Canister Handling Facility Criticality Safety Calculations
 
190-00C-CH00-00100-000-00B

April 2005

1. PURPOSE 
This design calculation revises and updates the previous criticality evaluation for the canister 
handling, transfer and staging operations to be performed in the Canister Handling Facility 
(CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the 
calculation is to demonstrate that the handling operations of canisters performed in the CHF meet 
the nuclear criticality safety design criteria specified in the Project Design Criteria (PDC) 
Document (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in 
Project Requirements Document (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the 
functional/operational nuclear safety requirement in the Project Functional and Operational 
Requirements document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear 
criticality safety requirements described in the Canister Handling Facility Description Document 
(BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in 
this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in 
the Categorization of Event Sequences for License Application (BSC 2004 [DIRS 167268], 
Section 7). 
The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) 
spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level 
radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dualpurpose 
canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become 
available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety 
analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval 
Nuclear Propulsion Program) and will not be included in this document. In addition, this 
calculation is valid for the current design of the CHF and may not reflect the ongoing design 
evolution of the facility. However, it is anticipated that design changes to the facility layout will 
have little or no impact on the criticality results and/or conclusions presented in this document. 
This calculation is subject to the Quality Assurance Requirements and Description (DOE 2004 
[DIRS 171539]) because the CHF is included in the Q-List (BSC 2005 [DIRS 171190], p. A-3) 
as an item important to safety. This calculation is prepared in accordance with AP-3.12Q, Design 
Calculations and Analyses [DIRS 168413].

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2. METHOD 
2.1 CRITICALITY SAFETY ANALYSIS 
The criticality safety calculations presented in this document evaluate the DOE SNF canisters in 
the CHF to ensure they meet the criticality safety requirements under normal conditions as well 
as for Category 1 and 2 event sequences, in accordance with 10 CFR 63 [DIRS 173273]. Further, 
this calculation determines the minimum spacing for the canister staging racks in the CHF and a 
controlled moderator height for defense in depth. The off-normal and accident conditions are 
inclusive of Category 1 and 2 event sequences as defined in 10 CFR 63.2 [DIRS 173273]. 
Moderator and reflector conditions are varied to find the most reactive configuration. The 
process and methodology for criticality safety analysis given in the Preclosure Criticality 
Analysis Process Report (BSC 2004 [DIRS 172058], Section 2.2.7) will be implemented in these 
calculations, and the following method will be followed: 
 The design of the facility will be based on the most reactive fuel. 
 The multiplication factor (keff), including all biases and uncertainties at a 95 % confidence 
level, will not exceed 0.95 under all credible normal, and Category 1 and 2 event sequences 
(NRC 2000 [DIRS 149756], Section 8.4.1.1). 
 Conservative modeling assumptions will be considered leading to maximum reactivity for 
dimensional variables (e.g., pitch and manufacturing tolerances for canisters). 
 Conservative modeling assumptions will also be used regarding materials in the fuel 
including no accounting for burnable poisons in fuel, no credit for 234U and 236U in fuel, no 
credit for fission products or transuranic absorbers in fuel, and use of the most reactive fuel 
stack density. 
 Fixed neutron absorber used for criticality control can only be taken credit for up to 75% of 
the neutron absorbing material (NRC 2000 [DIRS 149756], Section 8.4.1.1). 
Note that the terms model(s) and modeling as used in this calculation document refer to the 
geometric configurations of the criticality cases analyzed and not scientific models per LPSIII.
10Q-BSC, Models [DIRS 172972]. 
These calculations use the qualified software MCNP (Briesmeister 1997 [DIRS 103897] and 
CRWMS M&O 1998 [DIRS 154060]). MCNP is a three-dimensional Monte Carlo particle 
transport code with the capability to calculate eigenvalues for critical systems. The Nuclear 
Regulatory Commission (NRC) accepts MCNP in NUREG-1567 (NRC 2000 [DIRS 149756], 
p. 8-10) for criticality calculations. 
2.2 ELECTRONIC MANAGEMENT OF INFORMATION 
Electronic management of information generated from these calculations is controlled in 
accordance with Section 5.1.2 of AP-3.13Q, Design Control [DIRS 167460]. The computer input 
and output files generated from this calculation are stored on a Compact Disc (CD), and 
submitted as an attachment to this document (Attachment II).

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3. ASSUMPTIONS 
3.1 DOE SNF FUEL (DROP SCENARIOS) 
Assumption: It is assumed that a drop of a DOE SNF canisters will not cause a breach in the 
CHF. 
Rationale: DOE SNF standardized canisters can withstand, without breaching, a drop in any 
orientation from a height of 23 ft (7 m) onto an essentially unyielding flat surface per the U.S. 
Department of Energy Spent Nuclear Fuel Canister Survivability document (BSC 2004 [DIRS 
168792], Section 6). Multicanister overpack (MCO) canisters can withstand, without breaching, 
a flat-bottom drop (3 degrees or less off vertical) from a height of 23 ft (7 m) and a drop in any 
orientation from a height of 2 ft (0.6 m) (individuallynot both in sequence) onto an essentially 
unyielding flat surface (BSC 2004 [DIRS 168792], Section 6). These lifting heights are 
considered through requirement PRD-013/T-014 (Canori and Leitner 2003 [DIRS 166275], p. 3- 
63) since drop test performance (i.e., canisters surviving flat bottom drops from a height of 23 ft 
and a drops in any orientation from a height of 2 ft) of the DOE SNF canisters must be 
considered. It is therefore reasonable to assume that the DOE SNF canisters will not breach 
during handling in the CHF. 
Confirmation Status: This assumption requires no further confirmation based on the stated 
rationale. 
Use in the Calculation: Sections 5.1.4, 6.3, and 6.4. 
3.2 WASTE PACKAGE COMPONENTS MODELED AS RIGHT CYLINDERS 
Assumption: It is assumed that the components, such as DOE SNF disposable canisters, tubes, 
end fittings, and fuel pins may be modeled as right prisms or right cylinders. In most cases, this 
was accomplished by conserving volume but changing geometry, i.e., replacing a region with an 
irregular shape of structural material with two cylindrical regions (one of structure and one of 
void) having the proper volumes. For a few other cases structural material was removed or fuel 
material was added to achieve a right prism or right cylinder (as in the case of dished fuel pellets 
modeled as right cylinders). 
Rationale: This assumption is conservative. Geometry of structural material does not 
significantly affect reactivity as long as the approximate thickness is conserved. Removing 
structural material is conservative because the structural material is composed mainly of neutron 
absorbers, and hence its absence provides a higher value for the keff of the system. Addition of 
fuel material is conservative because it adds fissile material and increases the reactivity of the 
system. This assumption is made to simplify the MCNP calculation but does not significantly 
impact the results of the calculation. 
Confirmation Status: This assumption requires no further confirmation based on the stated 
rationale. 
Use in the Calculation: Section 5.1.4.

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3.3 STRUCTURAL MATERIAL MAY BE NEGLECTED 
Assumption: It is assumed that some components of the DOE SNF disposable canisters (such as 
baskets or structural material) may be neglected. 
Rationale: This assumption is conservative since these components are composed primarily of 
neutron absorbers. 
Confirmation Status: This assumption requires no further confirmation based on the stated 
rationale. 
Use in the Calculation: Section 5.1.4. 
3.4 HYDRAULIC FLUID COMPOSITION 
Assumption: It is assumed that the hydraulic fluid used as an alternative moderator material was 
a conventional silicone fluid (polysiloxane fluid) with a degree of polymerization of four (Gelest 
2004 [DIRS 169915], p. 11). 
Rationale: The basis for this assumption is that the CHF design has not identified the hydraulic 
fluid for lubrication, but the candidate material used for lubrication is expected to be less 
effective moderator than water. The material used for this calculation to demonstrate criticality 
safety is a common silicone based hydraulic fluid (Gelest Inc. 2004 [DIRS 169915], p. 7). It 
should be recognized that this material might not represent the most reactive condition. Any 
change in the design selection of the hydraulic fluid would require re-evaluation of criticality 
safety to ensure that all the criticality design criteria are met. 
Confirmation Status: This assumption requires no further confirmation based on the stated 
rationale. 
Use in the Calculation: Section 5.1.4. 
3.5 PINS IN IDENT-69 CANISTER ARE SIMILAR IN GEOMETRY TO PINS IN 
TYPE 4.1 DFA 
Assumption: Exact dimensions for the fuel pins in the Ident-69 canister are not known. Only 
the outer cladding diameter is given in source references. It is assumed that the total volume of 
fuel in the source assembly is the same for the Ident-69 pins and the Type 4.1 DFA. It is further 
assumed that dimensions for the two fuel pins are similar: same length of each component 
including active fuel length and same cladding thickness. 
Rationale: Given the number of fuel pins in each source assembly, an assumption of same 
volume and active fuel length results in a calculated fuel pellet diameter that is reasonable given 
the outer diameter of the Ident-69 fuel rods (SNF.xls, worksheet FFTF). Other dimensions such 
as total pin length and dimensions for structural components are not expected to significantly 
impact the reactivity of the fuel. 
Confirmation Status: This assumption requires no further confirmation based on the stated 
rationale. 
Use in the Calculation: Section 5.1.4.

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3.6 ALUMINUM 3003 MAY BE MODELED AS ALUMINUM 6061 
Assumption: For Fermi SNF, it is assumed that the end cups of the 04 canister (made of 
aluminum 3003) are made of aluminum 6061 like the main body of the canister. 
Rationale: The volume of the end cups comprises a small fraction of the volume of the whole 
body of the 04 canister. The composition differences between aluminum 3003 and aluminum 
6061 are neutronically insignificant. In addition, aluminum is a weak neutron absorber. 
Confirmation Status: This assumption requires no further confirmation based on the stated 
rationale. 
Use in the Calculation: Section 5.1.4. 
3.7 DENSITY OF GdPO4 
Assumption: It is assumed that the density of GdPO4 (anhydrous gadolinium phosphate) is 
5 g/cm3. 
Rationale: No density is reported for GdPO4. However, gadolinium is a rare earth element, 
and compounds formed by gadolinium will have similar properties to other rare earth 
compounds. The density of monazite (anhydrous rare earth phosphate containing a mix of rare 
earth elements) is reported as 5  5.3 g/cc (Weast, R.C., ed. 1972 [DIRS 127163], p. B-195). 
The density of GdPO4 is expected to fall into this range. The lower bound of 5 g/cc was selected 
because this results in the smallest mass of Gd and is therefore conservative. 
Confirmation Status: This assumption requires no further confirmation based on the stated 
rationale. 
Use in the Calculation: Attachment II/SNF.xls, worksheet Fermi. 
3.8 COMPOSITION OF PINS IN IDENT-69 CANISTER 
Assumption: The exact isotopic composition for the fuel pins in the Ident-69 canister is not 
known. Only the Pu / U ratio is given in source references. It is assumed that the isotopic ratios 
of the isotopes of Pu and the isotopes of U are the same for the Ident-69 pins and the Type 4.1 
driver fuel assembly (DFA). It is further assumed that the densities of Ident-69 and Type 4.1 
fuel are the same. 
Rationale: Based on the data in (INEEL 2002 [DIRS 158820], Table 1), the isotopic 
breakdowns of Pu and U are expected to be approximately the same for all fuels. Type 4.1 fuel 
has the highest ratio of Pu-239, which is conservative. 
Confirmation Status: This assumption requires no further confirmation based on the stated 
rationale. 
Use in the Calculation: Attachment II/SNF.xls, worksheet FFTF.

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4. USE OF COMPUTER SOFTWARE 
4.1 BASELINED SOFTWARE 
4.1.1 MCNP 
The MCNP code (CRWMS M&O 1998 [DIRS 154060]) was used to calculate the multiplication 
factor, keff, for all systems presented in this report. The software specifications are as follows: 
 Program Name: MCNP (CRWMS M&O 1998 [DIRS 154060]) 
 Version/Revision Number: Version 4B2LV 
 Status/Operating System: Qualified/HP-UX B.10.20 
 Software Tracking Number: 30033 V4B2LV 
 Computer Type: HP 9000 Series Workstations 
 CPU Number: 700887 
The input and output files for the various MCNP calculations are contained on a CD (Attachment 
II) and the files are listed in Attachment I. 
The MCNP software used was: (1) appropriate for the criticality (keff) calculations, (2) used only 
within the range of validation as documented through Briesmeister (1997 [DIRS 103897]) and 
CRWMS M&O (1998 [DIRS 102836], Section 3.1), and (3) obtained from Software 
Configuration Management in accordance with appropriate procedures. 
4.2 COMMERCIAL OFF-THE-SHELF SOFTWARE 
4.2.1 MICROSOFT EXCEL 97 SR-2 
 Title: Excel 
 Version/Revision Number: Microsoft Excel 97 SR-2 
 This version is installed on a PC running Microsoft Windows 2000 with CPU number 
503009 
The file for the Excel calculation is contained on a CD (Attachment II) and the file is listed in 
Attachment I. 
Excel was used to calculate weight fractions and weight percent. The calculations can be 
reproduced and checked by hand. Excel is exempt from qualification per Section 2.1.6 of LPSI.
11Q-BSC, Software Management [DIRS 171923].

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5. CALCULATION 
All technical product inputs and sources of the inputs used in the development of this calculation 
are documented in this section. Attachment III features general arrangement drawings of the 
CHF as of the date of this calculation, and may not reflect the ongoing design evolution. The 
purpose of these drawings is to show the functional areas where the canisters will be handled and 
staged and to show the moderator controlled areas. 
5.1 CALCULATIONAL INPUTS 
5.1.1 Design Requirements and Criteria 
The design criteria for criticality safety analysis provided in Section 4.9.2.2 of the Project 
Design Criteria Document (BSC 2004 [DIRS 171599]) are used in these calculations. The 
pertinent criteria for CHF criticality safety include the following (BSC 2004 [DIRS 171599], 
Section 4.9.2.2): 
 The multiplication factor, keff, including all biases and uncertainty at a 95 percent confidence 
level, shall not exceed 0.95 under all normal conditions, and Category 1 and Category 2 
event sequences. 
 For fixed-neutron absorbers used for criticality control such as grid plates or inserts, no more 
than 75 percent credit of the neutron absorber content is used for preclosure criticality 
analyses, unless standard acceptance tests verify that the presence and uniformity of the 
neutron absorber are more effective. 
The Project Requirements Document seeks to prevent unplanned nuclear criticality events 
through requirement PRD-015/P-096 (Canori and Leitner 2003 [DIRS 166275], p. 4-206). From 
an operational/performance requirement point of view, the staging racks  shall be designed to 
maintain criticality safety per the Project Functional and Operational Requirements document 
(Curry 2004 [DIRS 170557], Section 1.1.6-4). The functional requirement 3.1.1.3.4.13 of the 
Canister Handling Facility Description Document (BSC 2004 [DIRS 168992], p. 3-22) states 
that the canister staging pit geometry shall prevent criticality. The criteria for this requirement 
is that the facility shall provide space, layout, and barriers as required to ensure moderator 
control in criticality sensitive areas and that piping containing moderator shall be routed 
around criticality sensitive area. Further, functional requirement 3.2.3.1 (BSC 2004 [DIRS 
168992], p. 3-28) states that the facility shall be designed and operated to prevent any credible 
criticality event from occurring. The basis for this requirement is to prevent nuclear criticality 
events, in accordance with DOE nuclear facility safety programs. 
5.1.2 DOE SNF Types 
Criticality evaluations are presented in this document for DOE fuel, which has been categorized 
into nine fuel groups (Mecham, D.C. 2004 [DIRS 170673], Section 4.2.4.1). A representative 
fuel type is chosen as bounding case for each group. Table 5.1-1 shows the nine DOE fuel 
groups and their corresponding representative fuel.

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Table 5.1-1 DOE Fuel Groups and Representative Fuel Types 
Group 
Number DOE Fuel Type a Representative Fuel 
1 Uranium Metal N-Reactor 
2 Uranium-Zirconium/Uranium-Molybdenum Enrico Fermi 
3 Uranium Oxide (High Enriched Uranium) Shippingport Pressurized Water Reactor 
(PWR) 
4 Uranium Oxide (Low Enriched Uranium) Three Mile Island (debris) 
5 Uranium-Aluminum Advanced Test Reactor (ATR) 
6 Uranium/Thorium/Plutonium Carbide Fort St. Vrain 
7 Mixed Oxide Fast Flux Test Facility (FFTF) 
8 Uranium/Thorium Oxide Shippingport Light Water Breeder Reactor 
(LWBR) 
9 Uranium-Zirconium-Hydride Training Research Isotopes General Atomics 
(TRIGA) 
a Source data from Mecham, D.C. 2004 [DIRS 170673], Section 4.2.4.1 
5.1.3 Upper Subcritical Limit 
The definition of upper subcritical limit (USL) is (BSC 2004 [DIRS 172058], Section 3.5): 
kS + .kS = USL (1) 
where kS is the MCNP calculated value for the system, .kS is an allowance for (a) statistical or 
convergence uncertainties, or both in the computation of ks, (b) material and fabrication 
tolerances, and (c) uncertainties due to the geometric or material representations used in the 
computational method [Note: allowance for items (b) and (c) can be obviated by using bounding 
representations]. 
Per ANS-8.7/ANSI N16.5 [DIRS 144741], the calculated keff for a fissile system is considered to 
be acceptable provided the calculated keff plus 2 sigma is less than a specified USL. The 
Benchmark and Critical Limit Calculation for DOE SNF (BSC 2002 [DIRS 161781], Section 
6.3) document establishes a USL of 0.9631 for DOE fuel based on the statistical averaging 
method. Applying a 0.05 administrative margin (BSC 2004 [DIRS 172058], Section 3.4.1), the 
final USL for the calculations presented in this document is 0.9131. 
5.1.4 DOE SNF Calculation Inputs 
The descriptions of the fuel loading in the MCO canisters and DOE SNF standardized canisters 
in the following subsections are taken from the most current criticality analyses and design 
documents (unless otherwise noted). Since the DOE SNF canisters arrive sealed at the CHF 
(Section 1) and are assumed to not breach (Assumption 3.1), the DOE SNF canisters are 
modeled as dry on the inside. For defense in depth, however, the most reactive DOE fuel types 
are modeled with moderator (partially and fully flooded) on the inside.

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5.1.4.1 N-Reactor Fuel and Canister Description 
The MCNP input model representing N-Reactor fuel in a canister utilized the existing MCNP 
input files 4zv0111 and 1zv0111 from the Intact and Degraded Component Criticality 
Calculations of N Reactor Spent Nuclear Fuel (CRWMS M&O 2001 [DIRS 153262]) document 
as a starting point for the present calculations. The previously existing MCNP input files were 
stripped of the waste package (WP) and surrounding HLW canisters so that only the DOE 
canister itself was contained and modeled in the present MCNP input files. 
Two types of fuel were considered: Mark IV fuel and Mark 1A fuel. Figure 5.1-1 presents the 
radial view of the DOE canisters containing Mark IV and Mark 1A fuel. One Mark IV fuel 
element contains a maximum of 16.0 kg of 0.947 % enriched uranium in the outer tube and 7.5 
kg of 0.947 % enriched uranium in the inner tube (DOE 2000 [DIRS 150095], Table 3-1). The 
multi-canister overpack (MCO) contains up to 54 fuel elements stacked in five layers (DOE 2000 
[DIRS 150095], pp. 23-25). The fuel tubes are 60.96 mm in diameter (DOE 2000 [DIRS 
150095], Table 3-1) and they are close packed in the disposal canister. 
The Mark 1A fuel element is enriched to 1.25 % U-235 in the outer tube (DOE 2000 [DIRS 
150095], Table 3-1) and there are 48 fuel elements in the canister with the center post being 
6.625 inch outer diameter (OD) versus 2.835 inch OD for Mark IV (DOE 2000 [DIRS 150095], 
Figure 4-2). The outer fuel tube is 6.096 cm diameter on the outside, 4.496 cm diameter on the 
inside and it is 53 cm long (DOE 2000 [DIRS 150095], Table 3-1). The cladding is 0.0635 cm 
thick on the outside and 0.0555 cm thick on the inside (DOE 2000 [DIRS 150095], Table 3-2). 
Similarly, the inner tube is 3.175 cm diameter on the outside, 1.118 cm diameter on the inside 
and it is 53 cm long (DOE 2000 [DIRS 150095], Table 3-1). The cladding is 0.1015 cm thick on 
the outside and 0.0635 cm thick on the inside (DOE 2000 [DIRS 150095], Table 3-2). 
The canister design (MCO with no neutron absorber) includes a nominal length of 4198.37 mm 
(165.29 in.) and a maximum outer diameter of 642.9 mm (25.31 in.) (CRWMS M&O 2001 
[DIRS 153262], p. 14). Beyond these basic dimensions, fuel-specific internals (also called 
baskets) have been designed for each canister based on the known maximum lengths of the fuels 
(Mark IV or 1A) contained therein. The MCOs are constructed out of 304L stainless steel having 
an outside diameter 60.92 cm (23.985 in.) and a wall thickness of 1.27 cm (0.5 in.) (CRWMS 
M&O 2001 [DIRS 153262], p. 14). The top portion of the MCO has a slightly larger diameter of 
64.29 cm (25.31 in.) than the overall tube, the overall length of the MCO is 422.707 cm (166.42 
in.) with an inner cavity height to the top of the stacked baskets of 356.545 cm (140.372 in.), and 
the bottom plate has a thickness of 5.11 cm (2.01 in.) (CRWMS M&O 2001 [DIRS 153262], p. 
14). There is a metal structure that adds another 57.91 cm (22.80 in.) to the top of the MCO 
above the top basket (CRWMS M&O 2001 [DIRS 153262], p. 14). This structure is not 
represented in these calculations and the bottom plate is represented with a thickness of 
4.4704 cm. 
It should also be mentioned that a central process post constructed out of 304L stainless steel is 
present in the MCOs. In the case of the Mark IV fuel baskets, the post outer diameter is 7.20 cm 
(2.835 in.) with a 1.37 cm (0.54 in.) thick wall. The Mark 1A fuel use a 16.83 cm (6.625 in.) 
diameter post with a 4.458 cm (1.755 in. [max.]) hole drilled in the center for a 6.18 cm (2.435 
in.) wall thickness (DOE 2000 [DIRS 150095], Figure 4-2).

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NOTE: Figure not to scale. 
Figure 5.1-1 Radial View of DOE Canisters Containing Mark IV and Mark 1A Fuel 
5.1.4.2 Enrico Fermi Fuel and Canister Description 
The MCNP input model representing Enrico Fermi fuel in a canister utilized the existing MCNP 
input file can3 from the Criticality Potential of Intact DOE SNF Canisters in a Misloaded Dry 
Waste Package (BSC 2004 [DIRS 172201]) document as a starting point for the present 
calculations. The boundary conditions were the only changes made to the previously existing 
MCNP input files (see Section 5.2 for description of boundary conditions). 
Fermi fuel is packaged as loose pins contained in two concentric DOE SNF shipping canisters. 
The fuel pins are made of uranium/molybdenum alloy (approximately 10 wt% molybdenum 
alloyed with uranium of 25.69% U-235 enrichment) (DOE 1999 [DIRS 104110], Section 3). 
The fuel is contained by zirconium cladding, and there are no gaps between the cladding and the 
fuel. Zirconium fuel pin tips are swaged onto the ends of the fuel pin. The Fermi fuel pins are 
placed inside canisters known as 04 canisters, which were in turn placed in 01 canisters. 
Each 01 canister contains one 04 canister, and each 04 canister contains 140 loose fuel pins 
with no supporting or spacing mechanism. The 01 and 04 canisters are made from 6061 
aluminum, except for the end plugs of the 04 canister which are fabricated from 3003 
aluminum. For this calculation, the end plugs of the 04 canister are also assumed to be of 6061 
aluminum (Assumption 3.6). The 01 shipping canisters with all their contents are placed into 
the short DOE standardized SNF canister with the twelve support tubes welded to a base plate 
(CRWMS M&O 1999 [DIRS 104118], Section 5). The plate-and-tube assembly is then placed 
into the DOE standardized SNF canister and the 01 canisters are placed into the tubes and the 
gaps are filled with iron shot containing a percentage of GdPO4. This shot is poured between the 
01 canisters and the tubes and between the tubes and the DOE standardized SNF canister. A 
second plate-and-tube assembly is placed in the DOE standardized SNF canister using the same 
method, for a total of 24 01 canisters per DOE standardized SNF canister. The two layers are

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topped with a spacer plate and a spacer before the DOE standardized SNF canister is sealed. 
Figure 5.1-2 shows the radial view of the Fermi DOE SNF canister. 
The geometry of the Fermi fuel and SNF contents has been simplified to adapt the geometry to 
right cylinders (Assumption 3.2). In addition, the fuel pins were rearranged into a hexagonalpitch 
array (a more conservative arrangement). A comparison of the actual dimensions of the 
Fermi fuel and canister with the dimensions used in the MCNP cases can be found in Table 5.1- 
2. Note that this calculation considers support tubes made of Ni-Gd alloy and a loading of 3% by 
volume GdPO4 in the iron shot. The compositions of the support tubes, Fermi fuel, and iron shot 
are calculated in SNF.xls, worksheets Ni-Gd alloy and Fermi. 
NOTE: Figure is not to scale. 
Figure 5.1-2 Radial View of the Enrico Fermi DOE SNF Canister

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Table 5.1-2 Fermi Fuel and Packaging Dimensions and Materials 
Component Material a Parameter Actual Dimension 
(mm) a 
Dimension Used (mm) 
Outer Diameter 3.7592 3.7592 
Fuel U/Mo Alloy 
Length 774.70 774.7000 
Inner Diameter 3.7592 3.7592 
Outer Diameter 4.0132 4.0132 
Inner Length 774.70 774.7000 
Cladding Zirconium 
Outer Length 781.9390 774.7000 
Entire Fuel Pin -------------- Pin Pitch Varies (loose) 4.0132 (hexagonal) 
Inner Diameter 66.548 66.5480 
-04 Canister Walls Aluminum 6061 
Outer Diameter 69.85 69.8500 
Total Length 50.8 50.8000 c -04 Canister Top 
Plug 
Aluminum 6061 
(Assumption 3.6) Thickness 1.651 1.6510 c 
Total Length 25.4 25.4000 c -04 Canister Bottom 
Plug 
Aluminum 6061 
(Assumption 3.6) Thickness 1.651 1.6510 c 
Inner Length 825.5 825.5000 
Entire -04 Canister -------------------- 
Outer Length 901.7 828.8020 c 
Inner Diameter 76.2 76.2000 
-01 Canister Walls Aluminum 6061 
Outer Diameter 82.55 82.5500 
Total Length 101.6 (estimated) 101.6000 -01 Canister Top 
Fitting Aluminum 6061 d 
Lid Thickness 12.7 12.7000 
Total Length 25.4 25.4000 -01 Canister Bottom 
Fitting Aluminum 6061 d 
Lid Thickness 12.7 12.7000 
Inner Length 952.5 (approx) 952.5000 
Entire -01 Canister -------------------- 
Outer Length 1079.5 977.9000 d 
Inner Diameter 92.0 f 92.0000 
Outer Diameter 101.6 f 101.6000 
Length 1100.0 f 1100.0000 Support Tubes Ni-Gd alloy e 
Pitch Not uniform 
(see Figure 5.1-2) f 
Not uniform 
(see Figure 5.1-2) 
Diameter 426 f 438.1500 
Base/Spacer Plate SS 316L 
Thickness 9.5 f 9.5000 
Spacer Void g Length 335.5 f 335.5000 
Void over Spacer Void Length 11 f 11.0000 
Inner Diameter 430 (min) b 438.1500 Short Standardized 
SNF Canister SS 316L b 
Inner Length 2540 (min) b 2575.0000 
a Source data from DOE 1999 [DIRS 104110], Section 3 and Section 4 unless otherwise stated. 
b Source data from DOE 1999 [DIRS 140225], Sections 3.2.3 (material) and 3.2.2 (dimensions). 
c The 04 canister end plugs were modeled as a lid with the same thickness as the plug material and an adjacent 
region of void for the remainder of the hollow plug length (Assumption 3.6). 
d The 01 end fittings were modeled as a solid lid and an adjacent void region for the remainder of the fitting length 
(Assumption 3.3). Iron shot is conservatively excluded from this void region. 
e Specified in DOE 1999 [DIRS 104110], Section 4.1 as steel. Ni-Gd alloy was substituted. 
f Source data from CRWMS M&O 1999 [DIRS 104118], Attachment 3. 
g Specified in CRWMS M&O 1999 [DIRS 104118], Attachment 3 as SS 316L. The spacer is hollow, so void was 
substituted (Assumption 3.3).

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5.1.4.3 Shippingport PWR Fuel and Canister Description 
As a starting point for the present Shippingport PWR calculations, the existing MCNP input file 
spdds00 was utilized from the Canister Handling Facility Criticality Safety Calculations (BSC 
2004 [DIRS 167614], which initially originated from the Intact and Degraded Criticality 
Calculations for the Codisposal of Shippingport PWR Fuel in a Waste Package (CRWMS M&O 
2000 [DIRS 144714]) document. 
One Shippingport PWR fuel cluster contains 19.5 kg of U-235 with an enrichment of 93.2% 
(DOE 1999 [DIRS 104940], Table 3-1). The fuel is a mixture of UO2-ZrO2-CaO (DOE 1999 
[DIRS 104940], p. 6). The most reactive fuel consists of 54.9 wt% UO2, 39.9 wt% ZrO2 and 5.2 
wt% CaO (DOE 1999 [DIRS 104940], Table 3-2). The canister contains only one fuel cluster 
and there is no added neutron absorber in the canister. Figure 5.1-3 presents the radial view of 
the DOE canister containing Shippingport PWR fuel. 
The DOE SNF canister is a right circular cylinder pipe made of stainless steel (Type 316L or 
UNS S31603) with an outside diameter of 457.2 mm (18 in.) and a wall thickness of 9.525 mm 
(0.375 in.) (CRWMS M&O 2000 [DIRS 144714], p. 15). The nominal internal length of the 
DOE SNF canister reserved for fuel loading is 268.09 cm with a 19 mm thick base plate and a 
9.5 mm thick top spacer plate (CRWMS M&O 2000 [DIRS 144714], p. 15). The nominal 
internal length of the empty space above the top spacer plate to the impact plate is 140.75 cm and 
the inner radius of the spacer cylinder in the empty space is 20.615 cm with a thickness of 0.635 
cm (CRWMS M&O 2000 [DIRS 144714], p. 15). The canister also contains 9.5 mm thick 
stainless steel (Type 316L) guide plates that are used to hold the Shippingport PWR SNF cluster 
(CRWMS M&O 2000 [DIRS 144714], p. 15). 
NOTE: Figure is not to scale. 
Figure 5.1-3 Radial View of DOE Canister Containing Shippingport DOE Fuel

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5.1.4.4 TMI-2 Fuel and Canister Description 
The MCNP input model representing TMI-2 fuel in a canister utilized the existing MCNP input 
file can9 from the Criticality Potential of Intact DOE SNF Canisters in a Misloaded Dry Waste 
Package (BSC 2004 [DIRS 172201]) document as a starting point for the present calculations. 
The boundary conditions were the only changes made to the previously existing MCNP input 
files (see Section 5.2 for description of boundary conditions). 
TMI fuel debris originates from TMI Unit 2 (TMI-2) where the typical fuel assembly installed in 
the reactor was a 15 x 15 Babcock & Wilcox PWR fuel assembly. Each fuel assembly contained 
an average of 208 fuel pins. The fuel was uranium oxide with a U-235 enrichment of 2.96%, 
2.64%, or 1.98% weight percent (DOE 2003 [DIRS 164970], pp. 19-23). The fuel was formed 
into dished, chamfered pellets that were then placed into Zircaloy-4 cladding to form fuel pins. 
The fuel pins were loaded into an assembly structure formed of stainless steels (304 or 304L or 
316), Zircaloy-4, and Inconel 718. During the TMI cleanup, debris from damaged fuel 
assemblies was packaged into canisters. The physical dimensions of the canisters are such that 
not more than a single, intact assembly could be stored in any one canister. The exact contents 
and fuel arrangement for each canister are unknown. For these criticality calculations, the 
amount of material in a TMI container is taken to be the fuel contained in one entire assembly at 
a beginning of life (BOL) enrichment of 2.96% (463.63 kg of U, including 13.72 kg of U-235, 
with a UO2 density of 95% theoretical density). Although structural materials are present in the 
canisters, it is conservative to neglect them. The presence of any assembly structural materials is 
ignored for this calculation (Assumption 3.3). 
Three types of canisters were used to package the TMI fuel debris: defueling, knockout, and 
filter. Based on the results of previous criticality analyses presented in the Intact and Degraded 
Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste 
Package document (BSC 2004 [DIRS 168935], p. 53), the knockout (KO) canister has been 
judged to be the most reactive and will be the only type considered in this calculation. The KO 
canister is fabricated from 300 series stainless steels, predominantly 304L. It is based around a 
14 in. Schedule 10 pipe with a reversed dish head forming the bottom of the canister. The top of 
the canister is a metal plate with penetrations and hardware for hydraulic loading and 
dewatering. The internal assembly for the KO canisters is designed to support five internal tubes 
filled with boron carbide (B4C) poison pellets: one large center A tube and four outer B rods. 
Seven intermediate support plates or spiders are held in place by the poison rods, which in turn 
rest on one bottom support plate. A cross-section of the KO canister is shown in Figure 5.1-4. 
Note that for these calculations, the fuel fills the available area of the KO canister up to a 
calculated height. The geometry of the TMI fuel and SNF contents has been simplified to adapt 
the geometry to right cylinders (Assumption 3.2). A comparison of the actual dimensions with 
the dimensions used in the MCNP cases can be found in Table 5.1-3. The fuel composition is 
calculated in SNF.xls, worksheet TMI.

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NOTE: Figure is not to scale. 
Figure 5.1-4 Radial View of DOE KO Canister Containing TMI-2 Fuel

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Table 5.1-3 TMI-2 Fuel and Packaging Dimensions and Materials 
Component Material a Parameter Actual Dimension 
(mm) a Dimension Used (mm) 
Fuel 
Outer Diameter 9.362 (0.3686 in.) b 
Length 11.049 (0.435 in.) b 
Geometry Dished / chamfered b 
Active Fuel Length / 
Rod 3601.7 (141.8 in.) b 
Number of Pellets / 
Rod ~ 326 (estimated) 
Number of Fuel 
Rods / Assembly 208 b 
Modeled as filling the 
available area of the KO 
canister for a height of 
59.609 cm, centered in 
the can. Calculations 
are shown in SNF.xls, 
worksheet TMI. 
Fuel Pellet UO2, 2.96 wt% 
U-235 
Mass Fuel / 
Assembly 
525.95 kg 
(1159.53 lb) b 525.95 kg 
Width 216.81 (8.536 in.) b 
Assembly Structure Various 
Length 4206.88 (165.625 in.)b 
Neglected per 
Assumption 3.3 
KO Canister Internals 
Inner Diameter 57.175 (calculated) 57.175 
Outer Diameter 73.025 (2.875 in.) 73.025 
Thickness 7.925 (0.312 in.) 7.925 
KO Canister Center 
Support 
SS 304L 
Length 3371.85 3371.850 
Inner Diameter 50.673 (calculated) c 50.673 
Outer Diameter 53.975 (2.125 in.) c 53.975 
Thickness 1.651 (0.065 in.) 1.651 
KO Canister Poison 
Tube A 
SS 316L 
Length Not explicitly given 3327.400 
Outer Diameter 49.657 (1.955 in.) KO Canister Poison 
Stack A 
B4C 
Length Not explicitly given 
Neglected for 
conservatism 
Inner Diameter 20.65 (calculated) 20.650 
Outer Diameter 33.35 33.350 
Thickness 6.35 6.350 
Length 3327.4 (min) 3327.400 
Number of Stacks 4 4 
KO Canister Poison 
Tubes B 
SS 316L 
Location 63.50 mm from center in 
x and y directions 
63.50 mm from center in 
x and y directions 
Outer Diameter 19.558 (0.770 in.) KO Canister Poison 
Stacks B 
B4C 
Length Not explicitly given 
Neglected for 
conservatism 
Outer Diameter Not explicitly given 
Thickness 12.7 
Number of Plates 7 
Intermediate Support 
Plates SS 304L 
Pitch 400.05 mm (15.75 in.) 
Neglected per 
Assumption 3.3 
Outer Diameter 340.52 340.520 
Thickness 31.75 31.750 Base Support Plate 
SS 304L 
Location 127 mm from bottom of 
KO canister 
127 mm from bottom of 
KO canister

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Table 5.1-3 TMI-2 Fuel and Packaging Dimensions and Materials (cont.) 
KO Canister 
Inner Diameter 342.9 (calculated) 342.900 
Outer Diameter 355.6 355.600 KO Canister Walls SS 304L 
Thickness 6.35 6.350 
Outer Diameter 355.6 335.600 
Thickness 101.6 101.600 KO Canister Top SS 304L 
Top Skirt Length 101.6 101.600 
Outer Diameter 355.6 335.600 
Thickness 9.525 9.525 KO Canister Bottom SS 304L 
Geometry Reverse dish Right cylinder with void 
region 
Outer Length 3803.65 3803.650 
Entire KO Canister ------------- 
Inner Cavity Length ~3475 (estimated) 3473.450 
DOE Standardized SNF Canister and Internals 
Inner Diameter 393.7 (calculated) 393.700 
Outer Diameter 406.4 (16 in.) 406.400 
Thickness 6.35 (0.25 in.) 6.350 
Centering Device Carbon Steel 
Length Not explicitly given 4117.000 
Inner Diameter 430 (min) d 438.150 Long Standardized 
SNF Canister SS 316L d 
Inner Length 4114.8 (min) d 4117.000 
a Source data from DOE 2003 [DIRS 164970], Sections 3.1, 3.2, and 4.1.4, except where specified. 
b Source data from DOE 2003 [DIRS 164970], Appendix B, column for Ref B-1. 
c The source document DOE 2003 [DIRS 164970] lists conflicting values of 2.125 in. for the inner diameter of poison 
tube A (Section 3.2.3.1) and 2.125 in. for the outer diameter of poison tube A (Figure 3). 2.125 in. was chosen as 
the outer diameter so that poison tube A would fit inside the center support tube. 
d Source data from DOE 1999 [DIRS 140225], Sections 3.2.3 (material) and 3.2.2 (dimensions) 
5.1.4.5 Advanced Test Reactor (ATR) Fuel and Canister Description 
The Advanced Test Reactor fuel is intended for disposal in the short DOE standardized SNF 
canister and the 5-DHLW/DOE SNF short waste package. The description of the ATR fuel is 
from the Specification for Advanced Test Reactor Mark VII Zone Loaded Fuel Elements report 
(INEEL 2003 [DIRS 171506]). The description of the loading of ATR fuel into the DOE 
standardized SNF canister is from Packaging Strategies for Criticality Safety for Other DOE 
Fuels in a Repository (DOE 2004 [DIRS 170071]). All ATR fuel and canister-related 
information is from these references unless otherwise noted. 
For this calculation, the ATR fuel elements denoted ATR-7F were used. This is the same fuel 
element used in Intact and Degraded Mode Criticality Calculations for the Codisposal of ATR 
Spent Nuclear Fuel in a Waste Package (BSC 2004 [DIRS 171926], Section 5.1.1) and is the 
ATR fuel element with the highest fissile loading (Paige, B.E. 1969 [DIRS 167978], pp. 29, 35, 
39, and 43). The specified fissile loading is 1075 g U-235 10 g (INEEL 2003 [DIRS 171506], 
p. 20). 
The ATR-7F fuel element consists of 19 concentric curved fuel plates held in place by aluminum 
6061 side plates and aluminum 356 end boxes to form a curved fuel element. Each plate consists 
of uranium aluminide (UAlx) fuel, with a uranium enrichment of 93  1 wt% U-235 and a

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uranium-to-aluminum ratio that varies from plate to plate. Some of the fuel plates also contain 
integrated B4C as a poison, but this is conservatively neglected for the current analysis. The fuel 
plates are clad in aluminum 6061. When the fuel plates are assembled, the angle of curvature of 
the fuel element is 45 degrees. Water gaps are present between each of the concentric plates. A 
diagram of the fuel element is provided in Figure 5.1-5, and dimensions are given in Table 5.1-4 
and Table 5.1-5. 
NOTE: Figure is not to scale. 
Figure 5.1-5 ATR Fuel Element 
Prior to disposal in the DOE standardized SNF canister, the upper and lower end boxes are 
removed. The remainder of the fuel element is placed into the DOE standardized SNF canister 
as follows (DOE 2004 [DIRS 170071], pp. 53-55). A basket assembly made of Ni-Gd alloy and 
SS 304L is placed into the SNF canister. This basket, shown in Figure 5.1-6, can accommodate 
10 ATR fuel elements. After the fuel elements are loaded, a circular plate of SS 304L is placed 
on top of the basket. A second basket assembly is then loaded into the SNF canister and filled 
with an additional 10 fuel elements. Dimensions of the basket assembly are given in Table 5.1- 
4. The fuel core composition is calculated in ATR.xls, worksheet Fuel Compositions.

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NOTE: Figure not to scale. 
Figure 5.1-6 Basket for ATR Fuel Elements

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Table 5.1-4 ATR Fuel and Packaging Dimensions and Materials 
Component Material Parameter Actual Dimension (mm) Dimension Used 
(mm) 
Fuel Element a 
Length 1200.404 (47.26 in.) to 
1238.504 (48.76 in.) 
1238.504 
(48.76 in.) 
Radius See Table 5.1-5 As in Table 5.1-5 
Width See Table 5.1-5 As in Table 5.1-5 
Fuel Plate Cores 
UAlx / 
Al powder/ 
B4C 
Thickness 0.508 (0.020 in.) c 0.508 (0.020 in.) c 
Length 1256.919 (49.485 in.) to 
1257.681 (49.515 in.) 
1257.300 
(49.500 in.) 
Radius See Table 5.1-5 As in Table 5.1-5 
Width See Table 5.1-5 As in Table 5.1-5 
Fuel Plate Cladding Aluminum 6061 
Thickness See Table 5.1-5 As in Table 5.1-5 
Fuel Plate Gap Void Radial Thickness See Table 5.1-5 As in Table 5.1-5 
Length 1256.919 (49.485 in.) to 
1257.681 (49.515 in.) 
1257.300 
(49.500 in.) 
Width 62.154 (2.447 in.) 62.154 
Inner Radius Not explicitly given 75.286 (2.964 in.) 
Side Plates Aluminum 6061 
Outer Radius Not explicitly given 140.030 (5.513 in.) 
SNF Internals b 
Inner Diameter 426.136 (calculated) 426.136 
Outer Diameter 429.25 429.250 Basket Sleeve SS 304L 
Length Not explicitly given 1260.000 
Length Not explicitly given 1260.000 
Thickness 9.525 9.525 
Horizontal Spacing 136.5 mm apart 136.500 mm apart 
Basket Plates C-4 Alloy 
(Ni-Gd Alloy) 
Vertical Spacing 101.1 mm apart 101.100 mm apart 
Outer Diameter Not explicitly given 438.150 
Spacer Plate SS 304L 
Thickness 6.35 to 9.525 9.525 
Inner Diameter 430.0 (min) d 438.150 d Short Standardized 
SNF Canister SS 316L d 
Inner Length 2540.0 (min) d 2575.000 d 
a Source data from INEEL 2003 [DIRS 171506], except where noted. 
b Source data from DOE 2004 [DIRS 170071], pp. 53-55, except where noted. 
c Source data from Paige, B.E. 1969 [DIRS 167978], p. 38 as referenced by p. 39. 
d Taken from Table 5.1-2.

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Table 5.1-5 Dimensions for Individual ATR Fuel Plates 
Plate # Inner Radius 
of Plate 
(cm) a 
Total Radial 
Thickness 
of Plate 
(cm) b 
Gap Between 
This Plate 
and Next 
(cm) b 
Arc Length of Fuel 
Plate Between 
Side Plates 
(cm) c 
Min. Arc Length 
From Side Plate to 
Fuel Core 
(cm) b 
Side N/A N/A 0.12954 
(0.051 in.) N/A N/A 
1 7.6581 
(3.015 in.) 
0.20320 
(0.080 in.) 
0.19812 
(0.078 in.) 
5.0641 
(1.994 in.) 
0.36830 
(0.145 in.) 
2 8.0594 
(3.173 in.) 
0.12700 
(0.050 in.) 
0.19812 
(0.078 in.) 
5.3793 
(2.118 in.) 
0.11430 
(0.045 in.) 
3 8.3845 
(3.301 in.) 
0.12700 
(0.050 in.) 
0.19812 
(0.078 in.) 
5.6347 
(2.218 in.) 
0.11430 
(0.045 in.) 
4 8.7097 
(3.429 in.) 
0.12700 
(0.050 in.) 
0.19812 
(0.078 in.) 
5.8901 
(2.319 in.) 
0.11430 
(0.045 in.) 
5 9.0348 
(3.557 in.) 
0.12700 
(0.050 in.) 
0.19812 
(0.078 in.) 
6.1455 
(2.419 in.) 
0.11430 
(0.045 in.) 
6 9.3599 
(3.685 in.) 
0.12700 
(0.050 in.) 
0.19812 
(0.078 in.) 
6.4009 
(2.520 in.) 
0.11430 
(0.045 in.) 
7 9.6850 
(3.813 in.) 
0.12700 
(0.050 in.) 
0.19812 
(0.078 in.) 
6.6563 
(2.621 in.) 
0.11430 
(0.045 in.) 
8 10.0101 
(3.941 in.) 
0.12700 
(0.050 in.) 
0.19812 
(0.078 in.) 
6.9116 
(2.721 in.) 
0.11430 
(0.045 in.) 
9 10.3353 
(4.069 in.) 
0.12700 
(0.050 in.) 
0.19812 
(0.078 in.) 
7.1670 
(2.822 in.) 
0.11430 
(0.045 in.) 
10 10.6604 
(4.197 in.) 
0.12700 
(0.050 in.) 
0.19558 
(0.077 in.) 
7.4224 
(2.922 in.) 
0.11430 
(0.045 in.) 
11 10.9830 
(4.324 in.) 
0.12700 
(0.050 in.) 
0.19558 
(0.077 in.) 
7.6757 
(3.022 in.) 
0.11430 
(0.045 in.) 
12 11.3055 
(4.451 in.) 
0.12700 
(0.050 in.) 
0.19558 
(0.077 in.) 
7.9291 
(3.122 in.) 
0.11430 
(0.045 in.) 
13 11.6281 
(4.578 in.) 
0.12700 
(0.050 in.) 
0.19558 
(0.077 in.) 
8.1825 
(3.221 in.) 
0.11430 
(0.045 in.) 
14 11.9507 
(4.705 in.) 
0.12700 
(0.050 in.) 
0.19558 
(0.077 in.) 
8.4358 
(3.321 in.) 
0.11430 
(0.045 in.) 
15 12.2733 
(4.832 in.) 
0.12700 
(0.050 in.) 
0.19558 
(0.077 in.) 
8.6892 
(3.421 in.) 
0.11430 
(0.045 in.) 
16 12.5959 
(4.959 in.) 
0.12700 
(0.050 in.) 
0.19558 
(0.077 in.) 
8.9426 
(3.521 in.) 
0.11430 
(0.045 in.) 
17 12.9184 
(5.086 in.) 
0.12700 
(0.050 in.) 
0.19558 
(0.077 in.) 
9.1959 
(3.620 in.) 
0.11430 
(0.045 in.) 
18 13.2410 
(5.213 in.) 
0.12700 
(0.050 in.) 
0.19558 
(0.077 in.) 
9.4493 
(3.720 in.) 
0.16510 
(0.065 in.) 
19 13.5636 
(5.340 in.) 
0.25400 
(0.100 in.) 
0.18542 
(0.073 in.) 
9.7027 
(3.820 in.) 
0.36830 
(0.145 in.) 
a According to (INEEL 2003 [DIRS 171506], p. 59), the inner radius of Plate 1 is 3.015 inches. All other radii are 
calculated from this value and the given thicknesses (ATR.xls, worksheet Dimensions). 
b Data from (INEEL 2003 [DIRS 171506]). 
c Calculated based on 45 degree fuel element, side plate thickness, and inner radii of fuel (see ATR.xls, worksheet 
Dimensions).

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5.1.4.6 Fort St. Vrain Fuel and Canister Description 
The MCNP input model representing Fort St. Vrain fuel in a canister utilized the existing MCNP 
input file can6 from the Criticality Potential of Intact DOE SNF Canisters in a Misloaded Dry 
Waste Package (BSC 2004 [DIRS 172201]) document as a starting point for the present 
calculations. The boundary conditions were the only changes made to the previously existing 
MCNP input files (see Section 5.2 for description of boundary conditions). 
Fort St. Vrain (FSV) fuel consists of small particles of uranium carbide, which are coated 
with pyrolytic carbon and silicon carbide and then bound in a carbonized matrix to form 
a solid substance. This solid material is shaped into fuel compacts, which are then 
placed in channels drilled into large hexagonal prisms made of graphite. Each graphite 
block, loaded with fuel compacts, comprises one fuel element (Taylor, L.L. 2001 [DIRS 
154726], Section 2.1.2). A radial view of a standard FSV fuel element is shown in Figure 
5.1-7. 
The FSV fuel elements are loaded into the DOE standardized SNF canister with no 
internal basket and no added neutron absorber. Each DOE standardized SNF canister 
can contain up to five FSV fuel elements, stacked axially. Dowels built into the fuel 
elements keep them aligned in the axial direction. 
The geometry of the FSV fuel and SNF contents has been simplified to adapt the geometry to 
right prisms and cylinders (Assumption 3.2). A comparison of the actual dimensions with the 
dimensions used in the MCNP cases can be found in Table 5.1-6. The fuel composition is 
calculated in SNF.xls, worksheet FSV. 
Figure 5.1-7 Radial View of Fort St. Vrain Canister

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Table 5.1-6 Fort St. Vrain Fuel and Packaging Dimensions and Materials 
Component Material a Parameter Actual Dimension 
(mm) a 
Dimension Used 
(mm) 
Outer Diameter 12.5 12.500 
Length 49.276 49.276 Fuel Compact 
Uranium or 
thorium carbide 
matrix Number per Fuel 
Channel 14 or 15 15 
Diameter 12.70 12.700 
Length b ~ 785.368 785.368 Fuel Channels Void 
Number 210 210 
Diameter 12.70 12.700 Top Plug for Fuel 
Channels 
H-327 or H-451 
graphite Length 12.7 12.700 
Diameter, 
innermost ring 12.700 12.700 
Diameter, 
standard 15.875 15.875 
Length Not explicitly given 792.988 
Void / Coolant 
Channels Void 
Number 108 108 
Diameter 12.7 12.700 
Length Not explicitly given 792.988 Burnable Poison 
Channels Void c 
Number 6 6 
Diameter 41.275 41.275 Central Fuel 
Handling Hole Void 
Length 381 381.000 
Diameter (across 
flats) 360.60 360.600 
Length 792.988 792.988 Fuel Element H-327 or H-451 
graphite 
Channel Pitch 18.796 18.796 
Inner Diameter 430 (min) d 438.150 d Long Standardized 
SNF Canister SS 316L d 
Inner Length 4114.8 (min) d 4117.000 d 
a Source data from (Taylor, L.L. 2001 [DIRS 154726], Section 2.1.2), except where noted. 
b The fuel channel length is the length of the non-plugged hole as measured from the top of the fuel element. 
c No credit is taken for burnable poison; these channels are modeled as empty. 
d Source data from Table 5.1-3. 
5.1.4.7 FFTF Fuel and Canister Description 
The MCNP input model representing FFTF fuel in a canister utilized the existing MCNP input 
file can2 from the Criticality Potential of Intact DOE SNF Canisters in a Misloaded Dry Waste 
Package (BSC 2004 [DIRS 172201]) document as a starting point for the present calculations. 
The boundary conditions were the only changes made to the previously existing MCNP input 
files (see Section 5.2 for description of boundary conditions).

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There are four distinct types of FFTF standard driver fuel. The most reactive type (based on 
prior calculations) is the Type 4.1 driver fuel (CRWMS M&O 1999 [DIRS 102842], Sections 
6.1.1 and 6.1.2). For this fuel type the fuel pin contains a mixed oxide fuel region of 70.72 wt% 
UO1.96 and 29.28 wt% PuO1.96 formed into dished pellets. Similar pellets made of natural UO2 
serve as an insulator at each end of the fuel region. On the outer ends of the UO2 insulator 
regions are regions of Inconel 600 reflector. Above the top reflector is a spring made of SS 302 
and a plenum of SS 316. The cladding and end caps are fabricated from SS 316 (INEEL 2002 
[DIRS 158820], pp.15-17). 
The FFTF standard driver fuel assembly (DFA) is hexagonally shaped and contains 217 
cylindrical fuel pins as described above. Each fuel pin is held in place in a triangular-pitch array 
by a wire-wrapped spacer. The DFA also includes the inlet nozzle, orifice plates, neutron shield 
assembly, diffuser block, load pads, and handling socket (INEEL 2002 [DIRS 158820], Figure 
3). 
In addition to the whole DFAs, some standard and non-standard DFAs have been disassembled 
into fuel pins. These fuel pins range in diameter from 5.8 mm to 12.8 mm, and their 
compositions span a wide range of enrichments. The fuel pins from the disassembled DFAs are 
packaged into an Ident-69 container. Pins from different DFAs may be mixed in one container. 
The Ident-69 container with the highest Pu loading is Storage Serial No. ID69-033 (INEEL 2002 
[DIRS 158820], Table A-2). The contents of this container are considered to be representative of 
the fuel type, and are chosen for use in this calculation. The 151 fuel pins stored in this container 
are 6.9 mm in diameter and had an initial enrichment of 31.2 wt% Pu/(Pu+U) with a surrounding 
insulator of depleted uranium oxide. For conservatism, this calculation replaces the depleted 
uranium oxide with natural uranium oxide. 
The most reactive version of the Ident-69 container is the compartmented model (INEEL 2002 
[DIRS 158820], p. 18), which can contain up to 217 fuel pins. The Ident-69 container is 
comprised of 5 in. SS 304L pipe with a transition to 2.5 in. pipe at one end of the container. 
Inside the pipe are a cylindrical central compartment and six radial compartments (BSC 2002 
[DIRS 164418], Section 5.1.1). The fuel pins are contained within these compartments and 
supported on a grid plate. For this calculation, the Ident-69 canister internals are neglected 
(Assumption 3.3) and the fuel pins are conservatively modeled as a minimum-pitch hexagonal 
array in the center of the canister. The FFTF canister internals consist of basket with a cylindrical 
center tube and five divider plates extending radially from the center to the DOE standardized 
SNF canister inner wall (BSC 2002 [DIRS 164418], Sections 5.1.1 and 5.1.2). 
This calculation considers an SNF basket made of Ni-Gd alloy and five DFAs loaded in the 
radial positions of the DOE standardized SNF canister. The geometry of the FFTF fuel and SNF 
contents has been simplified to adapt the geometry to right prisms and cylinders (Assumption 
3.2). A comparison of the actual dimensions with the dimensions used in the MCNP cases can 
be found in Table 5.1-7. The FFTF fuel compositions are calculated in SNF.xls, worksheet 
FFTF.

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Table 5.1-7 FFTF Fuel and Packaging Dimensions and Materials 
Component Material a Parameter Actual Dimension 
(mm) a 
Dimension Used 
(mm) 
FFTF Type 4.1 Driver Fuel Pins 
Outer Diameter 4.9403 4.9404 
Fuel 
70.72 wt% UO1.96 
and 29.28 wt% 
PuO1.96 Length 914.4 914.4000 
Outer Diameter 4.9403 4.9404 
Length, Top 20.32 20.3200 Insulator Natural UO2 
Length, Bottom 20.32 20.3200 
Outer Diameter 4.8133 4.9404 
Length, Top 144.78 144.7800 Reflector Inconel 600 
Length, Bottom 144.78 144.7800 
Wire Diameter 0.8052 
Spring SS 302 
Coiled Length 125.5 
Neglected d 
Outer Diameter 4.9022 
Wall Thickness 0.1397 Plenum SS 316 
Length 862.1 
Neglected d 
Inner Diameter 5.08 5.0800 
Outer Diameter 5.842 5.84520 Cladding SS 316 
Thickness 0.381 0.3810 
Outer Diameter 5.842 5.8420 
Length, Top 104.6 104.6000 End Caps SS 316 
Length, Bottom 35.6 35.6000 
Entire Fuel Pin -------------- Length 2372.36 2372.3600 
FFTF Type 4.1 Driver Fuel Assembly 
Length 2372.36 2372.3600 
Location 
(center of fuel) 
1663.7 mm from 
bottom of DFA 
1663.7 mm from 
bottom of DFA 
Pin Pitch 7.2644 7.2644 
Fuel Pin Region -------------- 
Number of Pins 217 217 
Diameter 1.4224 Wire-Wrapped 
Spacers SS 316 
Axial Pitch 304.80 
Neglected d 
Length 477.52 (calculated) 477.5200 
Lower Assembly 
Region -------------- Diameter (across 
points) 138.1125 (max) 134.1810 
Length 807.72 (calculated) 807.7200 
Upper Assembly 
Region -------------- Diameter (across 
points) 138.1125 (max) 134.1810 
Length 3657.6 3657.6000 
Diameter (across 
flats) 116.205 116.2050 
Diameter (across 
points) 131.064 
134.1810 
(calculated from 
diameter across flats) 
Entire DFA SS 316 
Wall Thickness 3.048 3.0480

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Table 5.1-7 FFTF Fuel and Packaging Dimensions and Materials (cont.) 
Component Material a Parameter Actual Dimension 
(mm) a Dimension Used (mm) 
Fuel Pins in Ident-69 Container 
Outer Diameter Not explicitly given 5.5558 e,f Fuel 68.8 wt% UO1.96 
and 31.2 wt% 
PuO1.96 
Length Not explicitly given 914.4000 e 
Insulator Natural UO2 g Outer Diameter Not explicitly given 5.5558 e 
Reflector Inconel 600 Outer Diameter Not explicitly given 5.5558 e 
Inner Diameter Not explicitly given 6.1380 e 
Outer Diameter 6.9 6.9000 
Cladding SS 316 
Thickness Not explicitly given 0.3810 e 
End Caps SS 316 Outer Diameter Not explicitly given 6.9000 e 
Remainder of Fuel 
Pin ------------ All other 
dimensions 
Not explicitly given Assumed same as Type 
4.1 fuel e 
Ident-69 Container 
Length Not explicitly given 
Inner Diameter 41.402 
Center Tube Not explicitly given 
Outer Diameter 44.450 
Neglected d 
Length Not explicitly given 
Divider Plates Not explicitly given 
Thickness 1.524 b Neglected d 
Length 3225.8 b 3225.8000 
Inner Diameter 135.763 135.7630 Ident-69 Canister, 
Top Portion SS 304L 
Outer Diameter 141.30 141.3000 
Length 431.8 431.8000 
Inner Diameter Not explicitly given 135.7630 Ident-69 Canister, 
Bottom Portion SS 304L 
Outer Diameter 73.02 141.3000 
Inner Length Not explicitly given 3652.0630 
Outer Length 3657.6 3657.6000 Entire Ident-69 
Canister SS 304L 
Number of Pins 151 151 
DOE Standardized SNF Canister 
Length 4125 b 4117.0000 
Inner Diameter 153 b 153.0000 Center Tube Ni-Gd alloy h 
Outer Diameter 173 b 173.0000 
Length 4125 b 4117.0000 
Divider Plates Ni-Gd alloy h 
Thickness 10 b 10.0000 
Inner Diameter 430.0 (min) c 438.1500 Long Standardized 
SNF Canister SS 316L c 
Inner Length 4114.8 (min) c 4117.0000 
a Source data from INEEL 2002 [DIRS 158820], Section 3 unless stated otherwise. 
b Source data from BSC 2002 [DIRS 164418], Section 5.1.1 and Section 5.1.2. 
c Source data from Table 5.1-3. 
d See Assumption 3.3. Spring region and plenum were represented as a void region 987.56 mm long. 
e These fuel pins were assumed to be similar to Type 4.1 fuel pins (Assumption 3.5). 
f Value calculated in SNF.xls, worksheet FFTF. 
g Specified in INEEL 2002 [DIRS 158820], Table A-2 as depleted UO2. Natural UO2 was substituted. 
h Specified in BSC 2002 [DIRS 164418], Section 5.1.2 as SS 316L doped with GdPO4. Ni-Gd alloy was substituted.

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5.1.4.8 Shippingport LWBR Fuel and Canister Description 
As a starting point for the present Shippingport LWBR calculations, the existing MCNP input 
file sldds78 was utilized from the Canister Handling Facility Criticality Safety Calculations 
(BSC 2004 [DIRS 167614], which initially originated from the Intact and Degraded Criticality 
Calculations for the Codisposal of Shippingport LWBR Spent Nuclear Fuel in a Waste Package 
(CRWMS M&O 2000 [DIRS 151722]) document. 
The fuel consists of a binary matrix of UO2-ThO2 (DOE 1999 [DIRS 105007], p. 16). The most 
reactive fuel consists of 5.202 wt% of U-233 and U-235 in the heavy metal and has a density of 
9.71 g/cc (CRWMS M&O 2000 [DIRS 151722], Table 5-2). The fuel pellet OD is 0.252 inch 
(DOE 1999 [DIRS 105007], Table 3-5) or 0.64 cm and the pin pitch is 0.369 inch (DOE 1999 
[DIRS 105007], Figure 3-3) or 0.937 cm. The cladding OD is 0.3063 inch and is 0.02217 inch 
thick (DOE 1999 [DIRS 105007], Table 3-8). There are 619 fuel rods per assembly (DOE 1999 
[DIRS 105007], p. 16). 
The DOE SNF canister is a right circular cylinder pipe made of stainless steel (Type 316L or 
UNS S31603) with an outside diameter of 457.2 mm (18 in.) and a wall thickness of 9.525 mm 
(0.375 in.). The nominal internal length of the DOE SNF canister reserved for fuel loading is 
411.7086 cm (162.09 in.). The top and bottom carbon steel (ASME SA-36) impact plates are 
50.8 mm (2.0 in.) thick at the centers (CRWMS M&O 2000 [DIRS 151722], p. 18). Dished 
heads seal the ends of the DOE SNF canister. The DOE SNF canister pipe extends several inches 
beyond the dished heads on each end to give a maximum external length of 456.9968 cm (179.92 
in.). Each DOE SNF canister will also contain a rectangular basket structure to hold the 
Shippingport LWBR fuel assembly (CRWMS M&O 2000 [DIRS 151722], p. 18). The basket 
plates are made of 9.5 mm (0.374 in.) thick stainless steel (Type 316L or UNS S31603) and the 
inner widths of the plates are 295 mm and 257 mm, respectively. Finally, a boxed shaped spacer 
will fit inside the basket to elevate the SNF above the canister bottom (CRWMS M&O 2000 
[DIRS 151722], p. 18). 
5.1.4.9 TRIGA Fuel and Canister Description 
The MCNP input model representing TRIGA fuel in a canister utilized the existing MCNP input 
file can7 from the Criticality Potential of Intact DOE SNF Canisters in a Misloaded Dry Waste 
Package (BSC 2004 [DIRS 172201]) document as a starting point for the present calculations. 
The boundary conditions were the only changes made to the previously existing MCNP input 
files (see Section 5.2 for description of boundary conditions). 
For this calculation, the TRIGA fuel type FLIP in a standard-streamline stainless steel rod was 
used. This is the same fuel rod used in TRIGA Fuel Phase I and II Criticality Calculation 
(CRWMS M&O 1999 [DIRS 135852], Section 5.1.4). 
TRIGA fuel is packaged as loose fuel elements held in place by support baskets. The FLIP fuel 
elements are made of uranium and zirconium hydride (approximately 91.5 wt% zirconium 
hydride (H/Zr ratio of 1.6) mixed with uranium of 70% U-235 enrichment). A hole is drilled 
through the center of the fuel, and a zirconium rod is inserted in the hole. The fuel and 
zirconium rod are contained in a SS 316 cladding. Graphite reflectors are present above and

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below the fuel inside the clad. Zirconium end fittings are swaged onto the ends of the fuel pin 
(DOE 1999 [DIRS 103891], Section 3.2). 
The intact TRIGA fuel rods are placed into the DOE standardized SNF canister by first loading a 
base plate into the DOE standardized SNF canister, followed by a support basket. The support 
basket is made of SS 316L and holds 37 fuel pins, one per tube. Two more baskets are loaded 
into the DOE standardized SNF canister (with a base plate under each one) for a total of 111 fuel 
pins per DOE standardized SNF canister. No base plate is placed on top of the three layers 
(CRWMS M&O 1999 [DIRS 135852], Section 5.1.3 and Attachment IV). A radial view of the 
basket is shown in Figure 5.1-8. 
Figure 5.1-8 Basket Assembly for TRIGA SNF Canister 
This calculation considers a support basket made of Ni-Gd alloy. The geometry of the TRIGA 
fuel and SNF contents has been simplified to adapt the geometry to right cylinders (Assumption 
3.2). A comparison of the actual dimensions with the dimensions used in the MCNP cases can 
be found in Table 5.1-8. The TRIGA fuel composition is calculated in SNF.xls, worksheet 
TRIGA.

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5.1-8 TRIGA Fuel and Packaging Dimensions and Materials 
Component Material a Parameter Actual Dimension 
(mm) a 
Dimension Used 
(mm) 
Outer Diameter 5.715 5.715 
Zr Core Zirconium 
Length 381.0 381.000 
Inner Diameter 6.35 6.350 
Outer Diameter 36.449 36.449 Fuel U-ZrH1.6 
Length 381.0 381.000 
Outer Diameter 36.449 36.449 
Upper Reflector Graphite 
Length 65.024 65.024 
Outer Diameter 36.449 36.449 
Lower Reflector Graphite 
Length 94.488 94.488 
Inner Diameter 36.525 36.525 
Outer Diameter 37.541 37.541 
Inner Length Not given 540.512 
Cladding SS 304L 
Outer Length 753.872 600.512 b 
Inner Diameter 49.3 d 49.300 
Outer Diameter 60.3 d 60.300 
Length 836 d 836.000 
Support Basket 
(each tube) SS 316L c 
Pitch 60.3 d 60.300 
Thickness 7.9 d Neglected e Basket Support 
Bracket 
SS 316L 
Length 150 d Neglected e 
Diameter 426 d 438.150 f 
Base Plate SS 316L 
Thickness 9.5 d 9.500 
Void over Top Layer Void Length 10.5 d 38.500 g 
Inner Diameter 430 (min)h 438.150 Short Standardized 
SNF Canister SS 316L h 
Inner Length 2540 (min) h 2575.000 
a Source data from DOE 1999 [DIRS 103891], Section 3.2 and Section 3.2.1, except where noted. 
b The end fittings were modeled as a right circular cylinder with the same volume and mass as the end plugs, and an 
adjacent region of void for the remainder of the fuel pin length (see Assumption 3.2). 
c Specified in CRWMS M&O 1999 [DIRS 135852], Section 5.2 as SS 316L with a coating of advanced absorber 
matrix. Ni-Gd alloy with no additional absorber was substituted. 
d Source data from CRWMS M&O 1999 [DIRS 135852], Attachment IV. 
e See Assumption 3.3. 
f Diameter increased to simplify modeling. 
g Length of void increased to match short DOE standardized SNF canister interior length. 
h Taken from Table 5.1-2

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Per Assumption 3.4, Polysiloxane fluid was chosen as an alternate moderator material (in the 
event hydraulic fluid/oil leak from a handling crane). The chemical formula for this fluid is 
CH3 CH3 
CH3 
CH3 
CH3 
CH3 
CH3 
CH3 
Si O Si Si O 
4 
Source: (Gelest 2004 [DIRS 169915], p. 11) 
Polysiloxane fluid was modeled with a density of 0.9 g/cm3 (Gelest 2004 [DIRS 169915], p. 11).

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5.2 CRITICALITY CALCULATIONS 
5.2.1 Dry Single DOE Canisters 
MCNP calculations were performed for the various DOE fuel types (see Section 5.1.2) modeled 
as single canisters. The canisters are modeled as dry, since they will remain sealed during 
handling in the CHF (see Section 1). The impact of no poison loading present in the basket 
structure (due to manufacturer errors, etc.) was also investigated for the applicable DOE SNF 
canisters. The results are presented in Section 6.1. 
5.2.2 Dry Arrays of DOE Canisters 
The DOE SNF canisters were modeled in an array configuration to investigate a minimum 
separation distance for the staging racks in the CHF. In order to find the most reactive 
configuration in the CHF, various potential outside reflector materials were evaluated. The 
reflector materials considered are void/air and water. Previous studies have shown that higher 
keff is produced with canister surrounded by air than concrete for DOE fuel in canisters when 
canister separation is equal to or less than 60 cm for array of canisters (there are some exceptions 
at 60 cm distance) (BSC 2004 [DIRS 171589], Table 5.2-1). Consequently concrete was not 
considered as a reflector material. The reflector is modeled with a 30 cm width/height in the 
radial and axial directions. An infinite array of casks was also evaluated to investigate neutronic 
isolation. The results are presented in Section 6.2. 
5.2.3 Category 1 and 2 Event Sequences 
The Category 1 and Category 2 event sequences applicable to the CHF have been identified in 
the Categorization of Event Sequences for License Application document (BSC 2004 [DIRS 
167268], Section 7). Only Category 2 events have been identified for the CHF and are presented 
in Table 5.2-1. The supporting calculations for the event sequences are provided in Section 6.3. 
They included rearranged fuel and moderator (water) flooded canisters for defense in depth. 
Also, hydraulic fluid/oil was also considered as a moderator (because it may leak from a 
handling crane) for defense in depth. 
Table 5.2-1 Category 2 Event Sequences for CHF 
Event Sequence 
identifier 
Criticality Event 
Description Reference 
2-01 Drop of a transportation cask without 
impact limiters in the CHF 
2-03 
Drop of inner lid of a transportation 
cask, MSC, or WP into a transportation 
cask, MSC, or WP in the CHF 
2-07 Drop of a canister during transfer by 
crane 
2-08 Drop of handling equipment onto a 
canister 
2-10 Drop of unsealed WP in CHF 
2-11 Drop of a WP with a known closure 
defect 
2-24 
Drop or collision of handling equipment 
into an open WP loaded with DOE 
canisters 
BSC 2004 [DIRS 167268], Section 7

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6. RESULTS AND CONCLUSIONS 
This section presents the results of the criticality calculations and makes recommendations for 
additional criticality safety design features as appropriate. The outputs presented in this 
document are all reasonable compared to the inputs and the results are suitable for the intended 
use. The uncertainties are taken into account by consistently using a conservative approach, 
which is the result of the methods and assumptions described in Sections 2 and 3, respectively. 
6.1 SINGLE DOE CANSITERS 
Table 6.1-1 presents the keff values along with their standard deviation (St. Dev.) of the various 
DOE fuel types with and without fixed neutron poison present. For the scenarios without fixed 
neutron absorber present (for defense in depth to account for manufacturer errors, etc.), the 
basket structures were replaced by void in the MCNP model. It can be seen from Table 6.1-1 that 
the neutron poison in a dry environment has an insignificant effect on the reactivity of the DOE 
SNF canisters. The changes in keff are due to the presence or absence of additional material in the 
canister. Overall, all of the keff values are much below the USL and poses no criticality concerns. 
Table 6.1-1 keff Values of Dry Single Canisters With and Without Poison 
With neutron poison Without neutron poison 
DOE Fuel 
keff St. Dev. MCNP files keff St. Dev. MCNP files 
FFTF 0.31462 0.00029 fftf5sG, 
fftf5sG.out 0.29702 0.00030 fftf5sG0, 
fftf5sG0.out 
Fermi 0.32351 0.00032 fermi-s, 
fermi-s.out 0.26127 0.00024 fermi-s0, 
fermi-s0.out 
Triga 0.38785 0.00075 triga-s, 
triga-s.out 0.40668 0.00083 triga-s0, 
triga-s0.out 
Fort St. Vrain N/A N/A N/A 0.03191 0.00018 fsv-s0, 
fsv-s0.out 
TMI N/A N/A N/A 0.21680 0.00031 
tmi-k0, 
tmi-k0.out & 
tmi-k0.out1 
ATR 0.02170 0.00003 atr-s, 
atr-s.out 0.02010 0.00002 atr-s0, 
atr-s0.out 
Shippingport 
PWR N/A N/A N/A 0.04193 0.00005 sh-ps0, 
sh-ps0.out 
Shippingport 
LWBR N/A N/A N/A 0.11399 0.00013 sh-ls0, 
sh-ls0.out 
N-Reactor 
(Mark 1A) N/A N/A N/A 0.23368 0.00028 nr1A-s0, 
nr1A-s0.out 
N-Reactor 
(Mark IV) N/A N/A N/A 0.28555 0.00032 nr4B-s0, 
nr4B-s0.out

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6.2 ARRAYS OF DOE CANISTERS 
Table 6.2-1 presents the keff values of an infinite (x and y directions) array of dry DOE SNF 
canisters surrounded by air. The canisters are either almost touching (0.001 cm separation) or are 
placed 30 cm apart. It can be seen that FFTF, Fermi, and TRIGA fuel types exceed the USL 
when they are almost touching. Consequently, the DOE fuel canisters must be placed 30 cm 
apart in the CHF staging racks to remain below USL. There are no limitations to the array size. 
Also note that these calculations do not take credit for the fixed neutron absorbers, except where 
noted. As mentioned in Section 6.1, the neutron absorber contribution in a dry environment is 
insignificant as demonstrated in Table 6.1-1. 
The same array configurations were also investigated with water as an outside reflector. The 
results are presented in Table 6.2-2 and it can be seen that water presence outside the canister 
reduces the reactivity of the canister array. 
Table 6.2-1 keff Values of Dry Canisters in an Infinite Array W/o Poison (Air Surrounding Canisters) 
0.001 cm separation between 
canisters 
30 cm separation between 
canisters DOE Fuel 
keff St. Dev. MCNP files keff St. Dev. MCNP files 
1.10418 0.00069 fftf5a0A, 
fftf5a0A.out FFTF 
1.06506 a 0.00066 fftP5a0A, 
fftP5a0A.out 
0.58884 0.00053 fft30a0A, 
fft30a0A.out 
1.05684 0.00064 fermiAa0, 
fermiAa0.out Fermi 
0.97844 a 0.00052 fermPAa0, 
fermPAa0.out 
0.54998 0.00046 fer30a0A, 
fer30a0A.out 
1.23663 0.00090 trigaAa0, 
trigaAa0.out Triga 
0.68213 a 0.00076 trigPAa0, 
trigPAa0.out 
0.83546 0.00104 tri30a0A, 
tri30a0A.out 
Fort St. Vrain 0.72281 0.00063 fsvAa0, 
fsvAa0.out 0.36730 0.00060 fsv30Aa0, 
fsv30Aa0.out 
TMI 0.35692 0.00040 tmiKAa0, 
tmiKAa0.out 0.26954 0.00033 tmi30Aa0, 
tmi30Aa0.out 
ATR 0.43270 0.00054 atrAa0, 
atrAa0.out 0.07974 0.00012 atr30Aa0, 
atr30Aa0.out 
Shippingport 
PWR 0.50674 0.00046 shPAa0, 
shPAa0.out 0.13513 0.00018 shP30Aa0, 
shP30Aa0.out 
Shippingport 
LWBR 0.43852 0.00038 shLAa0, 
shLAa0.out 0.22163 0.00024 shL30Aa0, 
shL30Aa0.out 
N-Reactor 
(Mark 1A) 0.38492 0.00033 nr1AaA0, 
nr1AaA0.out 0.32870 0.00033 nr1A30a0, 
nr1A30a0.out 
N-Reactor 
(Mark IV) 0.40684 0.00041 nr4BaA0, 
nr4BaA0.out 0.36811 0.00036 nr4B30aA, 
nr4B30aA.out 
a Calculated as designed with the neutron poison present.

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Table 6.2-2 keff Values of Dry Canisters in an Infinite Array W/o Poison (Water Surrounding Canisters) 
0.001 cm separation between 
canisters 
30 cm separation between 
canisters DOE Fuel 
keff St. Dev. MCNP files keff St. Dev. MCNP files 
0.75006 0.00074 fftf5a0, 
fftf5a0.out FFTF 
0.66975 a 0.00067 fftfP5a0, 
fftfP5a0.out 
0.48338 0.00061 fft30a0, 
fft30a0.out 
0.78956 0.00073 fermi-a0, 
fermi-a0.out Fermi 
0.50959 a 0.00050 fermiPa0, 
fermiPa0.out 
0.49676 0.00068 ferm30a0, 
ferm30a0.out 
0.82504 0.00096 triga-a0, 
triga-a0.out Triga 
0.49261 a 0.00071 trigaPa0, 
trigaPa0.out 
0.58690 0.00090 trig30a0, 
trig30a0.out 
Fort St. Vrain 0.37636 0.00059 fsv-a0, 
fsv-a0.out 0.19520 0.00047 fsv30a0, 
fsv30a0.out 
TMI 0.26660 0.00038 tmiK-a0, 
tmiK-a0.out 0.23468 0.00034 tmiK30a0, 
tmiK30a0.out 
ATR 0.36625 0.00060 atr-a0, 
atr-a0.out 0.17875 0.00042 atr30a0, 
atr30a0.out 
Shippingport 
PWR 0.24891 0.00050 shP-a0, 
shP-a0.out 0.13348 0.00039 shP30a0, 
shP30a0.out 
Shippingport 
LWBR 0.32144 0.00057 shL-a0, 
shL-a0.out 0.20036 0.00044 shL30a0, 
shL30a0.out 
N-Reactor 
(Mark 1A) 0.42652 0.00044 nr1A-a0, 
nr1A-a0.out 0.33741 0.00040 nr1A30aA, 
nr1A30aA.out 
N-Reactor 
(Mark IV) 0.45035 0.00048 nr4B-a0, 
nr4B-a0.out 0.37214 0.00046 nr4B30a0, 
nr4B30a0.out 
a Calculated as designed with the neutron poison present. 
6.3 CATEGORY 1 AND 2 EVENT SEQUENCES 
Table 6.3-1 presents the evaluation of the Category 1 and 2 event sequences for the CHF. As 
mentioned in Section 5.2.3, only Category 2 events have been identified for the CHF.

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Table 6.3-1 Evaluation of Category 2 Event Sequences for CHF 
Event 
Sequence 
identifier a 
Criticality Event 
Description 
Criticality Safety 
Evaluation 
2-01 
Drop of a transportation 
cask without impact 
limiters. 
Per Assumption 3.1, the DOE canisters will not breach due a drop 
and, consequently, a moderator cannot intrude into the canister to 
make the system more reactive. Also, see supporting defense in 
depth calculations below. 
2-03 
Drop of inner lid of a 
transportation cask, MSC, 
or WP into a transportation 
cask, MSC, or WP. 
Table 6.3-2 presents the results of rearranged DOE fuel inside of the 
canisters and this scenario does not pose any criticality concerns. 
Defense in depth evaluations have also been performed below for 
when moderators are present. 
2-07 Drop of a canister during 
transfer by crane. See evaluation for event sequence 2-01 
2-08 Drop of handling equipment 
onto a canister. See evaluation for event sequence 2-03 
2-10 Drop of unsealed WP. See evaluation for event sequence 2-01 
2-11 Drop of a WP with a known 
closure defect. See evaluation for event sequence 2-01 
2-24 
Drop or collision of 
handling equipment into an 
open WP loaded with DOE 
canisters 
See evaluation for event sequence 2-03 
a BSC 2004 [DIRS 167268], Section 7 
Loss of geometric structure, due to a drop, inside the canister was studied for the most reactive 
DOE SNF under dry conditions without any fixed poison present. Since the DOE canisters will 
remain unopened, the dry conditions represent undamaged canister containment. Further, single 
canisters were considered since it is not credible that two or more canisters would be dropped 
near each other due to that movement of canisters is limited to one canister at a time (BSC 2004 
[DIRS 168992], p. 2-3). Table 6.3-2 presents the calculated keff values for the DOE fuel types 
considered. The results show that a loss of geometric structure (e.g., broken/lost support tubes) 
inside the DOE canisters under dry conditions does not pose a criticality concern. 
Table 6.3-2 keff Values of Single Dry Canisters W/o Poison and Geometric Structure (Air Surrounding 
Canisters) 
Dry Single Canisters 
DOE Fuel 
keff St. Dev. MCNP files 
FFTF 0.31558 0.00031 fftf5sGr, fftf5sGr.out 
Fermi 0.27196 0.00028 fermiGs0, fermiGs0.out 
TRIGA 0.49726 0.00097 trigaGs0, trigaGs0.out 
Fort St. Vrain 0.07931 0.00031 fsvGs0, fsvGs0.out 
TMI 0.27964 0.00031 tmi-s, tmi-s.out 
For defense-in-depth, flooded conditions are considered for TRIGA fuel (most reactive fuel 
based on the scenarios described above) in the event the drop will puncture the canister 
containment and water is present in the facility. Table 6.3-3 presents the keff values as a function

Licensing Calculation 
Title: Canister Handling Facility Criticality Safety Calculations 
Document Identifier: 190-00C-CH00-00100-000-00B Page 42 of 50 
of water height. Hydraulic fluid/oil was also considered as a potential moderator in case it may 
leak from a handling crane. The results show that the moderator height, for defense in depth, 
should be controlled to 55 cm from the bottom of the canisters in order to be within the USL. 
Note that these conclusions are based on non-credit for neutron poison, which is very 
conservative. Table 6.3-3 shows that a fully flooded TRIGA canister with neutron poison 
present is well below USL. It can also be seen from Table 6.3-3 that the neutron poison is very 
effective in a wet environment. Note that the limiting moderator height could vary for other DOE 
SNF canisters. 
Table 6.3-3 keff Values of Flooded TRIGA Canister W/o Poison and Geometric Structure (Air Surrounding 
Canister) 
TRIGA fuel Moderator 
(water) height 
(cm) keff St. Dev. MCNP files 
192 a 1.13643 0.00095 trigGs0w, trigGs0w.out 
192 a, b 0.55482 0.00085 triPGs0w, triPGs0w.out 
192 a, b (oil) 0.51105 0.00084 triPGs0o, triPGs0o.out 
128 1.13199 0.00100 triGs0w1, triGs0w1.out 
64 1.01983 0.00110 triGs0w4, triGs0w4.out 
60 0.97158 0.00109 triGs0w7, triGs0w7.out 
55 0.87970 0.00104 triGs0w6, triGs0w6.out 
55 (oil) 0.82499 0.00113 triGs0o6, triGs0o6.out 
50 0.74743 0.00106 triGs0w5, triGs0w5.out 
a Fully flooded 
b Calculated with neutron poison present 
6.4 CONCLUSIONS AND RECOMMENDATIONS 
The processes for the CHF have been evaluated for criticality safety for normal operations, 
Category 1 and 2 event sequences. The results presented in this document lead to the following 
conclusions and recommendations: 
 All of the nine DOE SNF types considered (see Section 5.1.2) are safely below USL as single 
canisters under dry normal operations. 
 When the DOE SNF canisters are placed in a dry infinite array configuration, the canisters 
needs to be separated by 30 cm (canister surface to canister surface) to safely be below USL. 
 The identified evaluated Category 2 events for the CHF were found criticality safe under 
nominal conditions. 
 For defense in depth:

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Title: Canister Handling Facility Criticality Safety Calculations 
Document Identifier: 190-00C-CH00-00100-000-00B Page 43 of 50 
1. an infinite array of DOE SNF canisters placed 30 cm apart without neutron absorber 
(to account for manufacturer errors, etc.) under dry conditions is criticality safe. 
2. a fully flooded single canister (TRIGA fuel) with neutron absorber is criticality safe. 
3. an infinite array of DOE SNF canisters with dry inside and wet outside is criticality 
safe (with and without neutron absorber). 
In summary, normal operations in CHF prove to be criticality safe for all DOE SNF types 
considered in this calculation. The DOE SNF canisters can be placed in an infinite array size as 
long as the spacing between the surfaces of the canisters is 30 cm or greater. A fully flooded 
scenario does not pose a criticality concern when taking credit for the neutron poison. 
Consequently, the moderator controlled areas of the CHF, shown in Attachment III, are provided 
for defense in depth. It should also be mentioned that design requirements, as indicated by 
Assumption 3.1, need to be implemented to limit lift heights in the CHF to ensure no breaching 
of DOE SNF canisters. Further, it will be required when selecting the design basis hydraulic 
fluid (Assumption 3.4) that it is a less effective moderator than water.

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Document Identifier: 190-00C-CH00-00100-000-00B Page 44 of 50 
7. REFERENCES 
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Title: Canister Handling Facility Criticality Safety Calculations 
Document Identifier: 190-00C-CH00-00100-000-00B Page 45 of 50 
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Plan. 190-P10-CH00-00107-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
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A & B. 190-P10-CH00-00108-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
ENG.20040914.0020. [DIRS 171752] 
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C & D. 190-P10-CH00-00109-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
ENG.20040915.0001. [DIRS 171812] 
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E. 190-P10-CH00-00110-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
ENG.20040915.0002. [DIRS 172133] 
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F & G. 190-P10-CH00-00111-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
ENG.20040611.0015. [DIRS 170371] 
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H & J. 190-P10-CH00-00112-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
ENG.20040915.0003. [DIRS 171753]

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Title: Canister Handling Facility Criticality Safety Calculations 
Document Identifier: 190-00C-CH00-00100-000-00B Page 46 of 50 
BSC (Bechtel SAIC Company) 2004. Canister Handling Facility General Arrangement Section 
K & L. 190-P10-CH00-00113-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
ENG.20040915.0004. [DIRS 171813] 
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Application. 000-00C-MGR0-00800-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. 
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Misloaded Dry Waste Package. CAL-DSD-NU-000008 REV 00A. Las Vegas, Nevada: Bechtel 
SAIC Company. ACC: DOC.20041004.0008. [DIRS 172201] 
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Codisposal of ATR Spent Nuclear Fuel in a Waste Package. CAL-DSD-NU-000007 REV 00A. 
Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20041018.0001. [DIRS 171926] 
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Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package. CAL-DSD-NU-000004 REV 00A. 
Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20040329.0002. [DIRS 168935] 
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NU-000004 REV 04. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
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00100-000-003. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20041124.0001. 
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Survivability. 000-PSA-WHS0-00100-000-000 REV 00. Las Vegas, Nevada: Bechtel SAIC 
Company. ACC: ENG.20040719.0001. [DIRS 168792] 
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Carlo N-Particle Transport Code. CSCI: 30033 V4B2LV. DI: 30033-2003, Rev. 01. Las Vegas, 
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Intact Mode. BBA000000-01717-0210-00037 REV 00. Las Vegas, Nevada: CRWMS M&O. 
ACC: MOL.19990125.0079. [DIRS 104118]

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Title: Canister Handling Facility Criticality Safety Calculations 
Document Identifier: 190-00C-CH00-00100-000-00B Page 47 of 50 
CRWMS M&O 1999. Fast Flux Test Facility (FFTF) Reactor Fuel Criticality Calculations. 
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000001 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19991209.0195. [DIRS 
135852] 
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Shippingport LWBR Spent Nuclear Fuel in a Waste Package. CAL-EDC-NU-000004 REV 00. 
Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20000922.0093. [DIRS 151722] 
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Shippingport PWR Fuel in a Waste Package. CAL-EDC-NU-000002 REV 00. Las Vegas, 
Nevada: CRWMS M&O. ACC: MOL.20000209.0233. [DIRS 144714] 
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Spent Nuclear Fuel. CAL-EDC-NU-000003 REV 00. Las Vegas, Nevada: CRWMS M&O. 
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Specification for Department of Energy Standardized Spent Nuclear Fuel Canisters. 
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Criticality Analysis. DOE/SNF/REP-035, Rev. 0. Washington, D.C.: U.S. Department of Energy. 
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for Disposal Criticality Analysis. DOE/SNF/REP-051, Rev. 0. Washington, D.C.: U.S. 
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for Disposal Criticality Analysis. DOE/SNF/REP-040, Rev. 0. Washington, D.C.: U.S. 
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Criticality Analysis. DOE/SNF/REP-048, Rev. 0. Washington, D.C.: U.S. Department of Energy. 
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Criticality Analysis. DOE/SNF/REP-056, Rev. 0. [Washington, D.C.]: U.S. Department of 
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Title: Canister Handling Facility Criticality Safety Calculations 
Document Identifier: 190-00C-CH00-00100-000-00B Page 48 of 50 
DOE (U.S. Department of Energy) 2003. TMI Fuel Characteristics for Disposal Criticality 
Analysis. DOE/SNF/REP-084, Rev. 0. Idaho Falls, Idaho: U.S. Department of Energy, Idaho 
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Document Identifier: 190-00C-CH00-00100-000-00B Page 50 of 50 
8. ATTACHMENTS 
This calculation document includes three attachments: 
ATTACHMENT I Listing of Computer Files 
ATTACHMENT II One Compact Disc 
ATTACHMENT III Canister Handling Facility General Arrangement Drawings 
(Secondary references on these drawings are not relevant to this 
calculation.)

Licensing Calculation 
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Document Identifier: 190-00C-CH00-00100-000-00B Page I-1 of I-4 
ATTACHMENT I 
LISTING OF COMPUTER FILES 
All MCNP input and output files documented in this calculation were stored on an electronic 
medium (compact disc) as Attachment II. Also, the Microsoft Excel spreadsheets used to 
calculate input values are included on the compact disc. 
Date Time Size File 
02/25/2005 02:24p 52,736 ATR.xls 
02/25/2005 01:48p 57,344 SNF.xls 
02/15/2005 09:04a 20,308 DRY-S/atr-s 
02/15/2005 09:04a 686,365 DRY-S/atr-s.out 
02/15/2005 09:04a 20,244 DRY-S/atr-s0 
02/15/2005 09:04a 664,459 DRY-S/atr-s0.out 
02/15/2005 09:04a 16,297 DRY-S/fermi-s 
02/15/2005 09:04a 1,919,921 DRY-S/fermi-s.out 
02/15/2005 09:04a 16,077 DRY-S/fermi-s0 
02/15/2005 09:05a 1,881,146 DRY-S/fermi-s0.out 
02/15/2005 09:05a 20,403 DRY-S/fftf5sG 
02/15/2005 09:05a 1,737,079 DRY-S/fftf5sG.out 
02/15/2005 09:05a 20,355 DRY-S/fftf5sG0 
02/15/2005 09:05a 1,721,262 DRY-S/fftf5sG0.out 
02/15/2005 09:05a 16,372 DRY-S/fsv-s0 
02/15/2005 09:05a 505,463 DRY-S/fsv-s0.out 
02/15/2005 09:04a 7,963 DRY-S/nr1A-s0 
02/15/2005 09:04a 419,758 DRY-S/nr1A-s0.out 
02/15/2005 09:04a 7,868 DRY-S/nr4B-s0 
02/15/2005 09:04a 432,630 DRY-S/nr4B-s0.out 
02/15/2005 09:04a 12,214 DRY-S/sh-ls0 
02/15/2005 09:04a 489,045 DRY-S/sh-ls0.out 
02/15/2005 09:04a 28,313 DRY-S/sh-ps0 
02/15/2005 09:04a 774,170 DRY-S/sh-ps0.out 
02/15/2005 09:04a 53,739 DRY-S/tmi-ko 
02/15/2005 09:04a 815,141 DRY-S/tmi-ko.out 
02/15/2005 09:04a 594,545 DRY-S/tmi-ko.out1 
02/15/2005 09:05a 11,068 DRY-S/triga-s 
02/15/2005 09:05a 599,801 DRY-S/triga-s.out 
02/15/2005 09:05a 11,036 DRY-S/triga-s0 
02/15/2005 09:05a 594,868 DRY-S/triga-s0.out 
02/15/2005 09:04a 15,985 DRY-S/GEOM/fermiGs0 
02/15/2005 09:04a 1,880,828 DRY-S/GEOM/fermiGs0.out 
02/15/2005 09:04a 20,610 DRY-S/GEOM/fftf5sGr 
02/15/2005 09:04a 1,724,202 DRY-S/GEOM/fftf5sGr.out 
02/15/2005 09:04a 16,373 DRY-S/GEOM/fsvGs0 
02/15/2005 09:04a 507,079 DRY-S/GEOM/fsvGs0.out 
02/15/2005 09:04a 11,106 DRY-S/GEOM/tmi-s 
02/15/2005 09:04a 452,854 DRY-S/GEOM/tmi-s.out 
02/15/2005 09:04a 11,051 DRY-S/GEOM/trigaGs0

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Title: Canister Handling Facility Criticality Safety Calculations 
Document Identifier: 190-00C-CH00-00100-000-00B Page I-2 of I-4 
Date Time Size File 
02/15/2005 09:04a 682,577 DRY-S/GEOM/trigaGs0.out 
02/25/2005 02:51p 11,080 DRY-S/GEOM/MOD/trigGs0w 
02/25/2005 02:51p 686,672 DRY-S/GEOM/MOD/trigGs0w.out 
02/25/2005 02:51p 13,170 DRY-S/GEOM/MOD/triGs0o6 
02/25/2005 02:51p 777,979 DRY-S/GEOM/MOD/triGs0o6.out 
02/25/2005 02:51p 13,098 DRY-S/GEOM/MOD/triGs0w1 
02/25/2005 02:51p 702,087 DRY-S/GEOM/MOD/triGs0w1.out 
02/25/2005 02:51p 13,188 DRY-S/GEOM/MOD/triGs0w4 
02/25/2005 02:51p 775,771 DRY-S/GEOM/MOD/triGs0w4.out 
02/25/2005 02:51p 13,188 DRY-S/GEOM/MOD/triGs0w5 
02/25/2005 02:51p 775,877 DRY-S/GEOM/MOD/triGs0w5.out 
02/25/2005 02:51p 13,188 DRY-S/GEOM/MOD/triGs0w6 
02/25/2005 02:51p 775,877 DRY-S/GEOM/MOD/triGs0w6.out 
02/25/2005 02:51p 13,188 DRY-S/GEOM/MOD/triGs0w7 
02/25/2005 02:51p 775,994 DRY-S/GEOM/MOD/triGs0w7.out 
03/10/2005 02:17p 11,091 DRY-S/GEOM/MOD/triPGs0o 
03/10/2005 02:17p 694,233 DRY-S/GEOM/MOD/triPGs0o.out 
03/10/2005 02:17p 11,112 DRY-S/GEOM/MOD/triPGs0w 
03/10/2005 02:17p 691,884 DRY-S/GEOM/MOD/triPGs0w.out 
02/15/2005 09:00a 20,614 DRY-A/atr-a0 
02/15/2005 09:00a 668,961 DRY-A/atr-a0.out 
02/15/2005 09:00a 20,638 DRY-A/atr30a0 
02/15/2005 09:00a 669,441 DRY-A/atr30a0.out 
02/15/2005 09:00a 20,634 DRY-A/atr30Aa0 
02/15/2005 09:00a 668,407 DRY-A/atr30Aa0.out 
02/15/2005 09:00a 20,610 DRY-A/atrAa0 
02/15/2005 09:00a 677,119 DRY-A/atrAa0.out 
02/15/2005 09:00a 16,370 DRY-A/fer30a0A 
02/15/2005 09:00a 1,883,608 DRY-A/fer30a0A.out 
02/15/2005 09:00a 16,339 DRY-A/ferm30a0 
02/15/2005 09:00a 1,885,159 DRY-A/ferm30a0.out 
02/15/2005 09:00a 16,309 DRY-A/fermi-a0 
02/15/2005 09:00a 1,884,841 DRY-A/fermi-a0.out 
02/15/2005 09:00a 16,980 DRY-A/fermiAa 
02/15/2005 09:00a 1,924,381 DRY-A/fermiAa.out 
02/15/2005 09:00a 16,352 DRY-A/fermiAa0 
02/15/2005 09:00a 1,886,470 DRY-A/fermiAa0.out 
02/15/2005 09:01a 20,391 DRY-A/fft30a0A 
02/15/2005 09:01a 1,724,765 DRY-A/fft30a0A.out 
02/15/2005 09:01a 20,344 DRY-A/fftf30a0 
02/15/2005 09:01a 1,723,611 DRY-A/fftf30a0.out 
02/15/2005 09:01a 20,346 DRY-A/fftf5a0 
02/15/2005 09:01a 1,726,109 DRY-A/fftf5a0.out 
02/15/2005 09:01a 20,393 DRY-A/fftf5a0A 
02/15/2005 09:01a 1,736,809 DRY-A/fftf5a0A.out 
02/15/2005 09:01a 16,607 DRY-A/fsv-a0 
02/15/2005 09:01a 512,014 DRY-A/fsv-a0.out

Licensing Calculation 
Title: Canister Handling Facility Criticality Safety Calculations 
Document Identifier: 190-00C-CH00-00100-000-00B Page I-3 of I-4 
Date Time Size File 
02/15/2005 09:01a 16,625 DRY-A/fsv30a0 
02/15/2005 09:01a 511,797 DRY-A/fsv30a0.out 
02/15/2005 09:01a 16,621 DRY-A/fsv30Aa0 
02/15/2005 09:01a 509,269 DRY-A/fsv30Aa0.out 
02/15/2005 09:01a 16,603 DRY-A/fsvAa0 
02/15/2005 09:01a 514,605 DRY-A/fsvAa0.out 
02/15/2005 09:01a 8,347 DRY-A/nr1A-a0 
02/15/2005 09:01a 423,518 DRY-A/nr1A-a0.out 
03/21/2005 04:24p 8,328 DRY-A/nr1A30a0 
03/21/2005 04:24p 423,360 DRY-A/nr1A30a0.out 
02/15/2005 09:01a 8,325 DRY-A/nr1A30aA 
02/15/2005 09:01a 424,466 DRY-A/nr1A30aA.out 
02/15/2005 09:01a 8,345 DRY-A/nr1AaA0 
02/15/2005 09:01a 423,996 DRY-A/nr1AaA0.out 
02/15/2005 09:01a 8,204 DRY-A/nr4B-a0 
02/15/2005 09:01a 438,298 DRY-A/nr4B-a0.out 
02/15/2005 09:01a 8,229 DRY-A/nr4B30a0 
02/15/2005 09:01a 435,556 DRY-A/nr4B30a0.out 
02/15/2005 09:01a 8,231 DRY-A/nr4B30aA 
02/15/2005 09:01a 438,458 DRY-A/nr4B30aA.out 
02/15/2005 09:01a 8,250 DRY-A/nr4BaA0 
02/15/2005 09:01a 432,612 DRY-A/nr4BaA0.out 
02/15/2005 09:01a 12,447 DRY-A/shL-a0 
02/15/2005 09:01a 492,086 DRY-A/shL-a0.out 
02/15/2005 09:01a 12,469 DRY-A/shL30a0 
02/15/2005 09:01a 490,532 DRY-A/shL30a0.out 
02/15/2005 09:01a 12,464 DRY-A/shL30Aa0 
02/15/2005 09:01a 495,067 DRY-A/shL30Aa0.out 
02/15/2005 09:01a 12,442 DRY-A/shLAa0 
02/15/2005 09:01a 494,431 DRY-A/shLAa0.out 
02/15/2005 09:01a 28,689 DRY-A/shP-a0 
02/15/2005 09:01a 779,502 DRY-A/shP-a0.out 
02/15/2005 09:01a 28,712 DRY-A/shP30a0 
02/15/2005 09:01a 779,502 DRY-A/shP30a0.out 
02/15/2005 09:01a 28,706 DRY-A/shP30Aa0 
02/15/2005 09:01a 779,097 DRY-A/shP30Aa0.out 
02/15/2005 09:01a 28,683 DRY-A/shPAa0 
02/15/2005 09:01a 780,995 DRY-A/shPAa0.out 
02/15/2005 09:01a 54,124 DRY-A/tmi30Aa0 
02/15/2005 09:01a 1,046,663 DRY-A/tmi30Aa0.out 
02/15/2005 09:01a 54,108 DRY-A/tmiK-a0 
02/15/2005 09:02a 1,051,621 DRY-A/ tmiK-a0.out 
02/15/2005 09:02a 54,131 DRY-A/tmiK30a0 
02/15/2005 09:02a 1,046,425 DRY-A/tmiK30a0.out 
02/15/2005 09:02a 54,102 DRY-A/tmiKAa0 
02/15/2005 09:02a 1,047,807 DRY-A/tmiKAa0.out 
02/15/2005 09:02a 11,285 DRY-A/tri30a0A

Licensing Calculation 
Title: Canister Handling Facility Criticality Safety Calculations 
Document Identifier: 190-00C-CH00-00100-000-00B Page I-4 of I-4 
Date Time Size File 
02/15/2005 09:02a 597,615 DRY-A/tri30a0A.out 
02/15/2005 09:02a 11,288 DRY-A/trig30a0 
02/15/2005 09:02a 599,327 DRY-A/trig30a0.out 
02/15/2005 09:02a 11,270 DRY-A/triga-a0 
02/15/2005 09:02a 598,691 DRY-A/triga-a0.out 
02/15/2005 09:02a 11,267 DRY-A/trigaAa0 
02/15/2005 09:02a 596,873 DRY-A/trigaAa0.out 
03/10/2005 02:19p 16,537 DRY-A/fermiPa0 
03/10/2005 02:19p 1,925,666 DRY-A/fermiPa0.out 
03/10/2005 02:18p 16,581 DRY-A/fermPAa0 
03/10/2005 02:18p 1,928,445 DRY-A/fermPAa0.out 
03/10/2005 02:19p 20,576 DRY-A/fftfP5a0 
03/10/2005 02:19p 1,741,395 DRY-A/fftfP5a0.out 
03/10/2005 02:19p 20,623 DRY-A/fftP5a0A 
03/10/2005 02:19p 1,751,257 DRY-A/fftP5a0A.out 
02/15/2005 09:01a 28,689 DRY-A/shP-a0 
02/15/2005 09:01a 779,502 DRY-A/shP-a0.out 
02/15/2005 09:01a 28,712 DRY-A/shP30a0 
02/15/2005 09:01a 779,502 DRY-A/shP30a0.out 
02/15/2005 09:01a 28,706 DRY-A/shP30Aa0 
02/15/2005 09:01a 779,097 DRY-A/shP30Aa0.out 
02/15/2005 09:01a 28,683 DRY-A/shPAa0 
02/15/2005 09:01a 780,995 DRY-A/shPAa0.out 
03/10/2005 02:20p 11,302 DRY-A/trigaPa0 
03/10/2005 02:20p 602,648 DRY-A/trigaPa0.out 
03/10/2005 02:20p 11,299 DRY-A/trigPAa0 
03/10/2005 02:20p 603,502 DRY-A/trigPAa0.out