Fuel Handling Facility Criticality Safety Calculations Rev 00A, ICN 00

210-00C-FH00-00400-000-00

June, 2004

1. PURPOSE

The purpose of this design calculation is to perforin a criticality evaluation +
of the Fuel Handling

Facility (FHF) and the operations and processes performed therein. The current +
intent of the FHF

is to receive transportation casks whose contents will be unloaded and +
transferred to waste

packages (WP) or MGR Specific Casks (MSC) in the fuel transfer bays. Further, +
the WPs will

also be prepared in the FHF for transfer to the sub-surface facility (for +
disposal). The MSCs will

be transferred to the Aging Facility for storage. The criticality evaluation of+
 the FHF features the

following:

0 Consider the types of waste to be received in the FHF as specified below:

I . Uncanistered commercial spent nuclear fuel (CSNF)

2. Canistered CSNF (with the exception of horizontal dual-purpose canister +
(DPC)

and/or multi-purpose canisters (MPCs)

3. Navy canistered SNF (long and short)

4. Department of Energy (DOE) canistered high-level waste (HLW)

5. DOE canistered SNF (with the exception of MCOs)

Evaluate the criticality analyses previously perfortned for the existing +
Nuclear Regulatory

Commission (NRC)-certified transportation casks (under 10 CFR 71) to be +
received in the

FHF to ensure that these analyses address all FHF conditions including nonnal +
operations,

and Category I and 2 event sequences.

Evaluate FHF criticality conditions resulting from various Category I and 2 +
event sequences.

Note that there are currently no Category I and 2 event sequences identified +
for FHF.

Consequently, potential hazards from a criticality point of view will be +
considered as

identified in the Internal Hazards Analysis for License Application document +
(BSC 2004c,

Section 6.6.4).

o Assess effects of potential moderator intrusion into the fuel transfer bay +
for defense in depth.

The SNF/HLW waste transfer activity (i.e., assembly and canister transfer) that+
 is being canied

out in the FHF has been classified as safety category in the Q-Iist (BSC 2003, +
p. A-6). Therefore,

this design calculation is subject to the requirements of the Quality Assurance+
 Requirements and

Description (DOE 2004), even though the FHF itself has not yet been classified +
in the Q-Iist.

Performance of the work scope as described and development of the associated +
technical product

conform to the procedure AP-3.12Q, Design Calculations and Analyses.

It should also be mentioned that the facility description document or the +
system description

document for the FHF is not available at this time. Consequently, this +
calculation is valid for the

current design and may not reflect the ongoing design evolution of the FHF.

Nuclear Analysis Design Calculation

Title: Fuel Handling Facility Criticality Safety Calculations

Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 8 of 40

2. METHOD

2.1 CRITICALITY SAFETY ANALYSIS

The criticality safety calculations presented in this document evaluate the +
various waste fonns in

the transportation casks, WPs, canisters, and MSCs casks in the FHF to ensure +
they all meet the

criticality safety requirements under normal conditions as well as for Category+
 I and 2 events.

Moderator conditions are also varied to find the most reactive configuration. +
The process and

methodology for criticality safety analysis given in the Preclosure Criticality+
 Analysis Process

Report (BSC 2004e, Section 2.2.7) will be implemented in these calculations. +
The following

method will be pursued for each waste form and cask/canister configuration (BSC+
 2004e,

Section 2.2.7):

&bull; The design basis for the FHF relies on the most reactive fuel assemblies

&bull; The multiplication factor (kff) will not exceed 0.95, including all +
biases and uncertainties in

the data and method of the analysis, under all normal, and Category I and 2 +
event sequences

&bull; Conservative modeling dimensional variables will be used (e.g., assembly+
 pitch,

manufacturing tolerances for assemblies, etc.) in order to maximize reactivity

&bull; Conservative modeling assumptions will also be used regarding materials +
in fuel including

no accounting for bumable poisons in fuel, no credit for 231U and 231U in fuel,+
 and use of the

most reactive fuel stack density

&bull; Credit can only be taken for up to 75 % of the neutron absorbing +
material in criticality

controls (e.g., grid plates).

Moderator density will be varied over the range of 0.0 through 1.0 in order to +
include all

possible criticality conditions.

hi addition, the effect on criticality due to rearrangement of fuel inside the +
casks/canisters from

natural events (e.g., earthquakes) or incidents (e.g., dropped assembly) will +
be considered (BSC

2004e, Section 2.2.8).

These calculations use the qualified software MCNP (Briesmeister 1997 and CRWMS+
 M&amp;O

1998a). MCNP is a three-dimensional Monte Carlo particle transportation code +
with the

capability to calculate eigenvalues for critical systems. The Nuclear +
Regulatory Commission

(NRC) accepts MCNP in NUREG-1567 (NRC 2000, 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 AP-3.13Q, Design Control. 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 11).

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Nuclear Analysis Design Calculation

Title: Fuel Handling Facility Cn'ticality Safety Calculations

Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 9 of 40

3. ASSUMPTIONS

3.1 The nominal acceptable calculated value of kff is assumed to be 0.925 as a +
criticality

limit in order to meet the design criten'a specified in the PDC Document [i.e.,+
 kff can not

exceed 0.95 including uncertainties and bias at 95% confidence level (Minwalla +
2003,

Section 4.9.2.2.1)]. In other words, the nominal value provides a margin of +
0.025 (0.95 -

0.925) to account for code bias and uncertainties at 95% confidence level.

Rationale: Uncertainties and bias that need to be considered in this analysis +
pertain to

statistical uncertainties, dimensional uncertainties, code bias, and tolerance +
uncertainties.

Applicable code bias for the fuel type and enrichment range of this analysis is+
 typically

less than 0.5 % (CRWMS M&amp;O 1999c, Section 4). An allowance of 2% is +
provided to

account for the remaining uncertainties associated with statistical variation, +
dimensional

variables and tolerances. This allowance is similar to, and slightly greater +
than

(conservative), the value used for the SNF storage and transportation cask +
criticality

evaluations (General Atomics 1993, p. 6.4-7).

Usage: This assumption is used throughout this desigii calculation.

3.2 The naval waste packages will be designed in such a way as to make +
criticality not

credible.

Rationale: Criticality analyses for naval waste packages are the responsibility+
 of the U.S

Department of the Navy. Per the Naval Spent Nuclear Fuel Waste Package System

Description document, the &quot;sealed waste package shall provide criticality +
Qontrol&quot; to

reduce the probability of a criticality occurring (BSC 2004d, Section 3.1.1.3).

Usage: Sections 5.1 and 5.2.3.

3.3 The MGR Site specific Cask (MSC) is assumed to be similar in design, other +
than the

neutron poison loading/configuration, to the Multi Purpose Canister (MPC)-24 +
for PWR

fuel and the MPC-68 for BV~R fuel.

Rationale: Since the MSC is still being developed, the criticality control +
features will be

similar to the existing NRC-certified storage casks.

Usage: This assumption is used in Sections 5.1, 5.2.3, and 6. L

3.4 The Fort St. Vrain fuel is assumed to have a U-23 5 enrichment of 100%.

Rationale: This assumption was used to introduce conservatism into the +
calculation.

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Title: Fuel Handling Facility Criticality Safety Calculations

Document Identirier: 210-OOC-FHOO-00400-000-OOA Page 10 of 40

Usage: Section 5.1.

3.5 It is assumed that the fuel basket inside the canister (if present) remains+
 intact following a

canister drop.

Rationale: The canister intemals are designed to remain sufficiently intact +
that there

would be no criticality concem following any credible drop or handling mishap

(Minwalla 2003, Section 4.9.2.2.7).

Usage: Section 6.2.

3.6 For damaged fuel calculations, it is assumed that the inside of the +
canister is dry.

Rationale: The canister will be shipped to the repository dry. There is no +
credible

mechanism by which the inside of the canister could become flooddd since the +
canister is

designed to withstand any credible drop without breaching (Canori and Leitner +
2003,

p. 3-63).

Usage: Section 6.2.

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Title: Fuel Handling Facility Criticality Safety Calculations

Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 11 of 40

4. USE OF COMPUTER SOFTWARE

4.1 BASELINED SOFTWARE

4.1.1 MCNP

The MCNP code (CRWMS M&amp;O 1998a) was used to calculate the multiplication +
factor, kff, for

all systems presented in this report. The software specifications are as +
follows:

&bull; Program Name: MCNP (CRWMS M&amp;O 1998a)

&bull; Version/Revision Number: Version 4B2LV

&bull; Status/Operating System: Qualified/HP-UX B. 10.20

&bull; Software Tracking Number: 30033 V4132LV

&bull; Computer Type: HP 9000 Series Workstations

&bull; CPU Number: 700887

The input and output files for the various MCNP calculations are contained on a+
 CD (Attachment

11) and the files are listed in Attachment 1.

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) and +
CRWMS M&amp;O

(I 998b, 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

&bull; Title: Excel

&bull; Version/Revision Number: Microsofl@ Excel 97 SR-2

&bull; This version is installed on a PC running Microsofl Windows 2000 with +
CPU number

503009

The files for the various Excel calculations are contained on a CD (Attachment +
11) and the.files

are listed in Attachment 1.

The Excel software was used to illustrate results in Sections 5.2 and 6. Excel +
is exempt from

qualification per Section 2.1.6 of LP-SI. I IQ, Software Management.

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Calculation

Title: Fuel Handling Facility Criticality Safety Calculations

Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 12 of 40

5. CALCULATION

All technical product inputs and sources of the inputs used in thc: development+
 of this calculation

are documented in this section. It should also be mentioned that the terrns +
&quot;model(s)&quot; and

&quot;modeling&quot; as used in this calculation document refer to the +
geometnic configurations of the

cnticality cases analyzed.

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 (Minwalla 2003) are used in these calculations. The +
pertinent criteria

for surface facility criticality include the following (Minwalla 2003, Section +
4.9.2.2):

0 The multiplication factor (keff) will not exceed 0.95, including all biases +
and uncertainties in

the data and method of the analysis, under all nortnal and off-normal event +
sequences. This

criterion satisfies Requirement Number PRD-013/T-022 in the Project +
Requirements

Document (Canori and Leitner 2003, p. 3-76).

The facility design will utilize a favorable geometry and/or fixed neutron +
absorbers without

the use of bumup credit.

No moderator shall be present in any area where radioactive waste is being +
handled (fuel

transfer rooms, WP closure room etc.). Attachment III features draft sketches +
of the FHF as

of the date of this calculation (the general arrangement drawings have not been+
 finalized yet),

and may not reflect the ongoing design evolution. The purpose of the sketches +
is to show

functional areas where moderator control is engineered in the design for +
criticality safety.

These functional areas will remain with moderator control, even if design +
changes are made

to the FHF with respect to the layout.

5.1.2 Most Reactive Fuel Selection

In accordance with the requirements given in Preclosure Criticality Analysis +
Process Report

(BSC 2004e, Section 2.2.7), the design of the facility should be based on most +
reactive fuel

assemblies. The following evaluations were performed with MCNP to detennine the+
 most

reactive fuel assembly.

Commercial Fuel

The evaluations perforined in the Surface Facility Criticality Safety +
Calculations document show

that the W 17 x 17 OFA was found to be the most reactive PWR fuel assembly when+
 modeled as

a single fuel assembly and in a storage cask configuration (BSC 2004f, Section +
5.2.1). In

addition, an evaluation to detennine the most reactive fuel assembly was +
perfon-ned in the Final

Safety and Analysis Report for the Holtec International Storage and Transfer +
Operation

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Calculation

Title: Fuel Handling Facility Criticality Safety Calculations

Document Identirier: 210-OOC-FHOO-00400-000-OOA Page 13 of 40

Reinforced Module Cask System (HI-STORM 100 Cask Systeni) and the Westinghouse +
17xl7

OFA was selected (Holtec Intemational 2002, Section 6.2-2). The GE 8 x 8 array +
was selected

for the BWR fuel (Holtec Intemational 2002, Section 6.2-3 as featured in the +
Aging Facility

Criticality Safety Calculations document (BSC 2004a, Section 5.1.3). The +
results from the

Surface Facility Criticality Safety Calculations and Aging Facility Criticality+
 Safety Calculation

documents will be utilized later in this FHF evaluation where applicable.

DOE Fuel

DOE SNF has been categorized into nine fuel groups (Mecharn, D.C. 2004, Section+
 4.2.4.1):

I . Uranium Metal fuels (N-Reactor)

2. Uranium-Zirconium/Uranium-Molybdenum fuels (Enrico Fei-mi Liquid Metal +
Reactor)

3. Uranium Oxide fuels (high enriched uranium - Shippingport PWR)

4. Uranium Oxide fuels (low enriched uranium - Three Mile Island (TMI)-2 PVvR)

5. Uranium-Aluminum fuels (foreign research reactor - Melt &amp; Dilute)

6. Uranium/Thorium/Plutonium Carbide fuels (Ft. St. Vrain Gas Cooled Reactor)

7. Mixed Oxide fuels (Fast Flux Test Facility (FFTF) Reactor)

8. Uranium/Thorium Oxide fuels (Shippingport Light Water Breeder Reactor +
(LWBR))

9. Uranium-Zirconium-Hydride fuels (Training Research Isotopes General Atomics +
(TRIGA)).

The Canister Handling Criticality Safety Calculations document evaluates the +
DOE fuel types

listed above (both Mark IA and Mark IV type fuel are considered for N Reactor +
and type &quot;D&quot;

and type &quot;K&quot; canister are evaluated for TMI-2 fuel) (BSC 2004b, +
Section 5.1.2). The FFTF fuel

type was shown to be the most reactive in a single flooded canister under +
normal conditions

(BSC 2004b, Section 6.1). However, during off-normal conditions the TRIGA fuel +
is the more

reactive fuel type (BSC 2004b, Section 6.2). Additional calculations will be +
presented later in

this document for all DOE fuel types to show the effect of varying moderator +
density (Section

6.2.1) and effects of neutron poison variation for applicable DOE fuel types +
(Section 6.2.2). An

off-norrnal condition with TRIGA fuel will also be shown later in this document+
 (Section 5.2).

Naval Fuel Canisters

The design and safety analyses of the naval spent nuclear fuel VVT are the +
responsibility of the

U. S. Department of the Navy. No data on the naval waste package, including +
criticality

analyses, are publicly available. Section 3.1.1.3 of Naval Spent Nuclear Fuel +
Waste Package

System Description Document (BSC 2004d) states &quot;The sealed waste package +
shall provide

criticality control ... The tenn 'control' used in this requirement is intended+
 to mean that neutron

absorber materials are added and the waste package loading configuration desip +
reduces the

probability of a criticality occurring.&quot; From this description, it is +
assumed that the naval canister

and intemals will be designed to make criticality not credible (Assumption 3.2)+
 and no

evaluations will be featured in this document.

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Nuclear Analysis Design Calculation

Title: Fuel Handling Facility Criticality Safety Calculations

Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 14 of 40

5.1.3 Upper Suberitical Limit

In accordance with the requirements given in Preclosure Criticality Analysis +
Process Report

(BSC 2004e, Section 2.2.7), kff should not exceed 0.95, including all biases +
and uncertainties in

the data and method of the analysis. All evaluations featuring a cask and the +
MSC are performed

for the worst case combination of manufacturing tolerances with respect to +
criticality (Holtec

Intemational 2002, p.6.3-2). Evaluations were perforrned to deterTnine the +
effects of tolerances

(Holtec Intemational 2002, Tables 6.3-1 &amp; 6.3-2). It was determined that +
design pararneters

important to criticality safety are fuel enrichment, the inherent geometry of +
the fuel basket

structure and the fixed neutron absorbing panels (Boral) (Holtec Intemational +
2002, p. 6.3-3).

Further, the results referred to in Section 6 of this report are within the +
bounds of the kff values

demonstrated in the Final Safety and Analysis Reportfor the Holtec +
International Storage and

Transfer Operation Reinforced Module Cask System (HI-STORM 100 Cask System) to +
cover

uncertainties and bias. The remainder of the results either presented or +
referred to in this

document is designed to meet an upper subcritical limit of 0.925 per Assumption+
 3. 1.

5.1.4 Geometry Calculation Inputs

Physical inputs for the van'ous casks/canisters and waste fonns are described +
in this section.

Since criticality evaluations have been perfonned for some of the +
casks/canisters and waste

forms previously, the inputs will be referenced from these previous studies.

Commercial Fuel (Transportation Casks and MSCs)

A representative vertical cask is selected here for criticality calculations to+
 demonstrate

compliance with the cn'ticality safety requirements. The selected cask is +
HI-STORM 100, as this

system is currently qualified for high seismic requirements to ensure that the +
YMP seismic

spectrum will be enveloped (Cogema 2004, p.5). Since the MSC is still being +
developed, it was

assumed that it is similar in design to the MPC-24 of the HI-STORM 100 cask +
system for PWR

fuel and MPC-68 of the HI-STORM 100 cask system for BWR fuel (Assumption 3.3). +
A

criticality evaluation for these casks has already been perfon-ned in the Aging+
 Facility Criticality

Safety Calculations document, where the configuration and physical dimensions +
for the MPC-24

as well as fuel specifications (BSC 2004a, Section 5.1.5. 1) and the MPC-68 +
(BSC 2004a, Section

5.1.5.3) are given. Figure 5.1-1 illustrates the radial view of the MPC-24 cask+
 inside the

HI-STORM 100 overpack and Figure 5.1-2 illustrates the radial view of the +
MPC-68 cask inside

the HI-STORM 100 overpack. Figure 5.1-2 also displays the axial view of the MPC+
 inside the

overpack, which configuration is the same for both the MPC-24 and MPC-68 casks.

It should be mentioned that calculations of the MSCs only exist for commercial +
SNF, which was

valid at the time of the criticality safety calculation perforined for the +
Aging Facility. This

criticality safety calculation is in the process of being revised to include +
other waste types for

aging.

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ign Calculation

Title: Fuel Handling Facility Cniticality Safety Calculations

Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 15 of 40

Radial

Overpack

Figure 5.1 -1 Radial View of the MPC-24 Cask Inside the H [-STORM 100 Overpack

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Nuclear Aiialysis Design Calculation

'I'Itle- Fuel f faiidliiig Facility Criticality Safety Calculations

Document ldentifier: 2 1 O-OOC-FHOO-00400-000-OOA Page 16 of 40

MPC

MPC

),erpack

X O,e,pack

Active

F,pl

TRIM, &quot;f, A

Figure 5.1-2 Radial and Axtal View of the MPC-68 Cask Inside the HI-STORM +
1000verpack

DOE FLICI (Canisters

I'able 5.1-1 presetits the pliysical dimensioiis of ilie caiiisters atid Table +
5.1-2 stiows the DOE

Ctiel pariiiiieters. Figtire 5.1-3 displays the DOE caiiisters considered in +
this cvaltiation, as

describcd in Section5.1.2, inthera(lial %iew. ATiaxial represeTitation of the +
DOE SNFcaiiisters

is iilso iiiclticled in Figtirc 5.1-3. It should be iiiciitioiied iliat the +
MCNP iiiptit 1-ites f'toiii t[ic

Safeli CalculationsdOCUIrCut (BSC2004b) were used asastitlliiig

point for the ciilctilatiotis presciited in this doctiniciit. For inore details+
 regardiiig caiiister

physical (Iiiiiezisioiis, sec Section 5.1.4 (BSC 2004b) aiid Sectioii 5.1.2 +
(BSC 2004b) for more

specifics re.prding DOE fuel parameters.

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Title: Fuel Handling Facility Criticality Safety Calculations

Document Identiirier: 210-OOC-FHOO-00400-000-OOA Page 17 of 40

Table 5.1 -1 Physical Dimensions of DOE Canisters

DOE Fuel Type	Canister o.d.	Canister length	Canister Capacity	Reference

	(cm)	(cm)

			3360 fuel pins (2 sets of	CRWMS M&amp;O

Enrico Fermi	45.72		12 tubes each	2000a

p. 13

			containing 140 pins)	,

	45.72(0.95	456.90 (414.50 cm	1302 fuel pins (6	CRWMS M&amp;O

FFTF	cm wall	internal length)	assemblies with each	1999a, Figures 5-3 &amp;

	thickness)		217 fuel pins)	5-4

Fort St. Vrain	45.72(0.95

cm wall	457.0 (411.71 cm	5 fuel elements stacked

BSC 2001 a

p. 15

	thickness)	internal length)	vertically	,

	45.72(0.95	299.90 (254.0 cm	3 ingots stacked

Melt &amp; Dilute	cm wall	internal length)	vertically	BSC 2001 b, p. 11

	thickness)

			270 fuel elements (54	CRWMS M&amp;O 2001,

N Reactor	64.29	419.84	fuel elements stacked 5	p. 14

			high)	DOE 2000, pp. 23-25

				(canister capacity)

	45.72(0.95		7428 fuel rods (12	CRWMS M&amp;O

Shippingport LWBR	cm wall	457.0 (411.71 cm	assemblies with each	2000b, p. 18

	thickness)	internal length)	619 fuel rods)	DOE 1999b, p. 16

				(canister capacity)

Shippingport PWR	45.72(0.95

cm wall	268.09 (internal

1 fuel cluster	CRWMS M&amp;O

thickness)	length)		2000c, p. 15

b	35.56(0.64	380

37 (346

55 cm	1 fuel assembly (1 5xl 5	DOE 2003, pp. 21

TMI-2 (D canister	cm wall	.

.

internal length)	array having 204 fuel	(canister capacity),

	thickness)		rods)	25 &amp; 26

	45.72(0.95	254.70 (internal	111 fuel elements (37	CRWMS M&amp;O

TRIGA	cm wall	length)	fuel elements stacked 3	1999d

p. 13

	thickness) I		high)	,

a Mark 1A contains 48 fuel elements stacked 5 high, comprising a total of 240 +
fuel elements (DOE 2000,Fig. 4-2~

b The K canister has a large internal diameter over which fuel matrix material +
is not constrained (see Fig. 5.1-3)

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Title: Fuel Handling Facility Criticality Safety Calculations

Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 18 of 40

Table 5.1-2 DOE Fuel Parameters

	Max. fissile	Fuel	Clad	Clad	Pin Pitch	Fuel

DOE Fuel Type	enrichment

a	o.d.

b	i.d.	o.d.

(cm) c	length	Reference

	(%j	(CM)	(CM)	(cm)		(cm)

							DOE 1999a, p.8

Enrico Fermi	25.69	0.376	0.376	0.401	0.44	77.47	CRWMS M&amp;O

							2000a, p. 12

							(clad)

							INEEL 2002,

FFTF	25.95	0.495	0.508	0.584	0.726	237.24	p.15,17 (pin

							pitch) &amp; Fig. 3

							(fuel o. .)

	100.0						Taylor 2001, p. 21

Fort St. Vrain	(Assumption	1.245	-		1.880		&amp; Fig. 2-3 (pin

	3.4)						itch)

Melt &amp; Dilute	20.0	41.91	-	-	-	76.2	BSC 2001c, p.3

N Reactor - outer	e

1

25	6.096	6.096	6.223			DOff-2000,

fuel tube d	.	4.496 f	4.496	4.607	6.096	53.0	Tables 3-1 &amp; 3-2

							(clad)

N Reactor - inner	e

1

25	3.175	3.175	3.378			DOE 2000,

fuel tube d	.	1.118'	1.118	1.245	6.096	53.0	Tables 3-1 &amp; 3-2

							(clad)

							DOE 1999b, p. 16

							(enr.), Fig. 3-3

Shippingport LWBR	4.90	0.640	0.734	0.778	0.937		(pin pitch), Table

							3-5 (fuel o.d.) &amp;

							Table 3-8 (clad)

Shippingport PWR	93.2	-			-		DOE 1999c,

							Table 3-1

							DOE 2003, p. 19

TMI-2	2.96	0.936	0.958	1.092	1.443	360.12	(enr,), p. 21, p. 22

							(fuel length) &amp;

							p.23 (fuel o.d.)

TRIGA	70

0	3.480	3

490	592

3	3

480	38

10	CRWMS M&amp;O

	.	0.635 g ,	.	.	.	.	1999b, p. 19

b This is the total fissile content divided by the total heavy metal mass x +
100.

For fuel in the form of cylindrical rods, this is the fuel outside diameter

c For fuel in the form of cylindrical rods, this is the nominal pin pitch in +
the canister

d See Figure 5.1-3 for locations of outer and inner fuel tubes

' The enrichment for Mark IV (case B) is 0.95 %

f Inside diameters of fuel tubes

9 Inside diameters of fuel tube

<HHTML><A +
HREF="http://rms.ymp.gov/cgi-bin/record_image?rec=ENG.20040630.0012&img=18"><IM+
G SRC="/basisicons/image.gif" ALT="[Image +
(ENG.20040630.0012-NV:18)]"></A></HHTML>

NUClcar Analysis Dcsigii C'alculatioii

Title: Fuel Handling Facility Criticality Safety Calculations

Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 19 of 40

Radial V. of N Phil or Mark I A Crosier i Voee or N Pvactur Mark V Canister +
R,d.dJ Vd. of Pool, &amp; Dial, soon, It I r

P a dial Vowo of Fng St Vrain Canistpr

i View, of FFTF Cents,

Roodial Vew of Enrico Fermi Camotor

Rad,il View of Shippingport LWBR

Collate,

Fqadial View of 7Mli2 7ype K C a lilsier

Pedial Vipw of Shippingpnq PWP Canisier

i \/,. of TRIGA C col

Radial View or TMI I Typ, D C,fi

Anal View

or DIE

Cedeller

Figure 5.1-3 Radial and Axial View of the DOE Fuel Canisters

Navy Fuel

Per Assumption 3.2, no evaluations will be perfortned of navy fuel since it is +
the responsibility of

the U. S. Department of the Navy.

5.1.5 Material Compositions

Material compositions for the various casks/canisters and waste foryns are +
described in this

section. Since criticality evaluations have been performed for some of the +
casks/canisters and

waste forrns previously, the inputs will be referenced from these previous +
studies.

Commercial Fuel (PVVR &amp; BWR)

Since no additional calculations were perfonned in this document for commercial+
 fuel, the

material compositions can be found in the Aging Facility Criticality Safety +
Calculations

document for PWR (BSC 2004a, Section 5.1.5.2) and BWR (BSC 2004a, Section +
5.1.5.4) fuel.

DOE Fuel

Table 5.1-3 display's the relevant material properties for DOE non-fuel +
materials used in the

MCNP models. Table 5.1-4 presents the isotopic content of the fuel materials +
for each DOE type

fuel considered in this calculation.

Table 5.1-3 Material Properties for DOE Non-Fuel Materials

Material	Density

3

Weight Percent (wt %)	Reference/

	(g/CM

)		Remark

H20	1

0 a	H - 0.6666667

(throughout model)	.	0 - 0.3333333

		0:49.94 Ca:22.63 C:10.53 Mg:9.42

Magnuson Concrete	2.147	SiA.21 K0.9445 A1:0.7859 Fe:0.559,5	NRC 1997, Volume 3,

		Ti:0.148 Na:0.1411 H:0.3319 S:0.2483	p. M8.2.4

		C1:0.0523 Mn 0.0512

Type 304L	7

94	Fe:68.045 Cr:19.0 Ni:10.0 Mn:2.0	ASTM A 276-91 a 199 1, p. 2

Stainless Steel	.	Si:0.75 N:0.1 13:0.045 S:0.03 C:0.03	ASTM GI-90 1999, Table +
Xl

Type 316L		Fe:65.295 COTO Ni:12.0 Mn-2 * 0	ASTM A 276-91 a 199 1

p

2

Stainless Steel	7.98	Mo:2.5 Si:1.0 N:0.1 P:0.045 S:0.03	,

.

ASTM G 1 -90 1999

Table Xi

		C:0.03	,

Type 516		Fe:98.33 Mn:1

025 Si:0

275 P:0

035	ASME 2001, Sec IIA, SA-

Carbon Steel	7.85	.

.

.

S:0.035 C:0.3	516/SA-516M &amp; Sec IIA, SA-

			20/SA-20M, item 14

- I he moderator density was varied between 0.0 - 1 .0 glcm' to study moderator+
 density variations in Section 6

b Values given in atom fraction and not wt %

Nuclear

Table 5.1-4 Material Properties for Each DOE Fuel Type

DOE Fuel Type	Dens	Weight Percent (wt%)	Neutron Absorber (kg)

	(g/c )

Enrico Fermi	17.424	U7235:22.96 U-238:66-41	1.2-3.03

		Mo:10.63

		0:11.63 U-235:0.13

FFTF	10

02	U-238:62.37 Pu-239:22.54	0.4-19.26 b

	.	Pu-240:3.01 Pu-241:0.26

		Pu-242:0.06

		Th-232:25.69 C:64.81

Fort St. Vrain	1.991

U-235:3.54 Si:5.96	------

		U-235:3.64 U-238:14.56

Melt and Dilute	3.00	A1:77.97 Gd:0.50 H:0.37	0.0009-4.73 c

		0:2.96

N Reactor	18.39	U-235:1.25 U-238:98.75	------

		0:12.12 U-233:4.57

Shippingport LWBR	9.71	U-234:0.06 U-238:0.02	------

		Th-232:83.23

		U-235:45.04 U-238:3.29

Shippingport PWR - zone 1	6.36

Ca:3.72 Zr:29.54 0:18.41	------

Shippingport PWR -zone 2	6.36	U-235:32.98 U-238:2.41	------

		Ca:4.15 Zr:39.98 0:20.48

		U-235:21.74 U-238:1.59

Shippingport PWR - zone 3	6.36

Ca:4.57 Zr:49.67 0:22.43	------

		U-235:2.61 U-238:85.53

TMI-2	10.42

0:11.86	------

		U-235:5.94 U-238:2.56

TRIGA	6.58	Zr:89.91 H:1.59

SOURCE: BSC 2004b, Table 5-3. Also, see BSC 2004b, Section 5.1.2 for fuel +
description.

8 Neutron absorber (Gd) contents in canister were varied. 1 vol% corresponds to+
 3 kg (CRWMS M&amp;O 2000a, p. 12)

b Neutron absorber (Gd) contents in canister were varied. 5 wt% corresponds to +
19.26 kg, which is the

maximum amount of gadolinium (CRWMS M&amp;O 1999a, p.21)

c Neutron absorber (Gd) contents in ingots were varied. 0.5 wt% corresponds to +
4.73 kg (BSC 2001c, p.3)

Naval Fuel Canisters

Per Assumption 3.2, no evaluations will be perfonned of naval fuel canisters +
since it is the

responsibility of the U. S. Department of the Navy.

5.2 CRITICALITY CALCULATIONS

The process and methodology for criticality safety analysis given in the +
Preclosure Criticality

Analysis Process Report (BSC 2004e, Section 2.2.7) were implemented in these +
calculations.

The sections below feature calculations perfonned in this document. As +
mentioned earlier,

results from previously perfonned applicable criticality evaluations are +
utilized in Section 6 and

are not included in sections below.

5.2.1 DOE Fuel Moderator Density Variations

Moderator density, which could vary from dry to fully moderated conditions +
under accident

conditions, was varied in the MCNP model over the range of 0.0 to 1.0 g/cm 3 . +
The calculations

were perforTned for all DOE ftiels considered in this document. They were each +
modeled as

single ca&quot;nisters with their outsides surrounded by concrete.

5.2.2 DOE Fuel Neutron Poison Variations

Variations in neutron poison loading were studied to cover any errors of its +
integration in the

basket structure or fuel mixture. The DOE fuels that have neutron poisons as +
part of their

interior basket structure are Enrico Fenni and FFTF. Melt &amp; Dilute fuel has+
 integrated neutron

poison in its fuel. Each of these DOE fuels were modeled as single canisters +
with their outsides

surrounded by concrete. The mass of the neutron absorbers was varied in +
accordance with the

range shown in Table 5.1-4.

5.2.3 Category 1 and 2 Event Sequences

At the present time, no Category I and 2 events have been identified for the +
Fuel Handling

Facility. The Internal Hazards Analysis for License Application document (BSC +
2004c)

identifies potential events that could lead to a criticality accident. While +
these potential events

have not yet been categorized into Category I and 2 or beyond Category 2 +
events, all of the

identified potential events will be considered in this document for +
conservatism.

Table 5.2-1 describes the potential criticality events for the FHF entrance +
vestibule and the

applicable criticality safety evaluation performed for each event. Table 5.2-2 +
describes the

potential criticality events for the FHF preparation room and the applicable +
criticality safety

evaluation performed for each event. Table 5.2-3 refers to potential +
criticality events associated

with CSNF assembly transfer and the FHF main transfer room, the fuel transfer +
bay, and the fuel

transfer room. Table 5.2-4 refers to potential criticality events for canister +
transfer in the FHF

main transfer room. Table 5.2-5 describes the potential criticality events for +
FHF WP closure.

Table 5.2-6 describes the potential criticality events for WP loadout in the +
FHF main transfer

room, preparation room, and entrance vestibule. Table 5.2-7 refers to potential+
 criticality events

for loaded MSC removal operations in the FHF main transfer room, preparation +
room, and

entrance vestibule. The supporting calculations for the potential criticality +
events are provided in

the subsections.

Table 5.2-1 Potential Criticality Events for the FHF Entrance Vestibule

Potential Event a	Criticality Safety

	Evaluation

Criticality associated with a raifcar (holding

a loaded cask) derailment or collision	Regulatory compliance with 10 CFR 50, 71+
 and 72 provides

followed by a load tipover or fall and	assurance of criticality safety for this+
 event.

rearrangement of the cask internals.

Criticality associated with an overturning or

collision involving an LWT or an OWT	Regulatory compliance with 10 CFR 50, 71 +
and 72 provides

holding a loaded cask and rearrangement	assurance of criticality safety for +
this event.

of cask internals.

Criticality associated with a drop or	Regulatory compliance with 10 CFR 50, 71 +
and 72 provides

slapdown of a cask and a rearrangement	assurance of criticaliiy safety for this+
 event.

of the container internals.

	Per Assumption 3.3, the MSC is similar in design to a NRC-certified

Criticality associated with a drop or	cask. There is no effect on the +
criticality control features of the

slapdown of an MSC and a rearrangement	system as a result of this event shown +
by the cask handling

of the container internals.	accident evaluation in BSC 2004a, Table 5.2-1. +
Furthermore, there

	is no moderator intrusion to make the configuration more reactive.

Criticality associated with a gantry crane	Per Assumption 3.3, the MSC is +
similar in design to a NRC-certified

holding an MSC collision followed by a	cask. There is no effect on the +
criticality control features of the

load drop or tipover and a rearrangement	system as a result of this event shown+
 by the cask tip-over

of the MSC internals	evaluation in BSC 2004a, Table 5.2-1. Furthermore, there +
is no

.	moderator intrusion to make the configuration more reactive.

&quot; BSC 2004c, Table 23b

Table 5.2-2 Potential Criticality Events for the FHF Preparation Room

Potential Event a	Criticality Safety

Evaluation

	Regulatory compliance with 10 CFR 50, ~1 and 72 provides

Criticality associated with a loaded cask or	assurance of criticality safety +
for this event. Also, there is no effect

loaded MSC collision or trolley derailment	on the criticality control features +
of the system as a result of this

followed by a load tipover or fall and a	event shown by the tip-over evaluation+
 in BSC 2004a, Table 5.2-1.

rearrangement of the cask internals.	Furthermore, there is no moderator +
intrusion to make the

	configurati n more reactive.

' BSC 2004c, Table 24b

Nuclear

im Calculation

Title: Fuel Handling Facility Criticality Safety Calculations

Document Identifier: 210-OOC-FHOO-00400-000-OOA Page 24 of 40

Table 5.2-3 Potential Criticality Events for the FHF Transfer RooMS b

Potential Event a	Criticality Safety

	Evaluation

	Since the transportation cask, for the SNF assembly drop into a

	transportation cask scenado, is NRC-certified to be 10 CFR 71

Criticality associated with a drop of an	compliant, the criticality evaluation +
performed for the cask

SN F assembly from the spent fuel transfer	certification is adequate to cover +
this event. Per Assumption 3.3,

machine into a cask, MSC, or WP and a	the MSC is similar in design to a NRC +
certified cask, and will also

rearrangement of the cask, MSC, or WP	be in compliance with 10 CFR 71. The WP +
is designed to

internals.	withstand credible hazards without significant rearrangement of the

	iuel (Minwalia 2003, Section 4.9.2.2.7) for the potential event

	involving the WP.

Criticality associated with a drop of an

SNF assembly from the spent fuel transfer	The drop could cause reconfiguration +
of the CSNF assembly.

machine and a rearrangement of the fuel	Section 5.2.3.1 of BSC 2004f evaluates +
the keff of a reconfigured,

rods that comprise the assembly due to	fully flooded, CSNF and it remains +
safely below 0.9.

impact.

	Regulatory compliance with 10 CFR 50, 71 and 72 provides

	assurance of criticality safety for the event with the transportation

Criticality associated with the drop of	cask. Per Assumption 3.3, the MSC is +
designed to the same

heavy equipment onto a loaded, open	standards as the cask and consequently +
regulatory compliance

cask, MSC, or WP and a rearrangement of	provides assurance of criticality +
safety. The WP is designed to

the container internals.	withstand credible hazards without significant +
rearrangement of the

	fuel (Minwalia 2003, Section 4.9.2.2.7) for the potential event

	involving the WP.

	Fuel assembly misloading is not an issue for &quot;out-of-package&quot;

Criticality associated with a misload of a	criticality, as the criticality +
evaluations (BSC 2004a &amp; BSC 2004f)

WP or and MSC.	are based on 5% maximum enrichment and no burnup credit is

	taken.

	Criticality evaluations for fully flooded conditions of a MSC (BSC

Docking dng leaking water into an	2004a, Section 6.3) and 5% maximum enrichment+
 with no burnup

unsealed loaded cask or MSC leads to a	credit taken shows that there is no +
criticality concern. Further,

criticality.	moderator intrusion studies shows (BSC 2004f, Section 6.1) that

	fully flooded conditions are bounding.

a BSC 2004c, Table 26b

b The events are associated with CSNF assembly transfer and applies to all FHF +
transfer rooms (i.e., main transfer

room, fuel transfer bay, and fuel transfer room)

Table 5.2-4 Potential Criticality Events for the FHF Main Transfer ROOM b

Potential Event a	Criticality Safety

	Evaluation

Criticality associated with a drop of a

loaded cask or MSC from the main	Per Assumption 3.3, the MSC is designed to the+
 same standards as

transfer room overhead crane and a	the cask and consequently regulatory +
compliance with 10 CFR 50,

rearrangement of the cask or MSC	71 and 72 provides a.ssurance of criticality +
safety.

internals.

	Per Assumption 3.2 it is the responsibility of the U.S Department of

	the Navy to ensure criticality safety for the naval SNF canister.

Criticality associated with a drop of a DPC,	Regulatory compliance with 10 CFR +
50, 71 and 72 provides

a DOE SNF canister, a naval SNF	assurance of criticality safety for the event +
with the DPC. Section

canister, or a DOE HLW canister and a	5.2.3.1 of this document demonstrates +
that a rearrangement of fuel

rearrangement of canister internals.	for the DOE canisters does not increase +
the reactivity. In addition,

	Section 5.2.3.2 shows that losing the skirt, or impact limiter, of the

	canister due to a drop also not increase the reactivity.

	Regulatory compliance with 10 CFR 50, 71 and 72 provides

	assurance of criticality safety for the event with the transportation

Criticality associated with the drop of	cask. Per Assumption 3.3, the MSC is +
designed to the same

heavy equipment onto a loaded, open	standards as the cask and consequently +
regulatory compliance

cask, MSC, or WP and a rearrangement of	provides assurance of criticality +
safety. The WP is designed to

the container internals.	withstand credible hazards without significant +
rearrangement of the

	fuel (Minwalia 2003, Section 4.9.2.2.7) for the potential event

	involving the WP.

	Fuel assembly misloading is not an issue for&quot;out-of-package&quot;

Criticality associated with a misload of a	criticality (BSC 2004e, p. 1), as +
the criticality evaluations (BSC

WP or an MSC.	2004a &amp; BSC 2004f) are based on 5% maximum enrichment and

	no bumup credit is taken.

' BSC 2004c, Table 27b

b The potential events are associated with canister transfer

Table 5.2-5 Potential Criticality Events for FHF Waste Package Closure

Potential Event a	Criticality Safety

	Evaluation

Criticality associated with a trolley holding	The WP is designed to withstand +
credible hazards without

a sealed or unsealed WP derailment	significant rearrangement of the fuel +
(Minwalla 2003, Section

followed by a load tipover or fall and	4.9.2.2.7).

rearrangement of the container intemals.

Criticality associated with a drop of a	The WP is designed to withstand +
credible hazards without

loaded, unsealed WP from the main	significant rearrangement of the fuel +
(Minwalla 2003, Section

transfer room overhead crane and a	4.9.2.2.7).

rearrangement of the container intemals.

Criticality associated with the drop of	The WP is designed to withstand +
credible hazards without

heavy equipment onto a loaded, unsealed	significant rearrangement of the fuel +
(Minwalla 2003, Section

WP and a rearrangement of the container	4.9.2.2.7).

internals.

a BSC 2004c, Table 28b

Table 5.2-6 Potential Criticality Events for Waste Package Loadout

Potential Event	Criticality Safety

	Evaluation

Criticality associated with a trolley holding	The WP is designed to withstand +
credible hazards without

a sealed WP derailment followed by a load	significant rearrangement of the fuel+
 (Minwalla 2003

Section

tipover or fall and rearrangement of the	,

4.9.2.2.7).

WP internals.

Criticality associated with a drop or	The WP is designed to withstand credible +
hazards without

collision of a sealed WP and a	significant rearrangement of the fuel (Minwalia +
2003, Section

rearrangement of the container internals.	4.9.2.2.7).

Criticality associated with a slapdown of a	The WP is designed to withstand +
credible hazards without

sealed WP and a rearrangement of the	significant rearrangement of the fuel +
(Minwalla 2003, Section

container internals.	4.9.2.2.7).

oou zuu4c, i aDie zvD

b The events applies to the FHF main transfer room, preparation room, and +
entrance vestibule

Table 5.2-7 Potential Criticality Events for Loaded MSC Removal OperationS b

Potential Event a	Criticality Safety

	Evaluation

Criticality associated with an MSC trolley	Per Assumption 3.3, the MSC is +
similar in design to a NRC-certified

collision or trolley derailment followed by a	cask. There is no effect on the +
criticality control features of the

load tipover or fall and a rearrangement of	system as a result of this event +
shown by the cask tip-over

the MSC internals.	evaluation in BSC 2004a, Table 5.2-1. Furthermore, there is +
no

	moderator intrusion to make the configuration more reactive.

Criticality associated with a drop or	Per Assumption 3.2, the MSC is similar in+
 design to a NRC-certified

slapdown of a loaded MSC from an	cask. There is no effect on the criticality +
control features of the

overhead crane and a rearrangement of	system as a result of this event shown by+
 the cask handling

cask internals	accident evaluation in BSC 2004a, Table 5.2-1. Furthermore, +
there

.	is no moderator intrusion to make the configuration more reactive.

	Regulatory compliance with 10 CFR 50, 71 and 72 provides

	assurance of cdticality safety for the event with the transportation

Criticality associated with the drop of	cask. Per Assumption 3.3, the MSC is +
designed to the same

heavy equipment onto an unsealed MSC	standards as the cask and consequently +
regulatory compliance

and a rearrangement of the container	provides assurance of criticality safety. +
The WP is designed to

internals.	withstand credible hazards without significant rearrangement of the

	fuel (Minwalla 2003, Section 4.9.2.2.7) for the potential event

	involving the WP.

- t3,-jU ZUU4C, i anie 3ob

b The events applies to the FHF main transfer room, preparation room, and +
entrance vestibule

5.2.3.1 Rearrangement of DOE Canister Iniernals

Potential Category I and 2 event sequences includes drops of DOE canisters +
causing a

rearrangement of canister intemals (see Table 5.2-4). For this purpose fuel pin+
 pitch variations

(i.e., increased and deceased pin pitch) were modeled for the applicable DOE +
canisters in MCNP

to simulate this potential event. The DOE canisters were modeled as single +
canisters with

flooded inside conditions and concrete reflection outside, representing the +
most reactive

configuration. Table 6-1 of BSC 2004b demonstrates that for nearly all DOE fuel+
 types, k,ff

increases when the outside is surrounded by concrete instead of water. The +
reason for this is that

concrete is a better reflector than water in the enviromnent that is being +
modeled. Table 5.2-8

lists the keff as a function of pin pitch for each applicable DOE fuel types in+
 their respective

canisters. It can be seen that varying the pin pitch does not cause a +
criticality concem. Further,

the pin pitches that produce the maximum keff are used for criticality +
evaluations under nonnal

conditions (BSC 2004b, Section 6.1). The results presented in Table 5.2-8 are +
also illustrated in

Figure 5.2-1, which is generated in D0E_fue1.x1s. Note that the N Reactor fuel +
denoted &quot;A&quot; in

the figure refers to Mark I A fuel and &quot;B&quot; refers to Mark IV fuel.

Table 5.2-8 DOE Fuel Pin Pitch Variations

Pitch

(cm)	Koff

I	St. Dev

I	MCNP files a	Pitch

(cm)	Keff

I	St. Dev	MCNP files'

	Enrico Fermi		FFTF

0.01	0.81681	0.00077	efwdsOl.out	0.6	0.83516	0.00094	ffwds06.out

0.02	0.83621	0.00076	efwds02.out	0.8	0.82977	0.00097	ffwds08.out

0.03	0.85551	0.00076	efwds03.out	1.0	0.8512	0.00095	ffwdsl O.out

0.04	0.87361	0.00074	efwds04.out	1.2	0.86941	0.00091	ffwdsl 2.out

0.05	0.88478	0.00077	efwds05.out	1.4	0.87562	0.00102	ffwdsl4.out

0.06	0.88616	0.00078	efwds06.out	1.5	0.87592	0.00076	ffwdsl5.out

0.07	0.88564	0.00077	efwds07.out	1.6	0.87644	0.00103	ffwds16.out

0.08	0.88028	0.00079	efwds08.out	1.7	0.87629	0.0008	ffwdsl7.out

0.09	0.87757	0.00077	efwds09.out	1.8	0.87231	0.00094	ffwdsl8.out

0.1	0.87865	0.00085	efwdsl O.out	2.0	0.86517	0.00106	ffwds20.out

0.11	0.88474	0.00081	efwdsl 1.out		TRIGA

0.12	0.88278	0.0008	efwdsl2.out	6.03	0.83818	0.00112	trwds60.out

0.13	0.87078	0.0008	efwdsl 3.out	6.5	0.79473	0.00110	trwds65.out

0.14	0.85951	0.00083	efwds14.out	7.0	0.74618	0.00980	trwds70.out

0.15	0.85611	0.00075	efwdsl 5.out	7.5	0.69476	0.00100	trwds75.out

	N Reactor Mark IA Fuel		N R	eactor Mar	k IV Fuel

6.2	0.72258	0.00063	nrwds62a.out	6.2	0.71364	0.00058	nrwds62b.out

6.6	0.78310	0.00064	nrwds66a.out	6.6	0.77336	0,00060	nrwds66b.out

7.0	0.82639	0.00063	nrwds70a.out	7.0	0.81915	0.00058	nrwds70b.out

7.4	0.84729	0.00064	nrwds74a.out	7.4	0.84396	0.00059	nrwds74b.out

7.8	0.84929	0.00067	nrwds78a.out	7.8	0.85334	0.00056	nrwds78b.out

8.2	0.83357	0.00067	nrwds82a.out	8.2	0.84373	0.00058	nrwds82b.out

8.6	0.80056	0.00065	nrwds86a.out	8.6	0.82394	0.00057	nrwds86b.out

9.0	0.76490	1 0.00064	1 nrwds90a.out	9.0	0.80148 1	0.00055	1 nrwds90b.out

	TM	I-2 Type D Canister		TM	I-2 Type K Canister

1.1	0.61646	0.0009	tmwds 11 d.out	1.1	0.41605	0.0007	tmwdsl 1 k.out

1.3	0.77349	0.00097	tmwdsl3d.out	1.3	0.57354	0.00087	tmwdsl3k.out

1.5	0.85634	0.00097	tmwdsl5d.out	1.5	0.70058	0.00087	tmwdsl5k.out

1.7	0.85296	0.00091	tmwdsl 7d.out	1.7	0.77586	0.00099	tmwdsl7k.out

1.9	0.83542	0.00097	tmwdsl 9d.out	1.9	0.80811	0.00089	tmwdsl9k.out

2.1	0.78778	0.00096	tmwds2l d.out	2.1	0.78911	0.00097	tmwds2l Kout

2.3	0.74795	0.00086	tmwds23d.out	2.3	0.73864	0.00088	tmwds23k.out

2.5	0.69909	, 0.00084	, tmwds25d.out	2.5	0.68338 ,	0.0009	tmwds25k.out

a The input files to each run have the same name as the corresponding output +
file but without the out

b extension (e.g., the input file matching output file tmwdsl 1 d.out is tmwdsl+
 1 d).

Smallest possible physical spacing.

0.9

0.85

0.8

0.75

0.7

0.65

0.6

5.5

0.9

0.85

0.8

0.75

0.7

0.65

0.6

0.55

0.5

0.45

0.4

0

0.95

0.9

0.85

0.8

0.75

0.7

0.65

0.6

0

Figure 5.2-1 DOE Fuel Pin Pitches versus keff

6 6.5 7 7.5 8 8.5 9 9.5

Pitch (cm )

0.5 1 1.5 2 2.5 3

Pitch (cm)

0.05 0.1 0.15 0.2

Pitch (cm)

5.2.3.2 DOE Fuel Drop or Slapdown

In the event of a drop or slapdown, as discussed earlier in Section 5.2.3. 1, +
the lower skirts of the

canister might get damaged causing the interior fuel basket to sit directly on +
the bottom of the

canister. A calculation was perfonned with a single flooded TRIGA fuel canister+
 featuring a

neglected lower skirt (the area below the skirt was exchanged from water to +
concrete). Table 5.2-

9 shows the results and it can be seen by comparing this calculation to the +
calculation of an intact

TRIGA fuel canister that a lost or damaged lower skirt does not impact kff.

Table 5.2-9 TRIGA Fuel Canister With and Without Skirt

MCNP Model

Description	Kff ~	Standard

Deviation	-MCNP input &amp; output

files

TRIGA

Intact Lower Skirt	0.83804	0.00115	trwds60, trwds60.out

Damaged Lower Skirt	0.83804	0.00115	trwds60s, trwds6ft.out

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 CASK AND MSC CRITIALITY EVALUATION

Criticality evaluations performed in the Aging Facility Criticality Safety +
Calculations document

(BSC, 2004a) feature the Holtec HI-STORM 100 cask system as a representative +
cask. Per

Assumption 3.3, the MSC is similar in design to this NRC-licensed cask. The +
results presented

in Section 6 of BSC 2004a shows that there are no criticality concems +
associated with the Hi-

STORM cask or the MSC. The results lead to the following main conclusions (BSC +
2004a,

Section 6.5):

Both the PWR and BWR results consistently demonstrate that the conditions +
outside the Hl-

STORM overpack (e.g., spacing, moderation, and reflection) have no discemable +
impact on

the reactivity of the cask. This indicates that the casks are neutronically +
isolated and

consequently the cask orientation (e.g., vertical versus horizontal) will not +
matter.

9 Reactivity of the loaded casks decreases with reduction in moderator density.

Maximum reactivity is reached when the casks are fully flooded with water at +
full density

(1.0 g/CM3).

The scenarios considered in BSC 2004a covers the conditions featured in the +
FHF. For this

purpose, no additional criticality calculations were performed for a cask or +
MSC in the FHF.

6.2 DOE FUEL CANISTER CRITICALITY EVALUATIONS

Criticality evaluations were performed fo'r DOE fuel in the Canister Handling +
Facility Criticality

Safety Calculations document (BSC 2004b) were the results lead to the following+
 conclusions

applicable to FHF operations (BSC 2004b, Section 6.3):

Criticality is not a concem for any single DOE fuel canister under both nonnal +
and flooded

conditions.

Cn'ticality is not a concem for darnaged fuel resulting from a catastrophic +
drop of a single

fuel canister for any DOE type. This is based on the assumptions that interior +
basket, if

present, remains intact following a canister drop (Assumption 3.5) and that the+
 damaged fuel

inside the canister is dry (Assumption 3.6).

To verify that fully flooded conditions are the most reactive, additional +
calculations were

perforrned were the fuel moderator density was varied. The results from the +
calculations are

presented below in Section 6.2. L Further, the Melt &amp; Dilute has integrated+
 neutron poison (Gd)

in its ftiel while the Enrico Ferrni and FFTF fiiels have neutron absorbers in +
their interior baskets.

Variations in neutron poison loading were also studied to cover any errors of +
its integration in the

basket structure or fuel mixture. The results from the calculations are +
presented below in Section

6.2.2.

6.2.1 DOE Fuel Moderator Density Variations

Moderator density, which could vary from dry to fully moderated conditions +
under accident

conditions, has been varied over the range of 0.0 to 1.0 g/cm 3 for all DOE +
fuel types considered

in this document. Table 6.2-1 displays kff as a function of moderator density +
for the various

DOE fuel. It can be seen that the reactivity of the DOE fuel decreases with +
reduction in

moderator density. Consequently, it can be concluded that fully flooded +
conditions are the most

reactive. Figure 6.2- 1, generated in D0E_fue1.x1s, illustrates the results +
presented in Table 6.2- 1.

Table 6.2-1 K,.ff as a Function of Moderator Density Variations

Density

(g/cc)	K

-off I	St. Dev	MCNP fileS a

I	Keff	St. Dev	MCNP files a

I

	Enrico Fermi	FFTF

1.0	0.91063	0.00078	efwdsOft.out	0.92084	0.00074	ffwdsl5y.out

0.98	0.90684	0.00078	efwds698.out	0.91541	0.00078	ftdsWout

0.95	0.90194	0.00079	efwds695.out	0.90891	0.00082	ffwds95.out

0.90	0.88938	0.00079	efwds069.out	0.89624	0.00079	ffwds159.out

0.80	0.86651	0.00075	efwds068.out	0.87326	0.00075	ffwds158.out

0.50	0.78822	0.00076	efwds650.out	0.79055	0.00073	ffwds50.out

0.0	0.49484	0.00041 1	efddsOU.out	0.56406	0.00054 ,	ffwdsNy.out

	Fort St. Vrain	Melt &amp; Di	lute

1.0	0.84467	0.00093	fswdsOO.out	0.83129	0.00268	mdwdsOOz.out

0.98	0.84357	0.00092	fswds09B.out	0.82582	0.00253	mdwdsO98.out

0.95	0.84218	0.00095	fswds095.out	0.82558	0.00243	mdwds095.out

0.90	0.83603	0.00100	fswds009.out	0.81020	0.00261	mdwds009.out

0.80	0.82287	0.00101	fswds008.out	0.80444	0.00268	mdwds008.out

0.50	0.76258	0.00104	fswds05O.out	0.75754	0.0022	rndwds050.out

0.0	0.51526	0.00099	fsdds00.0ut	0.65896	0.00227	rnddds00z.out

Table 6.2-1 (cont.) K,ff as a Function of Moderator Density Variations

Density

(g/cc)

Keff

St. Dev

MCNP files'

K,.ff

St. Dev

T	MCNP fileS a

	N Reactor &quot;A&quot;	N Reactor &quot;B&quot;

1.0	0.84929	0.00067	nrwds78a.out	0.85334	0.00056	nrwds78b.out

0.98	0.84736	0.00062	nrwd98a.out	0.84978	0.00058	nrwd98b.out

0.95	0.84189	0.00066	nrwd95a.out	0.84547	0.00056	nrwd95b.out

0.90	0.83229	0.00064	nrwd789a.out	0.83613	0.00059	nrwd789b.out

0.80	0.81078	0.00060	nrwd788a.out	0.81506	0.00060	nrwd788b.out

0.50	0.70305	0.00056	nrwd50a.out	0.70516	0.00058	nrwd50b.out

0.0	0.34216	, 0.00038 ,	nrdds62a.out	0.34489	0.00034	nrdds62b.out

	Shippingpor	t LWBR		Shippingport PWR

1.0	0.87003	0.00101	slwds94.out	0.87972	0.00100	spwdsOO.out

0.98	0.85963	0.00112	slwds98.out	0.87777	0.00111	spwds98.out

0.95	0.84579	0.00110	slwds95.out	0.86895	0.00102	spwds95.out

0.90	0.82493	0.00105	slwds949.out	0.86277	0.00103	spwds009.out

0.80	0.77979	0.00104	slwds948.out	0.84687	0.00107	spwds008.out

0.50	0.61967	0.00096	slwds50.out	0.78996	0.00105	spwds50.out

0.0	0.22755	0.00045 ,	sldds78.out	0.23594	0.00044 ,	spddsOO.out

	TMI-2 &quot;	D&quot;	TMI-2 &quot;	K&quot;

1.0	0.85634	0.00097	tmwdsl5d.out	0.80811	0.00089	trnwdsl9k.out

0.98	0.84905	0.00091	tmwd98d.out	0.80082	0.00098	tmwd98k.out

0.95	0.83840	0.00094	tmwd95d.out	0.79464	0.00089	tmwd95k.out

0.90	0.81916	0.00106	tmwdl59d.out	0.78005	0.00093	tmwdl99k.out

0.80	0.77684	0.00100	tmwdl58d.out	0.74678	0.00097	tmwdl98k.out

0.50	0.60730	0.00094	tmwd50d.out	0.60890	0.00094	tmwd50k.out

0.0	0.21237	0.00035 ,	tmddsl 1 d.out	0.18203 ,	0.00032	trnddsl5k.out

		TRIGA

1.0	0.83818	0.00112	trwds60.out

0.98	0.83600	0.00105	trwds98.out	a The input files to each run have the

0.95	0.83750	0.00106	trwds95.out	same name as the corresponding

fil

i

h

h

0.90	0.83361	0.00104	trwds609.0ut	Output

e but w

t

out t

e out

extension (e

g

the input file matching

0.80	0.82977	0.00112	trwds608.out	.

.,

output file trwds6O.out is trwds60).

E5 0:::::

F	0.80371	0.00104	trwds50.out

0.0	0.70702	0.00093	trdds60.out

0.9

0.8

0.7

0.6

0.5

0.4

0.3

0.2

0.1

0

Figure 6.2-1 DOE Fuel Moderator Density Variations versus k,,ff

0.2 0.4 0.6 0.8 1

Density (g/CM3)

6.2.2 DOE Fuel Neutron Poison Variations

Variations in neutron poison loading were studied, to cover any errors of its +
integration in the

basket structure or fuel mixture, and Table 6.2-2 shows the variations in k,ff +
as a function of

neutron poison loading. Note that the Enrico Fenni and FFTF fuels have +
non-integrated neutron

poison (i.e., the neutron poison is fixed and a part of the basket structure) +
while the Melt &amp;

Dilute fuel has integrated neutron poison in its fuel. It can be seen ftom +
Table 6.2-2 that a large

decrease in neutron poison will not cause kff of a single canister to exceed +
the upper subcritical

limit.

Table 6.2-2 DOE Fuel Neutron Poison Variations

Neutron				Neutron

Poison	Keff	St. Dev	MCNP files a	Poison	Keff	St. Dev	MCNP files'

(Vol%)				(wt%)

	Enrico Fermi (non-integrated		FFTF (non-integrated neutron

	neutron poison)		poison)

1.0

0.88616

0.00078

efwds06

out	5.0

0

87592

0

00076

ffwdS15

out

(3 kg)			.	(19 .26 kg)	.	.	.

0.4	091063	0.00078	efwdsOft.out	0.1	0.92084	0.00074	ffwdsl5y.out

(1.2 kg)				(0-39 kg)

		Melt &amp; Dilute (integrated neutron

		poison)

'The input files to each run have the same	0 * 5	0.39017	0.00124	mdwdsOO.out

name as the corresponding output file but	(4.73 kg)

without the out extension (e.g., the input file	0.001	0

80902	00256

0	mdwdsOOx

out

matching output file mdwdsOO.out is mdwdsOO).	_(0.009 kg)	.	.	.

	0 , 0001	3129	0.00268	mdwdsOOz

out

	(9E-4 kg)	I		.

6.3 CATEGORY 1 AND 2 EVENT SEQUENCES ,

Category 1 and 2 event sequences were evaluated as presented in Section 5.2.3 +
and were found

to be within the criticality safety design limits.

6.4 CONCLUSIONS AND RECOMMENDATIONS

The FHF and its processes have been evaluated for criticality safety for normal+
 operations,

Category I and 2 event sequences. The results presented in this document lead +
to the following

conclusions and recommendations:

9 The criticality evaluations of the cask and MSC demonstrate that the +
conditions outside the

overpack (e.g., spacing, moderation, reflection) have no discemable impact on +
the reactivity

of the cask for both PWR and BVVR fuel (also noted in Section 6. 1).

Single DOE canisters are subcritical under both norinal and flooded conditions.+
 Concrete

surrounding the DOE canister is a better reflector than water and consequently +
produces a

higher keff-

The DOE fuel containing neutron poison (ie., Enrico Fermi, FFTF, and Melt &amp;+
 Dilute)

showed that losing a significant amount of neutron poison (due to manufacturing+
 errors etc.)

still promotes subcriticality.

Reactivity of the loaded cask, MSC or DOE canisters decreases with reduction in+
 moderator

density.

Maximum reactivity is reached when the cask, MSC or DOE canisters are fully +
flooded with

water at full density (1.0 g/cm 3).

Criticality events potentially occuning in the FHF do not compromise +
criticality safety. It

should be recognized that the potential criticality events have not yet been +
categon'zed into

Category I and 2 event sequences.

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available.

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Fuel and High-Level Radioactive Waste. Readily available.

AP-3.12Q, Rev. 2, ICN 2. Design Calculations and Analyses. Washington, D.C.: +
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Department of Energy, Office of Civilian Radioactive Waste Management. ACC:

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Idaho: U.S. Department of Energy, Idaho National Operations Office. TIC: +
252933.

LP-SI. I I Q-BSC, Rev. 0, ICN 0. Software Management. Washington, D.C.: U.S. +
Department of

Energy, Office of Civilian Radioactive Waste Management. ACC: +
DOC.20040225.0007.

Mecham, D.C., ed. 2004. Waste Package Component Design Methodology Report. +
000-30R-

WISO-00100-000-001. Las Vegas, Nevada: Bechtel SAIC Company. ACC:

ENG.20040617.0012.

Minwalla, H.J. 2003. Project Design Criteria Document. +
000-3DR-MGRO-00100-000-001. Las

Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20030402.0001.

NRC (U.S. Nuclear Regulatory Commission) 1997. SCALE, RSIC Computer Code +
Collection

(CCC-545). NUREG/CR-0200, Rev. 5. Washington, D.C.: U.S. Nuclear Regulatory

Commission. TIC: 235920.

NRC (U.S. Nuclear Regulatory Commission) 2000. Standard Review Plan for Spent +
Fuel Dry

Storage Facilities. NUREG-1567. Washington, D.C.: U.S. Nuclear Regulatory +
Commission.

TIC:247929.

Taylor, L.L. 2001. Fort Saint Vrain HTGR (ThlU Carbide) Fuel Characteristics +
for Disposal

Criticality Analysis. DOE/SNF/REP-060, Rev. 0. [Washington, D.C.]: U.S. +
Department of

Energy, Office of Enviromnental Management. TIC: 249783.

8. ATTACHMENTS

This calculation document includes three attachments:

ATTACHMENT I Listing of Computer Files (7 pages)

ATTACHMENT 11 One Compact Disk Containing All Files Listed in Attachment I (I +
of 1)

(0 pages)

ATTACHMENTIII Draft Sketches of FHF Moderator Control Areas (10 pages)

ATTACHMENT I

LISTING OF COMPUTER FILES

This attachment lists the input and output file names for the MCNP and Excel +
calculations. All

input and output are stored on an electronic medium (compact disc) in ASCII +
fonnat as part of

this attachment.

Date	Time	File Size	File Nwne

06/07/2004	04:23p	78,848	DOE - fuel.xls

06/08/2004	12:1 lp	5,961	trwds60

06/08/2004	12:1 lp	474,856	trwds60.out

06/08/2004	12:1 lp	5,999	trwds60s

06/08/2004	12:1 lp	476,188	trwds60s.out

02/26/2004	04:54p	11,90S	efwds068

02/26/2004	10:23p	654,776	efwds068.out

02/26/2004	04:54p	11,905	efwds069

02/26/2004	05:50p	654,776	efwds069.out

06/08/2004	12:09p	11,918	efwds650

06/08/2004	12:09p	654,956	efwds650.out

06/08/2004	12:09p	11,918	efwds695

06/08/2004	12:09p	654,642	efwds695.out

06/08/2004	12:09p	11,918	efwds698

06/08/2004	12:09p	654,743	efwds698.out

02/26/2004	04:55p	15,398	ffwds 15 8

02/26/2004	11:26p	676,100	ffwdsl58.out

02/26/2004	04:55p	15,398	ffwds I S 9

02/26/2004	06:51p	676,100	ffwdsl59.out

06/08/2004	12:10p	15,402	ffwds50

06/08/2004	12:10p	676,601	ffwds50.out

06/08/2004	12:10p	15,402	ffwds95

06/08/2004	12:10p	676,178	ffwds95.out

06/08/2004	12:10p	15,402	ffwds98

06/08/2004	12:10p	676,279	ffwds98.out

02/26/2004	04:57p	11,053	fswds008

02/26/2004	11:57p	520,020	fswds008.out

02/26/2004	04:57p	11,053	fswds009

02/26/2004	07:23p	520,020	fswds009.out

06/08/2004	12:10p	11,083	fswds05O

06/08/2004	12:10p	519,950	fswds05O.out

06/08/2004	12:10p	11,069	fswds095

06/08/2004	12:10p	520,066	fswds095.out

06/08/2004	12:10p	11,083	fswds098

06/08/2004	12:10p	519,950	fswds098.out

02/26/2004	04:59p	5,540	mdwds008

02/27/2004	12:00a	342,086	mdwdsOO8.out

Date	Time	File Size	File Name

02/26/2004	04:59p	5,540	mdwds009

02/26/2004	07:26p	342,086	mdwds009.out

06/08/2004	12: 1 Op	5,539	mdwds05O

06/08/2004	12: 1 Op	342,085	mdwds05O.out

06/08/2004	12: 1 Op	5,539	mdwds095

06/08/2004	12: 1 Op	342,085	mdwds095.out

06/08/2004	12: 1 Op	5,539	mdwds098

06/08/2004	12: 1 Op	342,220	mdwds098.out

06/08/2004	12: 1 Op	7,302	nrwd50a

06/08/2004	12: 1 Op	413,605	nrwd50a.out

06/08/2004	12: 1 Op	7,324	nrwd50b

06/08/2004	12: 1 Op	414,896	nrwd50b.out

02/26/2004	05:03p	7,302	nrwd788a

02/27/2004	12:20a	415,067	nrwd788a.out

02/26/2004	05:04p	7,324	nrwd788b

02/27/2004	12:39a	413,820	nrwd788b.out

02/26/2004	05:03p	7,302	nrwd789a

02/26/2004	07:45p	415,067	nrwd789a.out

02/26/2004	05:04p	7,324	nrwd789b

02/26/2004	08:04p	414,972	nrwd789b.out

06/08/2004	12: 1 Op	7,302	nrwd95a

06/08/2004	12: 1 Op	414,439	nrwd95a.out

06/08/2004	12: 1 Op	7,324	nrwd95b

06/08/2004	12: 1 Op	414,112	nrwd95b.out

06/08/2004	12: 1 Op	7,302	nrwd98a

06/08/2004	12: 1 Op	414,475	nrwd98a.out

06/08/2004	12: 1 Op	7,324	nrwd98b

06/08/2004	12: 1 Op	414,896	nrwd98b.out

06/08/2004	12: 1 Op	11,758	s1wds50

06/08/2004	12: 1 Op	491,718	s1wds50.out

02/26/2004	05:05p	11,741	slwds948

02/27/2004	01:07a	491,504	slwds948.out

02/26/2004	05:05p	11,741	s1wds949

02/26/2004	08:3 lp	491,504	slwds949.out

06/08/2004	12: 1 Op	11,758	slwds95

06/08/2004	12: 1 Op	491,765	slwds95.out

06/08/2004	12: 1 Op	11,758	slwds98

06/08/2004	12: 1 Op	491,718	slwds98.out

02/26/2004	05:06p	27,582	spwds008

02/27/2004	01:50a	796,106	spwdsOO8.out

02/26/2004	05:05p	27,582	spwds009

02/26/2004	09:15p	796,106	spwds009.out

06/08/2004	12: 1 Op	27,587	spwds05O

06/08/2004	12: 1 Op	796,371	spwds05O.out

Date	Time	File Size	File Name

06/08/2004	12: 1 Op	27,587	spwds095

06/08/2004	12: 1 Op	796,573	spwds095.out

06/08/2004	12: 1 Op	27,587	spwds098

06/08/2004	12: 1 Op	796,674	spwds098.out

02/26/2004	05:12p	6,457	tmwdl58d

02/27/2004	02:07a	445,923	tmwdl58d.out

02/26/2004	05:12p	6,457	tmwdl59d

02/26/2004	09:31p	445,916	trnwd159d.out

02/27/2004	06:54a	7,618	trnwd198k

02/27/2004	07:20a	467,153	tmwdl98k.out

02/27/2004	06:54a	7,618	tmwdl99k

02/27/2004	07:45a	467,153	tmwdl99k.out

06/08/2004	12: 1 Op	6,457	tmwd50d

06/08/2004	12: 1 Op	445,868	tmwd50d.out

06/08/2004	12: 1 Op	7,630	tmwd50k

06/08/2004	12: 1 Op	467,423	tmwd50k.out

06/08/2004	12: 1 Op	6,457	tmwd95d

06/08/2004	12: 1 Op	445,969	tmwd95d.out

06/08/2004	12: 1 Op	7,630	tmwd95k

06/08/2004	12: 1 Op	467,524	tmwd95k.out

06/08/2004	12: 1 Op	6,457	tmwd98d

06/08/2004	12: 1 Op	445,969	tmwd98d.out

06/08/2004	12: 1 Op	7,630	tmwd98k

06/08/2004	12: 1 Op	467,421	tmwd98k.out

06/08/2004	12: 1 Op	5,980	trwds50

06/08/2004	12: 1 Op	464,230	trwds50.out

02/26/2004	05:07p	5,989	trwds608

02/27/2004	02:22a	463,651	trwds608.out

02/26/2004	05:07p	5,989	trwds609

02/26/2004	09:46p	463,333	trwds609.out

06/08/2004	12: 1 Op	5,981	trwds95

06/08/2004	12: 1 Op	464,064	trwds95.out

06/08/2004	12: 1 Op	5,982	trwds98

06/08/2004	12: 1 Op	464,162	trwds98.out

06/08/2004	01:09P	12,243	efddsOIx

06/08/2004	01:09P	655,227	efddsOlx.out

06/08/2004	01:09P	11,878	efwdsOl

06/08/2004	01:09P	654,644	efwdsOl.out

06/08/2004	01:10P	11,878	efwds02

06/08/2004	01:10P	654,157	efwds02.out

06/08/2004	01:10P	11,878	efwds03

06/08/2004	01:10P	655,167	efwds03.out

06/08/2004	01:10P	11,878	efwds04

06/08/2004	01:10P	654,776	efwds04.out

Date	Time	File Size	File Name

06/08/2004	01:10P	11,878	efwds05

06/08/2004	01:10P	654,776	efwds05.out

06/08/2004	01:10P	11,878	efwds06

06/08/2004	01:10P	654,776	efwds06.out

06/08/2004	01:10P	11,879	efwds06x

06/08/2004	01:10P	654,776	efwds06x.out

06/08/2004	01:10P	11,878	efwds07

06/08/2004	01:10P	655,167	efwds07.out

06/08/2004	01:10P	11,878	efwds08

06/08/2004	01:10P	655,209	efwds08.out

06/08/2004	01:10P	11,878	efwds09

06/08/2004	01:10P	653,726	efwds09.out

06/08/2004	01:10P	11,877	efwds 10

06/08/2004	01:10P	654,458	efwds I O.out

06/08/2004	01:10P	11,878	efwds 11

06/08/2004	01:10P	654,442	efwds I Lout

06/08/2004	01:10P	11,878	efwds12

06/08/2004	01:10P	654,458	efwdsl2.out

06/08/2004	01:10P	11,878	efwds 13

06/08/2004	01:10P	654,458	efwdsl1out

06/08/2004	01:10P	11,878	efwds14

06/08/2004	01:10P	654,458	efwds14.out

06/08/2004	01:10P	11,878	efwds15

06/08/2004	01:10P	654,458	efwds I 5.out

06/08/2004	01:10P	15,490	ffdds06y

06/08/2004	01:10P	678,277	ffdds06y.out

06/08/2004	01:10P	15,378	ffwds06

06/08/2004	01:10P	509,731	ffwds06.out

06/08/2004	01:10P	15,378	ffwds08

06/08/2004	01:10P	509,731	ffwdsO8.out

06/08/2004	01:10P	15,366	ffwds 10

06/08/2004	01:10P	509,731	ffwdslO.out

06/08/2004	01:10P	15,378	fiRvds 12

06/08/2004	01:10P	509,731	ffwds 12. out

06/08/2004	01:10P	15,378	ffwdsl4

06/08/2004	01:10P	509,413	ffwdsl4.out

06/08/2004	01:10P	15,380	ffwds 15

06/08/2004	01:10P	675,782	ffwds 15. out

06/08/2004	01:10P	15,390	ffwds 15 y

06/08/2004	01:10P	676,100	ffwdsl5y.out

06/08/2004	01:10P	15,378	ffwds16

06/08/2004	01:10P	509,413	ffwds 16. out

06/08/2004	01:10P	15,380	ffwds 17

06/08/2004	01:10P	675,782	ffwds I Tout

Date	Time	File Size	File Name

06/08/2004	01:10P	15,378	ffwds 18

06/08/2004	01:10P	509,413	ffwds18.out

06/08/2004	01:10P	11,210	fsddsOO

06/08/2004	01:10P	522,282	fsddsOO.out

06/08/2004	01:10P	11,021	fswdsOO

06/08/2004	01:10P	520,066	fswdsOO.out

06/08/2004	01:10P	5,627	mdddsOOz

06/08/2004	01:10P	342,816	mdddsOOz.out

06/08/2004	01:10P	5,513	mdwdsOO

06/08/2004	01:10P	342,086	mdwdsOO.out

06/08/2004	01:10P	5,511	mdwdsOOx

06/08/2004	01:10P	342,086	mdwdsOOx.out

06/08/2004	01:10P	5,516	mdwdsOOz

06/08/2004	01:10P	342,086	nidwdsOOz.out

06/08/2004	01:10P	7,437	nrdds62a

06/08/2004	01:10P	404,105	nrdds62a.out

06/08/2004	01:10P	7,459	nrdds62b

06/08/2004	01:10P	417,210	nrdds62b.out

06/08/2004	01:10P	7,275	nrwds62a

06/08/2004	01:10P	401,851	nrwds62a.out

06/08/2004	01:10P	7,297	nrwds62b

06/08/2004	01:10P	414,972	nrwds62b.out

06/08/2004	01:10P	7,275	nrwds66a

06/08/2004	01:10P	401,867	nrwds66a.out

06/08/2004	01:10P	7,297	nrwds66b

06/08/2004	01:10P	413,784	nrwds66b.out

06/08/2004	01:10P	7,265	nrwds70a

06/08/2004	0-1: 1 Op	401,867	nrwds70a.out

06/08/2004	01:10P	7,287	nrwds70b

06/08/2004	01:10P	414,972	nrwds70b.out

06/08/2004	01:10P	7,275	nrwds74a

06/08/2004	01:10P	401,867	nrwds74a.out

06/08/2004	01:10P	7,297	nrwds74b

06/08/2004	01:10P	413,784	nrwds74b.out

06/08/2004	01:10P	7,274	nrwds78a

06/08/2004	01:10P	413,879	nrwds78a.out

06/08/2004	01:10P	7,296	nrwds78b

06/08/2004	01:10P	412,632	nrwds78b.out

06/08/2004	01:10P	7,275	nrwds82a

06/08/2004	01:10P	415,067	nrwds82a.out

06/08/2004	01:10P	7,297	nrwds82b

06/08/2004	01:10P	413,030	nrwds82b.out

06/08/2004	01:10P	7,275	nrwds86a

06/08/2004	01:10P	414,077	nrwds86a.out

Date	Time	File Size	File Name

06/08/2004	01:10P	7,297	nrwds86b

06/08/2004	01:10P	413,820	nrwds86b.out

06/08/2004	01:10P	7,265	nrwds90a

06/08/2004	01:10P	415,067	nrwds90a.out

06/08/2004	01:10P	7,287	nrwds90b

06/08/2004	01:10P	414,012	nrwds90b.out

06/08/2004	01:10P	11,919	sldds78

06/08/2004	01:10P	493,232	sldds78.out

06/08/2004	01:10P	11,699	s1wds94

06/08/2004	01:10P	492,126	slwds94.out

06/08/2004	01:10P	27,755	spddsOO

06/08/2004	01:10P	798,042	spddsOO.out

06/08/2004	01:10P	27,564	spwdsOO

06/08/2004	01:10P	796,573	spwdsOO.out

06/08/2004	01:10P	6,599	tmdds I I d

06/08/2004	01:10P	447,605	tmdds I I d. out

06/08/2004	01:10P	7,734	tmdds I 5k

06/08/2004	01:10P	461,523	tmddsl5k.out

06/08/2004	01:10P	6,449	tmwds 11 d

06/08/2004	01:10P	445,916	tmwds I I d. out

06/08/2004	01:09P	7,595	tmwds I I k

06/08/2004	01:09P	461,638	tmwds I I k. out

06/08/2004	01:09P	6,435	tmwds 13 d

06/08/2004	01:09P	445,923	tmwdsl3d.out

06/08/2004	01:09P	7,579	tmwds I 3k

06/08/2004	01:09P	459,192	trnwdsl3k.out

06/08/2004	01:09P	6,435	tmwds 15 d

06/08/2004	01:09P	445,916	tmwdsl5d.out

06/08/2004	01:09P	7,577	tmwds 1 5k

06/08/2004	01:09P	456,474	tmwdsl5k.out

06/08/2004	01:09P	6,435	tmwds I 7d

06/08/2004	01:09P	446,405	tmwds 1 7d.out

06/08/2004	01:09P	7,578	tmwds 1 7k

06/08/2004	01:09P	456,550	tmwdsl7k.out

06/08/2004	01:09P	6,435	tmwds 1 9d

06/08/2004	01:09P	446,412	tmwds I 9d.out

06/08/2004	01:09P	7,578	tmwds I 9k

06/08/2004	01:09P	469,228	tmwds I 9k. out

06/08/2004	01:09P	6,435	tmwds2ld

06/08/2004	01:09P	446,412	tmwds2 1 d.out

06/08/2004	01:09P	7,578	tmwds2 1 k

06/08/2004	01:09P	469,425	tmwds2lk.out

06/08/2004	01:09P	6,435	tmwds23d

06/08/2004	01:09P	446,503	tmwds23d.out

Date	Time	File Size	File Name

06/08/2004	01:09P	7,578	trnwds23k

06/08/2004	01:09P	469,432	tmwds23k.out

06/08/2004	01:09P	6,435	tmwds25d

06/08/2004	01:09P	446,412	tmwds25d.out

06/08/2004	01:09P	7,592	tmwds25k

06/08/2004	01:09P	467,160	tmwds25k.out

06/08/2004	01:09P	6,132	trdds60

06/08/2004	01:09P	466,348	trdds60.out

06/08/2004	01:09P	5,954	trwds65

06/08/2004	01:09P	463,914	trwds65.out

06/08/2004	01:09P	5,942	trwds70

06/08/2004	01:09P	463,811	trwds70.out

06/08/2004	01:09P	5,954	trwds75

06/08/2004	01:09P	464,333	trwds75.out