nit 2 (Continued)

Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category M53-M54 4 Q71-Q73 4 T42-T43 4 MSS-MSG 5 Q74-Q76 5 T44-T46 5 M57 4 Q77 4 T47 4 M58-M59 5 Q78 5 T48 5 MGO 4 Q79-Q80 4 T49-T51 4 M61 3 T52-T53 5 M62-M63 4 R01-R07 5 T54 4 M64 3 ROB 4 TSS 5 MGS 4 R09-R38 5 T56-T72 4 M66 5 R39 4 T73-T80 5 M67 3 R40-R43 5 M68 5 R44-RSO 4 M69-M71 4 R51-R69 5 U01-U04 5 M72 5 R70 4 uos 4 R71-R72 5 U06-U13 5 N01-N08 4 R73-R74 4 U14 4 N09-N12 5 R75-R79 5 U15-U16 5 N13-N14 4 R80-R81 4 U17-U21 4 N15-N16 5 R82 5 U22 5 N17-N23 4 R83-R85 4 U23 4 N24-N32 5 U24-U49 5 N33-N47 4 S01-S44 5 USO 4 N48 5 S45 4 USl 5 N49-N80 4 S46-S47 5 U52 4 S48 4 U53-U61 5 P01-P02 4 S49-S61 5 U62-U64 4 P03 3 S62 4 UGS 5 P04-P47 4 S63-S65 5 U66-U68 4 P48 5 S66 4 U69-U73 5 P49-P60 4 S67-S77 5 P61-P72 5 V01-V16 5 V17-V29 4 QOl-QGS 5 T01-T32 5 V30-V35 5 QGG 4 T33-T34 4 V36 4 Q67-Q68 5 T35-T36 5 V37-V38 5 Q69 4 T37 3 V39 4 Q70 5 T38-T41 5 V40-V41 5 N ET-28 091-000 3-0 1 , Re v i s ion 0 B-3 Ta bl e B.1: Fue l Assembly React i vity Categor i zation for Assembly IDs A through X for Unit 2 (Co n t inu ed) Assembly ID Category V42-V43 4 V44-V49 5 vso 4 V51-V54 5 VSS-V57 4 V58-V61 5 V62 4 V63 5 V64-V65 4 V66-V67 5 V68 4 V69-V77 5 V78-V79 4 V80-V81 5 V82 4 V83 5 V84 4 V85 5 V86 4 V87-V88 5 V89 4 V90-V91 5 V92 4 WOl-WlO 4 Wll 5 W 1 2-W15 4 W16 5 W17 4 W18-W19 5 W20 4

  • FRSB is the fuel rod storage ba s ket NET-28091-0003-01 , Revision 0 Indian Point Uni t 2 F uel Assembly ID Category Assembly ID Category W21 5 X01-X02 3 W22 4 X03-X04 5 W23 5 XOS-X37 4 W24 4 X38 5 W25 5 X39-X49 4 W26 4 XSO-XSl 5 W27 5 X52-X53 4 W28-W34 4 X54-XSS 5 W35 5 X56-X58 4 W36-W38 4 X59-X60 5 W39 5 X61-X62 4 W40 4 X63 5 W41-W43 5 X64-X65 4 W44-W45 4 X66 5 W46 5 X67 4 W47 4 X68-X69 5 W48-W49 5 X70-X73 4 wso 4 X74 5 WSl 5 X75 4 W52-WSS 4 X76 5 W56-W58 5 X77 4 W59-W60 4 X78 5 W61 5 X79 4 W62 4 X80-X93 s W63-W67 5 X94-X95 4 W68 4 X96 5 W69-W71 5 W72 4 FRss* 4 W73-W83 5 W84 4 W85-W93 5 B-4 Table B.2: F u e l Assembl y Reactivity Categ o rizat io n for Fuel Assembl y IDs A through AA fo r Unit 3 Indian Point Unit 3 Fue l Assembly ID I Category I Assembly ID I Category I Assembly ID I Category V43 I 3 I V48 I 3 I I All Other Indian Po i nt 3 fuel (Batches A through AA) are Category 4 N ET-28 091-000 3-01, R ev i s i o n 0 B-5 ENCLOSURE 3 TO NL-17-144 Indian Point Unit 2 NEI 12-16 Draft Revision 2c Checklist Entergy Nuclear Operations , Inc. Indian Point Unit 2 Docket No. 50-247 Criticality Analysis Checklist

-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation

1.0 Introduction

and Overview Purpose of submittal YES Section 1. Remove credit for BoraflexŽ Changes requested YES Section 1 and 10 Summary of physical changes YES Section 1. BoraflexŽ loss Summary of Tech Spec changes YES Section 10 Summary of analytical scope YES Section 1.2 2.0 Acceptance Criteria and Regulatory Guidance Summary of requirements and guidance YES Requirements documents referenced YES Section 1.3 Guidance documents referenced YES Section 1.2 and 10.1 Acceptance criteria described YES Section 1.3 3.0 Reactor and Fuel Design Description Describe reactor operating parameters YES Section 3.4 Describe all fuel in pool YES Section 3.2 Geometric dimensions (Nominal and YES Section 3.2 Tolerance)

Schematic of guide tube patterns YES Figure 6.1 Material compositions YES Section 3 Describe future fuel to be covered YES Section 3.2 Geometric dimensions (nomina l and YES Section 3.2 to l erance) Schematic of guide tube patterns YES Section 3.2 Material compositions YES Figu re 6.1 Describe all fuel inserts YES Section 3.3. Geometric dimens io ns (nominal and YES Section 3.3 Tolerances are not used tolerance) since Depletion An alysis uses nominal dimensions. Schematic (axia l/cross section) NO Standard Westinghouse Designs Material compositions YES Sect ion 3.3 Describe non-standard fuel YES Section 8.10-8.12 Geometric dimens io ns YES Section 8.11 Describe non-fuel items in fuel cells NO Limits given in Section 8.13 Nominal and tolerance dimensions NO Limit s given in Section 8.13 4.0 Spent Fuel Pool/Storage Rack Description New fuel vault and Storage rack description N/A No change i n current license needed Nominal and tolerance dimensions N/A Schematic (axia l/cro ss section) N/A Material compositions N/A Spent fuel pool, Storage rack description YES Section 3.1 Nominal and tolerance dimensions YES Section 3.1, Tab l e 3.1 Page 1 of 7 Criticality Analysis Checklist

-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes/ Explanation Schematic (axial/cross section) YES/NO Cross sections in Figures 3.2 and 3.3. No axial details given since there is no variation in the over relevant axial heights Material compositions YES Section 3.1 (SCALE elemental brea k down of SS 304 is u sed but not specified)

Other Reactivity Contro l Devices {Inserts)

N/A There are no rack inserts. Control rods inserted in the fuel assemb li es are credited and covered with the fuel inserts. Nom inal and tolerance dimensions N/A Schematic (axial/cross section) N/A Material compositions N/A 5.0 Overview of the Method of Analysis New fuel rack analys is description N/A Storage geometries N/A Bounding assemb ly design(s)

N/A Integral absorber cred it N/A Accident analysis N/A Spent fuel storage rack analy sis description YES Section 2.0 Storage geometries YES Figure 1.1 and Section 8.5 Bounding assembly design(s) YES Batch Groupings are used. Introduced in Section 5. Fuel designs given in Section 3.2 So lub le boron credit YES Soluble boron credit is taken by inference with the criteria se l ected in Section 1.3. Boron dilution analysis YES Section 9.6 which references the current approved analysis. Burnup credit YES Section 2.0 Decay/cooling time credit YES Section 10.2 Integral absorber cred it YES Section 10.2 Other credit YES Figure 1.1 shows the credited control rods in specific locat ions Fixed neutron absorbers N/A Not taking credit for any fixed neutron absorbers Aging management program N/A Accident analysis YES Section 9 Temperature increase YES Section 9.4 Assembly drop YES Section 9.3 Single Assembly misload YES Section 9.3 Multiple misload YES Section 9.5 Page 2 of 7 Criticality Analysis Checklist

-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Boron dilution YES Sect i on 9.6 which references the current approved analysis. Other YES Sect i on 9.2 (Misplaced Assembly), Sect i on 9.7 (Seism i c) Fuel out of rack analysis (Normal Operations)

Yes Section 9.1 Handling YES Section 9.1 Movement YES Section 9.1 Inspection YES Sect i on 9.1 6.0 Computer Codes, Cross Sections, and Validation Overview Code/Modules Used for Calculation of kett YES Sect i on 2.1 Cross section library YES Section 2.1 Descript i on of nuclides used YES Table 2.1 Convergence checks YES Section 6.5 and Section 6.6.2 Code/Module Used for Depletion Calculation YES Section 2.1 and Section 5.6 Cross sect i on library YES Section 2.1 and Section 5.6 Description of nuclides used YES Section 2.1 Convergence checks YES Section 5.6 Validation of Code and Library YES Sect i on 4 and Appendix A Major Actinides and Structural Materials YES Sections 4.1 and 4.2 Minor Actinides and F i ss i on Products YES 1.5% bias (NUREG/CR-7109) -Section 4 Absorbers Credited YES Sections 4.1 and A.2.3 7.0 Criticality Safety Analysis of the New Fuel Rack Not part of License Application Rack model N/A Boundary conditions N/A Source distribution N/A Geometry restrictions N/A Limiting fuel design N/A Fuel dens i ty N/A Burnable Poisons N/A Fuel dimens i ons N/A Axial blankets N/A Limiting rack model N/A Storage vault dimensions and materials N/A Temperature N/A Multiple regions/configurations N/A Flooded N/A Low density moderator N/A Eccentric fuel placement N/A Tolerances N/A Fuel geometry N/A Fuel pin pitch N/A Page 3 of 7 Criticality Analysis Checklist

-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Fuel pellet OD N/A Fuel clad OD N/A Fuel content N/A Enrichment N/A Density N/A Integral Absorber N/A Rack geometry N/A Rack pitch N/A Cell wall thickness N/A Storage vault dimensions/materials N/A Code uncertainty N/A Biases N/A Temperature N/A Code bias N/A Moderator Conditions N/A Fully flooded and optimum density N/A 8.0 Depletion Analysis for Spent Fuel Section 5 Depletion Model Considerations Time step verification YES Section 5.6 Convergence verification YES Section 5.6 Simplifications YES Section 5.6 Non-uniform enrichments YES Axial Blankets, Sections 5.6 and 6.2 Post depletion nucl i de ad j ustments YES Section 5.9 Cooling time YES Sect i on 5.9 Depletion Parameters YES Sections 5.1-5.5 Burnable Absorbers YES Section 5.2 Integral absorbers Yes Section 5.2 Soluble Boron YES Section 5.3 Fuel and Moderator Temperature YES Section 5.1 Specific power YES Section 5.4 Control rod insertion YES Section 5.5 Atypical Cycle Operating H i story YES Section 5.3 utilized full details of cycle lengths (some short cycles) for determining average ppm. Section 5.1 utilized most limit i ng temperatures in cycles where power changed (IP2 cycle 10 and IP3 cycle 12) 9.0 Critica l ity Safety Analysis of Spent Fuel Pool Storage Racks Rack model YES Boundary conditions YES Section 6.1 Source distribution Yes Section 6.5, Section 6.6.2 Geometry restrictions YES Sect i on 10.3 Page 4 of 7 Criticality Analysis Checklist

-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Design Basis Fuel Description NO Mul tiple batch groupings as well as assemb l y specific analyses are utilized.

The secti ons given be low for this to pic are where the batch va lu es are identified. Fue l dens ity YES Sect i on 3.2 Burnable Poisons YES Section 5.2 Fuel assembly inserts YES Section 5.2 Fuel dimensions YES Section 3.2 Axial blankets YES Section 3.2 C onfi gurations conside red Bo r ated YES Section 8.14 Un borated YES All but Section 8.14 Multiple rack designs YES Sections 8.2, 8.3 , and 8.4 Alternate storage geometry YES Section 8.5 Reactivity Control Devices Fuel Assembly Insert s YES Control Rod Credit , Sections 6.6 and 8.3 Storage Cell Inserts N/A No used. Storage Cell Blocking Devices YES Sect ion 8.7 Axial burnup shapes Uniform/Distributed YES Section 6.2.1 Nodalizat i on YES Section 6.2. Blankets mode l ed YES Section 6.2 Tolerances/Uncertainties Fuel geometry Fuel rod pin pitch YES Section 7.1 Fuel pellet OD YES Section 7.1 Cladding OD YES Section 7.1 Axial fuel pos ition NO Insig nificant reactiv i ty since the rack has no axial variation (Applies to racks crediting absorbers that are not full l ength which doe s not apply to Indian Point.) Fuel content En rich ment YES Section 7.1 Dens it y YES Sect ion 7.1 Assembly inser t dimens ions and mater ials NO Deplet i on uses nominal dimensions , Control rod densities are reduced a bounding 20% (Section 6.6) Rack geometry Flux trap size (width) NO Reduction of the cell pitch reduced the flux trap width Rack cel l pitch YES Section 7.1 Rack wall thickness YES Section 7.1 Page 5 of 7 \J Criticality Analysis Checklist

-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes/ Explanation Neutron absorber dimensions N/A No credit taken for rack absorbers Rack insert dimensions and materials N/A No rack inserts used Code val i dation uncerta i nty YES Section 7.5 Criticality case uncerta i nty (statistical)

YES Section 7.5 Depletion uncertainty YES Sect i on 7.2 Burnup uncertainty YES Section 7.2 Biases Design basis fuel design NO Used most limiting fuel in each batch grouping Fuel geometry Clad creep YES Section 7.2 Grid growth (pin pitch) YES Sect i on 7.2 Minimum grid volume NO Conservatively ignored grids. Minor actinides and fission product worth YES Section 7.2 Code bias YES Section 7.5 Temperature NO Analysis was performed at most limiting temperature. See Section 8.1 for determining most limiting temperature. Eccentric fuel placement YES Section 7.4. Include in full pool analysis rather than a bias. See Section 8.4.3. lncore thimble depletion effect NO Included in the analy s is rather than a bias. See Section 5.6. NRC administrative margin NO Rather than specify a bias for the NRC administrative margin , the k 95;59 is calculated showing at least 1% margin. Calculated k 95;59 in Sections 8.3.2 and 8.4.3 Modeling simplifications Identified and described YES Section 6.2 10.0 Interface Analysis Interface configurations analyzed Between dissimilar racks Section 6.6 Between storage configurations within a rack Section 6.6 Interface restrictions NO Categorization of cell rather than an interface restriction. 11.0 Normal Conditions Fuel handling equipment NO Fuel handling equipment can only handle one assembly at a time and therefo r e do not pose a criticality concern. Fuel handling operations are in Section 9.1. Administrative controls YES Section 9.1 Fuel inspection equipment or processes YES Section 9.1 Page 6 of 7 Criticality Analysis Checklist

-Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Fuel reconstitution YES Section 9.1 12.0 Accident Analysis Boron dilution YES Sect i on 9.6 which references the cur r ent approved analysis. Norm al conditions YES Sect i on 9.1. Accident conditions YES Section 9.6 which references the current approved analysis. Single assembly mislead YES Sect i on 9.3 Fuel assembly misplacement YES Sect i on 9.2 Neutron absorber insert mislead YES Sect i on 9.5 addresse s withdrawa l of required control rods. Multiple fuel mislead YES Section 9.5 Dropped assembly YES Sect i on 9.3 Temperature YES Section 9.4 Seismic event or other natural phenomena YES Section 9.7 13.0 Analysis Results and Conclusions Summary of results YES Sect i onlO Burnup curve(s) YES Section 10.2 Intermediate decay time treatment YES Section 10.2 New administrative co ntrols YES Section 9.1 Techn i cal Spec i fication markup covered Technical Specification markups YES in an a ttachment separate from the CSA report. 14.0 References Appendix A Computer Code Validation

Code validation methodology and bases YES Appendix A New Fuel YES Section A.2 Dep l eted Fuel YES Section A.3 MOX crit i cal YES Section A.3.2 HTC critical YES Section A.3.1 H i gh temperature crit i cals YES Section A.4 Convergence NO Convergence of the Cr i tical Exper i ments is cove r ed by the same discussion of convergence for all the analysis. See Section 6.5 Trends YES Sect i on A.2.5 Bias and uncertainty YES Section A.2.6 Range of applicability YES Section A.2.8 Analysis of area of appl i cability coverage YES Section A.2.8 Page 7 of 7
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