ML081050375

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Report No. NET-290-01, Rev. 1, Evaluation of the Nine Mile Point 1 Boraflex Spent Fuel Racks with 7x7, 8x8 and 9x9 Fuel Assemblies Taking No Credit for Boraflex for Reactivity Control
ML081050375
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 04/03/2008
From:
Constellation Energy Group, Northeast Technology Corp
To:
Constellation Nuclear, Nine Mile Point, Office of Nuclear Reactor Regulation
References
7706697 NET-290-01, Rev. 1
Download: ML081050375 (56)
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ATTACHMENT 2 REPORT NO. NET-290-01, REVISION 1 EVALUATION OF THE NINE MILE POINT 1 BORAFLEX SPENT FUEL RACKS WITH 7X7, 8X8 AND 9X9 FUEL ASSEMBLIES TAKING NO CREDIT FOR BORAFLEX FOR REACTIVITY CONTROL Nine Mile Point Nuclear Station, LLC' April 3, 2008

Report No. NET-290-01 Evaluation of the Nine Mile Point 1 Boraflex Spent Fuel Racks with 7x7, 8x8 and 9x9 Fuel Assemblies Taking No Credit for Boraflex for Reactivity Control October 2007 Prepared for Constellation Nuclear Corporation, LLC Prepared by:

Northeast Technology Corp.

108 North Front Street, 3 rd Floor UPO Box 4178 Kingston, New York 12401 Under Purchase Order: 7706697

v. Pepa Reviewed by: ApprovedIQA):

/ 1' t .,

NET-290-01 Table of Contents 1.0 Intro d u c tio n ........................................................................................................... 1 1.1 Fuel and Fuel Rack Design Description ................................................ 2 1.2 Design Basis and Design Criteria ......................................................... 6 2.0 Analytical Methods and Assumptions .............................................................. 8 3.0 Results of the Criticality Analyses ................................................................... 13 3.1 CASMO-4 and KENO V.a Reactivity Calculation Comparison ............. 13 3.2 Reactivity Calculations ................................. ...................................... 14 3.2.1 CASMO-4 Depletion Calculations ............................................. 14 3.2.2 Reference KENO V.a Model .................................................... 15 3.3 Effect of Tolerances and Uncertainties ................................................ 19 3.3.1 Tolerances and Calculational Uncertainties ............................. 19 3.3.2 Uncertainty Introduced by Depletion Calculations .................... 20 3.4 Summary of Reactivity Calculations ..................................................... 21 3.4.1 Reference Loading ................................................................... 21 3.5 Abnormal/Accident Conditions ...................... :..................................... 24 4 .0 C o nclusio ns ................................................................................................ . . 28 5 .0 R e fe re n ce s ................................................................................................... . . 29 Appendix: Benchmarking Computer Codes for Calculating the Reactivity State of Spent Fuel Storage Racks, Storage Casks and Transportation Casks.

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NET-290-01 List of Figures Figure 1: The Spent Fuel Storage Pool at the Nine Mile Point 1 Station .................... 3 Figure 2: A 4x4 Array of Fuel Storage Cells (shown with Boraflex)

Filled w ith 9X9 Fuel Assem blies ..................................................... 4 Figure 3: Depletion Characteristics of the Advanced Fuel Types ............................... 9 Figure 4: KENO V.a Model of NMP1 Racks with 9x9 Fuel ....................................... 12 Figure 5: U-235 Enrichment and Gadolinia Distribution of 8x8 Assemblies as Modeled in the NMP1 "South" Boraflex Module ................ 17 Figure 6: Reference Case Keno V.a Generated Plot of the NMP1 Boraflex Modules Loaded with 8x8 and 9x9 Fuel Assemblies and BORAL Modules Loaded with 10x1O Fuel Assemblies ............................. 18 Figure 7: Keno V.a Generated Plot of a Dropped Assembly Resting on Top of "North" Boraflex Module ........................................................................... 25 Figure 8: Keno V.a Generated Plot of a Dropped Assembly Alongside of the "North" and "South" Boraflex Modules ................................................................... 26 Figure 9: Keno V.a Generated Plot of a Reload Assembly (1Ox1 0)

Misloaded Adjacent to the "North" BORAL Modules .................... 27 iii

NET-290-01 List of Tables Table 1: Fuel Assembly Descriptions Nine Mile Point 1 Nuclear Power Station ...... 5 Table 2: CASMO-4/KENO V.a Reactivity Comparison in Standard Cold-Core Geometry at Zero Burnup ......... ......................... 13 Table 3: Reactivity Equivalent Fresh Fuel Enrichments and Limiting La ttice k. ........................................................................................... . . 14 Table 4: Summary of Criticality Calculation Results for the NMP1 Boraflex Spent Fuel; Racks with North Module Containing Peak Reactivity 9x9 at a REFFE of 2.35 w/o ................................................................. 22 Table 5: Minimum Gadolinia Loading as a Function of Initial Peak Planar E nrichm ent for 9x9 F ue l........................................................................... 23 iv

NET-290-01

1.0 INTRODUCTION

The Nine Mile Point Unit No. 1 (NMPI) spent fuel pool contains two types of spent fuel storage racks. One type, the majority of racks, utilizes the neutron absorber material BORAL for reactivity control; the other type utilizes Boraflex (only two modules). The Boraflex racks were originally licensed for unirradiated fuel assemblies with a peak lattice enrichment of 3.75 w/o U-235 111. The Boraflex racks were subsequently analyzed for un-irradiated fuel assemblies with initial enrichments up to 4.65 w/o U-235 with a minimum of 7 Gd 2 0 3 rods at 4.0 w/o Gd 2 03[2]. The BORAL racks were installed in late 2004 replacing all but two of the Boraflex modules with new BORAL rack modules [3, These two Boraflex modules are located in the southwest corner of the NMP1 spent fuel pool. There is unrestricted access to all 198 storage cells in the "North" module. The "South" Boraflex module contains 216 storage cells with cell access restricted by a tooling table. The tooling table is supported by four pedestals seated in four empty cells within the module[21 . This table precludes access to a significant portion of the storage cells beneath it.

This report documents criticality analyses of the two remaining Boraflex modules based on:

1) The actual inventory of assemblies loaded in the South module that are inaccessible due to the tooling table and; 2) loading of the North module with any 7x7 or 8x8 assembly at peak reactivity or any 9x9 assembly with a specified combination of maximum enrichment and minimum number of gadolinia rods. The analyses are based on the assumption that the adjacent BORAL racks are filled with maximum reactivity 10x1 0 fuel assemblies with a peak lattice enrichment of 4.6 or less. This corresponds to a 10x1 0 bundle with a k. _<1.31 in standard cold core geometry (SCCG)[5]. The analysis provides maximum flexibility with respect to future fuel storage utilization of the Boraflex modules and possible removal of fuel assemblies for dry cask storage.

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NET-290-01 From the analyses the maximum allowable enrichment and minimum required gadolinia rod combination such that keff *- 0.95 are determined. The maximum calculated keff includes fuel and rack allowances for as-built tolerances, model bias and calculational uncertainties, which when statistically combined, ensure that the true keff - 0.95 at a 95% probability and at a 95% confidence level.

1.1 Fuel and Fuel Rack Design Description The remaining two Boraflex spent fuel rack modules include a "North" module consisting of an 11 xl 8 array of cells and a "South" module consisting of a .12x1 8 array of storage cells located in the southwest corner of the NMP1 spent fuel pool as shown in Figure 1[3,4]. The individual storage cells utilize Boraflex in a flux trap configuration as shown in Figure 2.

The rack structural components are made from 304L stainless steel. The storage cells are asymmetric with two sheets of Boraflex forming a flux trap between assemblies in the E-W direction. In the N-S direction, the fuel assemblies are separated by the stainless steel rack structure.

Nominally, the Boraflex sheets are 134 inches long, beginning 10.79 inches above the base plate of the rack module and extending to 144.79" above the base plate. The active fuel region extends from elevation 7.22 inches to elevation 152.46 inches for all fuel assemblies. For these assemblies, the top and bottom six inches of the active fuel length are natural uranium. For the current analysis, Boraflex sheets were replaced with water.

All fuel is 145.24 inches long.

The fuel design parameters for the 7x7, 8x8 and 9x9 fuel types are shown in Tablel.

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NET-290-01 TOTAL SPENT FUEL POOL STORAGE = 3910 CELLS NORTH BORAL STORAGE = 3496 CELLS BORAFLEX STORAGE = 414 CELLS Figure 1: The Spent Fuel Storage Pool at the Nine Mile Point 1 Station 3

NET-290-01 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000

-0000 00ooo0000 0ooo 0000 0-0o00000 ooo 00000--OOO 0000-/0000 0000-o0000 0000o/0000 00000000 000000000 000000000 00000000 1000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000O 000000000 000000000 000 000000 00000P0000 000000000 000000000 000000000 000000000 00000000 000000000 00000000 000000000 00000000 000000000 000000000 000000 0000000O00 000000000 00000000 000000000 000000000 000000000 000000000 000000000 00000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 00000000 0000-0000 00000000 0000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 000000000 00000000 000000000 00000000 00000000 000000000 000000000 000000000 000000000 00000000 000000000 00000000 000000000 000000000 000000000 ggg -0gg U000000000

-00000000

00------00000 000000000 00000000 000000000 000000000 000000000 000000000 000000000 000000000

  • 00000 000000000 00 00)0000 000000000 0 O0000 0 O-O 000000000 0000O0OO 00000000 000O\---000 000000000 0 O000 0o0001-1000 O000,A 000000000 000ý00 000000J0 0000J0000 O000 O000O00 0 00000000o0 00000o00o0 0oooo0000 000000000000 Flux Trap Boraflex Fuel Stainless Bundle Steel Figure 2: A 4x4 Array of Fuel Storage Cells (shown with Boraflex)

Filled with 9X9 Fuel Assemblies 4

NET-290-01 Table 1 Fuel Assembly Description Nine Mile Point 1 Nuclear Power Station 8x8 8x8 9x9 lOxlO FUEL RODS (GE7) (GE8x8/R)

Cladding Material Zircaloy Zircaloy Zircaloy Zircaloy Zircaloy Cladding Tube OD, in. 0.563 0.483 0.483 0.440 0.404 Cladding Tube Wall Thickness, in. 0.032 0.032 0.032 0.028 0.026 Pellet Material Sintered Sintered Sintered Sintered Sintered U0 2 U0 2 U0 2 U0 2 U0 2 Pellet OD, in. 0.487 0.410 0.410 0.376 0.345 Pellet Density, gm/cm 3 (% theoretical) 10.412 10.412 10.5764 10.631 10.631 (95%) (95%) (96.5%) (97%) (97%)

Pellet-to-Clad Diametral Gap, in. 0.012 0.009 0.009 0.008 0.007 FUEL ASSEMBLIES 1 Number of Rods (# of water rods) 49 (0) 62 (2) 60 (4) 74 (2 large) 92 (2 large)

Rod Array 7x7 8x8 8x8 9x9 10x10 Rod-to-Rod Pitch, in. 0.738. 0.640 0.640 0.566 0.510 Assembly Dimensions 5.166 x 5.26 x 5.26 x 5.094 x 510x 5.10 (without fuel channel), in. 5.166 5.26 5.26 5.094 Maximum Assembly Planer Average Enrichment, WT percent 235U, in 2.5 3.20 3.60 4.60 4.60 Boraflex Modules Axial Fuel Loading (gms U-235/cm- 13.511 17.15 17.15 22.85 23.92 assembly) 5

NET-290-01 1.2 Design Basis and Design Criteria The analyses and evaluations described in this report demonstrate for the NMP1 Boraflex spent fuel racks keff<* 0.95 when completely loaded with the most reactive limiting fuel type under the most reactive conditions. The maximum calculated reactivity (keff) when adjusted for code biases, fuel and rack manufacturing tolerances and methodology/calculational uncertainties (combined in a root-mean-square sense) will be less than or equal to 0.95 with a 95% probability at a 95% confidence level.

All analyses and evaluations have been conducted in accordance with the following codes, standards and regulations as applicable to spent fuel storage facilities:

" American Nuclear Society, American National Standard Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants, ANSI/ANS-57.2-1983. October 7, 1983.

o Nuclear Regulatory Commission, Letter to All Power Reactor Licensees from B. K.

Grimes. OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications. April 14, 1978, as amended by letter dated January 18, 1979.

o USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, New Fuel Storage, and Section 9.1.2, Spent Fuel Storage.

o USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2, December 1981.

General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling.

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NET-290-01 o ANS/ANSI 8.12-1987, Nuclear Criticality Control and Safety of Plutonium - Uranium Fuel Mixtures Outside Reactor.

It is noted that the above USNRC and ANS documents refer to the requirement that the maximum effective neutron multiplication factor (keff) be less than or equal to 0.95. The analyses of the reference case fuel/rack configurations are based on an infinite repeating array in lateral extent and finite in the z-direction.

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NET-290-01 2.0 ANALYTICAL METHODS AND ASSUMPTIONS The reactivity state of the NMP1 spent fuel racks has been analyzed using KENO V.a from the SCALE-PC[61 package and CASMO-4[7 1. These computer codes have been validated and verified for spent fuel rack evaluations by benchmarking calculations of LWR critical experiments as described in the Appendix to this report. The computer codes (or their predecessors) have been previously reviewed and approved by the USNRC for spent fuel rack criticality evaluations[B].

To identify a most reactive fuel type that can be stored in the "North" Boraflex rack module the following approach was adopted. The current fuel loading configuration in the "South" Boraflex rack module has been assumed and the most reactive 9x9 bundle that can be stored in the "North" Boraflex module has been determined. The most reactive fuel lattice is defined for each fuel type, including the maximum planar average enrichment (w/o U-235) and minimum number of Gd 20 3 rods, each rod containing the minimum w/o Gd 2 03 loading. The depletion characteristics for this fuel assembly (k. versus burnup) both for the standard cold-core geometry (SCCG) and for fuel rack geometry were assessed with CASMO-4 to determine the burnup resulting in peak assembly reactivity (k-). In these calculations, the fuel assembly is depleted at hot full power conditions in core geometry using CASMO-4. At specified burnup levels, the assembly is brought to the cold zero power condition (no Xenon) and modeled in the rack geometry. Subsequently, the assembly is subjected to additional burnup in the hot full power condition in core geometry and the iterative process repeated. The depletion characteristics of a fuel assembly with gadolinia are shown in Figure 3 as well as the depletion characteristics of an assembly without gadolinia burnable poisons.

The base-case reference value of keff of the fuel and rack configuration has been determined with KENO V.a. The effect of depletion on storage rack reactivity has been determined using CASMO-4. The KENO V.a model of the NMP1 fuel and storage rack is an exact rendering of the fuel and rack geometry as shown in Figure 4. Due to asymmetries in the NMP1 rack, the CASMO-4 model contains some approximations. For this reason, the CASMO-4 results are applied on a relative and not absolute basis (relative to the exact KENO V.a model).

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NET-290-01 MARGIN TO THE 0.95 DESIGN LIMIT 0.95 ~MAXIMUM RACK REACTIVITY (95195)

ALLOWANCE FOR MODEL UNCERTAINTIES, BIASES, FUEL AND RACK TOLERANCES DESIGN AND UNCERTAINTIES POINT *}ALLOWANCE FOR DEPLETION DEPENDENT UNCERTAINTIES cc MAXIMUM ENRICHMENT BUNDLE, NO Gd2O3 MAXIMUM ENRICHMENT BUNDLE --

WITH MINIMUM NUMBER OF Gd20 3 RODS AT MINIMUM Gd2O3 LOADING BURNUP Figure 3: Depletion Characteristics of the Advanced Fuel Types 9

NET-290-01 To assure that the actual fuel/rack reactivity is always less than the calculated maximum reactivity, the following conservative assumptions have been applied to the analyses:

1. The fuel assembly design parameters for these analyses are based on the most reactive lattice for a given fuel type.
2. The maximum fuel enrichment is uniform throughout the assembly. The assumption of uniform enrichment results in a higher reactivity than the distributed enrichments in the actual assemblies[81.
3. The fuel assembly is channeled in the rack as this condition results in the highest reactivity.
4. The moderator is assumed to be demineralized water at full water density (1.0 gm/cm 3).
5. All available storage locations are loaded with assemblies of maximum reactivity. This is conservative since four locations in the "South" module contain the tooling table feet and cannot be loaded with fuel.
6. No credit is taken for neutron absorption in the fuel assembly grid spacers or upper and lower end fittings.
7. No credit is taken for any natural uranium or reduced enrichment axial blankets (fuel is assumed to be at maximum average planar enrichment).
8. The number of gadolinia rods is taken as the minimum number contained in any region of the fuel assembly (vanished regions typically contain one less gadolinia rod than dominant regions).
9. Gadolinia loading (w/o Gd 20 3) is assumed to be the minimum loading for assemblies with split gadolinia loadings.
10. BORAL racks contain 10x1O fuel at the reactivity equivalent fresh fuel enrichment (REFFE) that yield k,- = 1.31 in the standard cold core geometry (SCCG). The BORAL boron loading is at the minimum certified areal density 2

of 0.0150 gms b-1 0/cm . [6]

11. All fuel is assumed to have an active length of 145.2 inches.

10

NET-290-01 Based on the analyses described subsequently the maximum k- at a 95% probability with a 95% confidence level of the fuel/rack configuration is calculated as:

k. = kref +Akbia, + Ak2 where kref = Nominal KENO V.a k.- adjusted for depletion effects Akbias = Model bias Tolerances and Uncertainties:

Ak1 = U0 2 enrichment tolerance Ak2 = U0 2 pellet density tolerance Ak3 = Gd 20 3 loading tolerance Ak4 = Rack cell inner width tolerance Ak5 = Rack cell wall thickness tolerance Ak6 = Flux trap width tolerance Ak7 = Pellet diameter tolerance Ak8 = Cladding inside diameter tolerance Ak9 = Cladding outside diameter tolerance Akio = Cladding wall thickness tolerance Ak11 = Asymmetric assembly position tolerance Ak12 = Methodology bias uncertainty (95 x 95)

Ak13 = Calculational uncertainty (95 x 95) 11

NET-290-01 HALF INTERNAL DIVIDER BORA FLEX FUEL ASSEMBLY

/

.0000000001\ N

ý000000000O OOOO0 000OO STAINLESS N

000000000' STEEL CLAD 000000000 000000000 HALF POISON INSERT LHALF POISON BOX WALL FUEL BOX WALL Figure 4: KENO V.a Model of NMP1 Racks with 9x9 Fuel (All boundaries of the cell assume spectral reflection of neutrons) 12

NET-290-01 3.0 RESULTS OF THE CRITICALITY ANALYSES 3.1 CASMO-4 and KENO V.a Reactivity Calculation Comparison As a check of the two independent methods used for these analyses, the reactivity of each fuel type in the standard cold-core geometry (SCCG) at cold temperature conditions (68 0F) has been calculated both with KENO V.a and with CASMO-4 at zero burnup. Both models are exact renderings of the assemblies in core geometry. Table 2 contains the k. for each fuel lattice assembly with Gd 2 0 3 rods. The reported k- values include model biases which have been determined via benchmark calculations. These model biases are -0.0078 and -

0.0103 for KENO V.a and CASMO-4, respectively.

Table 2 CASMO-4 / KENO V.a Reactivity Comparison in Standard Cold-Core Geometry at Zero Burnup Gd 2O3 Enrichment k-(rods, w/o) (w/o U-235) KENO V.a CASMO-4 2@1.0,1 2 @105 2.50 1.17469 +/- 0.0004 2 @0.5 1.17560 8x8 (2 water rods) 7 @4.0 3.20 1.13409 +/- 0.0004 1.13405 8x8 (4 water rods) 8 @4.0 3.42 1.12932 +/- 0.0004 1.12875 9x9 9@4.0 4.00 1.15734+/-0.0004 1.15720 9x9 12@5.0 4.21 1.10254+/-0.0004 1.10120 9x9 14 @ 5.0 4.60 1.07947 +/- 0.0004 1.07951 The maximum difference between the Keno V.a and CASMO-4 eigen values is less than 0.0015 Ak.

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NET-290-01 3.2 Reactivity Calculations 3.2.1 CASMO-4 Depletion Calculations CASMO-4 was utilized to compute the reactivity of a limiting reactivity fuel lattice as a function of burnup for each fuel type. The limiting reactivity lattice with respect to planar average enrichment, number of gadolinia rods and gadolinia loading was determined from fuel assembly design reports 9l'1]. Sensitivity analyses demonstrated that average power density, average fuel temperature, saturation temperature of the moderator, and zero void results in the most reactive condition (peak bundle reactivity). Accordingly, the depletion calculations were conducted under these average conditions with 0% void.

For subsequent analyses of the Boraflex modules in finite three-dimensional models, the reactivity equivalent fresh fuel enrichment (REFFE) was utilized. The REFFE was determined by modeling each fuel type in cold core geometry with KENO V.a. This process is iterative in that the fuel enrichment is varied until the k. matches the k-0 at peak bundle reactivity as determined with CASMO-4. Table 3 contains the k- values for each lattice type for CASMO-4 and Keno V.a as well as the k. of the REFFE.

Table 3 Reactivity Equivalent Fresh Fuel Enrichments and Limiting Lattice k-No. of k.. I k.

Array w/o U235 No fGd203 kk.REFFE Gadolinia Rods (CASMO) (KENO V.a) 7x7 2.5 2,2 1.0, 0.5 1.2479 1.2485 2.06 8x8 3.2 7 4.0 1.2164 1.2169 1.95 8x8 3.42 8 4.0 1.2258 1.2255 2.02 I 8x8 3.6 9 4.0 1.2368 1.2379 2.10 9x9 3.41 7 4.0 1.2349 1.2363 2.13 9x9 4.0 9 4.0 1.2676 1.2687 2.35 9x9 4.21 12 5.0 1.2441 1.2478 2.2 1 9x9 4.6 14 5.0 1.2568 1.2588 2.27 14

NET-290-01 3.2.2 Reference Keno V.a Model A reference Keno V.a model was created based upon the actual fuel assemblies loaded in the "South" Boraflex module under the tooling table. All assemblies in the "South" module are 8x8 lattices and are conservatively assumed to be at a REFFE corresponding to a burnup of peak assembly reactivity. To simplify modeling of the South module, several assemblies were conservatively modeled at higher enrichments. With the following exceptions, all "South" module cells contain 8x8 fuel at 3.2 peak planar enrichment:

  • Cells 2A55 thru 2A58 and 2B72 thru 2C72 actually contain 8x8 assemblies at 2.82 w/o U-235 peak planar enrichment, however, these cells are conservatively modeled as 8x8 assemblies at 3.2 w/o peak planar enrichment.
  • Cells 2656, 2L56, 2671 and 2L71 are empty cells containing tooling table support legs, nevertheless are conservatively modeled as 8x8 fuel assemblies at 3.2 w/o.
  • Cell 2A71 and 2M72 contain 8x8 assemblies at 3.2 w/o; these cells are conservatively modeled as 3.42 w/o.enrichment assemblies.

" Cells 2D71 thru 2F71 contain 8x8 fuel assemblies at 3.2 w/o; these cells are conservatively modeled as 3.42 w/o fuel assemblies.

" Cells 2G72 thru 2K72 contain 8x8 fuel assemblies at 2.82 w/o. These cells are conservatively modeled as 8x8 fuel at 3.2 w/o enrichment.

  • Cells 2M59 thru 2M71 and 2K55 thru 2L55 contain 8x8 fuel assemblies at 3.42 w/o as currently loaded in the "South" module.

Figure 5 shows the fuel initial enrichment, gadolinia loading, and limiting k. (SCCG) of assemblies modeled in the "South" module.

In the "North" module there are two non-fuel components residing in cells 2D37 and 2L53.

For conservatism, all cells are assumed to include 9x9 fuel at peak reactivity. The analysis for the "North" module accounted for possible future reload enrichments up to 4.60 w/o U-235 with a minimum number of gadolinia rods at the minimum loading based upon reload assembly design reports[ 91' °]. In modeling the 9x9 assemblies in the "North" module, several conservatisms were included in the model. These include:

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NET-290-01

  • The number of gadolinia rods was taken at the minimum number in any zone (e.g.

vanished zones typically had one less gadolinia rod than did the dominant zones).

" For assemblies with split gadolinia loadings, the minimum loading was used.

The neighboring fuel racks to the East and North of the Boraflex modules contain BORAL as the neutron absorber material. Additional arrays of BORAL modules containing 10x10 fuel assemblies were added to the Boraflex modules to create a full pool model as shown in Figure 6. Although 10x1 0 assemblies are not currently loaded in the BORAL modules, the future use of this array type is possible. However, 10x1 0 assemblies may NOT be stored in the Boraflex modules. The following conservative assumptions were used to model the additional BORAL modules:

" As previous analyses had shown the 10x1 0 fuel type is more reactive than the 9x9 fuel type[21, the 10x10 fuel assemblies were modeled in the BORAL modules.

  • All fuel is at a REFFE of 2.55 w/o U-235, corresponding to the tech specification limit
k. < 1.31 in SCCG. [5 1
  • The areal density of the BORAL absorber is assumed to be at the minimum certified 2

value of 0.015 gms B-10/cm . [11]

The reference case is a full fuel height model with water albedoes in the axial directions.

The South and West boundary conditions both incorporate a 24-inch concrete albedo and the North and East boundaries along the BORAL modules incorporate a water albedo boundary condition.

The reference case Keno V.a model was executed using 3050 neutron generations and 5,000 neutrons per generation for a total of 15 million neutron histories. The first fifty neutron generations were omitted to attain source convergence.

16

NET-290-01 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 3.42 3.2 7rods@4w/o 7rods§4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rod§@4w/o 7, w/o@4olo 7mds@4w/o 7rodso4w/o Brods@4w/o 8rodsQ4w/o 7rods@4w/o 65 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1,2164 1.2258 1.2258 1.2164 3.2 3.2 3.2 3.2 3.2 3.2 32 3.2 3.2 3.2 3.2 3.2 7rods@4w/o 7rods@4w/o 7rods@4w/o ?rods@4w/o 7rods@4w/o 7rods@4wlo 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4wlo 7rods@4w/o 7rods@4w/o 56 1,2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 7rods@4w/o 7rodso4w/o 7mods@4wlo 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 57 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 7rods@4w/o 77odos4w/4 7mds@4w/o 7rods@4wo 7ro7s@4owo 7ods@4w1 7rods@4wlo 7rods@4wlo 7rods@4wlo 7rods@4w/o 7rodsa4wlo 7rods@4wto 58 1.2164 1.2164 1,2164 1,2164 1.2164 1.2164 1.2164 1.2164 1,2164 1.2164 1.2164 1.2164 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 7 rods@4wlo 7rods@4w l0 7 rod s@4w/o 7,ods@4 wlo 7rodos@4wlo 7rod5@4w/o 7 ods@4w/o 7roms@4w/o 7rodls@4w lo 7rodsQ4w lo 7rodso4w/o 8rodso 4 wo 59:

1.2164 1.2164 1,2164 1,2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2258 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 7ods@4w/o 7rods@4w/o 7rods§4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4wlo 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 8rods4w/o 60 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2258 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 7rods@4woo 7 o o/o77ds@4 7rodw@40r/o 7o044/o 7,044/0 7rs44w/o 7rods@4w/o 7rods@4w/o 7rods44w/o 7rods@4w/o 7rod,@4w/o 8r0dsl4w/o 61 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1,2164 1.2164 1.2164 1.2164 1.22586 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 7rods,0 4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/a 7rods@4w/o 7rods@4w/o 7rods@4w/o 6rods/4wlo 62 1.2164 1.2164 1,2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2258 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 7rodsC4w/o 7mds@4w/o 7rods@4w/o 7rods@4w/o 7rods404wo 7rods 7 w4w/o 7,0 4 w/o 7rods@4w/o 7 rods(4w/o 7rods@4w/o w/o,74w/a 0rods/4w/o63 1.2164 1.2164 1.2164 1.2164 172164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2256 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 7rods@4w/o 7rods@4w/o 70ods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rodso4w4 o 7rods@4w/o 7rod0@4wo0 Brods/4w/o64 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2258 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 7rods@4w/o 7ods@w4w/o 7wo/o04w/0 7 7rodo04wdo 7rod@4w/ w 7rodsQ4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rod,@4w/o arodo/4w/o 65 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2258 -

3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 7rods@4w/o 7rods@4w/o 7 ods@4w/o 7rods@44w/o 7rds@4w/o 7rods@4w/0 7 044/0 7rodw74w/0 7 044/0 7 0s@4wlo 7rods@4w/0 8rod044w/o 66 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2258 3.2 3.2 3.2 32 3 3.2 3.2 3.2 3.2 3.2 3.2 3.42 7rods@4w/o 7rod3@4w/o 7rods@4w/o 7rods@4w/o 70ods@4w/o 7lrods@4wo, 7ros4wl/ 7ro4s.4wlo 7rods@4wlo 7rods@4w/o 7rods@4w/o0 rods/4w/o 67 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1,2164 1.2258  %

3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 7rods@4w/o 7rod0@4w/o 7rods@4w/o 7,04044 7,044/0 lrodwo04w/o 7r04o040// 7r04ss4w/0 7rods@4w/o 7rodo04w/o 7rodoos@4w/o rodsl4w/o1/

1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1ý2164 1.2258 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 70ods@4w0o 7rods04w/o 7rod0@4w/o 7rods@4w/o 7rods@4w/o 7rods044w/o 7rods@4w/o 7rods@4w/o 7rods04w/o 7rods@4w/o 7rods@4w/o 8md0/4w/o 6 1.2164 1.2164 1.2164 1,2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 1.2258 -

3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.2 3.42 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 86odsl4w/0 70 1.2164 1.2164 1.2164 1.2164 1.2164 1.2164 .1.2164 1.2164 1.2164 1.2164 1.2164 1.2258 3.42 3.2 3.2 3.42 3.42 3.42 3.2 3.2 3.2 3.2 3.2 3.42 8 rodoO4w /o 7rods@ 4w/o 7 0ods@ 4w /o 8rods/4w/o 8rods/4w/o 80ds4/4w, o 7rods044w/o 7r o 4 wlo 7r0ods4 w/o 7r ods@4w/ o 7rods@ 4w/o 8ro4 wlo 71 1.2258 1.2164 1.2164 1.2258 1.2258 1.2258 1.2164 1.2164 1.2164 1.2164 1.2164 1.2258 3.42 3.2 3.2 3.42 3.42 3.42 3.2 3.2 3.2 3.2 3.2 3.42 8rods/4w0o 7rods@4w/o 7rods@4w/o 8rods04w/o 8rods/4w/o 8rods/4w/o 70ds@4w/0o 7rods@4w/o 7rods@4w/o 7rods@4w/o 7rods@4w/o 8rods/4w/o 72 1.2258 1.2164 1.2164 1.2258 1.2258 1.2258 1.2164 1.2164 1.2164 12164 1.2164 1.2258 2A  :-2B , = 2C  ; 2E I -2G-2HW' I '-2Jý'U- '-2K . 1 2 91 2M-Figure 5: U-235 Enrichment and Gadolinia Distribution of 8x8 Assemblies as Modeled in the NMP1 "South" Boraflex Module.

Key: w/o u-235 Gad rods, loading limiting k.

17

NET-290-01 Figure 6: Reference Case Keno V.a Generated Plot of the NMP1 Boraflex Modules Loaded with 8x8 and 9x9 Fuel Assemblies and BORAL Modules Loaded with 10x1O Fuel Assemblies.

18

NET-290-01 3.3 Effect of Tolerances and Uncertainties 3.3.1 Tolerances and Calculational Uncertainties To evaluate the reactivity effects of fuel and rack manufacturing tolerances in the "North" and "South" modules, CASMO-4 and Keno V.a perturbation calculations were performed.

The limiting reactivity fuel assembly 9x9 at peak burnup was used. This is conservative with respect to the tolerance effects from the lower reactivity assemblies. Additionally, tolerances were determined based on an infinite array of storage cells. The following tolerance and uncertainty components are addressed:

U-235 Enrichment: The enrichment tolerance of +/- 0.088 w/o U-235 variation about the nominal reference value of 4.60 w/o U-235 was considered 1 2 ].

U0 2 Pellet Density: A variation of -2.0%/+1.0 (absolute) about the nominal reference theoretical density of 97%[21.

Pellet Dishing: The pellets were assumed to be undished. This is a conservative assumption in that it maximizes the U-235 loading per axial centimeter of the fuel stack. No sensitivity analyses were completed with respect to the variations in the pellet dishing factor.

Pellet Diameter: A conservative tolerance of +/- 0.002 inches was assumedI12.

Clad Inside Diameter: A conservative tolerance of +/- 0.002 inches was assumed[ 12].

Clad Outside Diameter: Clad OD is bounded by the combination of wall thickness tolerance and inner diameter tolerance. Each of these is addressed separately, thus no further analysis is required.

12 Clad Thickness: A bounding tolerance of +/- 0.004 inch[ 1 19

NET-290-01 gd2 O3 Loading: While the maximum reactivity assembly at 11.5 GWD/MTU does not contain Gd 20 3, the burnup at peak reactivity depends on the initial Gd 2 0 3 loading. The tolerance of +/-10% (relative) in the Gd 20 3 loading has been assumed[2 ].

Cell Inside Dimension: The manufacturing tolerance of +/- 0.030 inch for the variations in cell wall inside dimensions was used[21.

Stainless Steel Thickness: A stainless steel sheet tolerance of_+/- 0.004 inches was used,2].

Flux Trap Width: The manufacturing tolerance on the flux trap width of -0.038 inch was used[2 ].

Cell-to-Cell Pitch: Cell-to-cell pitch is determined by the cell wall thickness, cell inside dimensions and flux trap width in the NMP1 fuel rack. Each of these is addressed separately and no further allowances are required.

Assembly Location: The reference KENO V.a reactivity calculations are based on a model with each assembly symmetrically positioned in each storage cell. The effect of four adjacent assemblies with minimum separation distance has been considered.

Calculational Uncertainty: The 95% probability / 95% confidence level uncertainty associated with the reference KENO V.a calculation has been applied.

Methodology Uncertainty: The 95% probability I 95% confidence level uncertainty of 0.0078 as determined from benchmark calculations (see Appendix) has been considered.

3.3.2 Uncertainty Introduced by Depletion Calculations Critical experiment data are generally not available for spent fuel and, accordingly, some judgment must be used to assess those uncertainties introduced by the depletion calculations. CASMO-4 and the 70 group cross section library used for these analyses have been used extensively to generate assembly average cross sections for core follow calculations and reload fuel design in both BWRs and PWRs. Any significant error in those 20

NET-290-01 depletion calculations would be detectable either by incore instrumentation measurements of core power distribution or cycle energy output or both. Significant deviations between the predicted and actual fuel cycle lengths and core power distributions using CASMO-4 generated cross sections are not observed.

For the purpose of assessing the effects of uncertainties introduced by depletion calculations, it is useful to estimate the magnitude of depletion uncertainties in k- and compare this uncertainty with margins inherent in the present calculation. It is assumed that depletion calculations introduce an uncertainty in k. which is a linear function of burnup such that at a burnup of 40,000 MWD/MTU the Akunc due to depletion effects is 0.02. So that for the limiting reactivity assembly at 11.5 GWD/MTU, the uncertainty introduced by depletion is 0.00575 in Ak. Conservatisms in the reference model outlined in Section 3.2 contain adequate margin to account for any uncertainties due to depletion effects.

3.4 Summary of Reactivity Calculations 3.4.1 Reference Loading The explicit model developed in Section 3.2 was executed to determine the reactivity of the two remaining NMP1 Boraflex spent fuel racks while taking no credit for Boraflex.

For the reference case with the "South" module conservatively modeled as described in Section 3.2 and the "North" module loaded with 9x9 fuel at maximum reactivity, the calculated keff was 0.91816. Additional cases were run with various U0 2 enrichment/gadolinia combinations to confirm that the reference case keff was bounding.

Table 4 contains a summary of the criticality analyses results for the NMP1 spent fuel racks. Table 5 contains the required minimum gadolinia loading as a function of enrichment for the 9x9 fuel in the "North" module.

21

NET-290-01 Table 4 Summary of Criticality Calculation Results for the NMP1 Boraflex Spent Fuel Racks with North Module Containing Peak Reactivity 9x9 Fuel at a REFFE of 2.35 w/o (corresponding to 4.0 w/o, 9 Gadolinia Rods @ 4 w/o)

Nominal keff (includes model bias) 0.91816 Tolerances and Uncertainties:

Fuel Material

- U-235 Enrichment +0.00422

- U02 Density +0.00213

-Gd20 3 Loading +0.00869

-Pellet Diameter +0.00275

-Clad Diameter/thickness +0.00434 Rack Construction

- Flux Trap Width +0.00653

- Cell ID +0.00208

- Cell Wall Thickness +0.00000 Assembly Placement +0.00046 Methodology Bias (95195) +0.00939 Calculational Uncertainty (95195) +0.00036 Square Root of Sum of Squares: +0.01613 Maximum k. (95x 95) 0.93429 Effect of Worst Case Accident +0.01260 Maximum k- (95x 95, Including Accident) 0.94689 Margin +0.00311 22

NET-290-01 The reactivity effects of tolerances and uncertainties were evaluated with an infinite array of 9x9 assemblies at peak reactivity and when combined in a root-mean-square sense yield Ak = 0.01612. The difference between these values and the 0.95 design limit represents margin, which would be available to accommodate reactivity increases as may be the result of postulated accidents.

Table 5 Minimum Gadolinia Loading as a Function of Initial Peak Planar Enrichment for 9x9 Fuel 9s'"l° w/o U-235 Number of Gadolinia w/o Gadolinia Limiting k 1 (SCCG) 3.41 7 4.0 1.2349 4.0* 9 4.0 1.2676 4.21 12 5.0 1.2441 4.6+ 14 5.0 1.2568

  • Reference Case/Limiting kI

+ Predicted for Future Reloads 23

NET-290-01 3.5 Abnormal/Accident Conditions The reactivity effects of the following abnormal/accident conditions have been conservatively evaluated:

  • Fuel Assembly Drop

" Fuel Assembly Inadvertent Positioning Alongside Rack

" Fuel Assembly Misload

  • Moderator Temperature Variations The drop of a 1Ox1 0 reload fuel assembly assumed to come to rest in a horizontal position on top of the "North" module has been evaluated with all assemblies in place as shown in Figure 6 (on page 19). The reactivity effect is negligible (Ak < 0.00021).

The inadvertent positioning or drop of a fuel assembly alongside of the Boraflex modules in the corner of the "North" module and above the "South" module and the pool wall as shown in Figure 7 has been evaluated. The increase in rack reactivity as determined by KENO V.a is negligible (Ak < 0.00053).

For both the assembly drop and inadvertent positioning, the reactivity effect is well within the margin inherent in the design of the NMP1 spent fuel racks assuming 100% Boraflex loss.

The misloading of a 10x1O fuel assembly in the "North" Boraflex module has been evaluated for multiple positions within the Boraflex module. The maximum reactivity effect was determined to occur when the 10x1O reload assembly is centered in the "North" module shown in Figure 8, with the resulting reactivity effect Ak = 0.00110. Under the conservative assumptions of these analyses, the maximum fuel rack keff (at a 95%

probability with a 95% confidence level) has been determined to be 0.93539.

The effect of variations in moderator density and temperature on the reactivity of the NMP1 fuel storage racks has been analyzed(1 ). These analyses were performed at 220 0 F, the point of boiling at the depth of the fuel racks and with 20% void. The maximum reactivity effect is +0.01260Ak. For these conditions, the maximum keff (at a 95% probability with a 95% confidence. level) is 0.94689. Therefore, within the moderator temperature variations analyzed, adequate subcriticality margin is maintained.

24

NET-290-01 Figure 7: Keno V.a Generated Plot of a Dropped Assembly Resting on Top of "North" Boraflex Module.

25

NET-290-01 Dropped Bundle Figure 8: Keno V.a Generated Plot of a Dropped Fuel Assembly Alongside of the "North" and "South" Boraflex Modules.

26

NET-290-01 Misloaded Bundle Figure 9: Keno V.a Generated Plot of a Reload Assembly (10x1O)

Misloaded Adjacent to the "North" BORAL Modules.

27

NET-290-01

4.0 CONCLUSION

S The reactivity state of the NMP1 spent fuel storage pool has been analyzed and evaluated.

This analysis is based on 1) no reactivity credit for Boraflex, 2) with the existing fuel loading configuration below the tooling table and 3) with bounding reactivity (9x9) fuel loaded in the "North" module. With respect to the latter, the bounding 9x9 fuel type is characterized by an initial enrichment of 4.00 w/o U-235 with a minimum Gd 2 0 3 loading of 9 rods at 4.00 w/o. Analyses have demonstrated that for the NMP1 spent fuel racks the maximum keff is less than 0.95, when including the reactivity effects of tolerances, uncertainties, code biases and the effects of postulated accidents.

The maximum keff of the NMP1 spent fuel racks satisfies the 0.95 limit provided that:

1. The "South" Boraflex module is loaded with 8x8 assemblies containing the enrichments and gadolinia contents as shown in Figure 5 with 8x8 fuel.
2. The "North" Boraflex module is loaded:
  • with existing 7x7 or 8x8 fuel, or

" with future 9x9 fuel types at the peak planar enrichment with a minimum number of Gd 20 3 bearing rods and at a minimum gadolinia loading as specified in Table 5.

Provided that these conditions are met, the worst case accident scenarios, and the total loss of the Boraflex can be safely accommodated in the NMP1 spent fuel Boraflex racks.

28

NET-290-01

5.0 REFERENCES

1. "Nine Mile Point 1 Spent Fuel Pool Modification, Supplemental Submittal," Docket No. 50-220, DPR-63, Niagara Mohawk Power Corporation; June 1983.
2. "Evaluation of the Nine Mile Point 1 Boraflex Spent Fuel Racks for the General Electric 9x9 and 10x10 Fuel Types", NET-110-01, Northeast Technology Corp.;

March 26, 1996.

3. Email from W. Carter (NMP) to M. Harris (NETCO), dated 5/13/05 subject: "NMP1 SFP.xls".
4. Email from W. Carter (NMP) to M. Harris (NETCO), dated 6/29/07 subject: "NMP1 Boraflex per April 8, 2007 Shuffieworks Fig File.MDI".
5. Email from W., Carter (NMP) to M. Harris (NETCO), dated 9/25/2007,

Subject:

"Words from NMP1 Technical Specifications."

6. "SCALE-PC: Modular Code System for Performing Criticality Safety Analyses for Licensing Evaluation, for Workstations and Personal Computers", Version 5, Parts 0 thru 3, RSIC Computer Code Collection CCC-545. Oak Ridge National Laboratory:

Oak Ridge, Tennessee; May 2004.

7. Edenius, Malte and Bengt H. Forssen. "CASMO-4: A Fuel Assembly Burnup Program - User's Manual," Version 2.05, Rev 3, SSP-01/400. Studsvik of America:

Newton, Massachusetts; July 2003.

8. Memorandum from L. Kopp, SRE, to Timothy Collins, Chief, Reactor Systems Branch, Division of Systems Safety and Analysis, "Guidance on the Regulatory Requirements for Criticality Safety Analysis of Fuel Storage at Light Water Reactor Power Plants", August 19, 1998.

29

NET-290-01

9. Letter from W. Carter (NMP) to K. Lindquist (NETCO) w/Attachments: Letter from Ed Gibbs (GE) to Paul Netusil dated 7/14/1997 w/Attached Bundle Design Report for NMP1 from Cycle 1 through Reload 12 for GE8B Fuel and Applicable Sections of Bundle Design Reports for Reloads 9 through 12.
10. Email from W. Carter (NMP) to M. Harris (NETCO) dated 10/2/2007,

Subject:

"Bundle Announcement Reports for R13 to R19 (Current Cycle)."

11. Email from Mr. Kristopher Cummings (Holtec) to W. Carter (NMP), dated 9/11/07,

Subject:

"Proprietary Information on Holtec Racks at Nine Mile Point Nuclear Station."

12."Evaluation of Unseated Bundles on the Reactivity State of the Nine Mile Point 1 Spent Fuel Racks including the Effects of Boraflex Gaps, Shrinkage and Width Dissolution," NET-256-01, Northeast Technology Corp.; September 30, 2005.

30

NET-290-01 Appendix:

NETCO Report No. 901-02-05 Benchmarking Computer Codes for Calculating the Reactivity State of Spent Fuel Storage Racks, Storage Casks and Transportation Casks.

NET-290-01 Report No. 901-02-05 Benchmarking Computer Codes for Calculating the Reactivity State of Spent Fuel Storage Racks, Storage Casks and Transportation Casks Northeast Technology Corp.

108 North Front Street Third Floor UPO Box 4178 Kingston, New York 12402 Review/Approval Record Rev. Date Prepared by: Reviewed/Approved by: Approved (QA) by:

Table of Contents Section Page 1.0 INT R O D UC T IO N .................................................................................................. 1 2.0 BENCHMARKING - STANDARD PROBLEMS AND CONFIGURATION CONTROL ......................................................................... 3 2.1 SCALE-5 and MCNP5 Configuration Control ........................................ 3 2.2 S am ple P roblem s ................................................................................. .. 3 2.3 CASMO-4 Configuration Control ............................................................ 4 3.0 BENCHMARK MODELING OF LWR CRITICAL EXPERIMENTS ................... 5 3.1 Benchmarking of SCALE-5 and MCNP5 ............................................... 5 3.2 Benchmarking of CASMO-4 ................................................................ 12 4 .0 C O NC LU S IO NS ............................................................................................ . . 14 5 .0 R E F E R E NC E S .............................................................................................. . . 15

NET-290-01 List of Tables Table Page Table 3-1: B&W[ 6 1 and CSNI [7] Critical Experiments - Design Parameters ................. 8 Table 3-2: B&W[6] and CSNI[ 7 1 Critical Experiment Results ....................................... 9 Table 3-3: B&W Critical Experiments as CASMO Infinite Arrays - Results ............... 13 ii

List of Figures Figure Page Figure 3-1: Variation of Bias (keff -1) with Moderator-to-Fuel Ratio .......................... 10 Figure 3-2: Variation of Bias (keff -1) with Absorber Strength .................................. 11 iii

1.0 Introduction This report documents the results of benchmark calculations of three computer codes used to compute the reactivity state of nuclear fuel assemblies in close-packed arrays.

Such close-packed arrays include high density spent fuel storage racks, dry storage casks and casks for transporting nuclear fuel. The three computer codes, which were benchmarked and validated are:

  • KENO V.a, which is a module of SCALE 5[1]
  • MCNP5[21 1
  • CASMO-4[3 Earlier versions of KENO and CASMO have been previously benchmarked and validated by NETCO. 45' ]

To benchmark and validate the codes for spent fuel racks and cask evaluations, KENO and MCNP were used to simulate a series of critical experiments. The calculated eigenvalues (keff) were then compared with the critical condition (keff = 1.0).to determine the bias inherent in the calculated values. For the KENO V.a calculation, the 238 energy group ENDF/B-V cross-section library was used. For the MCNP5 calculations, the continuous energy cross-section library based on ENDF/B-VI was used.

After determining the inherent biases associated with KENO V.a and MCNP5, both KENO V.a and CASMO-4 (with its own 70 energy group cross-section library) were used to model central arrays of select critical experiments. It is noted that CASMO-4 models an infinitely repeating array of fuel assemblies and is generally used to generate cross-sections for core simulator models. As such, it does not lend itself directly to finite arrays of fuel racks surrounded by a reflector, as is the case in the critical experiments considered. Accordingly, the central fuel arrays of five critical experiments were modeled as infinite arrays with both KENO V.a and CASMO-4. A comparison of the KENO V.a and CASMO-4 eigenvalues provides a means to determine the CASMO-4 bias.

1

NET-290-01 For the purposes of benchmarking, a set of five Babcock and Wilcox (B&W) critical experiments (XIII, XIV, XV, XVII, and XIX)[61, were selected because they closely represent typical fuel/rack geometries with neutron absorber panels. In addition, the International Committee on the Safety of Nuclear Installations (CSNI) identified a sequence of benchmark problems17] that closely replicate both fuel/rack and fuel/cask geometries, and included typical LW enrichments and H/ 235 U ratios. The resulting models are representative of most fuel storage rack and fuel cask configurations used today.

All work completed for the benchmarking calculations was carried out under NETCO's Quality Assurance Program[81 . The methods employed have been patterned to comply with industry accepted standards[9' 10'11 -and with accepted industry criticality references[ 12,13,14,15,16]

2

NET-290-01 2.0 BENCHMARKING - STANDARD PROBLEMS AND CONFIGURATION CONTROL 2.1 SCALE-5 and MCNP5 Configuration Control The binary executable codes and associated batch files were provided by RSICC on CD-ROM for use on Intel Pentium based micro-computers running under the Windows operating system. In this form, the programs can not be altered or modified. In addition to the binary executable codes, there are several supporting files which contain cross-section sets, etc. The file name, file size, and creation date for each executable file is given in Appendix A*. Prior to executing either code sequence, the user will verify the file names, creation dates, and sizes to insure that they have not been changed.

Appendix B contains two CD-ROMs, which include the as-received versions of all files required to execute these programs. In all applications described in this report and for all subsequent applications, the files listed in Appendix A are to be used. This appendix is not provided in the non-proprietary version of this report.

2.2 Sample Problems A suite of input files with their corresponding output files were provided with each code.

The input file names and batch files used to execute them are listed in Appendix A.

These were executed on NETCO's host computer via batch files provided by RSICC and the resulting output files compared to those provided by RSICC on CD-ROM.

Except for the date and time of execution stamps, the respective output files were identical. Each code uses a pseudo-random number generator that is initiated with a default seed-Value. Since the default value was used in each case, the sequences of random numbers were the same, leading to identical calculations. This verifies that the as-received versions of both codes are identical to the versions documented in the User's Manuals[1 '2 ]

  • (Appendices A, B, C, D and E are included in the proprietary version of this report.)

3

NET-290-01 Examination of the sample input decks shows that the run modules in batch files exercise all of the code options used by this benchmarking exercise. Before and after each subsequent use of each code, one set of sample input modules are executed and the output files compared to the sample output files to verify that no system degradation has occurred. (All of these files are contained in Appendix B* at the end of this report).

This appendix is not provided in the non-proprietary version of this report.

2.3 CASMO-4 Configuration Control The version of CASMO-4 used for these analyses was developed for a RISCC workstation. Version 2.05.01 of CASMO-4 was used for this benchmarking work and subsequent users of CASMO-4 for NETCO will verify that Version 2.05.01 is being used. CASMO-4 and all versions are controlled by Studsvik of America under their Quality Assurance Program1171. If a different version of CASMO-4 is used by NETCO for any subsequent analyses, the CASMO-4 analyses in Section 3.2 shall be repeated with the version in use.

  • (Appendices A, B, C, D and E are included in the proprietary version of this report.)

4

NET-290-01 3.0 BENCHMARK MODELING OF LWR CRITICAL EXPERIMENTS An index of input and output files for each experiment modeled is contained in Appendices C* and D*. For each experiment, the input and output files are on 3.5 inch 1.44 MB diskettes which are also contained in Appendices C and D. Appendix E*

contains the calculation notebook for this project and represents a permanent record of all hand calculations performed during input preparation. All input parameters are fully traceable to the appropriate source documents. These appendices are not provided in the non-proprietary version of this report.

3.1 BENCHMARKING OF SCALE-5 and MCNP5 The B&W experiments[61 include twenty (20) water moderated LWR fuel rod cores and close-packed critical LWR fuel storage arrays. Of these, five (5) used boron carbide/aluminum cermet poison plates (BORAL) in the closest possible packing geometry representing a 3x3 array of LWR fuel assemblies in high density fuel storage racks. These five (5) experiments have been modeled, as they most closely represent LWR fuel in high density fuel storage racks and cask configurations with neutron absorber panels. Table 3-1 summarizes some of the model parameters, including U-235 enrichment, moderator-to-fuel ratio and absorber macroscopic absorption cross-section.

The Committee on the Safety of Nuclear Installations (CSNI) has published a selection of critical experiments 71 , which are a sequence of exercises arranged in order of increasing complexity, introducing one new parameter into the geometry and materials at a time. They were selected specifically to validate calculational methods for criticality safety assessments. The fuel is designed to simulate LWR fuel, is water moderated, and the lattices include BORAL plates between assemblies when neutron poisons are

  • (Appendices A, B, C, D and E are included in the proprietary version of this report.)

5

NET-290-01 included. The sequence starts with Experiment 1-1, a single array of 20x18, 2.35 w/o 235U rods with a water reflector all around. Experiments 1-2-1 and 1-2-2 are also single water reflected arrays but are at a higher enrichment (4.74 w/o 235 U) and are at undermoderated (1-2-1) and optimum moderation (1-2-2) conditions. Experiment 2-1 235 has three square arrays of 2.35 w/o U fuel separated by BORAL neutron absorber 23 5 plates. Experiment 2-2 has a 2x2 array of four 4.74 w/o U rod arrays also separated by BORAL plates. Experiments 3-A-1 and 3-B-1 are similar to experiment 2-1 but include, respectively, lead and steel reflecting walls. Experiment 3-A-2 is similar to Experiment 2-2 but also has a lead reflecting wall.

In each MCNP5 model of the criticals, 4,000,000 neutrons in 2,000 generations were tracked. In each KENO model of the criticals, at least 20,000,000 neutrons in at least 10,000 generations were tracked. The output files were always checked to insure that the fission source distribution had converged. A summary of the distribution of keff over all generations is automatically plotted in the output files and shows them to be approximately normally distributed. Thus, normal one-sided tolerance limits with appropriate 95% probability / 95% confidence factors (95/95) can be used. The calculated results for each critical experiment are given in Table 3-2, including the calculated keff, the one-standard-deviation statistical uncertainty of keff, denoted by Y, and the bias with respect to the critical state keff = 1.0.

The overall bias between the calculation eigenvalue and the experiments is calculated as follows. First, the variance-weighted mean is calculated as N N km,= * (k/c-7)/l (1/o- (3-1) where N = 13 (for the 5 B&W and 8 CSNI criticals), ki is the SCALE-5 calculated keff for 6

NET-290-01 critical i, and ai is the SCALE-5 calculated standard deviation of the distribution of keff for critical i. The standard deviation around k, is given by N 11/2 (3-2)

N= 1 (ki _k.)2]

The bias is calculated as km - 1, and has the same standard deviation as km. Based upon the results shown in Table 3-2, it is recommended that the 238 energy group ENDF/B-V library be used in all criticality analyses. For SCALE-5, the resulting mean bias for this library is -0.00782 +/- 0.00361. For MCNP5, using the continuous energy cross-section library based on ENDF/B-VI, the resulting variance weighted mean bias is

- 0.00574 +/- 0.00509.

Correlations of bias with respect to moderator-to-fuel ratio (H / 235 U) number density ratio and absorber strength (Zath) were investigated and found to be not significant. The coefficient of determination for bias versus moderator-to-fuel ratio for the 238 group ENDF/B-V library was a negligible 2.6%, whereas for MCNP5 it was 4.1%, indicating that the method bias is not strongly dependent on moderator-to-fuel ratio. In all cases, the bias becomes less negative with decreasing moderator-to-fuel ratio (i.e., increasing enrichment). The coefficient of determination for bias versus absorber strength for the 238 Group ENDF/B-V library was an insignificant 6.1%, while for MCNP5, it was 37.1%.

In all cases, the bias becomes less negative with increased absorber strength. These results are illustrated in Figures 3-1 and 3-2, respectively.

7

NET-290-01 Table 3-1: B&W [6] and CSNI 17] Critical Experiments - Design Parameters Reference Experiment Absorber Absorber Enrichment H/235U Number Type [cm] w% Ratio 6 XIII BORAL 1.871 2.459 216.43 6 XIV BORAL 1.460 2.459 216.52 6 XV BORAL 0.475 2.459 216.52 6 XVII BORAL 0.293 2.459 216.54 6 XIX BORAL 0.129 2.459 216.54 7 1-1 none - 2.35 398.72 7 1-2-1 none - 4.75 109.44 6 1-2-2 none - 4.75 228.53 7 2-1 BORAL 30.6 2.35 398.72 7 2-2 BORAL 24.6 4.75 228.53 7 3-A-1 none - 2.35 398.75 7 3-B-1 none 2.35 398.75 7 3-A-2 BORAL 24.6 4.75 228.53 8

NET-290-01 Table 3-2 B&WE6 ] and CSNIm Critical Experiment Results Reference Experiment eI S E5 Keff_____ sigma IIbias IIKeffI sigma bias 6 XII1 0.99341 0.00017 -0.00659 0.99422 0.00035 -0.00578 6 X]V 0.98989 0.00018 -0.01011 0.98997 0.00035 -0.01003 6 A/ 0.98623 0.00017 -0.01377 0.98525 0.00035 -0.01475 6 XVII 0.98972 0.00016 -0.01028 0.98846 0.00034 -0.01154 6 XIX 0.99136 0.00018 -0.00864 0.99004 0.00035 -0.00996 7 1-1 0.99048 0.00017 -0.00952 0.99294 0.00032 -0.00706 7 1-2-1 0.99404 0.00020 -0.00596 1.00000 0.00030 0.00000 7 1-2-2 0.99774 0.00020 -0.00226 1.00000 0.00030 0.00000 7 2-1 0.98925 0.00017 -0.01075 0.99164 0.00032 -0.00836 7 2-2 0.99549 0.00020 -0.00451 1.00000 0.00030 0.00000 7 3-A-1 0.99390 0.00018 -0.00610 0.99012 0.00033 -0.00988 7 3-B-1 0.99287 0.00017 -0.00713 0.99590 0.00033 -0.00410 7 3-A-2 0.99904 0.00020 -0.00096 0.99746 0.00041 -0.00254 Arithmetic 0.99218 0.99426 Mean Variance -0.00782 -0.00574 Weighted 4_.00782 -0.00574 Standard Deviation + 0.00361 + 0.00509 9

NET-290-01 0.005

- SCALE5, r2=2.6%

MCNP5,r 2=4.1%

  • SCALE5 0 A MCNP5 S

-0.005 U,

rm

-0.01

-0.015

-0.02 100 150 200 250 300 350 400 H/23 5U RATIO Figure 3-1: Variation of Bias (keff -1) with Moderator-to-Fuel Ratio 10

NET-290-01 0.005 2

-- SCALE5, r =6.1%

MCNP5 ,r2=7.1%

0 SCALE5 A MCNP5 0 A AA

-0.005 _ ° - -

co A II °- °'

-0.01 ý i I i i ii i i I i t i I

-0.015 0 10 20 30 40 ABSORBER Y,.",cm" Figure 3-2: Variation of Bias (keff -1) with Absorber Strength 11

NET-290-01 3.2 BENCHMARKING OF CASMO-4 This section compares SCALE-5ý1] and CASMO-4[ 21 calculations for k4 of the same five B&W critical experiments[6 1 discussed in Section 3.1. CASMO-4 is limited in its ability to render a geometric model and can only be used for infinite arrays of assemblies. Thus, for this benchmark analysis, the central assembly of the 3x3 array of assemblies in the B&W critical experiments was modeled and then assumed to be infinitely reflected. The assembly pitch was preserved in the model, but the effect of the finite water reflector around the 3x3 array was lost, making the model supercritical.

SCALE-5 was also used to model the B&W critical experiments with exactly the same geometry as they were rendered in CASMO-4. Because the bias of SCALE-5 is known (see Section 3.1), it can be applied to the SCALE-5 result to obtain a best-estimate of the supercritical state of the infinitely reflected assembly model. The CASMO-4 result can then be compared with this best estimate to obtain a CASMO-4 bias.

The results of the SCALE-5 and CASMO-4 analyses are compared in Table 3-3. The CASMO-4 bias is calculated as biaSCASMO-4 = kCASMO kSCALE-5, best estimate where kSCALE-5, best estimate = kSCALE biaSSCALE-5 For CASMO-4 the resulting mean bias and standard deviation for the 238 Group ENDF/B-V library are -0.01028 and 0.00198 respectively.

12

NET-290-01 Table 3-3: B&W Critical Experiments as CASMO Infinite Arrays - Results SCALE-5 (bias corrected)

Experiment CASMO-4[ 238GROUPNDF5

______ _____ Keff ][sigma bias X)II 1.08816 1.10160 0.00050 -0.01423 XQV 1.08860 1.10175 0.00049 -0.01523 XV\ 1.09832 1.10961 0.00045 -0.01280 XVII 1.10740 1.11732 0.00045 -0.00945 XMX 1.11614 1.12330 0.00043 -0.00832 bias i -0.01028

_ I _ _

Sigma ++/-0. 00198 13

NET-290-01

4.0 CONCLUSION

S SCALE-5 and MCNP5 have been benchmarked by modeling five (5) Babcock and Wilcox critical experiments and eight (8) CSNI critical experiments representative of fuel storage rack and fuel cask geometries. At a 95% probability / 95% confidence level, the computed bias for SCALE-5 and MCNP5 are -0.01381 and -0.01460, respectively.

CASMO-4 has also been benchmarked by modeling the five (5) Babcock and Wilcox critical experiments as infinite arrays. Best estimates of the k4 for the exact same geometry were calculated using SCALE-5 and applying the mean bias reported above.

The CASMO-4 bias with respect to these values was calculated to be -0.01028

+ 0.00198 (1 sigma). The comparison of SCALE-5 and CASMO-4 serves to verify the results of each with respect to the other.

It is therefore concluded that these calculational methods have been adequately benchmarked and validated. They may be used individually or in combination for the criticality analysis of spent fuel storage racks, fuel casks and fuel casks in close proximity to fuel storage racks, provided the appropriate biases are applied.

14

NET-256-01

5.0 REFERENCES

1. "SCALE-PC: Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations and Personal Computers, Version 5," Volumes 0 through 3, RSIC Computer Code Collection CCC-725.

Oak Ridge National Laboratory: Oak Ridge, Tennessee; Draft May 2004.

2. "MCNP - A General Monte Carlo N-Particle Transport Code, Version 5",

Volumes 1 - 3, RSICC Computer Code Collection CCC-710, X-5 Monte Carlo Team, Los Alamos National Laboratory, Los Alamos, NM, April 24, 2003.

3. Ekberg, Kim, Bengt H. Forssen and Dave Knott. "CASMO-4: A Fuel Assembly Burnup Program - User's Manual," Version 1.10 STUDSVIK/SOA-95/1. Studsvik of America: Newton, Massachusetts; September 1995.
4. NETCO Report 901-02-03: "Benchmarking of the SCALE-PC Version 4.3 Criticality and Safety Analysis Sequence Using the KENO V.a Monte Carlo Code and of the Multigroup Two-Dimensional Transport Theory CASMO-4 Code",

Northeast Technology Corporation: Kingston, New York; May 1995.

5. NETCO Report 901-02-04: "Benchmarking of the SCALE-PC Version 4.3 Criticality and Safety Analysis Sequence Using the KENO5A Monte Carlo code and of the Multigroup Two-Dimensional Transport Theory CASMO-4 Code",

Northeast Technology Corporation: Kingston, New York; January 2000.

6. Baldwin, M. N., G. S. Hoovler, R. L. Eng, and F. G. Welfare. "Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel,"

BAW-1484-7. Babcock & Wilcox: Lynchburg, Virginia; July 1979.

7. Mancioppi, S., and G. F. Gualdrini. "Standard Problem Exercise on Criticality Codes for Spent-Fuel-Transport Containers," CNEN-RT/FI(81)25. Comitato Nazionale Energia Nucleare: Rome; October 1981, Performed by CNEN for Committee for Safety of Nuclear Installations (CSNI).
8. "Quality Assurance Manual, Rev. 0, Northeast Technology Corp: Kingston, NY; 27 June 2001.
9. "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," ANSI/ANS-8.1-1983, Revision of ANSI/N 16.1-1975. American Nuclear Society: La Grange Park, Illinois; Approved 7 October 1983.

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NET-256-01

10. "Quality Assurance Requirements of Computer Software for Nuclear Facility Applications," Part 2.7 of "Quality Assurance Requirements for Nuclear Facility Applications," ASME NQA-2-1989, Revision of ANSI/ASME NQA-2-1986.

American Society of Mechanical Engineers: New York; Issued 30 September 1989.

11. Natrella, Mary Gibbons. Experimental Statistics, National Bureau of Standards Handbook 91. U.S. Government Printing Office: Washington, D.C.; 1 Aug. 1963.
12. Cooney, B. F., T. R. Freeman, and M. H. Lipner. "Comparison of Experiments and Calculations for LWR Storage Geometries." Transactions of the American Nuclear Society: Vol. 39, pp. 531-532; November 1981.
13. Westfall, R. M., and J. R. Knight. "SCALE System Cross Section Validation with Shipping Cask Critical Experiments." Transactions of the American Nuclear Society: Vol. 33, pp. 368-370; 1979.
14. McCamis, R. H. "Validation of KENO V.a for Criticality Safety Calculations of Low-Enriched Uranium-235 Systems," AECL-10146-1. Atomic Energy of Canada Limited, Whiteshell Laboratories, Pinawa, Manitoba; February 1991.
15. Bierman, S. R., E. D. Clayton, and B. M. Durst. "Critical Separation Between Subcritical Clusters of 2.35 Wt% 2 3 5 U Enriched UO2 Rods in Water with Fixed Neutron Poisons," PNL-2438. Battelle Pacific Northwest Laboratories: Richland, Washington; October 1977.
16. Bierman, S. R., E. D. Clayton, and B. M. Durst. "Critical Separation Between 235 U Enriched U0 2 Rods in Water with Fixed Subcritical Clusters of 4.29 Wt%

Neutron Poisons," NUREG/CR-0073 RC. Battelle Pacific Northwest Laboratories: Richland, Washington; May 1978.

17. "Quality Assurance Program", SOA/REV 2. Studsvik of America: Newton, Massachusetts; Approved 16 August 1991.

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