Initial Radionuclide Inventories
ANL-WIS-MD-000020 REV 01
September 2004 

DOC.20040921.0003
QA: QA 

NOTICE OF OPEN CHANGE DOCUMENTS - THIS DOCUMENT IS IMPACTED BY THE LISTED CHANGE DOCUMENT AND CANNOT BE USED WITHOUT IT.
-----------------------------------------
1) ACN-001, DATED 09/27/2005 


Prepared for: 

U.S. Department of Energy 
Office of Civilian Radioactive Waste Management 
Office of Repository Development 
1551 Hillshire Drive 
Las Vegas, Nevada 89134-6321 

Prepared by: 
Bechtel SAIC Company, LLC 
1180 Town Center Drive 
Las Vegas, Nevada 89144 

Under Contract Number 
DE-AC28-01RW12101 


DISCLAIMER 

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither 
the United States Government nor any agency thereof, nor any of their employees, nor any of their contractors, 
subcontractors or their employees, makes any warranty, express or implied, or assumes any legal liability or 
responsibility for the accuracy, completeness, or any third partys use or the results of such use of any information, 
apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. 
Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, 
or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United 
States Government or any agency thereof or its contractors or subcontractors. The views and opinions of authors 
expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. 




Page iv 

OCRWM SCIENTIFIC ANALYSIS CHANGE HISTORY 
(CONTINUED) 1. Total Pages: 124 
2. Scientific Analysis Title 
Initial Radionuclide Inventories 
3. DI (including Revision Number) 
ANL-WIS-MD-000020 REV 01 
Change History (Continued) 
12. Revision No. 13. Description of Change 
01 
(Continued) 
Continued from Page iii: 
Issues from Audit OCRWMP-BSC-04-16 performed on Initial Radionuclide Inventories Rev 0 during the 
preparation of Rev01 have been addressed in Rev 01. (Note: input and output data have not changed from Rev 0 
to Rev. 01). As a result of the audit, CR2810D and CR2848B were filed. 
CR2810D: Assumptions are now documented in Section 5. The number of canisters per waste package is defined 
in Tables 4-2 and then through analysis in Table 6-3 in accordance with the referenced baseline document. The 
correlation between configurations and "bins" as been provided. "Arrival Cases" A-C and A'-C- have been 
defined (Table 6-6). Headers and titles for tables have been reconfigured for transparency. Direct input data are 
reported in Section 4 exactly as represented in source documents (e.g. format, inventory, fractional waste 
packages). Table 13 (now Table 4-14) are direct inputs from Radionuclide Screening (BSC 2002 [160059]). 
CR2848: Inconsistencies perceived between this analysis and Source Terms for HLW Glass Canisters, 
CAL-MGR-NU-000002 REV 01, both of which provide inputs to TSPA-LA, have been resolved. 
Condition #1:The Hamel 2003 and Allison 2004 documents are more recent sources of data than the Source 
Terms for HLW Glass Canisters document published in the 2000. Although 137Cs and 90Sr "capsule" data is not 
specifically called out in the Hamel or Allison document inventories, the inventory of these two isotopes 
previously reported are bounded by the data reported here. The number of Hanford and SRS canisters has been 
used to determine the weighted average of the radionuclide per canister. However, the total grams per package 
reported in the output DTN, is based on the total number of canisters using the canister per package data for 
codisposal packages shown in Tables 6-3. 
Condition 2: The "arrival methodologies" for the 63,000 MTU case were labeled A-C, and A'-D' and the cases 
were defined (Table 6-6) and the data tables labeled consistently to improve transparency of the analysis. 
Condition 3: Projected average burnup ratios (Table 6-5) used to develop uncertainty factors (Table 6-7) for 
reported the CSNF inventory are based on the number of packages associated with a given burnup, i.e., 
GWd/MTU. Using the MTU instead of the number of packages at a given burnup produces the same ratios and 
therefore there is no impact to reported values to TSPA-LA. See workbook inv_rev1C.xls, worksheet: new 
burnup. 
Condition 4: Hamel (source of Hanford's HLW inventory), Fowler (source of SRS HLW inventory) and Luptak 
(via substitution of DOE/SNF/REP-078 Rev. 01) have been qualified for intended use in accordance with 
AP-SIII.9Q. The qualifications are presented in Appendix IV of REV 01 of this analysis. The Fowler reference is 
now referred to as Allison 2004. 
These qualifications remove TBV- 5536 (Hamel) and TBV-5455 (Luptak). Qualifications are presented in 
Appendix IV. The Luptak memo was replaced with DOE/RW/REP-078 Rev.01 in accordance with the 
Resolution Plan noted in TBV-5455. This removed the TBV. The differences in inventories have been addressed 
in Sec. 6.6.2 and the analysis shows that the new data are within the bounds of those previously identified. 

CONTENTS 


Page 

1. 
PURPOSE.............................................................................................................................1-1 
1.1 
STUDY LIMITATIONS AND CONSIDERATIONS ................................................. 1-2 
1.2 
YUCCA MOUNTAIN PROJECT DOCUMENTS PROVIDING INPUT TO
THIS ANALYSIS.........................................................................................................1-3 
2. 
QUALITY ASSURANCE.....................................................................................................2-1 
3. 
USE OF SOFTWARE ........................................................................................................... 3-1 
3.1 
BASELINED SOFTWARE.......................................................................................... 3-1 
3.2 
COMPUTER SOFTWARE USED............................................................................... 3-1 
4. 
INPUTS................................................................................................................................. 4-1 
4.1 
DIRECT INPUTS ......................................................................................................... 4-1 
4.2 
CRITERIA .................................................................................................................. 4-18 
4.3 
CODES, STANDARDS, AND REGULATIONS...................................................... 4-19 
5. 
ASSUMPTIONS.................................................................................................................... 5-1 
5.1 
COMMERCIAL SPENT NUCLEAR FUEL ............................................................... 5-1 
5.2 
DOE-OWNED SPENT NUCLEAR FUEL..................................................................5-1 
5.3 
DOE-OWNED HLW (DHLW) .................................................................................... 5-4 
6. 
SCIENTIFIC ANALYSIS DISCUSSION............................................................................. 6-1 
6.1 
RADIONUCLIDES ...................................................................................................... 6-4 
6.2 
WASTES AND WASTE PACKAGES........................................................................6-5 
6.3 
FEATURES, EVENTS, AND PROCESSES CONSIDERED IN THE 
ANALYSIS.................................................................................................................6-11 
6.4 
APPROACH AND METHODS OF CALCULATIONS ........................................... 6-12 
6.4.1 
Radionuclide Inventory................................................................................... 6-12 
6.4.2 
Inventory Heterogeneity ................................................................................. 6-12 
6.4.3 
Radionuclide Analysis of Activated Mineral Deposits on Assemblies .......... 6-13 
6.5 
RADIONUCLIDE INVENTORY ANALYSES RESULTS...................................... 6-13 
6.6 
DISCUSSION OF UNCERTAINTIES ...................................................................... 6-14 
6.6.1 
CSNF Waste.................................................................................................... 6-14 
6.6.2 
DSNF Waste ................................................................................................... 6-19 
6.6.3 
HLW Glass Waste........................................................................................... 6-22 
6.7 
INTENDED USE OF OUTPUT DATA..................................................................... 6-25 
7. 
CONCLUSIONS.................................................................................................................... 7-1 
7.1 
INITIAL RADIONUCLIDE INVENTORIES ................................................................ 7-2 
7.2 
YUCCA MOUNTAIN REVIEW PLAN ACCEPTANCE CRITERIA ....................... 7-3 

CONTENTS 


Page 

8. 
INPUTS AND REFERENCES..............................................................................................8-1 
8.1 
DOCUMENTS CITED.................................................................................................8-1 
8.2 
CODES, STANDARDS, REGULATIONS, AND PROCEDURES............................ 8-5 
8.3 
SOFTWARE CODES...................................................................................................8-6 
8.4 
SOURCE DATA, LISTED BY DATA TRACKING NUMBER ................................ 8-7 
APPENDIX I. 
NEUTRON-ACTIVATION PRODUCTS OUTSIDE THE SPENTFUEL 
MATRIX..............................................................................................I-1 
APPENDIX II. 
DESCRIPTION OF FILES ON CD ............................................................. II-1 
APPENDIX III. 
WORKSHEETS AND PLOTS IN INV_REV1.XLS..................................III-1 
APPENDIX IV. 
QUALIFICATION OF UNQUALIFIED DATA USED AS DIRECT
INPUTS........................................................................................................IV-1 
APPENDIX V. 
DATA QUALIFICATION PLAN................................................................ V-1 

FIGURES 


Page 

6-1. Various Waste Package Types Containing Different Wastes..........................................6-7 
6-2. Some Waste Package Configurations for CSNF, DSNF, Naval, and HLW Glass..........6-8 
6-3. Histograms of Waste Package Configurations .............................................................. 6-11 
6-4. Average Grams per Package of Radionuclides for Each Waste Type .......................... 6-13 
6-5. Comparison of the Per-Package Inventories of the Five CSNF Configurations ........... 6-16 
6-6. Ratios of CSNF Configuration Inventories to the CSNF Weighted Average ............... 6-17 
6-7. Uncertainty in 2003 DSNF Radionuclide Inventory Estimates.....................................6-20 
6-8. 
Comparison of the Nominal Radionuclide Inventory for 2003 (old nom) and 2004 
(new nom) ...................................................................................................................... 6-21 
6-9. Comparison of the 2003 and 2004 DSNF Radionuclide Inventories ............................ 6-22 
6-10. 
Radionuclide Inventories of the Older and Newer Estimates for Average Short and 
Long HLW Glass Canisters ........................................................................................... 6-24 
6-11. SRS Glass Batch Inventories ......................................................................................... 6-24 

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TABLES 


Page 

4-1. Direct Inputs Used in This Report ................................................................................... 4-2 
4-2. 
Waste Package Configurations and Inventory Information - Nominal Quantity for 
LA...................................................................................................................................4-3 
4-3. Constants Used in Calculations ....................................................................................... 4-3 
4-4. 
Average Radionuclide Inventory (Curies/Assembly) for Each CSNF Waste Package 
Configuration at the Projected Year of Repository Emplacement (2033) for 
63,000 MTU, Case A.......................................................................................................4-4 
4-5. 
63,000 MTU Waste Package Burnup Characteristics Based on 1999 CSNF Waste 
Streams............................................................................................................................4-6 
4-6. 
CSNF Assembly Totals and Burnup Characteristics for the 63,000 MTHM Cases 
Based on 2002 CSNF Waste Streams..............................................................................4-6 
4-7. Summary of Percent Differences between Calculated and Computed Compositions.....4-7 
4-8. Nominal and Bounding Radionuclide Inventories for DSNF Canisters..........................4-9 
4-9. 
New Estimated Curies Per Canister for Savannah River Site (SRS) HLW (Batches 
1A, 1B, 2 and 3, and 4 to 10)......................................................................................... 4-11 
4-10. Estimated Radioactivity in the Hanford HLW Canisters at Decay Date of 2030.......... 4-12 
4-11. 
Hanford HLW Canister Production Estimates for Alternative Canister Wastes 
Loading and Canister Fill Levels...................................................................................4-12 
4-12. Radionuclide Half-Lives................................................................................................4-13 
4-13. Results of the Screening Analysis.................................................................................. 4-14 
4-14. 
Additional Radionuclides Needed for Accurate Accounting of Those Screened In 
for 10,000 Years............................................................................................................. 4-15 
4-15. 
Identification and Screening Status of CSNF PWR and BWR Hardware-Activation 
Products.........................................................................................................................4-17 
4-16. 99Tc and 14C Inventories of BWR Assembly at Specified Characteristics ................... 4-18 
5-1. Assumptions Related to DSNF ........................................................................................ 5-2 
6-1. Precursors Identified as Significant to the Regulatory Period for TSPA-LA..................6-5 
6-2. Included Radionuclides (32)............................................................................................6-5
6-3. 
Package Configurations Including New Waste Codisposal Packages to 
Accommodate DSNFa....................................................................................................6-10 
6-4. 
Features, Events, and Processes Included (Screened In) in TSPA-LA and Addressed 
in this Report.................................................................................................................. 6-12 
6-5. Average Burnups for the 5 CSNF Configurations, and Ratios to Average ................... 6-16 
6-6. Arrival Forecasts for the 63,000 MTU .......................................................................... 6-18 
6-7. Average Burnups for the Arrival Scenarios Projected For Years 1999 And 2002........6-19 
7-1. Nominal Grams per Waste Package of Radionuclides for Each Type of Waste.............7-2 
7-2. 
Uncertainty Multipliers for Grams per Package of Radionuclides for Each Waste 
Type ................................................................................................................................ 7-3 

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ACRONYMS 


AP absorber plates 

BOL beginning of life 
BSC Bechtel SAIC Company, LLC 
BWR boiling water reactor 

CDSP codisposal 
CSNF commercial spent nuclear fuel 
CR control rods 

DOE U.S. Department of Energy 
DHLW DOE owned high-level (radioactive) waste 
DSNF DOE owned spent nuclear fuel 

EIS Environmental Impact Statement 
EOC end of cycle 
EOL end of life 
EPRI Electric Power Research Institute 

FEIS Final Environmental Impact Statement 
FEP feature, event, or process 
FEPs features, events, and processes 

HLW high-level (radioactive) waste 

INEEL Idaho National Engineering and Environmental Laboratory 

LA License Application 

MCO multi-canister overpack 
MTHM metric tons of heavy metal 
MTU metric tons of uranium 

PWR pressurized water reactor 

SNF spent nuclear fuel 
SNFID spent nuclear fuel identification 
SR Site Recommendation 
SRS Savannah River Site 

TSPA Total System Performance Assessment 
TSPA-LA Total System Performance Assessment - License Application 


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1. PURPOSE 


The purpose of this analysis is to provide an initial radionuclide inventory (in grams per waste 
package) and associated uncertainty distributions for use in the Total System Performance 
Assessment for the License Application (TSPA-LA) in support of the license application for the 
repository at Yucca Mountain, Nevada. This document is intended for use in postclosure 
analysis only. 

Bounding waste stream information and data were collected that capture probable limits. For 
commercially generated waste, this analysis considers alternative waste stream projections to 
bound the characteristics of wastes likely to be encountered using arrival scenarios that 
potentially impact the commercial spent nuclear fuel (CSNF) waste stream. For TSPA-LA, this 
radionuclide inventory analysis considers U.S. Department of Energy (DOE) high-level 
radioactive waste (DHLW) glass and two types of spent nuclear fuel (SNF): CSNF and DOE-
owned (DSNF). These wastes are placed in two groups of waste packages: the CSNF waste 
package and the codisposal waste package (CDSP), which are designated to contain DHLW 
glass and DSNF, or DHLW glass only. The radionuclide inventory for naval SNF is provided 
separately in the classified Naval Nuclear Propulsion Program Technical Support Document for 
the License Application. 

As noted previously, the radionuclide inventory data presented here is intended only for 
TSPA-LA postclosure calculations. It is not applicable to preclosure safety calculations. Safe 
storage, transportation, and ultimate disposal of these wastes require safety analyses to support 
the design and licensing of repository equipment and facilities. These analyses will require 
radionuclide inventories to represent the radioactive source term that must be accommodated 
during handling, storage and disposition of these wastes. This analysis uses the best available 
information to identify the radionuclide inventory that is expected at the last year of last 
emplacement, currently identified as 2030 and 2033, depending on the type of waste. TSPA-LA 
uses the results of this analysis to decay the inventory to the year of repository closure projected 
for the year of 2060. 

The radionuclides of importance to TSPA-LA dose calculations were assessed in Radionuclide 
Screening (BSC 2002 [DIRS 160059]). The radionuclide screening analysis separately 
considered the following two postclosure time periods: the 10,000-year regulatory period for the 
repository at Yucca Mountain, and the period of up to one million years after emplacement 
identified in Final Environmental Impact Statement (FEIS) (DOE 2002 [DIRS 155970], 
Section 6.3). Four release scenarios were considered: (1) nominal, (2) human-intrusion, (3) 
intrusive igneous event, and (4) eruptive igneous event. Table 13 of the screening analysis (Table 
4-13) identifies those isotopes that contribute 95 percent of the dose for the different scenarios 
implemented in the TSPA-LA for the 10,000-year regulatory period (columns 2 and 4) and for 
the FEIS period of one million years (columns 3 and 5). The screening document also indicated 
that a separate Environmental Protection Agency (EPA) groundwater protection standard 
(40 CFR 197.30 [DIRS 165519]) required 228Ra be added to the list. Precursors of the screened-
in isotopes due to radioactive decay and in-growth were identified for consideration (Table 4-14) 
(BSC 2002 [DIRS 160059], Table 12). Precursors must be considered for inclusion in TSPA-LA, 
either by direct inclusion or by appropriate augmentation of the daughter product. Based on this 
premise, this analysis identifies seven precursors for the 21 radionuclides identified as significant 


to the regulatory period (Table 6-1). These have been added to the inventory for TSPA-LA 
postclosure analysis. Thus, a total of 28 radionuclides for the 10,000-year regulatory period and 
an additional four important to the million-year FEIS calculations are identified for TSPA-LA 
(Tables 6-2 and 7-1). The screening document notes that extending the screening to 20,000 
years adds no radionuclides to the lists (DOE 2002 [DIRS 155970]). 

This analysis was prepared in accordance with Technical Work Plan for: Regulatory Integration 
Modeling and Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 171583]). 
This analysis provides information and data to support the technical basis for the waste-form 
representation in the TSPA-LA as it relates to the initial radionuclide inventory at the projected 
time of emplacement. The technical work plan defines the scope of this report and identifies the 
regulatory and acceptance criteria to be addressed in this analysis. There were no deviations from 
the technical work plan. In addition, the guidance provided in Total System Performance 
Assessment-License Application Methods and Approach (BSC 2003 [DIRS 166296]) has been 
used in the preparation of this analysis as applicable. Acceptance criteria related to 
Section 2.2.1.3.4 of Yucca Mountain Review Plan, Final Report (NRC 2003 [DIRS 163274]) are 
addressed in this report, as identified in Table 3-1 of Technical Work Plan for: Regulatory 
Integration Modeling and Analysis of the Waste Form and Waste Package (BSC 2004 
[DIRS 171583]). 

1.1 STUDY LIMITATIONS AND CONSIDERATIONS 

As required by the Nuclear Waste Policy Act, as amended (NWPA 1987 [DIRS 100016]), the 
DOE must dispose of all under-contract CSNF and HLW and all government-managed SNF and 
HLW (BSC 2003 [DIRS 165990]). This analysis addresses the base case for CSNF currently 
allocated (63,000 MTHM CSNF and 7,000 MTHM government-managed (DOE-owned) waste) 
in accordance with the legal-limit imposed by the Nuclear Waste Policy Act, as amended 
(NWPA 1987 [DIRS 100016]). 

DSNF includes diverse fuels from various experimental, research, and production reactors and 
consists of several hundred different fuel types that have been stored at several sites (DOE 2003 
[DIRS 163377], p. 14). Because much of the available data for these fuels is historical, a process 
for creating a conservative estimate of the DSNF source terms was developed. The process relies 
on data used to provide radionuclide inventories for typical SNFs at a range of decay times. 
These results used ORIGEN-based calculational techniques to develop templates scaled to 
estimate the radionuclide inventory of other similar fuels. Several validation studies have been 
performed to demonstrate the validity of the model and underlying codes used (DOE 2004 
[DIRS 169354], p.14). The National Spent Nuclear Fuel Program consolidates DSNF 
information and makes it available to support DOE planning and scoping activities as well as 
design and licensing efforts to enable final repository disposal of DSNF. This source (used as 
input to this analysis) provides radionuclide inventory data for all DSNF destined for the 
repository except for naval SNF. 

Among the government-managed wastes, naval SNF (which is part of the DSNF MTHM 
repository allocation) represents a very small percentage of the nuclear waste expected for the 
repository (BSC 2003 [DIRS 165990] p. xix). Details regarding the design and performance of 
naval SNF are provided separately in the classified Naval Nuclear Propulsion Program 


Technical Support Document and are not part of this analysis. Naval SNF waste packages (total 
long and short canisters) have been reported as an upper bound of 300 (McKenzie 2001 
[DIRS 158051]) and have been included in the number of CSNF waste packages. 

The CSNF bare fuel in DOE's possession in not included in this analysis. Source Term Estimates 
for Spent Nuclear Fuels (DOE 2003 [DIRS 163377]) provides the inventory for DSNF. This 
analysis excludes that part of the inventory associated with bare fuel and assumes that the 
remaining inventory is conservative for TSPA-LA calculations. The basis for this assumption is 
provided in Section 5.2. The repository allotment for DSNF including naval fuel is 2,333 MTHM 
(Dreyfus 1995 [DIRS 104384]; Lytle 1995 [DIRS 104398]). The repository allocation for naval 
fuel is 65 MTHM, leaving an allotment of 2,268 MTHM for DSNF. The DSNF inventory (2,348 
metric tons of 238U), excluding the bare fuel inventory (48 metric tons of 238U), exceeds the 
repository allotment for DSNF (2,348 - 48 = 2,300, which is greater than 2,268). Therefore, 
using the DSNF radionuclide inventory reported in Source Term Estimates for Spent Nuclear 
Fuels (DOE 2003 [DIRS 163377]), excluding the bare fuel produces a conservative DSNF 
inventory for TSPA-LA. 

Information in this analysis is based on projections of future events. Therefore, it is recognized 
that there may be wastes with characteristics different from those described in this document. 
The DOE will accommodate, through operational adjustments and design flexibility, a wide 
range of transport and arrival sequences, packaging configurations, and waste forms. This 
analysis takes into consideration these parameters and uses the data as input, understanding that 
future analyses or models may define them differently in the future. 

1.2 YUCCA MOUNTAIN PROJECT DOCUMENTS PROVIDING INPUT TO THIS 
ANALYSIS 

Several documents produced by the Yucca Mountain Project provide direct inputs to this report 
(BSC 2002 [DIRS 160059]; BSC 2003 [DIRS 165990], Tables 2 and 4; CRWMS M&O 2000 
[DIRS 138239]). 


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2. QUALITY ASSURANCE 


This document was developed in accordance with Section 8 of Technical Work Plan for: 
Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package 

(BSC 2004 [DIRS 171583]). The technical work plan indicates that this analysis is subject to the 
Yucca Mountain Project quality assurance program (DOE 2004 [DIRS 171539]) because it will 
be used to support the TSPA-LA. This document is subject to Quality Assurance Requirements 
Description (DOE 2004 [DIRS 171539]) requirements. 

AP-SIII.9Q, Scientific Analyses, was primarily followed in developing this document. Data used 
as direct input to this analysis are qualified for intended use in accordance with AP-SIII.2Q, 
Qualification of Unqualified Data; AP-3.15Q, Managing Technical Product Inputs; and 
AP-SIII.9Q, Scientific Analyses. The qualification plans and reports are documented in 
Appendix IV. The signed originals of the data qualification plans are included in the record 
package for this analysis. 

Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and 
Waste Package (BSC 2004 [DIRS 171583]) includes an evaluation of current work processes 
and procedures in accordance with AP-SV.1Q, Control of the Electronic Management of 
Information. This evaluation determined that the work conducted under this technical work plan 
require controls for the electronic management of information. This analysis activity does not 
involve sensitive unclassified information and, therefore, controls for that purpose were not 
implemented. Controls to protect information from damage or destruction for its prescribed 
lifetime include dual electronic storage on individual hard drives and a network drive with 
remote backup each night to a secure drive. These systems also ensure that information is readily 
retrievable. 

The BSC Intranet includes several controlled databases for the storage of data, program 
documents, procedures, reports, models, analyses, etc. to ensure information is accurate and 
readily available (e.g., OCRWM Program Documents Database (OPDD), Document Information 
Reference System (DIRS), Records Information System (RIS), Automated Technical Data 
Tracking system (ATDT)). These systems, in conjunction with procedures, ensure that data or 
information inputs are from controlled sources and changes to input (data, documents, and 
design information) undergo the same level of control as the original. Access to controlled 
information is limited to authorized users through password protection. These system-specified 
design parameters define how the information will be stored with respect to media, conditions, 
location retention time, security, and access. They also provide controls to properly identify 
storage and transfer media as to source, physical and logical format, size, and relevant date. 

With respect to analyses, AP-SIII.9Q, Scientific Analyses ensures documents address content 
determined important to TSPA-LA. AP-17.1Q, Records Management, and AP-6.1Q, Document 
Control, ensure documents and records are controlled and complete, specify controls for changes 
to documents, and define retention requirements for records. To ensure transfers of information 
are error free or within a defined permissible error rate, sum checks after transfers or visual 
checks of file information (such as file size, date and time, and file names) are used. These 
checks and the following procedural controls provide adequate controlled access to maintain the 
security and integrity of the information and ensure transfers of information are error free or 


within a defined permissible error rate. AP-3.15, Managing Technical Product Inputs; 
AP-SIII.2Q Qualification of Unqualified Data; and AP-SIII.3Q, Submittal and Incorporation of 
Data to the Technical Data Management System are some examples. All data used as direct 
input to develop the results or conclusions in this analysis are qualified in accordance with 
AP-SIII.2Q, Qualification of Unqualified Data, AP-3-15Q, Managing Technical Product Inputs, 
or AP-SIII.9Q, Scientific Analyses. AP-SIII.2Q and AP-SIII.9Q provide criteria and processes 
for the qualification of data for its intended use within a technical product. 

Controls are also implemented that include the use of InfoWorks software. InfoWorks is a BSC 
project-wide control, for version control implemented during the preparation and editing of this 
analysis. Access to documents and subsequent drafts using InfoWorks is password protected. 
Visual checks were performed to ensure data used in this analysis has been documented 
accurately and completely in Design and Engineering (D&E) Information Exchange Drawings 
(IEDs). IEDs are prepared for data used in analyses and model reports that originate from Design 
and Engineering (D&E) calculations. This IED process includes a Technical Management 
Review Board review and approval of a decision proposal and copy of the IED. The data 
selected is at the author's discretion. 

This report does not directly impact structures, systems, or components classified in accordance 
with AP-2.22Q, Classification Analyses and Maintenance of the Q-List. 


3. USE OF SOFTWARE 


3.1 BASELINED SOFTWARE 

No controlled and baselined software subject to Section 2.2 of LP-SI.11Q-BSC, Software 
Management, was used in the creation of this document. 

3.2 COMPUTER SOFTWARE USED 

The calculations performed in this analysis used commercially available, off-the-shelf software 
(Microsoft Excel 97). Use of Microsoft Excel is considered appropriate because a) the 
calculations require simple mathematical expressions for deriving the final results, which are 
standard operations in Microsoft Excel, and b) Microsoft Excel has built-in graphical charting 
capabilities. No software routines or macros are developed in this use of Microsoft Excel. 
Calculations and the use of inputs are discussed in Sections 4 and 6. Details of the workbook 
and worksheet calculations and the corresponding mathematical expressions are given in 
Appendix II. Appendix III is a CD containing the Microsoft Excel file inv_rev1.xls. 
Workbook and worksheet names will be denoted in Courier font throughout this document. 


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4. INPUTS 


4.1 DIRECT INPUTS 

This section addresses the inputs directly used to develop the initial radionuclide inventory for input 
to TSPA-LA. Table 4-1 provides a brief description of data used as direct input, its location in this 
document, and the data source. Figure II-1 provides a graphical depiction of the flow of the analysis 
and the use of inputs. Equations that are used in this analysis are simple arithmetic operations 
(multiplication, division, addition, and unit conversions) and are discussed in the context of 
presenting the use of inputs for analysis development in Section 6. Appendices II and III provide 
further illustration of these operations. 

To determine the average grams per package for the radionuclide inventory expected at the end of 
emplacement, calculations using several data inputs are needed. For the purpose of this analysis, the 
inventory considers DHLW glass and the three types of SNF: CSNF, DSNF, and naval SNF. These 
wastes are distributed in two groups of waste packages: (1) the CSNF waste packages (which include 
naval waste packages for the purposes of the analysis only), and (2) the codisposal waste package, 
which contains DSNF and DHLW glass or only DHLW glass. 

Table 4-2 identifies waste package configurations and the nominal quantity of waste packages 
assigned to each configuration for TSPA-LA (BSC 2004 [DIRS 169472], Table 11). To facilitate the 
transparency of this analysis, waste package configurations identified in Table 11 of D&E/PA/C IED 
Typical Waste Package Components Assembly (BSC 2004 [DIRS 169472]) are described (type of 
waste package for each configuration is identified) and a number is assigned that correlates with the 
analysis performed as described in Section 6. 

The lists of radionuclides determined to be significant to TSPA-LA dose calculations were identified 
in Radionuclide Screening (BSC 2002 [DIRS 160059], Tables 12 and 13). Table 4-13 shows the 
20 radionuclides screened in that contribute to 95 percent of the dose for the scenarios 
implemented in the TSPA-LA. The radionuclides screened in for the 10,000-year regulatory period 
after emplacement are shown in columns 2 and 4. The radionuclides identified for the FEIS million-
year period are shown in columns 3 and 5 (BSC 2002 [DIRS 160059], Table 13). This list includes 
228Ra, which was added to comply with EPA's groundwater protection standard. Radionuclide 
Screening (BSC 2002 [DIRS 160059] identified three radionuclides that are precursors of screened 
in radionuclides for consideration, to be added directly to the inventory or added by augmentation of 
the daughter product (Table 4-14). This analysis identifies seven radionuclides (Table 6-1) that are 
precursors to screened in isotopes to be included in the inventory for the regulatory period. Thus, 
20 + 1 + 7 radionuclides are identified for the 10,000-year regulatory period. The addition of four 
radionuclides for the FEIS million-year period results in the final list of 32 radionuclides for 
TSPA-LA analysis (Tables 6-2 and 7-1). Table 7-1 shows the nominal radionuclide inventory in 
grams per radionuclide per waste package at the projected year of final emplacement) 
(DTN: SN0310T0505503.004) for 32 radionuclides. 

To determine g/Ci (grams per curie), Equation II-1 (Appendix II) is used. The mass per activity 
(mi /ai) in g/Ci for each of 32 radionuclides (i) from Table 6-2 is calculated using constants from 
Parrington et al. (1996 [DIRS 103896]). The latest half-life for 79Se was taken from Nuclear Data 
Sheets for A = 79 (Singh 2002 [DIRS 164741], p. 1). The molecular weights (MWi) (in g/mole) 
were taken from the radionuclide names because that value had enough significant figures for these 
purposes. 


Table 4-1. Direct Inputs Used in This Report 

Data/Technical Information Values/Table Units Source 
Waste 
Package 
Configuration 
Number of packages for each 
configuration Table 4-2 Packages BSC 2004 [DIRS 169472], Table 11 
Number of canisters or assemblies 
per package Table 4-2 Canisters / 
Assemblies BSC 2004 [DIRS 169472], Table 11 
Average CSNF radionuclide 
inventory in 2033 for each of the 5 
WP configurations (Bins 1-5) for 
Case A-63,000 MTU scenario 
based on 1999 CSNF waste 
streams. 
Table 4-4 
Curies per 
radionuclide 
per 
assembly 
(Bins 1-5) 
CRWMS M&O 2000 [DIRS 138239], 
from Disk I, Attachment III, File 
D:\BIN.EXE_Files\output_files\Case_ 
A_63K\average_nuclides.dat. Also, 
see (CD for this analysis): workbook 
CSNFcasea63.xls 
Average Initial Burnup values for 
63,000 MTU (63K) Cases A-C 
(Arrival scenarios) for the 5 WP 
configurations; number of waste 
packages for each waste package 
design 
Table 4-5 Burnup in 
GWd/MTU 
CRWMS M&O 2000 [DIRS 138239] 
Tables 10, 12, 14; p. 18 
Total Number of PWR&BWR 
Assemblies and burnup values for 
63K, Cases A' to D' (Arrival 
scenarios) 
Table 4-6 
Burnup in 
units of 
GWd/MTM 
BSC 2003 [DIRS 165990], Table 2 
BWR & PWR burnup values for the 
63K MTHM Cases (Arrival 
scenarios A' to D') 
Table 4-6 GWd/MTU 
BSC 2003 [DIRS 165990], Table 4 
Modeled Calculation of Uncertainty 
for CSNF 
Table 4-7 
Section 6.6.1 
None 
DeHart and Hermann 1996 
[DIRS 156084], Table 19 
Hermann and DeHart 1998 
[DIRS 106563], Table 24 
Isotopes Required by EPA & NRC 228Ra, 226Ra none 
40 CFR 197.30 [DIRS 165519] and 
10 CFR 63.331 [DIRS 156605] for 
groundwater protection 
DSNF inventory Table 4-8 Curies/ 
canister 
First eight worksheets of Total 
2030 D568-585.xls (DOE 2003 
[DIRS 163377]) 
DSNF canister count 2,500 to 5,000 Canisters DOE 2004 [DIRS 169354], p. 41 
SRS Glass Information Table 4-9 Curies/ 
canister Allison 2004 [DIRS 168734] 
Hanford estimated radiochemical 
inventory in HLW, Basis 1/1/2030 Table 4-10 Curies Hamel 2003 [DIRS 164947], Table 2 
HLW canister production estimates 
for Alt. Waste loading and canister 
fill levels 
Table 4-11 Canisters Hamel 2003 [DIRS 164947], p.1-9, 
Table 1-3 
Radionuclide half-lives, constants 
and conversion factors 
Tables 4-3 and 
4-13 
Years, 
grams, mass 
per activity 
(mi /ai) 
Parrington et al. 1996 [DIRS 103896] 
Half-life of 79Se Table 4-3 Years Singh 2002 [DIRS 164741] 
Half-life of 232Th Table 4-3 Lide 1991 [DIRS 131202], p. 11 to 
122 
List of Radionuclides 
Tables 4-14 
and 4-15 None 
BSC 2002 [DIRS 160059], Tables 12 
and 13; Parrington et al. 1996 
[DIRS 103896] 
Radionuclides in crud Appendix I.3 & 
I.4 Curies BSC 2004 [DIRS 171407], Table 2 
BSC 2004 [DIRS 171435], Table 8 
Radionuclide inventory in hardware Table 16 Curies BSC 2004 [DIRS 171435] 
BSC 2004 [DIRS 171436] 


Table 4-2. Waste Package Configurations and Inventory Information - Nominal Quantity for LA 

Configuration 
Number 
Waste Package 
Configuration 
Waste 
Package 
Type 
Number of Waste 
Packages 
(Nominal Qty. for 
LA) Configuration Description 
1 21-PWR AP CSNF 4,299 Waste package with absorber plates for criticality control 
that will hold up to 21 PWR assemblies 
2 21-PWR CR CSNF 95 Waste packages with control rods for criticality control 
that will hold up to 21 PWR assemblies 
3 12-PWR AP long CSNF 163 Waste package with absorber plates for criticality control 
that will hold up to 12 PWR assemblies 
4 44-BWR AP CSNF 2,831 Waste package with absorber plates for criticality control 
that will hold up to 44 BWR assemblies 
5 24-BWR AP CSNF 84 Waste package with thick absorber plates for criticality 
control that will hold up to 24 BWR assemblies 
6 5 HLW short / 
1 DSNF short CDSP 1,147 Waste package containing five short HLW glass 
canisters and one short DSNF canister 
7a 5 HLW long / 
1 DSNF long CDSP 1,406 Waste package containing 5 long HLW glass canisters 
and one long DSNF canister 
8a 5HLW long / 
1 DSNF short CDSP 31 Waste package containing 5 long HLW glass canisters 
and one short DSNF canister 
7b 5 HLW long only CDSP 679 Waste package containing 5 long HLW glass canisters 
8b 2 MCO/2 HLW CDSP 149 Waste package containing 2 multi-canister over packs 
and 2 long HLW glass canisters 
9 Naval short Naval 144 
10 Naval long Naval 156 
Total CDSP 3,412 
Total in CSNF + Naval 7,772 
Grand Total 11,184 

Source: BSC 2004 [DIRS 169472], Table 11. 

NOTE: The subset totals are derived from source data, but are presented here to enhance transparency of the 
analysis. 

Table 4-3. Constants Used in Calculations 

Constant Value Source 
Avogadro Number 6.02214  1023 Parrington et al. 1996 [DIRS 103896], p. 59 
Bq/Ci 3.70  1010 Parrington et al. 1996 [DIRS 103896], p. 58 
Seconds per Mean Solar Year 31,556,930 Parrington et al. 1996 [DIRS 103896], p. 55 
Half-Lives See Table 4-12 Parrington et al. 1996 [DIRS 103896], pp. 18 to 50 
Half-Life of 79Se 2.95  105 years Singh 2002 [DIRS 164741] 
232Th 1.40  1010 Decimal point left out in Parrington et al. 1996 [DIRS 103896]. 
Value confirmed by Lide 1991 [DIRS 131202], pp. 11 to 122 

In order to derive the radionuclide inventory for CSNF, information on assembly type, burnup, 
and MTU per assembly are used to develop arrival forecasts for the given repository receipt year. 
Table 4-4 provides the average CSNF radionuclide inventory as curies per assembly for each 
CSNF waste package configuration for the projected repository receipt year (year 2033). Waste 
package configurations 1 through 5 in Table 4-2 correspond to Bins 1 through 5 in Table 4-4. 

ANL-WIS-MD-000020 REV 01 4-3 September 2004 


The data for the 32 isotopes of interest to TSPA-LA, (taken from 61 radionuclides identified for 
CSNF, were collected based on the 63,000-MTHM inventory. The 63,000-MTHM inventory is 
90 percent of the 70,000-MTHM legislated repository legal limit. The curies per assembly 
were converted to grams per assembly by multiplying by the g/Ci for each radionuclide. The 
per-assembly inventories (in grams) for CSNF waste packages, multiplied by the number of 
assemblies per package for each configuration (e.g., 21 assemblies for configuration 1; 
44 assemblies for configuration 4, etc.) shown in Table 4-2, yield the grams per package. The 
number of packages in each configuration shown in Table 4-2 multiplied by that configurations 
per-package values yields the total grams for each waste configuration. The total number of 
grams for each CSNF configuration is then used to derive the weighted average. The weighted 
average is calculated using this grand total for the appropriate CSNF configuration divided by 
the total number of CSNF packages. 

Table 4-4. Average Radionuclide Inventory (Curies/Assembly) for Each CSNF Waste Package 

Configuration at the Projected Year of Repository Emplacement (2033) for 63,000 MTU, 

Case A 

Nuclide Bin 1a Bin 2a Bin 3a Bin 4a Bin 5a 
227Ac 1.32E-05 1.19E-05 1.48E-05 8.38E-07 2.45E-07 
241Am 1.68E+03 7.64E+02 2.11E+03 4.39E+02 7.64E+01 
242mAm 4.61E+00 1.02E+00 6.90E+00 1.43E+00 7.11E-02 
243Am 1.49E+01 1.55E+00 2.25E+01 3.94E+00 5.71E-02 
137mBa 2.70E+04 7.19E+03 4.34E+04 8.95E+03 1.07E+03 
14C 2.91E-01 5.94E-01 3.60E-01 1.42E-01 2.75E-02 
113mCd 4.90E+00 5.91E-01 8.86E+00 1.57E+00 5.92E-02 
36Cl 5.45E-03 2.16E-03 7.39E-03 2.36E-03 4.81E-04 
242Cm 3.82E+00 8.42E-01 5.69E+00 1.18E+00 5.87E-02 
243cm 6.34E+00 3.51E-01 1.08E+01 1.64E+00 1.05E-02 
244Cm 8.29E+02 1.83E+01 1.46E+03 2.07E+02 2.61E-01 
245Cm 1.99E-01 7.05E-03 3.06E-01 3.32E-02 6.10E-05 
246Cm 6.86E-02 1.02E-03 1.04E-01 1.50E-02 3.19E-05 
247Cm 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
60Co 2.51E+02 2.37E+01 5.23E+02 4.30E+01 7.64E-01 
134Cs 1.82E+02 5.81E+00 2.63E+02 6.11E+01 1.08E-04 
135Cs 2.71E-01 1.09E-01 3.86E-01 9.98E-02 2.20E-02 
137Cs 2.86E+04 7.61E+03 4.59E+04 9.48E+03 1.13E+03 
154Eu 4.38E+02 2.57E+01 8.67E+02 1.27E+02 1.04E+00 
155Eu 4.61E+01 1.96E+00 9.63E+01 1.79E+01 2.51E-02 
55Fe 8.49E+00 6.94E-01 1.52E+01 3.87E+00 2.31E-03 
221Fra 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
3H 7.47E+01 1.28E+01 1.37E+02 2.68E+01 1.58E+00 
129I 1.71E-02 6.62E-03 2.39E-02 5.60E-03 1.10E-03 
85Kr 7.50E+02 1.17E+02 1.41E+03 2.54E+02 1.15E+01 
93mNb 8.61E+00 1.70E+00 1.49E+01 3.49E-01 2.28E-01 
94Nb 6.60E-01 2.63E-01 9.15E-01 1.46E-02 4.25E-02 
59Ni 1.75E+00 1.81E+00 2.28E+00 4.07E-01 2.19E-01 


Table 4-4. Average Radionuclide Inventory (Curies/Assembly) for Each CSNF Waste Package 
Configuration at the Projected Year of Repository Emplacement (2033) for 63,000 MTU, 
Case A (Continued) 

Nuclide Bin 1a Bin 2a Bin 3a Bin 4a Bin 5a 
63Ni 1.97E+02 1.66E+02 2.74E+02 4.55E+01 1.92E+01 
237Np 1.93E-01 7.02E-02 2.69E-01 5.05E-02 8.15E-03 
231Pa 2.48E-05 2.16E-05 3.34E-05 7.33E-06 5.23E-06 
210Pba 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
107Pd 6.34E-02 1.82E-02 9.00E-02 2.02E-02 2.32E-03 
147Pm 3.37E+02 4.70E+01 6.10E+02 1.52E+02 1.31E-02 
218Poa 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
238Pu 1.57E+03 2.25E+02 2.42E+03 4.04E+02 1.33E+01 
239Pu 1.55E+02 1.13E+02 2.02E+02 4.49E+01 2.55E+01 
240Pu 2.57E+02 1.07E+02 3.48E+02 8.92E+01 1.67E+01 
241Pu 1.61E+04 2.66E+03 2.89E+04 4.43E+03 1.88E+02 
242Pu 1.18E+00 2.12E-01 1.71E+00 3.77E-01 1.26E-02 
226Raa 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
228Raa 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
106Ru 2.76E+01 2.69E-01 8.92E+00 6.92E+00 3.22E-10 
125Sb 2.35E+01 1.74E+00 4.36E+01 9.68E+00 6.55E-04 
79Se 3.59E-02 1.48E-02 5.02E-02 1.19E-02 2.60E-03 
147Sma 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
151Sm 1.66E+02 9.05E+01 2.38E+02 4.04E+01 1.80E+01 
126Sn 2.96E-01 9.96E-02 4.16E-01 9.66E-02 1.48E-02 
90Sr 1.91E+04 5.61E+03 3.07E+04 6.41E+03 8.79E+02 
99Tc 7.12E+00 3.09E+00 9.89E+00 2.40E+00 5.58E-01 
229Tha 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
230Th 1.70E-04 2.52E-04 1.71E-04 5.94E-05 1.04E-04 
232Tha 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
232U 1.39E-02 1.78E-03 2.17E-02 3.19E-03 1.11E-04 
233U 3.71E-05 2.12E-05 4.45E-05 5.77E-06 4.08E-07 
234U 6.03E-01 5.54E-01 7.73E-01 1.98E-01 1.94E-01 
235U 7.53E-03 1.42E-02 9.12E-03 2.29E-03 6.18E-03 
236U 1.39E-01 7.58E-02 1.92E-01 4.58E-02 1.52E-02 
238U 1.35E-01 1.18E-01 1.68E-01 5.64E-02 5.43E-02 
90Y 1.91E+04 5.61E+03 3.07E+04 6.41E+03 8.79E+02 
93Zr 7.05E-01 2.91E-01 9.83E-01 2.55E-01 5.68E-02 

Source: CRWMS M&O 2000 [DIRS 138239]. See Appendix III 
this document, file inv_rev1.xls, worksheet a63 for data. 

NOTE: a0.00E+00 is the value presented in the data input 
source. Bin numbers (1 to 5) are equivalent to 
Configuration numbers (1-5) in Table 4-2. 

To estimate uncertainty in the weighted average inventory for CSNF, the burnups of the 
radionuclides were examined. Table 4-5 and Table 4-6 present burnup characteristics for 
assembly types corresponding to the CSNF waste package configurations identified in Table 4-2. 
Cases A to C in Table 4-5 are based on 1999 waste stream cases (CRWMS M&O 2000 


[DIRS 138239]). Table 4-6, Cases A' to D' (BSC 2003 [DIRS 165990]) are based on four 2002 
waste stream cases. These cases represent different arrival case methodologies pertaining to 
forecasts on how waste packages will arrive at the repository (in terms of age of assembly at the 
time of arrival) and, therefore, affect the characterization of the inventory. The cases are further 
discussed in Section 6; however, it is noted here that the CSNF radionuclide inventory presented 
in Table 4-4 is based on the Case A 63,000 MTU scenario of CSNF for the 1999 forecast 
waste streams (CRWMS M&O 2000 [DIRS 138239]). The weighted average burnups were 
calculated using the three 1999 waste stream cases (A to C), and the four 2002 waste stream 
cases (A' to D'). These values were then compared to the nominal base case (Case A) in order to 
evaluate uncertainty associated with the inventory. The data are presented in Section 6.6.1; the 
calculations are documented in Appendix II and Appendix III CD files (687 kb workbook: 
inv_rev1.xls). The description of these files and the flow of data files or information are 
presented in Appendix II. 

Table 4-5. 63,000 MTU Waste Package Burnup Characteristics Based on 1999 CSNF Waste Streams 

Case 
63,000 MTU 
Waste Package 
Number of 
Packages 
Average Initial 
Burnup (GWd/MTU) 
21-PWR AP 4,299 41.485 
21-PWR CR 95 19.624 
A 12-PWR AP 163 46.326 
44-BWR 2,831 34.08 
24-BWR 84 8.065 
21-PWR AP 4,301 42.887 
21-PWR CR 93 19.544 
B 12-PWR AP 163 48.302 
44-BWR 2,836 35.279 
24-BWR 84 8.042 
21-PWR AP 4,298 40.536 
21-PWR CR 101 19.547 
C 12-PWR AP 163 39.67 
44-BWR 2,808 32.448 
24-BWR 97 8.466 
Table 4-6. CSNF Assembly Totals and Burnup Characteristics for the 63,000 MTHM Cases Based on 
2002 CSNF Waste Streams 



Source: CRWMS 2000 [DIRS 138239], Tables 10, 12, and 14. 

Average Burnup 
(GWd/MTH) Number of Assemblies 
Case BWR PWR BWR PWR 
A' 32 39 124,848 94,344 
B' 42 46 127,303 94,373 
C' 45 49 127,700 94,405 
D' 47 51 127,772 93,252 

Source: BSC 2003 [DIRS 165990], Tables 2 and 4. 


The uncertainty associated with CSNF inventory projections includes the uncertainty related to 
measured versus calculated burn-up for BWR and PWR fuel assemblies (Massie 2004 
[DIRS 170651], Summary Sheet) and the manner in which this affects the radionuclide 
inventory. 

Equation 4-1 evaluates the percent difference of calculated to measured burnup (P): 

P = 100 (calculated burnup - measured burnup)/(measured burnup) (Eq. 4-1) 

Table 4-7 presents additional data used to evaluate inventory projections. These data pertain to 
biases and uncertainties between calculated and measured isotopic concentrations. These data are 
used in Sec. 6.6.1 to develop the uncertainty in isotopic predictions. 

Table 4-7. Summary of Percent Differences Between Calculated and Computed Compositions 

PWR a BWR b 
Isotope c Element Ave. (%)d Isotope c Element Ave. (%)d 
243 Am -4.3 241 Am 4.1 
135 Cs 5.1 
137 Cs 1.3 137 Cs 1.5 
238 Pu 2.6 238 Pu -7 
239 Pu -2.1 239 Pu -2.1 
240 Pu -0.2 240 Pu -0.9 
241 Pu -1 241 Pu -4.5 
242 Pu 3.1 242 Pu 0.5 
90 Sr 8.9 
99 Tc 12.1 
234 U 6.1 234 U -0.2 
235 U 0.1 235 U -2 
236 U -2.8 236 U -1.2 
238 U 0.1 238 U -0.1 

NOTES: a DeHart and Hermann 1996 [DIRS 156084], Table 19 

b 

Hermann and DeHart 1998 [DIRS 106563], Table 24 

c 

Only isotopes important to TSPA-LA see Section 6.1 

d

 (Calculated/measured 1)  100%, average of 4 to 30 samples. 
This value is dependent on the cross section and fission yield 
libraries used. These values were calculated using the 
44GROUPNDF5 cross-section and fission yield libraries. 

Table 4-8 provides the summary of curies for the 32 radionuclides of interest for DSNF 
(DOE 2003 [DIRS 163377] Section 8). Similar to the mathematical operations described for the 
CSNF data, the DSNF inventory (in curies) for the radionuclides of interest is converted to grams 
for the nominal and bounding fuel inventories for each canister or assembly type. The total 
grams of each radionuclide for the total number of canisters or assemblies that comprise DSNF is 
then derived. These data are also used to develop a base case for DSNF that contains all 
wastes except the bare assemblies (see DSNF assumptions in Sec. 5.2). The uncertainty is 
computed from the ratio of the maximum to the nominal value for each isotope. Analyses of the 
data are presented in Section 6.6.2 and Appendices II and III. The description of the 
calculations, a list of the files containing data calculated, and an illustration of the flow of data 


and calculations are presented in Appendix II. The data corresponding to these files and 
calculations are documented on Excel worksheets in the Appendix III CD file inv_rev1.xls 
(687kb workbook: inv_rev1.xls). The data in Table 4-8 are used to develop the new 
codisposal configurations to accommodate DSNF shown in Table 6-3. 

Tables 4-10 and 4-11 provide the basis for the radionuclide content of the HLW glass for the 
license application. The curie content at the decay date of 2030 is used to perform the 
mathematical conversions similar to those described for CSNF and DSNF. The radionuclide 
inventory for the various batches in Table 4-9 is used to develop curves that show the 
heterogeneity of this DSNF inventory. Analysis of these data is presented in Sections 6.4 and 

6.6.3. The description of calculations, a list of the data files used in calculations, and a schematic 
illustrating the flow of calculations and data files is presented in Appendix II. The data contained 
in these files for calculations and the results are documented in Appendix III CD files (534kb 
workbook: inv_rev1.xls). 

Table 4-10 provides the summary of the estimated HLW inventory for canisters at the Hanford 
site as of January 1, 2001 for the radionuclides of interest decayed to January 1, 2030. Although137Cs and 90Sr capsules are not called out in the new data for Hanford, their new values bound the 
old values reported in Source Terms for HLW Glass (CRWMS M&O 2000 [DIRS 151947]. 
These values are compared and plotted in inv_rev1.xls, worksheet newHLW. Table 4-11 
provides the HLW canister production estimates for three production cases (program, planning, 
and technology) applied to three alternative canister-fill scenarios. The analysis of these data 
result in values that bound the expected range of HLW canisters to be produced at Hanford's 
Waste Treatment and Immobilization Plant. The data are used to calculate the inventory in grams 
and curies for the radionuclides important to dose at the time of expected emplacement. These 
data are provided so that emplacement of waste in the repository can be planned in light of the 
various waste package configurations designed for the various kinds of nuclear wastes. 

Table 4-11 provides a matrix for the number of canisters possible when three radionuclide 
loadings and three canister-filling cases are considered. Uncertainty factors for the HLW 
radionuclide inventory are developed from these data (Tables 4-9, 4-10, and 4-11) and a 
discussion can be found in Section 6.6.3. 

Table 4-12 provides the half-lives for the 32 radionuclides of interest for TSPA-LA. The isotopic 
composition of the inventory is a key factor in the design, licensing and operation of the 
repository (and related systems). Since this inventory is time dependent, radionuclide half-lives 
are used to evaluate and determine this inventory at the time of emplacement. 


Initial Radionuclide Inventories 

Table 4-8. Nominal and Bounding Radionuclide Inventories for DSNF Canisters 

Total Curies Ci/1810 Canisters Ci/1815 Canisters Ci/2410 Canisters Ci/2415 Canisters 
Ci/High-Integrity 
Canisters 
Ci/Multicanister 
Overpack Canisters 
Ci/Bare Fuel 
Canisters 
Radionuclide Nominal Bounding Nominal Bounding Nominal Bounding Nominal Bounding Nominal Bounding Nominal Bounding Nominal Bounding Nominal Bounding 
227Ac 2.96E+02 3.52E+02 3.31E+00 7.53E+00 2.64E+02 2.78E+02 1.95E-03 3.89E-03 2.74E+01 4.84E+01 1.18E+00 1.75E+01 4.89E-03 5.94E-03 4.49E-02 8.96E-02 
241Am 2.81E+06 4.01E+06 1.40E+06 1.91E+06 5.08E+05 7.93E+05 2.45E+03 4.90E+03 6.85E+01 1.21E+02 3.26E+04 3.68E+04 5.66E+05 6.96E+05 3.00E+05 5.72E+05 
243Am 5.44E+03 8.47E+03 2.60E+03 3.61E+03 1.31E+03 2.20E+03 1.44E+00 2.88E+00 8.77E-02 1.55E-01 8.98E+01 1.08E+02 5.05E+02 8.07E+02 9.29E+02 1.74E+03 
14C 2.92E+04 3.85E+04 2.26E+04 2.91E+04 3.97E+03 4.44E+03 5.53E-03 1.11E-02 2.60E+01 4.60E+01 2.40E+01 4.10E+01 4.86E+02 5.68E+02 2.13E+03 4.26E+03 
36Cl 5.09E+02 6.82E+02 4.18E+02 5.42E+02 5.00E+01 5.77E+01 1.27E-26 2.54E-26 5.10E-01 9.01E-01 3.43E-01 6.73E-01 3.48E-01 6.96E-01 4.02E+01 8.03E+01 
245Cm 7.43E+01 1.33E+02 9.20E+00 1.47E+01 3.34E+01 5.99E+01 2.96E-03 5.92E-03 1.96E-03 3.46E-03 2.25E+00 2.82E+00 8.53E+00 1.71E+01 2.10E+01 3.82E+01 
135Cs 4.00E+02 5.77E+02 9.74E+01 1.47E+02 1.96E+02 2.76E+02 3.34E+00 6.68E+00 8.06E+00 1.42E+01 1.42E+01 1.95E+01 5.73E+01 7.02E+01 2.29E+01 4.32E+01 
137Cs 2.98E+07 4.77E+07 1.07E+07 1.93E+07 1.06E+07 1.55E+07 1.39E+06 2.79E+06 4.16E+05 7.35E+05 4.93E+05 7.80E+05 4.71E+06 5.70E+06 1.49E+06 2.79E+06 
129I 2.23E+01 3.28E+01 5.20E+00 8.83E+00 9.51E+00 1.30E+01 7.30E-01 1.46E+00 4.46E-01 7.89E-01 4.44E-01 7.40E-01 4.78E+00 5.78E+00 1.21E+00 2.24E+00 
237Np 2.08E+02 3.22E+02 7.64E+01 1.29E+02 5.40E+01 8.29E+01 9.28E+00 1.86E+01 3.55E-02 6.28E-02 2.13E+00 2.52E+00 5.09E+01 6.21E+01 1.49E+01 2.75E+01 
231Pa 3.41E+02 4.08E+02 3.77E+00 8.57E+00 3.01E+02 3.18E+02 3.82E-03 7.63E-03 3.38E+01 5.98E+01 1.46E+00 2.17E+01 1.04E-02 1.23E-02 5.17E-02 1.03E-01 
210Pb 8.57E-02 9.90E-02 1.65E-02 2.02E-02 6.33E-02 6.60E-02 3.21E-04 6.42E-04 4.01E-03 7.09E-03 1.86E-04 2.58E-03 4.57E-04 5.81E-04 9.51E-04 1.87E-03 
238Pu 8.20E+05 1.37E+06 2.57E+05 4.75E+05 3.19E+05 5.10E+05 1.57E+04 3.14E+04 1.20E+02 2.11E+02 9.29E+03 1.15E+04 1.19E+05 1.59E+05 9.96E+04 1.80E+05 
239Pu 4.80E+05 5.86E+05 1.15E+05 1.31E+05 1.93E+05 2.44E+05 4.15E+02 8.30E+02 7.75E+00 1.37E+01 3.16E+03 8.30E+03 1.51E+05 1.75E+05 1.70E+04 2.77E+04 
240Pu 3.50E+05 4.54E+05 6.08E+04 8.66E+04 1.42E+05 1.80E+05 2.36E+02 4.72E+02 4.56E+00 8.05E+00 7.15E+03 7.60E+03 1.22E+05 1.44E+05 1.78E+04 3.59E+04 
241Pu 1.08E+07 2.33E+07 3.78E+06 1.04E+07 4.55E+06 7.41E+06 1.57E+04 3.15E+04 4.14E+02 7.31E+02 7.96E+04 2.10E+05 1.69E+06 2.08E+06 6.70E+05 3.15E+06 
242Pu 4.96E+02 7.26E+02 1.51E+02 1.88E+02 1.38E+02 2.34E+02 3.52E-01 7.04E-01 1.15E-02 2.03E-02 8.83E+00 1.07E+01 1.01E+02 1.39E+02 9.65E+01 1.53E+02 
226Ra 1.54E-01 1.78E-01 3.54E-02 4.36E-02 1.08E-01 1.12E-01 8.74E-04 1.75E-03 6.03E-03 1.07E-02 2.96E-04 3.91E-03 1.34E-03 1.66E-03 2.18E-03 4.29E-03 
228Ra 1.39E+01 1.68E+01 2.19E-01 4.29E-01 1.23E+01 1.32E+01 3.01E-08 6.01E-08 1.30E+00 2.30E+00 5.55E-02 8.33E-01 2.40E-06 4.71E-06 1.04E-02 2.07E-02 
79Se 3.68E+02 5.32E+02 9.21E+01 1.54E+02 1.71E+02 2.20E+02 1.25E+01 2.51E+01 9.97E+00 1.76E+01 5.17E+00 1.15E+01 6.06E+01 7.32E+01 1.63E+01 3.03E+01 
126Sn 3.88E+02 6.00E+02 8.62E+01 1.47E+02 2.25E+02 3.14E+02 1.12E+01 2.24E+01 1.12E+01 1.98E+01 1.36E+01 2.10E+01 1.09E+01 2.19E+01 2.92E+01 5.36E+01 
90Sr 2.52E+07 4.04E+07 9.95E+06 1.80E+07 8.90E+06 1.25E+07 1.31E+06 2.61E+06 4.20E+05 7.43E+05 3.07E+05 5.84E+05 3.29E+06 3.96E+06 1.05E+06 1.97E+06 
99Tc 9.53E+03 1.44E+04 2.98E+03 5.00E+03 3.34E+03 4.73E+03 4.09E+02 8.19E+02 9.16E+01 1.62E+02 1.59E+02 2.26E+02 2.02E+03 2.43E+03 5.31E+02 9.87E+02 
229Th 2.33E+02 2.74E+02 2.60E+00 5.88E+00 2.11E+02 2.24E+02 1.20E-05 2.41E-05 1.82E+01 3.21E+01 7.74E-01 1.16E+01 8.36E-05 1.53E-04 3.91E-02 7.81E-02 
230Th 8.32E+00 9.90E+00 2.32E+00 2.93E+00 5.28E+00 5.55E+00 8.09E-02 1.62E-01 3.22E-01 5.69E-01 1.70E-02 2.10E-01 1.40E-01 1.69E-01 1.57E-01 3.08E-01 
232Th 8.04E+00 8.40E+00 5.98E-01 6.31E-01 3.48E+00 3.77E+00 3.72E-08 7.44E-08 3.60E+00 3.62E+00 3.55E-01 3.56E-01 2.43E-06 4.74E-06 1.04E-02 2.08E-02 
232U 9.52E+04 1.16E+05 9.30E+02 2.20E+03 8.31E+04 8.84E+04 1.06E-01 2.12E-01 1.06E+04 1.88E+04 4.53E+02 6.81E+03 1.78E-01 3.56E-01 3.99E+00 7.91E+00 
233U 1.74E+04 2.16E+04 1.85E+03 1.96E+03 3.18E+03 6.22E+03 3.52E-03 7.03E-03 1.12E+04 1.22E+04 1.16E+03 1.20E+03 1.92E-02 3.14E-02 6.25E+00 1.25E+01 
234U 1.01E+04 1.26E+04 3.90E+03 5.19E+03 5.18E+03 5.52E+03 1.80E+02 3.60E+02 2.30E+02 4.07E+02 1.70E+01 1.56E+02 3.59E+02 4.26E+02 2.74E+02 5.38E+02 
235U 1.93E+02 2.66E+02 1.09E+02 1.51E+02 2.20E+01 3.15E+01 4.64E+00 7.28E+00 1.88E-02 3.13E-02 1.77E-01 3.83E-01 4.57E+01 5.25E+01 1.17E+01 2.31E+01 
236U 2.83E+02 4.19E+02 1.11E+02 1.82E+02 7.84E+01 1.06E+02 1.50E+01 3.00E+01 3.74E-02 6.60E-02 2.40E+00 2.67E+00 6.39E+01 7.51E+01 1.21E+01 2.28E+01 
238U 7.89E+02 8.00E+02 2.69E+01 3.34E+01 4.20E+01 4.30E+01 8.14E-02 8.55E-02 1.51E-03 1.59E-03 4.32E-01 4.95E-01 7.03E+02 7.05E+02 1.62E+01 1.85E+01 
Number of 
Canisters 1402.9 1446.1 165.3 27.0 162.4 403.0 166 

Source: First eight worksheets of Total 2030 D568585.xls, DOE 2003 [DIRS 163377]. 
NOTES: Only radionuclides of interest to TSPA with respect to dose are shown (see Section 6.1). Row for 245Cm hidden in source.


ANL-WIS-MD-000020 REV 01 4-9 September 2004 


Initial Radionuclide Inventories 

INTENTIONALLY LEFT BLANK 


ANL-WIS-MD-000020 REV 01 4-10 September 2004 


Table 4-9. New Estimated Curies Per Canister for Savannah River Site (SRS) HLW (Batches 1A, 1B, 2 
and 3, and 4 to 10) 

Radionuclide 
Batch (Curies) 
1Aa 1Bb 2 and 3c 4 to 10d 
227Ac 6.4196E-08 4.6920E-08 8.5283E-08 6.8448E-08 
241Am 4.1273E+00 4.5983E+00 2.2002E+01 7.1435E+01 
243Am 2.3323E-02 3.5501E-02 3.6707E-01 4.2980E-09 
14C     
36Cl     
245Cm 3.4999E-04  2.6618E-03 6.1466E-03 
135Cs 3.2400E-04 5.1600E-04 7.4399E-04 1.2400E-01 
137Cs 1.7272E+01 4.5063E+01 1.0212E+02 1.7020E+04 
129I  7.2900E-05 5.2200E-06  
237Np 8.6761E-03 1.0542E-02 9.3651E-03 2.1704E-02 
231Pa 1.6097E-07 1.3325E-07 2.5896E-07 2.0920E-07 
210Pb 1.1539E-08 4.4721E-08 8.1881E-09 7.7802E-09 
238Pu 2.7255E+01 4.6471E+01 2.1417E+01 5.4948E+02 
239Pu 4.1858E+00 3.3772E+00 5.2864E+00 1.1291E+01 
240Pu 1.1210E+00 1.1747E+00 1.6894E+00 5.3319E+00 
241Pu 3.6146E+00 6.8403E+00 5.2572E+00 9.6093E+01 
242Pu 9.8616E-04 1.9614E-03 3.1613E-03 1.1502E-02 
226Ra 4.1214E-08 1.4139E-07 3.5806E-08 3.8956E-08 
228Ra 6.9194E-13 1.1175E-04 2.4904E-05 9.0666E-04 
79Se 6.7775E-03 5.8681E-02 4.3088E-02  
126Sn 4.0090E-03 2.3095E-02 3.0894E-02  
90Sr 1.6010E+02 1.2159E+03 1.6155E+03 1.4104E+04 
99Tc 1.3598E-01 1.2399E-01 8.6792E-02 7.8193E+00 
229Th 1.3297E-04 1.7108E-04 1.8785E-05 8.3143E-05 
230Th 5.5935E-06 1.4574E-05 6.2142E-06 8.6012E-06 
232Th 9.0127E-13 1.1500E-04 2.5900E-05 9.4400E-04 
232U    1.2826E-04 
233U 1.5999E-02 3.0197E-02 7.3702E-03 3.2899E-02 
234U 1.9137E-02 2.7487E-02 2.6428E-02 5.8085E-02 
235U 2.1615E-04 2.1310E-04 4.5314E-04 3.6930E-04 
236U 5.1916E-04 7.3003E-04 6.5233E-04 1.5442E-03 
238U 6.7800E-03 5.3300E-03 1.7400E-02 1.6500E-02 
Total Canisters 495 726 705 3,134 

Source: aAllison 2004 [DIRS 168734], Appendix 1 (i.e., Fowler report), Table 5. 
bAllison 2004 [DIRS 168734], Appendix 1 (i.e., Fowler report), Table 6. 
cAllison 2004 [DIRS 168734], Appendix 1 (i.e., Fowler report), Table 7. 
dAllison 2004 [DIRS 168734], Appendix 1 (i.e., Fowler report), Table 8. 


Table 4-10. Estimated Radioactivity in the Hanford HLW Canisters at Decay Date of 2030 

Isotope Element 
Estimated 
Radiochemical 
Inventory in HLW 
Canisters, Basis 
January 1, 2030 
(Curies) Isotope Element 
Estimated 
Radiochemical 
Inventory in HLW 
Canisters, Basis 
January 1, 2030 
(Curies) 
227 Ac 52 240 Pu 12,262 
241 Am 138,408 241 Pu 30,931 
243 Am 15 242 Pu 1 
242 Cm 0 226 Ra 0.09 
243 Cm 7 228 Ra 2 
244 Cm 98 106 Ru 0 
60 Co 179 125 Sb 15 
134 Cs 1 79 Se 122 
137 Cs 23,464,803 151 Sm 2,797,968 
152 Eu 369 126 Sn 579 
154 Eu 28,084 90 Sr 25,114,788 
155 Eu 1 99 Tc 29,697 
129 I 48 229 Th 2 
137 mBaa 23,464,803 232 Th 8 
113 mCda 3,867 232 U 33 
93 mNba 1,138 233 U 510 
59 Ni 1,370 234 U 220 
63 Ni 104,703 235 U 9 
237 Np 141 236 U 6 
231 Pa 272 238 U 199 
238 Pu 3,880 90 Y 25,114,788 
239 Pu 69,042 93 Zr 4,810 
Table 4-11. Hanford HLW Canister Production Estimates for Alternative Canister Wastes Loading and 
Canister Fill Levels 



Source: Hamel 2003 [DIRS 164947], Table 2. 
a The m stands for metastable and is in the source table as shown. 

Canister Fill 
Scenario 
Canister 
(percent fill) 
Canister Glass 
Volume (m3) 
Program Case 
(assumes 
27.7 wt % waste 
loading) 
Planning Case 
(assumes 
36.3 wt % waste 
loading) 
Technology Case 
(assumes 
45 wt % waste 
loading) 
Canisters 
Maximum 100 1.19 11,484 8,744 7,071 
Contract 95 1.14 12,085 9,202 canisters 7,442 
Minimum 87 1.04 13,205 10,054 8,131 

Source: Hamel 2003 [DIRS 164947], p. 1-9, Tables 1 to 3. 


Table 4-12. Radionuclide Half-Lives 

Radionuclide Half-Life (years) Data Source 
227Ac 2.18E+01 Parrington et al. 1996 [DIRS 103896] 
241Am 4.33E+02 Parrington et al. 1996 [DIRS 103896] 
243Am 7.37E+03 Parrington et al. 1996 [DIRS 103896] 
14C 5.72E+03 Parrington et al. 1996 [DIRS 103896] 
36Cl 3.01E+05 Parrington et al. 1996 [DIRS 103896] 
245Cm 8.50E+03 Parrington et al. 1996 [DIRS 103896] 
135Cs 2.30E+06 Parrington et al. 1996 [DIRS 103896] 
137Cs 3.01E+01 Parrington et al. 1996 [DIRS 103896] 
129I 1.57E+07 Parrington et al. 1996 [DIRS 103896] 
237Np 2.14E+06 Parrington et al. 1996 [DIRS 103896] 
231Pa 3.28E+04 Parrington et al. 1996 [DIRS 103896] 
210Pb 2.26E+01 Parrington et al. 1996 [DIRS 103896] 
238Pu 8.77E+01 Parrington et al. 1996 [DIRS 103896] 
239Pu 2.41E+04 Parrington et al. 1996 [DIRS 103896] 
240Pu 6.56E+03 Parrington et al. 1996 [DIRS 103896] 
241Pu 1.44E+01 Parrington et al. 1996 [DIRS 103896] 
242Pu 3.75E+05 Parrington et al. 1996 [DIRS 103896] 
226Ra 1.60E+03 Parrington et al. 1996 [DIRS 103896] 
228Ra 5.76E+00 Parrington et al. 1996 [DIRS 103896] 
79Se 2.95E+05 Singh 2002 [DIRS 164741] 
126Sn 2.50E+05 Parrington et al. 1996 [DIRS 103896] 
90Sr 2.88E+01 Parrington et al. 1996 [DIRS 103896] 
99Tc 2.13E+05 Parrington et al. 1996 [DIRS 103896] 
229Th 7.30E+03 Parrington et al. 1996 [DIRS 103896] 
230Th 7.54E+04 Parrington et al. 1996 [DIRS 103896] 
232Th 1.40E+10 
Decimal point left out in Parrington et al. 1996 
[DIRS 103896]. Value confirmed by Lide 1991 
[DIRS 131202], pp. 11 to 122. 
232U 6.98E+01 Parrington et al. 1996 [DIRS 103896] 
233U 1.59E+05 Parrington et al. 1996 [DIRS 103896] 
234U 2.46E+05 Parrington et al. 1996 [DIRS 103896] 
235U 7.04E+08 Parrington et al. 1996 [DIRS 103896] 
236U 2.34E+07 Parrington et al. 1996 [DIRS 103896] 
238U 4.47E+09 Parrington et al. 1996 [DIRS 103896] 

Radionuclide Screening (BSC 2002 [DIRS 160059]) considered two postclosure periods: the 
10,000-year regulatory and the million-year FEIS and four release scenarios: (1) nominal, 

(2) human intrusion, (3) intrusive igneous event, and (4) eruptive igneous event. The screening 
analysis considered the two types of CSNF: spent boiling water reactor (BWR) fuel and spent 
pressurized water reactor (PWR) fuel, the U.S. Department of Energy spent nuclear fuel (DSNF), 
and high-level radioactive waste (HLW). Average and outlying (that is, high burnup, high initial 
enrichment, low age, or otherwise exceptional) forms of each waste-form type were considered. 
Table 4-13 lists the radionuclides identified in Table 13 of Radionuclide Screening (BSC 2002 
[DIRS 160059]) as those that may substantially affect repository performance for the exposure 


scenarios and times listed and should be considered in the total system performance assessment 
modeling for license application. It was also noted that extending the screening from 
10,000 years to 20,000 years adds no radionuclides to the list for the 10,000-year regulatory 
period (BSC 2002 [DIRS 160059], p. 38 and Table 10). 

The purpose of the screening analysis was to identify all of the radionuclides of major 
importance to dose and virtually all of the marginally important radionuclides. Further analysis 
determined that precursors to screened-in radionuclides (Table 4-14) should be considered. 

Table 4-13. Results of the Screening Analysis 

Radionuclide 
Nominal, Human-Intrusion, and Intrusive 
Igneous Scenarios Eruptive Igneous Scenario 
100 to 
20,000 years 
20,000 to 
1 million years 
100 to 
20,000 years 
20,000 to 
1 million years 
227Ac 227Ac 227Ac 227Ac 227Ac 
241Am 241Am 241Am 
243Am 243Am 243Am 243Am 243Am 
14C 14C 14C 
36Cl 36Cl 
135Cs 135Cs 135Cs 
137Cs 137Cs 137Cs 
129I 129I 129I 
237Np 237Np 237Np 237Np 
231Pa 231Pa 231Pa 231Pa 
210Pb 210Pb 210Pb 
238Pu 238Pu 238Pu 
239Pu 239Pu 239Pu 239Pu 239Pu 
240Pu 240Pu 240Pu 240Pu 240Pu 
242Pu 242Pu 242Pu 
226Ra 226Ra 226Ra 226Ra 
79Se 79Se 
126Sn 126Sn 126Sn 
90Sr 90Sr 90Sr 
99Tc 99Tc 99Tc 99Tc 
229Th 229Th 229Th 229Th 229Th 
230Th 230Th 230Th 
232Th 232Th 232Th 
232U 232U 232U 
233U 233U 233U 233U 233U 
234U 234U 234U 234U 234U 
236U 236U 236U 
238U 238U 238U 238U 
Counts 20 23 12 18 

Source: BSC 2002 [DIRS 160059], Table 13. 

NOTES: Extending the screening for the regulatory period to include the 20,000-year screening time adds 
no radionuclides to the screening lists (BSC 2002 [DIRS 160059], Section 6.3.2). See Section 6.1 
for a discussion of additional radionuclides that are added to the radionuclide inventory. 

ANL-WIS-MD-000020 REV 01 4-14 September 2004 


Table 4-14. Additional Radionuclides Needed for Accurate Accounting of Those Screened In for 10,000 
Years 

Nuclide 
Examined Discussion 
Additional 
Nuclides 
227Ac Daughter of 231Pa, which is screened in (see Table Note).  
241Am 245Cm . 241Pu . 241Am. 245Cm has a half-life much greater than that of 241Am and can 
provide a source of 241Am. Toward the end of the regulatory period, 241Am is in secular 
equilibrium with 245Cm. Also, 245Bk . 241Am, but 245Bk does not appear in the waste 
forms used for the screening analysis. 
241Pu 245Cm 
243Am 247Bk . 243Am, but 247Bk does not appear in the waste forms used for the screening 
analysis. Also, 247Cm . 243Pu . 243Am. 243Pu has a half-life less than one year and 
need not be included in the inventory because it is merely serving as a conduit for the 
decay of 247Cm to 243Am. 247Cm has a half-life of more than 15 million years and would, 
therefore, provide a nearly constant source of 243Am throughout the period of this 
analysis. However, 247Cm does not appear in the waste forms used for the screening 
analysis except in DSNF, where its activity is a very small fraction of the 243Am activity 
throughout the period of analysis; therefore, 247Cm need not be included in the 
inventory. 
 
14C Activation product; not produced by the decay of anything in the waste forms used for 
the screening analysis. 
 
36Cl Activation product; not produced by the decay of anything in the waste forms used for 
the screening analysis. 
 
135Cs Fission product; not produced by the decay of anything in the waste forms used for the 
screening analysis 
 
137Cs Fission product; not produced by the decay of anything in the waste forms used for the 
screening analysis 
 
129I Fission product; not produced by the decay of anything in the waste forms used for the 
screening analysis 
 
237Np Decay product of 241Am, which is screened in. Also produced by decay of 237U, which 
is produced by a minor branch of the decay of 241Pu. 237U has a half-life less than one 
year and need not be included in the initial inventory because it is merely serving as a 
conduit for the decay of 241Pu to 237Np. 241Pu is listed above as needed for the 
inventory. 
 
231Pa 235U . 231Th . 231Pa. 231Th has a half-life less than one year and need not be 
included in the initial inventory because it is merely serving as a conduit for the decay of 
235U to 231Pa. 235U is present in the waste forms used for the screening analysis for all 
waste forms and its inventory is needed to accurately project the inventory of 231Pa, 
especially for times beyond the regulatory period. 
235U 
210Pb 226Ra . 210Pb through a series of short-lived radionuclides that need not be included in 
the inventory. 226Ra is screened in wherever 210Pb is screened in. 
 
238Pu 242mAm . 242Am . 242Cm . 238Pu. 242Cm and 242Am have half-lives less than one 
year and need not be included in the inventory because they are merely serving as a 
conduit for the decay of 242mAm to 238Pu. Also 242mAm . 238Np . 238Pu. 238Np has a 
half-life of less than one year and need not be included in the inventory. 242mAm has a 
longer half-life than 238Pu, so could conceivably provide a source of 238Pu worth 
tracking in TSPA-LA. However, although 242mAm shows up in BWR, DSNF, HLW, and 
PWR waste forms, in each case, the 242mAm activity is negligible compared to that of 
238Pu. 
 
239Pu 243Am . 239Np . 239Pu. 239Np has a half-life less than one year and need not be 
included in the inventory. 243Am is screened- n. Also, 243Cm . 239Pu. 243Cm appears 
in BWR, DSNF, HLW, and PWR waste forms, but in each case, the 243Cm activity is 
negligible compared to that of 239Pu. 
 


Table 4-14. Additional Radionuclides Needed for Accurate Accounting of Those Screened In for 10,000 
Years (Continued) 

Nuclide 
Examined Discussion 
Additional 
Nuclides 
240Pu 244Cm . 240Pu. 244Cm has a half-life much shorter than that of 240Pu. For that reason, 
its initial activity would have to be much greater than that of 240Pu to significantly affect 
the activity of 240Pu. This claim can be verified as follows: A necessary condition for 
the inventory of the parent radionuclide to significantly affect that of the daughter is that 
the number of atoms in initial inventory for the parent Np be at least comparable to that 
of the daughter, Nd. That is, Np . Nd or Np > Nd. In this case, because the half-life of 
the parent is much less than that of the daughter, the relationship between the decay 
constants is .p >> .d. Given the required relationship of the numbers of atoms Np and 
Nd, the necessary condition stated in terms of activities (A=N.) is Np.p >> Nd.d. As it 
happens, the necessary condition is not met because the initial 244Cm activity in the 
waste forms that were used for the screening analysis is less than the initial 240Pu 
activity. Therefore, 244Cm need not be included in the inventory. 
 
242Pu 246Cm . 242Pu. 246Cm has a half-life much shorter than that of 242Pu. For that reason, 
its activity would have to be much higher than that of 242Pu to significantly affect the 
activity of 242Pu (See the discussion for 240Pu for a justification of this claim). As it 
happens, the initial 246Cm activity in the waste forms that were used for the screening 
analysis is less than the initial 242Pu activity. Therefore, 246Cm need not be included in 
the inventory. 
 
226Ra Decay product of 230Th, which is screened in  
228Ra Decay product of 232Th, which is screened in  
79Se Fission product; not produced by the decay of anything in the waste forms used for the 
screening analysis 
 
126Sn Fission product; not produced by the decay of anything in the waste forms used for the 
screening analysis 
 
90Sr Fission product; not produced by the decay of anything in the waste forms used for the 
screening analysis 
 
99Tc Fission product; not produced by the decay of anything in the waste forms used for the 
screening analysis 
 
229Th Decay product of 233U, which is screened in  
230Th Decay product of 234U, which is screened in  
232Th Decay product of 236U, which is screened in  
232U Decay product of 236Pu, which has a half-life of about 3 years. With such a short half-
life, by the time the waste is received at the repository, most of the 236Pu will have 
decayed to 232U. In any case, 236Pu has a half-life much shorter than that of 232U. For 
that reason, its activity would have to be much higher than that of 232U to significantly 
affect the activity of 232U (See the discussion for 240Pu for a justification of this claim). 
As it happens, the initial 236Pu activity in the waste forms that were used for the 
screening analysis is less than the initial 232U activity. Therefore, 236Pu need not be 
included in the inventory. 
 
233U 237Np . 233Pa . 233U. 237Np is screened in. 233Pa has a half-life less than one year 
and need not be included in the inventory. 
 
234U 238U . 234Th . 234Pa . 234U. 238U is screened in. 234Pa and 234Th have half-lives less 
than 1 year and need not be included in the inventory. 
 
236U Decay product of 240Pu, which is screened in  
238U Decay product of 242Pu, which is screened in  

Source: BSC 2002 [DIRS 160059], Table 12. 

NOTE: Radionuclide Screening (BSC 2002 [DIRS 160059], p. 38) states that the radionuclides shown in the far 
right column should be included for the inventory either by direct inclusion or by appropriate augmentation 
of the daughter product. 

See Parrington et al. 1996 [DIRS 103896] for decay relationships and half-lives. 

ANL-WIS-MD-000020 REV 01 4-16 September 2004 


Table 4-15 presents the nuclear reactions and half-lives of those elements that make up the 
composition of BWR and PWR assembly hardware that provide the basis for identifying 
activation products. 

Table 4-15. Identification and Screening Status of CSNF PWR and BWR Hardware-Activation Products 

Element 
Nuclear Reactions Considered 
(Half-Lives in Parentheses) a 
Al 27Al (n, .) 28Al (2 min) . 28Si (stable) 
B 11B (n, .) 12B(20 ms) . 12C (stable) 
C 13C(n, .) 14C (5.7  103 yr) 
Co 59Co (n, .) 60Co (5.3 yr) 
Cr 50Cr (n, .) 51Cr (27 days) . 51V(stable) 
54Cr(n, .) 55Cr (3 min) . 55Mn (stable) 
Cu 63Cu(n, .) 64Cu(13 hr) . 64Zn (stable), 64Ni (stable) 
65Cu (n, .) 66Cu (5 min) . 66Zn(stable) 
Fe 54Fe (n, .) 55Fe (2.7 yr) . 55Mn (stable) 
54Fe (n, a) 51Cr (27 days) . 51V (stable) 
58Fe (n, .) 59Fe (45 days) . 59Co (stable) 
Mn 55Mn (n, .) 56Mn (2.6 hr) . 56Fe (stable) 
Mo 92Mo (n, .) 93Mo (6.9 hr) . 93Nb (stable) 
98Mo (n, .) 99Mo (2.7 days) . 99Tc (2.1  105 yr) 
100Mo (n, .) 101Mo (15 min) . 101Tc (14 min) . 100Ru (stable) 
N 14N (n, p) 14C (5.7  103 y) 
15N (n, .) 16N (7 sec) . 16O (stable) 
Nb 93Nb (n, .) 94Nb (2.0  104 yr) 
Ni 58Ni (n, .) 59Ni (7.6  104 yr) 
62Ni (n, .) 63Ni (1  102 yr) 
64Ni (n, .) 63Ni (2.5 hr) . 65Cu (stable) 
O 18O (n, .) 19O (27 sec) . 19F (stable) 
17O (n,a) 14C (5.7  103 yr) 
P 31P (n, .) 32P (14 days) . 32S (stable) 
S 34S (n, .) 35S (87 days) . 35Cl (stable) 
36S (n, .) 37S (87 days) . 37Cl (stable) 
Si 30Si (n, .) 31Si (2.6 hr) . 31P (stable) 
Sn 112Sn (n, .) 113Sn (115 days) . 113In (stable) 
120Sn (n, .) 121Sn (1.1 days) . 121Sb (stable) 
122Sn (n, .) 123Sn (129 days) . 123Sb (stable) 
124Sn (n, .) 125Sn (9.6 days) . 125Sb (2.7 yr) 
Ti 50Ti (n, .) 51Ti (5.8 min) . 51V (stable) 
Zr 92Zr (n, .) 93Zr (1.5  106 yr) 
94Zr (n, .) 95Zr (64 days) . 95Nb (35 days) . 95Mo (stable) 
96Zr (n, .) 97Zr (17 hr) . 97Nb (1.2 h) . 97Mo (stable) 

Source: Parrington et al. 1996 [DIRS 103896]. 
NOTES: a Activation of reaction products ignored. 


ANL-WIS-MD-000020 REV 01 4-17 September 2004 


Table 4-16 provides the 14C and 99Tc inventories expected in 5-year-old and 25-year-old PWR 
and BWR assemblies (BSC 2004 [DIRS 171435]; BSC 2004 [DIRS 171436]). These tables are 
used in Appendix I to evaluate neutron-activation products outside the spent-fuel matrix and 
their contribution to the radionuclide inventory. The values are intended to be bounding. 

Table 4-16. 99Tc and 14C Inventories of BWR Assembly at Specified Characteristics 

Average BWR SNF 
Assembly 
Bounding BWR SNF 
Assembly 
Average PWR SNF 
Assembly 
Bounding PWR SNF 
Assembly 
(3.5 %, 40 GWd/MTU, 
25 years) 
(5.0 %, 75 GWd/MTU, 
5 years) 
(4.0%, 48GWd/MTU, 
25 years) (5.0%, 80GWd/MTU, 5 years) 
Fission 
Products Light Elements 
Fission 
Products Light Elements 
Fission 
Products Light Elements 
Fission 
Products Light Elements 
99Tc 14C 99Tc 99Tc 14C 99Tc 99Tc 14C 99Tc 99Tc 14C 99Tc 
Fuel 3.20100 1.4010-1 0.00100 5.35100 2.5210-1 1.3410-5 8.99100 3.3010-1 6.1410-4 1.34101 5.3110-1 9.8310-4 
Total nonfuel 0.00100 3.5410-2 0.00100 0.00100 6.3610-2 0.00100 0.00100 2.2610-3 4.8710-4 0.00100 3.6210-3 7.7410-4 
Bottom 2.4410-2 0.00100 4.3910-2 0.00100 0.00100 2.9110-4 0.00100 4.6310-4 
Plenum 4.1110-3 0.00100 7.3910-3 0.00100 0.00100 3.7810-5 0.00100 6.0410-5 
Top 6.8510-3 0.00100 1.2310-2 0.00100 2.2610-3 1.5810-4 3.6210-3 2.5110-4 
TOTAL 3.20100 1.7510-1 0.00100 5.35100 3.1610-1 1.3410-5 8.99100 3.3210-1 1.1010-3 1.34101 5.3510-1 1.7610-3 

Source: BSC 2004 [DIRS 171435], Table 6; BSC 2004 [DIRS 171436], Table 12. 

NOTE: 0.00x100 is the value presented in the data input source. 

4.2 CRITERIA 

Criteria addressed by Technical Work Plan For: Regulatory Integration Modeling and Analysis 
of the Waste Form and Waste Package (BSC 2004 [DIRS 171583], Section 3) and applicable to 
this analysis are presented in Sections 4.2 and 4.3. 

Section 3.4 of Project Requirements Document (Canori and Leitner 2003 [DIRS 166275]) 
contains requirements relevant to this analysis. The key requirements (referred to by their 
requirement identifier) are: 

1. 
PRD-002/T-014 (Canori and Leitner 2003 [DIRS 166275], p. 3-14) Performance 
Objectives for the Geologic Repository after Permanent Closure (traceable to 
10 CFR 63.113 [DIRS 156605]). 
This section specifies the repository performance objectives that must be met 
following permanent closure. It includes a requirement for multiple barriers and limits 
on radiological exposure. 


2. 
PRD-002/T-015 (Canori and Leitner 2003 [DIRS 166275], p. 3-14) Requirements for 
Performance Assessment (traceable to 10 CFR 63.114 [DIRS 156605]). 


This section specifies the technical requirements to be used in performing a 
performance assessment. It includes requirements for calculations, including data 
related to site geology, hydrology, variability in the models, and deterioration or 
degradation processes, including waste form degradation. 


Work described in this document support these requirements, but more specific criteria exist in 
Section 2.2.1 of Yucca Mountain Review Plan, Final Report (NRC 2003 [DIRS 163274]). 
Selected acceptance criteria are presented to supplement or clarify Project Requirements 
Document (Canori and Leitner 2003 [DIRS 166275]) citation. The following acceptance criteria 
were identified in the technical work plan as applicable to this analysis. Section 7.2 quotes the 
full text of the applicable acceptance criteria with pointers to the information within this report 
that pertains to the criteria. 

Radionuclide Release Rates Acceptance Criteria (NRC 2003 [DIRS 163274], 
Section 2.2.1.3.4.3): 

 
Acceptance Criterion 1System Description and Model Integration are Adequate 
 
Acceptance Criterion 2Data are Sufficient for Model Justification 
 
Acceptance Criterion 3Data Uncertainty is Characterized and Propagated Through the 
Model Abstraction. 


Section 3.3 of Technical Work Plan For: Regulatory Integration Modeling and Analysis of the 
Waste Form and Waste Package (BSC 2004 [DIRS 171583]) addresses other acceptance or 
completion criteria. Technical products are required to address and document the accuracy, 
precision, and representativeness of the work performed as part of the uncertainty analyses. In 
addition, technical products will meet the level of detail and accuracy needed to support the 
TSPA-LA model. 

4.3 CODES, STANDARDS, AND REGULATIONS 

Section 3.1 of Technical Work Plan For: Regulatory Integration Modeling and Analysis of the 
Waste Form and Waste Package (BSC 2004 [DIRS 171583]) states that this work will conform, 
as appropriate, to guidance provided in ASTM C 1174-97 [DIRS 105725], Standard Practice for 
Prediction of the Long-Term Behavior of Materials, Including Waste Forms, Used in Engineered 
Barrier Systems (EBS) for Geological Disposal of High-Level Radioactive Waste. This analysis 
conforms to this guidance in that it conforms to methods used to aid in the prediction of the 
long-term behavior of materials, such as engineered barrier system (EBS) materials and waste 
forms, used in the geologic disposal of high-level nuclear waste in the U.S. Government disposal 
site. 

Section 3.2 of Technical Work Plan For: Regulatory Integration Modeling and Analysis of the 
Waste Form and Waste Package (BSC 2004 [DIRS 171583]) states that 10 CFR Part 63 
[DIRS 156605] is applicable to this work. Certain sections of 10 CFR Part 63 [DIRS 156605] 
and 40 CFR Part 197 [DIRS 165519] are relevant to the use of the results from analyses 
performed as a part of the Total System Performance Assessment for the License Application 
(TSPA-LA). The output from this analysis will be used in the TSPA-LA postclosure analysis. 


INTENTIONALLY LEFT BLANK 



5. ASSUMPTIONS 


There are many assumptions that have been made in the development of the source documents 
used in this analysis. The assumptions relevant to the data used in this analysis are provided here. 

5.1 COMMERCIAL SPENT NUCLEAR FUEL 

The data presented in this analysis that was used to characterize the CSNF radionuclide 
inventory was based on calculations that compiled commercial waste stream information from 
several sources (CRWMS M&O 2000 [DIRS 138239]). In order to derive the number of waste 
packages for each waste package design configuration, the average initial burnup (GWd/MTU) 
and the radionuclide inventory in curies for the year of expected completion of emplacement 
(2033), assumptions were made. Source term information for a Babcock & Wilcox Mark B 
PWR assembly with initial heavy metal loading of 475 kg was chosen (to develop the inputs to 
this analysis) because it has characteristics that are considered generally representative of other 
PWR assemblies. Therefore, it was assumed that the source term values from the Mark B 
assemblies are representative and can be applied to other PWR assemblies of different classes 
and initial uranium loading by a simple ratio of the initial uranium loading. This assumption is 
verified in PWR and BWR Source Term Sensitivity Study (BSC 2004 [DIRS 167058]). 

Similarly, source term information was developed for a General Electric 2/3 BWR assembly with 
an initial heavy metal loading of 200 kg. This assembly was chosen because it has characteristics 
that are representative of other BWR assemblies. It was assumed that source term values from 
this assembly could be applied to other BWR assemblies of different classes and initial uranium 
loading by a simple ratio of the initial uranium loading. This assumption is also verified in PWR 
and BWR Source Term Sensitivity Study (BSC 2004 [DIRS 167058]). 

Assumptions were also made as to the loading of assemblies of different types into the various 
waste package configurations. Although the specific criteria for loading assemblies is not yet 
known, it was assumed that no thermal criteria will be violated so long as the mix of assemblies 
placed into a waste package are below the maximum allowable limit (CRWMS M&O 2000 
[DIRS 138239]). Since it is unknown which assemblies the utilities will send at a given time, 
arrival forecasts were developed based on three different arrival case methodologies to account 
for the most likely scenarios. These were presented in Section 4.1, and Tables 4-5 and 4-6. 

5.2 DOE-OWNED SPENT NUCLEAR FUEL 

The National Spent Fuel Program has noted that complete information is not available for many 
DSNFs. In the absence of some information needed to calculate scaling factors, assumptions that 
tend to err toward a more conservative result were used (DOE 2003 [DIRS 163377], p. 21). 
These assumptions are intended to cause a conservative bias such that the resulting estimate will 
predict a source term with a higher dose. Table 5-1 contains assumptions suggested when 
information is incomplete. 


Table 5-1. Assumptions Related to DSNF 

Unknown 
Parameter Conservative Assumption Basis 
Cladding If cladding is unknown, assume it is 
stainless steel 
Stainless steel is more conducive to the production 
of activation products than that of other typical 
cladding materials (e.g., aluminum, zirconium, 
graphite). 
Fuel compound If end-of-life plutonium exceeds 1% 
by weight, assume a mixed oxide 
fuel. 
If thorium is present at end-of-life, 
assume a U-Th oxide fuel. 
Otherwise, assume a uranium fuel. 
Because the majority of spent nuclear fuels (SNFs) 
are uranium fuels, this is assumed unless 
information provides evidence of other fuel 
compounds. 
BOL enrichment Assume the initial fissile mass equals 
the fissile mass depleted (i.e., 100% 
depletion). 
If needed, the initial uranium 
inventory can be estimated as the 
end-of-life heavy metal mass plus the 
initial fissile mass. 
Estimates the lowest possible enrichment (i.e., will 
underpredict the actual enrichment). 
These correlations assume uranium fuels. Uranium 
fuels comprise the majority of DSNFs. These 
correlations also provide reasonable 
approximations for other fuel types. 
Moderator Heavy water. Heavy water moderation produces a soft neutron 
spectrum that is generally more conducive to 
transmutation of heavy metals 
Reactor shutdown 
or fuel removal 
date 
Date for fuel shipping, storage, or any 
other activity that confirms the fuel 
was no longer in the reactor. 
Use of later date will produce a conservative result 
for all radionuclides of interest except 237Np and241Am because, for a period, they may increase 
rather that decrease. 

Source: DOE 2003 [DIRS 163377], Table 1. 

The CSNF bare fuel that is in DOE's possession is not included in this analysis. Source Term 
Estimates for DOE Spent Nuclear Fuels (DOE 2003 [DIRS 163377]) provides the inventory for 
DSNF. This analysis excludes the part of that inventory associated with bare fuel and assumes 
that the remaining inventory is conservative for TSPA-LA calculations. The basis for this 
assumption is as follows. The repository allotment for DSNF, that also includes naval fuel, is 
2,333 MTHM (Dreyfus 1995 [DIRS 104384]; Lytle 1995 [DIRS 104398]). The repository 
allocation for naval fuel is 65 MTHM, leaving an allotment of 2,268 MTHM for DSNF. The 
DSNF inventory (2,348 metric ton of 238U) excluding the bare fuel inventory (48 metric ton of238U) exceeds the repository allotment for DSNF (2,348 - 48 = 2,300, which is greater than 
2,268). Therefore, using the DSNF radionuclide inventory reported in Source Term Estimates for 
Spent Nuclear Fuels (DOE 2003 [DIRS 163377]), excluding the bare fuel, produces a 
conservative DSNF inventory for TSPA-LA. 

The nonnaval DSNF inventory of 2,348 metric tons of 238U (DOE 2003 [DIRS 163377]) includes 
48 metric tons of 238U of bare fuel assemblies. Bare fuel assemblies do not fit in codisposal waste 
packages. The remaining nonnaval inventory of 2,300 metric tons of 238U are allocated to the 
3,412 codisposal packages. The remaining 7,472 waste packages designated for CSNF are fully 
allocated to PWR and BWR assemblies and are, therefore, not available for bare fuel. 

This analysis distributes the remaining DSNF inventory into codisposal waste packages. 
TSPA-LA models 11,184 waste packages (see Table 4-2), of which 7,472 are CSNF, 300 are 


naval and 3,412 are codisposal (BSC 2004 [DIRS 169472], Table 11). If the codisposal 
packages are fully loaded, the DSNF and HLW glass repository allocations of 2,333 and 4,667 
MTHM respectively, will be exceeded (Dreyfus 1995 [DIRS 104384]); Lytle 1995 
[DIRS 104398]). Although the number of DSNF and HLW canisters implied by the number of 
codisposal waste packages cannot be emplaced in the repository using the existing limits, 
including these canisters within the TSPA results in a conservative inventory. 

As noted in Section 6.6.2, 38 percent of the DSNF radionuclide inventory (Table 4-8) comes 
from 0.31 percent of the total DSNF. The report states that 0.31 percent of the mass relied 
heavily on conservative assumptions to compensate for incomplete information. Uncertainties 
include those related to burnup, BOL (beginning of life) for the fuel, the BOL concentration in 
structural materials, the impurities in structural materials, the uncertainties related to 
methodologies, the statistical uncertainties, as well as measured versus calculated uncertainties. 
All of these attributes are considered and contribute to an overall uncertainty that is applied to 
the nominal and bounding inventory. If we can assume that the 99.69 percent of the fuels with 
sufficient information can represent the 0.31 percent of the fuel with insufficient information, the 
expected value for the radionuclide inventory (curies) for DSNF is more reasonably 62 percent 
of the nominal inventory estimated (i.e., 62 percent + 62 percent  0.31 percent . 
62 percent)(DOE 2003 [DIRS 163377]). This assumption is used in the analysis in 
Section 6.6.2. 

For DSNF, radiological inventories are estimated based on a validated ORIGEN code output 
(i.e., a fuel template) that was developed for that fuel type (DOE 2003 [DIRS 163377]. 
Approximately 10 percent of the DSNF inventory uses a template that, although based on 
another fuel type, shares parameters that dominate the model with respect to generation of 
radionuclides. For about 0.2 percent of the fuels in the inventory, the burnup information is 
uncertain because of missing or incomplete BOL (beginning of life) and burnup values. For 
these fuels, BOL heavy metal is assumed to be two times the EOL (end of life) value, which 
overestimates the burnup for over 96 percent of the DSNF. This assumption produces a very 
conservative estimate of the burnup, which results in an increased scaling of the template 
radionuclide inventories. The remainder of approximately 0.11 percent of the DSNF uses a 
Worst-Case template that was derived by taking the highest normalized (Ci/MTU) values for 
each radionuclide from all the available templates. Although such a fuel does not physically 
exist, this template is used to bound fuel materials in the DSNF inventory for which a template 
cannot be selected (DOE 2004 [DIRS 169354], p. E-18). There are two reasons for this: the 
unavailability of a template that adequately models the fuel and the unavailability of sufficient 
information to identify a proper template. For example, various different fuels and cladding 
types, in which much of the fuel has been destructively examined, and do not conform to any one 
template. Thus the radionuclide inventory for these fuel records are estimated using the worst-
case template, which maximizes the radionuclide production as a function of burnup and 
produces a conservative estimate. Since bounding burnup information is not available to support 
estimation of the inventory, the estimated nominal burnup is used and the bounding value is 
conservatively assumed to be two times (2) the nominal (DOE 2003 [DIRS 163377], p. 38). 
Use of the worst-case template provides a high degree of confidence that, for a given burnup, the 
resulting radionuclide estimate will bound virtually any fuel's actual radionuclide content 
(DOE 2003 [DIRS 163377], Section E-3.1). The reported inventory is derived using data that 


has been measured, reported, calculated, and modeled having applied statistical analysis to 
identify those most likely, as well as those that are bounding. 

5.3 DOE-OWNED HLW (DHLW) 

It is assumed that radionuclide inventory and waste package data from the Savannah River Site 
(SRS) and the Hanford site are sufficient to characterize the inventory of HLW glass inventory 
for TSPA-LA. The SRS data is from Allison (2004 [DIRS 168734], Appendix 1) and the 
Hanford data is from Hamel (2003 [DIRS 164947]). The basis for this assumption is, although 
earlier reports such as Inventory Abstraction (BSC 2001 [DIRS 157575]) and Source Terms for 
HLW Glass Canisters (CRWMS M&O 2000 [DIRS 151947]) analyzed data from Hanford, SRS, 
INEEL, and the West Valley Demonstration Project, further analyses showed that the SRS and 
Hanford glass packages dominated the radionuclide content of the glass inventory. This 
assumption is used in Section 6.6.3. 

As with DSNF and CSNF, assumptions pertaining to the reported inventory are also described 
here. It is noted that the actual number of canisters to be produced at the SRS and their curie 
concentrations are dependent on the waste loading achieved and the efficiency of cleaning. It is 
assumed that the glass compositions and curie content in canisters already produced at the SRS 
provides a suitable modeling basis for future production of SRS canisters. 

It is assumed that waste packages will be fully loaded for TSPA-LA purposes. Note that using 
the historic method for counting MTHM per glass canister (2.3 MTHM per West Valley 
canister and 0.5 MTHM per DHLW canister) (DOE 1985 [DIRS 103492], Table 1-1), the glass 
allocation of 4,667 MTHM (Dreyfus 1995 [DIRS 104384]; Lytle 1995 [DIRS 104398]) would 
equal about 8,329 glass canisters instead of the 16,614 canisters used in TSPA for codisposal. 
The TSPA was calculated assuming the full 16,614 canisters thus providing an extra margin of 
conservatism and preventing the underestimation of risk due to HLW glass. 

An assumption was made in the absence of confirming data to bound the percent waste loading 
of future HLW canisters based on Hanford's Waste Treatment and Immobilization Plant (WTP) 
technology case. Estimating a increase of 7 wt % in waste loading above that specified for the 
technology case and applying it to the 100 percent fill scenario yields an uncertainty factor of 
1.5, which when applied to the HLW inventory derives a maximum value. Since the technology 
case will require additional technology development to demonstrate the glass waste form and 
vitrification technology, it is reasonable to make an assumption to bound the waste for this case. 


6. SCIENTIFIC ANALYSIS DISCUSSION 


This section presents the results of this analysis. A description of the mathematical operations 
performed in this analysis is presented in more detail in Appendices I and II, and in Excel tables 
and graphs on the Appendix III CD. The outputs of this analysis, the initial radionuclide 
inventories (in grams per radionuclide per waste package) and uncertainty distributions 
associated with this inventory are presented in Tables 7-1 and 7-2, and 
DTN: SN0310T0505503.004. The outputs are reasonable compared to the inputs, and the 
results are suitable for the intended use. The uncertainties are taken into account by consistently 
using the most conservative approach; the calculations, therefore, yield a conservatively 
bounding set of results. The directly used inputs to this analysis are listed in Section 4.1. The 
qualification of unqualified data used as direct input is presented in Appendix IV. The following 
inputs are used as corroborating information: 

1. 
Support for the claim that Zircaloy is highly corrosion resistant under repository 
conditions (BSC 2003 [DIRS 168795]). The information from this source is addressed 
in Appendix I.4.3. 
2. 
Superseded report analyzing radionuclide inventory (BSC 2001 [DIRS 157575]). The 
information from this source is addressed in Sections 1, 6, 6.6.3, and Appendices I 
and II for an analysis of neutron-activation products in the crud and BWR or PWR 
hardware. 
3. 
Initial breach fraction of cladding in CSNF (BSC 2003 [DIRS 162153], p. 38). This 
information from this source is addressed in Appendix I.4.1. 
4. 
Fast release fraction of 99Tc (BSC 2004 [DIRS 167321], Table 8.1-1). This 
information from this source is addressed in Appendix I.4.1. 
5. 
No credit was taken for DSNF cladding (BSC 2003 [DIRS 163693], p. 50). This 
information from this source is addressed in Appendix I.4.1. 
6. 
DSNF degradation rate for LA (BSC 2003 [DIRS 163693], p. 65). This information 
from this source is addressed in Appendix I.4.2. 
7. 
Information from Source Terms for HLW Glass Canisters (CRWMS M&O 2000 
[DIRS 151947], Tables 6-1 and 6-2) is discussed in Section 6.6.3 and Appendix II.3. 
8. 
Information from Total System Performance Assessment for the Site Recommendation 
(CRWMS M&O 2000 [DIRS 153246]) is addressed in Appendix I.4.2. 
9. 
Materials specifications for CSNF assembly hardware, specifically stainless steel 
components were used from DOE 1987 [DIRS 132333] to corroborate information 
used by the analysis in Appendix I.4.2. 
10. 
Determination of the Accuracy of Utility Spent-Fuel Burnup Records (EPRI 1999 
[DIRS 164649]) provided a corroborative measure of the uncertainty related to 





measured versus calculated values of burnup CSNF. This information can be found in 
Section 6.6.1. 


11. 
Reactor Record Uncertainty Determination (Massie 2004 [DIRS 170651], Summary 
Sheet) provides 2004 information pertaining to the percent uncertainty related to 
measured versus calculated values of burnup of CSNF BWR and PWR assemblies. 
These data can be found in Section 6.6.1 and Appendix III. 
12. 
Source Term Estimates for DOE Spent Nuclear Fuels (DOE 2004 [DIRS 169354]) 
Revision 1 is used as corroborative input to provide updated DSNF radionuclide 
inventory data published January 2004. These data are compared with the DSNF 
radionuclide inventory data used in this analysis to validate the inputs used and 
validate the uncertainty factors developed. The data used in this analysis to develop 
the inventory comes from Revision 0 of the same report (DOE 2003 [DIRS 163377]), 
published in March 2003. These data can be found in Section 6.6.2 and Appendix III 
Excel workbook file inv_rev1.xls. 
13. 
2002 Waste Stream Projection Report (BSC 2003 [DIRS 165990], p.16 and 17) is the 
source for the definitions of Cases A' to D', an alternative set of methodologies used to 
develop arrival forecasts. Cases A' to D' were compared to Cases A to C from Waste 
Packages and Source Terms for the Commercial 1999 Design Basis Waste Streams 
(CRWMS M&O 2000 [DIRS 138239]), which is also a source of direct input 
(Table 4-1). The data are part of the uncertainty associated with arrival forecasts used 
to develop uncertainty factors for CSNF inventory. These definitions can be found in 
Table 6-6. 


14. 
An illustration from Waste Form Degradation Process Model Report 
(CRWMS M&O 2002 [DIRS 150704], Figure 3.1-1) used to provide an overview of 
waste package types containing different wastes in the inventory (Figure 6-1). 
15. 
An illustration from Yucca Mountain Science and Engineering Report (DOE 2002 
[DIRS 155943], Figure 3-5) is used to provide illustration of some waste package 
configurations for CSNF, DSNF, naval, and HLW glass. This illustration is shown as 
Figure 6-2. 
16. 
PWR and BWR Source Term Sensitivity Study (BSC 2004 [DIRS 167058]) is used to 
corroborate the assumption that the source term values from the Mark B assembly are 
representative and can be applied to other PWR assemblies of different classes and 
initial uranium loadings by a simple ratio of the initial uranium loading. It is also used 
to corroborate the assumption that the source term values from the GE 2/3 assembly 
are representative and can be applied to other BWR assemblies of different classes 
and initial uranium loadings by a simple ratio of the initial uranium loading. 


This analysis is a simple mathematical compilation of lists of radionuclide inventory data 
obtained from documented sources (i.e., addition, subtraction, division, multiplication of existing 
inventory data). No models have been used in this analysis and no alternative methods other 
than the one used were evaluated to provide a cumulative radionuclide inventory for TSPA-LA. 


The fact that the sources used as data input provide the most recent data needed and that these 
data are considered bounding, applying simple mathematical operations on the inventories 
reported for DSNF, CSNF, and HLW projections were considered the only reasonable method. 

Uncertainties applied to data were developed from uncertainties commonly used in the nuclear 
industry. These uncertainties pertain to errors in records (EPRI 1999 [DIRS 164649]), errors in 
measured versus calculated data (Hermann and DeHart 1998 [DIRS 106563]; DeHart and 
Hermann 1996 [DIRS 156084]), uncertainties related to the age and activity levels (burnup) of 
the waste and age-at-arrival using arrival scenarios. Use of known error or uncertainty factors, 
calculation of averages, and developing ratios as factors to establish ranges of values, are the 
methods used in this analysis. Where more recent data has become available, comparisons are 
identified or plotted, or both, to illustrate the validity of calculated uncertainties or show the 
more recent data is bounded by that used in this analysis, or both. Referenced documents have 
applied methodologies to developing uncertainties associated with the radionuclide inventory 
source terms reported and used in this analysis. No other methods of determining the inventory 
or developing the uncertainty were considered in this analysis. 

The sources of data used in this analysis provide qualified input data. Sources where the data 
require qualification in accordance with BSC procedures, the qualification is documented in 
Appendix IV of this analysis. This data qualification is intended to provide the desired level of 
confidence that the data are suitable for their intended use and the intended use is only for this 
analysis. AP-SIII.9Q, Section 5.2.1-l states, The extent to which the data demonstrate the 
properties of interest shall be addressed. One or more of the following factors shall be used when 
presenting the case that the data are suitable for intended use: 

 
Reliability of data source 
 
Qualification of personnel or organizations generating the data 
 
Prior uses of the data 
 
Availability of corroborating data. 


Data from the following sources are qualified for intended use in accordance with the above 
factors in AP-SIII.9Q and found in the Appendix IV as noted: 

IV.1. Projected Glass Composition and Curie Content of Canisters from the Savannah 
River Site (U) (Allison 2004 [DIRS 168734], Appendix 1 (Fowler report, 
Rev.2)). 

IV.2. Waste Treatment and Immobilization Plant (WTP) High-Level Waste (HLW) 
Canister Production Estimates to Support Analyses by the Yucca Mountain 
Project (Hamel 2003 [DIRS 164947]). 

IV.3. Validation of SCALE (SASH2H) Isotopic Predictions for BWR Spent Fuel 
(Hermann and DeHart 1998 [DIRS 106563]); An Extension of the Validation of 
SCALE (SASH2H) Isotopic Predictions for PWR Spent Fuel (DeHart and 
Hermann 1996 [DIRS 156084]). 


IV.4. Source Term Estimates for DOE Spent Nuclear Fuels (DOE 2004 [DIRS 
163377]; Source Term Estimates for DOE Spent Nuclear Fuels (DOE 2004 
[DIRS 169354]). 

For data not derived from external sources, AP-SIII.2Q Qualification of Unqualified Data has 
been used for data qualification. Inputs qualified in accordance with AP-SIII.2Q require a Data 
Qualification Plan and use of one or a combination of methods identified in Attachment 3 of 
AP-SIII.2Q. This plan is provided in Appendix V; the results this data qualification task are 
presented in Section IV.5. The data to be qualified in accordance with AP-SIII.2Q are from the 
following reference sources: 

IV.5 2002 Waste Stream Projections Report (BSC 2003 [DIRS 165990]). Data 
Qualification Plan for this source of data is in Appendix V; the signed copy is in 
the record package for this analysis. 

Assumptions have been incorporated into this analysis and are documented in Section 5. An 
assumption was made in the absence of confirming data to bound the percent waste loading of 
future HLW canisters based on Hanford's Waste Treatment and Immobilization Plant (WTP) 
technology case. Estimating an increase of 7 wt % in waste loading above that specified for the 
technology case and applying it to the 100 percent fill scenario yields an uncertainty factor of 
1.5, which when applied to the HLW inventory derives a maximum value. Because the 
technology case will require additional technology development to demonstrate the glass waste 
form and vitrification technology, it is reasonable to make an assumption to bound the waste for 
this case. 

6.1 RADIONUCLIDES 

The radionuclides of importance to TSPA-LA dose calculations were assessed in Radionuclide 
Screening (BSC 2002 [DIRS 160059]). This screening document separately considered both the 
10,000-year regulatory period and the million-year FEIS period. Four release scenarios were 
considered: (1) the nominal scenario, (2) the human-intrusion scenario, (3) an intrusive igneous 
event, and (4) an eruptive igneous event. These scenarios result in dose from groundwater and 
the atmospheric pathways. Table 4-13 shows the radionuclides that contribute to 95 percent of 
the dose for the different scenarios implemented in the TSPA-LA for the regulatory period of 
10,000 years (columns 2 and 4), and those added for the million-year period (columns 3 and 5) 
(BSC 2002 [DIRS 160059]). Additional radionuclides were included for groundwater protection 
based U.S. Environmental Protection Agency (EPA) regulations (40 CFR 197.30 
[DIRS 165519]) and 10 CFR 63.331 [DIRS 156605]. These standards set limits on: 1) combined 
activity of 226Ra and 228Ra in groundwater; 2) gross alpha activity (including 226Ra but excluding 
radon and uranium), and 3) dose from combined beta and photon emitting radionuclides in 
groundwater. 226Ra and 228Ra were added to the inventory for TSPA-LA. Compliance with the 
gross-alpha regulation requires inclusion of appropriate equilibrium activity contributions for 
short-lived decay products (BSC 2002 [DIRS 160059], Section 6.6.1). Short-lived decay 
products can be computed from the activities of their parents so it is not necessary to list or 
include them in the radionuclide inventory for this analysis (DOE 2002 [DIRS 160059], 
Section 6.6.1). However, Radionuclide Screening (BSC 2002 [DIRS 160059]) notes that the 
alpha activities of long-lived decay products that have been screened in, and those of their short-
lived daughters are to be counted in the gross-alpha calculation. The third item is concerned 


with dose from contaminated drinking water and because this pathway was included in the initial 
screening, no additional analysis was needed. 

Radionuclide Screening (BSC 2002 [DIRS 160059]) noted that some radionuclides that are 
precursors of important radionuclides were not identified by the screening analysis as potentially 
important. Table 4-14 illustrates the systematic examination of the radionuclides screened in 
according to the 95 percent cutoff and of 228Ra included for groundwater protection standard. 
Three additional radionuclides were identified for inclusion based either on the daughter product 
or augmentation of the precursor radionuclide screened in: 241Pu, 245Cm, and 235U (BSC 2002 
[DIRS 160059], Table 12). As a result of this analysis, seven precursors have been added for 
inclusion in TSPA-LA (Table 6-1). 

Table 6-1. Precursors Identified as Significant to the Regulatory Period for TSPA-LA 

Radionuclide Added Precursor 
241Am 245Cm and 241Pu 
231Pa 235U 
226Raa 230Th 
228Ra 232Th 
232Th 236U 
238U 242Pu 

Source: This table is a compilation of the preceding text. 

NOTE: 228Ra was screened in due to EPA groundwater 
protection standard 40 CFR Part 197 [DIRS 165519]. 

Table 6-2 includes the 21 radionuclides identified as significant to the regulatory period (20 
radionuclides identified in Table 4-13 screened in for dose, plus 228Ra included for groundwater 
protection), plus seven precursors identified in Table 6-1 as significant to the regulatory period. 
In addition, four additional radionuclides are included in the inventory and are identified for the 
FEIS million-year period. 

6.2 WASTES AND WASTE PACKAGES 

Figure 6-1 illustrates the four types of nuclear waste in the waste form inventory to be emplaced 
in the repository: CSNF, DSNF, DHLW glass, and naval spent nuclear fuel. For the repository 
layout and thermal calculations, ten waste package configurations were designed to 
accommodate the waste types listed in Table 4-2. (Configurations 7a, 7b and 8a, 8b represent 
two loadings of two configurations.) Figure 6-2 illustrates some of the waste package 
configurations. 

The CSNF configurations 1 through 5 in Table 4-2 contain the commercial fuel assemblies from 
PWRs and BWRs loaded with either neutron absorber plates (AP) or control rods (CR) for 
criticality control. The codisposal configurations 6, 7a, 8a, 7b, and 8b in Table 4-2 hold DSNF 
and DHLW in various sized canisters within the waste package. 

The DSNF inventory analyzed in Section 6.6.2 of this analysis includes all DSNF destined for 
the repository except for the bare fuel and the naval SNF. Bare fuel is CSNF in the custody of 
DOE and DHLW is analyzed in Section 6.6.3. 


Naval fuel will be packaged in either short or long canisters (Configurations 9 and 10 in 
Table 4-2). Naval fuel is a DSNF. It is a robust fuel and is included in the allocation for 
disposal of DSNF. For TSPA-LA postclosure analysis, the 300 naval packages are treated as 
CSNF packages. This approximation is justified for the following two reasons: 1) naval fuel is 
not codisposed with glass; 2) naval fuel is a robust fuel, more robust than CSNF, and much more 
robust than DSNF as modeled in TSPA (no cladding and/or instantaneous dissolution). Even for 
scenarios where CSNF cladding is not degraded by events, comparison studies have shown that a 
CSNF package release bounds a naval package release (BSC 2001 [DIRS 152059], 
Section 6.1.1). 

Table 6-2. Included Radionuclides (32) 

Radionuclides 
Important to Dose 
Screened in 
for Dose 
Precursor Screened 
in For Dose 
Screened In for 
Groundwater 
Protection (EPA) 
Additional Radionuclides for 
the FEIS Million-Year 
Calculation 
227Ac 227Ac 
241Am 241Am 
243Am 243Am 
14C 14C 
36Cl 36Cl 
245Cm 245Cm 
135Cs 135Cs 
137Cs 137Cs 
129I 129I 
237Np 237Np 
231Pa 231Pa 
210Pb 210Pb 
238Pu 238Pu 
239Pu 239Pu 
240Pu 240Pu 
241Pu 241Pu 
242Pu 242Pu 
226Rab 226Raa 
228Raa 228Rab 
79Se 79Se 
126Sn 126Sn 
90Sr 90Sr 
99Tc 99Tc 
229Th 229Th 
230Th 230Th 
232Th 232Th 
232U 232U 
233U 233U 
234U 234U 
235U 235U 
236U 236U 
238U 238U 
count 20 7 1 4 

Source: BSC 2002 [DIRS 160059], Section 6.1. 

NOTES: Radionuclide Screening (BSC [DIRS 160059]) shows that extending screening to 20,000 years after 
emplacement adds no new radionuclides. 

a 

Additional Isotopes required by 10 CFR 63.331b [DIRS 156605] for groundwater protection. 

b 

Additional Isotopes required by EPA 40 CFR 197.30a [DIRS 165519] for groundwater protection. 

ANL-WIS-MD-000020 REV 01 6-6 September 2004 


Figure 6-1. Various Waste Package Types Containing Different Wastes 


Source: Modified from CRWMS M&O 2000 [DIRS 150707], Figure 3.1-1. 

There are many types of DOE-owned fuel assemblies and parts. These fuels and parts come 
from a wide range of reactor types with various cladding materials and enrichments. The large 
number of DSNFs is indicative of the number and variety of different reactor types that once 
existed. The DOE-owned fuel assemblies and parts have been categorized by the size, shape, 
composition, condition of the assemblies, and the size and corrosion resistance of the canister 
into which wastes are to be loaded (DOE 2003 [DIRS 163377]): 

 
Short standard canister, 1.5 ft by 10 ft 
 
Long standard canister, 1.5 ft by 15 ft 
 
Wide short standard canister, 2 ft by 10 ft 
 
Wide long standard canister, 2 ft by 15 ft 
 
High-integrity can (HIC) (placed within a long standard canister), 1.5 ft by 15 ft 
 
Multicanister overpacks (MCO), 2.1 ft by 14 ft 
 
Bare PWR and BWR assemblies. 



Source: Modified from DOE 2002 [DIRS 155943], Figure 3-5. 
NOTE: For clarity, basket is excluded from this illustration. 
Figure 6-2. Some Waste Package Configurations for CSNF, DSNF, Naval, and HLW Glass 

ANL-WIS-MD-000020 REV 01 6-8 September 2004 


The wide canisters and the bare PWR and BWR assemblies from DOE (DOE 2003 
[DIRS 163377]) are not mentioned in Table 4-2. The DOE has possession of 79 PWR and 87 
BWR assemblies, which would fit into four 21-PWR packages and two 44-BWR packages. This 
trivial number of packages, compared to the number of other CSNF packages, is not included in 
the DSNF average for TSPA-LA. The wide packages, however, are included in TSPA-LA, as 
shown in the alternate set of loadings in Table 6-3. 

During production of this analysis, the codisposal loadings changed from those in Table 4-2 to 
those in Table 6-3. In Table 6-3, the CDSP configurations have been altered to accept the DSNF 
inventory that represents 2,300 metric tons of 238U (DOE 2003 [DIRS 163377]), while 
preserving the total number of codisposal packages (DOE 2003 [DIRS 167367]. DSNF data on 
nominal and bounding fuel inventories in total curies per category and the number of packages 
per category are shown in Tables 4-8 and 4-9 (DOE 2003 [DIRS 163377], Total 2030 
D568__585.xls). These data were used to develop the data for the DOE canister 
configurations in Table 6-3. The numbers of packages needed to accommodate the canisters in 
each category were identified. One canister per package was calculated for each configuration 
except multi-canister overpack (MCO), which has two canisters per package. Unneeded 
radionuclides were removed from this data and these data were placed in DSNF Ci Summary 
32 (Appendix III Microsoft Excel workbook: inv_rev1.xls, worksheet: Ci Summary 32 
for DSNF). The detailed process for calculating the old and new average inventory per package 
is shown in Appendix II-3. 

The largest difference between Tables 4-2 and 6-3 is the redefinition of Configuration 8a to 
accommodate the wide canisters within waste packages designed to hold the multicanisteroverpack-
type (MCO) canisters. Configuration 8a has 192 canisters to accommodate the 
165 wide-short and 27 wide-long canisters (DOE 2003 [DIRS 163377]). Configuration 7a has 
both long and high-integrity cans (HIC) in long-standard canisters. Configuration 7b has the 
seven waste packages without DSNF needed to match the total number of codisposal packages 
listed in Table 4-2. Note the move of glass canisters from Configuration 7b to the other 
configurations. The new configurations accommodate 195 more DSNF canisters and 543 fewer 
glass canisters. The deviation of Table 6-3 from Table 4-2 is justified, because Table 4-2 was 
provided for layout and thermal purposes and the codisposal packages represent a small fraction 
of the thermal load. In addition, during the preparation of this analysis, the maximum 
capacity/loading of CDSP waste packages (as well as CSNF) was documented as one of the 
functional and operational requirements and design solutions (DOE 2003 [DIRS 163377]). 

Figure 6-3 plots the number of waste packages of each type for the two alternate configurations. 
It can be seen that the CSNF configurations dominate the co-disposal and naval configurations. 


Table 6-3. Package Configurations Including New Waste Codisposal Packages to Accommodate DSNF 

Configuration 
Number 
Waste 
Package 
Type 
Number of 
Waste 
Packages 
Spent Nuclear 
Fuel Unit 
Max # of 
Units per 
Packagea 
# of Glass 
Canisters 
per 
Packagea 
Glass 
Unit Shorthand 
1 CSNF 4,299 PWR assembly 21 0 21-PWR AP 
2 CSNF 95 PWR assembly 21 0 21-PWR CR 
3 CSNF 163 PWR assembly 12 0 12-PWR AP 
4 CSNF 2,831 BWR assembly 44 0 44-BWR 
5 CSNF 84 BWR assembly 24 0 24-BWR 
6 Codisposal 1,403 Short Canister 1 DSNF 5 HLW short 1S/5S 
7a Codisposal 1,608 Long Canister 1 DSNF 5 HLW long 1L/5L 
8a Codisposal 192 Wide Canister 1 DSNF 3 HLW long 1W/3L 
7b Codisposal 7 None 0 5 HLW long 0/5L 
8b Codisposal 202 MCO 2 MCO 2 HLW long 2MCO/2L 
9 Naval 144 Canister 1 0 Naval 
10 Naval 156 Canister 1 0 Naval 
Total Number of Codisposal 
Waste Packages 
3,412 
Total Number of CSNF + Naval 7,772 
Waste Packages 
Grand Total 11,184 

Source: DOE 2003 [DIRS 163377], Total 2030 D568--585.xls. From Appendix III Microsoft Excel 
workbook: inv_rev1.xls, worksheet: new configs. 

NOTES: The CSNF assemblies in DOE custody (bare fuel) were not included in the DSNF inventory for 
TSPA-LA. See DSNF Assumptions, Sec. 5.2. 

a 

Maximum capacity for CDSP, naval, MCO, and BWR/PWR CSNF waste packages are defined in 
DOE 2003 [DIRS 167367]. 


Figure 6-3. Histograms of Waste Package Configurations 


Source: From Appendix III: Microsoft Excel workbook: inv_rev1.xls, worksheet chart config. 

DHLW glass will be delivered to the repository in short and long canisters. 

6.3 FEATURES, EVENTS, AND PROCESSES CONSIDERED IN THE ANALYSIS 

As stipulated in Technical Work Plan For: Regulatory Integration Modeling and Analysis of the 
Waste Form and Waste Package (BSC 2004 [DIRS 171583]), this analysis addresses key 
radionuclide inventory-related features, events, and processes (FEPs) for waste form and 
engineered barrier system components of the repository that are screened in (i.e., included FEPs) 
for TSPA-LA. These FEPs are identified in Table 6-4. There were no screened-out FEPs 
identified. 

Each FEP was evaluated for inclusion or exclusion in the total system performance assessment 
against three criteria, which are stated as regulatory requirements at NRCs rule 
10 CFR 63.114(d)(e)(f) [DIRS 156605]. The development of a comprehensive list of FEPs 
potentially relevant to postclosure performance of the potential Yucca Mountain repository is an 
ongoing, iterative process based on site-specific information, design, and regulations. The 
approach for developing an initial list of FEPs, in support of TSPA-SR (CRWMS M&O 2000 
[DIRS 153246]), was documented in The Development of Information Catalogued in REV00 of 
the YMP FEP Database (BSC 2001 [DIRS 154365]). The initial FEP list contained 328 FEPs, 
of which 176 were included in TSPA-SR models (CRWMS M&O 2000 [DIRS 153246], 
Tables B-9 through B-17). To support TSPA-LA, the FEP list was reevaluated in accordance 
with The Enhanced Plan for Features, Events, and Processes (FEPs) at Yucca Mountain 
(BSC 2002 [DIRS 158966]). 


Table 6-4 provides a list of FEPs that are evaluated in this analysis in accordance with their 
assignment in the LA FEP list (MO0407SEPFEPLA.000 [DIRS 170760]). Specific reference to 
the various sections within this document where issues related to each FEP are addressed is 
provided in the tables. The detailed discussion of these FEPs, their implementation in 
TSPA-LA, and the exclusionary arguments are documented in Miscellaneous Waste-Form FEPs 
(BSC 2004 [DIRS 170020]). 

Table 6-4. Features, Events, and Processes Included (Screened In) in TSPA-LA and Addressed in 
this Report 

FEP Number FEP Name Sections Where Disposition is Described 
2.1.01.01.0A Waste inventory All, summarized in Section 6.7 
Radionuclide Screening (BSC 2002 [DIRS 
160059], Tables 12 and 13) 
2.1.01.03.0A Heterogeneity of waste inventory Sections 6.4 and 6.6 
3.1.01.01.0A Radioactive decay and in-growth Section 6.1 

Source: DTN: MO0407SEPFEPLA6.000 [DIRS 170760]. 

6.4 APPROACH AND METHODS OF CALCULATIONS 

6.4.1 Radionuclide Inventory 

All direct inputs are listed in Table 4-1. The inputs identified in Table 4-1 and the new 
configurations defined in Table 6-3 were used to calculate the nominal radionuclide inventory in 
grams per package for representative CSNF and codisposal waste packages. The inventories and 
associated uncertainty ranges were calculated using standard functions of the Microsoft Excel 
Software. These calculations and electronic data files are described in Appendix II. The data are 
developed and illustrated in Appendix III that is a CD containing Microsoft Excel workbook: 
inv_rev1.xls. These calculations resulted from inventories identified for the year of projected 
completion of emplacement (e.g., 2033 for CSNF and 2030 for HLW and DSNF). TSPA-LA 
requires inventory for the year of repository closure and, therefore, takes the output of this 
analysis and decays the inventory to that date, which is 2060 for the TSPA-LA calculations. 
This report provides the inventories at years 2030 and 2033. Further discussion of the 
calculations supporting the uncertainty ranges is provided in Section 6.6. 

6.4.2 Inventory Heterogeneity 

The repository waste types are heterogeneous both in type (SNF versus glass) and in inventory 
per package. CSNF, DSNF, and HLW shipped to the repository will contain quantities of 
radionuclides that vary from waste package to waste package, fuel assembly to fuel assembly, 
and from canister to canister. The composition of each of these waste forms vary due to several 
factors that include but are not limited to initial uranium enrichment, possible plutonium 
enrichment, and fuel burn up. Given these differences in composition, the mass of radionuclides 
available for transport will vary significantly among waste packages. The heterogeneity is larger 
for DSNF and HLW glass inventories than for CSNF. This package-to-package inventory 
variability is not significant for postclosure TSPA-LA because it samples across many 
realizations, and each waste package is as likely to fail as any other one. The heterogeneity of the 


inventory for TSPA-LA postclosure performance is characterized with the uncertainty 
parameters (Section 6.6) for the average inventory within the CSNF and codisposal packages. 

6.4.3 Radionuclide Analysis of Activated Mineral Deposits on Assemblies 

Appendix I is an analysis of the radionuclides that exist on the outside of cladding and the 
assembly hardware. This analysis demonstrates that it is appropriate to neglect the radionuclide 
contribution from activated mineral deposits (crud) on CSNF assemblies in TSPA-LA. 
However, the Appendix I analysis identifies 99Tc and 14C as potentially significant hardware-
activation products in CSNF assemblies. As demonstrated in the analysis, a negligible fraction 
of the 99Tc in CSNF is due to hardware activation. Because the TSPA does not take credit for 
DSNF cladding, the 99Tc inventory from CSNF hardware is overwhelmed by the 99Tc from 
DSNF in codisposal waste packages. Therefore, combining the inventory of 99Tc from CSNF 
hardware-activation products together with the 99Tc in the CSNF matrix, as has been done in the 
supporting calculations of Radionuclide Screening (BSC 2002 [DIRS 160059]), is an acceptable 
approximation for TSPA-LA. The analysis in Appendix I estimates that approximately 
18 percent of the 14C inventory in CSNF originates from neutron activation of stainless steel 
hardware outside the fuel rods. This hardware will start to corrode and release 14C upon waste 
package breach. TSPA-LA will consider this contribution separately from the 14C in the fuel 
matrix. 

6.5 RADIONUCLIDE INVENTORY ANALYSES RESULTS 

Table 7-1 shows the expected nominal inventory in average grams per package for 
32 radionuclides for each waste type at year 2030 or 2033. The inventory is shown in Figure 6-4 
with the radionuclides listed in descending order of the mass in the CSNF packages. The 
numerical values are provided in Appendix III Microsoft Excel workbook: inv_rev1.xls, 
worksheet Chart gpp. Uncertainty in the inventory values is discussed in the next section. 


Source: From Appendix III Microsoft Excel workbook: inv_rev1.xls, worksheet Chart gpp. 
NOTE: Years of Projection: CSNF, 2033; DSNF, 2030; and HLW, 2030. 
Figure 6-4. Average Grams per Package of Radionuclides for Each Waste Type 


6.6 DISCUSSION OF UNCERTAINTIES 

A number of sources contribute to the uncertainty in the radionuclide inventories. The first 
source is due to the computational method and nuclear data used in predicting future 
radionuclide inventories (e.g., isotopic neutron cross section or decay half-life). The second 
source of uncertainty is the completeness of the records that are kept for the SNF and HLW 
materials (e.g., burnup history or HLW batch compositions). The third source, and most difficult 
to quantify, is the uncertainty about future decisions that will influence the creation, packaging 
and shipment of waste. These decisions influence the selection of waste types destined for 
emplacement for which the legal limit is defined as 70,000 MTHM. Because the waste is 
heterogeneous, the selection process can change the average waste characteristics. Uncertainty is 
discussed in greater detail in the following subsections for each of the three waste types. 

6.6.1 CSNF Waste 

The investigations of computational methods for the prediction of radionuclide inventory based 
on uncertainties in PWR and BWR burnup data have been reviewed. It has been shown that 
isotopic concentrations can be predicted with reasonable accuracy relative to measured data 
even in light of the complexities of BWR and PWR reactor design and operation. Records and 
nuclear data associated with burnup and nuclear reactor operation have been studied by DeHart 
and Hermann (1996 [DIRS 156084]) and Hermann and DeHart (1998 [DIRS 106563]). 
Validation studies have demonstrated the capability to predict the isotopic composition of PWR 
and BWR fuels as a function of burnup history using the computer code systems SCALE 
(Standardized Computer Analysis for Licensing Evaluation) and SASH2H, a multicode sequence 
that determines the isotopic composition of spent fuel using the ORIGEN-S code for depletion 
and decay calculations (DeHart and Hermann 1996 [DIRS 156084], p.1); Hermann and DeHart 
(1998 [DIRS 106563], p. xi). 

Differences between calculated and measured isotopic concentrations for any fuel result from 
biases and uncertainties in both the calculational method and the original experimental 
measurement. The uncertainty associated with the prediction of the concentration of a given 
isotope is estimated from a set of measured-to-calculated ratios. The reported average percentage 
difference between measured and computed isotopic concentrations are used in this analysis to 
evaluate the uncertainty associated with the CSNF radionuclide inventory. Uncertainties for 
other radionuclides of interest to TSPA-LA that are not part of nuclear reactor operations are 
evaluated by these methods as well in the absence of data specific to such isotopes. Table 4-7 
lists the average percent difference between measured and computed isotopic concentrations 
for those isotopes important to dose for 10,000-year regulatory period using Equation 6-1 as 
calculated by DeHart and Hermann (1996 [DIRS 156084], Table 19) and Hermann and DeHart 
(1998 [DIRS 106563], Table 24). The sources studied BWR and PWR fuels respectively. 

% Difference = (calculated/measured -1)  100% (Eq. 6-1) 

The maximum and minimum percent difference between measured and computed isotopic 
concentrations for specific isotopes shown in Table 4-7 is +12.1 percent for 99Tc and -7 percent 
for 238Pu. This percent difference is used to develop correction factors for CSNF in this analysis. 
The values (+12.1 percent and 7 percent) have been converted to correction factors of 0.89 and 


1.08, respectively using Equation 6-2 and have been applied to the CSNF inventory 
(Section 6.6.1). Qualification of data used as direct input to this analysis taken from DeHart and 
Hermann (1996 [DIRS 156084], Table 19) and Hermann and DeHart (1998 [DIRS 106563], 
Table 24) is provided in Appendix IV.3. 

Correction factor = 1/[1+(measure of error/100)] (Eq. 6-2) 

The use of burnup uncertainty for PWR and BWR fuel is well known in the nuclear industry. 
Utility data were used to evaluate the uncertainties associated with calculated burnup versus 
measured burnup, and the percent difference of calculated to measured burnup (P). Data for 
5,447 fuel assemblies at the end of cycle (EOC) in groupings of burnup: >10,000, 10,000 to 
30,000, and burnup > 30,000 MWd/MTU were evaluated and recently reported (Massie 2004 
[DIRS 170651], Summary Sheet). (P) is defined in Massie (2004 [DIRS 170651]) as the percent 
difference of calculated to measured burnup as shown by the Equation 6-3. It was determined 
that the largest uncertainty associated with P for BWR and PWR assemblies, was 4.2 percent. 

P = 100(calculated burnup - measured burnup)/(measured burnup) (Eq. 6-3) 

Determination of the Accuracy of Utility Spent-Fuel Burnup Records (EPRI 1999 
[DIRS 164649] p. vii) estimates the uncertainty in burnup history records to be 1.89 percent. 
Burnup history and uncertainties related to computational methods, records, and nuclear data 
were applied in the inventory calculations by DeHart and Hermann (1996 [DIRS 156084]) and 
Hermann and DeHart (1998 [DIRS 106563]). Therefore, these factors need not be applied 
separately in this analysis of DSNF radionuclide uncertainty. 

The heterogeneity in CSNF inventory can be seen in the difference between configurations as 
shown in Figure 6-5, which compares the average grams per package for the 32 radionuclides of 
importance to the TSPA-LA for five CSNF configurations. The ratio of configuration average 
inventory to the repository weighted average inventory for each radionuclide is shown in 
Figure 6-6. In Figures 6-5 and 6-6, Configurations 1 (21-PWR AP), 3 (12-PWR AP), and 
4 (44-BWR) are clustered around the weighted average. Configurations 2 (21-PWR CR) and 
5 (24-BWR) are lower for all but 14C, 230Th, and 235U. The maximum ratio of an isotope-
averaged grams per package, to the weighted average for a CSNF configuration is 1.3, with a 
minimum of 0.15. Calculations are shown in Appendix III Microsoft Excel workbook: 
inv_rev1.xls, worksheet config variability. 

The configuration inventory trends are consistent with the average burnup trends shown in 
Table 6-5. Note that the correlation of radionuclide inventory with burnup is strong for 
activation and fission products, but weak for 238U. 238U does not decrease much during reactor 
operations even with enrichments of 5 percent and maximum burnup as can be seen in the tight 
cluster of 238U inventories in Figures 6-5 and 6-6. The range of 238U inventory in Figures 6-5 
and 6-6 is dominated by the variation in the amount of fuel per package in each configuration. 


Table 6-5. Average Burnups for the 5 CSNF Configurations, and Ratios to Average 


Source: From Appendix III Microsoft Excel workbook: inv_rev1.xls, worksheet, NewSRSbatches. 
Figure 6-5. Comparison of the Per-Package Inventories of the Five CSNF Configurations 

Configuration Number of Packages 
Average 
Burnup 
(GWd/MTU) Configuration 
Ratio: Average 
Burnup/Weighted 
average 
1 4,299 41 1 1.09 
2 95 20 2 0.51 
3 163 46 3 1.21 
4 2,831 34 4 0.89 
5 84 8 5 0.21 
Total 7,472 
Weighted 
Average 38 
Minimum 0.21 
Maximum 1.21 

Source: CRWMS M&O 2000 [DIRS 138239]. See (this analysis) input data in Table 4-5 and Appendix III 
Microsoft Excel workbook: inv_rev1.xls, worksheet burnup. 


Initial Radionuclide Inventories 


Source: Appendix III Microsoft Excel workbook: inv_rev1.xls, worksheet config variability. 

NOTE: The weighted average is the grams per waste package divided by the weighted average grams per waste package. 
Figure 6-6. Ratios of CSNF Configuration Inventories to the CSNF Weighted Average 

ANL-WIS-MD-000020 REV 01 6-17 September 2004 


To investigate the uncertainty in the repository-average inventories due to heterogeneity of the 
waste, the average burnups from three 1999 arrival forecasts, A to C (CRWMS M&O 2000 
[DIRS 138239]) and four 2002 arrival forecasts, A' to D' (BSC 2003 [DIRS 165990], 
Section 3.2.1) were compared. The base case (i.e., Case A 63,000 MTHM) 
(CRWMS M&O 2000 [DIRS 138239]), was used to derive the CSNF inventory given in 
Table 7-1. Since it is unknown which assemblies the utilities will send at a given time, factors 
were calculated using the three 1999 arrival forecasts in relation to the four 2002 forecasts, all 
based on the base case of 63,000 MTHM commercial fuel allocation. For ease of comparison, the 
cases are defined in Table 6-6. 

Table 6-6. Arrival Forecasts for the 63,000 MTU 

Case Definition 
A Fuel selection begins with 10-year-old spent fuel 
B Fuel selection begins with 10-year-old spent fuel in strict order of age 
C Fuel selection begins with oldest fuel first. 
A' Oldest fuel first with no dry storage.(OFF) 
B' Youngest fuel first greater than 10 years (YFF10) 
C' Youngest Fuel first greater that 5 years old in strict age order (YFF5) 
D' Limiting YFF5: All youngest (5-year old) fuel is selected first followed by older fuel 
in increasing age order. 

Source: Cases A to C are from CRWMS M&O 2000 [DIRS 138239], Tables 10, 12, and 14 
Cases A' to D' are from BSC 2003 [DIRS 165990]. 

The actual arrival sequence will be determined by the utilities, but the forecasts were designed to 
bound the actual arrival sequence with waste streams that are the most and least stressing to the 
repository systems. The 2002 forecast arrival scenarios B', C', and D' have larger fractions of 
younger fuel than the 1999 forecast projections. The resulting higher average burnups for these 
scenarios can be seen in Table 6-7. Because reactor technology has advanced and the younger 
fuels have higher average burnups than older fuels, the average burnups from these arrival 
scenarios are higher. The minimum and maximum projected average burnup ratios for these, and 
all the cases over the 1999 Case A, are 0.95 and 1.29 (Table 6-7). When multiplied by the 
minimum and maximum correction factors (measured/calculated), of 0.89 and 1.08 for the 
computational method, a range of 0.85 to 1.40 was obtained. This range of ratios provides a 
reasonable estimate for the range of uncertainty present in the repository-average inventories for 
burnup-dependent radionuclides. Since the radionuclide inventories are burnup-dependent 
(except that of 238U for CSNF), they were not sampled independently. Accordingly, a single 
variable is defined to capture the uncertainty for all isotopes except 238U. In this case, an 
uncertainty multiplier was chosen that was sampled and then applied to all radionuclide 
inventories except 238U. In defining the probability distribution of this multiplier, the end points 
were known (i.e., 0.85 and 1.40), but the shape of the uncertainty distribution was unknown. In 
the absence of that information, a uniform distribution was chosen. Thus, an uncertainty 
multiplier with a uniform distribution between 0.85 and 1.40 was chosen for the TSPA-LA, for 
application to the nominal CSNF values (provided in Table 7-1) for all radionuclides 
except 238U. 


Table 6-7. Average Burnups for the Arrival Scenarios Projected For Years 1999 And 2002 

Case for 63,000 MTHM 
1999 Projected 
Average Burnup 
(GWd/MTU) 
Ratios To 
1999 Case A 
A 38 1.00 
B 39 1.03 
C 37 0.97 
2002 Projected 
Average Burnup 
GWd/MTU 
A' 36 0.95 
B' 44 1.16 
C' 47 1.24 
D' 49 1.29 
min 0.95 
max 1.29 

NOTES: The three 1999 arrival forecasts A to C (CRWMS 

M&O 2000 [DIRS 138239], Tables 10, 12, and 14) and 

four 2002 arrival forecasts, A' to D' (BSC 2003 

[DIRS 165990] 3.2.1 and Tables 2 and 4). 

6.6.2 DSNF Waste 

The inventory estimates given provide a nominal and a bounding radionuclide inventory for the 
many heterogeneous types of DOE-owned spent nuclear fuels (DSNFs) (DOE 2003 
[DIRS 163377]). Although the bounding radionuclide inventory estimates were provided to 
assess the preclosure risk associated with handling a worst-case canister, the values have also 
been used to represent a bounding inventory for assessing postclosure risk. In Appendix III, 
Microsoft Excel workbook: inv_rev1.xls, worksheet DSNF uncert, the nominal and 
bounding grams per package (gpp) for the weighted average of DSNF waste were analyzed, 
excluding the bare commercial PWR and BWR assemblies in DOE's possession (DOE 2003 
[DIRS 163377]). The source of inventory data for DSNF (DOE 2003 [DIRS 163377]) and the 
canister count (DOE 2004 [DIRS 169354] are qualified for intended use in Appendix IV. 

Source Term Estimates for DOE Spent Nuclear Fuels (DOE 2004 [DIRS 169354], p. 38) 
determined that bounding burnup information was not available to support inventory 
calculations. Therefore, the nominal burnup was estimated and the bounding burnup was 
conservatively assumed to be two times (2) the nominal burnup (DOE 2004 [DIRS 169354], 

p. 38). Thus, the ratio of the bounding to nominal grams per package for each isotope ranges 
from approximately one to two. Source Term Estimates for DOE Spent Nuclear Fuels 
(DOE 2003 [DIRS 163377], p. 39) reports that 38 percent of the DSNF inventory comes from 

0.31 
percent of the total DSNF. This 0.31 percent of the mass relied on conservative assumptions 
to compensate for missing information. If we can assume that 0.31 percent of the fuel can be 
represented by the 99.69 percent of the DSNF with sufficient information, we can conclude that 
the expected inventory is more reasonably represented as 62 percent of the nominal. See 
assumption in Section 5.2. 





In addition to the uncertainty of the DSNF inventory, there is the uncertainty associated with the 
loading of the DSNF waste into DSNF canisters. The 2004 DOE report states that the Spent 
Fuel Database provides point estimates of the number of U.S. DSNF canisters required for all 
DSNF (DOE 2004 [DIRS 169354] Appendix F, p. F-3). In 2003, Source Term Estimates for 
DOE Spent Nuclear Fuels REV 0 (DOE 2003 [DIRS 163377]) used a canister count of 3,607. 
Revision 1 of that report (DOE 2004 [DIRS 169354] p. 41) provides a DSNF canister count 
range of 2,500 to 5,000 to include uncertainty. Using the canister counts to examine the 
uncertainty used here, applying a factor based on the ratio of 3,607/2,500 to the maximum of 


1.99 
(DSNF_uncert worksheet) and 3607/5000 to the reasonable inventories (i.e., 0.62 of the 
nominal (DSNF_uncert worksheet), yields a maximum and minimum of 2.90 and 0.45, 
respectively). As noted in Source Term Estimates for DOE Spent Nuclear Fuels (DOE 2003 
[DIRS 163377]), the best estimate of the inventory is the inventory that is represented by 0.62 of 
the nominal inventory. The minimum, reasonable, and maximum DSNF radionuclide 
inventories associated with the 2003 data are shown in Figure 6-7. 


Figure 6-7. Uncertainty in 2003 DSNF Radionuclide Inventory Estimates 


Source: DOE 2003 [DIRS 163377]; Appendix III Microsoft Excel workbook: inv_rev1.xls, worksheet chart DSNF 
uncert. 

Since this analysis uses inventory data from Source Term Estimates for DOE Spent Nuclear 
Fuels (DOE 2003 [DIRS 163377]), a comparison has been made of the two inventories (i.e., 
against DOE 2004 [DIRS 169354]). The results are plotted on a log scale in Figure 6-8 and 
Figure 6-9. These data are provided in worksheets described in Appendix II and provided on the 
Appendix III CD. As can be seen in Figure 6-8, the 2003 data used in this analysis tracks closely 
with 2004 data. Both old max and old min inventory projections are bounded by the 
uncertainty calculated for the 2003 data (Figure 6-9). In fact, the 2004 report states that the net 
impact of the changes on the total estimated radionuclide inventory based on updates to the 
DOE spent nuclear fuel database was an ~2% decrease for the bounding case and an ~14% 
decrease for the nominal case. An assessment of the impacts concluded that these changes are 
not expected to impact the repository licensing basis (DOE 2004 [DIRS 169354], p. 7). 
Therefore, the analysis performed using the 2003 DSNF inventory is valid for TSPA-LA. 


Figure 6-8. Comparison of the Nominal Radionuclide Inventory for 2003 (old nom) and 2004 (new nom) 


Source: DOE 2003 [DIRS 163377] and DOE 2004 [DIRS 169354]; Appendix III Microsoft Excel workbook: 
inv_rev1.xls, worksheet, DSNF compare plots. 

Like CSNF, the uncertainties of the DSNF radionuclide inventories are correlated and an 
uncertainty multiplier is defined to capture the uncertainty for all radionuclides except 238U. The 
inventory of 238U is well known for DSNF and has much less relative uncertainty than the other 
radionuclides because it is present in the initial fuel and generally changes little during reactor 
operation. This is seen in Figures 6-7, 6-8, and 6-9. Defining a probability distribution for the 
DSNF multiplier, three points are provided: a minimum value, expected or most likely value 
and a bounding value. In this situation, using a triangular distribution is reasonable to best 
estimate the range of the data. Thus, the DSNF multiplier is defined as a triangular distribution, 
with a minimum of 0.45, a most likely value of 0.62, and a maximum value of 2.90. These 
factors have been applied to the nominal values for DSNF grams per package in Table 7-1 for all 
isotopes except 238U. Figures 6-8 and 6-9 plot the data applying these factors to the radionuclide 
inventory identified in the DOE 2003 and the DOE 2004 reports (old nom overlaid on new 
nom in Figure 6-8; old nom, old reas, old bound, old max overlaid on new nom, 
new reas, new bound, new max in Figure 6-9). 

For the new data plotted in Figures 6-8 and 6-9, it is noted in the 2004 report (DOE 2004 
[DIRS 169354]) that 17.4 percent of the 2004 radionuclide inventory comes from the results of 

0.2 percent of the total DSNF. This 0.2 percent of the mass relied heavily on conservative 
assumptions to compensate for missing information. Consequently, by assuming 0.2 percent of 
the DSNF with insufficient information can be reasonably represented by the 99.8 percent of 
DSNF inventory with sufficient information, one may conclude that the total radionuclide 
inventory associated with DSNF is more reasonably about 83 percent of the nominal inventory 
estimated. This is compared to the DOE 2003 report estimates that 38 percent of the DSNF 
radionuclide inventory comes from 0.31 percent of the DSNF inventory. The 0.31 percent with 
insufficient information is represented by the 99.69 percent of the DSNF inventory with 
sufficient information such that the expected value for the radionuclide inventory is 
reasonably about 62% of the nominal inventory estimated (DOE 2003 [DIRS 163377]). 


Figure 6-9. Comparison of the 2003 and 2004 DSNF Radionuclide Inventories 


Source: DOE 2003 [DIRS 163377] and DOE 2004 [DIRS 169354]; Appendix III Microsoft Excel workbook: 
inv_rev1.xls, worksheet, DSNF compare plots. 

6.6.3 HLW Glass Waste 

Earlier determinations of HLW inventories such as Inventory Abstraction (BSC 2001 
[DIRS 157575]) used information for glass from four sites: 1) Hanford, (2) INEEL, (3) Savannah 
River Site (SRS), and (4) the West Valley Demonstration Project. The estimated inventories per 
package from the four sites were heterogeneous. In that analysis, the SRS and Hanford glass 
packages dominated the radionuclide content of the glass inventory. Since that analysis, new 
information has been reported from SRS (Allison 2004 [DIRS 168734], Appendix 1) and 
Hanford (Hamel 2003 [DIRS 164947]). These sources are used in this analysis as the basis for 
the radionuclide content HLW and are expected to bound the HLW inventory for TSPA-LA. 
Qualification of data from Allison 2004 is provided in Appendix IV.1; data from Hamel (2003 
[DIRS 164947]) is qualified in Appendix IV.2) 

The new HLW calculations use the data for the Savannah River HLW glass inventory 
(Allison 2004 [DIRS 168734], Appendix 1) and are shown in Table 4-9. The Hanford canister 
production and waste loading estimates are shown in Tables 4-10 and 4-11 (Hamel 2003 
[DIRS 164947]). These data were used to compare the difference between the old 
(CRWMS M&O 2000 [DIRS 151947]) and new estimates for average short and long canister 
radionuclide inventories as shown in Figure 6-10. This comparison does not imply that the two 
inventories are intended to match each other. Source Terms of HLW Glass (CRWMS M&O 2000 
[DIRS 151947]) is used by TSPA-LA for heat calculations, whereas this document is used by 
TSPA-LA for the radionuclide inventory. 

Data for curies per SRS batch and curies for all of Hanford HLW for the year 2030 were 
collected and converted to curies per canister based on the inventory for 32 radionuclides from 
Table 6-2 and a weighted average is calculated based on the new number of canisters in new 
configurations. The average ratio of new/old inventories for isotopes reported with greater than 


1  10-6 grams was 3.4 for short canisters and 1.8 for long canisters. The new information for 
short canisters (Allison 2004 [DIRS 168734], Appendix 1, Tables 5 to 8) provide the curies per 
canister decayed to the year 2030 for batches of SRS glass already poured (batch 1A, year 1994; 
batch 1B, year 2000; and batch 2 and 3, year 2002) and for those in the future (batches 4 through 
10). Figure 6-11 shows the estimated radionuclide inventories for those batches, which show 
significant heterogeneity. It can also be seen that the future batches dominate the average. 

New long canister information for Hanford shown in Table 4-11 (Hamel 2003 [DIRS 164947], 

p. 1-9, Tables 1-3) shows the expected production of canisters for the program, planning, and 
technology cases based on three canister-fill scenarios: maximum fill (100 percent), WTP 
contract fill (95 percent), and minimum fill (87 percent). The data in Table 4-11 are used to 
develop the uncertainty distribution as discussed in Section 6.6.3. 

Canister production estimates take into account three assumed levels of waste loadings in 
weight-percent (wt %). The Program case assumes an effective waste loading of 27.7 wt %; the 
planning case assumes an effective waste loading of 36.3 wt %; and the technology case assumes 
an effective waste loading of 45 wt %. Therefore, for radionuclide loading, the program case 
with the lowest waste loading (27.7) results in the maximum number of HLW canisters for the 
three canister-fill scenarios. The planning case, with 36.3 percent waste loading yields the 
nominal, expected number of HLW canisters for the three canister-fill scenarios. The technology 
case, which requires new technology to increase the radionuclide waste-loading capability to 

45.0 wt %, yields the minimum the number of canisters needed to store the waste inventory for 
the three canister-fill scenarios (Hamel 2003 [DIRS 164947], p.1-9 and Tables 1 to 3). For 
canister-fill scenarios, the LA case is the contract case of 95 percent full. 

Like that of CSNF and DSNF, the uncertainties in HLW radionuclide inventories are not 
independent. In this case, the inventories are dependent on radionuclide loading (wt %) per 
canister; therefore, an uncertainty multiplier is used. With the current information, the nominal 
values reported in Table 7-1 are the most likely values and, thus, the most likely value for 
the uncertainty multiplier is one. The most likely value corresponds to the expected/nominal 
number of Hanford HLW canisters shown at center of the matrix listed in Table 4-11 (i.e., the 
planning case and contract fill scenario of 95 percent yielding 9,202 canisters per waste 
packages). The minimum uncertainty multiplier of 0.70 is obtained from the ratio of the nominal 
number of canisters per waste packages (9,202) to the maximum number of canisters per waste 
packages (13,205). An uncertainty ratio of 1.3 is derived from the nominal number of 
canisters/waste packages (9,202) to the minimum number of canisters per waste packages 
(7,071). However, because of the uncertainty in possible new vitrified waste forms at SRS with 
loading up to 50 percent and higher, it is prudent to be conservative in the estimate of the upper 
limit. Therefore, conservatively estimating increase in waste loading of 7 wt % above that 
specified for the technology case and applying it to the 100-percent fill scenario, this new 
maximum loading of 52 percent provides a margin, which corresponds to a ratio of 1.5. With the 
most likely, the maximum, and the minimum values defined, a triangular distribution is chosen 
for the HLW uncertainty multiplier used by TSPA. This multiplier is to be applied to the 
nominal HLW inventories shown in Table 7-1 for all isotopes (including 238U). The uncertainty 
multiplier using a triangular distribution has a minimum of 0.7, a most likely value of 1.0, and 
a maximum value of 1.5. 


Figure 6-10. Radionuclide Inventories of the Older and Newer Estimates for Average Short and Long 
HLW Glass Canisters 


Source: CRWMS M&O 2000 [DIRS 151947]; Hamel 2003 [DIRS 164947]; Allison 2004 [DIRS 168734], Appendix I; 
Appendix III Microsoft Excel workbook: inv_rev1.xls, worksheet - newHLW32. 

NOTE: Values less than 10-6 have been truncated to 10-6 for plotting purposes. 

Figure 6-11. SRS Glass Batch Inventories 


Source: Allison 2004 [DIRS 168734], Appendix 1; Appendix III Microsoft Excel workbook: inv_rev1.xls, worksheet 

newSRSbatches. 
NOTE: Values less than 10-6 have been truncated to 10-6 for plotting purposes. 


6.7 INTENDED USE OF OUTPUT DATA 

The output from this inventory analysis will be used as a direct input to the TSPA-LA and is 
limited to postclosure analysis. The calculations were performed in this report uses year-specific 
inventories from direct inputs described in Section 4 and identified in Table 4.1: 2033 for CSNF, 
and 2030 for HLW and DSNF. This report provides the inventories at years 2033 and 2030 that 
TSPA-LA will use to model that inventory at 2060. The data are reported with uncertainties due 
to many assumptions used to derive waste inventories. Several of these uncertainties have been 
discussed and applied both in the derivation of the source data used in this analysis and 
application of these data in this analysis. The effects of conservative assumptions used to 
compensate for missing information clearly show an inverse correlation between the available 
information used in the methodology and the resulting radionuclide concentrations estimated. 

In the selection of radionuclides for initial inventory analysis, radioactive decay, and in-growth 
were considered. The daughter radionuclides of these processes, along with parent radionuclides 
are included in the list of screened radionuclides for the TSPA-LA (BSC 2002 [DIRS 160059]). 
Radioactive decay and in-growth can be used to reduce the number of isotopes carried by the 
TSPA-LA calculations. For example, in the chain 245Cm . 241Pu . 241Am, 241Am is screened 
in as an important isotope, but 245Cm and 241Pu are not. To account for the in-growth of 241Am 
from these isotopes, they can be included in the TSPA-LA calculation, or they can be artificially 
decayed to 241Am to augment the 241Am at the start of the calculation. 

The output of this analysis is shown in Tables 7-1 and 7-2. The values from Table 7-1 are to be 
used as the nominal values. For scenarios where many packages breach, the uncertainties for 
most radionuclide inventories are modeled with the uncertainty multipliers shown in Table 7-2. 
Because the uncertainty multipliers specifically address the heterogeneity of the waste form 
inventory per package, the inclusion of these uncertainty multipliers in the TSPA-LA disposition 
provides for a direct implementation of FEP 2.1.01.03.0A regarding heterogeneity of the waste 
package in the TSPA-LA model. The isotopes 36Cl, 79Se, 126Sn, and 210Pb have been included in 
Table 7-1 to retain a capability to meet the needs of the million-year calculations, but are not 
required for the 10,000-year calculations. The list of radionuclides identified for the regulatory 
period (i.e., 10,000 years did not change when the screening analysis was extended to 
20,000 years). 

In Appendix I, the analysis of activation products show that 18 percent of the 14C inventory in 
CSNF is calculated to originate from neutron activation of stainless steel hardware outside the 
fuel rods. For the TSPA-LA model this inventory was considered to be an unbound inventory, 
available for immediate release after potential breach of the waste package. 


INTENTIONALLY LEFT BLANK 



7. CONCLUSIONS 


This analysis uses the most recent radionuclide information for CSNF, DSNF, and HLW to 
provide nominal radionuclide inventory in grams per package and uncertainty distributions for 
radionuclides important to dose calculations for TSPA-LA. A revision of a report providing 
source term estimates for DOE spent nuclear fuels (DSNF) was released by the U.S. DOE 
National Spent Nuclear Fuel Program in January 2004 (DOE 2004 [DIRS 169354]). The data in 
that revision (DOE 2004 [DIRS 169354]) was compared to the previous revision of the same 
report released in March 2003 (DOE 2003 [DIRS 163377]) that was the source of radionuclide 
inventory data used in this analysis. These comparisons are presented in Section 6.6.2, 
Figures 6-8 and 6-9. As noted in the revised report (DOE 2004 [DIRS 169354]): 

Subsequent to the original March 2003 issue of this report, the National Spent 
Nuclear Fuel Program Spent Fuel Database has been updated to Version 5.0.1. 
This revision incorporates information from Version 5.0.1 of the Spent Fuel 
Database. The net impact of the changes on the total estimated radionuclide 
inventory was an ~2% decrease for the bounding case and an ~14% decrease for 
the nominal case. An assessment of the impacts concluded that these changes are 
not expected to impact the repository licensing basis. 

Therefore, the DSNF radionuclide inventory analyzed and reported in Table 7-1 is valid and 
bounding. The sources of data for CSNF (CRWMS M&O 2000 [DIRS 138239]) and Waste 
Stream Projections Report (BSC 2003 [DIRS 165990])) and DHLW glass (memorandums by 
Hamel (2003 [DIRS 164947]) and Allison (2004 [DIRS 168734], Appendix 1)) are the most 
recent available. 

Data used as direct input to analyses must meet the definition of qualified data in accordance 
with AP-3.15Q, Managing Technical Product Inputs. Data (BSC 2003 [DIRS 165990]; 
Hamel 2003 [DIRS 164947]; and Allison 2004 [DIRS 168734]) have been qualified in 
Appendix IV in accordance with procedural requirements (AP-SIII.9Q and AP-SIII.2Q) that are 
part of Quality Assurance Requirements and Description (DOE 2004 [DIRS 171539]) definitions 
for Qualification of Data. 

The initial radionuclide inventory represents that inventory of radionuclides that has been 
identified as being important to dose. The quantity and activity of these radionuclides have been 
modeled and these data reported in referenced documents to the year of projected emplacement; 
2033 for CSNF, 2030 for DSNF and HLW. This analysis used the modeled inventory data 
provided by the referenced sources and calculated the total grams for each of the 
28 radionuclides identified for the 10,000-year regulatory period required for the TSPA-LA and 
four additional radionuclides for the FEIS million-year calculations. Based on the configuration, 
quantity, and capacity of waste package types, this analysis provides TSPA-LA with the nominal 
grams per waste package of these radionuclides for each type of waste. The total number of 
waste packages for each waste type (i.e., CSNF, DSNF, and HLW) is also provided. In addition, 
this analysis provides uncertainty multipliers for inventory of radionuclides for each waste type 
(CSNF, DSNF, and HLW). These data will be used by the TSPA-LA to model the inventory to 
the year 2060, the projected repository closure. 


7.1 INITIAL RADIONUCLIDE INVENTORIES 

The output of this analysis is provided in Tables 7-1 and 7-2 and reported in 
DTN: SN0310T0505503.004; only the three significant figures shown are to be used. The DTN 
numbers should not be reformatted to show more significant numbers, because these additional 
significant numbers are meaningless and have changed slightly. The numbers reported here were 
formatted by Microsoft Excel to three significant figures and were not rounded. See Table 6-2 for 
the list of radioisotopes directly screened in, and for the list of radionuclides that were not directly 
screened in by the analysis, but should be accounted for in the inventory, either by direct inclusion, 
or by appropriate augmentation of the daughter product. These results are provided for repository-
scale postclosure calculations and are restricted to that use. 

Table 7-1. Nominal Grams per Waste Package of Radionuclides for Each Type of Waste 

Radionuclide 
Grams per Waste Package 
CSNF DSNF HLW 
227Ac 2.50E-06 1.20E-03 2.07E-04 
241Am 8.28E+03 2.15E+02 4.07E+01 
243Am 1.26E+03 6.63E+00 6.24E-01 
14C a 1.37E+00 1.78E+00 0.00E+00 c 
36Clb 3.27E+00 4.17E+00 0.00E+00 c 
245Cm 1.77E+01 9.11E-02 5.89E-02 
135Cs 4.41E+03 9.59E+01 1.38E+02 
137Cs 5.97E+03 9.57E+01 3.28E+02 
129I 1.75E+03 3.51E+01 7.89E+01 
237Np 4.63E+03 8.02E+01 1.08E+02 
231Pa 9.28E-03 2.11E+00 1.66E+00 
210Pbb 0.00E+00 c 3.30E-07 3.69E-10 
238Pu 1.54E+03 1.23E+01 4.24E+01 
239Pu 4.37E+04 2.18E+03 6.06E+02 
240Pu 2.08E+04 4.28E+02 5.01E+01 
241Pu 2.69E+03 2.88E+01 1.32E+00 
242Pu 5.34E+03 2.97E+01 4.22E+00 
226Ra 0.00E+00 c 4.50E-05 2.63E-05 
228Ra 0.00E+00 c 1.49E-05 6.51E-06 
79Seb 4.24E+01 6.72E+00 7.61E+00 
126Snb 4.69E+02 9.26E+00 1.85E+01 
90Sr 2.52E+03 5.14E+01 1.89E+02 
99Tc 7.64E+03 1.56E+02 1.10E+03 
229Th 0.00E+00 c 3.19E-01 3.58E-03 
230Th 1.54E-01 1.16E-01 8.81E-04 
232Th 0.00E+00 c 2.14E+04 3.23E+04 
232U 1.03E-02 1.26E+00 4.43E-04 
233U 5.83E-02 5.30E+02 2.11E+01 
234U 1.77E+03 4.66E+02 2.53E+01 
235U 6.34E+04 2.47E+04 1.53E+03 
236U 3.89E+04 1.23E+03 6.50E+01 
238U 7.92E+06 6.74E+05 2.57E+05 

Output DTN: SN0310T0505503.004. 
Source: Appendix III Microsoft Excel workbook: inv_rev1.xls, worksheet inv32.
NOTES: Year of projection: CSNF, 2033; DSNF, 2030; and HLW, 2030 


Total number of waste packages is 11,184 (7,772 CSNF and 3,412 Codisposal) 

a 

18 percent of 14C for CSNF originates from activation of the hardware outside of the cladding. 

b 

Isotopes not needed for 10,000-year regulatory period, but included to retain ability to perform millionyear 
calculations. 

c 

Grams listed as 0.00E+00 is the value presented in the data input source. 

ANL-WIS-MD-000020 REV 01 7-2 September 2004 


Table 7-2. Uncertainty Multipliers for Grams per Package of Radionuclides for Each Waste Type 

CSNF DSNF HLW 
Isotopes All except 238U All except 238U All 
Distribution Uniform Triangular Triangular 
Minimum 0.85 0.45 0.70 
Most Likely N/A 0.62 1 
Maximum 1.40 2.90 1.5 

Output DTN: SN0310T0505503.004. 

Source: Appendix III Microsoft Excel workbook: inv_rev1.xls, worksheet inv32. 

7.2 YUCCA MOUNTAIN REVIEW PLAN ACCEPTANCE CRITERIA 

The following acceptance criteria from Section 2.2.1.3.4.3 of Yucca Mountain Review Plan, 
Final Report (NRC 2003 [DIRS 163274]) are based on meeting the relevant requirements of 
10 CFR 63.114(a) to (c) and (e) to (g) [DIRS 156605], as they relate to the radionuclide release 
rates and solubility limits model abstraction. For each applicable criterion, the criterion is quoted, 
followed by pointers to the information addressing the criterion. In some cases, the criterion is 
shared by more that one analysis/model report. These criteria will be fully addressed when 
related reports are complete. 

Acceptance Criterion 1System Description and Model Integration Are Adequate 

(1) TSPA adequately incorporates important design features, physical 
phenomena, and couplings, and uses consistent and appropriate assumptions 
throughout the radionuclide release rates and solubility limits abstraction 
process; 

The input to this analysis in Section 4.1 came from the cited documents. The physical process of 
decay and in-growth is accounted for. The discretization of the inventory and the form of the 
uncertainty were consistent with TSPA analysis. 

(2) The abstraction of radionuclide release rates and solubility limits uses 
assumptions, technical bases, data, and models that are appropriate and 
consistent with other related U.S. Department of Energy abstractions. 

For example, the assumptions used for this analysis are consistent with the 
abstractions of Degradation of Engineered Barriers (Section 2.2.1.3.1); 
Mechanical Disruption of Waste Packages (Section 2.2.1.3.2); Quantity and 
Chemistry of Water Contacting Engineered Barriers and Waste Forms (Section 
2.2.1.3.3); Climate and Infiltration (Section 2.2.1.3.5); and Flow Paths in the 
Unsaturated Zone (Section 2.2.1.3.6). The descriptions and technical bases 
provide transparent and traceable support for the abstraction of radionuclide 
release rates and solubility limits; 

The technical bases and data are appropriate but not always consistent. Differences between 
Tables 4-2 and 6-3 are justified in Section 6.2. 


(3) The abstraction of radionuclide release rates and solubility limits provides 
sufficient, consistent design information on waste packages and engineered 
barrier systems. 

For example, inventory calculations and selected radionuclides are based on the 
detailed information provided on the distribution (both spatially and by 
compositional phase) of the radionuclide inventory, within the various types of 
high-level radioactive waste ; 

This analysis is based on the detailed information provided on the distribution (both spatially and 
by compositional phase) of the radionuclide inventory, within the various types of high-level 
radioactive waste as described in Sections 4 and 6. 

(4) to (8) 

Not applicable. 

Acceptance Criterion 2Data Are Sufficient for Model Justification 

(1) 
Not applicable. 
(2) 
Sufficient data have been collected on the characteristics of the natural system 
and engineered materials to establish initial and boundary conditions for 
conceptual models and simulations of thermal-hydrologic-chemical coupled 
processes. 


For example, sufficient data should be provided on design features, such as the 
type, quantity, and reactivity of materials, that may affect radionuclide release for 
this abstraction; 

The data and its uncertainty are discussed in Sections 4.1 and 6.6, respectively. 

(3) to (4) 

Not applicable. 

Acceptance Criterion 3Data Uncertainty Is Characterized and Propagated Through the 
Analysis 

(1) 
Data used for parameter values, assumed ranges, probability distributions, 
and bounding assumptions that are technically defensible, reasonably account for 
uncertainties and variabilities, and do not result in an under-representation of the 
risk estimate; 
(2) 
to (9) 


Data uncertainty is described in Section 6.6. 

Not applicable. 


8. INPUTS AND REFERENCES 


8.1 DOCUMENTS CITED 

168734 Allison, J.M. 2004. Request for Referenceable Information on High-Level Waste 
(HLW) Radionuclide Inventories in Support of Preparation of the Yucca Mountain 
Project License Application (Your Letter, JCP-0445, 1/28/04). Memorandum from 

J.M. Allison (DOE/SR) to J. Arthur, III (OCRWM), February 26, 2004, 0303040661, 
with attachment. ACC: MOL.20040317.0265. 

170913 Bowman, S.M. 2002. Configuration Management Plan for the SCALE Code System. 
SCALE-CMP-001, Rev. 6. Oak Ridge, Tennessee: Oak Ridge National Laboratory. 
ACC: MOL.20040708.0339. 

170910 Broadhead, B.L. 1996. Verification and Validation Plan for the SCALE Code System. 
SCALE-CCV-001, Rev. 1. Oak Ridge, Tennessee: Oak Ridge National Laboratory. 
ACC: MOL.20040708.0341. 

157575 BSC (Bechtel SAIC Company) 2001. Inventory Abstraction. ANL-WIS-MD-000006 
REV 00 ICN 03. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: MOL.20020123.0278. 

152059 BSC 2001. Performance Assessment of U.S. Department of Energy Spent Fuels in 
Support of Site Recommendation. CAL-WIS-PA-000002 REV 00. Las Vegas, 
Nevada: Bechtel SAIC Company. ACC: MOL.20010627.0026. 

157450 BSC 2001. Technical Work Plan for: Design Basis Waste Input for License 
Application. TWP-CRW-SE-000005 REV 00. [Washington, D.C.]: Bechtel SAIC 
Company. ACC: MOV.20010814.0002. 

158966 BSC 2002. The Enhanced Plan for Features, Events, and Processes (FEPs) at Yucca 
Mountain. TDR-WIS-PA-000005 REV 00. Las Vegas, Nevada: Bechtel SAIC 
Company. ACC: MOL.20020417.0385. 

160059 BSC 2002. Radionuclide Screening. ANL-WIS-MD-000006 REV 01. Las Vegas, 
Nevada: Bechtel SAIC Company. ACC: MOL.20020923.0177. 

165990 BSC 2003. 2002 Waste Stream Projections Report. TDR-CRW-SE-000022 REV 01. 
Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20031020.0002. 

162153 BSC 2003. Clad Degradation  Summary and Abstraction for LA. ANL-WIS-MD000021 
REV 00. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20030626.0002. 

163693 BSC 2003. DSNF and Other Waste Form Degradation Abstraction. ANL-WIS-MD000004 
REV 02. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20030711.0002. 


168795 BSC 2003. In-Package Chemistry for Waste Forms. ANL-EBS-MD-000056 REV 00, 
with 1 errata. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: MOL.20010322.0490; DOC.20031014.0005. 

166296 BSC 2003. Total System Performance Assessment-License Application Methods and 
Approach. TDR-WIS-PA-000006 REV 00 ICN 01. Las Vegas, Nevada: Bechtel 
SAIC Company. ACC: DOC.20031215.0001. 

169472 BSC 2004. D&E/PA/C IED Typical Waste Package Components Assembly. 800-IEDWIS0-
00202-000-00C. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: ENG.20040517.0008. 

167321 BSC 2004. CSNF Waste Form Degradation: Summary Abstraction. ANL-EBS-MD000015 
REV 01, with 2 errata. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20030708.0004; DOC.20031224.0001; DOC.20040202.0002. 

167058 BSC 2004. PWR and BWR Source Term Sensitivity Study. 000-00C-MGR0-00300000-
00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040114.0003. 

171583 BSC 2004. Technical Work Plan For: Regulatory Integration Modeling and Analysis 
of the Waste Form and Waste Package. TWP-WIS-MD-000009 REV 00 ICN 01. Las 
Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20040910.0001. 

170020 BSC 2004. Miscellaneous Waste-Form FEPs. ANL-WIS-MD-000009 REV 02. 
Las Vegas, Nevada: Bechtel SAIC Company. 

171407 BSC 2004. D&E/RIT IED - PWR and BWR Source Term. 000-IED-MGR0-00101000-
00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040812.0032. 

171435 BSC 2004. D&E/RIT IED - PWR and BWR Source Term. 000-IED-MGR0-00104000-
00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040813.0003. 

171436 BSC 2004. D&E/RIT IED - PWR and BWR Source Term. 000-IED-MGR0-00106000-
00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040813.0005. 

171502 BSC 2004. D&E/RIT IED PWR AND BWR SOURCE TERM. 000-IED-MGR000102-
000-00A. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: ENG.20040813.0001. 

171503 BSC 2004. D&E/RIT IED PWR AND BWR SOURCE TERM. 000-IED-MGR000103-
000-00A. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: ENG20040813.0002. 

166275 Canori, G.F. and Leitner, M.M. 2003. Project Requirements Document. TER-MGRMD-
000001 REV 02. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20031222.0006. 


170914 Carlisle, G. 1999. SCALE 4.2 Retirement. Memorandum from G. Carlisle to T. 
Doering, and D. Lancaster, September 3, 1999. ACC: MOL.20020110.0090. 

151947 CRWMS M&O 2000. Source Terms for HLW Glass Canisters. CAL-MGR-NU000002 
REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20000823.0004. 

153246 CRWMS M&O 2000. Total System Performance Assessment for the Site 
Recommendation. TDR-WIS-PA-000001 REV 00 ICN 01. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL.20001220.0045. 

150707 CRWMS M&O 2000. Waste Form Degradation Process Model Report. TDR-WISMD-
000001 REV 00 ICN 01. Las Vegas, Nevada: CRWMS M&O. 
ACC: MOL.20000713.0362. 

138239 CRWMS M&O 2000. Waste Packages and Source Terms for the Commercial 1999 
Design Basis Waste Streams. CAL-MGR-MD-000001 REV 00. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL.20000214.0479. 

156084 DeHart, M.D. and Hermann, O.W. 1996. An Extension of the Validation of SCALE 
(SAS2H) Isotopic Predictions for PWR Spent Fuel. ORNL/TM-13317. Oak Ridge, 
Tennessee: Oak Ridge National Laboratory. ACC: MOL.19970930.0475. 

104384 Dreyfus, D. 1995. Proposed Mix of DOE-Owned High Level Waste and Spent 
Nuclear Fuel Memorandum from D. Dreyfus (DOE/OCRWM) to J.E. Lytle 
(DOE/OEM), November 9, 1995, with attachment. ACC: MOL.19990319.0341 

103492 DOE (U.S. Department of Energy) 1985. An Evaluation of Commercial Repository 
Capacity for the Disposal of Defense High-Level Waste. DOE/DP/0020/1. 
Washington, D.C.: U.S. Department of Energy. ACC: MOL.20010730.0387. 

132333 DOE 1987. Appendix 2A. Physical Descriptions of LWR Fuel Assemblies. Volume 3 
of Characteristics of Spent Fuel, High-Level Waste, and Other Radioactive Wastes 
Which May Require Long-Term Isolation. DOE/RW-0184. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: HQX.19880405.0024. 

103191 DOE 1994. Final Supplemental Environmental Impact Statement, Defense Waste 
Processing Facility. DOE/EIS-0082-S. Aiken, South Carolina: U.S. Department of 
Energy. ACC: MOL.20010730.0383. 

171371 DOE 2001. Quality Assurance Audit Report EM-ARC-01-13 of the National Spent 
Nuclear Fuel Program At Idaho Falls, Idaho. EM-ARC-01-13. [Washington, D.C.]: 

U.S. Department of Energy. ACC: MOL.20011206.0198. 

155970 DOE 2002. Final Environmental Impact Statement for a Geologic Repository for the 
Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca 
Mountain, Nye County, Nevada. DOE/EIS-0250. Washington, D.C.: U.S. Department 
of Energy, Office of Civilian Radioactive Waste Management. 


ACC: MOL.20020524.0314; MOL.20020524.0315; MOL.20020524.0316; 
MOL.20020524.0317; MOL.20020524.0318; MOL.20020524.0319; 
MOL.20020524.0320. 

171675 DOE 2002. Quality Assurance Checklist, National Spent Fuel Nuclear Fuel Program. 
Dates of Evaluation September 23-27, 2002. Activity No.: EM-ARC-02-10. 
Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management:. ACC: MOL.20030303.0018. 

158873 DOE 2002. Waste Acceptance System Requirements Document. DOE/RW-0351, Rev. 

4. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive 
Waste Management. ACC: MOL.20020326.0056. 

155943 DOE 2002. Yucca Mountain Science and Engineering Report. DOE/RW-0539, Rev. 

1. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive 
Waste Management. ACC: MOL.20020404.0042. 

163377 DOE 2003. Source Term Estimates for DOE Spent Nuclear Fuels. DOE/SNF/REP078, 
Rev. 0. Idaho Falls, Idaho: U.S. Department of Energy, Idaho Operations Office. 
TIC: 254275. 

167367 DOE 2003. Repository Design Asset Functional & Operational Requirements and 
Design Solutions. DOE/RW-0600, Rev. 1. Las Vegas, Nevada: U.S. Department of 
Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20040202.0005. 

171539 DOE 2004. Quality Assurance Requirements and Description. DOE/RW-0333P, Rev. 

16. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive 
Waste Management. ACC: DOC.20040907.0002. 

169354 DOE 2004. Source Term Estimates for DOE Spent Nuclear Fuels. DOE/SNF/REP078 
Rev. 1. Three volumes. Idaho Falls, Idaho: U.S. Department of Energy, Idaho 
Operations Office. ACC: MOL.20040524.0451. 

164649 EPRI (Electric Power Research Institute) 1999. Determination of the Accuracy of 
Utility Spent-Fuel Burnup Records. EPRI TR-112054. Palo Alto, California: Electric 
Power Research Institute. TIC: 254706. 

154365 Freeze, G.A.; Brodsky, N.S.; and Swift, P.N. 2001. The Development of Information 
Catalogued in REV00 of the YMP FEP Database. TDR-WIS-MD-000003 REV 00 
ICN 01. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20010301.0237. 

164947 Hamel, W.F. 2003. Waste Treatment and Immobilization Plant (WTP) High-Level 
Waste (HLW) Canister Production Estimates to Support Analyses by the Yucca 
Mountain Project. Memorandum from W.F. Hamel (DOE) to W.J. Taylor, June 26, 
2003, with attachment. ACC: MOL.20030828.0080. 


150704 Harris, J.A. and Birch, P. 1987. The Effects on Topsoil of Storage During Opencast 
Mining Operations. Journal of the Science of Food and Agriculture, 40, (3), 
220-221. [New York, New York]: Elsevier. TIC: 243316. 

106563 Hermann, O.W. and DeHart, M.D. 1998. Validation of SCALE (SAS2H) Isotopic 
Predictions for BWR Spent Fuel. ORNL/TM-13315. Oak Ridge, Tennessee: 
Oak Ridge National Laboratory. TIC: 245042. 

131202 Lide, D.R., ed. 1991. CRC Handbook of Chemistry and Physics. 72nd Edition. Boca 
Raton, Florida: CRC Press. TIC: 3595. 

104398 Lytle, J.E. 1995. Disposal of DOE-owned High Level Waste and Spent Nuclear 
Fuel. Memorandum from J.E. Lytle (DOE) to D.A. Dreyfus (DOE/OCRWM), 
October 26, 1995. ACC: HQO.19951116.0015. 

170651 Massie, H.L., Jr. 2004. Reactor Record Uncertainty Determination. 32-5041666-02. 
Las Vegas, Nevada: Areva. ACC: DOC.20040623.0002. 

158051 McKenzie, J.M. 2001. Thermal Data for Naval Nuclear Propulsion Program Spent 
Nuclear Fuel Canister. Letter from J.M. McKenzie (Department of the Navy) to S.P. 
Mellington (DOE/YMSCO), August 29, 2001, Ser 08U/01-13933, with enclosure. 
ACC: MOL.20011029.0285. 

163274 NRC (U.S. Nuclear Regulatory Commission) 2003. Yucca Mountain Review Plan, 
Final Report. NUREG-1804, Rev. 2. Washington, D.C.: U.S. Nuclear Regulatory 
Commission, Office of Nuclear Material Safety and Safeguards. TIC: 254568. 

170912 ORNL (Oak Ridge National Laboratory) 2002. Quality Assurance Plan for the 
SCALE Computational System. SCALE-QAP-005, Rev. 1. Oak Ridge, Tennessee: 
Oak Ridge National Laboratory. ACC: MOL.20040708.0340. 

103896 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. 

164741 Singh, B. 2002. Nuclear Data Sheets for A = 79. Nuclear Data Sheets, 96, (1), 
1-240. San Diego, California: Elsevier. TIC: 254728. 

8.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES 

156605 10 CFR 63. Energy: Disposal of High-Level Radioactive Wastes in a Geologic 
Repository at Yucca Mountain, Nevada. Readily available. 

165519 40 CFR 197. Protection of Environment: Public Health and Environmental Radiation 
Protection Standards for Yucca Mountain, Nevada. Readily available. 


AP-2.22Q, Rev. 1, ICN 0. Classification Analyses and Maintenance of the Q-List. 
Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: DOC.20030807.0002. 


AP-3.12Q, Rev. 2, ICN 2. Design Calculations and Analyses. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20040318.0002. 


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


U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20040510.0004. 
AP-SIII.2Q, Rev. 1, ICN 2. Qualification of Unqualified Data. Washington, D.C.: 


U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20040127.0008. 


AP-SIII.9Q, Rev. 1, ICN 6. Scientific Analyses. Washington, D.C.: U.S. Department 
of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20040805.0003. 

AP-SV.1Q, Rev. 1, ICN 0. Control of the Electronic Management of Information. 
Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: DOC.20040308.001. 

LP-SI.11Q-BSC, Rev. 0, ICN 0. Software Management. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20040225.0007. 

153446 ANSI/ASME NQA-1. 1986. Quality Assurance Program Requirements for Nuclear 
Facilities. New York, New York: American Society of Mechanical Engineers. 
TIC: 223750. 

103902 ASME NQA-1 1994. Retention of Quality Assurance Records. New York, New York: 
American Society of Mechanical Engineers. TIC: 223334. 

105725 ASTM C 1174-97. 1998. Standard Practice for Prediction of the Long-Term 
Behavior of Materials, Including Waste Forms, Used in Engineered Barrier Systems 
(EBS) for Geological Disposal of High-Level Radioactive Waste. West 
Conshohocken, Pennsylvania: American Society for Testing and Materials. 
TIC: 246015. 

100016 Nuclear Waste Policy Amendments Act of 1987. Public Law No. 100-203, 101 Stat. 
1330. Readily available. 

8.3 SOFTWARE CODES 

154059 Software Code: SCALE. V4.3. HP. 30011 V4.3. 


154134 Software Code: SCALE. V4.3 (Addendum). HP. 30011 V4.3 (Addendum). 
154394 Software Code: SCALE. V4.4A. HP. 10129-4.4A-00. 

8.4 SOURCE DATA, LISTED BY DATA TRACKING NUMBER 

170760 MO0407SEPFEPLA.000. LA FEP List. Submittal date: 07/20/2004. 

8.5 OUTPUT DATA, LISTED BY DATA TRACKING NUMBER 

SN0310T0505503.004. Initial Radionuclide Inventories for TSPA-LA. Submittal 
date: 10/27/2003. 

The DIRS number 168761 was issued for this DTN upon the approval of Rev.00 of this analysis. 
Neither the DTN nor the data have changed as a result of generating Rev. 01 of this analysis. 


INTENTIONALLY LEFT BLANK 



APPENDIX I 
NEUTRON-ACTIVATION PRODUCTS OUTSIDE THE SPENT-FUEL MATRIX



INTENTIONALLY LEFT BLANK 



I. NEUTRON-ACTIVATION PRODUCTS OUTSIDE THE SPENT-FUEL MATRIX 

This appendix has been taken from Inventory Abstraction (BSC 2001 [DIRS 157575]) with only 
minor updates and editing. 

I.1 PURPOSE 

Most of the radioactivity associated with CSNF assemblies is contained within the spent-fuel 
matrix. To the extent that the cladding is credited as a barrier to the release of radioactivity, the 
release of radioactivity from within the cladding is associated with cladding failure. However, 
during reactor operation, some materials outside the fuel rods become radioactive due to neutron 
activation. Activation products outside the fuel rods can become available for release after 
breach of the waste package but long before the bulk of the radionuclide inventory from the fuel 
matrix is released. Two kinds of materials are candidates for neutron activation: (1) accumulated 
mineral deposits (crud) on the surface of the cladding that can build up during reactor operation, 
and (2) assembly hardware (e.g., top and bottom tie plates or nozzles, grid plates, springs, end 
caps, guide tubes, instrument tubes, or the cladding). The radionuclide inventories that are inputs 
to this analysis include activation products from the hardware but not from the crud. Activation 
products from the hardware have been lumped together with the inventory of radionuclides 
within the fuel matrix in the radionuclide inventory presented in Table 7-1. The purpose of this 
appendix is to examine the appropriateness of (1) neglecting the radionuclides in the crud and 

(2) combining the inventory of hardware-activation products in CSNF together with 
radionuclides in the fuel matrix for repository performance assessment. 

I.2 INPUTS 

The data used in this appendix were derived from BSC Design and Engineering calculations that 
provide radionuclide inventories as a function of time for CSNF assemblies (BSC 2004 
[DIRS 171435]; BSC 2004 [DIRS 171436]). 

I.3 ACTIVATED MINERAL DEPOSITS (CRUD) 

In addition to the source terms for an assembly based on the irradiated assembly, the activated 
corrosion products from the coolant deposited on the surfaces of the assembly (crud) are 
estimated. These surfaces include all the areas of the assembly exposed to the flow of coolant. 
The composition of the crud is determined by the conditions within the reactor during operation. 
A number of studies have identified the following radionuclides in PWR (BSC 2004 
[DIRS 171435]) and BWR (BSC 2004 [DIRS 171407]) crud: 51Cr, 54Mn, 55Fe, 58Co, 59Fe, 60Co,
63Ni, 65Zn, and 95Zr. These data are based on bounding estimates of surface area and the 
radionuclide activity density. With the exception of 63Ni, these radionuclides need not be 
considered further because they have half-lives of less than 10 years and become insignificant by 
the time SNF arrives at the potential repository; they have been screened out on that basis. 63Ni, 
which has a much longer half-life (about 100 years), was screened out by the scenario analyses 
described in Section 6 of Radionuclide Screening (BSC 2002 [DIRS 160059]). However, 
because the crud contribution was not included as input to the radionuclide-screening analysis, 
63Ni is considered in the crud. It has been estimated that there is no 63Ni in the crud present on 
BWR fuel (BSC 2004 [DIRS 171407]). The amount of 63Ni present in 30-year-old PWR crud is 


estimated to be 0.55 Ci (BSC 2004 [DIRS 171435]). The amount of 63Ni in hardware is shown 
to be much greater. For example, for an average 30-year-old PWR assembly, 20.5 Ci of 63Ni is 
calculated for the top-region hardware such as the top nozzle, spring retainer, upper end plug, 
and upper nuts (BSC 2004 [DIRS 171436]). Thus, the crud would contribute less than 3 percent 
of the total curies from crud and hardware combined. Therefore, due to the small contribution 
from crud and the relative short 100-year half-life of 63Ni, it is appropriate for the screening 
analysis in particular and for performance assessment generally to neglect the radionuclides 
present in SNF crud. 

I.4 HARDWARE 

The elements included as constituents of the hardware in BWR and PWR assemblies (BSC 2004 
[DIRS 171435]; BSC 2004 [DIRS 171436]), 99Tc and 14C, have been screened in and are 
expected to be present in hardware as neutron activation products. Radionuclides that have been 
screened out but have half-lives greater than 20 years and are present in the hardware are 93Zr,
59Ni, 63Ni, and 94Nb. 

The radionuclide-inventory calculations divide the BWR and PWR assemblies into four regions: 
fuel, bottom, plenum, and top. The calculations preserve the distinction between activation 
products (called light elements in the computer output) and fission products. The activation 
products in the bottom, plenum, and top regions originate from hardware because there is no fuel 
in those regions. In the fuel region, the distinction between activation products from hardware 
and from constituents of the fuel is not maintained. However, by noting the relative amounts of 
an element in the hardware and fuel that activates to a radionuclide of interest, it is often possible 
to determine the primary location of the activation product of interest. 

This analysis of activation products in hardware, specifically 14C, cites Appendix 2A. Physical 
Descriptions of LWR Fuel Assemblies (DOE 1987 [DIRS 132333]) to corroborate data pertaining 
to the fact that stainless steels such as Stainless Steel Type 304 are present in hardware 
components. This analysis is not impacted by the technical errors in that DOE report (DOE 1987 
[DIRS 132333]). 

I.4.1 Technetium-99 

99Tc is an abundant fission product (Parrington et al. 1996 [DIRS 103896]). Therefore, the 
inventory of 99Tc that is produced as a fission product in the matrix expected to be much greater 
than the inventory of 99Tc that is produced by hardware activation. However, at least for CSNF 
and naval fuel, the corrosion-resistant Zircaloy fuel cladding will impede the release of the bulk 
of the radionuclides from the fuel. For codisposal waste packages, no credit is taken in the 
TSPA-LA for DSNF cladding (BSC 2003 [DIRS 163693], p. 50). 

For CSNF, the amount of radioactivity released shortly after the breach of the waste package 
depends on the fraction of fuel rods with perforated cladding and the fraction of a perforated fuel 
rods radionuclide inventory that can escape. The fraction of fuel rods whose cladding will have 
failed by the time the waste package is breached depends on the condition of the fuel as 
emplaced and on repository conditions. Clad Degradation  Summary and Abstraction for LA 
(BSC 2003 [DIRS 162153], p. 38) estimates the fraction of rods whose cladding will have failed 


by the time the waste package is breached as ranging from 0.01 to 1 percent for Zircaloy rods. 
To this fraction, the fraction of rods with stainless steel cladding (1.04 percent) is added. The 
maximum fraction, 2.04 percent will be used in this Appendix for illustrative calculations. 

The fast-release fraction of the fission-product inventory for a perforated fuel rod will migrate 
through cracks and gaps in perforated fuel rods soon after waste-package breach. The 
recommended fast-release fraction for a perforated rod is given as a triangular distribution 
between 0.01 and 0.26 (BSC 2004 [DIRS 167321], Table 8.1-1). Illustrative calculations in this 
appendix will use the value of 0.1. 

The estimated inventory of fission-product 99Tc, for the average 25-year-old PWR assembly is 

8.99 Ci (Table 4-16). Because PWR fuels have low concentrations of Mo (BSC 2004 
[DIRS 171435]), which activates to 99Tc, for the average PWR assembly (Table 4-15), nearly the 
entire 99Tc activation product in the fuel region originates in the fuel-region hardware. The 
estimated total neutron-activation inventory of 99Tc for an average 25-year-old PWR assembly is 

1.10  10-3 Ci (Table 4-16). This value is an order of magnitude less than the mean fast-release 
inventory of 99Tc that would be available at that time from fission products upon waste package 
breach, 0.0204  0.1  8.99 Ci = 1.8  10-2 Ci. Similarly, for a bounding 5-year-old PWR 
assembly, the inventory of 99Tc from activation is less than the fission product inventory that 
would be available from fast release of fission products (Table 4-16). Therefore, combining the 
CSNF hardware 99Tc with the CSNF matrix 99Tc is a reasonable approximation for TSPA-LA. 

I.4.2 Carbon-14 

14C is not produced in appreciable amounts as a fission product (Parrington et al. 1996 
[DIRS 103896]). It is produced by activation of nitrogen, carbon, and oxygen in the fuel and 
hardware (Table 4-15). To judge the relative importance of the three elements as sources of 14C, 
consider the product of the reaction cross-section and the relative abundance (see Parrington et 
al. 1996 [DIRS 103896], pp. 18 and 19 for the cross-sections and relative abundances). For 
oxygen, the 17O abundance is 0.04 percent and the thermal (n,a) cross-section is 0.24 barns, 
yielding 9.6  10-5 barns per atom. For carbon, the 13C abundance is 1.1% and the thermal (n,.) 
cross section is 1.4 millibarns, yielding 1.5  10-5 barns per atom. For nitrogen, the 14N 
abundance is 99.63% and the thermal (n, p) cross-section is 1.83 barns, yielding 1.8 barns per 
atom. Therefore, unless concentrations of oxygen or carbon exceed those of nitrogen by many 
orders of magnitude, nitrogen is the most important source of 14C. 

Hardware in the fuel region contains little nitrogen, carbon, and oxygen in comparison to the fuel 
(BSC 2004 [DIRS 171436]). Therefore, the entire 14C inventory in the fuel region can be 
considered part of the fuel. 

Hardware outside the fuel region does contain elements that activate to 14C. Typically, the top 
and bottom nozzles or tie plates are composed of Stainless Steel Type 304, though other stainless 
steels are found (DOE 1987 [DIRS 132333], tables throughout labeled Fuel Assembly 
Hardware Parts and Materials Report). The top and bottom tie plates or nozzles account for 
most of the mass of the top- and bottom-region hardware and contain carbon and nitrogen 
(BSC 2004 [DIRS 171436]). The Zircaloy components contain some oxygen, but this source of 
14C can be neglected because the mass of oxygen is comparable to or less than that of nitrogen 


(BSC 2004 [DIRS 171435]). The plenum springs and the plenum getters are the only other 
sources of 14C outside the fuel region that are accounted for in the radionuclide inventory 
calculations (BSC 2004 [DIRS 171436]). The plenum springs and getters are protected by the 
cladding, so they should be lumped with the fuel for the purposes of this section. 

For the average 25-year-old BWR assembly (Table 4-16), 14C in the fuel matrix and the plenum 
region is approximately five times greater than in the top and bottom hardware (1.40  10-1 + 

4.11  10-3) / (6.85  10-3 + 2.44  10-2) = 4.6. The ratio of the 14C inventory in the top and 
bottom hardware to that available by fast release (see Appendix I.4.1) in the fuel region is 
estimated illustratively for an average BWR assembly as (6.85  10-3 + 2.44  10-2) / (1.40  10-1 
 0.0204  0.1) = 104.3 (Table 4-16). Thus, there is much more 14C available from activation of 
exposed stainless steel components of the hardware than from fast release from the fuel. 
Similarly, for average PWR and for bounding PWR and BWR assemblies, there is much more14C available from activation of exposed stainless steel components of the hardware than from 
fast release from the fuel (Table 4-16). 

This observation may seem to require that the 14C from CSNF hardware activation be treated as 
separate from the fuel matrix. However, 14C from DSNF will be available soon after waste 
package breach because the TSPA-LA model does not take credit for the DSNF cladding in 
codisposal waste packages. In fact, the dissolution of DSNF is conservatively modeled in the 
TSPA-LA to be instantaneous (BSC 2003 [DIRS 163693], p. 65). Therefore, it is important to 
consider the 14C contribution from DSNF in codisposal waste packages. 

First, consider the fraction of the 14C in CSNF waste packages that originate from hardware 
activation. Based on the representative assemblies that were used in the radionuclide inventory 
calculations, the ratios of top and bottom region inventory to total inventory for BWR and PWR 
assemblies from Table 4-16 are: 

(6.85  10-3 + 2.44  10-2) / 1.75  10-1 = 0.1786 = 17.9 percent for average BWR 
assemblies, and 

(2.26  10-3 + 0) / 3.32  10-1 = 0.00681 = 0.7 percent for average PWR assemblies. 

Similar ratios are obtained for bounding assemblies. 

(1.23  10-2 + 4.39  10-2) / 3.16  10-1 = 0.1778 = 17.8 percent for bounding BWR 
assemblies; 

(3.62  10-3 +0.00) / 5.35  10-1 = 0.0668 = 6.7 percent for bounding PWR assemblies. 

The great difference between the BWR and PWR fractions appears because the Babcock and 
Wilcox PWR assembly used for the inventory calculations has top and bottom nozzles composed 
of Stainless Steel Type CF3M, which does not contain nitrogen (N) as a constituent of the alloy 
(BSC 2004 [DIRS 171502]). The top and bottom nozzles or tie plates of most other assemblies 
(including Westinghouse and Combustion Engineering PWR assemblies and the representative 
General Electric BWR assembly) are made of Stainless Steel Type 304 (DOE 1987 
[DIRS 132333], Fuel Assembly Hardware Parts and Materials Reports throughout), which 


contains N (BSC 2004 [DIRS 171503]). Therefore, the fraction computed for the General 
Electric BWR assembly is taken to represent CSNF in general. Thus, based on the bounding 
BWR assembly, approximately 18 percent of the 14C inventory for CSNF assemblies originates 
from hardware activation. 

The repositorys 14C inventory from CSNF can be estimated as the product of the 14C inventory 
in a CSNF waste package, 1.37 g per waste package (Table 7-1); and the number of CSNF + 
naval waste packages, 7,772 (Table 6-3): 1.1  104 g. The repositorys 14C inventory from CSNF 
hardware is estimated as 18 percent of the total (i.e., 2.0  103 g). 

Now consider the 14C contribution from DSNF in codisposal waste packages. The amount of 14C 
available soon after waste package breach from codisposal waste packages is estimated to be 
3,412 waste packages (Table 6-3) times 1.78 g per waste package from DSNF (Table 7-1): 

6.1  103 g. 

Because the amount of 14C available from CSNF hardware (2.0  103 g) is similar to that 
available from codisposal waste packages (6.1  103 g), the consequence of combining hardware 
and fuel-matrix contributions is less benign than it is for 99Tc. As modeled in the TSPA-LA the 
early release from 14C is proportional to the contribution from codisposal waste packages plus 
the fast release contribution from CSNF waste packages (0.0204  0.1  1.1  104g = 22 g): 

6.1  103 g + 22 g . 6.1  103 g. Note that combining the 14C from CSNF hardware with the fuel 
and applying protective effect of the cladding effectively removes 14C due to CSNF from 
consideration. 

A more realistic model would allow the portion of the 14C inventory that is contained in the 
hardware to be released independent of the failure of the cladding. Depending on the 
degradation rate of the material containing the 14C, hardware degradation could be a significant 
contributor to the early release of 14C. At one extreme, hardware inventory could be released 
soon after breach of the waste package so that it would be available at roughly the same time as 
the fast-release inventory from the spent fuel. This would yield an early release roughly 
proportional to 6.1  103 g + 2.0  103g = 8.1  103 g. The ratio of the early release 
contributions from the two approaches (8.1  103 g / 6.1  103 g) is a rough correction factor for 
early releases due to 14C equal to about 1.3. Thus, combining the inventory of 14C from CSNF 
hardware-activation products together with the 14C in the CSNF matrix may cause a slight 
underestimate in the early release of 14C. With time, more and more of the 14C from the fuel 
matrix would be released, so the magnitude of the underestimate would diminish with time. 

Although 14C is not one of the primary contributors to general dose rate in the TSPA-SR model 
(CRWMS M&O 2000 [DIRS 153246], Figure 6.1-8), it is first among the lesser contributors 
before 20,000 years (CRWMS M&O 2000 [DIRS 153246], Figure 4.1-7). As shown by 
Figure 6.1-7 of Total System Performance Assessment for the Site Recommendation 
(CRWMS M&O 2000 [DIRS 153246]), 14C, along with 99Tc, is an important contributor to the 
dose from photon- and beta-emitting radionuclides in groundwater shortly after the end of the 
10,000-year regulatory period. Although the TSPA-SR document has been canceled as the 
project has moved closer to the License Application and preparation of TSPA-LA, these cited 
statements are appropriate for the use intended here due to the reliability of the source. Due to 
the importance of 14C to the early dose and the moderate significance of the 14C contribution 


from CSNF hardware, this analysis concludes that the 14C from hardware should not be lumped 
with the 14C for the fuel matrix in the TSPA-LA. 

I.4.3 Zirconium-93 

93Zr is produced by activation of 92Zr (Table 4-15), which is a constituent of Zircaloy 
components, most notably the cladding, and is not present in other hardware (BSC 2004 
[DIRS 171502]; BSC 2004 [DIRS 171503]). Because Zircaloy is very resistant to corrosion 
under repository conditions (BSC 2003 [DIRS 168795], Appendix II), significant release of 93Zr 
will be delayed well beyond waste package breach. Moreover, 93Zr has been screened out as a 
potential contributor to TSPA-LA dose calculations (BSC 2002 [DIRS 160059], Table 13). 
Therefore, the 93Zr produced by hardware activation can be neglected in assessing repository 
performance. 

I.4.4 Nickel-59, Nickel-63, and Niobium-94 

59Ni, 63Ni, and 94Nb are not produced in appreciable amounts as fission products 
(Parrington et al. 1996 [DIRS 103896]). They are produced by activation of Ni and Nb in 
hardware (Table 4-15). Some of the 59Ni, 63Ni, and 94Nb will originate in Inconel components. 
Because Inconel, like Zircaloy, is very corrosion resistant, the release of some fraction of the59Ni, 63Ni, and 94Nb would be delayed well beyond waste package breach and should not be 
treated as available upon waste package breach. 59Ni, 63Ni, and even 94Nb are also found in 
stainless steel components, which would degrade more quickly than Inconel would and release 
the radionuclides sooner. However, because 59Ni, 63Ni, and 94Nb have been screened out by the 
inventory screening analysis (BSC 2002 [DIRS 160059], Table 13), they can be neglected in 
assessing repository performance. 

I.5 CONCLUSION 

Radionuclide contributions from the crud on CSNF assemblies were screened out based on the 
methodology used in the screening analysis performed in Radionuclide Screening (BSC 2002 
[DIRS 160059]. Further analysis here based on current assessment of CSNF (BSC 2004 
[DIRS 171407]) shows that radionuclides in the crud can be neglected in TSPA-LA dose 
calculations. 

A negligible fraction of the 99Tc in CSNF originates from neutron activation of hardware. The 
inventories of 99Tc from CSNF hardware-activation products and from the CSNF matrix have 
been combined for TSPA-LA calculations. 

Approximately 18 percent of the 14C inventory in CSNF originates from neutron activation of 
stainless steel hardware outside the fuel rods. Combining the inventory of 14C from CSNF 
hardware-activation products together with the 14C in the CSNF matrix is a nonconservative 
approximation for repository performance assessment. It is recommended that TSPA analysts 
consider the importance of treating the 14C in the hardware separately. 


APPENDIX II 
DESCRIPTION OF FILES ON CD 



INTENTIONALLY LEFT BLANK 



II. DESCRIPTION OF FILES ON CD 


The Appendix III CD contains the 687kb workbook: inv_rev1.xls dated 10/31/2003. 
Information was added to this workbook on 6/23/2004. Note workbooks and worksheets are 
shown in Courier font throughout the report. 

newSRSbatches 
HLW 
Input 9 
newHLWtotal 
newHLW32 
Hanford Table 2 
HLW 
Input 10 
Hanford Table 3 
HLW 
Input 11 
HLWtotal 
HLWpp 
oldHLW32 
oldHLWlong 
oldHLWshort 
HLW 
Old Input 
inv32 
DSNFuncert 
DSNFpp 
DSNF g 
DSNF Ci Summary 32 
DSNF 
Input 7, 8 
gpCi 
Radionuclide 
Input 12,13,14 
new configs 
CSNFtotal 
CSNFpp 
CSNF32 
CSNFcasea63 
CSNF 
Input 3 
burnup 
CSNF 
Input 4 
Configurations 
Inputs 1,2,7 
Figure II-1. Flow of Information Between Worksheets in Excel Workbook inv_rev1.xls 

The following sections describe the arithmetic operations and calculations performed in this 
analysis and the content of the Excel worksheets that contain the data in these files. The data is 
provided on a CD, which is Appendix III. Figure II-1 shows flow of the calculations and 
information in between the worksheets in inv_rev1.xls. Arithmetic operations are also 
shown on the Excel worksheets. The cumulative list of worksheet files and plot files that can be 
found on the Appendix III CD is provided in III.1. 

II.1 GENERAL INFORMATION SHEETS IN INV_REV1.XLS 

Configurations, newconfigs, gpCi 

The information discussed in Sections 6.1 and 6.2 and presented in Table 4-2 is presented in 
Microsoft Excel worksheets configurations, new configs and charted in Chart config. 
Information in Table 4-3 is found in gpCi. 

gpCi (grams per curie) 

Using Equation II-1, the mass per activity (mi/ai) in g/Ci for each of 32 radionuclides (i) are 
calculated using constants by Parrington et al. (1996 [DIRS 103896]): half-lives in years (ti), 
pp. 18 to 50; Avogadros constant (NA), p. 59; 3.7  1010 Bq/Ci, p. 58; and 31,556,930 sec/yr, p. 

55. 


The latest half-life for 79Se was taken from nuclear data sheets by Singh (2002 [DIRS 164741]). 
The molecular weights (MWi) in g/mol were taken from the radionuclide names because that 
value had enough significant figures for these purposes: 

10

mi ( g/Ci)=3.7 10 Bq/Ci  31,556,930 sec/yr  MWi  ti (Eq. II-1) 
a ln(2)  Na 

II.2 CSNF SHEETS: 

a63, CSNFcasea63, CSNF32, CSNFpp, CSNFtotal, burnup, burnup recalc, 
config variability 


The average CSNF radionuclide and curie per assembly data for the year 2033 
(CRWMS M&O 2000 [DIRS 138239]) were transposed into columns of sheet CSNFcasea63. 

Sheet CSNF32 collected the information for the 32 isotopes of interest from the 61 radionuclides 
in CSNFcasea63. Next, the curies were converted to grams by multiplying by the g/Ci for each 
radionuclide from gpCi. To prevent errors when plotting zero grams in log-space, a cutoff of 1  
10-6 was imposed. 

The worksheet CSNFpp (pp stands for per package) takes the per assembly inventories in grams 
from CSNF32 and multiples by the number of assemblies per package for the configuration. The 
number of assemblies per package is taken from new configs. The data is repeated with a 1  
10-6 cutoff for log plotting. The weighted average of the five configurations is provided in 
column N, but calculated in column Q of CSNFtotal, which is discussed below. 

The worksheet CSNFtotal calculates the total grams for each waste type for each configuration 
by multiplying the per package values from CSNFpp by the number of packages in that 
configuration. The configuration is listed in row 2; the number of packages for each 
configuration is listed in row 3 as taken from configurations. The grand total for the 
appropriate CSNF package type is calculated in column G by summing the values in columns B 
through F. The weighted average is calculated in column H as the grand total divided by the 
total number of CSNF packages. This information is repeated with a 1  10-6 cutoff for 
log-plotting. The weighted average, i, is carried forward to inv32 for use in TSPA-LA. 

To estimate uncertainty in the weighted average in inv32, the burnups were examined in 
worksheet burnup. Weighted average burnups were calculated for the three 1999 waste stream 
arrival scenarios (CRWMS M&O 2000 [DIRS 138239]) and the four 2002 arrival scenarios 
(BSC 2003 [DIRS 165990], Tables 2 and 4). The calculated weighted average burnups were 
then compared to the nominal base for use in Section 6.6.1. The data in sheet: burnup recalc 
compares two methods of determining the uncertainty factors; one based on the number of 
packages associated with a given burnup and the other uses the MTU. The results are the same 
for the two methods. 

Sheet config variability is the data used to graph Figure 6-6, which shows the ratios of 
CSNF configuration inventories to the CSNF weighted average. 


II.3 HLW SHEETS: 

oldHLWlong, oldHLWshort, oldHLW32, newSRSbatches, Hanford table 3, Hanford 
table2, HLWpp, HLWtotal, newHLW32, newHLWtotal 

Inventory Abstraction (BSC 2001 [DIRS 157575]) data was repeated here before new 
information was made available. This old information has been kept in this document in order to 
examine the volatility of the projected radionuclide inventory estimates. The old calculations are 
presented first. 

The HLW radionuclide and curie per canister data for the year 2040 in Tables 6-1 and 6-2 of 
Source Terms for HLW Glass Canisters (CRWMS M&O 2000 [DIRS 151947]) were cut and 
pasted into sheets oldHLWshort and oldHLWlong, respectively. From these, the relevant data 
was collected into convenient format in sheet oldHLW32 columns. Columns L through P of 
oldHLW32 performed two tasks: they summarized the short glass canister inventory into 
minimum, maximum and average, and converted curies to grams. 

During production of this analysis, the codisposal configurations were changed from those in 
Table 4-2 to those in Table 6-3. Sheets HLWpp and HLWtotal were done for the first set of 
configurations, and newHLWtotal was used for the new set of configurations (new calculations 
described below). 

The worksheet HLWpp takes the per-canister inventories in grams from oldHLW32 and multiplies 
by the number of cans per package for the configuration. The number of canisters per package is 
taken from configurations and listed in row 4 (columns B-F). The weighted average of the 5 
configurations is provided in column N that is linked to column Q of HLWtotal where it is 
calculated. The worksheet HLWtotal calculates the total grams for each waste type for each 
configuration by multiplying the per package values from HLWpp by the number of packages in 
that configuration. The configuration is listed in row 3; the number of packages for each 
configuration is listed in row 4 as taken from configurations. The grand total for the codisposal 
package type is calculated in column G by summing the values in columns B through F 
(configurations 1 through 5 for CSNF). The weighted average is calculated in column H as the 
grand total divided by the total number of CSNF packages. This information is repeated in 
columns with a 1  10-6 cutoff for easy log plotting. 

The new calculations start with newSRSbatches (Allison 2004 [DIRS 168734] and Hanford 
table 2 (Hamel 2003 [DIRS 164947]), where curies per SRS batch and curies for all of 
Hanford for the year 2030, were collected and converted to curies per canister. Data shown in 
newSRSbatches continues to the right with comparisons of the curies in the old short canisters 
and with preparations for plotting Figure 6-11. Data in newHLW32 converts the curies per 
canister to grams per canister for the two sets of HLW glass data and compares the new to old 
calculations. The data was prepared for plotting Figure 6-10. The file: newHLWtotal 
calculated the weighted average based on the new number of canisters in new configs. It also 
prepares and plots the old and new HLW weighted averages. Hanfordtable3 is used to 
develop the uncertainty distribution as discussed in Section 6.6.3. 


II.4 DSNF SHEETS 

The following is a list of data worksheets for the DSNF analysis: DSNF Ci summary 32, 
DSNF g, DSNFpp, DSNF uncert, Chart DSNF uncert, DSNF Ci summary 32 
new, DSNF g (2), DSNFpp. 

The following is a list of plot files for DSNF: chart total, chart gpp, inv32 for 
plotting, inv32. 

The following is a list of plot files comparing the 2003 DSNF data with the 2004 data: 

DSNF compare plots 
old min 
old nom 
old reas 
old bound 
old max 
new nom 
new bound 


 old nom 
new nom 


Compare REV 0 and REV 1 DSNF inventory 
old min 
old nom 
old max 
new nom 


Radionuclide 
old min 
old bound 
Old max 
new bound 


old bound 
new bound 


DSNF Ci summary 32 collected information from the first eight worksheets of Total 2030 
D568585.xls file from the NSNFP report REP-078 (DOE 2003 [DIRS 163377]) for the 
32 isotopes of interest. DSNF g converts the curies to grams, and adds a column for the base 
case which contains all wastes except the bare assemblies. DSNFpp converts the total grams to 
grams per package and DSNF uncert computes the ratio of the maximum to the nominal value 
for each isotope. 

This analysis uses, as a basis for DSNF wastes, a second set of DSNF files created to compare 
radionuclide inventory data provided by Revisions 0 and 1 of Source Term Estimates for DOE 
Spent Nuclear Fuels (DOE 2003 [DIRS 163377]; DOE 2004 [DIRS 169354]). The new files 
parallel the content and the calculations in the same way as the original files using Rev. 0 of the 
referenced report. 


Therefore, DSNF Ci summary 32 new collected information from the first eight worksheets of 
Summary_2030.xls file from Source Term Estimates for DOE Spent Nuclear Fuels 
(DOE 2004 [DIRS 169354]) for the 32 isotopes of interest. DSNF g (2) converts the curies to 
grams and adds a column for the base case, which contains all wastes except the bare 
assemblies. DSNFpp (2) converts the total grams to grams per package and DSNF uncert 
(which is the same as the same uncertainty is applied to the new data) computes the ratio of, in 
this case, the bounding instead of maximum, to the nominal value for each isotope. Columns 
new nom and new bound from worksheet DSNF uncert are used to create the plots 
presented in Section 6.6.2 comparing the radionuclide inventory from the 2004 data with the 
2003 data used in this analysis to prepare the radionuclide inventory for TSPA-LA. 

II.5 TSPA-LA INVENTORY SHEETS 

Inv32, inv32 for plotting, Chart total, and Chart gpp 


The worksheet inv32 collects the weighted average grams per package for the 32 radionuclides 
for the three waste types into a single sheet. Note that the inventory reported is for various dates: 
CSNF 2033, DSNF 2030, and HLW 2030 respectively. The number of CSNF and codisposal 
waste packages is listed at the top of the columns. Note that there are 7,472 CSNF waste 
packages, but the addition of 300 naval waste packages that will be treated as CSNF waste 
packages in the TSPA-LA, increases the total as listed to 7,772. inv32 for plotting, Chart 
total, and Chart gppwere included for plotting purposes. 


INTENTIONALLY LEFT BLANK 



APPENDIX III 
WORKSHEETS AND PLOTS IN INV_REV1.XLS



INTENTIONALLY LEFT BLANK 



III. WORKSHEETS AND PLOTS IN INV_REV1.XLS


The following worksheets and plot files are in workbook inv_rev1.xls on the CD. The use 
and content of these is described in Appendix II and in the analysis. 

Worksheets: 

README 
chart total 
chart gpp 
inv32 for plotting 
inv32 


Plot files: 

DSNF compare plots 
old min 
old nom 
old reas 
old bound 
old max 
new nom 
new bound 


 old nom 
new nom 


Compare REV 0 and REV 1 DSNF inventory 
old min 
old nom 
old max 
new nom 


Radionuclide 
old min 
old bound 
Old max 
new bound 


old bound 
new bound 


Worksheets: 

DSNF uncert 
DSNFpp (2) 
DSNFpp 
DSNF g (2) 
DSNF g 
DSNF Ci summary 32 
DSNF Ci summary 32 new 
new HLW total 
newHLW32 
HLWtotal 
HLWpp 
Hanford table 2 



Hanford table 3 
newSRSbatches 
oldHLW32 
oldHLWshort 
oldHLWlong 
CSNF uncert 
burnup 
burnup recalc 
CSNFtotal 
config variability 
CSNFpp 
CSNF32 
CSNFcasea63 
a63 
gpCi 
Chart config 
new configs 
configurations 



APPENDIX IV 
QUALIFICATION OF UNQUALIFIED DATA USED AS DIRECT INPUTS 



INTENTIONALLY LEFT BLANK 



IV. QUALIFICATION OF UNQUALIFIED DATA USED AS DIRECT INPUTS 

Data from references used as direct input to this analysis must meet the definition of qualified 
in accordance with AP-3.15Q, Managing Technical Product Inputs. Input data that does not 
meet the definition qualified data require qualification in accordance with AP-SIII.9Q, 
Scientific Analyses, Section 5.5.1 k-m, requirements. Where data is obtained from outside 
sources that are not established facts, the data must be demonstrated to be suitable for the 
specific application. When appropriately justified these data are considered qualified for use 
within the technical product. This data qualification is intended to provide the desired level of 
confidence that the data are suitable for their intended use and the intended use is only for this 
analysis. AP-SIII.9Q, Section 5.2.1-l states, The extent to which the data demonstrate the 
properties of interest shall be addressed. One or more of the following factors shall be used when 
presenting the case that the data are suitable for intended use: 

 
Reliability of data source 
 
Qualification of personnel or organizations generating the data 
 
Prior uses of the data 
 
Availability of corroborating data. 


Data from the following sources are qualified for intended use in accordance with the above 
factors in AP-SIII.9Q: 

IV.1. Projected Glass Composition and Curie Content of Canisters from the Savannah 
River Site (U) (Allison 2004 [DIRS 168734], Appendix 1 (Fowler report, 
Rev.2)) 

IV.2. Waste Treatment and Immobilization Plant (WTP) High-Level Waste (HLW) 
Canister Production Estimates to Support Analyses by the Yucca Mountain 
Project (Hamel 2003 [DIRS 164947]) 

IV.3. Validation of SCALE (SASH2H) Isotopic Predictions for BWR Spent Fuel 
(Hermann and DeHart 1998 [DIRS 106563]); An Extension of the Validation of 
SCALE (SASH2H) Isotopic Predictions for PWR Spent Fuel (DeHart and 
Hermann 1996 [DIRS 156084]) 

IV.4. Source Term Estimates for DOE Spent Nuclear Fuels (DOE 2004 
[DIRS 163377]; Source Term Estimates for DOE Spent Nuclear Fuels (DOE 
2004 [DIRS 169354]). 

For data not derived from external sources, AP-SIII.2Q, Qualification of Unqualified Data, has 
been used for data qualification. Inputs qualified in accordance with AP-SIII.2Q require a Data 
Qualification Plan and use of one or a combination of methods identified in Attachment 3 of 
AP-SIII.2Q. These plans are provided in this appendix along with the results of the data 
qualification task. The data to be qualified in accordance with AP-SIII.2Q are from the 
following reference sources: 

IV.5 2002 Waste Stream Projections Report (BSC 2003 [DIRS 165990]). 


IV.1 Qualification of Data from Allison 2004 ( [DIRS 168734], Attachment I, Addendum 

I: Projected Glass Composition and Curie Content of Canister from the Savannah 
River Site (U) 

The following data qualification was performed in accordance with AP-SIII.9Q, Scientific 
Analyses, Section 5.2.1l. The following factors have been used to show that data are suitable for 
intended use: 

 
Reliability of data source 
 
Qualification of personnel or organizations generating the data 
 
Extent to which the data demonstrate the properties of interest 
 
Prior uses of the data. 


The data used in this analysis is from a report by J.R. Fowler, Projected Glass Composition and 
Curie Content of Canister from the Savannah River Site (U) attached as a memorandum from 

J. M. Allison, Manager, Department of Energy (DOE) Savannah River Operations Office (SR), 
to John Arthur III, Deputy Director, Office of Civilian Radioactive Waste Management, Las 
Vegas, NV (Allison 2004 [DIRS 168734], Attachment I and Appendix I). This report is the 
source of input data for Savannah River DHLW glass. The memorandum has been attached to 
this report. 

The memorandum states that these data have been reviewed and determined by the Deputy 
Director, Office of Civilian Radioactive Waste Management (OCRWM) as technically adequate 
and referenceable information on HLW glass. It has been determined that the use of these data 
are appropriate for this analysis and characterization of the radionuclide inventory for HLW 
forms/glass for the SRS. The memorandum included in this qualification notes that the 
Westinghouse Savannah River Company endorses the report demonstrating the reliability of the 
source. 

The data used in this analysis includes: the HLW radionuclide inventory and their source terms 
(i.e., curies per canister, and the number of canisters for each batch(es).) These data are shown 
in Table 4-9 New Estimated Curies Per Canister for the Savannah River Site (SRS), HLW 
(Batches 1A, 1B, 2&3, 4-10 of this analysis. The data are found in the Appendix 1 of the source 
document (Allison 2004 [DIRS 168734]): Tables 5-8 corresponding to Decay of SRS Sludge 
Batches 1A, 1B, 2&3, 7 and 4-10, the corresponding number of canisters identified as needed to 
contain the HLW in that batch where the curies per canister are decayed to the year 2030 (i.e., 
the projected year of emplacement of these wastes). The data are appropriate for the type of data 
under consideration and demonstrate the properties of interest. The properties of interest include 
the HLW radionuclide inventory in curies, the number of canisters projected as necessary for 
disposal of these wastes, and the curie levels projected to the year of projected emplacement of 
these wastes (currently 2030). These specific data are needed in order to produce the necessary 
input to TSPA-LA and, therefore, the data are considered qualified for intended use in that they 
demonstrate the properties of interest. 

Any comprehensive characterization of the inventory of radionuclides planned for geologic 
disposal must include HLW forms. Westinghouse Savannah River Company (WSRC) is the 
DOE contractor responsible for HLW form production and the Savannah River Defense Waste 


Processing Facility (DWPF) processes such waste for disposal. Having been informally reviewed 
(by John Arthur III) and concluded (by him) to be technically adequate for (the programs) needs, 
the WSRC has endorsed the report and formally entered it into their document control system 
and is submitting the report to the Office of Science and Technical Information. 

Attachment I to the memorandum states the appended (Fowler) report was prepared by WSRC 
to provide referenceable information on SRS HLW forms planned for geologic disposal 
(Allison 2004 [DIRS 168734]). The radiological data requested by DOE OCRWM and will be 
used in preparation of the preclosure and postclosure safety analyses for the Yucca Mountain 
Project (YMP) and License Application. This attachment also states that a design check of 
Appendix I (the Fowler report) has been performed by WSRC in accordance with the SRS 
procedure for design verification and checking and this design check is documented by 
signature of the reviewer on the Appendix 1 document (Allison 2004 [DIRS 168734]). 

It was also determined that the data cited were obtained from testing and analyses conducted at 
the Westinghouse Savannah under the requirements of ASME NQA-1 [DIRS 103902], Retention 
of Quality Assurance Records. This was confirmed by discussions with Joe Yanek, QA Manager 
for the Savannah River Site, where the analyses of the sludges were made. An NQA-1 quality 
assurance (ASME NQA-1 [DIRS 103902]) is a nuclear quality assurance program instituted by 
many DOE-contracted facilities and consists of 18 criteria that are the basis of the QARD that 
defines the quality assurance for the Yucca Mountain project. It contains requirements for the 
qualification of personnel for the work they do, the calibration of measuring and test equipment, 
requirements for the content and retention of records to name a few. 

The qualification team contacted the Savannah River site quality assurance manager, Mr. Joe 
Yanek, who confirmed the subject report was contracted out by the Westinghouse Savannah 
River Company. The author of the report, John R. Fowler, researched and collected data from 
existing documentation and reports containing analytical results from testing the composition of 
HLW glass and curie content of canisters based on these measured values and projections. The 
updated data in this report are based on either actual production of waste glass or the remaining 
high-level waste inventory stored at the SRS. Accordingly, the radiological content for the 
various glasses described provide actual concentrations for canisters already produced into a 
reasonable and realistic estimate for future production of SRS canisters (Allison 2004 
[DIRS 168734], Appendix 1, p.5). 

With respect to the analytical data in the Fowler report that characterized the sludge and waste 
forms, it was determined through interviews with the QA Manager of Westinghouse SRS, 
Joe Yanek, that Westinghouse SRS has had an NQA-1 Quality Assurance Program for many 
years. Mr. Yanek confirmed that one of the authors/co-authors (N.E. Bibler) of a referenced 
report providing analytical data on DWPF glass and sludge samples was known to him and that 
the laboratory analyses performed by Mr. Bibler would have been conducted under an NQA-1 
QA program. Such a program would be equivalent to the QA program that BSC is currently 
under. Therefore, based on this information, these data may be considered qualified for intended 
use based on qualification of personnel or organizations generating the data. 

As noted in this report, the inventory data needed is for projections of the curie levels at the year 
of projected emplacement. For HLW glass, this is currently 2030. The report notes that the 


values provided are bounding numbers and that, the exact number of canisters to be produced in 
the future is, as always, uncertain. The Fowler report states that the number of canisters from 
future production estimates is based on a 2002 forecast used by many within the DOE complex. 
Therefore, based on this information, the data used from this source is qualified for intended use 
based on prior uses of the data. 

The fact that the curie content and the number of canisters is formulated to be bounding using 
estimates based on actual canister production and current waste inventories and, therefore, they 
more accurately reflect the expected curie content and isotopic mix of DWPF glass canisters than 
the curie content the data used from this source meets data requirements for this analysis. In 
light of the fact that the data from sludge analyses is reported in a current report (Fowler report is 
dated 1/6/04), and these data have been reviewed and determined by the Deputy Director, 
OCRWM as technically adequate and referenceable information on HLW glass, it has been 
determined that the use of these data are appropriate for this analysis and characterization of the 
radionuclide inventory for HLW forms or glass for the SRS. 

These data are combined with the HLW data obtained from Hanford (Hamel 2003 
[DIRS 164947]). Uncertainty values are derived to show the range of values reported. The 
Fowler report states that the SRS production of canisters for HLW glass and the storage and 
filling of canisters is conducted in accordance with the environmental impact statements for the 
DWPF, such as Final Supplemental Environmental Impact Statement, Defense Waste Processing 
Facility (DOE 1994 [DIRS 103191]. 

Conclusion: Based on the information obtained and examined (and noted in this report), the data 
used from the referenced document (Allison 2004 [DIRS 168734]) meets the data qualification 
requirements in accordance with AP-SIII.9Q, and as such, provides the desired level of 
confidence that the data are suitable for intended use. The intended use is only for this work 
product. Therefore, no action is required under AP-SIII.3Q, Submittal and Incorporation of Data 
to the Technical Data Management System. However, the data and its source have been 
identified in the document input reference system under AP-3.15Q, Managing Technical Product 
Inputs. 


IV.2 QUALIFICATION OF DATA USED FROM WASTE TREATMENT 
IMMOBILIZATION PLANT (WTP) HIGH-LEVEL WASTE (HLW) CANISTER 
PRODUCTION ESTIMATES TO SUPPORT ANALYSES BY THE YUCCA 
MOUNTAIN PROJECT (HAMEL 2003 [DIRS 164947]) 

The following data qualification was performed in accordance with AP-SIII.9Q, Scientific 
Analyses, Section 5.2.1l. The following factors have been used to show that data are suitable for 
intended use: 

 
Reliability of data source 
 
Qualification of personnel or organizations generating the data 
 
Extent to which the data demonstrate the properties of interest 
 
Prior uses of the data. 


The source of the data used in this analysis comes from a memorandum from W.F. Hamel, 
Director, Waste Treatment Project Engineering & Commissioning Division of DOE Richland 
Operations Office, dated June 26 2003. The subject of the memorandum is Waste Treatment 
Immobilization Plant (WTP) High-Level Waste (HLW) Canister Production Estimates To 
Support Analyses By the Yucca Mountain Project. The memorandum has an attached report: 
Estimates of Immobilized High-Level Waste (HLW) to be Produced in the Hanford Waste 
Treatment and Immobilization Plant (WTP). 

The memorandum states that it transmits estimates of the HLW canister production 
requirements for the WTP and that these estimates have been derived from existing WTP 
Project information and have been prepared to support analyses by the Office of Civilian 
Radioactive Waste Management (RW) for the Yucca Mountain Project. 

The memorandum describes three Cases that bound the expected range of HLW canisters to be 
produced in the WTP; the WTP Program Case, the WTP Planning Case, and the WTP 
Technology Case. A description of these cases is provided in the memorandum with the 
associated wt % waste loading. These data in the memorandum along with the data in the 
attached report are used in this report and are the subject of this data qualification. 

The data from the Hamel memo and attached report that are used in this analysis are: 

1. 
The HLW canister production estimates that bound the range of HLW canisters 
produced in the WTP. 
2. 
The data from Table 1 of this reference summarizing the chemical composition of the 
HLW glass provided and the associated waste loading wt % for three cases: WTP 
program case, the planning case, and the technology case. 
3. 
Table 2, which provides the estimated radioactivity in the HLW canisters at decay 
dates of which only the data for 2030, and the waste loading percent. 
4. 
The data from Table 3 of this reference entitled: HLW Canister Production Estimates 
for Alternative Canister Waste Loading and Canister Fill Levels providing the 
percent canister fill levels (100%, 95% and 87%), corresponding to the three canister 





fill scenarios, maximum fill case, WTP contract fill case, and minimum fill case; the 
number of canisters for each case based on fill levels. 


5. 
The wt % of waste loading corresponding to the three cases from Table 2 of the 
reference is used to develop uncertainty related to Hanford HLW. The inputs from this 
reference are shown in Table 4-11. 


The estimates of canister production were prepared to support scoping studies to be conducted by 
DOE Richland Operations Office (RW). It is stated however that these scoping studies were not 
prepared under Quality Assurance Requirements and Description (DOE 2004 [DIRS 171539]). 

This qualification analysis has determined that there is confidence in the data acquisition and 
development of results; that results are appropriate, reasonable, and suitable for their intended 
use and the information is transmitted by sources or professionals considered reliable and 
qualified. 

The transmittal letter and report are prepared by an individual (William J. Hamel) prominent in 
the organization responsible for managing and assessing the production and content of HLW at 
the Hanford facility in Richland, WA.; he is the Director of the Waste Treatment Project 
Engineering & Commissioning Division. The memorandum states that the estimates of HLW 
canister production are stated to bound the expected range of HLW canisters to be produced by 
the Waste Treatment and Immobilization Plant (WTP). Although the estimates provided were 
prepared to support scoping studies that will be conducted by RW and have not been prepared 
under Quality Assurance Requirements and Description (DOE 2004 [DIRS 171539]), all three 
cases are believed to be compliant with the current Waste Acceptance System Requirements 
Document (WASRD) (DOE 2002 [DIRS 158873]). The memorandum states that other DOE 
organizations within RW will be using these estimates in addition to the fact that the WASRD is 
a well-known and used requirements document used for planning the management of HLWs. 

The data provided in Table 2 of the reference summarizes the radiochemical inventory in curies 
of the HLW canisters for the decay date of 2030 needed in the analysis report. Table 3 of the 
reference provides data with respect to canister fill levels for the three cases (A through C) that 
provide bounding estimates on the range of HLW canisters to be produced by the WTP. The 
memorandum states that: All three cases result in a HLW form that is believed to be compliant 
with RW's acceptance requirements identified in Waste Acceptance System Requirements 
Document (WASRD). These data, chemical composition of HLW glass, the total curies per 
radionuclide for the radionuclides important to dose, and number of canisters for the three cases 
are intended to bound the range of HLW canisters and content for the year of projected 
emplacement, 2030. These data are needed for this report under revision by the RIT in order to 
provide necessary data on the radionuclide inventory to be used as input by TSPA-LA. 

Conclusion: This data qualification analysis shows that the data used from the Hamel reference 
are qualified for intended use. The basis for this conclusion is that both the transmittal letter and 
the report are prepared by an individual (William J. Hamel) who is prominent in the organization 
responsible for managing and assessing the production and content of HLW at the Hanford 
facility in Richland, WA. Moreover, the estimates of HLW canister production are stated to 
bound the expected range of HLW canisters to be produced by the Waste Treatment and 


Immobilization Plant (WTP). Although the estimates provided were prepared to scoping 
studies that will be conducted by RW and have not been prepared under Quality Assurance 
Requirements and Description (DOE 2004 [DIRS 171539]), all three cases are believed to be 
compliant with the current Waste Acceptance System Requirements Document (WASRD). 
The fact that other DOE organizations within RW will be using these estimates in addition to the 
fact that the WASRD is a well known and used requirements document used for planning the 
management of HLWs, it is believed that the data reported in this reference source is qualified 
for the intended use in the this report. In addition, the data provided in Table 2 of the reference 
summarizes the radiochemical inventory in curies of the HLW canisters for the decay date of 
2030 needed in the analysis report. Table 3 of the reference provides data with respect to canister 
fill levels for three fill scenarios (maximum, (100 percent), WTP contract fill (95 percent), and 
minimum fill case ((87 percent), correlated to the canister production numbers. These data allow 
uncertainty calculations to be applied to the inventory projected for 2030. 

In addition, the information obtained and examined provides the desired level of confidence that 
the data used from the subject reference are suitable for intended use, and the intended use is 
only for this work product. Therefore, no action is required under AP-SIII.3Q, Submittal and 
Incorporation of Data to the Technical Data Management System. However, the data and their 
sources have been identified in the Document Input Reference System under AP-3.15Q, 
Managing Technical Product Inputs. 

IV.3 QUALIFICATION OF DATA USED FROM AN EXTENSION OF THE 
VALIDATION OF SCALE (SAS2H) ISOTOPIC PREDICTIONS FOR PWR 
SPENT FUEL (DEHART AND HERMANN 1996 [DIRS 156084]) AND 
VALIDATION OF SCALE (SAS2H) ISOTOPIC PREDICTIONS FOR BWR 
SPENT FUEL (HERMANN AND DEHART 1998 [DIRS 106563]) 

A data qualification effort has been conducted for the following data used as direct input to this 
analysis and referenced from the reports: 

 
Validation of Scale (SAS2H) Isotopic Predictions for BWR Spent Fuel, Hermann, O.W. 
and DeHart, M.D. 1998. ORNL/TM-13315; Table 24: Summary of percentage 
differences between measured and computed composition in this BWR validation study 
as averages and spreads. Specifically, the minimum and maximum values (spread) 
associated with the identified BWR radionuclides of interest to TSPA-LA were used in 
this analysis. The equation used to determine these uncertainties and referenced in the 
Table 24 is: (Calculated/measured - 1)  100%. The values from Table 24 are presented 
in Table 4-7 of this analysis. 
 
An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent 
Fuel, DeHart, M.D. and Hermann, O.W. 1996, ORNL/TM-13317; Table 19: Summary 
of percentage differences between measure and computed composition in this study as 
averages and spreads. Specifically, the minimum and maximum values (spread) 
associated with the identified PWR radionuclides that are of interest to TSPA-LA were 
used in this analysis. The equation used to determine these uncertainties (referenced in 
the Table 24) is: (Calculated/measured - 1)  100%. The values from Table 19 are 
presented in Table 4-7. 



The data from these two sources were used to identify the maximum and minimum correction 
factors associated with computational methods. When combined with the maximum and 
minimum projected average burnup ratios developed from arrival scenarios in Section 6.6.1 of 
this analysis, the overall uncertainty for the CSNF inventory was determined. 

The following data qualification was performed in accordance with AP-SIII.9Q, Scientific 
Analyses, Section 5.2.1l. The following factors have been used to show that data are suitable for 
intended use: 

 
Reliability of data source 
 
Extent to which the data demonstrate the properties of interest 
 
Prior uses of the data. 


The average percent difference between measured and calculated radionuclide isotopes was 
derived using the SCALE (Standardized Computer Analyses for Licensing Evaluation) Code 
System. The SCALE Code System has been developed by the Nuclear Analysis Methods and 
Applications (NAMA) Group of the Nuclear Science and Technology Division at the Oak Ridge 
National Laboratory (ORNL). This work has been ongoing since 1976 for the Nuclear 
Regulatory Commission (NRC) Office of Nuclear Material Safety and Safeguards (NMSS) and 
NRC Office of Research. The SCALE code system is composed of control modules 
(e.g., SAS2H (Shielding Analysis Sequence 2H)), which drive functional modules (e.g., 
ORIGEN-S) to perform nuclear-related calculations, (e.g., isotope generation and depletion in 
the case of SAS2H/ORIGEN-S and as those used by this analysis). The major analysis modules 
and libraries implemented for the SCALE 4.0 release in 1990 had already been used for 
production quality analyses for at least 10 to 25 years then, and were already widely accepted as 
qualified by the nuclear industry. 

In 1988, during the development of the SCALE 4.0 release, the Department of Energy (DOE) 
Office of Civilian Radioactive Waste Management (OCRWM) funded the development of a 
quality assurance program for SCALE, as described in SCALE-QAP-005, Rev.1 (ORNL 2002 
[DIRS 170912]). This plan is based on requirements from: DOE Order 5700.6C, Quality 
Assurance; and ASME NQA-1, Quality Assurance Requirements for Nuclear Facilities 
(ASME 1986 [DIRS 153446]; as well as on ORNL Standards Based Management System QA 
requirements. 

The SCALE QA Plan is complemented by Configuration Management Plan for the SCALE Code 
System (Bowman 2002 [DIRS 170913]), which provides configuration control of software for 
use at ORNL and for release through the Radiation Safety Informational Computational Center 
to users such as the Yucca Mountain Project. Formal verification and validation (V&V) of the 
SCALE code system is controlled by Verification and Validation Plan for the SCALE Code 
System (Broadhead 1996 [DIRS 170910]). 

The versions of the SCALE code system that have been baselined at the Yucca Mountain Project 
include: SCALE 4.2 (CRWMS M&O 1996 [DIRS 170914]); SCALE 4.3 (SCALE V4.3, 
CSCI: 30011 [DIRS 154059]; SCALE V4.3, CSCI: 30011 Addendum [DIRS 154134]) and the 
current version, SCALE 4.4a (SCALE V4.4a, STN: 10129-4.4A-00 [DIRS 154394]). All of 


these were developed at ORNL under NRC sponsorship, following some approved version of 
these QA, Configuration Management, and V&V plans. 

Therefore, there is ample evidence as to the reliability of the sources of these data and use 
throughout the nuclear community. The authors of the subject references and data were among 
the developers as well as authors of the validation studies for SCALE and were members of the 
NAMA Group at ORNL. Over 20 publications using SCALE (SAS2H) are available for viewing 
and download from the on-line Oak Ridge National Laboratory (ORNL) Comprehensive 
Publications and Presentations Registry (http://lib1.isd.ornl.gov). 

The data meets the needs of this analysis for determination of uncertainty factors associated with 
the CSNF inventory calculations, as it provides isotope depletion statistics provided using well 
known methods to account for differences between measure and computed composition given as 
averages and spreads. 

Conclusion: Based on an evaluation of the factors identified from AP-SIII.9Q, it has concluded 
that the data are qualified for its intended use in this report; that the sources for the data and the 
data themselves provide the desired level of confidence needed for this analysis. These data are 
qualified here only for this work product and, therefore, no action required by AP-SIII.3Q. 
However, actions have been taken to update the Document Input Reference System in 
accordance with AP-3.15Q. 

IV.4 QUALIFICATION OF DATA USED FROM SOURCE TERM ESTIMATES FOR 
DOE SPENT NUCLEAR FUELS (DOE 2003 [DIRS 163377]) AND SOURCE TERM 
ESTIMATES FOR DOE SPENT NUCLEAR FUELS (DOE 2004 [DIRS 169354]) 

A data qualification effort has been conducted for the following data used as direct input to the 
this analysis and referenced from the reports: 

 
Source Term Estimates for DOE Spent Nuclear Fuels. DOE/SNF/REP-078, Rev. 0. 
Idaho Falls, Idaho: U.S. Department of Energy, Idaho Operations Office (DOE 2003 
[DIRS 163377]); 2030 Summary Tables from pages D-568 through D-575, estimates of 
the radiological source terms for spent nuclear fuels owned by DOE. Section 8, 
Uncertainty and Error to provide the uncertainty for the data. 
 
Source Term Estimates for DOE Spent Nuclear Fuels. DOE/SNF/REP-078 Rev.1. Three 
volumes. Idaho Falls, Idaho: U.S. Department of Energy, Idaho Operations Office 
(DOE 2004 [DIRS 169354]): Page 41, estimate of DOE spent nuclear fuel canisters to 
be used. 
 
Source Term Estimates for DOE Spent Nuclear Fuels (DOE 2004 [DIRS 169354]) 
Revision 1 is used as corroborative input to provide updated DSNF radionuclide 
inventory data published January 2004 and to provide a canister count that includes the 
uncertainty of the data. These data are compared with the DSNF radionuclide inventory 
data used in this analysis to validate the inputs used and validate the uncertainty factors 
developed. The data used in this analysis to develop the inventory comes from 
Revision 0 of the same report (DOE 2003 [DIRS 163377]). 



The following data qualification was performed in accordance with AP-SIII.9Q, Scientific 
Analyses, Section 5.2.1l. The following factors have been used to show that data are suitable for 
intended use: 

 
Reliability of data source 
 
Qualification of personnel or organizations generating the data 
 
Extent to which the data demonstrate the properties of interest. 


DSNF includes diverse fuels from various experimental, research, and production reactors and 
consists of several hundred different fuel types that have been stored at several sites over the 
years. As a result, the availability and completeness of the radionuclide inventories and 
associated documentation varies considerably for DSNF. A process for creating a conservative 
estimate of these SNF source terms was developed by a team of experts representing each of the 
DSNF storage sites. The process relies on precalculated results that are used to provide 
radionuclide inventories for typical SNFs at a range of decay times. These results are used to 
develop templates that are scaled to estimate the radionuclide inventory of other similar fuels. 
The templates were generated using ORIGEN-based calculational techniques. Several validation 
studies, referenced in the report, have been performed to demonstrate the validity of the model 
and underlying codes used. 

The report was produced by the Idaho National Engineering and Environmental Laboratory 
(INEEL) under contract number DE-AC07-99ID13727 to U.S. Department of Energy for use in 
the Yucca Mountain Project. The Quality Assurance section states that procedures that 
implement Quality Assurance Requirements and Description (DOE 2004 [DIRS 171539]) 
principles were applied to the report. Audits have been performed on the Quality Assurance 
programs at INEEL by the Office of Civilian Radioactive Waste Management (OCRWM). 

OCRWM performed an audit of selected QA program elements of INEELs Quality Assurance 
program September 17-21, 2001 (DOE 2001 [DIRS 171371]). Included in the program elements 
evaluated was software. The results of the audit was satisfactory and effective implementation of 
the examined portions of the Office of Civilian Radioactive Waste Management (OCRWM) QA 
program in accordance with Quality Assurance Requirements and Description (QARD) 
(DOE 2004 [DIRS 171539]) and applicable implementing procedures. 

OCRWM performed an audit of selected QA program elements of the INEEL quality assurance 
program September 23-26, 2002 (DOE 2002 [DIRS 171675]). Included in the program elements 
evaluated was software. The results of the audit were satisfactory and effective implementation 
of the examined portions of the Office of Civilian Radioactive Waste Management (OCRWM) 
QA program with the exception of deficiencies written during the audit. Deficiency Reports EM 
(O)-03-D-004 through 007 were written because of the audit and closed by 06/25/03. 

The reports establish a process for a conservative estimate of DSNF source terms. The process 
relies on precalculated results that provide radionuclide inventories for typical SNFs at a range of 
decay times. These results are used as templates that are scaled to estimate radionuclides for 
other similar fuels. To estimate a SNF source term, an appropriate template is selected to model 
the production of activation products and transuranics by matching the reactor moderator, fuel 
cladding and compound and beginning-of-life enrichment. Precalculated radionuclide inventories 


are extracted from the appropriate template for the desired decay period and then scaled to 
account for differences in fuel mass and specific burnup. By modeling various combinations of 
reactor moderator, fuel enrichment, fuel compound and cladding, templates have been developed 
to reasonably model a broad range of DSNFs. The template methodology enables a source term 
estimate to be completed for virtually any DSNF for decay dates up to 100 years following 
reactor shutdown. 

The inventory estimates from the reports provide both a nominal and a bounding radionuclide 
inventory estimates for each of the many and heterogeneous types of DSNFs. 

Conclusion: Based on an evaluation of the factors identified from AP-SIII.9Q, it has been 
concluded that the data are qualified for its intended use in this report; that the sources for the 
data and the data themselves provide the desired level of confidence needed for this analysis. 
These data are qualified here only for this work product and, therefore, no action required under 
AP-SIII.3Q. However, actions have been taken to update the Document Input Reference System 
under AP-3.15Q. 

IV.5 QUALIFICATION OF DATA USED FROM 2002 WASTE STREAM 
PROJECTIONS REPORT (BSC 2002 [DIRS 160059]) 

The following data qualification was performed in accordance with AP-SIII.2Q, Qualification of 
Unqualified Data. The qualification method used is technical assessment and is included as part 
of this analysis. The technical assessment approach may be taken when it is determined that an 
independent evaluation of the data by a subject matter expert is needed to raise the confidence of 
the data to a proper level for the intended use. The rationale for choosing this method is that this 
report develops project data that cannot be verified at this time. The criteria for this qualification 
are based on the evaluation of attributes and the determination that the data can be qualified for 
intended use as it is applied in this report. This qualification does not qualify the input for other 
analyses or for general use, but is limited to its use in this analysis. 

The qualification attributes include: (1) the extent to which the data demonstrate the properties of 
interest; (2) prior peer reviews and uses of the data and their results and associated verification 
processes; (3) extent and reliability of the documentation associated with the data; (4) the 
importance of the data to showing that the repository design meets the performance objectives of 
10 CFR Part 63 [DIRS 156605]. 

A data qualification effort has been conducted for data used by this analysis taken from 2002 
Waste Stream Projections Report (BSC 2002 [DIRS 165990]). The data are: 

1. 
Total number of BWR and PWR assemblies for the 63K Case A, 63K Case B, 
63K Case C, and 63K Case D arrival scenarios and methodologies shown in 
Table 2: CSNF Assembly Total for Discharge Projections of the subject reference. 
These data are shown in Table 4-6 of this analysis: CSNF Assembly Totals and 
Characteristics for the 63,000 MTHM Cases and referred to as Cases A' to D'. 
2. 
The average burnup (GWd/MTU) for BWR and PWR assemblies for waste steam 
scenarios: Case A through Case D from Table 4: Repository Arrival CSNF Assembly 
Characteristics for the 63K MTHM Cases of the reference also shown in Table 4-6 of 



this analysis: CSNF Assembly Totals and Characteristics for the 63,000 MTHM 
Cases. 

The qualification was conducted to determine if these data from 2002 Waste Stream Projections 
Report (TDR-CRW-SE-000022 REV 01) (BSC 2003 [DIRS 165990]) can be used as a direct 
input in Initial Radionuclide Inventories (ANL-WIS-MD-000020 REV01). The data used from 
this reference is used to derive the uncertainty factors to be applied to the average radionuclide 
inventory for CSNF as calculated in Section 6.6.1. The following justification is provided for this 
determination. 

The reference document, 2002 Waste Stream Projections Report (BSC 2003 [DIRS 165990]), 
was produced under the Office of Civilian Radioactive Waste Management for the Yucca 
Mountain Project. The document was not produced as QA:QA and although an Activity 
Evaluation was conducted in accordance with the Technical Work Plan for Design Basis Input 
for License Application (BSC 2001 [DIRS 157450]), it was determined that the report was not 
subject to the requirements based on Quality Assurance Requirements and Description. The 
inherent nature of the data does not lend itself to direct qualification because it is based on 
projections using data gathered from a variety of unqualified sources that include various nuclear 
power utilities. Raw data is originally submitted to the Department of Energy are compiled and 
Waste Steam Projection Reports are developed as necessary to establish the design basis. Office 
of Civilian Radioactive Waste Management (OCRWM) has accepted this process for design 
basis using data provided in these projection reports. The data contained in the reports is 
generally considered the best available and only available data because it originates directly from 
the utilities and has traceability to its source. 

2002 Waste Stream Projections Report (BSC 2003 [DIRS 165990]) considers scenarios based on 
the design-basis legal limit of 63,000 MTHM CSNF and 7,000 MTHM government-managed 
wastes. As noted in the report, the purpose of the report was to provide information on waste 
stream characteristics and bounding and average waste stream, transportation characteristics as 
well as waste arrival data at the repository for use in the CRWMS design and development of 
system level requirements. 2002 Waste Stream Projections Report (BSC 2003 [DIRS 165990]) 
contains data sets that span the range of possible delivery options for CSNF, government-
managed wastes and HLW although the data apply primarily to CSNF. It is noted in the report 
that these data capture probable limits of what might occur in the future with regard to fuel 
selection by the utilities. The report addresses the influence of arrival sequences of CSNF 
intended to provide insight into design margins for parameters that are affected by the waste 
stream (BSC 2003 [DIRS 165990], p. 1). 

Because much of the data are projections far into the future and that there is no factual data to 
use (i.e., it is recognized that waste characteristic may be different from those described in this 
document), the data are technically correct in this respect. The use of burnup data from 
projections based upon current inventory data and applying arrival scenarios to develop 
uncertainty factors related to the CSNF inventory at the year of projected emplacement is 
appropriate for the intended use in this model report. The information and data provided in 2002 
Waste Stream Projections Report support the conclusion that the data and the methodologies 
used for this analysis are appropriate. The data demonstrate the properties of interest (burnup 
projections for the year of projected emplacement; the importance of the data to showing that the 


proposed repository design meets the performance objectives of 10 CFR 60 (uncertainty 
associated with the CSNF inventory derived using current data and a range of arrival scenarios) 
all support the conclusion that the data used from this reference are qualified for intended use. 
Moreover, the report was reviewed and approved for use by the BSC Manager of Projects, 
Nancy Williams on 4/7/2003 (BSC 2003 [DIRS 165990]). 

In conclusion, the Data Qualification Team has concluded that the source for the data and the 
data themselves provide the desired level of confidence needed for this analysis and based on the 
criteria and attributes used to evaluate the subject data, these data are qualified for intended use 
in this report and fulfill the performance objectives of the intended use. The technical expert, 
who is the technical checker for this analysis, has reviewed this data qualification and confirms 
that it meets all the requirements of AP-SIII.2.Q. 

These data are qualified here only for this work product and, therefore, no action is required in 
accordance with AP-SIII.3Q. However, actions have been taken to update the Document Input 
Reference System under AP-3.15Q. 


APPENDIX V 
DATA QUALIFICATION PLAN 



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V. DATA QUALIFICATION PLAN 




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