ML092310649

Calculation No. CAL-DSD-NU-000004, Rev. 00A, Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package

ML092310649
Person / Time
Site:	Three Mile Island, 06300001
Issue date:	03/23/2004
From:	Montierth L US Dept of Energy, Office of Civilian Radioactive Waste Mgmt (OCRWM)
To:	NRC/NMSS/DHLWRS/LID/PMBB
Shared Package
ML092310639	List: ML092310649 ML092310650 ML092310651 ML092310653 ML092310659 ML092310672 ML092310674
References
DOC.20040329.0002 CAL-DSD-NU-000004, Rev 00A
Download: ML092310649 (75)

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OOC.20040329.0002 1.QA: QA OCRWM DESIGN CALCULATION OR ANALYSIS COVER SHEET

2. Page I of?4

3. System 14. Document Identifier DOE Spent Nuclear Fuel CAL-DSD-NU-OOOOO4 REV OOA

5. Tille Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package

6. Group Risk and Criticality

7. Document Status Designation 181 Preliminary o Anal o Cancelled

8. Notes/Comments Attachments Total Number of Pages Attachment I: CD 1 CD-ROM Attachment n: 1 RECORD OF REVISIONS,

11. 12. 13. 14. 15. 16.

9. 10. 17.

Total # Last Originator Checker aER Approved/Accepted No. Reason For Revision ofPgs. Pg.# (Print/Sian/Date) (PrlntlSlgn/Oate) (PrintlSlgn/Oate) Date (PrintlSianl OOA ~ilia1 issue 75 11-1 I ~ Mootierth Haria R. Radulescu Darrell K. Svalstad Daniel A. Thomas

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1) ECN-001, DATED 06/01/2005

2) ECN-002, DATED 07/27/2005 ,

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 3 of74 CONTENTS Page I. PURPOSE 7

2. METHOD 8

3. ASSUMPTIONS 9

4. USE OF COMPUTER SOFTWARE AND MODELS 11 4.1 SOFTWARE 11 4.1.1 MCNP 1I

5. CALCULATION 12 5.1 WASTE PACKAGE COMPONENTS DESCRIPTION 12 5.1.1 TMI-2 Spent Nuclear FueL 13 5.1.2 TMI-2 Fuel Canisters 14 5.1.3 Description of DOE SNF Canister. 17 5.1.4 High-Level Waste Glass Pour Canister 19 5.1.5 Chemical Description of Borosilicate Glass 19 5.1.6 Waste Package Description 19 5.2 MATERIALS DESCRIPTION 21 5.3 FORMULAS 26 5.4 INTACT MODE CRITICALITY CALCULATIONS 27 5.4.1 Treatment ofTMI Fuel and Canisters 27 5.4.2 TMI Fuel Pellet and Other Modeling Details 29 5.5 DEGRADED MODE 32 5.5.1 Components Degrades Within the Intact SNF Canister 34 5.5.2 Internal Components of the Waste Package Outside SNF Canister Degrade 35 5.5.3 All Components Have Degraded 40

6. RESULTS 41 6.1 INTACT MODE , 42 6.1.1 Loose TMI Fuel Pellets in TMI Fuel Canisters and Other Modeling Details 42 6.2 DEGRADED MODE 54 6.2.1 Inner Components of the SNF Canister Degrade First.. 55 6.2.2 Outer Components of the Waste Package are Degraded (Outside SNF Canister)58 6.2.3 TMI Fuel Surrounded by Post-breach Clay in the Waste Package 69 6.3 SUMMARy 70

7. REFERENCES 71 7.1 DOCUMENTS CITED 71 7.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES 72 7.3 SOURCE DATA 73

8. ATTACHMENTS 74

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 4 of74 TABLES Page Table 1. TMI Maximum Uranium Mass (both Total Uranium and U-235) 17 Table 2. Waste Package Dimensions a and Material Specifications 20 Table 3. Composition and Density of Stainless SteeI304L. 21 Table 4. Composition and Density of Savannah River Site High-Level Waste Glass 22 Table 5. Composition and Density of Stainless Steel 316L. 23 Table 6. Composition and Density of Alloy 22 23 Table 7. Composition and Density of Carbon Steel 516 Grade 70 23 Table 8. Composition and Density of Dry Tuff.. 24 Table 9. Pre-Breach Clay Compositions a,b .24 Table 10. Post-Breach Clay Compositions 25 Table 11. Alternative Post-Breach Clay Composition .25 Table 12. Cylindrical TMI Fuel Pellets in the Knockout (KG) Canister. .42 Table 13. Spherical TMI Fuel Pellets in the Knockout (KG) Canister a without Internals 45 Table 14. Spherical TMI Fuel Pellets in the Knockout (KG) Canister a with Internals 47 Table 15. Spherical TMI Fuel Pellets in Dry Knockout (KG) Canister a without Internals .49 Table 16. Variations of Cases with Spherical Fuel Pellets in the Knockout (KG) Canister 50 Table 17. Cylindrical TMI Fuel Pellets in the Fuel Canister (D-type) 53 Table 18. Homogenized TMI Fuel in the Knockout (KG) Canister (without Internals) 54 Table 19. TMI Fuel Pellets in SNF Canister (Degraded Sleeve and KG Canister) 56 Table 20. TMI Fuel Pellets in Intact SNF and TMI Canisters Surrounded by Pre-breach Clay 60 Table 21. TMI Fuel Pellets in Intact SNF Canister Surrounded by Dry Pre-breach Clay (Degraded TMI Canister and Sleeve) 61 Table 22. TMI Fuel in Intact SNF Canister Containing Goethite Mixed with Water Surrounded by Dry Pre-breach Clay 63 Table 23. TMI Fuel in Intact KG Canister with Degraded SNF Canister Surrounded by Pre-breach Clay 65 Table 24. TMI Fuel Pellets Form Array Surrounded by Layers of Goethite and Pre-breach Clay in the Waste Package : 66 Table 25. TMI Fuel Pellets are Surrounded with the Post-breach Clay 69 Table 26. Degraded TMI Fuel Mixed with Post-breach Clay 70

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 5 of 74 FIGURES Page Figure 1. Typical Pressurized Water Reactor Fuel Assembly 14 Figure 2. Schematic Cross-section of Three Mile Island Unit 2 Canister Types 15 Figure 3. Cross-sectional Schematic ofthe Fuel Canister (D-type) Inserted in an Outer Can .16 Figure 4. DOE Spent Nuclear Fuel Standard Canister 18 Figure 5. Cross-sectional View of Typical Waste Package 21 Figure 6. Cross-sectional View of Knockout (KO) Canister at Different Elevations 29 Figure 7. Pellet Array Partially Fills KO Canister's Cross-section 30 Figure 8. Canisters Are in a Gravity Position in the waste package .31 Figure 9. Sleeve Is Collapsed in the SNF Canister. 31 Figure 10. Cross-sectional View of the TMI Fuel Canister 32 Figure 11. Cross-sectional View of the SNF Canister. 35 Figure 12. Cross-sectional View of an Intact SNF Canister Centered in Clay Formed from the Degradation of Components External to the Canister 36 Figure 13. Intact TMI Canister Surrounded by Goethite Trapped in Pre-breach Clay in the Waste Package 38 Figure 14. Fuel Pellets Surrounded by Goethite and Pre-breach Clay from Completely Degraded Components in the Waste Package 39 Figure 15. Similar Configuration as Shown in Figure 14 but Fewer Pellet Stacks in Fuel Array 40

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 7 of 74

1. PURPOSE The objective of these calculations is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) Three Mile Island - Unit 2 (TMI-2) spent nuclear fuel (SNF) in canisters. This analysis evaluates codisposal in a 5-Defense High-Level Waste (5-DHLW/DOE SNF) Long Waste Package (Civilian Radioactive Waste Management System Management and Operating Contractor [CRWMS M&O] 2000b, Attachment V), which is to be placed in a potential monitored geologic repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (kerr) for both intact and degraded mode internal configurations of the waste package.

These calculations will support the analysis that will be performed to demonstrate the technical viability for disposing of low-enriched spent nuclear fuel at yucca mountain. There are no limitations on the use of the results of this calculation.

This calculation is subject to the Quality Assurance Requirements and Description (DOE 2003a) as it addresses the codisposal viability ofTMI-2 SNF at Yucca Mountain. This document is prepared in accordance with AP-3.12Q, Design Calculations and Analyses, and AP-3.l5Q, Managing Technical Product Inputs.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 8 of74

2. METHOD The method to perform the criticality calculations consists of using MCNP Version 4B2LV (CRWMS M&O 1998a, CRWMS M&O 1998b) to calculate the effective neutron multiplication factor of the waste package. The calculations are performed using the continuous-energy cross section libraries, which are part of the qualified code system MCNP 4B2LV (CRWMS M&O 1998a, CRWMS M&O 1998b). All calculations are performed with the most reactive fissile concentration that bounds the beginning-of-life (BOL) and end-of-life (EOL) TMI fuels.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 9 of 74

3. ASSUMPTIONS 3.1 For the degraded mode criticality calculations, it is assumed that the iron in the stainless steel degrades to goethite (FeOOH) rather than hematite (Fe203). The basis of this assumption is that it is conservative to consider goethite rather than hematite since hydrogen (a moderator) is a component of goethite. All the other constituents of stainless steel are neglected since they are neutron absorbers, and hence their absence provides a conservative (higher) value for the kerr of the system. This assumption is used throughout Section 5.

3.2 Ba-138 cross sections are used instead ofBa-137 cross sections in the MCNP input since the cross sections of Ba-137 are not available in the MCNP 4B2LV cross section libraries. The basis of this assumption is that it is conservative since the thermal neutron capture cross section and the resonance integral of Ba-137 (5.1 and 4 bam, respectively [Parrington et aI.

1996, p. 34]) are greater than the thermal neutron capture cross section and the resonance integral of Ba-138 (0.43 and 0.3 bam, respectively [Parrington et aI. 1996, p. 34]). This assumption is used throughout Section 5.

3.3 The most reactive fissile enrichment of 3 wt% is used for the TMI fuel to bound the enrichment of the most highly loaded fuel canister. This selected fuel enrichment (3 wt.%) is larger than the enrichments used in any intact fuel assembly. The basis of this assumption is that the selected enrichment is conservative since it maximizes the fissile isotope (U-235) content while minimizing the effect of neutron absorption (U-238). This assumption is used throughout Section 5.

3.4 Al cross sections are used instead of Zn cross sections in the MCNP input since the cross sections of Zn are not available in the MCNP 4B2LV cross-section libraries. The basis of this assumption is that it is conservative since the thermal neutron capture cross section and the resonance integral of Zn (Parrington et aI., 1996, p. 24) are greater than the thermal neutron capture cross section and the resonance integral of Al (Parrington et aI., 1996, p. 21). This assumption is used throughout Section 5.

3.5 Water is always assumed present and is assumed to fill void spaces in the fuel pellet arrays and canisters. The basis of this assumption is that it is conservative since this allows optimal moderation and therefore more reactive configurations. This assumption is used throughout Section 5.

3.6 An interim critical limit of 0.97 is assumed. The basis for this assumption is that once the criticality model has been validated for TMI-2 SNF, the critical limit is expected to be higher than 0.97. This value is used in determining whether the criticality concerns of any scenario are satisfied but is not used directly in the calculations. This assumption is used throughout Section 6.

3.7 The composition of the Hanford HLW glass is assumed to be the same as the Savannah River Site glass composition. The basis for this assumption is that the predicted composition of

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 10 of74 Savannah River Site glass is known (CRWMS M&O 1999a, p. 7). This assumption is used throughout Section 5.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 11 of74

4. USE OF COMPUTER SOFTWARE AND MODELS 4.1 SOFTWARE The commercial off-the-shelf software MS EXCEL Version 2000 SR-I installed on a personal computer (PC) Dell Optiplex GX260 operating under Windows 2000 operating system, was used for performing arithmetical manipulations in a spreadsheet type environment. Microsoft EXCEL Version 2000 SR-I is an exempt software application in accordance with LP-SI.1IQ-BSC, Section 2.1.1. The developed spreadsheet files are included in Attachment I. The spreadsheets contain sufficient information to allow an independent check to reproduce or verify the results.

4.1.1 MCNP The MCNP code is used to calculate the keff of the waste package. The software specifications are as follow:

• Program Name: MCNP

• Version/Revision Number: Version 4B2LV

• Status/Operating System: QualifiedlHP-UX B.I 0.20

• Computer Software Configuration Item Number: 30033 V4B2LV

• Computer Type: Hewlett Packard (HP) 9000 Series Workstations

• Computer processing unit number: Software is installed on the INEEL workstation "bigdog" whose INEEL Tag number is 336829 The input and output files for the various MCNP calculations are included in Attachment I, (Attachment II gives the list of the files on Attachment I). The calculation files described in Sections 5 and 6 are such that an independent repetition of the software use may be performed.

The MCNP software used is: (a) appropriate for the application of research and commercial reactor keff calculations, (b) used only within the range of validation as documented in CRWMS M&O (1998a), (c) obtained from the Software Configuration Management in accordance with Administrative Procedure AP-SI.IQ, Software Management.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 12 of74

5. CALCULATION This section describes the calculations performed to calculate the keff of an intact and a degraded waste package containing high-level waste material and TMI spent nuclear fuel. Section 5.1 describes the waste package and its contents. Section 5.2 gives the composition of the materials used in this calculation. The basic formulas used in this calculation are listed in Section 5.3. The different intact configurations of a waste package are outlined in Section 5.4. Section 5.5 describes calculations performed to characterize the degraded configurations of a waste package. The MCNP input and output files developed for this section are presented in Attachment I. The spreadsheet used to prepare the MCNP input files is given in Attachment I, file "tmi_cals.xls." The results of the calculations are presented in Section 6.

The Savannah River Site high-level waste glass degraded (pre-breach clay) compositions are from CRWMS M&O (2000a) and BSC (200la). The composition from CRWMS M&O (2000a) is for the Shippingport LWBR fuel. Since these fuels (Shippingport and TMI) share the same waste package externals, i.e., components external to the SNF canister, the pre-breach compositions would be the same. The Savannah River Site high-level waste glass composition and density are from CRWMS M&O (1999a) and Stout and Leider (1991), respectively. The Savannah River Site high-level waste glass canister dimensions are from Taylor (1997).

Avogadro's number is from Parrington et al. (1996). Atomic weights are from Parrington et al.

(1996) and Audi and Wapstra (1995).

The description of the TMI fuel and canisters is from the TMI Fuel Characteristics for Disposal Criticality Analysis report (DOE 2003b). All fuel and canister-related information is from this reference unless otherwise noted.

The tuff composition and the tuff density are taken from a previous calculation (CRWMS M&O 2001, Attachment II, spreadsheet "Tuff Composition.xls").

This calculation is based in part on technical information given in DOE (2003b). The fuel group is identified by the National Spent Nuclear Fuel Program, and a 'representative' fuel type within that group is used to establish limits, e.g. burnup, fissile content, weights, dimensions.

The number of digits in the values cited herein may be the result of a calculation or may reflect the input from another source; consequently, the number of digits should not be interpreted as an indication of accuracy.

The metric units used in this document are calculated using the English units as given in DOE (2003b). The differences that might exist between the metric units calculated and the metric units cited in DOE (2003b) have no effect on the calculation and should not be interpreted as an indication of accuracy.

5.1 WASTE PACKAGE COMPONENTS DESCRIPTION

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 13 of74 5.1.1 TMI-2 Spent Nuclear Fuel The typical fuel assembly used in the TMI-2 reactor was a Babcock & Wilcox 15 x 15-rod array inside a 216.81 mm (8.536 in.) square envelope (see Figure 1 for a depiction of a typical pressurized water reactor [PWR] assembly). Although this array gives a total count of225 rods per assembly, the manufacturer specification indicates that typically only 208 of the rods are filled with uranium oxide pellets. While the total assembly is 4206.88 mm (165.625 in.) long, and the individual rod lengths are 3903.47 mm (153.68 in.), the active rod length is listed as 3601.72 mm (141.8 in.). The uranium oxide pellets stack constitutes the active rod length. The uranium fuel matrix is contained in pellets that are 9.398 mm (0.37 in.) in diameter and 11.049 mm (0.435 in.)

long. If the active length is divided by this pellet length of 11.049 mm (0.435 in.), this gives almost 326 pellets per rod, which is reduced to 325 pellets to give a (conservative) slightly larger fissile mass per pellet. This gives a total number of (208 x 325=) 67,600 pellets per assembly. The maximum beginning-of-life (BOL) U-235 content of 13.72 kg in any TMI-2 assembly provides the basis for the criticality analysis. The maximum enrichment of the uranium fuel is given as 2.96 wt%. Enrichments of 2.64 wt% and 1.98 wt% are reported for pellets in some of the PWR assemblies. The void fraction in the fuel matrix is given as 7.5%. (The terms "TMI-2" and "TMI" are used inter-changeably in this report.)

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 14 of74 03-GA50387-13 Figure 1. Typical Pressurized Water Reactor Fuel Assembly (overall dimensions may differ slightly based on manufacturer and are not necessarily those used in this analysis).

5.1.2 TMI-2 Fuel Canisters The TMI-2 canisters were fabricated for use in recovery and cleanup of the reactor core after the TMI-2 accident. The design of the TMI-2 canisters is such that not more than a single commercial 15 x 15 Babcock & Wilcox PWR assembly could be installed in anyone canister. The highest reported fissile loading in any TMI-2 canister is only 73.3% of a BOL fissile load for the most highly loaded PWR assembly. The defueling operations of the TMI-2 core used three types of canisters (see Figure 2), each with its own particular designation and numerical code.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 15 of74 The canisters with the "D" designator (defueling) were used to contain the bulk of core materials of a size that enabled grappling, could be picked up, or remotely handled. The "K" designator relates to the knockout (KO) canisters that were used in association with wet-vacuuming operations to remove loose debris that did not lend itself to physical grappling. There are only a total of 12 KO canisters, though one is reported to contain the highest total uranium loading of any TMI-2 canister.

Subsequent and downstream collection of the vacuumed debris stream then passed through a filter canister with an "F" designator. These filter canisters represent the least heavily loaded for either total uranium or U-235 per package and, in several cases, are reported to have no reportable uranium. A schematic representation of each type TMI-2 canister is shown in Figure 2.

Upper closure head with bolts Drain connector In Out Screen Drain tube Drain tube Tie rod Inlet pipe 149.75" Low density Support concrete mix Poison spiders (8) rods Tie rod Poison rod

~

Module end caps Fuel (F-Type) Knockout (K-Type) Filter (F-Type) 03-GA50387-14 Diagram of the the TMI-2 canister types Figure 2. Schematic Cross-section of Three Mile Island Unit 2 Canister Types.

The basic structure of the TMI canister centers on a l4-in. Schedule 10 (0.25-in. wall thickness) pipe. The bottom of the canister is a reversed-dish head with a 0.375-in. thickness. The top of the canister is a 4-in.-thick metal plate with penetrations suitable for hydraulic loading and dewatering.

All structural canister materials used Type 304L stainless steel for canister construction.

The D-type canisters, also referred to as "fuel canisters," have a box structure located in the center of the canister that has internal dimensions of 231.78 mm (9.125 in.) square and 3465.51 mm (1367/16 in.) long. The chord sections between the internal box and the inside of the TMI canister are filled with LiCon'. This is a low density (1 g/cm3) concrete mixture consisting of 60% Alcoa

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 16 of74 CA-25C refractory cement, 11 % glass microspheres, and 29% water by weight. This mixture was intended to create a solid filler with an approximate density of 1 g/cm3

• Its composition is given in Attachment I, spreadsheet "tmi_calcs.xls," sheet "Materials." A cross-sectional schematic of the canister is seen in Figure 3.

Fuel Pellets and H20 Gap between Fuel Canister and Outer Canister Figure 3. Cross-sectional Schematic of the Fuel Canister (D-type) Inserted in an Outer Can.

Materials of construction employ a variety of the 300 series stainless steels, and those in tum are dominated by the use of 304L type stainless. The internal sleeve is a seal-welded sandwich of 304L stainless steel that completely encompasses the BoraFM (boron aluminide) layer.

For the K-type canisters the internal assembly is designed to support five internal tubes with B4C poisoning. The larger center "A" tube consists of a 73-mm (2.875 in.) diameter tube with a 7.9-mm (0.312 in.) wall thickness and a length of3371.85 mm (132.75 in.). This center tube has an internal tube of 53.975 mm (2.125 in.) with a 1.6-mm (0.063 in.) wall thickness and is filled with B4 C pellets. The four outer "B" rods are centered approximately 63.50 mm (2.50 in.) each way on an X-Y plane from the canister centerline. These peripheral tubes have a 33.35-mm (1.313-in.) outer diameter with a 6.35-mm (0.25-in.) wall thickness with a minimum length of 3327.4 mm (131 in.).

The seven intermediate support plates (referred to as "spiders" in Figure 2) are held in place by the poison rods, and the plates are spaced approximately 406.4 mm (16 in.) apart with a 12.7-mm (0.5-in.) plate thickness. The single, bottom support plate has a 340.52-mm (13-13/32-in.) diameter and a 31.75-mm (1.25-in.) thickness.

From shipping data, the five canisters with the highest reported uranium and U-235 weights are given in Table 1. From this data, the nominal enrichments of the canisters containing the highest uranium and U-235 masses are 2.13 wt% (K506) and 2.67 wt% (Dll9), respectively. The total debris mass at time of packaging and the corresponding masses for a new assembly are also listed.

The loading of a new assembly (BOL) is also included in the table.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 17 of74 Table 1. TMI Maximum Uranium Mass (both Total Uranium and U-235).

Interim Nine Highest Storage Nine Highest Interim Uranium Uranium U-235 Storage U* Corresponding Canister 10 Masses, Mass, Masses, 235 Mass, Total Pu, Debris Mass, Number kg a kg b kg a kg b kg ka c d K506 441.9 + 99.9 441.90 9.42 +/- 2.13 9.42 0.900 842.18 +/- 168.3 0330 402.7 +/-42f 402.70 8.58 +/- 0.92 8.58 0.820 767.35 +/- 9.1 c

0283 400.5 + 103 400.50 7.57 + 1.95 7.57 0.898 0+0 c

0331 399.3 + 42.3 399.30 8.51 + 0.91 8.51 0.813 761.00 +/- 9.1 c

0361 398.8 + 42.3 398.80 8.50 + 0.91 8.50 0.812 760.09 +/-..9.1 d

0119 376.5 +/- 97.3 376.50 10.06 +/- 2.6 10.06 0.539 O+/-O d

0260 373.8 +/- 96.4 373.80 9.66 +/- 2.42 9.66 0.571 115.64+/- 102.9 d

0299 353.9 + 104.0 353.90 9.37 + 2.60 9.37 0.516 O+/-O d

0193 351.9 + 93.0 351.90 9.41 + 2.49 9.41 0.504 O+/-O 1 1 New assembly e 463.63 - 13.72 - - -

SOURCE: OOE 2003b, Table 4.

NOTES: a Reported masses at time of shipment from TMI to the INEEL.

b Masses reported for TMI-2 canisters transferred to interim, dry storage; these values provide the basis for criticality safety analysis for up to 12 TMI-2 canisters (per position) in Nuclear Regulatory Commission-licensed dry storage.

c Five highest uranium masses.

d Five highest U-235 masses.

e Comparison masses are for an intact, beginning-of-life fuel loading.

1 Oue to the rounding up of the conversion factor from pounds to kilograms, a slightly smaller total uranium mass of 463.55 kg was used in these calculations. The calculations are still conservative since the enrichment was rounded up to 3 wt% giving a larger than actual U-235 mass of 13.906 kg.

A discussion of previous criticality analyses done to support various systems or conditions is presented in DOE (2003b), which highlights the differences in assumptions and/or values used in the various TMI-2 models. There are differences in modeling assumptions because of the different conditions expected in storage and transportation versus those expected in monitored geologic repository disposal.

5.1.3 Description of DOE SNF Canister The description of the 15-ft DOE SNF canister (also referred to as the 18 in.-diameter DOE SNF canister) is taken from DOE design specifications (DOE 1999, p. 5, A-2 and A-3). The DOE SNF canister is a right circular cylinder made of stainless steel pipe (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.). A nominal internal length of the DOE SNF canister used for fuel loading is 4117.086 mm (162.09 in.);

minimum length dimension is 4114.8 mm (162 in.). (This nomina11ength is insignificantly different from the minimum length.) A sketch of the canister is shown in Figure 4.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 18 of74 I

i 1

Impact Plate 10-ft Canister:

(118.11:in [2999- I 118.01-1n mm 15-ft Canister:

(179.92-in [4569- )

179.82-in mm 162-in [4115-mm)

Impact Plate

- 18.aO-in

[457.2-mm)

(Nominal) 03-GA50387*15 Figure 4. DOE Spent Nuclear Fuel Standard Canister A sleeve of 16 in. (40.64 cm) outside diameter and 0.5 in. (1.27 cm) thickness is used as a spacer inside the SNF canister. The sleeve could be made of either stainless or carbon steel depending on desired structural and corrosion properties. The sleeve would utilize some sort of standoff structure that would center it in the SNF canister, though this detail is not modeled. The TMI canister is placed directly in the sleeve.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 19 of74 5.1.4 High-Level Waste Glass Pour Canister There is no long Savannah River Site high-level waste (HLW) glass canister. Therefore, the expected Hanford l5-foot HLW glass canister is used in the TMI canister waste package. Because the specific composition of the Hanford HLW glass has not yet been specified, it is assumed to be the same as the Savannah River Site glass composition. The Hanford l5-foot HLW glass canister is a 4,572-mm (180-in.)-10ng stainless steel Type 304L canister with an outer diameter of 610 mm (24 in.) (Taylor 1997). The wall thickness is 10.5 mm (0.4134 in.). These parameters are the same as the Savannah River Site canister, except that it is longer. The maximum loaded canister weight is 4,200 kg, and the fill volume is 87% (Taylor 1997).

5.1.5 Chemical Description of Borosilicate Glass The borosilicate glass intended for the Hanford canister has yet to be specified in terms of composition, physical properties, etc. However, given similar characteristics of the waste produced in the fuel dissolution and neutralization before tank farm storage, the resulting glass composition should be similar to that produced at Savannah River. Trace quantities of these materials provide a minimal impact on overall chemistry behavior and mobilization of fissile materials.

5.1.6 Waste Package Description The waste package contains five HLW glass pour canisters spaced radially around an 18 in. DOE SNF canister (CRWMS M&O 2000b, pp. 30-32 and Attachment V). The waste package description used to generate the calculated results in this report is based on recent design changes (unless noted otherwise), where the dimensions are slightly different than those given in this section. For example, there is now a 5 mm gap between the inner and outer shells resulting in a 10 mm increase in the outer shell diameter, and the details of the closure lids have also changed slightly. These dimensional changes have an insignificant effect on the calculated results based on a selected number of cases evaluated with the dimensions given in this section. The waste package barrier materials are typical of those used for commercial spent nuclear fuel waste containers. The waste package is designed to accommodate five HLW canisters surrounding a single SNF canister in the center position. The length of the waste package varies depending on whether it is to accommodate a 10 or l5-ft HLW/DOE SNF canister. Figure 5 depicts a cross-sectional view of the waste package and its internals.

The barrier materials of the waste package are typical of those used for commercial SNF waste packages. The inner barrier is composed of 50-mm (1.969-in.)-thick Type 316 NG stainless steel.

The outer barrier comprises 25 mm (0.984 in.) of high-nickel alloy (Alloy 22). The SNF canister support tube and basket plates are constructed out of carbon steel (ASTM A 516). A summary of pertinent dimensions and material specifications is provided in Table 2.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 20 of74 Table 2. Waste Package Dimensions a and Material Specifications.

Dimension Component Material Parameter mm)

Outer barrier shell S8-575 (Alloy 22) IThickness 25 Outer diameter 2030 Inner barrier shell SS 316 NG IThickness 50 Inner length 4618 Extended outer shell lid S8-575 (Alloy 22) IThickness 25 Outer shell flat closure lid p8-575 (Alloy 22) !Thickness 10 Inner shell lid pS 316 NG !Thickness 105 Closure lid to extended outer lid gap ~ir IThickness 30 Inner shell lid to closure lid gap ~ir IThickness 30 Support tube ~STM A 516 Grade 70 puter diameter 565 Inner diameter 501.5 Length 4607 Inner bracket ~STM A 516 Grade 70 IThickness 25.4 Length 4607 Outer bracket ~STM A 516 Grade 70 !Thickness 12.7 Length 4607 SOURCE: CRWMS M&O 2000b, pp. 30-32 and Attachment V NOTE: a More recent, proposed dimensions (used to generate the MCNP results) are: outer diameter of outer barrier shell is 2040 mm; inner length of inner barrier shell is 4617 mm; thickness of outer shell lid is 25.4 mm; thicknesses of inner shell lids are 50.8 mm (top) and 50 mm (bottom); thickness of gap between inner shell lid and closure lid is 47.23 mm; and thickness of gap between bottom inner and outer shell lids is 70 mm.

The outside diameter of the waste package is 2,030 mm (79.92 in.), and the inside cavity length is 4,618 mm (181.8 in.), which is designed to accommodate Hanford 15-ft HLW glass canisters. The lids of the inner barrier are 105 mm (4.134 in.) thick; those of the outer barrier are 25 mm (0.984 in.) thick. There is a 30-mm (1.181-in.) gap between the inner and outer barrier upper lids. Each end of the waste package has a 225-mm-Iong (8.858-in.) skirt. Note that some of the details concerning the lids are simplified, e.g., a I cm flat closure lid and 3 cm gap are neglected for the upper lid.

The DOE SNF canister is placed in a 31.75-mm (1.250-in.)-thick support tube with a nominal outer diameter of 565 mm (22.244 in.). The support tube is connected to the inside wall of the waste package by a web-like structure of basket plates to support five long HLW glass canisters. The support tube and the plates are 4,607 mm (181.378 in.) long.

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 21 of74 OUTER BRACKET

- - INNER BRACKET 01880mm INNER SHELL INNER DIAMETER 01980mm INNER SHELL OUTER DIAMETER 02030mm OUTER SHEll OUTER DIAMETER 0565mm SUPPORT TUBE OUTER DIAMETER 0501.5mm SUPPORT TUBE_

INNER DIAMETER

---613.5mm TYP-Figure 5. Cross-sectional View of Typical Waste Package.

5.2 MATERIALS DESCRIPTION Tables 3 through 11 give the composition of the materials used in this calculation. The number densities used in the inputs are calculated in Attachment I, spreadsheet "tmi_cals.xls."

Table 3. Composition and Density of Stainless Steel 304L.

Element Composition (wt %)a Value Used (wt %)

C 0.030 (max) 0.030 Mn 2.000 (max) 2.000 p 0.045 (max) 0.045 S 0.030 (max) 0.030 Si 0.750 (max) 0.750 Cr 18-20 19.000 Ni 8-12 10.000 N 0.100 (max) 0.100 Fe Balance 68.045 Densityb = 7.94 g/cm 3 NOTES: a ASTM A 240/A 240M-99b, p. 2.

bASTM G 1-90, Table X1.1.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 22 of 74 Table 4. Composition and Density of Savannah River Site High-Level Waste Glass Element I Isotope CompositionS (wt %) Element I Isotope CompositionS (wt %)

0 4.4770E+01 Ni 7.3490E-01 U-234 3.2794E-04 Pb 6.0961E-02 U-235 4.3514E-03 Si 2.1888E+01 U-236 1.0415E-03 Th 1.8559E-01 U-238 1.8666E+00 Ti 5.9676E-01 Pu-238 5.1819E-03 Zn d 6.4636E-02 Pu-239 1.2412E-02 8-10 5.9176E-01 Pu-240 2.2773E-03 8-11 2.6189E+00 Pu-241 9.6857E-04 Li-6 9.5955E-02 Pu-242 1.9168E-04 Li-7 1.3804E+OO Cs-133 4.0948E-02 F 3.1852E-02 Cs-135 5.1615E-03 Cu 1.5264E-01 8a-137 c 1.1267E-01 Fe 7.3907E+00 AI 2.3318E+00 K 2.9887E+00 S 1.2945E-01 Mg 8.2475E-01 Ca 6.6188E-01 Mn 1.5577E+00 P 1.4059E-02 Na 8.6284E+00 Cr 8.2567E-02 CI 1.1591E-01 Ag 5.0282E-02 - -

Density b at 25 *C = 2.85 g/cm 3 NOTES: a CRWMS 1999a, p. 7.

b Stout and Leider 1991, p. 2.2.1.1-4.

C See Assumption 3.2.

d. See Assumption 3.4

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 23 of74 Table 5. Composition and Density of Stainless Steel 316L Element Composition a (wt %) Value Used C 0.03 (max) 0.0300 N 0.10 (max) 0.1000 Si 1.00 (max) 1.0000 p 0.045 (max) 0.0450 S 0.03 (max) 0.0300 Cr 16-18 17.0000 Mn 2.00 (max) 2.0000 Ni 10-14 12.0000 Mo 2-3 2.5000 Fe Balance 65.2950 Densityb = 7.98 g/cm 3 NOTES: a ASTM A 276-91 a, p. 2.

bASTM G 1-90, Table X1.1.

Table 6. Composition and Density of Alloy 22 Element Composition (wt %) Value Used C 0.015 (max) 0.015 Mn 0.50 (max) 0.5 Si 0.08 (max) 0.08 Cr 20-22.5 21.25 Mo 12.5-14.5 13.5 Co 2.50 (max) 2.5 W 2.5-3.5 3.0 V 0.35 (max) 0.35 Fe 2.0-6.0 4.0 P 0.02 (max) 0.02 S 0.02 (max) 0.02 Ni Balance 54.765 Density = 8.69 g/cm 3 SOURCE: DTN: M00003RIB00071.000 Table 7. Composition and Density of Carbon Steel 516 Grade 70 Element Composition (wt %) Value Used C 0.28 0.30 Mn 0.85-1.20 1.025 P 0.035 (max) 0.035 S 0.035 (max) 0.035 Si 0.15-0.40 0.275 Fe Balance 98.33 Density = 7.85 g/cm 3 SOURCE: DTN: M00003RIB00072.000

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 24 of74 Table 8. Composition and Density of Dry Tuff Mineral Composition (wt %) Element Composition (wt %)

Si02 76.83 Si 0.359 AI203 12.74 AI 0.067 FeO 0.84 Fe 0.007 MgO 0.25 Mg 0.002 CaO 0.56 Ca 0.004 Na20 3.59 Na 0.027 K20 4.93 K 0.041 Ti02 0.1 Ti 0.001 P20S 0.02 P 0.0001 MnO 0.07 Mn 0.001

- - H 0.000

- - 0 0.492 3

Density=2.245 g/cm NOTE: CRWMS M&O (2001, Attachment II spreadsheet Tuff composltlon.xls")

Table 9. Pre-Breach Clay Compositions a,b Mass of Element after 59473 Mass of Element after 53241 Element Years of Emplacement C (kg) Years of Emplacement (kg) 0 1.55E+04 9.67E+03 H 8.07E+01 7.14E+01 Fe 1.98E+04 1.07E+04 AI 3.46E+02 3.36E+02 Ba 2.16E+01 u 2.15E+01 u Ca 1.69E+02 8.11E+01 F 2.60E+00 1.04E+00 P 1.27E+01 5.09E+00" K 1.62E+01 O.OOE+OO Mg 5.81E+01 9.05E+01 Mn 4.46E+02 1.67E+02 Na 1.39E+01 O.OOE+OO Ni 1.85E+03 3.87E+02" Si 4.47E+03 3.42E+03 Cr 8.14E+00 8.10E+00 U 2.69E+02 O.OOE+OO Total (kg) 4.31E+04 2.50E+04 3 8 Density (g/cm ) 4.23 3.88 NOTES: a Clay IS formed from DHLW glass degradation.

bBSC 2001a, p. 56 and CRWMS M&O 2000a, p. 41 C Composition same as used for Shippingport LWBR, (CRWMS M&O 2000a, p.

41).

d Values vary by one digit in least significant decimal place since masses are calculated from number of moles from reference using isotopic rather than elemental atomic masses.

e Value listed in reference is 4.235 which is insignificantly different from value used.

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 25 of74 Table 10. Post-Breach Clay Compositions Mass of Element Mass of Element after 72689 Years of after 378240 Years Element Emplacement (kg) of Emplacement (kg) 0 9.93E+03 1.24E+04 AI 3.35E+02 3.35E+02 Sa 2.14E+01 2.13E+01 Ca 6.58E+01 7.02E+01 Cr 8.09E+00 8.08E+OO F 5.51 E-01 O.OOE+OO Fe 1.13E+04 1.65E+04 H 6.94E+01 7.10E+01 C O.OOE+OO O.OOE+OO p 5.41E+00 8.80E+00 Mg 9.93E+01 9.99E+01 Mn 1.82E+02 3.43E+02 Mo 1.60E+01 2.51E+01 Ni 4.18E+02 4.10E+02 Si 3.43E+03 3.53E+03 Th 3.23E+00* 5.39E+01

• U-235 4.48E-01

• 7.25E+00*

Total (kg) 2.59E+04 3.39E+04 Density (g/cm 3 ) 3.98 b

4.16 b SOURCE: BSC (2001a, p. 59 and Attachment III, file fm2t1011.60)

NOTES:

• These thorium and U-235 masses are not used since they are for Fort Saint Vrain SNF; uranium masses from one TMI assembly and the plutonium from the HLW glass (see Table 4) are included in the post-breach composition used here, see Attachment I, spreadsheet "tmi calcs.xls," sheet "Materials."

3 b Densities of 4.05 and 4.21 g/cm are used for the clay compositions at 72689 and 378240 years, respectively.

Table 11. Alternative Post-Breach Clay Composition Mass of Element after 74818 Element Years of Emplacement (kg) 0 9.96E+03 AI 3.35E+02 B O.OOE+OO Ba 2.14E+01 Ca 6.58E+01 Cr 8.09E+00 F 4.80E-01 Fe 1.13E+04 H 6.94E+01 p 5.44E+OO Ma 9.93E+01 Mn 1.84E+02 Mo 1.77E+01 Ni 4.19E+02 Si 3.43E+03 Th 5.39E-01

• Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 26 of74 Table 11. Alternative Post-Breach Clay Composition (Continued)

Mass of Element after 74818 Element Years of Emplacement (kg) 8 U-235 7.18E-02 Total (kg) 2.59E+04 3 b Density (g/cm ) 3.92 SOURCE: SSC 2001a, p. 62 and file fm2i1021.60.

NOTES: 8 These thorium and U-235 masses are not used since they are for Fort Saint Vrain SNF; uranium masses from one TMI assembly and the plutonium from the HLW glass (see Table 4) are included in the post-breach composition used here, see Attachment I, spreadsheet "tmLcalcs.xls,"

sheet "Materials."

3 b A density of 3.99 g/cm is used.

5.3 FORMULAS The basic equation used to calculate the number density values for materials composed of one or more elements/isotopes is shown below. It is used in the spreadsheet included in Attachment I, and in the cases described throughout Section 5:

3 th where: N i is the number densi7 in atoms/cm of the i element/isotope, note that in Attachment I, N i is multiplied by 10-2 cm2/bam and is therefore in units of atoms/(bam*cm) mi is the mass in grams of the ith element/isotope in the material m is the mass in grams of the material; note that m = ~ mj Na is the Avogadro's number (6.022 E+23 atoms/mole, Parrington et ai. 1996, p. 59)

M is the atomic mass in g/mole of the ith element (Parrington et aI., 1996)/isotope (Audi and Wapstra, 1995)

M is the atomic mass in g/mole of the material Vi is the volume in cm3 of the ith element/isotope in the material V is the volume in cm3 of the material; note that V = ~Vi Pi is the density of the ith element/isotope pis the density of the material; note that p = ~pj*(ViN)

(aj)i is the atom fraction of the ith isotope of the element; note that M = ~(aj)i M Volumes of horizontal, cylinder segments (volume = area of circle segment x length of the cylinder) are also calculated throughout Attachment I. Given a specified volume of material in a cylinder, the following equation is solved iteratively for the material height (h) inside the package.

These calculations are based on the equation for the segment of a circle shown below (Beyer 1987,

p. 125):

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 27 of74 Area of a segment of a circle ~ ( R' cos-'( R ~ h) - (R - h),/2Rh - h' J where: R is the cylinder radius, and h is the height of the segment.

Alternatively, the height of the material in a degraded waste package can be calculated with the following parametric formula for the area of the segment and the angle e (Beyer 1987, p. 125),

which is defined such that h=R*[1-cos(t9)] and 0 ~ B~ 1t Area of a segment of a circle = R 2 ( B - ~ sin(2B)) .

The equation to calculate the fuel array void fraction, Vr, for each fuel array is shown below. For example, if the pellet array fills the entire cross-section of the canister then A cs is the canister cross-sectional area.

where: paxial is the axial pitch, N s is the number of pellet stacks, V p is the volume of the fuel pellet, and A cs is the effective cross-sectional area of the pellets (in the plane perpendicular to the axis of the cylinder).

5.4 INTACT MODE CRITICALITY CALCULATIONS In this section, the intact mode of the DOE SNF canister is analyzed. These configurations represent a waste package, which has been breached allowing inflow of water, but the internal components of the waste package are as-loaded, i.e., intact. Though, unless noted otherwise, unoccupied spaces in the SNF canister and waste package are modeled as void, but the TMI canister is water flooded. Modeling of the end structure of the DOE SNF canister treats both the impact plate and the dished head as a single piece that serves as an end reflector. The curved gap between the two pieces is conservatively modeled as filled with carbon steel. A sleeve is positioned between the SNF and TMI canisters and reduces the amount of "rattle-room" in the SNF canister. The waste package is reflected by tuff. Variations of the intact configurations are examined to identify the configuration that results in the highest calculated kerr value within the range of possible conditions.

5.4.1 Treatment of TMI Fuel and Canisters A single PWR fuel assembly is used as the basis for criticality scenarios associated with the TMI canisters. The enrichment of the fuel is assumed to be 3 wt% (Assumption 3.3), giving a U-235 mass of 13.906 kg per assembly for the new assembly total uranium mass given in Table 1. These values of enrichment and fissile mass are in excess of those values for the actual fuel in the canisters since they contain, in general, a blend of the three different enrichments given in Section

Engi nec[~.sYlil~2.::..S .t.P.:.:;ro:::.!*.::::cc~*l,--____ ea lculation Till ': Intact and Degraded Mode Critical.ity Calculations 'tor the Codisposal of TMI,-2 SperH uclcar Fuel in a Waste Package Document Identifier: AL*DSD-NU-000004 Rev OOA Pag' 28 of 74 5.1. t (all Ie s than 3 wtG;(). (See DOE 200 b {Table C-l] for a detailed accounting or uranium and plutonium content for ctlch TMT cani tcc.) The void frdction ,inside the fucl (7.5%. is assumed satur ted with waleI' (Assumption 3.5). The fuel is modeled as individual pellcts wiJh the dimension' given in ScHon 5.1.1 thougb spheriCi.u pellets with different diameters are al 0 inv *tigated. The individual fuel pellets (either cylindri al or spherical) remain axially aligned in what arc rdorred to here as "p~lIct stacks." Most results given below arc for cylindrical pcllet~.

In many ca Ct\ intact TMI canisters are modeled completely filled with pellets. Depending on the spa ing of the pellets tbi is typicall)' more th~m one as. embly's worth of fuel pellets or 67,600 pelle ,unle" 'tated otherwise. No zirconium is included in the canisters, but if present would act fl . a moderator displ.acer. Tb' RIC j~ replaced by water in tbe poi 'on tubes of the KO callister. and the annular up in the center tube cont.ains a mixture of fuel and waler. The canisters arc assumed water flooded (A' umption 3.5). While Born]'" was installed as an integral part of the box lincr for the fuel t pc canisters, no c.redit is taken tor it presence.

Only a portion aU of the KO canister internal' arc modeled. The fuel sits on a botton'! plate inside th cl.lllister, and this i always modeled. The B4 C in the poison tub is repine d with wafer; and the annulus bctW(;CD the central (larger) poi~ n tube and it liUpport. tube is as umed to contain hlel and wah,t. Since the poison tubes do not extend over the entire inne,f length of the canister, this upper pol1ioo of the Callister i open (o oth r uttemals are m deled. in tbis portion of the canister).

The p Uets in tll' pellet stacks tbat bapPCtl to panially intersect with the poison tubes arc modeled as partial pellets; the  ! P code allows a partial pellet to be modeled. In other words, the partial pellets consist of that portion of the intersecting pellets that are external to the poison lube. Because no internals arc modeled in the upper portion of the canister, the fuel pellets alway' completely till Ihis upper (open) ponion. unl . noted otherwi '0. These derails are shown in Figure 6 at different elevations of the canister. lo somccasc'. as identified in the tables, tbe poison lubes in the KO canister are neglected (thi is referred to as simply "without internal ,*.

A) The Upper Portioo of theKO B) Lower Portion oftheKO CanIster Above the POlson Tubes Canister Showing Poison TUbes

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 29 of74 Figure 6. Cross-sectional View of Knockout (KO) Canister at Different Elevations.

5.4.2 TMI Fuel Pellet and Other Modeling Details In this section, an exhaustive study of the spacing of cylindrical (same dimensions as intact TMI pellets) and spherical pellets in the KO canister is conducted to determine the most reactive KO canister configuration. As part of this study, the diameter of the spherical pellets is also varied.

The two most obvious choices for the spherical radius, R, are: (1) the diameter of the sphere equal to the smallest cylindrical pellet dimension, in this case the cylindrical diameter; and (2) the radius chosen such that the volume of the sphere and cylinder are equal. The former and latter choices give the smallest and largest radii, respectively, that are considered here. A total of four values of radius are investigated, these are 0.46990 cm, 0.52000 em, 0.54500 cm and 0.56772 cm, the two intermediate values are chosen to be about midway between the minimum and maximum radii, and then again about midway between this value and the maximum. For modeling convenience, the pellets (spherical and cylindrical) are aligned with the axial direction of the canister to form pellet stacks. The pellet stacks are arranged in a circular array in the canister. The array is referred to as being circular, because it is generated by positioning the centers of the pellet stacks on concentric circles (each a row) while maintaining as uniform a spacing (a pitch) as possible between all adjacent rods (not just those on the same row). This produces an irregular lattice that is neither completely triangular nor square. In practice, circular arrays have been shown to be as or more reactive than triangular arrays with the same pitch and number of rods (DOE 2002, p. 41). This spacing between stacks is referred to as a "radial" pitch, and the axial spacing is characterized by the gap between pellets (cylindrical pellets) or an axial pitch (spherical pellets). (Note that an axial pitch could also be used for the cylindrical pellets and is equal to the sum of the pellet length and gap.) In actuality, the pellets in the canisters are generally not axially aligned or uniformly spaced.

If the localized pellet orientation and separation are reasonably uniform throughout the fuel, then the actual pellet spacing may be better characterized by the array void fraction as calculated in Section 5.3.

Results for the spacing study of cylindrical pellets are given in Table 12. The radial and axial spacing is increased from the pellets touching until the most reactive configuration is found. In the first set of cases, the canister contains one assembly's worth of pellets, and the poison tubes are neglected. In the second set, the canister internals (poison tubes) are modeled, and the internal canister length is filled with pellets. In all cases, this is more pellets than are in an assembly (67,600 pellets).

Similar cases studying pellet spacing but for spherical pellets are studied in Tables 13 and 14 for the canister without and with internals, respectively. Each table is divided into four sets, one for each value of spherical radius. As for the cylindrical pellet cases, the canister contains either an assembly's worth or greater of pellets for the internals neglected or modeled, respectively. Cases with no water in the KO canister are investigated in Table 15. For these cases the canister contains an assembly's worth of spherical fuel pellets with the maximum radius, and the canister internals are neglected. The pellets are touching in the radial and axial directions since this configuration is anticipated to be most reactive for a dry canister. To confirm this, a limited number of cases with increased radial and axial separation are also considered.

En .iQ\;lcrcd, ~tem~s..:.P...:.r.::.olc:*c~c",-t~~ _ Calculalion

Title:

lnwet and Degraded Mode rhicalily Calculations for the Codisposal of TMI-2 SpeTlt Nuckof Fuel in a Waste Pacbgc Document Identifkr: CAL-DSD-NU-OOOOO4 Rev OOA Page 30 of74 In the next table. able 16 tb fuel loading in the canisters (with internals) and other details arc investigated. These cases arc variations of cases in Tables 13 and 14. In the first l()Ur sets of results

( ne tor each valu~ or spherical radius), the cani ICf contains an assembly' worth of fuel. Since Ihis only partially fill the canister length, the pellcts are modeled in the upper part of the c3nititcr since this end i open above the poison tubes. Thi is done t) maximize the pellct array cro '.

,celional area even though this open volume is not sufticient to contain all tbe fuel. [I) the tirst set of the tnblc, tbe effect of redu 'lng the cross-sectional area of the pellet array. i.., redu'ing the number of pellet tacks, i investig~ltcd For cases with reduced em s- cel;on, the canister is rolaled 4 in order to change the relative positioning of the poi 'on tubes and pellet array as shown in I,)

l";"igurc 7. Simihlr cases in the fourth and sixth cts investigate different fuel loadings (both sets) and n::dlclced .may eros - ctional area (sixth ct for a cani 'ter with and without internals, rc pcctively.

Other detail invest; ,. ted. in this table are: water replaces the void lYdweCJl the SNF and TMI canister; location of these canisters in the waste packa 'c' t'ilc1locatcd. in the bottom rather than top pcrtion of K canister; variations in the thickn s . nd compo, ition of the sleeve surrounding the TMI clltlister; no waler saturation of voids in the. fuel pellets; aU c~njsters (including HLW canister) in the waste package have hiflcd downward due to the effect of gravity, see Figure 8; the sleeve between th' SNF and TMI canisters is no longer centered ill the SNF canister, see Figure 9; the earlil.:r design dimension' of the waste package are used, sec Sc~tion 5.1.6; and the enrichment and plutonium contel of th fuel are varied, The enrichment and fuel m 'used here must be shown to bound dIe ffcctJvc enrichment of the most highly loaded KG canister givcn in Table I (2. 13 WI%

enrichment) and its plutonium content which is less than one kg of lotal plutonium.

A) Canister Contains One Assembly'S B) Canister Containing One Assembly's WOrl.h of Pellets Worth at Pellets s Rotated A5" Figure '7. PeUet Array Partially Fills KO Canister's Cross.-seclion.

En rinccn::d Systems Project Calculation

Title:

lutar.:t and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package

[)ocumcntldcntificr: CAL-DSD*NU-000004 Rev OOA Pu 'C 31 of74 Figure 8. Canisters Are in a Gravity Position in the waste package.

SNF Canister COllapsed SI eve TMI Canisler Figure 9. Sleeve Is Collapsed in the SNF Canister.

A spacin SlU(Jy of cylindrical pellcts in a TMr fuel canister (D.*type) L given in Table 17. This is donI,; 0 determine which canister is the 010 t reactive. In the tirs'! set of the table, the canister contain an assembly. worth of fuel pellct that are in a hexagonal array. In these casco only whole pdlcts arc modeled in the canister. In the ond set of cast.'S if there i sufficient space around the insid> perimeter of th > canister) then partial pellets arc positioned in the a.rrn.y positions as shown in i ure 10. incc these c<\ cs contain the same number of wholcpellclS, the nllmhcr of panial

~b!!.~~~~~:.!...!..~~~,--- ..Gf.llcW.[!i()n

[or th Codisposal of TMI-2 Spent Page 32 of74 pellets is the amount of tllcl ill excess of one assembly's worth. For all cases, the Boral lM layer is replaced by wnlcr, and th ~ inner square box structure remains intact.

Agul'e 10. Cross-sectional View of the TMI Fuel Canister.

III tb fi.nal t ble of resufl of this section, cases arc considered where the KO c.misler contains one assembly's worth of tuel homogenized with varying amounts of water. This is done to demon Ifate that.l1 adding hctcrogcncou fuel pellets is more rcactivethao a homogeneous mixture for TMI fuel (i.e., low-enriched fuel. For simplicity the canister internals arc neglected. and the water volume fra~tiofl (wvf) in the fuel i* incrca. 'cd from completely dry until the most fc.active fuel mixturc is fbund. The remainder of the canister is assumed (Assumption 3.5) to cont.ain water (Assumption 3.5) except for a cQuple orca cwhcre it is dry.. These rsults are giv n in Table 18, and the most fC tive cas can be compared to the equivalent heterogeneous caiiCS givcn in Tables 12 and 13.

5.5 DEGRADED MODE The criticality caJculatiotls .Qnducted for the degruded cases are discu5~ed in the following sections.

, 1,;vcral C nfigUntlions are considered. Ddnilcd descriptions of the C configurations arc given on pllgcS 27 through 37 of CRWMS M&O (l999b. In Scctic>o 5.5.1, configurations arc analyzed re'tJlting front the. degradation scenarios in which the components illside the SNF canister (sleeve and TMI-2 canister) degrade (CRWMS M&O 1999b, pp. 27-29). III Section 5.5.2, configurations rc ulting [r rn the degrnd,Hion of the high-level waste lass arc 'investigated (sec AWlchmcnt 1,

~prcad 'heet "tmi_cals.xls," sheet "WP" for detailed de criplion of ease) tor sensitivity to changll1g wast" pa 'kagc pammelcrs. Configuration where all the internal components of the waste p. ckagc hit e dC 1 radcd nrc discus cd in S ction 5.. (C0 Attachment J, preadshcct "tmi_cals.xls" sheet "WP" ftlr detailed description of cas.e. , and are also investigated tor dirfercnt waste package conditi r*. 111 configurations rculling from water flowing through the waste package, it has been shown (If C 2001a, p. 64) Ihat the fissile material will likely be nushed out of the waste p<lckagc.

As ttl* nmount of tissile material decrease, the risk of intemal criticality i diminished.

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 33 of 74 The degraded configurations are analyzed as a sequence of progressing degradation rather than an immediate transition from intact to completely degraded. These degraded configurations are expected to be more reactive than intact configurations. Water is always assumed present and is responsible for the degradation considered here. The details of this process depend on degradation rates and how the water enters the waste package and interacts with its contents. Except for the scenario where everything is degraded in the waste package, no degradation of the fuel pellets is considered since low-enriched (less than approximately 5 wt%) uranium fuel is more reactive when in a heterogeneous configuration (DOE 2003b, p. 50). Typically the first components to degrade would be the HLW canisters, the waste package basket and the inner lining of the waste package.

The degraded components form pre-breach clay which surrounds the intact SNF canister. The sleeve inside the SNF canister could also be degraded at this time, or might degrade prior to the formation of the pre-breach clay. Details of the degradation process would determine whether the SNF or TMI canister degrade first leaving fuel pellets in the surviving canister surrounded by pre-breach clay and the other degradation products. A near final configuration is all components degraded and the fuel pellets surrounded by layers of different materials with little mixing between layers. The presence of water in the layers is an added complication that must be considered. A possible final configuration, though uninteresting from a criticality point of view (as shown below in Table 26), would be for the entire contents of the waste package to be completely degraded and mixed together. There may be no identified mechanism for such complete homogenization of the degraded waste package contents.

For degraded mode calculations, the iron content of stainless steel degrades to goethite (FeOOH)

(Assumption 3.1), and the degradation products from the other steel constituents are neglected. In determining the amount of goethite formed, only the canister walls are considered, and the internal canister components are neglected. The void fraction in the degradation products is expected to be around 40% though in some cases values of 60% and larger are used. In an aqueous environment, water can saturate the material filling these voids.

The TMI pellets remain intact except for the scenario of complete internal degradation. At least one assembly's worth of pellets is used in the analysis. The loose pellet arrays are modeled as stacks of axially aligned pellets with a uniform separation between stacks and pellets, i.e., characterized by radial and axial pitches (two parameters), respectively. In actuality, the loose, randomly positioned pellets are not, in general, axially aligned or uniformly spaced. If the pellet separation is reasonably uniform throughout the array, then the pellet spacing can be characterized by (one parameter) an array void fraction (see definition in Section 5.4.2) which is also listed in the results. Degradation products and/or water can fill these voids between fuel pellets.

While this addresses the spacing of the pellets, the axial loading of the pellets must also be addressed. For an un-breached TMI canister, the maximum loading is limited by the maximum number of pellets that would fit in the canister and is determined here for whichever canister type has the largest cross-sectional area, i.e., the KO canister. Once the TMI canister is breached and if there is no axial redistribution of fuel, the axial fuel loading is also limited by the canister's maximum loading. Another consideration is that there is no identifiable mechanism to elevate any pellet against gravity. This means that each pellet should be located somewhere between its intact position and the bottom of the medium containing the entire array. For a realistic array, this imposes an upper limit on the radial pitch for any given axial fuel loading.

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of 1MI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 34 of74 In these cases, the terms "fraction of water" or "percent of water" refer to a volume fraction (vf) or to a percentage of volume, respectively. The percentages listed for the other components of the degradation products are volume fractions. As was done for the intact cases, the waste package is reflected by tuff.

5.5.1 Components Degrades Within the Intact SNF Canister In this section, cases are investigated where components inside the SNF canister degrade while it remains intact. This most closely corresponds to scenario IP-l A from YMP (2000, pp.3-13 and 3-

14) and from CRWMS M&O 1999b, p. 27). Results for this section are given in Section 6.2.1.

For this scenario, the sleeve and 1MI canister have degraded leaving the fuel pellets inside the intact SNF canister surrounded by goethite and/or water. Unless noted otherwise, the SNF canister contained a stainless steel sleeve. Components external to the SNF canister are also intact.

Cylindrical pellets are considered and the results examine a range of pellet pitches and axial fuel loadings. This includes cases where the entire cross-section of the SNF canister is filled as shown in Figure 11. This case is unrealistic because some of the pellets at the top of the array have been moved against gravity (gravity is downward in the plane of the figure) from their intact positions that can be seen in Figure 8. Results are shown in Table 19. In the first set of the table, goethite is neglected and the pellets are surrounded by water. In the next two sets, water is above dry goethite that is at the bottom of the canister, see Figure 11. The level of goethite is determined by the amount produced and by the number of pellet stacks that are displaced. Cases in these sets examine the effects of increased goethite from a carbon steel sleeve and of completely filling the canister with dry goethite. This latter case is done solely as a comparison. In the third set, the number of pellet stacks is decreased giving a reduced axial fuel loading for some of the more reactive cases of the second set. This is accomplished by maintaining the radial pitch and axial spacing between pellets while decreasing the number of pellets stacks and appropriately increasing the number of pellets per stack so as to maintain the same total amount of fuel. In the last two sets of the table, goethite is mixed with sufficient water to completely fill the SNF canister except for the last case of the fourth set which has a smaller wvf. This case illustrates the effect of decreasing the amount of water in the surrounding goethite mixture. Other cases in this set show the effect of changing pitch in the array. Cases in the last set show the effect of decreasing the number of pellet stacks for some of the more reactive cases of the previous set.

EfH~inccrL"d Sy~tcms Projec!::.!'l:--_______________ , _ _C~ICl!l.ation

Title:

Intacl and Degraded Mode Criticality alculations fol' the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document '1d<Dtificr:AL-DSD-NLJ-000004 Rev OOA Page 35 of74 SNF Water Above C"lIlisler Goethite Dry Goethite from Degraded Sleeve and TMI Canisler Figure 11, Cross-sectional View of the SNF Canister.

5.5.2 Internal Components of the Waste Package Outside SNF Canister Oegrade This 'celion de cribcs configurations resulting froD) the scenario lP-3 (YMP 2000, p. 3-13). Thc internal components of the waste package outside the SNF canister arc COluplctcly degraded.

including tbe iImer barrier shell of the waste package. The compositions of th slurry resulting from this de adation arc given in CRWMS M&O (2000a, p. 41) and BSC (200Ia, p. 56) and arc r Ii rred to as the pre-breach clay (Table 9). The amount of water mixed \vith this day is varied.

There i ' U-238 pre 'Clft in the 'lurry from the degraded glass, but iris lconscrvativcly) neglected in these calculations since it is a neutron absorber. The ca cs in this section (:3n be further divided inlo 2 catcgorie ' depending upon whether the SNF canister i~ trealed as bCiJlg intact or degraded. In the firs cat gory, de' Tibcd i.n 'e tioo 5.5.2.1, the SNF canister is intact and its corr.ents can either be intact, p,trtially r completely degraded. In the second category, described in Section 5.5.2.2, the SNF carli t r has dcrnded, but the degradalion products from the canister and its cont~nts remain s parat* and ha e not yet chemically reacted with the pre-breach clay.

5.5.2..1 I.ntact SNF Canister with Contents Kither Intact or l>eg:raded Tbe ~ NF eMi '1er CQl'ltlgul1ltions studied include intact and degraded cases and arc derived from the mosl [cacti cases identified in the previous sections (Sections 5.4 and 5.5.1). For the cases given in Table 20, the intact SNF canister containing intact components is surrounded by pre-bi' ach clay.

Tbe KO canisteT and leeve arc intact, exc.ept a not.ed below. and the ennis!cr (with internals) is filled with spheri a1 pellets. This configuration is nol tbe most reactive case of Section 5.4, but it is typt al. of the mo. t reactive ase of intere 1. in the lim! set of this ttlblc, the vertical height of the SNF cani. ler in dry clay is varied from just under the clay layer to r sting on the bottom of the

Engineered System - Project CaJculaiinn

Title:

Jnt,lct and Degraded Mode Criticality Calculation odisposal of TMJ-2 Spcnl Nuclear Fuel in a Wn t Package D 'tunent Identifier: AL-DSD-NU-OOOO04 Rev OOA Page 36 ofi4 astc pack'lgc. variation of the case with the NF canister restiJlg on the \V,lsle package bottom bUI with carlier waste package design dimensions is investigated. Otber ea-cs in this set examine the effect of filling the void between the sleeve and the canisters with clay mixed with varying amounts of water and of a carbon steel sleeve degraded to goethite. Figure 12 hows this configuration where the slccv has degraded to goethite, and the SNF cani 'tel' is cenlered in the pre-breach clay. In UIC second ct. the water content of the clay is incrca ed llotilthe waste packa,c is completely filled. Th' composition of the clay with the various volume fr'lctions of wateT is dctcnn.incd in Attachment 1. spreadsheet "lmi_calcs..xls," sheet "WP".

..01.-_-- Tuff Surrounding WP WPOuler Barrier Water lntactSNF Canister Con ered in Dry Clay CI y from the Degr dad C<ln ents of the WP Goethite from

'Degrade Sleeve _-..,;_...:

Betwoen canisters Figore 12. Cross-sectional View of an Intact SNF Canister Centered in Clay Formed from the 0 gradation of Components External to the Canister The next table of r 'ults, Table 21, considers cases wh rc the TMI canister and sl eve have de rradcdlcaving fuel p 'llet* and goethite in tbe SNf canister. The description of these en es very clo ely p mile)' tho c in Section 5.5.1. ummarizing, variations in pellet ::.pacing and rna rial compo ili n in tbe cani ter are invc:'tigatcd. Jo the first et. the goethite is nc,1 ctcd leaving only wat r in the C !lister. ea e' with fuel pellets urroulldcd by dry goethite covered with water aTC gi ' 11 in tb cond and third se '. A case in the ccond SCI evaluates the effect of using the carlier design ft te packag . dimen ions. In the third set of the table the number of p Het stocks is rcdu d tor me of the m r reactive case of the econd set. The remaining rc. ults for thi section arc given in Tab} 22. Here goethite mixed with sufficient water to completely till the F canJ tcr urrounds the fu I pellets. The'c ca es inv tigate tbe e1 eelS of pellet spacin . redu in the wvf in lhl:: thite and r ducing the number of pellet stack while incrca ing the number f pelle p r tack, and maintaining the arne radial pitch and ial spacing between pellet.

Engineered Systems Proiect Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-OOOOO4 Rev OOA Page 37 of74 5.5.2.2 Degraded SNF Canister with Non-reacted Pre-breach Clay For these configurations the SNF canister and sleeve have degraded to goethite, but the TMI canister is intact (Section 5.5.2.2.1) or fully degraded (Section 5.5.2.2.2). The goethite formed from the sleeve and canister walls has not yet chemically reacted with the pre-breach clay. The configurations in this section are similar to those in Sections 5.4 and 5.5.1, but now pre-breach clay and goethite, in separate layers, can surround the fuel pellets. The positioning of these configurations in the waste package is varied, and any unoccupied space in the waste package is filled with water. The results of these cases can be found in Section 6.2.2.2.

5.5.2.2.1 Intact TMI Canister with Degraded SNF Canister and Pre-breach Clay in the Waste Package In these configurations, the TMI canister is intact and surrounded by unmixed layers of pre-breach clay and goethite. The volume fraction of water in the materials is varied. Results for the cases described here are presented in Table 23. For these cases the KO canister contains spherical pellets with identical canister details as that described in Section 5.4. In the first set of the table, the goethite is neglected leaving the canister surrounded by dry pre-breach clay. Cases in this set investigate the effect of reducing the number of pellet stacks in the canister. The first few cases of the second set investigate the effect of changing the position of the canister in the dry clay, and of changing the wvf in the clay. In the next few cases the goethite forms an annulus surrounding the canister which in tum is surrounded by clay. This type of configuration could occur if the SNF canister and sleeve degrade to goethite after being trapped in the pre-breach clay and is shown in Figure 13. Several variations of this case are investigated by changing the volume fraction of water in the goethite surrounding the canister. In the last two cases of the table, the effect of homogeneously mixing pre-breach clay with the water inside of the canister is determined.

t En 'in~~!cd S *:m-:m':',...!.P..!..f(~)ju::c~c~t (alculalioll Tille: Intact and Degraded Mod Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD*NU-000004 Rev OOA Page 3R of 74

___ DryClay Intact KO Canister ~

Goethite and Waler Mixture

......-I-~~~ Surrounds the KO Canister Figure 1'3 Intact TMI Canister Surrounded by Goethite Trapped in Pre-breach Clay in the Waste Package 5.5.2.2.2 Degraded T 11 Canister, SIf'f"ve and SN'F Canister with Pre-breach Ctayin the Waste Package The, conligurotions are <I continuation of tho c in Section 5.5.2. I where the sleeve and T~lI cani ter bav' degraded. but now the S T canister is also degraded. These matcriab ellll IIJlm separate Qr mixed layers covering the fuel pellets at the boHorn of the waste package. Since water is availabl in the. waste packa 'c. different wvf in the layers must ulso be iovcsiigaled. Figure 14 sno\vs an c ample of these materials tonniug diiTerent layers in the waste packag. Fuel pellets or heap I at the bottom of the waste package and arc mostly surrounded by (l layer goetbite <lnd aler. Thc next I yer composed of pre-breach clay and watcr covers Ihc uppermost pellets. The TO t of the w:lsle package is filled with water. Configurations with these materials at lcast partially mi. d together would be more rea[isfic, tbough due to the high iron content of the pre.breach clay (s* c Table 9) even c mplclc mixing of the e matcrials would not significnntly alter the composition oftht day. The ro ults fOf these Crise:;. arc sbown in Table 24.

, tem' Pro'eel Calq~latio.!!

Title:

Inlacl and Degraded Mode Criticalityalculations for the COOt ipOsal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 39 of74 1:L-_- Tuff WPOuter Barrier Water Pre-breach

._-~*--'--- Clay Mixed wfth Water Goethite Mixed Wl h Water TMI Pellet Slaw Figure 14. Fuel Pellets Surrounded by Goethite and Pre-breach Clay from Completely Degraded Components in the Waste Package Variations of lh' waler volume fraction in the c materials goethite and pre-breach clay) and the spacing (pitch) of the pellet stacks arc inve ligated. Different pellet configumtions with the ,Ime pitch and covered with the same or similar material layers arc also considered. For exampl . the C nf1guration shown in Figure 15 has the same pitch and goethite/water composition liJr tbe boltom layer as that shown in Figllre 14, but since there arc fewer J'lCllet stacks (the n.umber of pelle . per Sl4.lck i' appropriately incrca' d) the pellets arc almost completely covere<). by the goethite layer, A ca. also compar the effect of modeling the earlier d ign wa -te package dimensions. Cases with a reduced number of pellct Slacks arc considered in the ccond sct or the table.

Calt:ulation alculalions for the Codi 'po al of TMI-2 Spent Page 40 of74 Tuff WPOuter Barrier

-..-_ _ Water Dry Pre-breach Clay Goofhile Mixed with Water TMI Pellet Slacks Figure 15. Similar Configuration as Shown in Figure 14 but Fewer Pellet Slacks in Fuel Array 5.5.3 All Components Have Degraded Th e configurations represent the final stage of degradation that would occur aftcr the scenario d' -(;ribcd in Section 5.5.2.2.2. Tbe composition or the clay resulting from the degradation or all components iHside the wa te package is taken from Tables 10 and 11, which corne- from B C (200 Ia, pp. 59 and 62). This clay is referred to as post-breach clay and is Ihe clay from the Fort aint Vrain HTGR S F analysis. it is u cd here after n 'glecting the thorium and utaniumcontcnt but includes the plutonillm from the HLW gla. S, S e Table 9, Compositions lor the clay, fucl UlldJor wat~r mix.ed in various proportions arc determil1cdin Attachment I, spreadsheet "tmi cals.xls" sheet "WP." R'sults for the cases dc.'cribed here Me presented in Tuble 25 where the fuel pellcts remain intact and arc surrounded by post-breach clay. The pdlct configurations are similar to those de cribcd in SccJlon 5.5.2.2.2, b\Jl only post-breach clay and waler arc in tbe wa te packa. aseinvc. Ii atc varying pitdll~S, water vol.ume fraction in the clay and pellet c nfi urations. A cas~ al 0 compares th~ effect of us.iog the earlier dcs.ign wa lC po 'kagc dimensions, For the next cases, rcsults giv 1l in Table 26, one assembly's worth of fucl pellets are as umcd to be completely degraded and the fuel is homogenized with the po I-breach clay. Tbree ca e arc

,onsidered. one for each of the composition' given in Tables 10 and II, so as to determine the mo t reactive composition. The clay is homogenized with varying amounts of water, and any remaining space in the a t package is filled with water. The percentage of water in the clay is increa cd until the ("'1lrire volume avajlablc in the waste package i' filled with the clay mixture. The cffe t of n\: Iccting the U-238 contcnt oflhc fuel is also examined.

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 41 of74

6. RESULTS This report documents the various calculations for intact and degraded mode configurations of the TMI-2 canister waste package. Sections 6.1 and 6.2 present the kerr for the intact and the degraded configurations, respectively. The kerr results represent the average collision, absorption, and track length estimator from the MCNP calculations. The standard deviation (0") represents the standard deviation of kerr about the average combined collision, absorption, and track length estimate due to Monte Carlo calculation statistics. The average energy of a neutron causing fission (AENCF) is the energy per source particle lost to fission divided by the weight per source neutron lost to fission from the "problem summary section" of the MCNP output. The MCNP input and output files developed for this calculation are included in ASCII format in Attachment I. (The output file name is derived by appending ".0" to the input file name.) The H/X ratio is the ratio of moles of hydrogen to moles of fissile materials (U-235) and is determined on the basis of a fuel pellet unit cell. Since the fuel pellets are typically in an irregular array, the ratio is the simple average of the values of HlX for square and hexagonal unit cells. The array void fraction is also listed for all cases that involve fuel pellets and is determined from the expression for Vr given in Section 5.4.2 using the effective cross-sectional area of the array, unless noted otherwise.

For convenience in referring to cases in a given table, groups of cases may be referred to as a "set,"

where the set of cases examine variations to a single or limited number of parameters. A preceding header (in bold type) within each table denotes the set. In all discussions of kerr in the text, it is implied that kerr is kerr + 20". For example, if it states that "kerr for the system is greater than 0.95," it is to be interpreted as "kerr + 20" for the system is greater than 0.95." Values of kerr in the tables are as labeled in the column headings.

A single PWR fuel assembly is used as the basis for criticality scenarios associated with the TMI canisters. The enrichment of the fuel is assumed to be 3 wt% (Assumption 3.3), giving a U-235 mass of 13.906 kg per assembly. The void fraction inside the fuel is assumed saturated with water (Assumption 3.5). The fuel is modeled as individual pellets with the dimensions given in Section 5.1.1 though spherical pellets with different diameters are investigated. The individual fuel pellet remain axially aligned in what are referred to as "pellet stacks." This applies for either cylindrical or spherical pellets. The results given below are for cylindrical pellets unless otherwise stated.

M While Boraf was installed as an integral portion of the box liner for the fuel assembly, no credit is taken for its presence.

For many cases intact TMI canisters are modeled completely filled with pellets. This is typically more than one assembly's worth of intact fuel pellets or 67,600 pellets. No zirconium is included in the canisters. Zirconium would act as a moderator displacer. The B4 C is replaced by water in the poison tubes of the KO canister, and the annular gap in the center tube contains a mixture of fuel and water. The canisters are assumed water flooded (Assumption 3.5).

For degraded mode calculations, the iron content of stainless steel degrades to goethite (FeOOH),

and the other steel constituents are neglected. For this process, only the canister walls are considered, and the internal canister components are neglected. The TMI pellets remain intact even

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 42 of74 for the completely degraded scenarios. For degraded TMI canisters, one assembly's worth of pellets is used in the analysis.

The interim critical limit (ICL) is assumed to be 0.97 (Assumption 3.6).

6.1 INTACT MODE This section gives the results of the calculations described in Section 5.4. For these cases intact means that the canisters (both TMI and SNF) and waste package internals (external to the SNF canister) are intact, but the TMI fuel is modeled as being most reactive. For low enrichment, this means the fuel is heterogeneous, i.e., pellets, but not necessarily with the same dimensions as the intact assembly fuel pellets. The following results are for TMI fuel in the KG canister inside an SNF canister in the center basket position of the waste package surrounded by five HLW canisters in the outer basket positions. The KG canister is water flooded, and the empty space in the SNF canister (between canisters and sleeve) is void. The internals of the waste package are intact, empty spaces in the waste package are modeled as void, and the waste package is reflected by dry tuff.

6.1.1 Loose TMI Fuel Pellets in TMI Fuel Canisters and Other Modeling Details Before evaluating details for TMI fuel, it is necessary to develop a conservative yet realistic model for the fuel in the TMI canister. This is also useful because the characteristics important to the criticality of the fuel can be learned. The first step is to find the most reactive fuel pellet by investigating fuel pellet shape, size and spacing. Here only spherical and cylindrical fuel pellets are considered though it is generally expected that spherical pellets would be as or more reactive than any other shape.

The results shown in Table 12 are for cylindrical pellets in a KG canister. The dimensions of the pellets are the same as those of an intact fuel assembly, i.e., 0.9398 cm diameter and 1.1049 cm length. The spacing between pellet stacks is characterized by a (radial) pitch, and the axial separation between pellets in the stack is characterized by a gap or axial pitch which is simply equal to the length of the pellet plus gap.

Table 12. Cylindrical TMI Fuel Pellets in the Knockout (KG) Canister Axial Void Radial Pellets Fraction, keff + AENCF H/X Pitch, cm Gap, cm V, k eff +/- 0 20 , keV Ratio File Name Cylindrical Pellets (R = 0.4699 cm) in Canister" without Internals 0.94° O. 0.188 0.6790 +/- 0.0009 0.6808 607.5 25.6 cv-I .94.0 1.1 O. 0.400 0.8915 +/- 0.0011 0.8936 328.6 68.3 cy-I 1.1.0 1.2 O. 0.494 0.9749 +/- 0.0011 0.9771 253.1 98.4 cy-I 1.2.0 1.3 O. 0.570 1.0256 +/- 0.0010 1.0276 205.0 131.1 cy-I 1.3.0 1.4 O. 0.623 1.0536 +/- 0.0009 1.0555 175.2 166.4 cy-I 1.4.0 1.5 O. 0.669 1.0703 +/- 0.0010 1.0722 151.7 204.3 cv-I 1.5.0 1.6 O. 0.710 1.0724 +/- 0.0009 1.0742 134.7 244.9 cy-I 1.6.0

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 43 of74 Table 12. Cylindrical TMI Fuel Pellets in the Knockout (KO) Canister (Continued)

Axial Void Radial Pellets Fraction, kelt + AENCF H/X Pitch, cm Gap, cm V, kelt+/- <1 2<1 , keV Ratio File Name 1.7 O. 0.739 1.0683 +/- 0.0010 1.0704 123.6 288.0 cy-I 1.7.0 1.8 O. 0.767 1.0582 +/- 0.0011 1.0603 113.2 333.8 cy-I 1.8.0 1.9 O. 0.789 1.0462 +/- 0.0010 1.0482 105.4 382.2 cy-I 1.9.0 2.0 O. 0.809 1.0250 +/- 0.0010 1.0269 98.4 433.2 cy-I 2.0 0.94° 0.5 0.441 0.9303 +/- 0.0011 0.9324 297.4 77.9 cy-I .94 .5.0 0.94° 1.0 0.574 1.0238 +/- 0.0010 1.0258 209.3 130.2 cy-I .94 1.0 b cy-0.94 1.25 0.619 1.0415+/-0.0010 1.0436 186.6 156.4 I .94 1.25.0 0.94° 1.5 0.656 1.0505 +/- 0.0010 1.0525 170.1 182.5 cy-I .94 1.5.0 b cy-0.94 1.75 0.686 1.0485 +/- 0.0009 1.0503 157.0 208.7 I .94 1.75.0 0.94° 2.0 0.711 1.0398 +/- 0.0010 1.0417 148.2 234.8 cy-I .94 2.0 0.94° 2.5 0.751 1.0167 +/- 0.0009 1.0186 138.0 287.1 cy-I .94 2.5.0 1.2 0.2 0.571 1.0187 +/- 0.0009 1.0205 203.3 132.5 cy-I 1.2 .2.0 1.2 0.4 0.628 1.0492 +/- 0.0010 1.0512 172.8 166.6 cy-I 1.2 .4.0 1.2 0.6 0.672 1.0636 +/- 0.0010 1.0655 151.3 200.7 cy-I 1.2 .6.0 1.2 0.8 0.706 1.0695 +/- 0.0009 1.0713 136.7 234.8 cy-I 1.2 .8.0 1.2 1.0 0.734 1.0673 +/- 0.0009 1.0691 126.1 268.9 cy-I 1.2 1.0 1.2 1.2 0.757 1.0617 +/- 0.0009 1.0635 116.3 303.0 cy-I 1.2 1.2.0 1.3 0.2 0.636 1.0527 +/- 0.0011 1.0548 167.5 171.1 cy-I 1.3 .2.0 1.3 0.4 0.684 1.0680 +/- 0.0010 1.0700 143.5 211.1 cy-I 1.3 .4.0 1.3 0.6 0.721 1.0726 +/- 0.0010 1.0747 129.2 251.1 cy-I 1.3 .6.0 1.3 0.8 0.750 1.0682 +/- 0.0009 1.0701 118.3 291.1 cy-I 1.3 .8.0 1.4 0.2 0.681 1.0672 +/- 0.0010 1.0692 145.5 212.8 cy-I 1.4 .2.0 1.4 0.4 0.723 1.0718 +/- 0.0010 1.0738 128.2 259.2 cy-I 1.4 .4.0 1.4 0.6 0.756 1.0691 +/- 0.0010 1.0711 115.0 305.6 cy-I 1.4 .6.0 1.5 0.1 0.697 1.0712 +/- 0.0010 1.0731 138.9 231.0 cy-I 1.5 .1.0 1.5 0.2 0.720 1.0732 +/- 0.0010 1.0751 128.5 257.6 cy-I 1.5 .2.0 1.5 0.3 0.740 1.0699 +/- 0.0010 1.0718 120.3 284.2 cy-I 1.5 .3.0 1.5 0.5 0.772 1.0633 +/- 0.0010 1.0653 107.9 337.5 cy-I 1.5 .5.0 1.6 0.1 0.734 1.0705 +/- 0.0009 1.0724 123.4 275.2 cy-I 1.6 .1.0 1.6 0.2 0.754 1.0680 +/- 0.0010 1.0699 114.9 305.5 cy-I 1.6 .2.0 1.6 0.3 0.772 1.0625 +/- 0.0010 1.0644 107.4 335.8 cy-I 1.6 .3.0 Cylindrical Pellets =

(R 0.4699 cm) in Canister C with Internals 1.3 O. 0.570 0.9489 +/- 0.0010 0.9509 203.3 131.1 cy 1.3.0 1.4 O. 0.623 0.9545 +/- 0.0009 0.9563 175.4 166.4 cy 1.4.0 1.5 O. 0.669 0.9605 +/- 0.0010 0.9625 156.1 204.3 cy 1.5.0 1.6 O. 0.710 0.9586 +/- 0.0010 0.9606 139.3 244.9 cy_ 1.6.0 1.7 O. 0.739 0.9429 +/- 0.0009 0.9447 127.9 288.0 cy_ 1.7.0 1.3 0.2 0.636 0.9606 +/- 0.0010 0.9625 170.1 171.1 cy 1.3 .2.0 1.3 0.3 0.661 0.9646 +/- 0.0009 0.9665 158.3 191.1 cy 1.3 .3.0 1.3 0.4 0.684 0.9649 +/- 0.0009 0.9667 147.5 211.1 cy 1.3 .4.0 1.3 0.5 0.704 0.9615 +/- 0.0011 0.9637 139.5 231.1 cy_ 1.3 .5.0 1.3 0.6 0.721 0.9585 +/- 0.0009 0.9604 133.8 251.1 cy_ 1.3 .6.0 1.4 0.1 0.654 0.9555 +/- 0.0010 0.9575 162.0 189.6 cy_ 1.4 .1.0 1.4 0.2 0.681 0.9560 +/- 0.0010 0.9580 149.0 212.8 cy_ 1.4 .2.0 1.4 0.3 0.703 0.9535 +/- 0.0009 0.9554 140.2 236.0 cy_ 1.4 .3.0 1.4 0.4 0.723 0.9514 +/- 0.0011 0.9536 130.5 259.2 cy_ 1.4 .4.0

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 44 of74 Table 12. Cylindrical TMI Fuel Pellets in the Knockout (KG) Canister (Continued)

Axial Void Radial Pellets Fraction, keff + AENCF H/X Pitch, cm GaD, cm Vf keff+/- cr 2cr , keV Ratio File Name 1.4 0.6 0.756 0.9381 +/- 0.0010 0.9400 119.8 305.6 cy 1.4 .6.0 1.5 0.1 0.697 0.9575 +/- 0.0010 0.9595 143.6 231.0 cy 1.5 .1.0 1.5 0.2 0.720 0.9552 +/- 0.0010 0.9571 132.9 257.6 cy 1.5 .2.0 1.5 0.3 0.740 0.9516 +/- 0.0010 0.9535 125.1 284.2 cy_ 1.5 .3.0 NOTES: a Canister contains one assembly's worth of pellets.

b Pellets are essentially touching in the radial direction.

C Pellet stacks fill the entire length of the canister.

In the first set of Table 12, the pellet array is positioned in the KG canister where the internal poison tubes are neglected (referred to as "without internals"). The pitch of the pellet array and the gap between pellets are varied. For a given pitch, the pellet stacks are positioned so as to best fill the cross-section of the KG canister. The number of pellets per stack (length of stack) is adjusted to give one assembly's worth of pellets per KG canister. The results show the effect of these variations. For the first several cases of the first set, the pellets are axially touching ( 0 em gap) and the pitch is varied whereas in the remaining cases of the set the pitch is fixed and the gap is varied.

Similar results are given in the second set, but the internals of the KG canister are included. The portion of any pellet stack that intersects a poison tube is represented as being a partial (incomplete) pellet. In this set, unlike the previous set, the pellet stacks fill the entire length of the KG canister.

This models more than one assembly's worth of pellets in the KG canister, with the exact number depending on the pitch and gap. Because no internals are in the upper portion of the canister, i.e.,

the poison tubes do not extend over the entire inner length of the canister, the fuel pellets always fill this upper (open) portion. Cross-sectional views of the canister at these different elevations are shown in Figure 6.

For cases without canister internals, the results show that for a fixed value of pitch a most reactive case occurs at a specific value of gap.. As the pitch increases, these most reactive cases occur at a decreasing value of gap. The largest kerr, 1.0751, occurs for pitch and gap equal to 1.5 em and 0.2 em, respectively, though statistically identical values also occur for other values of pitch and gap.

For example, kerr for a gap of 0 em and pitch of 1.7 em is just a few cr smaller than this largest value. It is interesting to note that these values all have an array void fraction of about 0.72. Any increase in gap and/or pitch causes keff to further decrease. These values of kerr> 1 show that the degraded canister internals are needed for reactivity control. In the second set, the most reactive case, kerr = 0.9667, occurs for a pitch and gap of 1.3 em and 0.4 cm, respectively, though statistically equivalent results also occurs for other values of pitch and gap. The results show similar behavior as the previous set, though the most reactive cases occur for an array void fraction of about 0.68.

The results in the remainder of this section are for spherical pellets and are investigated to insure that the most reactive pellet configuration is modeled. A radial pitch characterizes the spacing of the fuel stacks, but a dimensionless axial pitch is used instead of an axial gap. This dimensionless

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 45 of74 pitch is defined to be (p axial - p min) / (p radial - P min), where p axial is the axial pitch, p min is the minimum pitch given by p min = 2R (R is the pellet radius) and p radial is the radial pitch. This is a relative measure of the axial to radial spacing between pellets.

The results in Table 13 are for spherical pellets in a KO canister with the poison tubes neglected.

The number of pellets is chosen to give an assembly's worth of fuel. The table is divided into four sets, one for each value of radius, and the cases within each set are organized with varying radial pitches for fixed values of the dimensionless axial pitch. Values of kerr for the most reactive case of each set are, 1.0742, 1.0749, 1.0756 and 1.0759 and occur for an array void fraction of about 0.73.

These values are statistically identical to each other and to the value for the cylindrical pellets.

Table 13. Spherical TMI Fuel Pellets in the Knockout (KG) Canister a without Internals Dimensio Radial n-Iess Void Pitch, Axial Fraction, AENCF, H/X cm Pitch V, keff+/- cr keff + 2cr keV Ratio File Name

=

Spherical Pellets (R 0.4699 cm) 1.1 O. 0.600 1.0324 +/- 0.0011 1.0345 187.9 147.4 wp1ae-1 Ut.o 1.2 O. 0.662 1.0605 +/- 0.0010 1.0624 156.0 192.6 wp1 ae-I 1.21.0 1.3 O. 0.713 1.0711 +/-0.0010 1.0732 132.0 241.6 wp1 ae-I 1.31.0 1.4 O. 0.749 1.0720 + 0.0010 1.0739 116.7 294.6 wp1 ae-I 1.41.0 1.5 O. 0.780 1.0608 +/- 0.0010 1.0627 104.4 351.5 wp1 ae-I 1.51.0 1.1 0.5 0.631 1.0473 +/- 0.0011 1.0496 171.0 167.7 wp1ae-1 1.1h.0 1.2 0.5 0.704 1.0694 +/- 0.0010 1.0714 135.1 231.7 wp1 ae-I 1.2h.0 1.3 0.5 0.759 1.0706 +/- 0.0010 1.0725 112.3 305.2 wp1ae-1 1.3h.0 1.4 0.5 0.798 1.0549 +/- 0.0010 1.0569 97.1 388.7 wp1ae-1 1.4h.0 1.1 1.0 0.658 1.0577 +/- 0.0010 1.0597 156.8 187.9 wp1ae-1 1.1.0 1.2 1.0 0.736 1.0722 +/- 0.0010 1.0742 121.4 270.8 wp1ae-1 1.2.0 1.3 1.0 0.793 1.0586 +/- 0.0010 1.0606 99.1 368.7 wp 1ae-I 1.3.0 1.4 1.0 0.831 1.0247 +/- 0.0009 1.0266 85.5 482.9 wp1 ae-I 1.4.0 1.1 1.5 0.681 1.0671 +/- 0.0010 1.0690 146.3 208.1 wp1ae-1 1.1+h.0 1.2 1.5 0.762 1.0682 +/- 0.0010 1.0701 110.8 309.9 wp1ae-1 1.2+h.0 1.3 1.5 0.818 1.0393 +/- 0.0009 1.0410 90.7 432.2 wp1ae-1 1.3+h.0 wp1ae-1.0 2.0 0.577 1.0186 +/- 0.0011 1.0209 202.7 131.4 1 1.0+2h.0 wp1ae-1.1 2.0 0.702 1.0708 +/- 0.0010 1.0727 137.1 228.4 I 1.1+2h.0 wp1ae-1.2 2.0 0.783 1.0617 +/- 0.0011 1.0638 103.1 349.0 1 1.2+2h.0 wp1ae-1.3 2.0 0.838 1.0176 +/- 0.0009 1.0195 83.9 495.8 1 1.3+2h.0

=

Spherical Pellets (R 0.5200 cm) 1.2 O. 0.590 1.0310 +/- 0.0010 1.0329 193.1 140.8 wp 1ac-I 1.21.0 1.3 O. 0.649 1.0596 +/- 0.0010 1.0615 163.0 180.8 wp 1ac-I 1.31.0 1.4 O. 0.697 1.0700 +/- 0.0011 1.0721 139.3 224.1 wp1 ac-I 1.41.0 1.5 O. 0.734 1.0718 +/- 0.0010 1.0737 123.9 270.5 wp1ac-1 1.51.0 1.2 0.5 0.619 1.0445 +/- 0.0011 1.0467 177.6 158.5 wp1 ac-I 1.2h.0 1.3 0.5 0.688 1.0696 +/- 0.0010 1.0715 142.9 214.6 wp1 ac-I 1.3h.0 1.4 0.5 0.742 1.0720 +/- 0.0009 1.0738 119.1 278.4 wP1ac-1 1.4h.0 1.5 0.5 0.783 1.0616 +/- 0.0010 1.0636 104.7 350.2 wp1ac-1 1.5h.0 1.2 1.0 0.644 1.0554 +/- 0.0010 1.0574 163.9 176.2 wo1ac-1 1.2.0 1.3 1.0 0.719 1.0730 +/- 0.0009 1.0749 130.1 248.5 wp1 ac-I 1.3.0

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 46 of74 Table 13. Spherical TMI Fuel Pellets in the Knockout (KO) Canister a without Internals (Continued)

Dimensio Radial n-Iess Void Pitch, Axial Fraction, AENCF, HIX cm Pitch Vf keff+/- U keff + 20- keV Ratio File Name 1.4 1.0 0.775 1.0656 +/- 0.0010 1.0676 107.3 332.8 wp 1ac-I 1.4.0 1.5 1.0 0.816 1.0404 +/- 0.0010 1.0423 91.7 430.0 wp1ac-1 1.5.0 1.2 1.5 0.667 1.0623 +/- 0.0010 1.0644 154.1 194.0 wp1 ac-I 1.2+h.0 1.3 1.5 0.744 1.0713 +/- 0.0010 1.0733 119.2 194.0 wp1ac-1 1.3+h.0 1.4 1.5 0.801 1.0492 +/- 0.0009 1.0511 97.8 194.0 wp1ac-1 1.4+h.0 1.5 1.5 0.840 1.0137 +/- 0.0009 1.0156 84.2 194.0 wp 1ac-I 1.5+h.0 wp1ac-1.2 2.0 0.686 1.0680 +/- 0.0010 1.0700 144.1 211.7 I 1.2+2h.0 wp1ac-1.3 2.0 0.766 1.0676 +/- 0.0010 1.0695 110.0 211.7 I 1.3+2h.0 wp1ac-1.4 2.0 0.821 1.0321 +/- 0.0010 1.0340 90.5 211.7 I 1.4+2h.0

=

Spherical Pellets (R 0.5450 cm) 1.1 O. 0.472 0.9514 +/- 0.0011 0.9535 271.8 86.5 wp1ab-1 1.11.0 1.2 O. 0.555 1.0115 +/- 0.0010 1.0135 215.1 120.1 wp1ab-1 1.21.0 1.3 O. 0.615 1.0460 +/- 0.0010 1.0479 179.9 156.5 wp1ab-1 1.31.0 1.4 O. 0.668 1.0665 +/- 0.0011 1.0687 153.3 195.9 wp1ab-1 1.41.0 1.5 O. 0.709 1.0724 +/- 0.0009 1.0742 135.3 238.2 wp1ab-1 1.51.0 1.6 O. 0.741 1.0720 +/- 0.0009 1.0739 121.6 283.4 wp1ab-1 1.61.0 1.1 0.5 0.474 0.9534 +/- 0.0010 0.9555 269.5 87.3 wp1ab-1 1.1h.0 1.2 0.5 0.577 1.0227 +/- 0.0010 1.0248 202.2 130.7 wp1ab-1 1.2h.0 1.3 0.5 0.649 1.0570 +/- 0.0010 1.0591 162.8 180.3 wp1ab-1 1.3h.0 1.4 0.5 0.709 1.0736 +/- 0.0010 1.0756 133.4 236.6 wp1ab-1 1.4h.0 1.5 0.5 0.755 1.0711 +/- 0.0011 1.0732 115.0 299.9 wp1ab-1 1.5h.0 1.6 0.5 0.790 1.0560 +/- 0.0009 1.0578 101.9 370.8 wp1ab-1 1.6h.0 1.1 1.0 0.477 0.9529 +/- 0.0012 0.9552 267.4 88.1 wp1ab-1 1.1.0 1.2 1.0 0.596 1.0300 +/- 0.0010 1.0319 189.8 141.3 wp1ab-1 1.2.0 1.3 1.0 0.677 1.0669 +/- 0.0010 1.0689 148.1 204.0 wp1ab-1 1.3.0 1.4 1.0 0.741 1.0722 +/- 0.0010 1.0742 120.0 277.2 wp1ab-1 1.4.0 1.5 1.0 0.789 1.0570 +/- 0.0010 1.0589 101.9 361.6 wp1ab-1 1.5.0 1.6 1.0 0.823 1.0298 +/- 0.0009 1.0316 89.4 458.1 wp1ab-1 1.6.0 1.2 1.5 0.614 1.0426 +/- 0.0011 1.0447 181.5 151.9 wp1ab-1 1.2+h.0 1.3 1.5 0.701 1.0715 +/- 0.0010 1.0735 138.1 227.8 wp1ab-1 1.3+h.0 1.4 1.5 0.767 1.0676 +/- 0.0010 1.0696 110.6 317.9 wp1ab-1 1.4+h.0 1.5 1.5 0.814 1.0388 +/- 0.0010 1.0407 93.4 423.4 wp1ab-1 1.5+h.0 wp1ab-1.2 2.0 0.630 1.0486 +/- 0.0011 1.0508 173.0 162.5 I 1.2+2h.0 wp1ab-1.3 2.0 0.722 1.0737 +/- 0.0009 1.0755 129.6 251.5 I 1.3+2h.0 wp1ab-1.4 2.0 0.788 1.0575 +/- 0.0010 1.0594 103.0 358.5 I 1.4+2h.0 Spherical Pellets (R =0.56772cm) 1.136 O. 0.462 0.9438 +/- 0.0011 0.9461 280.5 83.5 wp1a-1 tt.o 1.2 O. 0.519 0.9860 +/- 0.0010 0.9880 239.9 103.6 wp1a-1 1.21.0 1.3 O. 0.582 1.0257 +/- 0.0010 1.0278 199.3 137.2 wp1 a-I 1.31.0 1.4 O. 0.640 1.0563 +/- 0.0011 1.0585 167.0 173.5 wp1 a-I 1.41.0 1.5 O. 0.685 1.0699 +/- 0.0009 1.0717 145.9 212.5 wp1a-1 1.51.0 1.6 O. 0.719 1.0716 +/- 0.0010 1.0735 131.1 254.1 wp 1a-I 1.61.0 1.7 O. 0.751 1.0691 +/- 0.0009 1.0708 117.7 298.5 wp1a-1 1.71.0 1.2 0.5 0.532 0.9933 +/- 0.0011 0.9954 230.8 109.1 wp1a-1 1.2h.o

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 47 of74 Table 13. Spherical TMI Fuel Pellets in the Knockout (KO) Canister a without Internals (Continued)

Dimensio Radial n-Iess Void Pitch, Axial Fraction, AENCF, H/X cm Pitch V, keff+/- CJ keff + 2CJ keY Ratio File Name 1.3 0.5 0.610 1.0413 +/- 0.0010 1.0433 183.0 153.7 wp1a-1 1.3h.o 1.4 0.5 0.677 1.0699 +/- 0.0010 1.0720 147.5 204.2 wp1a-1 1.4h.o 1.5 0.5 0.729 1.0724 +/- 0.0010 1.0744 127.2 261.0 wp1a-1 1.5h.o 1.6 0.5 0.766 1.0683 +/- 0.0010 1.0702 110.5 324.5 wp1a-1 1.6h.o 1.2 1.0 0.544 1.0022 +/- 0.0010 1.0042 222.7 114.6 wp1a-1 1.2.0 1.3 1.0 0.635 1.0546 +/- 0.0009 1.0565 170.0 170.1 wp1a-1 1.3.0 1.4 1.0 0.708 1.0719 +/- 0.0010 1.0739 134.7 234.9 wp1a-1 1.4.0 1.5 1.0 0.762 1.0683 +/- 0.0009 1.0701 112.6 309.6 wp1a-1 1.5.0 1.6 1.0 0.800 1.0493 +/- 0.0009 1.0511 98.7 394.9 wp1a-1 1.6.0 1.2 1.5 0.556 1.0111 +/- 0.0010 1.0131 215.2 120.1 wp1a-1 1.2+h.o 1.3 1.5 0.657 1.0637 +/- 0.0010 1.0657 158.9 186.6 wp1a-1 1.3+h.o 1.4 1.5 0.733 1.0740 +/- 0.0010 1.0759 124.8 265.6 wp1a-1 1.4+h.o 1.5 1.5 0.787 1.0568 +/- 0.0009 1.0587 104.0 358.1 wp1 a-I 1.5+h.o 1.2 2.0 0.567 1.0193 +/- 0.0011 1.0214 208.3 125.6 wp1a-1 1.2+2h.o 1.3 2.0 0.676 1.0658 +/- 0.0010 1.0678 150.6 203.0 wp1a-1 1.3+2h.o 1.4 2.0 0.754 1.0700 +/- 0.0010 1.0721 116.6 296.3 wp1 a-I 1.4+2h.o 1.5 2.0 0.808 1.0428 +/- 0.0010 1.0447 95.5 406.7 wp1a-1 1.5+2h.o NOTE: a Canister contains one assembly's worth of pellets.

The results in Table 14 are for spherical pellets in the KO canister with internal poison tubes. The pellet stacks extend the entire length of the KO canister. As was done with the cylindrical pellets, this models more than one assembly's worth of pellets, with the exact number depending on the axial and radial pitches. The table is divided into sets, one for each value of the four radii of interest. Values of kerr for the most reactive case of each set are 0.9662, 0.9668, 0.9662 and 0.9665, and they occur for an array void fraction of about 0.67. These values are again statistically identical to each other and to the value of kerr for the most reactive case with cylindrical pellets.

Table 14. Spherical TMI Fuel Pellets in the Knockout (KO) Canister a with Internals Radial Dimension Void Pitch, -less Axial Fraction, AENCF, H/X cm Pitch V, keff+/- CJ keff + 2CJ keY Ratio File Name

=

Spherical Pellets (R 0.4699 cm) 1.0 O. 0.523 0.9188 +/- 0.0010 0.9208 232.9 106.2 wp1ae 1.01.0 1.1 O. 0.600 0.9521 +/- 0.0010 0.9540 188.4 147.4 wp1ae 1.11.0 1.2 o. 0.662 0.9643 +/- 0.0009 0.9662 157.7 192.6 wp1ae 1.21.0 1.3 O. 0.713 0.9610 +/- 0.0009 0.9628 134.1 241.6 wp1 ae 1.31.0 1.0 0.5 0.538 0.9246 +/- 0.0010 0.9265 224.3 112.5 wp1ae 1.0h.o 1.1 0.5 0.631 0.9574 +/- 0.0010 0.9594 173.3 167.7 wp1ae 1.1h.o 1.2 0.5 0.704 0.9640 +/- 0.0010 0.9660 139.3 231.7 wp1ae 1.2h.o 1.3 0.5 0.759 0.9448 +/- 0.0009 0.9467 116.7 305.2 wp1ae 1.3h.o 1.0 1.0 0.551 0.9304 +/- 0.0010 0.9324 215.1 118.8 wp1ae 1.0.0 1.1 1.0 0.658 0.9624 +/- 0.0010 0.9644 159.0 187.9 wp1ae 1.1.0 1.2 1.0 0.736 0.9584 +/- 0.0010 0.9604 125.2 270.8 wp1ae 1.2.0 1.3 1.0 0.793 0.9264 +/- 0.0009 0.9282 104.2 368.7 wp1ae 1.3.0 1.0 1.5 0.565 0.9343 +/- 0.0011 0.9364 207.3 125.1 wp1ae 1.0+h.o 1.1 1.5 0.681 0.9620 +/- 0.0010 0.9640 148.8 208.1 wp1ae 1.1+h.o

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 48 of74 Table 14. Spherical TMI Fuel Pellets in the Knockout (KG) Canister a with Internals (Continued)

Radial Dimension Void Pitch, -less Axial Fraction, AENCF, H/X cm Pitch Vf keff+/- C1 keff + 2C1 keV Ratio File Name 1.2 1.5 0.762 0.9493 +/- 0.0009 0.9511 115.7 309.9 wp1ae 1.2+h.0 1.3 1.5 0.818 0.9029 +/- 0.0010 0.9048 96.2 432.2 wp1ae 1.3+h.0 Spherical Pellets R = 0.520 cm 1.1 O. 0.520 0.9249 +/- 0.0010 0.9269 234.3 103.9 wp1ac 1.11.0 1.2 O. 0.590 0.9494 +/- 0.0010 0.9514 196.0 140.8 wp1ac 1.21.0 1.3 O. 0.649 0.9622 +/- 0.0010 0.9642 163.5 180.8 wp1 ac 1.31.0 1.4 O. 0.697 0.9565 +/- 0.0009 0.9584 143.0 224.1 wp1 ac 1.41.0 1.1 0.5 0.533 0.9300 +/- 0.0010 0.9320 225.9 109.5 wp1ac 1.1h.0 1.2 0.5 0.619 0.9573 +/- 0.0010 0.9592 179.5 158.5 wp1ac 1.2h.0 1.3 0.5 0.688 0.9648 +/- 0.0010 0.9668 145.1 214.6 wp1ac 1.3h.0 1.4 0.5 0.742 0.9481 +/- 0.0010 0.9501 124.5 278.4 wp1 ac 1.4h.0 1.1 1.0 0.546 0.9358 +/- 0.0009 0.9376 219.9 115.1 wp1ac 1.1.0 1.2 1.0 0.644 0.9619 +/- 0.0009 0.9636 167.8 176.2 wp1ac 1.2.0 1.3 1.0 0.719 0.9599 +/- 0.0011 0.9620 133.8 248.5 wp1ac 1.3.0 1.4 1.0 0.775 0.9300 +/- 0.0009 0.9319 112.6 332.8 wp1ac 1.4.0 Spherical Pellets (R = 0.5450 cm) 1.2 O. 0.555 0.9404 +/- 0.0010* 0.9423 214.2 120.1 wp1 ab 1.21.0 1.3 O. 0.615 0.9571 +/- 0.0010 0.9592 181.9 156.5 wp1 ab 1.31.0 1.4 O. 0.668 0.9588 +/- 0.0009 0.9606 156.0 195.9 wp1 ab 1.41.0 1.5 O. 0.709 0.9555 +/- 0.0010 0.9574 139.0 238.2 wp1 ab 1.51.0 1.2 0.5 0.577 0.9481 +/- 0.0009 0.9500 201.1 130.7 wp1ab 1.2h.0 1.3 0.5 0.649 0.9634 +/- 0.0010 0.9654 163.4 180.3 wp1ab 1.3h.0 1.4 0.5 0.709 0.9562 +/- 0.0009 0.9580 137.9 236.6 wp1ab 1.4h.0 1.2 1.0 0.596 0.9517 +/- 0.0009 0.9535 192.1 141.3 wp1ab 1.2.0 1.3 1.0 0.677 0.9643 +/- 0.0010 0.9662 151.3 204.0 wp1ab 1.3.0 1.4 1.0 0.741 0.9470 +/- 0.0010 0.9490 125.2 277.2 wp1ab 1.4.0 1.2 1.5 0.614 0.9555 +/- 0.0009 0.9573 183.6 151.9 wp1ab 1.2+h.0 1.3 1.5 0.701 0.9628 +/- 0.0010 0.9647 141.5 227.8 wp1ab 1.3+h.0 1.4 1.5 0.767 0.9362 +/- 0.0010 0.9381 115.3 317.9 wp1ab 1.4+h.0 1.1 2.0 0.481 0.9072 +/- 0.0011 0.9094 258.9 89.8 wp1ab 1.1+2h.0 1.2 2.0 0.630 0.9596 +/- 0.0010 0.9615 175.6 162.5 wp1ab 1.2+2h.0 1.3 2.0 0.722 0.9595 +/- 0.0009 0.9614 132.2 251.5 wp1 ab 1.3+2h.0 1.4 2.0 0.788 0.9207 +/- 0.0009 0.9225 108.2 358.5 wp1 ab 1.4+2h.0 Spherical Pellets (R = 0.56772cm) 1.2 O. 0.519 0.9258 +/- 0.0009 0.9277 232.9 103.6 wp1a 1.21.0 1.3 O. 0.582 0.9483 +/- 0.0010 0.9504 198.9 137.2 wp1a 1.31.0 1.4 O. 0.640 0.9587 +/- 0.0010 0.9606 169.5 173.5 wp1a 1.41.0 1.5 O. 0.685 0.9571 +/- 0.0010 0.9590 150.5 212.5 wp1a 1.51.0 1.6 O. 0.719 0.9620 +/- 0.0009 0.9638 133.8 254.1 wp1a 1.61.0 1.2 0.5 0.532 0.9317 +/- 0.0009 0.9336 226.2 109.1 wp1a 1.2h.0 1.3 0.5 0.610 0.9569 +/- 0.0009 0.9588 183.3 153.7 wp1a 1.3h.0 1.4 0.5 0.677 0.9602 +/- 0.0010 0.9621 152.9 204.2 wp1a 1.4h.0 1.5 0.5 0.729 0.9523 +/- 0.0009 0.9541 131.2 261.0 wp1a 1.5h.0 1.2 1.0 0.544 0.9351 +/- 0.0010 0.9372 220.6 114.6 wp1a 1.2.0 1.3 1.0 0.635 0.9622 +/- 0.0009 0.9640 171.7 170.1 wp1a 1.3.0 1.4 1.0 0.708 0.9576 +/- 0.0010 0.9596 139.0 234.9 wp1a 1.4.0 1.5 1.0 0.762 0.9363 +/- 0.0009 0.9382 117.9 309.6 wp1a 1.5.0 1.2 1.5 0.556 0.9400 +/- 0.0010 0.9419 213.0 120.1 wp1a 1.2+h.0 1.3 1.5 0.657 0.9645 +/- 0.0010 0.9665 161.6 186.6 wp1a 1.3+h.0

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 49 of74 Table 14. Spherical TMI Fuel Pellets in the Knockout (KG) Canister" with Internals (Continued)

Radial Dimension Void Pitch, -less Axial Fraction, AENCF, HfX cm Pitch V, keff+/ keff + 20- keV Ratio File Name 1.4 1.5 0.733 0.9525 +/- 0.0010 0.9544 128.6 265.6 wp1a 1.4+h.o 1.5 1.5 0.787 0.9216 +/- 0.0010 0.9236 109.2 358.1 wp1a 1.5+h.o 1.2 2.0 0.567 0.9457 +/- 0.0010 0.9477 205.7 125.6 wp1a 1.2+2h.o 1.3 2.0 0.676 0.9640 +/- 0.0009 0.9658 152.7 203.0 wp1a 1.3+2h.o 1.4 2.0 0.754 0.9444 +/- 0.0011 0.9465 121.3 296.3 wp1a 1.4+2h.o NOTE: a Pellet stacks fill the entire length of the canister.

In summary, the intact cylindrical pellets are as reactive as the four spherical pellet types considered provided the pellet spacing is optimal. The most reactive cases occur for array void fractions of about 0.72 and 0.67 for canisters without and with internals, respectively, regardless of whether the pellets are cylindrical or spherical. This shows that this parameter is useful in characterizing the pellet array.

In the next table, Table 15, the TMI canister is dry (the fuel is still water saturated), the poison tube inserts are neglected and the canister contains one assembly's worth of fuel. (These are variations of cases in Table 13, but water in the canister is neglected.) In the first case, the pellets are touching in the axial and radial directions, while in the rest of the cases the pellets are separated in either the axial or radial directions. Any increase in spacing between pellets is seen to decrease kerr, though there are no criticality concerns for any of these cases. .

Table 15. Spherical TMI Fuel Pellets in Dry Knockout (KG) Canister" without Internals Radial Pitch, cmf Void AEN Fractio keff + CF, HfX n keff+/ 20- keV Ratio Comment File Name No water in KO canister containing Spherical Pellets (R=O.56772 cm)

Variation of case wp1a-Ut.o; pellets 1.136/ 1406.

0.3069 +/- 0.0005 0.3079 7.3 touching in radial and axial wp1a-Ut-w.o 0.462 0 directions 1.136/ 1461. Axial gap between pellets is 0.2 em; wp1a-U+*2-0.2733 +/- 0.0004 0.2741 7.3 0.543 6 pellets touching in radial direction w.o 1.2/ 1445. Variation of case wp1 a-'-1.2t.o; wp1 a-'-1.2t-0.2838 +/- 0.0004 0.2847 7.3 0.519 3 pellets touching in axial direction w.o 1.4/ 1531. Variation of case wp1 a-'-1.4t.o; wp1a-U*4t-0.2283 +/- 0.0004 0.2290 7.3 0.640 1 pellets touching in axial direction w.o NOTE: " Canister contains one assembly's worth of pellets.

The results in Table 16 are variations of cases in the previous tables and show the effects of different canister and/or pellet configurations and loadings. In the first four sets of the table, approximately one assembly's worth of fuel pellets are loaded in the KG canister with poison tube inserts. For cases where the entire canister length is not filled with pellets and the poison tube inserts are present, the fuel always fills the upper (open) portion of the canister, unless noted

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 50 of74 otherwise. Values of kerr for these cases are 0.9637,0.9608,0.9616 and 0.9605 for each of the four different radii. The difference between the most reactive of these cases and that for the canister completely filled with fuel is not statistically significant, indicating that an assembly's worth of fuel in the canister is neutronically infinitely long. Also in the first set are cases where the canister contains an entire assembly's worth of pellets, but fills only 90% and 80% of the canister's cross-section. This is done by reducing the number of pellet stacks. For a horizontally positioned canister it would appear more likely that the cross-section would be partially rather than completely filled with fuel. Each reduction in cross-section decreases kerr by 0.026 and 0.031. Cases where the canister is rotated 45° are included since this gives a different pellet array configuration with respect to the outer poison tubes, see Figure 7. This effect on kerr is insignificant.

Table 16. Variations of Cases with Spherical Fuel Pellets in the Knockout (KO) Canister Radial Pitch, cm/ AEN Void keff + CF. H/X Fraction keff+/- cr 2cr keV Ratio CommentS File Name Cross-sectional Area Varies for Array of Spherical Pellets (R=O.4699 cm) (Internals Modeled) 1.2/ 0.9616 +/- Case wp1ae_1.2h.o but contains one 0.9637 139.0 231.7 wae_1.2hL 1as.o 0.704 0.0010 assembly's worth of pellets 1.2/ 0.9360 +/- Case wae_1.2hL1as.o but only 90%

0.9379 138.2 231.7 wae_1.2hL 1_.9.0 0.704 0.0010 as many pellet stacks (607/674) 1.2/ 0.9359 +/- wae_1.2hL 1_.9a.

0.9378 137.6 231.7 Previous case but canister rotated 45° 0.704 0.0009 0 1.2/ 0.9047 +/- Case wae_1.2hL1as.o but only 80%

0.9067 141.2 231.7 wae_1.2hL 1_.8.0 0.703 0.0010 as many pellet stacks (539/674) 1.2/ 0.9061 +/- wae_1.2hL 1_.8a.

0.9079 140.0 231.7 Previous case but canister rotated 45° 0.703 0.0009 0 Spherical Pellets (R=O.520 cm) in Canister with Internals 1.3/ 0.9588 +/- Case wp1 ac_1.3h.o but contains one 0.9608 146.5 214.6 wac_1.3h1as.o 0.688 0.0010 assembly's worth of pellets 1.3/ 0.9378 +/- Case wp1 ac_1.3h.o but water 0.9398 149.0 214.6 wac_1.3h+w.o 0.688 0.0010 between SNF and KO canisters Spherical Pellets (R=O.5450 cm) in Canister with Internals 1.3/ 0.9596 +/- Case wp1ab_1.3.0 but contains one 0.9616 151.7 204.0 wab_1.3as.o 0.677 0.0010 assembly's worth of pellets Spherical Pellets (R=O.56772 cm), Cases are Variations of Case wp1a 1.3+h.o 1.3/ 0.9586 +/- Contains one assembly's worth of 0.9605 163.0 186.6 wL 1asa.o 0.657 0.0009 pellets Case wL 1asa.o but KO canister 1.3/ 0.9574 +/- moved to top of SNF canister, which in 0.9592 162.3 186.6 wL 1asaR.o 0.657 0.0009 turn is moved to top (right) of waste package 1.3/ 0.9568 +/- Case wL 1asa.o but KO canister filled 0.9587 161.2 186.6 wL 1Lasa.o 0.657 0.0010 from bottom end up 1.3/ 0.9593 +/- Thickness of sleeve reduced by 1/2 0.9613 162.4 186.6 hslv.o 0.657 0.0010 (0.25" thick) 1.3/ 0.9549 +/-

0.9568 163.0 186.6 Sleeve is neglected noslv.o 0.657 0.0009 1.3/ 0.9650 +/- Stainless steel sleeve replaced with 0.9670 160.4 186.6 cstl.o 0.657 0.0010 carbon steel sleeve 1.3/ 0.9557 +/- Case cstl.o but fuel voids contain no 0.9577 165.9 179.3 cstl-sat.o 0.657 0.0010 water (fuel completely dry)

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 51 of74 Table 16. Variations of Cases with Spherical Fuel Pellets in the Knockout (KG) Canister (Continued)

Radial Pitch, cml AEN Void CF, H/X Fraction keV Ratio Comment a File Name 1.31 0.9639 +/-

0.9659 161.8 186.6 Pellets fill 87.5% of KO canister length wL-1.0 0.657 0.0010 1.3/ 0.9620 +/-

0.9641 162.0 186.6 Pellets fill 75% of KO canister length wL-2.0 0.657 0.0011 1.3/ 0.9615 +/-

0.9636 161.4 186.6 Pellets fill 62.5% of KO canister length wL-3.0 0.657 0.0010 1.3/ 0.9550 +/-

0.9569 161.7 186.6 Pellets fill 50% of KO canister length wL-4.0 0.657 0.0010 1.3/ 0.9534 +/- Contains 84.2% of an assembly's 0.9555 162.4 186.6 wL.842.0 0.657 0.0010 worth 1.3/ 0.9485 +/- Contains 68.7% of an assembly's 0.9506 165.2 186.6 wL.687.0 0.657 0.0011 worth 1.3/ 0.9586 +/-

0.9604 162.5 186.6 HLW canisters in gravity position grav1.0 0.657 0.0009 1.3/ 0.9609 +/- 0.9629 161.8 186.6 HLW canisters in gravity position, grav2.0 0.657 0.0010 sleeve collapsed 1.3/ 0.9613 +/- Previous case, but earlier waste 0.0010 0.9634 162.9 186.6 packaQe desiQn dimensions used grav2wp.o 0.657 Spherical Pellets (R=O.56772 cm), Cases are Variants of Case wL1asa.o 1.3/ 0.9844 +/- 09863 1640 1765 Fuel enrichment is 2.96 wt% and wL 1asa+Pu.o 0.657 0.0010 . . . contains one kg of Pu-239 per canister 1.3/ 0.9248 +/- 0.9267 167.2 212.0 Fuel enrichment is 2.64 wt% wL 1asaiE.o 0.657 0.0010 (intermediate enrichment) 1.31 0.9602 +/- 09622 1679 1963 Fuel enrichment is 2.64 wt% and wL 1asaiE+Pu.o 0.657 0.0010 . . . contains one kQ of Pu-239 per canister Spherical Pellets (R=O.56772 cm), Variants of Case wp1 a-I_1.4+h.o (without Internals) 1.4/ 1.0803 +/- 1.0821 123.5 265.6 Entire length of canister filled with wa-L1.4+hal.o 0.733 0.0009 pellets 1.4/ 1.0797 +/- Previous case but earlier waste wa-0.733 0.0010 1.0817 124.4 265.6 packaQe desiQn dimensions are used I 1.4+halwp.o 1.4/ 1 0522 +/- Canister contains one assembly's 60010 1.0542 123.9 265.6 worth of pellets with 90% as many wa-L1.4+h_.9.0 0.733

. pellet stacks (444/493) 1.4/ 1.0212 +/- 1.0232 124.6 265.6 Case wa-I 1.4+h .9.0 but 80% as 0.732 0.0010 manv oellet stacks (394/493) 1.4/ 0.9868 +/- 0.9889 126.3 265.6 Case wa-I 1.4+h .9.0 but 70% as 0.732 0.0010 many pellet stacks (345/493)

Case wa-I_1.4+h_.9.0 but 60% as 1.4/ 0.9407 +/-

0.9426 128.8 265.6 many pellet stacks (296/493); entire wa-L1.4+h_.6.0 0.731 0.0010 lenQth filled NOTE: a Ratio in parentheses is the reduced number to the original number of pellets stacks.

In the second set, a case is included where the entire SNF canister (outside the TMI canister) is water flooded. This reduces keff by 0.027.

In the second case of the fourth set of Table 16, the top of the KG canister is positioned next to the top of the SNF canister, which in tum is moved laterally to the top of the waste package. In the

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 52 of74 third case of this same set, the KO canister is filled from the bottom up with an assembly's worth of pellets so that the upper (open) portion of the canister does not contain any fuel. Both of these effects are negligible. In the next three cases of the set the stainless steel sleeve is reduced in thickness, neglected and replaced with a carbon steel sleeve of the same dimensions. Neglecting the sleeve reduces kerr by about 0.01, reducing the sleeve thickness by half has the same proportional effect, and replacing the stainless steel with carbon steel has no (statistical) effect. In the next case of the set, any water content in the fuel voids is neglected (for the case with the carbon steel sleeve), reducing kerr by 0.009. In the next 6 cases of the set, the length (and amount) of fuel in the KO canister is reduced until the canister contains less than an assembly's worth of fuel.

(Note that an assembly's worth of fuel fills about 51% of the length of the canister.) The difference in keff between a completely full canister and one containing 68.7% of an assembly's worth of fuel is only 0.016. Finally, in the second and third to the last cases of the set the KO, SNF and HLW canisters are positioned in what would be a gravity position for a horizontally oriented waste package. In the first of these cases, the sleeve is centered in the SNF canister, whereas in the second it is at the bottom of the SNF canister so as to simulate a collapsed sleeve, see Figure 9. The value of kerr for the first of these cases is reduced by 0.006, whereas the second is statistically unchanged. The last case of the set compares the earlier waste package design to the previous case in the set and shows that the results are insensitive to these dimensional changes.

In the fifth set of Table 16, variations in fuel enrichment and plutonium content are investigated.

These cases are variations of case wLlasa.o (base case) of the fourth set and contain an assembly's worth of fuel pellets. In the first case, the enrichment is reduced slightly to 2.96 wt% (actual value) and one kg of Pu-239 is added to the fuel. This increases kerr by 0.026 from the base case and exceeds the ICL, though this combination of enrichment and plutonium content is unrealistic. In the next two cases the enrichment is further reduced to that of the intermediate value (2.64 wt%)

and one of the cases also contains one kg of Pu-239. The value of kerr for the case without plutonium is reduced by 0.034 as compared to the base case, whereas kerr for the case with plutonium is statistically identical. This demonstrates that the fuel composition modeled here is conservative since the enrichment, U-235 content and plutonium content for this latter case exceeds that of any TMI canister.

In the last set of Table 16 the fuel loading is investigated for a KO canister that does not include the poison tube inserts. The base case for this set is case wpla-I_1.4+h.o of Table 13. In the first case of the set the canister is completely filled with fuel. This slightly increases keff by a few cr further indicating that an assembly's worth of fuel is neutronically infinitely long. The next case is a variation of the first case but uses the earlier waste package design dimensions. Results for these cases are statistically identical. For the rest of the cases of this set the canister contains an assembly's worth of fuel, but the fuel array fills only a portion, as indicated in the table, of the canister's cross-section. As the cross-section of the array decreases its length increases thus keeping the amount of fuel constant until the length of the canister is filled (this occurs at a cross-section of 60%) The ICL is only satisfied when the array occupies less than 70% of the canister's cross-section.

Before proceeding with further analysis of the TMI fuel, it is necessary to evaluate TMI fuel in the D-type fuel canister to ensure that it is not more reactive than the KO canister. As such the KO canister has been replaced with a fuel canister, and the reactivity of the system is evaluated. The

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 53 of74 fuel is modeled as intact cylindrical fuel pellets that are axially aligned to form pellet stacks that are positioned in a hexagonal array. No axial gap between pellets is considered since it was shown in the first set of Table 12 that a most reactive configuration can be achieved with axially touching pellets as long as the radial pitch is optimal. The fuel canister is water filled, and the empty space in the SNF canister (outside the fuel canister) is modeled as void.

Results are given in Table 17 for the fuel canister (D-type). In the first set of the table, the cross-section of the fuel canister is best filled with whole pellets in stacks, and the length of the stacks is chosen to give one assembly's worth of pellets. In the second set, partial pellet stacks are included around the perimeter of the fuel array. These stacks are where the fuel and inside wall of the canister intersect and this contributes fuel in addition to one assembly's worth. (All cases in the second set contain more than one assembly's worth of fuel.) The pitch of the stacks is varied for both sets. The most reactive case occurs for a pitch of 1.6 cm and an array void fraction of 0.69.

These cases are seen to be less reactive than those for the KG canister. Thus the KG canister is used for evaluating the TMI fuel.

Table 17. Cylindrical TMI Fuel Pellets in the Fuel Canister (D-type)

Axial Void Radial Pellet Fracti AENCF, H/X Pitch, cm GaD, cm on, V, keff+/- cr keff + 2cr keV Ratio File Name

=

(Whole) Cylindrical Pellets (R 0.4699 cm) 0.9048 +/-

1.4 O. 0.608 0.0011 0.9069 195.7 148.0 fh_1.4.0 0.9239 +/-

1.5 O. 0.651 0.0010 0.9259 171.8 183.2 fh_1.5.o 0.9298 +/-

1.6 O. 0.693 0.9318 151.0 220.8 fh_1.6.o 0.0010 0.9331 +/-

1.7 O. 0.721 0.9351 137.3 260.9 fh_1.7.o 0.0010 0.9216 +/-

1.8 O. 0.759 0.0010 0.9237 125.0 303.4 fh_1.8.o

=

(Whole and Partial) Cylindrical Pellets (R 0.4699 cm) 0.9087 +/-

1.4 O. 0.608 0.0010 0.9107 197.8 148.0 fh_1.4a.o 0.9278 +/-

1.5 O. 0.651 0.0011 0.9300 169.9 183.2 fh_1.5a.o 0.9366 +/-

1.6 O. 0.693 0.0011 0.9387 151.5 220.8 fh_1.6a.o 0.9355 +/-

1.7 O. 0.721 0.0010 0.9375 137.5 260.9 fh_1.7a.o 0.9285 +/-

1.8 O. 0.759 0.9306 124.3 303.4 fh_1.8a.o 0.0010 The final results in this section, see Table 18, are for homogenized TMI fuel in the KG canister that neglects the canister internals. For these cases, one assembly's worth of fuel is mixed with various amounts of water as listed in the "comment" column of the table. The remainder of the canister not occupied by the fuel mixture is water filled unless noted otherwise. Comparison of the result for the most reactive water moderated case here with those for the heterogeneous cases, Tables 12 and 13, shows that homogeneous fuel is less reactive.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 54 of74 Table 18. Homogenized TMI Fuel in the Knockout (KG) Canister (without Internals)

AEN keff + CF, H/X keff+/- cr 2cr keY Ratio Comment a File Name 0.3403 +/- 1787. No water in fuel or TMI canister; compare to 0.3411 0.0 hom_wO.o 0.0004 3 case wp1 a-I tt-w.O of Table 15 0.4476 +/- 1219. Voids in fuel are water saturated; remainder 0.4488 7.3 hom_sat-w.o 0.0006 8 of TMI canister is dry 0.4763 +/- 1064.

0.4776 7.3 Voids in fuel are water saturated hom_sat.o 0.0007 7 0.6732 +/- hom_w.2.0 0.6750 593.3 31.6 Fuel has 20% wvf 0.0009 0.8514 +/-

0.8534 339.9 72.1 Fuel has 40% wvf hom_w.4.0 0.0010 0.9806 +/-

0.9828 191.1 153.2 Fuel has 60% wvf hom_w.6.0 0.0011 1.0169+/-

1.0190 135.8 234.2 Fuel has 70% wvf hom_w.7.0 0.0010 1.0140 +/-

1.0159 89.6 396.3 Fuel has 80% wvf hom_w.8.0 0.0010 0.9826 +/-

0.9849 69.6 558.4 Fuel has 85% wvf hom_w.85.0 0.0012 0.9024 +/-

0.9051 50.1 882.5 Fuel has 90% wvf hom_w.9.0 0.0013 NOTE: a Water volume fraction (wvf) listed is in addition to the water in the fuel voids; canister contains one assembly's worth of fuel; unless noted otherwise, remainder of TMI canister is water flooded 6.2 DEGRADED MODE This section gives the results of the calculations described in Section 5.5. Section 6.2.1 presents the results of the calculation where the inner component of the SNF canister degrades before the high-level waste canisters, the SNF canister, and the waste package basket (see Sections 5.5.1). Section 6.2.2 gives the results for the calculation where the internal components of the waste package (but external to the DOE SNF canister) degrade first (see Section 5.5.2). Section 6.2.3 presents the results for a waste package with its internal components fully degraded (see Section 5.5.3).

The results in the previous tables show that TMI fuel modeled as intact cylindrical pellets is just as reactive as any of the spherical pellets that are considered. Thus cylindrical pellets with a diameter and length of 0.9398 cm and 1.1049 cm, respectively, are used for much of the degraded mode analysis. Typical degraded analysis would assume little if any axial redistribution of fuel.

Unfortunately, the initial axial distribution of the fuel is unspecified being somewhat random since it depends on the initial loading and subsequent handling of the TMI canister. As such the TMI fuel is assumed to have been in KO canisters because the unoccupied (open) volume above the poison tubes at the top of the canister has a larger cross-sectional area than that available in the fuel canisters. It was empirically determined that 1083 touching pellet stacks could fit into this open cross-section. Thus no more than 1083 pellet stacks are ever present in any degraded analysis. A larger number would only be possible if there were axial redistribution of fuel after the breach of the TMI canister which is not considered here. The length of the stacks is chosen to give one assembly's worth of fuel pellets. This choice for the maximum number of pellet stacks is very

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 55 of74 conservative as illustrated by the following considerations. If this limit were determined by assuming the pellets uniformly filled the entire axial length of the KG canister, then the number of pellet stacks could be as small as 214 for a canister that contained one assembly's worth of pellets.

This would reduce the fissile linear loading by more than a factor of five leading to a much less reactive configuration. The number of pellet stacks could easily be determined from a realistic, though conservative, estimate of the void fraction in the TMI canister. This is easily done with the expression for the array void fraction, Vr, given in Section 5.4.2. For intact, axially touching cylindrical pellets in the KG canister, 1083 pellets stacks corresponds to a void fraction of 18.6%.

Considering that the pellets are randomly packed and the canister contains other inert materials, e.g., zirconium cladding, this value seems unrealistically small.

Since a maximum of just less than 700 pellet stacks fit in the square box structure of the fuel canister, the results for the KG canister easily bound those for the fuel canister.

6.2.1 Inner Components of the SNF Canister Degrade First This section presents the results of the calculation described in Section 5.5.1. Here the KG canister and sleeve have degraded leaving degradation products and fuel in the SNF canister. Unless noted otherwise, there are 1083 pellet stacks in the canister and the sleeve was stainless steel. The iron components of the KG canister and sleeve degrade to goethite, though the degradation products from the nonferrous components have been released from the waste package and are neglected.

The SNF canister is flooded with water, unless noted otherwise, and the other waste package components external to the SNF canister are intact. The HLW canisters are positioned at closest approach to the SNF canister in a non-gravity position, though this detail is unimportant. to the criticality of the waste package.

Also listed in the results of this section is the height of the pellet array, H. This is defined as the distance from the inside bottom of the SNF canister to the top of the pellet array. If the array fills the entire cross-section of the canister, then this height is simply equal to the inside diameter of the canister, 43.8 cm. If there is no plausible mechanism to raise pellets against gravity, then the maximum values for H would be 36.2 cm and 37.8 cm for a collapsed and centered sleeve, respectively. Arrays with heights greater than these values are unrealistic since some of the pellets have been raised against gravity. This in effect places an upper limit on the maximum radial pitch for an array with a given number of pellet stacks. Nevertheless, results for such arrays are listed and are shown to be among the most reactive for this scenario.

In the first set of Table 19, the goethite is neglected leaving only fuel surrounded by water in the SNF canister. The first three cases have radially touching pellet stacks and varying axial separation between pellets. For a sufficiently large gap, keff is greater than critical. In the next case, keff is also greater than critical for a radial pitch of 1.6 cm and axially touching pellets. In the final case of the set, water is neglected in the canister and the pellets are touching in the radial and axial directions, showing that there are no criticality concerns for a dry canister.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 56 of74 Table 19. TMI Fuel Pellets in SNF Canister (Degraded'Sleeve and KG Canister)

Radial Pitch x Axial Gap AEN H/X (cm)1 Void keff+ CF, Ratio a b Fraction keff+/- 0' 20' keV Comment File Name Water in SNF Canister; Goethite from Degraded KO Canister and Sleeve is Neglected 0.94 x 0 1 0.7000 +/-

0.7018 550.2 25.6 H=25.7 sn_.94_0.0 0.183 0.0009 0.94 x 11 1.0104+/-

1.0123 205.3 130.2 H=25.7 sn_.94_1.0 0.571 0.0010 0.94 x 1.51 1.0343 +/-

1.0365 167.3 182.5 H=25.7 sn_.94_1.5.0 0.654 0.0011 1.1524 +/- Canister contains 631 pellet 1.6xOIO.710 1.1542 132.6 244.9 sn_1.6_0.0

° 0.0009 stacks; H=43.8 0.94 x 1 0.3727 +/- 1341. Void in SNF canister (water is 0.3737 7.3 sn_d_.94_0.0 0.183 0.0005 8 neQlected); H=25.7 Dry Goethite (from Degraded KO Canister and Sleeve) and Water Surround Fuel 0.94 x 01 0.5450 +/-

0.5465 766.3 15.2 H=25.7 sn_9_*94_O.o 0.183 0.0007 0.94 x 11 0.7282 +/- 60.5 0.7301 293.6 H=25.7 sn_9_*94_1.0 0.571 0.0009 130.2 0.94 x 1.51 0.7761 +/- 83.1 0.7779 228.5 H=25.7 sn_9_*94_1.5.0 0.654 0.0009 156.4 0.94 x 2.0 1 0.7877 +/- 105.8 0.7896 196.6 H=25.7 sn_9_*94_2.0 0.709 0.0009 182.5 0.94 x 2.51 0.7760 +/- 128.4 0.7778 177.8 H=25.7 sn_9_*94_2.5.0 0.750 0.0009 234.8

° 1 x 10.275 0.5924 +/-

0.0007 0.5938 637.2 21.8 H=28.5 sn_9_1_O.0 0.8973 +/- 46.7 1.2 x 010.494 0.8993 291.4 H=41.9 sn_9_1.2_O.0 0.0010 98.4 1.0305 +/- 76.2 Canister contains 816 pellet 1.4 x 0 10.625 1.0324 191.9 sn_9_1.4_0.0 0.0010 166.4 stacks; H=43.8 1.0636 +/- 110.1 Canister contains 631 pellet 1.6xOI0.710 1.0656 147.0 sn_9_1.6_O.0 0.0010 244.9 stacks; H=43.8 1.0718+/- 128.8 Canister contains 560 pellet 1.7 x 010.742 1.0737 131.2 sn_9_1.7_O.0 0.0010 288.0 stacks; H=43.8

° 1.8 x 10.768 1.0639 +/-

0.0009 1.0657 121.4 148.6 333.8 Canister contains 505 pellet stacks; H=43.8 sn_9_1.8_O.0 0.7313 +/- Case sn_9_1.6_O.0 but canister 1.6xO/O.710 0.7329 234.4 110.1 sn_9+_1.6_0.0 0.0008 filled with dry Qoethite; H=43.8 Case sn_9_1.6_O.o but canister 1.0132+/- 110.1 1.6xO/O.710 1.0152 154.1 contained carbon steel sleeve; Sn_9cs_1.6_0.0 0.0010 244.9 H=43.8 Cross-sectional Area of the Pellet Array is Reduced for Some of the Cases in the Previous Set

° 1.7 x 10.746 1.0490 +/-

0.0009 1.0509 131.7 128.8 288.0 Canister contains 527 pellet stacks (527/560); H=39.9 sn_9_1.7b_O.o 1.0331 +/- 128.8 Canister contains 508 pellet 1.7 x 010.746 1.0350 132.8 sn_9_1.7c_0.o 0.0010 288.0 stacks (508/560); H=37.9 Canister contains 573 pellet 0.9599 +/- 110.1 stacks (573/631); canister 1.6xO/O.713 0.9619 157.3 sn_9cs_1.6c_O.o 0.0010 244.9 contained carbon steel sleeve; H=37.9

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 57 of74 Table 19. TMI Fuel Pellets in SNF Canister (Degraded Sleeve and KO Canister) (Continued)

Radial Pitch x Axial Gap AEN H/X (cm)1 Void keff+ CF, Ratio a

Fraction keff+/- cr 2cr keV Comment b File Name Canister contains 508 pellet 0.9661 +/- 128.8 stacks (508/560); canister 1.7 x 0 I 0.746 0.9680 141.1 sn_9cs_1.7c_0.o 0.0010 288.0 contained carbon steel sleeve; H=37.9 Canister contains 458 pellet 0.9656 +/- 148.6 stacks (458/505); canister Sn_9cs_1.8c_0.

1.8 x 0 I 0.771 0.9675 130.1 0.0010 333.8 contained carbon steel sleeve; 0 H=37.9 C

Goethite from KO Canister and Sleeve (carbon steel) is Mixed with Water 0.7118 +/-

1 x 0 I 0.275 0.7135 505.9 32.9 H=28.5 sn_9.42_1_0.0 0.0009 0.8979 +/-

1 x 1/0.620 0.8999 217.8 123.3 H=28.5 sn_9.42_1_1.0 0.0010 0.8946 +/-

1 x 1.2/0.653 0.8965 202.1 141.4 H=28.5 sn_9.42_1_1.2.0 0.0009 0.8903 +/-

1 x 1.4/0.680 0.8922 189.5 159.4 H=28.5 sn_9.42_1_1.4.0 0.0009

° 1.2 x I 0.494 0.9066 +/-

0.0010 0.9086 293.6 76.8 H=41.9 sn_9.42_1.2_0.0 0.9683 +/- Canister contains 816 pellet 1.4 x 010.625 0.9702 208.0 128.8 sn_9.42_1.4_0.0 0.0010 stacks; H=43.8

° 1.5 x I 0.671 0.9739 +/-

0.0009 0.9757 183.2 157.7 Canister contains 716 pellet stacks; H=43.8 sn_9.42_1.5_0.0 0.9673 +/- Canister contains 631 pellet 1.6xOIO.710 0.9691 162.6 188.7 sn_9.42_1.6_0.0 0.0009 stacks; H=43.8 0.9308 +/- Canister contains 505 pellet 1.8 x 010.768 0.9323 139.7 256.6 sn_9.42_1.8_0.0 0.0008 stacks; H=43.8 0.9107 +/- Canister contains 631 pellet 1.6xOIO.710 0.9125 176.0 170.2 sn_9*55_1.6_0.0 0.0009 stacks; 44.6% wvf; H=43.8 Cross-sectional Area of the Pellets is Reduced for Some of the Cases in the Previous Set 1.4 x ° I 0.625 0.9704 +/-

0.0010 0.9724 207.3 128.8 Canister contains 802 pellet stacks (802/816); H=41.9 sn_9.42_1.4a_O.

0 0.9643 +/- Canister contains 777 pellet sn_9.42_1.4b_O.

1.4 x 010.626 0.9661 205.3 128.8 0.0009 stacks (777/816); H=39.9 0 0.9597 +/- Canister contains 747 pellet sn_9*42_1.4c_O.

1.4 x 0 / 0.626 0.9614 206.5 128.8 0.0009 stacks (747/816); H=37.9 0 1.4 x 0.21 0.9596 +/- Canister contains 747 pellet sn_9.42_1.4c_.2 0.9614 175.4 164.2 0.684 0.0009 stacks (747/816); H=37.9 .0 1.4xO.41 0.9518 +/- Canister contains 747 pellet sn_9.42_1.4c_.4 0.9535 155.1 199.6 0.726 0.0009 stacks (747/816); H=37.9 .0 0.9603 +/- Canister contains 656 pellet sn_9.42_1.5c_O.

1.5 x 0 / 0.672 0.9623 182.1 157.7 0.0010 stacks (656/716); H=37.9 0 1.5 x 0.21 0.9500 +/- Canister contains 656 pellet sn_9.42_1.5c_.2 0.9517 156.9 198.4 0.722 0.0009 stacks (656/716); H=37.9 .0 1.5 x 0.41 0.9321 +/- Canister contains 656 pellet sn_9.42_1.5c_.4 0.9338 140.2 239.1 0.759 0.0009 stacks (656/716); H=37.9 .0 NOTES: a Second value of H/X (if 9iven) is for the pellet stacks surrounded by water.

b Unless noted otherwise, the SNF canister contains 1083 pellet stacks and contained a stainless steel sleeve; ratio in parentheses is the reduced number to the ori9inal number of pellets stacks; H is the hei9ht (cm) of the pellet array.

C Water volume fraction (wvf) of mixture is 58.3% which just fills the entire SNF canister.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 58 of 74 In the next set of results, dry goethite in the canister is covered by water. Cases investigate varying radial pitch and axial separation. The level of goethite in the canister is determined from the number of fuel pellets submerged in it and the amount produced from degradation. In the first five cases of the set the fuel pellets are touching in the radial direction and are completely submerged in the dry goethite. For these cases an increasing axial gap causes kerr to increase and then reach a maximum value that is much less than the ICL. In the next six cases of the set, an increasing radial pitch places more of the pellets in the water causing kerr to increase. The most reactive of these cases is much greater than critical and occurs at a pitch of 1.7 cm. In the next to the last case of the set, dry goethite replaces all the water in the canister for the case with a pitch of 1.6 cm. This greatly reduces kerr by 0.333, showing that dry goethite is a worse moderator than water though it does have some moderating properties (compare to the last case of the first set of this table). In the last case of the set, the effect of replacing the stainless steel sleeve with a carbon steel sleeve is investigated. This increases the goethite in the canister since carbon steel has a higher iron content than stainless steel. This reduces kerr by 0.050.

In the third set of the table, the cross-sectional area that the fuel array occupies (and thus the number of pellet stacks) is reduced for some of the more reactive cases of the second set. The stainless steel sleeve is also replaced by carbon steel for some of the cases. As seen, this is particularly effective in reducing the system reactivity below the ICL.

In the next to the last set of Table 19, the fuel is surrounded by a mixture of goethite and water.

The amount of goethite is based on the canister having contained a carbon steel sleeve, and the volume fraction of water is chosen to be 58.3%. This is the maximum water volume fraction possible because the corresponding mixture volume when combined with the fuel pellet volume just fills the entire SNF canister. The cases in this set examine different combinations of radial pitch and axial separation. A few cases are found that violate the ICL. In the last case of the set, the wvf is reduced to 45% for one of the more reactive cases, and k eff decreases by 0.057 showing that any decrease in wvf also decreases kerr.

Finally in the last set of Table 19, the number of pellet stacks is reduced for some of the more reactive cases of the previous set. A decreasing number of pellet stacks and hence a decreasing linear loading produces a decreasing keff.

6.2.2 Outer Components of the Waste Package are Degraded (Outside SNF Canister)

This section gives the results of the calculations described in Section 5.5.2. In the configurations studied in this section, the high-level waste canisters and the waste package basket degrade before the inner components of the SNF canister. The results in this section are divided depending upon whether the SNF canister is intact or degraded and are given in Sections 6.2.2.1 and 6.2.2.2, respectively.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 59 of74 6.2.2.1 Intact SNF Canister with Contents Either Intact or Degraded Surrounded by Pre-breach Clay in the Waste Package The description of the cases evaluated in this section is given in Section 5.5.2.1. The results are for an intact SNF canister whose contents are also intact (case wpla_1.3+h.o of Table 14 is used as the base case with keff = 0.9665), and the canister is surrounded by pre-breach clay. The rest of the waste package is water flooded. The results in Table 20 investigate the positioning of the canister in the clay, the effect of water content in the pre-breach clay and the effect of the SNF canister containing clay and/or water. In the first three cases of the first set, the waste package contains dry clay and the location of the canister is positioned just under the surface of the clay, in the center of the clay, and at the bottom of the clay positioned on the waste package bottom. Values of kerr for these cases only differ from the base case value when the canister is just below the surface of the clay, reducing kerr by 0.010. The next case of the set is a comparison with the SNF canister resting on the waste package bottom and the earlier design dimensions used for the waste package. The results are insensitive to this change. In the next four cases, voids in the SNF canister are replaced with clay and/or water. Dry clay between the KG and SNF canisters increases the base case value of kerr by 0.013, but as the water volume fraction of the clay increases kerr decreases until the canister contains only water, where kerr is reduced by 0.028 (relative to the base case). In the last case, dry goethite from the degraded carbon steel sleeve fills the SNF canister, reducing kerr by 0.011. In the second set, the canister is located in the center of the clay, and the volume percent of water in the clay is increased until the waste package is completely filled. For an increasing wvf, kerr decreases demonstrating that dry clay is a more reactive reflector. This occurs because the water 'thermalizes' neutrons away from the canister increasing the probability of capture in the pre-breach clay.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 60 of74 Table 20. TMI Fuel Pellets in Intact SNF and TMI Canisters Surrounded by Pre-breach Clay AEN keff + CF, H/X keff+/- cr 2cr keV Ratio Comment File Name Following Cases are Variations of Case wp1a_1.3+h.o a (R =0.56772 cm) 0.9540 +/-

0.9562 162.2 186.6 Top of SNF canister just below surface of dry clay w_1.3+h_cla.o 0.0011 0.9628 +/- SNF canister centered in dry clay w_1.3+h_clb.o 0.9647 160.9 186.6 0.0010 0.9657 +/- SNF canister at bottom of waste package below dry w_1.3+h_c1c.o 0.9675 159.8 186.6 0.0009 c1av 0.9645 +/- Previous case but waste package uses earlier design w_1.3+h_clcwp.o 0.9665 160.1 186.6 0.0010 dimensions 0.9778 +/- Case w_1.3+h_clb.o, but SNF canister contains dry w_1.3+h_c1b+c.o 0.9799 158.5 186.6 0.0010 clav 0.9709 +/- Case w_1.3+h_clb.o, but SNF canister contains clay w_1.3+h_clb+c.1.

0.9729 158.8 186.6 0.0010 with a 10% water volume fraction 0 0.9655 +/- Case w_1.3+h_clb.o, but SNF canister contains clay w_1.3+h_clb+c.2.

0.9674 159.7 186.6 0.0010 with a 20% water volume fraction 0 0.9367 +/- Case w_1.3+h_clb.o, but SNF canister contains 0.9386 164.0 186.6 w_1.3+h_clb+w.o 0.0009 water 0.9535 +/- Case w_1.3+h_clb.o, but SNF canister contains dry 0.9554 161.9 186.6 w_1.3+h_c1b+g.o 0.0010 Qoethite from deQraded carbon steel sleeve Following Cases are Variations of Case w_1.3+h_clb.o a, b 0.9431 +/-

0.9451 163.4 186.6 Clay contains 20% water by volume w_1.3+h_clb.2.o 0.0010 0.9277 +/-

0.9296 166.5 186.6 Clay contains 40% water by volume w_1.3+h_clb.4.o 0.0009 0.9214 +/- Clay contains 55% water by volume (waste package 0.9235 169.1 186.6 w_1.3+h_clb.55.o 0.0011 is full)

NOTES: aThe radial pitch is 1.3 cm, the dimensionless axial pitch is 1.5 and the void fraction, Vr, is 0.790.

b The SNF canister is centered in the clay mixture.

For the results in Table 21 the KO canister and sleeve have degraded leaving degradation products and fuel in the intact SNF canister. The outer waste package components have degraded to pre-breach clay leaving the SNF canister centered in dry pre-breach clay. These results closely parallel those in Table 19 of Section 6.2.1. Unless noted otherwise, there are 1083 pellet stacks in the canister, the sleeve was stainless steel and the SNF canister is flooded with water.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 61 of74 Table 21. TMI Fuel Pellets in Intact SNF Canister Surrounded by Dry Pre-breach Clay (Degraded TMI Canister and Sleeve)

Radial Pitch x Axial Gap AEN H/X (cm) IVoid kelt + CF, Ratio a b Fraction kelt+/- 0 20 keV Comment File Name Water in SNF Canister; Goethite (from Dearadation of TMI Canister and Sleeve) is nealected 0.94 x 0 1 0.7152 +/- 0.717 526.4 25.6 H=25.7 sc_.94_0.o 0.183 0.0009 1 0.94 x 11 1.0142 +/- 1.016 204.8 130.2 H=25.7 sc_.94_1.o 0.571 0.0010 3 0.94 x 1.51 1.0365 +/- 1.038 165.9 182.5 H=25.7 sc_.94_1.5.o 0.654 0.0010 5 1.1568 +/- 1.158 Canister contains 631 pellet stacks; 1.6xOI0.710 130.7 244.9 sc_1.6_0.o 0.0011 9 H=43.8 0.94 x 0 1 0.4290 +/- 0.430 1060. Void in SNF canister (water is 7.3 sC_d_.94_0.o 0.183 0.0006 2 5 neglected); H=25.7 Dry Goethite (from Degraded KO Canister and Sleeve) and Water Surround Fuel 0.94 x 0 1 0.5682 +/- 0.569 717.0 15.2 H=25.7 SC_9_*94_O.o 0.183 0.0008 7 0.94 x 1 1 0.7373 +/- 0.739 60.5 287.8 H=25.7 SC_9_*94_ 1.o 0.571 0.0009 1 130.2 0.94 x 1.51 0.7826 +/- 0.784 83.1 SC_9_*94_ 1.5.

221.9 H=25.7 0.654 0.0009 4 156.4 0 0.94 x 2.0 1 0.7895 +/- 0.791 105.8 192.6 H=25.7 SC_9_*94_2.o 0.709 0.0008 2 182.5 0.94 x 2.51 0.7789 +/- 0.780 128.4 SC_9_*94_2.5.

175.8 H=25.7 0.750 0.0009 7 234.8 0 0.6126 +/- 0.614 1 x 0 10.275 602.6 21.8 H=28.5 SC_9_1_O.o 0.0008 2 0.9200 +/- 0.921 46.7 1.2 x 0 10.494 280.8 H=41.9 SC_9_1.2_O.o 0.0010 9 98.4 1.0416 +/- 1.043 76.2 Canister contains 816 pellet stacks; 1.4 x 0 10.625 188.0 SC_9_1.4_O.o 0.0010 5 166.4 H=43.8 1.0778 +/- 1.079 110.1 Canister contains 631 pellet stacks; 1.6xOI0.710 143.8 SC_9_1.6_O.o 0.0010 7 244.9 H=43.8 1.0749 +/- 1.076 110.1 Previous case but earlier waste SC_9_1.6_Owp.

1.6xOI0.710 144.2 0.0009 7 244.9 packaoe desion dimensions are used 0 1.0776 +/- 1.079 128.8 Canister contains 560 pellet stacks; 1.7xOI0.742 129.6 SC_9_1.7_O.o 0.0009 4 288.0 H=43.8 1.0692 +/- 1.071 148.6 Canister contains 505 pellet stacks; 1.8xOI0.768 119.3 SC_9_1.8_O.o 0.0009 0 333.8 H=43.8 0.7422 +/- 0.743 Case SC_9_1.6_O.o but canister filled 1.6xOI0.710 226.0 110.1 sc_9+_1.6_0.o 0.0008 8 with dry goethite; H=43.8 Case SC_9_1.6_O.o but canister 1.0248 +/- 1.026 110.1 sc_9cs_1.6_0.

1.6xOI0.710 151.4 contained carbon steel sleeve; 0.0009 6 244.9 0 H=43.8 Cross-sectional Area of the Pellet Array is Reduced for Some of the Cases in the Previous Set 1.0530 +/- 1.054 128.8 Canister contains 527 pellet stacks 1.7xOI0.746 130.3 SC_9_1.7b_0.o 0.0010 9 288.0 (527/560); H=39.9 1.0366 +/- 1.038 128.8 Canister contains 508 pellet stacks 1.7xOI0.746 131.8 SC_9_1.7c_0.o 0.0009 4 288.0 (508/560); H=37.9

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 62 of74 Table 21. TMI Fuel Pellets in Intact SNF Canister Surrounded by Dry Pre-breach Clay (Degraded TMI Canister and Sleeve) (Continued)

Radial Pitch x Axial Gap AEN H/X (cm) I Void kef! + CF, Ratio a b Fraction kef!+/- 0 20 keV Comment File Name Canister contains 573 pellet stacks 0.9672 +/- 0.969 110.1 sc_gcs_1.6c_0 1.6xOI0.713 156.4 (573/631) and canister contained 0.0010 2 244.9 .0 carbon steel sleeve; H=37.9 Canister contains 508 pellet stacks 0.9717 +/- 0.973 128.8 sc_gcs_1.7c_0 1.7 x 0 I 0.746 140.6 (508/560) and canister contained 0.0009 6 288.0 .0 carbon steel sleeve; H=37.9 Canister contains 458 pellet stacks 0.9709 +/- 0.972 148.6 sc_gcs_1.8c_0 1.8 x 0 I 0.771 128.4 (458/505) and canister contained 0.0009 7 333.8 .0 carbon steel sleeve; H=37.9 NOTES: a Second value of H/X (if given) is for the pellet stacks surrounded by water.

b Unless noted otherwise, the SNF canister contains 1083 pellet stacks and contained a stainless steel sleeve; ratio in parentheses is the reduced number to the original number of pellets stacks; H is the height (cm) of the pellet array, see Section 6.2.1.

In the first set, the goethite from degradation of the KO canister and sleeve is neglected and the pellets are surrounded by water. The SNF canister details for this set are identical to those of the first set of Table 19, and the results vary by only a few cr for the more thermalized cases (lower values of AENCF). For the least thermalized case (dry SNF canister), kerr is 0.057 larger for the canister reflected by dry pre-breach clay.

Again the SNF canister details in the second and third sets are the same as those used in Table 19, i.e., dry goethite covered by water is in the SNF canister, and a variety of pellet array spacings and configurations are examined. The results are roughly 0.005 larger than those of Table 19. A case is also given in the second set of the table comparing the earlier waste package design, showing that the results are insensitive to this design change.

The results in the next table, Table 22, are a continuation of the same configurations as in the previous table, Table 21, except the goethite from the degradation of the KO canister and sleeve is mixed with sufficient water (58.3% wvf) to completely fill the rest of the SNF canister (unless noted otherwise). These results parallel those of the last two sets of Table 19, except the waste package internals are degraded rather than being intact. As indicated by the results in the previous table, the dry pre-breach clay is seen to be a better reflector than the intact waste package internals with the most reactive case of this set being 0.01 larger than the equivalent case of Table 19.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 63 of74 Table 22. TMI Fuel in Intact SNF Canister Containing Goethite Mixed with Water Surrounded by Dry Pre-breach Clay Radial Pitch x Axial Gap AEN (cm) 1 Void keff + CF, H/X Fraction keff+/- cr 2cr keV Ratio Comment a File Name Goethite from KO Canister and Sleeve (Carbon Steel) is Mixed with Water b 1 x 0 I 0.275 0.7259 +/- 0.727 485.9 32.9 SC_9A2_1_0.0 H=28.5 0.0009 6 1 x 0.81 0.580 0.8978 +/- 0.899 236.0 105.2 H=28.5 0.0009 6 sc QA2 1 .8.0 1 x 1/0.620 0.9037 +/- 0.905 216.1 123.3 H=28.5 0.0010 6 sc QA2 1 1.0 1 x 1.2 10.653 0.9002 +/- 0.902 200.1 14104 SC_9A2_1_1.2.

H=28.5 0.0009 0 0 1 x 104/0.680 0.8946 +/- 0.896 187.3 15904 SC_9A2_1_1A.

H=28.5 0.0009 5 0 1 x 1.6/0.704 0.8851 +/- 0.886 '177.8 177.5 SC_9A2_1_1.6.

H=28.5 0.0009 9 0 1 x 1.8/0.724 0.8693 +/- 0.871 171.5 195.6 SC_9A2_1_1.8.

H=28.5 0.0009 0 0 1.2 x 0/00494 0.9209 +/- 0.922 284.5 76.8 SC_9A2_1.2_0.

H=41.9 0.0009 8 0 104 x 0 I 0.625 0.9800 +/- 0.981 203.1 128.8 Canister contains 816 pellet SC_9A2_1A_0.

0.0010 9 stacks; H=43.8 0 1.5 x 0 I 0.671 0.9822 +/- 0.983 179.4 157.7 Canister contains 716 pellet SC_9A2_1.5_0.

0.0009 9 stacks; H=43.8 0 1.6xO/0.710 0.9734 +/- 0.975 161.4 188.7 Canister contains 631 pellet sc_g.42_1.6_0.

0.0008 0 stacks; H=43.8 0 1.8 x 0 I 0.768 0.9356 +/- 0.937 138.0 256.6 Canister contains 505 pellet SC_9A2_1.8_0.

0.0009 3 stacks; H=43.8 0 1.6xO/0.710 0.9174 +/- 0.919 174.2 170.2 Canister contains 631 pellet SC_9*55_1.6_0.

0.0009 2 stacks; 44.6% water volume 0 fraction; H=43.8 Cross-sectional Area of Pellet Array is Reduced for Most Reactive Cases of Previous Set 1.4 x 0 I 0.625 0.9779 +/- 0.979 202.2 128.8 Canister contains 802 pellet SC_9A2_1Aa_0 0.0009 7 stacks (802/816); H=41.9 .0 104 x 0 I 0.626 0.9703 +/- 0.972 202.6 128.8 Canister contains 777 pellet SC_9A2_1Ab_0 0.0010 2 stacks (777/816); H=39.9 .0 1.4 x 0 I 0.626 0.9629 +/- 0.964 202.7 128.8 Canister contains 747 pellet SC_9A2_1Ac_0 0.0009 8 stacks (747/816); H=37.9 .0 104 x 0.21 0.9615 +/- 0.963 172.8 164.2 Canister contains 747 pellet SC_9A2_1Ac_.

0.684 0.0009 3 stacks (747/816); H=37.9 2.0 1AxOAI 0.9551 +/- 0.956 153.9 199.6 Canister contains 747 pellet sC_9 A2_1Ac_.

0.726 0.0009 9 stacks (747/816); H=37.9 4.0 1.5 x 0 I 0.672 0.9647 +/- 0.966 179.0 157.7 Canister contains 656 pellet sc_9*42_1.5c_0 0.0009 4 stacks (656/716); H=37.9 .0 1.5 x 0.21 0.9565 +/- 0.958 155.5 198.4 Canister contains 656 pellet SC_9A2_1.5c_.

0.722 0.0010 4 stacks (656/716); H=37.9 2.0 1.5 x 0041 0.9365 +/- 0.938 13904 239.1 Canister contains 656 pellet SC_9A2_1.5c_.

0.759 0.0008 2 stacks (656/716); H=37.9 4.0 1.6 x 0 I 0.684 0.9654 +/- 0.967 161.3 188.7 Canister contains 597 pellet SC_9A2_1.6b_

0.0009 1 stacks (597/613); H=39.9 0.0 NOTES: a The SNF canister contained a carbon steel sleeve and, unless noted otherwise, 1083 pellet stacks; ratio in parentheses is the reduced number to the original number of pellets stacks; H is the height (cm) of the pellet array, see Section 6.2.1.

bWater volume fraction (wvf) of mixture is 58.3% which just fills the entire SNF canister.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 64 of74 In the next set, the number of pellet stacks is reduced so that only a portion of the cross-sectional area of the SNF canister is filled for those cases in the previous set that exceed the ICL. For a reduction in the number of stacks, the number of pellets per stack is appropriately increased to maintain an assembly's worth of pellets per canister. The results show that an 8.4% reduction in the number of pellet stacks is sufficient to meet the ICL for the most reactive case of the first set.

Additional cases are included for this configuration to show that increasing axial gap further decreases kerr.

6.2.2.2 TMI Fuel Surrounded by Degradation Products with (Non-reacted) Pre-breach Clay In this section the SNF canister and sleeve have degraded to goethite that has mixed, but not chemically reacted, with the pre-breach clay formed from the degradation of the outer components of the waste package. These cases are described in Section 5.5.2.2, and the results are presented in Sections 6.2.2.2.1 (intact TMI canister) and 6.2.2.2.2 (degraded TMI canister).

6.2.2.2.1 Intact KO Canister Surrounded by Degradation Products from the SNF Canister and Sleeve Mixed with Pre-breach Clay The results in this section are for the KO canister and its contents intact and the rest of the components in the waste package degraded as described in Section 5.5.2.2.1. The cases considered are based on some of the more reactive cases of Table 14 with the description of the KO canister and its contents being identical. Unless noted otherwise, the canister is centered in dry pre-breach clay. The results for these cases are given in Table 23. In the first set of the table, the goethite from the degraded SNF canister and sleeve is neglected. The first two cases demonstrate that a canister containing one assembly's worth of fuel is no more reactive than a canister completely filled with fuel. Values ofkefrfor these cases are 0.014 larger than the base case value (kerr = 0.9660), showing that dry pre-breach clay in the waste package is more reactive than intact internals. In the last two cases of the set the number of pellet stacks is reduced filling a smaller portion of the canister cross-section (as indicated in the table). The ICL is satisfied for even a 10% reduction in the number of stacks.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 65 of74 Table 23. TMI Fuel in Intact KG Canister with Degraded SNF Canister Surrounded by Pre-breach Clay AEN keff + CF, H/X keff+/ 20- keV Ratio Comment File Name Variations of Case wp1 ae_1.2h.o a (R =0.4699 em) of Table 14 and Goethite Neglected 0.9780 +/- 0.0010 0.9800 135.3 231.7 Spherical pellets fill canister wae-s_c1b.o 0.9747 +/- 0.0009 0.9765 137.7 231.7 Canister contains one assembly's worth wae-of fuel pellets s 1.2has clb.o 0.9450 +/- 0.0010 0.9470 137.1 231.7 Previous case, but pellet array fills 90% of wae-sas_c1b_.9.o canister cross-section 0.9135 +/- 0.0009 0.9154 139.7 231.7 Previous case, but pellet array fills 80% of wae-sas_c1b_.8.o canister cross-section

=

Following Cases are Variations of Case wp1a_1.3+h.o b (R 0.56772 em) of Table 14 0.9767 +/- 0.0010 0.9788 158.0 186.6 Canister centered in clay; 90ethite w-s_c1b.o neqlected 0.9782 +/- 0.0010 0.9802 158.1 186.6 Previous case, but canister at bottom of w-s_c1c.o WP 0.9519 +/- 0.0010 0.9539 161.6 186.6 Case w-s_c1b.o, but 20% (by volume) w-s_c1b.2.o water in clay 0.9381 +/- 0.0010 0.9401 163.1 186.6 Case w-s_clb.o, but 40% (by volume) w-s_clb.4.o water in clay 0.9499 +/- 0.0010 0.9520 161.8 186.6 Canister surrounded by dry 90ethite W-S_9_clb.o centered in dry clav 0.9405 +/- 0.0010 0.9425 162.4 186.6 Previous case, but 20% (by volume) W-S_9*8_c1b.o water in qoethite 0.9311 +/- 0.0011 0.9332 164.2 186.6 Previous case, but 40% (by volume) W-S_9*6_clb.o water in qoethite 0.9459 +/- 0.0010 0.9478 163.4 186.6 Case w-S_9_clb.o, but SNF canister W-S_9CS_clb.o contained a carbon steel sleeve 0.9582 +/- 0.0009 0.9600 162.4 178.5 Case w-s_clb.o but 5% (by volume) clay w-s_c1b+c.05i.o mixed with water in TMI canister 0.9491 +/- 0.0009 0.9509 165.7 174.5 Case w-s_clb.o but 7.5% (by volume) clay w-s_c1b+c.075i.o mixed with water in TMI canister NOTES: a Void fraction is 0.704.

b Void fraction is 0.657.

In the first four cases of the second set, the goethite is again neglected and the positioning of the canister and the water content of the pre-breach clay are varied. The addition of water to the clay decreases ketr, but the reactivity is unchanged for the canister at the center or bottom of the dry clay.

For the rest of the cases of the set the canister is centered in dry clay since this is most reactive. In the next four cases of the set, the goethite mixed with varying amounts of water (as indicated in the table) forms an annulus around the canister. The amount of goethite is based on a stainless steel sleeve, though one case considers an increased amount of goethite based on a carbon steel sleeve.

Results for these cases show that an increase in either goethite or water in the goethite decreases ketr. In the last two cases of the set, the goethite is neglected, the canister is centered in the dry pre-breach clay layer (shown to be the most reactive configuration) and a small amount of pre-breach clay is homogeneously mixed with the water inside the KG canister. The results for these cases show that an increase in the amount of clay inside the canister causes a decrease in ketr.

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 66 of74 6.2.2.2.2 Degraded SNF Canister, TMI Canister and Sleeve Mixed with Pre-breach Clay Surround TMI Fuel The results in this section are for cases described in Section 5.5.2.2.2 where pre-breach clay and goethite from the degraded SNF canister, sleeve (carbon steel) and TMI canister surround the TMI fuel pellets at the bottom of the waste package. These materials are assumed to form separate layers that may also contain water and/or fuel. In general, some of the pellets are in each of the different layers depending on the volume of each material, the order in which the materials are layered and the height of the pellet stack. In determining the height of each layer the displacement of pellets is considered. Pellet stacks of cylindrical pellets with no axial gap between pellets are modeled for these cases. The composition of each layer is given in the table as well as the volume fraction of water in the materials. These results are given in Table 24. The values of H/X listed in the table are an average of the values for square and triangular lattices and are also weighted by the number of pellet stacks in each layer (for those cases where the stacks are in different layers).

Table 24. TMI Fuel Pellets Form Array Surrounded by Layers of Goethite and Pre-breach Clay in the Waste Package Bottom Layer Next Layer Radial Pitch (em) a/

Contents by Contents by Comment / Void AENCF HIX Volume ('Yo) Volume ('Yo) Fraction b kelt +/- <r kelt + 2<r (keV) Ratio File Name The Following Cases Have 1083 Pellet Stacks (Unless Noted Otherwise) 21.8 Goethite 100 Clay 100 1.0 0.5673 +/- 0.0007 0.5687 692.3 wp_1a_g_c.o 10.6 1 / Previous case but water replaces 40.8 Water 100 Clay 100 0.7119 +/- 0.0009 0.7138 512.9 wp_1a_w_c.o goethite (goethite 10.6 neglected)

Goethite 40 77.7 Clay 100 1.2 0.8759 +/- 0.0009 0.8777 30804 wp_1.2a_904_c.o water 60 16.3 1.2 / Previous case Goethite 40 77.7 Clay 100 but different pellet 0.9039 +/- 0.0009 0.9057 290.2 wp_1.2b_go4_c.o water 60 16.3 stack configuration Goethite 40 130.3 Clay 100 1.4 0.9459 +/- 0.0009 0.9476 22004 wp_104a_go4_c.o water 60 23.1 104 / Previous case Goethite 40 130.3 Clay 100 but different pellet 0.9707 +/- 0.0010 0.9726 205.7 wp_104b_go4_c.o water 60 23.1 stack configuration 104/ Same Goethite 40 Clay 80 130.3 configuration as case 0.9720 +/- 0.0010 0.9739 205.2 wp_104b_go4_c.8.o water 60 water 20 51.7 wp 104b _go4 c.O 104 / Similar Goethite 40 configuration as case 130.3 Clay 100 0.9723 +/- 0.0009 0.9742 197.9 wp_1.4bL_go4_c.o water 60 wp_104b_9o4_c.o but 23.1 onlv 946 stacks Goethite 40 159.6 Clay 100 1.5 0.9584 +/- 0.0009 0.9602 194.2 wp_1.5a_9*4_c.o water 60 26.8 1.5/ Previous case Goethite 40 159.6 Clay 100 but different pellet 0.9783 +/- 0.0009 0.9802 180.0 wp_1.5b_g.4_c.o water 60 26.8 stack configuration 1.5/ Same Goethite 40 Clay 80 159.6 configuration as case 0.9768 +/- 0.0009 0.9786 180.6 wp_1.5b_9 04_c.8.o water 60 water 20 62.3 wp 1.5bg.4 c.O

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 67 of74 Table 24. TMI Fuel Pellets Form Array Surrounded by Layers of Goethite and Pre-breach Clay in the Waste Pckage (Continued)

Bottom Layer Next Layer Radial Pitch (cm) a/

Contents by Contents by Comment / Void AENCF H/X Volume (%) Volume (%) Fraction b k eff +/- 0" keff + 20" (keV) Ratio File Name 1.5/ Same Goethite 40 Clay 60 159.6 configuration as case 0.9798 +/- 0.0009 0.9816 176.5 wp_1.5b_g.4_c.6.0 water 60 water 40 97.8 wp 1.5b g.4 C.o 1.5/ Same Goethite 40 Clay 47 159.6 configuration as case 0.9813 +/- 0.0010 0.9833 176.7 wp_1.5b_g.4_c.47.0 water 60 water 53 120.7 wp 1.5b g.4 C.o c Previous case but Goethite 40 Clay 47 159.6 earlier WP design 0.9817 +/- 0.0008 0.9834 175.9 wp_1.5b_g.4_c.47p.o water 60 water 53 120.7 dimensions used 1.5/ Same Goethite 60 Clay 46 137.3 configuration as case 0.9136 +/- 0.0009 0.9155 195.3 wp_1.5b_g.6_c.46.0 water 40 water 54 122.8 WP 1.5b 0.4 C.o c 1.5/ Same Goethite 80 Clay 45 114.9 configuration as case 0.8787 +/- 0.0009 0.8805 206.3 wp_1.5b_g.8_c.45.0 water 20 water 55 123.8 wp 1.5b _g.4 C.o c 1.5/ Same Clay 45 92.6 Goethite 100 configuration as case 0.8696 +/- 0.0009 0.8714 212.5 Wp_1.5b_g_c.45.0 water 55 124.3 wp 1.5bg.4 C.o c 1.5/ Same 92.6 Goethite 100 Clay 100 configuration as case 0.6713 +/- 0.0008 0.6729 301.9 wp_1.5b_g_c.o 26.8 wp 1.5b g.4 C.o 1.5/ Similar Goethite 40 configuration as case 159.6 Clay 100 0.9738 +/- 0.0009 0.9756 174.6 wp_1.5bL_g.4_c.o water 60 wp_1.5b_g.4_c.o but 26.8 onIv 954 stacks Goethite 40 191.0 Clay 100 1.6 0.9578 +/- 0.0009 0.9596 174.9 wp_1.6a_g.4_c.o water 60 30.9 1.6/ Same Goethite 40 Clay 80 191.0 configuration as case 0.9665 +/- 0.0008 0.9682 172.3 wp_1.6a_g.4_c.8.0 water 60 water 20 73.7 WP 1.6a 0.4 C.o 1.6/ Similar Goethite 40 configuration as case 191.0 Clay 100 0.9615 +/- 0.0009 0.9634 161.3 wp_1.6aL_g.4_c.o water 60 wp_1.6a_g.4_c.o but 30.9

- - .. _- ~ - ~ -- 0r'ly_729~tacks_ - - - -- - --

1.6/ Case

~ ~

Goethite 40 191.0 Clay 100 wp_1.6a_g.4_c.o but 0.9688 +/- 0.0009 0.9706 163.9 wp_1.6b_g.4_c.o water 60 30.9 different configuration 1.6/ Same Goethite 40 Clay 47 191.0 configuration as case 0.9738 +/- 0.0009 0.9755 158.3 wp_1.6b_g.4_c.47.0 water 60 water 53 144.0 wp 1.6b g.4 C.o Goethite 40 259.7 Clay 100 1.8 0.9316 +/- 0.0009 0.9333 149.8 wp_1.8a_g.4_c.o water 60 39.7 1.8 / Previous case Goethite 40 259.7 Clay 100 but different pellet 0.9338 +/- 0.0008 0.9354 141.5 wp_1.8b_g.4_c.o water 60 39.7 stack configuration Variations of Cases in the Previous Set but with Number of Pellet Stacks Reduced (less than 1083) 1.5/ Case Goethite 40 Clay 47 wp_1.5b_g.4_c.47.0 159.6 0.9728 +/- 0.0009 0.9747 175.6 wp_1.5bb_g.4_c.47.0 water 60 water 53 but 998 pellet stacks 120.7 (998/1083)

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 68 of74 Table 24. TMI Fuel Pellets Form Array Surrounded by Layers of Goethite and Pre-breach Clay in the Waste Pckage (Continued)

Bottom Layer Next Layer Radial Pitch (cm) a/

Contents by Contents by Comment / Void AENCF HIX Volume (%) Volume (%) Fraction b keff +/- 0" keff + 20" (keV) Ratio File Name 1.5/ Case Goethite 40 Clay 47 wp_1.5b_g.4_c.47.0 159.6 0.9615 +/- 0.0009 0.9632 175.1 wp_1.5bc_g.4_c.47.0 water 60 water 53 but 932 pellet stacks 120.7 (932/1083) 1.6/ Case Goethite 40 Clay 47 wp_1.6b_g.4_c.47.0 191.0 0.9729 +/- 0.0008 0.9746 158.7 wp_1.6ba_9.4_c.47.0 water 60 water 53 but 1065 pellet 144.0 stacks (1065/1083) 1.6/ Case Goethite 40 Clay 47 wp_1.6b_9*4_c.47.0 191.0 0.9651 +/- 0.0009 0.9670 157.8 wp_1.6bb_g.4_c.47.0 water 60 water 53 but 998 pellet stacks 144.0 (99811083)

NOTES: a Cylindrical pellets with no axial gap; ratio in parentheses is the reduced number to the original number of pellets stacks.

b Void fractions are equal to the simple average of the void fractions for a square and hexagonal unit cell; values are: 0.253, 0.481,0.619,0.668,0.708 and 0.769 for pitches of 1.0 cm, 1.2 cm, 1.4 cm, 1.5 cm, 1.6 cm and 1.8 cm, respectively.

C The waste package is completely filled by fuel and layers of degraded materials.

In the first set of results in the table, there are 1083 pellet stacks, unless noted otherwise, this being the maximum number of axially aligned stacks that fit in a KO canister. The results are arranged in order of increasing radial pitch. For each value of pitch the water volume fractions of the layers are varied and/or different pellet configurations are investigated. The bottom layer is composed of goethite and/or water, and this layer is covered by a layer of pre-breach clay and water. Any remaining space in the waste package is water flooded. A case in this set using the earlier waste package design dimensions shows that the results are insensitive to this change.

Examination of the results in this set shows that the reactivity for each case is dependent on the pellet array configuration, pitch and water content of the material. The configurations can be categorized according to the surface to volume (SN) ratios of the entire pellet array with a more compacted array having a smaller SN ratio than an array where the pins are spread out. For all else being equal (fuel, pitch and composition of material), the larger SN ratios promote more neutron leakage and a decreased kef[. Smaller SN ratio configurations would appear to be less probable since any disturbance of the waste package would tend to spread out the pins increasing the SN ratio, increasing leakage and decreasing kerr. For increasing pitch, the results show that kerr initially increases, reaches a maximum value and then decreases. An increasing water content in the degradation materials has the greatest effect when applied to the bottom layer where a significant amount of the fuel is found. This is further illustrated by those cases where decreasing the number of pellet stacks either has no effect on or slightly increases the reactivity. This illustrates the interplay between the pitch, fuel pellet configuration and water content of the materials.

In the second set of the table, the number of pellet stacks for some of the more reactive cases of the first set is sufficiently reduced so that the ICL is satisfied.

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 69 of74 6.2.3 TMI Fuel Surrounded by Post-breach Clay in the Waste Package This section presents the results of the calculations described in Section 5.5.3. Table 25 gives the values of kerr for these calculations where the waste package containing post-breach clay (composition given in Table 10 at 72689 years) surrounds the TMI fuel pellets. For most of the cases in the table, the post-breach clay is mixed with sufficient water to completely fill the waste package. The radial pitch of the pellet stack is varied, and the largest value of keff occurs at a pitch of 1.6 cm. For this most reactive case, the volume fraction of water is decreased until the post-breach clay is dry. Decreasing the amount of water in the clay is seen to decrease kelT. Other cases investigate a different pellet configuration and reducing the number of pellet stacks for this most reactive case, these reduce kelT below the ICL. A case is also included showing that the earlier waste package design dimensions produces statistically identical results.

Table 25. TMI Fuel Pellets are Surrounded with the Post-breach Clay (Clay Composition is for 72689 years after emplacement, see Table 10)

Content of Fuel Layer by AENCF H/X Radial Pitch (cm) I Commene I Volume (%) keff+/- U keff + 2u (keV) Ratio Void Fraction b File Name Clay 43.4 0.6812 +/- 0.0009 0.6829 537.5 27.7 1C psa.43_1.0 Water 56.6 0.8473 +/- 0.0010 c Clay 43.4 0.8492 322.2 62.8 1.2 psa.43_1.2.0 Water 56.6 Clay 43.4 0.9411 +/- 0.0010 0.9432 228.1 104.2 1.4 c psa.43_1.4.0 Water 56.6 c

Clay 43.4 0.9639 +/- 0.0009 0.9656 196.8 127.3 1.5 psa.43_1.5.0 Water 56.6 Clay 43.4 0.9759 +/- 0.0010 0.9778 176.1 152.0 1.6 c psa.43_1.6.0 Water-- 56.6

-173:5-- 152.0 Clay 43.4 0.9787 +/- 0.0010 0.9807 Previous case but earlier waste psa.43_1.6wp.o Water 56.6 I package desian dimensions are used Clay 43.4 0.9461 +/- 0.0009 0.9479 176.2 152.0 1.6 c I Previous case but different psa.43_1.6b.o Water 56.6 pellet stack configuration Clay 43.4 0.9654 +/- 0.0009 0.9673 177.8 152.0 1.6 C I case psa.43_1.6.0 but 998 psa.43_1.6ab.o Water 56.6 pellet stacks (998/1083)

Clay 60 0.8774 +/- 0.0009 0.8791 207.3 115.9 1.6 psa.6_1.6.0 Water 40 Clay 80 0.7305 +/- 0.0009 0.7322 270.1 I 72.9 1.6 psa.8_1.6.0 Water 20 I Clay 100 0.5182 +/- 0.0007 0.5195 419.2  : 30.0 1.6 psa_1.6.0

-_... . .... __. . -f - - - - -

1,ic Clay 43.4 0.9399 +/- 0.0009 '0.9417' 161.2 : 178.3 psa:43_U.o' Water 56.6 Clay 43.4 0.9659 +/- 0.0009 0.9677 147.7 206.1 1.8 c psa.43_1.8.0 Water 56.6 NOTES: a The waste package contains cylindrical pellets with no axial gap in 1083 pellet stacks, unless noted otherwise; ratio in parentheses is the reduced number to the original number of pellets stacks.

b Void fractions are equal to the simple average of the void fractions for a square and hexagonal unit cell; values are: 0.253, 0.481, 0.619, 0.668, 0.708, 0.741 and 0.769 for pitches of 1 em, 1.2 em, 1.4 em, 1.5 em, 1.6 em, 1.7 em and 1.8 em, respectively.

cThe waste package is completely filled by fuel and post-breach clay mixture.

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 70 of74 For the results in Table 26, an assembly's worth of TMI pellets are assumed to be completely degraded and homogeneously mixed with the post-breach clay compositions given in Tables 10 and

11. The first three cases of Table 26 are for each of the compositions given in these tables. For these cases the clay layers are dry though water fills the rest of the waste package. Clearly, there are no criticality concerns for any of these cases. The water volume fraction is increased in the next three cases, showing a slightly peak in reactivity for a 20% wvf, though keff is so small that it is of no concern. In the last case, the U-238 content of the TMI fuel is neglected for the 20% wvf case.

This increases keff though its value is still too small to be of any concern.

Table 26. Degraded TMI Fuel Mixed with Post-breach Clay Content of Fuel Layer by AENCF H/X Volume ('Yo) keff+/- cr keff + 2cr (keV) Ratio Comment a File Name Clay 100 0.1200 +/- 0.0001 0.1202 73.4 997.5 Post-breach clay at 72689 years, see psa.o Table 10 Clay 100 0.0890 +/- 0.0001 0.0892 81.8 1020.5 Post-breach clay at 378240 years, psb.o see Table 10 Clay 100 0.1191 +/- 0.0001 0.1193 73.9 997.4 Post-breach clay at 74818 years, see pSC.o Table 11 Clay 80 0.1264 +/- 0.0001 0.1266 45.7 3610.9 Post-breach clay at 72689 years, see psa.8.o Water 20 Table 10 Clay 60 0.1167 +/- 0.0001 0.1169 33.4 7966.8 Post-breach clay at 72689 years, see psa.6.o Water 40 Table 10 Clay 43.4 0.1010 +/- 0.0001 0.1011 24.9 14728.1 Post-breach clay at 72689 years, see psa.43.o Water 56.6 Table 10; waste package complete filled Clay 80 0.1321 +/- 0.0001 0.1323 5.6 3610.9 Case psa.8.o, but U-238 content of psa.8-U.o Water 20 fuel is neglected 6.3

SUMMARY

The results throughout Sections 6.1 and 6.2 present a large number of parametric evaluations for intact configurations and a wide range of degraded configurations for the codisposal ofTMI-2 SNF.

All outputs are reasonable compared to the inputs and the results of this calculation are suitable for their intended use.

Engineered Systems Project Calculation

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Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 71 of74

7. REFERENCES 7.1 DOCUMENTS CITED Audi, G. and Wapstra, A.H. 1995, Atomic Mass Adjustment, Mass Listfor Analysis. Upton, New York: Brookhaven National Laboratory, National Nuclear Data Center. TIC: 242718.

Beyer, W.H., ed. 1987. CRC Standard Mathematical Tables. 28th Edition. 3rd Printing 1988.

Boca Raton, Florida: CRC Press. TIC: 240507 BSC (Bechtel SAIC Company) 2001a. EQ6 Calculationfor Chemical Degradation ofFort Saint Vrain (Th/U Carbide) Waste Packages. CAL-EDC-MD-000011 REV 00. Las Vegas, Nevada:

Bechtel SAIC Company. ACC: MOL.20010831.0300.

CRWMS M&O 1998a. Software Qualification Reportfor MCNP Version 4B2, A General Monte Carlo N-Particle Transport Code. CSCI: 30033 V4B2LV. DI: 30033-2003, Rev. 01. Las Vegas, Nevada: CRWMS M&O. ACC: MOL. 19980622.0637.

CRWMS M&O 1998b. Software Code: MCNP. V 4B2LV. HP, HPUX 9.07 and 10.20; PC, Windows 95; Sun, Solaris 2.6.30033 V4B2LV.

CRWMS M&O 1999a. DOE SRS HLW Glass Chemical Composition. BBAOOOOOO-01717-0210-00038 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19990215.0397.

CRWMS M&O 1999b. Generic Degradation Scenario and Configuration Analysis for DOE Codisposal Waste Package. BBAOOOOOO-01717-0200-00071 REV 00. Las Vegas, Nevada:

CRWMS M&O. ACC: MOL.19991118.0180.

CRWMS M&O 2000a. EQ6 Calculationfor Chemical Degradation ofShippingport LWBR (Th/U Oxide) Spent Nuclear Fuel Waste Packages. CAL-EDC-MD-000008 REV 00. Las Vegas, Nevada:

CRWMS M&O. ACC: MOL.20000926.0295.

CRWMS M&O 2000b. Design Analysis for the Defense High-Level Waste Disposal Container, ANL-DDC-ME-000001 Rev 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20000627.0254.

CRWMS M&O 2001. Evaluation ofCodisposal Viability for U-Metal (N-Reactor) DOE-Owned Fuel. TDR-EDC-NU-000004 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC:

MOL.20010314.0004.

DOE (U.S. Department of Energy) 1999. Design Specification. Volume 1 of Preliminary Design Specification for Department ofEnergy Standardized Spent Nuclear Fuel Canisters.

DOE/SNF/REP-011, Rev. 3. Washington, D.C.: U.S. Department of Energy, Office of Spent Fuel Management and Special Projects. TIC: 246602.

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 72 of74 DOE 2002. Criticality Scoping Analysis ofa Dual Canister/Waste Package Disposal Strategy.

DOE/SNF/REP-080, Rev. O. Idaho Falls, Idaho: U.S. Department of Energy, Idaho Operations Office. ACC: MOL.20031014.0018.

DOE 2003a. Quality Assurance Requirements and Description. DOEIRW-0333P, Rev. 13.

Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management.

ACC: DOC.20030422.0003.

DOE 2003b. TMI Fuel Characteristics for Disposal Criticality Analysis. DOE/SNF/REP-084, Rev. O. Idaho Falls, Idaho: U.S. Department of Energy, Idaho Operations Office. ACC:

MOL.20031013.0388 Parrington, J.R.; Knox, H.D.; Breneman, S.L.; Baum, E.M.; and Feiner, F. 1996. Nuclides and Isotopes, Chart of the Nuclides. 15th Edition. San Jose, California: General Electric Company and KAPL, Inc. TIC: 233705.

Stout, R.B. and Leider, H.R., eds. 1991. Preliminary Waste Form Characteristics Report. Version 1.0. Livermore, California: Lawrence Livermore National Laboratory. ACC: MOL.19940726.0118.

Taylor, W.J. 1997. "Incorporating Hanford 15 Foot (4.5 Meter) Canister into Civilian Radioactive Waste Management System (CRWMS) Baseline." Memorandum from W.J. Taylor (DOE) to J.

Williams (Office of Waste Acceptance Storage and Transportation), April 2, 1997. ACC:

HQP.19970609.00 14.

YMP (Yucca Mountain Site Characterization Project) 2000. Disposal Criticality Analysis Methodology Topical Report. YMP/TR-004Q, Rev. 01. Las Vegas, Nevada: Yucca Mountain Site Characterization Office. ACC: MOL.20001214.0001.

7.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES AP-3.12Q, Rev. 2, ICN 2. Design Calculation and Analyses. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Managment. ACC: DOC.20040318.0002.

AP-3.15Q, Rev. 4, ICN 2. Managing Technical Product Inputs. Washington, D.C.: U.S.

Department of Energy, Office of Civilian Radioactive Waste Management. ACC:

DOC.20030627.0002.

ASTM (American Society for Testing and Materials) A 2401 A 240M-99b. 2000. Standard Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet, and Strip for Pressure Vessels. West Conshohocken, Pennsylvania: American Society for Testing and Materials. TIC: 248529.

ASTM A 276-91a. 1991. Standard Specification for Stainless and Heat-Resisting Steel Bars and Shapes. Philadelphia, Pennsylvania: American Society for Testing and Materials. TIC: 240022.

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 73 of 74 ASTM A 516/A 516M-90. 1991.* Standard Specification for Pressure Vessel Plates, Carbon Steel, for Moderate-and Lower-Temperature Service. West Conshohocken, Pennsylvania: American Society for Testing and Materials. TIC: 240032.

ASTM G 1-90 (Reapproved 1999). 1999. Standard Practice for Preparing, Cleaning, and Evaluating Corrosion Test Specimens. West Conshohocken, Pennsylvania: American Society for Testing and Materials. TIC: 238771.

LP-SI.llQ-BSC, Rev.O. Software Management. Las Vegas, NV: BSC. ACC: DOC.20040225.0007.

7.3 SOURCE DATA M00003RIB00071.000. Physical and Chemical Characteristics of Alloy 22. Submittal date:

03/13/2000.

M00003RIBOoon.000. Physical and Chemical Characteristics of Steel, A 516. Submittal date:

03/13/2000.

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD-NU-000004 Rev OOA Page 74 of74

8. ATTACHMENTS Attachment I: One Compact Disk (CD) containing MCNP input and output files and the EXCEL spreadsheet used in the calculation process.

Attachment II: Description of file contained in Attachment I (1 page).

Engineered Systems Project Calculation

Title:

Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package Document Identifier: CAL-DSD -NU-000004 Rev OOA Attachment II Page II-I of II-I ATTACHMENT II This attachment contains a listing and description of the zip file contained on the attachment CD of this calculation. The zip archive was created using WinZip 8.1. The zip file attributes are:

Archive File Name File Size (bytes) FileDate File Time TMl.zip 75,224,537 10/15/2003 12:58 PM There are 871 total files contained in a unique directory structure. Upon file extraction, 870 MCNP input and output files along with one Excel spreadsheet will be found.